ML15329A246

Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 - Request for Additional Information Related to the License Amendment Request to Transition to Areva Fuel

ML15329A246
Person / Time
Site:	Dresden, Quad Cities
Issue date:	01/19/2016
From:	Brown E A Plant Licensing Branch III
To:	Bryan Hanson Exelon Generation Co
Brown E A, NRR-DORL 415-2315
References
CAC MF5736, CAC MF5737, CAC MF5738, CAC MF5739
Download: ML15329A246 (12)

Text

Mr. Bryan C. Hanson Senior Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 19,2016 Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO) Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 SUBJECT: DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 AND QUAD CITIES NUCLEAR POWER STATION, UNITS 1AND2 -REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE LICENSE AMENDMENT REQUEST TO TRANSITION TO AREVA FUEL (CAC NOS. MF5736, MF5737, MF5738, AND MF5739) Dear Mr. Hanson: By application to the U. S. Nuclear Regulatory Commission (NRC) dated February 6, 2015, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15043A489), supplemented by letter dated September 1, 2015 (ADAMS Accession No. ML 15251A381), Exelon Generation Company, LLC (EGC, the licensee) submitted a license amendment request for Dresden Nuclear Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2, to transition to AREVA ATRIUM 10XM fuel. The NRC has been informed by AREVA that a core simulation code known as MICROBURN-B2 (MB2) used in the analysis of this submittal contains errors. The potential exists that MB2 code errors will impact this submittal since the EGC fuels transition analysis is based on assumptions the MB2 code is error free and the analyzed results are valid. Mr. Timothy Byam of your staff was advised by the NRC that EGC should supplement the submittal justifying use of the MICROBURN-B2 code. Supplemental information may enhance the NRC's understanding of any associated impacts which need to be factored into the staffs safety review. Furthermore, during the NRC staff review it was determined that additional information is required to complete its safety evaluation. The specific information being requested is addressed in the enclosure to this letter. In an email addressed to Mr. Timothy Byam on December 23, 2015, and a follow-up conversation with him on January 7, 2016, it was agreed that EGC would provide the NRC with responses to the request for additional information (RAI) no later than January 22, 2016. A proprietary (non-public) version of this cover letter/enclosure is under ADAMS Accession No. ML 15329A240. The NRC staff considers that timely responses to the RAls ensures sufficient time is available for staff review and to contribute toward the NRC's goal of efficient and effective use of staff resources.

B. Hanson -2 -If circumstances result in the need to revise the requested response date, please contact Mr. Russell Haskell at (301) 415-1129, or by email at Russell.Haskell@nrc.gov. Docket Nos. 50-237, 50-249, 50-254 and 50-265 Enclosure: Request for Additional Information cc w/encl: Distribution via Listserv Sincerely, fq-Y __:-,.--"" . Eva A. Brown, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation REQUEST FOR ADDITIONAL INFORMATION DRESDEN NUCLEAR POWER STATION (DNPS), UNITS 2 AND 3, QUAD CITIES NUCLEAR POWER STATION (QCNPS), UNITS 1 AND 2, TRANSITION TO AREVA FUEL DOCKET NOS. 50-237, 50-249. 50-254, AND 50-265 This document contained proprietary information pursuant to Title 10 of the Code of Federal Regulations Section 2.390. Redacted Proprietary information is identified by text enclosed within double brackets [[ ]]. 1. In Attachment 1 to the submittal dated February 6, 2015 (Agencywide Document Access Management System (ADAMS) Accession No. ML 15043A489), the licensee states that: ... the application of the AREVA methodologies to a representative transition core design of ATRIUM 10XM and OPTIMA2 fuel (i.e., QCNPS Units 1 and 2) is sufficient to demonstrate the applicability of the methodologies for AREVA ATRIUM 10XM fuel at both stations (i.e., DNPS and QCNPS). Provide a description demonstrating that the non-cycle specific accident analyses are applicable to the transition cores for each unit and provide a justification that the representative cycle values for thermal limits and transients are characteristic of typical values expected for the actual transition cores for each unit. 2. Tables 3.5 through 3.8 of Attachment 9 of the submittal appears to indicate that the critical power ratio (CPR) for OPTIMA2 fuel is consistently lower than that of ATRIUM 10XM fuel. For the transition core, address whether the OPTIMA2 fuel will be the limiting fuel from the standpoint of operating limit maximum CPR values calculated for the anticipated operational occurrences (AOOs). 3. Page 2-1 of Attachment 9 to the submittal, appears to indicate that under rated conditions the ATRIUM 1 OXM assembly has a lower flow compared to an OPTIMA2 assembly, at the same power. However, tables 3.5 through 3.8 in Attachment 9 appear to show that the CPR for OPTIMA2 fuel remains consistently lower than that of ATRIUM 10XM fuel. Given the identified flow for the ATRIUM 1 OXM assembly, address why the data appears to indicate CPR values for OPTIMA2 fuel are lower than that of the ATRIUM 10XM fuel. Enclosure

