ML20064L200

Final Rept EPRI TR-103591, Burnup Verification Measurements on Spent-Fuel Assemblies at Oconee Nuclear Station

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Issue date:	01/31/1994
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Keywords: EPRI TR-103591 Spent fuels Project 3290-07 Waste transport Final Report Electne Power Criticality January 1994 Research Institute Burnup 4

Burnup Verification Measurements on Spent-Fuel Assemblies at Oconee Nuclear Station Prepared by Sandia National Laboratories Albuquerque, New Mexico and Los Alamos National Laboratory Los Alamos, New Mexico 9403240190 940317 PDR ADOCK 05000269 P PDR ,

gqy b::;lr R E P O R T

SUMMARY

Bumup Verification Measurements on Spent-Fuel Assemblies at Oconee Nuclear Station The application of bumup credit to the design of spent fuel casks

- results in significantly reduced costs and risks in the transport and storage of spent fuel assemblies. A measurement system to verify reactor records of spent fuel bumup can permit the bumup credit savings to be realized, This report describes the demonstration of a practical, accurate method of verifying reactor records for the exposure of spent fuel. ,

BACKGROUND NRC regulations require a substantial margin of safety to ensure that spent nuclear fuel cannot reach criticality (support a sustained nuclear reaction)

INTEREST CATEGORIES as a result of unforeseen accident or abnormal shipping conditions. A practical measurement can permit use of realistic properties instead of conservative fresh-Light water rea0 tor fuel fuel properties. Storage, transportation, and disposal designs based on realistic Radioactive waste spent fuel composition (bumup credit calculations) can result in significantly more-management efficient arrays of assemblies, reduce the need for expensive neutron absorbers, and decrease the risk in transporting a given quantity of fuel by permitting higher KEYWORDS payload in each shipment and fewer total shipments. EPRI cosponsored this work with Sandia National Laboratories and Los Alamos National Laboratory to perform bumup venfication measurements applying the fork detector system, used by the Spent fuels Intemational Atomic Energy Agency to venty reactor records for safeguard apph,ea-Waste transport tions. Duke Power Company's Oconee Nuclear Station served as host utility.

g Burnup OBJECTIVES To establish a database using the fork detector system at an oper-ating nuclear utikty; to determine measurement compatibility with utility operating procedures; to develop in operational plan for implernenting verification measure-ments wi*h utihty input.

APPROACH investigatoc used the fork detector system to examine spent-fuel assemblies. First, they measured neutron and gamma ray emissions from individual spent fuel assemblies in the storage pool. Next, they performed tests that demon-strated the abihty of the system to verify reactor records for bumup and cooling times and detect deviations from those records. Finally, they examined 93 assem-

- bhes, measuring bumup variation in two assemblies.

RESULTS The fork detector system measures the passive neutron and gamma-ray emissions from individual spent fual assemblies while in the storage pool. After five years of decay, the predominant neotron emitterin spent fuelis curium 244, formed by successive neutron captur9 btginning with uranium 238. The major gamma emitter after several years o' cool.ng is cesium 137, produced as a fission i product. The shorter lived isotopes of curium and cesium are activation products, l which are insignificant after a few yeert of cookng.

in testing, the fork detector system rraasurements correlated well with the Oconee .

Nuclear Station records. The average dviation of the reactor burnup records from (

l EPRi TR 103591s Electnc Power Research Institute

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th3 cilibration was 10% without corrections for initial anrichmInt and 2.2% ' ,

• with corrections. The derived calibration indicated that the neutron signal )

was proportional to the bumup raised to the 3.81 power. The gamma ray signals were also in general agreement (15%) with the burnup records. ,

Two of the 93 assemblies measured proved anomalous, producing much higher neutron signals than the bumup would explain. In a verification '

campaign, these two assemblies would require further study or be excluded from the acceptable fuel for a bumup credit cask.

EPRI PERSPECTIVE The fork detector system performed quite well and proved relatively easy to set up and operate, it could provide an accept-able means for venfying burnup of fuel assemblies before loading into a i bumup credit cask or canister. Detector system measurements should be  :

used to screen for gross errors in reactor records, such as inadvertent assignment of the bumup of one assembly to another. Ultimate qualifica-tion of a fuel assembly for loading should be based on the verified reactor records for bumup, since the records are likely to have less uncertainty in isotopic composition. j PROJECT RP3290-07 Project Manager: R. F. Williams  !

Nuclear Power Division Contractors: Sandia National Laboratories; Los Alamos National Laboratory -

For further information on EPRI research programs, call EPRI Technical Information Specialists (415) 855 2411.

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Bumup Verification Measurements on Spent-Fuel Assemblies at Oconee Nuclear Station TR 103591 Research Project 3290-07 Final Report, January 1994 Prepared by SANDIA NATIONAL LABORATORIES

• 1515 Eubank S.E.

Albuquerque, New Mexico 87115 Principal investigator R. l. Ewing

'A U.S. Depertment of Energy facility, operated under contract DE AC04 94ALC000 by Martin Marietta Corporation.

ALAMOS NATIONAL LABORATORY" -

Post Office Box 808 SM-30 Bikini Atoll Road Los Alamos, New Mexico 87545 Pnncipal Investigators G. E. Besler R. Siebehst "A U.S. Department of Energy facility, operated under contract W 7405 ENG 36 by the University of Califomia.

DUKE POWER COMPANY .

