ML12263A300

San Onofre Nuclear Generating Station, Units 2 and 3 - Response to Request for Additional Information Regarding License Amendment Request for Permanent Use of Areva Fuel and for Permanent Exemption to Use M5 Cladding

ML12263A300
Person / Time
Site:	San Onofre
Issue date:	09/14/2012
From:	St.Onge R J Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME6820, TAC ME6821, TAC ME6822, TAC ME6823
Download: ML12263A300 (42)

Text

SOUTHERN CALIFORNIA Richard J. St. OngeEDISON Director, Nuclear Regulatory Affairs and,J E ISONEmergency PlanningAn EDISON INTERNATIONAL CompanySeptember 14, 2012U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001Subject: Docket Nos. 50-361 and 50-362Response to Request for Additional Information Regarding LicenseAmendment Request for Permanent Use of AREVA Fueland for Permanent Exemption to Use M5 Cladding(TAC Nos. ME6820, ME6821, ME6822, AND ME6823)San Onofre Nuclear Generating Station, Units 2 and 3Reference: Letter from N. Kalyanam (NRC) to P. T. Dietrich (SCE) dated August 1,2012; Subject: San Onofre Nuclear Generating Station, Units 2 and 3 -License Amendment Request RE: Use of AREVA Fuel (TAC Nos.ME6820, ME6821, ME6822, AND ME6823)Dear Sir or Madam:By letter dated August 1, 2012, the Nuclear Regulatory Commission issued a Requestfor Additional Information (RAI) regarding use of unrestricted usage of AREVA fuel andpermanent exemption to use M5 cladding.The RAI letter requested a response within 30 days of receipt of the letter. NRC staffagreed by phone on September 4, 2012, that SCE may submit the response bySeptember 14, 2012.Enclosure 2 of this submittal contains information that is proprietary to SCE or AREVA.SCE requests that this proprietary Enclosure be withheld from public disclosure inaccordance with 10 CFR 2.390(a)(4). Enclosure 1 provides notarized affidavits fromSCE and AREVA which set forth the basis on which the information in Enclosure 2 maybe withheld from public disclosure by the Commission and addresses with specificity theconsiderations listed by paragraph (b)(4) of 10 CFR 2.390. Enclosure 3 provides thenon-proprietary version of Enclosure 2.P.O. Box 128San Clemente, CA 92672 Document Control Desk-2-September 14, 2012There are no new regulatory commitments contained in this letter. If you have anyquestions or require additional information, please contact Ms. Linda T. Conklin,Licensing Manager, at (949) 368-9443.Sincerely,Enclosures:1. NOTARIZED AFFIDAVITSProprietary Enclosures2. Response to Request for Additional Information (RAI) regarding use ofunrestricted usage of AREVA fuel and permanent exemption to use M5claddingNon-Proprietary Enclosures3. Response to Request for Additional Information (RAI) regarding use ofunrestricted usage of AREVA fuel and permanent exemption to use M5claddingcc: E. E. Collins, Regional Administrator, NRC Region IVR. Hall, NRC Project Manager, San Onofre Units 2 and 3G. G. Warnick, NRC Senior Resident Inspector, San Onofre Units 2 and 3 ENCLOSURE 1NOTARIZED AFFIDAVITS AFFIDAVITSTATE OF CALIFORNIA )) SS.CITY OF SAN CLEMENTE)1. My name is Owen J. Thomsen. I am employed by Southern California EdisonCompany ("SCE"). My present capacity is Manager, Nuclear Fuel Management, for the SanOnofre Nuclear Generating Station ("SONGS"), and in that capacity I am authorized to executethis Affidavit.2. SCE is the operating agent for SONGS. I am familiar with the policiesestablished by SCE to determine whether certain SCE information is proprietary andconfidential, and to ensure the proper application of these policies.3. I am familiar with SCE information in the document entitled "San Onofre NuclearGenerating Station, Units 2 and 3, Proposed Permanent Exemption Request and ProposedChange Number (PCN) 600, Amendment Application Numbers 261 and 247, Request forUnrestricted Use of AREVA Fuel," (referred to herein as "Document") submitted to the NRC inJuly 2011.4. SCE has classified the information contained in the Document as proprietaryand confidential in accordance with SCE's policies.5. Specifically, SCE applied the following criteria to determine that theinformation contained in the Document should be classified as proprietary and confidential:(a) SCE has a Non-Disclosure Agreement (NDA) with Westinghouse Electric LLC("Westinghouse") and AREVA NP ("AREVA") (the NDA is referred to as the"Westinghouse-AREVA-SCE NDA"), under which Westinghouse and AREVAhave provided to SCE certain proprietary and confidential information containedin the Document.

(b) The information reveals details of Westinghouse's, SCE's, and/or AREVA'sresearch and development plans and programs, or the results of these plans andprograms.(c) The information includes test data or analytical techniques concerning a process,methodology, or component, the application of which results in a competitivecommercial advantage for Westinghouse, SCE, and/or AREVA.(d) The information reveals certain distinguishing aspects of a process,methodology, or component, the exclusive use of which provides a competitivecommercial advantage for Westinghouse, SCE, and/or AREVA on productoptimization or marketability.(e) The unauthorized use of the information by one of Westinghouse's, SCE's,and/or AREVA's competitors would permit the offending party to significantlyreduce its expenditures, in time or resources, to design, produce, or market asimilar product or service.(f) The information contained in the Document is vital to a competitive commercialadvantage held by Westinghouse, SCE, and/or AREVA, would be helpful to theircompetitors, and would likely cause substantial harm to the competitive positionof Westinghouse, SCE, and AREVA.6. The information contained in the Document is considered proprietary andconfidential for the reasons set forth in Paragraph 5. In addition, the information contained in theDocument is of the type customarily held in confidence by AREVA, Westinghouse, and SCE,and not made available to the public. Based on my experience in the nuclear industry, I amaware that other companies also regard the type of information contained in the Document asproprietary and confidential.

