ML19249A431
| ML19249A431 | |
| Person / Time | |
|---|---|
| Issue date: | 05/31/1977 |
| From: | Ting P NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | |
| References | |
| NUREG-0230, NUREG-230, NUDOCS 7908220641 | |
| Download: ML19249A431 (23) | |
Text
NUREG-0230 VERIFICATION OF PRIOR MEASUREMENTS BY NONDESTRUCTIVE ASSAY 9 P'S fr220 4 S'/
Office of Standards Development-g,; 7
- 5 '1 U. S. Nuclear Regulatory Commission
Available from National Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $3.50 ; Microfiche $3.00 847 255
NUR EG-0230 VERIFICATION OF PRIOR MEASUREMENTS BY NONDESTRUCTIVE ASSAY P. Ting Date Published: May 1977 h
Division of Siting, Health,and Safeguards Standards Office of Standards Development U. S. Nuclear Regulatory Commission Washington, D. C. 20555
TABLE OF CONTENTS (Continued)
FMe 12 BIBLIOGRAPHY.
APPENDIX A - Major Gamma-Ray Signatures for the Fissionable 15 isotopes.
APPENDIX B - Assigned Half-Lives and Specific Powers of Plutonium 16 Isotopes and Americium-241........
APPENDIX C - Neutron Emission Pates of Some Nuclear Materials........
17 IV
TABLE OF CONTENTS Page AC KNOWL EDuM EN T S....................
v 1.
INTRODUCTION.
I 2.
FACTORS AFFECTING MEASUREMENT ACCURACIES............
1 2.1 Operational Characteristics....
I 2.1.1 Operational Stability......
1 2.1.2 Geometric Effects.....
2 2.1.3 Stimulating Radiation Unifcrmity.....
2 2.1.4 Energy of the Stimulating F.adiation...
2 2.2 Isotopic Composition of SNM Sample.
2 2.2.1 Multiple Gamma-Ray Sources.
2 2.2.2 Multiple Spontaneously Fissioning Isotopes...
3 2.2.3 Multiple Thermal Power Source.....
3 2.2.4 Multiple Fissile and Fertile Isotopes..........
3 2.3 Response Dependence on the SNM and Matrix Materials Within Sample..
3 2.3.1 Attenuation of Emitted Gemma Rays...
3 2.3.2 Neutron Multiplication.
3 2.3.3 Alpha-Neutron (2-n) Reactions......
4 2.3.4 Sample Self-Shielding.....
4 2.3.5 Isotopic Equilibrium....
4 2.3.6 Effect of Matrix Materials....
4 2.4 Composition and Dimension of tre Container.
4 3.
NDA TECHNIQUE SELECTION.......
4 3.1 SNM Applications...
4 3.1.1 Plutonium............
4 3.1.2 Uranium.
5 3.2 Examples of SNM Measurements by NDA.
5 3.2.1 Gamma Spectroscopy....
5 3.2.2 Calorimetry Method..
7 3.2.3 Neutron Coincidence Counting.
7 3.2.4 Active Assay....
8 4.
ACCURACY OF NDA MEASUREMENTS.
8 5.
CONCLUSION.
8 5.1 Technique Selection........
8 5.2 NDA System Specifications.
9 5.3 Other Considerations.....
9 REFERENCES.....
10 tii
ACKNOWLEDGMENTS The author wishes to express his appreciation to Mr. William J. Gallagher of ERDA for his contribution in the preparation of this report.
Acknowledgment also is extended to the following individuals for their valuable comments and suggestions:
Mr. Louis W. Doher of Rockwell International; Dr. Raymond Gunnink of Lawrence Livermore Laboratory; Dr. P. D. Randolph of Aerojet Nuclear t.ompany; Dr. David E. Rundquist of Science Application, Inc.; Dr. Darryl B. Smith of Los Alamos Scientific Laboratory; Dr. Walter W.
Strohm of Mound Laboratory; Mr. C. P. Sutter of Atlantic Richfield Hanford Company; and Mr. Erick L.
May, Mr. Dean M. Kunihiro, and Dr. Willard B. Brown of NRC.
The author also would like to thank Mrs. L. Gallagher of NRC for her editorial assistance and careful review of the final draft of this report.
4915'1 v
1.
INTRODUCTION Paragraph 70.51(f)(3)(ii) of 10 CFR Part 70, "Special Neclear Material," requires that ICC licensees subject to 70.51(e) establish procedures for verification by remeasurira previously measured, non-tamper-safed special nuclear material (SNM) in process. The purpose of this report is to delineate nondestructive assay (NDA) techniques that may be used to fulfill the above requirement.
Since NDA methodology and applications of the associated instrumentai. ion are discussed extensively in the literature included in the bibliography, emphasis in this re yrt is placed only on the factors affecting NDA measurement accuracies and the selection of the appropriate NDA techniques for certain memrements based on these factors.
Verification of previously measured materials, for which tamper-safing was not employed, is needed for demonstrating at physical inventory (Ref.1) that the quantities of SNM previously measured are still present. To ensure that there has been no apparent removal of the SNM since the previous measurement, all containers may have to be first reweighed and then scanned by an appropriate NDA technique.
By reweighing each previously measured non-tamper-sated container, followed by NDA scanning of an attribute characteristic of the contained SNM, any significant removal of such material since the previous measurement can be detected. The attribute is selected for beirg uniquely characteristic of the contained SNM so that alteration of the material to avoid detection of the SNM removal would be difficult. Such unique attributes include the heat output of a plutonium sample, the induced radiation from a fissile sample undergoing active assay, the isotopic com-position of a low-enrichment uranium sample, the gama spectroscopic analysis of a U-235 sample, or the spontaneous fission neutrons from a Pu-240 sample.
It is hoped that the infomation presented in this report will be helpful in establishing an acceptable framework for ising nondestructive assay for the verification measurement of SNM in non-tamper-safed containers.
2.
FACTORS AFFECTING MEASUREMENT ACCURACIES Certain sample characteristics and measurement practices can affect the response of NDA systems. If these are not controlled or properly accounted for in determining the quantity of SNM from the measured activity, errors will be introduced into the final results. This section identifies important sources of measurement error and recor_. ends possible solutions to reduce or eliminate them.
