Regulatory Guide 5.21

(Task SG 044-4), Revision 1, Nondestructive Uranium-235 Enrichment Assay by Gamma Ray Spectrometry

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Issue date:	12/31/1983
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RG-5.21, Rev 1
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Revision 1 December 1983 U.S. NUCLEAR REGULATORY COMMISSION

REGULATORY GUIDE

OFFICEOF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 521 (Task SG 0444)

NONDESTRUCTIVE URANIUM-235 ENRICHMENT ASSAY

BY GAMMA RAY SPECTROMETRY

A. INTRODUCTION

1. BASIS FOR GAMMK.RAY MEASUREMENT OF URA

Section 70.51, "Material Balance, Inventory, and Records Requirements," of 10 CFR Part 70, "Domestic Licensing of NIUM ENRICHMENT

Special Nuclear Material," requires, in part, that licensees 31 The alpha decay of 2 3 5 U to 2 Th Is accompanied by authorized to possess and use at any one time more than the emission of a prominent gamma ray at 185.7 keV

one effective kilogram of special nuclear material (SNM)

(4.3 x 104 of these 185.7-keV gamma rays are emitted per determine the inventory difference (ID) and its associated second per gram of 2 3 5 U). The relatively low energy and standard error (estimator) of inventory difference (SEID)

consequent low penetrating power of these gamma rays for each element and the fissile isotope for uranium con implies that most of the rays that are emitted in the tained in material in process. Such a determination is to be interior of the sample are absorbed within the material based on measurements of the quantity of the element and Itself. Thick2 materials therefore exhibit a 185.7-keV

of the fissile isotope for uranium.

gamma ray emission characteristic of an infinite medium;

The majority of measurement techniques used in SNM Le., the 185.7-keV gamma flux emitted from the sample surface does not depend upon the size or dimensions of accountability are specific to either the element or the the material. Under these conditions the 185.7-keV

isotope but not to both. A combination of techniques Is intensity Is directly proportional to the U enrichment.

therefore required to determine the ID and SEID by element and by fissile isotope for uranium. Passive gamma ray A measure of this 185.7-keV intensity with a suitable

.2 detector forms the basis for an enrichment measurement spectrometry is a nondestructive method for measuring the technique.

enrichment

235 or relative concentration of the fissile isotope U in uranium, but this technique is used in conjunction The thickness of the material with respect to the mean with an assay for the element uranium in order to deter free path of the 185.7-keV gamma ray is the primary mine the amount of 235 U.

characteristic that determines the applicability of passive gamma ray spectrometry for the measurement of isotope This guide describes conditions for 235U enrichment enrichment. The measurement technique is applicable measurements using gamma ray spectrometry that are only If the material Is thick. However, in addition to the acceptable to the NRC staff and provides procedures for thickness of the material, other conditions must be operation, calibration, error analysis, and measurement control.' Examples of 2 3SU enrichment assays using port satisfied before the gamma ray measurement technique able and in-line instruments based on the techniques out can be accurately applied. An approximate analytical expression for the detected 185.7-keV activity is given lined in this guide may be found in References 1 through 4.

below. This expression has been separated into several indi Any guidance in this document related to information vidual terms to aid in identifying those parameters that may interfere with the measurement. Although approximate, collection activities has been cleared under OMB Clearance No. 3150-0009. this relationship can be used to estimate the magnitude of

2 The terms "thick" and "thin" are used throughout this guide to refer to distances in relation to the mean free path of the. I5.7-keV

Calibration error analysis, and measurement control are dis gamma ray in the material under consideration. The mean free path cussed in Regulatory Guide 5.53, "Qualification, Calibration, and Isthe I/e-foldlng distance of the gamma ray flux or, in other terms, Error Estimation Methods for Nondestructive Assay." A proposed the average distance a gamma ray traverses before Interacting.

revision to this guide has been issued for comment as Task SG 049-4.

USNRC REGULATORY GUIDES Comments should be sent to the Sectetary of the Commission, U.S. Nuclear Regulatory Commission. Washington, D.C. 20555, Regulatory Guides are Issued to describe and make available to the Attention: Docketing and Service Branch.