7. In accordance with the Westinghouse-AREVA-SCE NDA, the Document hasbeen made available to the NRC in confidence, with the request that the information containedin this Document be withheld from public disclosure. The request for withholding the informationfrom public disclosure is made in accordance with 10 CFR 2.390. The information qualifies forwithholding from public disclosure under 10 CFR 2.390(a)(4) "Trade secrets and commercial orfinancial information."8. In accordance with SCE's policies governing the protection and control ofproprietary and confidential information, the information contained in the Document has beenmade available, on a limited basis, to others outside Westinghouse, SCE and AREVA only asrequired in accordance with the Westinghouse-AREVA-SCE NDA.9. SCE's policies require that proprietary and confidential information be kept ina secured file or area and distributed on a need-to-know basis. The information contained in theDocument has been kept in accordance with these policies.10. The foregoing statements are true and correct to the best of my knowledge,information, and belief, and if called as a witness I would competently testify thereto. I declareunder penalty of perjury under the laws of the State of California that the above is true andcorrect./w/n.homseOw'eri fhsen.i.....,SUBSCRIBED before me this " d (tfm no antlftu of " -I IAday of _2011. __ _ 7.t~o be #W W" ho"WNW#WNOTARY PUBLIC, STATE OF CALIFORNIAMY COMMISSION EXPIRES: .....RULPLi'TRVEReg. #: .Commission # 1936995Notary Public -California zOrange County -My Comm. Expires Jun 14, 2015 AFFIDAVITCOMMONWEALTH OF VIRGINIA) ss.COUNTY OF CAMPBELL )1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVANP Inc. (AREVA NP) and as such I am authorized to execute this Affidavit.2. I am familiar with the criteria applied by AREVA NP to determine whethercertain AREVA NP information is proprietary. I am familiar with the policies established byAREVA NP to ensure the proper application of these criteria.3. I am familiar with the AREVA NP information contained in a letter from RickWilliamson (AREVA NP) to Owen Thomsen (Southern California Edison) with subject "ProposedResponse to RAI on License Amendment Request for Use of AREVA VQP Fuel," FAB12-439,dated September 5, 2012 and referred to herein as "Document." Information contained in thisDocument has been classified by AREVA NP as proprietary in accordance with the policiesestablished by AREVA NP for the control and protection of proprietary and confidentialinformation.4. This Document contains information of a proprietary and confidential natureand is of the type customarily held in confidence by AREVA NP and not made available to thepublic. Based on my experience, I am aware that other companies regard information of thekind contained in this Document as proprietary and confidential.5. This Document has been made available to the U.S. Nuclear RegulatoryCommission in confidence with the request that the information contained in this Document bewithheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure isrequested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financialinformation."6. The following criteria are customarily applied by AREVA NP to determinewhether information should be classified as proprietary:(a) The information reveals details of AREVA NP's research and developmentplans and programs or their results.(b) Use of the information by a competitor would permit the competitor tosignificantly reduce its expenditures, in time or resources, to design, produce,or market a similar product or service.(c) The information includes test data or analytical techniques concerning aprocess, methodology, or component, the application of which results in acompetitive advantage for AREVA NP.(d) The information reveals certain distinguishing aspects of a process,methodology, or component, the exclusive use of which provides acompetitive advantage for AREVA NP in product optimization or marketability.(e) The information is vital to a competitive advantage held by AREVA NP, wouldbe helpful to competitors to AREVA NP, and would likely cause substantialharm to the competitive position of AREVA NP.The information in the Document is considered proprietary for the reasons set forth inparagraphs 6(b) and 6(c) above.7. In accordance with AREVA NP's policies governing the protection and controlof information, proprietary information contained in this Document has been made available, ona limited basis, to others outside AREVA NP only as required and under suitable agreementproviding for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a securedfile or area and distributed on a need-to-know basis.9. The foregoing statements are true and correct to the best of my knowledge,information, and belief.SUBSCRIBED before me thisday of 012.Kathleen A. BennettNOTARY PUBLIC, COMMONWEALTH OF VIRGINIAMY COMMISSION EXPIRES: 8/31/2015Reg. #110864K ATHLEEN ANN BENNETTNotay Public ICommonwealth of Virginia110064MY Commission Expires Aug 31, 2015 ENCLOSURE 3Response to Request for AdditionalInformation (RAI) regarding use ofunrestricted usage of AREVA fueland permanent exemption to useM5 cladding (Non-Proprietary)

