Prompt Gamma Neutron Activation Analysis

Library of Congress Cataloging-in-Publication Data Prompt gamma neutron activation analysis / edited by Zeev £3. A l f ...

Author: Zeev Alfassi | Chien Chung

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Library of Congress Cataloging-in-Publication Data Prompt gamma neutron activation analysis / edited by Zeev £3. A l f w i and Chien Chung. p. cm. Includes bibliogrilphical references m d index. ISBN 0-8493-5 149-9 1. Nuclear activation analysis. I. Alfassi, Zeev B. 11. Chung. Chien, 1950- .

QDGOG.P76 1995 543'.08824c20

94-24929

CIP

This book contains information obtaincd from aulhcntic imd highly regardcd sources. Reprinted material is quoted with permission, and sources are indicated. A wide variety of references are listed. Reasonable efforts have been made to publish reliable data and information, but the author and the publisher cannot assume responsibility for the validity of all materials or for the consequences of their usc. Neither this book nor any part may be reproduced or transmitted in any Corn1 or by any means, electronic or mechanical, including photocopying, microfilming, and rccording, or by any information storage or retrieval system, without prior permission in writing from the publisher. All rights reserved. Authorization to photocopy items for internal or personal use. or the personal or internal use of specific clients, may be granted by CRC Press, Inc., provided that $.SO per page photocopied is paid directly to Copyright Clcnrilnce Ccntcr, 27 Congress Strecl, Salc~n.MA 01970 USA. Thc fce code for uscrs of the Transactional Reporting Scrvicc is ISBN 0-8493-5 149-9/95/$0.00+$.50. The fee is subject to change without notice. For organizationsthat have been granted a photocopy license by the CCC, a separate systenl of payment has been arranged. CRC Press, Inc.'s consent does not extend to copying for general distribution, for promotion, for creating new works, or for resale. Specific permission must be obtained in writing from CRC Press for such copying. Direct all inquiries to CRC Press, Inc., 2000 Corporate Blvd., N.W., Boca Raton, Florida 33431. O 1995 by CRC Press, Inc.

No claim to original U.S.Government works International Standard Book Number 0-8493-5149-9 Library of Congress Card Number 94-24929 Printed in the United States of America 1 2 3 4 5 6 7 8 9 0 Printed on acid-free paper

CONTENTS

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.

2 iTI~tsu~en and t s Shielding ;. :,: - $ ;=:c~z& Chung

.............................................................................................................

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Chapter 3 ~ e & hDamage n and Induced Effects on Nuclear Instruments Used for PGAA chi% Chung

dh+$p- 4

pr&ipt Gamma Neutmn Activation Analysis with Reactor Neutrons Zeev B. Aljiassi

.................

...................................

Chapter 5 PGNAA with Radioisotopic Sources, Neutron Generators, and Charged Particle . Acceler{ators ................................................................................................................................. Zeev B. Avassi Chapter 6 Prompt Gamma Actitration Analysis with Guided Neutron Beams Riclrfird M. Lirrdstron? arid Chus.hirn Yoneznrvn CZmptcr 9 . . Prompt G,zmma Activation Analysis

.F.t;l Irivt:,

.........................................

..............................................................................

C ' I p k l chliq

(CBI::H)~CT

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I n Sitrt 'rrk~qrlicationls......................................................................................................................... Jiunn-h'sihR Chao Chapter 9 On-Line Applications Jiunn-Hsing Chao

..................................................................................................................

Appendix P - Thermal Neutron Capture Gamma-Rays J. K. TLili

.......................................................... ......................

Appendix 31 Otlaer Rndiation-ReZated Properties of Prompt Gamma Activation Facility Chien Chrng

.........................

THE EDITORS Zeev B. Alfassi, Ph.D., is professor of Radiochemistry and the Chairman of the Nuclear Engineering Department in the Ben Gurion University, Beer Sheva, Israel. Professor Alfassi received his B.Sc. and M.Sc. degrees from the Hebrew University in Jerusalem in 1964 and 1965, respectively, in the fields of chemistry and biochemistry. He received his Ph.D. from the Weizmann Institute of Science and the Soreq Nuclear Research Center in 1970. Professor Alfassi is a member of the council of the Israel Nuclear Society. He has published more than 150 scientific papers and edited the CRC books Chemical Kinetics of Small Organic Radicals, Activation Analysis, and Elemental Analysis by Naclear Methods and Prcconcentration Techniques of Trace Elements. Last year he edited Chetnical Analysis by Nuclear Methods (John Wiley & Sons) and Determination of Trace Elements (VCH). His current research interests include chemical analysis by nuclear methods, radioisotope production and uses, radiation chemistry and.chemica1kinetics of radicals . in solution, and solubility of electrolytes in water-miscible organic-solvents mixture.

Chien Chung, Ph.D., is chairman of the ~ e ~ h e ofn Nuclear t Science, National Tsing Hua University, Taiwan, ROC. He is also the adjunct professor of the National Defense Medical College in Taipei. Dr. Chung obtained his B.Sc. degree in 1972 from the Department of Nuclear Engineering, NTHU and his Ph.D. degree in 1980 from the Department of Chemistry, McGilI University, Montreal, Quebec, Canada. Dr. Chung is also a permanent member of the American Nuclear Society, divisional member of the American Chemical Society, and life-time member of the Chinese Nuclear Society. Dr. Chung served the NTHU as Institutional Director, Research Center Director, and the Dean of the Student Affairs. Before he came to NTHU as Associate Professor in 1983, he served both at the-Brookhaven National Lab as visiting scientist and the University of Maryland as research associate. Dr. Chung was promoted to full professor in 1986 at the age of 36. Dr. Chung has published more than 100 original papers in scientific journals recognized by the Citation Index; he has also presented over 50 invited papers at international conferences to whkh he served as chairman in numerous cases. His current major research interests include nuclear instrumentation, nuclear medical science, health physics, nuclear and radiochemistry, and environmental monitoring -of radiation.

Richard M. Lindstrom Inorganic Analytical Research Division Nationa! Institute of Science and Technology Gaithersburg, Maryland

J. K. Tuli Development Center ational Tsing Hua University Hsinchu, Taiwan

National Nuclear Data Center Brookhaven National Laboratory Upton, New York

Chien Churng

Chushiro Yonezawa

Department of Nuclear Science Natiorlal Tsing Hua University Hsinchu, 'Ijiwan

Eepartrnent of Chemistry Japan Atomic Energy Research Institute 'lbkai-mum, Ibaraki-ken, Japan

Chapter 1

Introduction Zeev B. Alfassi CONTENTS 1. Introduction ................................................................................................................................. 1 11. Neutron Sources .......................................................................................................................... 3 A. Research Nuclear Reactor ................................................................................................. 3 1. Neutrons from Accelerators and Neutron Generators .................................................... 4 2. Radioactive Neutron Sources ..........................................................................................4 111. Gamma-Detectors ....................................................................................................................5 A. Scintillation Detectors NaI(T1) and BGO ............................................................................ 5 B. Solid-state Ionization Detector ............................................................................................ 6 IV. The Shape of the y-Spectrum ................................................................................................. 7 V. Detection Systems .......................................................................................................................10 A. Anticoincidence Compton Suppression ................................................................................ 10 B. Coincidence Double-Escape Counting ................................................................................. 11 References ........................................................................................................................................ 12

f. INTRODUCTION All nuclear techniques of chemical alialysis involve the interaction between an incoming projectile (neutron, charged particle, or y-photon) and a target nucleus. As a consequence of this interaction, there is usually formation of two products-the light one and a heavier product. The light one is mainly yphoton, neutron, or small charged particle. We can write this interaction in the chemical form of reactions: projectile C target nucleus

+ light product + heavy product

or, it can be written in the shorter formalism of nuclear physicists:

target (projectile, light product) heavy product This formal writing does not exclude the possibility that :he reacting target projectile is forming initially a compound nucleus which later disintegrates into the two products. However, it limits the length of time between the interaction and the formation of products. For activation analysis, as long as this time is shorter than 1 p s (in most cases, it can even be 1 s or 1 min), we can treat the final products as our products, neglecting the intermediate steps. The rate of the nuclear interaction, R (number of interactions per unit time), is given by an equation similar to that of a bimolecalar chemical reaction:

where N = the density of the target nuclei (target nuclei per unit volume) I = the intensity of the incident projectiles (number per unit time)

X

-

=

a =

thickness of the target the reaction cross-section (a measure of the reaction probability)

0-84911-5149-9/951$0.00+$.50 0 1995 by CKC Press. Inc.

.

Cross-section has the dimensions of area and, since most cross-sections arc of the order of cm2, the cross-sections are reported in units of barns, defined as: ,"2 .

1 barn = lo-" cm2

Equation 1 shows that R is proportional to N, and hence by measuring R, we can calculate N-the number of target nuclei. This is the basis for all methods of activation analysis. R is measured by measuring the number of newly formed products species, either of the heavy product or that of the light one; and from it, we deduce the number of nuclei of that type present in the sample. Since the reaction is with the nucleus, and the surrounding electrons have no effect, the analysis tells only about the existing atoms, but not their chenlical speciation. It is therefore an elemental analysis. The heavy product can be a stable nuclide or a radioactive one. In cases where the heavy product is unstable, the number of nuclei of the heavy products formed can be measured by following the decay of the heavy product. This measurement is done after the end of the nuclear activation of the target. Consequently, this type of analysis is called delayed analysis. This method of analysis is limited to cases of unstable (either excited or radioactive nuclide) heavy product, with appropriate half-life. If the half-life is too short, either its saturation activity can be too low or it is technically too short for measurement. If the half-life is too long, activation during a reasonable length of time does not produce sufficient activity. Even in cases where the formed heavy product has a half-life of reasonable length, it cannot be used in delayed analysis if it is only a P- emitter due to the difficulties in distinguishing between p- particles emitted from different radionuclides. These problems do not exist in the measurement of the light particles. Another phase of measurement of the number of formed particles is by measurement, during the interaction, of the emitted light product. These light products have to be measured during the experiment and cannot be delayed, since they either disappeared in a very short time (as is the case of y-photons or neutrons) or can be measured only due to their high kinetic energy. If we wait, they will lose their high kinetic energy and can no longer be measured. Their amounts cannot be measured chemically due to the small amount formed (in the case of light product which is a small nucleus as e.g., protons or a-particles). Chemical measurement is not as sensitive as measurement of decaying nuclei or high kinetic energy charged particles. The measurement of a-particles by counting is several orders of magnitude more sensitive than chemical measurement of He gas. The same is true for accelerated protons vs. hydrogen molecules. Since in this case, the measurement of radiation (yphotons or small particles) occur during the nuclear interaction, this form of analysis is referred to as prompt analysis. In the case of prompt y-neutron activation analysis (PGNAA), the projectile is a neutron and the light product ia a y-photon. The reaction is written T(n,y)P, where the T is the target nucleus and P stands for the product. This reaction is called radiative capture of neutrons. There is one exception to the usual prompt y-measurement, in which the y measured is not due to the (n,y) reaction, and this is the case for boron. Boron (like lithium-6) does not react with neutrons by emission of photons, but rather by emission of a-particles, i-e., via the reaction 'OB (n,a) 7Li. However, the lLi formed is only partially (6.5%) in the ground state. Most of the 7Li nuclei (93.5%) are formed in thc excited state, which is decxcilcd with a half-life of 7.3 X 10'" s by emission of 477-keV yrays. Considering our agreement-that everything that occurs within less than 1 p,s is included in the interaction process-we can write this process as 'OB (n,cry) 'ki. The cross-section for this reaction is 0.935 of the conventional 'OB (n,a) 7Li cross-section. In summary, PGNAA offers a nondestructive, relatively rapid method for determination of trace and major elements that cannot be determined by conventional activation methods due to either only Pemission or inconvenient half-life (either too short or too long, including stable nuclides). The energy of neutron capture y-radiation ranges from about 50 keV to about 10 MeV and the spectra of most nuclides are fairly complex. PGNAA has found many uses in research, medicine, and industry. It was reported that in the coal industry in Canada and the U.S., there are 26 commercial PGNAA gauges.' However, this is not limited only to the coal industries but PGNAA gauges are found also in cement and other mineral industries. The main technical factors leading to the success of these gauges in recent years are:2

1. The development of a very stable high count rate y-spectroscopy 2. The availability of large radiation-hardened G e detectors (although most gauges use NaI(T1) detectors)

3, The availability of inexpensive, small but reliable and powerful multichannel analyzers (PC-based) 4. careful study and understanding of the transport processes of neutrons and the associated 7-rays in various bulk media

11. NEUTRON SOURCES ain sources of neutrons are available: (I) research nuclear reactor (nuclear reactors used as n sources); (2) ion and electron accelerators including neutron generators; and (3) radioactive es. Research nuclear reactors have the highest neutron fluxes, but are limited concerning on-site emination, price, and availability. Consequently, nuclear reactors are used predominantly for delayed neutron activation analysis of very minute amounts or for sensitive neutron radiography. However, it is used also for PGNAA. When site irradiation is important, neutron generators or radioactive sources are used.

A. RESEARCH NUCLEAR REACTOR Research nuclear reactors are usually Iarge devices in which fissionable material, almost exclusively is fissioned into two nuclides with simultaneous emission of neutrons that induce further fissions jn a chain reaction. The fission-produced neutrons are very energetic. The cross-section for neutroninduced fission of fissionable nticlides increases with decreasing energy of the neutrons; in order to increase the neutron activity, moderators that slowed down the neutrons are added to the reactor. To reflect back some of the neutrons that leaked from the reactor core, reflectors are used. The fission process releases large amounts of energy, mainly due to the stopping of the recoiling two fissioned and the system is cooled by a coolant (either liquid or gas). The nuclear reactors are categorized according to their fuel, moderator, coolant, reflector, and configuration. Almost all research nuclear reactors (neutron sources) are heterogeneous reactors in which the fuel is in the form of rods. The fuel (natural uranium has or!ly 0.7% *'TJ-the fissile material). Most research reactors have is enriched 235U 93 to 99% W. Many of the reactors have rods that are U-A1 alloys; however, some of the newer designs (mainly those converted to 20% 2'5U) are of the uranium-silicide type. Triga reactors operate with uranium-zirconium hydride fuel that, due to its large negative temperature coefficient of reactivity, allow the operation of the reactor in pulses. In the Light Water Reactor (LWR), ordinary water (H20) is used both as a moderator and as a coolant. The reflector is mainly graphite, but there are also Be or H20 reflected reactors. The construction is either pool type or tank-in-pool type. Due to the relatively high cross-section for capturing thermal neutrons by H atoms, the flux of neutrons in LWRs always contains large fractions of fast and epithermal neutrons. f i e available power is in the range of 10 to 5000 kW,with neutron fluxes of 5 X lOI4 to 1.5 X lot8d m 2 . s. The neutrons are usually divided into three groups according to their energy: (1) thermal energy with most expected energy of kT = 0.025 eV at room temperature; this group is usually extended up to the Cd (thermal neutrons absorber) cutoff of 0.5 eV (2) neutrons with energy between 0.5 eV and 1 MeV are called epithemal neutrons; and (3) neutrons with energies above 1 MeV are called fast neutrons. Many reactors are unique in their design; however, there are some commercial types that are more common-the American TRIGA and the Canadian Slowpoke. The TRIGA reactor is a popular multipurpose research reactor. About 50 of them are operating with power levels of 18 kW to 3 MW (fluxes of 7 X lot5to 3 X lot7d m 2 . s). The most common types are those of 250 kW and 1 MW. They are of the pool type, graphite reflected with uranium-zirconium hydride fuel, and 23SUenrichment of 10 to 70%. The Slowpoke reactor is a low-power (20 k W ) reactor designed specifically as a teaching aid, with additional purposes of activation analysis and production of small amounts of radioisotopes. The system is designed to operate remotely. It can be provided with up to five irradiation sites in the core with a flux of 1016 d m 2 s and five further tubes outside the reflector with half that flux. Heavy-water research reactors are tank types. They usually have enriched uranium fuel, heavy-water moderator and cooled, and heavy-water and graphite reflected. Due to the low cross-section for thermal neutrons, absorption by D and 0 , they are characterized by well-thermalized neutron flux (very little epithermal and fast neutrons fluxes except the core). Due to the lower moderation power of D compxed to H, the physical size of heavy water reactors is larger and hence they have a large available irradiation volume. Their power is usually bstween 10 and 26 MW (fluxes of up to 2 X loi8dm? - s).

2 3 5 ~ ~

a

1. Neutrons from Accelerators and Neutron Generators Charged particle accelerators can produce neutron fluxes by (d,n), (p,n), or (a,n) rcactions. Electrona&$dlerating machines can produce neutron fluxes via (y,n) reactions, where the y-flux is obtained by stopping the electrons in a high-Z material. The main reactions for producing neutron fluxes are D(~,II)~H T(d,n)4He, ~, and 9Be(d,~~)10B. The first two reactions are exothermic and require very little acceleration of the deuteron beams; this is the reason why they are used in the neutron generators, which are very small accelerators with accelerations of 150 to 500 keV. The disadvantage is that the or T2) adsorbed onto metal. If the target is heated, too much of the gas will target material is gas (D2 be desorbed from the metal target; this limits the current of the bombarding deuteron beam and consequently the neutron flux. In the case where a higher flux of neutrons is required, a target of Be is used; however, the deuterons should have an energy of a few meV. The most common use of the reaction T(d,n)4He is due to the low acceleration needed and Lhe higher cross-section of this reaction, as compared to D(d,t~)~He. Different high-voltage generators are used in various neutron generators: Cockroft-Walton, insulating core transformer (ICT), Van de Graaff, and electrostatic rotor machines. A deuteron beam is produced by various ion sources and accelerated to one to few hundred keV. The beam may also be a mixture of 50% tritium and 50% deuterium. Qpical beam intensities are of the order of a few milliamps, but intensities as high as 150 mA have been used. Neutrons of 14 MeV are produced from the interaction of the beam with a large surface-tritiated target. The target consists of a few hundred microns of titanium, zirconium, or palladium evaporated on a 30-mm diameter backing disk and saturated with tritium gas. The disk is water cooled and made of heat-conducting material, silver, or copper. Rotating targets are sometimes used to ensure heat dissipation. Commercially compact sealed-tube neutron generators are also available. The usual neutron generators have fluxes up to 5 X 10" rds; however, special generators with outputs of 5 X lOI3 n s-' were also constructed. 2. Radioactive Neutron Sources Radionuclide neutron sources are composed either of a radionuclide emitting a- or y-rays, together with an appropriate surrounding material, or from a radionuclide decaying by spontaneous fission. The y-emitter in photoneutron sources is surrounded by beryllium or deuterium (such as DzO) and neutrons are emitted due to the (y,n) reaction: y

+ 9 B e - + 2 a + n - 1.67 MeV

y

+ D + H + n - 2.23 MeV

Very few radionuclides with reasonable half-lives emit y-rays with energies that high. Due to this, together with the disadvantagc of the long range of y-rays, the use of that kind of neutron source is very limited. The most c o m m o n l ~used photoneutrons source is '"Sb-Be, which emits neutrons of 2 6 -T 1.5 eV. If4Sb is produced l!&g$actor irradiation of natural antimhy. IZ4Sbemits sevex'd y-rays'f2including 1.692-MeV photons (with intensity of 48%), and has half-life of 60.9 d. The neutron source is composed of two parts, the core (a sphere or a cylinder) nmadc of irradiated antimony metal and a shell of beryllium metal about 2 cm thick. The practical yield is about lo7 n/s per I Ci '"Sb. Most neutron sources use an a-interaction with YBe: 9Be

+ a -+ I2C + n + 5.91 MeV (I2C is in the 4.43-MeV excited stare)

The spectra of these neutrons spread up to 10 to 12 MeV. The common a-emitters used are Ra, 210Po, n9Pu, and 241Am.The main properties of the sources are summarized in Table 1. Table I Properties of Neutron Sources from 9Be(~,n)i2CReaction a emitter +

Neutron yield (n s-' Ci-I) Half-life

Approximate size (cm3Ci-') Heating ( m W Ci-') y-dose (rad h-' Ci-I)

210P~

2.5 X lo6 138.4 d 0.1 32 0.11 (4.43 MeV)

23gP~ 1.7 X lo6 24,360 yr 12 31 0.08 (4.43 MeV)

2.2 X lo6

458 yr 3 33 10 (60 eV), 0.1 (4.43 MeV)

1

fie po-Be and Am-Be sources are prepared by mixing fine powders of beryllium with polonium metal jar americium oxide. The Pu-Be source is an intermetalIic compound, Pu-Bel3. The mixture or -the intermetallic compound is doubly encapsulated, first in an inner capsule of tantalum and then in an outer capsule of stainless steel. In 80% of the (a,n) interactions, the I2Cnucleus is left in the 4.43MeV excited state, which decays with an emission of 4.43 MeV y-photons. The y-dose of these three = . - q ~ ~ ( ~I4C , n sources ) is lower than with the initially used source of Ra-Be due to the y-rays from the Ra source. Lower doses of y-nys are associated with the neutron source of spontaneous fission. The kource of spontaneous fission is "*Cf. .spontaneousfusion ZS2Cf two fission products 3.8n 200 MeV .,

tln=2.65 yr

+

+

The half-life for spontaneous fission is 85.5 years, and for alpha emission is 2.73 years, and the effective half-life is 2.65 years:

The "'Cf sources are very compact. The neutron spectrum is very similar to that of neutron-induced fission, with a mean energy of 2.348 MeV. The neutron yield is 2.3 1 X lo6n/s pg. In order to increase with 23SUthan to use larger amounts of 252Cf. the flux of neutrons, it is cheaper to surround the ZS2Cf These devices are called neutron mulripliers. 252Cf(1 tng) combined with 1.4 g 2)5U (93.4%) and polyethylene as moderator is equivalent to a neutron source containing 33 mg 252Cf.This device has thermal and fast neutron fluxes of 4 X lo8and 6 X 10" n/s . cm2.

-

111. GAMMA-DETECTORS In activation analysis, we need a detection system for the y-photons that will measure the total number of photons together with their energy distributions; i.e., we need to have a y-ray spectrometer. These measurements are done mainly by the use of two solid-phase detectors working o n different phenomena. A. SClNTlLLATlON DETECTORS Na!(TI) AND EGO A scintillation radioactivity detector consists of a scintillator or phosphoc optically coupled to a photomultiplier tube. The most common scintillator for y-ray measurements is a large crystal of NaI activated with 0.1 to 0.2% TI. The y-photon is reacting with the detector, ejecting electrons. These electrons produce excitation or ionization in the scintillator crystal. De-excitation of the scintillator occurs via fluorescence in about 0.2 p s by the TI+ activator (visible light). The small percentage of TI, the "activator", is added to shift the wavelength of the emitted light by the detector to longer wavelength for two reasons: ( I ) IRorder to reduce the self-absorption of the emitted de-excitation light by the NaI crystal; and (2) the shift from UV light to visible light increases the sensitivity of the photomultiplier to the emitted light. The most popular size for routine y-ray spectrum measurements is a 7.5-cm diameter, 7.5-crn high cylinder. It requires approximately 30 eV of energy depositeci in NaI crystal to produce one light photon. The light photon ejects an electron from the photocathode and this electron is accelerated toward the dynodes in the photomultiplier by electric voltage. In each dynode, the electron ejects more electrons (typically 3 to 4), mu!tiplying the electron current. Usually, there are 10 dynodes and each electron ejected from the photocathode is producing about lO%electrons. It takes, on the average, about 10 light photons to release one photoelectron at the photocathode of the multiplier. Thus, it takes approximately 300 eV of y-energy deposited in NaI to release one photoelectron. Since the number of photoelectrons is proportional to the y-energy, the number of final electrons in the pulse, the voltage of the electric pulse, is proportional to the energy of the y-photon. The pulse voltage obtained from the photomultiplier is quite high (50 to 1000 rnV) and only modest further amplification by a pulse amplifier is required. The electric pulses are sorted according to their energies by a multichannel pulse height analyzer, which usually sorts pulses of 0 to 10 V. Each energy range is fitted to one channel. Nowadays, most multichannel analyzers are ADCs on an electronic card fitted to a PC. The PC is used also to analyze t h e obtained spectrum. The spectrum of counts per channel is actually an energy spectrum of the ?-rays.

LOO

Figure 1

1 00

.

1

.

1 ZOO

charnel

ray spectrum of a radioactive sample (a) taken with a 3" x

1

L

1

1600

3" Nal(T1) detector and (b) taken

with a Ge(Li) detector. Another scintillation detector used i s a crystal of bismuth germanate (Bi4Ge3OI2; abbreviated BGO) due to its higher efficiency for absorption of y-rays (since it has a high Z-element). However, BGO crystals are more expensive and are less frequently used.

6. SOLID-STATE IONIZATION DETECTOR The simplest idea for measurement of radioactive decay is by the use of the main property of the emittcd particlcs or photons as ionization radiation. When an ionizing radiation is striking a nonconducting or semiconducting material, it forms in it electrons and holes or cations. The amount of electrons formed is proportional to the energy of the striking photon or charged particle. The amount of energy required to raise an electron from the valence band to the conduction band in a semiconductor is considerably smaller than the same energy in insulators, or the energy required to form an electron-ion pair in the liquid or gaseous phase. Thus, for the same amount of energy absorbed in the detector, more electrons are formed in a semiconductor material than in an insuIator. The larger number of electrons reduces the statistical fluctuation in the number of electrons and hence reduces the width of the radiation signal in the detector (measured as FWHM = full width at half maximum of the peak). Figure 1 shows the peaks obtained from the same radioactive sample by NaI(T1) (a scintillator) and a Ge detector (an ionization detector). The smaller FWHM leads to much higher resolution. The currents formed by the electrons produced in the ionization detector by the radiation is smaller than those formed in the scintillation detector due to the high multiplication factor in the photomultiplier. Hence, before measurement of the current from the ionization detector, it has to be amplified. In order to reduce losses in the pulse current and to reduce the rise time of the pulse, a preamplifier located as close as possible to the ionization detector is used for initial amplification of the detector output signal. Since the capacitance of the solid-state detector depends upon a high-voltage bias, a charge-sensitive preamplifier should be

-.

"7.

used. The voltage pulse produced at the output of the preamplifier is proportional to collected charge :Y-andindependent of detector capacitance. However, the output pulse from the preamplifier is too low ,for sorting by the multichannel analyzer (MCA). Further amplification is done by the main amplifier, serves to shape the pulse. In order to reduce the noise due to the leakage current of the -elecbonsin the conduction band agitated by thermal excitation, the semiconductor crystal is cooled to nitrogen temperature. Of all semiconductor materials. germanium is exclusively used for modem -- 1,y-ray spectrometry since only for Ge and Si can adequately pure material be prepared. Si is suitable only for X-ray measurements, due to its low atomic number, which reduces the interaction cross-section. .. n e impurities that are not tetravalent will produce either almost a free electron in the conduction band or a "hole", leading to increased leakage current (a noise), a current not due to ionizing radiation. In the past, even germanium could not be prepared pure enough to be used as a y-ray detector. However, its impurities could be small enough to be compensated by the lithium ion drifting method developed in 1960. The method was used to effectively compensate p-type (acceptor type containing an excess of trivalent impurities) grown crystals of germanium. The small lithium ions were pulled into the crystals by electric field and high temperature. Due to its very low ionization potential, lithium acts as donor impurities compensating the excess of holes. The lithium ions, having high mobility, were drifted under the influence of the local fields in such a way that the number of lithium donor atoms compensated everywhere in the crystal exactly the acceptors of the original ~naterials.These lithium-drifted detectors are called extrinsic detectors and are written as Ge(Li) (pronounced "jelly"). Lithium ions have high mobility in germanium at room temperature; thus, Ge(Li) detectors must always be kept cooled at liquid nitrogen temperature (7710 immediately after the desired compensation is obtained. In the mid 1370s, advances in germanium purification technology made available high-purity germanium that could be used for y-ray spectrometry detection without lithium drifting. These intrinsic germanium detectors are usually called HPGe (abbreviation for high-purity germanium). The outstanding feature of these detectors is that they do not have to be kept at liquid nitrogen temperature constantly. HPGe detectors must be cooled to 77K only during the measurement in order to reduce the leakage current due to thermal excitation of electrons to the conducting band. HPGe and Ge(Li) detectors are virtually identical from the point of view of measurement. However, due to the more convenient use of HPGe, they completely replace Ge(Li) detectors in contemporary y-ray spectroscopy. The main performance characteristic of a nuclear detector is its resolution, expressed as the full width of the peak at half its maximum (abbreviated RVI-IM). The narrower the peak (lower FWHM), the better is the ability of the detector to separate two close peaks, i.e., better resolution. The Ge detectors feature very high resolution as compared to NaT(Tl), as can be seen from the narrower peaks in Figure 1. At 1332 keV (this is one of the y-rays of T o and is used usuaIly to characterize the FWHM of a detector), the resolution of a good HPGe detector is about 1.8 keV, while the resolution of 3" X 3" NaI(T1) detector is about 60 keV. The resolution is better for a Ge detector; however, the efficiency (the fraction of y-photons hitting the detectors that appear in the photopeak of the spectrum) of Nd(T1) is larger. The efficiency of the Ge detectors is given usually relative to NaI(Tlj. The first Ge(Li) detectors were small and their efficiency relative to NaT(Tl) were less than 10%(for 1332 keV). However, nowadays, HPGe detectors with about 100-mm diameter are constructed and their relative efficiency to Na(T1) 3" X 3" are almost 60% at 1332 keV. The relative efficiency decreases with increasing energy due to the higher Z of iodine compared to germanium. For most applications, the resolution is more important than efficiency and Ge detectors are the more commonly used detectors. A schematic diagram of an HPGe/Ge(Li) detector and the measurement system.is given in Figure 2. ,

%?

-e

IV. THE SHAPE OF THE y-SPECTRUM In order to understand the observed shape of the y-spectrum, it is imperative to know the different processes of interaction of radiation with matter. A photon of light in the IR, visible, or UV range can be absorbed totally or not at all, but it cannot lose only part of its energy. The situation is difFerent for high-energy y-rays. Low-energy X-rays interact with matter mainly by the photoelectric process. In this process, the y-photon interacts with one electron of the material, losing all its energy to this electron. The transferred energy is higher than the binding energy of the electron, and thus the electron is ejected from the atom with kinetic e n e s y equal to this difference (hv+,,<>,,- binding energy of the electron). Since the range of high kinetic energy electrons is considerably shorter than that of y-rays, the ejected electrons will lose their kinetic energy inside the detector material; from the point of view of the

.

Figure 2 A schematic diagram of HPGe or Ge(Li) detection sys-

tem. (1) HPGe or Ge(Li) crystal, (2)end cap, (3)Field Effect Transistor preamplifier, (4) metal housing, (5) vacuum/pressure release, (6) liquid nitrogen filling tube, (7) cold finger to cool the crystal, (8) gas absorbant in order to keep vacuum, (9) liquid nitrogen, (10) vacuum heat insulation. detector, all the photon energy has been absorbed in it, resulting in the same energy absorbed for all photons (assuming that the photons are monoenergetic). The situation is different for higher energy yrays. The cross-section (probability of reaction = rate constant) for the photoelectric process decreases with increasing energy of the pb~ton.For higher energy photons, the main interaction with matter is via Compton scattering. In ge&rdiurn, the photoelectric effect is the dominant process for y-interaction up to about 200 keV; whereas:%&n 200 keV and up. the Compton scattering becomes more important. In the Compton scattering, the photon interacts with what might be called a "free electron" (mainly electrons from outer shells, whereas a photoelectric process occurs with inner shell electrons, mainly from the most inner one, the K shell). The photon is transmitting only part of its energy. The energy that the electron received is more than sufficient to eject it from the atom and it moves with the excess energy as kinetic energy, losing it by collisions in a very short-distance. The photon retains part of its -energy by changing its frequency since E = hv, and is scattered in a different direction. In the Compton process, the energy of the original photon is shared between two particles (the ejected electron and the scattered photon) and, consequently, there is a continuous distribution of energies of the scattered photon. In order to conserve both energy and momentum, the y-photon cannot lose all its energy (in the photoelectric effect the momentum conservation is compensated by the recoiling of the atom from which the electron was ejected, while in Compton scattering it is an interaction with the "free" electron). The maximum energy the photon can lose is given by the expression:

is the electron rest mass. moc2 is the rest mass where E,. is the energy of the initial photon and energy of an electron, which is equal to 0.51 1 MeV. The main difference between the photoelectric and Compton processes with respect to their responses in the detector material is that while in both processes the electrons lose all their energy in the detector, due to their short range, the scattered photon

might escape from the crystal without further interaction, thus leaving in the detector less energy than that of original photon hitting the detector. In y-ray spectrometry, we are measuring the full-energy peak, called the photopeok, as this energy will be the only one to appear if the only interaction process is the photoelectric absorption. The photopeak results either from the y-photons losing all their energy by a photoelectric absorption or by a Compton scattering followed by photoelectric absorption of the scattered lower energy y-photons in the detector (the photoelectric absorption can be after one scattering step or several scattering steps, all of them within the detector crystal). However, the scattered y-photons from the Compton process might escape from the detector crystal, leaving in the detector less energy than the full-energy peak. This escape of the scattered photon not only reduces the photopeak, but its main disadvantage is the formation of a background for lower energy peaks since part of the energy (that given to the electron and some of y photons absorbed ip the detector) is absorbed in the detector. For a monoenergetic source, the Compton scattering produces a conti~iuumbackground ranging from zero up to a maximum, called the Compton edge. This results from the fact that the minimum energy that the scattered photon can have is not zero, but rather is given by the equation:

For E,

>> mc2, E,,,ini,,,,,approaches a value of

*, 2

i.e., 256 keV. Thus, for high-energy y. the Cornpton

edge will be separated from the photopeak by about 256 keV. For a monoenergetic y-emitter radioactive source, the energy range between the photopeak and the Compton edge is almost free of counts. However, most real samples have many different y-energies and the Compton continuum stretches from zero up to the Compton edge of the highest y-energy. If the energy of the measured photon is above 1.022 MeV, the photon can react also in a third process-a process called pair production. In this process, the photon energy is transformed, under the influence of the field of a nucleus, into matter in the form of a pair electron-positron (transformation of energy into matter-antimatter pair). Since the rest mass of either electron or positron is 0.51 1 MeV, the threshold of this reaction is 1.022 MeV to conserve energy. The cross-section for pair production is very low, below 1.5 to 1.6 MeV. The excess energy is shared between the kinetic energies of the electron and the positron. Since the ranges of positrons andelectrons are very short, this excess energy (EL, - 1.022 MeV) will be deposited in the detector. When the positron loses all its kinetic energy, it reacts with an electron in an annihilation reaction to form two y-photons of 0.51 1 MeV each (transformation of matter-antimatter into electromagnetic energy). Each of these 51 1-keV photons can escape from the detector without any interaction, can lose part of its energy by Compton scattering, or can lose all its energy in the detector (either by photoelzctric absorption or by successive Compton scattering and photoelectric absorption). If one of the 51 1-keV photons deposits all its energy in the detector, while the other one totally escapes from the detector, the energy absorbed in the detector will be 5 11 keV less than the full-energy peak. With y-photons of above 1.6 MeV, one usually sees also a peak of E, - 5 11 keV, where E, is the photopeak energy. This peak is called single-escape peak, due to one 5 11keV photon escaping from the detector. Another peak in the spectrum is the double-escape peak, arising from the two 51 1-keV photons escaping from the detector. The energy of this peak is E, - 1.022 MeV. 24Nahas two y-lines at 2.75 and 1.39 MeV. One expects to have four lines in the y-ray spectrum at 2.75, 2.24 (single-escape), 1.73 (double-escape), and 1.39 MeV. Although 1.39 MeV is above the threshold of 1.022 MeV for pair-production, its cross-section for this process is quite low and pairproduction contributes very little to the intkraction of the 1.39 MeV photon with the detector. Hence, in most cases, one will not see the single and double escape peaks of this photon. Figure 3 gives the measured y-ray spectra of 24Na. The ratio of escape peaks to the photcpeak depends on the detector size (influencing the probability of the y-escape) and on the energy of tile y-photon (affecting the chance that the photon will interact by pair production). Many of the y-lines in PGNAA have very high energies, resulting i n the main interaction with the detector being pair-production and hence large single- and double-escape peaks.

Figure 4 Schematic diagram of detection system for pair spectroscopy and Compton's suppression: (1) an HPGe detector with liquid nitrogen cooling, and (2) Nal(TI) detectors with photomultipliers.

B. COINCIDENCE DOUBLE-ESCAPE COUNTING For high-energy y-rays, the cross-section (probability of reaction) of pair-production is higher than that for Compton scattering. Since, the chance of the two 511-keV y-photons produced in the annihilation of the positron to escape from the germanium is quite high, the counts of the double-escape peak are quite high. In early PGNAA, some people used smaller detectors to increase the escape of the two annihilating quanta, obtaining mainly only double-escape peaks. This choice decreases the efficiency of the photopeaks and single-escape peaks, leading to a simpler spectra in the high-energy range. However, the choice of a smaller Ge detector also reduced considerably the efficiency'for pair-production, leading to a smaller peak. A better snlution to obtain an only double-escape spectrum is the use of triple coincidence or 1.022-MeV coincidence counting. In triple coincidence, the large Ge detector is surrounded by two (or sometimes four) NaI or BGO detectors. These scintillation detectors are connected to single-channel analyzers, al!owing an output signal only if it due to 511-keV y-rays. These singlechannel analyzers are connected to a coincidence unit, together with the Ge detector. Only when two simultaneous pulses come from two different scintillators does the coincidence unit'allow the recording of the signaI from the Ge detector. The two simultaneous (within a fixed time range) 51 1-keV photons indicate that the event occurring in the Ce detector is a pair-production process, rejecting all Compton scattering events together with pair production events where one or two 5 11-keV photons were absorbed in the detector. Figure 4 gives a schematic description of a Ge detector surrounded with scintillators. The same system is used both for coincidence double-escape counting and Compton suppression anticoincidence counting, as can be seen in the electronic scheme in Figure 5. The system measured also the spectra of singles (photopeaks). Thus, using different ADCs (usually connected to the same PC), the three different spectra are measured simultaneously. In some systems, Lhc scvernl surrounding scintillators arc rcpl~lcctlby onc nllnr~ltlrscinlillntor ~rnd the triple coincidence is replaced by double coincidence, in which the absorption of 1.022 MeV in the annular detector is used to open the gate for recording the signals from the germanium detector. It is worse than the triple coincidence since the 1.022 MeV must not be due to two 0.51 1-MeV photons.

ADC- l 1

Ge C

Singles

1 Anti

stan

TAC - SCA A stop

b

Coicidence Unit

-

ADC-2 Compto~~

Suppression

Scintillation

Specuum

-

detector

-+ -

f

Coinadence Unit

ADC-3 L.

double

Figure 5 The electronic scheme for simultaneous measuring of the three different spectra: CFDConstant f;action discriminator used as a single-channel analyzer; OR-A logical unit for which one pulse input is sufficient to get output; TAG-SCA-Time-to-Amplitude Converter plus single-channel analyzer to limit the time interval between the two pulses. The anticoincidence unit is usually part of the ADC and not a separate unit. The detector boxes inciude their high voltage supplies, amplifier, and preamplifiers.

In some system$ the annul dence me:tsurements.

tillator is split into optically isolatdh secton to allow triple cainci$4

1. Sowerby, B. D., On-line nuclear tecllniques in the coal industry, Nucl. Geoplrys., 5. 491, 199 1. 2. Gozani, T., Physics of recent applicatior~sfor on line analysis of bulk minerals, in Capture Gomnm-Ruy Spectmscopy and R e l a r d Topics-1 984, Raman, S., Ed., kmeric:tn Institute of Physics, New Yorl:, 1985,82G.

Chapter 2

instruments and Shielding Chien Chung CONTENTS I. Introduction ................................................................................................................................. 13 11. .Prompt y -Ray Spectroscopy ....................................................................................................... 14 A. Semiconducting Detectors .......................................~............................................................ 15 B. Scintillation Detectors ...........................................................................................................16 111. Advanced Instrumentation ...................... . ................................................................................. 17 A. Anti-Compton and Pair Spectrometers ................................................................................19 B. Field Instruments and Spectroscopy .................................................................................... 23 1V. Shielding of the PGAA Facility .................................... ............................................................. 25 A. Requirements for Shielding .................................................................................................. 26 B. Biological and Detector Shields ........................................................................................... 29 V. Discussion ................................................................................................................................... 35 ............................................................................................................... 35 References ........................... .

As mentioned earlier in this book, the PGAA method demands that neutrons bombard the target and utilizes a spectrometric detector to count the emitting prompt y-rays for quantitative analysis; therefore, a y-ray spectrometer and radiation shields, together with the neutron beam and sample handling device, are the major parts of a PGAA facility. Since the PGAA facility is usually attached to a radiation control zone such as a nuclear research reactor, the researchers are inevitably required to perform the on-site sample changing and troubleshooting while the intense neutron beam as well as scattered prompt yrays are all around. Hence, pcrsonal radialion safety must bc observcd for PGAA operation. A y-ray spectrometric detecting system contains complex electronics in addition to the y-ray detector. These electronic signal processing units, and associated computer with analyzing software, should be placed in a counting room as part of the PGAA facility and preferably have temperature and humidity control. Since hundreds of prompt y-rays appear in the multichannel spectrum and many of them are overlapping one another, software with automatic data reduction and sophisticated photopeak analyzing ability are mandatory. Many software packages with a prompt y-ray library are commercially available for upgrading the current PGAA facility. A prominent part of the prompt y-ray spectrum is the Compton background. The background of the Compton continuum not only obscures the identification of minor photopeaks, but also increases the uncertainty of the position and intensity of observable photopeaks. Reduction of this Compton continuum can be achieved using the Compton-suppressed spectrometer with detector shield. Several types of detector shields have been used as anti-Compton annuli: sodium iodide detectors, plastic scintillators, and more recently bismuth germanate detectors, resulting in great improvement of the spectral analysis. Numerous reactor-based PGAA facilities have now adopted such systems to improve their spectrometric performance. The y-rays and, in particular, neutrons are ionization radiations, causing human tissue damage and organ dysfunction; prolonged exposure to radiation, in particular to the high-energy neutrons and yrays, may lead to instant injury or even fatality. An upper limit of accumulated radiation dose of 50 milli-Sivert (mSv, a radiation dose equivalent unit) each year is recommended for radiation workers.' A 1-mSvh dose rate is approximately equal to an exposure to thermal neutron flux of 26,000 nls . cm2 or to fast neutron flux typically of 670 n/s cm2, or to 8 MeV high-energy y-flux of 12,000 photonst cm2 s. In the previous chapter, a thermal neutron flux on the order of 107n cm-*s-' from nuclear reactor is delivered to the samples: m5ilental exposure to these neutrons for 8 min can cause the researchers to exceed the annual dose allowance, facing the risk of rad~ationinjuries. Therefore, in the

-

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HPGe PREAMP DETECTOR

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Figure 1 Basic electronic flow chart of (A) scintillation and (R) semiconducting detector systems used for PGAA facility.

PGAA facility, maximum effort shall be put to shield off the radiation and facility itself for biological protection.

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II. PROMPT y-HAY SPECTROSCOPY

'

'

To detect and analyze the prompt y-ray in PGAA experiments, a photon tlcrector is required. 'Thc choice of a particular detector type for PGAA application depends upon the y-cnergy rangc of Iitterest and the application's resolution apd efficiency requirements. 'i%e detector must have s:~Ificientnlateriai to absorb a large f~actionof tfiehy-ray energy. Thus, any gas-filled counter is not suitable for de~ecting full-energy y-rays because the probability of absorbing all the y-ray energy is too low. Further, the higher y-ray energies, as frequently encountered in PGAA, are more effectively absorbed by highcr Z materials. Other considerations are count rate capability, resolution, and if timing applications are involved, pulse rise time. The kinds of detectors commonly used in PGAA can be categorized as scintillation detectors and semiconducting detectors. Scintillation detectors are used in conjunction with a photomultiplier tube (PMT) to convert the scintillation light pulse into an electric pulse. Solid crystal scintillators such as sodium iodide or NnI(TI) are commonly used in PGAA. Semiconducting detectors, made from single crystals of very pure germanium, are the highest performance detector type. The superior resolution of these detectors has revolutionized data gathering for prompt y-ray measurements. Mmy configurations of scintillation detectors are cornmcrcially available, ranging from small bismuth germanate (BGC)) to large NaI(T1) with multiple phototubes, a typical electronic flow chart of a scintillating detecting system is shown in Figure 1A. On the other hand, the semiconducting detector, which exhibits the highest resolution of all y-detectors, virtually requires the use of a multichannel analyzing system for spectrum analysis. A typical semiconducting y-ray spectroscopic syslem consists of the germanium detector, high-voltage power supply, preamplifier (PA), amplifier, analog-to-digital converter (ADC), and multichannel analyzer (MCA), as illustrated in Figure lB. Details of these photon detection systems used for PGAA experiments are given below.

Table 1 Photopeak Energy Resolution for Various Detectors Used in PGAA Experiment Resolution: FWHM (keV) at Detector type and size (Diameter x Height)

HPGe (2" X 2.3") NaI (2" X 2") BGO (2" x 2")

122 keV

662 keV

1332 keV

1.2

1.8 75 125

0.8 23

53

40

85

10829 keV

6.0 214 450

A. SEMICONDUCTING DETECTQRS A semiconductor is a material that can act as an insulator or as a conductor. It is fabricated from either elemental or compound single-crystal materials having a band gap in the range of approximately 1 to 5 eV. The Group IV elements, in particular germanium, are by far the most widely used semiconductors. Semiconducting detectors have a P-I-N diode structure in which the intrinsic (I) region is created by depletion of charge camers when a reverse bias is applied across the diode. When photons interact within the depletion region, charge carriers (holes and electrons) are freed and swept to their respective collecting electrode by the electric field. The resultant charge is integrated by a charge-sensitive PA and converted to a voltage pulse with an amplitude proportional to the original photon energy. The band gap signifies the temperature sensitivity of the materials and the practical ways in which these materials can be used as detectors. As a practical matter, high-purity germanium (HPGe) detectors must be cooled in order to reduce the thermal noise to an acceptable level. The most common medium for detector cooling is liquid nitrogen (LN,); however, recent advances in electrical cooling systems have made an electrically refrigerated cryostat a viable alternative for many field applications. In LN2cooled detectors, the detector element is housed in a clean vacuum chamber attached to or inserted in a LN2 dewar. The detector is in thermal contact with the liquid nitrogen, which cools it to around 77K. At this temperature, reverse leakage currents are low enough to be ignored. Performance of any photon detection system can be rated by its resolution of spectral photopeak and efficiency to detect the full-energy y-ray. Semiconducting detectors provide greatly improved energy resolution over other types of radiation detectors for many reasons. Fundamentally, the resolution advantage can be attributed to the small amount of energy required to produce a charge carrier and the consequent large "output signal" relative to other detector types for the same incident photon energy. At 3 eV per electron-hole pair, the number of charge carriers produced in Ge is about 1 and 2 orders of magnitude higher than in gas counters and scintillation detectors, respectively. The charge multiplication that takes place in the PMT associated with scintillation detectors, resulting in large signals, does nothing to improve the f~nd~amental statistics of charge production. The photopeak energy resolution, in terms of full-width-at-half-maximum (FWHM) for several prompt y-ray energies, are listed in Table 1 for various types of detectors for comparison. The energy resolution of the y-ray spectrum taken from an HPGe detector is explicitly superior to those taken from scintillating detectors with similar siye, with reduction factors on FWWM ranging from 30 to 75 at various photon energies. In Figure 2A, the spectrum of the prompt y-rays emitted from the Dy sample in the PGAA facility attached to the I-MW Tsing Hua Open-pool Reactor (THOR) in Taiwan is illustrated; the neighboring photopedcs are clearIy identifiable using a high-resolution HPGe detector, while the multiplet photopeaks, such as the 466-keV of Dy and 472-keV promptay-rays from Na of the NaI(T1) shield, are easily resolvable. On the other hand, the detector efficiency to absorb the full energy of the incoming prompt y-ray is closely related to its atomic number and density of the detector. At low energies, detector efficiency is a function of cross-sectional area and window thickness while at high energies, total active detector volume more or less determines counting efficiency. Coaxial HPGe detectors are specified in terms of their relative full-energy peak efficiency compared to that of a 3-in diameter by 3-in long (3" X 3") NaI(T1) scintillation detector at a detector-to-source distance of 25 cm. Detectors of greater than 100% relative efficiency have been fabricated from germanium crystals ranging up to about 75 mm in diameter. Approximately 2 kg germanium are required for such a detector. Typical efficiencies, in terms of counts per photon emitted from the source is shown in Figure 3, where the efticiency curves for the MPGe detector, as well as those for 13GO and NaI(T1) scintillation detectors with similar detector size, are

also displayed.

CHANNEL No., keV Figure 2 Excerpt of (A) normal and (6)Compton-suppressed prompt y-ray spectra measured by a highresolution HPGe detector attached to the P G W H O R facility.

8. SCINTILLATION DETECTORS The y-ray interacting with a scintillator produces a pulse of light that is converted to an electric pulse by a PMT. The PMT consists of a photocathode, a focusing electrode, and 10 or more dynodes that multiply the number of electrons striking them several times each. The anode and dynodes are biased by a chain of resistors typically located in a plug-on tube base assembly. Complete assemblies including scintillator and PMT are commercially available. The properties of a scintillator material required for good detectors dre transparency, availability in large size, and large light output proportional to y-ray energy. Relatively few materials have such good properties for detectors. Thallium-activated NaI and BGO crystals are commonly used. The high Z of iodine in NaI(T1) gives good efficiency for y-ray detection. A small amount of TI is added in order to activate the crystal, so that the designation is NaI(T1) for the crystal. The best resolution achievable is about 8.0% for the 662-keV y-ray from I3'Cs for a 2" X 2" crystal, and is slightly better for smaller sizes. Although the spectral rc:sohtion of the NaI(T1) detector is inferior to that of the HPGe detector, as indicated in Table 1, its detecting efficiency, particularly in the high-energy region, is several times better than that of the high-resolution HPGe detector. The light decay time constant in NaI(T1) is about 0.25 ks. Typical charge-sensitive PAS translate this into an output pulse rise time of about 0.5 p-s. The NaI(T1) detector is the dominant material for y-detection, used as either spectrometer or detector

-

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Figure 3 Absolute efficiency curves of the prompt y-ray detecting systems using the BGO, Nal, and HPGe

detectors with size indicated; the curve for the pair spectrometer using the HPGe and a 9 X 1 0 Nal(TI) shield is also plotted.

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shield for Compton-suppressed and pair spectroscopic systems; this is due to its relatively economical cost in terms of the benefits for both spectral resolution and detecting efficiency. The bismuth germanate (Bi4Ge3O,J detector, commonly abbreviated as BGO, became commercially available in the late 1970s. It has been utilized as a y-ray counter or spectrometer in a rapidly growing number of applications. A major advantage of BGO is its high density (7.13 g/cm3) and even larger effective atomic number, resulting in a high detecting efficiency for high-energy y-rays. Despite its lower Iight output and poor spectroscopic resolution relative to the more popular NaI(T1) scintillator, the BGO detector has been widely applied to either the main detector or the guard scintillator in the anti-Compton spe~trometer.~ For y-ray energy above a few MeV in the medium-energy region, observation using NaI(T1) and HPGe detectors suffered from inferior detecting efficiency and insufficient sensitivity for high-energy photons. The detecting efficiency of the BGO detector is at least 10 times higher than that of the Nal(T1) scintillator with identical detector size for photon energy above 5 MeV, as explicitly observable in Figure 3, further encouraging its application in high-energy y-ray experiments. In Figure 4, the prompt y-ray spectra taken by the BGO, NaI(TI), and HPGe detectors with similar size in an in vivo PGAA scan using the 0.1-W Tsing Hua Mobile Educational Reactor (THMER) in Taiwan are shown.3 Although the high-resolution HPGe detector yields excellent quality of the prompt photon spectrum, in particular for the 10.83-MeV high-energy prompt y-ray emitted from the N(n,r) reaction in the sample, it takes, however, a long counting period of 50,000 s to collect enough counts. On the other hand, only 1800 s is needed by use of the BGO detector to acquire an identifiable 10.83MeV prompt photopeak due to its superior counting efficiency-almost 100 times higher than that of the HPGe detector, as illustrated in Figure 3. With the same counting period of 1800 s, the NaI(T1) detector is impractical for on-line measurement of nitrogen prompt y-rays because the accidental sum in the 9- to 11-MeV region, as shown in Figure 48,obscures any qualitative andquantitative identification of the 10.83-MeV prompt photopeak. Hence, the scintillating detectors, in particular the NaI(T1) crystal for medium-energy and BGO for high-energy prompt y-rays, are useful in a PGAA experiment provided that the counting efficiency on sample is preferentially demanded.

! I ! . ADVANCED INSTRUMENTATION y-Ray spectroscopy using eithzr high-rcsoltttion serniconcluc~ingdetectors or high-efficiency scintillating detectors has been widely app!ieri in PCA A experiments. Since a high-quality prompt spectrum with

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Figure 4 Prompt y-ray spectra taken by the (A) BGO, (B) Nal(TI), and (C) HPGe detector in IVPG4A scan using THMER facility. (Reprinted with permission from Appl. Radiat. Isot. 547,42(6), Lea, C.J., Chao. J. H., and Chung, C., High energy gamma ray spectrometer using BGO detector. Copyright 1991; and Appl. Radiat. /sot 44(6), Chung, C., Wei, Y Y., and Chen, Y. Y, Determination of whole-body nitrogen and radiation assessment using IVPGAA techpique, 941, Copyright ?993,Pergarnun Journak Ud.)

minimal neutron damage to the detector is preferable on the one hand, and a complex fail-save alarming device is required to link up the real-time analytical result on the other, advanced instrumentation in pGAA experiment is in great demand. In this section, advanced prompt y-ray spectroscopy such as anti-Compton and pair spectrometry, as well as those used in field applications, are briefly introduced below and their distinguishing features are discussed.

A. ANTI-COMPTON AND PAlR SPECTROMETERS A prominent part of the prompt y-ray spectrum is the Compton background arising from the incomplete deposit of original photon energy owing to the escape of incoherently Compton-scattered y-rays from the detector. The background of the Compton continuum is not only obscuring the identification of minor photopeaks, but is increasing the uncertainty of the position and intensity of observable photopeaks. Reduction of this Compton continuum can be achieved by surrounding the HPGe detector with a large detector shield that is used in an anticoincidence mode. Scattered photons escaping from the HPGe detector and leaving part of the original photon energy behind are eventually captured by the surrounding detector shield. When signals generated from both detectors occur simultaneously, storage of the HPGe event is blocked while full-energy photons are stored; thus, the Compton continuum is suppressed accordingly. In the past 30 years, several types of detector shields have been used as anti-Compton sinnuli: NaI(T1) detectors, plastic scintillators, and more recently.BG0 detectors. These have resulted in great improvements of the spectral analysis. In Table 2, some representative anti-Compton prompt y-ray spectrometers using NaI(T1) shields are listed." The suppression factor of the anti-Compton spectrometer, usually quoted as the major feature of a Compton-suppressed y-ray spectrum, is the ratio of the height of the Compton continuum without suppression to the height with suppression; in Table 2, the Compton suppression factor for low-energy y-rays is around 5 to 8 and greatly improves the quality of the lowenergy part of the y-ray spectrum. The anti-Comptodpair spectrometer used at the P G M H O R facility is described here as a typical advanced instrument for PGAA expe~iments.~ The main y-ray detectors are an HPGe detector and an annular NaI(TI) detector shield to perform Compton suppression and a pair of 51 I-keV y-ray detectors. The 145-cm3 n-type HPGe detector has 30% relative efficiency, a resolution of 2-keV FWHM at 1332 keV, and a peak-to-Compton ratio of 46: 1 at the said photope'ak energy. The detector shield surrounding the primary HPGe detector is an annulus 9" X 10" NaI(T1) detector. A 3"-diameter lateral hole can

Table 2 Dimensions a n d Performance of Representative Anti-Compton Prompt y-ray Spectrometers Using Nal(T1) As a Shield Dimensions of Mal(TI)

vpe

Efficiencya

Diarnaier (cm)

Thickness (cm)

Well size (cm3)

Compton suppression factor" Well typeb

%o

13'Cs

Ge (Li) Ge (Li) Ge (Li) HPGe HPGe Ge (Li) Ge (Li) Ge (Li) Ge (Li) Ge (Li) Ge (Li) HPGe Efficiency relative to that of 3" X 3" Nd(T1) detector. L = asymmetric arrangement of L-shaped well; T = symmetric arrangement of through-side-hole. 'Compton suppression factor taken around 80% of the full-energypeak just below the Compton edge of the photopeak. (Reprinted with permission from Nrtcl. Inslr. and Meth. A243, Chung, C. and Yuan, L.J., Performance of a HPGeNdII(T1) Campton suppression spectrometer in high-level radioenvironmerllal studies, 102, Copyright 1986, NorthHolland Physics Publ~shing.) a

accommodate the 38-cm long HPGe detector can. Eight 3" PMTs, four on each optically isolated half, view the NaI(T1) detector. The sample ;o; PGAA iri-adiaiion can be placed as close as 25 cm to the HPGe detector head if the interference from scattering neutrons is less intensive. The electronic block diagram h e d for this spectrometer is shown in Figure 5. The basic electronic set-up is for the two optional data accumulation modes. In the Compton-suppressed mode, y-ray events collected in thc HPGc detcctor (which has coincidcnt events in annular NaI(T1) detcctor within 0.1 ps) are rejected at ADCl input. For pair simuttaneously in both halves of the NaI(T1) crystal is treated as an allowable gate for an HPGe pulse at ADC2. Digitized signals for either mode and single spectrum are further processed in data acquisition of a multiparaineter system in which four 409b or two 8192multichannel spectra, each with different effects, can be obtained simultaneously. The useful detecting y-ray range of the data accumulation modes covers the energy from 0.1 to 11 MeV: the Compton suppression data are accumulated within the range of 0.1 to 4 MeV, while the pair spectra are collected from 3 to 11 MeV. An off-line spectrum analytical code based on the software package of modified SAMPO is available. The collected spectra can be transferred to either a PDP 11/34 computer or a personal computer for further analysis. The effect of Compton suppression, judging at 662 keV of the '37Cspho:ope&s peak-to-Compton ratio, is improved by a factor of 5.2 as indicated in Table 2; at 2754 keV of the "Na photopeak, the peak-to-Compton ratio is further improved by a factor of 7 under anti-Cornpton operation. The improvement effect of such Compton suppression is shown in Figure 6. At complex radiocnvironmenta1 tcst. the effect of Compton suppression not only yields a much better ratio of the Compton continuun~ between unsuppressed and suppressed spectra, but also eliminates most contaminating and interfering y-rays. The effect of Compton suppression in a PGAA experiment is demonstrated more dramatically in Figurc 7, which shows y-ray spectra taken duriilg the full-power operation of the THOR facility. In additiun to the background photopeaks from natural radioriuclidcs and isotope productic,rl, ihe experimen- . tal reactor floor contains scattering neutrons originating from the PGAA facility itself. Thc anti-Compton operatic?. yields a strikingly'good ratio of the Compton continuum among normal, unsuppressed, and suppressed spectid (1 16:11:1). The reductions of the continuum background in y-ray spectra for such high-level counting were evaluated at varioas photon energy ranges and the results are shown in Figure 8. For high-level counting, the average reduction of background is improved by a factor o r 9 lo 12 on all energy ranges, as indicat4,Jn the figure. The results indicate that th%&tein is adequately sensitive to many $menis of inte&. The detectin$: sensitivity can be divided into kroups. The most sensitive elements ire B, Gd, Cd, Srn, and He. f$<* which the detection limits based upon a 1-g sample for 20-h irradiation are in sub-ppm range. A typical anti-Compton y-my spectrum to investigate a Gd sarnpIe in tfle PGAA experiment using the THOK facility is illustrated in Figure 2B. As numerous prompt y-rays used for PGAA identification have plmaton cnergics nlmvc 3 MeV, a different approach to simplifying thc rccorded high-resolution spectrum is to sclecL only thc doubleescape peak. If the y-ray energy is sufficiently high, a significant fraction of all interactions will correspond to pair-production in which both photons produced by positron annihilation escape from the primary detector. Inasmuch as these annihilation photons are always emitted in opposite directions, the dctector shield shown in Figure 5 can intercept them with a reawnable c11a-m. IC coincidence is demarded among all three detectors, the selection of double-escape events will be highly srccific afid most of the continuum will be suppressed. A considerable saci-ifice in counting efficiency must always be made, but the isolation of the double-escape peak and suppression of utller backgrouiids can be very effective. The counting efficiency of this set-up is shown in Figure 3. At a medium energy of 3 MeV, the counting efficiency of a pair spectrometer is 10 times lower than that in normal operation; however, as prompt y-ray energy increases, the efficiency difference of the two modes reduces and the doubleescape peaks dominate the high-energy part of the pair spectrum. Both Compton-suppressed and pair spectrometers are a complementary part of the normal prompt y-ray spectrum; !heir effectiveness can be enhanced by assigning thc energy range of each mode to gain as much spectroscopic information as possible. A typical energy range for each collecting mode, as mentioned in the previous example, is 0.1 to 11 MeV for a normal spectrum, 0.1 to 4 MeV for a Compton-suppressed spectrum, and 3 to 11 MeV for a pair spectrum to cover all prompt y-rays in the best possible mar.[ ~er.

'

TIMING AMPL I F I E R O R T E C 4 60

DETECTOR

O R T E C 4 ZOA

SUPPLY

n-TYPE Ge D E T E C T O R

'

REAMPLIFIER C 4-r d

SUPPLY

C

J

Na 1 DETECTOR

DELAY L I N E AFIPI.IF1F.R O R T E C 4 60

PREAMPL I F I El i

A

-

i

,

TINING SCA ORTEC4 20A

"

FAST C O l N C l OENCE ORTEC414A

LJ

,

Figure 5 Block diagram of the electronic and logic nuclear instrument modules used with Compton-suppressed and pa

Figure 7 High-level radioenvironmental measurements by HPGe detector during full-power operation of the THOR facility: (A) no shielding, normal spectrum with high count rate of 2700 countsls; (8) shielded with Nal(TI) detector, unsuppressed spectrum with count rate of 153 counts/s; and (C) Comptonsuppressed spectrum with count rate of 33 countsls. Counting period for each spectrum is 2000 s. (Reprinted with permission from Nucl. Instr. and Meth. A243, Chung, C. and Yuan L. J., Performance of a HPGe-Nal(TI) Compton suppression spectrometer in high-level radioenvironmental studies, 102, Copyright 1986, North-Holland Physics Publishing.)

5. FlELD INSTRUMENTS AND SPECTROSCOPY As the PGAA is similar to the conventional INAA in that neutrons are utilized as projectiles and multielemental concentration can be analyzed in a nondestructive way, the flexibility of PGAA, however, has some advantages over the INAA since it requires only a nominal neutron flux, at least lod6lower than that utilized in INAA, and analytical results are immediately available during the irradiation without long waiting for the radioactive decay. Field applications using either fixed or mobile PGAA units have been reported. Examples of these field investigations are in vivo (IVPGAA) medical diagnosis using isotopic neutron source6 or a mobile reactor;' in situ (ISPGAA) geological and environmental surveys using a portable neutron generator" or isotopic s o ~ r c e and ; ~ on-line (OLPGAA) interrogation using isotopic neutron source." The instruments and spectroscopic systems used in field applications are preferably shock and vibration resistant, rugged proved, and stable at various temperatures, pressures, and humidities. Commercial units for PGAA field application further demand easy handling with minimal maintenance, user-friendly operation procedures, and maximum protection from the source of radiation, in particular the neutrons. Among numerous methods to acquire the prompt y-ray spectrum in field applications, they nll demand a rapid, quantitative determination of elemental concentration in the sample with proper analytical results as real-time output. 1-Iere, the field instrument of OLI'GAA used for explosive detection in airport security inspection is briefly described as an example."

W JG H-LEVEL CGUII T I N G

GAMMA RAY ENERGY ( k e V )

-

-

1

Figure 8 Background reduction in the anti-Compton mode for spectra shown in Figure 7. (Reprinted with permission from Nucl. Instr. and Meth. 8243, Chung, C. and Yuan, L. J., Performance of a HPGeNal(TI) Compton suppression spectromster in high-level radioenvironmental studies, 102, Copyright 1986, North-Holland Physics Publishing.)