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interfering effects in order to establish limits on the range of applicability and to determine the associated uncer material with a characteristic length called the critical tainties introduced into the measurement. This relationship distance xo, where x is defined as the thickness of material that produces 99.5 percent of the measured 185.7-keV

is:

activity:

effective source of 185.7-keV

gamma raysfseen by the detector x0 = -A ln(0.005) = S.29A (2)

where

? i""=( -..

u A +W£ (Q/47r) exp(-pclicd)

uuj, I ,,a A (1)

'/A= 1UPU + z Pipi (3)

I

enrich- physical riea] otrica container ment constants Composi- efficiency absorption don I Calculated values of xo for several common materials are given in Table 1.

area defined detector by collimator efficiency Where Table 1I

CALCULATED VALUES OF x AND MATERIAL

C = detected 185.7-keV activity COMPOSITION fERM

E = enrichment of the uranium (<I)

Material Com PU,Pi, PC= density of.the uranium (U), matrix material position Term (i), and container wall (c), respectively, in Critical g/cm 3 Density Distance Material (g/cm 3 ) x 0 (cm) 1

11U'

I

PC = mass attenuation coefficient for 185.7-keV

gamma rays in uranium (U), matrix material U (metal) 18.7 0.20 1.000

(i), and container wall (c) in cm 2 /g UF

6

4.7 1.08 1.040

a = specific 185.7-keV gamma ray activity of UO 10.9 0.37 1.012

23 UA 8 7.3 0.56 sU

= 4.3 x 104 gamma rays/sec-g Uranyl. Nitrate 2.8 2.30

1.015

1.09S K.

.Values of the mass attenuation

= net absolute detector full energy peak effi coefficient, pa, may be found References 6 and 7. in ciency for detecting 185.7-keV gamma rays

(< 1) Other nondestructive assay (NDA) techniques are capable of detecting SNM distributed within a containe

r. The enrich

= solidanglesubtendedby the detector(SI < 2ir)

ment measurement technique, however, is inherently a surface measurement. Therefore, the "sample" observed, A = cross-sectional area of material defined by I.e., the surface, must be representative of all the material in the detector collimator the container. In this respect the enrichment measurement is more analogous to chemical analysis than are other NDA

d = container wall thickness techniques.

A derivation of this expression, as well as other necessary background information on the theory of enrichment mea

2.2 Material Composition surements, may be found in Reference 5. As evident in Equation 1, the activity (C) is proportional to the enrich If the gamma ray measurement is to be dependent only ment (E) but is affected by sqveral other characteristics as on the enrichment, the term related to the composition of the matrix should be approximately equal to one, Le.,

well.

2. MATERIAL AND CONTAINER WALL EFFECTS ON Pi ~

MEASUREMENT (4)

2.1 Material Thickness This condition ensures that the enrichment measurement will be insensitive to variations in the matrix composition.

In order for Equation 1 to be applicable, the material However, if this matrix term differs significantly from must be sufficiently thick to produce strong attenuation of

185.7-keV gamma rays. To determine whether this criterion unity, the enrichment measurement can still be performed provided the matrix composition of the standard and is met, it is useful to compare the actual thickness of the samples remains reasonably constant.

5.21-2

to account for attenuation of the 185.7-keV gamma rays Calculated values of this quantity for common materials (see Equation 5). Commercial equipment is available to are given in Table 1. The deviations of the numbers in measure wall thicknesses ranging from about 0.025 to Table I from unity indicate that a bias can be introduced 5.0 cm with relative precisions of approximately 1.0 per

> by ignoring the difference in material composition. cent to 0.1 percent, respectively.

