ML11215A090

Proposed Permanent Exemption Request and Proposed Change Number (PCN) 600, Amendment Application Numbers 261 and 249, Request for Unrestricted Use of Areva Fuel

ML11215A090
Person / Time
Site:	San Onofre
Issue date:	07/29/2011
From:	Bauder D Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML11215A090 (295)

Text

Proprietary Information Douglas R. Bauder I EDISONION SOUTHERN CALIFORNIA An EDISON INTERNATIONAL9' Company Withhold from Public Disclosure Site Vice President & Station Manager San Onofre Nuclear Generating Station 10 CFR 50.12 July 29, 2011 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

San Onofre Nuclear Generating Station, Units 2 and 3 Docket Nos. 50-361 and 50-362 Proposed Permanent Exemption Request and Proposed Change Number (PCN) 600, Amendment Application Numbers 261 and 249, Request for Unrestricted Use of AREVA Fuel

References:

1. Letter from NRC (Hall) to SCE (Ridenoure) dated December 15, 2009, "SONGS Units 2 and 3 - Issuance of Amendments Revising Technical Specification 5.7.1.5, Core Operating Limits Report (COLR)" (TAC Nos. ME0604 and ME0605)" (ADAMS Accession Number ML093220105)

2. Letter from NRC (Hall) to SCE (Ridenoure) dated December 17, 2009, "San Onofre Nuclear Generating Station, Units 2 and 3 -

Temporary Exemption from the Requirements of 10 CFR Part 50, Section 50.46 and Appendix K for Lead Fuel Assemblies (TAC Nos. ME0602 and ME0603) (ADAMS Accession Number ML090860415)

Dear Sir or Madam:

Pursuant to 10 CFR 50.12, "Specific Exemptions," Southern California Edison Company (SCE) is requesting an exemption from the requirements of 10 CFR 50.46 "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," and 10 CFR 50, Appendix K, "ECCS Evaluation Models." In addition to the exemption, pursuant to 10 CFR 50.90, SCE is requesting approval of Amendment Application Numbers 261 and 249 for San Onofre Units 2 and 3, respectively, which consist of Proposed Change Number (PCN) 600: Request for Unrestricted Use of AREVA Fuel. The SCE evaluation of this request under the standards set forth in 10 CFR 50.92(c) is provided in Enclosure 2, and has determined that a finding of "no significant hazards consideration" is justified.

P.O. Box 128 Proprietary Information San Clemente, CA 92672 (949) 368-9275 PAX 89275 Withhold from Public Disclosure Fax: (949) 368-9881 Decontrolled Upon Removal From Enclosures Doug.Bauder@sce.coni

Proprietary Information Withhold from Public Disclosure Document Control Desk July 29, 2011 References 1 and 2 provided certain SCE requested licensing actions that were evaluated and approved by the NRC to support the Lead Fuel Assembly (LFA) program. The LFA program is the starting basis for the Vendor Qualification Program (VQP) to support continued San Onofre Nuclear Generating Station (SONGS) reload fuel supply initiatives.

This request continues to support the use of AREVA fuel, and is being submitted to implement license, Technical Specification, and reload analysis methodology updates that are required to facilitate an unrestricted use of AREVA fuel at SONGS Units 2 and 3, starting with Unit 3, Cycle 17. The changes consist of:

1) A permanent SONGS license exemption from the requirements of 10 CFR 50.46 and 10 CFR 50 Appendix K, to the extent to which these regulations limit acceptable cladding materials to only Zircaloy and ZIRLOTM. The requested exemption will allow the use of fuel rods clad with AREVA M5TM cladding. A temporary exemption from these requirements has been granted to SONGS by the NRC in Reference 2 for license changes needed to support the Lead Fuel Assembly (LFA) program involving AREVA M5TM cladding (NRC ADAMS Package ML090860392). The NRC has also recently approved a permanent exemption for Calvert Cliffs regarding the use of the same cladding material and very similar fuel design from the same fuel vendor (NRC ADAMS Accession Numbers ML101760523 and ML110190632).

2) Changes to SONGS Technical Specification 5.7.1.5 (CORE OPERATING LIMITS REPORT (COLR)) methodology reference list that are required to support the core design and implementation of unrestricted use of the potential alternative vendor fuel assemblies.

3) The SONGS Technical Specification 4.2.1 (Fuel Assemblies) description of clad materials will also be updated to include AREVA M5TM cladding.

4) In addition, the SONGS Technical Specification 2.1.1.2 (Reactor Core Safety Limits) will be updated to specifically identify a fuel centerline melt safety limit with corresponding adjustments made to account for burnable absorber fuel rods.

5) Incorporation of fuel burnup limits consistent with AREVA M5TM for the AREVA fuel assemblies (generically approved by the NRC in prior fuel vendor submittals) into the SONGS licensing basis.

Each of these requested changes is described in detail in Enclosure 2. Enclosure 2 provides detailed justifications supporting these changes and addresses in detail the No Significant Hazards and Environmental considerations.

Enclosures 2, 3, and 4 of this submittal contain information that is proprietary to SCE or AREVA. SCE requests that these proprietary Enclosures be withheld from public Proprietary Information Withhold from Public Disclosure Decontrolled Upon Removal From Enclosures

Proprietary Information Withhold from Public Disclosure Document Control Desk July 29, 2011 disclosure in accordance with 10 CFR 2.390(a)(4). Enclosure 1 provides notarized affidavits from SCE and AREVA which set forth the basis on which the information in Enclosures 2, 3, and 4 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed by paragraph (b)(4) of 10 CFR 2.390. Enclosures 5, 6, and 7 provide non-proprietary versions of Enclosures 2, 3, and 4, respectively.

Approval of the proposed amendments is requested by October 1, 2012.

Implementation of this license amendment requires revision to the current Technical Specifications. Implementation of these amendments will also necessitate revision to various sections of the SONGS Units 2 and 3 Updated Final Safety Analysis Report (UFSAR). SCE requests that these amendments be issued effective as of the date of issuance, to be implemented within 180 days from the date of issuance.

Implementation of these license amendments will supersede the AREVA Lead Fuel Assembly (LFA) commitments specified in References 1 and 2.

This request contains no new commitments.

If there are any questions or if additional information is needed, please contact Ms. Linda T. Conklin at (949) 368-9443 and/or linda.conklinDsce.com.

I declare under penalty of perjury that the foregoing is true and correct Executed on I I d*i(

Date Sincerely, Proprietary Information Withhold from Public Disclosure Decontrolled Upon Removal From Enclosures

Proprietary Information Withhold from Public Disclosure Document Control Desk July 29, 2011

Enclosures:

1. NOTARIZED AFFIDAVITS Proprietary Enclosures

2. LICENSEE'S EVALUATION Attachments to Enclosure 2 A. Listing of Acronyms B. Technical Specification and Bases Changes C. SONGS - Summary of Impact on UFSAR Chapter 15 Events D. Compilation of Calvert Cliffs RAI's - Application to SONGS and Responses E. Evaluation of AREVA SER and TER Methodology Limitations

3. ANP-2975(P), Revision 0, "San Onofre Nuclear Generating Station Unit 2 and Unit 3 Realistic Large Break Report" June 2011

4. ANP-2974(P), Revision 0, "San Onofre Nuclear Generating Station Unit 2 and Unit 3 Small Break Report" July 2011 Non-Proprietary Enclosures

5. LICENSEE'S EVALUATION Attachments to Enclosure 5 A. Listing of Acronyms B. Technical Specification and Bases Changes C. SONGS - Summary of Impact on UFSAR Chapter 15 Events D. Compilation of Calvert Cliffs RAI's - Application to SONGS and Responses E. Evaluation of AREVA SER and TER Methodology Limitations

6. ANP-2975(NP), Revision 0, "San Onofre Nuclear Generating Station Unit 2 and Unit 3 Realistic Large Break Report" June 2011

7. ANP-2974(NP), Revision 0, "San Onofre Nuclear Generating Station Unit 2 and Unit 3 Small Break Report" July 2011 cc: E. E. Collins, Regional Administrator, NRC Region IV R. Hall, NRC Project Manager, San Onofre Units 2 and 3 G. G. Warnick, NRC Senior Resident Inspector, San Onofre Units 2 and 3 S. Y. Hsu, California Department of Public Health, Radiologic Health Branch Proprietary Information Withhold from Public Disclosure Decontrolled Upon Removal From Enclosures

ENCLOSURE 1 NOTARIZED AFFIDAVITS

AFFIDAVIT STATE OF CALIFORNIA )

) SS.

CITY OF SAN CLEMENTE)

1. My name is Owen J. Thomsen. I am employed by Southern California Edison Company ("SCE"). My present capacity is Manager, Nuclear Fuel Management, for the San Onofre Nuclear Generating Station ("SONGS"), and in that capacity I am authorized to execute this Affidavit.

2. SCE is the operating agent for SONGS. I am familiar with the policies established by SCE to determine whether certain SCE information is proprietary and confidential, and to ensure the proper application of these policies.

3. I am familiar with SCE information in the document entitled "San Onofre Nuclear Generating Station, Units 2 and 3, Proposed Permanent Exemption Request and Proposed Change Number (PCN) 600, Amendment Application Numbers 261 and 247, Request for Unrestricted Use of AREVA Fuel," (referred to herein as "Document") submitted to the NRC in July 2011.

4. SCE has classified the information contained in the Document as proprietary and confidential in accordance with SCE's policies.

5. Specifically, SCE applied the following criteria to determine that the information 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 AREVA have provided to SCE certain proprietary and confidential information contained in the Document.

(b) The information reveals details of Westinghouse's, SCE's, and/or AREVA's research and development plans and programs, or the results of these plans and programs.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive commercial 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 competitive commercial advantage for Westinghouse, SCE, and/or AREVA on product optimization 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 significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(f) The information contained in the Document is vital to a competitive commercial advantage held by Westinghouse, SCE, and/or AREVA, would be helpful to their competitors, and would likely cause substantial harm to the competitive position of Westinghouse, SCE, and AREVA.

6. The information contained in the Document is considered proprietary and confidential for the reasons set forth in Paragraph 5. In addition, the information contained in the Document 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 am aware that other companies also regard the type of information contained in the Document as proprietary and confidential.