Post Offee Box 1006 Charlotte, North Carolina 272011006 Principal Investigator G. R. Walden Prepared for Electric Power Research Institute 3412 Hillview Avenue Palo Alto, Cahfomia 94304 EPRI Project Manager R. F. Wdhams Fuel Rehabihty Storage and Disposal Program Nuclear Power Division w

DISCLAIMER OF WARRANTIES AND LIMITATION OF UABILITIES TmS REPORT WAS PREPARED BY THE ORGANIZATION (S) NAMED BELOW AS AN ACCOUNT OF WORK SPONSO COSPONSORED By THE ELECTRIC POWER RESEARCH INSTITUTE,INC. (EPRI) NEITHER EPRI. ANY MEMBER OF EPRI, ANY COSPONSOR, THE ORGANIZAtl0N(S) NAMED BELOW, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM.

( Al MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED,(1) WITH RESPECT TO THE USE ANY AFORMATION. APPARATUS. METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS REPORT, INCLUDING MERCHANTABILITY AND FITNESS FOR A PARTICULAR PURPOSE, OR (ll) THAT SUCH USE DOES NOT INFRINGE ON OR INTERFERE WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTYS INTELLECTUAL PROPERTY, OR (Ill) THAT THIS REPORT is $Ulf ABLE TO ANY PARTICULAR USER'S CIRCUMSTANCE; OR (B) ASSUMES RESPONSIBILITY FOR ANY DAMAGES OR OTHER LIABILITY WHATSOEVER (INCLUDING ANY CO DAMAGES. EVEN IF EPRI OR ANY EPRI REPRESENTATIVE HAS BEEN ADVISED OF THE POSSIBILITY OF SUCH DAMAG RESULTING FROM YOUR SELECTION OR USE OF THIS REPORT OR ANY INFORMATION, APPARATUS, METHOD, PROCESS. OR StMILAR ITEM DISCLOSED IN THl3 REPORT, ORGANIZATION (S) THAT PREPARED THIS REPORT:

SANDIA NATION AL LA80RATORIES LOS ALAMOS ET10NAL LABORATORY ORDERING INFORMATION Requests for copes of this repod should be directed to the EPRI Distribution Center,207 Coggins Drive.

P O Box 23205, Pleasant Hill, CA 94523, (510) 934-4212. There is no charge for reports requested by EPRI rnember utilmes.

Etectnc Power Research institute and EPRI are registered sehnce marks of Electnc Power Research Institute,Inc.

Copynght C 1994 Electnc Power Research Insttute, Inc. AE nghts reserved.

ABSTRACT The FORK measurement system has been used to examine spent fuel assemblies at the Oconee Nuclear Station of Duke Power Company. The neutron and gamma-ray emissions from individual spent fuel assemb..ies were measured in the storage pool after the assemblies were partially raised out of the storage rack. The tests were designed to demonstrate the ability of the FORK system to verify reactor records for bumum and cooling time, to detect deviations from those records, and to develop arocec ures for the use of the system that are compatible with utility o perations. .

Ninety-three assemblies were examined in 31/2 days of operation. The variation in .l burnup along the length of the assembly was measured for two assemblies. The FORK measurements correlated satisfactorily with the Oconee reactor records. The average deviation of the burnup measurements from the calibration was 10% without corrections for initial enrichment, and 2.2% with corrections. Two anomalous assemblies were detected well outside these values. The system proved to be compatible with storage pool operations, and could be used most effectively to verify reactor records in a campaign involving a large number of assemblies. The test program was a cooperative effort involving Sandia National Laboratories, Los Alamos National Laboratory, Duke Power Company, and the Electric Power Research Institute.

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CONTENTS Section Page 1 Introduetlon. . ...... . ......

.....................................................................1-1 B u rn u p C r e d i t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Radlation From Spent Fuel . . ...................................1-3 Fork System.

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..........................1-4 2 Verification Tosting at Oconee Nuclear Statlon. .... ............. ....... . .. ........... .... 2-1 Oconee Fuel. .. .. . . .. . . . . . .

...........................................2-1 Procedure.. . . . . . ....

...............................................2-1 Results and Analysis. .. .. . . . . . . . . . .. .........................2-1 Utility Comments. .. . . . . . . .. ..... ......................2-4 ,

3 Conclusions. . . . .

. . . . . . . . . . . . . . . . . . . . .. 3-1 Acknowledgments . . . .. . . . . . . . . . . . . .

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References . .. .. . . . . . ....................................,R-1 Appendix A: A s sem bly a n d FO R K Data .. .. . . .. . . . . . . . .. . ...... .... .. . . . .. ... .. .. ... ... . . .... A-1 Appendix B: E n ric hm ent C o rrection Fa ct or....................... .......................... . B-1 ~

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ILLUSTRATIONS  :

Figure Page 1 -1 Burnup Credit Loading Curve 12 1-2 Fork Detector and Control Electronics 1-5  ;

1-3 Dissassembled Fork Detector Head 1-6 1-4 Fork System Arrangement in Spent Fuel Pool 1-7 2-1 Neutron Data and Calibration 2-3 2-2 Gamma-Ray Data 2-4 B-1 Relative Cm-244 Production B2 l

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l TABLES Table Page 1-1 Gamma-Ray and Neutron Emitting Nuclides 13 2-1 Gamma-Ray Data 2-4 i

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1 INTRODUCTION The nuclear properties of spent reactor fuel depend on the initial composition of the fuel and the burnup that the fuel experienced in the reactor. The nuclear reactivity of a spent fuel assembly and the criticality (the ability to sustain a fission chain reaction) of an array of spent fuel assemblies can be calculated from the initial enrichment, burnup, and cooling time of the assemblies 1.' An average burnup value is assigned to each fuel <

assembly at the time of discharge from the reactor based on the operating history of the reactor and the distribution of the neutron flux as monitored by in-core measurements during operation. In this report, burnup will refer to the average burnup value assigned to an assembly. Burnup is commonly expressed as the time integral of the thermal power (e.g., gigawatt-days) origir. ally contained in the assembly.per metric ton A need for verification of uranium arises measurements (GWD/MTU) from metal the incorporation of burnup credit concepts in the design of storage and transport systems for spent reactor fuel. A verification measurement can contribute to the acceptibility of burnup credit by preventing criticality problems due to miscalculation or misidentification of assemblies. The purposes of the measurement operation described here were to establish a database with the FORK decector system at an operating nuclear utility, to determine compatibility with utility operating procedures, and to develop an operational plan for imp ementing verification measurements with ,

utility input.