Inhomogeneities in matrix material composition, uranium Using standardized containers to hold the sample mate density, and uranium enrichment within the measured rial in order to minimize uncertainties and possible errors volume of the material (as characterized by the depth xo associated with container-to-container wall thickness and the collimated area A) can produce changes in the corrections is strongly recommended.

measured 185.7-keV activity and affect the accuracy of an enrichment calculated on the basis of that activity. Varia 3. DETECTOR-RELATED FACTORS

tions in the content of low-atomic-number (Z < 30) matrix materials and inhomogeneities in uranium density in such 3.1 Area and Geometrical Efficiency matrix material produce a small to negligible effect on measurement accuracy. Care is necessary, however, in The area of the material viewed by the detector and the applying this technique to materials having high-atomic geometrical efficiency are variables that may be adjusted, number matrices (Z > 50) or materials having uranium within limits, to optimize a system. Two important factors concentrations less than approximately 75 percent. Signifi must be noted:

cant inaccuracies can arise when the uranium enrichment itself varies throughout the sample. 1. Once these variables are fixed, changes in these parameters will alter the calibration of the instrument and The above conclusions about the effects of inhomogene invalidate subsequent measurement results.

ities are based on the assumption that the thickness of the material exceeds the critical distance, xo, and that the 2. The placement of the material within the container inhomogeneities exist within this depth. In the case of will affect the detected activity. It is important that there extremely inhomogeneous materials such as scrap, the are no void spaces between the material and the container condition of sufficient depth may not always be fulfilled or wall.

inhomogeneities may exist beyond the depth xo; i.e., the sample is not representative. Therefore, this technique 3.2 Net Detector Efficiency is not applicable to such inhomogeneous materials.

Thallium-activated sodium iodide, Nal(TI), lithium-drifted

2J 2.3 Container Wall Thickness germanium, Ge(Li), and high-purity germanium, HPGe (also referred to as intrinsic germanium, IG), detectors have been Variations in the thickness of the container walls can used to perform these measurements. The detection systems significantly affect the activity measured by the detector.

are generally conventional gamma ray spectrometry systems The fractional change in the activity AC/C due to a small that are commercially available in modular or single-unit change Ad in the container wall thickpess can be expressed:

construction. Some useful guidelines for the procurement

(5) and setup of a solid-state-detector-based system are given in AC =_*lPcPcAd Regulatory Guide 5.9, "Specifications for Ge(Li) Spectros3 copy Systems for Material Protection Measurements."

Calculated values of AC/C corresponding to a change in Factors that influence detector selection and the control container thickness Ad of 0.0025 cm for common con required for accurate results are discussed below.

tainer materials are given in Table 2.

Table 2 3.2.1 Background CALCULATED VALUES OF AC/C 3.2.1.1 Compton Background. This background is pre dominantly produced by the 765-keV and lO01-keV

2 4 2 38 AC gamma rays of 3 mPa, a daughter of U. Since in most Density cases the Compton background behaves smoothly in Material (g/cm*) C

the vicinity of the 185.7-keV peak, it can be readily sub;

Steel 7.8 -0.003 tracted, leaving only the net counts in the 185.7-keV

Aluminum 2.7 -0.0009 full-energy peak.

Polyethylene 0.95 - 0.0004

3.2.1.2 Overlapping Peaks. The observable peak from certain gamma rays may overlap that of the 185.7-keV peak Therefore, the container wall thickness must be known owing to the finite energy resolution of the detector; i.e.,

(e~g., by measuring an adequate number of the containers

3 J before loading). In some cases, an unknown container wall A proposed revision to this guide has been issued for comment Spectros as Task SG 042-2 with the title "Guidelines for Germanium Material."

thickness can be measured using an ultrasonic technique copy Systems for Measurement of Special Nuclear after which a simple correction can be applied to the data

5.21-3

the difference in energies may be less than twice the full 1-mm-thick cadmium filter will reduce x-ray interference, width of the spectrum peak at half its maximum height eliminating this source of count-rate losses. Note that (FWHM). This problem is common in enrichment measure present-day counting electronics are capable of handling ments of recently separated uranium from a reprocessing high negative count rates without significant losses from

3lant. The peak from a strong 208-keV gamma ray from either pileup or system dead time. However, if a measure

37U (half-life of 6.75 days) can overlap the 185.7-keV

ment situation arises in which count rates are excessive, peak when a Nal detector is used. Analytical separation of tighter collimation of the opening on the front face of the the two unresolved peaks, i.e., peak stripping, may be detector is a simple method for reducing count rates to applied. An alternative solution is to use a Ge(Li) or HPGe tolerable levels at which complicated loss corrections are detector so that both peaks are dearly resolved. The2 3U not essential.

activity present in reprocessed uranium will depend on the amount of 241pu present before reprocessing and also on 3.2.3 Instability in Detector Electronics the time elapsed since separation.

