ML11342A122

Millstone Power Station, Unit 2 - Response to Request for Additional Information Regarding the Cycle 21 Core Operating Limits Report

ML11342A122
Person / Time
Site:	Millstone
Issue date:	11/30/2011
From:	MacManus R K Dominion, Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME6365
Download: ML11342A122 (19)

Text

Dominion Nuclear Connecticut, Inc. jEIhf.4:Lt.Millstone Power Station DRope Ferry Road NOV 3 0 20flWaterford, CT 06385U. S. Nuclear Regulatory Commission Serial No. 11-620Attention: Document Control Desk NSSLA/WDC ROWashington, DC 20555 Docket No. 50-336License No. DPR-65DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 2RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THECYCLE 21 CORE OPERATING LIMITS REPORT (TAC NO. ME6365)Dominion Nuclear Connecticut, Inc. (DNC) submitted the Millstone Power Station Unit 2(MPS2) Cycle 21 Core Operating Limits Report (COLR) to the Nuclear RegulatoryCommission (NRC) in a letter dated May 19, 2011. The COLR includes the values ofcycle-specific parameter limits and is submitted to the NRC for information. In a letterdated October 24, 2011, the NRC transmitted a request for additional information (RAI)to DNC related to the MPS2 Cycle 21 COLR. DNC agreed to respond to the RAI byNovember 30, 2011.Attachment 1 provides DNC's response to the NRC's RAI. Attachment 2 provides the10 CFR 50.59 Screen supporting the Cycle 21 COLR as requested in RAI Question 5.The AREVA calculation requested in RAI Question 4 contains proprietary information.A non-proprietary version of the calculation is being prepared by AREVA NP. Thecalculation and the non-proprietary version of the calculation will be submitted byJanuary 31, 2012, as discussed with the NRC project manager.If you have any questions regarding this submittal, please contact Wanda Craft at (804)273-4687.Sincerely,R. K. MacManusDirector, Nuclear Station Safety and Licensing -MillstoneAttachments:1. Response to Request for Additional Information Regarding the Cycle 21 CoreOperating Limits Report2. 10 CFR 50.59 Screen Supporting the Cycle 21 Core Operating Limits ReportCommitments made in this letter:1. None Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRPage 2 of 2cc: U.S. Nuclear Regulatory CommissionRegion I475 Allendale RoadKing of Prussia, PA 19406-1415C. J. SandersProject Manager -Millstone Power StationU.S. Nuclear Regulatory CommissionOne White Flint North11555 Rockville PikeMail Stop 08-B3Rockville, MD 20852-2738NRC Senior Resident InspectorMillstone Power Station Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRATTACHMENT 1RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDINGTHE CYCLE 21 CORE OPERATING LIMITS REPORTDOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 2 Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRAttachment 1, Page 1 of 8Dominion Nuclear Connecticut, Inc. (DNC) submitted the Millstone Power Station Unit 2(MPS2) Cycle 21 Core Operating Limits Report (COLR) to the Nuclear RegulatoryCommission (NRC) in a letter dated May 19, 2011. The COLR includes the values ofcycle-specific parameter limits and is submitted to the NRC for information. In a letterdated October 24, 2011, the NRC transmitted a request for additional information (RAI)to DNC related to the MPS2 Cycle 21 COLR. This attachment provides DNC'sresponse to the NRC's RAI.INTRODUCTIONMPS2 has a 'fixed' Incore Instrument (ICI) system. The ICI system consists of 45arrays and each array consists of four levels of Rhodium detector segments withnominal positioning at 20%, 40%, 60% and 80% of the core height. Within the core, theICIs are located within Zircaloy thimble tubes. The thimble tubes are conduits whichprovide a means for quick removal and reinsertion of the ICls during refueling outagesand for centering and cooling of the ICIs within them.The industry has experienced radiation induced growth of Zircaloy instrument thimbletubes. Dominion contracted Westinghouse to replace the 45 instrument thimble tubeswith tubes that are 10.5 inches shorter than the original design. The shorter,replacement thimble tubes are necessary to ensure that the thimble tubes do notcontact the fuel assembly lower end fitting due to radiation induced growth at the end ofplant life. The replacement of the thimble tubes took place during the fall 2009 refuelingoutage (2R19) with Cycle 20 being the first cycle of operation with the replaced thimbletubes.During field fabrication of the replacement tubes in 2R19, Westinghouse cut 26 of the45 thimble tubes shorter than intended by 1.375 inches. By design, the ICIs should be'free hanging' within the thimble tubes. However, the shortened thimble tubes raisedthe possibility that some of the ICI strings were bottomed out and slightly misalignedfrom the ideal location.While some of the ICIs may still have been free hanging in the shortened thimble tubes,Dominion conservatively instructed AREVA to quantify the potential impact on theindications of core power distribution by assuming that the 26 affected ICI strings weremisaligned by the maximum amount of 1.375 inches. Any potential impacts wereaddressed in the AREVA cycle-specific setpoint analysis. For Cycle 20 operation, nochange was needed to the acceptable operation regions as defined in the COLR figures(i.e., tents) and the impact on FQN (or Linear Heat Generation Rate (LHGR)) wasaccommodated within the known conservatism of the methodology.

Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRAttachment 1, Page 2 of 8For Cycle 21 operation, a slight change in the Linear Heat Rate limiting condition foroperation (LCO) monitoring tent (COLR Figure 2.5-1, used only when monitoring withexcore detectors) and the use of a FQN penalty factor (used when monitoring with incoredetectors) were needed to account for the maximum possible misalignment of the ICIs.An associated 1.0025 penalty factor was included in COLR Section 2.5 for Cycle 21.Question IPlease provide a detailed description of the methodology used to determine the linearheat rate measurement. Is this methodology approved by the NRC, and is it describedin the documents referenced in TS 6.9.1.8b?DNC ResponseReference 1-1, which is listed as Reference 1 in MPS2 Technical Specifications (TS)6.9.1.8b, contains the approved methodology used to validate the INPAX-II methodusing PRISM results. In the NRC Safety Evaluation (SE) for the Reference 1-1 topicalreport, the use of INPAX-I1 for SAV95 application is identified as one of the SErestrictions for incore monitoring of Combustion Engineering design plants that use fixedincore detectors, and thus is appropriate for MPS2. A detailed description of theINPAX-II method which converts measured signals to power distributions is cited inReference 1-1 as Reference 11 (denoted here as Reference 1-2).REFERENCES1-1 EMF-96-029(P)(A) Volumes 1 and 2, "Reactor Analysis System for PWRs,Volume 1 -Methodology Description, Volume 2 -Benchmarking Results,"Siemens Power Coirporation, January 1997.1-2 XN-NF-83-01(P), "Exxon Nuclear Analysis of Power Distribution MeasuredUncertainty for St. Lucie Unit 1," Exxon Nuclear Company, January 1983.Question 2Describe the methodology used to generate a penalty factor to account for the impact ofthe offset ICI detectors on the linear heat rate measurement. Is this methodologyapproved by the NRC, and is it described in the documents referenced in TS 6.9.1.8b?

Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRAttachment 1, Page 3 of 8DNC ResponseA summary of the analytical procedure used to compute the misaligned ICI penaltyfactor, applied to the uncertainty on the FQN (or LHGR) power distribution peakingfactor, is presented below:* The NRC-approved core simulator code PRISM (Reference 2-1) was used togenerate predicted nodal power and activation rate information specific to theMPS2 Cycle 21 reactor core. Reference 2-1 is listed as Reference 1 in MPS2 TS6.9.1.8b. Nodal power and activation rate information was generated atnumerous axial points for each instrumented fuel assembly and at numeroustimes during core life.* The PRISM-generated activation rate information was used to generate pseudo-measured (or simulated) incore detector signals at both "nominal" and "offset" ICIdetector conditions throughout core life. The "nominal" detector configurationswere centered at the standard positions of core height. In the "offset" detectorconfiguration, the 26 identified incore detectors were conservatively offset by themaximum amount of 1.375 inches. For each incore detector, a pseudo-measured signal was generated in the nominal and offset configurations. Atvarious times in core life, using the INPAX-II methodology that is cited inReference 2-1, synthesized signals were used to create two power distributions.The "nominal" detector signals were used to generate a nominal pseudo-measured 3-D power distribution. This power distribution represents what thereconstructed power distribution would be if all detectors were in properalignment. The "offset" detector signals were used to generate an offset pseudo-measured 3-D power distribution. This power distribution represents what thereconstructed power distribution would be if all 26 identified detectors weremisaligned by the maximum amount." The relative difference between the reconstructed "nominal" and "offset" nodalpower distributions represents the potential error due to the misaligned detectors.This error was calculated for limiting reactor core locations which areinstrumented. The maximum under-prediction difference for limiting measuredlocations during any time in core life defines the maximum potential error due tothe offset detectors. This maximum error was applied to the uncertaintycalculated in Reference 2-1 and the amount over the TS measurement-calculational uncertainty factor was the additional penalty applied for this reload." The TS measurement-calculational uncertainty factor for FQN (or LHGR) is 1.07for the INPAX-II core monitoring system installed at MPS2 (Reference 2-1).Therefore, the additional penalty factor of 1.0025 will be applied to peak Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRAttachment 1, Page 4 of 8measured FQN (or LHGR), as determined by the INPAX-I1 core monitoringsystem, to account for the potentially misaligned incore detectors.The total uncertainty factor for FRT, including the impact of the offset ICI