Burnup Credit Spent fuel assemblies must be stored and transported so that criticality is not possible, I

even under theoretically optimized conditions. Calculations of criticality have '

traditionally assumed that the assemblies are immersed in pure water, and that the  !

composition of the fuelis unchanged from its original (fresh) state. Calculations using realistic spent fuel composition (burnua credit calculations) can result in significantly ,

more efficient arrays of assemblies, anc. can reduce the need for expensive neutron absorbers. Burnup credit calculations make use of the fact that the nuclear reactivity of '

the spent assembly is reduced by the depletion of fissile material and the production of l neutron absorbers by activation and fission reactions. The use of burnup credit calculations to replace " fresh fuel" calculations in the design of casks for transporting spent fuel can increase the number of assemblies that can be safely loaded into a cask by .

as much as a factor of four. The application of burnup credit to the design of spent fuel  !

casks results in significantly reduced costs and risks in the transport and storage of spent fuel assemblies 2,

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Introductwrr Spent fuel casks designed using burnup credit are restricted to accept only assemblies that meet certain minimum burnup restrictions, to limit maximum theoretical criticality to less than 95% The characteristics of fuel acceptable for loading into a burnup credit cask can be specified by a loading curve, an example of which is shown in Figure 1-1.

This loading curve is for illustrative purposes only. The design of each cask or storage arrangement would generate its own specific loading curve. The curve delineates the minimum burnup credit required for a particular initial enrichment and separates the assemblies with acceptable characteristics from those that are unacceptable. If unacceptable ossemblies are present in the spent fuel pool, the possibility exists that some unacceptable fuel could be misloaded due to misapplied reactor records or an error in assembly identification.

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Figure 1-1 Burnup Credit Loading Curve 1-2

1 Introductwn Studies have concluded that the utility-supplied data on burnup are of greater accuracy and reliability than could be provided by additional radiation measurements on spent-fuel assemblies 3. The role of a burnup measurement after discharge is to demonstrate the consistency of the reactor records, detect possible misidentification of assemblies, and detect anomalous assemblies that might affect criticality.

Radiation From Spent Fuel The emission rate of neutrons and gamma-rays produced by natural radioactive decay of the radioisotopes in the spent fuel can be related to the b6rnup of the assembly.

Radioisotopes are produced in the fuel elements during operation of the reactor by activation and fission reactions, and decay with a wide range of half-lives after the assembly is discharged from the reactor. Some important gamma-ray and neutron emitting nuclides are listed in Table 1-1.

Table 1 1 Gamma Ray and Neutron Emitting Nuclides Nuclide Half Life (yr) Radiation 242Cm 0.45 n, spontaneous fission 244Cm 18.1 n, spontaneous fission 134Cs 2.06 y,605,796 kev I37Cs 30.0 y,662 kev In the application of bumup credit, the fuel assemblies have been cooled for over I 5 years, which greatly simplifies the analysis of the emitted radiation. For shorter cochng times many more isotopes are significant emitters, but most have decayed to insignificance after several years because of the predominance of short half-lives in the fission and activation products. After 5 years the predominant neutron emitter is l curium-244, which is formed by successive neutron capture beginning with uranium-238. The neutron emission is found to follow a power law relationship with burnup in which the neutron signal increases with about the fourth power of the burnup. The neutron signal is therefore very sensitive to burnup. An additional advantage to the neutron measurement is to reduce the problem of self-shielding of the internally generated radiation. The attenuation is greater for gamma-rays than it is for neutrons, so that neutrons that reach the detector can originate from rods deeper inside the assembly than could be sampled by gamma-rays alone. The major gamma emitter, after several years of cooling, is cesium-137 which is produced as a fission product so that its production is essentially a linear function of burnup. The shorter-lived isotopes of curium and cesium are activation products that are insignificant after a few years of cooling. The combination of the gamma and neutron measurements allows both the bumup and the cooling time of each assembly to be checked. The purposes of this 4 verification operation at Oconee Nuclear Station were to generate a database of measurements with the FORK detector at an operating nudear utility, to examine the interfaces between the requirements of the measurement and the utility operations, and to obtain utility input to the development of an operational plan for implementing such a measurement.

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Introductwn FORK System The FORK system measures the passive neutron and gamma-ray emission from individual spent fuel assemblies while in the storage pool. The system, designed at Los Alamos National Laboratory, has been used for many years by the International Atomic Energy Agency (IAEA) to verify reactor records for safeguard applications. The results of those measurements are summarized, and publications cited, m Reference 4.

