ML13094A077
| ML13094A077 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/02/2013 |
| From: | St.Onge R J Southern California Edison Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC ME9727 | |
| Download: ML13094A077 (35) | |
Text
ISOUTHERN CALIFORNIAEDISONAn EDISON INTERNATIONAL& CompanyRichard I. St. OngeDirector, Nuclear Regulatory Affairs andEmergency PlanningProprietary InformationWithhold from Public DisclosureApril 2, 201310 CFR 50.4U.S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001Subject:Docket No. 50-361Response to Request for Additional Information (RAI 11), Revision 1Regarding Confirmatory Action Letter Response(TAC No. ME 9727)San Onofre Nuclear Generating Station, Unit 21. Letter from Mr. Elmo E. Collins (USNRC) to Mr. Peter T. Dietrich (SCE), datedMarch 27, 2012, Confirmatory Action Letter 4-12-001, San Onofre NuclearGenerating Station, Units 2 and 3, Commitments to Address Steam GeneratorTube DegradationReferences:2. Letter from Mr. Peter T. Dietrich (SCE) to Mr. Elmo E. Collins (USNRC), datedOctober 3, 2012, Confirmatory Action Letter -Actions to Address SteamGenerator Tube Degradation, San Onofre Nuclear Generating Station, Unit 23. Letter from Mr. James R. Hall (USNRC) to Mr. Peter T. Dietrich (SCE), datedDecember 26, 2012, Request for Additional Information Regarding Responseto Confirmatory Action Letter, San Onofre Nuclear Generating Station, Unit 24. Letter from Mr. Richard J. St. Onge (SCE) to NRC Document Control Desk,dated January 21, 2013, Response to Request for Additional Information(RAI 11) Regarding Confirmatory Action Letter Response, San Onofre NuclearGenerating Station, Unit 2Dear Sir or Madam,On March 27, 2012, the Nuclear Regulatory Commission (NRC) issued a Confirmatory ActionLetter (CAL) (Reference 1) to Southern California Edison (SCE) describing actions that the NRCand SCE agreed would be completed to address issues identified in the steam generator tubesof San Onofre Nuclear Generating Station (SONGS) Units 2 and 3. In a letter to the NRC datedOctober 3, 2012 (Reference 2), SCE reported completion of the Unit 2 CAL actions andincluded a Return to Service Report (RTSR) that provided details of their completion.By letter dated December 26, 2012 (Reference 3), the NRC issued Requests for AdditionalInformation (RAIs) regarding the CAL response. SCE provided the response to RAI 11 in aProprietary InformationWithhold from Public DisclosureDecontrolled Upon Removal From Enclosure 2IP.O. Box 128San Clemente, CA 92672 Proprietary InformationWithhold from Public DisclosureDocument Control Desk-2-April 2, 2013letter dated January 21, 2013 (Reference 4). The response to RAI 11 was revised to addressquestions raised by the NRC during the public meeting on February 27, 2013. Enclosure 2 ofthis letter provides Revision 1 to the RAI 11 response.Enclosure 2 of this submittal contains proprietary information. SCE requests that thisproprietary enclosure be withheld from public disclosure in accordance with 10 CFR 2.390(a)(4).Enclosure 1 provides a notarized affidavit from Westinghouse, which sets forth the basis onwhich the information in Enclosure 2 may be withheld from public disclosure by the NRC andaddresses with specificity the considerations listed by paragraph (b)(4) of 10 CFR 2.390.Enclosure 3 provides the non-proprietary version of Enclosure 2.Enclosure 4 provides a list of commitments identified in this submittal. If you have anyquestions or require additional information, please call me at (949) 368-6240.Sincerely,~ ,4WLt-Enclosures:1. Notarized Affidavit2. Response to RAI 11, Revision 1 (Proprietary)3. Response to RAI 11, Revision 1 (Non-Proprietary)4. List of Commitmentscc: A. T. Howell Ill, Regional Administrator, NRC Region IVJ. R. Hall, NRC Project Manager, SONGS Units 2 and 3G. G. Warnick, NRC Senior Resident Inspector, SONGS Units 2 and 3R. E. Lantz, Branch Chief, Division of Reactor Projects, NRC Region IVProprietary InformationWithhold from Public DisclosureDecontrolled Upon Removal From Enclosure 2 ENCLOSURE 1Notarized Affidavit O WestinghouseU.S. Nuclear Regulatory CommissionDocument Control Desk11555 Rockville PikeRockville, MD 20852Westinghouse Electric CompanyNuclear Services1000 Westinghouse DriveCranberry Township, Pennsylvania 16066USADirect tel:Direct fax:e-mail:Proj letter:(412) 374-4643(724) 720-0754greshaja@westinghouse.comNF-SCE-13-10CAW- 13-3680April 1, 2013APPLICATION FOR WITHHOLDING PROPRIETARYINFORMATION FROM PUBLIC DISCLOSURESubject: Proprietary Content for, "Follow-on Response to NRC Confirmatory Action Letter RAI #11 forSONGS Unit 2" (Proprietary)The proprietary information for which withholding is being requested in the above-referenced report isfurther identified in Affidavit CAW-13-3680 signed by the owner of the proprietary information,Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basison which the information may be withheld from public disclosure by the Commission and addresses withspecificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission'sregulations.Accordingly, this letter authorizes the utilization of the accompanying affidavit by Southern CaliforniaEdison.Correspondence with respect to the proprietary aspects of the application for withholding or theWestinghouse affidavit should reference CAW- 13-3680, and should be addressed to James A. Gresham.,Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 428, 1000 WestinghouseDrive, Cranberry Township, Pennsylvania 16066.Very truly yours,Thomas Rodack, DirectorLicensing and Engineering ProgramsEnclosures CAW-13-3680AFFIDAVITCOMMONWEALTH OF PENNSYLVANIA:ssCOUNTY OF BUTLER:Before me, the undersigned authority, personally appeared Thomas Rodack, who, being by meduly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf ofWestinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in thisAffidavit are true and correct to the best of his knowledge, information, and belief:Thomas Rodack, DirectorLicensing and Engineering ProgramsSworn to and subscribed before methis Ist of April 2013Notary PublicCOMMONWEALTH OF PENNSYLVANIANotarial SealAnne M. Stegman, Notary PublicUnity Twp., westmoreland CountyMy Comrnmlion Expires Aug. 7,2016MEMBER, PENNSYLVANIA ASSOCIATION OF NOTARIES 2CAW- 13-3680(1) I am Director, Licensing and Engineering Programs, in Nuclear Fuel, Westinghouse ElectricCompany LLC (Westinghouse), and as such, I have been specifically delegated the function ofreviewing the proprietary information sought to be withheld from public disclosure in connectionwith nuclear power plant licensing and rule making proceedings, and am authorized to apply forits withholding on behalf of Westinghouse.(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of theCommission's regulations and in conjunction with the Westinghouse Application for WithholdingProprietary Information from Public Disclosure accompanying this Affidavit.(3) 1 have personal knowledge of the criteria and procedures utilized by Westinghouse in designatinginformation as a trade secret, privileged or as confidential commercial or financial information.(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations,the following is furnished for consideration by the Commission in determining whether theinformation sought to be withheld from public disclosure should be withheld.(i) The information sought to be withheld from public disclosure is owned and has been heldin confidence by Westinghouse.(ii) The information is of a type customarily held in confidence by Westinghouse and notcustomarily disclosed to the public. Westinghouse has a rational basis for determiningthe types of information customarily held in confidence by it and, in that connection,utilizes a system to determine when and whether to hold certain types of information inconfidence. The application of that system and the substance of that system constitutesWestinghouse policy and provides the rational basis required.Under that system, information is held in confidence if it falls in one or more of severaltypes, the release of which might result in the loss of an existing or potential competitiveadvantage, as follows:(a) The information reveals the distinguishing aspects of a process (or component,structure, tool, method, etc.) where prevention of its use by any of CAW- 13-3680Westinghouse's competitors without license from Westinghouse constitutes acompetitive economic advantage over other companies.(b) It consists of supporting data, including test data, relative to a process (orcomponent, structure, tool, method, etc.), the application of which data secures acompetitive economic advantage, e.g., by optimization or improvedmarketability.(c) Its use by a competitor would reduce his expenditure of resources or improve hiscompetitive position in the design, manufacture, shipment, installation, assuranceof quality, or licensing a similar product.(d) It reveals cost or price information, production capacities, budget levels, orcommercial strategies of Westinghouse, its customers or suppliers.(e) It reveals aspects of past, present, or future Westinghouse or customer fundeddevelopment plans and programs of potential commercial value to Westinghouse.(f) It contains patentable ideas, for which patent protection may be desirable.There are sound policy reasons behind the Westinghouse system which include thefollowing:(a) The use of such information by Westinghouse gives Westinghouse a competitiveadvantage over its competitors. It is, therefore, withheld from disclosure toprotect the Westinghouse competitive position.(b) It is information that is marketable in many ways. The extent to which suchinformation is available to competitors diminishes the Westinghouse ability tosell products and services involving the use of the information.(c) Use by our competitor would put Westinghouse at a competitive disadvantage byreducing his expenditure of resources at our expense.