SOUTHERN CALIFORNIA EDISONRESPONSE TO REQUEST FOR ADDITIONAL INFORMATIONLICENSE AMENDMENT REQUEST FOR PERMANENT USE OF AREVA FUELAND FOR PERMANENT EXEMPTION TO USE M5 CLADDINGDOCKET NOS. 50-361 AND 50-362TAC NOS. ME6820, ME6821, ME6822, AND ME6823SUBJECT PAGERA I#1 ............................................................................................................................................. 2RA I #2 ............................................................................................................................................. 3RA I #3 ............................................................................................................................................. 4RA I #4 ............................................................................................................................................. 6RA I #5 ............................................................................................................................................. 8RA I#6 ........................................................................................................................................... 10RA I #7 ........................................................................................................................................... 15RA I #8 ........................................................................................................................................... 16RA I #9 ........................................................................................................................................... 18RA I #10 ......................................................................................................................................... 20RA I #11 ......................................................................................................................................... 24RA I #12 ......................................................................................................................................... 26RA I #13 ......................................................................................................................................... 28RSB RA I #1 .................................................................................................................................. 32 RAI #11. Page 3(a) SCE states that, "The number of fuel assemblies in the initial batch will bebetween eight fuel assemblies and approximately a half core of fuel assemblies."Please specify the number of AREVA fuel assemblies that are to be inserted intothe SONGS core during the next cycle. Also specify which components of theAREVA fuel assemblies will be of M5 alloy material.(b) It is stated that the "exact reload core fuel management has not been defined" atthe time of the LAR submittal. Please provide the details of the core reloadmanagement for the next cycle.RESPONSE:(a) SCE currently plans to introduce 36 AREVA CE-HTP fuel assemblies into the Unit 3 Cycle17 core. The AREVA assembly design is described in Enclosure 2 Section 5.1.2 of PCN-600. The fuel assembly components made of M5TM alloy material are [(b) Due to the unscheduled shutdown of Unit 3 Cycle 16 by a steam generator tube leak, forcingan extended mid-cycle outage, the final core pattern for Unit 3 Cycle 17 has still not beenfinalized. The 36 AREVA CE-HTP assemblies will be implemented into the pattern withsome assemblies placed in limiting or near limiting locations. The core loading patternsreferenced in PCN-600 (Enclosure 2 Section 7.1, Figures 7.1.2 and 7.1.3) present potentialUnit 3 Cycle 17 and Cycle 18 core designs to demonstrate and exercise all aspects of themethodology associated with transitioning to AREVA fuel. Using the SONGS fuelmanagement guidelines, Cycle 17 was designed as a transition core with half core freshAREVA fuel and the other half Westinghouse once burned fuel to fully exercise the mixedcore process. Cycle 18 was subsequently designed with a full core of AREVA fuel (note thecenter assembly exception in Section 4.1.3) to reflect the end state AREVA assembly core.2 RAI #22. Section 3.2.1Provide justification for continuing the use of CE methodologies, CENPD-382-P-Aand CENPD-275-P-A in support of the fuel Safety Limits (SLs) in SONGS TechnicalSpecification (TS) 2.1.1.2 in the specification of fuel centerline temperature variationwith burnup and its adjustments with burnable poison.RESPONSE:PCN-600 (Enclosure 2, Section 4.10) provides justification for continuing the use of CEmethodologies, CENPD-382-P-A and CENPD-275-P-A in support of the SONGS Units 2 and 3fuel Safety Limit (SL) 2.1.1.2 in the specification of fuel centerline temperature variation withburnup and its adjustments with burnable poison.Per PCN-600, the selection of the Westinghouse methodology to represent the fuel meltingtemperature is based on [1.Complementing the justification presented in PCN-600 are the findings presented in the NRCSafety Evaluation in support of gadolinia burnable poison Topical Report CENPD-275-P-A,which specifies a fuel centerline temperature adjustment for the gadolinia burnable poison.CENPD-275-P-A (Revision l-P, Supplement 1-P-A, Section 2.2.1) states that the slightly moreconservative correlation for erbia additions may be used for gadolinia additions. Theacceptability of the use of the erbia correlation for gadolinia is addressed in the NRC Staff SafetyEvaluation for CENPD-275-P, Revision l-P, Supplement 1-P. The Safety Evaluationacknowledges that the gadolinium and erbium are closely-related rare earth elements, whichform oxides of the same structure. Because of the similarities of these oxides, the measuredvalues of specific additions of gadolinia or erbia on urania properties would be similar. TheSafety Evaluation states use of the slightly more conservative erbia correlation (as done in theSCE process) as the gadolinia correlation is acceptable.The measured values of specific additions of gadolinia and erbia on urania properties are similar.The CE methodologies for burnable poison adjustments have been found to be conservative forgadolinia and erbia. The fuel melt correlation was developed using data which is independent offuel manufacturer. Therefore, it is justified to use CE methodologies, CENPD-382-P-A andCENPD-275-P-A, in support of the fuel Safety Limits in SONGS Technical Specification 2.1.1.2in the specification of fuel centerline temperature variation with burnup and its adjustments withburnable poison.3 RAI #33. Section 4.1.3(a) Explain why there may be a need to retain the center assembly from the oldvendor.