7. In accordance with the Westinghouse-AREVA-SCE NDA, the Document has been made available to the NRC in confidence, with the request that the information contained in this Document be withheld from public disclosure. The request for withholding the information from public disclosure is made in accordance with 10 CFR 2.390. The information qualifies for withholding from public disclosure under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

8. In accordance with SCE's policies governing the protection and control of proprietary and confidential information, the information contained in the Document has been made available, on a limited basis, to others outside Westinghouse, SCE and AREVA only as required in accordance with the Westinghouse-AREVA-SCE NDA.

9. SCE's policies require that proprietary and confidential information be kept in a secured file or area and distributed on a need-to-know basis. The information contained in the Document 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 declare under penalty of perjury under the laws of the State of California that the above is true and correct.

0 enJ. ho SUBSCRIBED before me this m. "f.Uaab "-o"" -.. ~,t day of _2011. _ _ ,,

pmusdism tob go*ngmm* theIb Who "WW f90.%"64 i M A NOTARY PUBLIC, STATE OF CALIFORNIA MY COMMISSION EXPIRES: T'1I .IMI PTI M...

Reg. #: Commission # 1936996 Notary Public - California z Orange County My Comm. Egiaes Jun 14,2015

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP 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 whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report ANP-2975(P), Revision 0, entitled "San Onofre Nuclear Generating Station Unit 2 and Unit 3 Realistic Large Break LOCA Report," dated June 2011 and referred to herein as "Document."

Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld 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 is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secret and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly 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 a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(c) and 6(e) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file 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 this ____

day of 2011.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/14 Reg. # 7079129

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP 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 whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report ANP-2974(P), Revision 0, entitled "San Onofre Nuclear Generating Station Unit 2 and Unit 3 Small Break LOCA Report," dated July 2011 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in

this Document be withheld 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 is requested qualifies under 10 CFR 2.390(a)(4)

"Trade secret and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly 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 a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(c) and 6(e) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been

made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file 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 this_____

dayof

• uA 2011.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/14 Reg. # 7079129 NotarV.Public b Commonwealth of Virginia 7079129 My Commission Expires Oct 31, 2014r

ENCLOSURE 5 LICENSEE'S EVALUATION (Non-Proprietary)

SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

ENCLOSURE 5 LICENSEE'S EVALUATION (Non-Proprietary)

Permanent Exemption Request and SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

SUMMARY

2.0 VENDOR QUALIFICATION PROGRAM (VQP) - OVERVIEW 2.1 LFA Demonstration Program 2.2 Licensing Basis Changes 2.3 Demonstration of Updated SCE Reload Analysis Methodology for AREVA Fuel Design 3.0 CHANGES TO SONGS LICENSING BASIS 3.1 Permanent Exemption - 10 CFR 50.46 and 10 CFR 50 Appendix K for M5TM 3.2 Technical Specifications / LCS COLR Changes 4.0 CHANGES TO SCE RELOAD ANALYSIS PROCESS AND METHODOLOGY 4.1 Physics Analysis 4.2 Core Thermal Hydraulic Analysis 4.3 Fuel Rod Behavior Analysis 4.4 Vendor Qualified Maximum Fuel Burnup Limits 4.5 Non-LOCA Transient Analysis 4.6 LOCA Analysis - Interface with Fuel Vendor 4.7 Fuel Mechanical Design Analysis - Interface with Fuel Vendor 4.8 COLSS/CPC Setpoints Analysis 4.9 Alternative Source Term Dose Consequence 4.1Q Fuel Centerline Melting Temperature 5.0 1 AREVA SCOPE OF RELOAD ANALYSIS 5.1 Fuel Mechanical Design Analysis 5.2 LOCA Analyses 6.0 CHANGES TO REACTOR CORE DESIGN AND MONITORING PROGRAM 6.1 Fuel Management Guidelines for SONGS Units 2 and 3 6.2 Reactor Core Design Review Team 6.3 Reload Ground Rules

6.4 Vendor Fuel Design Change Interface 6.5 Vendor Reload Analysis Computer Codes and Methodology Interface 6.6 SCE Engineer Training Qualification Guide 6.7 Core Reload Analyses and Activities Checklist 6.8 Source Verification and Vendor Fuel Fabrication Interface 6.9 Fuel Vendor Engineering Interface 6.10 Site Program Impact 6.11 Licensing and Design Basis Document Updates 6.12 Design Process Flow and Controls 6.13 COLSS/CPC Products 6.14 Low Power and Power Ascension Testing 6.15 Core, Spent Fuel Pool and Dry Cask Storage Requirements 6.16 Core and Fuel Monitoring 7.0 IMPACT OF CHANGES ON RELOAD ANALYSES 7.1 AREVA Fuel Reload Core Design Comparison of Key Physics Parameters 7.2 AREVA Fuel Reload Core Thermal Hydraulic Analysis 7.3 FATES3B w/M5TM Fuel Behavior Predictions 7.4 VQP Comparison of Selected Non-LOCA Transient Analysis

8.0 REFERENCES

ATTACHMENTS A. Listing of Acronyms B. Technical Specification and Bases Changes C. SONGS - Summary of Impact on UFSAR Chapter 15 Events D. Compilation of Calvert Cliffs RAI's - Applications to SONGS and Responses E. Evaluation of AREVA SER and TER Methodology Limitations

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

SUMMARY

The San Onofre Nuclear Generating Station (SONGS) Unit 2 and Unit 3 cores each consist of 217 fuel assemblies. Each fuel assembly consists of 236 fuel rods. The rods are arranged in a square 16 x 16 array (CE16). The current fuel vendor (Westinghouse-CE, or simply, Westinghouse; formerly ABB/CE) fuel rods consist of slightly enriched uranium dioxide cylindrical ceramic pellets, encapsulated within a cylindrical Zircaloy or ZIRLOTM tube.

Southern California Edison (SCE) is currently evaluating the use of alternative fuel vendor (AREVA NP, or, simply AREVA) fuel assemblies in order to eliminate grid to rod fretting fuel failures. Unlike current fuel assemblies, the AREVA fuel assemblies will contain M 5 TM alloy cladding material. A Lead Fuel Assembly (LFA) program to demonstrate fuel design compatibility has been reviewed and approved by the NRC (References 8.2 and 8.6), and is currently underway at SONGS.

The next step in the licensing process for AREVA fuel is the implementation of licensing basis changes for SONGS Units 2 and 3 that will enable the smooth transition between the LFA program and unrestricted fuel use. These proposed changes to the SONGS licensing basis have individually been previously reviewed and approved by the NRC for fuel vendor use and/or other utility licensee usage. This SONGS PCN 600 change request will coordinate and implement the approved AREVA methodologies, parameters and correlations into the approved SCE Reload Analysis Methodology (Reference 8.4) process and the SONGS licensing basis.

Section 4 of the SCE Reload Analysis Methodology (Reference 8.4) document describes the Reactor Core Design and Monitoring Program that ensures that there is a thorough engineering and safety evaluation performed to support each reactor core design and its subsequent operation. Section 3.2 of the NRC Evaluation accompanying the Reload Analysis Methodology approval letter describes the NRC audit of the process and controls that SCE applies to the reload process, including procedures, checklists, design calculations and associated reviews, facility change evaluations (now referred to as Nuclear Engineering Change Packages (NECPs)), training materials, SCE training records, and correspondence between the fuel vendor (then, ABB/CE, now Westinghouse) and SCE. Section 3.2.7 of the NRC Evaluation of Reference 8.4 states:

Page 1 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 3.2.7 Change in Fuel Vendor The models and methods discussed in the SE and reviewed during the audit were approved by the NRC for use by ABB/CE and have been transferredto SCE via the RTT (Reload Technology Transfer) Program. Any change in fuel vendor would require an evaluation of changes requiredto the physics and safety analysis methodology to accommodate that vendor's particularfuel designs. Changes of this type would require a thorough engineering evaluation, verification, and validation priorto use of the new fuel design. If necessary,a topical report covering significant changes in models and/or methods would be requiredby the NRC for review and approval.

Specifically included in the Section 4 of the SCE Reload Analysis Methodology (Reference 8.4) scope are Section 4.4 - Fuel Vendor (ABB/CE) Fuel Design Change Interface, Section 4.5 - Fuel Vendor (ABB/CE) Reload Analysis Computer Codes and Methodology Interface, Section 4.9 - Fuel Vendor (ABB/CE) Engineering Interface, and Section 4.11 - Licensing and Design Basis Updates. These NRC reviewed and approved components of the SCE Reload Analysis Methodology have been applied to the proposed transition to AREVA fuel supply. This process has produced a thorough identification, engineering evaluation, verification and validation of the necessary changes associated with this potential fuel vendor change. The results of this change management and control process are presented in several sections of this license amendment request.

Section 2.0 of this document provides an overview of the VQP program. Section 3.0 provides the details of the PCN 600 changes to the SONGS licensing basis. Sections 3.1 and 3.2 provide the regulatory evaluations required to be provided with the request for permanent exemption from 10 CFR 50.46 fuel cladding material listing, and for the SONGS specific Technical Specification changes. Section 4.0 provides the description and justifications for the SCE controlled reload analysis methodology updates required to incorporate approved AREVA methodologies, parameters and correlations into approved SCE Reload Analysis Methodology (Reference 8.4). Section 5.0 describes the revised fuel vendor interface (process and responsibility) relationships. Section 6.0 presents the summary of the SONGS review of reactor core design and monitoring processes and the changes required for the use of AREVA fuel. Section 7.0 provides selected demonstration summaries of the impacted SONGS reload design analyses.

Page 2 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel This license amendment requests NRC approval of a number of items required to enable SCE to analyze reload cores utilizing the AREVA fuel design. The fuel design to be used at SONGS is the same as the AREVA LFA design irradiated at SONGS except for minor modifications to enhance the seismic characteristics of the fuel assembly.

The differences between the reload fuel design and the LFA design are discussed in Section 5.1. The number of fuel assemblies in the initial batch will be between eight fuel assemblies and approximately a half core of fuel assemblies. The fuel design used in the SONGS reactors will be analyzed with the analytic methods described in this enclosure. The exact reload core fuel management has not been defined at this time.

The reload cores that the reload fuel design is inserted into will also be analyzed with the analytic methods described in this enclosure. The analytic methods described in this enclosure have either been previously approved by the NRC or are based on SCE and AREVA analytic methods that have been approved by the NRC and either modified to integrate the AREVA methods with SCE methods or modified to reflect previous NRC comments on the AREVA methods.