The factors contributing to the overall uncertainty associated with an NDA measurement are identified in the following four areas: (1) The operational characteristics of the NDA systems, (2) the isotopic composition of the SNM sample, (3) the SNM and matrix materials within a sample, and (4) the composition and dimensions of the container.
2.1 Operational Characteristics The operational characteristics of the NDA system, together with the cbility of the system to resolve the desired response from a composite signal, determine the ultimate usefulness of the system. These operational characteristics include (1) operational stability, (2) geometric effects, (3) stimulating radiation uniformity, and (4) energy of the stimulating radiation.
2.1.1 Operational Stability The ability of an NDA system to reproduce a given measurement may be sensitive to fluctua-tions in the operational environment. Temperature, humidity, and line voltage variations affect NDA systems to some extent. These effects may be manifested through the introduction of spurious electronic noise or changes in the high voltage applied to the detectors or amplifiers, thereby changing Ae detection efficiency.
The environment must be controlled if such fluctuations result in severe NDA response variations that cannot be eliminated through calibration and operational procedures. The sensi-tivity to background radiations can be monitored and controlled through proper location of the system and the use of radiation shielding, if required.
2.1.2 Ge_ometric Effects The concern in the geometry of counting is: What portion of the radiations that are emitted by an SPN sample actually enters the sensitive volume of the detector? This portion ranges from a very small fraction, when the distance from sample to detector is large, to 100%
for some 4w-detectors, when the SNM sample being counted is completely inside the detector.
Also, for some detectors such as the well-type gamma and neutron counters, the response of the gross count rate is a function of the position of a point source within the well. This implies that the size and distribution of the sample being counted within the well may have an important bearing on the detector response. To make the necessary error corrections, therefore, it is important to predetermine the ceometry effect of a counting system.
2.1.3 Stimulating Radiation Uniformity It is important that the stimulating radiation field (i.e., interrogating neutron or gamma-ray flux) in active NDA systems be designed to be uniform in intensity and energy spectrum throughout the volume of the irradiation chamber. The residual effect can be measured by using an SNM sample that is small in comparison with the dimensions of the irradiation chamber. The response can then be measured with the S?N sample positioned in different locations within the irradiation chamber.
Various methods, including rotating and scanning the container, scanning the source, distributing sources, and combinations of these methods, have been investigated to reduce the response uncertainty attributable to a nonuniform stimulating radiation,ield. Scanning a rotating container while the detector and source positions are fixed appears to offer an advan-tage in response uniformity and is therefore recommended.
2.1.4 Energy of the Stimulating Radiation If the energy of the stimulating radiation is as high as practicable but below the threshold of any interfering reactions such as the neutron-induced fission in U-238, the pene-
' ration of the stimulating radiation will be enhanced throughout the volume of the irradiation chamber. A high-energy source providing neutrons above the energy of the fission threshold for fertile constituents such as U-238 and Th-232 can be employed to assay the fertile content of a container.
The presence of extraneous materials, particularly those of low atomic number, lowers the energy spectrum of the interrogating neutron flux in active neutron NDA systems. Incorporating a thermal neutron detector to monitor this effect and thereby provide a basis for a correction to reduce the response uncertainty caused by this effect is recommended.
Active neutron NDA systems with the capability L noderate the interrogating neutron spectrum can provide increased assay sensitivity for samples containing small amounts of fiss' material (<100 grams). To enhance the range of usefulness of the system, it is best that thi<
moderation capability be removable.
2.2 Isotopic Composition cf SNM Sample The sample response observed by an NDA system is usually a composite of contributions from more than a single isotope of uranium or plutonium. These sources of radiation are described below.
2.2.1 Multiple Gamma-Ray Sources The plutonium isotopes Pu-238 through Pu-241, as well as Am-241 (a daughter product of Pu-241),
are spontaneous gama-ray emitters. The observed spontaneous gama-ray spectrum from plutonium represents the contribution of gamma rays from these isotopes.
847 M, 9 The uranium isotopes U-233 and U-238 spontanecusly emit a gamma-ray spectrum and one gama ray from U-238 (s766 kev, which is really from the Pa-234 daughter nuclide) is nearly of the same energy as one emitted by Pu-238. Uranium-232, occurring in combination with U-233, has a series of prolific gamma-ray-emitting daughter products, including Th-228.
2.2.2 Multiple Spontaneously Fissioning Isotopes Plutonium-240 is the principal spontaneously fissioning isotope in the plutonium fuel now used in the nuclear fuel cycle. A few percent of the total fission signal observed is due to Pu-238 and Pu-242.
Uranium has a much lower spontaneous fission rate than flutonium, but in mixed-oxide fuels the spontaneous fission yield from U-238, which is the predominant isotope present, may approach one percent of the Pu-240 yield.
2.2.3 Multiple Thermal Power Source Each calorimetric measurement response from plutonium is a summation of the spontaneous power emission (in watts per gram) from the isotopes Pu-238, Pu-239, Pu-240, and Am-241. The contribution from each isotope varies as the amount of each isotope varies in the plutonium.
Plutonium-238 and Am-241 have much higher specific power (5.6716x10-1 and 1.1423x10-1 watts /
gram) than the other plutonium isotopes.
2.2.4 Multiple Fissile and Fertile Isotopes In active NDA assay systems, the observed fission response may come from more than one radioisotope. When assaying uranium, if the energy of the interrogating neutrons is above the fission threshold energy for U-238, the response will include a contribution from the fertile U-238 present in he sample, in addition to the response from the fissile U-235, which has a much lower fission threshold.
Wnen a sample of plutonium or uranium mixed-oxide fue activel: interrogated, the measured response ulll include contributions from the plutonium anc uranium fissile isotopes and, if the energy of the interrogating neutrons is high enough, the response will include contributions from both the fissile and fertile isotopes of plutonium and uranium.
2.3 Response Dependence on the SNM and Matrix Materials Within Sample If an NDA system is sensitive to the geometry of the detection system and the field of interrogating radiation, variations in the observed sample response may be attributed to the attenuation of emitted gamma rays, neutrnn multiplication, alpha-neutron reactions, atte nuation of the interrogating radiation, and isotopic equilibrium. These factors are discussed i 60w.