(b) Explain why the core will [].RESPONSE:(a) The SONGS reactor core has an odd number of fuel assemblies (i.e., 217) and thusthe center assembly is typically retained for a third cycle of operation or re-insertedfrom the spent fuel pool. For fuel economics, SCE would like to retain the ability toreinsert the center assembly from the old fuel vendor without having to invoke all ofthe mixed core processes or requirements: e.g. mixed core compatibility and LOCAanalyses.(b) In the SONGS checkerboard fuel management patterns, the high burnup centerassembly is in a low power, high RCS flow location and thus not limiting from apower peaking or thermal hydraulic perspective. Table 1 below shows powerpeaking from the center assembly for the cycle 17 and 18 patterns described in PCN-600. As seen from the table, the center assembly is not, nor is it likely to be, at ornear limiting for SONGS cores. Enclosure 2, Figure 7.2.4 shows the relatively highflow of the center assembly location.4 Table 1: VQP* Cycles 17 and 18 Center Assembly Power PeakingParameter Cycle 17 Cycle 18BOC I EOC BOC I EOCCenter AssemblyRelative PowerCenter Assembly FrCenter Assembly FqCore FrCore Fq ]* VQP is an acronym for the Vendor Qualification Plan, i.e. the PCN-600submittal for the unrestricted use of AREVA fuel.5 RAI #44. Section 4.1.4(a) Provide details of how the two commitments that AREVA made to the NRC in theTopical Report, XN-NF-85-92(P)(A), have been implemented. The commitmentsare: (i)(b) Provide details of how the []. (i.e., the details of the nuclear design analysis fora typical cycle).RESPONSE:(a) AREVA has committed to ensuring that the Gadolinia bearing fuel rod will not be thelimiting rod in the core. On a cycle-specific basis, []. The commitmentmade by AREVA to the NRC is embedded in the reload process as an automatic checkperformed by the fuel rod analysis code. For each batch of fuel analyzed as part of the reloadanalysis, the code checks to make sure that the maximum gas pressure predicted for theGadolinia rods is less than that predicted for the UO2 rods. If this criterion is violated, thenthe code flags this occurrence as a failure to meet the gas pressure criterion. Typically, themaximum predicted gas pressure for gadolinia rods is less than that predicted for the U02rods. In case this criterion is failed, modifications will have to be made to the core designsuch that the criterion can be met for the upcoming cycle.In the SCE fuel management guidelines SCE has adopted the standard enrichment cutbackused by AREVA for gadolina assemblies. []. By reducing the enrichmentin accordance with this formula, the power in the gadolina rod will always be less than thepeak power in the assembly and consequently always less that the peak power in the core.As an example for PCN-600 cycle 17 demonstration analysis, the peak enrichmentin PCN-600 Figure 7.1.2 is 3.5 w/o U235.In this assembly the8% gadolinia rod has been to 2.1 w/o U235.In this way the gadolinia rod willalways remain non-limiting for power peaking.(b) Based on the process described in response (a) above, the gadolinia bearing fuel rod willalways be [ ] if the fuel enrichment cutback is employed. However, in thesituation that SCE does not employ the requisite gadolina enrichment cutback, the followingstep by step process is employed to ensure that the peak power of any gadolinia rod will notbe the peak rod power in the core:6

i. The SIMULATE-3 physics model for a reload cycle is depleted in steps of 1 GWD/T fromBOC to EOC. SIMULATE-3 calculates a pin-by-pin power distribution at each 1GWD/T bumup point. The SIMULATE-3 output contains a summary of maximumpeaking factors and the assembly in which the maximun occurs.ii. At each SIMULATE-3 bumup point, a utility computer program (e.g. MCEDIT) produces apin-by-pin power distribution for every fuel assembly in the reactor core. In eachassembly pin-by-pin power picture, the maximum fuel rod power and the maximum fuelrod location in that assembly are identified. Therefore, the location of the peak fuel rodin the core is identified.iii. Final verification that the peak fuel rod does not contain Gadolinia is done by manualcomparison to the enrichment zone pattern for the assembly in which the peak rodoccurs.7 RAI #55. Section 4.21I.RESPONSE:The SCE TORC computer code [ ] was implemented, validated, and testedin accordance with SCE computer code control procedures. Test cases have been run and theoutput checked to verify that the correlation has been implemented correctly and to ensure thatthe installation of the CHF correlation has not affected the existing code capabilities, analyses oroutput. The results have been validated by comparison to [IThe SCE formal software update process defined in Section 4.5.2 of Reference 8.4 of PCN-600was applied to [The TORC modeling scheme at SONGS as discussed in SCE-9801-P-A [8 During a review of calculations perfonned in support of PCN-600, [I9 RAI #66. Section 4.2.1Provide typical calculations where the system parameter uncertainties and stateparameter uncertainties are statistically combined to obtain the minimum DNBR limitI ]. Provide a list ofall parameters with uncertainties that are used in the calculation which leads to theminimum DNBR limit.RESPONSE:The SONGS design Specified Acceptable Fuel Design DNBR limit (SAFDL DNBR) forSONGS is 1.31 (Technical Specification LCO 2.1.1.1). This SAFDL DNBR Limit wascalculated based on the NRC-approved Modified Statistical Combination of Uncertainties(MSCU) methods.IThe MSCU process considers two groups of uncertainties: [IThe first group, [1, includes the following uncertainties: [I10 The second group, I.] The remaining uncertainties arepotentially impacted and are addressed below.Engineering & Systematic Factors Uncertainties]. The results are compared to the existing SONGS values in Table 4.2.Table 4.2System Parameter Uncertainties Used As Inputs in SONGS MSCU Analysis [11