Page 3 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 2.0 VENDOR QUALIFICATION PROGRAM - OVERVIEW In a manner similar to the SCE Reload Technology Transfer program described and implemented under the Reference 8.4 SCE Reload Analysis Methodology submittal, the potential SONGS transition from Westinghouse to AREVA fuel supply involves carefully planned and coordinated (with both fuel vendors) technical and licensing basis activities.

For the purpose of this submittal, "Westinghouse" is used to denote all prior nuclear vendors that have been acquired by Westinghouse or their current subsidiaries (e.g.,

ABB, Combustion Engineering (CE), ABB/CE, etc.). Similarly, "AREVA" is used to denote all prior companies that have been acquired by AREVA or their current subsidiaries (e.g., AREVA-NP, Exxon Nuclear, Siemens, etc.). See Attachment A for more information.

2.1 LFA Demonstration Program The LFA demonstration program is described in detail in Reference 8.1. In addition to the physical demonstration of the acceptability of this alternative fuel vendor design at the SONGS facility, the LFA program also provided a demonstration and training opportunity for the SCE-vendor relationships and interfaces, the SCE and vendor controlled tools and design models, and the training and qualification needs for the SCE staff that are needed to implement AREVA as a fuel vendor for SONGS.

2.2 Licensing Basis Changes The SONGS licensing basis changes are more limited than changes requested and approved for other licensees who have made, or who are considering a change from Westinghouse to AREVA fuel supply. Other licensees (such as Calvert Cliffs) have requested a conversion from their existing fuel supplier methodology to the proposed new fuel supplier methodology for the full scope of reload fuel design and analysis. The SONGS licensing basis changes are limited to the addition of AREVA Loss Of Coolant Analysis (LOCA) methods, mechanical design methods, manufacturing data, and the addition of selected AREVA correlations and parameters (including adoption of such items as AREVA specific Departure From Nucleate Boiling Ratio (DNBR) correlations) into the existing NRC-approved Westinghouse-CE based SONGS reload design analysis methodology.

The SONGS specific licensing basis changes are described in detail in Section 3.0.

These changes were identified during the application of the SCE Reload Analysis Methodology (Reference 8.4) change management and control process described in parts of Section 4.0 of that methodology. The approved reload analysis process was applied during the LFA demonstration program to identify any program, process, modeling requirement, or referenced licensing basis that was impacted. An impact assessment of the effects on the SCE Reload Analysis Methodology Section 4.0 Page 4 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel (Reactor Core Design and Monitoring Program) was performed and is documented in Section 6.0.

2.3 Demonstration of Updated SCE Reload Analysis Methodology for AREVA Fuel Design The SCE objective is to put in place approved licensing bases, engineering and vendor interface processes, analysis tools, personnel training and qualification packages, and supporting benchmark documentation that will permit SCE to smoothly and effectively transition from Westinghouse fuel to unrestricted use of AREVA fuel. One measure of success of these objectives is the SCE staff's ability to independently perform all required reload analysis activities for the AREVA fuel design, and to achieve results of comparable quality to those obtained from the currently approved reload analysis process for Westinghouse fuel. Accordingly, the applicable scope of the approved SCE Reload Analysis Methodology (Reference 8.4) was exercised for a "Vendor Qualification Program (VQP) Licensing Basis Fuel Cycle" as a demonstration of this capability. The VQP fuel cycle(s) are demonstration basis fuel cycles.

Changes to the SCE reload analysis tools and models are documented in Section 4.0.

Some of the changes involve the addition of portions of the NRC-approved AREVA licensing basis fuel design parameters and correlations as user-selectable options within the SCE Reload Analysis Methodology. Section 4.0 mirrors Section 3.0 of the approved SCE Reload Analysis Methodology, which provides an overview of the SCE Reload Analysis Process.

Selected analysis results using the proposed updates to the SCE reload analysis process are provided in Section 7.0. This documentation provides comparisons to the results of the similar calculations using the currently approved process for the current fuel designs, or to reference results obtained from the fuel vendor(s), such as LOCA and fuel mechanical design.

Page 5 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 3.0 CHANGES TO SONGS LICENSING BASIS 3.1 Permanent Exemption - 10 CFR 50.46 and 10 CFR 50 Appendix K for M5TM The proposed unrestricted use of AREVA fuel at SONGS Units 2 and 3 requires a permanent exemption from the requirements of 10 CFR 50.46 "Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors," and 10 CFR 50, Appendix K, "ECCS Evaluation Models." Reference 8.2 documents the NRC approval of a temporary exemption from these requirements to support the SONGS Units 2 and 3 Lead Fuel Assembly (LFA) program. The requested permanent exemption will replace that approved temporary exemption.

Part 50.46(a)(l)(i) of Title 10 of the Code of Federal Regulations (10 CFR 50.46(a)(l)(i))

states in part:

"Eachboiling or pressurizedlight-water nuclearpower reactorfueled with uranium oxide pellets within cylindrical Zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculatedcooling performance following postulated loss-of-coolant accidents conforms to the criteriaset forth in paragraph(b) of this section.

ECCS cooling performance must be calculatedin accordance with an acceptable evaluation model and must be calculatedfor a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated. "

10 CFR 50.46 continues with a delineation of specifications for peak cladding temperature, maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and lonýterm cooling. Since 10 CFR 50.46 specifically refers to fuel with Zircaloy or ZIRLO cladding and doesn't list AREVA MSM cladding, the use M5TM cladding requires a permanent exemption from this section of the regulations.

10 CFR 50, Appendix K, paragraph I.A.5, states in part:

"The rate of energy release, hydrogen generation, and cladding oxidation from the metal/waterreactionshall be calculated using the Baker-Justequation."

The Baker-Just equation presumes the use of Zircaloy or ZIRLOTM cladding. The routine use of M5 cladding requires a permanent exemption from this section of the regulations.

Page 6 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Pursuant to 10 CFR 50.12, "Specific Exemptions," SCE is requesting a permanent exemption from the requirements of 10 CFR 50.46 "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors" and 10 CFR 50, Appendix K "ECCS Evaluation Models" for SONGS Units 2 and 3.

The permanent exemption will allow the unrestricted use of fuel assemblies manufactured by AREVA with M5TM alloy clad fuel rods, consistent with NRC-approved SCE and AREVA design and analysis methodologies.

Currently, numerous US pressurized water reactors (PWRs) have used M5TM alloy cladding fuel assemblies in full batch applications, including Arkansas Nuclear One Unit 1, Crystal River Unit 3, Three Mile Island Unit 1, Davis Besse, North Anna Units 1 and 2, Sequoyah Units 1 and 2, Oconee Units 1, 2 and 3, Ft. Calhoun, Palisades and Calvert Cliffs Units 1 and 2. M5TM alloy clad fuel assemblies have also been used extensively in European plants.

10 CFR 50.12. Specific Exemption The standards set forth in 10 CFR 50.12 provide that the Commission may grant exemptions from the requirements of the regulations for reasons consistent with the following:

• The exemption is authorized by law;

• The exemption will not present an undue risk to the public health and safety;

• The exemption is consistent with the common defense and security; and 0 Special circumstances are present.

This exemption is authorized by law. The remaining standards for the permanent exemption are also satisfied, as described in the following paragraphs.

The exemption will not present an undue risk to public health and safety. The NRC-approved M5TM topical reports (References 8.9 and 8.10) demonstrate that predicted chemical, mechanical, and material performance characteristics of the M5TM alloy cladding are within those approved for Zircaloy under anticipated operational occurrences and postulated accidents. Normal reload design and analysis methodologies in use at SCE and at the fuel vendor will evaluate the M5TM clad fuel susceptibility to failure during normal operation, anticipated operational occurrences and postulated accidents for each core design. M5TM fuel performance (as well as any co-resident Zircaloy and/or ZIRLOTM fuel) in each core design will be evaluated and any predicted fuel failures will be limited such that dose consequence impacts are within the applicable regulatory limits. Therefore, the use of M5TM clad will not present an undue risk to the public health and safety.

Page 7 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The exemption is consistent with the common defense and security. The use of M5TM clad fuel assemblies enables the implementation of a more robust design to eliminate grid to rod fretting fuel failures. This change in fuel material used in the plant has no relation to security issues. Therefore, the common defense and security are not impacted by this exemption request.

Special circumstances are present. As set forth in 10 CFR 50.12(a)(2)(ii), which states that special circumstances are present whenever "Applicationof the regulationin the particularcircumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule..."

10 CFR 50.46 identifies acceptance criteria for ECCS system performance at nuclear power facilities. The effectiveness of the ECCS in SONGS Units 2 and 3 will not be affected by the use of M5TM clad fuel assemblies. Due to the similarities in the material properties of the M5TM alloy to Zircaloy or ZIRLOTM as identified in the AREVA M5TM alloy topical reports (References 8.9 and 8.10) it can be concluded that the ECCS performance would not be adversely affected.

The intent of paragraph I.A.5 of Appendix K to 10 CFR 50 is to apply an equation for rates of energy release, hydrogen generation, and cladding oxidation from metal-water reaction that conservatively bounds all post-LOCA (Loss of Coolant Accident) scenarios. The supporting documentation for the AREVA M5TM topical reports (References 8.9 and 8.10) shows that due to the similarities in the composition of the M5TM alloy cladding and Zircaloy or ZIRLOTM, the application of the Baker-Just equation will continue to conservatively bound all post-LOCA scenarios.

A strict interpretation of 10 CFR 50.46 and 10 CFR 50, Appendix K, would not allow the use of M5TM alloy cladding fuel rods in reload fuel assemblies since the cladding material does not fall within the strict definition of Zircaloy or ZIRLOTM even though the AREVA M5TM topical reports show that the intent of the regulations is met. Application of these regulations in this particular circumstance would not serve the underlying purpose of the rule and is not necessary to achieve the underlying purpose of the rule, so special circumstances exist.

Page 8 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 3.2 Technical Specifications Changes The proposed unrestricted use of AREVA fuel at SONGS Units 2 and 3 require the following Technical Specifications changes:

1) SONGS Technical Specification 2.1.1.2 (Reactor Core Safety Limits) will be updated to specify a fuel centerline melt safety limit with corresponding adjustments made to account for gadolinia as a burnable absorber.