A weight r&io up to 82% nitrogen content is found in comnonly used expln~ivcs,'~ and an excited nitrogen nuclide emits 10.83-MeV prompt y-rays when a neutron interacts wilh explosive hiddan in r rogc:ll the inspecting luggage. By counting such characteristic proinpt photopeak areas, the amount of nGt inside the luggage can be revealed and the weight of the possible explosive can be leadily estimated. The BGO deteclcr is the best choice for such applications as it can work under an intense neutron dose. This detector has high photon detection e%ciency at high photon energy; in addition, it has been proved to outperform the NaI detector in various high-energy photon detection applications. Iil order to identify explosive-like materials readily, an array detector, consisting of ten 2" X 2" BGO detectors coupled to associated electronics is used. The block diagram of the counting system is shown in Figure 9. Once the proper gain of each BGO detector (derived from the attached CANBERRA 2007 PA) has been reached, the specific energy photon in the energy window can be registered to a summed spectrum. Unfortunately, this gain level is I~ard, if not impossible, to reach, since the energy resolution of the BGO is poor. In order to solve the probiern, CANBERRA 2015A amplifierltiming single-channel analyzers (TSCA) coupled to a home-made sund mixer are chosen. The low-level (E) discriminator and energy window (AE) for each TSCA can be adjusted to the proper energy range to detect the 10.83-MeV full-energy and single-escape peaks. That is to say, a specific photon energy falling in the adjusted energy range, E and E + AE, may trigger a logic output to the sum/mixer, which is subsequently added together to fit the CANBERRA 2071A dual counterltimer for count rate assessmect. In order to check the performance of each BGO detector output, the PA linear signal output can be connected to the CANBERRA-85 MCA system to visually evaluate the y-ray spectral quality and spectral background. Both MCA and cocn?erltimer output were linked to a personal computer for background subtraction while a relay from the computer may be set up to trigger a preset alarm. If the count rate in either the energy spectrum or the counter is over the preset alarm level (equivalent to a significant amount of nitrogen hidden in the luggage), the alarm may be set off immediately. Current airport security inspection of check-in or carry-on luggage allows 5 to 10 s to scan each item; the explosive detection system, wlth a count rate of 40 cpslkg explosive, has a detection limit for a typical HMX chemical explosive arolml 0.5 kg, about the fatal amount that rnay destroy a

I:lslyl I J + T ~ ~I ~ M PA DGO 2

TSCA

TIMER TSCA

Figure 9 Electronic block diagram of the prompt y-ray counting system designed for explosive detection in airport security inspection. commercial airplane in mid-air if planted pr~perly.'~ There are already several explosive detection devices based on the OLPGAA technique installed in some international airports for security inspection. Alternatively, the PGAA instruments for field applications can be very compact and simple without sacrificing the data collection requirements. Here, a simple PGAA instrument using a high-resolution HPGe detector, designed for in siru water pollutant survey, is described.I4 The instruments used for such purposes contain a prompt y-ray detecting system jn addition to the submerged neutron source and associated devices. In ISPGAA operation, a neutron-resisting and high-resolution y-ray spectrometer is preferred, dictating the use of the HPGe detector. A commercially available portable HPGe detector with 10% relative efficiency is used to detect the prompt y-ray emitted from the interaction of neutrons with water contents. The system resolution FWHM is 3.9 keV at 2223 keV. The signal of prompt yrays collected by the HPGe detector is transmitted through a built-in PA, then fed into a portable, DCpowered CANBERRA MCA 10-plus MCA system containing the HVPS, amplifier, ADC, and MCA modules. The MCA system, together with the portable personal computer for data handling, can be placed on-board the survey motorboat. Two systems of this ISPGAA operation have been developed, the backscattering and transmission systems, as illustrated in Figure 10. In the backscattering system, shielding material is enclosed in an aluminum probe and the 252Cfneutron source, together with an HPGe detector, are positioned on opposite sides of a lead block. The lead, and the neutron shield surrounding the detector, reduce the intensity of the fission neutrons prompt y-rays directly from the neutron source at the detector crystal, therefore suppressing the interfering y-rays in the spectra collected. The signal and power cables, passing through a waterproof cable outlet on top of the probe, are fixed to the steel cable that is then attached to the cable wench at the stem of the motorboat. On the other hand, the neutron source in the transmission detection system is submerged in water, whereas the HPGe detector is placed above the water line; the water in between serves as both sample being analyzed and y-raylneutron shield.-A higher intensity neutron source is necessary to match the detection sensitivity achieved using the backscattering system. The detection sensitivity of the PGAA systems can be evaluated using the chlorine prompt y-ray peaks. Figure 11 shows the chlorine-capture prompt y-ray spectra collected in the salt water using both backscattering and transmission systems. Since chlorine generates a number of intense prompt capture y-rays over a wide energy range up to 9 MeV, it is ideally suitable for the evaluation of the PGAA performance for in situ application.15 Both geometric arrangements yield a high-resolution prompt yray spectrum, and numerous elements emitting prompt y-rays with energies above 4 MeV can be identified down to parts per million with these systems.16

For radiation safety and health concerns, as well as the reduction of neutron damage on sophisticated components such as spectrometric detectors, neutrons should be collimated and shielded properly such

I I

: I I

,

I

Operating

Figure 10 ISPGAA probe arrangement for (A) back-scattering and (6) transmission counting; here, positions of HPGe prompt y-ray spectrometer (a) and neutron sourco (b) are indicated.

that only the sample is exposed to thc mtense bombarding neutrons. There are two necessary confinements for neuiians in t!re PGAA facility: a coilimation of neutrons prior to the bornb~cln~ent on the sarnplc and the shielding of scattered neutrons afterward. A neutson collimator is a devicc s w h that neutrons from one end are allowed to pass through a channel and come out at the other end in a well-defined direction toward the irradiation position. Neutrons striking the material other than tl~chollow channcl will be absorbed eventually. 4 ne.utron shield is a device whercin neutrons can be absorbed when they impinge on it and generate rn$$nal prompt y r a y s as well as inducedi radioactivity. The design of a neutron s%c& is performed f i s t by setting up th4$naximurn permissible radiation. dose rate at various locations o r the PGAA facility. This implies a knowledge of the physicai layout of the PGAA facility and its associated equipment. The neutron flux intensity at various locations is required in order to determine the source distribution. With the knowledge of the PGAA facility available, a preliminary choice can be made of collimator and shield layout, so that neutron attenuation and transport calculation can be carried out. In addition to the scattering neutrons, y-rays originating within the neutron source itself, as well as prompt and decayed ?-rays produced by neutron interactions with materials other than the sample, have to be shielded properly such that their interference to the y-ray spectrometer is minimal. Multiple layers of shielding materials are usually adopted to surround the y-ray spectrometer to stop these interfering y-rays. Since the sample to be investigated is exposed to high neutron flux and emits intense prompt yrays itself, it is preferably handled by an atomatic control device. Most PGAA facilities have an interlock of sample handler to neutron beam shutter in order to ensure no accidental exposure of researcher to intense neutron beams. A shutter device contains a driven unit to push the neutronabsorbing material into the neutron beam, or into the collimator. Alternatively, if the sample is handled by manual procedures, the neutron source should be shut off by a mechanical device to cope with the manual handling of the sample. In addition, the sample to be analyzed can be fixed on the stand to which the criteria of choosing the stand material are the same as those fcr neutrm collimator and shield-minimal induced prompt y-rays and residual radioactivity.

A. REQLJIREMENTS FOR SHIELDING The most significant radiations for which shielding is required for a PGAA faci!ity are the neutrons and y-rays emitted from the neutron source as well as the prompt and decayed -y-rqs prciluccd by .I

G A M M A E N E R G Y . MeV

Figure 11 Chlorine prompt y-ray spectra collected ii.l water containing 1920 pprn CI using (A) a backscattering system with 44 pg 252Cf;counting time: 18,000 s ; dead time: 10.5%; and (B) a transmission system with 260 pg 252Cfat an operation depth of 105 crn; counting time: 36,000 s; dead time: 8.5%. neutron interactions with surrounding materials other than the sample itself. The shield design involves choosing suitable shielding materials with proper thickness such that the sum of neutron and y-doses is in an acceptable range. As mentioned earlier, a maximum permissible radiation dose rate of 0.025 mSv/h, for both neutron and y-radiations together, is adopted outside the PGAA facility. This is calculated from the recommended level of accumulated annual doses of 50 mSv for a radiation worker who conducts on-site PGAA work 40 h a week and 50 weeks a year. For both neutron collimation and absorption design, a material that can slow down neutrons with a large scattering effect is required first, followed by thermal neutron absorption with minimal induced radioactivity. Ideal slowing-down materials should contain light elements with large scattering and small absorption cross-sections for fast neutrons; the best choices that include economical consideration are deionized water, heavy water, beryllium, graphite, and zirconium hydride of nuclear grade. Thermal neutron absorption materials are those containing elements with high capture cross-section, low emission rate of prompt y-rays, and negligible amounts of induced radioactivity. Among the best choices are materials containing 6Li,'OD,"'Cd, and some rare earth elements such as Sm, Eu, Gd, or Dy. The neutron collimator and neutron shield can be made from the combinations of both slowing-down and absorption materials mentioned above. Radiation safety for the researcher, convenience, and economic

GAMMA R A Y EtdERGYv k@V

Ftguro 12 (A) Cd(n,r) reaction cross-section and neutron fiux at the sample station of PGAA/THfviER facility as a function of neutron energy; (B) law-energy prompt ?-ray spnctrurn of Cd sample.

considerations are prime concerns of PGAA operation. 'l'hc &yskni art3 nucleax properties of somc sclected shielding materials are listed in Arixmdix I1 at the cnd t - thc hook. One of the most popul'ar shielding materials for the rieumn is cadmium, freqtlcntly used as the construction materia: for the rieuiron sh.i;:cr or even ihc col;t,ol biaiz icjr a nucleu r c d x *.c)re. The most significant feature of the cadmium to btop neutrons is thxt one of its stable isotopzs can absorb L2lermal neutrons with huge cross-sections uy the '13Cd(n,rj "'Cd rca~tiun.As shown in Figure 128, the weighted cross-sections of the Cd(n,.-) reaction are morc ihan 1CXl barns in thz thcnnnl energ;, range, with resonance at the epithermal energy range, and dect-eased to several barns, equivalent to its gcomctric cross-section, in the fast neutron energy range. iiwi,*c,cdmwni is a very eflcciive shielding material to absorb thermal neutrons. However, the Cd(n,r) reaction also emits numerous prornpt y-rays, as illustrated in Figure 12B.Thus, cadmiui'i done c'mnot be utilized as the only shield and 1s usually incorporated with a y-attenuator, such as lexl, to completely st(?;? both thcimal neutrons and attenuete the induced prompt y-rays. On the otherhand, neutrons emitted from the source of thePGAA facility are not completely thcrrrialized. In the PGAmHMER facility mentioned previously, the neutron flux energy spectrtird at the san~plestation is also illustrated in Figure 12A, showing that the neutrons vary across 9 orders of magnitude from the thermal to the fast neutron energy range. Hcnce, a shielding material containing cadmiunl with lead can only absorb thermal neutrons and cannot ever! slow down neutrons at other energy ranges. A moderator such as deionized water has to be applied iirst to slow dowl i:clliihermal neuii-ons prior to h e apglicaiion

Table 3 Shielding Materials Used for PGAA Facility

Fast neutron moderator

Material Deionized water Heavy water Beryllium oxide Graphite (nuclear grade) DiphenyI (200 O F ) ZrH, (nuclear grade)

u,b = 71 a, = 755

Thermal neutron absorber

Lithium Boron Rhodium Silver Cadmium Indium Gold Mercury Most Tare earth elements Iron Copper Lead Bismuth

pc = 7.86 p = 8.94 p = 1 1.35 p = 9.75

Shielding function

Remark MR. = 58

a, =

149

+r +r

a, = 63 + r a, = 2,450 + r u, = 191 + r a, = 99 + r a, = 380 + r u, = large + r

-

y-Ray attenuator

M R = Moderating ratio for fast neutrons. ua= Absorption cross-section for thermal neutrons, barn; with accompanying prompt y-ray r, emission. p = density, g ~ m - ~ .

of a thermal neutron shield and finally the y-shield. In Table 3, the shielding materials frequently used for a PGAA facility are given according to their specific functions. As both fast neutron and y-ray(s) can only be slowed and attenuated, any combination of shielding materials at the best arrangement cannot stop the radiation completely. In the PGAA/TMMER facility with neutron shutter closed," the y-ray spectrometer can still record interfering y-rays and neutrons from the surroundings. The background y-ray spectrum observed in the PGAA experiment has some features that reflect the characteristic construction materials close to the reactor core. A background y-ray spectrum with an energy range of 0.25 to 11.4 MeV obtained from the water-filled phantom is shown in Figure 13. The most intense photopeak in the spectrum is the 2223-keV prompt y-rays from the H(n,r)D reaction. This is because the phantom blocking the external neutron beam contains about 3 X loz7hydrogen nuclides. Above the 2223-keV region, the spectrum is dominated by prompt y-rays as well as their singleldouble-escape peaks from construction materials near the reactor core such as carbon (710 kg graphite in thermal column and reflector), iron (425 kg steel as inner reactor tank), lead (10-kg thick lead wall as y-shield in the vertical beam tube), and aluminum (8 kg aluminum as reactor fuel tank). In the low-energy region, the background spectrum has four major components: (1) Compton scattering background from high-energy y-rays, (2) prompt y-rays from construction materials as well as the single- and double-escape peaks from 2223 keV, (3) annihilation y-rays of 51 1' keV from the pair production of high-energy prompt y-rays, and (4) prompt y-rays from the Ge(n,r) reaction induced by leakage neutron within the HPGe detector. Hence, selection of shielding materials for absorbing neutrons and attenuating interfering y-rays should be performed carefully and, in turn, be dictated by the detection limit of elements of interest in the prompt y-ray spectrum.

B. BIOLOGICAL AND DETECTOR SHIELDS Despite the wide range of field appIications, most PGAA facilities are reactor based. On the reactor floor, the irradiation can be conducted internally in the core or externally outside the reactor by an extracted neutron beam through a collimator. The trade-offs between high tlux with low detecting efficiency in internal irradiation and low flux with high detecting efficiency in external irradiation approximately cancel out in t h e two configurations; however, the external geometry is superior due mainly to the following advantages:

PROMPT

GAMMA RAY ENE6G.Y ( keV

Figure 13 Background spectrum of IVPGAA measurement obtained from a water phantom; gamma peaks of specific elements are labeled; single-escape peak (S) and double escape peak (D) for those underlined photopeaks are also shown. (Reprinted with permission from Appl. Radiat. Isot. 36(5),Chung, C., Yuan. L. J., Chen, K. 0.. Weng, I? S.. Ckang, f? S., and Ho, Y H. A feasibility study of the IVPGAA using THMER, 357, Copyright 1985, Pergamon Journal Ltd.)

-

Lower interferences from reactor core fission y-rays Minimal detector deterioration due to heating and radiation damage Little induced radioactivity in the detzctor and shields Flexible neutron beam profiles using neutronly-filter

Hence, most reactor-based PGAA facilities are in external geometry and the one attached to

c.

"0'

"a 3 X

3 A LL.

Z

lo7 lo6

lo5

0

a: k-

10'

3

w lo3 z 1o2 10

Figure 14 General shielding layout for P G W H O R facility and thermal neutron flux distribution within.

THOR in Taiwan, a typical PGAA set-up using external reactor beam together with associated radiation shields, is introduced briefly as an example. Neutrons from the THOR facility are extracted horizontally from the H 2 0 reflector region by means of the through-port approximately 30 cm from the nearest fuel elements, as shown in Figure 14. Two-section beam tubes are inserted in the through-port to collimate and provide the neutron beam at targets with a cross-section of 5 cm in diameter. The thermal neutron flux measured in the beam using gold foil is 10" n/cm2 - s at the region closest to the reactor core with a cadmium ratio of 11: 1; 1.3 X lo6n/crn2 s at the target with a cadmium ratio of 26.4: 1; and further dropped to 100 n/cm2 s at beam end. The beam passing through the target is captured with a beam catcher consisting of 40 wt % B02 and polyethylene matrix blocks surrounded by heavy concrete blocks, as illustrated in Figure 14. The concrete blocks also serve as a biological shield for those who have to work around the facility. Outside the biological shield, the dose rate measured by health physics instruments yields 18 pSv/h for y-radiation and 5 pSv/h for fast neutrons, further assuring the safety of the PGAA operation. The closed-in shield design to protect the prompt y-ray spectrometer should eliminate most scattering neutrons and attenuate interfering y-rays. The radiation shields to protect the antiCompton spectrometer used in the PGAA/THOR facility is described here.4 Lead is chosen as the major y-ray detector shielding material and a layer of 8-cm thickness is used to surround the spectrometer. To eliminate the X-rays induced by any lead-photon interaction, an annular NaI(T1) detector is further covered by absorption layers of lucite (I-cm thick) and with an outer ring of copper (1.5-cm thick). In addition, to prevent neutrons from reaching the detectors, the front-end window of the HPGe detector is also covered by a 1-cm layer of 6LiF powder; outside the lead shielding of the spectrometer is an additional 5-cm layer of B02-loadedpolyethylene blocks serving as a first-line scattering neutron absorber. The detector assembly is placed on a trolley such that the 1-ton assembly can be moved around the PGAA facility with easy adjustment in all directions, as shown in Figure 15.

-

There are scores of IVPGAA facilities attached to hospitals and research institutes for medical diagnosis, their numbers in operation are next to the reactor-based facilities that are dedicated for the analysis of elemental concentration in small samples. The radiation shield design for an IVPGAA

Figure 15 (A) Side view of the detector assembly and (€3) cross-sectional view of the 7-ray spectrometer and shielding of the P G W H O R facility.

facility is less complicated than that of 2 reactor-based facility due to the neutron flux used in IVPGAA being at least 100 times less than that of reactor external beam. The biological and detector shields of three IVPGAA facilities using a 23RPu-Be neutron source,'* *2Cf fission neutron s o u r ~ e ,and ' ~ mobile reactop are described briefly as follows. Alshough the neutron capture reactions of interest inside human organs and tissues occur predominantly with low-energy neutrons (below 0.5 eV), it is necessary to use fast neutrons for the irradiation of an extended object like Ule human trunk because of the poor penetration of the body by slow neutrons. The fast neutrons penetrate the body and are slowed down to thermal energies by elastic and inelastic collisions with body elements (mainly elastic scattering by body hydrogen). The neutron source used in Reference 18 is 85 Ci Wu-Be; this sourcc. emits neutrons with a mean energy of about 4.5 MeV. The irradiation facility is shown in Figure 16A. The neutron source is housed in a collimator made of epoxy resin heavily doped with Li2C0, and %iF. It is designed to provide a rectangular beam 13 X 20 cm2at the level of the bed, situated 50 cm above the source. The center of the neutron beam is offset 8 cm from the midline of the bed. The fast neutron flux at the level of the bed is calculated to be 7.2 X 104 n/cm2 . s. The neutron shield consists of boxes filled with polyester resin mixed 25% by weight with Li2CO3and 5 wt % with 6LiE These compounds are used here to minimize thermal neutron intensity. The entire facility is covered with a 10-cm thick layer of lead. This layer reduces the intensity of the y-rays cmitted from the source as well as those produced in the neutron shielding material due to neutron capture and inelastic scattering. Each of the two Ge(Li) detector:; i s shielded from direct vicw of the source by a bismuth annulus 5.5-cm thick and tungsten alloy (98% W, 2% Cu) blocks 10-crn thick positioned underneath the annulus. To reduce the neutron capture and inelastic scattering, thc detector is further surrounded by a cup, 1.5-cm thick, made of paraffin heavily doped with 6LiF. The layout of the californium neutron source, radiation shielding, and the detector in the IVPGAA facility is shown in Figure 16B.I9 The 100 pg Z52Cfis housed in a steel conical collimator, which is placed in a cylindrical-shaped, certified transportation container made of berated polyethylene. The top of the cylinder is covered with a 5 cm-thick layer of lead. A 1.0-cm thick lead disc is placed just above the source within the collimator in order to reduce the fission y-rays originating from 252Cf.Tfie y-rays are measured with two Ge(Li) detectors, each having a 100-cm3 volume and 25% efficiency. The detectors are shielded with cups containing a mixture of wax, 6LiF, and polyethylene-boron-lead blocks. The choice of shielding materials used for the detectors depends on the prompt y-ray energy of the particular elements of interest. m e IVPGAA station described in Reference 20 is on top of the reactor tank of the THMER critical assembly, as shown in Figure 16C. A vertical neutron beam, extracted from the reactor core, is collimated by boron-loaded pdyethylene-lucite and filtered by a 95% enriched 6Li2C07 thermal neutron absorber at the outlet. After the THMER is started up and reaches its full power,

Nwfrtn

A

6 C 5 E F

G

Slruclure moleriala E p x y resin Leod Polyester resin Tungsten olloy Bismuth

Figure 16 Biological and detector shields for IVPGAA facilities using (A) 238Pu-Beisotopic neutron source, (B) =ICf fission neutron so (Reprinted with permission from Phys. Med. Biol. 22(6), Vartsky, D.. Ellis, K. J., Chen, N. S.. and Cohn, S. H., A facility for h vivo meas PGAA, 1085, Copyright 1977, Institute of Physics Publishing; from Nucl. Instr. Meth. 213, Varkky. D., Ellis, K. J., Wielopolski, L.,an backaround in IVPGAA, 437, Copyright 1983. North-Holland Physics Publishing; and from Appl. Radiat. Isof. 39(2). Chung, C., In vivo . - " filtered neutron beam, 93, Copyright 1988, Pergamon Journal Ltd.)

Figure 17 (A) Layout and (B) cross-sectional view of the OLPGAA facility designed for airport security inspection to detect hidden explosives in luggage. (Reprinted with permission from Appl. Radiat. lsot. 44(12), Chung, C., Liu, S. M., Chao, J. H., and Chan, C. C., Feasibility study of explosive detection for airport security using a neutron source, 1425, Copyright 1993, Pergamon Press Ltd.)

neutrons from the 10.5-cm diameter beam tube can be shut off by pumping the boric acid into the beam tube in 45 s without affecting the reactor power level. In the IVPGAA station, the neutron beam is shut off during loading and unloading of patient. Once the patient is placed in a supine position, where the target organ is centered at the beam tube, the neutron beam is initiated by pumping off the boric acid solution. The HPGe detector and neutrody shields are also positioned next to the irradiated patient; lead and Li2CO3cups are used to attenuate scattering y-rays and neutrons, respectively. Since most of neutrons extracted form the reactor core are effectively slowed down and absorbed by the irradiated phantom, only minor neutron doses are measured in the THMER facility, with no more than 3 pSv h-I outside the reactor tank.21Scattering y-rays from the THMER operation, however, dominate the radiation exposures in the facility. y-Ray dose rate equivalents are 30 pSv h-' around the reactor tank, and drop only to 10 pSv h-I at the edge of THMER trailer 4 m away. This is because scattering y-rays, originating from the reactor core and penetrating all y-shields, are high-energy photons with little attenuation in the environment. In another field application, workers may be stationed near the PGAA facility, therefore demanding a mere stringent radiation shield. Here, the OLPGAA facility used for airport security inspection, together with its radiation shield, are described." In this set-up, an explosive detecting assembly that can accommodate two separate neutron sources (at ?Z-axis), as illustrated in Figure 17, is set up. Two commercial neutron shield casks (Reactor Experiments, Inc., Model C-258F), ruggedly constructed of heavy-gauge steel, as well as the boron-doped filter with exceptionally high hydrogen content, are used as primary thermal neutron moderator and absorber. The hydrogen moderates the fast neutrons while the boron absorbs them, and the steel cask attenuates any induced prompt y-rays within the C-258F neutron shield. A 10-cm thick lead wall is assembled around the C-258F neutron shield to further suppress any y-rays reaching the detectors other than those directly from inspectedhadiated luggage. A layer of 6LiFis placed in front of each 2" X 2" BGO detector to avoid scattering thermal neutrons interacting with the scintillator. Fast neutrons emitted from the 252Cfsource are collimated and directed to a luggage conveyor positioned 15 cm above. The average fast neutron fluence for a 30-s scan in luggage is estimated ~, to a dose equivalent of 0.1 mSv. This is well below to be around 7 X lo5 nl ~ m - corresponding the maximum level (0.7 mSv) recommended by local authorities on airport security irradiation.

-

The d i a t i o n level outside the surface of the explosive detecting assembly, surveyed by health physics probes, are 10 and 30 JLSVh-' for y-rays and neutrons, respectively. Hence, the assembly must be placed in a radiation control area unless an extra neutrordy shield is added to further reduce the leakage of neutrons and y-rays.

The performance of PGAA investigation of elemental composition in sample has been dictated by the instrumentation and shields associated with the facility. In this chapter, current set-ups of prompt yray spectrometric systems and various radiation shield designs are reviewed. Instruments and radiation shields used in current PGAA facilities may be further improved by the following modifications. The suitability of a PGAA facility is indicated by the low detection limit of the element of interest with a minimal dose, in addition to the requirements of a reliable rapid analytical process. The PGAA system still can be improved substantially by further lowering both the detection limit of elements of interest and dose equivalents received during irradiation. Despite the fact that higher neutron flux and longer irradiation time yield better statistics for lowering detection limits, unfortunately, they also increase the doses proportionally. Thus, increasing the neutron intensity or irradiation period cannot improve the overall PGAA performance. Explicit improvement can be made by employing a larger semiconducting detector Bndlor a multidetector array using a summed matching multiplexer. The experimental configurations in Figures 14 and 16 allow for two to four portable HPGe detectors with high detecting efficiency positioned on opposite sides of the sample, thus increasing the overall efficiency substantially. The same counting statistics would be achieved with only a fraction of the irradiation time, thereby reducing the dose equivalents accordingly. The accuracy of PGAA measurement and detection limits are governed also by spectral background, which can be suppressed by using an anti-Compton coincidence system. In consideration of the limited space at the detector position along the irradiation plane, the high-density BGO detector may be selected as a detector shield to suppress the Compton background originating irom high-energy y-ray scattering off the main HPGe detector. The Compton suppression spectrometer applied to high-level radioenvironments can effectively suppress the spectral background, thus providing an alternative of selecting a multidetector system. Although the radiation doses in PGAA are small, they are nevertheless not insignificant, and every possible effort should be made to reduce the unnecessary doses and confine the irradiation to the sample only. The radiation doses can be further reduced by a well-collimated neutron beam of much smaller neutron fluence, or shorter irradiation period. On the other hand, y-ray doses caused by high-energy scattering photons can be attenuated effectively by adding a lead cover around the detector; with a 4cm thickness of lead, the y-ray doses may be further reduced by 70%. Nevertheless, from all pioneering studies and investigations of instrumentation and radiation shields for the PGAA set-up, the prompt y-activation, using either a reactor-based facility or n field assembly, is highly feasible and technically promising.

REFERENCES 1. International Commission on Radiological Protection, Recommendation of the ICRP, Publication ICRP26, Pergamon Press, New York, 1977. 2. Lee, C. J., Chao, J. If., and Chung, C., High energy gamma ray spectrometer using BGO detector, Appl. Radial. Isot. 42(6). 547, 1991. 3. Chung, C., Wei, Y. Y., and Chen, Y. Y., Determination of whole-body nitrogen and radiation assessment using IVPGAA technique, Appl. Radiat. Isot. 44(6), 94 1 , 1993. 4. Chung, C. and Yuan, L. J., Performance of HPGe-NaI(T1) Compton suppression spectrometer in high-level radioenvironmental studies, Nucl. h t r : and Meth. A243, 102, 1986. 5. Chung, C. and Yuan, L. J., PGAA using beam from THOR facility, in Proc. 1st Asian Symp. Research Reactors, Institute for Atomic Energy, Rikkyo University, Tokyo, 1986, 3 10. 6. Cohn, S. H., Brennan, B. I,., Yasurnura, S., Vartsky, D., Vaswani, A. N., and Ellis, K. H., Evaluation of body composition and nitrogen content:; on chronic dialysis as determined by total body neutron activation,

Am. J. Clin. Nut,: 38, 52, 1983. L. J., Determination of Ca, C1, N,

7 . Chung, C. and Yuan,

beam, Appl. Radiat. Isot. ~ 7 9977, , 1988.

and P by IVPGAA

using mobile reactor neutron

8. Vartsky, D., Wielopolski, L., Ellis, K. J., and Cohn, S. H., High countrate problems in elemental analysis u~ingpulsed neutron inelastic scatlering, Nucl. Instr. Meth. 206, 575, 1983. 9. Chung, C. and Tkeng, T. C., ISPGAA of pollutant in seawater using a shallow neutron probc, in Capture Gamma Ray Spectroscopy 1987, Br
Chapter 3

Neutron Damage and induced Effects on Nuclear Instruments Used for PGAA C h k n Chung CONTENTS I. Introduction .............................................................................................................................. 37 11. Neutron Flux Distribution Around Instruments .........................................................................37 A. Neutrons Around Reactor-Based PGAA Instruments ........................................................ 38 B. Neutrons Around PGAA Field Instruments ....................................................................... 40 111. Neutron Damage on Detectors ................................................................................................... 43 A. Neutron Damage on Semiconducting Detectors ................................................................. 4 4 B. Neutron Damage on Scintillation Detectors .................................................................... 46 IV. Neutron-Induced Effects on Detectors ................................................................................... 49 A. Using an HPGe Detector as a Fast Neutron Monitor ......................................................... 50 B. Using an HPGe Detector as a Thermal Neutron Monitor ..................................................54 V. Discussion ................................................................................................................................ 56 References ............................................................................................................................................. 57

I. INTRODUCTION A prompt y-ray spectrometric system, containing either a scintillation or semiconducting detector, is the helm of a PGAA facility. Both kinds of radiation detectors are subject to neutron irradiation during the prompt y-ray analysis and various degrees of damage can be mounted by prolonged exposure. The NaI(T1) crystal generates intense internal radioactive 15-h 24Naand 25-min I2'I when it is exposed to even low flux neutrons, severely interfering with the collection of low-energy prompt y-rays. The BGO crystal may have less induced radioactivity; however, prolonged irradiation with thermal neutrons may disable it as a spectrometer due to intense radioactivity in the detector case and PMT base. On the other hand, sustained exposure of the HPGe detector to fast neutrons risks the high-resolution performance and may even become unusable due to distorted hole-trapping ability. Therefore, a neutron shield surrounding the spectrometer is mandatory in order to suppress the neutrons scattering into the detector. Since accumulated neutron lluence in the detector is the index of neutron damage, on-line monitoring of neutron flux within the detecting system is desirable. Fortunately, the prompt and decayed y-rays emitted from the neutron interaction with detector material provide information on neutron flux therein. Preliminary studies of using internally generated prompt y-rays from an HPGe detector as a neutron monitor have been performed, indicating its advantage in the real-time measurement for both thermal and fast neutron fluxes. Hence, the semiconducting detectors can monitor the low-flux neutron during PGAA measurements, thereby providing timely warning for possible neutron damage in its continuous service for PGAA measurement.

11. NEUTRON FLUX DISTRIBUTION AROUND INSTRUMENTS As illustrated in figures in the last chapter, each prompt y-ray spectron~etricsystem used in a PGAA facility is protected by neutronly-shields to avoid neutron damage on the one hand, and suppress the interference from scattering y-rays and neutrons on the other. Materials used for this purpose are typically a 'jLi-contained thermal neutron absorber, a lead y-shield, and some fast neutron moderating materials such as graphite if the space is available. Even with the best possible radiation protection, the spectrometric units still suffer from the bombardment by scattering and penetrating neutrons, inducing unwanted nuclear reactions such as (n,nlr) and eventually leading to crystal lattice damage and deterioration of spectroscopic performance, and generating numerous prompt and decayed y-rays to interfere with the collecting spectrum.

For instance, in the reactor-bascd PGAATTKMER facility,' the HPGe detector, with phantom skinto-dctcctor distance of 1.5 cm, is protected by a 1-cm thick lucilc disc fillet; with 95%-cnriched "$ as thermal neutron absorber. The scattered and slowed-down thern~alneutron flux in the HPGe crystal is measured around 200 n&m2 s; this is less than 3000 n1dcmZ s at the irradiated organ. Even with such low neutron flux, the syslem resolution of the spectrometer deteriorates to 2.3 keV at 1332 keV, or 12% worse than the expected performance. This deterioration is due mainly to the neutron interaction with Ge crystal lattice as well as the high courlt rate of the prompt y-rays. Hence, the neutron flux distribution around :he prompt y-ray spectrometiic unit has to be mapped first in order to know the influence and damage inflicted by the scattered neutrons in PGAA measurements.

.

.

A, NEUT.SONS AROUND REACTOR-BASED PGAA INSTRUMENTS Reactor-based PGAA facilities are permanently set up as a powerful nuclear analytical tool, demanding a high-quality neutron profile-proper energy distribution with the highest possible flux delivered to the sample from the reactor core. In order to avoid neutrons scattered into the prompt y-ray spectrometer, the detector is protected by carefully selected neutronly shields with ample sample-to-detector distance. Here, the reactor-based PGAA set-up attached io the THOR facility2 and neutrons around the prompt y-ray spectromete9 are described for the evaluation of the neutron-induced effect and damage on the instruments introduced in the latter part of this chapter. Neutrons from the THOR core are extracted horizontally from the reflector region by means of a though-port, which is also acting as the collimaior lo providc Uic neutron >cam at the sample with a cross-section of 5-cm diameter. The thermal neutron flux measured in the beam axis, as illustrated in Figure lA, is 1.3 X lo6 nlh/cm2 s at the sample position with a cadmium ratio of 26:l using the AuKd foil activation technique. The main y-ray detectors are a 25% HPGe detector and a 9" X 10"

CADMIUM RATIO 1 I

METER

I

Figure 1 Thermal neutron flux at (A) transitional (A-A) axis along the neutron beam direction and (B) cross-sectional (B-B) axis along the sample-detector direction around the prompt y-ray spectrometric unit of the P G W H O R facility.

annular NaT(T1) detector shield coupled together as an anti-Compton spectrometer. The geometric of the two detectors, together with shielding materials attached to them, are illustrated in Figure 1 ~Lead . is chosen as the major y-ray detector shielding material and a layer of 8-cm thick lead is'used to surround the spectrometer. To eliminate the X-rays induced by lead-photon interaction, the Gnular NaI(T1) detector is further covered by absorption layers of 1-cm thick lucite and 1.5-cm thick =Apper. Jn addition, to prevent the neutrons from reaching the detectors, the front-end window of the HpGe detector, as well as the entrancz hole immediately next to the sample, are also covered by a layer of 1-cm thick 6LiFpowder discs. Outside, the lead shielding of the spectrometer is again surrounded by an additional layer of 5-cm BOp loaded with polyethylene blocks as a first-line scattering neutron absorber. The thermal neutron flux, along with the sample-detector axis, are measured by AuKd and a set of 400-mg In-foils. Using the foils activation technique, neutron flux measurements are shown in Figure 1%. The thermal flux is continuously dropped while scattering into the detector from the sample along the sample-detector (B-B) axis. Sudden depress appears at the 6LiF disc as a result of effective thermal neutron absorption; the slight build-up of thermal neutron flux behind each 6LiF disc is due to the themalization of penetrating nonthermal, scattering neutrons. The thermal flux in the HPGe crystal, averaged over the In-foil measurements around the detector can, is about 29 10 n,Jcm2 . s, or 1145,000 of that at the sample 25 cm away. Although the thermal neutron bombardment on the HPGe detector has a low rate, the integrated neutron fluence, however, is sufficiently too large to be ignored. For instance, a prolonged PGAA measurement for 1 week will accumulate 1.8 X lo7 n,,,/cm2 in the HPGe detector, generating 100 Bq activity from all radioactive germanium isotopes within. Since the fast neutron flux at the HPGe detector is well below the detection limit of the In-foil activation technique (estimated around 3400 nflcm2 s), the fast flux in the HPGe crystal can only be estimated by the cadmium ratio available for the neutron beam profile. The fast neutron flux, assessed by the experimental cadmium ratio? is around 20 nf/cm2 - s; a week-long continuous operation for PGAA measurement will accumulate a damaging fast fluence around 1.2 X lo7 n/cm2, a level that may cause explicit performance deterioration. If the sample-to-detector distance extends further away, the neutron flux in the HPGe crystal may be reduced by inserting more neutron slowing-down material and thermal neutron absorber; however, the detecting efficiency of the spectrometer is also lowered in proportion to the inverse square of the distance. The trade-off and the choice of proper sample-to-detector distance depend upon the severity of neutron interference in the HPGe detector. For instance, in the PGAA/NBS facility," the anti-Compton prompt y-ray spectrometer is positioned further away from the sample station where the thermal neutron flux is 2 X 10' n&m2 s, or 150 times higher than that of PGANTHOR facility, with 40-cm sampleto-detector distance. Having an extra 12.7-cm thickness of paraffin and lQBPLi-contained material in between, the thermal neutron flux in the Ge(Li) detector was estimated to be around 2 n,,,/cm2 . s. On the other hand, some PGAA applications demand close-in geometry in which the detector is positioned immediately next to the irradiated sample. In the IVPGAA facility using the THMER critical a s ~ e m b l y ,the ~ filtered neutron beam, extracted from the reactor core, is delivered to the patient in supine position for in vivo medical scan to which a 25% HPGe detector is placed right next to the patient. A 1-cm thick lucite disc filled with 95%-enriched "iE; is buckled to the window of the HPGe detector to stop thermal neutrons from scattering into its active volume. Furthermore, in order to reduce the prompt y-ray background from the construction materials, 4.5-cm thick lead bricks are placed around the HPGe detector. With such a configuration, Compton background in the y-ray spectrum can be reduced by a factor of 10 in the low-energy region; this IVPGAA arrangement is illustrated in Figure 2. The foil activation technique is again used for the thermal neutron flux measurements around the prompt y-ray detector and within the liquid phantom; results of thermal flux along the detector-phantom axis are also shown in Figure 2. Although the thermal flux at the irradiated organ is rather low, around 3000 n&m2 . s, the thermalized and scattering neutrons after penetrating the 6LiF disc in the HPGe crystal is relatively high, around 200 n,,Jcm2 . s. This is due mainly to the close-in geometry between the HPGe detector and the irradiated phantom-only 1.5 cm away from the phantom skin and 5 crn away from the edge of neutron beam tube. The nonthermal neutron flux in the HPGe detector, calculated by neutron transport cock6 is around 560 nl/cm2 - s. A week-long IVPGAA operation will accumulate thermal neutron fluence to the detector mound 1.2 X lo8 n&m2 with 700 Bq induced radioactivity from all germanium isotopes, as well as n fast neutron fluence of 3.4 x lo8 nf/cm2in the germanium

+

Figure 2 The geometric arrangement and ther.rnal neutron flux distribution along the detector-phantom, along the X-axis in the I V P G M H M E R facility.

srystal, 2 dangerous rrccumulation level such that the detector may no longer be used as a prompt yray spectrometer.

B. NEUTRONS AROUND PGAA FIELD INSTRUMENTS Field applications using PGAA techniques include geological and environmental ISPGAA surveys, process-line OLPGAA measurements, and medical IVPGAA scans. In all cases, the detecting instruments are close to the neutron source well as the sample matrix in order to acquire rapid response of prompt y-rays. The popular nuclear instruments used for PGAA are HPGe semiconductitlg detectors, and NaI(T1) and BGO scintillation detectors as well as their associated electronics. Several neutron flux measurements around the prompt y-ray spectrometric unit have been mapped and their results are described in this section. In a water pollutant survey, a shallow 252Cf-HPGeprobe based on the ISPGAA technique was designed and constructed by Tsing Hua's group? Since the details of such a submerged PGAA probe have been described in later chapters, only the parts associated with neutron protection are mentioned here briefly. In the backscattering design, the probe, illustrated in Figure 3, is designed to operate vertically from the stem of a 5-ton motorboat for in situ measurement. Using 4-mm aluminum as construction material to sustain underwater pressure, the probe has an 18.2-cm inside diameter and a 66.9-cm height, providing enough space to accommodate the 2.7-pg 2S2Cfneutron source, neutron and y-ray shields, and a 10% portable HPGe semiconducting y-ray detector. To maintain the water-tight status during field operation, the cable outlet at the top and the removable part of the 252Cf housing at

FIgure 3 Sectional neutron f h x contour map showing ISPGAA application in water designed by Tsing Hua group. Numbers indicated for each curve represent tho average thermal

neutron flux in n/cm2 . s. (Reprinted with permission from Nucl. Instr. Meth. A267, Chung, C. and Tseng, T. C., ISPGAA of water pollutants using a shallow 252Cf HPGe probe, 223, Copyright 1988. North-Holland Physics.) the bottom are sealed by O-rings. Three mgged eye-hooks, attached to the top of the 32-kg probe, provide the necessary steel cable attachment to enable the probe to be submerged and recovered in field operation to depths of 25 rn. As shown in Figure 3, lead is chosen as the y-ray shielding material and is placed between the 2S2Cf neutron source and the y-ray detector. With a maximum thickness of 10 cm at the center and a total weight of 22 kg, the lead y-ray shield can attenuate 99.95% of the prompt fission y-rays and fission product decay y-rays from the neutron source, thereby avoiding interference to the prompt y-ray spectrum originating from the interactions of neutrons with the surrounding water contents. To suppress direct interaction of neutrons with the y-ray detector, two layers of neutron attenuators--one 2.6-cm thick, 95% enriched 6Li2C03as neutron absorber and one 2.6-cm thick lucite as fast-neutron moderator-are placed around the detector as a neutron shield. The neutron flux is monitored using the activation foil technique. A total of 72 high-purity indium foils, each weighing about 0.4 g, are activated at various axial and radial positions around the probe while submerged in the test tank. Isoflux contour curves are subsequently mapped in water around the probe based upon the experiinental neutron flux data, and their results are also shown in Figure 3. The thermal neutron intensity decreases isotropically as the distance from the probe increases, while the neutron intensity near the lower part of the probe is higher than that at the upper part. The most intense neutron flux in water, 15,000 n,Jcm2 s, is near the bottom of the probe; however, the neutron flux just off the HPGe detector drops rapidly to 1000 nlh/cm2 s due to the thermalization and absorption of neutron by water as well as by the neutron absorber inside the probe. On the other hand, the fast neutron flux in the HPGe detector can be calculated by neutron transport code, resulting in 350 nf/cm2 . s at the detector head. The neutron effect on the exposed FIPGe is severe, the system resolution in the ISPGAA survey is raised to 3 keV at 1332 keV, or 67% worse than expected performance. The high exposure rate to the and fast neutron fluence of 2 X scattering neutrons, with thermal neutron fluence of 6 X 10%lh/~m2 10' n,/cm2 as well as 1400 Bq internal activity in the HPGe crystal itself after 1-week continuous underwater operations, is due primarily to the compact arrangement between the detector and neutron source in the submerged probe. Service life for such backscattering probes is estimated only up to a Yew with continuous operation; the accumuIated neutron bombardments at the end of such service may inflict enough damage to disable the HPGe detector as y-ray spectrometer.

x THERMAL NEUTRON FLUX MAPPING,

100

nth / cm2. s

500

Ftgure 4 Thermal neutron flux dis3ibution around the explosive detecting assembly, designed for airport security inspection, which contains two 100 pg 252Cfisotopic neutron sources and ten 2" x 2" GGU detectors.

In other field applications, the scintillation detectors are utilized instead. For instance, the NaI(T1) detectors have been used for in vivo PGAA aedical scan: and the BGO detector-array has been used for on-line PGAA measurement in search of explosives in Iuggage? In the latter application, neutron flux is mapped in around both the luggage sad the detectors. In the OLPGAA work? an explosive detecting assembly that can accommodate two separate neutron sources (at k Z-axis) and 10 BGO detectors (five in each side of the X-axis), as illus~atedin Figurc 4, is sct up. Two commercial neutron shi~ldcasks, ruggcdly coiistructed of heavy-gaugc steel, as we11 as the boron-doped filter with exceptionally high hydrogen content, are used as primary thermal neutron moderator and absorber. The hydrogen moderates the fasl neulrons while the boron absorbs them, and the steel cask attenuates any induced prompt y-rays within the neutron shield. Two neutron sources, each with a 100-bg *2Cf source having neutron strength of 2.3 X lo8 n/s, are placed at the "_ Z cask. A 10-cm thick lead wall is assembled around the neutron shield to further suppress any y-rays reaching the detectors other than those directly from inspected luggage. In order to identify explosive-like materials readily, ;m m y detector, consisting of two sets of five 2" X 2" BGO detectors (positioned at the +, X-axis) with associated electronics, are used with Sf'-thick Li2C03cup coupled to the front end of each BGO detector to absorb scattered neutrons. The explosive detecting assembly is tested by measuring the thermal neutron flux distribution around inspected luggage. Since the luggage may contain food items and other personal belongings that have high neutron cross-sections, (e.g., a chlorine-containing item such as salted food), the neutron flux that reaches the luggage should be suppressed to a level low enough such that no dangerous level of radiation hazard or significant radioactivity will be induced. On [he other hand, the neutron flux should be kept at the level such that a fast scan of hidden explosives inside the luggage is feasible. Using the foil activation technique again, the thermal neutron flux can be evaluated by measuring induced radioactivity of In-foils. The thermal neutron flux distribution around the inspected Iuggage as well as around the BGO positions is also illustrated in Figure 4. As shown in the flux mapping figure, the irradiated luggage is encircled by thermal neutron flux with range from 2,500 to 15,000 nh/cm2 s; however, the BGO detector array, despite the protection I order of 250 of a 5"-thick Li2C03cup, still suffers from the exposure to a thermal neutron flux G ~ the n,&rn2 - s. Fast neutron flux, assessed by neutron transport code, is on the order of 10 n,/cm2 s; this low fast flux is due to the fact that the 252Cfsource is loaded well within the neutron shield with only

,narrow collimator opened toward the luggage, thus, only a small fraction of fast neutrons can be s&ttered by the luggage into the BGO detectors. After 1 year of continuous operation, the neutron fluence accumulated on each 2" X 2" BGO detector would be 8 X lo9nJcm2 and 3 X lo8n&m2 with 410 Bq internally activated radioactivity on the BGO crystal. This is far below the critical level that may disable the EGO detector used as the y-ray spectrometer."' If the BGO detectors are replaced by NaI(T1) detectors with identical size, similar to those used in [be commercial TNA unit," the activated Na, I, and TI isotopes may yield a total activity of 2500 Bq, &tting both g- and y-rays internally; it may hamper the spectroscopic performance in the low-energy r;lnge up to 3 MeV. In some IVPGAA medical diagnoses, large NaI(T1) crystals with dimensions of 6" x 6" and neutron source with similar strength are used.I2 If the scattered and thermalized neutrons in the NaI(T1) crystal are of the same order, the induced internal activity after 1-week operation would become 68,000 Bq with self-emitting decayed y-rays all over the low-energy part of the prompt y-ray spectrum, further hampering its usefulness as a spectrometer below 3 MeV. In all these field applications, the nuclear instruments are positioned as close to the irradiated sample as possible, while the sample is placed right on the neutron beam. With the constraint of limited space, radiation protection around the detector has preference to accommodate the y-shield first, with even less space left over to insert the neutron slowing-down material and thermal neutron absorber. These unique geometric arrangements, with less effective neutron protection around the nuclear instruments, yield much larger neutron fluence in the prompt y-ray detector with respect to that in the reactor-based PGAA facility. Hence, the detectors used for PGAA field applications may have shorter service life than those used in the reactor where ample space is available for the insertion of a neutron shield around the PGAA nuclear instruments.

Ill. NEUTRON DAMAGE ON DETECTORS All prompt y-ray detectors, semiconducting or scintillation crystals, will respond to some extent if exposed to scattered neutrons in PCJAA measurement; such interfering neutrons may generate pulses as undesirable background. These pulses fall into three general categories: prompt pulses that are produced within a few nanoseconds after the neutron enters the detector; capture y-pulses that are produced by the interaction of detector material with slowed-down neutrons at the time up to 100 ps after the neutron hits the detector; and pulses resulting from the induced radioactivity long after the neutron strikes the detector. In NaI(T1) and BGO scintillators, the internaI prompt pulses are principally due to the detection of y-rays produced in inelastic scattering interilctions of the fast neutron with the scintillators." The BGO detector has a better detection efficiency corresponding to the prompt y-ray pulse than to neutrons in comparison with NaI(T1) detector. Similar effects in HPGe semiconducting detectors may generate spurious peaks that arise mainly due to the excitation of germanium nuclei by inelastic neutron scattering, followed by the emission of de-excitation y-rays, internal conversion electrons, or X-rays. The neutroninduced peaks are usually identifiable in the spectrum because their width is normally larger than that for y-ray-induced events. The peak broadening takes place because a fraction of the excitation energy goes into the recoiling germanium nucleus, which subsequently~contributesa variable yield of electronhole pairs, adding to those created by the de-excitation radiation. The second type of pulse is generated from the capture reaction of thermal neutrons', or thermalized neutrons after slowing-down within the detector, interacting with nuclei in detector crystal. Prompt yray photopeaks originating from the Ge(n,r) reaction in the HPGe detector can be easily identified in the spectrum, although similar pulses generated in the NaI(T1) and BGO detectors are not explicitly identifiable due to poor spectral resolution. The last type of pulse is generated from the induced radioactivity long after the neutron bombardment. The internal activities are induced by (n,r) reactions on stable isotopes of Na, I, and TI for the NaI(T1) detector, Bi, Ge, and 0 for BGO detector, and Ge for the HPGe detector. Nuclear properties of these radionuclides activated by scattered neutrons in the detector are listed in Table 1. Other interfering pulses registered in the prompt y-ray detector may originate from the interactions of scattering neutrons with surrounding equipment and materials such as detector housing, vacuum enclosure, cryostat, and LN?dew= for IIPGe spectl-ometers;detector can, PMT, tube base, and packing materials for scintillation detectors; neighboring neutrody-shields; and sample handling devices. Their

Table 1 Properties of Radionuclides Activated by Scattered Neutrons in the Prompt 7Ray Detectors Stable Isotope [detector]

180 [I3GO]

,

Radioisotope (reaction)

Cross-sectlon, barn

Half-life (decay mode)

Intense y-ray, keV(Ir)

'go

0.00016

26.8 s

197 (90.7%) 1357 (55.0%)

0.13

21 1 (30.2%) 264 (53.0%)

2,200 520

160 (11.5%) 216 (21.2%) 443 (16.0%) 526 (1.5%)

3,660 860

nGe (n,4

.

(F)

(n.r)

0.06

11.3 h

(P-1

'

Relatlve responsea

76Ge

n'"Ge (n.4

0.10

52.9 s (P-,EC)

'"I

1281

6.1

25 min

[ N a T111

(n,r)

[HPGe,BGO]

(P',Ec)

-

--

--

- --

-

--

~,oOO.OOO ..

--

- - --

Note: Decay properties and physics constants are taken from Reference 14. ' Relative response is the relative activity after 1-week exposure to the same thermal neutron flux with identical

detector size. pulse intensity, depending on the distance of the neutron reaction spot to the detector, creates the unwanted background in the prompt y-ray spectrum.

A. NEUTRON DAMAGE ON SEMICONDUCTING DETECTORS The most popular prompt y-ray spectrometer used in PGAA measurement is the HPGe detector. Germanium has five stable isotopes, each with a different neutron reaction cross-section, yielding prompt y-rays with various energies and intensities. The relative response of each prompt y-ray in the thermal capture (n,,,,r) reaction and the radiative inelastic (n,nlr) reactions, estimated from the isotopic abundance, the reaction cross-section, the y-intensity as well as from the detector efficiency and the spectral background, are Iisted in Table 2. With more than 126 and 42 prompt y-rays emitted from the Ge(n&,r) and Ge(n,n'r) reactions, respectively, the most intense y-rays with the highest relative response are 691.2 and 596.4 keV, representing fast inelastic Ge(n,n'r) reaction or the first type of neutron-induced pulse, and thermal capture Ge(n,r) reaction, or the second type of neutron-induced pulse, respectively. A "background" prompt y-ray spectrum is collected with the neutron beam opened to the reactorbased PGANTHOR facility and illustrated in Figure 5. With scattered neutron fluxes estimated around 30 n&m2 s and 20 nl/cm2 s, the prompt y-rays emitted from Ge(n,r) and Ge(n,nrr) reactions do

2 Prompt Photopeak Properties Emitted from the Interaction of Various Germanium a,ot~peswith Neutrons Isotope Most intense abundance8 prompt yRelative $.lucllde atom, % rayb E,, keV responseC Thermal capture reactlon Ge(n,,,,r)

<

i L

20.5

175.1

500.2 7.8 596.4 36.5 253.5 7.8 159.5 Fast Inelastic reactlan Ge(n,ntr), wftk E, > 0.5 MeV 20.5 1039.6 27.4 69 1.2 7.8 500.2 36.5 . 596.4 7.8 562.8 27.4

Data taken from Reference 14. I~Datataken from Reference 16. Normalized to the relative count rate in the prompt y-spectra of HPGe detector in Reference 1 for thermal and fast neutron fields. ,&printed with permission from Nucl. Znstr. Meth. A301, Chung, C. and Chen, Y. R., Application of a germanium detector as a low flux neutron monitor, 328, Copyright 1990, Elsevier Science. I

*appeareverywhere from 0.01 to 9 MeV in the spectrum. Some single- and double-escape photopeaks -associated with germanium prompt y-rays with energy greater than 4 MeV are also identifiable in the high-energy part of the spectrum. About 70 of the 168 prompt y-rays of Ge(n,X) reactions are identified and labeled in the spectrum, and major photopeaks of 596, 691, 1098, and 1294 keV emitted from the -.inelastic scattering reaction of Ge(n,nlr) are broadened at the high-energy side of the photopeaks; this :is due to the extra contribution from the recoil energy of the scattered germanium. Some sharp, high$resolutionphotopeaks at 596, 868, and 8733 keV are clearly observable in the spectrum, indicating the contribution from the capture reaction of Ge(n,r) with various neutron energies. In addition, photopeaks from surrounding materials, scch as the 478-keV peak from the I0B(n,a) reaction on boron-contained neutron shield, as well as prompt (7724 keV) and decayed (1779 keV) y-rays from the aluminum-made detector can, are also observable in the figure. Decayed y-rays from radioactive germanium isotopes, such as those listed in Table 1, do not appear in the spectrum; this is due to the high background of the Compton continuum obscuring the observation of these low-energy (E, < 265 keV) photopeaks. As discussed in this section, semiconducting detectors are relatively sensitive to the performance degradation caused by damage created within the detector by incident neutrons. The large volume and long charge collection paths in germanium detectors make them susceptible to such degradation. Because the amount of damage created by fast neutrons of a given fluence is large compared with the damage from an equivalent fluence of y-rays, the most significant effects often arise in PGAA applications, in particular for those arrangements where the I-IPGe detector is positioned right next to the neutron beam. The principal consequence of radiation damage is to increase the amount of hole trapping within the active volume of the HPGe detector. In a damaged detector, the amount of charge collected is subject to a loss due to this trapping that will vary from pulse to pulse depending on the position of the interaction. Measured peaks in the pulse height spectrum will then show a tailing toward the lowenergy side. T h e spectra shown in Figure 6 illustrate a gradual broadening of the 2223-keV prompt yphotopeak measured by a 10% HPGe detector as fast neutron exposure is increased to 168,000 n,/cm2. Exposure of a thick-planer HPGe detector to a fluence of 10%Jcm2 is sufficient to risk measurable eventually, it may become totally unusable as a spectrometer after change in the detector res~lution;'~ exposure to a fluence of 10'O nf/cmz. In coaxial detectors, the specific detector configuration can have a strong influence on the measured Spectral effects. In an HPGe coaxial detector fabricated from high-purity p-type germanium, holes are the catrrier type that are drawn inwafd to the p' contact near the cylindrical axis. For the HPGe coaxial

PROMPT GAk4MA RAY ENERGY, MeV Figure 5 Background prompt y-ray spectrum taken by a 25% HPGe detector in the PGAAlTHOR facility with neutron beam opened and sample withdrawn. detector made froiL high-purity n-type germanium, the electrode polarity is reversed, and holes are instead collected dt the p+ contact fabricated an the outer cyimlrical surface. It is found that these ntypc WIJGe detectors show 90% lesS performance degradation from ratlintion damage when compared with the more commonp-type detectors. The difference lies in thc fact that the damage sites preferentially t r q holes rather than electrons. Becalisc of the cylindrical geometry anti &,: ::nuation of thc incident yrays, more interactions occur at large-volume detectors than at small ones. As a result of their superior performance in the presence of neutron damage, n-type HPGe coaxial detectors hsve become the preferable choice as the PGAA spectrometer.

B. NEUTRON DAMAGE ON SClNTiLLATlON DETECTORS Fast neutron bombarding effects on both NaI(T1) and BGO detectors, frequc~ilyused in PGAA measurements and facing the scattered neutrons all around, have been investigated and r e p ~ r t e d . ' ~ . In ' ~ -a~ ~ more recent study,'O the performance of a BGO scintillation dcteclor is tested after themla1 neutron bombardment. The drift of the photopeak channel as well as change of absolute counting efficiency and photopeak energy resolution become sever: as the BGO detector is bombarded with thermal neutrons

1

I

3

I

2

=*

DAMMA ENERGY r MeV

Figure 6 Prompt y-ray spectra taken by Tsing Hua group with a 10%HPGe detector in ISPGAA measurement under fast neutron bombardment on the detector with flux and fluence of (A) 0.1 nJcm2 s and 2900 nJcm2; (B) 1 nJcrn2 - s and 32,400 nJcrn2; and (C) 170 nJcm2 - s and 168,000 nJcm2.

-

to lot3n,,.,/cm2.Irradiation with thermal neutrons beyond lOI5 nJcm2 disables the BGO detector as a yray spectrometer due to the intense interference from internal radioactivity induced by neutron reactions. In this section, the example of neutron damage and performance degradation of the BGO detector under sustained neutron bombardment, as eventually happened in its application for PGAA measurement, is described. A 1.5" X 1.5" BGO detector is selected for such destruction study.'0 The BGO scintillation crystal is mounted in an aluminum, hermetically sealed housing, a 2" PMT with external p-shielding is coupled behind the BGO crystaI as an integral assembly. The BGO detector subject to thermal neutron bombardment is positioned in the thermal co'lunk of the THMER facility, whose thermal neutron flux ranges up to 10,000 n,,,/cmZ s, and irradiations with accumulated thermal neutron fluence from lo7 to lo9 n&m2 are performed. For higher neutron intensity, the BGO detector is subsequently relocated to the thermal column of the TMOR facility whose thermal neutron flux can be raised to 10' n,&m2 S; eight irradiations with accumulated neutron fluence from lo9to IOl5 n,Jcm2 are performed. Indium foil activation technique is again used to monitor the thermal neutron fluence. The BGO detector, having 12.8% resolution at the 662-keV y-ray level, is immediately tested by a set of monoenergetic y-sources after each neutron bombardment to check the drift of photopeak channel, change of energy resolution, and counting effkiency in the y-ray spectrum. Typical BGO spectra taken before and after the thermal neutron bombardment are illustrated inyigure 7.With accumulated thermal neutron fluence to 2.5 X 1011n,&m2, the photopeak channel, net counts, and FWHM of each photopeak are either shifted or changed slightly in the spectrum. With thema1 neutron fluence less than 10" nJcm2, drift of photopeak channel as well as change of energy resolution and counting eEciency measured by the irradiated BGO detector 'are small. When neutron fluence increases to lot3n&rn2, however, the recovery time becomes longer. The long postirradiation effect can be explicitly observed on the 137Cs peak measured by the irradiated BGO detector with accumulated fluence to 1.2 X 10" n,Jcm2. In Figure 8A, the energy resolution of the 137Cspeak, measured 11 d after the irradiation and thereafter, is plotted against the post-irradiation delay time. There is no data collection during the first 11 d owing to interference by intense internal radioactivity. The energy resolution, 24% (FWHM = 159 keV), occurs 1I d after irradiation; it is a factor of 2 worse than the resolution prior to irradiation. Even after 24 d, the energy resolution of the '"Cs photopeak recovers but levels off at 21%. A similar effect can also be observed for the absolute counting efficiency of the 662-keV photopeak in Figure 8B. The absolute efficiency slawly rises in the first 3 weeks, but levels off around 3.5 counts per 1000 photons emitted from the t37Cssource after the 4th week. The detector never recovers the original energy resolution of 12.8% and absolute efficiency of 25 counts per 1000 photons observed prior to the irradiation.

'"I

70

90

100

110

GAMMA

487

554

621

6 8

755

130

384

636

894

1146

1460

RAY ENERGY, keV

Figure 7 ?-Ray spectra of monoenergetjc photon sources taken by the BGO detector (A) before and (B) after the thermal neutron bombardment with 2.5 X 10li nl&m2 fluence. (Reprinted with permission from Appl. Radiat. /sot 42(8), Lee, C.,I. and Chung, C., Performance of gamma ray measurement using EGO detector after thermal neutron bombardment, 729, Copyright 1991, Pergamon Press.)

The changes of energy resolution and counting erficicrlcy for each reicience y-ray, mcastircd after the rccovery from each bombardment, are shown in Figure 9 as a function of thermal neutron fluence impinging on the BGO detector. The counting efficiency of all low-energy y-rays, as illustrated in Figure 9A, drops sharply as the BGO detector is bombarded with fluence over 10'' nIh/cmZ.Only the absolute efficiency of 6129-keV high-energy photons remains on the same order of magnitude in all neutron bombardments. The resolution of all investigated y-rays, similarly, begins a clear deterioration as illustrated in Figure 9B. with lOI3 n,h/~m2, With accumulated neutron fluence beyond 10'' n&m2, the irradiated BGO detector is no longer useful as a y-ray spectrometer because of the intense internal radioactivity. The induced radionuclidcs are produced by thermal capture reactions on the BGO crystal, packing materials, and the PMT; sornc of them are very long-lived, yielding intense background radiation. For instance, 570 d after irradiation with loi5n,,,/cm2, a total of 2.9 X lo5 Bq are estimated for internal radioactivity of the irradiated BGO detector. Major radioisotopes in both the irradiated BGO crystal and the PMT (such as Ag, Cs, Co, Mn, and Zn) are shown in the high-resolution y-ray spectrum illustrated in Figure 10, which is measured by a high-purity germanium detector. Thesz long-lived radionuclides within the irradiated BGO detector prevent its use as a y-ray spectrometer. Radioactive Bi, Ge, and 0 isotopes induced within the BGO crystal, however, are already decayed 19 months after the neutron bombardment and therefore are not observed in the spectra. In a similar study: an NaI(T1) detector shield is exposed to a thermal neutron flux of 30 qh/crn2 s for 30 min in PGAA measurement using the THOR facility. Some 3 h after the shut-off of the neutron beam, the radioisotopes of 25-min '"1 and 15-h "Na, induced by the interaction of neutrons with the NaI(T1) crystal, are clearly observable in the delayed y-ray spectrum measured by a high-resolution germanium detector, as illustrated in Figure 1I. With longer exposure time, the internally generated radioactivity may interfere with the prompt y-ray spectrum of energy less than 3 MeV. There are several factors that may affect the drift of peak channel as well as the change of absolute efficiency and energy resolution in the y-ray spectrum taken by the irradiated detector. Among these factors, the most prominent one is high counting rate due to internal radioactivity in the scintillation detector. Drift of the photopeak channel is usually associated with the photomultiplier tube and can be severe if the scintillation detector is subject to large changes in counting rate. Internal radioactivity imlnediately after thermal neutron bombardment in the scintillation detector with accumulated lOI3 n&m7

-

rfgure 8 Post-irradiation recovery of the i ( ~ j ' e n e resolution r~~ and (B) absolute effi&ncy for the 662-keV y-ray using the tested BGO detector. (Reprinted with per,mission from Appl. Radiat. Isot. 42(8),Lee, ,C.J. and Chung. C., Performance of gamma ray measurement using EGO. detector after thermal neutron bombard,merit, 729, Copyright 1991, Pergamon 'Press.) &.

21.9

. e

..:G ;5 1.5 ' 2 2 1.1

E:

-

o Q

II

13

15

17

19

21

23

Post Irradiation Time. Day

fluence is high, causing -20% drift of.the 662-keV y-ray in the spectrum. As the radionuclides decay, .the photopeak recovers toward the original channel registered prior to the neutron bombardment. Changes of absolute counting efficiency and energy resolution are also caused by the high counting rate variation at neutron fluence beyond 10'' n,,,/cm2.The resolution loss after neutron bombardment is 'due largely to the deterioration of charge collection statistics under high counting rate as well as the drift during counting. On the other hand, quantitative determination of the photopeak area is obscured by the intense spectral background of internal radiation, thus reducing the counting efficiency of the irradiated scintillation detector.

IV. NEUTRON-1NDLDCEQ EFFECTS OH DETECTORS As stated in Section 111, the HPGe semiconducting detector is more sensitive to fast neutron damage than the scintillation detector. A fast neutron fluence of 1OY nf/cm2impinging on the HPGe detector may risk the photopeak resolution change and become unserviceabk as a spectrometer after 10'O nf/ cm2 bombardment. On the other hand, the scintillation detector may sustain a thermal neutron fluence of lo1)n&m2 and become useless as a y-ray spectrometer with thermal neutron bombardment beyond lOlSnth/cm2.Hence, a timely measurement of neutron fluence around the more expensive HPGe detector is highly recommended in order to avoid the risk of overexposure. A more restricted control of neutron fluence accumulation in an HPGe detector, and to a lesser extent in scintillation detectors as well, is desired by an on-line neutron monitoring device. In the PGAA measurement, there is only a limited space surrounding the HPGe detector for the insertion of a neutron monitoring device. So far, only the foils are inserted around the detector.and they car1 only provide the neutron fluence data long after the PGAA measurement. In a previoun study,''

7

8

g

11

10

12

13

14 15

7

8

LOG ( Thermal neijtron fluence , n,/

9

I0

11

i2

13

14

c d

Figure 9 (A) Absolute counting efficiency and (B) energy resolution of y-rays measured by tho BGO detector long after bombardments with,various accumulated thermal neutron fluenc~s.(Reprintod with permission from Appl. Radiat. Isot. 42(8), Lee, C. J. and Chung, C., Performance of gamma ray moasurement using BGO detector after thermal neutron bombardment, 729, Copyright 1991, Pergarnon Frsss.)

the radiation damage that may result from a given exposure to neutrons can be measured using the G e ( n a reaction. The neutron-induced 691-keV peak is suggested for such an estimate by:

+f

=

300

X

(count rate in 691-keV peak) , nf/cm2- s detector volume in cm3

so that a quick determination of the photopeal! area can indicate whether further exposure of the detecictr to the neutron environment is advisable. However, this rough estimate does not yield unceriaint:i nor the contributions from neutrons with energies other than fast range (1 5 En< 10 MeV). In more rccsnt seven HPGe detectors with various active volumes using their prompt photopeaks enlitled from Ge(n,X) reactions are reported. These detectors are irradiated by neutrons with various energies. The semiconducting detectors emit prompt photopeaks at 596 and 691 keV from th- Gc(n,r) and Ge(n,nlr) reactions, respectively. The count rate of these index photopeaks is used to monitor the ncutron flux, resulting in a detection range of 2 to 3000 n/cm2 . s for thermal and epithermal neutrons ar:d 0.7 to 70 n/cm2 . s for fast neutrons. Advantages of using the HPGe detector as a neutron momtor bshile sewing primarily as a prompt y-ray spectrometer in a mixed neutronly-field are given below. The HPGe detector is widely used in the PGAA facility, however, facing constant, low-flux ncutron bombardment. Neutrons with various energies may interact with the germanium and induce sonic ptompt y-rays ir. the spectrum of interest. The 596-keV prompt y-rays from 73Ge(n,r)and '"Ge(n,nlr) reactions are identified and often used as signatures in the y-ray spectrum for the presence of neutrons in -yfield^.'^*'*'.^ In this section, a quantitative on-line detection of both thermal and fast neutron flux impinging on the HPGe detector is described on the basis of these preliminary investigations.

A. USING AN HPGe DETECTOR AS A FAST NEUTRON MONITOR To measure the prompt y-ray response of the HPGe detector for fast neutrons, each of the seven HPGe detectors with a crystal volume of 58 to 114 cm3 is irradiated in turn by a 2S2Cfneutron source in a calibration room. The neutron emission rate is calibrated using a long-counter coupled to a 226Ra/Be standard source. The observed neutron energy spectrum, at the surface of a "2Cf capsule with an energy range of 0.01 to 15 MeV, is adopted for flux conversion.

(s)PMT

113

511

109

lQ07

1205

Gamma Ray Energy, keV Figure 10 Qualitative identification of induced radionuclides 570 d after neutron bombardment in (A) the EGO crystal and packing materials, and (5)the associated photomultipliertube with housing, observed in the high-resolution y-ray spectra taken by the high-purity germanium detector. (Reprinted with permission f'rom Appl. Radiat. Isot. 42(8), Lee, C. J. and Chung, C., Performance of gamma ray measurement using EGO detector after thermal neutron bombardment, 729, Copyright 1991, Pergamon Press.)

GAMMA RAY ENERGY

keV

Figure 1 1 y-Ray spectrum, taken by a high-resolution germanium detector, of a 9 detector 3 h after the 30-min exposure to a thermal neutron flux of 30 n,,/crn2 s.

X

10" annular Nal(TI)

Figure 12 Excerpt of a 6800-s

I .