The gain of a photomultiplier tube is sensitive to changes

3.2.1.3 Ambient Background. The third source of in temperature, count rate, and magnetic field. Provision background originates from natural sources and from other can be made for gain checks or gain stabilization for enrich uranium-bearing materials located in the vicinity of the ment measurement applications. Various gain stabilizers measuring apparatus. This source can be particularly that automatically adjust the system gain to keep a refer bothersome since it can vary over time within wide limits ence peak centered between two preset energy limits are depending on plant operating conditions. available.

3.2.2 Count-Rate Losses . REGULATORY POSITION

Calculation of the detector count rates for purposes of Passive gamma ray spectrometry constitutes a means making dead-time 4 estimates requires calculation of the acceptable to the NRC staff for nondestructively determin total count rate, not only that due to 2 3 1U. Total count ing U enrichment, if the conditions identified below are rate estimates for low-enrichment material must therefore satisfied.

take into account the relatively important backgrounds of gamma rays from 238V daughters. If other radioactive I. RANGE OF APPLICATION

materials are present within the sample, their contributions to the total count rate must also be considered. All material to be assayed under a certain calibration Count-rate corrections can be made by determining the should be of similar chemical form, physical form, homo I'l

4 geneity, and impurity level.

dead time or by making measurements for known live-time intervals. The pileup or overlap of electronic pulses is a The critical distance o&f the material should be determined.

problem that also results in a loss of counts in the full Only those items of the material having dimensions greater energy peak for Ge(Li) systems. An electronic pulser may than this critical distance should be assayed by this technique.

be used to monitor and correct for these losses. However, a more reliable method involves the use of a radioactive The material should be homogeneous in all respects on a source fixed to the detector in an invariant geometry. macroscopics scale. The material should be homogeneous A photopeak area from the spectrum of this source is with respect to uranium enrichment on a microscopics counted along with a uranium peak area. The source peak scale.

area can then be compared with an earlier value taken without uranium present, and the dead time for the assay The containers should all be of similar size, geometry, measurement can be inferred. (Part of the regular measure and physical and chemical composition.

ment control would then involve uranium-free measurement of

241 the source peak area.) One possible source could be

2. SYSTEM REQUIREMENTS

Am, whose 60-keV gamma ray peak would be easily resolved from the uranium lines by either a Ge- or Nal-based NaI(TI) scintillation detectors having a resolution of system. If filtering of ambient low-energy gamma radiation FWHM less than 16 percent at the 185.7-keV peak of 2 3 5 U

is used, the 24 1 Am source can be placed between the are generally adequate for measuring the enrichment of detector and the absorber used for the filtering. If a high uranium. Crystals with a thickness in the range of 1.3 to resolution system is used, the recommended source for 1.8 cm are recommended for optimum efficiency. If other this purpose is 10 9 Cd, which emits only an 88-keV peak, radionuclides that emit significant quantities of gamma well below the uranium (185.7-keV) region, and has a radiation in an energy region E = 185.7 keV +/- 2 FWHM at half-life of 453 days. Radiation that provides no useful 185.7 keV are present, one of the following should be used:

information can be selectively attenuated by filters; e.g., a

4

"Dead time" refers to that portion of the measurement period a. A higher resolution detector, e.g., Ge(Li) or HPGe, or during which the instrument Is busy processing data already received and cannot accept new data. "Live time" means that portion of the measurement period during which the instrument can record detected K

events. To compare different data for which dead times are appreci 5 lMacroscople refers to distances greater than the critical distance;

able, compare counts measured for equal live-time periods, Le.,

(actual measurement period) - (dead time) = live time. microscopic to distances less than thi critical distance.

5.214

Calibration) should be determined and the position of b. A peak-stripping procedure to subtract the interfer ence. In this case', data 'should be provided to show the the 185.7-keV peak and neighboring peaks noted. The range of concentration' of the interfering radionuclide and threshold and width of each energy region should then be the accuracy and precision of the stripping technique over selected to avoid including any neighboring peaks and to this range. optimize the system stability and the signal-to-background ratio.