4CAW- 13-3680(d) Each component of proprietary information pertinent to a particular competitiveadvantage is potentially as valuable as the total competitive advantage. Ifcompetitors acquire components of proprietary information, any one componentmay be the key to the entire puzzle, thereby depriving Westinghouse of acompetitive advantage.(e) Unrestricted disclosure would jeopardize the position of prominence ofWestinghouse in the world market, and thereby give a market advantage to thecompetition of those countries.(f) The Westinghouse capacity to invest corporate assets in research anddevelopment depends upon the success in obtaining and maintaining acompetitive advantage.(iii) The information is being transmitted to the Commission in confidence and, under theprovisions of 10 CFR Section 2.390, it is to be received in confidence by theCommission.(iv) The information sought to be protected is not available in public sources or availableinformation has not been previously employed in the same original manner or method tothe best of our knowledge and belief.(v) The proprietary information sought to be withheld in this submittal is that which iscontained in, "Follow-on Response to NRC Confirmatory Action Letter RAI # I forSONGS Unit 2" (Proprietary), dated April 1, 2013, being transmitted by SouthernCalifornia Edison letter and Application for Withholding Proprietary Information fromPublic Disclosure, to the Document Control Desk. The proprietary information assubmitted by SCE to the NRC is that associated with a response to NRC RAI #1 Iwithrespect to fuel-clad modeling and may be used only for that purpose.This information is part of that which will enable Westinghouse to:(a) Support for SONGS Unit 2 enabling SCE to responds to NRC RAIs.
CAW- 13-3680Further this information has substantial commercial value as follows:(a) Westinghouse can sell support and defense of analyses involving Westinghousefuel-clad modeling to other licensees, as necessary.(b) The information requested to be withheld reveals the distinguishing aspects of amethodology which was developed by Westinghouse.Public disclosure of this proprietary information is likely to cause substantial harm to thecompetitive position of Westinghouse because it would enhance the ability ofcompetitors to provide similar calculations and licensing defense services for commercialpower reactors without commensurate expenses. Also, public disclosure of theinformation would enable others to use the information to meet NRC requirements forlicensing documentation without purchasing the right to use the information.The development of the technology described in part by the information is the result ofapplying the results of many years of experience in an intensive Westinghouse effort andthe expenditure of a considerable sum of money.In order for competitors of Westinghouse to duplicate this information, similar technicalprograms would have to be performed and a significant manpower effort, having therequisite talent and experience, would have to be expended.Further the deponent sayeth not.
PROPRIETARY INFORMATION NOTICETransmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRCin connection with requests for generic and/or plant-specific review and approval.In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning theprotection of proprietary information so submitted to the NRC, the information which is proprietary in theproprietary versions is contained within brackets, and where the proprietary information has been deletedin the non-proprietary versions, only the brackets remain (the information that was contained within thebrackets in the proprietary versions having been deleted).
ENCLOSURE 3SOUTHERN CALIFORNIA EDISONRESPONSE TO REQUEST FOR ADDITIONAL INFORMATIONREGARDING RESPONSE TO CONFIRMATORY ACTION LETTERDOCKET NO. 50-361TAC NO. ME 9727Response to RAI 11, Revision 1(NON-PROPRIETARY)
RAI 11Please submit an operational impact assessment for operation at 70% power. The assessmentshould focus on the cycle safety analysis and establish whether operation at 70% power iswithin the scope of SCE's safety analysis methodology, and that analyses and evaluations havebeen performed to conclude operation at 70% power for an extended period of time is safe.The evaluation should also demonstrate that the existing Technical Specifications, includinglimiting conditions for operation and surveillance requirements, are applicable for extendedoperation at 70% power.RESPONSE -Revision INote: This response includes information requested in RAI 14 associated with the operationalimpact assessment for operation at 70% power. RAI 14 states: "Provide a summary dispositionof the U2C17 calculations relative to the planned reduction in power operation."SCE has evaluated the extended reduced power operation for its impacts on the Unit 2 Cycle 17reload core design and safety analysis. The power levels evaluated range from 50% to 100%rated thermal power, which bounds the planned operation at the 70% power level. Theassessments were performed in accordance with NRC approved SONGS reload methodologyand topical reports referenced in the UFSAR and Technical Specification (TS) 5.7.1.5, and theSONGS Core Reload Analyses and Activities Checklist procedure.The impacts of extended reduced power operation on Unit 2 Cycle 17 core design and reloadanalyses, including UFSAR Chapter 15 safety analyses are summarized in Table 1, the impactassessment table. The impact assessment table is organized consistent with the SONGS CoreReload Analyses and Activities Checklist procedure. For each analysis, the Reload Checklistitem number is listed in the second column from the left; when applicable, the second columnalso lists the UFSAR Chapter 15 safety analysis section number. The determination of impactfor each analysis is summarized in the right column of the table.Tables 1 and 2 were revised to provide additional details and clarification to address issuesraised during the February 27, 2013 public meeting. Revisions are annotated in the tables bychange bars.Safety Analysis MethodoloqyThe NRC approved safety analysis methods, as described in TS 5.7.1.5, are used to establishthe core operating limits specified in the Core Operating Limits Report (COLR) whichencompass from Mode 6 up to Mode 1 operation at the rated thermal power. Therefore,operating at the 70% power level is within the scope of SCE safety analysis methodology. Nochange to the safety analysis methodology is required for extended reduced power operation.Safety AnalysisThe reload and safety analyses determined to be impacted by extended reduced poweroperation were re-analyzed. The conclusions of the reload analyses, including safety analyses,for extended reduced power operation are as follows: (1) All safety analyses results meet theestablished acceptance criteria, and (2) The radiological dose consequences for all safetyanalyses remain bounded by the dose consequences reported in the UFSAR.ENCLOSURE 3Page 2 of 23 Technical SpecificationsThe existing TS, including limiting conditions for operation (LCO) and surveillance requirements,are applicable for extended operation at 70% power. The impact assessment for TSsurveillance requirements is described in the following section.Impact Assessment for Technical Specification Surveillance RequirementsThe TS surveillance requirements were evaluated for the impacts of reduced power operation.The evaluation concluded all TS surveillance requirements under the reactor core design andmonitoring program that would have been performed at approximately 82% power or at fullpower will be performed with the plant operating at approximately 70% power. The evaluation issummarized in Table 