-2 -4. Page 4-5 of Attachment 5 to the supplement dated September 1, 2015 (ADAMS Accession No. ML 15251 A381 ), discusses that the low pressure coolant injection (LPCI) and low pressure core spray (LPCS) systems inherently have leakage flows which do not reach the intended injection or target location. Due to LPCl/LPCS flow leakage; Provide the following information: a. The percentage(%) of assumed leakage rate for each the LPCI and LPCS system flows; b. Address whether the assumed leakage rate for each the LPCI and LPCS system flows is conservative and bounds the actual leakage through each system; and, c. Provide the basis for the LPCI and LPCS system flow leakage. 5. On page 1-1 of Attachment 5 to the supplement dated September 1, 2015, states: [t]he break spectrum analyses documented in this report were performed for a core composed entirely of ATRIUM 10XM fuel at beginning-of-life (BOL) conditions. The actual transition core will include both ATRIUM 10XM and OPTIMA2 fuels. Assuming that there are appreciable geometrical differences between the two fuel designs causing different thermal-hydraulic conditions to exist within the assemblies in a transition core, and that the actual transition core will include both ATRIUM 10XM and OPTIMA2 fuels. Discuss the following: a. The assumption made in the los-of-coolant accident (LOCA) analysis that a core composed entirely of ATRIUM 10XM will conservatively bound the peak clad temperature (PCT) reported for the limiting break analysis, and, b. The conservatism of the limiting PCT of 2127 degrees Fahrenheit (° F), and whether this is bounding for either of the fuel types, i.e., ATRIUM 10XM or OPTIMA2. 6. The transition from OPTIMA2 to ATRIUM 10XM fuel on LOCA analysis is described in Attachment 5 to the supplement dated September 1, 2015. a. Provide a comparison of the values of PCT, limiting break-size (small or break LOCA), and break location for the current licensing-basis LOCA to that of the predicted values for the transition core. b. Address whether there are any significant changes in the parameters specified above in 6.a between the current licensing basis (CLB) and the representative transition core. If so, explain.