2) SONGS Technical Specification 4.2.1 (Fuel Assemblies) descrition of clad materials will be updated to specify that SONGS may use M5T clad.

3) Changes to the SONGS Technical Specification 5.7.1.5 (CORE OPERATING LIMITS REPORT (COLR)) methodology reference list that are required to support the core design and implementation of reload quantities of the potential AREVA fuel assemblies.

Technical Specification changes are handled in consideration of the requirements contained in 10 CFR 50.90. A facility license change, such as the requested permanent exemption from certain 10 CFR 50.46 and 10 CFR 50 Appendix K requirements to allow for the use of AREVA M5TM cladding, is handled in consideration of the requirements contained in 10 CFR 50.12, "Specific exemptions."

Similar Technical Specification changes are requested for both SONGS Units 2 and 3.

Mark-ups of the proposed changes are provided in Attachment B.

3.2.1 Description of Technical Specification Changes 3.2.1.1 Technical Specification 2.1.1.2 - Reactor Core Safety Limits The restrictions of Safety Limit (SL) 2.1.1.2 prevent overheating of the fuel and cladding and possible cladding perforation that would result in the release of fission products to the reactor coolant. The melting point of the fuel is dependent on fuel burnup and the amount and type of burnable poison used in the fuel. It is based on physical fuel material properties that are independent of fuel geometry and therefore independent of fuel manufacturer. SONGS will continue to use the approved Westinghouse methodology. The peak centerline temperature must be maintained less than 5080°F decreasing by 58°F per 10,000 MWD/MTU and adjusted for burnable poisons per CENPD-382-P-A (erbia) and per CENPD-275-P-A including Supplement 1-P-A (gadolinia). The adjustment for burnable poisons is proprietary and cannot be included in the Technical Specifications.

SONGS Units 2 and 3 Technical Specification 2.1.1.2 (Reactor Core Safety Limits) currently presents the peak fuel centerline temperature Safety Limit with a burnable poison adjustment based on Topical Report CENPD-382-P-A (Reference 8.32). This Topical Report specifically addresses erbia as a burnable poison. Technical Page 9 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Specification 2.1.1.2 does not currently cite a Topical Report that addresses peak fuel centerline temperature adjustments for gadolinia as a burnable poison. For Westinghouse/CE fuel designs, NUREG-1432 (Standard Technical Specifications -

Combustion Engineering Plants, Revision 3, Volume 2) describes the application of a gadolinia fuel rod adjustment to fuel centerline melt temperature based on CENPD-275-P-A (Reference 8.46). To utilize gadolinia burnable absorber fuel rods, Technical Specification 2.1.1.2 and its Bases are proposed to be modified to address temperature limits for all fuel types resident in SONGS cores. The modification to Technical Specification 2.1.1.2 will add specific references for the methodology documents that describe the application of a gadolinia fuel rod adjustment to fuel centerline melt temperature based on CENPD-275-P-A and its Supplement 1-P-A (References 8.46 and 8.47, respectively).

The inclusion of Supplement 1-P-A to CENPD-275-P-A updates Technical Specification 2.1.1.2 for consistency with the current SONGS licensing basis. SCE is currently licensed to use gadolinia as a burnable poison per Technical Specification 4.2.1. The SCE Reload Topical (Reference 8.4, Section 3.1.1) addresses the use of gadolinia as a burnable absorber through its citation of CENPD-275-P-A (Supplement 1-P-A to CENPD-275-P-A was approved by the NRC in April 1999 just prior to the approval of the SCE Reload Topical in June 1999). In 2009, the NRC-approved a supplement to the SCE Reload Analysis Methodology to incorporate the Studsvik Scandpower CASMO4 and SIMULATE-3 computer codes and models, including the modeling of gadolinia as a burnable absorber, into the SCE reload design process (Reference 8.5).

3.2.1.2 Technical Specification 4.2.1 - Fuel Assemblies SONGS Units 2 and 3 Technical Specification 4.2.1 (Fuel Assemblies) requires revision to fully implement unrestricted use of AREVA fuel. The permanent exemption to 10 CFR 50.46 and 10 CFR Appendix K will include the AREVA M5TM cladding material in the population of approved materials described by this Technical Specification.

3.2.1.3 Technical Specification 5.7.1.5 - Core Operating Limits Report (COLR)

This section provides a list of the analytical methods used to determine the core operating limits. The methods have been previously reviewed and approved by the NRC. With the change in fuel design, an accompanying change in methods used to derive and validate the core operating limits is necessary. These methods must be listed in Technical Specification 5.7.1.5. SCE is proposing to add the appropriate AREVA methods to the list and keep the existing Westinghouse methods. The different methods will be applied to the correct fuel rod cladding or grid design. SER limitations and restrictions for AREVA methods are provided in Attachment E. The added references are:

Page 10 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors" EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based" BAW-10240(P)-A, "Incorporation of M5TM Properties in Framatome ANP Approved Methods" Letter dated ____, -(NRC)to (SCE), "Issuance of Amendments for Unit 2 and for Unit 3, Request for Unrestricted Use of AREVA fuel, San Onofre Nuclear Generating Station."

3.2.2 Technical Evaluation The Technical Evaluation of the acceptability of these Technical Specification changes is described throughout Sections 4.0, 5.0, 6.0 and 7.0.

3.2.3 Regulatory Evaluations 3.2.3.1 Applicable Regulatory Requirements The NRC-approved SCE reload analysis methodologies, described in References 8.4 and 8.5, meet applicable regulatory requirements, including the following 10 CFR 50 Appendix A general design criteria:

Criterion 10 - Reactor design. The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

Criterion 11 - Reactor inherentprotection. The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

Criterion 12 - Suppression of reactorpower oscillations. The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

SCE evaluated the proposed changes, and determined that the specified acceptable fuel design limits will not be exceeded during normal operation, including the effects of anticipated operational occurrences. The characteristics of the reactor core, including prompt inherent nuclear feedback and suppression of power oscillations, will continue to Page 11 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel be ensured. The proposed changes do not change any assumptions previously made in evaluating radiological consequences or adversely affect any fission product barriers, nor does it increase the consequences of events described and evaluated in Chapter 15 of the UFSAR.

3.2.3.2 No Significant Hazards Consideration This license amendment request continues to support the use of AREVA fuel, and is being submitted to implement license, Technical Specification, and reload analysis methodology updates that are required to facilitate an unrestricted use of AREVA fuel at the Southern California Edison (SCE) SONGS Units 2 and 3, starting with Unit 3, Cycle

17. The changes consist of:

1) A permanent SONGS license exemption from the requirements of 10 CFR 50.46 and 10 CFR 50 Appendix K, to the extent to which these regulations limit acceptable cladding materials to only Zircaloy and ZIRLOT . The requested exemption will allow the use of fuel rods clad with AREVA M5TM cladding. A temporary exemption from these requirements has been granted to SONGS by the NRC in Reference 8.2 for license changes needed to support the Lead Fuel Assembly (LFA) program involving AREVA M5TM cladding (NRC ADAMS Package ML090860392). The NRC has also recently approved a permanent exemption for Calvert Cliffs regarding the use of the same cladding material and very similar fuel design from the same fuel vendor (NRC ADAMS Accession Numbers ML101760523 and ML110190632).

2) Changes to SONGS Technical Specification 5.7.1.5 (CORE OPERATING LIMITS REPORT (COLR)) methodology reference list that are required to support the core design and implementation of unrestricted use of the potential alternative vendor fuel assemblies.

3) The SONGS Technical Specification 4.2.1 (Fuel Assemblies) description of clad materials will also be updated to include AREVA M5TM cladding.

4) In addition, the SONGS Technical Specification 2.1.1.2 (Reactor Core Safety Limits) will be updated to specifically identify a fuel centerline melt safety limit with corresponding adjustments made to account for burnable absorber fuel rods.

5) Incorporation of fuel burnup limits consistent with AREVA M5TM for the AREVA fuel assemblies (generically approved by the NRC in prior fuel vendor submittals) into the SONGS licensing basis.

SCE has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment", as discussed below.

Page 12 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The reactor fuel and the analyses associated with the fuel are not accident initiators. The response of the fuel to an accident is analyzed using conservative techniques and the results are compared to the approved acceptance criteria.

These evaluation results will show that the fuel response to an accident is within approved acceptance criteria for both cores loaded with the new AREVA CE-HTP (High Thermal Performance) fuel and for cores loaded with both AREVA and Westinghouse design fuel. Therefore, the change in fuel design does not affect accident or transient initiation or consequences.

The proposed change to Technical Specification 2.1.1.2 (Reactor Core Safety Limits) does not require any physical change to any plant system, structure, or component. The change to establish the peak fuel centerline temperature is consistent with existing approved analysis methodology.

The proposed change to Technical Specification 4.2.1 (Fuel Assemblies) includes M5TM cladding. The change in cladding materials and fuel assembly design such as grids has been evaluated in this submittal and all acceptance criteria are met.

Topical Reports have been reviewed and approved by the NRC for use in determining core operating limits. The core operating limits to be developed using the new methodologies will be established in according with the applicable limitations as documented in the appropriate NRC Safety Evaluation reports.

The proposed change to Technical Specification 5.7.1.5 (Core Operating Limits Report (COLR)) enables the use of appropriate methodologies to analyze accidents. The proposed methodologies will ensure that the plant continues to meet applicable design criteria and safety analysis acceptance criteria.

The proposed change to the list of NRC-approved methodologies listed in Technical Specification 5.7.1.5 has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident.

The proposed change will update the listing of NRC-approved methodologies to allow analysis of both AREVA and Westinghouse fuel designs. Changes to the calculated core operating limits may only be made using NRC-approved methods, must be consistent with all applicable safety analysis limits and are controlled by the 10 CFR 50.59 process. The list of methodologies in Technical Specification 5.7.1.5 does not impact either the initiation of an accident or the mitigation of its consequences.

Page 13 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Use of AREVA CE-HTP fuel in SONGS reactor cores is consistent with the current plant design bases and does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigations. The operational characteristics of AREVA CE-HTP fuel are bounded by the safety analyses. The AREVA CE-HTP fuel design performs within fuel design limits and does not create the possibility of a new or different accident.

The proposed change to the Technical Specification 2.1.1.2 does not require any physical change to any plant system, structure, or component, nor does it require any change in safety analysis methods or results. The existing analyses remain unchanged and do not affect any accident initiators that would create a new accident.