I

1

500

700

1

I

800

1100

Gamma Ray Energy, keV

prompt y-ray spectrum nieasured by a 10% n-type HPGe detector in front of an unmoderated 252Cffast neutron source with 4, = 70 n/crn2 . s. Photopeaks emitted from the Ge (n,nl r ) and the (n,r) reaciions (underlined) are indicated. (Reprinted with permission from Nucl. Instr. Meth. A301, Chung, C. and Chen, Y R., Application of germanium detector as a low flux neutron rnonitos 328, Copyright 1991, North-Holland Physics.)

Each HPGe detector is placed at various distances away from the 252Cfneutron source while the prompt y-ray spectrum is collected. Since the spectrometric system dead-time is above 25% at sourceto-detector distances less than 300 cm (corresponding to a fast neutron flux >70 nf/cm2. s), the prompt y-ray spectrum can only be collected at a greater source-to-detector distance. A typical prompt y-ray spectrum collected 300 cm away from the 252Cfsource is shown in Figure 12. The prompt photopeaks of the Ge(n,nlr) and Ge(n,r) reactions are indicated in the figure. Major prompt y-rays of 563, 596, 691, 834, and 1040 keV in the Ge(n,nlr) reactions are broadened in the spectrum illustrated in the figure; this is due to the additional contribution from the recoil energy of inelastically scattered germanium. Some sharp, high-resolution photopealq at 500, 596, 609, and 868 keV also exist in the prompt spectrum, indicating the contribution from the capture reaction of Ge(n,r) with various neutron energies. The unmoderated neuron flux at a distance X cm from a point source of 252Cfwith neutron strength S nds, can be expressed as:

+f

where C represents the macrosc&'c removal cross-section in air. Using the Maxwell-Boltzmann distribution of fission neutrons emitted from 252Cfwith En = 1.416 MeV, the count rate of the 691.2-keV prompt photopeaks from the Ge(n,ntr) reaction at various source-to-detector distances can be correlated to a fast neutron flux; the correlation between count rate R(691 keV) and fast neutron flux +c is shown in Figure 13. Due to high spectrometric system dead-time, the 691.2-keV photopeak is washed out by the Compton continuum at >> 70 nf/cm2 s; this is considered the upper limit of fast neutron detection using the HPGe detector. When an HPGe detector is used for y-ray spectrometry in PGAA measurement, a mixed neutron field with neut~onenergy spread up to 9 orders of magnitude is usually encountered. For the IiPGc detector used as neutron monitor, the neutron response is directly proportional to the product of the isotopic abundance of the target nuclide and the reaction cross-section leading to both 596- and 691keV prompt photopeaks. There are, however, no absolute measurements of these (n,r) axi (n,ntr) reaction cross-sehlons; only the relative cross-sections of 72Ge(n,n'r)and 73Ge(n,r), together with the 74Ge(n,ntr)reactions, are reported.23These cross-sections are multiplied by the isotopic abundance as the relative neutron response, and the results, represented by prompt y-response, are plotted in Figure 14 as a function of neutron energy. The neutron response of 72Ge(n,n'r) leading to the emission of the 691-keV prompt photopeak has a reaction threshold at 0.7 MeV, leveling off above the threshold. Therefore, it is an excellent choice to use the 691-keV prompt photopeak to monitor fast and relativistic neutron flux. An empirical estimate of the neutron flux with an energy range of 0.7 to 15 MeV can be obtained from the y-ray count rate [R(691 keV)] in cps, weighted from each HPGe detector performance such as the one shown in Figure 13, and detector active volume (V in cm3) as follows:

+,

.

Flgure 13 Count rate of 691.2-keV prompt y-ray using a 10% n-type HPGe detector in front of an unmoderated 252Cf neutron source with a fast neutron range of 0.7 to 70 n,/cm2 s. (Reprinted with permission from Nucl. Instr. Meth. A301, Chung, C. and Chen, Y. R., Application of germanium detector as a low flux neutron monitor, 328, Copyright 1091, NorthHolland Physics.)

Fast Neutron Flux, nf/cn?-s

I-

$c

5-

0

a

2-

E E

m C3

1 7

sE

Figure 14 The relative neutron response of

o,

Ge(n,X) reactions represented by the relative prompt y-ray intensity as a function of neutron energy. (Reprintedwith permission from Nucl. Instr. Meth. A3Ol. Chung, C. and Chen, Y. R., Application of germanium detector as a low flux neutron monitor, 328, Copyright 1991, North-Holland Physics.)

a

.= +0

2

-

Ul

r

-

7 2Ge(n, rrr)

091 keV

.2 -

~h.~rnsl-~~~itl~.rmal--~lnl.~.-~f~1~-~.l~~~~1sllc-

.I-= f

l

-8

l

l

l

l

l

l

l

l

l

l

0 2 -6 -4 -2 Log (Neutron Energy), MeV

l

l

4

Here, the uncertainty of bfincludes the standard deviation of the weighted average values among the least-squares fits for each detector's R(691 keV), as well as the error due to anisotropic neutron emission from the encapsulated 252Cfneutron source. The fast neutron flux impinging on the I-IPGe detector used in PGAA measurement can be readily assessed by the count rate of the 691-keV photopeak using Equation 3. A comparison with those estimated by Equation 1 reveals results that are 60% more at 0.1 cps, and 150% more at 7 cps, as

Figure 15 Fast neutron flux in n-type HPGe detectors with varit

0.1

I

I

I

I , ,

0.5 R(691 keV I.

I

1

cps

I

I

I

5

I ,

ous relative efficiency at different count rates of the 691-keV prompt y-ray peak using Equations 1 (Ref. 21) and 3 (Hof. 22).

shown in Figure 15. This is due mainly to the more effective fast neutron removai by larger detector volume; at extremely low count rate in large HPGe crystals, the estimati~nsof fast neutron flux usi~rg both Equations 1 and 3 may converge with each other.

B. USING AN HPGe DETECTOR AS A THERMAL NEUTRON MONITOR In the investigation of semiconductor response to the thermal neutron bombasdnient, thc 1-TX3Gcdctirctors are first placed in the thermal column of the THOR facility. The flux of therrnalizcd neutrons, o:ig~ri;iic~! from the THOR core and thermalized in a graphite pile, is adjusted within thz r.ulge of 1 io 2(10,0W n,Jcm2 . s by tuning the stainless steel neutron shutter. Indium foils weighing around C.4 g each i l w used to monitor the relative neutron flux, while thermolurninescent dosimeters TLD-600 and TLD-70) are used to cross-check the absolute thermal flux, which in turn is calibrated by the effective neutron temperature at the thermal column. A typical prompt y-ray spectrum taken by a 10% rz-type KPCe detector is shown in Figure 16. Some i-ltense prompt y-rays emitted from the Ge(n*,r) reactions are clearly observed; however, the 691-keV photopeak is not presented, indicating no fast componeni irr the neutron field. During each prompt y-ray measurement, sets of flux monitoring devices, incl~~dlllg two In-foils and two pairs of TLD-600/700 dosimeters are attached to each side of the cylindrical W G e detector. The absolute thermal neutron flux is obtained using the average Ii6In aclivily in In-foil arid the associated reaction cross-section corrected by the effective neutron temperature. The 95.6%-enriched 6LiF in the TLD-600 is chosen because of its high thermal neutron sensiiivitv. Since hie radiation mixed field in the thermal column has a high y-component, the response of the 99.99%-enriched 'LiF in the TLD-700 is used to subtract the y-ray component in the TL-600 dosimeter. The responses of these irradiated TLD-6001'700 dosimeters as well as the control group are mcasured with a TLD reader. The ratio of the TLD-600 response to that of TLD-700at various neutron fluxes monitored by In-foils is shown in Figure 17. Below a flux intensity of 600 n&m2 . s, the response of TLD-600 relative to that of TLD-700 levels off, indicating its neutron detecting threshold. The response of the TLD-600 rapidly increases as the neutron flux increases beyond the threshold intensity. Since both the TLD-600 and TLD-700 dosimeters are sensitive to y-rays, their response ratio below the detection thresho!d is 0.692 -C 0.025, as illustrated in Figure 17. The HPGe detector response to thermalized neutrons can be correlated to the flux using the net response of the TLD-600 dosimeter down to 600 n,,,/cm2 S, or by using the In-foil monitor further down to 2 n,,,/cm2 S. Results for the count rate of the 596.4-keV prompt photopeak of the Ge(n*,r) reaction are plotted as the monitored thermal neutron flux in Figure 18. In the experimental set-up, the detection limit of' the 596-keV photopeak count rate is 0.1 cps, corresponding to 2 n&m2 . s; 011 the other hand, the

Gamma Ray Energy, keV

ure 16 Excerpt of a 2000-s prompt y-ray spectrum measured by a 10% n-type HPGe detector in the rmal column of the THOR facility, with cbul = 10 n/cm2 s. Photopeaks emitted from Ge(n,, r) reactions %indicated.(Reprinted with permission from Nucl. lnstr: Meth. A301, Ghung, C. and Chen, Y R., cation of germanium detector as a low flux neutron monitor, 328, Copyright 1991, North-Holland ics.)

.

Thermal Neutron Flux, nth/&*s jigure 17 Thermal neutron flux in the thermal column of the THOR facility monitored by In-foils with espect to the response ratio of TLD-600/700 reading. (Reprinted with permission from Nucl. Instr. Meth. WM, Chung C. and Chen, Y R., Application of germanium detector as a low flux neutron monitor, 328, 2opyright 1991, North-Holland Physics.)

'PPer limit is restricted by the spectrometric system dead-time loss of 2590 due to intense interference jf Prompt y-rays induced by the (nh,r) reactions with surrounding materials, and corresponding to 3000 @cm2 P i - s. The feasibility of using the HPGe detector as o convenient thermal neutron monitor while re~orrninga y-spectrum measurement is demonstrated in an ultralow neutron flux with range of 2 to $9,900n,,,/cm2 S. - In a complex neutron field presenting both thermal and nonthermal neutrons, as frequently encountered !lthe PGAA measurement, the fast neutrons rnay be scattered and slowed clown in air; some moderated crystal in the PGAA ieutmns with cpithermal and thermal energies rnay interact with the iet-u~.The count rate of the 5964-keV prompt y-ray at the surface of the encapsulated 252Cf neutron

100 1

a

50

I " ,

-

I

I

/

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/

.- 10 U

0

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a*.

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2m

/.

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d

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0.2

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Figure 18 Count rate of the 596.4keV prompt y-ray using a 10% n-type

-,

HPGe detector in a thermal neutron field at the range of 10 to 3000 n,J cm2 s. (Reprinted with permission from Nucl. Instr. Meth. A301, Chung, C. and Chen, Y R., Application of germanium detector as a low flux neutron monitor, 328, Copyright 1991, North-Holland Physics.)

+ +y

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0

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I

I

10

I

,

....I

ia2

103

Thermal Neutron Flux, n,Jcma-s

source therefore represents only the fast capture reaction of Ge(nf,r) without the component of neutrons with En < 10 keV. The count rate ratio of 596-keV to that of 691-keV prompt y-rays at various sourceto-detector distances can be measured, and a ratio of R(596 keV)/R(691 keV) = 0.44 -C 0.04 can be extrapolated back to the surface of the 252Cfneutron source. This ratio represents the fast neutron component that leads to the emission of 596-keV prompt y-ray by the 73Ge(n,r) reaction. The factors that determine the sensitivity of counting the 596-keV photopeak are difficult' io predict by means of theoretical calculations. The count rate R(596 keV) is given by Equation 4:

where 1,

= = = =

32 prompt y-rays of 596 keV per 100 neutrons capture position-dependent detecting efficiency of 596-keV y-ray neutron flux at neutron enzrgy E 73Ge(n,r) reaction cross-section at neutron energy E

(r) 4, (E) a (E) The neutron flux is not uniformly distributed within the crystal due to self-absorption and scattering and evidently varies with detector size, geometry, and surrounding matcrials. It is expected that with the same count rate on the 596-keV peak, the thermal neutron flux is higher for a large-volume HPGe crystal than the small-volume ones. The ~ ( r is ) position dependent within the detector, in which tllc HPGe crystal is the 596-keV y-source as well as the detector itself. The prompt y response N(73Ge)a(E)1-t, shown in Figure 14, is also encrgy dependent and difficult to resolve. As a result, an empirical formula based upon the experimental count rate of K(596 keV) is uscd to estimate the thermal and epithermal neutrons flux impinging on the HPGe detector with active volume V: E

+(E, < 0.7 MeV) = 25.5 X exp(2.52 X log(61N X [R(596 keV)

(5)

Here again, the uncertainty of neutron flux is derived from the standard deviation of the weighted average values of R(596 keV) among different HPGe detectors; and the 596-keV y-rays induced by fast neutron-induced 73Ge(nf,r)have been excluded. Thus, in a relatively short counting period, the thermal neutron flux around the HPGe detector can be readily determined for the purpose of damage control.

V. DISCUSSION y-Ray spectrometers utilizing scintillation and semiconducting detectors are widely adopted around a nuclear facility, especially for in-beam measurements. When these detectors and associated electronic

inswments are used in the PGAA set-up, neutrons with various energies, spread over 9 orders of magnitude from lo-' to 10MeV, may penetrate the neutron shield, scatter into the detector, and eventually interact with the detector material and associated instruments. Although the neutron flux impinging on fie detector is less than 1000 n/cm2 s and usually around 1 to 100 n/cm2 s, much less than the flux impinging on the sample in PGAA irradiation, it nevertheless induces unnecessary nuclear reactions that in turn emit prompt and decayed y-rays. Hence, neutron and y-shields should be carefully selected to surround the detector and instruments in the best possible manner. In order to collect the prompt y-rays effectively, the y-ray spectrometer in PGAA measurement is psitimed near the irradiated sample; however, the shorter sample-to-detector distance also makes the detector face higher flux of scattering neutrons; this is further complicated by the constraint of limited space available for inserting a radiation shield between the neutron beam and the detector. Both the NaI(Tl) and BGO scintillation detectors, as well as the high-resolution HPGe semiconducting detector, are subject to neutron damage after prolonged irradiation by neutrons. With accumulated neutron fluence to the level of lo9 nn,/cm2,the HPGe detector reveals the deterioration of its highcharacteristics. With accumulated neutron fluence to the level of 1015nh/cm2, the scintillation detector becomes totally useless due to the high intensity of induced radioactivity in the detector crystal, detector can, PMT, and the tube base. In order to extend the service life of the y-ray spectrometer in PGAA measurement, the neutron exposure of the detector should be planned with maximum neutron protection. For instance, if the detector is expected to have a 10-year service life, the neutron flux impinging on the HPGe detector should be less than 3 nf/crn2 s for continuous service in PGAA scan. To avoid overexposure in a neutron field, the y-ray spectrometer in PGAA measurement is preferably monitored on-line about the neutron flux therein. An ideal neutron monitor should be easy to handle in rugged conditions, have a fast readout, be free from interference of other radiations, fading-free, nontoxic, and inexpensive. Although each of the neutron dosimeter, counter, and survey instruments currently used can fulfill some of these requirements, none of them can satisfy all. The nontoxic HPGe detector used as a neutron monitor takes advantage of its application as a y-ray spectrometer in a mixed neutrody-field by providing information to distinguish prompt y-rays emitted form the Ge(n,X) reactions that can be characterized reasonably well for the neutron flux measurements for thermal, epitherrnal, and fast neutrons. The major disadvantages in using the HPGe detector as a neutron monitor are, however, its operation in a specific temperature and humidity range as well as its high sensitivity to mechanical shock and vibration. Due to the intense y-interference at high neutron flux, the HPGe detector has its upper limit of detection at about 25% of the spectrometric system dead-time loss in the y-ray spectrometer. The upper s and 70 nficm2 s in thermal and fast neutron fields, respectively. In field limit is 3000 n&m2 applications, however, one should be careful to interpret the 596- and 691-keV photopeaks emitted from Ge(n,X) reactions. When the HPGe detector is used as a y-ray spectrometer near a neutron beam, it is heavily shielded by neutron-moderating and -absorbing materials; such an arrangement may greatly reduce the applicability of the HPGe detector as a neutron monitor and underestimate the neutron flux outside where an NaI(TI) or BGO detector shield is employed. In addition, ferrous material is frequently used in an experimental set-up around a nuclear facility, its intense y-ray emitted from the Fe(n,r) reaction, five 692.1-keV prompt y-rays emitted in every 100 thermal capture reactions, and may interfere with the 691.2-keV prompt photopeaks emitted from the 72Ge(n,n'r)reaction; therefore the fast neutron flux may be overestimated. One of the disadvantages in using PGAA, as compared to the conventional INAA; is the exposure of the y-ray spectrometer to scattering neutrons. In this chapter, the neutron flux distribution around the prompt y-ray spectrometer in the PGAA set-up, the inflicted damage on the detector due to prolonged neutron bombardments, and using prompt y-rays emitted from the Ge(n,X) reactions to monitor lowflux neutrons on-line are briefly reviewed. With proper arrangement of neutrody-shields around the Prompt y-ray spectrometer and timely acquisition of neutron ilux data using an HPGe detector while collecting prompt y-rays emitted from the sample, the service life of the spectrometer in PGAA measurement can be properly extended and the inflicted damage can be kept to a minimum.

-

-

REFERENCES 1. Chung, C.,Yuan, L.J., Chtn, K. D., Weng, P. S., Chang, P. S., and Ha, Y. H., A feasibility study of the in vivo prompt gamma activation analysis using n mobile nuclear reactor, Appl. Rndiat. Isot. 36(5), 357, 1985.

2. Chung, C., Yuan, L. J., and Chen, K. B., Performance of a HPGe-NaI(T1) Compton suppression spectrometer in high-levcl radioenvironmental studies, Nucl. Itistr. Meth. A243, 102, 1986. 3. Chung, C. and Yuan, L. J., PGAA using beam from THOR facilily, in Proc. 1st Asian Symp. Research Reactors, Institute for Atomic Energy, Rikkyo University, Tokyo, 1986, 310. 4. Failey, M. P., Anderson, D. L., Zollar, W. H., Gordon, G. E., and Lindstrom, R. M., Neutron-capture PGAA for multielement determination in complex samples, Atzal. Chem. 51(13), 2209. 1979. 5. Chung, C., In vivo partial body activation analysis using filtered neutron beam, Appl. Radiat; Isor. 39(2), 93, 1988. 6. Chung, C. and Chen, C. Y., Modification of a mobile nuclear reactor for medical diagnosis, Nucl. Tech., 92, 159, 1990. 7. Chung, C. and Beng, T. C., In situ PGAA of water pollutants using shallow Wf-HPGe probe, Nucl. Instr: Meth. A267, 223, 1988. 8. Ellis, K. J., In vivo neutron activation analysis, in Activation Analysis, 11, Alfassi, Z . B., Ed., CKC Press, Boca Raton, FL, 1990, 407. 9. Chung, C., Liu, S. M., Chao, J. H., and Chan, C. C., Feasibility study of explosive detection for airport security using a neutron source, Appl. Radiat. Isot. 44(12), 1425, 1993. 10. Lee, C. J. and Chung, C., Performance of gamma ray measurement using BGO detector after thennal neutron bombardment, Appl. Radiat. Isor. 42(8), 729, 1991. 11. Shea, P., Gozani, T., and Bororgmanesh, H., A TNA explosive-detection system in airline baggage, Nucl. Instr: Meth. ,1299, 444, 1990. 12. Larrson, L., Alpsten, M., and Mattson, S., In vivo analysis of nitrogen using a *'Cf source, in Proc. 7tt-1 Int. ConJ Modem Trend Activation Analysis, Isotope Division, Riso National Lab, Roskilde, Denmark, 1986, 1425. 13. Knoll, G . F., Radiation Detection and Metxurement, 2nd ed., John Wiley & Sons, New York, 1989, chap. 11 and 12. 14. Lederer, C. M. and Shirley, V. S., Table of isotopes, 7th ed., John Wiley & Sons, New York, 1978. 15. Chung, C. and Chen, Y. R., Application of a germanium detector as a low flux neutron monitor, Nucl. Insic Meth. A301, 328, 1991. . 16. Lone, M. A., Leavitt, R. A., and Harrison, D. A., Prompt gamma rays from thermal neutron capture, Ar. Data Nucl. U i t a Tables, 26, 5 11, 1981. 17. Hausser, O., Lone, M. A., Alexander, T. K., Kushneriuk, S.A., and Gascon, J., The prompt response of bismuth germanate and NaI(T1) scintillati~t~ detectors to fast neutrons, Nucl. Instr. Merh. 213, 3 10, 1983. 18. Kobayshi, M., Kondo, K., Hirabayashi, H., Kurodawa, S., Taino, M., and Yamamoto, A., Radint~on damage of BGO crystals due to low energy gamma rays, high energy protons, and fast neutrons, Nucl. I t ~ s r r . Meth. 206, 107, 1983. 19. Bieler, Ch., Burkart, D., Marks, J., Ricbesell, M., Spitzer, H., Wittenburg, K., and Winter, G., Radratmn damage of BGO and CsI(T1) crystals, I ' J ~ Instr. . Meth. A234, 435, 1985. 20. Kubota,S., Motobayashi, T., Ogiwara, M., Idurakami, H., Ando, Y., Ruan, J., Shirato, S., and ~Vurakarnl, T., Response of BaF2, BaF2-plastic,and BGO scintillators to neutrons with energies between 15 and 45 MeV, Nucl. Instr. Merh. A285, 436, 1989. 21. Bell, R. A. I., Table for Calibration of Radiation Detectors, Australian National Universiiy Kcport ANU-PI 606, Australia, 1974. 22. Chno, J. H. and Chung, C., Low level neutron monitoring by use of germanium detectors, N d . Kn..iir: Afeth. A321, 535, 1992. 23. Gorber, D. I. and Kinsey, R. R., Neu:rov Cross Section, Vol. 11, BNL Report-325, Upton, NY, 1976, 193.

prompt Gamma Neutron Activation Analysis with t~eaactorNeutrons Zeer/ B. Alfassl

i

:-

I. .;Jntroduction .................................................................................................................................59 .-'

n.

.

m.

Techniques with PGNAA with Nuclear Reactors ................................................................ A. Filters and Collimators ......................................................................................................... B. Detection ............................................................................................................................... Practical Applications .................................................................................................................

references .............................................................................................................................................

$3

59 62 62

65 71

.Delayed Gamma Neutron Activation Analysis (DGNAA), in which the y-activity induced in the activated &hPle (by irradiation with neutrons) is measured after the end of irradiation, is performed mainly with nuclear reactors. On the other hand, PGNAA (Prompt Gamma Neutron Activation Analysis) is performed {mainly with other neutron sources such as neutron generators or neutron isotopic sources, either 9Be (a,n) I2C (commonly nowadays with 24'Amas the or-source) or a fission source such as 252Cf.There is -a reason for this difference. DGNAA is used for the analysis of very small concentrations of trace '&rnents; and for some elements, in spite of various new analytical techniques such as ICP-MS, it is &6 most sensitive method. Although a comparison of theoretical sensitivity limits shows that PGNAA is 'in most cases more sensitive than DGNAA (Isenhour and Morrison' showed this to be true for 61 bof.63 elements they calculated for), it is true only if an equal neutron flux and equal sample-detector distance are used. However, in PGNAA, since the measurements are done simultaneously with the irradiation, usually the samples are either put further from the reactor core leading to lower neutron :fluxes in PGNAA or the samples are close to the reactor core but the sample-detector distance is much laiger, thus resulting in lower count rates. The resulting consequence is that PGNAA is not used in 'most cases to determine the concentrations of the trace elements in the measured sample, but rather is -used in medium-concentration or even major elements. These measurements are usually required to be done in situ, excluding the possibility to use a nuclear reactor, although there is one mobile reactor used for this p u ~ p o s e In . ~ most cases, this work is done with an isotopic (a,n) neutron source or 252Cf fission source or a neutron generator, yet there are some elements for which PGNAA is the most sensitive nondestructive analytic method, and they are determined mainly using nuclear reactors.

11. TECHNIQUES WITH PGNAA W1TH NUCLEAR REACTORS A problem in the use of PGNAA, especizlly with neutrons coming from a nuclear reactor, is the necessity to differentiate the prompt y-rays emitted by the sample from y-rays coming from other sources. There are various sources for y-photons with a nuclear reactor: (1) prompt y-rays emitted from the surrounding -materials and the detectors; (2) y-rays due to inelastic scattering of the small fraction of fast neutrons; :and (3) reactor y-rays scattered by both the sample and the surroundings. Hammermesh and Hm-me13 ,solved this problem by taking four different spectra using various combinations of the sainple and a ;cadmium shutter in and out of the neutron Learn. The cadmium shutter absorbs thermal neutrons while ;transmitting fast neutrons and reactor y-rays. This approach, however, ignores the build-up of delayed ;Y-radiation and will be correct only if the half-lives of the activated nuclides in the sample are considerably :longer than.,the counting interval. Another disadvantage is that it assumes stable neutron flux, which 'is not always true. In order to circumvent these two disadvantages. it is necessary to make short measurements. Isenhour and Morrison' accomplished this by performing a series of measurements of ' Short duration. They used modulation of a collinmted neutron beam using a high-speed thermal neutron

-

'&93-5 149-91951SO 00+$.50 '995 by CRC ~ r c s s ~. n c .

59

chopper (Cd absorber on an iron ring), synchronized with a multichannel analyzer, thereby allowing recording of separate spectra for the two half-cycles of thc chopper. The cl-ropper was turning at about 1000 cps, allowing better than 99% cancellation of delayed y-rays from reactions with half-lives longer than 1 ms. Tney studied, for example, the PGNAA of a vanadium sample, with and without the modulated synchronized half-cycle chopper.*Withoutmodulation, the 1.434-MeV peak of the 3.75-mnin 52V[ W (n,y) 52V]is clearly seen in the spectra, whereas it is completely suppressed when the spectra is taken with the modulation, indicating that the modulation cut the delayed y-radiation. The cycle duration is assigced as 27, of which exactly half is open (when the thermal neutrons are impinging on the sample) and exactly half is closed (the thermal neutrons from the reactor are absorbed by the Cd chopper). Let us assign the number of delsyed y-photons emitted during the open half-cycle of tile i' cycle as Oi and that of the closed half-cycle to Ci. In the first open half-cycle, the radionuciides are formed and disintegrated simultaneously, leading to the number of disintegration during this halfcycle, 0,:

where R is the rate of production of the radionuclides emitting the delayed y-rays and N, is the number of atoms in saturation (R = h - N, = N . a - +). O1( 0 stands for the open half-cycle) is equal to the number of radionuclides formed, minus the one left at the end of irradiation. CI (C stands for the close half-cycle) is the number of decays during time T of the second half-cycle due to decay of those remaining in the end of the first half-cycle.

In the second cycle, there are delayed y-rays due both lo the radionuclides formed in the second cycle and those left from the first cycle. The number of y-rays from nuclides formed in rhe second nuclide are equal to that calculated for the first cycle (O1). The disintegration of radionuclides left from the first cycle is equal to C, multiplied by the decay factor (e+ for 0, and e-A'2T for C,):

Similarly, for the following cycles

The net y-ray spectra is the difference between the spectra of the two half-cycles. Thus, the net yphotons of decayed emission Ni is given for the n cycle by

Looking at the equation for C,, it is a sum of a geometric series. Thus,

For large n such that 2(n - AT >> 1 then e-2(n-')AT + 0 substituting 1 - e-2AT= (1 e-;"') the equation for D, is transformed to

-

e-A') (1 4-

The amount of prompt y-rays produced in each open half-cycle is R ratio of delayed to prompt y-rays on the cycle n R, is given by

-

T =

A

. N, .

,T. Thus, the

For R, to be lower than 0.01, XT should be lower than 0.35 or 2r < Tin. They used a chopper that can work up to 1000 cps, allowing better than 99% cancellation of the delayed y-rays from nuclides with half-lives longer than 1 ms. For most of the radionuclides, XT << 1 and eX7can be approximated by 1 h ~leading , to R, -- XV2. Lombard et aL4 used a chopper with a more useful frequency range. However, the use of Ge(Li) detectors enables the differentiation between y-rays from prompt emission and those from delayed emission due to their different energies. None of the modem set-ups of PGNAA is using this absorption modulation; nevertheless, in cases of too complex spectra, the use of a modulation technique gives a simpler spectra although requiring longer analysis time. Usually, the neutron beams used for PGNAA have a flux of neutrons Iower by 106 than in the core. Higher external thermal neutron fluxes are available by the use of a curved neutron guide. A special chapter in this book is devoted to "cold neutrons guides" and only a short description will be given here. Henkelman and BornS described the curved neutron guide of the high-flux reactor in Grenoble, France. The curved neutron guides are made from plates of glass thinly coated with nickel (-1000 A), forming a channel of rectangular internal cross-section 3 X 20 cm2.The thermal neutrons are propagated in the guide by total reflection on the walls and are separated from the reactor y-rays and epithermal and fast neutrons by the curvature of the neutron guide. This neutron guide usually leads to a 104 degradation factor in the flux. as compared to 106 without neutron guide, this in addition to the separation from the other y-rays and fast neutrons. Neutron guides are operating usually with "cold neutrons", which are neutrons passing through liquid He or liquid DZ,or through carbon cooled with liquid H2. (Liquid Hzis not used for cooling due to its high cross-section for absorption.) The use of subthermal neutrons leads both to higher transmission through the neutron guide and to higher (n,y) cross-sections. However, it should be remembered that the instrumentation involved is very expensive and only very few laboratories can afford them. The use of "cold neutrons" and neutron guides are described in detail in another chapter of this book. Henkelman and Born differentiate the measurement of the y-rays in PGNAA to low- and highenergy y-rays by the use of different detectors. Since the sample is placed quite far from the detector (17 cm), it should be possible to use simultaneously the two detectors. For low-energy y-rays, they used a large Ge(Li) detector to increase the efficiency. For the high-energy y-rays, they used a small Ge(Li) detector in order to reduce the three peaks from each y-line (photopeak, single escape, and double escape) to a single peak. With a small detector, high-energy y-rays lead only to the doubleescape peaks. However, this simpler spectra is on expense of lower efficiency. Gladney and co-workersb9 used the same relatively large Ge(Li) detector for both energy ranges. However, the NaX annulus detector surrounding the Ge(Li) was used in an anticoincidence mode for suppression of Compton scattering in the low-energy range; whereas, for the high-energy range, the NaI detector was used in the coincidence mode to ascertain that only double-escape peaks were monitored. Only if 1022-keV (= 2 X 511 keV) y-rays are detected in the Nd annulus, the coincidence unit opens the gate for the signal from the Ge(Li) detector. This method eliminates virtually all the photopeaks and single-escape peaks, together with the Compton scattering; however, it, leads to a Iower rate of counting. For relatively simple spectra, a bare Ge detector may be sufficient and more sensitive. Since the detector cannot be put together with the sample inside the reactor core where there is high flux of neutrons, there are two possible arrangements. One can either place the sample inside the reactor, where the flux is large, and 'observe the y-rays outside with a large sample-detector distance (internal set-up), or bring a neutron beam outside to a sample, mounted close to the detector (external set-up). The tint possibiIity has high neutron flux but poor counting geometry, whereas the second possibility has just the opposite. Gladney and c o - ~ o r k e r s ~ ~ used the first possibility, while Henkelmannt0and Failey et al." preferred the second choice. Faiiey et

+

al." summarized the advantages of the second method. From an efficiency point of view: the flux and geometry of counting balance each other. They gave four advantages of the external set-up. ( I ) In the intemal set-up, the samples are subjected to mating, large radiation damage, and there is l&e residual radioactivity in the sample; whereas, in the external method, there is no heating and very little radiation damage or induced residual radioactivity, which are important for nondestructive testing. (2) Large and fragile samples thnt cannot be inserted i f 1 1the ~ core can be measured with the external set-up. (3) The extcrnal set-up aIlows the use of filters to remove epithermal and fast neutrons or y-rays. There is also the possibility of adding an extra moderator to thermalize the neutron beam. (4) There is also the possibility of uslng more than one detector, either for coincidence measurements or for different energy ranges. The fourth advantage is not true, and several detectors can be used also for the internai set-up. Systems that were built especially for PGNAA purposes are always of the external set-up type. The internal set-ups are usually those that were constructed to perform nuclear structure investigations and used later for anaiytical applications.

A. FILTERS AND COLLIMATORS When using the external set-up, it is best to put inside the beam, before the sample, filters for epitllermal and fast neutrons and for y-rays from the reactor. A survey of materials for the filtersi2suggested the use of single crystals of bismuth or quartz having random orientation to the beam, due to their low cross-section for thermal neutron absorption. Bismuth crystal was found to be the best filter material due to its enhanced absorption of y-rays. Lombard et alhtudied the attenuation of thermal neutrons, fast neutrons, and y-rays by sir;gle crystals of quartz and ordinary (not single crystals) lead and bismuth. It was found that the attenuation coefficient (cm-I) is about the same for fast ne:ltrons, while much lower in the case of quartz for thermal neutrons and y-rays. Thus, using the same thickness of the three filters will lead to the same fast neutron interference, but in the case of Bi or Pb there will be considerably less thermal neutrons. For the same thermal flux, single-crystal quartz is better than ordinary Pb or Bi for removing fast neutrons but worse for y-ray removal. Single-crystal Pb or Bi will be better from the point of view of fast neutron discrimination, but they are considerably more expensive. To suppress further the reactor y-rays, a precollimator leaci filter was positioned before the quartz filter, with smaller size than the filter. To obtain small-sized beam impinging only on the sample, the lateral neutrons should be absorbcd by a collimator. Lombard et aL4 used 12-in. thick graphite collimators with a 0.5-in. hole through the center. Hanna et al.13 used lead and masonite for the collimation.

B. DETECTION The first experiments were done with an NaI(T1) detector. Immediately after the developmc~itof the Ge(Li) detecto:, they r e p l a ~ e d lthe ~ ' ~NaX(T1) detectors due to their superior resolution although at the expense of lower efficiency. Later, a11 PGNAA systems used a Ge(Li) detector, mostly surroundcd t y an annular NaI(T1) detector or several smaller NaIflI) detectors in an anticoincidence lt~odelo suppress the Compton-scattered interaction. Modern systems are using HPGe detectors (hyper-prc germamurn) and sometimes the NaI(Tl) detector is replaced by n BGO detector (bismuth germanate) duc LO its higher efficiency for y-ray detection, and hence better suppression of Lhe Compton-scaitered cvcilrs However, BGO detectors are more expensive, cspecially the annular one, than Nal(T1) detecto;:, ariu hence are used only in a few systems. The detector is susceptible to damage by either fast or thermal neutrons; thus, the detector should be protected with a material that absorbs neutrons with high cross-section without emitting prompt yrays. The only nuclide fulfilling these requirements is 6Li, which has a very high cross-section (950 b j for (n,a) reaction. '9 has even higher cross-section for (n,cx) reaction, but the (n,a) reaction with l0B leads to excited-state 7Li, which emits a y-ray on de-excitation. 6Li (n,a) reactions produce tritium in the ground state and hence there is no y-emission. Lithium cannot be used as a metal due to its reactivity, and it is usually used as 6LiFdue to the low cross-section of neutron absorption by 1E However, Li2C03 can be used also and even has a lower (n,y) cross-section. At most modem laboratories, the HPGe-NaI(T1) or HPGe-BGO spectrometer is equipped to simultaneously acquire three modes of spectra: single mode, Compton suppression mode (anticoincidence), and pair mode (coincidence with absorption of 1.02 MeV in the annular detector or triple coincidence with two 511-keV photons).18

LEAD

a

CONCRETE BORATE0 PARAFFIN

N a I (Compton Suppr

Flgure 1 Schematic layout of PGNAA facility at MURR (Missouri University Research Reactor). (Reprinted from Nucl. Instr. Meth. 188, Hanna, A. G., Brugger, R. M., and Glascock, M. D., The prompt gamma neutron activation analysis facility at MURR, 619, Copyright 1981, with permission.) The most extensive descriptions of a nuclear reactor-based PGNAA facility was given by Glascock and co-workers,I3describing the system installed at the University of Missouri Research Reactor (MURR)', .and by Anderson et al.,19 describing the system developed at the National Institute of Standards and Technology (NIST, then called NBS). Both systems were designed and constructed optimally for PGNAA for analytical purposes. The beam is extracted horizontally (MURR)or vertically (NIST). The reactor in Missouri has more fast neutrons (the reactor in NIST is D20moderated and it is more thermal in character); to thermalize them, two aluminum cans filled with helium gas at atmospheric pressure are placed in the reactor pool adjacent to the core and through the beryllium and graphite reflectors. (Helium has lower cross-section for neutron capture or scattering than N2 or 02,present in air.) Due to the location of the neutron source at the core face, the initial fast neutron and y-rays associated with the beam is quite high. Four single crystals of silicon (total length, 50 cm) were placed deep inside the beamport to filter the fast neutrons and y-rays. Although cooled Si crystal will transfer better the thermal neutrons, it is not feasible physically due to space problems. Two collimators made of lead and masonite are used at MURR to form a relatively narrow beam of neutrons (diameter of 6.5 cm with t 5% uniformity). At NIST, the collimators are constructed from lead and stainless steel for y-absorption and from borated polyethylene and 6Li2C0,-Ioadedpolystyrene for thermal neutron absorption. The detection system consists of a Ge(Li) detector placed at the center of a cylindrical NaI annulus (24-cm outside diameter and 32-cm length). In the case of NIST, they used a split annulus NaI detector to enable the measurement of triple coincidence. The systems used several ADCs to measure single, Comptonsuppressed, and double-escaped spectra. Figure 1 gives the schematic layout of the facility at MURR. Figure 2 gives the detection system and the shielding of the NIST facility. Anderson et al. summarized the data for some reactor-based PGNAA facilities. A portion of their table is given in Table 1. The mass of an element cannot be measured from its capture y-spectrum by a simple comparison with a low-mass, highly concentrated standard chemical compound of this element subjected to the same neutron flux. Due to the size of the samples, it is practically impossible in many cases to obtain -.a homogeneous irradiation with a directional flux of thermal neutrons. For example, it was found for simulated biological tissues that the LIterrnal neutron flux is attenuated by a factor of 2 per centimeter of tissue penetrated." Comparison of the y measured in a sample and n standard is quantitative only if the major element contents of both are similar and the geometry of both are identienl. The best way

BEAM TUBE AND DETECTOR SYSTEM -Teflon

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Figure 2 Detection system and shielding configuration of the PGNAA facility at NET (National Institute for Standards and Technology): a) vertical view, b) horizontal view. (Reprinted from J. Radioanai. Chem., 63, Anderson, D. L., Failey, M. I?, Zoller, W. H.. Walters, W. B., Gordon, G. E., and Lindstrom, R. M., 97, Copyright 1981, with permission.)

1 Some Reactor-Based PGNAA Facilities

-

6th

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1-0s Alamos -scientific Laboratory, U.S. U. of MarylandNitional Bureau of Standards, U.S. p. of Missouri, U.S. U. of Michigan,

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1000

ratio

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Detectorsample distance (rn)

Sample site or

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25 X 75 mm

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5 X lo8 (beam)

3 x lo7 (beam)

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1.5 X 1Ol0

(beam) Grenoble, France Instituto de Energia, Brazil Instituto Venezolano de Investigations Cientificas, Venezuela Pakistan Institute of Nuclear Science and Technology McMaster University, Canada

3.7 X lot2

(internal)

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4.75 x lo7

(beam)

1.2 X 107

(beam)

a) 3 X

loS

(beam) b) 3 X lot2 (internal)

n.r. = not reported. Taken from Anderson, D. L., Zoller, W. H.,Gordon, G . E., Walters, W. B., and Lindstrom, R. M., Prompt NeutronCapture Gamma-Ray Spectroscopy and Related Topics, 655, Copyright 1981, Conf. Ser. NO. 62, 4th (n,y) Int. Syrnp., Grenoble 1981.

a

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is to use the internal standard method as proposed by Isenhour and Morrison for the determination of boron in boric acid.20 In this method, a known amount of element, absent in the sample, is mixed thoroughly with the sample, and this element is used as internal reference. It is preferable to producing a standard with similar composition to the sample. The ratio between the reference element and the studied elements are measured in mixtures of known amounts of these elements.

!?I. PRACTICAL AQPLICAT1QMS Isenhour and Morrison2' determined boron by PGNAA using an NaI(T1) detector and the modulation technique to subtract the delayed y-activity. Boron is the best suited element for PGNAA since it has

a high cross-section for thermal neutron absorption, but it does not lead to a radionuciide and hence cannot be measured by DGNAA. Although the thermal neutron absorption is a (n,a) reaction rather than a (n,y) one, 92.5% of the 7Li formed in the I0B ( n p ) 'Li reaction is formed in the excited state, which decays with a half-life of 5 X 10-l4 s, and thus it can be looked upon as prompt y. The excited 'Li emits a 477-keV y-line that is used for quantitative measurement of the boron content of the sample. Greenwood" used a small Ge(Li) detector (volume 1 cm3 compared to 100 cm3 and larger Ge detectors used nowadays) for measuring a geological sample by PGNAA. Using a 30-g sample and 10 h of counting, Greenwood could measure zll the major elemental constituents (>0.1%). They measured sodium (2.5%), aluminum (6.7%), silicon (34%), potassium (3.7%), calcium (0.9%), titanium (0.2%), manganese (0.06%), and iron (1.7%). Rasmussen and Hukai2' measured the elemental concentration of coal by PGNAA using a 30-cm3 Ge(Li) detector and 30-g coal samples; each counting was for 25 h. They measured 10 elements in the samples: H, C, Si, N, S, Fe, C1, Ti, Al, and K. For calibration, they used coal sdnyles with known concentrations of these elements (determined by other analytical methods). Lombard et a1.4 measured boron and cadmium using an NaI(T1) detector with modulation technique; however, they did not use real samples and measured only standards. Cd is, beside B, the most st~itable element to be studied by PGNAA. The 'I3Cd (n,y) '14Cd reaction has a huge cross-section of 19896 b. Since Il3Cd has an abundance of 12.22%, it makes the elemental effective cross-section of 2430 barns still a huge cross-section. However, Il4Cd is a stable nuclide and DGNAA is done by other reaclions (isotopes) having much lower cross-section, whereas PGNAA can use this high cross-section reaction. The determination of Cd by PGNAA was reviewed recently by Grazman and Sch~eike1-t.~~ Greenwotrd et al.23 measured, by PGNAA using an NaI(T1) detector, a meteorite and terrestrial ruck samples, tIscb following elements: Fe, Si, H, Ni, Co, Mg, Na, K, Al, S, Gd, and samarium. Elkadi and Guffcy2' measured gold in a synthetic mixture resembling geological samples by PGNfiA using neutrons from a graphite thermal neutron column of a 10-kW pool reactor. They could measure 13 mg Au it1 20-g SiOl samples. Comar et al.I7 used a PGNAA with reactor neutrons to measure biological sarnpks both in vivo and in vitro. They measured H, C1, Na, K, N, S, and B in samples of bone, blood, and hair using the curved neutron guide of the Saclay reactor. Actually, they could not separate B Crom Na since they said that they have the same peak of 477 keV. They could separate B in other tissues having smaller NarB ratios. In order to medsure quantitatively in large biological samples, they used the internal standard method by mixing known amounts of glucose, red mercurous oxide, sodium chloride, and boric oxide. Using this standard, they measured boron, chlorine, and hydrogen in two different plant tissues. They estimated the sensitivity as 1 pg B, 1 mg C1, and 10 to 20 mg H (in a 2-g sample counted for 10 min). They also made in vivo studies of human tibia and measured in 10-min runs the concentralion of Ca and C1. They claimed that, for in vivo studies. PGNAA is preferable to DGNAA, since in DGNAA there is the problem of some of the radioisotopes leaving the irradiated zone during the neutron bombardment, due to their exchangeability in tissues, leading ta different density patterns during the following count~ng. Wiggins et d." measured manganese by PGNAA. Although their studies include the use of re;icinr neutrons as well as neutrons from the 252Cfsource, the real measurements were done only witii the 252Cfsource. Henkelmann and Born: using a nuclear reactor with thermal neutron guides, measured six rare earth elements (La, Nd, Sm, Eu, Gd, and Dy) and thorium in 1-mg monazite samples by 10-min counts using only low-energy y-ray (0 to 800 keV) mezsurements. They also analyzed rock samples (0.5-g sampicj, and by analyzing the high-energy y-rays, they measured Si, Fe, Al, Ca, Na, K, W, Mn,and Ti in one measurement. The lowest concentration was of Ti: 0.10%. Gladney et aL8 used PGNAA with nuclear reactor neutrons to measure trace quantiiies of 13 and Cd in industrial and standard materials. Samples of 0.1 g were placed in the reactor thermal column and the y-rays were measured with a detector 6 m distant through a series of collimators to exclude extraneous radiation. This method provides a rapid nondestructive analysis for >50 mg boron. For most samples studied, 5 to 15 rnin counting was sufficient. However, for materials with lower concentration of B, each sample was measured for several hours. The largest interference for the 477-keV line of B comes from the 472-keV line of Na. This interference is due to the wide peak of B caused by Doppler broadening. Samples with NalB ratios of lo3can be measured accurately; however, at Na/B 1 OS, the precision of B measurements was relatively poor. Repetitive analyses show the method to have a precision of 5%. They found that although Cd can be measured by PGNAA in these materials, the

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measurement is not sufficiently sensitive to provide a practical alternative for atomic absorption analysis except in cases where nondestructive methods are required. Heurtebise and L u b k o w i t ~used ~ ~ a collimated neutron beam from a nuclear reactor and impinged it on 1- to 5-g alloy samples. A flux of 4.75 X lo7 dcm2s, thermal neutrons (Cd ratio, 30.5) were used to measure the contents of several metals in various alloys by 30-min counting. In one steel sample, they measured Fe, Ni, Mn, Cr, and Al. In another steel sample, they measured Fe, Ni, and Co. In synthetic samples, they measured Ni, Cu, Co, Fe, and Ti (one mixture) or Zn, V, and Mn (another mixture) with errors of usually less than 4%. Only for Ni did they find a 9% error from the true composition. Heurtebise et aLn used the same PGNAA system to analyze the metal contents of 12 hydrodesulfurization catalysts. Co and Mo required 10-min counting for sufficient statistics. If Ni was present, they measured an additional 30 min per sample to obtain good counting statistics for the Ni. Titanium oxide was used as an internal standard. Relative standard deviations of 5% were typical for concentrations of 2 to 15%. Heurtebise et aLn applied PGNAA to determine the water content (by measuring the 2.232 MeV 7-line due to hydrogen) simultaneously with the determination of Co, Mo, and Ni in hydrodesulfurization catalysts. They used 0.1 to 1.0 g catalyst, pulverized and mixed with 50 to 100 mg Ti02; in order to use Ti as an internal standard. Samples were irradiated for 10 min with a thermal flux of 4.75 X lo7 n/cm2 s using 96 cm3 Ge(Li) detector at 23 cm from the sample; 10 min was found sufficient to obtain good statistics. This method was found to be the easiest and most rapid way to measure moisture content in the catalysts. It should be mentioned that this method can be used only if the sole compound containing hydrogen is H20.Large contamination with organic materials will be translated into exaggerated H@ content. To check the reproducibility of the method, they analyzed one catalyst eight times and obtained a relative standard deviation of 3.6% for 12.6% water content. Jurney et aL7 used PGNAA with a nuclear reactor to measure sulfur in complex materials. Previous studies for measuring sulfur were performed with a 2"Cf source. They determined sulfur in nine NBS standard reference materials and six standard coals. Comparing their results to measurements done by other methods, they obtained agreement within 1.0 -L 0.04. The standard deviation of repeating results are usually better than 5%. Cakium and potassium are the most likely sources of direct interference due to their close y-lines (S, 841 keV; K, 843 keV, Ca, 837 keV). However, unless the ratios are very small in S, the peaks can be separated. The main background is due to Compton from 2223 keV of H in the polyethylene container of the sample, which causes a detection limit (twice the'standard deviation of the background) of 100-pg S in 3.5-h irradiation and 3 0 pg in 20-min irradiation. As a standard, they used 100 mg elemental S, assuming that the main absorbing material is the polyethylene capsule that is identical for sample and standard. Gladney et aI? used the same system, measuring simulhneously nine elements of major and minor abundances: Si, Al, Fe, Na, K, Ca, Ti, Mg, and P. A1 was measured both by the prompt y-rays and the delayed y-rays. For the delayed y-rays, they started the countings 15 min after the start of irradiation. For each sample, they irradiated 3 h. All samples give results that agree with other analytical methods to better than 5% for Si and Ti and between 5 and 10% for A), Fe, Na, K,and Ca. This worse agreement is due to a single result that falls off the correlation of the others and might be due to larger interferences. Neglecting this outlier, the correlation is better than 5% for all these elements. The correlation for Na, Mg, and P is worse, probably due to the concentration being close to the detection limit. With the exception of Si, Fe, and Al, the results obtained by measuring the double-escape peak in coincidence with 1.02 MeV in the annular NaI(T1) were inferior to those obtained by measuring photopeaks with anticoincidence mode. Gladney and co-workers9 also measured nitrogen, carbon, and hydrogen simultaneously in 1-g samples. For N, they used the 5269- and 5298-keV lines, which were found to be more sensitive than the other lines. Since these lines lie very close together with no discernible interferences, they were summed to yield lower detection limits. For carbon, they used the 4945-keV line since the 3684-keV line has interferences from N. Hydrogen has only one y-line at 2223 keV. Irradiating for 1 h, the following detection limits were obtained: N, 500 ppm; C, 10%; and H, 5 ppm. Using 14 samples of environmental standards the correlation between the measured value and the previously known value was 1.02 + 0.08 for N, 1.02 4 0.04 For C, and 0.97 r+_ 0.07 for H. The same Los qlamos Group (GIadney and co-workers) used its system to measure the concentration of boron in meteorite^.'^ Islam et aL3' measured the content of S, Fe, and Si in n lunm rock sample with a 1-MWswimming Pool reactor. The system was of the internal type, irradiating in the reactor core with thermal flux of

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3 X 1012n/cm2 s and with a Ge(Li) detector far from the sample. Only double-escape peaks, using Ge(Li)-NaI(T1) triple coincidence pair-prr,3uction, were measured. The container capsules were trom carbon or beryllium, due to their low (n,y) cross-section. Very long runs of 50 to 260 h were performed. The 5420-keV line of sulfur is interfered with by an iron line, and correction should be done using other Fe lines and the ratio between Fe !ines. Using the previously measured values of Si and Fe content of this rock, the S content was measursd. Proteins were measured by measuring the N contents of samples (usually grains and other food) and using an asumed constant N fraction in proteins. The usual conversion factor is 6.75 (1 g nitrogen = 6.75 g protein). This is one of the main uses of PGNAA neutron reactors nowadays. T i w d 2 used the 10.82-MeV capture y-rays from the I4N (n,y)I5N reaction to measure N content, and hence protein content using an Am-Be neutron source; whereas, Andras et al.j3 used ILL nuclear reactor with a neutron guide. This was the first experiment to study the protein contents of single seeds, which is very important for genetic studies and for finding the right seeds for larger crops. A carefully shielded measuring chamber was connected to the exit window of a neutron guide where the flux is 8 X 10%/cn12 * s. They put the seeds on a moving plastic strip of 100 m, with 10-cm spacings behv ,en each seed. A 5" X 3" NaI detector was placed 20 cm from the sample. I4N was determined by the 10.82-MeV 7-line without any interference, and H was determined by the 2.23-MeV y-line. Due to the small mass of each seed, the contribution from the atmospheric nitrogen should be eliminated either by evacuating the measuring chamber or filling it with helium gas. Background for 9.5 to 11.3 MeV used for the nitrogen channel was measured with a pure C-H-0 compound such as glucose. They used 3 0 - s irradiation for each seed. The ratio between the nitrogen and the hydrogen (2.1 to 2.3 MeV) C O U ~ I L ~ K : ~ was used for the calculation of the nitrogen (H was the internal standard and compared to glucose). Kobayashi and Kandd4 used the Kyoto University Reactor (KUR) with a neutron guide lube LC) measure 1°B concentrations (ppm range) in tissues. This measurement is important for measuring 11 concentrations in tissues after injection of boron for the purpose of tumor neutron therapy via the I O U (n,a) 7Li (boron neutron capture therapy). 'fie detection limit of the system is 0.1 to 0.5 pprn 'OU.In practical measurements, 10 ppm O ' B can be measured in 1-g samples within 10% accuracy in less tiian 30 s. Hydrogen in the tissue is used as internal standard. Glascock et al.35used reactor-based PGNAA to measure Pb, Zn, and Sn, together with Cu in copperbased metallic bronzes and coins. All four elements are measured nondestructively in one ineasuremeni. Fairchild et al.36measured boron in patients pretreated by injection of boron, for boron neutron capture therapy (BNCT), using PGNAA with both the high-flux beam reactor and the medical research reactor to provide "on line" boron analysis for BNCT. Spychala et ale3' used a PGNAA facility with internaltarget geometry that was installed at a tangential tube of the research reactor at the GKSS research center in Germany. Tiley measured only the double-escape peaks using only triple coincidence rneasure~nents.A central HPGe detector is surrounded in nearly 4 n geometry by a large NaI(T1) scintillator split into four optically isolated sectors to allow triple coincidence measurement (it is called a pair spectrometer). Concentrations of the major and the minor characteristic elements of the sediments Na, Al, Si, K, Ca, Ti, Mn, Fe, Zn, C1, and S are determined simultaneously, as well as the trace conslituents Cr, NI,Cu, Cd, and Hg. The major and minor elements can characterize the sediment, and CJ and Hg w e irace elements that belong to the key elements in environmental research and protection. Ward et aL3*studied 35 elements in coals (standards and various samples) by DGNAA and added to it the measurement of boron by PGNAA. The y-line used for boron is the 478-keV y-line from "13 Y (n,y) 7Li* _) 7Li. Due to the recoiling of the 7Li, the 478-keV line is Doppler-broadened, and due to the broadening it has interferences from the 472-keV line of Na. A graphical interpolation correction was needed to correct for the 472-keV Na line. They measured only single spectra since they need only low-energy measurement. They found it was sufficient to use a "bare" HPGe detector without Compton suppression. Matsumoto and A i ~ a w aused ~ ~ a reactor-based PGNAA system to measure I0B concentration for BNCT. They used the thermal column of a Triga I1 (100-kW) reactor. A silicon single crystal has been used as a filter inserted into a horizontal beam hole to extract a relatively pure thermal neutron beam. The beam area was about 7 cm2 (defined by polyethylene and lead collimators). The neutron fluence ~ ~ the 2.23-MeV line of H for internal standard. is 5 X lo5 n/cm2 s. Kobayashi and K a r ~ d aused This is a good choice for large samples with uniformly distributed hydrogen, but unsuitable for a

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onhomogenized sample or small samples. Matsumoto and A i ~ a w used a ~ ~ a small silicon detector with 6 ~ i Fto measure the neutron h e n c e , rather than using an internal standard due to the small samples. ~ o g u et s a1.@built a prompt y-neutron activation analysis facility at the 5-MW MITR-I1 (MassachuPSsctts Institute of Technology Research Reactor), to support the neutron capture therapy program by ;devising a sensitive method for nondestructive measurement of B, independent of the chemical binding .. Eand surroundings. This facility is unique in that it uses a diffracted neutron beam. This facility has :jower thermal neutron flux than will be obtained using the direct beam; however, it has a considerably Flower fraction of fast neutrons and photons. This system has lower flux than those facilities using :neutron guides, which also have a low fraction of fast neutrons and photons; however, the building of a neutron guide facility is very expensive, existing only in very few places. The construction of the kdiffracted neutron beam is relatively inexpensive. The use of the relative high flux of thc PGNAA facilities is used mainly for measutement of carbon, hydrogen, boron, nitrogen, sodium, chlorine, potassium, sulfur, calcium, and cadmium in biological and gfood sarnple~P'.~~ - ~ o r i s o vet studied the prompt y-ray emission from a living tissues simulator in order to calculate ,the maximum neutron dose equivalents for in vivo analysis by PGNAA using a nuclear reactor. The used was a 25-crn diameter cylindrical tank filied 20-cm deep with an aqueous solution of Lufea (3% N). Various filters (Si, Gd,Fe, Al, S, and B) were used. Their recommendations were: for in ',viva analysis of human tissues at a depth of 2 to 5 cm, it is advisable to use neutrons of 20 to 40 keV $mean energy with a beam area of at least 45 crn2.The analysis sensitivity per unit neutron dose equivalent '.can be more than 1 order of magnitude compared to a 238P~-Be neutron source. 2. Riley and L i n d s t o d measured boron in borosilicate glasses by PGNAA using a boric acid sample !as the comparator. No correction was necesmy for overlap from the sodium peak at 472.4 keV since iB was found to be 3500 times more sensitive than Na (per equal weights). Due to the high cross-section for neutron capture by boron, high concentrations of boron can lead to appreciable neutron self-shielding ( l f e thickness of 2.4 mg B/cm2). Consequently, the amount of the boron was limited to I mg, providing an optimal balance between sufficient count rate for acceptable counting statistics within a convenient Jrradiatiodcounting time and minimal error from the effects of high count rate. Approximate calculations of the self-shielding show that 1 rng I3 Ieads to a 3.1% decrease in the neutron flux. This e p r was made constant for all samples by making all samples of the same-size pellets and 1 mg B, via dilution .with boron-free graphite. However, the error of the self-shielding is not large even if not taken care of. Changing the mass of boron in the pellets from 0.5 to 5% causes only 7% deviation from linearity. Kim and B W measured the boron content in reactor-grade aluminum by PGNAA. The experimental detection limit was found to be 5.7 ppm with 10% uncertainty using a beam of neutrons with 2.54-cm diameter and flux of 1.2 X lo6 rr/crn2 s (Cd ratio for gold, 13). For most reactor-grade materials, the boron content is much lower than this detection limit. In order to lower the detection limit, boron was preconcentrated using a cation exchange column. The eluate, containing boron, was evaporated to reduce the volume, after addition of mannitol to prevent the loss of boron by evaporation. The preconcentration technique allows the measurement of 0.57 ppm for 30-g A1 samples. Ward46used a PGNAA apparatus situated at the 0"-180" through-tube of the Impend College Consort 11 reactor with a thermal neutron flux at the target position of approximately 2 X lo6 n/cm2 s to measure the boron content of biological material. The target is about 23 cm from the Ge(Li) detector surrounded by Nal crystal. The Cornpton suppression reduces the Compton continuum in the region of the boron peak by a factor of 5.8. 50 ng boron can be detected with a lo4-s period' of measurement. Corrections for the interference of the' Na peak of 472 keV to the Doppler-broadened 478 keV peak of B was done by graphical interpolation between the interference-free portions of the boron peak. Boron contents were measured for bone and tooth samples of healthy and rheumatoid arthritic individuals. James et a1.47evaluated the use of PGNAA for determination of an intestinal rare earth marker, samarium, in comparison with TNAA. For PGNAA, 400-mg samples were measured at MURR for 30 min each. PGNAA has significant advantages with respect to its higher cross-section, leading to shorter turnaround time. Samples were measured both by PGNNAA and INAA and in only one sample did results from the two methods differ by more than the experimental error. Glascock et al.48discussed the use of reactor PGNAA for geochemical analysis. All of the major constituents (Si, Na, Al, Fe, K,Mg, S, iind Ca) usually measured by geochemists except oxygen and PhoVhonrs can be measured by PGNAA, Besides, PGNAA ciln be used to measure R and three REEs (Nd, Stn, and Gd).

M i ~ h a e l i compared s~~ the analytical d a b obtained by PGNAA, INAA, and ICP (inductively coupled plasma) for a sediment sample from the Elbe River. A total of 15 elements were measured, though PGNAA was not used for Mg. For most elements, the agreement between the methods wai quite good, except for Cr and Cd, for which ICP gives 40 to 50% higher results. Many studies used a nuclear reactor both for INAA and PGNAA, measuring most elements by INAA and complementing that by measuring others via PGNAA: predominantly those that could not be measured by INAA, but not only those. Germmi et d.'O measured I-g pellets of coal by irradiation in [he NIST reactor for 12 to 20 h. They measured H, B, C, N, Si, S, Cd, and Gd, which could not be measured by INAA in the same reactor; they obtained also the concentrations for Na, Al, C1, K, Ca, Ti, Mn, Fe, Nd, and Sm, which were also measured by INAA. Graham et aL5'used both INAA and PGNAA to measure the elemental concentrations in NBS standard reference materials using the MURR reactor. For PGNAA, they irradiated 1-g samples (geological samples) for 12 h and measured B, Na, Mg, Al, Si, C1, K, Ca, Ti, Mn, Fe, Sm, and Gd. Vogt and SchlegelSzdetermined 23 elements in ash by INAA and 10 by PGNAA; 6 of the 10 (Al, K, Ca, Ti, Fe, and Sm) were determined by both methods, but four elements (B, Si, Cd, and C;d) could be measured only with PGNAA. The interference of Na 472-keV peak with the 478-keV boron peak was subtracted by measuring the 90.9-keV peak of Na using the constant ratio of the two pc&s of Na. Simons and LandsbergeP measured 30 elements in reference material by nondestructive neutron activation methods. Cd and S were determined by PGNAA using 2 h for measurement. Ccl was dctenined ~~ 35 elements in by the 558.2-keV line and S by the 840.4-keV line. Ward and c o - w ~ r k e r smeasured human placenta by INAA (mainly instrumental) using three irradiations and four countings; they also measured boron by PGNAA. The detection limit was 0.01 ppm for 1- to 2-g samples measured for i04 s. A similar study was performed by Ward and Masods on brain tissue from Alzheimer's disease patients and control individuals: 30 elements were determined by INAA. Si was determi~tedby fast neutron activation analysis, and B was determined by PGNAA. For lo4-s measurement, a detection limit of SO ppb was found. Gokmen et aLS6measured trace element content in blood by both INAA and PGNAA. Using PGNAA, they measured H, C, N, 0, S, C1, and K. From those, only C1 and K were measured also by INAA. The C1 results obtained by PGNAA were lower than those obtained by MAA. The authors said that this often happens for matrices containing high levels of hydrogen, but no further explanatiorr fur Lhis effect was given. Chung2 used a mobile small reactor to measure Cd in vivo and in vitr-o. Absolute measurement of PGNAA can be done either by comparison to standards for each element or by one standard for all elements. For elemental standards, the mass is calculated using the equatron

where m stands for the mass of the measured element and C stands for the counts in the y-ray spectrum's peak due to this element. PGNAA often requires long irradiations and handing many standards in multielement analysis is time consumicg. One way to overcome this long irradiation time of many standards is to use multielement standads. This can be done by thoroughly mixing different compounds, or using standard reference materials or certified reference materials prepared by various agencies. These reference materials are not always available and difficulties may arise when one has to anaiyze a new type of material that has a different elemental composition than that of the reference material. Thus, it.is preferable to use only one elemental standard material to measure the neutron fluence, calculating from it the masses of all observed elernents using the available information on cross-section, abundances, y-line intensities, and detector efficiency. Using only one or two standards is called thc "absolute method" and it requires accurate knowledge of a11 nuclear and system parameters involved, besides having reproducible experimental csnditions. Khrbish and SpyrouJ;'discussed the possibility of using the absolute method for PGNAA. They compared the ratio of sensitivities for various elements to the sensitivity of boron in three different set-ups (theirss7and those of References 13 and 19). Having similar ratios of sensitivities will indicate thz possibility of using an absolute method; whereas, different ratios of sensitivities for two elements in the three set-ups show that the absolute method cannot be used. They found that S, C1 (by 517-keV line), K, Ca, Cd, and Au can be determined by the absolute method, but not C1 (by the 788-keV and 1164-keV lines), Fe, and Hg. There is much better agreement between Reference 13 and Reference 19 data than with those of Reference 57. This might be due to the better thermalization of the neutrons in the set-ups of References 13 and 19, as can be judged by

the cadmium ratios (CR). (Reference 13, CR = 42.0; Reference 19, C R = 55.0; Reference 57, C R = 9.0.) Thus, there is very good agreement for C1 (1164-keV line) between References 1 3 and 19, but a very large discrepancy with Reference 57. T h e same applies also for Fe. These results indicate that the absolute method can b e used only for a restricted range of cadmium ratios.

REFERENCES 1. Isenhour, T. L. and Morrison, G. H., Modulation technique for neutron capture gamma ray measurement in activation analysis. Anal. Chem. 38, 262, 1966. 2. Chung, C., Activation analysis with a mobile reactor in Activation Analysis, Vol. 2, Alfasi, Z. B., Ed., CRC Press, Boca Raton, FL, 1990, pp. 2: Chen, W. K. and Cl~ung,C. In vivo and in v i m medical diagnoses of toxic cadmium in rats, J. Radioanul. N~icl.Chem. 133, 349, 1989. 3. Hammermesh, B. and Hnmmel, V., Neutron capture y-ray spectra from elements Z = 17-30 and Z = 45-57, Phys. Rev. 88, 916. 1952. 4. Lombard, S. M., Isenhour, T. L., Weintz, P. H., Woodruff, G. E., and Wilson, W. E., Neutron-capture gamma-ray activation ana!ysis. Design of apparatus for trace analysis. Int. J. Appl. Radiat. Isor. 19, 15, 1968. 5. Henkelmann, R., and Born, M. J., Analytical use of neutron-capture gamma-rays, J. Radioanal. Chem. 16, 473, 1973. 6. Gladney, E. S., Jumey, E. T.,and Crarliss, D. B., Nondestructive determination of boron and cadmium in environmental materials by thermal neutron-prompt y-ray spectrometry, Anal. Chem. 48, 2139, 1976. 7. Jurney, E. T., Curtiss, D. R., and Glniiney, E. S., Determination of sulfur in environmental materials by thennal neutron capture prompt gamma ray spectrometry. Anal. Chem. 49, 1741, 1977. 8. Gladney, E. S., Curtiss, D. B., and Jumey, E. T., Multie!en~ent analysis of major and minor elements by thermal neutron induced capture gamma-ray spectrometry, J. Radioanal. Chem. 46, 299, 1978. E. T., Simultaneous determination of nitrogen, carbon and 9. Gladney, E. S., Curtiss, I). B., and J~~rney, hydrogen by thermal neutron prompt y-ray spectrometry, Anal. Chim. Acta, 110, 339, 1979. 10. Henkelmann, R., Analytische Verwertung der prompten gammastrahlung nach dem neutroneneinfang, Radiochim. Acta, 15, 169, 197 1. 11. Failey, M. P., Anderson, D. L., Zoller, W. II., Gordon, 6. E., and Lindstrom, R. M., Neutron-capture prompt gamma ray activation analysis for multielement determination in complex samples, Anal. Chem. 51, 2209, 1979. 12. Rustad, B. M., Als-Nielsen, J., Bahnsen, A., Christensen, C. J., and Nielsen, A., Single-crystal filters for attenuating epithermal neutrons and gamma rays in reactor beams, Rev. Sci. Instrnm. 36, 48, 1965. 13. Hanna, A. G., Brugger, R. M., and Glascock, M. D., The prompt gamma neutron activation analysis facility at MURR, Nucl. Instr. Meth. 188, 6 19, 1981. 14. Greenwood, R. C., Elemental analysis using the neutron capture gamma-ray technique with a Ge(Li) detector, Trans. Am. Nucl. Soc. 10, 28, 1967. 15. Orphan, V. J. and Rasmussen, N. C., A Ge(Li) detector for studying neutron capture gamma rays, Nucl. Instr. Meth. 48. 282, 1967. 16. Lombard, S. M. and Isenhour, 7'. L., Neutron capture gamma-ray activation analysis using lithium drifted gennanium semiconductor detectors, A n d . Chem. 40. 1990, 1968. 17. Comar, D., Crouzel, C., Chasteland, M., Riviere, R., and Kellershohn, C., The use of neutron-capture gamma radiation for the analysis of biological samples, Nucl. Appl. 6, 344, 1969. 18. Yonezawa, C., Haji Wood, A. I<., Hoshi, M., Ito, Y., and 'hchikawa, E., The characteristics of the prompt gamma-ray analyzing system at the neutron beam guide of JRR-3M, Nucl. Instr. Meth. Phys. Res. A329, 207, 1993. 19. Anderson, D. L., Failey, M. P., Zoller, W. H., Walters, W. B., Gordon, G. E., and Lindstrom, R. M., Facility for non-destructive analysis for major and trace elements using neutron capture gamma-ray spectrometry, J. Radioanal. Chem. 63, 97, 1951. 20. Isenhour, T. L. and Morrison, G . H., Determination of boron by thermal neutron activation analysis using a modulation technique, A n d . Chem. 38, 167. 1966. 2 1. Rasmussen, N. C. and I-Ivkni, Y., The prompt activation analysis of coal using neutron capture gamma rays, Trans. Am. Nucl. Soc. 10, 29, 1967. 22. Grazman, B. L. and Schweikert, E. A., A brief review of the determination of cadmium by prompt gammamy neutron activation analysis, J. Itadioanal. Nucl. Chern. 152, 497, 1991. 23. Greenwood, R. C., Reed, .l.H., Kol;ir, R. D., and Terrell, C. W., IJse of neutron-capture gamma ray for a lunar-surface compositional analysis, 7'inns. Am, Nucl. Soc. 9, 179, 1966. 24. Elkadi, A. and Duffey, D., Idenrificntion of gold in rnixn~recby nputron capture gamma rays, Trans. Am. N L ~SOC. . 12, 42, 1969.