2.Two surveillance procedures related to monitoring Reactor Coolant System (RCS) flow wererevised to (1) reduce the mi.m .u.mpwer required to perform the sur-eillances from 85% to68% power, and to (2) acco~unt for the slightly inrGeased RCS flow uncertainty at reduc~ed poweraperat4iethat had been required to be performed only above 85% power were revised to requireperformance of the surveillances at 70% power. No other surveillances were identified to beimpacted by plant operation at 70% power.ConclusionsExtended reduced power operation at 70% power has been evaluated and determined to beacceptable with respect to Unit 2 Cycle 17 reload core design and safety analysis. Reloadanalyses needed to support reactor startup and operation at 70% power have been completed.All TS LCO and surveillance requirements under the reactor core design and monitoringprogram normally performed at or above 70% power will be performed with the plant operatingat approximately 70% power. The above evaluations demonstrate that the existing TSs,including limiting conditions for operation and surveillance requirements, are applicable forextended operation at 70% power.ENCLOSURE 3Page 3 of 23 Table 1 -Revision ISONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM CHECKLIST ITEM D# (UFSAR SECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT1 0.1 Reload Ground Rules (RGR) Review No change to analysis is required. No change to Rated Thermal Power (RTP). RGRstill addresses 0% to 100% RTP operation. RGR addresses the full range of powerindependent and power dependent operating parameters, including those applicable atreduced power. The RGR Analysis Value defines the maximum or minimum valuewhich must be bounded in the safety analysis. The number is not necessarilyequivalent to the value used in an analysis (or Technical Specification) but will beconservative with respect to that value. The RGR Analysis Value includes applicableuncertainties and margins for which the safety analyses must be bounding.2 1.1.3 Design Models and Depletions Re-analysis was performed to determine impact, and all results were acceptable.Calculation revised to document depletion at 50% power from Beginning of Cycle(BOC) to End of Cycle (EOC) and comparison to 100% power.The results are expected, since radial power distributions are primarily a function of fuelburnup distribution, burnable poison loading, and control rod configuration. Depletion atreduced power level and corresponding reduced moderator temperature has anegligible impact on core average radial power distributions. Since the radial powerdistributions changed negligibly at 50% power compared to 100% power, an extendedoperation at any power level between these two points would also yield insignificantchanges to these parameters.The lead fuel assembly (LFA) integrated radial power peaking factors remain below95% of the core maximum integrated radial power peaking factor at all times in life. Themaximum pin burnup remains below the peak pin burnup limit (60,000 MWD/T).As the radial poerve- distributions an;d- dis-tortio-n facntors havef been determinend to bevalid, no downstream analyses areimatdImpact of extended reduced power operation on generic axial shapes and scram curvesis addressed in Item 10 (1-D HERMITE model.)ENCLOSURE3Page 4 of 23 Table I -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM 1CHECKLIST ITEM FT# CAECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)3 1.1.4 Design Parameters and FR Versus No change to analysis is required. Radial power distributions and generic axial shapesPower remain applicable. Individual Control Element Assembly (CEA) worth, CEA bank worth,scram worth, peaking factors, distortion factors that are strongly dependent on the radialpower distribution remain applicable. Extended reduced power operation results in lessPu-239 inventory. As such, generic bounding parameters (i.e., Fuel TemperatureCoefficient (FTC), Moderator Temperature Coefficient (MTC), kinetics parameters)remain applicable. Critical Boron Concentrations (CBC) at Beginning of Cycle (BOC)are not affected. CBC at End of Cycle (EOC) is similar. Therefore, bounding boronconcentration requirements and Inverse Boron Worths (IBW) are not impacted.Representative design parameter and Fr values for Reload Analysis Report (RAR) arenot impacted.4 1.1.5 Physics Input to LOCA, TORC, and No change to analysis is required for the physics inputs to LOCA analysis and TORCFATES Analysis (including Pin code analysis. BOC, limiting boron concentration, reactivity are not affected. RadialCensus) power distribution and peaking data remain applicable. Generic LOCA and TORC inputparameters remain applicable.Re-analysis was performed for the physics input to Fuel Performance Analysis (FATES)code analysis. Radial fall-off curves, Fr, and fast flux data were regenerated forreduced power operation. Generic axial shapes remain applicable. The impact of therevised input on FATES is addressed in Item 19.5 1.1.6 Physics Input to Fuel Mechanical Re-analysis was performed to determine impact, and all results were acceptable.Design Calculation revised to provide power history data for AREVA Lead Fuel Assembly (LFA)mechanical design analysis. Also updated maximum core residence time forWestinghouse analysis. Other generic parameters for Westinghouse mechanicaldesign analysis remain applicable due to similar radial power distribution.6 1.1.7 Physics Input to ASGT No change to analysis is required. Physics Input to Asymmetric Steam GeneratorTransient (ASGT) is performed at EOC with most negative Technical SpecificationMTC. Calculations performed at multiple power levels (90%, 70%, 50%, and 20%).Due to similar power distributions, results remain applicable.7 1.1.8 Physics Input to Post-Trip Steam Line No change to analysis is required. Analysis performed at EOC. Radial powerBreak Analysis distributions (at the same power level and burnup) are essentially identical. The MTC istuned to the most negative Tech Spec value (-3.7E-4 Ak/k/°F). Cooling down addsreactivity. More reactivity is added cooling from 100% power (higher T-fuel and T-mod)than reduced power to lower temperatures (e.g., 5450F, 3000F, 2000F, 68°F)ENCLOSURE3Page 5 of 23 Table 1 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM CHECKLIST ITEM DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)8 1.1.9 Physics Input to CEA Ejection No change to analysis is required. Physics data in this analysis were generated atAnalysis multiple power levels and the reduced power operating range is covered. Since thereduced power operation results in power distributions essentially identical to thosefrom 100% power operation, the data generated from the original analysis areapplicable to reduced power operation.9 1.1.10 Physics Input to CEA Withdrawal No change to analysis is required. Calculations performed at multiple power levels.Radial power distributions (at the same power level and burnup) are essential identical.CEA worth remains applicable since it is strongly dependent on power distribution.Limiting axial power shapes from axial shape index (ASI) search remain applicable.10 1.1.11 1-D HERMITE Model Re-analysis was performed to determine impact, and all results were acceptable.Analysis is revised to establish applicability of the generic axial shapes used in thedesign analyses and applicability of the SCRAM curves used in the design analyses.Analysis also shows that depletion at reduced power leads to essentially the samelimiting shapes from ASI search as those selected for the analyses of the designdepletions.11 1.1.12 Physics Input to Steam Line Break No change to analysis is required. This EOC event begins at 0% power. Radial