The proposed change to Technical Specification 4.2.1 does not create any new accident initiators. For example, postulated pipe breaks and valve motions are unaffected by the fuel design. Possible impacts such as postulated CEA motions are unaffected because the interface between the fuel assembly and the CEA has been designed to be unchanged.

The proposed change to the list of NRC-approved methodologies listed in Technical Specification 5.7.1.5 has no impact on any plant configuration or system performance. It updates the list of NRC-approved topical reports used to develop the core operating limits. There is no change to the parameters within which the plant is normally operated. The possibility of a new or different accident is not created.

Therefore, the proposed change does not create the possibility of a new or different kind of accident form any accident previously evaluated.

Page 14 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No. The proposed change does not involve a significant reduction in a margin of safety. The margin of safety as defined in the basis for any technical specification will not be reduced by the proposed change to the computer programs used for physics calculations for nuclear design analyses.

Use of AREVA CE-HTP fuel in SONGS reactor cores is consistent with the current plant design bases and does not adversely affect any fission product barrier, nor does it alter the safety function of safety systems, structures, or components, or their roles in accident prevention or mitigation. The operational characteristics of AREVA CE-HTP fuel in SONGS reactor cores are evaluated by the safety analyses and meet the safety analysis criteria. The AREVA CE-HTP fuel in SONGS reactor cores performs within fuel design limits. The proposed changes do not result in exceeding design basis limits. Therefore, all licensed safety margins are maintained.

The proposed change to Technical Specification 2.1.1.2 does not require any physical change to any plant system, structure, or component, nor does it require any change in safety analysis methods or results. Therefore, by changing the peak fuel centerline temperature adjustment for burnable poisons, the margin as established in the current licensing basis remains unchanged.

The proposed change to Technical Specification 4.2.1 has been evaluated in this submittal and all acceptance criteria are met.

The proposed change to the list of NRC-approved methodologies listed in Technical Specification 5.7.1.5 has no impact on any plant configuration or system performance. Topical Reports have been reviewed and approved by the NRC for use in determining core operating limits. The proposed methodologies will ensure that the plant continues to meet applicable design criteria and safety analysis acceptance criteria.

Page 15 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 3.2.4 Conclusions In conclusion, based on the considerations discussed in this section, (1) There is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) Such activities will be conducted in compliance with the Commission's regulations, and (3) The issuance of the amendment will not be inimical the common defense and security or the health and safety of the public.

3.2.5 Environmental Consideration Based on the above considerations, the proposed license change does not involve and will not result in a condition which significantly alters the impact of San Onofre on the environment. Thus, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR Part 51.22(c)(9), and pursuant to 10 CFR Part 51.22(b),

no environmental assessment need be prepared.

Page 16 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.0 CHANGES TO SCE RELOAD ANALYSIS PROCESS AND METHODOLOGY Section 3.0 of Reference 8.4 provides an overview of the SCE reload analysis process and methodology. In general, the requested licensing changes will not significantly change the flow paths or the relationships between SCE and fuel vendor processes and procedures. The detailed analyses and their inputs will change in some cases, but the overall process remains the same, except as noted in the following paragraphs.

There will be small changes in the process, and in the fuel vendor interface responsibilities. The Reference 8.4 Figure 3.0-2 Simplified Diagram of Reload Analysis Network has been updated to reflect this license amendment request, and is provided as Figure 4.1. Two areas of SCE responsibility are directly affected by this potential change to AREVA fuel supply. The Core Thermal Hydraulic Analysis scope and the Fuel Rod Performance Analysis scope within the SCE responsibilities will change. The indirect impacts on non-LOCA transient analyses performed by SCE are discussed in Section 4.5.

The current SCE Core Thermal Hydraulic Analysis scope will be modified to the extent that [

]for compliance with Technical Specification limits on DNBR. Section 4.2 discusses the specifics of the new correlation implementation. There will be no change made to the existing DNBR technical specification limit.

The current SCE Fuel Rod Performance Analysis scope will be modified [

] between SCE and AREVA for AREVA supplied reloads. Currently, SCE fully implements [

].

SCE will continue to apply that portion of the fuel rod performance analysis process

[ I Implementation of [

] will enable SCE to continue to perform these functions. Section 4.3 of this LAR discusses the specifics of the new analysis modules and their implementation.

As a result of the intended implementation of the [

Page 17 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

] Henceforth, the SCE Fuel Rod Performance analysis process will be referred to as the SCE Fuel Rod "Behavior" analysis process to differentiate it from the AREVA scope.

The NRC-approved SCE methodology also considers the impact of non-LOCA analyses on fuel behavior, specifically regarding the amount of fuel damage that may result from these transients. [

] The approved SCE evaluation process for DNBR propagation and statistical convolution will be retained based on the discussion and justifications provided in Section 4.5.

Page 18 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.1 Simplified Diagram of Reload Analysis Network - AREVA Fuel Supply AA11 oal ~

i .............

ICOR <t.w~

Stanup l~.N P~p;Ij

\

~tjiu C;

f71

/TS B, G,,m Ls &

L Ar' f' fC]!;[P( "'

/ f~Af.;- I

(~LCC CP(

1' SCENendor Interface SCE Scope SCE Scope-impacted by iAREVA Fuel Page 19 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.1 Physics Analysis The SCE Reload Analysis Methodology (Reference 8.4) Section 3.1 has been supplemented with Reference 8.5 (approved in Reference 8.6) to incorporate the Studsvik Scandpower CASMO-4/SIMULATE-3 computer codes and models into the SCE reload design process. The CASMO/SIMULATE code package consists of general purpose nuclear analysis computer codes, which were designed to be able to model a wide variety of fuel vendor designs, including the CE/Westinghouse designs and the AREVA designs planned for use at SONGS. No changes were needed to the physics processes described in Reference 8.4 to model AREVA cores. To ensure applicability of AREVA designs to the SONGS process CE generated transition and full core AREVA fuel cycles. Details of these VQP cycles are shown in Section 7.1.

4.1.1 CASMO-4 CASMO-4 (Reference 8.5) is a multi-group, two-dimensional neutron transport theory lattice physics code with depletion capability and the ability to generate cross-sections and discontinuity factors for both boiling water reactor (BWR) and pressurized water reactor (PWR) diffusion theory core analysis. Since Zircaloy-4, ZIRLOTM and M5TM cladding are neutronically the same, they are all modeled as Zircaloy-4 in CASMO-4. In addition, CASMO-4 is capable of accurately modeling both Westinghouse and AREVA burnable poisons including A12 0 3 -g 4C, Erbia, Gadolinia and Zirconium Diboride (Reference 8.5).

4.1.2 SIMULATE-3 SIMULATE-3 (Reference 8.5) is a two-group, three-dimensional (3-D), coarse mesh nodal diffusion theory reactor simulator computer program that employs both thermal-hydraulic and Doppler feedback. The nodal thermal-hydraulic properties are calculated based on the inlet temperature, RCS pressure, coolant mass flow rate, and the heat addition along the channel.

The pin-by-pin power distributions, on a 2-D or 3-D basis, are constructed from the inter- and intra-assembly information from the coarse mesh solution and the pin-wise assembly power distribution from CASMO-4. SIMULATE-3 performs a macroscopic depletion and individual uranium, plutonium, and fission product isotope concentrations are not computed. In addition, microscopic depletion of iodine, xenon, promethium, and samarium is included to allow modeling of typical reactor transients. No modifications were needed to the SIMULATE-3 code or models to implement AREVA fuel.

Page 20 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.1.3 SONGS Core Center Assembly Typical core reloads are no more than half core reloads. The SONGS reactor core has an odd number of fuel assemblies (i.e., 217) and thus the center assembly is typically retained for a third cycle of operation or re-inserted from the spent fuel pool. After a transition to a new fuel vendor, the need to use an assembly from the old vendor for the center assembly may be necessary. This assembly is typically high burnup and not limiting with respect to power peaking and thermal performance. Use of this type of center assembly from a previous vendor [

] core from the new fuel vendor). However, the core physics analysis will model this center assembly to ensure that the burnup limitations for each fuel assembly type are not exceeded and the peak integrated radial peaking factor for this assembly will be maintained at 0.95 or less of the core maximum integrated radial peaking factor at all times in core life to ensure that this assembly does not become limiting during cycle operation.

4.1.4 Gadolinia Burnable Absorber Limitations Gadolinia has been approved by the NRC for use as a burnable absorber for both Westinghouse and AREVA fuel. The impact of the use of gadolinia is discussed in other sections of this document (e.g. Section 4.10). In Reference 8.57, the NRC I

]. SCE fuel management guidelines [

I.

In Reference 8.57, AREVA also made several commitments to the NRC regarding gadolinia bearing fuel. These requirements are as follows:

(ii)

To ensure meeting the above requirements, SCE fuel management uses [

I.

Page 21 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.2 Core Thermal Hydraulic Analysis The approved SCE Reload Analysis Methodology (Reference 8.4) Section 3.2 describes the use of the TORC computer code for core thermal hydraulic analysis.

Implementing AREVA fuel into the SCE Thermal Hydraulic Analysis process [

The SCE formal software update process defined in Reference 8.4 Section 4.5.2 was applied to [

Page 22 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

[

I Figure 4.2

[

I Table 4.1 - Overall CHF Benchmark Statistics Comparison Standard Mean Computer Code Deviation Predicted/Measured

__________________r___________

Page 23 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel I

I Following the demonstrated [

], these methods were then demonstrated for the Unit 3 Cycle 17 VQP mixed core, and the Cycle 18 VQP full core to ensure that the process linked successfully with the rest of the SCE reload analysis methodology.

[d The SCE Core Thermal Hydraulic Analysis process [

] See Section 7.2 for results of this application.

Page 24 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.2.1 Thermal-Hydraulic Modified Statistical Combination of Uncertainties Verification The SONGS design Specified Acceptable Fuel Design DNBR limit (SAFDL DNBR) for SONGS is 1.31 (Technical Specification LCO 2.1.1.1). This SAFDL DNBR Limit was calculated based on the NRC-approved Modified Statistical Combination of Uncertainties (MSCU) methods from Reference 8.26.