The detection system gain should be stabilized by The net response attributed to 185.7-keV gamma rays monitoring a known reference peak.

should be the accufnulated counts in the peak region minus The system clock should be in live time. The system a multiple of the counts accumulated in a nearby back should provide a means of determining the count-rate losses ground region. A single upper background region may be based on the total counting rate, or provide additional monitored, or both a region above the peak region and one collimation to reduce the count rate. below may be monitored. If only an upper background region is monitored, the net response, R, is giyen by The design of the system should allow reproducible positioning of the detector or item being assayed. R - G abB.

The system should be capable of determining the gamma where G and B are the gross counts in the peak region and ray activity in at least two energy regions to allow subtrac the background region, respectively, and b is the multiple tion of the background. One region should encompass of the background to be subtracted. This net response, R,

185.7 keV, and the other should be above this region but should then be proportional to the enrichment, E:

should not overlap it. The threshold and, width of the regions should be adjustable. If dead-time corrections are E =CIR - C(G - bB)

measured with a pulser or source peak, a third and fourth region will have to be defined to establish the additional where C1 is a calibration constant to be determined (see peak area and its background. Regulatory Position 4, Calibration). The gross counts, G

and B, should be measured for all the standards. The The system should have provision for filtering out quantities G/E should then be plotted as a function of the low-energy radiation from external sources. quantities B/E anda straight line through the data determined:

3. DATA ACQUISITION G/E =b(B/E) + I/C1 Initial preparation of the assay instrumentation for data The slope of this line is b, the multiple of the upper back acquisition should involve careful determination of the ground region to be subtracted. The data from all the system energy gain, the position of key photopeak and standards should be used in determining this slope.

background regions, and the instrument response to cali bration. However, after the proper instrument settings are If both an upper and a lower background are monitored, established, routine operation can involve a less detailed the counts in each of these regions should be used to check of the peak positions. This verification can consist of determine a straight-line fit to the background. Using this either a visual check of the gamma ray spectrum on a straight-line approximation, the area or number of counts multichannel analyzer or a brief scan of the 140- to 200-keV under this line in the peak region should be subtracted from energy region with a single-channel analyzer. Verification the gross counts, G, to obtain the net response. An adequate that the 185.7-keV peak position correspondi to its~value at, technique based on this principle Is described in Reference 8.

calibration ensures that the instrument is still biased properly. On a number of recently developed portable gamma ray Verification of the 185.7-keV count rate with a uranium spectroscopy instruments, these calibration procedures can check source can also demonstrate continued validity of the be performed automatically by means of a microprocessor response calibration. In some cases it may be useful to based computational capability built into the instrument or check the position of two peaks in the tammanray spectrum, by a calculator. In such cases, the more reliable procedure in which case a 5 7Co gamma ray source (with a photopeak of complete calibration of the instrument before each assay at 122 keV) would be convenient. session may be practical.

If the total counting rate is determined primarily by the

4. CALIBRATION

185.7-keY gamma ray, the counting rate should be restricted (e.g., by absorbers or decreased geometrical efficiency) Calibýation 6 standards should be obtained by:

below those rates requiring correction. The system sensitivity will be reduced by these measures, and, if the sensitivity is 1. Selecting items from the production material. A

no longer adequate, separate calibrations should be made in group of the items selected should, after determination of two or more enrichment regions.

6 To determine the location and width of the 185.7-keV 'None of the calibration techniques or data reduction procedures peak region and the background regions, the energy spectrum discussed precludes the use of automated direct-readout systems for operation. The procedures described In this guide should be used for from each calibration standard (see Regulatory Position 4, adjustment and calibration of direct-readout instruments.

5.21-5

the gamma ray response, be measured by an independent, S. OPERATIONS

more accurate technique, e.g., mass spectrometry, that is traceable to or calibrated with National Bureau of Standards . The measurement of enrichment involves counting the (NBS) standard reference material. The other items should 185.7-keV gamma ray intensity from an infinite thickness be retained as working standards. of uranium-bearing material in a constant counting geometry.

A schematic of the counting geometry is given in Figure 1.