Comparison tests of this technique with more complex active and high-resolution measurement techniques have indicated essentially equal effectiveness 5 The FORK detector and its associated electronics are shown in Figure 1-2. Figure 1-3 is a disassembled view of the detector head. Each of the two arms of the FORK detector contains two fission chambers (the outer steel cylinders in Figure 1-3) to measure the yield of neutrons, and one ion chamber (the inner brass cylinders shown between the fission chambers in Figure 1-3) to measure gross gamma-ray emission. One fission chamber (the epithermal detector) in each arm is imbedded in a polyethylene cylinder that is surrounded by a thin sheet of cadmium. The other fission chamber is outside the cadmium cover and is sensitive to thermal neutrons and the boron content of the water in the spent fuel pool. The polyethylene cylinders containing the detectors are inserted into the polyethylene outer cover shown in Figure 1-3. The epithcrmal detectors provide the primary data used in the FORK technique. In the original (IAEA) application, the thermal neutron detectors were used to check the variation of the boron content among the spent fuel pools at different locations. In the present use, the thermal detectors serve as a back-up measurement to the epithermal data. The system is diagrammed in an operational arrangement in Figure 1-4. The detector is moved in the storage pool to the location of the spent fuel assembly to be examined. The detector head is positioned several feet above the top of the storage rack so that the radiation shielding provided by the water of the storage poolis adequate to ensure that the measurement is not influenced by radiation from nearby assemblies. The assembly is raised in the storage rack so that its midpoint is located at the detector head, the detector is moved into contact with the assembly, and the neutron and gamma-ray data are collected for 100 seconds. A burnup profile can be obtained by performing the measurements at various points along the length of the assembly. A battery-powered electronics unit and microprocessor are used to supply all power to the detectors, collect and analyze the detector outputs, and perform necessary calculations and documentation.

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2 VERIFICATION TESTING AT OCONEE NUCLEAR  :

STATION i t

Oconee Fuel Oconee Nuclear Station is a three unit generating site utilizing Babcock and Wilcox 2568 MW (thermal) Pressurized Water Reactors. Duke Power Company began commercial operation on the site in 1973. The FORK measurements were performed in the spent fuel storage pool that serves Units 1 and 2. The spent fuel assembly design is a Babcock and Wilcox 15 x 15 array that accepts separate control components such as control rods, burnable poison rods, and neutron source rods. Each assembly contains 208 fuel rods and 16 guide tubes. The maximum cross section is 8.54 inches, and overall length is 165.6 inches. ne nominal uranium weight is 464 kilograms.

Procedure De FORK detector was suspended from a moveable carriage on the fuel handling bridge over the spent fuel pool. The demineralized water in the pool contained ap proximately 2000 parts per million boron. The top of the storage rack is about 25 feet below the water level. During testing, the fuel assemblies were lifted in the storage racks by means of an auxiliary hoist mounted on the Stearns-Roger fuel handling r bridge. No assembly was completely removed from the rack. The detector head was fixed at a location about 6 feet above the top of the storage rack in the spent fuel pool.

The shielding provided by the 6 feet of water was adequate to produce the lowest background reading. Each selected assembly was raised in its rack until the detector a was at the center point of the asssembly. The detector was placed in contact with the assembly, and data were accumulated for 100 seconds to ensure that more than 10,000 counts were obtained in the epithermal neutron detectors. The ion chamber (gamma) current reaches its maximum value in about one second. The assembly was then .

lowered back into its rest position in the rack. Background data (no raised assembly) were taken each time the location of the detector was changed appreciably.

Results and Analysis Ninety-three assemblies were measured in about 31/2 working days of operation. The initial enrichment of the assemblies ranged from 2.91 to 3.92 weight percent uranium-235. De range in assembly average burnup was from 20.3 to 58.3 GWD/MTU. De .

cooling times varied from 4.2 to 14.8 years. Background data were found in all cases to be less than 1% of the signal from the assembly. Appendix A lists the data and analysis values for all assemblies.

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, Verufscation Testing at Oconer Nuclear Station The approach used in the analysis described here is to accumulate data from a large number of assemblies and generate an internal calibration by comparing each assembly to the best derived fit to all the data. This self-calibration eliminates the uncertainties that are introduced by external calibration techniques, while retaining the sensitivity to detect measurements that are inconsistent with the burnup from the reactor records.

The neutron data were extrapolated back to the date of discharge using an exponential factor of half-life 18 years, the half-life of the principal neutron emitter, curium-244. The extrapolated data for the epithermal neutron detectors are shown in Figure 2-1, a log-reactor record) for each assembly. The data log plot of neutron signal versus burnup (for the initial enrichment of the assemblies.

are shown with and without a correction The relationship of the neutron signal to burnup depends on the initial enrichment since curium-244 is produced by activation of uranium-238 rather than by fission reactions.

The " uncorrected data" (uncorrected for initial enrichment) for 91 assembly meas-urements can be fit by a power law curve determined by a least squares fit such that the average absolute deviation in burnup is about 10% Tius would be the best fit to the data if the initial enrichments were unknown. A factor to adjust the observed count rates for the variation in initial enrichment among the assemblies was calculated as described in Appendix B. The enrichment correction factor is normalized to an arbitrarily chosen enrichment of 3.0 weight percent uranium-235. For the Oconee data, the correction factor for initial enrichment varied from -7% to +53%

The " Enrichment Corrected Data" are fit by the calibration curve shown in Figure 2-1, for which the analytical expression derived from a least squares fit to the data is N = C

• B3 81 (e9 1) where N is the neutron count rate in counts per second, B is the burnup in GWD/MTU, and C is a fitted constant whose value is 0.000788. The neutron signalis proportional to the 3.81 power of the burnup. This value closely matches the values observed in earlier operations with the FORK system. With the enrichment correction applied, the data have an average absolute deviation in burnup from the calibration curve of about 2.2%

Among the 91 assemblies fit by the calibration curve, only one assembly deviated by more than 6%

The two data points marked " Outliers-Not Explained" in Figure 2-1 indicate two assemblies that exhibited much higher neutron signals than expected from the burnup records. These two data were not included in fitting the calibration curve. Both sets of neutron detectors indicated anomalous data for these two assemblies, but the corresponding gamma signals were not anomalous. The anomalies were noted at the time of measurement and the assemblies were remeasured with the same results. Since the objective of this operation was to build a substantial database of measurements with the FORK detector, it was necessary to measure as many assemblies as possible in the 2-2

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Subsequent examination of the reactor records and histories for these assemblies did not reveal any explanation for the anomalous results. The two anomalous assemblies would require further study in a verification exercise to thoroughly eliminate the possibility of an instrumentation problem.