2.3.1 Attenuation of Emitted Gamma Rays The intensity of gamma rays is attenuated as they pass through matter. The magnitude of this attenuation depends on the energy of the gamma rays, as well as the thickness, density, and atomic number of the absorbing material. Low-energy gamma rays are much more severely attenuated than those of high energy. Measuring gamma rays of energies below approximately 80 kev is difficult and usually impractical because of severe attenuation problems.
Uraniun and plutc,nium have high mass absorption coefficients; therefore, self-absorption of gamma rays by these two elements can severely limit the application of passive gamma-ray spec-trometry of relatively d L se samples.
2.3.2 Neutron Multiplication Neutrons from spontaneous or induced fission reactions can be absorbed by other nuclei, subsequently causing these nuclei to fission and release additional neutrons. This nultiplica -
tion results in an increasad sample response and therefore should be considered in all active and passive NDA systems.
r 847 76U 2.3.3 Alpha-Neutron (a-n) Reactions Neutrons are randomly generated through interaction of alpha particles with light elements such as oxygen, fluorine, nitrogen, carbon, and beryllium. Isotopes of uranium and olutonium ertit large quantities of alpha particles, which can cause (a-n) reacticns and produce neutrons th2t increase the response of fission measurements.
2.3.4 Sample Self-Shielding When performing an active assay of dense SNM such as metals or heterogeneous samples con-taining lumps of SNM, self-shielding of SNM against the interrogating neutrons will occur to varying degrees. Penetrability of the interrogating neutrons into a sample is dependent on the energies of these neutrons, the density of the sample, and the type and amount of matrix material present. Matrix materials containing neutron moderators such as hydrogen will atte nate the interrogating neutrons to lower energy levels and increase the sample self-shieldins ffect.
2.3.5 Isotopic Equilibrium Some gamma-ray signatures involve gamma ra): emitted from a daughter product of the fissioning isotope of the sample. These signatures can be accurately measured only when the isotopes of the sample and its daughter are in equilibrium. For example, the ingrowth of U-237 in freshly separated plutonium allows one to use the 208-kev gamma ray of U-237 to measure Pu-241 after equilibrium is reached.
2.3.6 Effect of Matrix Materials Matrix materials, which ran be considered as all materials present in the sample other than SNM, can have various effects on the NDA results.
Matrix materials of low atomic number, particularly hydrogenous materials, reduce the energy of neutrons by elastic collision. The lower-energy neutrons are more likely to interact with other nuclei in the sample and thus affect the observed spontaneous fission response.
Matrix materials of high atomic number cause attenuation of gamma rays, particularly gamma rays of low energies. This can cause decreased response during an NDA measurement.
2.4 Composition and Dimension of the Container Variations in the thickness of the sample container walls and sti,ctural composition can significantly affect the accuracy of NDA measurements. Therefare, the sa..M type of container should be used, if possible, for the standard and the unknown samples. Otherwise, a calculated or measured ccrrection must be applied to the measured data to take into account the difference in container wall thickness and structural composition.
3.
NDA TECHNIQUE SELECTION The NDA techniques selected for verification of prior measurements should reflect a con-sideration of the accuracy requirements for the assay and the type of SNM and containers in the invento ry.
No single NDA method is capable of universal application, and frequently more than one NDA technique is required to perform the job.
In this section, general applications of SNM measurements by NDA and e<amples of different techniques are discussed.
3.1 SNM Applications 3.1.1 Plutonium Calorimetry determinations are the least sensitive to matrix effects but rely on a detr' led knowledge of the Am-241 content and the plutonium isotopic composition to transform the measured heat flux to grams of plutonium (Ref. 2).
d-P
- 4 84 7 t6T m"
gG3
e Ganina-ray spectroscopy systems complement the potential of other assay methods by providing the capability to nondestructively determine, or verify, the Am-241 content and the plutonium isotopic composition (with the exception of Pu-242). High-resolution gama-ray systems are capable of extracting the maximum amount of information (isotopic composition, isotopic content, presence of extraneous gansna-ray sources) from an assay, but content density severely affects the accuracy of quantitative predictions t,ased on that assay methud.*
Passive coincidence detection of the spontaneous fission yield of Pu-bearing systems provides an indication of the combined Pu-238, Pu-240, and Pu-242 sample conter.t. With known isotopic composition, the Pu content can be computed (Ref. 3). Neutron multiplication effects become severe at high Pu sample loadings. Plastic scintillation coincidence detection systsms are often designed in conjunction with active neutron interrogation source systems. Op< rated in passive and active modes, such systems are able to provide an assay of both the spontaneously fissioning content and the fissile content of the sample. The spontaneous background can be subtracted from an active NDA response to provide a yield attributable to the fissile SNM content of the container.
3.1.2 Uraniun!
Active neutron systems can p ovide for both high-energy and moderated interrogation spectrum capabilities. Operation with the high-energy neutron source will decrease the density dependence and neutron self-shielding effects and will therefore significantly enhance the uniqueness of the observed response. To extend the applicability of such a system to small fissile loadings, a well-moderated interrogating spectrum can be used to take advantage of the increased U-235 fission probability for neutrons of low energy.
The number and energy of the gamma rays emitted from the uranium isotopes, with the excep-tions of the minor isotopes U-232 and U-237, are generally lower than those for plutonium. The 185-kev transition observed in the decay of U-235 is frequently used in uranium applications.
However, the penetration of this U-235 primary ganea ray is so poor that the ganvr.a-ray NDA technique is applicable only with low-density, homogeneous samples.**
Calorimetry is not applicable to the assay of uranium enriched in the U-235 isotope because of the low specific power. Also, the intense activity of the daughter products of U-232 imposes a severe complication on the use of calorimetry.
3.2 Examples of SNM Measurements by NDA The following examples are some typical NDA procedures used frequently in measuring SNM.
Table l provides a cross-reference of various NDA techniques used with eatisfactory results in measuring some common forms of SNM.
3.2.1 Gamma Spectroscopy 1.
Thallium-activated sodium iodide detectors are used for measuring uranium-235 enrichment (Ref. 4); the sample is measured through an appropriate collimator and the 185-kev activity is detected with a Nal(Tt) crystal to obtain the U-235 enrichment of the sample. Using the colli-mator and a simple method for proper subtraction of the Compton continuum permits measuring a signal that is truly proportional to the U-235 enrichment. The net counts in the 185-kev peak are proportional to the percent enrichment of the sample.