2. Fabricating standards that represent the material to The detector should be collimated and shielded from be assayed in chemical form, physical form, and impurity ambient radiation so that, as much as possible, only the level. The 235U enrichment of the material used in the radiation from the sample container is detected.

fabrication of the standards should be determined by a technique, e.g., mass spectrometry, that is traceable to or The detection system and counting geometry (i.e.,

calibrated with NBS standard reference material. collimator opening area, A, and collimator depth, x), the data reduction technique, and the count-rate loss corrections, The containers for the standards should have a geometry, if included, should be Identical to those used in the calibration.

dimensions, and a composition that approximate the mean of these parameters in the containers to be assayed. However, Data from all measurements should be recorded in an it should be emphasized that the best procedure is to appropriate log book.

standardize the, sample containers to minimize, if not eliminate, container-to-container differences. At least two working standards should be measured during each eight-hour operating shif

t. The measured

3. The values of enrichment for the calibration standards response should be compared to the expected response should span the range of values encountered in normal (value used in calibration) to determine if the difference

7 operation. No less than three separate standards should be exceeds three times the expected standard deviation. If used. (Good calibration practice dictates the use of at least this threshold is exceeded, measurements should be repeated two standards to determine the linear calibration constants to verify that the response is significantly different and that and a third standard to check the calibration computations.) the system should be recalibrated. In the event of a significant However, if the assay response (after application of appro change in the instrument response, every effort should be priate corrections) can be shown to be highly linear and to made to understand the underlying cause of the change have zero offset (i.e., zero response for zero enrichment), and, if possible, to remedy the cause rather than simply it may be more advantageous to avoid using standards with calibrate around the problem.

low enrichment because the low count rates would reduce the calibration precision. In such a case, calibration in the Prior to counting, all containers should be agitated. If upper half of the range of expected enrichments combined this is not possible, the material should be mixed by some with the constraint of zero response for zero enrichment method. One container from every ten should be measured can produce a higher precision calibration than a fitting of at two different locations on the container. The others may K

standard responses over the full range of expected enrich be measured at only one location. (If containers are scanned ments, including values at low enrichment. If such a cali to obtain an average enrichment, the degree of inhomogeneity bration procedure is used, careful initial establishment of should still be measured by this method.)

the zero offset and instrument linearity, followed by occasional verification of both assumptions, is strongly The difference between the measurements at different recommended. Such verification could be accomplished by locations on the container should be used to indicate a lack an occasional extended measurement of a low-enrichment of the expected homogeneity. If the two responses differ standard. It should be noted that if the measurement by more than three times the expected standard devia-.

system exhibits a nonzero offset (i.e., a nonzero response tion (which should include the effects of the usual or for zero sample enrichment), this is an indication of a expected inhomogeneity), measurements should be repeated background problem that should be corrected before assays to verify the existencen of an abnormal inhomogeneity. If are performed. the threshold is exceeded, the container should be rejected and investigated to determine the cause of the abnormal Each standard should be measured at a number of inhomogeneity.8 different locations, e.g., for a cylinder, at different heights and rotations about the axis. The mean of these values The container should be viewed at such a position that should be used as the response for that enrichment. The an infinite thickness of material fills the field of view dispersion in these values should be used as an initial defined by the collimator and detector (see Figure 1). The estimate of the variance due to material and container procedure for determining the fill of the container should inhomogeneity. be recorded, e.g., by visually inspecting at the time of filling and recording on the container tag.

In general, the data from the standards, i.e., the net responses attributed to the 185.7-keV gamma rays from the known uranium enrichments, can be employed in a simple 7 The user can always have a stricter criterion. This is a minimum.

linear calculation of the two calibration constants as described in Appendix 3 of Reference 5. If desired, more SThe difference may also be due to a large variation in wall involved least-squares techniques can also be used. thickness.

5.21-6

SCHEMATIC OF ENRICHMENT MEASUREMENT

SETUP

(

FIGURE 1 A schematic of a typical detector/collimator arrangement for a uranium enrichment measurement. The collimator depth (crucial in the calibration of the enrichment instrument) is denoted by x, the distance from the container surface to the collimator opening by r, and the container wall thickness by d. As long as an infinite thickness of assay material is contained

2 in the field of view of the detector, the distance r is not crucial. However, the preferred enrichment measurement setup is with the collimator opening in contact with the container surface (i.e., r = 0).