Figure 2-2 is a plot of the gamma-ray signal for each assembly divided by its burnup versus the cooling time. If the gamma-ray signal were due solely to fission products (like cesium 137), there ideally would be a single value for each cooling time. The average deviation from the mean value of the data at each cooling time is about 15%

Since the neutron data is a far more sensitive indicator of burnup, the gamma-ray data is used only as a general confirmation of cooling time and burnup. The batch discharge of spent fuel assemblies is evident in this display from the clustering of data around certain cooling times. The two assemblies that produced the lowest gamma-ray readings at about 1900 days cooling time did not produce anomalous neutron data and, for purposes of this exercise, were not considered to be significant deviations.

To investigate the capability of the FORK detector to measure the variation of burnup along the .ength of an assembly (burnup profiles), measurements were performed at  :

several locations on two assemblies. The locations of the measurements were not determined precisely, but were approximately midway between structural bands on the assemblies. The gamma-ray data are shown in Table 2-1. The results are similar to other PWR profiles.

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Bumup (GWD/MTU) 29.42 32.70 Utility Comments In general, the FORK detector performed quite well and proved relatively easy to set up  !

and operate. It could arovide an acceptable means for verifying burnup of fuel assemblies before loac ing into a burnup credit cask or canister. Measurements should ,

be used to screen for gross errors in reactor records, such as inadvertently assigning the i burnup of one assembly to another. Qualification of a fuel assembly for loading should ,

be based on the verified reactor records for burnup since the records are likely to have less uncertainty than the measurement3 . While the equipment is simple and straight-  !

forward when used by itself, its use could potentially interfere with loading operations if measurements were performed at the time of cask or canister loading. The preferred mode of operation,if 100% verification is required, would be to verify a large number of 2-4 l

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Venficatson Testing at Oconee Nuclear Station 6

assemblies in a single campaign and to administratively control access to qualified assemblies until the loading operation. Tne loading operation could then proceed efficiently without interruption or delay for measurements and decisions. An example of one possibility of administrative control would be to physically segregate qualified assemblies in a special section of the spent fuel pool. Some utilities may prefer to have the verification campaign performed by a certified vendor rather than to commit utility resources and personnel to an additional training, certification, and maintenance program specifically to perform the measurements. Additionally, site-specific safety reviews should be completed in advance of the campaign and the utility should be assisted in specifying the radiation dose to the FORK operator, including worst case scenarios. A number of specific recommendations concerning operations, interfaces, shielding, radiation protection, decontamination, etc., have been noted and will be integrated into further tests of the FORK system.

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CONCLUSIONS .

t The FORK measurements correlated well with the Oconee reactor records. The average deviation of the reactor burnup records from the calibration was 10% without corrections for initial enrichment and 2.2% with corrections. The derived calibration indicated that the neutron signal was proportional to the bumup raised to the 3.81 .i power. The gamma-ray signals were in general agreement (15%) with the burnup records. Ninety-three assemblies were measured in about 31/2 days of operation in the ,

spent fuel pool. Two anomalous assemblies were detected that produced much higher i neutron signal, than the burnup would explain. In a verification campaign these two assemblies would require further study or be excluded from the acceptable fuel. The system is capable of generating bumup profiles with very shot t measuring time. The effectiveness of the FORK system is due to the sensitivity of the epithermal neutron yield to burnup, the self-calibration generated by a series of measurements, and the redundancy provided by three detection systems. The system proved to be compatible  ;

with utility operations, and appears to be adequate :o verify reactor records for assemblies to be loaded into burnup credit casks.

Acknowledgments The authors appreciate the essential help of the staff at Oconee Nuclear Station and Duke Power Company in the training, planning, and execution of this operation.  !

Sandia National Laboratories managed the project with the sponsorship of Sandia's Laboratory Directed Research and Development Program (dtscretionary funds). Los Alamos National Laboratory made available a FORK system and the experienced i personnel to operate it. The Electric Power Research Institute (EPRI) sponsored the ,

utility measurement and publication program 3. Duk: Power Company made its l facilities and personnel available in support of the EPXI proposal to perform the l measurements.

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REFERENCES

1. C. V. Parks, ed.," SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200, vols.1-3, revision 3, Oak Ridge National Laboratory (1984).

2. T. L. Sanders. R. M. Westfall, R. H. Jones," Feasibility and Incentives for the Consideration of Spent Fuel Operating Histories in the Criticality Analysis of Spent Fuel Shipping Casks," Sandia Report SAND 87-0151, Sandia National Laboratories (1987).

3. E. R. Johnson Associates,"The incentives and Feasibility for Direct Measurement of Spent Nuclear Fuel Characteristics in the Federal Waste Management System,"

ORNL/Sub /86-S A094/3, J AI-296, Oak Ridge National Laboratory (1988).

4. G. E. Boster and P. M. Rinard,"Burnup Measurements with the Los Alamos FORK Detector / Nuclear Materials Management, vol XX, p. 509,(1991).

5. P. M. tunard, G. Bignan, J. Capsie, J. Romeyer-Dherbey," Comparison of the FORK and Python Spent-Fuel Detectors," Report LA-11867-MS, UC-700, Los Alamos National Laboratory Guly 1990).

6. W. R. Cobb and W. J. Eich, "A New Cell Depletion Code," Transactions of the American Nuclear Society, vol. 24, p. 442 (1976).

