ML042370178
| ML042370178 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire, 07003098  |
| Issue date: | 08/20/2004 |
| From: | Martin R NRC/NRR/DLPM/LPD2 |
| To: | Barron H Duke Energy Corp |
| Martin R, NRR/DLPM, 415-1493 | |
| References | |
| -RFPFR | |
| Download: ML042370178 (16) | |
Text
August 20, 2004 Mr. H. B. Barron Executive Vice President Nuclear Generation Duke Energy Corporation 526 South Church Street Charlotte, NC 28202
SUBJECT:
FINAL SAFETY EVALUATION FOR DUKE TOPICAL REPORT DPC-NE-1005P, NUCLEAR DESIGN METHODOLOGY USING CASMO-4/SIMULATE-3 MOX
Dear Mr. Barron:
Enclosed is a copy of the U.S. Nuclear Regulatory Commission staff's Safety Evaluation (SE) for Topical Report DPC-NE-1005P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX.
A draft of this SE was provided to you by letter dated February 20, 2004. By letter dated March 9, 2004, you provided comments on the draft SE. This final SE responds to those comments and issues the SE in final form. Your letter also stated that the draft SE contained no proprietary information.
In the event of any comments or questions, please contact me at (301) 415-1493.
Sincerely,
/RA/
Robert E. Martin, Senior Project Manager Project Directorate II-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-369, 50-370, 50-413, 50-414
Enclosure:
As stated cc w/encl: See next page
August 24, 2004 Mr. H. B. Barron Executive Vice President Nuclear Generation Duke Energy Corporation 526 South Church Street Charlotte, NC 28202
SUBJECT:
FINAL SAFETY EVALUATION FOR DUKE TOPICAL REPORT DPC-NE-1005P, NUCLEAR DESIGN METHODOLOGY USING CASMO-4/SIMULATE-3 MOX
Dear Mr. Barron:
Enclosed is a copy of the U.S. Nuclear Regulatory Commission staff's Safety Evaluation (SE) for Topical Report DPC-NE-1005P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX.
A draft of this SE was provided to you by letter dated February 20, 2004. By letter dated March 9, 2004, you provided comments on the draft SE. This final SE responds to those comments and issues the SE in final form. Your letter also stated that the draft SE contained no proprietary information. Please identify any such errors or concerns within 15 working days of the date of this letter.
In the event of any comments or questions, please contact me at (301) 415-1493.
Sincerely,
/RA/
Robert E. Martin, Senior Project Manager Project Directorate II-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-369, 50-370, 50-413, 50-414
Enclosure:
As stated cc w/encl: See next page DISTRIBUTION: (See attached list)
- No Significant Changes to SE input.
ADAMS ACCESSION NO: ML042370178 NRR-106 OFFICE PDII-1/PM PDII-1/LA(A)
DSSA:SRXB PDII-1/SC(A)
NAME RMartin DClarke (w/com)
FAkstulewicz MJ Ross-Lee DATE 8/17/04 8/16/04 8/17/04 8/20/04 OFFICIAL RECORD COPY
SUBJECT:
FINAL SAFETY EVALUATION FOR DUKE TOPICAL REPORT DPC-NE-1005P, NUCLEAR DESIGN METHODOLOGY USING CASMO-4/SIMULATE-3 MOX Dated: August 20, 2004 DISTRIBUTION:
PUBLIC PDII-1 R/F RidsNrrDlpmPdii-1 (MRoss-Lee)
RidsNrrLACHawes (Hard Copy)
RidsNrrPMRMartin RidsOgcRp RidsAcrsAcnwMailCenter RidsRgn2MailCenter (RHaag)
EGuthrie AAttard UShoop
McGuire Nuclear Station Catawba Nuclear Station cc:
Ms. Lisa F. Vaughn Duke Energy Corporation Mail Code - PB05E 422 South Church Street P.O. Box 1244 Charlotte, North Carolina 28201-1244 County Manager of Mecklenburg County 720 East Fourth Street Charlotte, North Carolina 28202 Mr. C. Jeffrey Thomas, Manager Regulatory Compliance Duke Energy Corporation McGuire Nuclear Site 12700 Hagers Ferry Road Huntersville, North Carolina 28078 Anne Cottingham, Esquire Winston and Strawn 1400 L Street, NW.