-3 -7. In Table 6.1 of Attachment 5 to the supplement dated September 1, 2015, the limiting PCT was reported as 2127 ° F. In Table 2.1 of Attachment 14 of the submittal dated February 6, 2015, the PCT was reported as 2138 ° F. Clarify why the higher PCT value of 2138 ° Fis not the limiting PCT. 8. Table 2.1 of Attachment 12 of the submittal dated February 6, 2015, provides a summary of the disposition of events for ATRIUM 10XM Fuel Introduction at QCNPS. Discuss whether any significant changes are observed in the transients and accidents analyzed, including potentially limiting events, with ATRIUM 10XM fuel in the core as opposed to OPTIMA2 fuel in the core. If so, discuss the specific events and reasons for the change. 9. Page 7-1 of Attachment 12 to the submittal dated February 6, 2015, indicates that the limiting American Society of Mechanical Engineers (ASME) over pressurization event was the feedwater controller failure with turbine bypass valves out-of-service and high neutron flux scram. For most of the boiling-water reactors, typically main steam isolation valve (MSIV) closure with high neutron flux scram is the limiting ASME over pressurization event. a. Address why the MSIV closure with high neutron flux scram is not the limiting over pressurization event at DNPS/QCNPS, and b. Discuss any significant changes in the results of the limiting ASME over pressurization analysis due to transition from OPTIMA2 to ATRIUM 10XM fuel. 10. Describe the methodology used to perform the anticipated transient without scram analysis, its applicability to mixed core of ATRIUM 10XM and OPTIMA2 fuels, and whether the computer code employed was NRG-approved. 11. Page 4-2 of Attachment 12 to the submittal dated February 6, 2015, states that QCNPS has implemented the Boiling Water Reactor Owners Group (BWROG) Long Term Stability Solution Option Ill (oscillation power range monitor (OPRM)). It was further stated that in cases where the OPRM system is declared inoperable backup stability protection (BSP) is provided in accordance with OG02-0119-260, "Backup Stability Protection for Inoperable Option Ill Solution," General Electric Nuclear Energy, July 17, 2002. Discuss the long-term stability solution option and BSP for DNPS. 12. Chapter 4.2 of NUREG-0800, "Standard Review Plan (SAP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," states that the fuel system safety review should provide assurance that the fuel system is not damaged as a result of normal operation and AOOs.

-4 -Appendix A to Chapter 4.2 indicates that earth quakes and postulated pipe breaks in reactor coolant system would result in external forces on the fuel assembly. It further states that fuel system coolability should be maintained and that the damage should not be so severe as to prevent control rod insertion when required during these low probability accidents. Attachment 9 to the supplement dated September 1, 2015 (ADAMS Accession No. ML 15251A381), summarizes the evaluation of fuel handling loads and structural evaluation of faulted conditions to confirm the structural integrity of the fuel assembly components. Regarding the structural integrity of the fuel assembly components under normal and faulted conditions, provide the following information: a. Details of the stress calculations performed to confirm the design margin to establish a baseline for adding accident loads. b. Details of the analysis and testing that determined the maximum axial handling of the load by the fuel assembly without yielding. c. Details of the rod bow analysis for ATRIUM 10XM fuel design and its impact on thermal margin assessment. d. Discuss the results of the rod bow analysis for the OPTIMA2 fuel design for each core and the impact of rod bow on thermal margin assessment. 13. Section 3.2.3 of Attachment 10 to the supplement dated September 1, 2015, addresses the overheating of fuel pellets. In relation to this section: a. Explain how radial depression of the thermal neutron flux is accounted for in defining the local volumetric heat generation rate. b. Explain the term "neutronic fuel assembly types." c. Explain how linear heat generation rate (LHGR) uncertainties are calculated using the stated methodology. d. Explain how fuel rod power histories are created using the stated methodology. Specifically discuss an equilibrium core design and the upcoming cycle. e. Explain why transients are randomly selected using the stated methodology. f. Discuss the methodology used for power measurement and operational uncertainties. Address whether this a deviation from the RODEX4 methodology.

-5 -14. Page 3-3 of Attachment 19 to the submittal dated February 6, 2015 (ADAMS Accession No. ML 15043A489), discusses the thermal-hydraulic compatibility methodology. Specifically, it discusses the methodology used to determine fuel assembly groupings. In relation to this section: a. Explain the process of [[ ]]. b. Describe how the core power distributions are input to the compatibility analysis code for the case where the [[ ]]; and, c. Address the assumption that [[ ]]. 15. Provide the dimensional data associated with the resident OPTIMA2 fuel, include; fuel channel dimensions, thickness, and length. 16. With the introduction of ATRIUM 10XM fuel to the Dresden and Quad Cities units, discuss for each unit how the licensee has accounted for sufficient clearance for the control rods and in-core instrumentation. 17. With the introduction of ATRIUM 10XM fuel to the Dresden and Quad Cities units, discuss for each unit the impacts of channel bow (as a function of burnup) on the clearances for control rods, in-core instrumentation, and adjacent assemblies. 18. Section 4 of Attachment 8 to the supplement dated September 1, 2015, discusses the AREVA CHF/CPR correlations for the co-resident fuel (OPTIMA2) used for thermal margin analysis. a. Page 4-1 states that [[ ]] 1. Provide a list of SPCB range of applicability with respect to pressure, inlet mass velocity, inlet sub-cooling, design local peaking and tested local peaking. 2. Compare the range of applicability of SPCB correlation with the range of applicability of ACE [AREVA advanced critical power correlation] ATRIUM 1 OXM correlation. For any parameter where the range of applicability is