]

The MSCU process considers two groups of uncertainties (Reference 8.26):

[

I The first group, [ ], includes the following uncertainties:

[

I The second group, [

I Page 25 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

[

] The remaining uncertainties are potentially impacted and are addressed below.

4.2.1.1 Engineering & Systematic Factors Uncertainties

[

]. The results are compared to the existing SONGS values in Table 4.2.

Table 4.2 MSCU Engineering Factor Uncertainties for AREVA HTP Fuel

[

.1 4.2.2 Thermal-Hydraulic Treatment of Rod Bow Penalty Another thermal-hydraulic model impact of the unrestricted use of AREVA fuel is the evaluation of the potential impact of rod bow on downstream calculations. The current rod bow methodology in SCE Reference 8.4 is based on the Reference 8.16 Westinghouse topical report. The NRC-approved AREVA rod bow methodology is given Page 26 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel in Reference 8.18. The AREVA rod bow methodology was approved for fuel rod burnups of 62 GWd/MTU in Reference 8.17 and Reference 8.14.

The SCE methodology provides for the rod bow penalty [

] AREVA calculated the Rod Bow Penalty for SONGS, and determined that the Rod Bow Penalty for the AREVA CE-HTP fuel design to be applied at SONGS is:

Table 4.2 SONGS Rod Bow Penalty for AREVA Fuel

[

I Rod bow penalty is [ .

4.3 Fuel Rod Behavior Analysis 4.3.1 Fuel Rod Behavior Analysis Introduction The current approved SCE Reload Analysis Methodology (Reference 8.4) Section 3.3 describes the use of the Westinghouse FATES3B computer code for fuel rod behavior analysis. Reload fuel rod behavior design and safety analyses are performed at SONGS with the FATES3B computer code. FATES3B is used for thermal performance evaluations under normal operation considering steady-state and anticipated transient conditions. Compliance with power-to-centerline melt and maximum rod internal pressure criteria are evaluated using FATES3B. Additionally, FATES3B is used to generate initial fuel/clad conditions for other design analyses, transient analyses, and accident analyses.

[

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LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel I

A summary of the FATES3B w/M 5 TM licensing applications and division of reload analysis scope is provided in Section 4.3.2. Section 4.3.3 presents the [

1, describes the verification and validation of the FATES3B w/M 5 TM code, and provides fuel rod behavior [

4.3.2 Summary of Reload Licensing Applications and Fuel Rod Behavior Scope Per Reference 8.34, currently SCE uses FATES3B to provide predictions of the steady-state response of fuel rods, and to model internal conditions of the fuel rods within the core from insertion to discharge. With the appropriate modeling of mechanical design data, power levels, and power distributions, these internal conditions serve as input for transient analyses (i.e., LOCA and non-LOCA) where initial conditions are required.

FATES3B results are also used to confirm compliance with design criteria and are input to reactor system setpoints analyses.

FATES3B is used to support the [

I SCE uses FATES3B to [

Although all of the same design and licensing calculations that are currently performed for a Westinghouse reload will need to be performed for AREVA fuel, the division of scope responsibility with AREVA will be different. In the general sense, AREVA will be performing all fuel mechanical design and LOCA analyses, including the fuel rod initial conditions for the AREVA LOCA analyses. SCE will be generating fuel rod behavior analysis data per Reference 8.25 to support the non-LOCA safety analyses and Page 28 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel calculations that support SCE setpoints analyses (per Reference8.4). Specifically, [

I AREVA uses their approved fuel behavior codes and methodology (see Section 5.1 and Enclosures 3 and 4) to perform all of the fuel mechanical design calculations and fuel rod initial conditions for input to their LOCA analysis for the SONGS units. [

4.3.3 M5TM Modeling in FATES3B Similar to the approach used for the [

] The conclusions of the V&V effort are addressed in Section 4.3.3.2, including comparison cases to measurements.

Page 29 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.3.3.1 M5TM Properties and Correlations The following sections describe the M5TM cladding thermal, mechanical, and irradiation material property models [

] No modifications have been made to the previously approved uranium, erbia, or gadolinia fuel properties and correlations.

4.3.3.1.1 M5TM Axial Growth The axial growth phenomenon, also called irradiation growth or axial clad growth, is when the fuel cladding length increases in the presence of neutron irradiation, even in the absence of applied loading. The nominal projection of M5TM cladding fuel rod growth is per the generic bounding empirical model that is a function of fluence as defined in the following equation (Reference 8.10, Section 6.1.7.1, page 6-6):

[

The use of a nominal projection of M5TM cladding fuel rod growth is consistent with the use of a nominal projection in the currently approved FATES3B methodology, because the local fast neutron flux is biased to the low side in licensing calculations (Reference 8.34, Response to Question A, page 3-5). Therefore, the local fast neutron flux will continue to drive the cladding growth in the FATES3B w/M5TM cladding model and the overall calculation of cladding growth will be conservative.

4.3.3.1.2 M5TM Creep Cladding creep is defined as the time dependent deformation of material under constant loading. Under the influence of the external coolant pressure, the cladding will creep down and make contact with the fuel. The M5TM cladding creep is per the following equations (Reference 8.10, Section 4.1.1, pages 4-1 through 4-3, as corrected by Reference 8.48).

The generalized creep rate (&c, h- ) is the sum of the thermal (&p,/ h-1 )and irradiation (l h' ) components:

Page 30 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel tc = i +tp The thermal and irradiation creep component rates are evaluated using the following equations:

I I

The values of the creep model constants used for M5TM cladding are:

[

I Page 31 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.3.3.1.3 M5TM Elastic Modulus The cladding elastically responds to the differences between the fuel rod cladding external and internal pressures. Young's modulus for cladding elastic deformation (in units of Pascal (Pa)) is per the following equation (Reference 8.10, Section 4.1.8, page 4-6):

[ I where T = cladding temperature (K) between [ I

[

Page 32 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.3.3.1.4 M5TM Poisson's Ratio Poisson's ratio is the ratio of transverse (lateral) contraction strain to longitudinal (axial) extension strain in the direction of the stretching force. Tensile deformation is considered positive and compressive deformation is considered negative. The M5TM Poisson's ratio is the following [ ] (Reference 8.10, Section 4.1.9, page 4-6):

POIS=[ ]

4.3.3.1.5 M5TM Thermal Conductivity The heat transfer from fuel pellet to the reactor coolant depends on the thermal conductivity of the cladding. The FATES3B algorithm assumes the thermal conductivity of the cladding to be a linear function of temperature. The M5TM thermal conductivity (W/m-K) is per the following quadratic equation (Reference 8.10, Section 4.1.3, page 4-4):

Due to the linear function of temperature requirement of the algorithm in FATES3B, the M5TM quadratic correlation for thermal conductivity was approximated over a limited temperature range by the following M5TM linear equation:

A plot of the M5TM quadratic and linear equations is shown in Figure 4.3.1 over the applicable temperature range. Over this temperature range, the difference between the quadratic and linear equations is no more than [ ]. This difference is acceptable given that the FATES3B code also models a simplified linear thermal conductivity equation for Zircaloy-4 and ZIRLO TM cladding.

Page 33 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.3.1 M5TM Thermal Conductivity Correlations

[

I Page 34 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.3.3.1.6 M5TM Thermal Expansion Coefficient The radial thermal expansion of the fuel cladding is evaluated at the arithmetic mean clad wall temperature, for each axial node. The average coefficient of radial thermal expansion used for steady state fuel thermal analysis assumes the cladding has not experienced an alpha to beta phase change at any prior time in its operating history.

The M5TM thermal expansion strain (in/in) in the axial direction is represented by the following equations that assume a reference temperature of 200C (Reference 8.10, Section 4.1.7, page 4-5):

[

]

The M5TM thermal expansion strain (in/in) in the tangential direction is represented by the following equations (Reference 8.10, Section 4.1.7, page 4-5):

[

The AREVA M5TM thermal expansion model is based on the low temperature range to I 7500C (Reference 8.10, Section 4.1.7, page 4-6). Therefore, only the equations corresponding to the temperature range of 200C to 7500C (i.e., 68 0 F to 1382°F) are programmed into the FATES3B w/M5TM code.

The M5TM thermal expansion strain (in/in) in the axial direction is converted to English units per the following equations:

[]

The M5TM thermal expansion strain (in/in) in the tangential direction is converted to English units per the following equations:

[

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LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel I

4.3.3.2 [ ] Verification and Validation I

I.

4.3.3.2.1 V&V Acceptance Criteria The verification and validation testing of [

]was performed utilizing the following three acceptance criteria:

Page 36 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

]. As discussed in the following paragraphs, the test cases include [

I Page 37 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.3.3.2.2 V&V Test Results and Outputs The following sub-sections present the results and outputs of the verification and validation testing. Each section contains a summary of the test case(s) used, testing performed, and the comparison of applicable output data to the established acceptance criteria.

V&V Criterion 1 - Zircaloy-4 and ZIRLO TM Computations Results Acceptance Criterion 1 specifies that the results of the calculations for [

]cladding materials.

] licensing cases.

Output from each of the Acceptance Criterion 1 test cases using [

]. Therefore, Acceptance Criterion 1 is satisfied.

V&V Criterion 2- M5TM Creep Behavior Results The cladding radial creep model is used to calculate the [

Page 38 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

] Therefore, Acceptance Criterion 2 is satisfied.

Page 39 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.3.2

[ ]Calculated Radial Creep versus Measured Radial Creep for the M5TM Test Rods

[

I V&V Criterion 3 - M5 TM Axial Growth Behavior Results During irradiation, the fuel rod grows axially due to a [

Page 40 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 3 Therefore, Acceptance-Criterion 3 is satisfied.

Page 41 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.3.3 Measured and Calculated Axial Growth as a Function of Fluence

[

I 4.3.3.3 [ .] Fuel Rod Behavior Comparisons This section provides fuel rod behavior comparisons of [

I Page 42 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.3.3.3.1 Fuel Temperature The fuel [

4.3.3.3.2 Power-to-Centerline Melt (PTM)

The power-to-centerline melt comparison of the [

I 4.3.3.3.3 Internal Hot Gas Pressure The rod internal pressure history comparison of the [

Page 43 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Page 44 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.3.4 Steady-State and Transient [

I Page 45 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.3.5 Axial Power Distributions

[

I Page 46 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.3.6 Fuel Average Temperature Comparison

[

I Page 47 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.3.7 Power-to-Centerline Melt (PTM) Comparison

[

I Page 48 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.3.8 Rod Internal Pressure Comparison

[

Page 49 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel I

4.3.3.4 M5 TM Cladding Modeling [ ] - Conclusions The successful Verification and Validation process described in the preceding sections demonstrates that SCE has successfully implemented the [

4.3.4 [ ] Fuel Behavior Methodology and Process Fuel rod behavior analyses are currently performed at SCE using the FATES3B computer code as described in various Fuel Evaluation Model Topical Reports (References 8.35, 8.36, 8.37, and 8.38). The methodology applied with FATES3B is described in additional Topical Reports (References 8.7, 8.24, 8.32, 8.34, 8.46, and 8.47) and [ ]clad fuel.