detectorstrings, was also evaluated and remains bounded by the criteria of 1.06 (6.0%),documented in Table 2.1 of Reference 2-1.Note that Reference 2-1 does not include a discussion regarding the development ofpenalty factors associated with the potential misalignment of the incore detectors.Since the development of penalty factors is not described in Reference 2-1, it is not partof the Reference 2-1 NRC approved methodology. The physical location of the incoredetectors relative to the core is considered an input to an analysis utilizing theReference 2-1 methodology which determines the predicted nodal power and activationrate information specific to the reactor core. As such, the physical location of the incoredetectors relative to the core is not considered an element of the Reference 2-1methodology.As discussed above, the Cycle 21-specific calculation includes separate cases thatutilize the Reference 2-1 methodology. The first case determined the predicted nodalpower and activation rate information specific to the reactor core assuming the incoredetector strings were at their nominal locations. The second case determined thepredicted nodal power and activation rate information specific to the reactor coreassuming 26 of the incore detector strings were offset by 1.375 inches. The maximumdifference in predicted nodal power and activation rate between these two casesprovides the basis for the penalty factors applied to the Cycle 21 core.In summary, the analysis to determine a conservative penalty on peak measured FQN (orLHGR) to account for ICI misalignment was performed using NRC-approved codes andmethods (PRISM, INPAX-II) described in the documents referenced in MPS2 TS6.9.1.8b. The approved methods do not preclude calculations and application of apenalty factor to address the location of ICI detectors.REFERENCE2-1 EMF-96-029(P)(A) Volumes 1 and 2, "Reactor Analysis System for PWRs,Volume 1 -Methodology Description, Volume 2 -Benchmarking Results,"Siemens Power Corporation, January 1997.

Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRAttachment 1, Page 5 of 8Question 3Is the impact of the offset ICI detectors on the revised acceptable operating region(Figure 2.5-1) conservative? Is the revised penalty factor specified in item 2.5conservative?DNC ResponseThe penalty factor of 1.0025 for the offset ICI detectors, which is discussed in theresponse to RAI Question 2, was applied to the setpoint verification calculations as aconservative bias on FQN. The setpoint verification calculations were performed inaccordance with AREVA topical report EMF-1961(P)(A), Statistical Setpoint/TransientMethodology for Combustion Engineering Type Reactors (Reference 3-1), which isReference 13 in MPS2 TS 6.9.1.8b.The operating region provided in Figure 2.5-1 of the Cycle 21 COLR was updated toprovide adequate margin with the application of the penalty. The impact of the offset ICIdetectors on the revised acceptable operating region is conservative.The penalty factor represents the additional margin needed to account for misalignedincore detectors, and ensures the INPAX-II measured nodal powers are not under-predicted.A conservative penalty factor was determined based upon the following conditions:" All 26 affected incore detector strings were conservatively assumed to be offsetby the maximum amount of 1.375 inches relative to the nominal configuration." The relative difference in the power reconstruction due to the offset detectorswas assessed using the INPAX-I1 methodology for limiting incore detectorlocations at multiple burnup intervals throughout the cycle." For Cycle 21, the maximum under-prediction in determining the "measured"nodal power as a result of the misaligned detectors is used to formulate thepenalty. The histogram in Figure 3-1 shows the resulting distribution of the"relative error in nodal power" between the misaligned and normal incoredetector configurations.REFERENCE3-1 EMF-1961(P)(A), "Statistical Setpoint/Transient Methodology for CombustionEngineering Type Reactors," Siemens Power Corporation, July 2000.

Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRAttachment 1, Page 6 of 8Millstone Unit 2 Cycle 21 Relative Error Distribution1600 -1400CA 1200 -UClo 000o 8000.60E Maximum under-:predictionZ 400 calculated200o0R[0Relative Error in Nodal Power [(offset -nominal) I offset]Figure 3-1Histogram showing the distribution of "Relative Error in Nodal Power" betweenthe misaligned and normal incore detector configurationsQuestion 4Provide documentation (e.g. any analysis or evaluation) of the impact of the offset ICIdetectors on the linear heat rate measurement.DNC ResponseThe analysis methodology, inputs, and results from the evaluation of the offset ICIdetectors are described in the responses to RAI Questions 2 and 3 above. Thetechnical document that calculates the penalty factor and provides the basis for theabove RAI responses is a proprietary AREVA engineering calculation.

Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRAttachment 1, Page 7 of 8Typically, documents such as calculations, cycle specific vendor deliverables, and 10CFR 50.59 screens/evaluations used to support licensing actions are made available forNRC review either at a Dominion facility, the vendor's local office, or at a vendor'sfacility. However, in this case, the information will be provided as requested.Accordingly, a non-proprietary version of the requested AREVA engineering calculationis being prepared by AREVA. The proprietary calculation, the associated non-proprietary version of the calculation, and the affidavit supporting withholding of theproprietary version from public disclosure will be provided by January 31, 2012.Question 5Please provide any Title 10 of the Code of Federal Regulations, Part 50.59 designchange screening and/or evaluations related to the offset ICI detectors.DNC ResponseAs discussed in the introduction, the penalty factor included in the Cycle 21 COLR wasthe result of an ICI thimble fabrication error that was introduced in the plant during2R19, prior to Cycle 20 operation. A summary of the 10 CFR 50.59 evaluationperformed for this change is included in the Reference 5-1 report. As part of the NRC'sinspection of MPS2 and MPS3 changes, tests, or experiments and permanentmodifications in November 2010, the NRC reviewed the Cycle 20 Reload SafetyEvaluation (RSE) prepared by Dominion (Reference 5-2) and the 10 CFR 50.59evaluation supporting the thimble tube error. Reference 5-3 was issued by the NRC atthat time, and both the Cycle 20 RSE and the 10 CFR 50.59 evaluation are listed asreviewed documents in Attachment A of Reference 5-3. NRC review of the fabricationerror is also mentioned in Reference 5-4.As mentioned above, documents such as calculations, cycle specific vendordeliverables, and 10 CFR 50.59 screens/evaluations used to support licensing actionsare typically made available for NRC review either at a Dominion facility or at a vendor'sfacility. In this case as well, the information is provided as requested. Accordingly,included in this response as Attachment 2 is a copy of the 10 CFR 50.59 screen (withnames and signatures redacted) which was performed in support of the Cycle 21 RSE(Reference 5-5), and which incorporated the penalty factor calculated by AREVA intothe Cycle 21 COLR. The 10 CFR 50.59 screen identified the use of NRC-approvedmethods.

Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRAttachment 1, Page 8 of 8REFERENCES5-1 Letter from R. K. MacManus (Dominion) to USNRC, "Dominion NuclearConnecticut, Inc. Millstone Power Station Units 1, 2, 3, and ISFSI 10 CFR 50.59,10 CFR 72.48 Change Report for 2008 and 2009, and the Commitment ChangeReport for 2009," June 30, 2010.5-2 Dominion Reload Safety Evaluation No. EVAL-ENG-RSE-M2C20, Rev. 1,"Reload Design for Millstone Unit 2 Cycle 20-Revised," approved November 12,2009.5-3 Letter from USNRC to Mr. David Heacock, "Millstone Power Station -NRCEvaluation of Changes, Tests, or Experiments and Permanent Modification TeamInspection Report 05000336/2010010 and 05000423/2010010," December 22,2010.5-4 Letter from USNRC to Mr. David Heacock, "Millstone Power Station -NRCIntegrated Inspection Report 05000336/2009005 and 05000423/2009005,"February 3, 2010.5-5 Dominion Reload Safety Evaluation No. EVAL-ENG-RSE-M2C21, Rev. 0,"Reload Design for Millstone Unit 2 Cycle 21," approved April 5, 2011.

Serial No. 11-620Docket No. 50-336RAI Response for MPS2 Cycle 21 COLRATTACHMENT 210 CFR 50.59 SCREEN RELATED TO SUPPORTING THE CYCLE 21 COREOPERATING LIMITS REPORTDOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 2 NDominion' 50.59/72.48 ScreenC5A00 -. 3 Pg 1Applicable Station Applicable Unit(s) Parent Document/ RevisionM North Anna Power StationE1 Surry Power Station E] Unit 1 [ Unit 2 EVAL-ENG-RSE-M2C21, Rev. 0[ Millstone Power Station E] Unit 3 El ISFSI (Reload Safety Evaluation forC] Kewaunee Power Station Millstone Unit 2, Cycle 21)Part I -Describe the Proposed Activity and Document Search ResultsA. Describe the proposed activity and scope of activities. Appropriate descriptive materials may be referenced or attached.This 50.59 Screening applies to the following activities / documents:.EVAL-ENG-RSE-M2C21, Rev. 0 (Reload Design for Millstone Unit 2 Cycle 21), and.LBDCR 1 I-MP2-004Reload Safety Evaluation (RSE) EVAL-ENG-RSE-M2C21, Rev. 0 (Reload Design for Millstone Unit 2 Cycle 21)supports Cycle 21 operation in all MODES (MODE I to MODE 6) as defined by the Millstone Unit 2 TechnicalSpecifications (Reference 1). The RSE is a Dominion fleet wide evaluation that is used to document the reloaddesign changes at Millstone. The RSE has been used at Millstone starting with Millstone Unit 2 Cycle 19 as perDominion Procedure