powerReturn-to-Power-f9-%Gye N 1 distributions (at the same power level and burnup) are essentially identical.12 1.1.13 FR Versus Temperature for Cooldown No change to analysis is required. Bounding distortion factors were determined basedEvents on multiple CEA configurations, temperature ranges at BOC and EOC. Radial powerdistributions (at the same power level and burnup) are essentially identical.13 1.1.14 Boron Requirement for SITs and No change to analysis is required. The case run for this calculation is performed at hotBAMU Tanks zero power (HZP). The Xenon starting condition is Hot Full Power (HFP) which isconservative.14 1.1.15 LOCA and Non-LOCA Source Term No change to analysis is required. This analysis tests the Cycle 17 conditions ofinterest against the parameters required for applicability of the LOCA and AlternativeSource Term (AST) source terms. The power level is used as a maximum not to beexceeded. Running Cycle 17 at reduced power results in less "short half-life" nuclides.Increase in "long half-life" nuclides due to extended calendar time is bounded by thelower production from extended reduced power.ENCLOSURE 3Page 6 of 23 Table 1 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM CHECKLIST ITEM(USA SECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)15 1.1.16 Tritium Production No change to analysis is required. Reduced power results in a decrease in tritiumproduction. The analysis at 100% power is conservative.16 1.1.17 STAR Physics Verification No change to analysis is required. This analysis uses BOC (HZP) conditions (Mode 3)for an assessment for $2C17 inclusion in the Startup Test Activity Reduction (STAR)program.17 1.1.18 Digital Setpoints Physics Data No change to analysis is required. The case sets encompass LCO and Limiting SafetySystem Settings (LSSS) ASI ranges. Power level does not impact axial shapessignificantly, so reduced powers are covered by the case set.18 1.1.19 Physics RAR Inputs Re-analysis was performed to determine impact, and all results were acceptable. RARhas been updated to reflect actual Cycle 16 EOC burnup and Cycle 17 reduced poweroperation.ENCLOSURE3Page 7 of 23 Table 1 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM {CHECKLIST ITEM 1I ( AECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)19 1.2.1 Fuel Performance Analysis (FATES) Re-analysis was performed to determine impact, and all results were acceptable.Reduced power results in fuel performance data that is not bounded when compared tothe Generic Fuel Performance data generated for ZIRLOTM in Cycle 14 (data used inLOCA Analysis). A revision to the Fuel Performance and Setpoints Analyses wasperformed to determine the appropriate penalty factors such that the Generic FuelPerformance data remained bounding.Operation at reduced power impacts several of the fuel modeling parameters andmechanisms within the FATES fuel performance code.Fuel Performance results are utilized in the ECCS LOCA Analysis (item #71).The Fuel Performance results are also used in the CEA Ejection Analysis (item #44).New data were transmitted and addressed in that analysis.20 1.2.2 T-H Input Summary No change to analysis is required. Calculation is a collection of input data that are notimpacted by reduced power.ENCLOSURE 3Page 8 of 23 Table 1 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM D ICHECKLIST ITEM T SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)21 1.2.4 T-H Limiting Assembly and CETOP No change to analysis is required. Power is not an input. Calculation is a benchmark ofBenchmarking Analysis CETOP to TORC computer codes at reference departure from nucleate boiling (DNBR)points rather than a benchmark at a given power. This benchmark is mainly driven bypower distributions from physics. Physics Models & Depletions has validated the powerdistributions used in the original calculation.22 1.2.5 Mechanical Design Analysis (Fuel Re-analysis was performed to determine impact, and all results were acceptable.Vendor) Westinghouse performed calculations to determine the impact of reduced power on thefuel mechanical design.AREVA performed calculations to determine the impact of reduced power on the LeadFuel Assembly fuel mechanical design.23 1.2.6 Power Operating Limit Partial No change to analysis is required. The calculation is driven by a large family of axialDerivative Verification shapes, which are not impacted by the power reduction.24 1.2.7 Setpoints Input Summary Re-analysis was performed to determine impact, and all results were acceptable.Calculation has been revised to address the increased reactor coolant system (RCS)flow uncertainty at reduced power.25 1.2.8 RCS Flow Uncertainties Re-analysis was performed to determine impact, and all results were acceptable. Hasbeen reanalyzed. RCS flow uncertainty increases due to reduced delta-temperatureand increased secondary calorimetric power uncertainty. More details of this analysisare provided in the RAI 12 Response.26 1.2.9 Fuel Mechanical Design Verification No change to analysis is required. The objective of the fuel mechanical designverification calculation is to document the design of the fuel based on the fuel vendorBill of Materials, Design Drawings and the design and material specificationstransmitted from the fuel vendor. Reduced power operation has no impact on thisanalysis.27 1.2.11 Secondary Calorimetric Power No change to analysis is required. Intermediate powers were explicitly analyzed in theUncertainty original calculation.ENCLOSURE 3Page 9 of 23 Table I -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM] CHECKLIST ITEM# ( AECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)28 1.2.12 Delta-T/Turbine Power Uncertainties No change to analysis is required. The analysis uses a reference power error of 1.3%at full power. The increase in reference power (i.e., secondary calorimetric power)associated with performing delta-t/turbine power calibrations at reduced power wouldincrease the uncertainties. The bounding results include -0.50% of conservatism;therefore, the analysis of record (AOR) remains bounding. Intermediate powers wereexplicitly analyzed in the original calculation.29 1.2.13 Cycle Independent Data and No change to analysis is required. CIDSAL provides cycle independent values to use orSetpoints Assumptions List (CIDSAL) to be verified in downstream analyses. Reduced power operation does not impact therequirements for downstream analysis verification. None of the calculations explicitlyperformed in the analysis section are dependent upon nominal plant operatingconditions or the power shapes/distributions at reduced power operation.30 1.2.16 Core Protection Calculator (CPC) No change to analysis is required. Intermediate powers were explicitly analyzed in theCalibration Allowances original calculation. Due to less decalibration, full power bounds lower power levels.31 1.2.17 Fuel Duty Index No change to analysis is required. Full power bounds lower power levels.32 1.2.18 T-H MSCU Verification No change to analysis is required. Power is not an input. Calculation is a verification ofresponse surface at reference DNBR points rather than a benchmark at a given power.33 1.2.19 CEA STAR Verification No change to analysis is required. Radial power distributions (at the same power leveland burnup) are essentially identical. At reduced power the plan is to continue tooperate with all rods out. The duration and depth of lead bank CEA insertion beyondthe typical all-rods-out position is monitored per the core follow procedure withnotification/action to review the conservative CEA life analysis when insertion exceedsan insertion assumption within the analysis.34 1.3.1 Summary of Transients Re-analysis