.1The standard reload process at SCE does not involve recalculating the DNBR limit each cycle.Instead, the 1.31 DNBR limit is maintained and verified as conservative for each cycle. This isdone by evaluating the DNBR change for a given perturbation of input parameters. In the MSCUAOR, the DNBR limit is dominated by the DNBR changes primarily due to perturbation of twofactors:[]In the normal SCE reload process, the assemblies considered DNBR limiting for the specificcycle are put through the same perturbation as was done in the AOR for Factors #1 and #2above. If the DNBR change for the cycle specific cases are less than the DNBR change for theequivalent AOR cases, then the cycle specific DNBR limit would be bounded by the AOR limitof 1.31 DNBR.To demonstrate the process, an MSCU analysis was performed for the PCN-600 cycle 17 mixedcore. The resultant probability distribution function (pdf) for both a CE 16 (CE-I ) limitingassembly and an AREVA (BHTP) limiting assembly versus the AOR (1.31 DNBR limit isshown in Figure 6-1.To provide easier comparison, Figure 6-2 shows the pdfs all shifted to the same mean. []12 Figure 6-113 rFigure 6-214 RAI #77. Section 4.2.2Section 4.2.2 states that] This section makes reference to Westinghousemethodology, AREVA methodology, and SCE methodology. Table 4.2 lists theSONGS rod bow penalty for AREVA fuel. Does this mean that the AREVA fuel typeis the most limiting based on the thermal-hydraulics analysis? The staff would likethe licensee to clearly provide the details about which fuel type is most limiting, theappropriate methodology used to calculate the rod bow penalty, and how the penaltyis applied [ ].RESPONSE:The SCE thermal-hydraulic analysis process considers ALL assemblies in the core to determinewhich assemblies are potentially DNBR limiting. A wide range of operating conditions andpower profiles are considered. Therefore, depending on the specific core design, it may bepossible for a Westinghouse fuel assembly to be DNBR limiting at one condition and anAREVA fuel assembly to be DNBR limiting at some other condition during the cycle.The SCE methodology provides for the rod bow penalty [15 RAI #88. Section 4.3.2The licensee states that [I(a) Due to the fact that [] please explain how the results willbe consistent.(b) The first paragraph of Section 4.3.2 states, "Per Reference 8.34 (CEN 193(B)Supplement 2-P), currently SCE uses FATES3B to provide predictions of thesteady state response of fuel rods, and to model internal conditions of the fuelrods within the core from insertion to discharge. With the appropriate modelingof mechanical design data, power levels, and power distributions, these [] This statement appears to conflict withthe last paragraph on page 28, which states, [] Please clarify the apparently ambiguous or conflictingstatements.RESPONSE:(a) Although the fuel performance data used in the fuel mechanical design, LOCA, non-LOCA and setpoints analyses are not originated by the same code or method, theseanalyses are originated using codes and methods that have been NRC approved fortheir intended purposes.[16 There is no inconsistency introduced into the reload analysis effort because [] In all cases, regardless of how the fuelperfonnance data is generated, the reload analysis effort will be performed usingcodes and methods that have been NRC approved for their intended purposes.(b) The first paragraph of Section 4.3.2 was intended to describe the current SCE reloadprocess (i.e. before PCN-600 submittal). Subsequent paragraphs of Section 4.3.2describe the division of fuel rod behavior scope between SCE and AREVA as itrelates to the proposed reload licensing applications. As noted in the fourth paragraphof Section 4.3.2, AREVA will be performing all fuel mechanical design and LOCAanalyses, including the fuel rod initial conditions for the AREVA LOCA analyses,and SCE will be generating fuel rod behavior analysis data to support the non-LOCA safety analyses and calculations that support SCE setpoints analyses.17 RAI #99. Section 4.3.3The licensee has used the CE/Westinghouse legacy code, FATES3B for their fuelrod behavior analyses for generating input to non-LOCA transient and setpointanalyses. Specifically, the FATES3B code has been used to modelProvide the details of the results from the [RESPONSE:The FATES3B code [I] is used to generate input to non-LOCA transient and setpoints analyses.IAREVA developed M5TM cladding materials and correlations that were approved by the NRC inAREVA M5TM topical report BAW-10240(P)-A. Similarly, Westinghouse developed ZIRLOTMcladding materials and correlations that were approved by the NRC in Westinghouse ZIRLOTMTopical Report CENPD-404-P-A. [The FATES3B code is used for thermal performance evaluations under normal operationconsidering steady-state and anticipated transient conditions. Per ZIRLOTM Topical ReportCENPD-404-P-A (response to Question 7),[18 As discussed in PCN-600 (Enclosure 2, Section 4.3.3.2), verification and validation (V&V)testing of [ ] was performed through a combination of code modificationreview and test case evaluations. [] All reference material propertydata, models, assumptions and required modifications were also reviewed and checked forcorrect implementation.As discussed in PCN-600 (Enclosure 2, Section 4.3.3.2), detailed reviews were conducted toensure the accuracy of [Additionally, [The fuel temperature, power-to-centerline melt, and rod internal pressure history] are shown in PCN-600 Figures 4.3.6, 4.3.7 and 4.3.8, respectively. [IThe use of the [I19 RAI #1010. Fuel Thermal Conductivity (Section 5.1.4)An outstanding issue related to the mechanical and material design of U02 fuel isthe thermal conductivity of irradiated U02 fuel considering the effects of burnup.The thermal conductivity of irradiated U02 fuel is affected by changes that takeplace in the fuel during irradiation: solid fission product buildup (both in solution andas precipitates), porosity and fission gas-bubble formation.NRC Information Notice 2009-23 dated October 8, 2009, notified licensees ofnuclear power reactors of the thermal conductivity degradation (TCD) of uraniumfuel pellets with increasing burnup. The significance of this effect was not includedin the fuel thermal-mechanical performance codes approved prior to 