NF-AA-NAF-200 (Reference 2). The Reference 2 Nuclear Analysis and Fuel (NAF)procedure refers to the Dominion Design Change Procedure (Reference 3) for reviews and evaluations related tothe RSE. Supporting evaluations for Cycle 21 are provided by AREVA in References 4, 5, 6 and 7 and in EVAL-ENG-RSE-M2C21, Rev. 0.EVAL-ENG-RSE-M2C2 1, Rev. 0 documents and evaluates the Millstone Unit 2 Cycle 21 core in order todemonstrate that the Cycle 21 core design and operation are acceptable and safe. The Cycle 21 core will be fueledwith 80 fresh MIL2-21(AA), 68 once-burned MIL2-20(Z), 68 twice- burned MIL2-19(Y), and 1 twice burnedMIB-8(W) fuel assemblies (Reference 4). All fuel in the Cycle 21 core is provided by AREVA. Reload MIL2-21is the seventh reload to utilize the High Thermal Performance (HTP) spacers and FUELGUARD'm debris-resistantlower tie plates.Based on the Cycle 21 Safety Analysis Report provided by AREVA (Reference 4) and other supportingdocumentation provided by AREVA (References 5, 6 and 7), the Cycle 21 core design and Cycle 21 operation inall MODES (MODE 1 to 6) are determined to be acceptable and safe. The safety analyses documented inReference 4 supports Cycle 21 operation up to a nominal core power level of 2,700 MWt for up to 16,275MWd/MTU. The Cycle 21 safety analyses are based on a Cycle 20 shutdown between14,300 MWdIMTU and 16,000 MWd/MTU. A detailed evaluation is provided in Reference 4 to support thefollowing areas for Cycle 21:* Mechanical evaluation,* Neutronics evaluation,* Thermal-hydraulic evaluation,* Setpoints verification, and* Standard Reviw wPlakr(SRP) Chapter 15 Safety Ahalyses (i.e., MP2 Final Safety Ahialysis R~pbrt(FSAR) Chapter 14).Key: DBLFPB -Design Basis Limit for a Fission Product BarrierForm No. 730943(Feb 2011) bominionO 50.59172.48 Screen59b72.48; ' CMAA40 -tacmn 3., .aeReload MIL2-21 is the seventh reload to contain High Thermal Performance (HTP) spacers and FUELGUARDdebris-resistant lower tie. plates and the third reload with the Alloy 718 High Mechanical Performance (HMP)bottom spacer to provide additional rod restraint at end of life. The only significant mechanical design changefrom the MIL2-20 reload was the use of chamfered pellets in the fuel rods (Reference 6).The first use of the HMvfP bottom spacer was in Cycle 19 (Batch MIL2-19) where the bottom zircaloy HTP spacerwas replaced with the HMP bottom spacer to provide additional rod restraint at end of life. As reported byAREVA in the Reference 4 Safety Analysis Report, this change has been thoroughly evaluated in Reference 6, itmeets the applicable design criteria and it has no adverse effects on the Safety Analysis. A feed batch of 80,MI12-21 fuel assemblies is used for Cycle 21 (vs. the batch size of 68 fuel assemblies used in recent cycles) sothat all core peripheral locations will contain a fuel assembly with the HMP bottom grid. It is anticipated that theCycle 21 core will be more resistant to the spinning fuel rod/grid-to-rod fretting failures seen in peripheral fuelassemblies in Cycles 17, 18 and 19. Additionally, the MJL2-21 feed batch is unlike previous feed batches in thattwelve of the MIL2-21 fuel assemblies (sub-batch AA6) are of very low enrichment (2.2 w/o) and they areintended for a maximum of 2 cycles of operation.LBDCR 1 -MP2-004 proposes the following three (3) revisions to Core Operating Limits Report (COLR):* Change 1: Cycle specific editorial change ('Cycle 20' changed to 'Cycle 21')* Change 2: In addition to the three uncertainty factors related to the Incore Detector Monitoring Systemthat are already specified in COLR Section 2.5, an additional penalty factor is used for Cycle 21 andnoted (a 1.0025 penalty factor is applied to account for the impact of the misaligned ICI detectors onthe linear heat rate measurement). This penalty is conservative and bounds the anomaly for the Cycle21 core (Reference 4).* Change 3: Based on the misalignment of the ICIs and the supporting analysis for Cycle 21 (Reference4), Break Point "E" on the Local Power Density-Limiting Condition for Operation (LPD-LCO) barnmust be moved from (+0.30, 65) to (+0.25, 65) to create sufficient margin to the limits. COLR Figure2.5-1 is revised to reflect the revised analyses and provide sufficient operating margin for positivevalues of ASI. Figure 6.5 of the Reference 4 Safety Analysis report illustrates the revised barn for theLPD-LCO.The LBDCR noted above is fully supported by the Reference 4 and 7 AREVA documents.REFERENCES:I. Millstone Unit 2 Technical Specifications.2. Dominion Administrative Procedure NF-AA-NAF-200, "Reload Management Process."3. Dominion Administrative Procedure CM-AA-DDC-201, Rev. 6, "Design Changes."4. ANP-2979, Rev. 001 "Millstone Unit 2 Cycle 21 Safety Analysis Report," dated March 2011.5. AREVA Engineering Information Record No. 51-9142594-000, "Millstone Unit 2 Cycle 21 Final FuelManagement Plan", dated August 25, 2010.6. ANP-2980P, Revision 0, PWR Fuel Design.Criteria Review for Millstone Unit 2 Reload MIL2-21 andKey: DBLFPB -Design Basis Limit for a Fission Product BarrierForm No. 730943(Feb 2011)