This technique may even be used to measure samples of uranium-bearing materials that are opaque to the 185-kev gamma rays since the measured 185-kev activity per unit is proportional to the U-235 enrichment. The measurement is nearly independent of matrix material (Z 30) and 1
- Although techniques do exist that can provide quite accurate assay on isotopic ratios that are not sensitive to density and matrix effects, the techniques are still under active development to make them more generally applicable.
- It should be noted that by using the Segmented Gamma Scan (SGS) procedure, homogeneous matrices are no longer a limiting factor for the measurement of U-235 material.
847 R T Table 1 NDA TECHNIQUE SELECTION FOR SOME COMMON SNM SAMPLES Technique Gamma Sper'roscopy Neutron Coincidence SNM Calorimetry NaI(TE)
Ge(L1)
Counting Active Assay b
Low-enrichment U-235 Fuel Pellet (Ref. 4)a y
(Ref. 5)
UF Product (enriched)
(Ref. 6)
X 6
UO2 (>80%)
(Refs. 7, 8)
(Refs. 7, 8)
(Ref. 9)
U-235 Ash (Ref. 10)
U-238 Incinerator Ash X
(Ref. 11)
UO Power X
(Refs. 5, 12) 2 U/Pu Mixed 0xide (Ref. 13)
(Refs. 14, 15)
(Ref. 16)
Pu Metal Alloy X
(Refs. 15, 17, 18)
X Pu0 Feed Material (Ref. 19)
X (Refs. 20, 21) 2 Pu-239 Mixed Oxide X
X (Refs. 6, 22)
Pu-238 Metal and Metal 0xides X
X (Ref. 23)
Pu Mixed 0xide Scrap X
X (Ref. 15)
Pu Nitrate (Refs. 16, 24, 25)
(Ref. 26)
Pu-239 (Honogenously distributed in ash or graphite matrix X
(Ref. 27)
X (Refs.23,26)
Pu-238 and Pu-239 Scrap and Waste Cans X
X X
X (Ref. 28)
Pu High-Density., High-SNM-CC Content l Samples (Ref. 2)
X N
,The references represent some typical applications.
bThe NDA technique is being used routicely in the nuclear industry.
t )
G O
relative densities of sample components (Ref. 6). Corrections can usually be made, when neces-sarv. for effects of cladaing thickness and the relative densities of uranium and matrix material.
2.
Thallium-activitated sodium iodide detectors are used with an external transmission cechnique for measuring plutonium (Ref. 27); gamma rays in the 335-kev complex from plutonium are counted by an NaI(TA) detector. The window of a single-channel analyzer is centered at the energy complex. Transmission measurements are made to correct"for sample attenuation. The plu+ onium content of the sample is then determined from the measured gama activity.
3.
Segmented gamma scanners employing lithium-drifted germanium (Ge(Li)) detectors are used for measuring alutonium or uranium ash (Ref.10); the 414-kev Pu-239 gama activity, the 766-kev Pu-238 gamma activity, or the 185-kev U-235 gama activity is measured with a collimated Ge(Li) detector. The detector, collimated so as to view about a 1/2-inch slice of the sample at any one time, scans the sample in the vertical direction. The data are corrected for self-absorption by measuring the transmission of an external source (for U-235, use a Yb-169 source; for Pu-239, use an Se-75 source; for Pu-238, use a Cs-137 source) through the sample correction applied as a function of vertical height of the sample to minimize errors due to sample inhomo-geneities. The measured activities are used to determine the amount of fissile material in the sample.
4.
Automated Ge(Li) scanners are used for measuring uranium / plutonium mixed-oxide fuel pins (Refs.14 and 15); a Ge(Li) da ?ctor is used to measure the intensity of the 414-kev gamma rays from Pu-239, the 185-kev gamma rays from U-235, and gamma rays from other heavy ' -topes.
To increase the efficiency for measuring a large throughput of the same type of sample.
a automated sample changer scans the samples. The measured 414-kev gama activity determines the amount of Pu-239 content; other isotopes from plutonium can also be assayed by measuring their respective gamma activities (see Appendix A).
5.
High-resolution Ge(Li) detectors are us3d for measuring plutonium nitrate (Ref. 24);
a high-resolution Ge(Li) detector is used to measure the intensity of the 129-kev (Pu-239),
149-kev (Pu-241),152-kev (Pu-238),160-kev (Pu-240), and 414-kev (Pu-239) gamma rays from plutonium nitrate solutions. The ratio of the measured intensities of tFe 129-kev and 414-kev gama-ray liries is used to correct for attenuation in the sample. The mcasured activities determine the total plutonium content of the s wple and the Pu-240, Pu-241, and Pu-238 content relative to Pu-239.
3.2.2 Calorimetry Method 1.
Portable dry calorimeurs are used for measuring mixed-oxide fuel rods (Ref.13); -
calorimeter measures the amount of heat generated by the natural radioactive decay of nuclear materials. For example, Pu-239 nuclei decay by alpha-particle emission, with each decay pro-ducing about 5.5-MeV kinetic energy plus the energy associated with gama radiation. Most of this energy is absorbed in the sample itself and is converted into heat (for Pu-239, the natural rate of heat production is about 2 milliwatts per gram of material). The portable dry calo-rimeter treasures the amount of additional electrical heating power required to maintain a fuel rod at a specific temperature higher than that which could be maintained by the heat produced in radioactive decay of nuclear material within the rod. If the isotopic composition of the rod is known, the nuclear material content of the sample can be detemined from the difference in electrical heating power required to maintain a fuel rod at the same temperature as a dummy rod not containing fuel.
2.