was performed to determine impact, and all results were acceptable.Calculation was revised to perform an evaluation of all Updated Final Safety AnalysisReport (UFSAR) Chapter 15 events for extended reduced power operation. Individualevents are addressed in subsequent entries to this table.35 1.3.2 CENTS Cycle Update and Action No change to analysis is required. Calculation and associated computer files alreadyModules accommodate power levels from 0 to 100 percent.ENCLOSURE3Page 10 of 23 Table 1 -Revision ISONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM CHECKLIST ITEM DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)36 1.3.3 (15.10.1.3.1.1) Main Steam Line Break (MSLB) No change to analysis is required. Pre-trip SLB is analyzed @100% power (withPre-Trip uncertainty). The generic physics inputs remain unchanged. Since the VOPT isgenerated on the rate of change in power setpoint (DELSPV), the actual trip occurs atthe same power rise, independent of the starting power level. As this is a RequiredOver Power Margin (ROPM) event, the actual initial power level chosen is notsignificant to the event.37 1.3.4 (15.10.1.3.1.2) Main Steam Line Break (MSLB) Post- No change to analysis is required. This event is limiting at hot zero power (HZP). HZPTrip cases show greatest return to power since there is minimum initial stored energy, decayheat and scram worth at HZP conditions. There is no impact to the HZP cases sinceHZP physics inputs and initial conditions do not change. A reactivity balance forreduced power showed that net reactivity change remained negative.HZP cases result in greatest return to criticality since initial stored NSSS energy, decayheat, and scram worth are minimized while steam generator pressure and mass are atmaximum. This minimizes RCS mass and maximizes cooldown potential.Consequently, the HZP MSLB event bounds MSLB initiated from power conditions.Operation at intermediate power levels does not alter these key parameters that makeHZP limiting.38 1.3.5 (15.10.4.1.4) Chemical Volume Control System No change to analysis is required. This is a BOC event that is not analyzed in Mode 1.(CVCS) Malfunction -Boron Dilution The reactivity addition due to a boron dilution event is less adverse than the CEAWithdrawal event at Power and therefore Mode 1 and the higher power portion of Mode2 are not explicitly addressed.39 1.3.6 (15.10.4.1.1) CEA Bank Withdrawal from No change to analysis is required. Event is evaluated at subcritical conditions. NoteSubcritical (CEAW @ SC) that this event is being re-evaluated to address the extended shut down.40 1.3.6 (15.10.4.1.1) CEA Bank Withdrawal at Low Power No change to analysis is required. Event is evaluated at hot zero power conditions.(CEAW @ HZP)41 1.3.6 (15.10.4.1.2) CEA Bank Withdrawal at Power No change to analysis is required. CEAW at reduced power is enveloped by CEAW @(CEAW @ Power, 50% & 100%) 50% Power and CEAW @ 100% Power; and the results are acceptable.42 1.3.8 (15.10.1.1.3) Increased Main Steam Flow (IMSF) No change to analysis is required. The system response is the same as IMSF+SF.ENCLOSURE 3Page 11 of 23 Table 1 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM TCHECKLIST ITEM# C RECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)43 1.3.8 (15.10.1.2.3) IMSF with Single Failure (SF) No change to analysis is required. IMSF+SF (fast & slow) analyzed @100% power.The generic physics inputs remain unchanged. The fast case credits the VOPT whichis generated on the rate of change in power (DELSPV) setpoint, as such the actual tripoccurs at the same power rise, independent of the starting power level. Since the fastcase is a Required Over Power Margin (ROPM) event, the actual initial power levelchosen is not significant to the event. The limiting event is the slow trip, which isinitiated from a Power Operating Limit. As such, the actual initial power level chosen isnot significant to the event.44 1.3.9 (15.10.4.3.2) CEA Ejection Re-analysis was performed to determine impact, and all results were acceptable. Theevent is normally analyzed at multiple power levels. It was reanalyzed to addressreduced power data from the fuel performance analysis.The CEA Ejection event is normally analyzed at multiple power levels according to thepower dependent insertion limits (PDIL) using fuel performance data based on 100%power operation. As discussed in Item 19,used inthe CEA Ejection analysis was determined with a conservative approach, the event wasre-analyzed to address the impact on fuel performance from reduced power operation.The fuel performance analysis (Item 19) generated additional fuel performance databased on reduced power operation at 70% and at 50%. The CEA Ejection cases werere-analyzed using the fuel performance data based on 70% power operation, as well asthose based on 50% power operation. The re-analysis showed1.The re-analysis bounds the planned reduced power operation for Unit 2 and all resultsmeet the acceptance criteria.45 1.3.10 (15.10.3.3.1) Reactor Coolant Pump Shaft Seizure No change to analysis is required. Bounded by Reactor Coolant Pump Sheared Shaft(RCPSS).ENCLOSURE3Page 12 of 23 Table 1 -Revision ISONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM 1ICHECKLIST ITEM D SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)46 1.3.10 (15.10.3.3.2) Reactor Coolant Pump Sheared Shaft No change to analysis is required. This is a margin/fuel failure calculation event. The(RCPSS) thermal margin loss for this event is initiated by the loss of flow from one pump (eitherseized rotor or sheared shaft). The reduction of thermal margin due to the loss of flowfrom one pump is not a function of the initial power (i.e., is constant at any power level).In addition, at reduced power, the initial thermal margin is larger than at the 100%power condition. Therefore, the analysis at full power is bounding.47 1.3.11 (15.10.2.1.3) Loss of Condenser Vacuum (LOCV) No change to analysis is required. Bounded by LOCV+SF48 1.3.11 (15.10.2.2.3) LOCV with Single Failure No change to analysis is required. This event is driven by plant response and not bydetailed core physics. There are two criteria (peak RCS pressure and peak secondarypressure). At lower powers, there is less internal energy in the reactor core, whichtranslates into a slower RCS pressure transient that is more rapidly mitigated by mainsteam safety valves (MSSVs). The peak secondary pressure event is evaluated atmultiple power levels to establish the allowed power level as a function of the number ofgagged MSSVs (Tech Spec 3.7.1).49 1.3.12 (15.10.6.3.2) Steam Generator Tube Rupture No change to analysis is required. The SGTR is a slow event and not sensitive to initial(SGTR) power. Furthermore, at lower powers there is a higher secondary pressure thattranslates to lower primary-to-secondary rupture flow (i.e., lower activity release).50 1.3.13 (15.10.1.1.4) Inadvertent Opening of a Steam No change to analysis is required. See IOSGADV+SFGenerator Safety or an AtmosphericDump Valve (IOSGADV)51 1.3.14 (15.10.1.2.4) IOSGADV with Single Failure No change to analysis is required. The IOSGADV+SF is analyzed at a power level of 1MWt.52 1.3.15 (15.10.9.1.1) Asymmetric Steam Generator No change to analysis is required. The ASGT event was analyzed in the AOR atTransient (ASGT) multiple power levels (90%, 70%, 50%, and 20%).53 1.3.16 (15.10.1.1.1) Decrease in Feedwater Temp No change to analysis is required. Since feedwater heating is reduced at reduced(DFWT) power, the potential loss in feedwater heating is also reduced. Impact at reduced poweris also mitigated by increased mass in RCS and Steam Generators (SGs) andincreased recirculation in SGs at lower