7. T. R. England, W. B. Wilson, and M. G. Stamatelatos, " Fission Product Data for Thermal Reactors," Parts 1 and 2, Electric Power Research Institute report, EPRI NP-356, Parts 1 and 2; also published as Los Alamos Scientific Laboratory reports LA-6745-MS and LA-6746-MS,(December 1976).

l

8. W. B. Wilson, R. J. LaBauve, and T. R. Eng'and, "H.B. Robinson-2 Spent Fuel isotopics; Sensitivity Series 2," (T-2-1128), Los Alamos Scientific Laboratory memo to John Phillips and Gerald Bosler (October 9,1980).  !

I l

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APPENDlX A: ASSEMBLY AND FORK DATA Columns 1 through 4 were obtained from the reactor records for the assemblies tested.

Column 1 is the assembly identification, column 2 is the initial enrichment, column 3 is the average bumup in gigawatt days per metric ton of uranium, and column 4 is the time interval between discharge from the reactor and the date of the measurement in days. Column 5 is the observed epithermal neutron count rate in counts per second, background subtracted. Column 6 is the observed thermal neutron count rate in counts per second, background subtracted. Column 7 is the epithermal data of column 5 extrapolated to the date of discharge ( EAT, A = 0.0001048 d-1, T = data of column 4) using the 18 year half-life of curium-244, and is the " uncorrected (for enrichment) data" of Figur( 2-1. Column 8 is the correction factor far initial enrichment described in Appendix B. Column 9 is the epithermal neutron count rate of column 7 multiplied by - t the factor of column 8, and is the " corrected data" of Figare 2-1. Column 10 is the bumup value determined from the calibration line, which is derived from the best fit to the data of column 9. Column 11 is the absolute deviation (in percent) of the bumup determined from the calibration (column 10) and the reactor record bumup (column 3).

Column 12 is the observed gamma (ion chamber) signal in milliamperes. Column 13 is the gamma signal divided by the bumup of column 3, and is plotted in Figure 2-2.

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I n qh No 4 N Wal AB/B G 4 U factor (cps) (GW1'/MTU) (%) (ma) C/B >

ID (wi%) (GWD/MTU) (d) (cp,) (cps) (cps) 45 10 57 45 6930 0.9 il 6432 19 41 4 32 13135 6 47 NIOlQ1 ZW 20 287 4(N7 M79 438 67 49124 0.944 463 55 32 55 0.43 248 88 7 68 NiOl61 2908 32 409 341.10 2.912 M78 232.85 279 14 33531 0.944 316 63 29.45 0.77 219.74 752 NJ0159 29 229 3473 M7.00 456 00 49443 0.946 472.65 32.72 0M 246 50 7.M NIOl69 2.912 32.7M

• 281 40 298 26 0.946 282 07 2857 3 06 225.93 7 66 NIO15C 2.914 29 476 M79 207.10 3478 335 98 M O91 0.946 32233 29 60 1El 231 07 7 67 N10168 2 914 30 141 236.74 320 73 351 02 0.946 332.09 29 82 1 08 229 19 7 to N1014R 2.914 30.151 M79 243.73 3478 259.65 355fo 373 90 0.947 353.91 3033 2 09 234 47 757 NIO15Q 2.914 30 972 376 81 0 947 356.69 3039 2.05 235.87 7to NIOl67 2.914 31 025 3477 261.71 361.05 3473 408 75 0.947 387.09 31 05 2 25 24130 7 60 NIO15T 2 914 31.763 284 00 346 00

'M 23.092 4097 7320 98 85 112.48 0943 106.08 22.11 424 155 00 8 71 NJOI4Y 2955 1.93 229.75 7.63

!5 30.131 M78 235.14 328 75 338 61 0.947 32056 NJ015L 0.947 29.17 5.92 227.48 734

_ A 15 31.001 3479 223 (C 307.53 322.02 305 02 NJOI5I 31.68 0.48 24237 7.61 31.831 M79 30630 402.80 441.12 0.948 418 05 N]O15K 2.915 19 48 3.97 134.fo 6.63 2.916 4097 45.20 61.60 69.45 0.942 65 44 NJ0151 20.288 4102 5924 66 49 0.942 62.65 1926 5.06 13135 6 47 N1015X 2.916 20.288 43.25 2.916 20365 4102 45.18 57.77 69.46 0 942 65.45 19.48 433 127.00 6.24 NJ015U 28 43 338 221.05 7.51 2.916 29.424 M73 203(X) 277.50 292.17 0.947 276.67 NJO15F 29.47 5.19 227.82 733 2.916 31.082 3478 232.43 325.67 334.70 0.948 317.25 NJ015Y 32.06 0.99 243.46 7.52 2.916 32388 M79 320.43 412.60 461.48 0.949 437.79 NIO15W 30.19 125 174.62 5 86 3C54 2.997 29.820 5404 197.85 277.17 348.66 0.998 348.01 MM 8022 88.97 14137 0.999 141.28 23.84 0.80 132.28 559 3C57 2.999 23.M8 83.40 146.37 1.001 147.07 24.09 2.24 142.83 6 06 3C32 3.001 23.562 5404 13522 16759 1.001 167.70 24.93 2.06 146.00 5.98 3C03 3.001 24.430 5404 95.10 14751 13229 16034 233.13 1.001 233.28 27.19 3.43 15531 5.91 3C38 3.001 262 % 5404 273.23 1.001 273.40 2834 2.72 164.00 5.94 3C40 3.001 27.592