An isothermal twin-bridge calorimeter is used for measuring plutonium oxide feed material (Ref.19); this calorimeter measures the difference in heat output Wen the known outat of a standard or electric heater and the output from the unknown sampie. There are several different techniques ufed to make this measurement, all of which empicy a measurement of the temperature difference between the sample chamber and the chamber containing the kntwn sample or reference heater. This temperature difference is measured with thermoresistive el eents placej inside each chamber and connected in a Wheatstone Bridge configuration. If the icotopic cont ut rf the sample is known, the quantity of nuclear material in the sample can be deteained frm the measured heat output from the reference chamber and the Wheatstone Bridge medurement of the temperature difference between the chambers.
hkf 3.2.3 Neutron Coincidence Counting An He-3 neutron coincidence counter is used for measuring U-238 incinerator ash (Ref.11);
neutrons emitted spontaneously from fissionable material samples are moderated and detected by moderated He-3 neutron detectors. Detector coincidence logic, designed to distinguish between fission neutrons and (a,n) neutrons, operates on the principle that each fission event produces two to three neutrons while an (a.n) event only produces a single neutron. The fission neutrons detected from a single fission event are correlated in time; on the average, they will be detected within a time that is rougnly equivalent to the neutron lifetime in the detector moderator. The number of coincidence events
- is therefore related to the amount of spontaneously fissioning material in the sample. A correction is measured for the accidental coincidences that occur because of the detection of two (2,n) neutrons within the coincidence time interval. Additional measurement made with an external tission source placed adjacent to the sample can be used to correct for changes in the effective counter efficiency due to the moderating effects of different sample matrices. A known isotopic content allows the total amount of fissionable material in the sample to be determined from the measured neutron coincidence count rate.
3.2.4 Active Assay 1.
An automated active coincidence counting system is used for measuring Pu-239 mixed-oxide light-water reactor (LWR) fuel rods (Ref. 22,, a low-energy Pu neutron source or a moderated Cf-25? source is used to irradiate the fuel rod (the source used depends on the scan speeds required). The multihead plastic scintillating detectors count the emissions of induced fission gama rays and neutrons from neutron-induced fission primarily occurring in the fissile material in the rod. The detectors use nanosecond coincidence logic to sort out fission events (multiple emissions of gama rays and neutrons) from background radiation. The coincidence count rate is directly related to the amount of fissile material within the rod.
2.
An active isotopic source assay system is used for measuring high enrichment U-235 oxides (Ref. 9); a collimated beam of neutrons from an external source is used to stirrulate fissions in the assay sample. When a fission event occurs in the sample, two to three neutrons and about seven gamma rays are emitted simultaneously. An array of fast-response, plastic scintillation radiation detectors detects two or more of tte emitted fission neutrons and gamma rays in coincidence. The coincidence requirement discrimina ;es against other reactions occurring within the sample that produce fewer particles per event, as well as direct interactions of the source neutrons with the detectors. The coincident count rate ' eelated to the amount of fissionable material in the sample.
4.
ACCURACY OF NDA MEASUREMENTS The relative accuracy of NDA techniques used for attribute scanning may vary widely 90ending on the sample type and the measurement system. However, it is iv ortant to sele-t the appropriate NDA technique for an application and to optimize the instrument-operating conditians so that the best relative accuracy of the measurement may be obtained.
As shown by investigators referenced in Table 1, relative accuracy of 10.17% is obtainable fcr measuring low-enrichment U-235 fuel pellets with a 2-;n x 2-in sodium iodide detector, while relative accuracies of 4 to 5% are obtainable for assaying plutonium nitrate solution with a high-resolution lithium-drif ted gennanium detector. Relative accuracies of less than +1% are obtainable for assaying plutonium with high-performance calorimeters. As for the neutron coincidence counting technique, relative accuracy of a few percent is obtainable for assaying plutonium oxides, while relative accuracy of about 3% is obtainable for assaying uranium ash.
When using active assay techniques, relative accuracies of 11% and 10.5% are obtainable for assaying LWR plutonium mixed-oxide and LWR U-235 fuel pellets, respectively.
5.
CONCLUSION General conclusions regarding the considerations in selecting and using NDA techniques for verification of the quantity of SNM stored in non-tamper-safed containers are given below.
5.1 Technique Selection It is important that the NDA technique selection be based on attainable accuracy after considering the characteristics of the SNM sample. The measurement performance required may be such that, when the uncertainty corresponding to the remeasurement is combined with the
- A coincidence event is one in which a pair of neutrons are detected within a specified time interval.
847 M h 3p
ur. certainties corresponding to the other inventory components, the quality constraints on the total limit of error of th
- rial unaccounted for (LEMUF) will be satisfied. However, it should be realized that in some cases no single technique appears capable of meeting all the requirements. When more than one type of information is required to separate a composite response, more than one NDA technique may be required to provide that information.
5.2 NDA System Specifications It is important that NDA systems selected for verification measurements be designed to meet the following objectives:
1.
Operational characteristics are essentially independent of external background radiation, temperature, humidity, and line voltage, or these variables must be closely controlled.
2.
System response is essentially independent of the isotopic concentration, matrix of the SNM sample, measurement geometry, and the stimulating source intensity and energy.
3.
Systems are designed to allow repu ducible positioning of the detector and item being assayed.
4.
With a choice of high-resolution ie(Li) gamma-ray detection systems, the system has the performance specifications described in Re gulatory Guide 5.9, " Specifications for Ge(Li) Suc-troscopy Systems for Material Protection Measurements."
5.
With a choice of hish-efficiency l'aI(Tt) gama-ray detection systems, the system has a resolution of Full-Width-Half ~ ximum less than 16% at the 185.7-kev energy peak of uranium-235, which is generally considered be adequate Mr measuring uranium enrichment.
6.
With a choice of neutron coincidence counters, the system has high-detection efficiency and is capable of operating in the presence of gama radiation. In addition, it is important that the instrument supplier provide the performance information that is required for estimating errors regaruing the precision of the coincidence response as a function of the real coincidence counting rat 3, the background to real coincidence ratio, and the relative coincidence Nsponse from a point of source of fission radiation as a function in the counting chamber.
5.3 Other Considerations To obtain the optimum results in verification measurements, it is also important that:
1.
SNM containers be selected consistent with the capability and sensitivity of the NDA system, with uniformity of wall thickness and material composition.
2.
Adequate collimators be used for both gamma and neutron coincidence counting systems.
3.