Wiggins, P. E,Duffey, D., and Elkady, A. A., Neutron capture gamma ray studies of marine manganese nodules using a nuclear reactor and a californium-252 source, Anal. Chim. Acra, 61, 421, 1972. Heurteblse, M. and Lubkowib, J. A., Determination OF metals in alloys by neutron cupture gamma-ray spectrometry, J. Radioanal. Chem. 31, 503, 1976. Heurtebise, M., Buenafama, H., and Lubkowitz, J. A., Determination of cobalt molybdenum and nickel in hydrodesulfurization catalysts by neutron capture y-ray spectrometry and neutron activation analysis, Anal. Chem. 48, 1969. 1976. Heurtebise, M. and Lubkowitz, J. A., Determination of water hydrodesulfurization catalysts by neutron capture-y ray spectrometry, Anal. Chem. 48, 2143, 1976. Pouraghabagher, A. R. and Profio, A. E., Neutron capture gamma ray spectrometry for determination of sulfur in oil, Anal. Chem. 46, 1223, 1074. Curtis, D., Gladney, E., and Jurney, E., A revision of the meteorite based cosmic auundance of boron, Geochim. Cosmochim. Acta, 44, 1945, 1980. Islam, M. A., Kennett, T. J., Prestwich, W. V., and Rees, C. E., Detemlination of the sulphur content of a lunar sample by neutron capture gamma ray spectrometry, J. Radioanal. Chem. 56, 123, 1980. Tiwari, P. N., Rapid and non-destructive determination of protein in grain samples by prompt (n,y) technique, Radiochem. Radioanal. Lett. 6, 363, 1971. Andras, L, Baiint, A., Csoke, A,, and Nagy, A. Z., Selection of single grain seeds by "N(n,y) "N nuclear reaction for protein improvement, Radiochem. Radioanal. Lett. 40, 27, 1979. Kobayashi, T. and Kanda, K., Microanalysis system of ppm-order "'B concentrations in tissue for neutron capture therapy by gamma-ray spectrometry, Nucl. Instr: Meth. 204, 525, 1983. Glascock, M. D., Spalding, T. G., Biers, J. C., and Cornman, M. F., Analysis of copper-based arietaiic artifacts by prompt gamma-ray neutron activation analysis, Archaeometry, 26, 96, 1984. Fairchild, R. G., Gabel, D., Laster, B. H., Greenberg, D., Kiszenick, W., and Micca, P. L.,Microanalytical techniques for boron analysis using the I0B (n,a) 7Li reaction, Med. Phys. 13, 50, 1986. Spychala, M., Michaelis, W., and Fanger, H. U., Prompt y-ray neutron activation analysis of river sediments, Nucl. Geophys. 1, 309, 1987. Ward, N. I., Kerr, S. A., and Otsuka, T., Multielement content of coal by neutron activation analysis technique, J. Radioanal. Nucl. Chem. 1 14, 1 13, 1987. Matsumoto, T. and Aizawa, O., Prompt gamma-ray neutron activation analysis of boron-10 in braiogicd materials, Appl. Radiat. Isot. 41, 897, 1990. Rogus, R., Harling, 0. K., Olmez, I., and Wirdzek, S., Boron-10 prompt gamma analysis using a diffracted neutron beam, in Progress in Neutron Capture Therapy for Cancer; Allen, B. J . and Harrington, U. V., his., Plenum Press, New York, 1992, 301-304. 41. Anderson, D. L. and Cunningham, W. C., Determination of boron and other elements in food and agriculturai products by PGAA, Trans. Am. Nucl. Soc 65, 140, 1992. 42. Rossbach, M., Prompt gamma cold neutron activation analysis applied to biological materials, Frcsenius J: Anal. Chem. 344,59, 1992. 43. Borisov, G. I., Komkoy, M. M., and Lconov, V. F., Prori~pt-gammaspectrometry for the c y t i n ~ i ~ i ~oft i v ~ ~ reactor neutron beams in biomedical rcscurch, Sov. Atom. Energy, (English tmnslutioa) 6'3. 921, 1988. 44. Riley, J. E. and Lindstrom, R. M., Detzmination of boron in borosiiicate glasses by neutron capture prornyl gamma-ray activation analysis, J. Radioarml. Nucl. Chem 109, 109, 1987. 45. Kim, N. B. and Bak, H., A study of boron determination in high purity aluminum t j capture gamma-ray measurement, J. Korean Nucl. Soc. 13, 229, 1981. 46. Ward, N. I., The determination of boron in biological materials by neutron irradiation and prompt gammaray spectrometry, J. Radioanal. Nucl. Chem. 1f 0, 633, 1987. 47. James, W. D., Arnold, F. F., Pond, K. R., Glascock, M. D., and Spalding, T. G., Application of prompt gamma activation analysis and neutron activation analysis to the use of samarium as an ~ntestinalmarker, J. Radioanal. N ~ c l Chem. . 83, 209, 1984. 48. Glascock, M. D, Coveney, R. M., Jr., Tittle, C. W., Gartner, M. L., and Murphy, R. I)., Geochemical applications for prompt gamma neutron activation, Nucl. Instr. Meth. Phys. Res. B lO/I 1, f 042, 1985. 49. Michaelis, W., Multielement analysis of environmental samples by total-reflection X-ray fluorescence spectrometry. neutron activation analysis and inductively coupled plasma optical emission spectroscopy, Fresenius Z Anal. Chem. 324, 662, 1986. 50. Germani, M. S., Gokmen, I., Sigleo, A. C., Kowalczyk, G. S., Olmez, I., Small, A. M., Anderson, D. L., Failey, M. P., Gulovali, M. C., Choquette, C. E., Lepel, E. A., Gorden, G. E., and Zolder, W. H., Concentrations of elements in the National Bureau of Standard's bituminous and subbituminous coal standard reference materials, Anal. Chem. 52, 240, 1980. 5 1. Graham, C. C., Giascock, M. D., Carni, J. J., Vogt, J. R., and Spalding, T. G., Determination of elements in National Bureau of Standards geological standard reference materials by neutron activation analysis, Aml. Chem. 54, 1623, 1982.

52. Vogt, J. R. and Schlegel, S. C., Elemental determinations in NBS 1633A fly ash standard reference material using INAA and PGNAA, J. Radioanal. Nucl. Chem. 88, 379, 1985. 53. Simons, A. and Landsberger, S., Analysis of marine biological certified reference material by various nondestructive neutron activation methods, J. Radioarral. Nucl. Chem. 110, 555, 1987. 54. Ward, N. I., Madlahon, T. D., and Mason, J. A., Elemental analysis of human placenta by neutron irradiation and gamma-ray spectrometry (standard, prompt and fast-neutron), J. Radioanal. Nucl. Chem. 1 13,501, 1987. 55. Ward, N. I. and Mason, 3. A., Neutron activation analysis techniques for identifying elemental status in Alzheimer disease. J . Radioanal. Nucl. Chem. 113. 515, 1987. 56. Gokmen, I. G., Gordon, G. E., and Arm, N. K., Application of different activation analysis techniques for determination of trace elements in human blood, J. Uadionnal. Nucl. Chem. 113, 453, 1987. 57. Khrbish, Y. S. and Spyrou, N. M., Prompt gamma-ray neutron activation analysis by the absolute method, J. Radioanal. Nucl. Chem. 151, 55, 1991.

Chapter 5

pGNAA with Radioisotopic Sources, Neutron Generators, >andCharged Particle Accelerators Zeev B. AIfassi CONTENTS I. Radionuclide Sources ................................................................................................................. 75 A. 241Am-BeSources .................................................................................................................. 75 ........................................................................................... 77 B. 1 - B e Sources ....................... . . C . DZCfSources .......................................................................................................................... 82 11. 14-MeV Neutron Generators ...................................................................................................... 87 HI. Accelerators ............... 89 References .............................................................................................................................................89

I. RADlONUCLlDE SOURCES Three radionuclide sources are used for PGNAA. Tko are of the types of a-emitter embedded in Be, using the 9Be(cw,n) 12Creaction, the a-emitting source being either 24'Amor 23k"9P~. The third source is the z2Cf spontaneous.fission source. Several factors influence which radionuclide source will be used. Matthews and Spyrou' chose to use an (a,n) neutron s urce rather than a spontaneous fission source since they were analyzing bulk samples and stated th t higher energy neutrons will lead to better uniformity of the activating flux. 252Cfspontaneous fission nentrons have neutrons with average energy of 2.35 MeV. The most abundant energy is about 1 to 2 MeV. The most abundant neutrons from the Be(a,n) source are at about 3 MeV, with considerable flux at higher energy up to 12.1 MeV and an average of 4.4 MeV. The higher energy of the (a,n) source permits also the investigation of reactions with higher threshold energies; however, this is not important for many PGNAA studies where mostly thermal neutrons are used. However, for 0 and C determination, it is important. An important disadvantage of the 2S2Cfsource is its relative short half-life (effective tln = 2.65 y), which leads to a considerable economic disadvantage. On the other hand, an important advantage of the 252Cfspontaneous fission source is the low y-ray dose rates. The y-ray dose rate from a 24'Am-Besource is about 10 times higher than that from a 252Cfsource with the same neutron emission rate.2 Especially problematic is the region between 3 and 4.5 MeV due to screening by the 4.43-MeV line of the excited I2C formed in the gBe(a,n) 12Creaction. From the point of view of y-ray dose rate, u8Pu-Be neutron sources have less than half the y-ray dose rate of "'Am-Be sources. However, 2 3 8 Psources ~ are less available, and this led Matthews and Spyrou to use a 241Am-Besource. 23Tuis more available than 23XP~; however, due to the very long halflife of 239P~, its size is much larger (16 g/Ci compared to 0.3 g/Ci of 24'Am.The size of the sources are 12 cm3/Ci for 239Pu-Beas compared to 3 cm3/Ci for 241Am-Be).However, most of the published work was performed with Pu-Be sources. In some cases, it is specifically written.that it is 238F'u,while in others the isotope used was not mentioned.

P

A. '''Arn-Be SOURCES Matthews and Spyrou' constructed a system for in vivo analysis of human subjects using a 5-Ci 24'AmBe source with a total neutron emission rate of 1.1 X lo7 nls. The source is located inside n shield and it is transferred to the irradiation position by compressed air. The detector is shielded with LiZO, and boron-loaded clay around it, and a Perspex cap filled with Li2C03powder on it. The irradiation site is surrounded by boron-loaded clay to prevent the neutrons from 'reaching the water shielding. This leads to increased background below 478 keV (due to boron), but reduced background in the range 0.478 to 2.222 MeV (due to H),which is more important. For measurements of 0.5 h of a phantom. they found detection limits of 6 ppm Cd (Ey= 0.559 keV), 160 ppm C1 (E, = 6.111 MeV), 250 ppm Na (E, = 1.164 MeV), and 1.25% for W (10.S28MeV). 0-8493-5 14&9/95/50.00+~.50 8 1995 by CRC Press. Inc.

for monitoring coal on a conveyor belt. A 5-Ci "'Am source was used together with a lead reflector behind the source to Scatter-back neutron into the sample. A Iead shadow shield was placed between the source and the sample in order to attenuate the y-rays formed inside the neutron source; mainly the 4.43 MeV of the decay of I2Cand those produced from I2C(n,n'y). Between the coal and the detector, they placed a flexible boron-loaded rubber thermal neutron absorber in order to prevent the neutrons penetrating the coal sample from reaching the detector. Reasonably good results (compared to chemical assay) were obtained for C, H,0,N, S, C1, and Si. Worse results were obtained for AI, Fe, Ti, K, Mg, and Na. However, even for these elements, the agreement is still satisfactory.

B. Pu-Be SOURCES The Z3BPu-Be sources for PGNAA were mostly used for the determination of in vivo nitrogen, a measure for the total mass of the protein. The first nuclear method for direct determination of total body nitrogen (TBN) in vivo was done with delayed y-activation analysis using the 14-MeV neutrons I4N(n,2n) I3N reacti~n~ measuring -~ the 51 1-keV annihilation line of the P-decay of I3N (9.96 min). However, as this line is due to all positron emitters, the lack of specificity poses problems of interference from other radionuclides. Another interference is due to reaction of protons formed by knock-on by the neutrons. Thus, it was shown that the reaction 160(p,n) I3N contributes as much as 19% to the nitrogen counts.'' Another problem associated with this method is the lack of uniformity of the fast neutron fluence in the body. The difficulties in this method and the advantages of in vivo PGNAA determination of TBN by the '4N(n,y) ISN reaction is discussed in detail by Cohn." The first system to measure TBN by in vivo PGNAA used neutrons produced by (p,n) reactions with accelerated protons from a cyclotron and will be discussed later. The first facility to measure TBN with a 238Pu-Besource was constnicted in Toronto and describedz by Mernagh et al." The excited I5N, produced by the thermal neutron capture, decayed 15% directly to ground-state I5N, emitting y-photons of 10.83 MeV. Since no other body element has a neutron capture y-ray of such high energy, it is possible to determine body nitrogen by measuring the 10.83MeV photons. Mernagh et aLizconstructed a PGNAA facility for in vivo TRN measurement using four 5-Ci 238P~Be sources, two above the patient and two below'so that the patient is irradiated bilateralIy. Each of the two sources are placed inside a collimator, a solid block of wax (40 X 40 X 54 cm3) with a 15cm wide hole from one side to the center. The wax block is surrounded by 5 cm lead for y-ray shielding to protect the patient and the workers. Surrounding this is a I-cm layer of boric acid to capture any thermal neutrons that may have escaped from the wax and lead shielding. Another 5-cm layer lead is added to the front of the collimators to add extra shielding for the detectors from the y-rays of the sources (mainly the 4.43 MeV due to the de-excitation of the I2Cproduced in the gBe(ol,n) I2C*reaction). The y-ray photons are measured by two 5" X 4" NaI(T1) detectors placed at the side of the patients. ' The detectors are surrounded by S cm boric acid to absorb most of the neutrons scattered into the detectors from the subject or other scatterer. The detection of nitrogen was studied by irradiation of cuboid phantoms (plastic vessel of 30 X 30 X 30 cm3) containing either water or an aqueous solution of NH40H containing approximately 3 kg N, i.e., 11% weight. The difference in the spectra of the two phantoms is very clear after 10-min irradiation. The N phantom has higher counts from 8.7 up to 11 MeV. The lower energy signals are due to the singleescape and double-escape peaks and to Compton scattering. The highest observed peak is the doubleescape peak at 9.81 MeV, whereas the photopeak (10.83 MeV) is quite small. For nitrogen determination, they took the integrated counts of the 9.5- to 11.0-MeV region. For 10-min irradiation, they observed reproducibility of 2.5% (1 standard deviation), which is consistent with statistical error. Use of phantoms with saline solutions of physiological concentrations showed that the chlorine y-rays increased the counts in the 9.5- to 11.0-MeV region by less than 3%. However, this amount of N is larger than exists in human beings; for human beings, the error is slightly higher. 10-min irradiation of the chest area (area 20 X 20 cm2) of a normal volunteer shows that the statistical error of N concentration is 23%. This method turned out to be a routine hospital pr~cedure.'~ Some 10 years later, McNeill and coworkersi4discussed the possibility of using this system for smaller size infants or laboratory animals of around 3-kg mass. The typical counts of the 9.5- to 11.0-MeV region for 1000 s in the case of an adult human is 3000, while the background is about 2500,leading to a staristical error of 3%.Positioning . errors tend to raise the actual measured error to 4 to 5%. as wns determined by repeated measurements.

During the measurement, the paticnt receives a dose equivalent of 0.3 mSv (30 mrem) to the trunk of the body. Reducing the total mass of the body by a factor of 20 to 30 will decrease by the same factor the difference between the measured counts for the patients and the background, leaving almost the same background. This will mean an error of 30 to 50% in the N content determination. In order to reduce the error, the background should be reduced. The major change made for small animal investigation was to insert a further 10-cm layer of lead between the source house and the NaI ciystal. This extra shielding left too little room for bilateral irradiation. The bilateral irradiation was done in order to improve the uniformity of the neutron flux in the subject. The nonuniformity is caused by absorption and scattering of h e neutrons within the subject; however, in small animals, this absorption and hence nonuniformitjr is very small. Consequently, all four 5-Ci UsPu-Besources were put inside one collimator, below the subject. The observed statistical error for approximately 3-kg animals with 3% nitrogen is about lo%, leading to results of 3.0 +- 0.3% nitrogen. This error can be reduced with more expensenamely, increasing the number of the detectors from one to four, and/or increasing the dose, either by stronger neutron sources or longer irradiation. Investing in these improvements can lead to statistical errors of 3 to 4%. Wang and Wanna15 studied the major contributors to the background of the I5N* peak (9.5- to 11.0MeV region) with an NaI detector. It was found that the main contributors are the random coincidence counting-also called "pile-up" (contributing about 65%)-and Gaussian spreading (contributing about 25%). Nitrogen in the air was found to contribute less than 4% of the background. This result indicates that there is no advantage in increasing the neutron source strength (neutron flux) since doubling the neutron flux will double the signal but triple the background. Doubling the neutron flux is quadrupling the random coincidence counting, while only doubling all the other contributors (such as Gaussian spreading and the signa1,due to air's nitrogen). The total increase in the background is (0.65 X 4) (I - 0.65) X 2 = 3.3. Thus, increasing the neutron flux will decrease the SIN ratio. If possible economically, it is preferable to double the measurement time while halving the neutron source. A similar system with a stronger neutron source of 85-Ci u8Pu-Be was constructed at Brookhaven by Cohn and co-workers.I6 In order to reduce the background, they avoided the use of steel in the support structure for the detectors so as to reduce the 10-MeV y-line from 57Feneutron capture. The use of hydrogenous material was avoided in the construction wherever possible. Instead of using boron for neutron absorption in Toronto's system (References 12 to 15). they used 6LiF and Li2C03since lithium is the only element capturing neutron without emission of y-rays. Although also for 'OB, the reaction is lOB(n,c~)~Li rather than the (n,y) reaction; Li is advantageous to B since the 7Li is fonned in an excited state decaying with a 478-keV photon emission. The tritium formed in the 6Li(n,cx)T reaction is formed in the ground state. The boron y-iine (478 keV) is very far from the measured region (9.5 to 11.0 MeV); however, it contributes to tile background duc to random coincidence counting (pilcup). Thus, for the cases where rigorous moderation of neutrons was necessary (as iil the collimator or iil the detector shieiding) resin heavily doped with 6LiF or Li2C03was used. An 85-Ci URPu-Resource (2.3 X 10' nls) is housed in a collimator made of epoxy resin doped with 6LiF and Li2C03.The source is located 50 cm from the motorized bed on which the patient lies. The neutron shield consists of boxes filled with polyester resin mixed with Li2C03and 'jLiF. The whole system is covered with a 20-cm thick layer of lead with additional 10-cm layer under the detectors. The dose rate is about 1 m r c d h at a distance of 1 m, and thus the medical staff may stand near the patient during measurement. Premoderation is required before the neutrons znter the patient's body in order to attain better uniformity of the neutron flux through the body. Since they measured both nitrogen and hydrogen in the subject body, the premoderation cannot be done by hydrogenous material and it is done with 5-cm thick aluminum box filled with D20. Two 6" X 6" NaI(T1) crystals are used for y-ray detection. They are surrounded by cups made of resin doped with Li compounds. Due to the high count rate, the multichannel analyzer is working in the fractional charge collection methodi7 in order to reduce the "pile-up" background. The pile-up was also reduced by selective pulse integration. With a bed scanning rate of 0.23 cmls, the absorbed dose is 1.25 mrad from neutrons and 0.6 mrad from y-rays per scan. Using RBE (relative biological efficiency) of 10 for fast neutrons, the dose is approximately 26 mrem per scan. This is the dose for the skin, and it is lower in the body due to attenuation. It was found that there is an initial build-up of slow neutrons in the beginning of a phantom reaching maximum 4-cm deep and then the flux starts to decline. Placing the detector above the body, opposite the irradiated side, will partially compensate for the declining thermal neutron flux since the deeper in the body the

+

prtial volume is the closer it is to the detector and hence the attenuation of the neutrons is compensated by both detector efficiency being higher and absorption of the y-photon in the body (self-attenuation) being smaller. Similar to the results of Vartsky et a1.,'8 they found that for a constant N/H solution in the phantom, almost similar results were obtained for NIH counts, irrespective of the geometry except the thickness. With appropriate correction for different thickness of body section^,'^ the coefficient of variation of N/H counts from shoulders to knees was found to be 23%. Thus, the counts of H peak can be used as an internaI standard in in vivo nitrogen measurement by PGNAA, compensating for variations in nitrogen counts due to body size and habitus of each subject. Using a nitrogen-free phantom, it was found that the background under the 10.83-MeV peak car? be approximated quite accurately by a combination of two exponentials (two lines in a semilogarithmic plot); thus, each pritient's background is obtained by fitting a line on each side of the nitrogen peaks. The total body hydrogen (TBII) used as internal standard is determined from body water, fat, and weight (determined by other methods), based on the known fraction of hydrogen in water (O.11), fat (0.12), and protein (0.07). TBH = (0.1 1 X water)

+ (0.12 X

fat)

+ 0.052[body weight - (water + fat)]

The authors measured nitrogen in 14 male volunteers and obtained 2.7 C 0.2% of body weight. Cohn and co-workers used this system for several studies. In one study, they measured19 the total body level of nitrogen (as well as of potassium, water, and fat) as a function of combined nutritional support given to cancer patients unable to maintain normal dimentation. Similar measurements were done with 15 renal patients on maintenance dialysis.20 The total body fat was originally obtained by a skinfold technique. However, this measurement of fat content was shown to lead to large error. The following measurements used a correlation between total body potassium ('IBK) (measured by total body counting of %) and the Iean body mass (LBM). The difference between body weight and lean body mass was taken as fat weight (TBF, total body fat). However, the correlation between LBM and TBK holds only for normal subjects but not for patients with wasting diseases or cancer patients. Vartsky et a1.2' suggested another method, in which TBN, TBH, and TBF are derived sin~ul~aneously from data obtained by neutron capture y-ray analysis. The background is taken from a phantom filled with a nitfogen-free tissue equival&t liquid. The body minerals are calculated from total body calcium measured by delayed gamma neutron activation analysis. Assuming the total weight (TW) to be

TFV

= TBF

+ TBW + TBP + BMA

(1)

where TBP = total body protein = 6.25 X TBN. BMA the body mineral ash (1) is calculated from total body calcium. Equation 1, together with Equation 2 for TBH and Equation 3 for TBN, leads to an explicit equation for TBN. '-

TBH = aTBW

+ bTBF + cTBF

TBN = ~.(CN/C,)-TBI-I

(2) I

(3)

a, b, c, and k are constants (a = 0.11, b = 0.12, c = 0.07; and k depends on the PGNAA system and is calcuIated from phantom with known concentrations):

TBN =

b(TW

- RMA) 4 (a - b)TBW

[k(CNICH)]-' + d(b - c)

Beddoe and co-workers in the School of Medicine at the University of Auckland, Australia constructed a PGNAA facility for in vivo analysis using a 23Vu-Beneutron source in the intensive care unit of the h o ~ p i t a l . ~Their ~ - ~ " main purpose was to measure TBN in surgical and critically ill intensive care patients in order to study nutritional effects and metabolic problems. Although the main purpose of the facility is to measure nitrogen, in the design of the facility consideration was also given to future possible decerminarion of carbon through the jnelastic neutron scattering I2C(n,n'y) I2C induced by fast neutrons. This reaction has a threshold energy of 4.3 MeV, which is about the same as the mean neutron energy

of a Pu-Be source. The excited carbon emits a 4.43-MeV y-ray that can be used for its determination. In order to enable the determination of carbon, the design tries to minimize the use of carbon-moderating materials, although it interferes with the moderation since the highest hydrogen concentration is for hydrocarbon compounds. The neutron beam is produced by two 7.6-Ci (281 GBq) 23RPu-Besources, each emitting 2.0 X lo7 nls, doubly encapsulated in steel, in cylinders 5-cm long and 2.5 cm in diameter. The neutron sources are located 1 m apart during the analysis and in the center of shielding tanks. At other times, the sources are retracted into 50-cm thick solid concrete buttresses. The main moderator is water in an aluminum tank. Aluminum was preferred to iron as the main constructional material due to its main y-line being at lower energy (7.7 MeV for A1 and 8.7 MeV for Fe) and interfering less with the N peaks (photopeak, single- and double-escape peaks stretched from 9.5 to 11.0 MeV). As a precaution against corrosion, all aluminum surfaces are coated with a bituminous paint. To reduce the 2.2-MeV peak from the water hydrogens (but incxcasing the 478-keV boron peak), the water in the moderating tank contains 5.7% boric acid 4.6% borax. The mixture of the two boron compounds is used in order to buffer the wdtex to pH = 7.5 to decrease aluminum corrosion. Bismuth collimators (minimal radial thickness = 5 cr?~ 1. in which the foci of the neutron sources are located, are used to increase the rieutron flux at the k v c i of the patient. First-order scattering calculations show that the collimator increases the neutron flux by 18 to 24%. The collimator also reduces the y-ray noise originating from the sources. The detector faces are shielded from thermal neutrons generated in the patient by LiF powder. The detector is surrounded by a 1.5-cm thick lead cylinder surrounded by a container with water and boric acid. The main observed peaks from nitrogen in the y-ray spectrum are the 9.81- and 10.32-MeV firstand double-escape peaks of the 10.83-MeV y-photons. The usual measurement of nitrogen is by collecting all the counts between 9.5 and 11 MeV. :12 found that the signal-to-noise ratio (S/N ratio) in the Despite the extensive shielding, Beddoe et ' nitrogen region (9.5 to 1 1 MeV) was less than satisfactory. It was found that the noise was predominantly due to fast neutrons reaching the detector and reacting with the NaI(T1) crystals. In order to reduce this noise, every available space around the detector was packed with paraffin to slow down the neutrons. This use of paraffin almost eliminates the possible use of the system for carbon determination, since paraffin has 85% carbon; however, it was essential for nitrogen determination.' The packing with Lhe wax paraffin was found to improve the S/N ratio from 0.46 to 1.30, with a 5% N phantom. Two 5" X 4" NaI(T1) crystals are used for detection. In order to reduce the dead-time, biased amplifiers are used to cut all pulses below 1.5 MeV. The patients are lying on a couch scanned at a speed of 2.6 cmdmin, and the measurement is done for 2000 s. The hydrogen peak is clearly the predominant one. The next clear peaks are due to N. Additional observed peaks are those due to C, C1, Pb, and Fe, due to construclion and shielding materials. The optimum energy ranges for integral counts were found to be 2.04 to 2.41 MeV for hydroger. and 9.5 to 11 MeV for nitrogen. Beddoe et al. found that for the cuboid acrylic phantom containing aqueous solution, the total counts in the 9.5- to 11-MeV region are linear with N concentration in the range 1 to 8% (correlation coefficient, 0.9998). The phantom size affects t l ~ota1 ~ counts for the same solution. However, the ratio of nitrogen counts to hydrogen coua: for a constant solution varies only 2.1% from linearity with phantom width within the range of mean subject widih. Thus, the N/H ratio r-an be taken as almost independent of phantom thickness and a linear function of width. The N/H ratio counts are used for in~crnalcalibration of the neutron flux since percentage hydrogen compositions in the human major compartments are not much different. liydrogen is mainly due to water, and total body water (TBW) can easily be measured Lo 4 1.5% by the tritiim dilutiun technique. Beddoe at al. estimated the precision of the TBN to 24.2%. The limiting factor in rile precision are the counting statistics. From the TBP weight, assuming the weight of protein to be nitrogen weight multiplied by 6.25, and estimating the relative small weight of minerals and glycogen from skeletal size, the total fat weight (TFW) can be obtained by subtraction from the TBW.

+

a

total body fat = total body weight

- (6.25

X TBN

+ TBW 1- minerals + glycogen)

Beddoe et a1.l' found that there is a 26.3% error in total body fat measurement. Beddoe et al.23 measured TBN and TBW for 41 normal subjects and 56 surgical ward patients receiving intravenous nutritional therapy by PGNAA and tritium dilution. They found that both protein depletion and increased hydration of the fat-free body mass accompany surgical illness. The calibration technique is discussed extensively in Reference 24. Chan and B e d d ~ adapted e~~ a Monte Carlo code to model the radiation

transport in the in vivo PGNAA facility. The results for slow neutrons were in broad agreement with those obtained experimentally with activation foils. Streat et measured total body fat (TBF) and fat-free mass (FFM) by the combination of in vivo PGNAA and tritium isotope dilution, and compared it to the more traditional methods of densitornetry and skinfold anthropornetry. The values of TBF for normal volunteers were in agreement for the three methods. However, for sick patients, skinfold anthropornetry leads to 19% lower values (-3 kg) than in vivo PGNAA/uitium dilution. They concluded that PGNAAItritium dilution is a suitable method for measuring TBF and FFM, parti_cularlywhen body composition is abnormal. Mcneill et aLZ7measured TBN in three groups of subjects: volunteers of different ages, patients with liver ailments, and patients on peritoneal dialysis. It was shown that TBN measurements gave information (in accord with clinical findings) that was not given by indirect methods of estimating LBM by skinfoId measurements. They also measured I( by 40Ktotal body counting and Ca by DGNAA. Beddoe et aLZ8used the Auckland systemz2for TBN measurement to measure total body chlorine (TBCI) using the W ( n , y ) 36C1reaction, and measuring the 36Cl*gammas of 5.6 and 6.1 MeV. The main problem is the accuracy of extracting elemental data from these poor signal-to-noise peaks. As in the case of nitrogen, hydrogen is used as an internal standard. The chlorine signal was extracted by calculating the background hy fitting an exponential function to the natural logarithm of the counts below the 5.6-MeV peak, between the 5.6- and 6.1-MeV peaks, and above the 6.1-MeV peak. The chlorine counts were calcnlnted from the 5.5- to 6.3-MeV region after subtracting this fitted background. The precision of the measurement was found to be 29.0% from 25 repeating scans. It was proven that the NcJNHcounts ratio is linear with C1 concentration from 0.5 to 8 times the normal physiological concentration. rather than the 5.6- and 6.1Mitra et alez9measured TBCI by measuring the 8.57-MeV y-line of MeV peaks, due to interference from fast neutron interaction with oxygen that is in the 6.1-MeV region. The abundance of the 8.57-MeV y-photons are lower, but it has less interference, the only interferences being due to Compton scattering of the nitrogen 10.8-MeV photons. This background was measured by phantoms filled with a nitrogen-only solution and the measurement of the ratio of N counts in C1 region (8.3 to 8.8 MeV) to N counts in N region (9.5 to 1 L.0 MeV). It was found to be independent of phantom size, and thus this ratio and the N counts were used for background correction due to N. Another interference is from random summing of lower energy pulses, measured from water phantom. The precision was found to be 4.9%. as measured from 20 repeating scans. Franklin et used a 238P~Be s o ~ ~ r for c e in vivo PGNAA measurement of cadmium in liver and kidney in order to improve the accuracy. The precision (detection limits) are 6.5 ppm for liver (with skin dose of 0.5 mSv) and 6.4 mg for kidney (with skin dose of 0.9 mSv). However, the accuracy is not such a good one since the calibration is dependent on geometry and thus depends on patient buildup and health. It was shown that the depth of the liver and the kidney affect considerably the counts. Hassan et aL3' described a PGNAA system with 2.5-Ci Pu-Be source for elemental analysis using a Ge(Li) detector. A domestic limonite sample was studied. The neutron source was put in one end of aluminum tube of 70-cm length (12-cm inner diameter and 2-mm wall thickness), whiIe a Ge(Li) detector was positioned at the other end of the tube. The Ge(Li) detector was covered by a closed-end jacket of boron carbide (25-mm thick) lined with a 2-mm thick cadmium sheet. A block of lead was designed especially as a shadow shield between the neutron source and the Ge(Li) detector. The neutron source and the shadow shield were positioned in a cylindrical paraffin wax block.

shadow

shield

utmn source

\

\AI tube

The neutron source detector distance was about 45 cm. The studied samples welt: inserted to fill a wooden box of 70 X 70 X 70 cm3, in which a central cylindrical hole, about 13 cm, exists for the source-detector tube.

With this system, they could measure Fe, Si, Mg, Al, C1, Na, and Mn in the linionite sample.

C. 252CfSOURCES Most PGNAAs studies were done with 252Cfdue to its low y-ray fluence. Another advantage of 252Cf sources is higher integrated exposure before the onset of neutron-induced damage in Ge(Li) detectors. Girratano et al.32performed a feasibility study of a PGNAA facility with a 252Cfsource for analysis of the constituents of bone. They tried to use a Ge(Li) detector, but found that its efficiency was too low for the high-energy y-lines and also the (n,y) in Ge leads to a 1953-keV y-line of Ge that overlapped the 1952-keV line of Ca. Consequently, they used a 5" X 5" Na(T1) detector. Neutrons from the 252CC source were partially moderated by D20 ccntained in a moderator-collimator Lucite tube. The source is surrounded for shielding by heavy water contained in a rectangular polyethylene tank. The tank is shielded by boric acid boxes on all sides of the tank. The amount of D,O in the moderator-collimator tube is changed to obtain an optimal signal-to-noise ratio. The detector is situated 45" lo the sourcesample line. The region between sample and detector is built from a lead, boron-polyethylene, and

source lithium polyethylen.e shield having a tunnel for the prompt photon to reach the detector. The background was measured with a 118"-thick boron-polyethylene plate under the sample to absorb thermal neutrons before hitting the sample. Subtracting this background Ied to a resultant net prompt y-ray spcctrum that was produced almost entirely by slow neutrons. They used this system to measurc elemental contents in bones in vitro. For CayNa, and C1, the minimal detectable amounts of elements were lower for the case of delayed y-rays than prompt y-rays. However, H and P could be detected only by prompt y-rays. Cumrnins et a1.33described the system buiii in Swansea, Wales, U.K. A 200-pg 2S2Cfsource (4.6 X lo8 nls) was housed in a cylindrical steel vessel (90-crn height and 90-cm diameter) filled with borated

beam. They studied the effect of various materials in the back of the source in order to increase neutron flux by reflection. W and Fe were found to be better reflectors than Pb; but since the difference was small, it was preferable to choose Pb since it serves simultaneously as a neutron reflector and as a yray shield. W can also do both but it is considerably more expensive. The primary collimator is a 4mm thick, 15.5-cm diameter steel tube, containing along part of its length a 5-cm thick lead annulus. On top of this collimator. there was another collimator made of borated wax with an internal diameter of 6 cm that provided optimum S2m values for measurement of cadmium (S, 559-keV photopeak counts due to Cd; B, background uncler this pee&). At the end of the collimator, there was a 25-mm thick bismuth filter. The bismuth reduced the y-flux, while having very little effect on the neutron spectrum and flux. The detector is shielded by borated wax and a layer of tungsten sandwich between two layers of lead. This sandwich used the better y-absorption of tungsten compared with lead, but avoided the interference of capture y-rays formed with higher probability in tungsten. The detector was also shielded w i t . natural and enriched 6LiF. The detector itself was a coaxial Ge(Li) detector. It was found that the response is strongly dependent on kidney depth. For a 7.0-mSv dose, it was found that the detection limits me 19 ppm for liver and 3 mg for kidney. Vartsky et al.3' studied the shape of the PGNAA background from a 252Cffacility and techniques to reduce the background. Additionally, they determined the limits of detection for a number of elements. The system is the same system used wich Pu/Be. A 100-pg 2S2Cfsource is housed in a steel conical collimator, which is placed in a cylindrical-shaped, certified transportation container made of borated polyethylene. The top of the cylinder is covered with a 5-cm thick layer of lead. A 1.0-cm thick lead disc is placed just above the source within the collimator in order to reduce the low-energy y-rays originating in the "Cf. The prompt y-rays are measured with two Ge(Li) detectors. The detectors are shielded with a TiF wax cup and polyethylene-boron-lead bricks. They measured mercury through its 368-keV line. Detection limits are 5.4 ppm Hg in the liver and 6.0 mg in the kidney with skin dose of 450 mrem (4.5 mSv). The highest peak for Ca is 1943 keV, but it is located on a high background due to the hydrogen 2230-keV line. Thus, it is preferable to use the 6420- and 4420-keV y-rays (together with the escapes of the first one). The detection limit in the spine is 1.8 g for a skin dose of 1 rem. Nitrogen in the liver was found to have a detection limit of 8.8 g with skin dose of 1 rem (10 mSv). These detection limits are not suficient for detection of I-Ig in unexposed subjects, but it is sufficient to measure Hg in the liver of exposed subjects and can be used to screen exposed individuals who are exposed to Hg-as, for example, dentists who are exposed to mercury amalgams. The main background for Ca peaks is due to neutron capture in the detectors. To reduce it, a 5-cm borated .wax layer is placed between the phantom and the detector. Larsson et built a PGNAA set-up with a 252Cfneutron source to measure body nitrogen in viva The 252Cfsource, emitting 2.5 X 10' nls, is contained in a polyethylene block that forms a collimator surrounded by a water tank. The patient is irradiated from below by a 15 X 80 cm2rectangular neutron field. The NaI(T1) detector is used for detection of the 10.8-MeV photons. To reduce pile-up interferences the pulses are kept short (1.5 ps) by a 1.5-ps fix dead-time ADC. A measuring time of 3 to 5 min is needed to collect 60 to 200 net counts. Ryde et a1.363konstructeda clinicaI instrument for in vivo multielement analysis by neutron activation with a 252Cfsource in Swansea, U. K. In principle, it can be used for prompt y-rays and delayed yrays36 but actual measurements were done only with prompt y-rays in order to measure nitrogenJ7 and c a l c i ~ m ? ~ . ~ ~ The source storage safe is a lead cube of side 140 mm, drilled to accept an aluminum sleeve of bore 16 mm. The safe is surrounded by a composite of shielding material consisting of poly-cast and poly-lead-boron bricks, borated wax, lead, and barytes concrete. For irradiation, the 252Cfsource is transferred through the A1 sleeve to the irradiation position by pneumatic transport using a vacuum/ compression pump. Two irradiation ports are available, differing in the distance from the scanning couch and in the length of the collimator. Two 252Cfsources (each of 4 GBq.) are contained in a small transport capsule fabricated from steel. The irradiation port is chosen by a diverter that is an integral part of the pneumatic transport system. The two-collimator facility was constructed in order to measure calcium with an HPGe detector and nitrogen with a an NaI(T1) detector. Due to the very high energy of the nitrogen photons (10.83 MeV), the efficiency with an 2-IPGe detector will be too low. The collimator is built from bismuth, together with iron for bnckscntterer. These materials were chosen according to extensive studies on the quality and profile of the emergent neutron beam from different combinations of collimator and reflector materials. For each port, further collimators can be used to

reduce the size of the neutron beam, as needed for measurement of cadmium in special organs. Ca is measured with the short collimator, while nitrcgcn and cadmium are measured with the long collimator. Tho HPGe detectors are used simultaneously. each shielded by caps of 'Li-enriched (95.3%) LiF of rr-inimum thickness (4 mm, 1% thcrmal neutron transmission). Depending on the element to be measured, the detectors are further shielded with a composite of borated wax, lead, and bismuth. For nitrogen measurement, two 6" X 6" NaI(T1) detectors were used, with high-stability, high-current capacity photomultiplier assemblies designed for negative bias operation to reduce pulse pile-up. The NaI(T1) detectors are shielded with enriched 6Li-LiF powder and 1-mm thick lead sheets. It was shown that the instrument can measure N, Ca, and Cd in vivo. It was optimized for N and Ca measurements since these elements were seen as the major clinical applications. Although the use of hydrogen as internal standard for N measurement is quite satisfactory for patients with normal body composition, it does not apply to all patients. For this reason, Borovnicar et aL4" measured the spatial sensitivity of an IVNAA facility with lS2Cf.They presented a method for measuring the spatial sensitivity by observing the prompt y-rays of C1 from a CC14 sample. C1 was chosen since its cross-section for thermal neutron absorption is of medium value. Using very high cross-section elements such as Gd2O3will lead to all neutrons being absorbed on the surface of the sample. Consequently, the surface area of the sample, rather than its concentration, will determine the prompt y-ray counts. Nitrogen has too low a cross-section to allow accurate measurement; thus, Cl with a mediurnvalue cross-section (33.2 b) allows for reasonably accurate measurement of a relatively small sample (cylinder of 4-cm diameter and 4-cm height). The y-energy of C1 is lower than that of N (6.1 vs. i0.E MeV), but the absorption coefficient in water is quite close. Their results demonstrate the eSSeclivencss of CC14 for measuring the overall spatial sensitivity of a PGNAA facility for TBN measurement. Several PGNAA systems with a 252Cfneutron source have been built for on-line analysis of coal. Wilde and Henog4' built an experimental system with a 1-mg lSZCfsource (2.3 X lo9 11.1s). The neutron source is moved up and down inside a zircaloy tube by a step motor drive. For analysis, the source is moved into the center of a cylindrical sample holder containing up to 150 kg coal. The y-photons arc measured with an HPGe detector. The detector is shielded from neutrons by paraffin loaded with 95%enriched 6LizC03from the front, and from the back with a thicker layer of paraffin wlth natural Li2C03. The detector is surroslnded by a multilayer decreasing-Z absorber to reduce low-energy y-background arid consequently to reduce the dead-time and pile-out. To further reduce the dead-time, the first quarter of the y-spectrum is suppressed. The systems were optimized by changing the source-d: cctor distance, and to a lesser extent by the thickness of the sample. A 400-s measurement enables the measurement of H, C, Al, Si, S, C1, Ca, Ti, and Fe in coal samples. S and Fe analyses agree with wet chemical ~~ a system for PGNAA analysis with a T f source and analyses within 10 to 20%. H e r z ~ gdescribed an HPGe detector to determine major and minor elements in large samples-up to 130 kg--of hard coal, raw lignite, and raw glass mixtures. Within 1000-s measurement, the system can determine the concentration of H, C, N, S, Na, Al, Si, C1, K, Ca, Ti, and Fe in hard coal and raw lignite. A 400-pg lS2Cfsource is used. The sample is positioned between the neutron source and the y-detector. Thus, the sample itself acts as a neutron moderator, and the primary and secondary y-rays emitted by the source and source housing are reduced. The system is similar to those described previously by Wilde and Herzog."' They compared the PGNAA of hard coal for Al, Si, Fe, Na, K, Ca, and S wilh chemical analysis; quite good agreement was found. Thc sample mass was 103 to 122 kg and was measured for 1000 s. In the case of raw lignite, samples of mass 86 to 124 kg and 1000-s measurement times were used. Very good agreement between PGNAA results and chemical analysis for H, C, Al, Si, S, Ca, Fe, and Ti was observed. It was found that PGNAk results could be used quite accurately for calculation of ash and water content in lignite. For raw glass mixtures, Herzog4' obtained quite good correlation between PGNAA and chemical analysis for S, Na, and Ca. Clayton, Hassan, and Wormald6 described the use of PGNAA with a z2Cf source to measure the concentration of several elements in coal during borehole logging. They found that although a high statistical counting accuracy can be obtained for several elements (Al, Si, S , and Fe) in a meisuring period of 5 min, the derived concentrations are limited by the low neutron capture cross-section of carbon (3 mb). The high concentration of carbon leads to measurable peaks (at 4945 and 3684 keV), but with large statistical uncertainties. They concluded that it is strongly preferable to employ a highenergy neutron source, such as "'Am-Be, in order to include oxygen in the analysis and to obtain a higher precision of measurement of C by using the higher cross-section, high threshold energy scattering

reaction '2C(n,n'y) I2C. They did it in a following paper: but most other workers continue to use s*Cf sources. Senftle and c o - ~ o r k e r sstudied ~ ~ . ~ ~in situ analysis of coal by a PGNAA borehole sonde. The borehole sonde consists of a 252Cfneutron source and an HPGe detector. While previous experiments were done in a borehole only slightly larger than the sonde, later experiments4' were done with a smaI1-sized sonde in an oversized borehole. The sonde had a diameter of 5.1 cm and a length of 122 cm, not including the neutron source. The sonde includes the detector, cryostat, preamplifier, amplifier, power supply, a microprocessor, a memory, and an ADC. The ADC was added in order to avoid the degradation of spectra that often results from transmission of fast analog signals over long cables. The borehole was 25.4 cm in diameter and was studied up to 229 m below surface. Various 2S2Cfsources were used (0.23 to 96 pg), and it was found that better results are obtained the larger the source is. The borehole sonde was found very useful to determine the depth and width of a coal seam and certain aspects of the lithology above and below the seam. However, spectral analyses in oversized boreholes were found to be inferior to those in a close-fitting borehole. The poor results are primarily due to the excess water in the borehole. Senftle and co-workersdsused a PGNAA sonde with an intrinsic germanium detector mounted in a cryostat cooled by a removable canister of frozen propane. However, they measured the delayed y of low energies from 233Paand 239Np. Vourvopoulos and c o - w ~ r k e r developed s ~ ~ ~ ~ a PGNAA faci!ity for on-line elemental analysis of coal, using 252Cf,to study mainly three points: (1) the simultaneous determination of the three elements, Al, C1, and S; (2) the effect of the variations of the CI concentration on the linearity of the results for the other elements; and (3) the utilization of an anti-Compton shield for the suppression of the Compton background. A 10- to 15-pg z2Cf source was placed inside a paraffin container to thermalize the neutrons. Further thermalization occurs in the coal sample (-13 kg). A Pb shield was placed between the neutron source and the coal sample in order to absorb the high-energy Fe y-rays emanating from the stainless steel neutron source casing. A borated epoxy and Pb shield was placed around the HPGe detector to absorb the nonthermalizecl neutrons reaching the detector when measurement was done without anti-Compton shield. For measurement with an anti-Compton shield, a Pb collimator and a sheet of borated epoxy were placed between the coal simple and the HPGe detector. The HPGe detector is surrounded with an annular NaI(T1) crystal of inner diameter 6.5 cm, outer diameter of 19.1 cm, and length of 12.7 cm. The NaI(TI) detector is split axially into four optically isolated quadrants, each one coupled to a photomultiplier. They found that A1 cannot be determined due to the high contribution of the A1 can of the HPGe detector. For CI, they found that PGNAA gave results within 10%of the actual C1 content over a large range of concentrations (300 to 1500 ppm). Sulfur determination by PGNAA agrees with chemical andysis in the range of 0.5 to 6.0% in weight.47Sulfur determinations were done by 3-h measurements. For chlorine, 20-h measurements were performed. It was found that increasing the C1 concentration by a factor of 5 (from natural 300 to 1500 ppm) does not change the results obtained for S6s4' by more than 5%. The anti-Compton annulus shield reduces the Compton background by a factor of 6 in the vicinity of 5.5 MeV, where S is determined (5420 keV). More significantly, it reduces the double-escape peak by a factor of 6. The anti-Compton suppression of the background, as well as the suppression of the single- and double-escape peaks, allow the S photopeak to become more pronounced. However, due to the very tight collimation needed in front of the HPGe detector in the case of an anti-Compton shield, coupled with the loss of the first and second escape peaks, a factor of 10 reduction in S counts results. The factor of 6 suppression of the Compton background was found to be inadequate for a meaningful improvement of the y-ray fits. They suggested that a drawback in the design of the anti-Compton shield was the absence of an NaI(T1) detector in the forward direction, in which the Compton cross-section is the largest. In the present system, the y-ray entry and the HPGe detector were along the same axis of the annulus. They suggested that a design with the HPGe detector at 90" to the incident y-rays should greatly enhance the Compton suppression. Marshall and Zumberge4*described application of PGNAA with a 252Cfsource to on-line measurement of coal quality. They described a prototype of an instrument installed in several plants, mainly in a coal-cleaning plant. The washed low-ash coal moves at a rate of 6 to 10 tonneslh. Two capsules, each of 100 pg 2s2Cf,were used as the neutron source. The two capsules are separated to improve the neutron flux uniformity. PoIyethylene spheres surround the capsules in order to moderate partially the neutrons before entering the measurement volume. The capsules and the slowing-down polyethylene spheres are

enclosed in bismuth to absorb the y-rays from the source and the moderator. The detector is shielded by borated polyethylene and an encapsulated 'LiH y-ray window, whereas the whole instrument is shielded by steel tanks containing water, ethylene glycol, and a basic boron compound. The boron compound serves also to prevent tank corrosion. The 6LiH should be carefully sealed from moisture. The detector is a 6" X 7" right circular cylinder of NaI(T1) that is thermostated at 54 "C to obtain a faster output pulse than that at room temperature. The photomultiplier (PMT) is connected to electronics that process the PMT output into digital signals that are transmitted over cables to the logic and power assembly. They prefer an NaI(T1) detector to a Ge detector due to the higher efficiency of NaI(T1) and since the important elements in coal (C, H, S, Fe, Si, and Al) have prominent prompt lines that can be measured even with the low-resolution NaI(T1). Another advantage of NaI(T1) is the high percentage of the photopeaks. NaI(T1) suffers less from degradation due to neutron interaction than does Ge. It should be mentioned that the resolution of Ge detectors is lower in industrial plants than in laboratories, due to the presence of high levels of vibrations and electromagnetic interference. However, due to the low resolution of NaI(T1) detectors, special care must be taken in processing and interpreting the detector outputs. The output signal is entered into slow and fast amplifiers, which are used both for pile-up rejection and energy resolution. In order to reduce the dead-time, the pulses are not analyzed by a multichannel analyzer, but by 36 parallel discriminators that define 36 levels of energy from 1.24 to 11.17 MeV. The pile-up rejection logics remove most of the pile-up signals, but 5 to 15% still remain. These pile-up pulses are stripped by calculation of pile-up formation, measured from measurements of one- and two-source spectra. These measurements lead to the coefficients (C,w)for the correction formula Nj = S j + 2 Cjkl St S,, where Sj are the pile-up free counts in the J energy window and Nj is the raw U

counts in this window. The summation is done over the instrument's 40 energy windov;~. The obtained spectra (counts in 40 energy windows) are calculated as a linear combination of 23 "calibration" spectra. These are the spectra of prompt y-lines from 14 pure elements: H, C1, C, S, Fe, Si, N, Al, Ti, Na, Ca, K, Ni, and Cr, together with two inelastic scattering spectra from C and 0, four background spectra, and three spectra reflecting neutron interaction with the detector. It was found that the instrument gives good results for sulfur content, ash content, and the calorific value (based primarily on the measurement of carbon and hydrogen). A very good agreement between this PGNAA instrument's results and laboratory analysis was found for concentration of C, 1-1, N, S, Al, Si, Fe, and Ca. Zaghloul et a1.49 used a PGNAA facility to measure Sm, Gd, and Mn contents in phosphate and monazite rock samples. The neutron source was 27 p g 252Cf(6 X lo7111s) moderated and shielded with paraffin and plexiglass. The detector was Ge(Li) covered by a close-end jacket of 'Li2C03 (95.6% en-ichment) of 1-cm thickness. The results show a variation of about 25% from the values determined by radiochemical neutron activation analysis. Thk was explained as due to the low signal-to-background ratio in the PGNAA measurements. It was suggested in order to increase the 252Cfstrength, to reduce the background, and to increase the detection solid angle. Yuren et aL50developed a 252Cf-basedPGNAA facility for the aluminum industry. The source is 100 to 200 p g 252Cffilled in a fm metal tube acd inserted into the measurement piping. Between the neutron source and the Ge detector there is a Bi plug, neutron moderator, and Cd layer. The detcctor is large-volume (40% relative efficiency) n-type HPGe connected to very fast preamplifier and amplifier with shaping time of 0.25 p s to reduce pile-up. The system is calibrated for the main components in Al-pulp and the counts due to each element are calculated by multivariate linear regression. The PGNAA data for H20, SiO, Fe203,A1203,CaO, and Na20 were compared to chemical analysis and very good agreement wzs obtaiced except for Fe203where disagreement of 10 to 20% was found, due to the relatively low concentration of Fe203(2.4 to 2.9%). Mikesell et aL5' employed a PGNAA system with 252Cfsource and HPGe detector to obtain a partial elemental analysis of a diabase rock down a borehole without taking a core sample. The concentrations of Al, Fe, Ca, Na, K, Ti,Mn, Cr, V, Ni, and B were measured from the counts under their peaks in the prompt y-spectra, using silicon as a calibration standard. Mg was calculated as residual to complete for 100%. The source was 52 pg z2Cf, and the source-to-detector distance was 49 cm with a lead shadow shield interposed between the source and the detector. No B neutron shield was used. Wiggins et a1."2 measured manganese nodules in seawater using a U2Cf-basedPGNAA facility with a Ge(Li) detector. They used 120 p g z2Cf contained in a platinum shield and doubly enclosed in a stainless steel capsule.

Grigorev et al." described a FGNAA facility with 252Cfthat can work in a ship-borne laboratory for the purpose of analysis of ocean-bottom deposits, ferromanganese concretions. The source and sample are embedded in a large box (100 X 80 X 80 cm3) filled with paraffin. The box is covered with yshield (100-mm thick lead) and neutron-stop blocks (60 to 80 mm). The %?Cf source has a neutron of lo8 nls. The sample with a mass of about 1 kg is inserted into a 10 X 10 X 10 cm3 lucite cassette and inserted through a vertical channel. The source-sample distance was chosen to be 3 to 5 cm since in this range the product f X E is maximal; f is the flux of neutrons and E is the geometrical yield of the y-rays in the detector. The Ge(Li) detector is shielded from neutrons by cadmium. The sample-detect~rdistance and the size of the collimators and y-filters used in front of the detector were optimized experimentally to obtain the best SIN ratio. They found that detection limits are 1.6%, I S % , and 0.15% for Mn, Fe, and Co, respectively. These figures are considerably lower than the average concentration of these elements in the ferromanganese concretions. Chung and c o - w ~ r k e r s developed ~~~j a '"Cf-PGNAA facility to determine in situ pollutants in lake water. It is discussed broadly in Chapter 7. Charbucinski et al.56developed a PGNAA facility for the coal industry, for salinity determinations, and for the mining industry. As an example for the mining industry, they studied the determination of manganese content in manganese ore. The neutron-y probe used in the field logging contained 2 pg 252Cfseparated 15 cm from a 2" X 2" NaI(T1) detector. Between the detector and the neutron source there is 5 cm bismuth shielding. The total length of the probe is 1 m. In order that the thermalization of the neutrons will not be very dependent on the field condition, a thermalizing nose cone was added to the probe; it was fabricated from high-density polyethyIene. The spectra were taken for several energies windows. The concentration of the manganese was calculated as an average from its severaI peaks.

11. 14-MeV NEUTRQN GENERATORS Oxby and co-workers" studied the possibility of in vivo PGNAA with a small pulsed neutron generator. They used a seated (d,t) neutron tube mounted inside a steel canister, which emits lo8 neutrons in a pulse width of 15 ps, at a rate of 1 to 3 pulses per second. They placed the neutron tube about 60 cm away from a 25-1 polyethylene container containing an aqueous solution of 6.5 kg urea. The neutron tube is shieIded with a 30-cm layer of iron blocks surrounded by a 30-cm layer of concrete blocks. The prompt y-rays are detected by a 4" X 2.5" NaT(T1) detector adjacent to the container but at 90° to the neutron beam. The detector is shielded by 5 cm lead. For measurement of nitrogen, the detector was set to acquire events starting about 70 p s after the beginning of the neutron-pulse and finishing about 300 p s later. These times were chosen in order to avoid counting while there was a significant fast neutron fluence, and to count only during the period of moderation of neutrons to thermal energies and capturing of the neutrons, as the life of a neutron in water is about 200 ps. They found that the amount of activation of nitrogen was too low, causing too low an activity i n the 10.8-MeV region. They suggested that this can he overcome by using a pulsed neutron generator with higher repetition rate. The optimum time between pulses is approximately 200 ps, which allows for a repetition rate of 5000 s-I. Zamenhof et aLS8studied, hy Monte Car10 simulation, the feasibility of in vivo measurement of various elements by PGNAA. They found that spectral interference of nonradiative capture origin can be completely eliminated by pulsing the detectorlspectrorneter system in anticoincidence mode with the neutron source. Calculations based on the results of the Monte Carlo simulation and on the efficiency characteristics of a Ge(Li) detector suggest that the primary limitation of the~proposedtechnique would be interelement spectral interference rather than inadequate detector sensitivity (efficiency). They designed an experimental facility for in vivo PGNAA using a 14-MeV neutron generator. The generator can be operated either continuously or in pulses. For their studies, they used a 20-1 phantom of water or aqueous solution. They used pulse mode with a repetition rate of 2000 s-'. The 500-ps cycle was set to four stages: 5 ps irradiation (neutron pulse); t p s waiting time; 5 ps counting time; and (490 t) p s waiting time. It was found that there is a large decrease in the total y measured when t is increased from 10 to 50 ps (decrease by a factor of 40), whereas the prompt y-rays (as measured by the 2.2MeV peak from H) is maximilf for t = 50 ws. Comparison ot measurement of water phantom both in pulsed mode (10 ps irradiation, 10 ps wait, 300 p s count, 10 ps wait) and continuously, normalizing to the same number of counts at 2.2MeV, show that the pulse operation decreases the background as well as peaks of inelastic scattering. Using a 20-1 Standard-Man-equivalent phantom, they were able

to detect [in a Ge(Li) detector] the y-lines from N, C1, Ca, K, C, Fe, Al, and Pb. Part of the C y-rays is due tp background-prompt y-rays from construction and shielding materials. They concluded that Id, C1, Ca, and N could be simultaneously measwed in human subjects at very acceptable doses (100 rnrad). If the energy resolution could be improved by a factor of 2, sulfur and carbon could perhaps also be measured. The main reason for the ION resolution was the high count rate in the low-energy range. This can be overcome by a low-energy discriminator between the preamplifier and the amplifier. Kyere et aLS9studied the feasibility of measuring total body carbon by PGNAA using the inelastic scattering reaction rather than the radiative capture. They used a neutron generator with a maximum output of 2.5 X lo9 s-' that is incident laterally upon the supine patient. The size of the beam is 30 X 180 cm2. The y-rays are detected by 6" X 5" NaI(T1). The detector is mounted inside a rectangular enclosure having a 10-cm thick iron wall facing the neutron generator and a 5-cm thick lead wall on the opposite side to attenuate y-radiation generated in the walls of the room. The space between the detector and the enclosed walls was filled with boric acid contained in plastic bags. Using phantoms with sugar solutions, it was found that the relative standard deviation is less than 2%. The dependence on body geometry was found to be less than 5%. Kehayias et a1.* employed a 14-MeV neutron generator in the pulse mode at 10 kHz to measure total body carbon and hydrogen in a two-phase (two interval times) counting system. Yankuba et aL6' studied the errors caused by random summing in in vivo PGNAA. They described a technique for estimation of losses and gains caused by random summing effects in spectra acquired at high counting rate. The correction is based upon studying the effect of two shaped pulses from a double-pulse generator, varying the time between the two pulses applied to the input of the nlain amplifier. The method developed was used in in vivo PGNAA. They used a neutron generator operated at a continuous output of approximately lo9 nls. Two heavily shielded NaI(T1) detectors (5" X 4") are positioned above the supine subject and three more underneath. A 70-kg anthroponlorphic phantom containing the bulk elements of the body in aqueous solution was irradiated and thc effect of the correction for the summing up events was studied. It was found that the correction increases the number of counts in the carbon peak by 8%. In the peak of nitrogen, the correction decreases the number of counts by 54%. This result indicates that lower counting rates should be used. Sutcliffe et aL61studied the use of a pulsed neulron generator for simultaneous measurement of C , H, and 0 , in vivo. They used the pulsing and measurement over different time intervals to discririijna:~ between y-rays from inelastic scattering and radiative capture. They modified a continuous high-outpur neutron generator to operate in a pulse mode with a frequency of 1 kHz. The neutron beam is collimated with a collimator having rectangular aperture 3 . ~ wide ~ 1 and 10-cm deep. The collimator is composed of 30 cm iron, 24 cm wax and polyethylene, 2 cm lead, and again 10 cm iron and 20 cm wax and polyethylene. The neutron generator is enclosed in a shield of concrete blocks. A 6" X 5" NaI(T1) detector was used for y-measurement. The dstector was located inside a lead castle. The phantom was completely surrounded with lead except for entry and exit ports for the neutron beam. The neutron pulse was 15 ps. During 18 p s from the beginning of the neutron pulse, y-rays due to inelastic scattering from 0 and C were measured. After a delay of 26 ps, the counts of the prompt y-rays due to radiative capture were measured for 600 ps. The main pe& is the 2.2-MeV peak from H. The counts in that time in the 2.2-MeV region are 81% of the total counts in that region (including the other t m e intervals). A small peak in this time is the one at 6.13 MeV from oxygen. However, the counts due to oxygen in this time interval are 10 times less that those in the first 18 ps. The'dose per measurement was 5 pSv and the reproducibility was t8.3% for 0 and t9.3% for C, for measurement of 0.5 h. Better reproducibility will be obtained for longer measurements. Sutcliffe et al.63 studied the variation with depth of the inelastic scattering by measuring the counts from carbon after varying the thickness of water. It was found that the half-value thickness of water for the 14.4-MeV neutron is 9.2 2 1.1 cm. By using various waiting times between the beginning of the neutron pulse and the beginning of collecting radiative y-rays from hydrogen (2.223 MeV), they found that the half-life for radiative capture is 113 + 7 ps. This indicates that about 500 ps is required between successive neutron pulses to allow for separation of radiative capture and inelastic scattering events. Grau et aLMused a PGNAA logging sonde, incorporating a pulsed neutron generator and NaI(T1) detector, for measurements of geological for ma ti or,^. They found that the most difficult task 11. interpreting these measurements is to separate signals generated by the elements in the formation rock and fluid from those due to elements inside the well bore. They studied the effect of porosity and water salinity

on this interference. They discussed methods for enhancing the spectral contribution from the earth formation, relative to that from the borehole.

Ill. ACCELERATORS Many of the preliminary studies on PGNAA were done with accelerators; however, very few were done later. In order to make this chapter more comprehensive, a discussion of the use of accelerators will be included. Fremlin's grouf5 used a 10-MeV proton beam from a 152-cm cyclotron impinging on a lithium target to produce neutrons with an average energy of 2.6 MeV with flux of 3 X 10" n/cm s for in vivo measurement of total body nitrogen. Using a pulsec! neutron beam and gated circuits,66they reduced considerably the pile-up of y-rays from lower-energy gammas. By collimating the neutrons, pulsing the neutron beam and shielding the detectors, they obtained an accuracy of 22% with an incident dose of 100 rnrem. The neutrons emitted vertically from the target passed through a 3-m concrete shaft fitted with a series of steel and lithium-doped polyester collimators to give a defined neutron beam of 25cm diameter at the level of the subject. The prompt y-rays are detected by three 5" X 5" NaI(T1) detectors. The neutron beam is delivered in 10-ps pulses at a rate of about 6 kHz. Counting is done only between neutron pulses. Biggin et aL6' used the same system to measure cadmium in viva About half of human cadmium is concentrated in the liver and kidneys where it is bound to a specific protein. The concentration of cadmium in normal liver and kidney tissues is in the region of 1 to 5 ppm and 10 to 30 ppm, respectively. They found that the minimum detected level of cadmium is 2.0 ppm for a dose of 1 rem. Better results can be obtained using larger Ge detectors or, preferably, using several detectors. It was not possible to use lead to shield the detector because they could not find cadmium-free lead. The same g r o ~ pdeveloped ~ ~ . ~ ~ the use of whole-body hydrogen as an internal standard for wholebody nitrogen, using the 10.8- and 2.2-MeV peaks ratio. Hydrogen mass is estimated from body parameters. Phantom and cadaver studies indicated that nitrogen mass can be estimated to 2 4 % or better. Chettle et aL7" found that nitrogen in clothing or skin gives only about one third the signal of the same amount of nitrogen inside the body. This is due to the fact that the high-energy neutrons were not thermalized on the surface. If the N/L-I ratio is about the same in the skin and total body, this lower thermal flux on the surface will not significantly change the data. The thermal flux on the body surface can be increased by the use of graphite neutron reflectors. Fletcher et aL7' studied the Cd content of liver and kidney by PGNAA, using both 10.2-MeV protons on Li and 238PU/Beneutron sources. From the data, estimates for the critical concentration of cadmium in the renal cortex and the biological half-life of cadmium in liver were obtained, Kacperek et alSn used a 2-MeV deuteron beam from a 2-MV Van de Graaf generator to produce a pulsed beam of 5.2-MeV neutrons by impinging on a deuterated titanium target using the d(d,n) 3He reaction. Silicon is determined by the inelastic scattering reaction 2RSi(n,n'y)*%i that emits 1779-keV y-rays. The pulsing separated the inelastic scattering y-rays from those due to thermaI neutron capture. Thus, the 1779-keV y-rays from 28Siare measured only during the 10 p s of the neutron pulse. Between pulses there was 300 to 700 p s waiting time. The interference from the 1779-key y-ray of 28A1due to "P(n,~i)~~Al was minimized by proper choice of the incident neutron energy.