Dom~inioW 50.59/72.48 Screen.4 ,I , , .p Cycle 21 Assemblies (transmitted in FAB 11-43, dated January 14, 2011).7. M. M. Ruhland (AREVA) letter to R. W. Sterner (Dominion), "Review of Millstone Unit 2 COLR",FAB10-822, dated December 3, 2010.8. Millstone Unit 2 Final Safety Analysis Report (FSAR).9. EMF- 1961(P)(A) Revision 0, "Statistical Setpoint/Transient Methodology for Combustion EngineeringType Reactors," Siemens Power Corporation, July 2000.10. NRC Letter Dated March 29, 200 1, "Millstone Nuclear Power Station Unit No. 2 -Issuance ofAmendment [No. 255] RE: Fuel Centerline Melt Linear Heat Rate Limit (TAC No. MA9646)."11. SPC Report XN-NF-82-06(P)(A), Revision I and Supplements 2, 4, and 5, "Qualification of ExxonNuclear Fuel for Extended Burnup," Exxon Nuclear Company, October 1986.B. Search the Technical Specifications and SAR including documents 'Incorporated by Reference." Describe relevant SAR-described function(s), performance requirements, and methods of evaluation of the affected SSCs, and where this information isin the Technical Specifications and SAR, including documents "Incorporated by Reference."The MP2 Technical Specifications (Reference 1) Sections 2.1 (Safety Limits), 2.2 (Limiting Safety SystemSettings), 3/4.1 (Reactivity Control Systems), 3/4.2 (Power Distribution Limits and 6.9 (Reporting Requirements)were reviewed. The relevant design functions include the ability of the Cycle 21 core to satisfy the FSAR Chapter14 Safety Analysis requirements as well as the associated IOCFR50 Appendix A general design criteria. Thedesign functions / design bases noted below are also potentially affected by a reload core design:1. Fuel assembly mechanical design bases, including mechanical loads and fuel assembly handling.2. Reactor core design.3. Reactor internals design bases.4. Control element drive design bases.5. Nuclear design bases, including fuel burnup limits, reactivity coefficients, power distribution, andshutdown margin.6. Thermal and hydraulic design bases, including core coolability, hydraulic stability, fuel pellet andcladding temperature limits, departure from nucleate boiling, and hot channel factorsThe FSAR (Reference 8) Chapter 3 (Reactor) and Chapter 14 (Safety Analysis) were reviewed..C. Does the Activity Involve a change to the Operating Ucense or Technical Specifications? .Yes I NoIf the answer is YES, process Operating License or Technical Specification change according to the appropriate procedure.If the answer is NO, describe the basis for the conclusion.Basis:The analytical methodologies that support the RSE and the related LBDCR have been previously applied toMillstone Unit 2 and therefore, no NRC approvals and no changes toTechnical Specification Section 6.9(references to approved methods) are required. Also, no changes to the Millstone Unit 2 Operating License arenecessary to support Cycle 21 operation.Key: DBLFPB -Design Basis Limit for a Fission Product BarrierForm No. 730943(Feb 2011) 2iDominion050.59/72.48 ScreenI £A400 £-Atahe 3 t. P ag ofIn summary, no changes to the Millstone Unit 2 Operating License or Technical Specifications are required due tothe implementation of the MIL2-21 fuel product for the operation of Cycle 21 as described in EVAL-ENG-RSE-M2C2 1, Rev. 0 or the associated LBDCR.Part II -Identify Areas Requiring Written Documentation1. Does the proposed activity involve a change to a Safety Analysis?2. Does the proposed activity involve a change to an SSC(s) credited in the Safety Analyses3. Does the proposed activity involve a change to an SSC(s) that support SSC(s) credited in theSafety Analyses?4. Does the proposed activity involve a change to an SSC(s) whose failure could initiate a transient(e.g., reactor trip, loss of feedwater, etc) or accident?5. Does the proposed activity involve a change to SAR-described SSC(s) or procedure controls thatperform functions that are required by or otherwise necessary to comply with regulations, licenseconditions, orders or Technical Specifications?6. Does the activity involve a change to a method of evaluation described in the SAR?7. Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)8. Does the activity exceed or potentially affect design basis limit for a fission product barrier (DBLFPB)?If the answers to all of the questions are NO, answer PART III as Not Applicable, and proceed to PartIV. An Evaluation is not needed. IF any of the above questions are checked YES, continue to"2art Ill.0 Yes0 YesEl No[f NoEl Yes ED No[3 Yes 2 No0l YesC1 Yesl Yes0J YesZ NoONoONo0 NoPart Ill -Determine Whether the Activity Involves Adverse EffectsIf all the questions in Part II were answered NO, then N/A this block El N/AOtherwise, identify below the specific SAR-described design function (YES from questions 1-5), method of evaluation (YES fromquestion 6), test or experiment (YES from question 7), or DBLFPB (YES from question 8).111.1 SAR-Described Design FunctionsIf the activity does not involve a SAR-described design function, then N/A this blockDoes the activity have an adverse effect on the SAR-described design function?If the answer is YES an Evaluation is required. If the answer is NO, describe the basis for the conclusion.[_ N/A0l YesZ NoThe proposed activities, documented by EVAL-ENG-RSE-M2C21, Rev. 0 and the related LBDCR support thefollowing:1. The loading of the Cycle 21 core with 80 fresh (MIL2-21) fuel assemblies, and2. COLR changes related to Cycle 21.These changes do not adversely affect any Structure, System, or Component (SSC) and do not adversely affect theirrelated design functions as described in the Millstone Unit 2 FSAR (Reference 8).As noted previously, the fresh, MIL2-21 fuel is the seventh fuel batch/reload to utilize the High ThermalPerformance (HTP) spacers and FUELGUARD M debris-resistant lower tie plates. This fuel design was firstutilized at Millstone Unit 2 in the Cycle 15 reload. Additionally, the Cycle 21 loading pattern meets all applicablemechanical and functional design criteria for operation in all MODES. The HMP bottom spacer satisfactorilymeets all the design criteria of the previously-used HTP bottom spacer.Ke:D LP esg ai ii fora-issio Prdc Bare Form" No .... .....3(eb20Key: DBLFPB -Design Basis Limit for a Fission Product BarrierForm No. 730943(Peb 2011) i)Dominion 50.59/72.48 Screen50.524W tac -The design functions / design bases noted below are potentially affected by a reload core design. Based onevaluations provided by AR.EVA in References 4 and 6, these areas have been determined not to be adverselyaffected by the proposed change for MODE 1 to 6 conditions. :I. Fuel assembly mechanical design bases, including mechanical loads and fuel assembly handling.2. Reactor core design.3. Reactor intemals design bases.4. Control element drive design bases.5. Nuclear design bases, including fuel bumup limits, reactivity coefficients, power distribution, andshutdown margin.6. Thermal and hydraulic design bases, including core coolability, hydraulic stability, fuel pellet and claddingtemperature limits, departure from nucleate boiling, and hot channel factors.Reference 4 demonstrates the ability of the Cycle 21 core to satisfy the FSAR Chapter 14 Safety Analysisrequirements.As noted in Section 6.0 of Reference 4, AREVA performed a trip setpoint verification in accordance with theReference 9 NRC approved methodology. This trip setpoint verification is routinely performed on a cycle specificbasis in accordance with that methodology. As part of this cycle specific setpoint verification effort for Cycle 21,AREVA updated the TM/LP and LPD related uncertainty calculations that are inputs for this setpoint verificationeffort. This uncertainty calculation update was based on lessons learned from work performed on other CE NSSSsister plants. Incorporation of updated uncertainties into the cycle specific trip setpoint verification effort is notconsidered to be a reanalysis of an existing safety analysis event that would require the development of a10CFR50.59 evaluation in accordance with the NEI 96-07 guidance.As discussed in Section 5.12 of the Cycle 21 RSE, 26 of the 45 new ICI thimble tubes are 1.375 inches (1 3/8") tooshort relative to the remaining 19 ICI thimble tube locations, resulting in the potential for misalignment of the ICIs.The cycle specific trip setpoint verification supporting the Reference 4 AREVA Safety Analysis Report specificallyaddresses this ICI misalignment similar to that done for Cycle 20. Explicitly including the ICI misalignment in theCycle 21 trip setpoint verification is also not considered a reanalysis of an existing safety analysis event that wouldrequire the development of a 10CFR50.59 evaluation in accordance with the NEI 96-07 guidance. The change tothe LPD LCO barn resulting from this setpoint verification effort was necessary to ensure the 15.1 kw/ft initialcondition of the existing FSAR Chapter 14 safety analysis is maintained while monitoring linear heat rate using theexcore detector monitoring system.In summary, there are no aspects of the RSE or LBDCR that cause an adverse effect on a Design Function.Key: SSC -Structures, Systems, and ComponentsFnrm No. 730943(Feb 2011) 50.59/72.48 ScreenCM-AA-400 -Attachment 3, Page 6 of 7.11.2 