Washington, DC 20005 Senior Resident Inspector c/o U. S. Nuclear Regulatory Commission 12700 Hagers Ferry Road Huntersville, North Carolina 28078 Mr. Peter R. Harden, IV VP-Customer Relations and Sales Westinghouse Electric Company 6000 Fairview Road 12th Floor Charlotte, North Carolina 28210 Dr. John M. Barry Mecklenburg County Department of Environmental Protection 700 N. Tryon Street Charlotte, North Carolina 28202 Mr. Richard M. Fry, Director Division of Radiation Protection North Carolina Department of Environment, Health, and Natural Resources 3825 Barrett Drive Raleigh, North Carolina 27609-7721 Ms. Karen E. Long Assistant Attorney General North Carolina Department of Justice P. O. Box 629 Raleigh, North Carolina 27602 Mr. R. L. Gill, Jr.
Manager - Nuclear Regulatory Issues and Industry Affairs Duke Energy Corporation 526 South Church Street Mail Stop EC05P Charlotte, North Carolina 28202 NCEM REP Program Manager 4713 Mail Service Center Raleigh, North Carolina 27699-4713 Mr. T. Richard Puryear Owners Group (NCEMC)
Duke Energy Corporation 4800 Concord Road York, South Carolina 29745 Mary Olson Director of the Southeast Office Nuclear Information and Resource Service 729 Haywood Road, 1-A P. O. Box 7586 Asheville, North Carolina 28802
McGuire Nuclear Station Catawba Nuclear Station cc:
Lee Keller Regulatory Compliance Manager Duke Energy Corporation 4800 Concord Road York, South Carolina 29745 North Carolina Municipal Power Agency Number 1 1427 Meadowwood Boulevard P. O. Box 29513 Raleigh, North Carolina 27626-0513 County Manager of York County York County Courthouse York, South Carolina 29745 Piedmont Municipal Power Agency 121 Village Drive Greer, South Carolina 29651 Saluda River Electric P. O. Box 929 Laurens, South Carolina 29360 Henry Porter, Assistant Director Division of Waste Management Bureau of Solid and Hazardous Waste Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201-1708 North Carolina Electric Membership Corp.
P. O. Box 27306 Raleigh, North Carolina 27611 Senior Resident Inspector 4830 Concord Road York, South Carolina 29745 Mr. Dhiaa Jamil Vice President Catawba Nuclear Station Duke Energy Corporation 4800 Concord Road York, South Carolina 29745 Mr. G. R. Peterson Vice President McGuire Nuclear Station Duke Energy Corporation 2700 Hagers Ferry Road Huntersville, North Carolina 28078
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT DPC-NE-1005P NUCLEAR DESIGN METHODOLOGY USING CASMO-4/SIMULATE-3 MOX FOR CATAWBA NUCLEAR STATION, UNITS 1 AND 2 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS: 50-413, 50-414, 50-369 AND 50-370
1.0 INTRODUCTION
The Duke Power Company (Duke or licensee) submitted by letters dated August 3 (Proprietary), and August 6, 2001 (Non-proprietary), the Topical Report DPC-NE-1005P, Revision 0, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX [mixed oxide],
for review by the U.S. Nuclear Regulatory Commission (NRC) staff. Duke is the license for the Catawba Nuclear Station (Catawba), Units 1 and 2, and the McGuire Nuclear Station (McGuire), Units 1 and 2. Duke submitted additional letters dated September 12 and November 12, 2002, and June 26, August 14 and December 2, 2003 (References 2, 3, 4, 5 and 6).
The Topical Report addresses the use of the Studsvik Core Management System (Studsvik/CMS) code package to support the reload design analyses for Catawba and McGuire.
The Studsvik/CMS code package primarily consists of the CASMO-4 and SIMULATE-3 MOX computer codes. The Topical Report demonstrates the validity and accuracy of the Studsvik/CMS code package at Catawba and McGuire for core reload design, core follow, and calculation of key core parameters for reload safety analysis. The NRC staffs review of the topical report considered the topical reports applicability for the use of low-enriched uranium (LEU) fuel at Catawba and McGuire and the use of up to four MOX lead test assemblies (LTAs) in one of the Catawba units. The NRC staffs review findings are based, in part, on licensee commitments included by Duke in Reference 4 as follows:
1.
For a lead assembly program containing four MOX fuel assemblies, Duke will place at least two of the MOX fuel lead assemblies in core locations that are measured directly by the movable incore detector system for the first and second cycles of lead assembly irradiation.
2.
Duke will perform the physics test program defined in Table 1 [of Reference 4] for all MOX fuel lead assembly cores and for each unit operating with partial MOX fuel cores until the equilibrium cycle defined [in Reference 4] is reached. Core power levels at which low and intermediate power escalation power distribution maps are taken will be consistent from cycle to cycle for each unit (within +/- 3% rated thermal power). Core power level at which power distribution maps are taken may vary among units and between McGuire and Catawba.
3.