5.21-7

The container wall thickness should be measured. The The measurement-to-measurement variance should be wall thickness and location of the measurement should be determined by periodically observing the net response indicated if the individual wall thickness measurements and from the standards and repeating measurements on selected the gamma ray measurement are made at this location. If process items. Each repeated measurement should be made the containers are nominally identical, an adequate sampling of these containers should be sufficient. The mean of the at a different location on the container surface, at different times of the day, and under different ambient conditions.9 K

measurements on these samples constitutes an acceptable The standard deviation should be determined and any measured value of the wall thickness that may be applied to trends (e.g., trends due to time or temperature) corrected all containers of this type or category.

for.

The energy spectrum from a process item selected at The item-to-item variance due to the variation in wall random should be used to determine the existence of thickness should be determined. The variance in the con unexpected interfering radiations and the approximate tainer wall thickness should be determined from measure magnitude of the interference. This test should be per ments of the sample container wall thickness, either during formed at a frequency that will ensure testing: the course of the assays or from separate measurements of randomly selected samples. The computed variance in the

1. At least one item in any new batch of material.

samples should be used as the variance of wall thickness.

This variance should be multiplied by the effect of a unit

2. At least one item if any changes in the material variation in that thickness on the measured 185.7-keV (see, processing occur. e.g., Table 2) response to determine its contribution to the total measurement variance.

3. At least one item per two-month period.

Item-to-item variations other than those measured, e.g.,

If an interference appears, either a higher resolution wall thickness, should be determined by periodically (see detector should be acquired or an adequate peak-stripping guidelines in Regulatory Position 5) selecting an item and routine applied. In both cases, additional standards that determining the enrichment by an independent technique include the interfering radiations should be selected and the traceable to, or calibrated with, NBS standard reference system should be recalibrated. material. A recommended approach is to adequately sample and determine the 2 3SU enrichment by calibrated mass No item should, be assayed if the measured response spectrometry. In addition to estimating the standard devia exceeds that of the highest enrichment' standard by more tion of these comparative measurements, the data can also than twice the standard deviation in the response from this be used to verify the continued stability of the instrument standard.

calibration. If any significant deviation of the calibration is noted from these comparisons, the cause of the change K

6. ERROR ANALYSIS,

should be identified before further assays are performed.

A regression or analysis-of-variance technique should be used to determine the uncertainty in the calibration con 9 The variance due to counting (including background) and variance due to lnhomogenelty, ambient conditions, etc., will the be stants. included In this measurement-to-measurement variance.

K1

5.21-8

REFERENCES

1. R. B. Walton et al., "Measurements of UF 6 Cylinders 5. L. A. Kull, "Guidelines for Gamma-Ray Spectroscopy with Portable Instruments," Nuclear Technology, Vol. 21, Measurements of 2 3 sU Enrichment," Brookhaven p. 133, 1974. National Laboratory, BNL-50414, March 1974.

2. T. D. Reilly et al., "A Continuous In-Line Monitor for 6. J. H. Hubbell, "Photon Cross Sections, Attenuatim UF Enrichment," Nuclear Technology, Vol. 23, p. 318, Coefficients, and Energy Absorption Coefficients from

19A4. 10 keV to 100 GeV," National Bureau of Standards, NSRDS-NBS 29, 1969.

3. P. Matussek and H. Ottmar, "Gamma-Ray Spectrom etry for In-Line Measurements of 2 3 5 U Enrichment 7. E. Storm and H. I. Israel, "Photon Cross Sections from in a Nuclear Fuel Fabrication Plant," in Safeguarding .001 to 100 MeV.for Elements I through 100," Los NuclearMaterials, IAEA-SM-201/46, pp.223-233, 1976. Alamos Scientific Laboratory, LA-3753, 1967.

Available from the International Atomic Energy Agency, UNIPUB, Inc., P.O. Box 433, New York, New York

10016. 8. G. Gunderson and M. Zucker, "Enrichment Measure ment in Low Enriched 2 3 SU Fuel Pellets," in "Proceed

4. R. B. Walton, "The Feasibility of Nondestructive Assay ings: 13th Annual Meeting," Journal of the Institute Measurements in Uranium Enrichment Plants," Los of Nuclear Materials Management, Vol. 1, No. 3, p. 221, Alamos Scientific Laboratory, LA-7212-MS, 1978. 1972.