1999.NRC Information Notice 2011-21 notified the licensees of the impact of irradiation onfuel thermal conductivity and its potential to cause errors (specifically, in predictedpeak clad temperature) in realistic ECCS evaluation models.10 CFR Part 50, Appendix K, "ECCS Evaluation Models",Section I.A.1, stipulatesthat, "The steady-state temperature distribution and stored energy in the fuel beforethe hypothetical accident shall be calculated for the burn-up that yields the highestcalculated cladding temperature (or, optionally, the highest calculated storedenergy.) To accomplish this, the thermal conductivity of the U02 shall be evaluatedas a function of burn-up and temperature, taking into consideration differences ininitial density, and the thermal conductance of the gap between the U02 and thecladding shall be evaluated as a function of the burn-up, taking into considerationfuel densification and expansion, the composition and pressure of the gases withinthe fuel rod, the initial cold gap dimension with its tolerances, and cladding creep."The CE/Westinghouse FATES3B code has been used for the evaluation of fuelthermal-mechanical performance at SONGS Units 2 and 3. [.1Thermal conductivity of U02 fuel degrades with burnup, and as such, the staffbelieves that each fuel vendor must have an explicit model to generate burnupdependent fuel thermal conductivity in their analyses to simulate transients andaccidents.(a) Explain how the licensee applied the TCD with burnup in the FATES3B code forthe fuel performance evaluation, addressing factors such as, fission gas release,power-to-melt evaluation, and clad strain and fatigue. Please provide details ofthe fuel temperature calculations that are dependent on the effects of burnup asdescribed above.20 (b) Explain how the impact of TCD with burnup has been addressed in the analysesof non-LOCA transients and postulated accidents, specifically but not limited to,the spectrum of control rod ejection accident analyses.RESPONSE:a)the selectedapplications of the FATES3B code are not significantly impacted by the effects of TCD.As discussed in PCN-600 Enclosure 2, Section 4.3.2, AREVA will be performing all fuelmechanical design and LOCA analyses, including the fuel rod initial conditions for theAREVA LOCA analyses, based on AREVA's approved computer codes RODEX2 &RODEX3A. SCE will be generating fuel rod behavior analysis data to support thenon-LOCA safety analyses (i.e., CEA Ejection analysis) and calculations that support SCEsetpoints analyses.Effect of TCD on Fission Gas Release for Rod Internal Pressure Calculations (i.e., no-clad liftoff)The FATES3B code is not used in the AREVA no-clad liftoff calculations, which will bedone by AREVA based on their approved RODEX2 computer code.Effect of TCD on Power to Fuel Centerline MeltThe FATES3B code [] conservatism in the fuelbehavior analysis results is accomplished by the means described in the NRC Question3.A response as presented in CEN-l 93(B)-P Supplement 2-P. The response states thatthe code results [21

] it is acceptable to not consider theeffects of TCD on power to fuel centerline melt.Effect of TCD on Clad Strain and FatigueThe FATES3B code is not used in the clad strain and fatigue analyses of AREVA fuel,which will be performed by AREVA using their RODEX2 computer code as part of theirfuel rod mechanical design.b) Effect of TCD on Non-LOCA Transient AnalysesSCE will use FATES3B [] for input to the transient analyses, and [] the CEA Ejection transient analysis.Depending on the non-LOCA transient being evaluated, [22

[ ]The CEA ejection transient analyst selects [ITherefore, it is acceptable to [] At BOL, the onset of TCD has not yet materialized. Therefore, TCD doesnot affect the CEA ejection transient analysis [I23 RAI #1111. The NRC staff intends to run FRAPCON-3.4 (Reference 2) benchmark calculationsof the resident CE 16x1 6 fuel rod design and the new AREVA HTP fuel rod design.Please provide the following input for both co-resident fuels at SONGS, Units 2 and3.A. Rod Power History, KW/ft as a function of GWd/MTU1. Bounding thermal-mechanical operating envelope (e.g., radial falloff curve)2. Discuss any application of rod power uncertainties3. Include power histories for different pellet designs (U02, Gadolinium).B. Axial Power Distribution (Fz at each axial node)1. Include axial power distributions (AXPDs) for different axial blanket configurations.C. Fuel Rod Design Specifications and Manufacturing Tolerances1. Outer diameter2. Inside diameter3. Pellet diameter4. Stack length5. Plenum length6. Pellet height7. Dish radius8. Dish depth9. Spring outside diameter10. Spring wire diameter11. Number of spring turns12. Maximum U-235 enrichment (%)13. Average U-235 enrichment (%)14. Maximum gadolinia content (%)15. Water in pellet (ppm)16. Nitrogen in pellet (ppm)17. Pellet density (%TD)18. Open porosity (%)19. Pellet surface roughness (microns)20. Expected density increase (gms/cc)21. Sintering temperature (OF)22. Cladding Alloy = (Material name)23. Final thermal treatment = (RXA or ?)24. Cladding surface roughness (microns)25. Cladding texture factor26. Cladding Hydrogen content (ppm)27. Fill gas pressure28. Fill gas composition29. Rate of CRUD accumulation factor (mils/hr)30. CRUD thermal conductivity24 D. Coolant conditions1. Coolant inlet temperature (TF)2. Coolant mass flux (Ibm/hr-ft2)3. System pressure (psia)RESPONSE:As agreed, a response to this item will be provided by September 30, 2012.25 RAI #1212. The NRC Standard Review Plan, Section 15.0.1, "Radiological ConsequenceAnalyses Using Alternative Source Terms," (ADAMS Accession NumberML003734190) states: "The analysis methods and assumptions used by thelicensee in determining the core inventory should be reviewed to ensure that theyare based on current licensing basis rated thermal power, enrichment, and burnup."Enclosure 2 to the submittal states that AREVA fuel is approved for CombustionEngineering pressurized water reactor's for a maximum peak burnup of 62,000Megawatt-Day(s) per