Duke will prepare a startup report for each operating cycle with MOX fuel lead assemblies and for each unit operating with partial MOX fuel cores until the equilibrium cycle defined above [in Reference 4] is reached. Each startup report will contain comparisons of predicted to measured data from the zero power physics tests and the power distribution maps taken during power escalation. The reports will include discussions of any parameter that did not meet acceptance criteria. Duke will provide each report to the NRC within 60 days of measurement of the final power distribution map.
4.
Duke will prepare an operating report for each operating cycle with MOX fuel lead assemblies and for each unit operating with partial MOX fuel cores until the equilibrium cycle defined above [in Reference 4] is reached. Each operating report will contain comparisons of predicted to measured monthly power distribution maps and monthly boron concentration letdown values. Duke will provide each cycle operating report to the NRC within 60 days of the end of the fuel cycle.
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 34, Contents of Applications; Technical Information, requires that safety analysis reports be submitted that analyze the design and performance of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents. As part of the core reload process, licensees perform reload safety evaluations (SE) to ensure that their safety analyses remain bounding for the design cycle. Licensees confirm that the analyses remain bounding by ensuring that the inputs to the safety analyses are conservative with respect to the current design cycle. They check these inputs by using core design codes and methodologies.
The objective of the nuclear design review for the fuel assemblies, control systems, and reactor core is to aid in confirming that fuel design limits will not be exceeded during normal operation or anticipated operational transients. The NRC staff acceptance criteria are based on Chapter 4.3, Nuclear Design, of the Standard Review Plan.
3.0 TECHNICAL EVALUATION
Currently, Catawba and McGuire use the CASMO-3/SIMULATE-3 analytical computer codes and various methodologies. In its submittal, Duke requests replacing its current codes with the newer Studsvik/CMS code package. The CASMO-4, CMS-LINK, SIMULATE-3 MOX, and SIMULATE-3K MOX computer codes comprise the Studsvik/CMS package.
The CASMO-4 computer code is the Studsvik Scandpower, Inc., lattice code. The CASMO-4 computer code, a multi-group two-dimensional transport theory code for depletion and branch calculations for a single assembly, is used to generate the lattice physics parameters. These parameters include the cross sections, nuclide concentrations, pin power distributions and other nuclear data used as input to the SIMULATE-3 MOX program for core performance analyses.
New features of CASMO-4 over CASMO-3 are the incorporation of the microscopic depletion of burnable absorbers into the main calculations, and the introduction of a heterogeneous model for the two-dimensional calculation. Also new in CASMO-4, is the use of the characteristics method for solving the transport equation. When MOX fuel is detected in the input, the code automatically uses a more detailed internal calculation to accommodate the larger variation of plutonium (Pu) cross sections and resonances. Studsvik also supplies the SIMULATE-3 MOX code. This code is a two-group, three-dimensional nodal program based on the NRC staff-approved QPANDA neutronics model that employs fourth-order polynomial representations of the intranodal flux distributions in both the fast and thermal neutron groups.
The code is based on modified coarse mesh (nodal) diffusion theory calculational technique, with coupled thermal hydraulic and Doppler feedback. The program explicitly models the baffle/reflector region, eliminating the need to normalize to higher-order fine mesh calculations.
It also includes the following modeling capabilities: solution of the two group neutron diffusion equation, fuel assembly homogenization, explicit reflector cross-section model, cross-section depletion and pin power reconstruction. The SIMULATE-3 MOX code uses a more refined solution technique to account for steeper flux gradients that exist between the MOX and LEU fuel interfaces.
In order to insure flux continuity at nodal interfaces and perform an accurate determination of pin-wise power distributions, SIMULATE-3 MOX uses assembly discontinuity factors that are pre-calculated by CASMO-4. These factors are related to the ratio of the nodal surface flux in the actual heterogenous geometry to the cell averaged flux in an equivalent homogeneous model, and are determined for each energy group as a function of exposure, moderator density and control-rod-state.
The two group model solves the neutron diffusion equation in three dimensions, and the assembly homogenization employs the flux discontinuity correction factors from CASMO-4 to combine the global (nodal) flux shape and the assembly heterogeneous flux distribution. The flux discontinuity concept is also applied to the baffle/reflector region in both radial and axial directions to eliminate the need for normalization, or other adjustments at the core/reflector interface.
The SIMULATE-3 MOX fuel depletion model uses tabular and functionalized macroscopic or microscopic, or both cross-sections to account for fuel exposure without tracking the individual nuclide concentrations. Depletion history effects are calculated by CASMO-4 and then processed by the CMS-LINK code for generation of the cross-section library used by SIMULATE-3 MOX.