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J., Prompt gamma-ray neutron activation analysis facility testing by Sm, Gd and Mn determination in rock samples. J. Radioanal. Nucl. Chem. 109,309, 1987. 50. Yuren, L., Yanxin, L., Yali, X., Yonghai, W., Youling D., Jin, T ., Ronyan, M., and Seymour, R., Development and applications of an on-line thermal neutron prompt-gamma element analysis system, J. Radioanal. Nucl. Chem. 151 , 83, 1991. 51. Mikesell, J. L., Senftle, F. E., Anderson, R. N., and Greenberg, M., Elemental concentrations in diabase determined by high resolution borehole gamma-my spectrometry, Nucl. Geophys. 3, 501, 1980. 52. Wiggins, P. I?., Duffey, D., and El Kady, A. A., Neutron-capture gamma ray studies of marine manganese , nodules using a nuclear reactor and a californium-252 source, Anal. Chim. Acta, 6 1, 421, 1972. 53. Grigorev, A. I., Ivanenko, V. y,and Kustov, V. N., Neutron-radiation analysis of ferromanganese concretions in the ocean, Sov. Atom. Energ., (English translation) 62, 338, 1987. 54. Chung, C. and Tseng, T. C., I n sit11 prompt gamma-ray activation analysis of water pollutants using a shallow 2-'2Cf-HPGe probe, Niicl. Insw Meth. A 267, 223, 1988. 55. Chao, J. H. and Chung, C., Optimization of in siru prompt gamma my analysis of lake water using a 1-IPGe252Cfprobe, Nucl. Instr: Meth. A 299. 65 1. 1990. 56. Charbucinski, J., Aylner, J. A., Eislet; P. L., and Borsarx~,M., Quantitative and qualitative applications of the neutron-gamma, Mtcl. Geophys. 3, 475, 1989: J. Charbucinski, J., Eider, P. L., and Borsaru, M., Quantitative borehole logging based on neutron excited gamma-reactions. Nucl. Geophys. 2. 137, 1988.

57. Oxby, C. B., Oldroyd, B., and Carthy, I. D. M., Experiences with a small pulsed neutron generator, Plrys. Med. Biol. 25. 137, 1980. 58. Znmenhof, H. G., Deulsch, 0. I., nnd Murray, B. W., A feasibility study of prompt capture gamrda in vivo neutron activation analysis, Med. Phys. 6, 179, 1979. 59. Kyere, K., Oldroyd, B., Oxby, C. B., Burkinshaw, L., Ellis, R. E., and Hill, G. L., The feasibility of measuring total body carbon by counting neutron inelastic scatter gamma rays, Phys. Med. Bid. 27,805, 1982. 60. Kehayias, J. J., Ellis, K. J., Cohn, S. H., Yqumura, S., and Weinstein, J. H., Use of a pulsed neutron generator for in vivo measurement of body calrbon "In vivo Body composition Studies", IPSM3, Ellis, K. J., Yasumura, S., and Morgan, W. D., Eds., 1986, 427. 41. Yankuba, S., Oldroyd, B., Metcalfe, S. C., Godfrey, J. E., and Oxby, C. B., Examination of errors caused by random summing in in vivo prompt gamma-ray analysis, Phys. Med. Biol. 34, 1089, 1989. 62. Sutcliffe, J. F., Waker, A. J., Smith, A. H., Barker, M. C. J., and Smith, M. A., A feasibility study for the simultaneous measurement of carbon, hydrogen and oxygen using 14.4 MeV neutrons, Pltys. Med. d i d . 36, 87, 1991. 63. Sutcliffe, J. F., Smith, H. A., Waker, A. J., Barker, M. C. J., and Smith, M. A., Inelastic scattering of pulsed 14.4 MeV neutrons as a function of depth in tissue and the half-life for radiative capture, Phys. Met!. Biol. 36. 643, 1991. 64. Grau, J. A., Antkiw, S., Hertzog, R. C., Manente, R. A., and Schweitzer, J. S., in siru neutron capture spectroscopy of geological formations, in Neutron-Capture Gamma-Ray Speclroscopy and Related Topics, Von Egidy, T., Gmnenwein, F., and Maier, B., American Institute of Physics 1985, 799. 65. Biggin, H. C., Chen, N. S., Ettinger, K. Y, Fremlin, J., Morgan, W. D., Nowotny, R., Chamberlain, M. J., and Harvey, T. C., Determination of nitrogen in living subject, Nature (London), 236, 187, 1972. 66. Harvey, T. C., Jain, S., Dykes, P. W., James, H., Chen, N. S., Chettle, D. R.,'Ettinger, K. V., Fremlin, J. H., and Thomas, B. J., Measurement of the whole body nitrogen by neutron activation analysis, Lancet, 2, 395, 1973. 67. Biggin, H. C., Chen, N. S., Ettinger, K. V., Frernlin, J. H., Morgan, W. D., Nowotny, R., Chamberlain, M. J., and Harvey, T. C., Cadmium by in vivo neutron activation analysis, J. Radioanal. Chem. 19,207, 1974. 68. Dabek, J. T., Vartsky, D., Dykes, P. W., Hardwicke, J., Thomas, B. J., Fremlin, J. H., and James, H. M., Prompt gamma neutron activation analysis to measure whole body nitrogen absolutely; Its application studies of in vivo changes in body compositions in health and disease, J. Radioanal. Chem. 37, 325, 1977. 69. Vartsky, D., Prestwich, W. V., Thomas, B. J., Dabek, J. T., Chettle, D. R., Fremlin, J. H., and Stammers, K., The use of body hydrogen as an internal standard for the measurement of nitrogen in vivo by prompt neutron capture gamma ray analysis, J. Radioanal. Chem. 48, 243, 1979. 70. Chettle, D. R., Fletcher, J. G, Downey, S. P. M. J., Scott, M. C., James, H. M., and Higgens, S. C., Measurement of nitrogen in vivo by neutron activation analysis: Further developments and applications, J. Radioanal. Chem. 7 1, 533, 1982. 71. Fletcher, J. G., Chettle, D. R., and Al-Haddad, I. K., Experience with the use of cadmium measurements of liver and kidney, J. Radioanal. Chem. 7 1, 547, 1982. 72. Kacperek, A., Evans, C. J., Dutton, J., Morgan, W. D., and Sivyer, A,, A system for the determination of silicon in the human lung using neutrons from a 2 MV Van de Graaf generator, J. Hadioanal. Nucl. Chern. 114, 165, 1982.

Chapter 6

Prompt Gamma Activation Analysis Neutron Beams Richard M. Lindstrorn and Chwshiro Yonezawa

CONTENTS I. Introduction .............................................................................................................................. 93 11. Advantages of Guided Neutron Reams .....................................................................................93 LII. Production of Guided Beams ..................................................................................................... 93 IV. Apparatus .................................................................................................................................... 95 V. Results and Applications ............................................................................................................. 96 VI. Accuracy ................................................................................................................................... 96 VII. Facilities for PGAA with Guided Seams ................................................................................ 97 VIII. Trends ..........................................................................................................:..................... 9s References ......................................................................................................................................... 99

.

f . INTRODUCTION The use of guided neutron bennls from research reactors has opened new applications in prompt gamma activation analysis (PGAA) because of the extraordinary freedom from background radiation, both fast neutrons and y-rays, that these beams afford, As a consequence, the y-ray counting efficiency, and hence the analytical sensitivity, can be high. This new technology promises substantially broadened applicability of reactor-based PGAA ant! applications to new areas of materials and nuclear science.

II. ADVANTAGES OF GUlDED NEUTRON BEAMS In recent years, neutron guides have been installed in over a dozen research reactors throughout the world. The basic physics have been understood for decades; a concise summary of the concepts has been given by Maier-Leibnitz.' Using reflective neutron optics, these devices transport desirable slow . neutrons with high efficiency to experiments located many tens of meters from the reactor. If the guide is curved or contains an efficient filter, the spectrum at the outlet contains only slow neutrons, with no fast neutrons (which would need to be thermalized by heavy hydrogenous material before being absorbed) or y-radiation (which would require massive high-Z shielding). Consequently, the y-ray detector may be located close to the sample without an unacceptably high background counting rate; this leads to improved efficiency for detecting the capture y-rays. This in turn gives correspondingly improved measurement sensitivity for all elements, and the possibility of greatly improved specificity by y-y coincidence counting. Since the local shielding need not contain hydrogenous thermalizing material, the background counting rate of the hydrogen capture y-ray can be low, and the detection limit for this important element is thereby greatly improved. The improved quality of the data that can be obtained with guided beams may be seen by comparing Figures l a and Ib. Figure l a is the low-energy portion of the capture y-spectrum of coal fly ash, irradiated in a guided cold neutron beam at the JRR-3M reactor at JAERL2 The peaks are distinct from the baseline, even down to X-ray energies. By contrast, the spatrum of the same material (Figure lb) irradiated in a collimated (but unfiltered) thermal neutron PGAA beam3 at the NBSR at NIST is dominated by a continuum of y-rays, originating from the reactor core and from neutron capture in the inner end of the beam tube and scattered from the sample into the detector.

I!!.

PRODUCTION OF GUIDED BEAMS

A neutron guide is 'an evacuated or helium-filled tube, generally rectangular in cross-section and made from polished glass slabs coated with a reflecting film of nickel. Just as with light in an optical fiber, 0.8493.5 1d9-9/951$0.O0+$.50 0 1905; by CRC Prcss. Inc.

Gamma-ray energy, keV

Gamma-ray energy, keV Figure 1 The low-energy portion of the capture-y-spectrum of NlST SRM 1633a Fly Ash. In Figure 1a .~ (upper), the sample was irradiated in a cold neutron beam from a bent guide at the JRR-3M r e a c t ~ rBy

contrast, in Figure 1b (lower), an unfiltered thermal beam at NlST was used.3 Both spectra were measured with Compton suppression.

neutrons striking the inner surface of the guide at an angle of incidence less than a critical angle are totally reflected. Thus, a beam can be transported through the guide for long distances with little loss of flux. Inasmuch as the refractive index for neutrons is a few parts per million less than unity, external reflection occurs at a very small critical angle. The efficiency of a guide (i.e., the fraction of the neutrons entering the guide that are reflected and transported to the outlet) is proportional to the square of ihc critical angle 0,. The critical angle in turn is proportional to the neutron wavelength; numerically, 0,

A);

= 0.00173h for Ni (8, in radians, h in this is 0.18"for thermal neutrons (h = 1.8 A at velocity v = 2200 m/s) and 0.40" at 4 A, the most probable wavelength of a 30-K Maxwellian distribution. Thus, the slower, longer wavelength neutrons are collected and transported more efficiently. The largest critical angle for a single substance suitable for a mirror coating is obtained with 5aNi, for which ec = 0.47' at 4 A. Multilayer Ni-Ti supermirrorsJ have been prepared with critical angles more than three times greater than 58Ni.Producing the large areas of supermirrors required for use in long guides is technicaIly difficult at present. Most guided beams are associated with a cold neutron source, which is a low-temperature moderator placed adjacent to the core of the reactor. Liquid hydrogen is the most common moderator, but supercritical H2,D20 ice, liquid methane, and solid mesitylene [(CH3)3CbH31 have also been used. Reactor neutrons are slowed to near thermal equilibrium with the cold moderator, so that the peak of the energy distribution is shifted to lower energy and thus longer wavelength. The average neutron temperature may be reduced from 300 to 30 K so that the average neutron wavelength is tripled. In addition to the greater transport efficiency of guides at long wavelength, cold neutrons are desirable for PGAA and other nuclear reaction measurements because of their increased capture probability. Most slow neutron absorption crosssections are proportional to the wavelength (inversely proportional to velocity: the "llv law") so that each cold neutron is three times as effective in producing the analyticaI signal as is a thermal neutron. Thermal neutrons can be guided as well, though with lower efficiency than cold neutrons. Since the critical angle is smaller at shorter wavelength, a smaller fraction of the neutrons entering the guide emerge from the exit. On the other hand, the neutron flux density in a thermal guide is still higher than it would be with a nonreflecting collimator. In addition, with a thermal guide, the substantial engineering difficulties of operating a low-temperature source adjacent to the reactor core are avoided. In addition to the desirable slow neutrons, other radiation enters the reactor end of a neutron guide: fast (epithermal and fission) neutrons, and y-rays from fission in the fuel and neutron capture in the reactor and beam tube components. For a simple guide shaped as a straight open tube, the slow/fast neutron flux ratio and the slow-nly ratio increases as the square of the distance from the source. Two methods have been used to remove the remaining undesirable components and produce beams of purely slow neutrons. In the first approach (used at KI;A5 and NfST,"for example), a filter is inserted into the beam. Cold bismuth removes fast neutrons and y-rays from the beam without greatly affecting the slow neutron spectrum. The filter at NTST is 127 mm Be and 178 rnm single-crystal Bi, both at liquid nitrogen temperature. Nearly all the neutrons with wavelength shorter than the Be Bragg cutoff at 3.95 A (energy E > 5 meV; velocity v > 1000 d s ) are removed. This cutoff wavelength is longer than optimum for PGAA, SO some analytically useful slow neutrons are- removed as well and the capture rate is lower than it would be without the filter. The resulting beam has a cadmium ratio of lo4 or more and only a few m R h of direct y-radiation. 4 second solution (used for example at JAER17 and Cornel18) is to include a curved section of neutron guide, approximated by several straight segments, between the reactor and the PGAA instrument. Slow neutrons are reflected around the curve while fast neutrons and y-rays, which travel in straight lines, are removed from the beam. A difficulty is that the resulting flux distribution in the guide is nonuniform, with more neutrons at the outside of the curve. This gradient can be reduced (but not removed entirely) by adding a straight guide after the bend, long enough for several reflections?

IV. APPARATUS The equipment used for guided neutron PGAA is the same as that described elsewhere in this volume for thermal neutrons, with the exception that the final neutron collimator, the bearnatop, and the sample shielding can be made of a pure absorber such as 6Li or 'OR without the need for a bulky thermalizing medium. The beam stop need only weigh a few tens of grams, not a ton. This feature of high-quality beams was first exploited at. Kyotoioto construct a compact target-detector assembly shielded by LiF tiles; these tiles were later used effectively at JAERI? Metallic 6Lislabs,'i and later s m d l 'LiF tiles." have been used at ~ ~ l i cand h , fused 'Li,CO, platest3and 'Li loarlcd glass'' at NIST. When small quantities of hydrogen are to be measured, this element should be avoided as much as possible in the materials of construction for the appara!us. Activc Compton suppression is ;t valurtble method of improving the signa1:background ratio. If the snmp!e-detector distance is very short, the y-speckometer system must be able to manage high count rates.

V. RESULTS AND APPLICATIONS Gains in sensitivity with cold neutrons, compared with thermal neutron beams, come from two concurrent effects. Because of the efficiency of neutron guides, the flux density of a guided beam is greater than in a collimated bean1 at the same distance from the reactor. Since most cross-sections have a llv dependence on the neutron velocity v, cold neutrons are several times more effective in inducing analytically useful capture reactions than thermal neutrons, so that the effective flux on target can be high. In addition, since the shielding can be thinner and more compact, the y-ray detector can be closer to the sample and thus the detection efficiency can be better. These two effects combine to give substantially better sensitivity with cold neutrms when compared to a thermal neutron system at the same reactor. In addition, the purer spectrum from a guided beam leads to lower background, so detection h i t s are further improved. This advantage is particularly pronounced for hydrogen. Comparing the two PGAA systems at NIST, the hydrogen background in the absence of a sample is about 1 mg for the thermal neutron instrument2 and a few micrograms for cold neutrons.ls This makes practical the nondestructive analysis of H in a varied suite of sample materials at much lower concentration than has been previously possible. As a few examples from work at NIST, s e ~ e r a samples l of C60fulierene and its derivatives liavc been analyzed, with the samples sealed in an aluminum can for subsequent measurement by neutron scattering (e.g., Neumann et a1.I6). Hydrocarbon contamination was found to be common in the earliest macroscopic samples, as much as 0.9 wt % H being found in one. One specimen of superconductor with nominal composition R b & , was determined to contain 0.9 0.2 hydrogen atoms per rubidium; apparently, this specimen had been exposed to moisture. Sulfur, hydrogen, and carbon were all measured in a sample of sulfonated fullerene C60(S04H)8,confirming the expected composition. A 1-pm film o i hydrated glass on a silicon wafer was found to contain 1.0 1.3 k g H/cm2, which corresponds to 6 r 9 wt % water in the film.

+

+

VI. ACCURACY Despite the improved measurement quality obtainable with guided neutron beams, there are also disadvantages, particularly for cold neutrons. The higher capture cross-section at long wavelength brings better sensitivity, but also a proportionally higher scattering cross-section. Scattering of thermal neutrons in hydrogenous samples IS well known to lead to analytical errors of as much as 20%, depending on sample size and shape.17J8Although the use of near-spherical samples minimizes the effects of s c a t ~ e r i n g ,the ~~.~~ phenomenon is more serious and less well studied for cold neutrons with their larger cross-sections. A fur;her complication is that cold neutrons gain energy by elastic scattering from a room-tcmperature sample, and therefore the capture rate depends on the temperature of the sample; we have seen a 20% higher capture rate in a 1-g sample of organic material when the sample was freshly cooled with liquid nitrogen, compared with the same sample at room temperature. Until a thorough study of cold neutron scattering is perfomsd, cold neutrons will be of more value in the analysis of metals and other inorganic materials than of highly scattering materials such as biological tissues and hydrocarbons. Nearly all analytical biases in PGAA disappear, however, if elemental ratios are measured; indeed, internal standards have been used in several laboratories to improve a c ~ u r a c y . ~The ' - ~basis ~ for accuracy is shown as f a l l o ~ s For . ~ a given set of experimental conditions (i.e., reactor power, con~position,and dimensions of sample, sample-detector distance), the net peak area Ax (in counts) of a capture y-ray from mass mx of element x in the spectrum of a homogeneous sample of volume V irradiated for a time T is in general given by:

where No is Avogadro's number, 8 the abundance of the capturing isotope, I' the y-ray yield in photons per capture, M the atomic weight, a the capture cross-section, and K a correction factor near unity LO account for pulse pile-up and other rate-related effects. The neutron flux density 4 varies over the volume of a finite sample because of beam heterogeneity, neutron scattering, and absorption (selfshielding). The counting efficiency E is also a function of position within the sample; numerically, the

gadient is about 1% per millimeter at a sample-detector distance of 25 cm. We assume that the shape of the E(E~) function is invariant over the small sample volume V (i.e., that y-self-absorption is small), and replace the integral over volume by a function of E, alone multiplied by a geometrical factor f . Similarly, we assume that the neutron spectrum and time parameters are separable. Normalizing to the corresponding exljression for the peak area of a monitor element s (almost any element) in the same sample, the space and time factors cancel.

We now consider the speciaI case of cold neutrons. At an energy much lower than the first resonance, the capture (or scattering) cross-section a in the laboratory system is inversely proportional to the neutron velocity v24and for nearly all nuclides a = uo vJv, where uois the 2200fmls cross-section. Therefore, the ratio of the reaction rate integrals becomes simply the ratio of values of uo.Making this substitution, we then obtain the relation between the ratio of experimentally measured quantities and the corresponding ratio of physical constants.

.

The final term is entirely ana!ogous to the k,-,factor used in oligomonitor neutron activation a n a l y s i ~ . ~ By measuring ratios of peaks in a single spectrum, a11 biases except the error in relative efficiency cancel. The k formulation thus provides the framework for the multielement calibration of a PGAA instrument using ratios of measured counting rates per gram of target element. As with radioactivity, the inclusion of neutron spectral parameters makes the k values transportable among cold and thermal neutron instruments. The model has been tested23by measuring the capture rates of cold neutrons by K and Br, in a series of 13-mm disks of varying thickness made from a homogeneous mixture of stoichiometric KBr with In these samples, the sensitivities (counts s-'g-I) of'the elements were urea or deuterourea (CD40N2). nearly a factor of 2 lower in the thickest 'H-urea samples because of self-shielding, scattering, and geometry differences. However, as Equation 3 predicts, the K/Br sensitivity ratio was found to be independent of sample mass and 'H content. A recent example of the use of elemental ratios at NIST is the determination of hydrogen concentrations in different locations on an entire titanium compressor blide from a gas turbine engine, where the irregular geometry made the sample mass ill-defined; consequently, W i ratios were the only measurements possible. In other cases, such as the measurement of stoichiometry of Cnoderivatives mentioned above, molar ratios have been in fact the quantity of interest to those who supplied' the samples.

-

VII. FAClblTIES FOR PGAA WITH GUIDED BEAMS Prompt y-activation analysis has been performed with guided beams at a number of reactors throughout the world. The characteristics of each installation are given in this section, and sensitivities for boron analysis are compared in Table 1. Also included in the table for comparison are two thermal beam facilities. Institut Laue-Langevin (ILL), Grenoble, France Two groups have reported PGAA measurements at the I-WR reactor in Grenoble. Henkelmann and Born2' were the first to demonstrate the utility of cold neutrons in PGAA, using a high flux of cold neutrons and a small Ge detector. Later, Kerr et al." reported the multielement analysis of numerous materials with a guidec! thermal beam. Neither of these experiments has led to the establishment of a permanent PGAA facility at ILL. Kyoto University Research Reactor Institute (KURRI), Tokai. Japan Kobayashi and Kandaio used a guided thermal beam to construct a PGAA system at the 5-MW KUR, optimized for the measurement of boron in tissue samples at concentrations below 1 pg/g.

Table 1

Comparison of Selected PGAA Facilities

Facility

UMd-NBS MURR ILL KFA

NIST CNRF JAERI JAERI

KUR ILL

Beam type

Thermal beam Thermal beam Cold guide Cold guide Cold guide Cold guide Thermal guide Thermal guide Thermal guide

B sensitivity, countfs mg

1060 760 (I$ =

3400 2600 1480 212

203 2700

Ref.

2

1.5 X 10'0nlcm2. s)

33 26 11 6 3 3 10

27

Forschungszentrum (KFA) Jiilich, Germany Lindstrom et al." reported the first combination of cold neutrons with a modem large y-ray detector, and demonstrated a fourfold increase in sensitivity over the thermal neutron PGAA instrument at NIST. The KFA system, with improvements, has continued in regular use at the FlU-2 reactor for the analysis of a wide variety of materia1s.12 National Institute of Standards and Technology (NIST), Gaithersburg, Maryland, USA The Cold Nerltron Research Facility (CNRF) at the Neutron Beam Split-Core lieactor (NBSR) includes a permanent, full-time instrument for PGAA with cold n e ~ l r o n sThe . ~ ~first ~ ~ measurements were made in late 1990. Japan Atomic Energy Research Institute (JAERI), Tokai, Japan A permanent PGAA instrument is installed at the JRR-3M (upgraded JRR-3) research reactor. which is operable at either a cold or thermal g ~ i d e . ~ .Both ~ ' guides include a curved section to remove fast neutrons and y-rays. The y-detector system incorporates a BGO Compton suppressor, which is also operable as a pair spectrometer. The instrument has been in operation since 1991. Cornell University (CU), Ithaca, New York, USA A mesitylene cold source is being installed in the 500-kW TRlGA reactor at Cornell Uni~ersity.'.~ Neutrons for PGAA are to be transported from the reactor hall to a low-background measurermilt room with a segmented curved guide. Central Institute for Physics (KFKI), Budapest, Hungary The recently restarted Budapest 10-MW reactor facility will include a PGAA instrument on a curved cold-neutron guide, which is scheduled to begin operation in 1993." Bhabha Atomic Research Centre (BARCj, Trombay, India A prompt y-instrument is under construction for use with a cold neutron beam from the Dhruva reactor.32

VI111. TRENDS As the use of neutron guides becomes more common at reactors, we may expect that more PGAA systems may be built to exploit these high-grade beams. The success of neutron guides has stimulated research into additional means of using neutron optics to improve the characteristics of neutron beams for elemental analysis. Unlike neutron scattering measurements where a low-dispersion parallel beam is preferred, methods such as PGAA that involve nuclear reactions are able to exploit increased reaction rate at the expense of increased angular divergence. Converging using supermirrors are attractive for condensing a neutron beam imo a smaller area. Multiplate beam benders have been investigated to steer a slow neutron beam away from fast neutrons and from interfering objects.37J8 Pyrolytic graphite crystals have been used to diffract much of the useful flux from a neutron beam onto an external sample for PGAA.39 Neutron focusing is being investigated with the goal of producing a very small, intense beam for ~ ~ convert a large-area beam into a analytical applications. A shaped stack of bent r n i c r o g ~ i d e scan narrow line. Another approach has been shown at Moscow and NIST to provide a true lens for both X-rays and neutron^.^'*^^ In this technique, an array of very many small capillaries has been used to

focus a beam 2 cm in diameter into a 1-mm spot. As this technology becomes developed, it may be p s i b l e to direct a high-intensity neutron beam onto a sampIe well removed from any sources of interfering radiation, and to perform spatially resolved PGAA on an extended sample in a manner reminiscent of the electron probe or PIXE.

NOTE ADDED IN PROOF Many of the points made in this review were discussed in a pioneering paper by D. Comar, C. CrouzeI,

M. Chasteland, R. Riviere, and C. Kellershohn, entitled 'The Use of Neutron Capture Gamma Radiations for the Analysis of Biological Samples," in Modern Trends in Activation Analysis (NBS Special Publ. J. R. De Voe, Ed., National Bureau of Standards, Washington, DC, 1969, p. 114.

REFERENCES Maier-Leibnitz, H., Neutron conducting tubes, in Neutron Capture Gamma-Ray Spectroscopy (STIIPUBI 235), Ryde, N., Ed., IAEA, Vienna, 1969, 93. Yonezawa, C., Wood, A. K. M., Hoshi, M., Ito, Y., and Tachikawa, E., The characteristics of the prompt gamma-ray analyzing system at the neutron beam guides of JRR-3M, Nucl. Instrum. Meth., A329.207, 1993. , and Lindstmm, Anderson, D. L, Failey, M. P., Zoller, W. H., Walters, W. B., Gordon, G. E R M., Facility for non-destructive analysis for major and trace elements using neutron-capture gamma-ray spectrometry, J. Radioanal. Chem., 63, 97, 1981. Majknak, C. F., Interference coatings for neutrons, Appl. Optics, 23, 3524, 1984. Werner, K, Springer, T., and Duppich, J., A new neutron guide laboratory at the FRJ-2(DIDO) reactor in Juelich, in Use and Development of Low and Medium F l u Research Reactors, Harling, 0.K., Clark, L., Jr., and von der Hardt, P., Eds., Karl Thiemig, Munich, 1984, 783. Lindstrom, R M., Zeisler, R., Vincent, D. H., ~ r e e n b e r R. ~ , R., Stone, C. A., Anderson, D. L., Clnrk, D. D., and Mackey, E. A., Neutron capture prompt gamma-ray activation analysis at the NIST Cold Neutron Research Facility, J. Radionnal. Nucl. Chem., 167, 121, 1993. Kawabata, Y, Suzuki, M.,Takahashi, H., Onishi, N., Shimanuki, A., Sugawa, Y., Niino, N., Kasai, T., Funasho, K., Hayakawa, S., and Okuhata, K., Construction of nentron guide tubes in upgraded JRR-3, J. Nucl. Sci. Tech., 27, 1138. 1990. Clark, D. D., Emoto, T., Outlet, C. G., Pekrul, E., and Rerg, 3. S., The Cornell University cold neutron beam facility, in 50 Years With Nuclear Fission, Behrens, J . W. and Carlson, A. D., Eds.. Am. Nucl. Soc., 1989. 855. Copley, J. R. D. and Mildner, D. F. R., Simulation and analysis of the transmission properties of curvedstraight neutron guide systems, Nucl. Sci. Eng., 110, 1, 1992. Kobayashi, T. and Knnda, K., Microanalysis system of ppm-order '"B concentrations in tissue for neutron capture therapy by prompt gamma-ray spectrometry, Nucl. Instrum. Meth., 204, 525, 1983. Lindstrom, R. M., Zeisler, K., and Rossbach, M., Activation analysis opportunities using cold neutron beams, J. Radioanal. Nud. Chem., 1 12, 321, 1987. Rossbach, M., Multielement prompt y cold neutron activation analysis of organic matter, Anal. Chenz.. 63. 2156, 1991. Rustad, B. M., Christensen, C. J., and Skytte Jensen, B., Fused Li6,C03for absorbing slow neutrons in collimating systems, Nucl. Inst. Meth., 33, 155, 1965. Stone, C. A., Zeisler, R., Blackburn, D. H., Kauffmann, D. A., and Cranmer, D. C., 6Li-doped silicate glass for thermal neutron shielding, in NIST Tech. Note 1285, U.S. Government Printing Office, Washington, D.C.. 1990, I 16. Lindstrom, R. M., Paul, R L., Vincent, D. H., and Greenberg, R. R., Measuring hydrogen by cold-neutron prompt-gamma activation analysis, J. Radioanal. Nucl. Chem., 180, 27 1, 1994. Neumann, D. A., Copley, J. 13. D., Cappelletti, R. L., Kamitakahara, W. A., Lindstrom, R. M., Creegan, K. M., Cox, D. M., Romanow, W. J., Coustel, N., McCauIey, J. P., Jr., Maliszewskyj, N. C., Fischer, J. E., and Smith, A. R., 111, Coherent quasielastic neutron scatrering study of the rotational dynamics of Cm in the orientationally disordered phase, Phys. Rev. Lett., 67, 3808, 1991. Copley, J. R. D. and Stone, C. A., Neutron scattering and its effect on reaction rates in neutron absorption experiments, Nucl. Insmtm. Melh., A28 1 , 593, 1989. Mackey, 33. A., Gordon, 6.E., Linclstmm, R. M., and Anderson, D. L., Effects of target shape and neutron scattering on element sensitivities Ihr neutron-capture prompt y-ray activation analysis, Anal. Chern., 63,

288. 1991.

19. Copley, J. R. D., Scattering effects within an absorbing sphere immersed in a field of neutrons, Nucl. Instrunr Mcflr., A307, 389, 1991. 20. Mackey, E. A., Gordon, G. E., Lindstrom, R. M., and Anderson, D. L., Use of spherical targets to insure accuracy in neutron capture prompt y-ray activation analysis of hydrogenous materials, Anal. Chem., 64. 2366, 1992. 21. Stone, C. A., Stone, S. F., Blackman, M. J., Olin, J. S., Myers, E., and Clark, D. D., Application of e of archaeological and art historical materials, prompt gamma-ray activation analysis to ~ t characterization Trans. A m Nucl. Soc.. 62, 224, 1990. 22. Rossbach, M. and Hiep, N. T., Prompt gamma cold neutron activation analysis applied to biological materials, Fresen. J. Anal. Chem., 344, 59, 1992. 23. Lindstrom, R. M., Fleming, R. E,Paul, R. L., and Mackey, E. A., The ko approach in cold-neutron promptgamma activation analysis, Proc. Int. kc Users Workshop, Rijksuniversiteit Gent. 1992, 121. 24. Foderaro, A,, The Elements of Neutron Interaction Theory, MIT Press, Cambridge, MA, 1971. 25. De Corte, F., Simonits, A,, De Wispelaere, A., and Hoste, J., Accuracy and reproducibility of the k, standardization method, J. Radioanal. Nucl. Chem., 113, 145, 1987. 26. Henkelmann, R. and Born, H. J., Analytical use of neutron-capture gamma-rays, J. Radioanal. Chern., 16. 473, 1973. 27. Kerr, S. A., Oliver, R. A., Vittoz, P., Vivier, G., Hoyler, F., MacMahon, T. D., and Ward, N. I., Elemental concentrations in geochemical reference samples by neutron capture prompt gamma-ray spectroscopy, J. Radioanal. Nucl. Chem., 113, 249, 1987. 28. Lindstrom, R M., Prompt-gamma activation analysis, J. Res. NIST, 98, 127, 1993. 29. Yonezawa, C., Prompt y-ray analysis of elements using cold and thermal reactor guided neutron beams, Analyt. Sci., 9, 185, 1993. 30. Clark, D. D., Outlet, C. G., and Berg, J. S., On the design of a cold neutron source, Nucl. Sci. Eng., 110. 445, 1992. 31. Molnar, G., Rbvay, Z., Veres, A., Simonits, A., and Rausch, H., Cold neutron facility for prompt gamma neutron activation analysis, J. Radioanal. Nucl. Chem., 167, 133, 1993. 32. .Gangadharan, S., personal communication, 1988. 33. Hanna, A. G., Brugger, R. M., and Glascock, M. D., The prompt gamma neutron activation analysis facility at MURR, Nucl. Instrum. Meth., 188, 619, 1981. 34. Mildner, D. F. R., Neutron gain for converging guide tubes, Nucl. Instrum. Meth., 200, 167, 1982. 35. Rossbach, M., Scharpf, O., Kaiser, W., Graf, W., Schirmer, A., Faber, W., Duppich, J., and Zeisler, R., The use of focusing supermirror neutron guides to enhance cold neutron fluence rates, Nucl. it~strum.Meth.. B35, 181, 1988. 36. Copley, J. R. D. and Majkrzak, C. F., Calculations and measurement of the performance of converging neutron guides, in Thin-film Neutron Opr~calDevices (Proc. SPIE 983). Majkrzak, C . E, Ed., Soc. PhotoOptical Instrum. Eng., Bellingham, WA, 1989, 93. 37. Alefeld, B., Duppich, J., Scharpf, O., Schirmer, A., Springer, T., and Werner, K., The new neutron guide laboratory at the FRJ-2 Reactor in the KFA Jiilich and its special beam forming devices, in Thin-film Neutron Optical Devices (Proc. SHE 983). Majkrzak, C. F., Ed., Soc. Photo-Optical Instrum. Eng., Bellingharn, WA, 1989. 75. 38. van Well, A. A., Stacked neutron guide system for neutron beam research, Nucl. Sci. Eng., 110, 10, 1992. 39. Harling, 0. K., Yanch, J. C., Choi, J. R., Soiares, G. R., Rogus, R. D., Moulin, D. J., Johnson, L. S., Olmez, I., Wirdzek, S., Bernard, J. A., Zamenhof, R. G., Nwanguma, C. I., Wazer, D. E., Saris, S., Madoc-Jon~s,H., Sledge, C. B., and Shortkroff, S., Boron neutron capturc therapy and radiation synovectomy research at the Massachusetts Institute of Technology Research Reactor, Nucl. Sci. Eng., 110, 330, 1992. 40. Mildner, D., Chen, H., Downing, G., and Sharov, V., Focused neutrons: A point 10 be made, J. Neictn>n -Res., 1, 1, 1993. 41. Kumakhov, M. A. and Sharov, V. A., A neutron lens, Nature, 357, 390, 1992. 42. Chen, H., Downing, R. G., Mildner, D. F. R., Gibson, W. M., Kumakhov, M. A., Ponomarev, I. Y., and Gubarev, M. V., Guiding and focusing neutron beams using capillary optics, Nature. 357, 391, 1992.

Chapter 7

In Vivo Prompt Gamma Activation Analysis Chien Chung CONTENTS

I. Introduction ........................................................................................................................-...... 101 11. IVPGAA Facility ...................................................................................................................... 102 A. General Layout .................................................................................................................... 103 B. Phantom Calibration ............................................................................................. ............. 110

111. Clinical Applications ................................................................................................................ 1 12 A. Partial-Body Scan ............................................................................................................... 112 1. Cadmium ........................................................................................................................ 1 12 2. Mercury .......................................................................................................................... 1 15 3. Other Elements .............................................................................................................. 1 16 B. Whole-Body Scan ............................................................................................................... 116 1. Nitrogen .......................................................................................................................... 1 16 . 2. Other Elements .............................................................................................................. 118 .. IV. Radiation Doses to Patients ......................................................................................................119 A. Neutron Flux Distribution in Body ................................. . . ............................................... 119 B. Doses for Partial-Body Scan ..............................................................................................122 C. Doses for Whole-Body Seem ............................................................................................ 125 V. Discussion ................................................................................................................................. 127 References ......................................................................................................... ................................ 128

Among the nuclear methods that have proved very useful in biological and medical analyses is the PGAA, in addition to conventional INAA. While the PGAA technique has been developed using various neutron sources and the system is usually installed around a nuclear facility such as a research reactor, many demands create field applications such as in vivo (IVPGAA) medical diagnosis of the sick patient. These real-time, on-line techniques do not rely on the induced radioactivity, thus, without the lengthy period of waiting long after the irradiation. The knowledge of neutron reaction and prompt y-rays emitted from it are prerequisite for the IVPGAA method; the neutron source and y-ray spectrometer are indispensable as integrated parts of the IVPGAA facility. With the high-resolution y-ray spectrometer commerdalized in the 1960s and prompt y-rays well documented in the 1 9 7 0 IVPGAA ~~ was rapidly adopted by the medical community as a reliable nuclear method in 1980s. Future expansion of IVPGAA with commercial units available for the medical community is conceivable due to its mobility, convenience, and versatility. T h e recent development of IVPGAA has opened a new era of insight into the elemental composition of the human body. Absolute measurements of some environmental contaminants, such as cadmium. mercury, and silicon in organs, as well as of vital constituents, such as calcium, nitrogen, and phosphorus in either the whole body or body parts, have been studied for therapy evaluation, clinical diagnosis, and investigation into the modeling of body composition. Detailed clinical application of IVPGAA is summarized by Cohn.' The principle of IVPGAA, designed for in vivo studies, is to determine elemental concentrations by measuring prompt y-rays emitted during nuclear reactions rather than y-rays emitted from radioactive decay. Nuclear reactions on target nuclei induced by neutrons, as illustrated in Figure 1, create excitedstate reaction products which in turn de-excite promptly to the ground state. The de-excitation is usually accomplished by emitting prompt y-rays that are detected external to the body. For instance, the thermal capture reaction of N(n,,r) involves a reaction cross-section of a,h= 0.075 b, emitting high-energy, 10,829-keV y-rays with a high yield of 14 photons per 100 neutrons captured. Thus, the emission rate of the 10,1129-keVprompt y-rays is nearly 500 rls during whole-body irradiation of a patient with a

r~U

I I I ~ R

photon

C23

gamma detectorC3 3

Figure 1 Stage of neutron capture reaction with body nitrogen a s target and emission of prompt y-ray, which is subsequently measured externally by a detector.

low thermal neutron flux of 10,000 n,,,/cmZ- S. On the other hand, total delayed y-rays, emitted at the end of a 500-s activation to form the radioactive isotopes of 7-s I6N, yield only 0.001 rls; this is far below the detection limit of even the most advanced whole-body, low-background y-ray spectrometer. Despite various analytical instruments having been applied to medical diagnosis, not a single method (instrumental or chemical) has been found to satisfy all of the sensitivity requirements for the elements of interest to all medical disciplines. Each method is complementary to others, for example, considering the instrumental methods of neutron activation analysis (INAA) and photon activation analysis (IPAA) in comparison to PGAA for the application to analyze biologically important samples such as human tissue, tht: elemental concentrations therein that can be measured by LNAA, IPAA, and PGAA techniques are displayed in Figure 2A. One can certainly observe the superiority of the PGAA method for the determination of some 20 elements without significant amounts of induced radioactivity; essential elements related to human balance of nutrition such as C, H, and N as well as the intake of trace level of toxic elements such as Cd and Hg can only be analyzed by IVPGAA. As a supplement, yet very distinguishable from the more conventional INAA, the IVPCiAA technque also provides clinical and biomedical applications in constructing major elemental profiles of body composition as well as trace elemental contamination for toxicological considerations. These elements, that can be analyzed by either in vivo INAA or in vivo PGAA, such as essential elements of N, Ca, and P as well as trace level of Cd, Hg, and Si, are illustrated in Figure 2B with organ of accumulation indicated. The body elements measured by IVPGAA at National Tsing Hua University, Taiwan, as listed in Table 1, are typical of measurements made at many research centers and hospitals. Although other in vivo neutron reactions not listed in the table can also occur in the living body, the accuracy of such measurement and its detection limit may restrict their useful clinical and biomedical applications at ihe present time. In this chapter, features of IVPGAA and its clinical applications are described. Although ncutrons used in IVPGAA for medical diagnosis can be extracted from any nuclear facility, such as accelerator, neutron generator, research reactor, and isotopic neutron source, as described elsewhere in this book, charactenstics of the in vivo medical diagnosis and radation sarety are emphasized in order to elucidate the utilization of a rather simple nuclear faci:ity for a specific in vivo activation analysis.

II. IVPGAA FACILITY Any in vivo PGAA facility contains a neutron source with safety device, a bed for palient irradiation, a prompt y-ray detecting system with analytical software, and necessary radiation shielding that keeps both patient and essential equipment away from radiation exposure. In addition, a phantom is required for system calibration. The selection of a neutron source for IVPGAA is based on the elements to be measured, the deg;ee of irradiation uniformity, the level of accuracy, and the induced radiation dose that must be below regulatory concern. The type of neutron sources used for IVPGAA are accelerators, reactors, neutron generators, spontaneous fission sources, and isotopic neutron sources. The design of an IVPGAA facility is usually influenced by the prompt y-counting system, in addition to the availability of the neutron source. To detect and analyze the prompt y-ray in IVPGAA diagnosis, a photon detector is required. The choice of a particular detector type for NPGAA application depends upon the y-energy range of interest and the application's resolution and efficiency requirements. Other considerations are count rate capability and, if timing applications are involved, the pulse rise time. The kind of detectors

Brain

, ( 7 ,

Brain

Figure 2 (A) Measurable elemental concentrations in human organs using PGAA, INAA, and lPAA techniques, and (B) sketch showing trace and essential elements with organs of their accumulation that can be analyzed by in vivo PGAA and INAA.

commonly used in IVPGAA are scintillation NaI(T1) detectors and semiconducting HPGe detectors, as mentioned in earlier chapters. In TVPGAA measurements, the quantitative determination is more difficult than in conventional activation analysis using a standard cornparator because of differences in organ size and human habits among individuals. The comparison of a human with a phantom in in vivo measurements must be justified by providing a uniform neutron fhix in the irradiated organ. A steady, nonthermal neutron beam is essential for in vivo diagnosis.

A. GENERAL LAYOUT The neutron sources used for IVPGAA are mainly an isotopic neutron source and mobile reactor in consideration of their convenience of maintenance, radiation safety, portability, and mobility. TOa lesser extent, a pennanent nuclear facility such as a neutron generator, accelerator, and research reactor is

Table 1 Prompt y-Rays Emitted from (n,r) Reaction with Elements Important to Medical Diagnosis and Their N

Element

Body organ at risk

Medical diagnosis

Normal contenta

(n,r) Reaction cross-section, barn

Asso prompt

Trace Level Cd

'% Fe Si

Kidney, liver Kidney, liver, brain Heart, liver Lung

Kidney dysfunction Itai-Itai disease Heart disease Lung disease

Major Constituent H C Ca C1

N P

Whole body Whole body Skeleton Whole body Whole body Whole body

Fat, water composition Energy expenditure Bone disease Transport disturbance Nutritional disturbance Metabolic disturbance

10.00% 20.8% 1SO% 0.13% 2.59% 1.12%

0.33 0.0034 0.43

33.2 . 0.0747 0.18

' Unless otherwise indicated, the normal content is in weight percentage of whole body of Reference Man, in Reference 2. Single- and double-escape peaks from the main photopeak are aiso analyzed in order to attain better counting result.

22

49 64 61 10,8 39

Nol(TI)

/I

EPOXY RESIN ( L i DOPED1

/DETECTORS

HEAVY WATER

2'8~"e,

/

NEUTRON SOURCE

Flgure 3 General layout of IVPGAA facility using 238Pu/Beisotopic neutron source. (Reprinted with permission from J. Nucl. Med. 20(1I ) , Vartsky, D., Ellis, K. J., and Cohn, S. H., In vivo measurement of body nitrogen by analysis of prompt gammas from neutron capture, 1158, Copyright 1979, The Society of Nuclear Medicine.)

DETECTORS

I

- POLYESTER

RESIN Li

- EPOXY RESIN +

+ Li

only occasionally providing neutrons for in vivo activation due to their complexity of facility operation and shielding problems for their intense radiation. The typical neutron flux delivered to the irradiated patient is in the range of lo3 to lo6 n/cm2 s for in vivo diagno~is,~ at least 6 orders of magnitude less than those provided by research reactors ar particle accelerators. Although the IVPGAA technique was first demonstrated using cyclotron-delivered neutrons: other research institutes soon followed the concept; however, use of the more convenient isotopic neutron sources for such medical diagnosis is anti~ipated.'.~ Details of some representative IVPGAA facilities using various isotopic neutron sources and mobile reactor are given below. The isotopic neutron sources are limited in choices and based upon either nuclear reaction between a target material and a radioactive precursor emitting particles, or a spontaneous decay process emitting fissioning neutrons. The popular neutron source of these kinds are 23BPu/Be and ''Cf sources, re~pectively.~ The self-contained 238P~/Be is fabricated by mixing the 87.7-year 2 3 8 Pa-emitting ~ isotope with beryllium target material; the energetic &-isotope may induce (ry,n) reactions with neutron emission rates of 2.4 X lo9 n/s/1000 Ci 238Pu.On the other hand, the most commonly used spontaneous fission source of the 2.65-year 252Cf,however, decays 252Cfemits fissioning neutrons at the rate of 2.3 X lo9dslmg 252CE; predominantly by a-emission. Since isotopic neutron sources are radioactive, they can be used in an IVPGAA facility long enough to be reasonably convenient; however, they cannot be "turned off' as those neutrons delivered from a neutron generator, accelerator, or reactor. In addition to providing neutrons, the isotopic neutron source also emits other lethal radiations such as a-particles, fissioning y-rays, and decayed y-rays interfering with in vivo measurement. At 1 m away from an unshielded isotopic neutron source, the dose rate caused by y-rays alone is in the range of 1.3 mSv/h/1000 Ci 23XPu/Be and 1.03 mSvlhlmg W f , or at least 10,000 times over background radiation level. Hence, a radiation shield surrounding both patient and y-ray detector becomes indispensable to attenuate these unnecessary and useless radiations emitted from the isotopic neutron sources. The IVPGAA facility using the 238Pu/13eisotopic neutron source was first developed by the research group at Brookhaven National Laborz~tory.~ In their facility, an 85-Ci 23XPu/Besource with average neutron energy of 4.5 MeV is used. As shown in Figure 3, the source is housed in a collimator made of epoxy resin doped with Li,CO, a ~ t l6LiF compounc!~.The collimator is designed to provide a

rectangular beam measuring 13 X 60 cm2 at the level of the bed, situated 50 crn above the neutron source. The collimation of the neutrons can be changed to provide smaller beam fields for partial body irradiation of a selected organ such as the liver. The fasl neutron flux at the level of the bed is 7200 n/Cm2 . S. The neutron shield consists of boxes filled with polyester resin mixed with Li,CO, and %iF. These compounds minimize radiative capture in the resin's hydrogen. The whole system is covered with a 20-cm thick layer of lead; the layer under the detectors is 30-cm thick. The lead reduces the intensity of y-photons emitted from the neutron source as well as those produced in the neutron shielding material by neutron capture and inelastic scattering reactions. The neutron shield reduces the radiation dose, allowing clinician or nurse stands near the patient during the measurement. The movable bed consists of a 5-cm thick aluminum plate driven by a motor so that the patient's body passes continuously through the neutron beam. A certain d e g r e of premoderation of the neutrons is necessary in order to provide uniformity of composite sensitivity through the thickness of the human body. The premoderation is provided by a 5cm thick box made of aluminum filled with D28. The box is positioned in the neutron team and pressed against the aluminum bed, with a thin Teflon sheet in between to reduce friction. The deuterium in the neutron field produces tritium by the D(n,r)T reaction; calculations of the tritium yield showed production of only 5 n Wyr. Since the uaPu/Be is pcmmanently sealed in the position indicated in the figure, the majority of unused neutrons are largely absorbed by the doped lithium surrounding the source, gencrating again tritium by the 'Li(n,a)T reactions in large quantity. Total activity of tritium in such an IVPGAA facility, 3 yr after neutron source installation, may reach 1 mCi and eventually saturaie to 6 mCi. One has to be reminded that the annual limit of intake of tritium in air, for patient and medical personnel around the IVPGAA facility, is 2.4 mCi/m3 by regulation. The induced prompt y-rays emitted from the body are detected by two 6" X 6" NaL(T1) cryslals positioned above the body, with aluminum for structural support. The detector geometry is designed to minimize the spatial variation of sensitivity for high-energy prompt y-rays. To reduce the neutron flux reaching the detectors, they are surrounded by cups made of resin doped with Li compounds. 'I'he detectors are coupled to photomultiplier tubes, 5" in diameter, connected with the caihode at high negative potential and the anode close to ground potential. In this mode, the signal can be taken directly from the anode without the need of a blocking capacitor. The method of pulse-height analysis is that the fractional charge collection is necessary since the total detector count rate is high (73,000 cps), thus providing a high probability of a pulse pile-up. The signals from the two anodes are passed through a pole-zero cancellation filter. This element shortens the duration of the anode current pulse from approximately 1 p s to 70 ns. The anode signal is then amplified and split into two branches: one feeds a linear gate and stretcher through a suitable delay, while the second triggers a fast discriminator :hat provides a logic signal to the h e w gate. Only signals that are above the discrimination level can be integrated in the linear gate and analyzed. However, recent development arid applications of a new scintillation detector of bismuth germanate (BGO) may simplify the counting electronics since its counting efficiency for high-energy prompt y-rays is about 10 times better than that of an NaI(T1) detector of similar size, as demonstrated in PGAA measurements e l s e ~ h e r e . ~ Another advanced IVPGAA facility using a 252Cfspontaneous fission source was coristructed anti demonstrated by the Swansea research group in the U.K."he main components o f this fac~lity,as illustrated in Figure 4, comprise the shielding material, a safe storage position, a pneumatic trmsport system to deliver the source to one of the two collimator positions, a scanning bed, and a y-ray detection system. A detailed description of each of thzse components and their main features are summarized below. A simple, inexpensive yet versatile pneumatic transport system was designed and built specifically for this application. It is based upon a small vacuum/compression pump that operates continuously while the instrument is in use. A small transport capsule is fabricated from steel and contains two U2Cf sources, each of 200 mg. The source storage safe is a lead cube of external dimensions 14 cm drilled to accept an aluminum sleeve. The storage safe is surrounded by a composite of shielding material consisting of poly-cast and poly-lead-boron bricks, borated wax, lead, and barytes concrete. A minimum thickness of 80 crr, of neutronly-shielding material surrounds the neutron source on the two sides of the instrument. The longer collimator is bolted to the underside of the upper shelf. Above the upper shelf on the platform of poly-lead-boron bricks, the second irradiation port for the shorter collimator is placed. All of the remaining space within the framework of the instrument is filled with shielding

Detectors posilioned above or beside the patient

Supporting frame Retnforced pvc hose . Source diverter Neutron source s t oroge posit ion

Figure 4 General layout of lVPGAA facility using 252Cfspontaneous fission source. (Reprinted with permission from Phys. Med. Biol. 32(10), Ryde, S. J. S., Morgan, W. D., Sivyer, A., Evans, C. J., and Dutton, J., A clinical'instrument for multi-element in vivo analysis by prompt, delayed, and cyclic neutron activation using 252Cf,1257, Copyright 1987, Institute of Physic Publishing Ltd.) material. A layer consisting of 7.5 cm Pb/Bi is placed on top of the bulk shielding aro6nd the collimator, as shown in Figure 4. In order to achieve adequate shielding and a proper height for the patient bed, the greater part of the main instrument structure and shielding is housed below floor level. Two collimators of different source-skin distance and each having an exit aperture width sufficient to accommodate the width of a patient are designed. Iron and bismuth for the backscatter and collimator, respectively, are used. Two iron blocks, bored at 45O to the vertical to accept the transport tube, form the irradiation ports. A large bismuth collimator suitable for total body analysis using the scanning geometry is placed upon each port. For each collimator, the beam dimensions are reduced to a size suitable for partial body analysis. A bed frame on wheels rests on aluminum rails and accommodates different bed tops depending upon the element to be measured. The frame is driven by a variable-speed reversible motor attached, via a reduction gearbox and magnetic clutch, to a 2.5-m power screw. A thyristor unit permits bed speed control within a range from 0.2 to 4.8 mmls. TWOdifferent detector systems are employed. The prompt measurement of Ca and Cd requires the use of a high-resolution senucond~tctingdetector, whereas the higher energy of the prompt y-ray from "N permits the use of an Nal(T1) detector. The two detector systems are outlined below. Two 20% ntype HPGe detectors of closed-end coaxial configuration with 1.9-keV or better resolution at 1332 keV are used. The cryostats are portable, with 24-h holding time, allowing the detectors to be placed around the patient in an optimum position for each element to be measured. Each detector is connected to a high-quality spectroscopic system incorporating a 440-MHz ADC. The detectors are shielded by caps of 6Li-enriched (95.3%) LiF of minimum thickness 0.4 cm. Depending upon the element to be measured, these shields are further surrounded by a composite of borated wax, lead, and bismuth. On the other hand, for the initial assessment of prompt nitrogen measurement, a 6" X 6 NaI(T1) detector of 10% resolution at 662 keV is available. Using the modified electronics as Vartsky designed,$ the system signal-to-background ratio in the high-energy region is improved by 20% due to reduced pulse pile-up, but the modified system has decreased resolution and greater gain instability is observed as the counting rate is varied. Subsequently, two new Nal(T1) detectors of better than 8% resolution with high stability and high current capacity PMT assemblies designed for negative bias operation were applied. These are configured into the same spectroscopy systems used with the HPGe detectors. Shielding of the NaI(T1) crystals consisls of LiF powder (95.3% 6Li) and I-mm thick lead sheets. This IVPGAA facility is certainly improved from the viewpoint of radiation safety since the neutron source can be stored, or turned off, to a well-shie!d locker when not in use, resulting in a dose equivalent rate

Figure 5 Modified THMER facility for IVPGAA medical diagnosis: (A) polyelhylene-1°B,O, disk, (8)lead bricks, (C) neutron beam tube, (D) BLi,CO, neutron absorber, (E) neutron beam shutter system, (F) x-y mobile bed, (Gjphantom. arid

(H) HPGe detecting system. (Reprinted with permission from Activafim Aoaiysrs, II, Chung, C., Activation analysis with small mobile reactor, Alfassi, Z. B., Ed., 301. Copyright 1990, CRC Press, Boca Raton, FL.)

at floor level around 1 pSv/h, or 10 times higher than the background level but 10 times lower than that at the Brookhaven installation? Although an IVPGAA facility using an isotopic neutron source may be easily installed in a hospital or research institute as desired, patients, in particular for those illed ones who have great problems to move, are inevitably transported to and from such a facility with difficulty. A mobile IVPGAA facility conceivably reduces the complexity of bringing sick patients around. Recently, an IVPGAA facility was assembled on a mobile nuclear reactor and has demonstrated its versatility for in vivo medicai diagnosi~."'~A brief summary of the mobile IVPGAA facility is given below. The Tsing Hua Mobile Educational Reactor (THMER), designed and converted for IVPGAA measurement, is a critical assembly with sustained thermal power of 0.1 W.I3 The facility contains three major parts: the reactor on trailer, the in vivo scan station, and the prompt y-counting system. The THMER core utilizes 20%-enriched '"U3O8 fuel, providing maximuni thermal neutron flux at core center of about 5 X lo6 n,,,/cm2 s. Since the critical assembly has a negative temperature rise and operates ai ambient temperature, no coolant is provided. A 10-cm thick lead wall is placed immediately outside the graphite reflected core; about S-ton saturated H3BO3 solution in the stainless steel tank further surrounds the lead block as a biological shield. As shown in Figure 5, on top of the reactor tank a thick layer of boron-doped polyethylene and lucite are piled up with another 8-cnl thick layer of lead block overhead as neutrody-shields for the in vivo scan station. The 13-ton THMER reactor, 2.1 m in diameter, has its bottom screwed to the main beams of ar: 11.5-m mobile trailer. The 28-ton trailer, accommodaling the THMER tank, control console, air conditioning system, and IVPGAA medical accessories, can be towed by a standard tractor on most major routes in Taiwan with a speed limit of 40 kmh. Its frequenl calls idandwide rcduces the complexity of bringing sick patients to the university. The IVPGAA station is on top of the reactor tank of the THMER facility, as illustrated in Figure 5. A vertical neutron beam is extracted from the reactor core and collimated by boron-doped polyethylene/ lucite. When the THMER is started up and reaches its full power, neutrons fram the 10.5-cm diameter beam tube can be shut off by pumping the saturated H3B03solution into the beam tube in 45 s without interfering with the control of the reactor power level. At the IVPGAA station, the neutron beam can be shut off during loading and unloading of the patient. Once the patient is placed in a supine position where the preselected organ is centered at the beam tube, the neutron beam is initiated by pumping off the H3B03solution. Since 90% of the thermalized neutron are absorbed by the skin and the first 8 cm of tissue, they contribute very little toward activating an element deep in the body organ but, rather,

they deliver unnecessary neutron doses. In order to eliminate the thermal neutrons extracted from the THMER facility, a 5-cm thick, 95%-enriched "i2CO3 absorber in a lucite cup is inserted on top of the neutron beam tube. This absorber can completely absorb thermal neutrons because of the high crossfor the 'Li(n,a)T reaction. To detect and analyze the prompt y-rays in IVPGAA experiments, a portable germanium detector with a 24-h liquid nitrogen holding time and supported by data collectinglanalyzing electronics and a neutronly-shield, is used. This n-type HPGe detector, having 25% relative efficiency, is placed next to the patient. The system resolution with ,an estimated scattering neutron flux of 200 n/cm2 . s is 2.25 keV at 1332 keV. Prompt y-rays collected by the detector are stored in a 4096-channel spectrum and subsequently analyzed, sorted, printed, and plotted using an MCA coupled to an IBMPC-printer-plotter. Before and after each IVPGAA measurement, the HPGe detector efficiency and energy resolution are calibrated and checked using sets of standard sources to ensure the satisfactory performance of the HPGe detector in a hostile environment of intense neutron flux. A I-cm thick lucite disc filled with 95%-enriched 6LiF is attached to the window of the I-IPGe detector to stop thermal neutrons from scattering into the detector's active volume. Another 3-cm layer of natural Li2C03is placed between the liquid nitrogen dewar and the 4-cm lead cylindrical shield to stop the thermal neutrons from scattering into the 1.2-1 liquid nitrogen, thus avoiding the interference with the partial-body nitrogen diagnosis. In addition to utilizing neutrons from the isotopic source or small reactor, other IVPGAA facilities use neutrons delivered from fixed nuclear installations such as cyclotrons, accelerators, and neutron generators. In Table 2, operational IVPGAA facilities using various neutron sources, together with their representative partial-body scan time and detection limit of toxic cadmium in critical organ, are listed.

Table 2 Comparison of the Capability of Operational IVPGAA Facilities Ref. (year)

Neutron source and strength

Detection limit (DL) Cd In body organ

Fast neutrons from cyclotron Fast neutrons from cyclotron 85 Ci 23RPu-Re

2 pg/g in liver

Irradiation time,

IVPGM

s

Index, Cda

0.5 pglg in liver

2.5 mg in kidney

1.8 pg/g in liver

20 Ci 24'Am-Be Fast neutron from cyclotron

30 pg/g in liver 3.25 mg in kidney

10 Ci 238Pu-Be

10 pg/g in liver

10 Ci 238h-13e

10 mg in kidney

10 pglg in liver 2.2 mg in kidney 1.5 pg/g in liver 9 pglg in live? 6.4 mg in kidney 6.5 pglg in liver 2.4 mg in kidney

25 pglg in liver

THMER reactor neutrons

1.3 mg in kidney

' IVPGAA index (Cd) = I/[DL (kidney) X skin dose (SD)], or 0.68/[DT2(liver)X SD]; *for facility without mobility. Estimated from the data providedin the cited reference. Reprinted with permission from Appl. Rtldiat. Isor. 36(5), Chung, C., Yuan, L. J., Chen, K. B., We%, P. S., ChaW P. S., and 1-10, Y. H., A feasibility study of the IVPGAA using rnohile nuclear reactor, 557. Copyright 1985. Pergamon Ress Ltd., with update infom;rtion added.

An NPGAA index, revealing the ability to lower the radiation dose and elemental detection limit, is also indicated in the table for comparison:, some high-index facilities, unfortilnately, lack mobility; others, including those illustrated in Figures 3 through 5, have reasonably good index performance.

B. PHANTOM CALIBRATION The elemental concentration in the human body diagnosed by the IVPGAA technique can be evaluated only if the neutron capture cross-section a, neutron flux
with N, the Avogadro's number and m, the atomic weight of the sought element. Thus, the unknown elemental weight in the sample can be evaluated as:

Figure 6 (A) Top, side, and transparent views of the man-like liquid phantom, anc! (8)top and side views of the female-like phantom; impact point of the neutron beam from IVPGAA measurement and deployment of dosimeters inside both phantoms are indicated. (Reprinted with permission from Appl. Radiat. /sot. 44(6), Chung, C., Wei, Y. Y., and Chen, Y Y , Determination of whole-body nitrogen and radiation assessment using IVPGAA technique, 941, Copyright 1993, Pergamon Press Ltd.; Activation Analysis, II, Chung, C., Activation analysis with small mobite reactor, Alfassi, 2. B., Ed., 307. Copyright 1990, CRC Press, Boca Raton, FL.)

If the irradiation condition of neutrons is unchanged (i.e., neutron flux intensity and neutron cnergy spectrum have no variation during the phantom and patient measurements), the quantitative determination can be further simplified as:

A flux monitor, such as a neutron counter, is usually positioned behind the IVPGAA set-up to give online, real-time readings of the flux intensity, thereby providing necessary correction if indeed a variation occurs in the irradiation condition.

Ill. CLINICAL APPLICATIONS The recent development of IVPGAA techniques has been extensively applied to medical diagnosis and clinical research and opened new insight into the elemental composition of the human body. Absolute measurements of some environmental contaminants such as Cd, Hg, and Si in organs requires a partialbody (PBIVPGAA) scan, while the measurements of vital constituents such as N, P, and Ca in either the total body or the body part requires a whole-body (WBIVPGAA) scan to which the patient is moving relative to the neutron beam during irradiation. Both PBIVPGAA md WBIVPGAA demand high precision and high accuracy of the measurement with minimum induced radiation doses for patients; information on elemental concentrations in the body is valuable for medical and clinical purposes. The measurement of cadmium represents the first PBEVPGAA application to the study of a toxic metal. Although only trace amounts of cadmium are normally present in the body, its in vivo detecGon is feasible due to the large radiative neutron capture cross-section. The thermal capture reaction of 113Cd(nh,r)"4Cdinvolves the naturally occurring stable isotope "3Cd of 12.2% abundance, with high reaction cross-sections of a, = 19,800 b and emitting 559-keV y-rays with a high yicld of 72.7 y-rays per 100 neutrons captured. On the other hand, the whole-body nitrogen content, related to the nutritional disturbance of the patients, has been measured by the WBIVPGAA technique. Although the thernlal neutron cross-section for radiative capture on nitrogen is small, it is countered by the large mass of nitrogen in the body. In the reaction '4N(n,r)15N,approximately 14% of the I5N de-excitations occur through a direct transition from the highest excited level of lSNto its ground state. This process produces the immediate emission of a 10,829-keV y-ray. In this section, the partial-body and whole-body scans for medical applications are described for each element of clinical interest.