In licensing applications, results are conservative when predicted temperatures (or stored energy) and internal gas pressure are higher than what the actual operating fuel rod is expected to experience for conditions at the start of a transient or results conservatively represent conditions during a transient. This is accomplished by [

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LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

.]

The FATES3B User's Manual discusses [

.1 The NRC Staff concluded that the biased reduction in solidus temperature is adequately conservative. Topical Report CENPD-275-P, Supplement 1-P-A NRC Staff Safety Evaluation (Reference 8.47, Section 2) [

Page 51 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Page 52 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.4 Vendor Qualified Maximum Fuel Burnup Limits The proposed unrestricted use of AREVA M5TM fuel at SONGS Units 2 and 3 requires that appropriate burnup limits for the AREVA fuel be established and incorporated into the SCE Reload Analysis Methodology controlling documents, procedures, and analysis tools. Westinghouse and AREVA burnup limits and their applicability into the reload process will be discussed in this section.

4.4.1 Westinghouse Fuel Burnup Limits As described in Reference 8.56 as modified by Reference 8.7, the peak fuel rod burnup limit for SONGS CE 16x16 fuel ZIRLOTM cladding is 60,000 MWD/MTU. During the transition to AREVA fuel and for all Westinghouse fuel in a core cycle (including situations where the center assembly is a Westinghouse fuel assembly as described in Section 4.1.3), SCE will design fuel cycles to ensure that all Westinghouse fuel will not exceed the 60,000 MWD/MTU maximum pin burnup limit.

4.4.2 AREVA Fuel Burnup Limits On September 9, 1991, the NRC issued a Safety Evaluation Report (SER) (Reference 8.41) which concluded that Topical Report ANF-88-133(P)(A) (Reference 8.14) provides an acceptable basis for all ANF (now AREVA) PWR fuel rod design to a burnup of 62 GWd/MTU.

On May 5, 2004, the NRC issued a Safety Evaluation Report for the Framatome ANP topical report BAW-10240(P)-A (Reference 8.10) that concluded that the Framatome ANP (now AREVA) PWR design methods were acceptable with M5TM material properties for peak rod exposures of 62 MWd/kgU. This approval explicitly included application to Westinghouse and Combustion Engineering designed PWRs.

As such AREVA fuel is approved for CE PWR's for a maximum peak rod burnup of 62,000 MWD/MTU.

4.4.3 Application of AREVA Burnup Limits for M 5 TM fuel SONGS In order to determine the applicability of the AREVA fuel peak burnup limit for SONGS Units 2 and 3, SCE reviewed (1) the SER restrictions and limitations associated with the AREVA burnup topical reports described in Section 4.4.2 and (2) the NRC-approved licensing basis of the SCE physics and fuel rod behavior codes and methods related to burnup limits.

Page 53 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The SER restrictions and limitations associated with the AREVA burnup topicals are included as Attachment E. As shown in Attachment E, AREVA's fuel mechanical design and LOCA analyses for SONGS Unit 2 and 3 meet the restrictions and limitations for the maximum AREVA burnup limit of 62,000 MWD/MTU.

SCE uses the NRC-approved CASMO-4/SIMULATE-3 to perform physics calculations for SONGS Units 2 and 3 (Reference 8.5). Description of the use of CASMO-4/SIMULATE-3 for AREVA fuel is described in Section 4.1. The Reference 8.6 topical report SER has no limitation or restriction on the use of these physics codes for maximum peak burnups up to 62,000 MWD/MTU.

SCE reload methodology as described in Reference 8.4 uses the FATES3B code to model fuel rod behavior. The FATES3B topical report SERs have no limitation or restriction on the use of this code for maximum peak burnup up to 62,000 MWD/MTU.

In References 8.53 and 8.55, 16x16 Next Generation Fuel Core Reference Report and RAIs, Westinghouse reviewed the use of FATES3B to extended burnups of 62 MWd/KgU. As described in References 8.53 and 8.55 (SER in Reference 8.54),

FATES3B has the capability to accurately model U0 2 and treat the impacted burnable absorbers. Standard U0 2 was accurately modeled in Reference 8.53. The effects of gadolinia burnable absorbers are discussed in Reference 8.47. The addition of gadolinia will impact the thermal conductivity and melt temperature, but only incrementally from beginning of life and is not burnup dependent. No model or methods changes are required to extend the burnup limits of these fuels (i.e., U0 2 and gadolinia either modeled separately or in combination) to 62,000 MWD/MTU. Therefore, the models and methodology for the FATES3B are applicable to 62,000 MWD/MTU peak pin burnup and the fuel rod configurations containing U0 2 and gadolinia.

The NRC identified the potential for fuel performance codes to under predict the thermal conductivity of fuel as a function of burnup in the NRC Information Notice (IN) 2009-23 (Reference 8.52). Westinghouse evaluated the potential for FATES3B to under predict thermal conductivity as a function of burnup and concluded that [

I.

4.4.4 Fuel Burnup Limits Conclusion Based on a review of the NRC-approved topical reports for Westinghouse and AREVA fuel and the capability of the computer codes and methods SCE will use to model AREVA M5TM fuel, SCE proposes to use a maximum peak pin burnup of 60,000 MWD/MTU for Westinghouse fuel and 62,000 MWD/MTU for AREVA fuel for SONGS 2 and 3.

Page 54 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.5 Non-LOCA Transient Analysis The currently approved SCE Reload Analysis Methodology (Reference 8.4) Section 3.4 (as modified by Reference 8.7, Reference 8.21, and Reference 8.5) describes the non-LOCA Transient Analysis process, with subsections describing those events normally analyzed, those events not normally analyzed, and those events related to setting or verifying transient related CPCS and COLSS constants. All UFSAR chapter 15 events were reviewed, as discussed in Attachment C. Per Attachment C, the only changes required to account for use of AREVA fuel were to CEA Ejection (as discussed in Section 4.5.4), Departure from Nucleate Boiling (DNB) propagation (as discussed in Section 4.5.1, and use of statistical convolution to determine fuel failure (as discussed in Section 4.5.2). There were no modifications to the SCE reload design process required for evaluating Non-LOCA transients that were made to evaluate AREVA fuel.

4.5.1 DNB Propagation 4.5.1.1 Analysis of M5 TM Application DNB propagation could occur when individual fuel pins are in DNB (i.e., DNBR<DNBR SAFDL) and the internal fuel pin pressure is greater than the RCS pressure. Under these conditions, degraded heat removal from the fuel pin may cause the fuel pin to balloon, resulting in flow blockage. This flow blockage could then cause additional fuel pins to enter DNB conditions and fail.

The proposed DNB propagation methodology in this LAR submittal is for SONGS to retain the Reference 8.24 CE methodology, and continue to verify that non-LOCA design basis events do not exceed the clad strain limit. The following discussion demonstrates that the current Reference 8.24 CE DNB propagation methodology is applicable to AREVA fuel with M5TM clad.

Per Reference 8.24, the CE 14x14 fuel represents a limiting case for DNB propagation resulting from cladding strain. This CE 14x14 limiting case has become the bounding values of strain for fuel in CE designed NSSS. For all CE fuel using ZIRLOTM cladding DNB propagation does not occur, provided that cladding strain is [ ] (Reference 8.24). This limit also applies to SONGS 16x16 assemblies.

Page 55 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The approved M5 TM cladding AREVA topical (Reference 8.9) was reviewed for potential M 5 TM specific requirements that would impact SCE methodologies or applications. The topical states in Section 4.1,

"...for DNB related events there is no consequence of a switch from Zircaloy-4 to M5, other than an improved ability to control the fuel assembly performance....

For those accident evaluations that produce cladding temperature responses that exceed the phase transition range, approximately 7000 C, a small impact on temperature response is expected and a revised calculation with M5 specific materialproperties should be performed for batch licensing. The results of those calculationsare not expected to differ substantially from Zircaloy-4 based calculations and no limiting criteria are expected to be challenged.......

The analysis that confirmed the [ ] in Reference 8.24 was performed using the INTEG utility code. INTEG determines cladding strain rate based upon cladding stress, temperature and material properties. INTEG performs an integration of the cladding strain rate over time to yield an overall strain.

An analysis was performed to demonstrate that [

]. The NRC has concluded that "...the single-rod strain data collected in the EDGAR facility are in general more conservative than the single-rod data used in NUREG-0630 and, therefore, are acceptable for use in developing M5 TM cladding ballooning and flow blockage models (Reference 8.9). Tabular results are given in Tables 4.5.1 and 4.5.2. The results are presented graphically in Figures 4.5.1 and 4.5.2.

Page 56 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PON 600 Request for Unrestricted Use of AR EVA Fuel Table 4.5.1 Table 4.5.2 Page 57 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.5.1 Clad Rupture Time @ [ I

[

Page 58 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.5.2 Clad Rupture Temperature @ [ I

[

Page 59 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.5.1.2 Conclusion For Non-LOCA transients that experience DNB pin failure, [

.1 Section 7.4.1 presents the Pre-Trip Steam Line Break event for the VQP core design (Section 7.1) to illustrate the SCE DNB propagation methodology for AREVA fuel.

4.5.2 Statistical Convolution of Uncertainties - Fuel Failure Prediction The approved DNB statistical convolution methodology (Reference 8.23 as updated by References 8.20 and 8.21) estimates the number of failed fuel rods by applying the probabilistic definition of DNBR SAFDL (i.e., the SAFDL represents DNBR value, above which there is a 95-percent probability, with a 95-percent confidence level, that the fuel rod will not experience DNB). The probability of DNB decreases as DNBR increases (e.g., above the DNBR SAFDL), and increases as DNBR decreases (e.g., below the DNBR SAFDL).