In implementing the verification or prior measurement program, the statistical aspects of the program also be considered. Various sampling plans and statistical tests may be used to ensure the validity of prior measurements within the original limits of error. Verification measurements should be performed at 3 level of confidence equivalent to the original measurement for the type of material; for example, a plan for sampling fuel pellets should be equivalent to the original sampling plan. The application of such statistical principles and concepts to inventory verification is discussed in a proposed Regulatory Guide entitled " Statistical Aspects of Prior Measurements Verification."
847 M6 h
REFERENCES 1.
'legulatory Guide 5.13. " Conduct of Nuclear Material Physical Inventories," U.S. Nuclear Regulatory Corrrnission (Nov.1973).
2.
O'Hara, F., et al., " Calorimetry for Safeguards Purposes," Mound Laboratory Report MLM-1798 (Jan. 1972).
3.
Sher, R., " Operating Characteristics of Neutron Well Coincidence Counters," Brookhaven National Laboratory Report BNL-50332 (Jan. 1972).
4.
Gunderson, G., et al., " Enrichment Measuremant in Low Enriched 235U Fuel Pellets,"
Proc.13th Annual INMM Meeting (1972).
5.
- Miller, H., et al., "Results of Typical Nondestructive Measurements of Fissionable Material in Varicus Forms Using NNC Machines," Proc.12th Annual INMM Meeting (1971).
6.
Reilly, T.
D., et al., " Nuclear Safeguards R&D." Los Alamos Scientific Laboratory Report LA 4705-MS(1971), p. 16.
7.
Reilly, T.
D., et al., " Nuclear Safeguards rad," Los Alamos Scientific Laboratory Report LA-4605-MS(1970), p.19.
8.
Reilly, T.
D., et al., " Nuclear Safeguards R&D," Los Alamos Scientific Laboratory Report LA-4605-MS(1970), p. 37.
9.
Gallagher W. J., " Evaluation of ISAS Under Corrnercial Operating Conditions," Brookhaven Office Report BH0-67-4.
10.
Reilly, T.
D., " Assay Experienca with Mcial at Oak Ridge," Proc.13th Annual INMM Meeting (1972).
- 11. Foley, J. E., " Nuclear Safeguards R&D." Los Alamos Scientific Laboratory Report LA-5705-MS(1970), p. 18.
12.
Bramblett, R.
L., et al., " Applications of Photoinduced Reactions to Nuclear Materials Safeguards Problems," General Atomic RT Report Gulf-RT-A-10914 (Dec.1971).
- 13. Beyer, N.
S., et al., "A Portable Dry Calorimeter for the No.7 destructive Assay of Mixed 0xide Fuel Rods," Pyc.13th Annual INMM Meeting (1972).
14.
Beyer, N. S., et al., "Four Passive Assay Techniques Applied to tiixed Oxide Fuel," Argonne National Laboratory Report ANL-1907 (1972).
- 15. Perry, R.
B., et al., " Measurement Reliability of Destructive and Nondestructive. Techniques of Assaying a Reactor Fuel Inventory," Proc. Ilth Annual INMM Meeting (1970).
- 16. Lovett, J.
E., et al., " Plant Instrumentation Program," Third Quarterly Report, Brockhaven Office Report BHO-69-3 (April 1970).
17.
Perry, R.
B., et al., " Nondestructive Assay of ZPPR Fuel," Proc. 9th Annual INMM Meeting (1968).
- 18. Beyer, N. S., et al., " Controlling Large Quantities of Plutonium Fuel," Proc.10th Annual INMM Meeting (1969).
- 19. O'Hara, F.. et al., " Calorimetric Determination of Plutonium in Balk Reactor fiaterials,"
Mound Laboratory (June 1971).
20.
Bishop. D. M., et al., " Plant Instranc.sa ion Program," Third Quarterly Repert, General Electric Company Report GEAP-12114, Oct.-Dec.1970.
~~
- 21. Wadelsamper, D. C., et al., " Integrated Safeguards Experiment," Fourth Quarterly Repor',
General Electric Company Report GEAP-12114-7, April-June 1971.
- 22. Miller, H., " LWR Fuel Rod Nondestructive A.say Measurement Using NNC Production and Portable Systems," Pmc.13th Ar.nual INMM Meeting (1972).
9,i !
- 23. Menlove, H. O., et al., " Nuclear Safeguards R&D," Los Alamos Scientific Laboratory Report LA-4456-MS (1970), p. 26.
- 24. Cline, J.
E., et al., "A Technique for Astay of L-10 Bottles of Plutonium Nitrate," Idaho Nuclear Corporation 'leport IN-1433 (1973).
25.
Cline, J.
E., "A Nondestructive Technique for Assay of Load-out Bottles of Plutonium Nitrate Solution," International Meeting on Nondestructive Measurements and Identification Techniques in Nuclear Safeguards, Italy, Sept.1971.
26.
Limpert, J. L., " Plant Instrumentation Program," Final Guarterly Progress Report, Westing-house Cheswick Report WCAP-7562-4 (May 1971).
- 27. Lawless, J. L., and Chanda, R., "A Plutonium Waste Counter," Trans. Am. Nuc' Soc.13,
- p. 748 (June 1970).
28.
Bramblett. R.
L.,
et al.,
application of Photoinduced Reactions to Nuclear Material Safeguards Problems," General Atomic RT trart Gulf-RT-A 10826 (1971).
29.
Reilly, T.
D., and Parker, J.
L., "A Guiue to Gamma-Ray Assay for Nuclear Material Accountability," Los Alamos Scientific Laboratory Manual LA-579. - (March 1975).
a 84I LVW '
BIBLIOGRAPHY General References American National Standard " Guide to Calibrating Nondestructive Assay Systems," ANSI N15.20-1975.
American National Standard, " Calibration Techniques for the Calorimetric Assay of Plutonium-Bearing Solids Applied to Nuclear Materials Control," ANSI N15.22-1975.
Auguston, R. and Reilly, T., " Fundamentals of Passive Nondestructive Assay of Fissionable Material," Ios Alamos Scientific Laboratory Manual LA-5651-M (Sept.1974).
Bingham, C.
D., et al., " Nondestructive Assay Mea:urements Can Be Traceable," Nuclear Materials Management, Vol. V. No. II, p. 32 (1976).