SIMULATE-3 MOX can be used to calculate the three-dimensional pin-by-pin power distribution in a manner that accounts for individual pin burnup and spectral effects. SIMULATE-3 MOX also calculates control rod worth and moderator, Doppler and xenon feedback effects.
SIMULATE-3K MOX is an extension of SIMULATE-3K, which is used for analysis of core transients. The spatial neutronics models in SIMULATE-3K MOX are identical to those in SIMULATE-3 MOX. SIMULATE-3K MOX solves the transient neutron diffusion equation incorporating effects of delayed neutrons, spontaneous fission in fuel, alpha-neutron interactions from actinide decay, and gamma-neutron interactions from long-term fission product decay. For the applications reviewed in Topical Report DPC-NE-1005P, SIMULATE-3K MOX is used only as part of the dynamic rod worth measurement (DRWM) methodology.
3.1 Model Benchmarking The licensees submittal, dated August 3, 2001, compares the CASMO-4/SIMULATE-3 MOX predictions of key physics parameters against plant data and critical experiments. For CASMO-4, this benchmarking encompassed criticality and pin power predictions for LEU and MOX fuel. As part of the development of the Catawba and McGuire models, the licensee compared CASMO-4/SIMULATE-3 MOX calculation predictions to plant and/or experimental data for reactivity worth for soluble boron, burnable poison rods, silver-indium-cadmium control rods, Isothermal temperature coefficient, and core power distribution. The licensee provided documentation that contained the results of benchmarking CASMO-4 results to Monte Carlo code calculations and critical experiments for LEU and MOX fuel assembly designs (References 5 and 6).
The licensee performed comparisons between CASMO-4 MOX predictions and data from three MOX critical experiments: Saxton, EPICURE, and ERASME/L. The results of these comparisons were used in the development of the fuel pin power uncertainties that are part of the overall nuclear uncertainty factors. The Saxton critical experiment used Pu that had an isotopic content that is close to current weapons grade Pu fuel. EPICURE used fuel pins that are similar to current 17x17 pressurized-water reactor fuel pins and emulated the hot condition fuel to moderator ratio. ERASME/L used a fissile Pu concentration of 8.28 percent that bounds the fissile Pu content expected in the Duke reactors. SIMULATE-3 MOX could not model the experiments because of their small configurations; therefore, theoretical problems were developed to test the ability of SIMULATE-3 MOX to replicate the CASMO-4 calculations. This provides greater assurance that the CASMO-4/SIMULATE-3 MOX suite of codes will predict the core parameters for a core containing four MOX LTAs with acceptable accuracy.
The comparison of CASMO-4/SIMULATE-3 MOX predictions to measured data incorporates bias and uncertainty for both the predictions and the measured data. The licensee then used statistical methods to account for these uncertainties. For MOX fuel, these methods accounted for the uncertainty from the CASMO comparisons with data and the uncertainty from the CASMO-4 to SIMULATE-3 MOX comparisons for the theoretical problems. Duke also used the CASMO-4/SIMULATE-3 MOX predictions in combination with the normalized flux map reaction rate comparisons to determine appropriate peaking factor uncertainty factors.
Duke intends to use the CASMO-4/SIMULATE-3 MOX programs in licensing applications, including calculations for core reload design, core follow, and calculation of key core parameters for reload safety analyses of Catawba and McGuire. The licensee used data from the Catawba, Unit 1, operating cycles 11 through 13, Catawba, Unit 2, operating cycles 9 through 11, and McGuire, Units 1 and 2, operating cycles 12 through 14, to benchmark the CASMO-4/SIMULATE-3 MOX models for LEU fuel. Duke also used data from the St. Laurent B1 reactor in France, cycles 5 through 10, to benchmark the CASMO-4/SIMULATE-3 MOX models for MOX fuel. These cycles cover core design changes over 17 cycles of operation.
Comparison of the St. Laurent parameters to the Catawba and McGuire reactor parameters were provided and demonstrated that the fuel and core parameters important to predicting the core physics response were similar. Loading pattern variations include out-in and low-leakage designs. For model benchmarking, the licensee used critical boron concentration measurements, startup physics testing data, and flux maps. The good agreement between the measured and the calculated values presented in the August 3, 2001, submittal, is used to validate the Duke application of these computer programs for analysis of Catawba and McGuire for LEU and MOX LTAs (maximum four LTAs in one of the Catawba units) fueled cores.
For the parameters compared, the licensee calculated a sample mean and standard deviation of the observed differences. They also determined bias to describe the statistical difference between predicted and reference values.