-6 -different between the two correlations, specify the impact on the safety limit minimum CPR determination. b. Page 4-2 states that a penalty is applied to bring the [[ ]] Address how the Adder is developed and is implemented in the SPCB correlation. 19. Appendix G of Attachment 8 to the supplement dated September 1, 2015, states that [[ ]] a. Provide a basis for the quadrant flow uncertainty listed in Appendix G. b. Describe the expressions used in calculations demonstrating how the total pressure drop uncertainty and sub-assembly flow uncertainty are obtained. 20. Page 5-2 of Attachment 8 to the supplement dated September 1, 2015, states that [[ ]] a. Address whether the computed fluence gradient for each unit is exceeded in the channel measurement database. If yes, provide a typical calculation demonstrating the impact on SLMCPR calculations; b. Provide a quantitative summary of how the uncertainty due to the water cross in the OPTIMA2 design impacts the SLMCPR calculations; and, c. Address whether there any limitations associated with fuel channel bow uncertainty if the computed fluence gradient is not bounded by the channel measurement database. 21. Section 7 and Appendix C of Attachment 8 to the supplement dated September 1, 2015, addresses Core Neutronics and Neutronic Methods, respectively. Using these references address the following;

-7 -a. Page C-1 of the report states: [m]odels for nodal [[ ]] are used to improve the accurate representation of the in-reactor configuration. Explain how these identified models are used improve the accurate representation of the in-reactor configuration. b. Provide details of the interpolation methodology employed to produce Figures C-10 through C-14 for all sets of intermediate void conditions. c. Page C-4 of the report states that the MICR08URN-82 (M82) methodology models a wide range of thermal hydraulic conditions including EPU [Extended Power Uprate] and extended power/flow operating map conditions. Address how the M82 methodology models the various thermal hydraulic conditions and extended range power/flow operating conditions and provide the range. Indicate whether the use of M82 methodology, in this analysis, is free for the entire power/flow range. If not, provide the basis why the use of M82 methodology is acceptable for this fuels transition. 22. Page 1-1 of Attachment 11 to the supplement dated September 1, 2015, Version 2, of MICR08URN-82 (M82) is listed as part of the cycle design analysis. a. Address whether Version 2 of MICR08URN-82 is the same version listed in "Reactors: Evaluation and Validation of CASM0-4/MICROBURN-82," Siemens Power Corporation (dated October 1999). b. Discuss the methodology used in the following M82 modeling features; 1 . Explicit control blade modeling. 2. Explicit neutronic treatment of spacer grids. 3. Explicit thermal-hydraulic modeling of water rod flow. 23. Section 7.1 of Attachment 8 to the supplement dated September 1, 2015, states that for each transition cycle, shutdown margin is computed by performing restart solutions based on a shuffled core from a short window previous cycle condition. Discuss the impact on the shutdown margin for future cycles if the plants were to operate for either nominal or long cycles.