power.ENCLOSURE 3Page 13 of 23 Table I -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM TCHECKLIST ITEMI ( AECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)54 1.3.17 (15.10.1.2.1) DFWT with Single Failure No change to analysis is required. Since feedwater heating is reduced at reducedpower, the potential loss in feedwater heating is also reduced. Impact at reduced poweris also mitigated by increased mass in RCS and Steam Generators and increasedrecirculation in SGs at lower power.55 1.3.18 (15.10.1.1.2) Increase in Feedwater Flow (IFF) No change to analysis is required. Primary to secondary heat transfer is dominated byheat of vaporization (Hfg) which is considerably greater than steam generator enthalpyrise resulting from sensible heat. Consequently, cool downs resulting from Increases inFeedwater Flow events are limited by Increases in Main Steam Flow events. Further,Increases in Steam Flow events occur more rapidly as changes in Feed Water aremitigated by the liquid mass and recirculation flow in the steam generators. Furtherfactors that mitigate Increasing Feedwater Flow events at reduced power includegreater RCS I SG mass, increased recirculation flow in the steam generators, greatersteam generator pressure and earlier reactor trip from increased feedwater flow -steamflow mismatch.56 1.3.18 (15.10.1.2.2) IFF with Single Failure No change to analysis is required. The most adverse single failure postulated for IFF isthe opening of all Steam Bypass Control System (SBCS) valves. Because the Increasein Main Steam Flow (IMSF) event postulates the opening of all SBCS valves andassumes that Main Feedwater flow increases to match steam flow, the IFF with SingleFailure is the essentially the same event as the IMSF event. Therefore, conclusionsregarding IMSF are applicable to IFF with Single Failure.57 1.3.19 (15.10.2.1.1) Loss of External Load (LOL) No change to analysis is required. The system response to the Loss of External Load,Turbine Trip, and the Loss of Condenser Vacuum are essentially the same. Therefore,the relationship between the events will remain the same at reduced power. As suchthese events remain bounded by LOCV.58 1.3.19 (15.10.2.2.1) LOL with Single Failure No change to analysis is required. The system response to the Loss of External Loadwith single failure, Turbine Trip with single failure, and the Loss of Condenser Vacuumwith single failure are essentially the same. Therefore, the relationship between theevents will remain the same at reduced power. As such these events remain boundedby LOCV+SF.59 1.3.19 (15.10.2.1.2) Turbine Trip (TT) No change to analysis is required. The system response to the Loss of External Load,Turbine Trip, and the Loss of Condenser Vacuum are essentially the same. Therefore,the relationship between the events will remain the same at reduced power. As suchthese events remain bounded by LOCV.ENCLOSURE 3Page 14 of 23 Table 1 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM fCHECKLIST ITEM 1ITEM SECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)60 1.3.19 (15.10.2.2.2) TT with Single Failure No change to analysis is required. The system response to the Loss of External Loadwith single failure, Turbine Trip with single failure, and the Loss of Condenser Vacuumwith single failure are essentially the same. Therefore, the relationship between theevents will remain the same at reduced power. As such these events remain boundedby LOCV+SF.61 1.3.20 (15.10.2.1.4) Loss of Normal AC Power (LONAC) No change to analysis is required. See LONAC+SF62 1.3.20 (15.10.2.2.4) LONAC with Single Failure No change to analysis is required. Operation at lower power level is less challengingwith respect to maintaining an adequate heat sink.63 1.3.21 (15.10.2.2.5) Loss of Normal Feedwater (LONF or No change to analysis is required. See LOFW+SFLOFW)64 1.3.21 (15.10.2.3.2) LOFW with Single Failure No change to analysis is required. Operation at lower power level is less challengingwith respect to maintaining an adequate heat sink.65 1.3.22 (15.10.2.3.1) Feedwater System Pipe Breaks No change to analysis is required. Peak primary and secondary pressure events were(FSPB or FWLB) analyzed at the least negative MTC value and main feedwater enthalpy correspondingto full power. The slightly higher MTC corresponding to reduced power is offset by thelower main feedwater enthalpy at reduced power. Operation at lower power level isless challenging with respect to maintaining an adequate heat sink. The energy in theplant is less at reduced power relative to full power, and therefore pressurizer overfill isbounded by the full power response.66 1.3.23 (15.10.5.1.1) CVCS Malfunction No change to analysis is required. See CVCS Malfunction+SF.67 1.3.23 (15.10.5.2.1) CVCS Malfunction with Single Failure No change to analysis is required. The energy in the plant is less at reduced powerrelative to full power, and therefore pressurizer overfill is bounded by the full powerresponse. Operation at lower power level is less challenging with respect tomaintaining an adequate heat sink.68 1.3.24 Pressurizer Spray Malfunction No change to analysis is required. See Core Protection Calculator (CPC) DynamicFilter Analysis.69 1.3.25 (15.10.4.1.5) Reactor Coolant Pump (RCP) -Start No change to analysis is required. Modes 1 and 2 were not analyzed becauseUp of an Inactive Loop operation in these Modes is only allowed with all 4 RCPs running.ENCLOSURE 3Page 15 of 23 Table 1 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM CHECKLIST ITEM# C RECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)70 1.3.27 (15.10.4.3.2) CEA Ejection (peak pressure No change to analysis is required. The event is limiting at hot zero power (HZP).analysis)71 1.4 (15.10.6.3.3) Emergency Core Cooling System Re-analysis was performed to determine impact, and all results were acceptable.(ECCS) Analyses including LBLOCA, Impact assessment addressed in analyses performed by Fuel Vendors.SBLOCA and LTC Details are provided in the response to RAI 13.72 (15.10.5.1.2) Inadvertent Operation of ECCS at No change to analysis is required. The system response to the IOECCS and CVCSPower (IOECCS) malfunction events are essentially the same. Therefore, the relationship between theevents will remain the same at reduced power. As such this event continues to bebounded by CVCS malfunction event.73 (15.10.5.2.2) IOECCS with Single Failure No change to analysis is required. The system response to the IOECCS with singlefailure and CVCS malfunction with single failure events are essentially the same.Therefore, the relationship between the events will remain the same at reduced power.As such this event continues to be bounded by CVCS malfunction event with singlefailure.74 (15.10.6.3.1) Primary Sample or Instrument Line No change to analysis is required. Mass releases are driven by energy in the primaryBreak (PSILB) system which is highest following operation at HFP. The event does not fail fuel, andthere is no ROPM requirement.75 (15.10.6.3.4) Inadvertent Opening of a PSV No change to analysis is required. The IOPSV event is bounded by small break LOCA.