The St. Laurent reactor uses reactor grade MOX fuel and though similar in composition to the weapons grade MOX fuel, the isotopic composition slightly differs. The Saxton critical experiment uses a Pu isotopic composition that is very close to the weapons grade MOX (90 percent fissile Pu composition.) Both benchmarks demonstrate that the CASMO-4/SIMULATE-3 MOX code can provide close predictions and provides confidence that the code will provide a close prediction of the MOX LTAs. To support future batch implementation, Duke provided a committment in Reference 4 that at least two of the MOX LTAs will be placed in instrumented core locations so that the results from the startup physics tests can be compared to the CASMO-4/SIMULATE-3 MOX predictions to demonstrate the applicability of the codes to analyze LEU and MOX fueled cores. The results of these benchmarks will be submitted to the NRC for review and approval.
The licensee demonstrated that the CASMO-4/SIMULATE-3 MOX models, in conjunction with the indicated reliability factors adequately represent the operating characteristics of Catawba and McGuire. Additionally, Duke did not change key aspects of their core design and analysis methodology, and maintains code and quality assurance practices that provide assurance that future changes to the core, fuel, and burnable poison design will be modeled with accuracy and conservatism. Since the Studsvik/CMS package adequately represents the operating characteristics, the NRC staff finds the use of the Studsvik/CMS package acceptable for Catawba for LEU fuel and up to four MOX LTAs and for McGuire with LEU fuel.
3.2 Statistics The NRC staff reviewed Dukes application for statistical content. The statistical issues revolved around the 95/95 (probability/confidence) tolerance limit calculations for each parameter of interest. The calculations give 95 percent assurance that at least 95 percent of the population will not exceed the tolerance limit.
The procedure used in the tolerance limits depended on whether the data could be assumed to be distributed normally. The licensee used an established technique for testing normality and assumed normality only if the technique validated that assumption. This approach is acceptable to the NRC staff.
When the normal distribution was applicable, the licensee used the traditional one-sided tolerance calculations. Otherwise, they used a nonparametric method to determine a conservatively large uncertainty (References 9, 10, 11 and 12). Both the parametric and the nonparametric approaches in their proper context are acceptable to the staff.
3.3 Dynamic Rod Worth Measurement DRWM provides a methodology for the licensee to measure the reactivity worth of the individual control rod banks without changing the boron concentration. The DRWM methodology takes the neutron flux signal from the excore detectors and conditions the excore detector signal through the use of analytical factors to convert the signal into the corresponding rod worth. The SE that approved the Westinghouse DRWM methodology required that anyone applying to use the methodology with their own codes perform calculations comparing their code results to the Westinghouse generated results and that the results must agree within 2 percent or 25 percent mille (pcm) for individual banks, and 2 percent for total bank worth. The acceptance criteria were developed to demonstrate that other parties that used the methodology were applying the codes and methodology correctly. The final test of using the methodology correctly is developing analytical factors that are consistent with the corresponding Westinghouse computations. This consistency is demonstrated by the measured rod worth comparisons.
Duke used the CASMO-4/SIMULATE-3 MOX/SIMULATE-3K MOX codes to generate comparisons to the Westinghouse generated results that used the ALPHA/PHOENIX/ANC codes per the DRWM topical requirements. Dukes analysis showed that 3 percent of the computational results did not meet the criteria. All of the comparisons that did not meet the criteria were for predictions of the rod worth. The comparisons between the measured rod worth CASMO-4/SIMULATE-3 MOX /SIMULATE-3K MOX and the Westinghouse results demonstrated that the analytical factors developed using the CASMO-4/SIMULATE-3 MOX/
SIMULATE-3K MOX code very closely mirror the Westinghouse results. All of the measured rod worth comparisons met the acceptance criteria.
When the underlying causes of the computational results which did not meet the criteria were investigated, it was noted that the predicted and measured rod bank worth deviations were consistent with the differences in the predicted radial Hot Zero Power (HZP) power distribution between Westinghouse and Duke. Relative to the Westinghouse calculation, Duke under-predicts the power of the assemblies on the core periphery which results in a calculated lower rod worth for the associated rod banks (banks SA, CD, SD, and SC) and over-predicts the power of the assemblies in the center of the core which results in a calculated higher rod worth for the associated rod banks (banks CC, CA, and SB.) In all cases where the predicted rod worth computational results did not meet the criteria, Duke predicted a lower bank rod worth that was consistent with the radial power distribution difference between Westinghouse and Duke. Likewise, the impact of the radial distribution caused Duke to consistently calculate a lower total bank worth relative to the Westinghouse calculation since a greater number of rod banks are on the periphery.