-8 -24. Section 7.2 of Attachment 8 to the supplement dated September 1, 2015, addresses LHGR monitoring of advanced fuel designs. From this discussion: a. Explain the explicit low power range monitor (LPRM) model, b. Explain the in-core monitoring system and how the model is used to account for perturbations to the local peaking factors of the rods surrounding the LPRM, and c. Explain how the rod power biases due to the presence of LPRM detectors are accounted for in the monitoring of LHGRs. 25. Appendix B of Attachment 8 to the supplement dated September 1, 2015, states that even though the multi-rod database used in the [[ ]] was obtained through third party organizations, the database and prediction uncertainties are not available to AREVA. Explain how the correlation has been independently validated by AREVA against public domain multi-rod data and proprietary data from ATRIUM-10 and ATRIUM 10XM test assemblies. 26. Figures 8-3 and 8-4 of Attachment 8 to the supplement dated September 1, 2015, exhibits comparisons of [[ ]]. In both these curves, there has been a shift in the data points outside the band for -0.05 (predicted -measured) for a range of void fraction 0.40 to 0.80. Provide the basis which validates the (Ohkawa-Lahey using ATRIUM-10 and ATRIUM 1 O XM Void Data) database considering the scatter of the data points. 27. Appendix C.2 of Attachment 8 to the supplement dated September 1, 2015, states the correspondence between the assembly powers of adjacent assemblies is quantified by a conservative multiplier as listed on Page C-5. Additionally, this multiplier is based on the correlation coefficient that is statistically calculated and shown in Figure 9.1 and 9.2 of EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASM0-4/MICROBURN-82," (ADAMS Accession No. ML003698495).

-9 -a. Provide the basis of the calculations used to derive the conservative multiplier shown on page C-5. b. In Section 8.2 of EMF-2158(P)(A), the report states a combination of uranium oxide (U02) and plutonium oxide (Pu02) bundles are used. For each unit address the application from a measurement with U02 and Pu02 to each core containing only U02 fuel. Describe the process used in this analysis. 28. Appendix D of Attachment 8 to the supplement dated September 1, 2015, describes the 1 Y2 group diffusion equation (page D-4). Address whether the first term of the equation should be and the v in the second term should be v2 to reflect fission rates for groups 1 and 2. 29. Appendix D of Attachment 8 to the supplement dated September 1, 2015, states that there are [[ ]] two-group cross sections. Explain what is meant by the [[ ]]. 30. Appendix F of Attachment 8 to the supplement dated September 1, 2015, summarizes the impact and treatment of fuel thermal conductivity degradation (TCD) with fuel burnup for licensing safety analyses such as AOOs, LOCA analyses. a. Address whether TCD was applied to the models provided in letters dated July 14, 2009 and April 27, 2012 (ADAMS Accession Nos. ML092010157 and ML 121220377, respectively). b. For each unit, discuss how AREVA intends to implement TCD models.

B. Hanson -2-If circumstances result in the need to revise the requested response date, please contact Mr. Russell Haskell at (301) 415-1129, or by email at Russell.Haskell@nrc.gov. Docket Nos. 50-237, 50-249, 50-254 and 50-265 Enclosure: Request for Additional Information cc w/encl: Distribution via Listserv DISTRIBUTION: PUBLIC LPL3-2 R/F RidsNrrDorlLpl3-2 Resource RidsNrrPMDresden Resource RidsNrrLASRohrer Resource RidsAcrsAcnw_MailCTR Resource RidsRgn3MailCenter Resource Sincerely, /RA BVaidya for/ Eva A. Brown, Senior Project Manager Plant Licensing Branch LPL 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNRRDssSnpb Resource RidsNRRDssSrxb Resource RidsNrrDorlDpr Resource MRazzaque, NRR MPanicker, NRR EBrown, NRR ADAMS Accession No.: ML 15329A246 *by email OFFICE DORL/LPL3-2/PM DORL/LPL3-2/PM DORL/LPL3-2/LA DSS/SRXB/BC(A) NAME RHaskell EBrown* SRohrer EOesterle* DATE 12/22/2015 12/22/2015 1/13/16 1/8/2016 OFFICE DSS/SNPB/BC DORL/LPL3-2/BC(A) DORL/LPL3-2/PM NAME JDean* JPoole (BVaidya for) EBrown DATE 1/8/2016 1/14/16 1/15/16 OFFICIAL RECORD COPY