Metric Ton Uranium (MWD/MTU) (Section 4.4.2). The currentfuel is designed to ensure the fuel does not exceed 60,000 MWD/MTU (Section4.4.1).A modification to the licensing basis fuel type can have the potential to change thecore isotopic distribution and inventory assumed in post-accident conditions. Theimpacts regarding the core inventory due to changes other than the cladding (i.e.burnup) are not discussed in the proposed amendment. Please provide ajustification to support that changes in the fuel design parameters do not significantlychange the core isotopic distribution and magnitude (source term) for the designbasis accidents analyzed.RESPONSE:The SONGS fission product inventory is calculated using the guidance in Section 3.1 ofAlternative Source Term (AST) Regulatory Guide 1.183. Table 4.1-1 of the SONGS ASTLicense Amendment Request (LAR) (ADAMS Accession No. ML043650403) summarizes theparameters modeled in the evaluation of the reactor core activity inventory. As detailed inSection 4.1.1 of the LAR, the core inventory of fission products is based on the maximumfull-power operation of the core with, as a minimum, currently licensed values for fuelenrichment, fuel burnup, and an assumed core power equal to the current licensed rated thermalpower times the emergency core cooling system evaluation uncertainty. These parameters wereexamined parametrically to maximize the fission product inventory. The ORIGEN-S code wasexecuted for the various combinations of core average burnups (0, 10, 20, 30 and 40 Gigawatt-Days per Metric Ton Uranium (GWD/MTU)) and enrichments. For each isotope, the maximumcurie value from the ORIGEN-S code runs was chosen to represent the inventory of that isotopein the composite fuel assembly. The SONGS AST License Amendment, based on this boundingsource term, was issued in December 2006 (ADAMS Accession No. ML063400359).The core average burnup range of 0 to 40 GWD/MTU conservatively bounds fuel managementscenarios up to 24-month operating cycles irrespective of the peak pin burnup limit. As such, anincrease in maximum peak burnup from 60 to 62 GWD/MTU will not increase the currentbounding source term.Section 4.9 of PCN-600 acknowledges the current reload analysis process for verifying that thecurrent bounding source term and current radiological dose analyses are applicable to the new26 fuel cycle. If the current source tenns or radiological dose analyses are invalidated, then thecurrent reload analysis process addresses the need for new cycle-specific source terms to begenerated for use in the accident radiological dose analyses. The methodology to calculate thebounding source terms will remain unchanged during and after the transition to AREVA fuel.27 RAI #1313. Enclosure 2, Section 7.4.2.3, Table 7.4.9, and Attachment C (Table C.1) of thesubmittal provide text and tables describing events analyzed in the Updated FinalSafety Analysis Report (UFSAR), the acceptance criteria for these events, and theimpact of the use of AREVA fuel on these analyzed events. The NRC staff has thefollowing questions concerning this information.A. The NRC staff compared this information to the current UFSAR discussion andnoted several differences. For example, some of the events described in thesubmittal have different acceptance criteria from those stated in UFSAR Table15.0-8 (i.e., 10 CFR Part 100 limits vs. 10 CFR 50.67 limits). Explain and justifywhy the acceptance criteria for certain events described in the submittal differfrom those in the UFSAR.B. For some events, Attachment C states that the event is bounded by anotherevent. The UFSAR is not consistent with some of these statements in AttachmentC. For example, Attachment C states that the UFSAR Section 15.1.2.1 event isbounded by the Section 15.1.2.3 doses. UFSAR Section 15.1.2.1.5 states thatthe doses for this event are bounded not by Section 15.1.2.1, but by Section15.1.2.4 events. Please explain why Attachment C is inconsistent with thedescriptions of the bounding events provided in the UFSAR and state which iscorrect.C. In the column labeled "Impact of AREVA Fuel" of Attachment C to Enclosure 2 (forUFSAR Sections 15.7.3.4, and 15.7.3.9) it states: "As all pins in both the droppedand impacted assemblies are assumed to fail, there is no difference with use ofAREVA fuel." A review of these UFSAR sections shows that the UFSAR analysisassumes 226 fuel pins fail which is less than all the fuel pins in 2 assemblies (472fuel pins). Please resolve this inconsistency.D. Many of the evaluations of the impact of the AREVA fuel only address the impactof the change on fuel failure (source term). Per Appendices E-H of RegulatoryGuide 1.183, "Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Plants," dated July 2000 (ADAMS Accession No.ML003716792), the dose consequences of several accidents are dependent onboth the source term and how the radioactivity is transported to the environment.A change in fuel has the potential for changing the release rate and the totalamount of steam needed to cool down the plant after an accident. For accidentanalyses this steam is assumed to transport radioactivity to the environment. Forthose analyses that consider more than the source term (i.e. the main steamlinebreak, steam generator tube rupture, locked rotor and rod ejection accidents)please address any impact of the AREVA fuel on the transport of radioactivity tothe environment.28 RESPONSE:A. Per UFSAR Section 15.0, UFSAR Section 15.10 was added to present updated fuel cycleand unit specific data and consequences for the events presented in Sections 15.1 through15.9. When comparing UFSAR Section 15.10 to UFSAR Sections 15.1 through 15.9 thefollowing should be noted:(a) UFSAR Sections 15.1 through 15.9 are consistent with the latest information that hasbeen reviewed and approved by the NRC. These sections are intended to be updatedwhen information has been submitted to and approved by the NRC.