The parameter of greatest interest for correct application of DRWM is the calculation of the analytical factor. Correct determination of the analytical factor is shown by close agreement in the measured rod worth comparisons. All of the measured rod worth comparisons met the acceptance criteria. Since all of the measured rod worth comparisons met the acceptance criteria and the deviations in the predicted rod worth comparisons were consistent with the radial power distribution predictions, the NRC staff finds the use of the CASMO-4/SIMULATE-3 MOX/SIMULATE-3K MOX code acceptable for use with the DRWM methodology.
The NRC staff finds the use of the CASMO-4/SIMULATE-3 MOX/SIMULATE-3K MOX methodology acceptable for use with the DRWM methodology for McGuire with LEU fuel and for Catawba with LEU fuel and up to four MOX LTAs.
4.0 RESPONSE TO DUKES COMMENTS ON DRAFT SE Dukes letter dated March 9, 2004, provided comments on the NRC staffs Draft SE. Those comments, and the NRC staffs response to them, are provided below.
Duke Comment No. 1 The NRC has restricted approval of the methodology for the use of up to four MOX fuel LTA at Catawba only. Duke has the following comments concerning this restriction.
(i)
NRC has chosen not to provide Duke with the approval that was sought for application of the methodology to partial MOX fuel cores. It is noted that NRC has not provided a technical basis for this action. If NRC restricts the methodology approval to four MOX fuel lead assemblies, Duke requests that NRC indicate in the SE what is considered necessary for extending that approval to larger-scale use of MOX fuel. For example, it could be assumed that NRC anticipates extending the approval to partial MOX fuel cores, provided that the results of the MOX fuel lead assembly program are satisfactory. If NRC has additional expectations beyond the lead assembly program, it is requested that NRC should make those expectations clear. As currently written, the SE provides no clarity on the application of the methodology to partial MOX fuel cores.
NRC Staff Response Dukes comment, as written, is not completely correct. The NRC staffs approval of the Topical Report extended to the use of the methodology for LEU fuel at Catawba and McGuire and to the use of four MOX LTAs in one of the Catawba units. The basis for the approval of the methodology for MOX LTAs in one of the Catawba units, and not in the McGuire units, is Dukes letter dated September 23, 2003, wherein Duke removed McGuire from the MOX LTA program and indicated that MOX LTAs would be used in one of the Catawba units.
The NRC staff is aware that industry core reload design and analysis practices are continually evolving. Considering the potential changes that may take place between the time of Dukes submittal of the Topical Report and the time of potential use of partial MOX cores, and other information that may be developed on the predictive capabilities of the code package, such as discussed in the sixth paragraph of Section 3.1 above and the following paragraph, the NRC staff elects to delay approval of the methodology for partial MOX fuel cores until such more specific information on the design of partial MOX cores becomes available.
Duke committed to place two LTAs in instrumented locations for the first and second core cycles as required by condition one. The purpose of taking the incore measurements of the LTAs is to be able to compare the measured results with the CASMO-4/SIMULATE-3 MOX calculated results to demonstrate the impact of using weapons grade material versus reactor grade material and to demonstrate that the reactor grade MOX database for calculating core reload design is appropriate for use with weapons grade MOX.
(ii)
Duke believes that the methodology approval for MOX fuel lead assemblies should not be constrained to one unit at Catawba only. As a practical matter, Duke intends to use MOX fuel lead assemblies at one Catawba unit only. However, this is not a nuclear analysis methodology issue. Furthermore, it is conceivable (though not likely) that MOX fuel lead assembly circumstances could change. The Duke report has justified application of the CASMO-4/SIMULATE-3MOX methodolgoy to MOX fuel at either McGuire or Catawba, assuming that the other necessary regulatory approvals are in place to support the loading of MOX fuel lead assemblies. Duke believes that the SE for the analytical methodology is an undesirable place for restrictions on the use of MOX fuel for reasons that have nothing to do with the methodology. At a minimum, if the Catawba-only restriction is retained, the SE should make it clear that the restriction on location of MOX fuel lead assembly use has no basis related to the analytical methodology, but is due to other considerations. In, Duke has included as markups that would make the MOX LTA approval applicable to all four units.
NRC Staff Response As noted above, Duke has removed McGuire from the MOX LTA program. Therefore, an explicit approval of the methodology for LTAs at McGuire would constitute approval of a methodology for a usage that the licensee indicates will never be exercised. As a matter of policy, the NRC staff elects not to issue such approvals. However, the NRC staff has not identified any technical issues that would preclude approval and use of this methodology for McGuire, had MOX LTAs been chosen for McGuire.
Duke Comment No. 2 The cc list should include McGuire Nuclear Station as well.
NRC Staff Response This report will also be distributed to the McGuire Mailing list.