BIBLIOGRAPHY

Alvar, K., H. Lukens, and N. Lurie, "Standard Containers This report contains a wealth of information on for SNM Storage, Transfer, and Measurement," U.S. nondestructive assay techniques and their asso Nuclear Regulatory Commission, NUREG/CR-1847, 1980. ciated instrumentation and has an extensive Available through the NRC/GPd Sales Program, U.S. treatise on gamma ray enrichment measurements.

Nuclear Regulatory Commission, Washington, D.C. 20555.

This report describes the variations of container Sher, R., and S. Untermeyer, "The Detection of Fission properties (especially wall thicknesses) and their able Materials by Nondestructive Means," American Nuclear effects on NDA measurements. A candidate list Society Monograph, La Grange Park, Illinois, 1980.

of standard containers, each sufficiently uniform to cause less than 0.2 percent variation in assay results, is given, along with comments on the This 1Iook contains a helpful overview of a wide value and impact of container standardization. variety of nondestructive assay techniques, including enrichment measurement by gamma ray Augustson, R. H., and T. D. Reilly, "Fundamentals of spectrometry. In addition, it contains a rather Passive Nondestructive Assay of Fissionable, !Material," extensive discussion of error estimation, measure Los Alamos Scientific Laboratory, LA-5651-M, Albuquerque, ment control techniques, and measurement New Mexico, 1974. statistics.

5.21-9

VALUE/IMPACT STATEMENT

1. PROPOSED ACTION 1.3.4 Public

1.1 Description No impact on the public can be foreseen.

Licensees authorized to possess at any one time more 1.4 Decision on Proposed Action than one effective kilogram of special nuclear material (SNM) are required in § 70.51 of 10CFR Part 70 to The guide should be revised to reflect improvements in determine the inventory difference (ID) and the associated technique and to bring the guide. into conformity with standard error (SEID) for each element and the fissile current usage.

isotope of uranium contained in material in process. The determination is made by measuring the quantity of the element and of the fissile isotope for uraniu

2. TECHNICAL APPROACH

It is not usually possible to determine both element Not applicable.

and isotope with one measurement. Therefore, a combina tion of techniques is required to measure the SNM ID and the SEID by element and by fissile isotope. Passive gamma

3. PROCEDURAL APPROACH

ray spectroscopy is a nondestructive method for measuring the relative concentration of the fissile isotope 2 3 5 U in Of the alternative procedures considered, revision of uranium. This technique is then used in conjunction with the existing regulatory guide was selected as the most an assay for the element uranium to determine the amount advantageous and cost effective.

of 2 3 5 U.

Regulatory Guide 5.21 describes conditions for 23SU 4. STATUTORY CONSIDERATIONS

enrichment measurements using gamma ray spectroscopy that are acceptable to the NRC staff. The proposed action 4.1 NRC Authority will revise the guide to conform to current usage and to add information on the state of the art of this technique. Authority for the proposed action is derived from the Atomic Energy Act of 1954, as amended, and the Energy

1.2 Need Reorganization Act of 1974, as amended, and implemented through the Commission's regulations.

The proposed action is needed to bring Regulatory Guide 5.21 up to date.

4.2 Need for NEPA Assessment

1.3 Value/Impact Assessment The proposed action is not a major action that may

1.3.1 NRC Operations significantly affect the quality of the human environment and does not require an environmental impact statement.

The experience and improvements in technology that have occurred since the guide was issued will be made available for use in the regulatory process. Using these S. RELATIONSHIP TO OTHER EXISTING OR

updated techniques should have no adverse impact. PROPOSED REGULATIONS OR POLICIES

1.3.2 Other Government Agencies The proposed action is one of a series of revisions of existing regulatory guides on nondestructive assay Not applicable. techniques.

1.3.3 Industry

6. SUMMARY.AND CONCLUSIONS

Since industry is already applying the techniques discussed in the guide, updating these techniques should Regulatory Guide 5.21 should be revised to bring it up have no adverse impact. to date.

5.21-10