• M03 155 06 199.02 9857 1%51 1.002 1 %.93 26.01 10.85 139.73 5.96 3C19 3.003 23.460 MM 11153 6.06 MG4 157.69 17025 1 002 17058 25.04 3.09 147.17 3C01 3.003 24295 96 61 3.003 24311 Mot 103.48 123.45 18236 1.002 182.71 2550 4.89 143.75 5.91 3C02 26.100 5405 111.63 169.97 196.75 1.002 197.12 26.01 034 15522 5.95 3C29 3.003 3090 31858 50535 440.47 1.038 457.26 32.43 0.08 2M57 8.77 NJ01Q5 3.064 32.460 725.24 704.84 1.036 73054 36.67 3.40 320 62 9.Gt NJOll'Q 3.064 35.4M 3090 509.79 637.43 735.80 1.036 762 41 37.08 3.25 31324 8.72 NJOIPT 3.0M 35.917 3092 532.07 3090 35553 461.73 49156 1.039 51059 3338 2.M 280 60 8M NJ01Q4 3.065 32.463 1.037 M3.15 3547 1.02 294 00 8 37 NJ01QE 3.065 35.108 3090 448.47 612.94 620 06 I

1 u

l 10 B T n nth No Ei N Scal AB/B G G/B Ei (wt%) (GWD/MTU) (d) (cpe) (cpe) (cps) factor (cps) (GWD/MTU) (%). (ou) ~

NjUlll 3 065 39 624 J090 73951 1046.27 102Z.46 1 033 10%H 4039 1.94 355.93 8.98 3.065 33 645 3090 73655 991 62 1018 37 1 033 1052 02 4035 1.78 349 M 8 80 NJO112 3 066 33369 3090 344.45 534.75 476.24 1 039 494 73 33.11 0 78 285.24 855 NJ01QB 3 066 33377 3090 356.80 492.71 49332 1 039 512.46 3342 0.12 278.83 835 NJOlQ2 NJOIPS 3 066 33.481 3090 367.00 505 22 507.42 1 039 527.08 33 66 055 293 85 8.78 3.066 33.913 3090 343 60 47637 475 07 1.039 49336 33 09 2.44 290.15 856 NJ01P6 3 066 34.438 3090 362.24 426.22 500 0 1 038 519.97 3354 2.60 283.87 824 NJO1QC 3.066 35.116 3090 448.45 583.90 620.0.3 1 038 643 47 35 47 1.01 29232 832 NJ01Q3 8.68 3.066 35589 3091 511.45 639.70 707.21 1.038 733.73 36.71 3.16 308.76 NJOIPA 36 88 8.67 3 066 35.590 3090 520.61 660.72 719.21 1 038 746.80 3 64 308.61 NJOIPP 280.69 8.41 3 067 33368 3091 374.48 460,44 517.82 1.039 538.22 33 85 1.44 NJOIPD 3232 286.95 8.46 3.067 33.932 3090 321.66 470.04 444.73 1.039 462.11 4.15 NJOIPV 34.683 3090 446.63 55553 61752 1.039 64137 35.44 2.19 294.18 8.48 NJOIPE 3.067 3.068 32.454 3090 34130 521.59 471.89 1.010 490.99 33 04 1.82 284.18 8.76 NJO150 3.068 33.667 3090 418.73 670.22 578.94 1.040 601.99 34.86 354 300.77 8.93 NJ01Q7 32.94 280.20 829 3.068 33.820 3090 33757 416.27 466.73 1.040 485.27 2.60 NJOIPU 3.068 34.682 3090 450.04 588.99 622.23 1.039 646.62 3552 2.41 288.44 832 NJ01QD 3.068 34.991 30%) 458.60 675.94 634.07 1.039 658.80 35.69 2.00 305.10 8.72 NJOIPG 3.068 35.106 3090 464.00 585.60 64153 1.039 66651 35.80 1.98 29356 836 NJOIPC 3633 857 3.068 35342 3092 490.70 622.~7 678.59 1.039 704.91 2.79 302.85 NJO1Q9 8.69 3.068 35354 3091 462.25 683.74 639.18 1.039 663.97 35.76 1.16 307.20 NJOIPH 38.47 8.73 3 069 35.117 3090 497.47 636.60 687.81 1.039 714.97 3.84 306.44 NJOIPZ 275.26 8.25 3.070 33383 3090 36950 446.64 510.88 1.041 531.89 33.74 1.08 NJOIPW 3.070 33.969 3090 321.08 3 %.95 443.93 1.041 462.04 3252 4.26 275.90 8.12 NJOIPL 285.81 8.41 3.070 33.989 3090 364.91 445.75 50453 1.041 525.10 33.63 1.05 NJOIPJ 3.070 35.909 3090 546.69 752.77 755.86 1.040 785.73 3738 439 325.47 9.06 NJOIPM 3.072 34.699 3090 418.73 670.22 578.94 1.041 602.95 34.87 050 300.77 8.67 NJ01PY 3.072 34.703 3090 462.11 574.60 838.92 1.041 665.42 35.78 3.12 296.65 855 NJOIPX NJO1QO 3.073 35.112 3091 430.65 628.07 595.49 1.042 62035 35.13 OM 296.50 8.44 180.52 233.17 219.76 1.145 251.70 27.73 3.49 308.60 10.74 NJO3AS 3.238 28.738 1876 3.238 28.845 - 1876 183.02 251.77 222.80 1.145 255.16 27.83 351 325.62 11.29 NJO3AR 1339 3.238 38.093 1876 74835 664.00 911.02 1.125 1024.83 40.07 520 509.99 NJO3BY 3.242 37.480 1877 516.00 817.00 62823 1.130 709.64 3639 2.90 515.00 13.74 NJO3BD 1359 3.242 37.672 1876 556.62 749.74 677.61 1.129 764.90 37.12 1.48 511.88 l NJO3BC 3242 37.679 1876 622.25 70037 75751 1.129 855.07 38.22 1.43 517.50 13.73 3 NJO33P

NJO3BG 3.242 38.073 1877 ~ 572.80 811.70 69738 1.127 786.01 3738 1.81 52936 13.90 g NJO3BF 3.242 38.250 1877 567.97 82233 69150 1.126 778.82 37.29 250 521.23 13.63