Duke Comment No. 3 With respect to Sections 3.0, 3.3, and 4.0, of the Draft SE, the SIMULATE-3K MOX computer code is an integral part of the methodology for DRWM. In order to ensure clarity, the SIMIULATE-3K MOX code should be specifically mentioned. Duke has included clarifying markups in Attachment 2.
NRC Staff Response Duke proposes the addition of the following paragraph at the end of Section 3.0:
SIMULATE-3K MOX is an extension of SIMULATE-3K, which is used for analysis of core transients. The spatial neutronics models in SIMULATE-3K MOX are identical to those in SIMULATE-3 MOX. SIMULATE-3K MOX solves the transient neutron diffusion equation incorporating effects of delayed neutrons, spontaneous fission in fuel, alpha-neutron interactions from actinide decay, and gamma-neutron interactions from long term fission product decay. For the applications reviewed in Topical Report DPC-NE-1005P, SIMULATE-3K MOX is used only as part of the dynamic rod worth measurement (DRWM) methodology.
The first three sentences of the above paragraph appear in the Topical Report, Section 2.4, SIMULATE 3K MOX, as a description of the SIMULATE-3K MOX codes capabilities. The NRC staff issued a request for additional information (RAI) on Section 2.4 and Duke responded on September 12, 2002. The information in the last sentence of the proposed paragraph above is included in that RAI (No. 5) response. The NRC staff finds this description of the SIMULATE 3K codes capabilities to be acceptable for inclusion in the SE.
Dukes proposed changes to SE Sections 3.3 and 4.0 on this matter consist of adding SIMULATE-3K MOX, to the code package name. The NRC staff finds this to be consistent with the Topical Report and the NRC staffs review and, therefore, acceptable.
Duke Comment on Section 3.3, Paragraph 1 As currently written, the beginning of Section 3.3 could give the impression that meeting the criteria for comparison to Westinghouse results (e.g., 2%/25 pcm) is an absolute requirement for applying DRWM with non-Westinghouse codes. As noted in Dukes December 2, 2003 letter on DRWM (Canady to U.S. NRC), the absolute need to meet those criteria was modified by the Safety Evaluation Report on WCAP-13360. This point should be clarified in the current SE to avoid creating an impression that Duke has failed to meet the appropriate DRWM requirements. By addressing those limited instances in which the acceptance criteria were not met, Duke has satisfied the pertinent requirements. Duke has included a markup addressing this point in Attachment 2.
NRC Staff Response Duke proposes to add the following to the first paragraph of the SE, Section 3.3:
A subsequent Safety Evaluation of Westinghouse WCAP-13360 accepted the clarification that deviations from the above acceptance criteria (comparison to Westinghouse generated results) may be acceptable if appropriately justified.
Dukes proposed clarification is accurate and the NRC staff finds it acceptable for inclusion into the SE.
Duke Comment on Section 3.3, Paragraph 3 Duke requests that NRC provide proper context for the discussion of Duke under-predictions in the second-to-last paragraph of Section 3.3. The under-predictions are relative to another analytical method (Westinghouse calculations), not data. Duke has included a clarifying markup in Attachment 2.
NRC Staff Response The licensees submittals include (a) comparison of predictions of control rod worths made by Westinghouse analytical methods to predictions made by Duke methods and, (b) comparison of measurements of control rod worths determined by Westinghouse to those determined by Duke. Duke proposes to add the words Relative to Westinghouse, to the second sentence of the third paragraph to clarify that the discussion refers to a comparison of two analytical methods and not to a comparison of measured data. Dukes proposal is consistent with its discussion of the issue in its letter dated December 2, 2003, and is consistent with the NRC staffs understanding of the issue, and is, therefore, acceptable.
Duke Comment on Section 3.3 Duke considers it essential that the SE clarify that the DRWM methodology is approved for application to cores including, at a minimum, four MOX fuel lead assemblies. Duke has included a clarifying markup in Attachment 2.
NRC Staff Response Duke proposes the inclusion of the following paragraph in Section 3.3:
The NRC staff finds the use of the CASMO-4/SIMULATE-3 MOX/SIMULATE-3K MOX methodology acceptable for use with the DRWM methodology for McGuire and Catawba with LEU fuel and up to four MOX LTAs.
The NRC staff approves the CASMO-4/SIMULATE-3 MOX/SIMULATE-3K methodology for DRWM methodology for McGuire and Catawba with LEU fuel and for Catawba with up to four MOX LTAs and has added a clarifying statement to Section 3.3 to this effect.
As noted above, Duke has removed McGuire from the MOX LTA program. Therefore, an explicit approval of the methodology for LTAs at McGuire would constitute approval of a methodology for a usage that the licensee indicates will never be exercized. As a matter of policy, the NRC staff elects not to issue such approvals. However, the NRC staff has not identified any technical issues that would preclude approval and use of this methodology for McGuire, had MOX LTAs been chosen for McGuire.