- Dragnev, T., "Non-Destructive Assay Techniques for Nuclear Safeguards Measurements," Atomic Energy Reviews ll, No. 2, p. 341 (1973).
George, R. S., et al., " Inspector Measurement Verification Activities," _ Proc.16th Annual INMM Meeting (1975).
Kull, L. A., " Catalogue of Nuclear Material Safeguards Instruments," Brookhaven iktional Laboratory Report BNL-17165 (August 1972).
Reilly, T. D., and Evans, M.
L., "Measuren.er.it Reliability for Nuclear Material Assay," Los Alamos Scientific Laboratory Report LA-6574 (1976).
Smith, D. B., and Forster, R. A., "NDA Measurement Error Due to Instrument Calibration," Proc.
15th Annual INMM Meeting (1974).
Gama-Ray _ Spectrometry Cline, J.
E., et al., " Gamma Rays Emitted by the Fissionable Nuclides and Associated 'sotopes,"
Aerojet Nuclear Company Report ANCR-1069 (Supplement To IN-1448) (July 1972).
Cline, J.
E., "Gama Rays Emitted by the Fissionable Nuclides and Associated Isotopes," Idano Nuclear Corporation Report IN-1448 (Jan.1971).
- Dragnev, T., and Scharf, K., " Nondestructive Gamma Spectrometry Measurement of 239Pu/ 40Pu and 2
Pu/ 40Pa Ratios," Int. J. App. Rad. and Isotopes, 26, p.125 (1975).
Gunnink, R., and Tinney, J.
F., " Analysis of Fuel Rods by Gama-Ray Spectroscopy," Lawrence Radiation Laboratory Report UCRL-51086 (Aug.1971)
- Gunnink, R., " Plutonium Isotopic Meaeurements by Gamma-Ray Spectrometry," Proceedings of the Symposium on Calorimetric Assay of Plutonium, Mound Laboratory Report MLM-2177, p. 45 (1973).
Gunnink, R., "A Simulation Study of Plutonium Gamma-Ray Groupings for Isotopic Ration Deter-minations," Lawrence Livermore Laboratory Report UCRL-51605 (1974).
- Gunnink, R., et al., "A System for Plutonium Analysis by Gamma-Ray Spectrometry: Part 1:
Techniques for Analysis of Solutions. Part 2: Computer Programs for Data Reduction and Interpretation," Lawrence Livermore Laboratory Report UCRL-51577, Parts 1 and 2 (1974).
Gunnink, R., " Status of Plutonium Isotopic Measurements by Gamma-Ray Spectrometry," Lawrence Livermore Laboratory Report UCRL-76418 (1975).
Gunnink, R., et al., "A Re-evaluation of the Ga.T.3-Ray Energies and Absolute Branching Intensities of 2 37U, 2 38,239,240,241Pu, and 241Am," L:.wrence Livermore Laborat ry Report UCRL-52139(1976). (This is an update of UCRL-51007,1971.)
Haas, F.
X., "iiondestructive Measurement of Plutonium Isotopic and Americium Concentrations Using Gama-Ray Spectrometry," Proceedings of the Symposium of Calorimetric Assay of Plutonium, Mound Laboratory Report MOM-2177, p. 60 (1973).
<n 847,n.o r Helmer, R. G., et al., "Gama-Ray Energy and Intensity Measurements with Ge(Li) Spectrometers,"
The Electromagnetic Interaction in Nuclear Spectroscopy Chapter 17 North-Holland (1975),
Higinbotham, W., " Passive Gama-Ray Scanning of Plutonium Recycle Fuel Rods," Brookhaven National Laboratory Report BNL-50351 (June 1972).
Kraner, H.
W., "On the Use of Gamma-Ray Spectroscopy To Determine Pu Isotopic Abundances in Plutonium Sources," Brookhaven htional Laboratory Report BNL-50237 (May 1970).
Kull, L. A., et al., " Guidelines for Gama-Ray Spectroscopy Measurements of 2 nU Enrichment,"
Brookhaven N,tional Laboratory Report BNL-50414 (March 1974).
Kull, L.
A., "An Introduction to Ge(L1) and Na! Gamma-Ray Detectors for Safeguards Applications,"
Argonne National Laboratory Report ANL/ACEA-103 (Sept..374).
Reilly T.
D., and Parker, J.
L., "A Guide to Gamma-Ray Assay for Nuclear Material Accountability "
Los Alamos Scientific Laboratory Manual LA-5794-M (March 1975).
Umezawa, H., et al., "Gama-Ray Spectrometric Determination of Isotopic Ratios of Plutonium,"
Japanese Atomic Energy Research Institute Report JAERI-6295 ((1975).
Calorimetry Brumbach, S.
B., et al., " Plutonium Calorimetry and SNM Holdup Measurements," Argonne National Laboratory Report ANL-77-8 (NUREG-0182) (1976).
Ditmars, D.
A., " Precision Measurement by Bunsen Ice Calorimetry of the Total Decay Power of a Plutonium Standard Heat Source," a paper presented at the 30th Annual Calorimetry Conference (1975).
(D. A. Ditmars, National Bureau of Standards Washington, D.C. 20234).
Fellers, C. L. and Seabaugh, P.
W., " Predication of Calorimeter Equilibrium," Proc. 17th Annual
!NMM Meeting (1976).
Gunn, S.
R., " Radiometric Calorimetry: A Review," Nuclear Instruments and Methods, Vol. 29, pp. 1-24 (Sept.1964).
Hansen, L. D., et al., "A Heat Conduction Calurimeter for the Measurement of Solid Heat-Producing Samples," a paper presented at the 31st Annual Calorimetry Conference (1976); (L. D.
Hansen,Tronac,Inc.,Orem, Utah).
O'Hara, F., et al., "The use of Calorimetry in Nuclear Materials Management," Mound Laboratory (June 1970).
O'Hara, F., et al., " Calorimetry for Safeguards Purposes," Mound Laboratory Report MLM-1798 (Jan. 1972).
O'Hara, F., et al., " Calorimetric Determination of Plutonium in Bulk Reactor Materials " Mound Laboratory (June 1971).
Perry, R. B., et al., "An Analytical Model for a Fast Response Calorimeter - With Applications,"
Proc.17th Annual INMM Meeting (1976).
Stephens, H.