(b) Section 15.10 presents the current plant configuration. This section includes data that hasbeen added in accordance with 10 CFR 50.59 since the last approval of the eventinformation by the NRC. This section is intended to be updated under the requirements of10 CFR 50.59.UFSAR Section 15.10.0.5 surmnarizes the assumptions, parameters, and calculationalmethods used to determine the doses that result from postulated accidents. As discussed inthis section, UFSAR Appendix 15G provides a list of the accidents modeled usingAlternative Source Term methodology (based on 10 CFR 50.67 limits), and UFSARAppendix 15B. 1 provides a list of the accidents modeled using pre-AST methodology(based on 10 CFR Part 100 limits).B. UFSAR Section 15.1.2.1.5 contained an inconsistency which has been corrected per theSONGS corrective action program. The noted UFSAR text stated that the radiologicalconsequences of this UFSAR Section 15.1.2.1 event are less severe than the results of theincreased main steam flow event with a concurrent loss of offsite power discussed inparagraph 15.1.2.4.5. However, the increased main steam flow event with a concurrentloss of offsite power event is discussed in paragraph 15.1.2.3.5 (not 15.1.2.4.5).As discussed in the response to Part "A", UFSAR Section 15.10 was added to presentupdated fuel cycle and unit specific data and consequences for the events presented inSections 15.1 through 15.9. The discussion as to which events are bounded by which otherevents is addressed in the UFSAR Section 15.10 subsections. Consistent with AttachmentC, UFSAR Section 15.10.1.2.1 correctly states that the UFSAR Section 15.10.1.2.1 eventdoses are bounded by the UFSAR Section 15.10.1.2.3 doses.29 C. As discussed in the response to Part "A", UFSAR Section 15.10 was added to presentupdated fuel cycle and unit specific data and consequences for the events presented inSections 15.1 through 15.9. Per UFSAR Sections 15.10.7.3.4 and 15.10.7.3.9, the numberof fuel pins that fail during a fuel handling accident is 472 (i.e., all the fuel pins in twoassemblies).D. The transport of radioactive material to the environment is dependent on:-Cladding integrity (Fuel Failure)-Primary to secondary leakage-Contaim-ent leakage-Reactor Coolant System (RCS) leakage-Engineered Safety Feature (ESF) leakage-The steaming rate from secondary (Mass Release)The cladding integrity (Fuel Failure) portion of the analysis remains unaffected. As discussed inSection 4.5.2 (page 61 of 166, Enclosure 2 to SONGS PCN 600) of the submittal, the NRC staffsis quoted as stating "the statistical convolution technique is conservative and acceptableprovided that the probability distribution for DNB is acceptable".As discussed in Section 4.2.1.1 (page 26 of 166, Enclosure 2 to SONGS PCN 600) of thesubmittal, the Modified Statistical Combination of Uncertainties (MSCU) analysis [IFor CEA ejection, the STRIKIN code [The primary to secondary leakage, containment leakage, RCS leakage and ESF leakage portionsof the analysis remain unaffected since they are independent of fuel type and cladding material.Therefore, []30 The steaming rate (Mass Release) portion of the analysis remains unaffected. The mass releaseis dependent on the core sensible heat, the RCS sensible heat, the core decay heat, and the heatremoval systems. As discussed in Section 4.5.3 (page 62 of 166, Enclosure 2 to SONGS PCN600) of the submittal, the M5TM cladding thermal conductivity, hgap, and the cladding specificheat [31 RSB RAI #1(Section 5.2) Please confirn that the Steam Generator 8% tube plugging assumption will remainbounding with respect to the number of tubes expected to be plugged.RESPONSE:The replacement steam generators (RSG), currently installed in both SONGS units, weredesigned and analyzed (including LOCA and non-LOCA events) assuming up to 8% pluggedtubes per steam generator. Consistent with the RSG design basis, the LOCA and non-LOCAevents preformed for and presented in PCN 600 were analyzed with an input value of up to 8%plugged tubes per steam generator. These are benchmark analyses to demonstrate themethodology to transition to AREVA fuel.For reload analyses, the number of plugged tubes per steam generator is a procedurallycontrolled input into LOCA and non-LOCA analyses. The Reload Groundrules (RGR)documents the number of plugged tubes per steam generator (currently RGR Item IV.005) to beused for LOCA and non-LOCA events. The RGR is reviewed and updated for each SONGSunit and cycle reload analysis campaign per procedure "Reload Groundrules (RGR) ControlMethodology." This procedure requires that all plant parameters used in the safety analyses bereviewed and updated to reflect the current or planned plant conditions applicable for theSONGS unit and cycle of interest. The RGR process is discussed in PCN 600 Section 6.3 and isunchanged from SONGS established process described in SONGS Reload AnalysisMethodology Topical Report SCE-9801-P-A, Section 4.3.Due to steam generator inspections during a refueling outage, the actual number of plugged tubescould change and must be confirmed to be in compliance with the value in the RGR prior tostartup. The "Core Reload Analysis and Activities Checklist" procedure performed every cycle(Step 6.1.7 and documented in the procedure's Attachment 3, Table 3.5) requires that this valuebe confinred prior to startup. Should the actual number of plugged tubes exceed the value in theRGR, LOCA and non-LOCA events would be reanalyzed/evaluated using a new bounding inputvalue prior to startup. This process is identical to that used previously by SONGS for theOriginal Steam Generators.Since Unit 3 steam generator inspections have not been completed at the time of this response,we cannot confirm that the 8% plugged tubes per steam generator input value used in the PCN600 benchmark analyses will remain bounding. However, we can confirn that if the number ofplugged tubes exceeds 8% plugged tubes per steam generator, the Unit 3 LOCA and non-LOCAevents will be reanalyzed/evaluated in accordance with SONGS' procedures prior to unit startup.32