Method of EvaluationIf the activity does not involve a change to a method, then NIA this block LI N/ADoes the activity result in an adverse change to a method of evaluation as described in the SAR that is usedin establishing the design bases or in the safety analyses? EQ Yes []NoIf the answer is YES an Evaluation is required. If the answer is NO, describe the basis for the conclusion (attach additionaldiscussion as necessary).Basis:The methodology used by AREVA in References 4 and 6 for Cycle 21 are unchanged from those used by AREVAin Cycle 20. The methodologies used in these evaluations are consistent with those listed in Section 6.9.8.1 b. ofthe Technical Specifications and Section 3 of the Core Operating Limit Report (Appendix 8.1 of the TechnicalRequirements Manual). Addressing the potential ICI misalignment and revising the TM/LP and LPD relateduncertainties are considered design inputs to the cycle specific trip setpoint verification that does not involve arevision to the Reference 9 setpoint methodology or any other method of evaluation described in the FSAR.Thus, the Safety Analysis supporting Cycle 21 operation (Reference 4) and the associated RSE, and LBDCR (seePart A) do not involve new or revised methodologies.111.3 Design Basis Limits for a Fission Product Barrier (DBLFPB)If the activity does not involve a change to a DBLFPB, then N/A this block [I N/ADoes the activity change or exceed a DBLFPB? C1 Yes 0 NoIf the answer is YES an Evaluation is required. If the answer is NO, describe the basis for the conclusion (attach additionalliscussion as necessary).Basis:It is verified in Reference 6 that the Batch MIL2-21 fuel product meets all applicable design criteria. The analysesreported in Reference 4 confirmed that Cycle 21 operation complies with all applicable Safety Limits and LimitingSafety System Settings. There are no changes to the DNBR, fuel temperature, fuel enthalpy, clad strain, or fuelburnup fuel design basis limits. In accordance with the Reference 10 NRC approval, the fuel centerline melt linearheat rate limit is calculated on a cycle by cycle basis for Millstone 2 using the Reference I 1 NRC approvedmethodology listed in the Technical Specifications, Core Operating Limits Report and FSAR Section 3.5. Theresults of that cycle specific fuel centerline melt linear heat rate limit assessment are contained in Section 6.2 ofReference 4. Since the cycle specific implementation of this limit is fully in accordance with the Reference 10NRC approval, this is not considered a change to the fuel centerline melt linear heat rate design basis limit. Inaddition, there are no changes to any RCS boundary or containment fission product barrier design basis limits.Therefore, the activities described in Part A of this Screen do not change or exceed a DBLFPB.111.4 Tests or ExperimentIf the activity does not involve a test or experiment, then N/A this block [ N/AIs the proposed test or experiment not described in the SAR AND does it utilize an SSC outside the reference bounds for design or isinconsistent with the analyses and description in the SAR? El Yes El NoIf the answer is YES an Evaluation is required. If the answer is NO, describe the basis for the conclusion.Basis:The proposed activities (EVAL-ENG-RSE-M2C21, Rev. 0 and the associated LBDCR):do not involve any tests orKey: SSC -Structures, Systems, and ComponentsForm No. 730943(Jun 2010)

Daominion' 50.59/72.48 ScreenMAS0-tabet Pag 7I *o 7-experiments not previously described in the FSAR. As such, the proposed change does not involve a test or expefimentwhere an SSC in a manner outside the design bases or inconsistent with the FSAR.PART IV ConclusionCheck all that apply1. An Evaluation is N NOT REQUIRED El REQUIRED (Provide 50.59/72.48 Evaluation in accordance with Subsection 3.3)2. A change to the SAR and/or any document "Incorporated by Reference" is:El NOT REQUIRED E REQUIRED (Process change in accordance with applicable procedure)Additional CommentsEVAL-ENG-RSE-M2C21, Rev. 0 documents the acceptability of the Cycle 21 feed fuel product (Batch MIL2-21)and Cycle 21 operation. As part of the development of the RSE, it was determined that an FSARCR update/changeis required. The implementation of the FSARCR is tracked by CA as noted in EVAL-ENG-RSE-M2C21, Rev. 0-Attachment 2 (Documents Required to be Updated List).There are no aspects of EVAL-ENG-RSE-M2C21, Rev. 0 or the related LBDCR that require a 50.59/72.48Evaluation.Thecompteted Screen is part of the documenOnetlvlt Iehange packagePreparer Name (Print) Prep .......... "Date03/31/11"3o-signer (only if Preparer is not qualified (Print) Co-signer Signature DateReviewer (Print) Revie er Si nature Date00ý ai iw ý03131111CKey: SSC -Structures, Systems, and ComponentsForm No. 730943(Feb 2011)