(IOPSV)76 1.5.1 Applicability Evaluation of Source No change to analysis is required. There is no change to core activity inventory sourceTerms in Dose Analyses term.77 1.5.2 Cycle Specific Dose Analysis No change to analysis is required. No Cycle 17 event-specific dose analysis wasperformed, therefore no impact for reduced power.ENCLOSURE3Page 16 of 23 Table I -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM CHECKLIST ITEMTI ( AECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)78 1.5.4 Applicability Evaluation of Dose Re-analysis was performed to determine impact, and all results were acceptable.Analyses Revised to document that the currently modeled radial peaking factors areconservatively greater than the increased radial peaking factors at reduced power.The transient analyses and mass release analyses are evaluated at the current 8%steam generator (SG) tube plugging limit. The dose calculation uses mass releasedata per the transient analyses and their assumed 8% SG tube plugging models. Thecalculation is revised with discretionary conservatism to model 20% SG tube plugging inthe calculation of the RCS dilution volume and mass considered for non-LOCA eventswhich have clad damage. Evaluated RCS dilution mass at RCS temperatures for both50% and 100% power, which envelopes powers between 50% and 100%.The mass release calculations are evaluated for a core inlet temperature (Tcold) of560F, which maximizes core average temperature (Tave). Currently modeled massrelease values in the Summary of Transients (SOT) correspond to full power operation.The SOT did not identify an increase in the amount of steam released from thesecondary side because it remains more limiting compared to operation at lower powerlevel due to lower sensible heat in the RCS and lower post trip decay heat.Technical Specification Action Statement Figure 3.4.16-1 allows for larger short termelevated primary coolant dose equivalent iodine-131 (1-131) activity for plant operationat 70% Rated Thermal Power (RTP). For UFSAR chapter 15 radiological safetyanalyses that model a pre-existing iodine spike, the SONGS licensing basis is an initialprimary coolant concentration of 60 pCi/gm dose equivalent 1-131 at 100% RTP.SCE commits to administratively control the RCS dose equivalent 1-131 specific activitydescribed in TS LCO 3.4.16 Action Al to no more than 60 pCi/gm (see Enclosure 4).No changes to these analyses are required for reduced power operation.ENCLOSURE 3Page 17 of 23 Table 1 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM] CHECKLIST ITEM FTI (SECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)79 n/a Fuel Corrosion and Oxide Thickness No change to analysis is required.(BOA Code) analysis The Westinghouse BOA code analysis for cycles 15, 16 and 17 was performed as partof the Zinc Injection project. This calculation compared predicted values for corrosionand oxide thickness, Fuel Duty Index and crud dryout to the Westinghouse ChemistryGuideline limits.Maximum values of Fuel Duty Index and Crud Dryout are driven by fresh fuel operatingat high power. Operation at reduced power would be bounded by the 100% powercases run in the analysis of record (AOR).Maximum values of corrosion and oxide thickness are driven by both power level andeffective full power days (EFPD). The AOR assumed a core operating strategy whichwould maximize corrosion and oxide; running fuel for three full cycles, a total of 1830EFPD. Table 2-1 of the AOR showed that the maximum predicted oxide thickness forU2C17 is 28.4 microns, well below the 100 micron limit. Operation at reduced power forlonger time would not significantly change the fuel rod corrosion rate, and there issubstantial margin to the 100 micron limit.80 n/a AREVA Lead Fuel Assembly (LFA) Re-evaluation was performed to determine impact, and all results were acceptable.compatibility Compatibility was verified by AREVA as documented in revised U2C1 7 Reload AnalysisReport (RAR).81 n/a WEC Lead Fuel Assembly (LFA) Re-evaluation was performed to determine impact, and all results were acceptable.compatibility Compatibility was verified by Westinghouse as documented in revised U2C17 RAR.82 n/a AREVA and WEC Chemistry Re-evaluation was performed to determine impact, and all results were acceptable.concurrence Concurrence for reduced power operation was performed by Westinghouse andAREVA as documented in revised U2C17 RAR.83 1.6.1 Reload Analysis Report (RAR) Re-analysis was performed to determine impact, and all results were acceptable.Revised to address extended operation at reduced power.84 1.6.2 Engineering Change Package (ECP) Re-evaluation was performed to determine impact, and all results were acceptable.and 1OCFR5O.59 Review 10CFR50.59: New 10CFR50.59 review issued to address the extended operation atreduced power.ECP: Affected Section Change (ASC) issued to address the extended operation atreduced power.ENCLOSURE3Page 18 of 23 Table I -Revision ISONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM 1CHECKLIST ITEM# (SECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)85 2.1.2 Physics Input to FLCEA Drop No change to analysis is required. Power distributions at the same power level andAnalysis and PFDTME Verification burnup are essentially identical. Analysis performed at multiple power levels.86 2.1.3 Physics Input to PLCEA Drop No change to analysis is required. Power distributions at the same power level andAnalysis burnup are essentially identical. Analysis performed at multiple power levels.87 2.1.5 Physics Input to CEA Deviation Within No change to analysis is required. Power distributions at the same power level andCPC Deadband burnup are essentially identical. Analysis performed at multiple power levels.88 2.1.9 Refueling Boron Concentration No change to analysis is required. Analyzed at BOC, Mode 6.89 2.1.10 CIDSAL Physics Verification No change to analysis is required. Radial power distributions (at the same power leveland burnup) are essentially identical. T-inlet program remain unchanged.90 2.2.1 (15.10.4.1.3) CEA Misoperation -Deviation within No change to analysis is required. Power distributions at the same power level andDead Band (DWDB) burnup are essentially identical. Analysis performed at multiple power levels.91 2.2.2 (15.10.4.1.3) CEA Misoperation -PLR Drop -No change to analysis is required. Power distributions at the same power level andPower _ 50% burnup are essentially identical. Event scenario is defined at _< 50% Power. Scenariosat >50% power are discussed in "CEA Misoperation -Single Part Length CEA Drop(PLR Drop) -Power > 50%."92 2.2.3 (15.10.4.1.3) CEA Misoperation -Single Full No change to analysis is required. Power distributions at the same power level andLength CEA Drop (FLCEA Drop) burnup are essentially identical. Analyzed at multiple power levels.93 2.2.3 (15.10.4.1.3) CEA Misoperation -Single Part No change to analysis is required. Power distributions at the same power level andLength CEA Drop (PLCEA Drop) -burnup are essentially identical. Analyzed at multiple power levels.Power > 50%94 2.2.3 (15.10.4.1.3) CEA Misoperation -Sub Group CEA No change to analysis is required. Power distributions at the same power level andDrop burnup are essentially identical. Analyzed at multiple power levels.95 2.2.4 AOPM Analysis No change to analysis is required. Power distributions at the same power level andburnup are essentially identical. Analyzed at multiple power levels.