A. PARTIAL-BODY SCAN 1. Cadmium A toxic contaminant such as cadmium (Cd) is concentrated in the liver and kidneys. The Stan
Flgure 7 Excerpt of prompt y-ray spectrum of 9.4 mg kidney cadmium in the PBIVPGAA scan. (Reprinted with permission from Appl. Radiat. Isot. 36(5),Chung, C., Yuan, L. J., Chen, K. B., Weng, I? S., Chang, I? S., and Ho, Y H., A feasibility study of the in vivo PGAA using a mobile nuclear reactor, 357, Copyright 1985, Pergamon Press Ltd.)

channels in the MCA. The total counts in the photopeak area of interest are directly proportional to the amount of Cd prepared for the bcidr~eyand therefore served as an indicator of the concentration of Cd in the kidney. In Figure 7, a portion of the PBIVPGAA spectrum from this kidney investigation is shown. The 559-keV prompt y-rays from the H3Cd(n,r)'i4Cd reaction are clearly observed, in addition to the 511-keV annihilation y-rays and 596-keV prompt y-rays from the Ge(n,r) reaction. The photopeak area of the 559-keV y-rays in a 500-s counting/irradiation period is 1613 -C- 233 counts, corresponding to the inserted Cd in the irradiated kidney. The detection limit of the toxic Cd concentration is 1.3 mg in kidney under a 500-s irradiatiodcounting period. In order to demonstrate the accuracy of organ Cd measurement using the IVPGAA technique, Chen and C h ~ n exposed g ~ ~ rats to toxic Cd, both chronically and acutely for the IVPGAA scan and subsequently sampled for in vitro INAA measurement. The relationship of the 559-keV photopeak count rate diagnosed

Figure 8 Relationship between IVPGAA 559-keV prompt y-count rate and cadmium concentration determined by lNAA in the intoxicated organ in rats. (Reprinted with permission form J. Radioanal. Nucl. Chem. Articles, 133(2), Chen, W. K. and Chung, C., In vivo and in vitro medical diagnosis of toxic cadmium in rats, 349, Copyright 1989, Elsevier Sequoia, S. A,)

f Ykps) = 0 1875x(rng1gkO 009

A 5 Cd concrnlrollon in both I'mand b

0

1

2

3

4

6

7

y ,mglg

b

Cadmium concentration (mgig)

Figure 9 Relationship between overall cadmium concentration in rats' body to (A) albumin (iull), (B) aggregated protein amount (iul I), and (C) ALT (glutamic pyruvic transaminase, GPT, iull) functions; damage levels in rats with organ dysfunction are also indicated by dashed lines. (Reprinted with permission from J. Radioanal. Nucl. Chem. Articles, 132(2),Chen, W. K. and Chung, C., In vivo and In vitro medical diagnosis of toxic cadmium in rats, 349, copyright 1989, Eisevier ,e,,oial S.

by IVPGAA and the Cd concentration determined by INAA in intoxicated rats is shown in Figure 8; from the data, a linear correlation of both in vivo PGAA and in vitro LNAA methods is well established. In their rats biochemical diagnoses are also performed simultaneously to correlate with thc results from an WPGAA scan. Cd poisoning may cause renal tabular damage, osteoporosis, and liver dysfunction. If renal tabular damage and liver dysfunction appear in the body, s~gnificantamounts of small molecular proteins also appear in urine. If thcre is serious damage to the liver and kidney, there will be increasing amounts of protein in the serum. The serum, liver, and kidney functions were analyzed and bioassayed and their results are shown in Figure 9. In this figure, it can be found that glutamic pyruvic transarninase (ALT) for the poisoned rats are explicitly higher than those of the normal rats. This implies that the liver is damaged with Cd concentrations above 160 pglg in liver, similar to those fatal concentrations in the human liver. Albumin and protein are, however, within the normal range to show that the damage due to Cd poisoning cannot be identified by such biological check-ups. On the other hand, total protein does not increase and shows only minor symptoms. Sensitivity and reproducibility of PBIVFGAA scans for organ cadmium in man have also been investigated.I6 Using a phantom filled with tissue equivalent liquid with various Cd concentrations in both kidneys and liver, the 559-keV prompt y-ray emitted from the Cd(n,r) reaction is identified and

C=workers exposed to Cd

Cd concentralion in liver ! p g / g )

Cd content in the left. kidney (mgl

Figure 10 Photopeak counts of 559-keV prompt y-rays against cadmium content in (A) liver and (8)left kidney of the Alderson phantom. (Reprinted with permission from Phys. Med. Biol. 22(6),Vartsky, D., Ellis, K. J., Chen, N. S., and Cohn, S. ti., A facility for in vivo measurement of kidney and liver cadmium by PGAA, 1085, Copyright 1977, Institute of Physics Publishing Ltd.)

its photope'ak area is integrated for quantitative evaluation. In Figure 10A, the relationship between the number of counts vs. Cd concentration i n the liver is shown. The upper limit of concentration, 300 pgl g, is the highest value quoted in the literature for workers exposed to cadmium.26As can be observed, the relationship is linear within these coucentration limits. Figure 10B also shows the same relationship for the kidney. Since both relationships are linear, a simple comparison with the liquid phantom is possible. The Cd detection limit in the study'6is 2.5 mg Cd in the kidney during the 30-min measurement. For the same scan time to the liver, the limit of detection is 1.8 pglg. 2. Mercury Mercury (Hg), the third member of the zinc and cadmium family, is a very toxic cumulative poison. In the body, Hg is normally only present in minute amounts, with a total soft tissue content of 13 mg. The clinical symptoms of toxicity depend on the chemical form of the mercury (metallic, inorganic, or organic compounds) and on whether the exposure is through ingestion or inhalation. Toxicologically, the accumulation of organic mercury compounds is, like cadmium, preferentially retained in the kidney.30 Because of the convenience of sampling, the current monitoring method for Hg involves in vitro measurement of Hg levels using aton-iic absorption spectrophotometry, spectrography. or conventional neutron activation analysis. The samples of these determinations are usually urine, blood, hair, and even breast milk. However, the results of in virro measurements do not indicate directly the organ concentration of Hg; therefore, they cannot be used as indicators of food poisoning. Since the kidney is the principal organ for accumulation, it would be valuable to monitor kidney levels directly. In the case of Hg, the normal abundance in organ is quite low. Although the range is wide, 0.43 mg mercury in each kidney is the representative figure for investigation. Hg is a weaker absorber than cadmium for low-energy neutrons; the thermal neutron capture cross-section of Hg is 376 b. The most intense neutron capture y-ray of Hg is the 368-keV transition from the first excited state of *OIHgby the 'O0Hg(n,,,, r)20'Hgreaction, with a high yield of 81 photons per 100 neutrons captured. However, the abundance of the most prominent y-ray in Hg is lower than for the corresponding y-ray in Cd, and the 368-keV prompt y-ray peak is interfered more severely by higher spectral background. Hence, the overall sensitivity for y-ray detection is worse for Hg than for Cd. A typical prompt y-ray spectrum taken from an IVPGAA scan for Hg content is illustrated in Figure 11. In some recent papers, Al-Hiti et aL3' reported a detection limit for brain mercury concentration of 100 pglg (1000 times more than the normal content); Sntith et aL3! reported a detection limit of kidney mercury around 16.7 mg (19 times more than the normal content) using cyclotron-induced secondary neutrons with an energy range of 2.5 to 3 MeV; the detection limit can be further reduced to the 2-

Ge- 326.0 k e V

-352.3 k e V

C

U

6 4

---

.

-

--

.

-

3 x i 03' 420

I

I

4 60

I

I

500 Chanml Number

11 Excerpt of the prompt y-ray spectrum from a 1800-s bilateral IVPGaA scan for kidney mercury using the THMER facility. (Reprinted with permission from Appl. Radial. Isot. 39(2), Chung, C., In vivo partial body activation analysis using filtered neutron beam, 93, Copyright 1988, Pergamon Press Ltd.)

Figure

Hg-367.8 k t v

I

I

I

54 o

k

mg level in kidney using fissioning neutron^,'^.^^ but still twice the normal content. The quantitative determinations of Hg in various organs using IVPGAA facilities are listed in Table 3 for comparison. Unfortunately, the detection limit for Hg in any target organ with prolonged scan periods is much higher than the normal content in patients, limiting its clinical applications to screening exposed individuals with high concentrations of toxic mercury.

3. Other Elements Several other elements have been measured by the PBIVPGAA technique, but immediate applications are not available. Among the elements with potential clinical interests is iron, which has been measured in vivo using neutron activation techniques, but only in patients with excessive overloads of the e l e m c n t ~ . ~ These techniques are unlikely to be used for the examination of other types of patients because of the poor sensitivity for this metal and the much higher doses required. Another potential application for medical diagnosis is lung silicon determination with PBIVPGAA scan. Silicon, in the form of free crystalline silica, is a potentially hazardous lung contaminant; normal contents up to 0.1 g per lung have been found in adults with no occupational exposure to silica-bearing dust. A study was made to assess the feasibility of determining the silicon level in human lungs in vivo by measuring the 1779keV y-rays arising from the neutron inelastic scattering reaction of 28Si(n,n'r)2%iwith neutron energies in the range of 5 to 8 MeV.35In measurements with an anthropomorphic phantom, a detection limit of 0.6 g silicon is obtained. This method is currently limited only to diagnose occupationals who are exposed to high concentration of silica-bearing dust.

B. WHOLE-BODY SCAN 1. Nitrogen As the nitrogen, directly proportional to the body protein content, is central to the structure of living matter, measurement of nitrogen concentrations in body parts or whole body may provide a quantitative estimate of patients for metabolic balance and nutritional assessment. The IVPGAA was first applied by the Birmingham University group on humans to measure the body nitrogena4Since that time, numerous Table 3 Detection Limit of Mercury in Organ Using Various IVPGAA Facilities Ref. (year)

Neutron source and strength

31(1980) 32(1982) 33(1983)

20 CP4'Am-Be 3 MeV from cyclotron 100 pg 252Cf

IO(1988)

THMER reactor at 0.1 W

a

Target organ

Brain Kidney Kidney Liver Kidney Liver

The normal content of mercury in organs is taken from Reference 2.

Detection limit

Normal contentd

100 PPm 16.7 mg 1.9 mg 5.4 Pdg 2.3 mg

0.10 ppm 0.87 mg 0.87 mg 0.30 pg/g 0.87 mg

2.2 Pgk

0.30 pg/g

-

Prompt gamma-ray energy

Figure 12 Prompt y-ray spectra for samples of (A) liquid nitrogen and (8)urea solution in a phantom as well as (C) pork medium. Each sample is irradiated by an external thermal neutron beam from the THMER facility and measured by a portable HPGeIMCA-10 detecting system for 9000 s. (Reprinted with permission from J. Radioanal. Nucl. Chem. Articles, 169(2),Chung, C. and Chen, Y Y, Determination of nitrogen in pork meat using IVPGAA techniques, 333, Copyright 1993, Elsevier Sequoia, S. A.)

IVPGAA facilities have been set up and large numbers of body nitrogen measurements have been performed on a regular basis. The weight percentage of nitrogen in the whole body of the Reference Man is 2.59%, or 25.9 g N per 1 kg human weight on the average? indicating that whole-body nitrogeli is around 1.8 kg for a person weighing 70 kg. Metabolic or nutritional problems in the human body may lead to the variation of body nitrogen content; hence, high-precision measurement of nitrogen becomes essential for quantitative determination of nitrogen using the IVPGAA technique. The basic principle of NPGAA determination of body nitrogen is well known; the nuclide may reach the excited state by the I4N -b n --+ ISN* reaction through neutron capture. Since the average lifetime of the excited I5N* is short (less than 10-l2 s), the residual excited nuclide emits the prompt y-rays readily and de-excites subsequently to the ground state of the stable 15N.The prompt y-ray of 10.829 MeV has an intensity of 14 photons for every 100 neutrons captured in the nuclei; these highenergy photons are free from interference by any known captured and decayed y-rays and can be explicitly identified in the y-spectrum measured by spectrometers such as the BGO scintillation and HPGe semiconducting detectors. A general layout of an IVPGAA facility for whole-body scan is illustrated in Figure 3, to which the patient is moving relative to the neutron beam during irradiation. Many research centers use NaT(T1) detectors with modified electronics to measure the 10.829-MeV high-energy prompt y-ray, whi!e others use the high-resolution MPGe detector, however, with less counting efficiency to identify it. For instance, the concentration of nitrogen in pork meat is determined by WBIVPGAA scan using Their high-resolution spectra were obtained in the CANBERRA 10-plus portable the THMER fa~ility.'~ system. A typical prompt y-ray spectnim of LN, in the phantom is shown in Figure 12A. The prompt y-ray peaks of nitrogen, together with those from shielding materials of the THMER facility (Pb and H), are clearly observable. The high-energy portion of the prompt spectrum of urea solution in the phantom, with a nitrogen ~ ~ n c e f i t r a t i oofn 4 0 g NllOO g, is illustrated in Figure 12B. The 10.829-MeV

Table 4 The WBIVPGAA Facilities Used for Body Nitrogen Measurement Reference (year)

Neutron source and strength

Cyclotron (p,n) Neutron generator Neuwon gcncrator X 2 75 Ci 238P~/Be 7.6 Ci Z3RPu/Be X 2 5.6 p g ='Cf 100 pg =?Cf 5 Ci "8Pu/Be X 4 200 pg =Vf

THMER reactor 15 Ci X 2

Scan time, Detector and size

s

-

Performance indexa

6" X 5" NaI(T1) X 2 6" X 4" NaI(T1) X 8 1 1.5" X 4" NaI(T1) X 2 6" X 6" Nd(T1) X 2 5" X 4" NaI(T1) X 2 8" X 6" NaI(T1) X 1 6" X 6" NaI(T1) X 1 5" X 4" NaI(T1) X 4 6" X 6" NaI(T1) X 2 HPGe X 1 YX2"BGOX 1 8" NaI X 4

' Performance index

= l/[scan time X radiation dose]. Information not available. Reprinted with permission from Appl. Radiat. Isot. 44(6), Chung, C., Wei, Y. Y., and Chen, Y. Y., Determination of whole-body nitrogen and radiation assessment using IVPGAA technique, 941, Copyright 1993, Yergamon Press Ltd., with update information added.

photopeak, together with its single-escape peak of 10.318 MeV and double-escape peak of 9.807 MeV, can be quantitatively identified. The high-energy portion of the prompt spectrum for pork n~edium irradiated for 9000 s with thermal neutrons is illustrated in Figure 12C. The integration of the 10.829MeV photopeak and its associated escape peak areas are enhanced. Hence, detemcnation of whole.body nitrogen using a high-resolution prompt y-ray spectrum is conlparatively straightforward although the scan time is longer than that using the NaI(T1) detector. In Table 4, the recent PGAA facilities used for whole-body in vivo nitrogen rncasurements i n published works are listed, with information on their neutron source, spectrometric detector, scan time, and performance index. More recently, CANBERRA INDUSTRIES, INC. commercialized the Body Protein Monitor, which contains four 8" NaZ(T1) detector arrays coupled to an MCA-100 multichannel analyzer and two isotopic neutron sources, each of 15 Ci; the total body protein, evaluated from nitrogen content, can be determined in 400-s scans.45 Conceivabiy, the WBIVPGAA measurement of body nitrogen will soon be widely accepted by the research community and hospitals.

2. Other Elements The initial application using the in vivo INAA technique is the measurement of aclivaled calcium in the body with various metabolic diseases. Since Ca in the body is normally located in the skeleton wholc-body, a WBIVPGAA rneasurcment of Ca can also provide direct examination of the total skeletal m a s 6 On the other hand, disorder of phosphorus in patients with renal failure has been obscrvcd to which the total body phosphorus detemined by WBIVPGAA gives the clinician useful data for the design of dialysis therapy.' The PGAA measurements of chlorine and carbon to date have not been applied to the same extent as those of body nitrogen and calcium. The WBIVPGAA measurement of body chlorine however, can be used to obtai'n accurate estimates of the intra- and extracellular compartments of body water.46 Prospectively, the body carbon measurement using WBNPGAA techniques should be an accurate index of the fat and lean tissue compartments of the body, providing the information of body nitrogen is available.47Longitudinal measurements of body carbon may even provide a direct technique for monitoring total energy expenditure in the patient.48 Since all the elements of interest in the, whole-body scan using IVPGAA technique, as those listed in Table 1, are either major or minor constituents in the body, both detection limit and precision are important for clinical diagnosis. The precision of determining body content of Ca, C1, P, C, N, and H in a WBIVPGAA scan with nominal radiation doses is in the range of 0.4-5.1% using the Brookhaven facility: while the commercialized IVPGAA monitor can measure the body protein content with a precision of 4% using the CANBERRA facility.45

Figure 13 The minimum detectable amount of body composition as afunction of scan time using WBIVPGAA technique with the THMER facilih].

Scan time,

1000 s

The detection limit of essential elements in the body using the WBIVPGAA technique with the

THMER facility is illustrated in Figure I3 at various irradiation times." Even with the 500-s scan period, the detection limits of N, P, Ca, and C1 are far less than the normal contents listed in Table 1. Thus, essential elements can be determined practically using the WBIVPGAA technique with high precision.

?IV. RADlP.TfON DOSES TO PATIENTS In most IWGAA facilities mentioned in the previous section, the dose equivalents for irradiated patients at the skin, depending upon the neutron flux intensity, dimension of irradiated area, length of irradiating period, 'and particular elements and orens in such medical diagnosis, ranged from 500 to 10,000 pSv per scan. Although the radiation doses at skin were reported in each and every previous IVPGAA work, only those reported by Tsing Hua's g r o ~ pincluded ~ . ~ ~ measured data and evaluation of dose equivalents according to health physics guidelines. In the NPGAA scan, the neutron beams are aimed at the kidney, liver, or lung where toxic contaminants accumulated, or irradiating whole-body for the determination of essential constituents. Organs and tissues sensitive to radiation, such as gonads, red bone marrow, bone surface, thyroid, breast, and lung, are either the irradiated organ itself or are not far from the neutron beam. Both neutrons and y-rays can scatter into the sensitive organs or tissues and interact with nuclei to cause radiation damage. It is the radiation dose in sensitive organs and tissues, not at the skin, that the radiation safety and protection in IVPGAA diagnosis should concern. In most IVPGAA reports, the neutron doses were simply assessed using a quality factor of 10 to convert the absorbed dose into dose equivalents; this is further confused by the radiation unit exchange between SI units of Sivert (Sv) and the more popular unit of rentgen-equivalent-man (rem). A simple measurement of thermal neutron flux inside the irradiated patient cannot resolve the complex evaluation of neutron dose equivalents, for which full knowledge of the neutron energy spectrum distributed totalbody is indispensable. The correct evaluation of induced radiation doses is further complicated by the internal dose caused by radionuclides activated therein. Since the majority of radiation doses are contributed from neutrons with various energies, the neutron flux distribution in the irradiated patient, together with radiation doses evaluated for partial-body and whole-body scans, are given in this section.

A. NEUTRON FLUX DISTRlEUT!ON IN BODY Since 90% of thermalized neutrons are absorbed by the skin and the first 8 cm of tissue?0 they contribute very little toward activating an element deep in the body organ; rather, they deliver unnecessary neutron doses. In order to eliminate the therrnal neutrons extracted from the IVPGAA facility, most researchers added a thermal neutron absorber (such as %contained material) as a filter to alter the neutron energy spectrum in the best possible manner. The uniformity of neutron Ilux in either irradiated organ or totalbody can be demonstrated by adding a thermal neutron filter with the bilateral scan technique.'' The thermal neutron flux in the phantom along the heam direction using the THMER facility, dcterrnined by activation foils and normalized to fission track data," is illustrated in Figure 14. Although the unfiltered neutron beam gives a high yield of neutron flux in the phantom, it is attenuated very

W y Limits -4

1,

Figure 14 Thermal neutron flux distributions at the posterior, or neutron beam tube direction, under = average neutron flux in various irradiation and neutron beam conditions using the THMER facility: body and & = average neutron flux in kidney. (Reprinted with permission from Appl. Radiat. Isot. 39(2), Chung, C., In vivo partial body activation analysis using filtered neutron beam, 93, Copyright 19013, Pergamon Journal Ltd.)

+,

rapidly, by a factor of 10 for every 8 cm of tissue-equivalent material, in a unilateral irradiation. It gives an uneven neutron flux distribution in s body organ such as the kidney. In a bi:d:eral irradiation using an unfiltered neutron beam, however, there is only limited improvement as far as the uniformity of the neutron flux is concerned. In contrast, the filtered neutron beam, with the thermal neutron absorber in position, gives a much more uniform distribution, although the mean neutron flux is only 40% in the total body, and 66% in the kidney of that in the unfiltered beam irradiation. Standard deviations for the mean neutron fluxes in the irradiated body, another indication of the uniformity of the neutron flux distribution, are also shown in Figure 14. The bilateral irradiation, using the thermal neutron absorber in the beam tube, yields the smallest standard deviation of the mean neutron flux in the organ. Thus, the greatest uniformity in in vivo activation is achieved by this arrangement; the thermal neutron flux generated in the phantom by moderation of the nonthernlal

COORDINATE cm Figure 15 Distribution of thermal neutron flux in a phantom along the (A) z-axis, (6)y-axis, and (C) xaxis in anterior and posterior irradiations using the IVPGAA technique. The coordinates are illustrated in Figure 5. (Reprinted with permission from Appl. Radiat. Isot. 36(5), Chung, C., Yuan, L. J., Chen, K. B., Weng, t? S., Chang, l? S., and Ho, Y H., A feasibility study of the in vivo PGAA using a mobile nuclear reactor, 357, Copyright 1985, Pergamon Press Ltd.)

neutrons is fairly uniform throughout the body. This arrangement is very suitable for IVPGAA to investigate the elemental composition of a body organ since most of the elements of interest are evenly activated by thermal neutrons. A more detailed thermal neutron distribution along the beam direction inside the irradiated phantom is mapped using the THMER facility? In Figure 15, the distribution of neutron flux in the phantom, with the neutron beam trained on the left kidney as the critical organ, is determined using the activation foil technique. As shown in Figure 15A, the thermal neutron flux distributes quite evenly in the kidney for both anterior and posterior irradiations. The evenly distributed thermal flux along the Z-axis is due to the moderation of nonthermal neutrons. In both the transversal (Y-axis) and longitudinal (X-axis) directions, as shown in Figures 15B and 15C, respectively, the thermal flux distributes symmetrically to the beam line and has a 10-cm field width of 50% of the maximum flux. The modification indeed localizes the neutron beam to the target organ; fewer neutrons are scattered in other organs, thus reducing the interorgan interference to <10%. The neutron energy spectrum and converted dose equivalents at all critical organs during IVPGAA scan cannot be properly measured since the neutron flux, only up to 10,000 n/cm2 s, is too weak to be detected whole-body. The neutron flux having energy Ei at distance r to the impact point with the angle 8 to the neutron beam direction, as illustrated in Figwe 68, without slowing down or absorption, is calculated by the following equation in neutron transport code:

where Sa(EiJ, Y) =

R

Sn(G) dx

*

dy/(*

+ y@, the average value of

measured source term i f

neutrons with energy Ei at the skin surface with coordinates X and Y to the impact point = empirical build-up factors in phantom, taken from Reference 49 =. exp [-P]IPdP, first-order exponential integral function

A& Ei(W X(EJ

1;:f, 1

.

= macroscopic renloval cross-section of neutron at energy Ei in phantom = (xi y;)ln, radius of neutron disc source at skin surface

+

Equation 4 is used to calculate the attenuated neutron flux at organs of interest for neutrons of the original incident energy Ei. For the slowing-down and diffusion of neutrons from initial energy Ei to final energy Ef, the neutron flux +,(Er,Ei,r) in organs at the distance r from the impact point can be represented as:

It', .;g

neutron age from Ei to Ef E' diffusion coefficient of neutron flux with energy E in phantom average log-energy decrement per collision of neutron in phantom macroscopic scattering cross-section of neutron at energy El in phantom

Thus, for a specific neutron energy El, the neutron flux &,(E,,r) in organ at distance r from the impact point becomes:

The first term on the right-hand sidc of the equal sign in Equation 6 is the neutron flux attenuated from

Figure 16 Integrated neutron flux as a function of neutron energy at various organs and tissues in the phantom using IVPGAA/THMER facility: (A) impact point an the skin surface, (0)kidney, (C) liver, (D) red bone marrow, (E) bone surface, (F) heart, (G) stomach, (H) lungs, (I) breasts, (J) intesiine, (K) gonads, and (L) thyroid. Converted dose rate equivalents (DRE) for neutrons are also labeled for each organ and

tissue in pSv/h. the impact point with the same energy; the second term is the sum of all neutron fluxes diffused from the impact point with different initial energy Ei that eventually slow down to Ef. Results of neutron flux distribution in organs and tissues of interest are calculated as above; the dose rate equivalents are converted using the conversion factors from ICRP.sl Calculated neutron energy spectra and corresponding dose rate equivalents for each organ are illustrated in Figure 16. From the shape of neutron energy spectra, it is seen that the thermal component dominates the neutron flux in organs far away from the impact point; this is because the thermal neutrons are building up from the slowing-down of nonthermal neutrons in the overlying tissues. Another feature is the rapid drop of dose rate equivalents as distance from the organ to the impact point increases. This indicates the wellcollimated neutron beam penetrates the phantoni upwardly and scatters little to other directions.

B. DOSES FOR PARTIAL-BODY SCAN As neutron dose equivalents have to be calculated using neutron transport code, the y-radiation dose, however, can be easily measured using either thermoluminescent detector (TLD) dosimetric chips or a miniature y-probe inserted into the phantom. Again, the PBIVPGAA scan using the THMER facility is taken as an example for dose equivalents as~essrnent.~~ Dose rate equivalents of y-rays in PBIVPGAA diagnosis were measured and evaluated for organs and tissues of interest, and their values are illustrated in Figure 17. As shown in this figure, y-ray dose rate equivalents drop less drastically than those of neutrons since the majority of y-ray incidents on the phantom originate from the scatterings oC high-energy y-rays that escape from the reactor core. It is only near the neutron beam line that the prompt y-ray flux is more than that of the scattering y-rays. At a distance of more than 40 cm away from the impact point transversely, the y-ray dose rate equivalents

4

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dose rate equivalents (ORE) in various positions inside the phantom for partial-body scan using IVPGAA technique with the THMER facility. Open points are measured with various foils and dosimeters; closed points are calculated results.

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in the phantom level off at 30 $3v/h. The dose rate equivalents of neutrons, converted from the calculated neutron energy spectrum at organs and tissues of interest, are also shown in the same figure as a function of both the distance from the impact point and the angle to the neutron beam direction. For comparison, dose rate equivalents converted from thermal neutron flux measured by indium foils positioned inside the phantom along the beam line are also shown. The few data points agree well with the neutron transport calculation and therefore the validity of the calculation is established. As shown in Figure 17, intermediate energy and fast neutrons with En 2 10 keV dominate the neutron doses everywhere in the phantom; neutrons with nonthermal energy are attenuated and slowed down rapidly inside the phantom, reducing the strength by a factor of 10 in every 8 cm of tissueequivalent material. The only exception is the build-up of thermal and epithermal neutron fluxes owing to the slowing-down of neutro~~s from nonthermal origins; this is expected in the THMER facility since the irradiated organ right over the neutron beam line can therefore interact with,a maximum thermal neutron flux for PBIVPGAA diagnosis. The lucite phantom, as well as the human body, is a good moderator for nonthermal neutrons and a modest absorber for thermal neutrons. As interpolated from Figure 17, the neutron dose rate equivalents for skin where neutrons leave the body after penetrating the 20-cm Iucite phantom i s 40 pSv/h; this is only 0.7% of the dose rate equivalents of 5780 p S v h at the impact point 20 cm down below.

Table 5 Neutron and ?-Ray Dose Equivalents (DE) per Scan for Organs at Risk in a 1800-s Partial-Body in vivo Diagnosis Using the Tsing Hua Mobile Educational Reactor Organ at risk Gonads Breast Red bone marrow Lung Thyroid Bone surface Remaindersb Weighted total

Weighing factor'

Dm), PSV

DE(r), CLSV

DE (all), PSV

0.25 0.15 0.12 0.12 0.03 0.03 0.30 1.00

0.45 2 0.22 0.63 1.26 7.47 t 3.74 1.00 1.99 0.06 2 0.03 4.76 -+ 2.38 153.16 2 76.58 47.53 + 22.98

13.91 rl: 1.39 17.24 1.72 21.03 2 2.10 16.18 1.62 13.16 1.32 19.18 1.92 33.14 r 3.31 21.44 1.13

14.36 1.41 18.50 1.83 28.50 14.29 18.17 1.90 13.22 1.32 23.94 t 3.06 186.30 -C 76.65 68.97 23.00

+

+

+

+ + + +

+ + + +

+

' Parameters taken from Reference 52. Remaining organs include liver, kidney, heart, intestine, and stomach. Reprinted with permission from Activation Analysis, 11, Chung, C., Activation analysis with small mobile reactor. Alfassi, Z. B., Ed., 314. Copyright 1990, CRC Press, Boca Raton, FL.

The internal dose for PBIVPGAA diagnostic purposes using neutrons is subject to a systcm of justification, optimization, and limitation of doses as far as the radiation exposure is concerned. The dose equivalents measured and evaluated ror those sensitive organs and tissues listed in ICRPS2are tabulated in Table 5. The weighted total dose equivalents per scan from both neutrons and y-rays is 69 pSv; this value is certainly well below the recommended annual limit of 5000 kSv and even less than most nuclear medical diagnoses using external ionization radiations (90 to 9000 pSv per scan) or radiopharmaceuticals (950 to 500,000 pSv per scan). Although most IVPGAA facilities reported only skin doses, their actual dose equivalents can be prorated to the results of the Tsing Hua group, to which simple skin doses and detailed dose equivalents are both available. In Table 6, the representative nuclear facilities using various/neutron sources with reported skin doses per scan are listed. If the neutron profile from various neutron sources is similar, their induced d x e equivalents for patient may be s f the same order of magnitude. Taking 3205-pSv skin doses corresponding to 69-pSv dose equivalents using the THMER facility, the dose equivalents for PBIVPGAA scans using neutron sources other than the reactor range from 11 to 215 pSv per scan; these again are far less than those doses induced by X-ray, CT, or MRI check-up and most of nuclear medical diagnoses such as SPECT and PET scans.

Table 6 Radiation Skin Doses and Dose Equivalents per Scan on Patient for Partial-Body and Whole-Body IVPGAA Measurements Using Various Neutron Sources Ref. (year)

Neutron source facility

Radiation dose pet scan

Partial-Body IVPGAA Measurements 100 mrem Z4'Am/Be 50 mrem Accelerator (p,n) 1000 mrem "ZCf 500-900 psv z3spu/ge 3205 pSv THMER reactor THMER reactor 69 pSv

37(1977) 39(1978) 40(1984) 42( 1986) 54(1986) 44(1993)

Whole-Body IVPGAA Measurements 100 mrem Cyclotron (p,n) Neutron generator 1000 mrem 50 mrem 238PdBe lS2Cf 2500 pSv ' THMER reactor 6410 FSV 63 pSv THMER reactor -

Remarks Skin doses Skin doses Skin doses Skin doses Skin doses Dose equivalents

Whole-body doses Whole-body doses Whole-body doses Whole-body doses Wli~le-bodydoses Dose equivalents

C. DOSES FOR WHOLE-BODYSCAN Although in the whole-body in vivo scan, the irradiated patient on the bed is moving relative to the neutron source, the measurement and evaluation of radiation dose are the same as those in a partialbody in vivo scan. All WBIVPGAA facilities, again except the Tsing Hua group, quoted unfortunately the radiation doses by either whole-body total doses or partial-body skin doses; no radiation dose evaluation for sensitive organs has been reported. Organs and tissues sensitive to ionization radiations, such as gonads, red bone marrow, bone surface, thyroid, breasts, lungs, liver, kidneys, heart, intestine, and stomach, are not far from the neutron beam path. Both neutrons and prompt y-rays can scatter into the sensitive organs or tissues and interact with nuclei to cause radiation damage. It is not at the skin nor is it the total doses that the radiation safety in WBIVPGAA diagnosis should concern. Here, the dose equivalents measurement and evaluation for whole-body in vivo scan performed by the Tsing Hua group are described for the health concern of the irradiated patient.44 In their investigation, neutron doses were measured and assessed at irradiated skin using the energy spectroscopic technique and calculated by two-group neutron transport code for all sensitive organs and tissues. y-Ray doses at positions where neutron doses were measured and calculated have been determined using dosimeters with background response subtracted. Features of dose equivalents are discussed under the context of radiation safety for such in vivo clinical application. The integrated neutron flux per energy range at the impact point on the skin surface was measured by the activation foil technique and calcu!atecl by neutron transport code; the neutron dose rate equivalents, converted from fluence-to-dose parameters, is 5780 pSv/h with the THMER facility operating at 0.1 W. However, it is the radiation dose at sensitive organs and tissues, rather than the skin surface, that causes health concern in the clinical diagnosis using the IVPGAA technique. Results of neutron flux distribution and corresponding dose rate equivalents in organs and tissues of interest are calculated using Equations 4 through 6. The neurron dose rate equivalents in the position inside the phantom, represented by the coordination r ,and 8 defined in Figure 6B, are calculated for thermal, epithemal, intermediate energy, and fast neutrons; the results are shown in Figure 18, for which some measured values are also indicated in thc figure. From both experimental and calculated data points, it is seen that the fast component dominates the neutron dose in the phantom; the dose rate equivalents drop rapidly as the distance to the impact point increases indicating that the well-collimated neutron beam penetrates the phantom upwardly and scatters little to other directions. Dose rate equivalents of y-rays in a WBIVPGAA scan were measured for organ's and tissues of interest by TLD devices and their values are also displayed in Figure 18. As shown in the plot, y-ray dose rate equivalents drop less drastically than those of neutrons; this is due largely to the dominant and homogeneous leakage y-rays originating from the THMER core. The dose rate at the-beam line increases to 630 p S v h since the y-rays emitted from the reactor core can strike th6 phantom directly. The geometric centers of thyroid, lungs, heart, liver, kidneys, stomach, intestine, female breasts, and gonads were selected to evaluate both neutron and y-ray dose rate equivdents. Selective spots for red bone marrow and bone surface were also assessed for the dispersion of radiation doses. The total dose equivalents could be deduced with uncertainty estimated at about 5 0 % for neutron doses and t10% for y-ray doses. The dose equivalents for 1-h whole-body scan, weighted among sensitive organs and tissues, are listed in Table 7, resulting in 63 -t- 8 p S v h for both neutrons and y-rays. This is certainly well below the national guidelines of 20,000 ~ S per V scan for the public and even less than most of the nuclear medical diagnoses using external ionizing radiations (at least 90 pSv per scan) or radiopharmaceuticals (e-g., 5000 pSv for PET scan). Again, if the neutron profile &om various neutron sources is similar to that extracted and filtered from the THMER facility, their neutron dose equivalents for irradiated patient may be of the same order of magnitude as those evaluated in the THMER facility. Taking 6410 pSv whole-body doses corresponding to 63 pSv dose equivalents for WBIVPGAA scan using neutron sources other than reactor, as those listed in Table 6, their dose equivalents range from 5 to 100 pSv per scan; these numbers are even less than those partial-body in vivo scans where the neutron beam is trained at the target orgnn of the irradiated patient. Unlike other nuclear medical methods, neutrons are exclusively used in IVPGAA medical diagnoses, and some nuclei interacting with neutrons eventually become radioactive. Induced radioactivities have to be calculated in order to evaluate the internal doses in the post-diagnostic period. Using the elemental compositions in Reference Man and assuming their hornogeneous distribution in the phantom, the activity for each activated radionucIide nt the end of a 1-h IVPGAA diagnosis using the THMER facility is calculated according to the neutron flux distribution inside the phantom. Results of the 15 activated

Figure 18 Neutron and y-dose rate equivalents (DRE) in various positions inside the phantom for wholebody scan using IVPGAA technique with THMER facility. Open points are measured with various foils and dosimeters; closed points are calculated results. (Reprinted with permission from Appl. Radiat Isot. 44(6), Chung, C.,Wei, Y. Y., and Chen, Y Y, Determinationof whole-body nitrogen and radiation assessment using IVPGAA technique, 941, Copyright 1993, Pergamon Press Ltd.)

Table 7 Neutron and y-Ray Dose Rate Equivalents (DRE) for Organs and Tissues at Risk in the Whole-Body in vivo Diagnosis Using Tsing Hua Mobile Educational Reactor Tissue at risk Gonads Breasts Red bone marrowb Lungs Thyroid Bone surfaceb Remainder' Weighted total

Weighlng factora

DRE (n), pSv/h

6.25 0.15 0.12 0.12 0.03 0.03 0.30 1.00

35.2 f 17.6 14.0 t 7.0 11.0 5.5 90.0 t 45.0 7.5 f 3.8 13.8 6.9 21.4 f 10.7 30.1 t 7.8

+

+

DUE (0,

pSv/h 33.0 32.4 32.4 35.0 26.8 43.9 31.1 32.6

t 3.3 fr 3.2 f 3.2 3.5 2 2.7 t 4.4 t 3.1 t 1.5

+

DRE (all), PSW 68.2 46.4 43.4 125.0 34.3 57.7 52.5 62.7

2 17.9

t 7.7

+ + +

6.4 2 45.1 4.7

8.2

t 11.1 -t-

7.9

' Parameters taken from Reference 52. Four spots from shoulder to legs are selected to represent the weighted average sections for red bone marrow; seven spots from skull to legs are selected to represent the weighted average sections for bone surface. Remaining organs include liver, kidneys, heart, intestine, and stomach. Reprinted with permission from Appl. Radiat. Isot. 44(6), Chung, C., Wei, Y. Y., and Chen, Y. Y., Determiriafiron of whole-body nitrogen and radiation assessment using IVPGAA techniques, 941, Copyright 1993, Pergamon Press Ltd.

Table 8 Activated Radionuclides in 1-h Whole-Body in vivo Diagnosis Using the Tsing Hua Mobile Educational Reactor -Activlty after Form of Reaction U, RadloHalfActivation -

-

nucllde

life 7.13 s

11 s 15 h

0.02s 9.46 min 2.24 rnin

-

reactions

barn

diagnosls, Bq

radiatlons

I5N(n,r ) '9F(n,a) IYF(n,r) ''Na(n, r ) nNa(n,r) " M g h r) nAl(n,r) "P(n,a)

2.62 h 14.3 d

37.3 rnin

86mRb

0.72 s 12.36 h 8.72 rnin 56 rnin 17.7 rnin 1.02 min

''P(n.p) "S(n,a) 3'P(n,r) "S(n,p) "Cl(n, r) W(n,a) "Cl(n, r ) 4'K(n,r) 4 T a ( n ,r) 68Zn(n,r) r hr) "Rb(n.r)

P-,r r

C

e-. r ,'P 0.047

0.01

P-, r P-, r EC, r IT, r

Note: Nuclear properties are taken from Reference 55.

radionuclides with induced activity greater than 0.1 Bq are listed in Table 8. From the tabulated values, 55% of the initial activity is from the decay of 0.02-s ""'Na; after its rapid decay, whole-body induced activity is reduced to 154 Bq. The total radioactivity is further decreased to 13% of initial activity I h after the diagnosis and eventually reduced to 2 Bq a day after the irradiation. Integrated internal doses from these radionuclides for 1 year are estimated to be 0.8 ~ S Vor, 1.7% of the internal doses caused by the natural 40K in the body; the induced radioactivity is considered insignificant compared to the internal natural radioactivity, and its health implication can be ignored.

Various neutron sources have been utilized to study the human body composition in vivo; the IVPGAA technique provides extra and indispensable information about the body composition and elemental concentration in tissue and organ. These partial-body and whole-body in vivo scans have been used in both clinical research and diagnostic applications; among numerous elements of interest, the measurement of organ cadmium and body protein wi!h nitrogen content are accepted most widely at the present. The methodologies have been advanced to commercial stage where facilities are exclusively designed and built for clinical applications. The use of mobile facilities to examine the critical organ cadmium and mercury burden in industrial workers should provide unique. information on the body's response to these'toxic metals. Routine monitoring of these populations at a regular interval has been proposed as a method to identify workers before the critical kidney burden level is e~ceeded.~ On the other hand, the measurement of body nitrogen by the whole-body IVPGAA technique is the only direct method to determine body protein mass. In the near future when the whole-body carbon measurements are developed further, a single measurement is anticipated to determine directly the fat, water, and protein compartments of the body. Such measurements should find applications in both the evaluation of nutritional status of the individual and in the monitoring of patients receiving modified dietary intake.3 Currently, the IVPGAA facilities all perform a reasonable scanning period for either partial-body or whole-body in vivo measurements, resulting in a dose equivalents range of 5 to 215 pSv per scan. These radiation doses are low compared to common rndiological examinations and nuclear medical diagnoses. The IVPGAA set-up yields a safe operation with negligible risk probability of radiation injury to the patient.

In this chapter, various IVPGAA facilities and their clinical applications are reviewed. The IVPGAA scan provides unique insight into elemental concentration in target organ and body compo_sition for clini2&l purposes. In vivo monitoring with minimal radiation doses of a patient's progress in the hospital as well as for the out-patient can be performed on a regular basis without discomfort and radiation risk. The recently commercialized IVPGAA units are expected in service for hospitals; routine uses are anticipated as another beneficial application of nuclear technology.

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26. Friherg, L., Piscntor, M., Nordherg, G. F., and Kjelistmm, T., Cadnriltm in Environment, 2nd ed., CRC Press, Cleveland, 1974. 27. Ishizabi, A., Fukushima, M., and Sakamoto, M., Cadmium in Japanese patients, Jpn. J. Hyg. 25, 86, 1970. 28. Ellis, K. J., Morgan, W. D., Zanzi, I., Yasumura, S., Vartsky, D., and Cohn, S. H., In vivo measurement of critical level of kidney cadmium: Dose effect studies in cadmium smelter workers, Am. J. Ind. Med. 1, 339. 1980. 29. Chen, W. K. and Chung, C., In vivo and in vitro medical diagnoses of toxic cadmium in rats. J. Radioanal. Nucl. Chem. Articles. 133(2). 349, 1989. 30. Schilling, R S. F., Occitpationnl Henlth Practice. 2nd ed., Butterworths, London. 1982, 106. 3 1. Al-Hiti, K., Al-Sidi, I. H., and Albedri, M. B., Determination of mercury levels in the brain by NAA, Appl. Radiat. Isot. 31, 563, 1980. 32. Smith, J. R, Athwal, S. S., ChettIe, P. R., and Scott, M.C., On the in vivo measurement of mercury using neutron capture and X-ray fluorescence, Appl. Radiat. Isot. 33, 557, 1982. 33. Vartsky, D., Ellis, K. J., Wieloposki, L., and Cohn, S. XI., The spectrum shape and the components of the background in IVPGAA, Nucl. Itzsrr. Meth. 2 13,437, 1983. 34. Cokn, S. H. and Parr, R. M., Eds., Nuclear-based techniques for the in vivo study of human body composition, Clin. Phys. Physiol. Meas.. 6, 275, 1985. 35. Ettinger, K. V., Morgan, W. D., Miola, U.J., Vnrtsky?D., Ellis, R.J., Wielopolski, L., and Cohn, S. H., Silicon measurement in a lung phantom by neutron inelastic scattering, Med. Phys. 9(4), 550, 1982. 36. Chung, C. and Chen, Y. Y.,Determination of nitrogen in pork meat using IVPGAA techniques, J. Radioanal. Nucl. Chem. Articles, l6!l(2), 333, 1993. 37. Dabek, J. T., Vartsky, D, Dykes, P. W.,Hanlwicke, J., Thomas, B. J., Fremlin, 3. H., and James, H. M., Prompt gamma neutron 2::tivation analysis to measure whole body nitrogen absolutely: Its application to study of in vivo changes in body composition in health and disease, J. Radioanal. Chem. 37. 325. 1977. 38. Oxby, C. B., Appleby9D. B., Bmoks, K., Burkinshaw, L., Krupowicz, D. W. K., McCarthy, I. D., Oldroyd, B., and Ellis, R. E., A technique for measuring total-body nitrogen in clinical investigations using the "N(n,2n)I3N reaction, Int. J. Appl. Radiat. Isot. 29, 205, 1978. 39. Williams, E. D., Boddy, K., Rarvey, I., and Baywood, d . K., Calibration and evaIuation of a system for total body in vivo activation analysis using I4 MeV neutrons, Phys. Med. Biol. 23, 405, 1978. 40. Boddoe, A. H., Zuidmeer, EL, nnd IIill, G. L., A prompt gamma in vivo neutron activation analysis facility for measurement of total body nitrogen in the critically ill, Phys. Med. Biol. 29(4). 571. 1984. 41. Allen, B. J., Blagozevic, N., Gaskin, X., Souter, V., and Howman-Giles, k,I: vivo determination for protein by prompt neutron capture in fibrocystic disease, in Capture Gamma-ray Spectroscopy and Related Topics, AIP, New York, 1985, 820. 42. Larson, L., Alpsten, M., and Mattson, S., In vivo analysis of nitrogen using a "Cf source, in Proc. of the 7th In?. Con$ Modem Trend Activation Analysis, Isotope Division, Riso National Lab, ~oskitde.Denmark. 1986, 1425. 43. McNeil, XC. G., Harrison, J. E., and Xrishnan, S. S., In vivo measurement of nitrogen by NAA, J. Radioanal. Nucl. Chem. Articles, 1 10. 655. 1987. 44. Chang, C., Wei, Y. Y., and Chen, Y. Y., Determination of whole-body nitrogen and radiation assessment using IVPGAA technique, Appl. Radiar. Isor. 44(6), 941, 1993. " 45. Canberra Nuclear Instrument Catalog, 9th ed., Canberra Industries, Meriden, 1993. 46. Yasumura, S., Brennon, B. L., Letteri, J. M., Zonzi, I., Ellis, K. J., Roginsky, M. S., and Cohn, S. H., Body electrolyte composition in normal subjects and hypertensive patients on therapy, Miner. Electrol. Metab. 2, 94, 1979. 47. Kehayias, J. J., Ellis, K. J., Cohn, S. H., Yasumura, S., and Weintein, J. H., Use of a pulsed neutron generator for in vivo measurement of body carbon, in In Vivo Body Composition Studies, Ellis, K . I., Yasumura, S., and Morgan, W.D.,Eds., Tile Institute of Physical Sciences in Medicine, London, 1987,427. 48. Burkinshaw, L., Models of the distribution of protein in the human body, in In Vivo Body Composition Studies. Ellis, K. J., Yasumura, S., and Morgan, W. D., Eds.. The Institute of Physical Science in Medicine, London, 1987, 15. 49. Chung, C., Chen, C. P., and Chang, P. S., Radiation dose from medical IVPGAA using a mobile nuclear reactor, Health Phys. 55(4), 67 1 , 1988. 50. Boddy, K., East, 13. W., and Robertson, I., Geometric factors in partial body in vivo activation analysis using epithermal and fast neutrons, Appl. RadiaL Isor. 21, 500, 1970. 51. International Commission on Radiological Protection, Data for protection against ionizing radiation from external source, ICRP Report 21, Pergamon Press, Elmsford, NY, 1973. 52 International Commission on Rndiological Protection, Recommendation of the ICRP, Publication 26, Pergamon Press, Elrnsford, NY, 1977.

53. Franklin, D. M., Chettle, D. R., and Scott, M. C., Studies relating to the accuracy of in vivo measurements of liver and kidney cadmium, J. Radioanal. Nucl. Chem. Articles, 114, 155, 1987. 54. Chung, C., Tseng, T. C., and Yuan, L. J., Determination of Ca, CI, N, and P by in vivo actidiion using mobile reactor neutron beam, in Proc. 7th In(. Car$ Modern Trend Activution Analysis, Isotope Division, Riso National Laboratory, Roskilde, Denmark, 1986, 1393. 55. Lederer, C. M. and Shirley, V. S., Table of Isotopes, 7th ed., John Wiley & Sons, New York, 1978.

Chapter 8

In Situ Applications Jiunn-Using Chao CONTENTS I. Introduction

...............................................................................................................................

131 131

11, ISPGAA Instrumentation ........................................................................................................ A. ISPGAA Probe ................................................................................... 131 B. Neutron Sources ...................................',..........................:.............................................. 132 C. Detectors .............................................................................................................................. 134 111. Oil Formation In Situ Analysis ................................................................................................ 135 A. Bulk Properties of Rock Formation ...............................:................................................ 137 B. Multielemental Analysis ..................................................................................................... 138. IV. Coal Mine In Situ Analysis ...................................................................................................... 139 A. Coal Ash Delineation Using Scintillation Detectors ......................................................... 140 B. Multielemental Analysis Using High-Resolution Detectors ............................................142 1. Analysis Using lsotopic Nentron Sources ............................................................. 143 2. Analysis Using Neutron Generators .............................................................................. 145 V. Environmental Water Body In Situ Survey ............................................................................. 145 A. Performance Test ................................................................................................................. 146 B. Field Survey Using the ISPGAA Probe ........................................................................... 148 C. Radiation Safety Concerns ................................................................................................. 149 VI. Planetary In Situ Exploration ............................................................................................. 150 VII. Seabed Mineral In Sit!{ Survey ............................................................................................... 153 VIII. Conclusions .............................................................................................................................. 154 References ...................................................................................................................................... 155

..................... . . .

I. INTRODUCTION In situ prompt y-ray activation analysis (ISPGAA) has been employed as an important well-logging technique in the petroleum industry since the 1950s. Investigators have expanded this technique to other applied fields over the last 30 years. In contrast to reactor-based PGAA, the ISPGAA7uses portable neutron sources with low-intensity neutron flux below IOXn/cm2 S; on the other hand, it acquires yray information from neutron-interacting samples of considerably Iarge volume. It is convenient for scanning samples in large qu'mtity, samples that are impossible to move to the lab for analysis. In addition, more representative elemental compositions can be obtained in a large-size sample. In sitic operation of a PGAA detecting system poses many challenges that are not encountered in the laboratory PGAA. Sensitivities of the elements of interest are severely limited by the neutron output from the neutron sources; background radiations induced by neutrons are complex, as are collecting prompt y-signals; hostile in situ environments of wind, precipitation, temperature, and humidity hamper the performance of detecting systems; and inevitable radiation damage to the equipment may occur due to prolonged neutron irradiation. The rapid development of nuclear instrumental technology has made the ISPGAA technique more versatile and flexible in applied fields, satisfying increaqing demands in analytical aspects. In this chapter, application using the ISPGAA technique in petroleum and coal well-logging, elemental and pollutant concentration monitoring in environmental water bodies, sea bed mineral surveys, and planetary surface explorations are introduced. Specification of the neutron sources, y-ray detectors, and nuclear electronics used in ISPGAA are described, and future research and development in ISPGAA are discussed.

!I.

ISPGAA INSTRUMENTATION

A. ISPGAA PROBE Applications of the ISPGAA technique to oil.and coa! borehole logging as well as water pollutant survey demand rapid acquisition of information regarding the distribution of elemental concentrations 0-8493-5 149-9lg51fiO.W+$.SO Q 1995 by CRC Press. Inc.

- ELECT'RONICS -

SCINTILLATION DETECTOR

SHOCK MOUNTING i NEUTRON SHIELD

- ISOTOPIC NEUTRON SOURCE

Figure 1 Typical neutron-y logging probe. (Reprinted with permission from Nucl. Instr. Meth. 158, Eisler, I? L. and Huppert, f?, A nuclear geophysical borehole logging system, 581, Copyright 1979, Elsevier Science Publishers B. V.)

from place to place at various time periods. Therefore, a small, compact, and mobile detecting probe is needed. The basic configuration of an ISPGAA probe consists of neutron suurcc, y-ray detector, radiation shielding materials placed in between, and the associated electronics with cables connecting to the analyzer above ground, as illustrated in Figure 1. The space between neutron source and detector must be optimized according to neutron intensity to improve counting sensitivity and signal-to-noise ratio, and minimize the neutron irradiation damage. A short source-to-detector distance can improve vertical resolution of the rocWsoil formation profile in a borehole, but inevitably increasing interferences in the y-spectrum and radiation damage to the detector. There are various types of neutron sources and detectors used in borehole analysis, offering several combinations for different borehole characteristics and different functioning analytical capabilities.

6. NEUTRON SOURCES There are two types of neutron sources utilized for ISPGAA measurement: neutron generators and sealed isotopic neutron sources. A neutron generator has the advantage of producing high neutron yields, being operated in pulse mode to distinguish indastic and capture prompt y-rays, and can be switched off when not in use. However, its complex electronics and ion source occupies the bulk space in an ISPGAA probe and the remaining room of the probe must be reserved for the attached power supply and the detecting system. The most popular type of neutron generators are based on the reaction of 3H(d,n)4He,producing monoenergetic neutrons of 14 MeV, by which an inelastic scattering reaction with surrounding sample is readily initiated and the subsequent prompt y-rays can be utilized to detect the elements that can hardly interact with thermal neutrons. However, the detector damage caused by fast neutrons is severe,'-' while the lifetime of a typical neutron generator is limited to several hundred hours, making frequent replacement of both generator and detector a necessity. A typical D-T neutron generator "Zetatron" is designed and dedicated to borehole logging applications,4 mainly consisting of two parts: ion source and ion accelerator, as illustrated in Figure 2, with a diameter of only 38 mm. The use of a pulsed generator for detection of inelastic prompt y-rays is a common practice in borehole logging for the determination of carbon and oxygen contents as well as other minor ele~nents.

MPOUT

BULATION PLASMA CUP

H V GLASS

3

\

'-

T

ION ACCELERATOR

ION SOURCE

ZETATRBN Figure 2 Zetatron neutron tube. (Reprinted with permission from IEEE Nucl. Sci. 28, Slope, L. A. et at., Operation and life of the Zetatron: A small neutron generator for borehole logging, 1696, Copyright 1981, The Institute of Electrical and Electronics Engineers, Inc.)

Figure 3 shows a brief operating cycle of the pulsed system in borehole 10gging.~The 18-ps interval of neutron burst is followed by a relatively long interval between pulses. Three spectra1 measurements by an NaI(T1) detector are timed in sequence and thus inelastic, prompt and delayed y-ray spectra can be recorded separately for mrlysis. Unfortunately, the detector pulse pile-up during the neutron burst restricts its usefulness only i n low count rates. Vartsky et al. demonstrated a solution to this problem using a multi-analog-to-digital converter (ADC) system to process more than one detetor pulse per neutron burst, permitting operation of the generator at high neutron burst output without suffering from detector pulse pile-up during the neutron bursL6 In contrast to the neutron generator, the isotopic neutron source has the advantage of occupying minimal space and independence of power supply, popular in the applications demanding a light-weight, compact probe. There are several types of isotopic neutron sources commercially available; among them, the 2.65-year 25ZCfspontaneous fission source is ideal for thermal neutron capture reactions due to its relatively low average neutron energy of 2.2 MeV. It has the highest specific emission rate of SPECTROSCOPY TIMING FOR INELASTIC AND CAPTURE RESPONSE

-

NEUTRON BURST -

I

I

TIME ( p ~ INELASTIC 6 CAPTURE

0

-

Y RAY COUNT RATE

SPECTRAL GATES

I

,

100

1

/CAPTURE

\

0

36

10

1

IBURST BKGNO n

-n

1

90

CAPTURE

Figure 3 Inelastic mode spectrometry timing scheme. (Reprinted with permission from Nucl. Geophys. 1, Grau, J. A. and Schweitzer, J. S., Prompt yray spectral analysis of well data obtained with Nal(TI) and 14 MeV neutrons, 158, Copyright

BURST

BACKGROUNO

INELASTIC

1987, Pergarnon Press, Ltd.)

GATE

GATE

SPECTRUM

I

BURST

CAPTURE SPECTRUM

2.34 X lo6 nls . pg and the lowest interfering y-ray yield. It was recommended by Senftle that the 252Cfneutron source with intensities of 2 to 40 pg is suitable for ISPGAA borehole logging.' An isotopic neutron source, producing neutrons by means of the (a,n) reaction, is also frequently used in borehole analysis. Such a neutron source is, in essence, a combination of a thin a-emitter as well as a light element of Li, Be, B, or F. The neutron energies of these isotopic neutron sources (e.g., the 433-yr 24'Am-Be)are sufficiently high (>4 MeV) for exciting. nuclides in the way of inelastic scattering, having the same effect as a neutron generator. On the other hand, the thermalized neutrons in the sample matrix may undergo capture reactions followed by emission of capture photons, which simultaneously appear with the inelastic prompt y-rays in the spectrum.

C. DETECTORS Both scintillation and semiconducting detectors have been employed as a counter or a spectronieter for ISPGAA. An ideal detector should fulfill the following properties: High detecting efficiency High count rate Good energy resolution Resistance to neutron damage Free from cryogen Ruggedness and nonhydroscopic Independent of temperature variatibn Among these detectors, the NaI(T1) scintillator with energy resolution of 5 to 10% has been used extensively in ISPGAA applications. Its important advantages of moderate scintillation efficiency and mild resistance to neutrons make it viable in hostile environments. Nevertheless, a variety of scintilhtors other than the NaI(T1) detector have been introduced as altel-natives for irr situ measurements in different conditions. The bismuth germanate (Bi4Ge3OI2) detector; comrnoniy abbreviated BGO, has high detection efficiency and nonhydroscopic properties, enabling it to perform a rapid survey in hostile environment^.^ Although the energy resolution of the BGO detector is much broader than NaI(Tl), the high-energy detecting efficiency of E, > 5 MeV makes it an ideal counter for high-energy y-ray detection in ISPGAA. The cadmium tungstate (CdW04) detector shows some promise for ISPGAA applications due to its high detection efficiency and reasonable light output, but its slow response limits its use to low count rate measurement^.^ The recently introduced barium fluoride (BaF2) scintillator has a fast response component of less than 1 ns, indicating its potential for high count rate m e a s ~ r e m e n t The . ~ ~ ~intensity ~ of the emission wavelength at 220 nm for the BaF, detector is independent of temperature variation although it is weak for coupling to a PMT.8 The unique property of the cerium-doped gadolinium orthosiIicate (GSO) detector can achieve a higher counting efficiency by varying cerium concentration without cooling as other scintillators do in nuclear well-logging measurements." The detection efficiency of inorganic detectors decreases with increasing temperature due to variations of the light output response time constant and the light emission spectrum,1° as shown in Figure 4. Such degradation and spectrum shift should be well understood prior to practical use in the field. For example, the emission spectrum of a scintillating material is more critical for matching a photocathode response spectra, since the greater restriction on photocathode materials is imposed in the logging industry. In addition to y-detectors, a variety of associated electronics capable of operating at high-temperature (up to 175 OC) borehole environments were developed." The energy resolution for detecting a y-peak is degraded with increasing temperature of the scintillation detector, owing to the reductims both in the light output of the scintillator crystal and in the photocathode sensitivity of the PMT. The adaptability of PMT optically coupled to the scintillator to the borehole environments has been investigated. A ceramic PMT was developed and proved to have excellent performance at temperatures up to 200 OC,a long operating life, and extreme mechanical ruggedness.13 With the enhancements in detector assembly packaging, the impacts of thermal, mechanical vibration, and magnetic field on the system can be minimized. In comparison with inorganic scintillation detectors, the semiconducting detectors exhibit an excellent energy rssolution that enables detailed elemental analysis in situ. However, disadvantages of the semiconducting detector listed below limit its usefulness in a borehole environment:

-

Figure 4 The emission spectrum of BGO shifts approximately 11 nm between -20 and +85 "C. (Reprinted with permission from lEEE Nucl. Sci.35, Melcher, C. L. and Schweitzer, J. S., Gamma-ray detector properties for hostile environments, 877, Copyright 1988,The Institute of Electrical and Electronics Engineers, Inc.)

350

450

650

850

760

Wavelength (nm)

Operation at liquid nitrogen temperature Prone to neutron damage Inferior detection efficiency Sensitive to temperature and moisture variation Liquid nitrogen, used as a cryogen for semiconducting detectors, suffers from the problem that the high-pressure conditions of boreholes result in an increase in temperature of liquid nitrogen, which can cause spectrometric drift, generate noise, and deteriorate spectral resolution. To countermeasure such inferiority, melting cryogens with large latent heats were proposed to replace liquid nitrogen as potential alternative^.'^ Liquid propane and fluorocarbon refrigerants, having properties of high boiling points and inert characteristics, were recommended as substitutes for operation in a borehole, outer space, and planetary surfaces.1616 A pressure vessel filled with propane was designkd by Boynton for such application, free from impurities for refill equipment,15 as shown in Figure 5.

El!. 81L FORMATION IN SlTU ANALYSIS The ISPGAA technique has been introduced to petroleum industries for exploration and evaluation of oil reserves for more than two decades. Early ISPGAA restricted its utilization to the determination of the presence of carbon contents, bulk properties of oil formations, and Today, apart from knowing where and how much oil is in the formation, realizing how to produce oil efficiently from current reservoirs is urgent!y demanded from an economical point of view. Hence, an understanding of the minerals in rock fornlation by means of ISPGAA multielemental analysis is of great In contrast to traditional electric, acoustic, and radioactivity logging methods, ISPGAA provides much more direct information regarding borehole environments by means of elemental analysis, which is derived from the measurement of characteristic prompt y-rays of concerned elements. More importantly, this method can be carried out in the cased boreholes with rapid readout of results while most of the traditional methods cannot. However, problems of interpretation arise when the desired measurement is perturbed by changes in other formation or sample chnntctcristics. The disturbances of the f~rmation,?~ which can invalidate a laboratory calibration of the probe, are:

Variations of formation constituents Density changes

Figure 5 Sketch of typical borehole probe using canister cryostat. (Reprinted with permission from Nucl. Instr. Meth. 123, Boynton, G. R., Canister cryogenic

system for cooling germanium semiconductor detectors in borehole and marine probes, 602, Copyright 1975, Elsevier Science Publishers B. V.) Moisture content Occurrence of elements of high neutron cross-section Interaction of measuring tool compound with neutrons Temperature changes The disturbances listed may perturb tool calibration by influencing the spatial and energy distributions of the neutron field, leading to important differences in interaction cross-sections and average probability of induced prompt y-rays reaching the detector. In order to simulate the actual statistical nature of the interaction processes, Monte Carlo techniques were developed for tracking the neutron and y-ray transport through matter, providing interpolation of results for the logging conditions difficult to simulate e~perirnentally.~~

Water -Saturated

O:.!i

Figure 6 Comparison of a 30-porosityunit sandstone formation oil- and watersaturated. (Reprinted with permission from J. Pet. Tech. July, Scott, H. D., New developments in remote analysis of rock formations, 712, Copyright 1986,Society of Petroleum Engineers.)

sat_u_ro_t_e_d-_d

Energy (MeV)

A. BULK PROPERTES QF ROCK FORMATIOM The economic effectiveness of an oil reservoir depends strongly upon its physical and chemical properties. The organic-rich source rock to generate oil and/or gas, consisting of a variety of hydrocarbons, is an essential subject for probing. The reservoir must have porosity to contain oil andlor gas and permeabiIity to permit fluid flow. Until the advent of high-resolution semiconductor detectors for in situ application, the NaI(T1) scintillator was the only c!~oice for y-ray spectroscopic analysis. Several elements can be quantitatively identified in the prompt y-ray spectrum with its mild resolution. Some essential parameters leading to the confirmation of the quality of oil reservoir, such as hydrocarbon content, can also be determined to some extent. The earliest application of the ISPGAA technique in oil well logging was to differentiate oil from water in formations having low salinities, offering information for evaluation of.thedegree of hydrocarbon saturation and subsequently the size of the reservoir. The technique is referred to as "carbon/oxygen l~gging";'~.~' the measured ratio of carbon to oxygen indicates the fraction of pore space occupied by oil and water. These two elements can be determined readily by measuring their inelastic prompt yrays: 4.43 MeV and 6.13 MeV emitted from I2C(n,n'y) and 'hO(n,n'y) reactions, respectively. An NaI(T1) detector coupled to a neutron generator operated in pulsed mode is sufficiently adequate for such m e a s ~ r e m e n t . ~The * ' ~prompt ~ y-ray spectra, as illustrated in Figure 6, show differences of the spectra coIlected in oil and water tanks in a laboratory test for the determination of optimal response of an ISPGAA system; the carbon-to-oxygen ratios for the oil and water tanks are 7.6 and 1.2, respe~tively.'~ The measurement of C/O logging can be significantly altered if the salinity of borehole formation is higher than 0.1%. As salinity increases, the y-ray response from chlorine (notably 6.1 1 MeV) also increases; this has the effect of reducing the actual C/O ratioI9 since chlorine capture y-rays are of the same energy as those from oxygen. On the contrary, the chlorine usually presents in the form of salt water and its prompt y-rays dominate the y-ray spectra; some investigators took early advantage of this to develop chlorine logs to measure the absence of saltwater underground and to infer the presence of hydrocarbons."*26These logs have largely been replaced by pulsed neutron capture logs, which are also dominated by chlorine response. The measurement of C/O logging is subject explicitly to the lithology of the formations surrounding boreholes. The carbon-dominated rock, for example the limestone, presents a false high C/O ratio to make the sand appear potentially oil productive. This can be corrected by knowing the content of major constituents or elements of formations. Thermal neutron capture prompt y-rays of 3.54 MeV in Si(n,y) and 1.94 MeV in Ca(n.7) reactions can he measured simultaneously to indicate the 1ithoIogyof formation. The ratios, denoted as SiKa or-Si/(Si -1- Ca), can be used as the lithology indicator^.'^.^^ A more explicit explanation of the environment of boreholes using both C/O and Si/Ca logging are illustrated in Figure 7, in which the rich oil sands correspond to high C/O and Si/Ca ratios, while limy sand or carbonate presents a low SiICa ratio with relatively high C/O ratios.19 The pore of formations is occupied by either water andlor oil; both contain hydrogen and can be measured using the capture prompt y-mys of 2.22 MeV emitted from the H(n,y) reaction. The ratio

S i / C a RATIO

Figure 7 Predicting formation productivity in limey/shaly sands from C/O and Si/Ca ratios. (Reprinted with permission from J. Pet. Tech. Sept., Lock, G. A. and Hoyer, W. A., Carbon-oxygen (C/O) log: Use and interpretation, 1047, Copyright 1974, Society of Petroleum Engineers.)

+

R(@) = W(Si Ca) was proposed as a porosity i n d i c a t ~ rin , ~ which ~ silicon and calcium are assumed to be the major constituents of rocks. The influence of porosity on the determination of the C/O ratio is predicted for limestone (CaC03), dolomite [CaMg(C03)2],and sandstone (SiOz) fomlations by theorectical calculation,24as illustrated in Figure 8. In combination with the C/O logging, the oil quality as well as quantity in formation are thus evaluated.

6. MULTIELEMENTAL ANALYSIS The detailed elemental analysis of formation constituents permits an understanding of the mineralogy of rock rather than only its lithology. This appiication expands rapidly, although it plays an insignificarrt role for oil exploration alone. It has pointed out the importance of elemental analysis of formation in interpretation of rock chemical compositions and properties relating to the oil produ~tion,~'-~~ as listed in Table 1. In resronse to this, an in situ logging is divided into three groups: direct, indirect, and inferred. Direct measurement is related to raw data of elemental concentrations for the fluids and minerals that comprise a formation. Indirect application, on the other hand, utilizes the output of the direct application, such as mineralogy, to estimate formation properties such as total clay content or cation exchange capacity (CEC). Finally, the inferred application uses the direct and indirect outputs to deduce or infer information about the history of the rock formation. The pernleability, anZiijx5iiant factor in determining oil productivity of a rock, is strongly affected by minerology, or clay type^.^'.^^ Clay minerals have a critical role in the production of hydrocarbons. For example, clays typically fill the space between sandstone grains, thus reducing the oil production, or oil flow from the formation ATOM lC RATIO

0a8-

0s7:

0.6

/-

DOLOM IT€ Ca M9 KO312

-LIMESTONE DOLOMITE

0.4-

0.2 -

Figure 8 Calculated dependence of atomic carbon-oxygen

-

+ are in fractional units. (Reprinted with permission from IEEE

0.1

<,

OFo:l

wtio on oil saturation(S,), porosity(+), and lithology. So and

, ' /'\SANDSTONE 0 2 - 3 04:

1.0

0.5

+

sioz POROSITY

Nucl. Sci. 26, Hertzog, R. C., Neutron-excited gamma-ray

spectrometry for well logging, 1564, Copyright 1979, The Institute of Electrical and Electronics Engineers, Inc.)

.

Table 1

Elements Important to Formation Analysis Common occurrences

Element

H C 0

Na

Mg A1

Si S

Cl K Ca Ti V

Fe Th

Minerals

Bound water Carbonates Most minerals Halite, feldspars, clays Dolomite, clays Clays, micas, feldspars Quartz, clays, micas, feldspars An!~ydrite,pyrite, gypsum Halite Mica, illite, feldspars Calcite, dolomite, anhydride, gypsum Resistates Clays Pyrite, siderite, clays Associated with clays

Fluids

Water, oil, gas Oil, gas

Water Saltwater

Saltwater

Heavy oil

~ e ~ r i $ dwith permission from J. Pet. Tech- July, Scott, H. D., New developments in remote elemental analysis of rock formations, 712, Copyright 1986. Society of Petroleum Engineers. rock. The presence of small amounts of clay minerals in formations can affect the permeability of the formation to fluid flow. Besides, clays with high CECs can affect traditional electric logging measurements designed to infer oil saturation. Identification of different clay minerals and their concentrations yields CEC corrections to the electrical logging mea~urernents.~'*"~~~~~' The presence of certain other minerals in formation reservoirs can cause large errors in the traditional interpretation of logging measurements. For example, the presence of pyrite will affect electrical and neutron absorption logging meas~rernents.'~ Some trace elemental concentrations, such as those of boron and gadolinium, may become significant in the interpretation of neutron absorption logging measurement^.^^^^' Table 2 provides a list of formation elements currently detectable by ISPGAA. These include measurements of inelastic, prompt y-rays, radioactive decay y-rays, and natural radioactivity. The preferred nuclear reactions and associated y-rays are also listed in this table. However, it is difficult to relate spectral information to elemental concentrations and its composition in the formation due to the lack of a sufficient number of elements describing the rock. In order to solve this problem, the measured spectra are decomposed into elemental standard spectra using the method of weighted leastsquares to explain elemental composition in Accurate estimates can be achieved by including more elemental spectra for analysis. Recently, with the expending app?ication of HPGe semiconducting detector in the logging instrument?8-29.3ec! more information with respect to borehole formation can be explained by knowing major and minor elemental contents. Ample efforts have concentrated on the use of a pulsed neutron generator for mulrielemental analysis in boreholes. Evans et al. used a timing gate technique to separate inelastic, capture, and decay y-rays operatec! in a mock-up borehole filled with dry or wet ample;^' spectra collected for wet samples are illustrated in Figure 9. The optimum time scheme of the detecting system was determined in terms of signal-to-noise ratio to improve the sensitivity of in siru well logging.