P., "A Refrigerant-ll Calorimeter for Measurement of Steady-State Powe, s," a paper presented at the 31st Annual Calorimetry Conference (1976); (H. P. Stephens, Sandia Laboratories, Albuquerque, N.M. 87115).
Strohm, W. W., and Havenstein, M. F., " Proceedings of the Symposium on the Calorimetric Assay of Piulentwa, October 24-25, 1973," Mound Laboratory Report MLM-2177 (Oct. 1974).
Neutron Measurements (Pastive and Act_ig Atwell, T. L., et al., " Assay of Plutanium Metal Buttons with a Portable Neutron Counter,"
Los Alamos Scientific Laboratory Rep:;rt LA-5431-PR, p. 20 (1973).
Bramblett, R.
L., " Fuel Rod Scanner for Quality Control and Safeguards Transaction," IEEE Nuclear Science Symposium (1974).
-847 M Bramblett, R.
L., " Passive and Active Measurement of 5fiM in 55-to 83-Gallon Drums *" Proc.16th Annual INMM Meeting (1975).
Foley, J. E., "4w-Neutron Counter for 55-Gallon Barrels,' Los Alamos Scient'.fic taboratory A-1 Progress Report, LA-4605-MS, p. 25 (1970).
Forster, R.
A., and Menlove, H. D., " Neutron Coincidence Detector for Fuel Pins: Description and Operating Procedures Manual," Los Alamos Scientific Laboratory Manual LA-5156-M (Feb.1973),
Menlov1 1.
0., " Nondestructive Assay of HTGR Fuel Pods," Proc. 16th Annual INMM Meeting (1975).
Menlen H. O., et al., "A Multi-Spectra Neutron Irradiation Technique for the Nondestructive Assay or "issionable Materials," Nucl. Technology 10, p. 366 (1971).
Samp on T.
E., " Neutron Yields from Uraniem Isotopes in Uranturn Hexafluoride," Nucl. Engr.
Sci _. 5_4_, No. 4, p. 470 (1974).
- Sher, R., " Operating Characteristics of Neutron Well Coincidence Counters," Brookhaven National Laboratory Report BNL-503?2 (Jan.1972).
Smith, D.
B., et al., "An Automated Nondestructive Assay System for the Measurement of Irradiated High-Enriched Uranium Fuel," Proc.16th Annual INMM Meeting (1975).
~
73 APPENDIX A MAJOR GAMMA-RAY SICNATURES TOR THE FISSIONABLE ISOTOPES
- Energy Intensity isotope (kev) ja-s)-l Coments 4
U i85.72 4.3 x 10 Only intense gamma ray. Resolved with Nal as well as Ge(Li). Useful for enrichment and quantitative measurements. Several much weaker peaks are seldom useful.
238 2
U 1001.10 1.0 x 10 These actually arise from the 234mPa daughter 766.40 3.9 x 101 of 23eU. After chemical separation, about 100 days are required for the activity to come into equilibrium at the levels stated. Plutonium-238 gives rise to the same 766.40-kev gama and would produce interference in U Pu mixtures. Useful for work viith Ge(L1) or Na!.
5 Pu 766.40 1.5 x 10 Most useful for quantitative assay. Ge(Li) or Na!.
152.77 6.5 x 10 Useful for isotopic determinations with Ge(L1).
4 Pu 413.69 3.4 x 10 The 413.69 usually provides the basis for Ge(Li) assays. The 413.69 plus the 375.02 and its weak neighbors fom a complex upon which Na! assays are based.
5 239Pu 129.28 1.4 x 10 Useful for isotopic determinations with Ge(Li).
Plutonium-239 has over 100 gama rays, some of which are useful for careful work with Ge(Li).
2O Several weak gamma rays, but all suffer bad inter-Pu ference from gammas of other isotopes. Requires very careful war.k with high-resolution detector to make use of any of them.
241 7
Pc 207.98 2.0 x 10 Actually from 237U daughter and requires about 25 days after chemical separation to come into equilibrium at stated value. May also have a few percent inter-ference from 241Am, which emits same gamma. Never-theless a good, clean stronc ga m useful with both Na! and Ge(L1).
6 164.59 1.8 x 10 Useful with Ge(Li). Also from 237U.
6 148.60 7.5 x 10 Useful with Ge(L1). Direct from 241Pu.
10 241 Am 59.54 4.6 x 10 Very stron gamma but attenuation problems. Useful with Ge(Li or Nal. Has several other much less intense gammas sometimes useful for Ge(L1) work.
242 No useful gamma rays at all. Nature failed us at Pu this point.
M f
This table is taken from Reference 29.
APPENDIX B ASSIONED HALF-LIVES AND SPECIFIC POWERS OF PLUT0NIUM ISOTOPES AND AMERICIUM-241 Half-Life
- Oecay Constant Specific Power
- Radionuclide (yea rs)
(day-1)
(mW/g)
Pu-238 87.79 +
0.08 (0.09%)
2.1617 x 10-5 56?.16 1 0.57 (0.10%)
Pu-239 24 082 1 46 (0.19%)
7.830 x 10-8 1.4293 1 0.0053 (0.21%)
Pu-240 6 537 1 10 (0.2%)
2.903 x 10-7 7.098 1 0.015 (0.2%)
Pu-241 14.35 1 0.02 (0.14%)
1.322 x 10-4 3.390 1 0.002 (0.06%)
Ar 241 434.1 1 0.5 (0.14%)
4.372 x 10-6 114.23 1 ).16 (0.14%)
In the table, the value of the constant is reported 1 the standard deviation. The standard deviation relative to the mean, in percent, is given in parentheses.
847 M APPEN0!X C NEUTRON EMISSION RATES OF SOME NUCLEAR MATERIALS Neutron Emission Rate (n.g-1.s-1 (a.n)
Isotope Spontaneous Fissions Reactions U-235 7.6 x 10-4 U-238 1.7 x 10-2 3
Pu-238 2.6 x 10 Pu-239 3.0 x 10-2 3
Pu-240 1.0 x 10 3
Pu-242
- 1. 7 x 10 7
Cm-242 2.3 x 10 3
(Pu-239) F 4.3 x 10 4
4 (Pu-238)0 1.4 x 10 2
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