*96 2.2.5 Transient Thermal Margin Summary No change to analysis is required. Analyzed at multiple power levels.ENCLOSURE 3Page 19 of 23 Table 1 -Revision ISONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM jCHECKLIST ITEM [II (SECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)97 2.2.6 (15.10.3.1.1) Partial Loss of RCS Flow (PLOF) No change to analysis is required. Bounded by TLOF.98 2.2.6 (15.10.3.2.2) PLOF with Single Failure No change to analysis is required. Bounded by RCPSS.99 2.2.6 (15.10.3.2.1) Total Loss of Forced Reactor Coolant No change to analysis is required. The total loss of coolant flow event was analyzed forFlow (TLOF) a bounding scenario at 100% power and a MTC of +0.5x10-4 Ap/°F. This scenariobounds all powers from 0 to 100%.100 2.2.6 (15.10.3.3.3) TLOF with Single Failure No change to analysis is required. Bounded by RCPSS.101 2.2.7 CPC Dynamic Filter Analysis No change to analysis is required. The bounding events considered include CEA(including the Pressurizer Spray Withdrawal, Excess Load events, etc. As the system response time for these eventsMalfunction) has not changed, the dynamic filter analysis remains conservative.102 2.3.4 MSOUA Database and Files No change to analysis is required. The impact of RCS flow uncertainty changes hasbeen captured in MSOUA Post-Processor.103 2.3.5 CPC Reload Data Block (RDB) No change to analysis is required. Reduced power has been implemented throughUpdate CPC Type 2 addressable constants, and not CPC RDB.104 2.3.6 MSOUA Post Processor Re-analysis was performed to determine impact, and all results were acceptable.Calculation has been revised for RCS flow uncertainty and the change in UNCERT fromthe FATES fuel performance analysis.The COLSS and CPC Departure from Nucleate Boiling-Ratio (DNBR) penaltiesincreased by approximately 2% due to the RCS flow uncertainty increase (see Item 25)and another 3% increase for licensee's discretionary conservatism; bringing the totalDNBR penalty increase to approximately 5%. The 70% RTP based COLSS LinearHeat Rate (LHR) penalty increased by approximately 8% due to an increase in the fuelperformance uncertainty factor from the revised fuel performance analyses (see Item19). Discretionary conservatism of 3% was added to the COLSS LHR penalty for atotal COLSS LHR penalty increase of approximately 11%. The CPC Linear PowerDensity (LPD) penalty increased by 3% due to the addition of 3% discretionaryconservatism.These COLSS and CPC overall uncertainty analysis based penalty increases areacceptable because adequate core DNBR and Local/Linear Power margins exist.ENCLOSURE 3Page 20 of 23 Table I -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM CHECKLIST ITEM D SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)105 2.3.7 Core Operating Limits Supervisory No change to analysis is required. Calculation is a prediction of operating margin at fullSystem (COLSS) & CPC Operating power. Reduced power increases operating margin.Margin Assessment106 2.3.8 COLSS Database No change to analysis is required. No changes are being made to the manner in whichCOLSS functions or responds. Therefore the cycle independent constants do notrequire change. The installed Primary AT power Block I constants were verified to bebounding. The cycle specific constants that are impacted by reduced power operationhave been addressed in the COLSS As-built Database and Test Cases calculation.107 3.1.1 Full Core Load Map No change to analysis is required. Fuel management not changed.108 3.1.3 As-Built Models and Depletions Re-analysis was performed to determine impact, and all results were acceptable.Calculation was revised to address extended reduced power operation and to verifyLead Fuel Assembly (LFA) compatibility operational requirements.109 3.1.4 CECOR Coefficients Impacted, and all results were acceptable. Calculation was revised to addressextended reduced power operation.110 3.1.5 As-Built Mini Depletion Re-analysis was performed to determine impact, and all results were acceptable.Calculation revised to address extended reduced power operation and to verify LFAcompatibility operational requirements.111 3.1.6 Decay Heat No change to analysis is required. Decay heat was evaluated at end of Cycle 16condition. The calculation specifically addresses outage times past 99 days.112 3.1.7 Simulator Data Re-analysis was performed to determine impact, and all results were acceptable.Calculation revised to address extended reduced power operation.113 3.1.8 Special Nuclear Material Database No change to analysis is required. The change to Cycle 17 operating power will have noUpdate effect on prior cycle spent fuel and its characteristics.114 3.1.9 Plant Physics Data Book Re-analysis was performed to determine impact, and all results were acceptable. DataBook has been revised to address extended reduced power operation.ENCLOSURE 3Page 21 of 23 Table 1 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofReload and UFSAR Chapter 15 Safety AnalysesITEM CHECKLIST ITEM 1# C RECTION) DESCRIPTION SUMMARY OF IMPACT ASSESSMENT# (UFSAR SECTION)115 3.1.10 Startup Physics Test Predictions Re-analysis was performed to determine impact, and all results were acceptable.Calculation has been revised to address changes to startup testing power plateaus.116 3.2.1 COLSS As-built Database and Test Re-analysis was performed to determine impact, and all results were acceptable.Cases Calculation has been revised to address extended reduced power operation impact onthe cycle specific COLSS reload constants for DNBR & Linear Heat Rate (LHR)penalties.117 3.2.2 CEFAST Database Analysis Re-analysis was performed to determine impact, and all results were acceptable.Calculation has been revised to address extended reduced power operation impact onthe cycle specific CPC reload constants for DNBR & Local Power Density (LPD)penalties.ENCLOSURE 3Page 22 of 23 Table 2 -Revision 1SONGS Unit 2 Cycle 17 Reduced Power Operation -Summary of Impact Assessment ofCore Design and Monitoring Technical Specification Surveillance RequirementsPower Applicability and Summary of Impact Assessment forSunv # Surveillance Topic Surveillance Frequency Performing at 68-70% Power3.1.3.1 Reactivity Balance Every 31 EFPD Steady state power (not full power) is required3.1.4.1 MTC within positive limit Prior to Mode 1 Performed at Hot Zero Power and projected to BOC 70%conditions3.1.4.2 MTC within negative limit Within 14 EFPD of peak Boron @ Peak boron occurs at BOC, -performed at Hot Zero PowerRTP and projected to HFP EOC conditions3.1.4.2 MTC within negative limit Within +/- 30 EFPD of Steady state power (not full power) is required; projected to2/3 of expected core burnup HFP EOC conditions3.2.2.1 CPC & COLSS Fxy > Between 40% -85% (i.e., prior to 68%-70% is within the power range required formeasured Fxy (CECOR) exceeding 85%) surveillance3.2.2.1 CPC & COLSS Fxy > Every 31 EFPD Steady state power (not full power) is requiredMeasured Fxy (CECOR)3.2.3.3 CPC Azimuthal Tilt > Every 31 EFPD Steady state power (not full power) is requiredMeasured Tilt (CECOR)3.3.1.2 RCS Flow in CPCs < Every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (not required until Procedure changed to perform surveillance at >-6870%Measured RCS Flow 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after power > 85% RTP) power3.3.1.5 RCS Flow by calorimetric Every 31 days (not required until Procedure changed to perform surveillance with additional12 hours after power > 85% RTP) margin at >--6870% power., and to require additional mr- giwAfhen suweillance is performed during extended perFation3.3.1.11 CPC Shape Annealing Prior to exceeding 85% A minimum ASI change, rather than a specific power level,Matrix (SAM) Verification is requiredN/A Startup Test Activity Normally performed after reaching Results are already adjusted from actual test conditions toReduction Program full power RTP conditions as a part of the test methodReactivity BalanceHZP -HFPENCLOSURE 3Page 23 of 23 ENCLOSURE 4List of Commitments
Enclosure
4List of CommitmentsThis table identifies an action discussed in this letter that Southern California Edison commits toperform. Any other actions discussed in this submittal are described for the NRC's informationand are not commitments.IOne- TimeTICommitment Only Sustainable Due DateOnly T DeDtAdministratively control the Prior to Unit 2RCS dose equivalent 1-131 X Cycle 17specific activity described Mode 2in Technical Specification operationLCO 3.4.16 Action Al to nomore than 60 pCi/gm.