IV. COAL MINE IN SIBU ANALYSIS The ISPGAA techniques app!ied to the field of coal analysis can be divided into two major categories: in situ borehole logging and on-line analysis. More attention has been paid to the latter due to stringent environmental protection codes, and this will be discussed in the next chapter. In situ coal analysis using the PGAA technique in a borehole is quite similar to that of geochemical logging for oil-bearing formation. In situ borehole logging for coal exploration yields information on the location, thickness, and quality of the coal seam. The c o d quality of ash content, heat capacity, sulphur content, and moisture can be determined readily by the ISPGAA technique.

Table 2 Elemental Analysis by Neutron-Induced ?-Rays Y

Reaction

u, mb

(thermal)

Wn,y) 2H '2C(n,y)"C 12C(n,n'y)'2C '60(n.n'y)'Q

322 3-4

-

'60(n,p)'6N

=Na(n,y)"Na 24Mg(n,y)"Mg 26Mg(n,y)27Mg 27Al(n,y)2"A1 27Al(n,y)28A1 "Al(n,n'y)27Al 28Si(n,y)?3i 28Si(n,n'y)28Si 32S(n,y)3'S 40K(natural) 4'K(n,y)42K 40Ca(n,y)4'Ca 48Ca(n,y)49Ca "V(n,y)S2V SsMn(n,y)S6Mn 'SMn(n.y)S6Mn Te(n,y)-"Fe MFe(n,n'y)S6Fe =Th(natural)

530 53 38 23 1 23 1

-

840

49 1

524 -

-

2100 430 1100 4880 13300 13300 2550

-

-

2223 4945 4439 6129 6129 1369 3539 844 7724 1779 844 3539 1779 84 1 1461

-

-

'

1942 3084 1434 847

-

T112

770

-

747

1,

keV

425 474 47

160

-

?'-ray energy,

m, rnb (14 MeV)

7244

7632-7645 847 2615

Prompt Prompt Prompt Prompt 7.13 s 15.0 h Prompt 9.46 rnin Prompt 2.24 min Prompt Prompt Prompt Prompt 1.28 X lo9yr Prompt Prompt 8.72 min 3.75 min 2.58 h Prompt Prompt Prompt 1.41 X IOl"yr

%

100

68 100

100 69

100 68 72 27 100

68 76 11 51 73 92

100

99 12 53

36

From Refirences 32 to 34.

A. COAL ASH DELINEATION USING SCINTILLATION DETECTORS The location and thickness of coal piles can be simply delineated by observing the prompt y-rays emitted from the borehole formati0n,4"~ as illustrated in Figure 10. The high spectral responses at an energy window from 3 to 10 MeV of the dirt bands relative to the coal bands is due primarily lo highenergy (>3 MeV) capture y-rays emitted from major elements of Si, Al, Fe, and Ti. In addition, the source-to-NaI(T1) detector distance of the probe explicitly determines spacial resolution of the borehole; reducing source-to-detector distance yields better delineation of coal piles and interseam sedimentary' strata. Rather than coal location exploration, the coal quality becomes important for economical reasons. ISPGAA techniques developed for coal ash delineation have been used successfully in Australia. The commercially available SIROLOG borehole logging as shown in Figure 11, was designed and developed in the CSIRO group, where two types of scintillators (NaI(T1) and BGO) were used to measure prompt y-rays from coal using a 'Cf neutron source. The parameter R, derived from the prompt y-ray spectrum, is defined as the ratio of the counts recorded in energy windows 2.6-5.2 to 2.0-2.4 MeV, representing the ash-composite elements (Si, Al) to hydrogen, respectively. This normalization is used to correct the neutron flux variations in different physical borehole conditions. Therefore, R can be directly related to ash content of coal from the chemical assay results. Howevx, the correlation between R and ash content varies at boreholes with different water contents, where neutron flux distribution depends on the therrnalizing effect of hydrogen in the ~ a t e r . ~ For ' . ~ ~dry and water-filled boreholes, corresponding calibration equations were derived using regression analysis, as listed

Ash(%) = - 18.96 + 74.499 R; for dry boreholes Ash(%) =

- 15.64 4- 79.2335 R;

for water-filled boreholes,

(1) (2)

Another method for correcting neutron flux perturbation in borehole measurements uses a boron-painted BGO detector to record boron capture y-counts at 478-keV peaks, which serve as a measurement of

Channel

Figure 9 Sample measured pulse height spectra for the wet sample. (a) The first time interval 0 to 100 FS, and (b) the second time interval 100 to 250 ps. (Reprinted with permission from Nucl. Instr. Meth. A219, Evans, L. G. et al., Inter-pulse high-resolution gamma-ray spectra using a 14 MeV pulse neutron generator, 239, Copyright 1984, Elsevier Science Publishers B. V.)

Figure 10 Geophysical profiles given by

o

200

400

600

800

counls/~cc

neutron capture for three source-to-detector distances: (a) 15 cm, (b) 23 cm, and (c) 33 cm. Coal beds are identified by the dark bands on the right side of the figure. (Reprinted with permission fr ~m Geophysics, 51, Charbucinski, J. et al., Prompt neutronaamma loaaina -- - for coal ash in water-filled goreholes, 1113, Copyright 1986, Society of Exploration Geophysicists.)

neutron flux intensity;49the ratio of the count rate in energy windows 2.6-5.2 to 0.38-0.60 MeV can also relate to ash content of coal. This method ;E successfully applied to coal seams with high variation in iron content, obtaining more accurate results than the y-y technique for coal ash determination. As mentioned in the previous section, scintillation detectors are prone to temperature change; gain stabilization at the 2.22-MeV capture prompt y-ray peak of hydrogen is required and usually performed automatically. This is particularly important for the BGO detector, due to its high dependence on temperature change.

B. MULTIELEMENTAL ANALYSIS USING HIGH-RESOLUTION DETECTORS The borehoie logging for determination of i d quality can be accomplished using high-reso1utio1.1 germanium detectors.5b55The composition of the coal can be derived by measuring the constituent

SIROMCP.

WINCH CONTROLLER

Sinplr board eomvu#r

HOST COMPUTER

.

c

Floppy disc Hard dlsc

Winch- MCA interface

SipnaL Anolog to digital convertor (ADC) board

1 pv SUP

Spactrat memory board

rings

El Key baard

Imprdonco matching Input/otr?put

I I / O ) board

P a r e r supply

E

Radiooetivo source //////I

Figure 11 Schematic of SlROLQG borehole logging system. (Reprinted with permission from Nucl. Geophys. 2, Charbucinski, F? L. et al., Quantitative nuclear borehole logging based on neutron excited gamma-reactions, 139, Copyright 7908, Pergamon Press, Ltd.) elements. The performance of this analysis is influenced by the neutron source employed in the probe; two types of neutron sources, the isotopic 2"Cf and 241Am-Re,and neutron generators, have been studied to establish their characteristics and feasibilities.

I . Analysis Using lsotopic Neutron Sources The in situ borehole logging of coal in terms of multielemental analysis was studied by Clayton et al. This analysis was conducted in using a 5-pg 252Cfneutron source coupled to an 11% HPGe dete~tor.~' a mock-up borehole in which five representative coal samples with ash content from 7 to 30% were measured by the ISPGAA technique. The relative concenlrations of 1I. major elements were determined in situ: Al, C, C1, Fe, H, K, N, Na, S, Si, 'and Ti. The bulk analysis of elemental concentrations has been compared to the laboratory chemical analysis, resulting in most major elements in coal being determined by this probe within 10% accuracy for a Zmin rapid field scan. However, this technique posed limitations in detection of carbon and oxygen due to their small thermal neutron capture crosssections, while the need for higher energy neutron sources to raise the reaction probability is recommended. In addition, high count rate electronics is necessary to obtain results of acceptable accurary in a short scanning time of a few minutes, a typical in sit11 borehole logging period. On the other hand, the 24'Am-Bcneutron source, providing a higher average neutron energy of 4 MeV, makes the determination of carbon and oxygen possible by means of inelastic scattering reactions of C(n,nly) and O(n,nfy) reactions wit!^ emission of 4.44- and 6.13-MeV prompt y-rays, respecti~ely.~'~' Figure 12 shows the spectra measured using 252Cfand 2J'Am-Be sources; it is explicit that the "'AmBe source is superior for measurements of carbon and oxygen alone. Using a "'Am-Be neutron source coupled to an NPGe detector, the measurement of sulphur content in coal was conducted by Underwood; the sulphur concentration can be determined down to 1% by weight, but the accuracy is intluenced by the chlorine content in the horehole e n v i r ~ n m e n t A ~ ~neutron ~~' transport code ANISN wns used to dencribe the influence of chlorine concentrations on the sulphur peak area of 2380 key

Energy

- keV

Energy

- keV

Figure 12 Prompt y-spectrum from (a) 252Cf, and (b) 241Am-Beneutrons interaction with coal. (Reprinted with permission from Int. J. Appl. Radiat. /sot., 34, Clayton, C. G. and Wormald, M. R., Coal analysis by nuclear methods, 5, Copyright 1983, Pergamon Press, Ltd.)

2. Analysis Using Neutron Generators Using a 14-MeV neutron generator probe coupled to a NaI(T1) detector or an HPGe detector has been investigated and proven capable of offering reliable results with respect to carbon and oxygen determination; this cannot be accomplished using low-energy isotopic neutron sources. The neutron generator can be operated in pulsed mode to fulfill an ultimate analysis of coal, in which both thermal capture and fast inelastic spectra are recorded by means of a time-segregated gating technique. The U.S. Geological Survey and U.S. Bureau of Mines have worked together since 1982 on a project to demonstrate the uses of a 14-MeV neutron generator coupled to a germanium detector for borehole logging.s6 For collecting more y-signals for analysis, timing window techniques were used in their in situ set-up. A practical method of obtaining an ultimate Ganalysisunder field conditions is to utilize the generator at pulse repetition rates of about 100 pulses/s and recording two spectra: the decay window for determination of oxygen and the prompt thermal neutron capture window for other elements.

V. ENVIRONMENTAL WATER BODY IN SITU SURVEY There is growing concern about the environmental poIlution problem, particularly in industrialized countries. The causes of water pollution can be the consequence of waste discharge from factories, mills, and refineries; immediare action should be taken before the toxic chemicals expand into the ecosystem. For this reason, a nuclear technique based upon ISPGAA was developed; it certainly offers potential for in situ, rapid, and continuous survey of such water pollutants. For in situ measurement in effluent body, water is the on!y major constituent, and serves as a neutron moderator as well as a neutron shield free from temperature variation. Thermalized neutrons are confined to a certain volume immediately surrounding the detecting probe where capture reactions are dominant and the prompt y-rays are used for the determination of pollutants. 'This is in contrast to the in situ borehole formation logging, where the determination of the absolute concentration of the essential element in various geological formations encountered some difficulties. For example, spatial and energy variations of the neutron fltlx in an underground matrix with a strong neutron absorber and dense neutron moderator distributed heterogeneously; adaptation of instruments to high-temperature conditions; and the equipment damage due to hostile environments and neutron radiation. The underwater probe used for in situ measurement of elemental concentrations and pollutants of environmental water bodies was first proposed and constructed by Chung and T~eng.~' There are two types of underwater ISPGAA probes developed with configurations referred to as backscatter and transmission geomettie~,5~~l as illustrated in Figure 13. In the hackscatter ISPGAA system, all equipment is enclosed in an aluminum probe and the 252Cfneutron source, together with the HPGe detector, are

Figure 13 Geometric arrangements for (A) backscatter and (B) transmission in situ PGAA assemblies. (a) HPGe detector, (b) %f neutron source, (c) neutronly-shield, and (d) aluminum probe. (Reprinted with permission from Nucl. Geophys. 7, Chao, J. H. and Chung, C., In situ elemental measurements in an environmental water body by prompt gamma-ray spectromatry, Copyright 1893, Pergamon Press, Ltd.)

Operating depth

d

positioned on opposite sides of a lead block. The lead, and the neutron shield surrounding the detector, rcduce the intcnsity of [he fission y-ray directly from the neutron source and the neutron flux intensity at the detector crystal, therefore suppressing the interfering y-rays in the collected spectra. As for the transmission system, the neutron source is submerged in water, whereas the HPGe detector is placed above the water-line; the water in between serves as both the sample being analyzed and y-raylneutron shield to remove neutrons as well as low-energy y-rays, but allows high-energy prompt y-rays to penetrate and reach the detector above the water surface. The basic concept for designing the ISPGAA probe is similar to that of logging sondes employed in surveys of geological formation. The submerged probe consists mainly of a W f neutron source, a portable HPGe detector, and neutron and y-shields in between, as illustrated in Figure 13. Signal and prompt y-rays collected by the HPGe detector were transmitted and analyzed in the battery-powered, portable CANBERRA-10 PLUS MCA with built-in amplifier, high-voltage power supply, and rechargeable batteries; all are placed above water-line.

A. PERFORMANCE TEST The performance of the submerged ISPGAA probe was evaluated in a 5.7-ton tank. The tank was filled with specific chlorine-concentrated water by adding sodium chloride, and the background measurements were performed in chlorine-free deionized water. Optimum operating depth, neutron flux distribution, HPGe detector eff~ciency,effective detecting range, and rninirnum detectable concentration (MDC) of pollutants were determined in this laboratory test.57*58 Numerous chlorine prompt y-rays with considerable intensity are appropriate for determining the efficiency of the ISPGAA probe. The detecting efficiency of the backscatter detection system is illustrated in Figure 14, as compared to a point source in front of the HPGe detector. In the transmission system, the detecting efficiency is relatively high for high-energy y-rays when using a strong neutron source operated at deeper position, where high-energy y-rays can penetrate the thick water layer and reach the HPGe detector. Figure 15 illustrates a chlorine capture y-spectrum collected in such a test.61 On the other hand, the measurement in demineralized water explains the possible origins of spectral background; the major background radiations for an underwater ISPGAA probe contah6'

L

POlnT SOURCE

-

1d2

>

-

U

C

-ill

2U U

W

16~-

-

n

VOLUME SOURCE

0

IU U!

I-

W

0

.

I ..... I 1

lo2

,.....I

GAMMA-RAV

1

a

I

.I-

10 3 ENERGY, K

la* V

Figure 14 Detector efficiency of the HPGe detector for both volume source in ISPGAA operation and point source with 6 cm sourceto-detector distance in normal application. (Reprinted with permission from Nucl. Instr. Meth. A267, Chung, C. and Tseng, T. C., In situ prompt gamma-ray activation analysis of water pollutants using a shallow 252Cf-HPGe probe, 227, Copyright 1988, Elsevier Science Publishers B. V.)

.

0

-

Double escape peak

Gamma-ray energy, M e V

Figure 15 Chlorine prompt capture ?-ray spectra collected in water containing 1920 ppm CI using a : s; dead transmission system with 44 pg 2sCf at an operating depth of 70 cm; counting time (1,)36,000 time (DT): 8.5%. (Reprinted with permission from Nucl. Geophys. 7, Chao, J. H. and Chung, C., In sifu elemental measurements in an environmental water body by prompt gamma-ray spectrometry, 471, Copyright 1993, Pergamon Press, Lfd.)

1. Neutron-induced y-rays from neutron interactions with water. The most prominent peaks at 2223 and 6129 keV result from the reactions 'H(n,y) and 160(n,n'y), or from I6N, produced through the '60(n,p)'6Nreaction with a half-life of 7.1 s. 2. Neutron-induced y-rays-from the HPGe detector. Neutrons that escape from the water can interact with the detector, resulting, for example, in y-rays of 596 and 691 keV from Ge(n,y) and Ge(n,nfy) reactions.62 3. Neutron-induced y-rays from the probe material. The major y-rays are 7724 keV from Al(n,y) and 7368 keV from the Pb(n,y) reactions. Following a long detection period, the decay y-rays of 1779 keV from 28A1are clearly observed in the low-energy part of the spectrum. For the transmission system, the optimum neutron operating depth can be determined according to the ratio of sensitivity to background ( S B ) or MDC of chlorine measured at the 6111-keV prompt yr a y ~ , 5 ~which , ~ ' served as index photopeaks for salinity determinati~n.~~ Figure 16 displays the sensitivity, spectral background, and the derived SIB as a function of neutron operating depth." The sampling range of the probe is a prime concern; it is dictated by the intensity and distribution of neutron flux in water around the probes. The neutron flux in water is measured by the indium foil activation technique, reflecting the probability of the emitted prompt y-rays in water around the probe. The actual number of prompt y-rays detected by the probe is a function of both neutron flux and water attenuation, which in turn vary with distance to the detector. The effective detecting volumes of the probes for the backscatter and transmission geometries were estimated to be 0.14 and 1.8 m3, respectively, using the accumulation of total counts contributed from water body at various distances from the HPGe detector. The sensitivities of some index elements of environmental concern are evaluated by comparison with chlorine, as below:

L"PCf source depth 1 cm

Figure 16 Variations of (A) detection sensitivity at 61 11 keV, (B) spectral background in 5 to 9 MeV range, and (C) SIB ratio with a neutron source at the operating depth for the transmission detection system with a 252Cf neutron source activity of: (a) 1.3, (b) 8.0, (c) 44, and (d) 260 ~ g(Reprinted . with permission from Nucl. Geophys. 7, Chao, J. H. and Chung, C., In situ elemental measurements in an environmental water body by prompt gammaray spectrometry, 473, Copyriy ht 1993, Pergamon Press, Ltd.)

where S(x) = the sensitivity of element x, cps/ppm S(C1)= the sensitivity of chlorine, cpstppm = detecting efficiency at prompt photopeak E I = prompt y-ray intensity A = atomic weight The MDC values of metallic elements in pollutants are evaluated in the same manner as that of a submerged probe for various-intensity neutron sources at the corresponding optimized source depth; some results are shown in Figure 17.58

B. FIELD SURVEY USING THE ISPGAA PROBE Unlike traditional techniques for in situ determination of salinity by measuring the properties of sampled salt water, the ISPGAA technique measures salinity in terms of chlorine content by detecing its prompt y-rays emitted from the Cl(n,y) reaction surrounding a neutron source. The salinity, defined as total dissolved salt in water in units of parts per th~usand(%o), can be simply derived by a linear relation as 1.80665 X Cl%o, with C1 representing chlorine content or ~ h l o r i n i t y In . ~ ~every form of water body, the chlorine content in water is proportional to the total dissolved mineral content, or salinity. For those elements other than chlorine dissolved in seawater (such as bromine, iodine, and potassium) with low neutron capture cross-sections, they yield negligible contribution compared to those with chlorine. Other elements contribute little to the prompt y-ray spectrum.

Figure 17 The MDC of elements as a function of 252Cfsource intensity at the corresponding optimum source depths determined in a 10-h measurement with a water surface to detector distance of 60 cm. Prompt y-rays (keV) of elements are also indicated. (Reprinted with permission from Nucl. Instr. Meth. A299, Chao, J. H. and Chung, C., Optimization of in situ prompt gamma-ray analysis of lake water using a HPGe-252Cfprobe, 654, Copyright 1990, Elsevier Science Publishers B. V.) The in situ field survey of river salinity has been performed in the Tamsui River in northern Taiwan. where a transmission probe equipped with a 44-pg ZS2Cf neutron source was used.55The measured salinity variation during tidal cycles and salinity distribution across the river were performed. A backscatter probe for lake pollutant surveys in fresh or low-salinity water bodies was also carried out in a campus lake.'O The chlorine concentrations in lake water were determined to range from 20 to 86 ppm.

C. RADIATION SAFETY CONCERNS A bare 252Cfneutron source is hazardous to humans due to its tremendous neutron dose rendered. Exposed to a 44-pg 252Cfsource in air at a distance 1 m away from the source, one suffers both neutron and y-doses at an equivalent rate of 1 mSv/h, which is far beyond the limit imposed to a radiation worker, 0.025 mSv/h. The 25?Cfneutron source is submerged in water during operation of the ISPGAA probe. In this condition, water acts as neutron moderator, thermal neutron absorbor, as well as a y-ray shield, reducing significantly neutron radiation and prompt y-rays near the water surface where radiation workers are stationed. On the other hnncl, the induced prompt y-rays of 2223 keV in H(n,y), and 61 !1, 7414, 8578 keV in CI(n,y) reactions become the main source of the y-dose. The distribution of dose rate on board, including neutron and y-ray contributions, is measured around 1 to 4 p S v h at the operating p o ~ i t i o n , ~ ~ w his i cwell h within the safety level recommended by the ICRP report." In addition to induced prompt y-rays, the thermalized neutrons in water inevitably produce activated nuclides, which emit decay P- and y-rays. !n seawater, the thermal neutrons are mainly absorbed by hydrogen, chlorine, and sodium; the residual activities induced by the 44-p,g 252Cfneutron source in seawater with salinity of 35% are calc~latedusing the

Table 3

Main Activated Products in Seawater lnduced by the 44-pg "*Cf Neutron Source

Nuclear reaction "Cl(n,y)38C1 "N~(II,~)'~N~ B'Br(n,y)82Br 35CI(n,y)MC1

Tin 37.3 min 15.0 h 35.3 h 3

X

IVyr

wtto

Abundance,

barns

%

0.428

24.23

0.53 2.7

100 49.3 1

43

75.77

Induced activity rate, Bq/s 35 8

4.3 x lo4 2.6 X lo-'

Reprinted with permission from Nucf. Geophys. 7, Chao, J. H. and Chung, C., In situ elemental measurements in an environmental water body by prompt gamma-ray spectrometry, 477, Copyright 1993, Pergamon Press, Ltd.

where

A = the saturated activity in Bq N = the number of nuclides per unit volume 4, = thermal neutron flux at distance (r) from source Table 3 lists major radioactive nuclides possibly generated in seawater as an ISPGAA system operates infinitely long.6' The production rate of each nuclide is estimated using Equation 4. Assuming that'the flowing rate of seawater is 1 m3/s and total activities induced convect uniformly within the effective sampling volume of 1.8 m3, the radioactive concentration would be less than 50 Bq/m3, which is much lower than the standard level of drinking water (37,000 Bq/m3), as well as the concentration of % in seawater (1 1,000 Bq/m3), as reported elsewhere.65

VI. PLANETARY IN S1TU EXPLORATION Chemical analysis of planetary surfaces is necessary for understanding the origin and evolutionary process of our solar system. Information on distribution of both major and trace elements on the planetary surface is essentizl for predicting the history of extraterrestrial activity as well as the existence of constituent materials of a planet. In 1964, NASA organized a team of investigators to develop the most reasonable techniques to measure the elemental compositions, hydr~gzncontent, and bulk density in the planetary surface.& Preliminary reports emphasized the use of an accelerator-type neutron source to implement detection of prompt and decay y-rays, as well as neutron "die away" measurements. Figure 18a envisions a simplified diagram of the apparatus for prompt y-ray detection on the lunar surface; the detection system can be arranged either parallel or perpendicular to the surface to be analyzed. A preliminary experiment of measuring inelastic prompt y-rays using a 2" X 2" NaI detector was performed on an artificial granite surface; y-ray spectra collected for horizontal and vertical probe orientations are shown in Figure 18b for comparison. Trombka et al. proposed the use of a 252Cfneutron source due to its small size, stable flux, and complete lack of power supply, developing a technique based on the measurement of capture prompt y-ray~.~'The detecting system is illustrated in Figure 19a, in which a Ge(Li) or NaI('l'1) detector was used to measure the high-energy capture prompt y-rays, capable of penetrating the soil and shield. The spectra collected in a mock-up granite surface are shown in Figure 19b. Similar to the borehole logging system, the in situ prompt y-ray analysis on a pIanetary surface requires a neutrcn source in connection with a y-ray detector on board a spaceship. Feasibility and limitations regarding such exploration using the ISPGAA technique have been estimated by Johnson and S e n f t l ~ , 6as ~ .concluded ~~ below. NaI(T1) detectors are preferable for space application due to their resistance to high-temperature conditions. In order to suppress spectral background due to activation on the detector crystal, especially three possible neutron sources were considered: (1) a the generation of 25-min IZBIand 15-h 24Na, neutron generator that can be turned off during space flight; (2) an isotopic (cx,n) source that is "unmixed" until it amves on the planetary surface; and (3) an isotopic source that can be kept physically separated from the detector during space flight.

ta)

i

r --1

I--f I I

I

GAMMA -RAY

CHANNEL NUMBER

Figure 18 (a) Geometrical arrangements of the neutron probe with respect to the lunar surface. (b) Gamma-ray spectrum from inelastic scattering for horizontal and vertical probe orientation. (Reprinted with permission from Science, 152, Caldwell. R. L. st at., Combination neutron experiment for remote analysis. 460461, Copyright 1966. American Association for the Advancement of Science.)

In response to the use of a semiconducting germanium detector for planetary exploration, the CO, solid-cryogen cooling system was designed by Nakano et al.'" The operating temperature of the gemanium detector is 130K, much higher than the liquid nitrogen temperature but somewhat lower than the criticaI temperature of 135K, above which the detector resolution begins to degrade appreciably." Senftle et al. used the Monte Carlo calculation technique to estimate the thermal neutron and induced y-ray fl uences at various distances from a "2Cf neutron source on typical lunar soil with water content up to 10% by weight.69The results inc!icated that the nonhydrogeneous condition encountered on the planetary surface restricts the use of captured and delayed y-ray analysis due to low concentrations of hydrogen or water. In that cme, inelastic prompt y-rays induced by fast neutron will dominate the yray spectra for analysis of the major and a few minor elements on the planetary surfh~e.'~

Figure 19 (a) 252Cfsource Ge(Li) detector experiment configuration. (b) Prompt capture and activation y-pulse-height spectrum. (Reprinted with permission from Nucl. Instr. Meth. 87, Trombra, J. I. et al., Neutron radiativecapture methods for surface elemental analysis, 37-38, Copyright 1970, Elsevier Science Publishers B. V.)

Neutron source

I

0

L

5

10

I L

15

20

I

I

I

25

30

35

1

4 0

45

crn

Figure 20 Sectional contour diagram showing sea bed contributions to ISPGAA probe countrate (from unit volume) a s percentages of the maximum value close to the surface. (Reprinted with permission from Int. J. Appl. Radiat. Isot. 34, Thornas, B. W. et at., Mineral exploration of the sea bed by towed sea bed spectrometers, 439, Copydght 1983, Pergamon Press Ltd.)

The seabed minerals are often formed as a result of a deposition process. Major interest centers on the potential of economic extraction of manganese oxide noduIes and strategic metal-bearing muds. The exploitation of these minerals is subject to knowledge of the locations and extent of seabed mineral deposits. Practical techniques for surveying mineral deposits and assaying value are underinvestigated. A variety of nuclear techniques have been developed and applied to mineral exploration. These include the measurement of natural y-radiation, a range of neutron interaction techniques, and energy dispersive X-ray fluorescence analysis. Among them, the natural y-ray towed seabed spectrometer is potentially feasible for the discovery and delineation of minerals containing radioactive isotopes. The towed spectrometer based on the ISPGAA technique can be used to identify the concentration and spatial distribution of many nonradioactive minerals currently of interest in marine exploration. An ISPGAA-based spectrometer consisting of a 'Cf neutron source and HPGe detector has been designed by Tharnos et al.," as shown schematically in Figure 20. The performance of this probe was demonstrated in a test tank filled with seawater and simulated seabed, consisting of silica sand seeded with known elements of interest. The capture prompt y-rays induced by a 100-pg W f source was collected by an HPGe detector. The performance of the probe, including the MDC for elements of economic interest, effective sampling depth below seabed, factors affecting distribution of neutron fluence in samples, and the influence of interfering elements on detection limits are reported in detail. Some pertinent results are described For a number of elements of economic interest, MDC values were calcuIated using the most intense capture y-rays indrmci by a 100-pg '*Cf neutron source for a counting period of 15 min, as listed in Table 4. The detector response to capture y-rays as a function of seabed depth was inferred using the contours of the percentage contribution to the total count rate, showing an effective sampling depth to 15 cm at the surface of h e seabed, and subject to the variation of seabed environment. Chlorine is the major interfering element in limiting detection sensitivity in neutron-induced y-ray spectrometly of the sealaed. This effect can he improved using a noncylindrical probe.

Table 4 Some Elements of Interest and Their Characteristics

Atomic mass Element Nickel Copper Zinc Molybdenum Silver Tin Gold Mercury Lead Chromium Cobalt Selenium Niobium Antimony Barium Tantalum Tungsten Platinum Vanadium Manganese Iron Zirconium Magnesium Titanium Cadmium

A 58.71 63.54 65.37

Thermal neutron absorption cross-section a.barns

4.6 3.9 1.1

intensity 1 (%), photons1 100 neutrons 42 28 12

Lowest limit of detection in Si02 matrix (counting time 15 min Sensitivity source 100 pg 262Cf), lulA wt% 3.2 1.7 0.2

0.2 0.3 2.0.

Reprinted with permission from Int. J. Appl. Radiat. Isot. 34, Thomas, B. W. et al., Mineral exploralion of the sea bed by towed sea bed spectrometers, 446, Copyright 1983, Pergamon Press, Ltd.

The Ocean Drilling Program (ODP) using nuclear logging techniques for ocean floor in situ s u r v q The ~najjor was implemented to determine the parameters characterizing the ocean floor geo~hemicalIy?~ advantages of nuclear measurements over traditional geophysical logs is the ability to deploy the technique for measuring the formation properties through drill pipe, providing a continuous record for analysis.

ISPGAA techniques have expanded to almost all aspects of bulk analysis for the last three decades. With the advent of innovative nuclear instrumentation and technology dedicated to these fields, sophisticated detection and detail evaluations became possible andlor practical, both from technological and economicai points of view. Large efforts still concentrate on the use of the ISPGAA technique in petroleum industries, keeping pace with the increased need in oil production in an economical way. The ISPGAA method performs elemental analysis for oil rock formation, permitting an understanding of mineral constituents, which is complementary to other techniques for identification of rocks with potential productivity, not just for the purpose of exploration. The economic and environmental importance of in situ borehole analysis of constituent elements in a coal mine before production result in the enhanced development in ISPGAA techniql-2s. Ash content evaIuation is always the prime concern for coal quality determination. Besides, due to stringent environmental protection limitations, sulphur content determination has become a necessity recently. An ISPGAA technique using an isotopic neutron source for surveying water polIutants and salinity has the advantages of large sampling volume and less radiation safety concerns; it is practical in

determining toxic elements in effluents from factories or a heavily polluted water body. Rapid survey and timely response to the accidental release can possibly be taken. Applications of ISPGAA in outerspace exploration and seabed mineral investigation show its unique &aracteristics for remote analysis; in such hostile environments, restrictions in detection instruments and samples to be analyzed should be considered.

REFERENCES 1. Stelson, P. H., Dickens, J. K., Raman, S., and Trammell, R. C., Deterioration of large Ge(Li) diodes caused by fast neutrons. Nucl. Irtstr. Meth. 98. 481. 1972.

2. Kraner, H. W., Pehl, R. H., and Maller, E. E., Fast neutron radiation damage of high-purity germanium detectors, IEEE Nucl. Sci. 22, 149, 1975. 3. Lee, C. J. and Chung, C., Performances of gamma ray measurement using bismuth germanate detector after thermal neutron bombardment, Appl. Radiat. Isor. 42. 729, 1991. 4. Shope, L. A., Berg, R. S., O'Meal, M. L., and Barnaby, R. E., Operation and life of the Zetatron: A small neutron generator for borehole logging, IEEE Nucl. Sci. 28, 1696, 1981. 5. Grau, J. A. and Scllweitzer, J. S., Prompt y-ray spectral analysis of well data obtained with NaI(T1) and 14 MeV neutrons, Nucl. Geophys. 1, 157, 1987. 6. Vartsky, D., Wielopolski, L., Ellis, K. J., and Cohn, S . M., High count rate proble&s in elemental analysis using pulsed neutron inelastic scattering, Nucl. Instx Meth. 206, 575, 1983. 7. Senftle, F. E,Macy, R. J., and Mikesell, J. L., Determination of the optimum-size californium-252 neutron source for borehole capture gammamy analysis, Nucl. Instr: Meth. 158, 293, 1979. 8. Rozsa, C., Dayton, R., and Raby, P., Characteristics of scintillators for well logging to 225 "C, IEEE Nucl. Sci. 37, 966, 1990. 9. Melcher, C. L., Manente, R. A., and Schweitzer, J. S., Applicability of barium fluoride and cadmium tungstate scintillators for well logging, IEEE N~rcl.Sci. 35, 1188, 1989. 10. Melcher, C. L. and Schweitzer, .I. S., Gamma-ray detector properties for hostile environments, IEEE Nucl. Sci. 35. 876. 1988. 1 1. Melcher, C. L, Schwdtzer, J. S., Manente, R. A., and Peterson, C. A., Application of GSO scintillators for well logging, IEEE NucL Sci. 38, 506, 199 1. 12. Traeger, R K. and Lysne, P. C, High temperature electronics application in well logging, IEEE Nucl. .Sci. 35, 852, 1988. 13. Melcher, C. L., Schweitzer, 3. S., Libeman, A., and Sirnonetti, J., Temperature dependence of fluoresence decay time and emission spectrum of bismuth germanate. IEEE Nucl. Sci. 32, 529, 1985. 14. Tanner, A. B., Moxharn, R. M., and Senftle, F. E., A probe for neutron activation analysis in a drill hole using 252Cfand Ge(Li) detector cooled by a melting cryogen, Nml. Instr: Meth. 100, 1 , 1972. 15. Boynton, G. R,Canister cryogenic system for cooling germanium semiconductordetectors in borehole and marine probes, Nucl. Instr. Mefh. 123, 599, 1975. 16. Mellor, D. W., Cryogenics for semiconductor detector neutron activation logging tools, Int. J. Appl. Rndiat. Dot. 36, 295, 1985. 17. Tittle, C. W., A history of nuclear well logging in the oil industry, Nucl. Geophys. 3, 75, 1989. 18. Schultz, W. E. and Smith, IB. I).,Laboratory and field evaluation of a carbonloxygen (C/O)well logging system, J. Pet. Tech. Oct., 1103, 1974. 19. Lock, G. A. and Iloyer, W. A., Carbon-oxygen (C/O) log: Use and interpretation. J. Per. Tech. Sept., 1044, 1974. 20. Culver, R B., Hopkinson, E. C., and Youmans, A. M., Carbontoxygen (C/O) logging instrumentation. Soc. Petrol. Engineers J. Oct., 463, 1974. 21. Kerr, S. A. and Worthington, P. F., Nuclear logging techniques for hydrocarbon, mineral and geological applications, IEEE N~rcl.Sci. 35, 794, 1988. 22. Schweitzer, J. S. and E l k , n.V., Review of nuclear techniques in subsurface geology. lEEE Nucl. Sci. 35. 800, 1988. 23. Sanders, L. G., The application of Monte Carlo comp~rtationsto formation analysis by neutron interactions, Int. J. Appl. Radiat. Isot. 34, 173, 1983. 24. Hertzog, R. C., Neutron-excited gamma-ray spectrometry for well logging, IEEE Nucl. Sci. 26, 1558, 1979. 25. Dewan, J. T., Stone, 0. L., and Morris, R. L., Chlorine logging in cased holes, J. Pet. Tech. June, 53 1. 1961. 26. Mckinlny, P. F. and Tanner, H. I,., The shaie-compensated chlorine log. J. Pet. Tech. Feb., 164, 1975. 27. Herrnn, M. M., Future i~pp!icalioriso f elcmcnti~lconcentr;~tionsI'rom geophysicnl logging, N ~ r c l .Grol~ilys. 3. 197, 1987. 28. HerZzog, R. C., ElemenraI conctmrations from neutron induced gamma ray spectroscopy, IEEE Nttcl. Sci. 35, 827. 1988.

29. Scott, H. D., New developments in remote elemental analysis of rock formations, J. Pet. Tech. July, 7 11, 1986. 30. Schweitzer, J. S., Developments in elemental concentration logging and applications, IEEE Nucl. Sci. 38. 497, 1991. 31. Karus, E. V. and Shimelevich, Yu. S., Nuclear geophysics in prospecting, exploration and development of oil and gas fields, Int. J. Appl. Radiat. Isot. 34, 95, 1983. 32. Senftle, F. E. and Mikesell, J. L., The nuclear ratio technique applied to borehole exploration for industrial metals and coal, Nucl. Geophys. 1, 227, 1987. 33. Lederer, C. M. and Shirley, V. S., in Table of Isotopes, John Wiley & Sons, New York, 1978. 34. Lone, M. A., Leavitt, R. A., and Harrison, D. A., Prompt gamma rays from thermal-neutron capture, Atomic Data and Nuclear Data Tables, 26, 5 I 1-559, I98 1. 35. Mills, W. R., Stmrnswold, D. C., and Allen, L. S., Advances in nuclear oil well logging, Nucl. Geophys. 5, 209, 1991. 36. Schweitzer, J. S., Nuclear techniques in the oil industry, Nucl. Geophys. 5, 65, 1991. 37. Schweitzer, J. S., Ellis, D. V., Grau, J. A., and Hertzog, R. C, Elemental concentrations from gamma-ray spectroscopy logs, Nucl. Geophys. 2, 175, 1988. 38. Grau, J. A. and Schweitzer, J. S., Elemental concentrations from thermal neutron capture gamma-ray spectra in geological formations. Nucl. Geophys. 3, 1. 1989. 39. Schweitzer, J. S., Hertzog, R. C., and Soran, P. D., Nuclear data for geophysical spectroscopic logging, Nucl. Geophys. 1,213, 1987. 40. Evans, L. G., Lapides, J. R., Trombka, J. I., and Jensen, D. H., In situ elemental analysis using neutroncapture gamma-ray spectroscopy, Nucl. Instr. Meth. 193, 353, 1982. 41. Evans, L. G., Trombka, J. I., Jensen, D. H., Stephenson, W. A., Hoover, R. A., Mikesell, J. L., Tanner, A. B., and Senftle, F. E., Inter-pulse high-resolution gamma-ray spectra using a 14 MeV pulsed neutron generator, Nucl. Instr: Meth. A219, 233, 1984. 42. EI-Kady, A. A., Abulfaraj, W. H., and Abulfattah, A. F., Evaluation of a prompt gamma ray sonde consisting of an isotopic neutron source and a high purity germanium detector, Nucl. Instr. Meth. A28 I , 236, 1989. 43. Charbucinski, J., Borsam, M., Eisler, P. L., and Youl, S. F., In situ borehole determination of ash content of coal using gamma-gamma and neutron-gamma techniques, in Proc. Advisory Group Meeting on Gummu, X-Ray and Neutron Techniquesfor the Coal Industry, IAEA Vienna, 1986, 60. 44. James, W. D., Activation analysis of coal and coal effluents, in Activation Analysis, LI, CRC Press, Boca Raton, FL, 1990, 364. 45. Charbucinski, J, Youl, S. F., Eisler, P. L., and Borsaru, M., Prompt neutron-gamma logging for coal ash in water-filled boreholes, Geophysics, 51, 11 10, 1986. 46. Eisler, P., Borsaru, M., Youl S. F, and Charbucinski J., Quantitative spectrometric borehole logging techniques for coal mining, Nucl. Geophyr. 2, 43, 1988. 47. Charbucinski, J., Eiler, P. L., and Borsasu, M., Quantitative nuclear borehole logging based on neutron excited gamma-reactions, Nucl. Geophys. 2, 137, 1988. 48. Borsaru, M., Charbucinski, J., Huppert, P., Youl, S., and Eisler, P., Coal ash determination in dry boreholes by the neutron capture technique, NucL Geoplrys. 2, 201, 1988. 49. Borsaru, M., Charbucinski, J., Eisler, P., and Ceravolo, C., Neutron gamma logging in coal seams of variable iron content, Nucl. Geophys. 5, I 17, 1991. 50. Zhao, J. Q., Liu, J. S., Ni, H. F., Liu, D., Wuru, G. S , Guo, T. C., Wu, S. Y, and Liu, J. L., Neutron capture gamma-ray spectral well logging ustng a %f-HPGe probe, Nucl. Geophys. 5, 123, 1991. 51. Clayton, C. G., Hassan, A. M., and Wormald, M. R, Multi-element analysis of coal during borehole logging by measurement of prompt y-rays from lhcrmaf neutron capturc, Inf. J. App1. Rudiar. Isof. 34.83, 1983. 52. Wormald, M. R. and Clayton, C. G., In situ analysis of coal by measurement of neutron-induced prompt y-rays. Int. J. Appl. Radiat. Isot. 34, 71, 1983. 53. Clayton, C. G. and Wormald, M. R., Coal analysis by nuclear methods, Int. J. Appl. Radiat. Isot. 34.3, 1983. 54. Underwood, M. C. and Petler, J. S., High resolution prompt (n,y) spectroscopy using a down hole logging tool, J. Radioanal. Nucl. Chem. Articles, 114, 379, 1987. 55. Underwood, M. C. and Petler, J. S., Bulk material analysis using energetic neutrons, Nucl. Geophys. 3, 235, 1987. 56. Mikesell, J. L., Senftle, F. E., and Tanner, A. B., Borehole elemental analysis of coal using fast neutrons, in Proc. Advisory Group Meeting on Gamma, X-Ray and Neutron Techniques for the Coal Industry, IAEA Vienna, 1986, 25. 57. Chung, C. and 'Ikeng, T. C., In situ prompt gamma-ray activation analysis of water pollutants using a shallow =Tf-HPGe probe, Nucl. Inst,: Meth. A267, 223, 1988. 58. Chao, J. H. and Chung, C., Optimization of in situ prompt gamma-ray analysis of lake water using a HPGeI5Tf probe, Nucl. Instr. Meth. A299, 651, 1990.

59. Chao, J. H. and Chung, C., In situ prompt gamma-ray measurement of river water salinity in Northern Taiwan using HPGe-U2Cfprobe, Appl. Radiat. Isot. A42, 723, 1991. 60. Chao, J. H. and Chong, C., In sit![ lake pollutant survey using prompt-gamma probe, Appl. Radiat. Isot. A42.735, 1991. 61. Chao, J. H. and Chung, C., In siru elemental measurements in an environmental water body by prompt gamma-ray spectrometry, Nucl. Geophys. 7, 469, 1993. 62. Chao, J. K., Neutron-induced gamma rays in germanium detectors, Appl. Radiar. Isor. 44, 605, 1993. 63. Wooster, W. S., Lee, A. J., and Dietrich, G., Redefinition of salinity, Deep Sea Res. 16, 321, 1969. 64. International Commission on Radiological Protection (ICRP) Report 21, Pergamon Press, Oxford, 1973. 59. 65. Eisenbud, M., Environmental Radioactivity, 3rd ed., Academic Press, Orlando, FL, 1987, 149. 66. CaIdwell, R. L., Mills, FY. R., Allen, L. S., Bell, P. R., and Heath, R. L., Combination neutron experiment for remote analysis, Science, 152, 457, 1966. 67. Trombka, J. I., Senft!e, E, and Schmadebeck, R., Neutron radiative capture methods for surface elemental analysis, Nucl. Instrr Meth. 87,37, 1970. 68. John, R. G., Evans, L. G., and Tromhka, 3. I., Neutron-gamma techniques for planetary exploration. IEEE Nucl. Sci. 26, 1574, 1979. 69. Senftle, F., Philbin, P., Moxham, R., Boynton, G., and Tkombka, J., Problems &countered in the use of neutron methods for elemental analysis on planetary surfaces, Nucl. in st^ Meth. 117, 435, 1974. 70. Nakaho, G. H., Imhof, W. I,., and Johnson, R. G., A satellite-borne high-resolution Ge(Li) gnmma-ray spectrometer system, IEEE Nrrcl. Sci. 2 1, 159 1974. 71. Nakaho, G. H. and Imhof, W. L., Performance of large germanium detectors at elevated temperatures for satellite applications, IEEE NucL Sci. 18, 258, 1971. 72. Englcrt, P., Bruckner, J., and Wanke, H., Planetary gamma-ray spectroscopy, a special form of prompt charged particle and prompt neutron activation analysis, J. Radioanal. Nucl. Chem. Articles. 112, 11, 1987. 73. Thomas, B. W., Clayton, C. G., Ranasinghe, V. V. C., and Blair, I. M., Mineral exploration of the sea bed by towed sea bed spectrometers. Int. J. Appl. Radial. Isot. 34. 437, 1983. 74. Lovell, M. A., Harvey, P. R., and Anderson, R. N., The application of nuclear logs in the ocean drilling program. IEEE Nucl. Sci. 37. 1386, 1990.

Chapter 9

On-Line Applications Jiunn-Using Chao CONTENTS I. Introduction ............................................................................................................................... 159 11. On-line Analysis of Coal .......................................................................................................... 159 A. Importance of On-line Analysis of Coal ......................................................................... 159 B. OLPGAA Coal Analysis .................................................................................................... 160 C. Development of OLPGAA Techniques .............................................................................. 160 D. Commercial OLPGAA Systems ......................................................................................... 164 111. Mineral On-line Process Control ........................................................................................ 164 A. Metal Industry ..................................................................................................................... 166 B. Geothermal Power Utilities ................................................................................................ 170 IV. Explosive On-line Interrogation ............................................................................................... 170 A. Detection of Explosives by OLPGAA .......................................................................... 170 B. Current OLPGAA System for Explosive Detection ..................................................... 172 C. OLPGAA Using Fast Neutrons .......................................................................................... 174 D. Chemical Warfare and Ammunition Identification ......................................................... 175 V. Conclusions ............................................................................................................................. 175 References ...........................................................................................................................................175

Applications of PGAA techniques to various industries for on-line bulk analysis of coal, mineral ore, nitrogen-contained explosives, as well as, geothermal water have proven to be an attractive alternative to conventional methods. It is not surprising that the penetrating neutron, passing through the matter being analyzed and inducing prompt y-rays through either thermal neutron capture or fast neutron composition from deep parts of the inelastic scattering processes, can carry information of eleme~~tal sample to be measured. As a result, hundreds of batch samples can be analyzed directly on the conveyor in a short period of time without tedious preparation prior to measurement. Recent investigations using on-!ine prompt y-ray activation analysis (OLPGAA) methods have concentrated on the coal industry for economic reasons. Potential applications to the fields of process control in the mineral industry, explosive interrogation for airport security, and water analysis in geothermal power plants were also studied and proven to be promising. For the ISPGAA measurement introduced in the previous chapter, the c!etection system moves from place to place to collect information from the geological samples with fixed positions; whereas, the OLPGAA analyzes the samples moving through a fixed detection system, to which the counting conditions, such as temperature, pressure, humidity, and shielding design, can be adjusted or optimized to meet the requirements of specified measurement. More complex and sophisticated instruments can be incorporated into the system to upgrade the detection performance, rather than a small, compact detecting probe in the cases of geological ISPGAA measurement.

!I. ON-LINE AMALYSPS OF COAL A. IMPORTANCE OF ON-LINE ANALYSIS OF COAL Thus far, coal is still one of the most important sources of energy; extensive reserves make it continue to control the economic and industria1 developments in the world. At the same time, growing demand for rapid on-line analysis of coal quality and compositions, which determine the effectiveness of combustion in a coal-fire(! power plant or process contro! in a coal tvashery, results in the development of various on-line analyzers. The most important parameters influencing coal quality are combustion heat, moisture, ash, and sulphur contents. For instance, for the purpose of electricity production in a O.ti493-5 149-9/95/$0.00+$.~0 8 1995 by CRC Press, Jnc.

power plant, the combustion heat of coal is of prime concern. Moisture and ash content of coal can also affect the combustion process and should be known prior to coal combustion. In addition, an understanding of ash concentrations and the ash composite elements can prevent heat exchangers in steam generators from slagging and fouling. Today, the strict regulatory standards for air pollution control call for the determination of sulphur content before coal combustion; more efforts have focused on this measurement as a requisite in coal analysis.

B. OLPGAA COAL ANALYSIS In response to the demand for on-line rapid determination of coal quality in an electricity power station, a variety of nuclear methods including prompt y-techniques were developed. The main function of these techniques is to determine ash content and then to estimate the carbon content, or combustion heat, of coal by subtracting the ash content frcrn total amount; this is usually accomplished by dual-y transmission, Compton backscatter, and y-pair-production techniques. For the measurement of coal moisture, microwave attenuation and fast neutronly techniques are popular. During the past 15 years, OLPGAA techniques with some unequaled advantages over other current on-line techniques for coal analysis were rapidly developed to meet the requirements of the coal industry, and to act as a supplement to other on-line analyzers. The newly developed OLPGAA systems are based on the requirements of economic and environmental aspects; up to 14 elements can be simutaneously detected to infer the parameters related to coal quality. Basic principles and performances of the OLPGAA technique are stated below:

1. The OLPGAA measurement can directly relate combustion heat to the caxllon content in coal, or in a more detailed manner, incorporate the composite elements H, 0, N, and S concentration for evaluation.' Some studies pointed out that the combustion heat can be calculated as the concentrations of C , H, 0 , N, and S are known, or for simplicity, by use of amount of C as an index element for evaluation of combustion heat. The relationship between combustion heat and C content of coal was obtained in an experimental measurement,' yielding the value of 0.443 M J k g %C. But the carbon-dependent value is subject to various types of coals with different ash and moisture contents. Corrections between heat combustion of coal with its ash and moisture contents were reported for coals from various mines. 2. The elemental content of coal ash can be determined by knowing some indicative elements (Al, Ca, Fe, and Si), which are readily measured by OLPGAA methods due to their high cross-sections for capture reactions, reflecting the quantity of corresponding ash constituents of A1203,CaO, Fe203,and SO2. Besides, minor minerals in coal, such as S, C1, Ti, and Na, can be measured simultaneously from the observed characteristic prompt y-ray spectrum. 3. OLPGAA provides more penetrating radiations than that of the traditional X-ray fluorescence method for determining sulphur content. Sulphur content in massive coal samples can be determined by measuring the prompt photon of 2380 keV in the S(n,y) reaction. 4. Similar to all other nuclear techniques, OLPGAA does not tell apart different chemical bonds of the elements, but allows only the determination of elemental concentration. Consequently, the microwave attenuation technique is still a requisite for determination of moisture in coal. However, OLPGAA techniques show some limited success in determining water content in coal produced from a known geologica! formation in which the ratio of carbon to organicbonded hydrogen is a constant.2J For instance, raw lignite in Germany has a mean C/H ratio of 12.73 for five samples with various water content^.^ Basically, the parameters mentioned above are derived from the knowledge of the elemental composition of coal. The characteristic prompt y-rays from the elements of interest in coal analysis are summarized in Table 1. C. DEVELOPMENT OF OLPGAA TECHNIQUES The OLPGAA technique for coal analysis has been continuously modified over the past 15 years. In the early stage, investigators performed 1aboram-y measurements on 50- to 150-kg coal samples to ensure the feasibility of this technique for coal a n a l y s i ~ .This ~ . ~ includes the identification of proper nuclear reactions and prompt y-lines for coal composite element analysis; the optimization of irradiation and detection geometry; the determination of effective coal thickness in the conveyor belt for various

Table 1 Prompt y-Rays from Elements Influencing Coal Quality Prompt y-ray energy, keV

Parameters of .-

Elements

coal quality

Neutron reaction

Combustion heat Ash content

Moisture

Sulphur content Note: Data. taken from References 1 and 2.

intensities of neutron sources; the selection of suitable neutron sources as well as detectors; and the improvement of accuracy in a short counting period. A pioneer OLPGAA detection system for coal analysis for laboratory testing, assembled by Sowerby? is illustrated in Figure 1. The fast neutrons emitted from the nonnmderated 238P~-Re neutron source irradiate the coal sample, resulting in the emission of 4445-keV y-rays for the determination of carbon, an element indicative of specific energy. The moisture content of coal was derived by measuring the

lScm

0x

IOcm

T h ~ c kN a l ( T t ) Det~tor

0 oron Trioxtde Shl el d

Tunqsttn /Lead Sh~cld

a

d m Scolr

(

mnt

Figure 1 Cross-sectional view of thc- backscatter gauge used for the determination of inelastic scatter

and capture y-ray count rates from bulk samples. (Reprinted with permission from Nucl. Instr: Meth. 160, Sowerby, 8. D.,Measurement of specific energy, ash and moisture in bulk coal samples by a combined neutron and gamma-ray method, 175, Copyright 1979, Elsevier Science Publishers B. V.)

2223-keV prompt y-ray, generated through the neutron capture reaction with hydrogen in coal, and also by the measured CIH ratio. The NaI(T1) detector was used to measure these prompt y-rays and a typical spectrum, measured by Sowerby: is shown in Figure 2. Another laboratory expcrimcnt using a W f isotopic neutron source and germanium detector to perform a multielemental analysis was also r e p ~ r t e d This . ~ system was optimized by compromising the sensitivity and background at various coal thicknesses. Up to 12 elements can be detected simultaneously in a 20-min irradiation and counting period. Most of the neutron sources used for OLPGAA are isotopic neutron sources; for example, 252Cfand 238Pu-Be.They are frequently used in OLPGAA applications due to their properties of high neutron yield, long-lasting output, and small size. However, using a D-T neutron generator in OLPGAA for determining carbon and oxygen contents was also in~estigated;~ their cross-sections for the (n,nly) reaction, producing inelastic prompt y-rays, are higher than those using isotopic neutron sources. The D-T neutron generator, unfortunately, has to be replaced after some hundreds of hours service due to the depletion of D/T ion source. Pair and Compton-suppressed y-ray spectrometric techniques were used in OLPGAA to improve spectral quality without loss of detection efficiency. A pair spectrometer is preferably used to detect high-energy y-rays, which tend to induce the pair-production effect in the primary detector; the subsequent emission of two annihilation photons (51 1 keV) in opposite directions are then dctected by the guard NaI(Tl) detectors surrounding the primary detector; this coincidence spectral technique improves the signal-to-noise ratio with reasonable detection efficiency. The Compton-suppressed technique, on the other hand, operates in anticoincidence mode to reject the Compton-scatrered photons escaping from the primary detector, thus reducing the spectral background intensity. Figure 3 shows the spectra collected in pair as well as Compton-suppressed mode when analyzing coal samples using 19 NaI(T1) detector Higher accuracy is achieved using PGAA than existing laboratory methods. The complexity of the hardware required is not expected to create major difficulties in an industrial environment based on present operating experience, which has been largely maintenance-free. During recent years, the German corporation Staatliches Materiaiprufungsamt has carried out, at the request of industrial customers, a number of feasibility studies based on PGAA on-line analyzers using

I

I 2 00

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I LOO

I

I

600

1

800

1001

CHANNEL NUMBER

Figure 2 Prompt y-ray spectrum with a ccal sample containing 76.4 wt.% carbon. (Reprinted with permission from Nucl. Instr. Meth. 160, Sowerby, B. D., Measurement of specific energy, ash and moisture in bulk coal samples by a combined neutron and gamma-ray method, 177, Copyright 1979, Elsevier Science Publishers 6.V.)

Gamma

energy ( M e V )

Figure 3 (a) Pair spectrum of a limestone sample as used in the manufacture of cement; (b) Compton-

suppressed spectrum for a coal from Easington colliery (Reprinted with permission from Nucl. Geophys. 3, Warmald, M. R., A bulk materials analyzer using pair and Compton-suppressed gamma-ray spectrome.. try, 465, Copyright 1989,Pcrgamon Press Ltd.)

Table 2

Summary of Present Commercial OLPGAA Systems in the Coal Industry

Manufacturer

Instrument

MDH-Motherwell Inc.

ELAN

Science Applications International Corporation (SAIC) Gamma-metrics

Nucoalyzer CONAC Gamma-metrics

Mineral Control Company

Coalscan

WCI)

Performances

Ash, moisture. sulphur, calorific value, elemental analysis Ash, moisture, sulphur, calorific value, elemental analysis Ash, moisture, sulphur, calorific value, elemental analysis Ash, moisture, sulphur, calorific value, elemental analvsis

Note: Data taken from References 8 and 9.

252Cfirradiator~.~ The main purpose of the experiments was the investigation of parameters influencing the accuracy and reproducibility of the analytical results. The work has concentrated on the analysis of hard coal, lignite, and raw glass mixtures.

D. COMMERCIAL OLPGAA SYSTEMS At the present time, there are 26 commercial OLPGAA analyzers installed or on order in the coal industry in the U.S. and Canada. The major suppliers,&'2 as listed in Table 2, are Gamma-metrics, MDH, SAIC, and MCI. Most of the gauges are based on the utilization of a z2Cf source and NaI(T1) detectors in a transmission geometry and analyze the coal as it passes down a vertical chute. The elemental analyzer (ELAN), designed and supplied by MDH-Motherwell Xnc.,'O has been installed in a coal-cleaning plam at Homer City, PA, since 1983, as shown in Figure 4a. Its configuration and system performances are briefly described below. As shown schematically in Figure 4b, two =%f neutron sources, each with 100 pg, are placed on the right-hand side of the coal chute opposite the 15 cm D X 18 cm H NaI(T1) detector.I0 As mentioned in the previous chapter, although the NaI(T1) detector suffers from poor resolution for y-ray detection, it is not so critical for coal analysis due to prominent lines produced from the composite elements of Fe, Si, and Al. Apart from this, the ratio of e s q x peak to full-energy peak of NaI(T1) is small relative to that of a germanium detector, making the y-lines clear for analysis. The NaI(T1) detector is teinperaturestabilized at 54 "C to obtain a faster output than that at room temperature, and the coupled PMT is capable of tolerating continuous counting rates of 400,000 cps. Borated polyethylene and a large-volume coal sample comprise the safety shield, reducing the radiation level at 1 nl from the instrument surface to below 7 p,Sv/h for personnel safety and licensing requirements; in addition, it also prevents interferences from capture y-rays induced by structure materials reaching the detector. In the signal processing stnge, thc nonlinear cffects duc to peak pile-up, energy nonlinearity, and density change should be minimized. Pile-up can be measured by subtracting the sum of two singlesource spectra from the spectrum that results from both sources together, resulting in a residual pileup spectrum. A theory based on measured pulse shapes and the hardware pile-up rejection mechanism was developed to predict the residual pile-up spectrum. The composite spectrum after pile-up stripping can be approximated as a linear superposition of the thermal capture element spectra, two inelastic scattering spectra, four background spectra, and lhree spectra reflecting neutron interactions in the NaI(T1) detector. Calibration of the ELAN includes deriving each of these determined spectra from measurements of various samples of known composition. Figure 5 is an envisioned composite spectrum after pile-up stripping for coal samples at the Homer City pIant. Nearly all major elements, including oxygen, which is inferred from other known elements, can be determined in order to infer the ash and heat contents of coal.

Ill. MINERAL ON-LINE PROCESS CONTROL The application of OLPGAA techniques to mineral processing plants is required to control the variations of constituent elements for economic reasons. Fast neutron OLPGAA is quite suitable for bulk analyses

(a) Installation of ELAN and (b) schematic view of ELAN for the on-line analysis of coal. (Reprinted with permission from Nucl. Geophys. 3, Marshall, J. H., Ill and Zumberge, J. F., On-line measurements of bulk coal using prompt gamma neutron activation analysis, 446-448, Copyright 1989,

Figure 4

Pergaman Press Ltd.)

-

80.478 (SHIELDING)

Figure 5 Typical coal spectrum after pile-up stripping, dual 65+g 252Cf sources. (Reprinted with permission from Nucl. Geophys. 3, Marshall, J. H., Ill and Zumberge, J. F., On-line measurements of bulk coal using prompt gamma neutron activation analysis, 456, Copyright 1989, Pergamon Press Lld.)

of metallic elements owing to their significant neutron inelastic cross-sections; on the other hand, first neutrons aie a more penetrating radiation to metallic minerah than other types of radiation. The detection sensitivities of characteristic prompt y-rays following an inelastic scattering (n,nfy) process are quite dependent upon the neutron bombarding en erg^.'^-'^ The incident neutron energy should be greater than thc lirst cxcilcd cncrgy lcvcl of un ubundunLshblc isotopc of thc clcrncnl. Figurc 6 r;hows lhc tipcclra of an iron sample bombarded with 1.5, 2.5, and 3.5-MeV neutrons, respectively. More high-energy prompt y-ray peaks can be seen in the spectrum with 3.5-MeV bombarding neutrons. However, highenergy neutrons also induced more background and interfering peaks; the optimum neutron energy for analyzing the iron sample was therefore determined to be 2.5 MeV.

A. METAL INDUSTRY In practical cases, isotopic neutron sources of 239Pu-Beor Z'OPo-Be,positioning in an annular geometric arrangement, as illustrated in Figure 7, are preferred to a neutron generator. The tungsten block positioned in between them serves as a y-shield for reducing y-radiation from the neutron source, and the Bz03 embracing the detector is used to minimize thermal neutron bombardment on the detector.'' Both highresolution germanium and NaI(T1) detectors used for detection of prompt y-rays were evaluated. Correction factors for various samples with different bulk densities were also estimated and normalized at y-energies of 0.5, 1.0, and 2.0 MeV, allowing measurement of ore samples with densities of 1.2 to 3.3 g/cm3, as shown in Figure 8. A pulsed height spectrum was obtained with a Pb3O4sample in Figure 9, indicating the possible elements existing in the ore and their relative abundances. The prompt y-rays induced through fast inelastic reactions for essentral elements in mineral industries and their normalized y-responses to HPGe and NaI(T1) detectors were determined to help identification of element^.'^ Five geometries used for the bulk analysis of lead sinter samples were investigated by Cunningham et a1.;I6 their properties are summarized in Table 3. Selecting the most suitable geometry depends on accuracy, counting time, convenience of installation, and radiation safety. The backscatter geometry

Channels

, encrgg

Fe(n,n'y)

2.5MeV neutrons 0

,

Channels energy

-t 10' c

C 0

L

Za

10'

~n

F3

10

0

U

1 200

600

1000

1400

1800 2200 2600 3060

3QMl

Channels ;energy

Figure 6 y-Ray spectra of an iron sample at three bombarding energies. (Reprinted with permission from J. Radioanal. Nucl. Chem. Articles, 46, Yates, S. W. et a!., Elemental analysis by gamma-ray detection following inelastic neutron scattering, 343, Copyright 1978, Elsevier Sequoias. A.)

Figure 7 Cross-sectional view of the annular geometry used to determine neutron inelastic scatter y-ray count rates from bulk samples. (Reprinted with permissionfrom Int. J. Appl. Radiat. Isot. 35, Cunningham. J. 8. et al., Bulk analysis of sulphur, lead, zinc and iron in lead sinter feed using neutron inelastic scatter y-rays, 638,Copyright 1984, Pergamon Press Ltd.)

Figure 8 Calculated density correction factors applied to the measured y-ray yields. (Reprinted with permission from Nucl. Instl: Meth. 166, Sowerby, B. D., Elemental analysis by neutron inelastic scatter gamma rays with a radioisotope neutron source, 572, Copyright 1979, Elssvier Science Publishers 8. V.)

100

200

2 BULK OENSllY I g cm']

1

3 00

LOO

500

600

3 )

700

CHANNEL NUMBER

Figure 9 Pulse height spectrum obtained with a Pb304sample. (Reprinted with permission from Nucl. I n s t ~Meth. 166, Sowerby, B. D., Elemental analysis by neutron inelastic scatter gamma rays with a radioisotope neutron source, 572, Copyright 1379, Elsevier Science Publishers B. V.)

Table 3 Details of the Five Geometries Used for the Bulk Anabsis of Lead Sinter Sam~les

Neutron source Source output (neutronds) NaI(T1) detector dimensions (cm) Sample thickness (cm) Total. count rate in S window (countds) Net count rate in S window (counts/s/wt% S) Counting statistical error (10 min) . (wt% S)

Backscatter (ignition layer)

Backscatter (maln layer)

24'Am-Be

U 8 PBe ~

lo7

7.7 X 106

2x

(GI5 X 20

410 X 10

$15 X 10

10 X 10 X 15

10

-20

-20

5

-40

725

2550

1400

1080

-2350

10.9

6.5

2.6

-8.5

0.27

0.33

0.73

0.33

Annular

Semi-annular

24'Am-Be

"' Am-Be

7.7 X 106

7.7

$10 X I0

5.3 0.29

.

X

lo6

Backscatter (brass box)

238P~-Be 2x

lo7

Reprinted with permission from Int. J. Appl. Radiat. Isot. 35, Cunningham, J. B. et aI.. Bulk analysis of sulphur, Iead, zinc and iron in lead sinter feed using neutron inelastic scatter y-nys,639, Copyright 1984, Pergamon Press Ltd. was chosen for plant testing. The S, Pb, Zn, and Fe in lead sinter feed can be determined to within about 0.3, 0.9, 0.2, and 0.3 wt %, respectively. The development and application of a thermal neutron OLPGAA technique to the aluminum production process was described by Liu et aLI7 The detection system includes a 100-pg z2Cf neutron source placed in the measurement piping and a germanium detector for detecting the capture prompt y-rays for determination of A120,, Fe02, CaO, and S i 0 2 in a pipeline, as illustrated in Figure 10. The system has 'Seen tested and was expected to reduce plant costs by over U.S.D 3,000,000 annually by saving energy consumption, rapid surveying of pulp quality, and reducing labor costs. The measurement error of A1203, Fe02, CaO, and SiOz is 0.3, 0.1, 0.4, and 0.4%, respectively.

Figure 10 Diagram of on-line measurement assembly for aluminum process control. (Reprinted with permission from J. Radioanal. Nwcl. Chem. Articles, 151, Liu, Y et al., Development and applications of an on-line thermal neutron prompt-gamma element analysis system, 86, Copyright 1991, Elsevier Sequoia

Although iron is an essential element for arra!ysis by thermal neutron capture because of its large cross-section and intense high-energy y-rays, the carbon and oxygen in steel have low concentrations and low capture cross-sections, making thermal neutron capture techniques unacceptable for determination of their concentrations in steel. Shirakawa evaluated the feasibility of fast neutron techniques for on-line determination of carbon and oxygen contents in steel by computer ~imulation.'~ The 14-MeV neutrons produced from a D--2 generator are used to excite carbon and oxygen with inelastic scattering crosssections of 0.35 and 0.43 barns, respectively. A high-efficiency BGO detector is preferred for the measurements of the 4.43-MeV y-rays emitted from carbon and 6.13 MeV from which both high energy photopeaks are well separated in the spectrum despite the poorer resolution of the BGO detector. Computed results based on Monte Carlo calculations showed that the accuracy of the measurement of carbon content was t O . l wt % in the range 0.4 to 2.5 wt % for 60-s irradiation with neutron source strengths of 0.5 to 1.0 X loSnls, using both transmission and the backscatter geometric arrangements. The accuracy of measuring oxygen was within f0.01 wt % in the range 0.1 to 0.2 wt % for a 60-s count following 60-s irradiation in both geometries.