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P D N u 2 ID B 1 n nth No Ei N Heal AB/B G G/B E Ei (w t*'.) (GWD/MTU) (d) (cps) (cps) (cps) factor (cps) (GWD/MTU) (%) (ma) [

3 245 38999 1877 656 43 97h 93 799 21 1.124 M9832 3311 0 73 566.23 1452 NJO3B2 3.245 41.489 1876 843 17 1151.26 1026 45 1.131 1160 46 41.40 0.21 58655 14.14 N]O3BO 3.253 41.755 1876 86334 1159 23 1051 00 1.143 1201.60 41.78 0 06 593.75 14.22 NJO3C4 NJO3C9 3 256 4 891 1877 513 40 770 63 625 06 1.139 712.11  % 43 1.26 50936 13.81 NJO3CA 3.256 36.948 1876 493.02 776 74 600 19 1.139 683 64 A04 2.46 50437 13 65 3.402 33 768 2029 345.98 448 61 428 00 1.235 52855 33.69 0.24 401.88 11.90 N10379 3.403 33 822 2030 35757 445 40 44238 1.235 54650 33.98 0.48 403.11 11.92 NJO37A 3.406 33 698 2029 369.18 462.71 456.70 1.238 565.26 34.29 1.75 407.95 12.11 NJO36P 3.408 E 995 2029 477.05 620 M 590.14 1.224 722.20 3656 1.17 45937 12.42 NJO37D 3.408 37399 2029 519 08 673 67 642.13 1.221 784.18 3736 0.11 474.99 12.70 NJO37E 3.410 37.660 2029 580.40 827.97 717.99 1.221 87633 38.46 2.13 490.94 13.04 NJO37G 3.412 37.460 2029 437.08 68534 540.69 1.223 66132 35.73 4.63 466.87 12.46 NJO368 3.412 37.691 2030 511.67 64937 633.03 1.221 773.22 37.22 1.25 47936 12.72 NJO36F NJO369 3.413 33.713 2029 331.25 487.27 409.78 1.242 508.96 3336 1.06 402.50 11.94 3.413  % 995 2029 512.25 71857 6n 68 1.227 77734 37.27 0.75 466 25 12.60 NJO365 NJO35Z 3.413 37.526 2029 513.78 701.97 63557 1.223 777.45 37.27 0.67 469.07 1250 3.417 37516 2029 517.75 79234 640.49 1.226 784.98 3737 039 491.24 13.09 NJO36E NJO37L 3.919 45.785 2029 863.55 1158.14 1068.26 1532 1636.92 4531 1.04 61438 13.42 3.919 46.105 2029 863.98 1208.47 1068.79 1530 16R 87 45.29 1.76 603.75 13.10 NJO37N NJO37K 3.919 58310 1541 2257.00 2913.94 2652.78 1.441 3823.85 56.60 2.94 1108.12 19.00 3.919 58310 1541 2241.94 296354 2635.07 1.441 3798 32 5650 3J.1 1101.87 18.90 NJO37K Outliers as g. = 2.24 3C09 3.004 23.815 5404 400.68 426.88 706.10 1.003 707.94 3637 52.72 132.97 558 3C28 3.004 23.911 54M 385.46 586.79 679.28 1.003 681.05 E00 5057 130 88 5.47 2.24 Profile Data - NJ0165 1&2 Grids 2.912 32.704 3473 29735 375.00 0.946 229.00 3&4 Grids 2.912 32.704 3473 347.00 456.00 0.946 24650 Profile Data -NJ015F Nozzle &1 2.916 29.424 3473 1834 3436 0.947 108.95 1&2 Grids 2.916 29.424 3473 142.00 209.80 0.947 199.80 2&3 Grids 2.916 29.424 3473 20650 285.00 0.947 22050 3&4 Grids 2.916 29.424 3473 203.00 27750 0.947 221.05

e APPENDIX B: ENRICHMENT CORRECTION FACTOR In the analysis of the neutron data the assumption is made that the source of the neutrons is the curium-244 in the assemblies. To adjust for the decay of curium-244 after discharge from the reactor, the observed neutron counting rates are extrapolated to the date of discharge using the 18 year half-life of curium-244, as described in Appendix A. An additional correction factor is needed because the neutron emission rate at a given bumup depends on the initial enrichment of the assemblies. This is due to the fact that the production of curium depends on the neutron flux in the reactor, rather than solely on the fission rate. The relationship between the production of curium and the bumup depends on the initial enrichment of the assembbes and the burnup. To adjust the observed neutron data for this dependence, an enrichment correction factor is applied to the extrapolated neutron count rates. The factor is determined from calculations of the production of curium-244 corresponding to the initial enrichment and bumup of the assembly. Curium aroduction calculations were obtained using a combination of the EPRI-CELL 6 ar.d E?RI-CINDER 7 codes. The calculations were validated by comparison to destructive chemical analysis of spent fuel from the H. B.

Robinson reactor.8 The curium-244 production calculations that were used in determining the initial enrichment correction are plotted in Figure B-1. The relative production rate for curium-244 is shown as a function of burnup for a family of curves covering the initial enrichments of interest in this report. Since the neutron data for each assembly are multiplied by a correction factor, only relative values are required.

For convenience the factors are normalized (=1) at an enrichment of 3.0 weight percent "

uranium-235. The correction factor for an assembly of a given bumup, taken from the reactor records,is defined as the ratio of the curium proc uction at an enrichment of 3.0 weight percent divided by the curium production at the initial enrichment of the l

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assembly. Values between the curves are interpolated using standard routines. The enrichment correction factors are listed for each of the Oconee assemblies in Appendix A, column 8.

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