Duke Comment on Section 4.0 In order to ensure clarity, the conclusion section should specifically address approval to use CASMO-4/SIMULATE-3 MOX/SIMULATE-3K MOX computer codes for DRWM. Duke has included clarifying markups in Attachment 2.
NRC Staff Response Duke proposed to add the following to Section 4.0:
In addition, the NRC staff concluded that the CASMO-4/SIMULATE-3 MOX/SIMULATE-3K MOX methodology can be applied to Catawba and McGuire DRWM.
This is essentially the same issue for SE Section 4.0 as discussed above for SE Section 3.3.
For the same reasons as discussed above, the NRC staff has added the same clarifying statement made in SE Section 3.3 to SE Section 4.0.
5.0 CONCLUSION
Duke submitted the Topical Report (Reference 1) and supplementary information in References 2, 3, 4, 5 and 6 for review by the NRC staff. The licensee performed extensive benchmarking using the CASMO-4/SIMULATE-3 MOX methodology. The licensees effort consisted of conducting detailed comparisons of calculated key physics parameters with measurements obtained from several operating cycles of Catawba and McGuire, the St. Laurent reactor in France, and several MOX critical experiments. These results were then used to determine the set of 95/95 (probability/confidence) tolerance limits for application to the calculation of the stated physics parameters.
Based on the review of the analyses and results presented in References 1, 2, 3 and 4, the NRC staff has concluded that the CASMO-4/SIMULATE-3 MOX methodology, as validated by Duke, can be applied to the Catawba and McGuire steady-state physics calculations for reload applications as described in the above technical evaluation. The NRC staff finds the use of the CASMO-4/SIMULATE-3 MOX/SIMULATE-3K MOX methodology acceptable for use with the DRWM methodology for McGuire with LEU fuel and for Catawba with LEU fuel and up to four MOX LTAs. The NRC staffs approval is limited to the range of fuel configurations and core design parameters as stated and referenced by the August 3, 2001, submittal. Introduction of significantly different fuel designs will require further validation of the above-stated physics methods for application to Catawba and McGuire by the licensee and will require review by the NRC staff. Additionally, the results of the LTA in-core performance and predictive capabilities of CASMO-4/SIMULATE-3 MOX/SIMULATE-3K MOX for weapons grade MOX will need to be demonstrated and submitted to the NRC for review and approval as part of any application for partial MOX cores.
This approval is subject to the conditions listed above in Section 1.0 that have been provided by Duke in Reference 4.
6.0 REFERENCES
1.
Letter from K. S. Canady, Duke Power to the U.S. NRC, Topical Report DPC-NE-1005P, Revision 0, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, August 3, 2001 (Proprietary). A non-proprietary version was submitted by letter dated August 6, 2001.
2.
Letter from K. S. Canady, Duke Power to the U.S. NRC, Response to Request for Additional Information - Topical Report DPC-NE-1005P, Revision 0, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, September 12, 2002.
3.
Letter from K. S. Canady, Duke Power to the U.S. NRC, Response to Request for Additional Information - Topical Report DPC-NE-1005P, Revision 0, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, November 12, 2002.
4.
Letter from M. S. Tuckman, Duke Power to the U.S. NRC, Physics Testing Program in Support of Topical Report DPC-NE-1005P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, June 26, 2003.
5.
Letter from K. S. Canady, Duke Power to the U.S. NRC, Topical Report DPC-NE-1005P, Revision 0, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, August 14, 2003 6.
Letter from K. S. Canady, Duke Power to the U.S. NRC, Additional Information Related to Duke Topical Report DPC-NE-1005P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX, December 2, 2003.
7.
David G. Knott, Malte Edenius, CASMO-4 Benchmark Against Critical Experiments, Proprietary, SOA-94/13, Studsvik of America, Inc., USA, 1994.
8.
David G. Knott, Malte Edenius, CASMO-4 Benchmark Against MCNP, Proprietary, SOA-94/12, Studsvik of America, Inc., USA, 1994.
9.
M. G. Natrella, Experimental Statistics, National Bureau of Standards Handbook 91, October 1966.
10.
U.S. Nuclear Regulatory Commission, Regulatory Guide 1.126, Revision 1, An Acceptable Model and Related Statistical Methods for the Analysis of Fuel Densification, March 1978.
11.
D. B. Owen, Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, SCR-607, Sandia Corporation, March 1963.
12.
ANSI-N15.15-1974, Assessment of the Assumption of Normality (Employing Individual Observed Values), October 1973.
Principal Contributors: A. Attard U. Shoop Date: August 20, 2004