B. GEOTHERMAL POWER UTlLlTlES The extraction of the heat from hot rock near the surface of the earth by circulating the coolant may generate electrical power. Instead of sending steam from the geothermal reservoir directly to the turbine, a primary heat exchanger to make secondary steam for the turbine is preferred; this arrangement isolates the mineralized solutions that will be at much higher temperatures and pressures than with current natural steam. Chemical problems with the water circuits (e.g., corrosion and solid precipitation) are foreseen and provide analytical information for planning the geothermal operations and for the plants. Nuclear spectroscopic methods were proposed to analyze the mineral composition in the circulated waters, which are difficult to access by other analytical sampling techniques due to their high temperature and pressure conditions. The OLPGAA technique based on the measurement of the characteristic prompt y-rays provides a solution to this problem. -4 laboratory assembly for this purpose was constructed to demonstrate its performance for on-line anaiysis of geothermal waters.I9 The on-line analyzing system neutron source and a 20-cm3 Ge(Li) detector was described, as illustrated in Figure using a 3.4-mg 252Cf 11, Both neutron source and detector are placed outside the pipes; the induced prompt y-rays can penetrate the wall and are measured for analysis. Spectra measured for distilled water and various spring waters are shown in Figure 12. Chlorine is responsible for corrosion problems and can be determined to very low concentrations. Valuable minerals to be extracted, such as copper, could be determined and the information fed to related process coritrols on chemical additions. The understanding of water minerals is helpful for the engineers to design and operate a geothermal power plant. The information on water content could assist in plant control. Sensing the change in water impurities is mandatory. IV. EXPLOSIVE ON-LINE INTERROGATION The detection of explosive materials in baggage by the OLPGAA technique is a newly proposed alternative to screeti suspected matter for airport security. The basic theory of such a PGAA method is to measure the unique prompt y-rays from nitrogen, it being a major element in explosives. The mean nitrogen density of various explosives is found up to 0.026 5 0.012 mo1/cm3, higher than the common nitrogen-contained materials such as foods, fabrics, and polymers;20therefore, the nitrogen content can be referred to as an indicator for explosives.

A. DETECTION OF EXPLOSWES BY OLPGAA From the detection point of view, the nitrogen caplure y-ray has the highest y-energy of 10.83 MeV; other neutron-induced prompt and decayed photopeaks in the spectrum do not superimpose and interfere with its full-energy md single-escape peaks. For such measurement, the use of a high-efficiency ydetector over a high-resolution detector is preferred. The NaI(T1) detector is still the most popular choice, but the investigation using a BGO detectw for detection of explosives demonstrated its superiority to NaI(Ti) in counting effi~iency.~'In Figure 13, a comparison of NaI(Tl) and BGO detectors for measuring nitrogen prompt y-peaks, with 2 kg urea as an explosive simulator irradiated by thermal neutrons from a "2Cf source, is illustrated. The peak at 10.83 MeV is obvious in the spectrum measured with the BGO detector. In a practical deteciim system, a multidetector array is necessary to improve the detection sensitivity, or shorten the counting duration. Although no other photopeaks can interfere

7 POLYETHYLENE

Ge ILi) DETECTOR

CAY05lAT

N

BLOCKS

I 1 Flgure 11 12Cf storage drum irradiation arrangement; all dimensions in inches. (Reprinted with permission from Nucl. Tech. 27, Duffey, D.,Analysis of geothermal power plant water using gamma rays from capture of californium-252 neutrons, 491, Copyright 1975, American Nuclear Society, Inc.)

CHANNEL

Figure 12 Prompt y-ray spectra obtained from (a) distilled water and (b) hot spring, irradiated with a

252Cf source. (Reprinted with permission from Nucl. Tech. 27,Duffey, D., Analysis of geothermal power plant water using gamma rays from capture of californium-252 neutrons, 495, Copyright 1975,American Nuclear Society, Inc-)

Gamma ray energy ( M e V )

la) 2.5"

1

2

3

4

5

a 2" BGO

6

7

8

9

1011

12

13

14

Qornma ray energy ( M e V )

Flgure 13 Prompt y-ray spectra for explosive-like material measured using (a) 2.5" x 2" BGO and (b) 3" x 3" Nal(T1) scintillator. (Reprinted with permission from Appl. Radiat. /sot 42, Lee, C. J. el al., tlighenergy gamma-ray spectrometer using bismuth germanate detectors, 550, Copyright 1991, Pergarnon Press, Ltd.)

with the nitrogen prompt y-ray of 10.83 MeV, background in this energy region is attributed mainly to random coincidence counting, and Gaussian spreading accounts for 65 and 2596, respectively, as estimated by Wang using an NaI(T1) detector.22It is concluded that improving shielding and using a high-resolution detector can reduce the background counting rate in the nitrogen region.

B. CURRENT OLPGAA SYSTEM FOR EXPLOSIVE DETECTION The capture of slow neutrons by nitrogen results in the highest energy y-ray normally observed in such reactions. SAIC made use of this fact to develop the first prototype thermal nitrogen analysis (TNA) system for airport security.23In this system, the luggage moves through a screen of thermal neutrons produced by a ZS2Cfsource; the detected 10.83-MeV y-rays are tomographically analyzed to give the spatial distribution of nitrogen. After extensive testing, the first SAIC system was put into service at JFK International Airport, New York, in Janua-y of 1990. Up to five units are scheduled to be operational by now at airports in the U.S. and E u r ~ p e . ~For ' radiation safety considerations, the leakage radiation from the system, induced radioactivities in thc irradiated luggage, potential hazard of malfunctions, and shipping and handling of the neutron source were carefully studied. The U.S. Nuclear Regulatory Committee (NRC) znd the relevant agencies found the TNA acceptable for airport lobby installations.

(A)

z N absorber Lead

1

-+!=='

PA BGO 1

TSCA

Figure 14 Layout of the explosive detection assembly designed for airport security inspection; and electronic block diagram of the y-ray counting system. (Reprinted with permission from Appl. Radiat. Isot. 43, Chung, C. eta!., Feasibility study of explosive detection for airport security using neutron source, 1426, Copyright 1993, Pergamon Press Ltd.)

TSCA

- + T s cPA ABGO + 5- ~

For normal operation of the system and anticipated abnormal events, the radiological impact to both the worker and the public is well below the dose guidelines imposed by the authorities. For increasing counting effkiency or reducing scanning time to make the PGAA explosive detection system more effective in practical use, the high-density BGO detector has been considered. An OLPGAA system, consisting of five 2" X 2" BGO detectors and a 20-pg 252Cf neutron source, was assembled for this purpose,24as shown in Figure 14. For each detector, an amplifier/TSCA was connected to select the proper energy range with respect to the 10.83-MeV nitrogen prompt y-ray; this included the 10.83MeV full-energy and single-escape peaks; urea with 42.7% nitrogen content was used as an explosivelike material in this test. Other prompt y-rays, originating mainly from hydrogen (2.2 MeV), chlorine (6.1 MeV), and iron (7.6 MeV), do not interfere with the designated photopeak. The net counting rate collected from the detecting system is around 0.4 countls/kg urea, and the derived detection limit is 0.5 kg for a 5-minscan. It is concluded that using a more intense 252Cf neutron source and bilateral detector arrays are necess~uyto improve the detection eficiency and shorten the counting period.

Elemental Composition, in weight %, of Chemical Warfares and an Explosive Element TNT Sarin VX Mustard Lewisite

Table 4

Reprinted with permission from IEEE Nucl. Sci. 39, Caffrey, A. J. et al., Chemical warfare agent and high explosive identification by spectroscopy of neutron-induced gamma rays, 1422, Copyright 1992, The Institute of Electrical and Electronics Engineers, Inc.

C . OLPGAA USING FAST NEUTRONS The thermal neutron-based OLPGAA will evolve and improve in the coming years. OLPGAA using fast neutron inelastic scattering facilitates the simultaneous determination of oxygen, carbon, and nitrogen by measuring their respective prompt y-rays at 6.13, 4.44, and 2.31 MeV, respecrively, yielding twoand three-dimensional images of the elements for. identifying a bomb and its position; this technique is now being developed by several groups.'O Recently, Science Applications International Corporation (SAIC) developed an Explosive Detection System (EDS) based on the Pulsed Fast Neuiron Activation (PFNA) technique." The induced prompt y-rays of 160,I2C,I4N,and 35Clcan be measured in the defined time intervals immediately after irradiation of a neutron burst; the determined ratios among these elements help specify explosives and even narcotics.

D. CHEMICAL WARFARE AND AMMUNITION IDENTlFlCATlON Inspection of chemical warfare agents and ammunitions may be possible by observing neutron-induced y-rays folkwing detection with a high-purity germanium detector.26The elemental conlposition of' some typical cheniical war';ue and high-explosive munitions are listed in Table 4. For Instance, the lewisite with a high content of arsenic can be identified clearly by the unique prompt y-rays. A 25'Cf neutron source and a high-resolution germanium deteztx are preferred for such measurement design, as shown in Figure 15. This system has proven successful in verifying declarcd chcinical warfare agellts and ammunitions.

V. CONCLUSIBMS The OLPGAA technique has spread into the coal industry for econon~icand environmental consideration. A number of commercial OLPGAA devices dedicated to coal analysis have been installed, for which a high-yield lS2Cfneutron source and scintillation detector array are usually coupied to chance counting efficiency. At the same time, sophisticated techniques, such as high-rate counting. coincidence measurement, and the spectrum stripping method, are incorporated to shorten the counting period without the loss of accuracy. The determination of major and minor constituents by the OLPGAA technique in many industrial process lines opens a new frontier for operational control, although it is still in the developmental stage. Fast neutrons are suihble for analysis of metallic matter due to its high penetrating ability and the reasonable reaction cross-section for these elements. The highest energy capture y-rays of 10.83 MeV from nitrogen plays an important role in on-line measurement of explosive material in baggage; this unique characteriStic y-ray is most discernible in the y-ray spectrum. With the improvement in radiation safety concerns for the public, such an OLPGAA device will soon be popular for the purpose of airport security. A pulsed'fast neutron activation technique allows determination of major elements in explosives, providing detailed information on explosive identification. Declared chemical warfare and ammunitions can be classjfied by their elemental compositions directly specified by the measured capture prompt y-rays as irradiated with a thermal neutron source.

Figure 15 Schematic neutron capture reaction for chemical warfare and ammunition identification. (Reprinted with permission from IEEE Nucl. Sci. 39, Caffrey, A. J. et al., Chemical warfare agent and high explosive identification by spectroscopy of neutron-induced gamma rays, 1422, Copyright 1992, The Institute of Electrical and Electronics Engineers, Inc.)

REFERENCES

1

Cywicka-Jnkiel, T. and Loskiewicz, J., Correlation Methods in Calorific Value Measurements of Coal, in Proc. Advisory Grorrp Meeting on Gamma,X-Ray and Neutron Techniques for the Coal Industry, IAEA, Vienna, 1986, 191. Sowerby, B. D., Measurement of specific energy, ash and moisture in bulk coal samples by a combined neutron and gamma-ray method. Nucl. Insrr. Meth. 160, 173, 1979. Henog, W., Prompt gamma neutron activation analysis of hard con]. raw lignite and a raw glass mixture, Nucl. Geophys. 3, 467, 1989. Wilde, H. R. and Ncrz,og, \K, On-line analysis of coal by neutron induced gamma spectrometry, J. Radioanal. Nucl. Chem. article.^, 7 1 , 253, !982. Henenberg, C. L., Use of small accelerators in coal analysis and coal sluny flow measurements, IEEE Nucl. Sci. 26, 1568, 1979. Wormald, M. R., A bulk materials analyzer using pair and Compton-suppressed gamma-ray spectrometry, Nucl. Geophys. 3, 461, 1989. Wormald, M. R., Pair spectrometer NaI(T1) array for neutron-induced prompt gamma-ray analysis. Nucl. Geophys. 3, 373, 1989. Sowerby, B. D., On-line nucleor techniques in the coal industry, ~Vucl.Geophys. 5, 491. 1991. Surman, P. L., Nuclear Techniques for the On-line Analysis of Coal in Electricity Generating Stations, in Proc. Advisory Group Meeting on Gamma. X-Rny and Neiitron Techniques for the Coal Industry, 1AEA. Vienna, 1986, 181. Marshall, J. H., 111 and Zumherge, J. F., On-line measurements of bulk coal using prompt gamma neutron activation analysis, Nucl. Geophys. 3, 445, 1989. Mcquaid, J. H., Brown, D. R., Gozani, T., and Bozorgmnnesh, H., High rate spectroscopy for on-line nuclear coal analyzer, IEEE Nucl. Sci. 28. 304, 1981. Gozani, T., Advances i n bu!k t:lemental analysis u s q neutroo interactions, Nucl. Geophys. 2, 163, 1988. Yates, S. W., filo, A. J., Chcn~,C. Y., and Coopc, D. E, Elemental analysis by gamma-my detection following inelastic neutron scnttering. J. R~dioarml.Nucl. CIlcin. Arficlcs. 46, 343, 1978.

14. Jiggins, A. H. and Habbani, F. I., Prompt gamma-ray analysis using 3.29 MeV neutron inelastic scattering, Int. J. Appl. Radiat. Isot. 27, 689, 1976. 15. Sowerby, B. D., Elemental analysis by neutron inelastic scatter gamma rays with a radioisotope neutron source, Nucl. Instr: Meth. 166, 571, 1979. 16. Cunningham, J. B., Sowerby, B. D., Rafter, P. T., nnd Greenwood-Smith, K., Bulk analysis of sulphur, lead, zinc and iron in lead sinter feed using neutron inelastic scatter y-rays, Int. J. Appl. Hadiat, h o t . 35, 635, 1984. 17. Liu, Y., Lu, Y., Xie, Y., Wang, Y., Du, Y., Tan, J., Uonian, M., and Seymour, R. S., Development and applications of an on-line thermal neutron prompt-gamma element analysis system, J. Radioanal. A1ucl. Chern. Articles, 151, 83, 1991. 18. Shirakawa, Y., A feasibility study to determine the carbon and oxygen contents of steel using fast neutron techniques, Nucl. Geophys. 5, 5 19, 1991. 19. Duffey, D., Balogna, J. P., and Wiggins, ?. F., Analysis of geothermal power plant water using gammarays from capture of californium-252 neutrons, Nucl. Technol. 27, 488, 1975. 20. Grodzins, L., Nuclear techniques for finding chemical explosives in airport luggage, Nucl. Instl: Merti. B56/ 57, 829, 1991. 21. Lee, C. J., Chao, J. H., and Chung, C., High-energy gamma-ray spectrometer using bismuth germanate detectors, Appl. Radial. Isot. 42, 547, 1991. 22. Wang, H. and Waana, C. M., IVNAA investigation of factors affecting the background in the measurement of nitrogen, J. Radioanal. Nucl. Chem. Articles, 151, 293, 1991. 23. Shea, P., Gozani, T., and Bozorgmanesh, H., A TNA explosives-detection system in airline baggage, Nucl. Instr. Meth. A299, 444, 1990. 24. Chung, C., Liu, S. M., Chao, J. H., and Chan, C. C., Feasibility study of explosive detection for airport security using neutron source, Appl. Radiar. Isot., 44, 1425, 1993. 25. Sawa, Z. P. and Gozani, T., PFNA technique for the detection of explosives, in Proc. 1st Inf. Synip. on Explosive Detection Technology, Virginia, 1992, 82. 26. Caffrey, A. .I., Cole, J. D., Gehrek, R. J., and Greenwood, R. C., Chemical warfare agent and high explosive identification by spectroscopy of neutron-induced gamma rays, IEEE Nucl. Sci. 39, 1422, 1992.

Appendix I

Thermal Neutron Capture Gamma-Rays

The energy and photon intensity of y-rays as seen in thennal-neutron capture are presented in two tables, one in as ending order of y energy and second organized by 2, A. Only those y-rays with E(y) 2 500 ke and intensity of 2 5 % of the strongest transition are included. In cases where there are more-than-one transitions for a nucleus within 0.1 keV of energy difference, the strongest one is kept. In the energy-ordered table, the three strongest transitions, if their intensities are 2 5 % of the strongest, are indicated in each case. Where the nuclide mass number is not indicated, the natural target was used. The nuclide given is the residual nucleus in the capture reaction. The y energies given ate in keV. The y intensities given are relative to 100 for the strongest transition. All data for A > 44 are talken from Evaluated Nuclear Str~rctureData File (8/91), a computer file of evaluated nuclear structure data maintained by the National Nuclear Data Center, Brookhaven National Laboratory, on behalf of' :he Nuclear Stnlcture and Decay Data network, coordinated by the International Atomic Energy Agency, Vienna. These data are published in Nuclear Data Sheets, Academic Press, San Diego, CA. The data for A 5 4 4 is taken from "Prompt Gamma Rays from Thermal-Neutrov Capture," M. A. Lone, R. A. Leavitt, D. A. Hamson, Atomic Data and Nuclear Data Tables 26, 51 1 (1% 1). This research was supported by the Office of Basic Energy Sciences, U.S. Department of Energy.

d

178 Table 1 Capture Gammas-Ordered

by Nucleus

Table 1 Capture Gammas-Ordered

by Nucleus (continued)

7282.2[ 8380.7

"I.

Table 1 Caprure Gammas-Ordered

by Nucleus (continued)

Table 1 Capture Gammas-Ordered

by Nucleus (continued)

"Yo

.'Yo 844.0 722.7 787.4 1023.0

1230.2 1432.3 6624.0

1

.'Yo

0.3

30.6 100.0 7.7

15.8 7.0 16.1

I

Table 1 Capture Gammas-Ordered by Nucleus (continued)

Table 1 Capture Gammas-Ordered

by Nucleus (continued)

Table 1 Capture Gammas-Ordered

by Nucleus (continued)

185

Table 1 Capture Gammas-Ordered by Nucleus (continued)

Table 1 Capture Gammas-Ordered

by Olucleus (continued)

--

Table 1 Capture Gammas-Ordered

7.0 18.9 31.1 7.0

13.1 7.8 11.3

6.2 ti. 7 7.1 5.2 8. 1 3.7

le.9

0.1 8.1 13.1 7.0

20.3 7.0 5.5 5.7 5.2 0.2

5.5 5.7 1.6

12.7 13.2 8.3 5.6 9.0 8.8 8.4 5.6 5.5 7.4

8.3 7.4

8.3 5.1 8.3 7.4 5.2

8.2 14.9 7.0 1.1 7.3 0.8 7.1 18.1 3.7

7.8 7.1

5. 5 13.1 18.0 6.0 22.1

74.a 32.3

by Nucleus (continued)

Table 1 Capture Gammas-Ordered

by Nucleus (continued)

Table 1 Capture Gammas-Ordered

by ~ucleus'(continoed)

Table 2

Capture Gammas-Energy

$3

M u c l .ua (Strongoat

Sorted

3 y'.)

Huc 1 o u a (Stroagoat 3 y e a )

'

,I

Table 2 Capture Gammas-Energy

Sorted (continued)

"'Yb

181,2.720.0.635.4

" * T s ~ ~ o ~ . 149.7.144.51 s.

Table 2 Capture Gammas-Energy Sorted (continued) # u c 1 *us (Strongest 3 7 . 8 )

Table 2 Capture Gammas-Energy

Sorted (continued)

Table 2 Capture Gammas-Energy

Sorted (continued)

195

Table 2 Capture Gammas-Energy Sorted (continued)

196 Table 2 Capture Gammas-Energy

Sorted (continued)

197 Table 2 Capture Gammas-Energy

Sorted (continued)

"'Yb 6614.7.804.7,776.4 1*1~m[167.8.62.1.1,147.6] 'aVf22.8.12S.1.8517-2)

Table 2 C a ~ t u r eGammas-Enerqv

Sorted (continued)

Table 2 Capture Gammas-Energy Sorted (continued)

sea.

8

970.2

200

Table 2 Capture Gammas-Energy Nuclaua ( S t r o n g a a t 3 y'a)

Sorted (continued)

Table 2 Capture Gammas-Energy

Sorted (a

Table 2 Capture Gammas-Energy

Sorted (continued)

Wuc 1 .us

(Strongeat 3 y'm)

Table 2 Capture Gammas-Energy

Sorted (continued)

204 Table 2 Capture Gammas-Energy N u c l bur ( S t r o n g e a t 3 7's)

Sorted (continued)

Table 2 Capture Gammas-Energy

Sorted (continued)

Table 2 Capture Gammas-Energy Sorted --- (continued)

Table 2 Capture Gammas-Energy

Lorted (continued)

208 Table 2 Capture Gammas-Energy N u c 1* u s ( S t r o n g ~ s t 3 1's)

Sorted (continued) Nua l - ~ u a (Strongeat

kr 3 1'0)

Table 2 Capture Gammas-Energy Sorted (continued) It?) (l-V)

(xl(Y))

Nuc l bus (slroogont 3 y 8 r )

Table 2

Capture Gammas-Energy

Sorted (continued)

Nucl eua ( S l r o n ~ m a l3 y'r)

"'RU

""Nd a''Th

1

174.0.130.1.284.4)

852.8.4838.Q.708.2 3473.0.586.8.472.31

Table 2 Capture Gamm~ts-Energ y Sorted (continued)

212

Table 2 Capture Gammas-Energy Soried (continued)

213

Table 2 Capture Gammas-Energy Nuclmur (Strontast 3 ? ' a )

Sorted (continued)

Table 2 Capture Gammas-Energy

Sorted (continued)

-

--

--- -

-

.

2.:.

Huolaus (ILrorgemL 3 7 ' 8 )

I " ' I "Kr

I

4573.8.l181.6,012.6 1475.9.3382.6.531.9

84Nb 5104.2.5491.0,6832.4) 21.1

18.2 LOO. 0

Table 2 Capture Gammas-Energy Nut l e u 8 ( S t r o n g o ~ t 3 7's)

Sorted (continued)

216 Table 2

Capture Gammas-Energy

Sorted (continued)

Table 2 Capture Gammas-Energy Nuolbua (Strongest 3

7's)

Sorted (continued)

21 8

Table 2 Capture Gammas-Energy

Sorted (continued)

Table 2 Capture Gammas-Energy Nuc 1our (strongom1 3 y'.)

Sorted (continued) Nuoleur (Srroagemt

3

7 . 8 )

220 Table 2

Capture Gammas-Energy

Sorted (continued)

Table 2 Capture Gammas-Energy N u c lbur (Strongest 3 7 ' 8 )

Sorted (continued) Nuc 1 r u t (3tro.go.t

3 7's)

"Po 810.8.883.6.1674.2) *eZn~7883.b,?Jf. 0 , 7 0 6 0 . 0 '7"Hf1125.7.5842.2.670.0

Table 2 Capture Gammas-Energy Nuolous (Strangont

3 7'8)

Sorted (continued)

Table 2

Capture Gammas-Energy

Sorted (continued)

Appendix /I

Other Radiation-Related Properties of Prompt Gamma Activation Facility Chien Chung CONTENTS

I. Introduction ..................................................................................... :.-........................................ 11. Construction Materials for PGAA Facility .............................................................................

III. Residual Radioactivity around PCAA Set-up ......................................................................... IV. Radiation Safety for the Operation of PGAA Facility ............................................................ References ...........................................................................................................................................

Although neutron behavior and prompt y-ray characteristics are certainly of major importance in the design work of a PGAA facility, the choice of satisfactory materials of construction is essential for a safe and practical PGAA set-up. Even if the structure materials with suitable physical and mechanical properties are available, they must have low neutron cross-section to avoid the interference of prompt y-ray measurement emitted from the irradiated sample. In the first section of this Appendix, emphasis will be placed on materials and material specifications that are unique to the construction and assembly of a PGAA facility. One of the major interferences for the PGAA measurements is the residual radioactivity induced by neutrons impinging on construction materials, neutronlgamma shields, prompt y-ray spectrometric unit, and the target sample as well. Activated short-lived radionuclei may obscure the identification of weak prompt y-rays in the spectrum; while longer-lived radionuclei may saturate the y-ray detector and disable it as a spectrometer. In the second section of this Appendix, the nuclear decay properties of frequently encountered radionuclei in PGAA measurement are tabulated. In PGAA measurement, scientists and researchers as well as the patients in in vivo medical scan are subject to radiation exposure. The radiation exposure may arise from neutron and y-sources external to the human body: from radionuc!ides that have entered the human body through inhalation and absorption through skin, or from the internally induced radioactivity by neutron irradiation. Although maximum safety efforts have been applied to the PGAA operation, scientists must be informed of the limit of radiation exposure and allowable intake in order to avoid excess exposure that may lead to radiation injury. In the last section of the Appendix, the allowable concentration of some radionuclei frequently encountered in PGAA operation as we!! as sensitive organs and tissues to radiation, recommended by regulatory authority, are listed. Unlike the conventional INAA, prompt y-ray activation measurements encounter unique radiationrelated problems although the PGAA operation has been widely recognized as an effective analytical method. Such problems include careful selection of construction material, radiation damage on the nuclear instruments, and radiation safety for both workers and irradiated patients. The Appendix provides necessary information and data for material selection, residual radioactivity evaluation, and radiation protection for both external and internal exposures. The values quoted in this Appendix may be used as references for preliminary calculation only and should not be used for design purposes; in the latter case, qualified engineers and licensed professionals should he consulted, in particular for the safety matter concerning the PGAA operation.

11. COMSTRUCYlON MABERIALS FOR PGAA FAClLITY Makrial requirements of PGAA sel-up can vary quite widely, in particular for the field applications such as in situ PGAA survey in a hostile geophysical environment, or in vivo PGAA diagnosis to which 0-8493-5 149-9/'?5/$0.00+5.50

O I995 by CRC Press, Inc.

I

the patient is irradiated by neutrons. In all c s e s , however, construction materials, neutron moderators and absorbers, and radiation shielding materials must be seiecred to meet appropriate requirements. The values and inforxation listed in the following tables, also to be found elsewhere,'-3 should b e ' k e d only for the preliminay design calculation.

Table 1 Properties of Construction Materials Used for PGAA Setup with Low Thermal Neutron Cross Sections --

--

-

-

-

Atomic number Element

(z)

Atomic weight (g/M)

Normal density (dcm3)

Thermal neutron cross section (barn)"

Beryllium Magnesium Zirconium Aluminum Niobium Yittrium Iron Molybdenum Chromium Copper Nickel

4(Bc) 12(Mg) 40(Zr) 13(A1) 4 l(Nb) 3907 26(Fe) 42(Mo) 24(Cr) 29(Cu) 28(Ni)

9.0 122 24.3 12 9 I -22 26.98 15 92.906 88.906 55.847 95.94 5 1.996 6334 58.7 1

1.85 1.74 6.50 2.70 8.57 5.5 1 7.87 10.20 7. I9 8.96 8.90

0.0092 0.063 0.185 0.230 1.15

a

1.28

2.55 2.65 3.10 3.79 4.43

Data taken from Reference 4.

Table 2 Austenitic Stainless Steel Used for PGAA Constructiona AlSi Type

Carbon, % (maximum)

Chromium, %

Nickel, %

Other Elements

" Data taken from Reference 5.

Table 3 Composition by Weight of Various Concretes for PGAA Setupa Concrete (Weight percent) Element (2)

Iron

Barytes

Regular

Aluminum (AI) Barium (Ba) Calcium (Ca) Iron (Fe) Hydrogen (H) Oxygen (0) Sulphur (S) Silicon (Si) -

Densitv( g/crn3)

" Data taken from Reference 6.

3.5

-

4.5

--

2.35

Table 4

Fast Neutron Removal Cross Section for Neutron Moderator Used in PGAA Setupa Atomic or

Material (2 or form)

Cross sections

molecular weight (g/M)

Density

27

9 10.8 2 12

1.70

Aluminum (Al) Beryllium (Be) Boron (B) Deuterium (D) Graphite (C)

,

Macroscopic

2.70

1.3 1

1.84

1.07 0.97 0.92 0.81

0.079 0.132

-

I 55.85 207.19 16 91.2

Hydrogen (H) Iron (Fe) Lead (Pb) Oxygen (0) Zirconium (Zr)

Microscopic (barns)

(gJcm3)

1.OO 1.98 3.53 0.92 2.36

7.87 11.34

-

6.5

--

18 20

Water (HzO) Heavy water (D20) Concrete, Regular Concrete, Barytes Concrete, Iron

Be0 B4C Diphenyl (C6HJC6HJ) Hydrocarbon oil (CH2) ZrH2

-

1.00 1.10 2.35 3.50 4.50

25 55.25 154

3-0 2.2 0.96

93.2

5.61

-

.

2.8

-

-

0.56 5.1 2.9 . 2.8 2.35

-

(1/cm)

-

-

0.065

0.168 0.118

-

0.101 0.103 0.092 0.089 0.105 0.159 0.111 0.12 0.01 1

-

0.085

' Data taken from Reference 1

Table 5 Thermal Neutron Absorbers Used in PGAA Setupa Nuclide

Natural abundance

Li-6 B-10 Rh- 103 Cd-1 13 a

Data taken from Reference 7. Reaction products labeled with

7.5% 19.8% 100% 12.2%

"*"

are radioactive.

Thermal cross section (barn)

+

942 2 3838 -t 2 134 It 4 19800 2 200

Nuclear reactionb "Li(n,a)'T* 10B(n.~)7Li '03Rh(n.r)'MRh* '"Cd(n,r)"'Cd

Table 6 Mass Attenuation Coefficient (cm2/g)for y-Ray Shielding Calculationa y-ray energy, MeV -

Material

W Pb Air

Hz0 Concrete

Human tissue a

-

0.1

0.5

1.o

1.5

2

4

6

4.2 10 5.290 ,151 .I67 .I69 ,163

.I250 .I450 .0868 .0966 .0870 .0936

,0640 ,0684 .0636 .0706 .Of535 .0683

.0492 .05 12 .05 17 .0575 .0517 .0556

.0437 ,0457 ,0445 ,0493 .OM5 ,0478

,0402 ,0420 .0307 ,0339 .0317 .0329

,041 .043 .025 .027 .026 .026

Data taken from Reference 8.

Fast Neutron Removal Cross Section for Neutron Moderator Used in PGAA Setupa

Table 4

Atomic or nlofecular weight (g/M)

Material

(2or form) Aluminum (Al) Beryllium (Be) Boron (B) Deuterium (D) Graphite (C)

Cross sections

Density (Cl1cm3)

Microscopic (barns)

Macroscopic (1Icm)

27 9 10.8 2 12

2.70 1.84

1.31 1.07 0.97 0.92 0.8 1

0.079 0.132

25 55.25 154

3.0 2.2 0.96

1.70

-

0.065

Hydrogen (H). Iron (Fe) Lead (Pb) Oxygen (0) Zirconium (Zr) Water (HzO) Heavy water (D20) Concrete, Regular Concrete, Barytes Concrete, Iron Be0 R4C Diphenyl (C6H,C6H5) Hydrocarbon oil (CH2) ZrH2

-

-

93.2

5.61

0.56 5.1 2.9 2.8 2.35

' Data taken from Reference 1.

Table 5 Thermal ~ e u t r o n Absorbers Used in PGAA Setupa Nuclide

Natural abundance

Li-6 B-10 Rh- 103 Cd-1 13

" Data taken from Reference 7. Reaction products labeled with

7.5% 19.8% 100% 12.2%

"*"

are radioactive.

Thermal cross section (barn)

Nuclear reactionb

942 t 2 3838 2 2 134 rt 4 19800 I+_ 200

bLi(n,a)'T* '"B(n,a)7Li 10IRh(n,r)lMRh* 113Cd(n,r)"4Cd

Table 6 Mass Attenuation Coefficient (cm2/g)for y-Ray Shielding Calculationa .. Material

H Be C N 0 Na Mg A1 Si P

S

K Ca Fe Cu Mo

W Pb Air H20 Concrete Human tissue

y-rs-j energy, MeV

0.5

1.O

I.5

2

4

6

.295 ,132 ,149 ,150 .151

.I730 ,0773 .0870 .0869 .0870

,1260 ,0565 ,0636 ,0636 .0636

,1030 ,0459 .05 18 ,0517 .05 18

,0876 ,0394 .OM .0445 .0445

,0579 ,0266 ,0304 ,0306 .0309

,044 .02 .024 ,024 .025

,151 ,160 .I61 .I72 ,174 ,188

.0833 .0860 ,0840 ,0869 .0846 ,0874

,0608 ,0627 .0614 ,0635 .0617 ,0635

.IN96 .05 12 .0500 .0517 ,0502 .0519

,0427 .0442 .0432 ,047 ,0436 ,0448

.0303 .03 15 .03 10 .0323 ,0316 ,0328

,02 .026 .026 ,027 ,027 ,028

.215 .238 ,344 .427 1.030

,0852 .0876 ,0828 .0820 ,0851

.06 18 ,0634 .0595 .0585 ,0575

.0505 .05 18 ,0485 .0476 .0467

.0438 .0451 ,0424 .0418 .0414

.0327 ,0338 .0330 ,0330 ,0349

.028 .030 .030 .030 .034

4.210 5.290 ,151 .I67 .I69 .I63

.I250 .I450 ,0868 .0966 ,0870 .0936

.0640 .0684 ,0636 ,0706 .0635 .0683

.0492 .05 12 ,0517 ,0575 .05 17 .0556

.0437 ,0457 ,0445 .0493 ,0445 ,0478

,0402

.M .043 .025 ,027 .026 .026

0.1

' Data taken from Reference 8.

,0420 .0307 .0339 .03 17 .0329

Table 9 Decay Properties of Radionuclides Induced by (nth,r)Reactions with Half-Lives 1 h < T,, 5 1 day Cross sectlon Half-life Decay Natural Nuclear reaction

(barn)

(hours)

0.10 5 0.02 0.108 2 0.002 0.64 f- 0.05 1.46 0.03 13.3 0.2

+

15.030 2.62 1.827 12.361 2.578

t 0.03 t 0.2 2 0.2 t 0.06 t 0.02

2.520 12.699 14.12 1.380 11.3

+ 0.4

4.42 1.83 4.48 1.134 2.805

IT, r IT, r

9.104 2.914 1.38 9.0 19.2

P-, r IT, r P-, r EC, P+, r @-, EC, r

abundance

23Na(n.r)24Na 30Si(n,r)3'Si 'OAI-(~,~)~'A~ 41K(n,r)42K ssMn(n,r)S6Mn

100% 3.1% 99.6% 6.73% 100%

MNi(n,r)6-'Ni 63Cu(n,r)"Cu 7'Ga(n.r)72Ga "Ge(n~)~'Ge "jGe(r~,r)~Ge

0.91 % 69.2% 39.9% 36.5% 7.8%

1.49 4.4 4.6 0.36 0.06

79Br(n,r)80"Br s2Kr(n,r)83mKr "Kr(n,r)8SmKr 84Sr(n,r)85mSr 86Sr(n.r)nmSr

50.69% 11.6% 57.0% 0.56% 9.8%

2.4 20 0.09 0.59 0.84

+

2 4

t 0.01 2 0.06

t- 0.05

mode P-, r P-. r P-. r

P-.

P-. P-, P', P-. P-. P-.

T-

r r EC, r r r r

P-. n'. I.

IT, EC, r IT. EC, r

P-.

r EC, r IT, r P-, r EC, r PI,

P-. r EC, P+, r EC, r P-. r 13-. r 3.684 5.5 23.85 16.98 13.10

P-. r IT. r P-. r P-. r IT, r

19.15 18.3 1.57 23.8 3.253

P-.

--

1931r(n,r)'9JIr '~Pt(n,r)"Tt '%Pt(n,r)'gmPt '%Hg(n,r)'"'"Hg 208Pb(n,r)2wPb

62.'7%

25.3% 25.3% 0.15% 52.3%

110 5 30 0.7 t 0.1 0.05 t 0.02 120 t 20 5.0 -1 0.1, x lo-4

r P-, r

IT, P-. r IT, EC, r

P-, r

Note: Decay properties are taken from Refep.xce 7. Decay mode: P-, beta; p+,positron; EC, electron capture; IT, isomeric transition: r, gamma transition.

Table 10 Decay Properties of Radionuciides Induced by 1 day < T,,, 11 month

Nuclear

Natural

reaction

abundance

31P(n,r)32P '6Ca(n,r)47Ca MCr(n,r)5'Cr 70Ge(n,r)7'Ge 7sAs(n,r)76As

100% 0.0035% 4.35% 20.5 %

100%

Cross sectlon (barn) 0.18 0.7 15.9 3.25 4.4

2 0.02 2 0.2

+ 0.4

2 0.13 2 0.2

(n,,,r) Reactions with Half-Lives Half-life (days) 14.282 4.540 1 27.701 11.15 1.097 1.4727 1A60 8.82 2.669 2.75 1 2.88 16.96 2.224 14.0 1.128

Decay mode

PP-, r EC, r EC

P-. P-.

r

r

P-, r P-. EC. r P-. r P-. r

EC.

EC, r

EC, r

P-.

r

IT, r

I3-

P-. P2.

r

P-.

m,

EC. r EC, r

IT, r IT, r

P-.

r

IT, r EC, r IT, EC, r 1.678 1.43 1.38 10.98 1.95

P-.

I.

IT, EC, r P-. r P-. r P-. r

P-. r P-. r

15.4 1.28 3.00 4.33 4.020

EC, r IT, r IT, r

2.697 2.672 5.013

EC, r P-. a,

Note: Decay properties are taken from Reference 7. Decay mode: a, alpha; @-, beta; capture; IT, isomeric transition; r; gamma transition.

P-. r

P+, positron; EC, electro

Table 9 Decay Properties of Radionuclides Induced by (n,,,r) Reactions with Half-Lives 1 h < TI,* 5 1 day Nuclear Natural Cross section Half-life Decay reactlon abundance (barn) (hours) mode

IT, r IT, r

P-, ITI r IT, EC, r IT, EC. r

'"Xe(n,r)'35Xe '33Cs(n,r)'"mCs '38Ba(n.r)'39Ba "Te(n~)'~~Ce 141Pr(n,r)142Pr

10.4% 100% 7 1.7% 0.19% 100%

0.25 2.5 0.4 6 7.6

t 0.03 t 0.4 0.1 2 1 t 0.7

9.104 2.914 1.38 9.0 19.2

'48Nd(n,r)'"9Nd '51Eu(n,r)'sZ"'Eu 's'E~(n,r)'S2m%~ 15BGd(n.r)'S9Gd '56Dy(n,r)'57Dy

5.7% 47.9% 47.9% 24.8% 0.057%

2.5 0.5 3200 200 4 2 1 2.4 Ifi 0.4 33 2 3

1.73 9.30 1.6 18.56 8.06

'64Dy(n,r)'6SDy '62Er(n,r)'63Er '64Er(n,r)'65Er '70Er(n,r)'7'Er 176Yb(n,r)'nYb

28.1% 0.14% 1.56% 14.9% 12.6%

900 19 13 5.7 2.4

+

"+

+ 300

2.334 1.25 10.34 7.52 1.88

Ifi 2 2 2 ? 0.2

t: 0.2

3.684 5.5 23.85 16.98 13.10 'q31r(n,r)'941r '"Pt(n,r)IqPt '"Pt(n,r)'"'"Pt IWHg(n,r)'97mHg 208Pb(n,r)209Pb

62.7% 25.3 % 25.3% 0.15% 52.3%

110 0.7 0.05 120

? 30

t 0.1 +- 0.02 ? 20

5.0 2 0.1, x ~ o - ~

Nore: Decay properties are taken from Reference 7. Decay mode: isomeric transition: r, gamma transition.

19.15 18.3 1.57 23.8 3.253

P-,

r

IT, r

P-. EC,

r

P+,

r

P-.

EC, r

P-,

r

flz, EC, r IT, r

P-:

r

EC, r

P-.

r

EC, P+. r EC, r

v. r B-, P-.

r r

IT, r

PA. r P-. r IT, r

P-. fi-,

r

IT, P-, r I?: EC, r

P-.

I .

P-,bera: P*, positron: EC, electron capture; IT,

Table 10 D e c q Properties of 1 day < TIl2 5 7 month

Radionuclides Induced by (n,,r) Reactions with Half-lives .4

Nuclear reaction

Natural abundance

Cross section (barn)

Half-life (days) 14.282 4.540 1 27.70 1 11.15 1.097

%e(~~,r)'~'*Xe '32Xe(n,r)133Xe

4.1% 26.9%

+

0.4 0.2 0.4 i: 0.1

100% 0.15% 100%

98.8 2 0.3 3000 t 200 3-01? It 0.002

Note: Decay properties are taken from Reference 7 . Decay mode: a, alpha; capture; IT, isomeric transition; r, gamma transition.

mode

P-

P-. r EC, r,.. EC B-. r

2.88 16.96 2.224 14.0 1.128

EC, r EC, r P-. r IT, r

9.625 2.68 1 16.78 1.25

P-. r P*, EC. r

0-

EC, r

P-,IT, r

8.89

IT. r

11.770 5.245 2.19 12.0

IT, r [3-, r IT,r EC.. r

1.43 1.38 10.98 1.95

IT,EC, r P-. r

1.118 9.40 4.19 6.7 1 3.777

P-, P-, P-. P-,

15.4 1.28 3.00 4.33 4.020

1*Au(n.r)'98A~ '%Hg(n~)'~~Hg 209B i(n.r)Z'oBi

Decay

2.697 2.672 5.0 13

.

.

P-, r B-. r r r r r

B-. EC. r

P-. r

P-.

r EC, r IT,r IT. r

P". r EC, r P-,

a. r

P-, beta; P+, positron; EC, electron

Table 13 Risk Factor and Weighing Factor for Human Tissues and Organs Sensitive to Ionization Radiations ICRP 26 Publication (1977) Risk

Tissue or organ

factor

Weighing factor

ICRP 60 Publication (1991)

Risk

factor

Weighing factor

Gonads Red bone marrow Colon Lung Stomach Bladder Breast Liver Esophagus Thyroid Skin Bone surface Remainder Total Note: Although the factors released in the recent publication (ICRP-60) are more conservative, the factors published in 1977 (ICRP-26) are still adopted by many authorities.

' Remaining organs include liver, kidneys, heart, intestine, and stomach. Remaining organs include adrenals, brain, small intestine, upper large intestine, kidneys, muscle, pancreas, spleen, thymus, and uterus.

REFERENCES 1. Lamarsh, J. R., Introducrion to Nmdear Engineering, 2nd ed., Addison-Wesley, Menlo Park., CA. 1983. 2. Cember, H., Inrrod~rctionto Healfh Physics, 2nd ed., Pergamon Press, New York. 1983. 3. Attix, E H., Introduction to Radiological Physics and Radiation Dosimetry, John Wiley & Sons, New York, 1986. 4. Garber, D. I. and Kinsey, R. R, Neutron Cross Section. Vols. I & II.3rd ed., BNL-325, Brookhaven National Laboratory, New York, 1976. 5. A. S. T. M. Special Publication 276, Materials in Nuclear Applications, ASTM, Washington D.C., 1960. 6. American Concrete Institute, Concrete for Radiation Shielding, 2nd ed., ACI, New York, 1962. 7. Lederer, C. M. and Shirley, V., Tclhle of Isotopes, 7th. ed., John Wiley & Sons, New York, 1978. 8. Templin, L. T., Reactor Physics Cotistam, Report ANL-5800, 2nd ed.. ANL, Chicago, 1963. 9. International Commission on Radiological Protection, Limits for Intakes of Radionuclides by Workers, ICRP publication 30, Parts 1, 2, 3, with associated Supplements and Addendum. Annals of ICRP 2-8, 19, Pergamon Press. Oxford, 1979-1988. 10. International Commission on Hndiological Protection, Recornmendation of the ICRe ICRP Publication 26, Annals of ICRP 1(3), Pergamon Press, Oxford, 1977. 1 1. International Commission on Radiological Protection, Age-Dependent Doses to Members of the Public from Intake of Radionuclides, ICRP Publication 56, Part 1. Annals of the ICRP 20(2), Pergamon Press, Oxford, 1989. 12. International Commission on Radiological Protection, 1990 Recommendations of the ICRP, ICRP Publication 60, Annals of the ICRP 21(1-3). Pergamon Press, Oxford, 1991. 13. International Commission on Radiological Protection, Annrral Lbnifson Intake of Radionuclides by Workers Based on the 1990 Recommentlations, ICRP Publication 61, Pergamon Press, Oxford. 1992.

INDEX A Absolute counting efficiency, 47, 49, 50 Absolute thermal flux. 54 ADC, see Analog-to-digital converter Airport security inspection of luggage. 24 OLPGAA facility used for, 34 Aluminum measuring of by PGNAA, 66 process control. 169 Alzheimer's disease, brain tissues from, 70 Analog-to-digital converter (ADC), 5. I I , 14,25, 133 Anticoincidence, 62 Anti-Compton annuli. 13 coincidence system, 35 mode, background reduction in. 24 shield, 85 spectrometer. 19, 39 Attenuation coefficient. 62

Backscattering. 26 gauge, 161 system, 25 Beam catcher, 31 heterogeneity. 96 Beryllium oxide, use of for PGAA facility. 29 BGO, see Bismuth gemanate Bilateral detector arrays. 173 Bilateral irradiation, 78 Biological shield. 3 1 Bismuth collimator, 80, 107 germanate (BGO). 14.15, 106 germanate detector, 17, 19, 24.43.47, 142 energy resolution of, 134 resolution of, 170 use of for PGAA facility, 29 Blood, measurement of trace clement content in, 70 BNCT, see Boron neutron capture therapy Body composition. 101 nitrogen determination of, 117 measurement. 118 Bone disease, 104

surface dose rate equivalents for, 126

sensitivity of to radiation, 119 Borehole measurements, neutron flux perturbation in, 140 probe, 136 Boron -doped filter, 34 determination of by PGNAA. 65 measurement of by NaI(T1) detector, 66 rneasurement of in plant tissues, 66 neutron capture therapy (BNCT), 68 use of for PGAA facility, 29 Borosilicate glasses, measurement of boron in, 69 Rrain mercury concentration. 115, 1 16

Cadaver studies. 89 Cadmium, 28 concentration, detection limit of. 113 measurement of by NaI(T1) detector, 66 ratio (CR),39, 71, 95 shutter. 59 use of for PGAA facility, 29 Calcium. mensuring of by PGNAA, 66 Calibration spectra, 86 Californium neutron source, 32 Carbon, determination of. 80, 118 Carbodoxygen logging, 137 Cation exchange capacity (CEC), 138, 139 CEC, see Cation exchange capacity 252Cf source, disadvantage of, 75 Chemical analysis, nuclear techniques of, 1 Chemical warfare, 175 Chlorine, 25 capture y-rays. 137 measmement of body. 118 measurement of in plant tissues, 66 prompt y-ray spectra, ?7 signal. 8 1 Clay minerals, 138 Clinical diagnosis, 101 Coal mdysis. 160 ash delineation, 140 in sitir analysis of. 85 rnensurement of sulphur content in, 143 neutron capture reaction with hydrogen in, 162 quality, elements influencing, 161 as source of energy. 159 Coaxid detectors, 45 Cold neutron(s), 61, 97 sensitivity with, 86 source, 95

Collimator(s), 26, 105, 110 lithium-dopd polycstcr, 89 measurement of Ca with, 84 at NIST. 63 Comparator. 110 Compton continuum background of, 19 reduction of, 13 Compton scattering, 8-10. 29, 77 Compton suppression, 1 1, 20, 21, 69 factor, 19 mode, 62 spectm, 22.23 spectrometer, 35 Constant fraction discriminator. 12 Construction materials choice of satisfactory, 225 y-ray background from, 39 properties of, 226 Converging guides, 98 Converted dose qoivalents, 121 Copper, use of for PGAA facility, 29 Counting efficiency, 20, 173 Counting statistics. 35 CR, see cadmium ratio Critical kidney burden level, 127 Cyclotron, fast neutrons from. 109

Deionized water, use of for PGAA facility, 29 Delayed analysis. 2 Delayed Gamma Neutron Activation Analysis (DGNAA), 59.79 Detection limits, 83 Detection sensitivity, 170 Detection systems. 10-12 anticoincidence Compton suppression. 10 coincidence double-escape counting, 11-1 2 Detector(s) crystal, 150 deterioration, 30 efficiency, 110. 146 -phnntom nxis, 39 photopeak energy resolution for. 15 radiation protection, 43 sensitivity, inadequate. 87 shield, 13, 19, 78 susceptibility of to damage, 62 DGNAA, see Delayed Gamma Neutron Activation Analysis Dialysis therapy, 118 Dje away measurements, 150 Diphenyl, use of for PGAA facility, 29 Dose rate equivalents, 122, 123 Double-escape peak, 9, 11.30, 68, 77, 85 Drinking water. 150

E ELAN, see Elemental analyzer Electrode polarity, 46 EIectron, binding energy of, 7 Elemental analyzer (ELAN), 164, 165 Elemental concentration, absolute measurement of. 110 Elements, detection limit of essential, I19 Energy nonlinearity, 164 resolution, post-irradiation recovery of, 49 Environ~nenlalprotection codes, 139 Epithermal neutrons, 3 Ethylene glycol, 86 Explosive detecting assembly, 35.42 Explosive Detection System. 174 Explosives. use of BGO detector for detecting, 170 Extrinsic detectors. 7

Fast analog signals, transmission of, 85 Fast neutron@),3, 32 bombarding, 46 energy, 28 fluence, 34.41.49 flux, 53.54 moderator. 29 removal, 54,227 Fat-free body mass, increased hydration of, 80 Fat-free mass (FFM). 8 1 F e d age theory, 76 FFM, see Fat-free mass Field Effect Transistor preamplifier, 6 Fluorocarbon refrigerants. 135 Flux density, 96 intensity. 54 monitor. 112 Foil activation technique, 39.42 Food poisoning. 115 Formation analysis, elements important to, 139 Free electron, 8 Full width at half rnaxianutn (FWHM), 6, 7, 25, 47

Gadolinium orthosilicate detector, 134 Gamma detectors, 5-7 scintillation detectors Na(T1) and BGO, 5-6 solid-state ionization detector, 6-7 Gas turbine engine. 97 Gaussian spreading, 78, 172 Genetic studies, 68 Geological formations, measurements of. 88 Geothermal waters, on-line analysis of, 170

Gemium detector, 1l,44 y-ray spectrum, 48.51 operating temperature of, 151 isotopes. 39, 45 purification technology, 7 recoil energy of. 45, 52 Gold, use of for PGAA facility, 29 Graphite collimator. 62 reflectors. 63 use of fof PGAA facility. 29 y-rayw attenuator. 29 capture of, 93 count rate. 52 detector, 105 spectrometry, 7.26 y-spectrum, shape of, 7-10

Heart disease, 104 dose rate equivalents evaluated in. 125 Heavy water, use of for PGAA facility, 29 High-flux reactor. 61 High-purity germanium (HPGe) detectors, 15 HPGe detectors, see High-pnrity germanium detectors Hydrocarbon contamination, 96 Hydrogen, liquid, 95 measurement of in plant tissues. 66

ICP-MS. 59 ICT. see Insulating core transformer INAA, see Instrumental methods of neutron activation analysis In-beam measurements, 56 Indium foils, 41 use of for PGAA facility. 29 Industrial workers, 127 Inelastic scattering. 79 interactions, 43 reaction. 89 In situ applications, 131-157 . coal mine in situ analysis. 139-145 coal ash delineation using scintillation detectors, 140-142 multielemental analysis using high-resolution detectors, 142-145 environmental water body in situ survey, 145-150 field survey using WPGAA probe, 148-149 performance test, 146-14,8

radiation safety concerns, 149-150 ISPGAA instrumentation, 131-235 detectors, 134-1 345 ISPGAA probe, 131-132 neutron sources, 132-134 oil formation in situ analysis. 135-139 bulk properties of rock formation, 137-138 multielemental analysis. 138-139 planetary in situ exploration, 150-152 seabed mineral in situ survey. 153-154 In situ PGAA (ISPGAA). 23. 25, 131. 135. 145 Instrumental methods of neutron nctivntion analysis (INAA), 102 Instnimentd photon activation analysis (IPAA), 102 Instruments, shielding and, 13-36 advanced instrumentation, 17-25 anti-Compton and pair spectrometers, 19-23 field instruments and spectroscopy. 23-25 prompt y-ray spectroscopy, 14-17 scintillation detectors, 16-17 semiconducting detectors, 15 shielding of PGAA facility, 25-35 biological and detector shields, 29-35 requirements for shielding, 26-29 Insulating core transformer (ICT). 4 Internal radiation. 49 Intestinal rare earth marker, 69 Intestine, dose rate equivalents evaluated in. 125 In vivo PGAAA (IVPGAA), 23 facilities comparison of capability of operational, 109 detector shields for. 33 . radiation shield design for, 31-32 index, 1 10 measurement, 30. 124 neutron sources used for, 103 station, 34 Ionization radiations, risk factors for human tissues sensitive to. 235 IPAA, see lnstrumentd photon activation analysis Iron. use of for PGAA facility. 29 Irradiation port, 83, 107 Isoflux contour curves, 41 Isotope capturing. 96 radioactive, 153 ISPGAA, see In situ PGAA Itd-itai disease. 104 IVPGAA, see In vivo PGAAA

Kidney cadrnirlni, prompt y-my spectrum of. 1 13 concentrntion of cndmiurn in, 89, 112 detection limit of mercury in, 116 dose rate equivalents evaluated in, 125

dysfunction. 104 mercury, detection limit of, 115 lrcutron flux distribution in, 120 KUR, see Kyoto University Reactor Kyoto University Reactor (KUR), 68 .

Lake pollutant surveys. 149 LBM, see Lean body mass Lead collinmtors. 68 -photon interaction. X-rays induced by, 39 sinter samples, bulk analysis of, 166 use of for PGAA facility. 29 Leakage radiation, 172 Lean body mass (LBM). 79 Light Water Reactor (LWR), 3 Light work condition, 233 Limestone, pair spectrum of, 163 Liquid hydrogen, 95 Liquid phantom. 39 Lithium polyethylene shield, 82 use of for PGAA facility, 29 Liver concentration of cadmium in, 89, 112 detection limit of mercury in, 116 dose rate equivalents evaluated in. 125 dysfunction. 114 Living tissues simulator, prompt y-ray emission from, 69 Lucite effective atomic number of, 110 phantom, 123 Lunar rock sample, 67 Lunar surface, geometrical arrangements of neutron probe with mspect to, 151 Lung@) . contaminant. 116 disease, 104 dose rate equivalents for, 125, 126 LWR, see Light Water Reactor

Manganese, measuring of by PGNAA, 66 Marine exploration, 153 Mass attenuation coefficient, 228 Mathematical model, for transport of neutrons, 76 MCA, see Multichannel analyzer MDC, see Minimum detectable concentration Medical community, IVPGAA adopted by, 101 Medical diagnosis elements importance to. 104 IVPGAA, 233 modified THMER facility for IVPGAA, 108

Mercury amalgams, dentists exposed to, 83 measured brain content of. 76 toxicity. clinical symptoms of, 115 use of for PGAA facility, 29 Metabolic disturbance, 104 Mineral processing plants, 164 Minimum detectable concentration (MDC),146, 147, 153 Mobile reactor, 32 Monazite rock samples, 86 Monte Carlo code. 80 simulation, 87 techniques, for tracking neutron, 136 Multichannel analyzer (MCA), 3, 7, 10, 14, 25 Multichannel pulse height analyzer, 5 Multidetector array, 35 Multielement analysis, by neuron activation, 83 MURR, see University of Missouri Research Reactor

National Institute of Standards and Technology (NET), 93,97 Neutron absorption, 78, 95 activation analysis, 97 amplifiers, 5 beam, 32 direction. 38 guided cold, 93 guided, 96 modulation of collimated, 59 profile, 39 profiles, 30 shutter system. 108 bombardment accumulated, 41 induced radioactivity after, 43 capture, 102 geophysical profiles given by, 142 logs, 137 reactions, 32 cold, 97 collimation. 27 counter. 112 cyclotron-induced secondary, 115 damage, reduction of, 25 detector damage caused by fast. 132 dosimeter, 57 emission anisotropic, 53 rate, 50 energy spectrum, 50, 121, 122 field mixed, 52

overexposure in, 57 fluence accumulated, 37 data. 49 flux. 78 contour map, 41 delivery of to irradiated patient, 105 measurements, 39.40, 57 nonthermal. 39 of research nuclear reactors, 3 uncertainty of, 56 unifonnity, 85. 119 focusing, 98 generator, 4, 143. 102 guides. 61. 96. 98 -induced y-rays, 140 inelastic scatterings. 76 moderator, 86 monitoring device, 49 premoderation of, 106 refractive index for, 94 scattering. 43, 96 shield, 26. 34. 37, 42 source, 3-5,33 detector distance. 82 : isotopic, 102 neutrons from accelerators and neutron generators, 4 properties of, 4- radioactive neutron sources, 4-5 specificiation of, 131 types of, 132, 143 thermalization of, 87 transport code. 42 wavelength, 94 Neutron beams, ~ r o m p gamma t activation analysis with guided, 93-100 accuracy, 96-97 advantages of guided neutron beams 93 apparatus. 95 facilities for PGAA with guic: .3 beams, 97-98 production of guided beams. 97 results and applications, 96 trends, 98-99 Neutron damage, induced effects of :clt nr *, ments used and, 37-58 neutron damage on detectors, 43- :g neutron damage on scintillation detec s. 4 6 4 9 neutron damage on semiconducting cle! ors, 44-46 neutron flux distribution around instruments, 37-43 neutrons around PGAA field instnlments, 40-43 neutrons around reactor-based PGAA instn~ments, 38 4 0 neutron-induced effec~son detectors, '19-56 ~

using an HPGe detector as fast neutron monitor. 50-54

using an HPGe detector as thermal neutron monitor. 54-56 NIST, see National Institute of Standards and Technology Nitrogen amount of inside body. 89 liquid, 135 measurement, 84 Nonthermal neutrons, slowing down of, 28 Nuclear interaction, rate of, 1 Nuclear reactor, y-photons with, 59 Nuclide mass number. 177 production of activated. 149 Nutrition, elements related to human balance of. 102 Nutritional disturbance. 104

Ocean-bottom deposits, analysis of, 87 Ocean Drilling Program, 154 Oil reservoir, 137 OLPGAA, see On-line PGAA On-line applications, 159-176 explosive on-line interrogation, 170-175 chemical warfare and ammunition identification, 175 current OLPGAA system for explosive detection, 172-1 73 detection of explosives by OLPGAA. 170-172 OLPGAA using fast neutrons, 174 mineral on-line process control, 164-170 geothermal power utilities. 170 metal industry, 166-170 on-line analysis of coal, 159-164 commercial OLPGAA systems, 164 development of OLPGAA techniques, 160-164 importance of on-line analysis of coal, 159-160 OLPGAA coal analysis, 160 On-line measurement assembly. 169 On-line PGAA (OLPGAA), 23.25, 159 facility, layout of. 34 technique, development of thermal neutron, 169 Organic materials, contamination with. 67 Osteoporosis, 1 14

PA, see Preamplifier Pair production, 9 spectrometer, 21, 162 spectroscopy, detection system for. 1 1 Palladium, 4 Peak-to-Compton ratio, 20 Performance degradation, 47

PET scan, 125 PGAA, sce Prompt gamma activation analysis PGNAA, sce Prompt gaxnrna neutron activation analysis Phmtom distribution of thennal neutron flux in, 120 female-like lucite, 110. 111 liquid. 39 Iucite, 123 man-like liquid, 110. 1 11 Phosphor, 5 Photo sources, monoenergetic, 48 Photomultiplier, 86 Photomultiplier tube (PMT), 14, 15, 43, 48, 107 Photopeak. 9. 11 area, quantitative determination of, 49 channel, drift of, 46 counts, 115 energy. 19.46 identification of, 13 Pile-up stripping. 166 Plastic scintillators, 13 Platinum shield. 86 PMT, see Photomultiplier tube Pole-zero cancellation filter, 106 Polyethylene borated, 164 detector shielded by baiated, 86 matrix blocks. 31 Pork meat, concentration of nitrogen in. 117 Potassium, measuring of bli PGNAA. 66 Preamplifier (PA), 14. 16 Prompt gamma activation analysis (PGAA), 93 disadvantages in using, 57 experiment, effect of Compton suppression in, 20 facilities neutrons emitted from source of. 28 reactor based, 13, 29 suitability of. 35 field applications, detectors used for, 43 measurement y-ray spectrometer in. 57 major interferences for, 225 systems. detection sensitivity of. 25 Prompt gamma activation andysis, in vivo, 101-130 clinical applications. 112-1 19 partial-body scan, 112-1 16 whole-body scan, 116-1 19 IVPGAA facility, 102-1 12 gmeraI layout, 103-1 10 phantom calibration, 110-1 12 radiation doses to patients. 119-127 doses for partial-body scan. 122-124 doses for whole-body scan, 125-127 neutron flux distribuson in body, 119-122 Prompt gamma activation fxility, other radiation-related properties of. 225-235

construction materials for PGAA facility, 225-228 radiation safety for operation of PGAA facility. 233-235 residual radioactivity around PGAA set-up. 229-233 Prompt gamma neutron activation analysis (PGNAA), 2, 3, 9, 59 facility, detection system of, 64 disadvantage of, 59 y-ray detection systems used for, 10 Prompt photopeak, 45, 52 Propane, liquid, 135 Protein depletion, 80 Pulse height spectrum, 45 analysis, 106 Pulses. categories of. 43 Pyrolytic graphite crystals, 98

Radiation damage, principal consequence of, 45 detectors. 37 exposure prime concerns of, 233 scientists subject to, 225 safety, 27. 103, 107. 172 skin doses, 124 tissues sensitive to, 119 Radiative capture, 2, 88, 112 Xadioactive decay, measurement of, 6 Kadioactivity, induced, 30 Radioisotope sources, PGNAA with neutron generators, charged particle accelerators. and, 75-92 accelerators. 89 14-MeV neutron generators, 87-89 radionuclide sources, 75-87 UIArn-Be sources, 75-77 252Cfsources, 82-87 Pu-Be sources, 77-82 Radionuclide(s). 60 decay properties c ; 229-233 derived air c: -.'ation of submerged, 234 disintegration : . , 60 encou%e,r:l i i i PGAA operation, 225 pr!'jtexxies of, 44 qtia:?a:ive identification of induced, 51 sourc:.. 'f.5 Random coinc ,.!ence counting, 78. 172 Rare earth eleilients, 29, 66 Raw glass mixtures, 84 Reaction core, 30, 32, 39 y-rays, suppression of, 62 high-flux, 6 1 power level, 108

probability. 1 thermal neutrons from. 60 Reactor neutrons, prompt gamma neutron activation analysis with, 59-73 practical applications. 65-71 techniques with PGNAA with nuclear reactors, 59-65 detection. 62-65 filters and collimators, 62 Red bone marrow dose rate equivalents for, 126 sensitivity of to radiation, 119 Reflection, increased neutron flux by, 83 Relative neutron flux. 54 Renal failure, disorder of phosphoms in patients with, 1 1 8 Renal tabular damage, 114 Research reactor, 102 Rhodium, use of for PGAA facility. 29

Salinity determination. 147, 148 Samarium. 69 Sample-to-detector distance, 39, 57 SAMPO, 20 Scattering inelastic, 88 neutron absorber, 31 peaks of inelastic, 87 Science Applications International Corporation, 174 Scintillation detector. 6, 49 Seawater, measurement of manganese nodules in, 86 Selective pulse integration, 78 Semiconducting detectors, 15, 134 Semiconductor materials. 7 Sensitivity to background, 147 Shadow shield, 81 Shielding materials. 32 Shutter device. 26 Signal-to-noise ratio (SM ratio), 80, 82, 132 Silica-bearing dust, occupational exposure to, 116 Silicon, measuring of by PGNAA, 66 Silver, use of for PGAA facility, 29 Single-channel analyzer, 12 Single-crystal quartz, 62 Single-escape peak, 9.30. 61. 85 Skin doses. 124 Slagging, 160 Slowpoke reactor, 3 SM ratio, see Signal-to-noise ratio Sodium iodide detectors, 13 measuring of by PGNAA, 66 Source -to-detector distance. 52 -detector tube, 82 storage safe, 83

Spectrometry timing scheme, inelastic mode. 133 Spontaneous fission source. 75 Stomach, dose rate equivalents evaluated in, 125 Sulfur content determination. 85. 154. 159. 160 measurement of in complex materials, 67 Surgical illness, 80

Target-detector assembly. 95 TBCI, see Total body chlorine TRF, see Total body fat TBH, see Total body hydrogen TBN measurement, 84 TBN, see Total body nitrogen TBW, see Total body water TFW, see Total fat weight Therapy evaluation, 101 Thermal capture reactions, 48 Thermal neutron absorber, 27, 29. 43. 84. 149 bombardment, semiconductor response to, 54 capture, 77 energy of y-rays in, 177 gamma-rays, 177 reactions of. 43 chopper. high-speed. 59-60 fluences, 41 flux. 31. 38, 39 distributian, 3 1, 40, 42 monitored, 54 of THOR facility, 55 moderator, primary, 42 Thermal nitrogen analysis (TNA), 172 Thermoluminescent dosimeters, 54 Thermoluminescent detector (TLD), I22 THMER, see Tsing Hua Mobile Educational Reactor THOR, see Tsing Mua Open-pool Reactor Thyroid, dose rate equivalents for. 125. 126 Time-to-Amplitude Converter, 12 Timing single-channel analyzer (TSCA), 24, 173 Titanium, 4 compressor blade, 97 measuring of by PGNAA, 66 oxide, 67 TLD, see Thermoluminescent detector TNA, see Thennal nitrogen analysis Total body carbon, measurement of, 88 chlorine (TBCI), 8 1 fat (THF), 79, 81 hydrogen (TRH), 79, 88 nitrogen (TRN), 77

watet (TWW). 80

Total fat weight (TFW), SO Total neutron emission rate, 75 Total weight (TW), 79 Transmission counting, 26 Transmission detection system. 25 Transport disturbance. 104 Triple coincidence. 12, 62 Tritium dilution, 80 TSCA. see Timing single-channel analyzer Tsing Hua Mobile Educational Reactor (THMER), 17, 108 critical assembly, 33 facility. 34, 116 Tsing Hua Open-pool Reactor (THOR), 15.38 neutrons from. 3 1 PGAA measurement using. 48 thermal column of, 47 Tungsten shielding. 76 TW,see Total weight Two-collimator facility, 83

U U.S. Nuclear Regulatory Committee (NRC), 172 University of Missouri Research Reactor (MURK), 63

W Waste discharge, 145 Water attenuation, 147 phantom, measurement of, 87 pollutant survey, 25, 40 shielding, 75 Wet chemical analysis, 84 Whole-body carbon measurements, I27

z Zetatron, 132, 133 Zirconium, 4 Zirconium hydride. 27