The DNB statistical convolution groups fuel rods with respect to their radial peaking factors; calculates the minimum DNBR in each radial peaking factor group, and determines the probability of DNB as a function of DNBR value. For any given value of DNBR, the number of fuel rods, within a radial peaking factor group, that are predicted to experience DNB and fail is the product of the number of fuel rods in the radial peaking factor group and the probability of DNB. The total number of predicted fuel rod failures is the summation of each group's failed fuel rods. The DNB statistical convolution methodology relies upon a probability distribution function (i.e., the probability of DNB as a function of DNBR).

Reference 8.23 contains probability distribution functions for the Combustion Engineering (CE) 14x14 and 16x16 rod assemblies. These probability distribution functions were established using the [

I.

Section 3.2.2 of the approved SCE Reload Analysis Methodology (Reference 8.4) describes the original SCE methodology for applying statistical convolution to evaluate DNBR consequences.

Page 60 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The implementation of Alternative Source Term (AST) dose consequence analysis for SCE (Reference 8.22, submittal; References 8.4, 8.20 and 8.21, NRC approvals) included a specific NRC review of the SCE statistical convolution of uncertainties and its impact on DNBR consequences. The NRC reviewed the specific application of this process as it relates to the determination of predicted DNBR fuel failures for non-LOCA events. As a result, the SCE reload analysis methodology now incorporates the statistical convolution of DNBR consequences, as described in the CE methodology of Reference 8.23, for all fuel failure events involving loss-of-flow.

The key to the use of the DNB statistical convolution methodology is developing the probability distribution function (pdf) of exceeding DNB with respect to DNBR. In the NRC Safety Evaluation Report (SER) approving Reference 8.23, the NRC staff stated:

"Since experimental evidence indicates that fuel cladding failure is not necessarily coincident with a short duration of DNB, we conclude that the statistical convolution technique is conservative and acceptable provided that the probability distribution for DNB is acceptable."

Reference 8.23 contained probability distributions for the Combustion Engineering (CE) 14x14 and 16x16 rod matrixes. Those distributions were established using the [

]. The NRC SER for the Reference 8.23 methodology document established a clear link between the computer codes, critical heat flux (CHF) correlation, and the probability distributions and requires NRC staff approval for any combination other than that specifically approved in the SER.

In Reference 8.20 and 8.21 the NRC-approved SONGS request to expand the use of fuel failure estimates by DNB statistical convolution methodology to all USFAR Chapter 15 non-LOCA events with fuel failure that assume a loss of flow. This approval is subject to the following restriction (Reference 8.20 as modified by Reference 8.21):

The use of any combination of critical heat flux correlation, or fuel design, other than that explicitly approved by CENPD-1 83-A (Reference 8.23), will require submittal of revised probability distributions for NRC staff review and approval.

As shown in Section 4.2.1, [

.1 Section 7.4.1 presents the Pre-Trip Steam Line Break event for the VQP core design (Section 7.1) to illustrate the SCE statistical convolution methodology for AREVA fuel.

Page 61 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.5.3 Summary of Cladding-Related Models in the Non-LOCA Transient Evaluation Models As discussed in the M5TM Topical (Reference 8.9 Section 4.1) and the ZIRLOTM Topical (Reference 8.7, Section 7.2), [

1.

Cladding Thermal Conductivity As described in Section 4.5.4.1, for use in CEA ejection, [

Cladding Specific Heat As evaluated in Section 4.5.4.1, for use in CEA ejection, the M5TM specific heat is

[

] for specific heat.

4.5.3.1 CENTS Code The CENTS computer code (Reference 8.33) is an interactive, faster than real time computer code for the simulation of the NSSS and related systems. It is capable of calculating the behavior of a PWR for both normal and abnormal conditions, including accidents. CENTS is the code currently used for transient analyses, replacing the original CESEC-III code.

A review of CENTS indicated that the cladding material properties employed are cladding thermal conductivity and specific heat. As discussed in the preceding paragraphs, [

.J This approach is consistent with [

] (Reference 8.7, Section 7.2). Consequently, no changes to CENTS to accommodate M5TM are needed.

4.5.3.2 HERMITE Code HERMITE (Reference 8.51) is a space-time kinetics computer code. HERMITE was developed for the analysis of design and off-design transients in PWRs by means of a numerical solution to the multi-dimensional, few-group, time dependent neutron Page 62 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel diffusion equation including feedback effects of fuel temperature, coolant temperature, coolant density and control rod motion. The heat conduction equation in the fuel pellet, gap and clad is solved by a finite difference method. Continuity and energy conservation equations are solved for the coolant enthalpy and density.

A review of HERMITE indicated that the cladding material properties employed are cladding thermal conductivity and specific heat. As discussed in the preceding paragraphs, [

]. This approach is consistent with [

] (Reference 8.7, Section 7.2). Consequently, [

] are needed.

4.5.4 CEA Ejection Non-LOCA Event Methodology Review The SONGS Reload Analysis Methodology Topical Report SCE-9801-P-A, Section 3.4.2.1.4 (Reference 8.4) addresses the Control Element Assembly (CEA) ejection event methodology. Each aspect of the CEA Ejection event methodology is evaluated below for potential impact by the change to AREVA fuel supply. Compared to the Westinghouse fuel currently in SONGS units, there is one major difference in AREVA fuel that might impact the CEA ejection event. M5 TM is used as the material for AREVA fuel rod cladding, while ZIRLOTM is used as the fuel rod cladding material in Westinghouse fuel.

See Section 7.4.2 for a summary comparison between the CEA Ejection event analysis for the VQP fuel cycle with Westinghouse and AREVA fuel.

4.5.4.1 Analysis for the Deposited Energy (Fuel Pin Enthalpy)

Acceptance Criteria The deposited energy acceptance criteria evaluation for this event is performed using the STRIKIN-II code (Reference 8.30) to determine the energy deposited in the fuel rods by an ejected CEA at various plant conditions and power levels. The analysis uses the basic methodology described in CENPD-190-A (Reference 8.31).

Page 63 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The Reference 8.31 methodology has been supplemented as required by CENPD-404-P-A Sections 7.2.1 and 7.3.1 (Reference 8.7) for the use of ZIRLOTM clad fuel rods.

The implementation of [ .]

To implement the AREVA M5TM cladding material, the material properties of the M5TM alloy were reviewed in order to determine the impact to the CEA Ejection analysis. The thermo-physical, mechanical, and corrosion properties of MSTM are discussed in the M5TM topical report (Reference 8.9). Consistent with the [

]are evaluated in the following paragraphs.

The thermal conductivity of the M5TM alloy was discussed in Appendix A of the M5TM topical report (Reference 8.9). In Section A.2.2.2 of Reference 8.9, the Zircaloy-4 thermal conductivity data used in RELAP-5/MOD2-B&W, together with the test data for pure zirconium, Zircaloy-2, Zircaloy-4, and M5TM alloy were compared. Section A.2.2.2 of Reference 8.9 concluded:

"there is a very good agreement between the thermal conductivity determinations regardless of alloy composition. Therefore, the RELAP5 (Zircaloy-4) data is considered adequate to represent the thermal conductivity of the M5TM alloy."

The heat capacity of the M5TM alloy was discussed in Section A.2.2.3 of the M5TM topical report (Reference 8.9). A comparison of specific heat for M5TM and Zircaloy-4 was shown in Figure 1-5-3 of Reference 8.9. [

.]

Section 1.5 and Figure 1-5-3 of Reference 8.9 showed that M5TM [

] For Zircaloy-4, Section A.2.2.3 of Reference 8.9 stated that the [

.] Based on these data, for transients in which the cladding temperature remains [

Page 64 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel However, for transients in which the cladding temperature may enter the phase change range the impact of M5TM must be considered due to the differences in heat capacities.

The M5TM topical report (Reference 8.9) identified CEA Ejection as the accident likely to result in cladding temperature that enters the phase change range of the M5TM material.

The heat capacities of M5TM are modeled in the CEA Ejection analysis [

] The specific derivation of the M5TM alloy's specific heat thermo-mechanical properties are provided below.

Section 1.5 of Reference 8.9 provides the following correlation for M5TM specific heat based on testing data:

[3 Where Cp = specific heat in J/g-K, and T= temperature in K.

Per Section 1.5 of Reference 8.9, the density of the M5TM alloy has been determined as I I The cladding specific heat data used in STRIKIN is volumetric specific heat (Btu/ft 3-OF) versus temperature (OF). The M5TM specific heat inputs for the STRIKIN code are developed based on the correlation and the M5TM alloy density above. Note that the M5TM volumetric specific heat data is generated using a [

] the STRIKIN topical report (Reference 8.30, Appendix I). The volumetric specific heat data of the [

]are compared in Table 4.5.3 and Figure 4.5.3.

Page 65 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 4.5.3 Comparison of the Specific Heat Data for [ ]

[

+ + +

• 1 +

I Page 66 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.5.3 Cladding Specific Heat Inputs to STRIKIN for [ I

[

I 4.5.4.2 Conclusion The STRIKIN code in CEA ejection will use [

I Section 7.4.2 presents the CEA Ejection event for the VOP core design (Section 7.1) to illustrate the SCE CEA Ejection methodology for AREVA fuel.

4.5.5 Non-LOCA Transient Methodology Review Conclusions The SCE Non-LOCA transient analysis methodology topical report (Reference 8.4) was reviewed for potential impacts resulting from the unrestricted use of AREVA fuel. No methodology or process changes, relative to the SCE Non-LOCA transient analysis methodology topical report (Reference 8.4), are required to implement AREVA fuel.

However, non-LOCA transient events will model [

] properties in all Non-LOCA events except CEA ejection.

Page 67 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The STRIKIN code in CEA ejection will use [

'I 4.6 LOCA Analysis - Interface with Fuel Vendor For SONGS mixed cores, both fuel vendors will continue to verify their respective large break LOCA and small break LOCA analyses under contract to SCE. For full cores both large and small break LOCA analyses will be verified by the full core fuel vendor.

The current Westinghouse post-LOCA Long Term Core Cooling analyses will remain applicable for both AREVA and Westinghouse fuel. The fuel vendor will need data and computer files to perform these analyses. In addition to the data presented in the Reload Ground Rules (Section 4.3 of Reference 8.4), physics and fuel behavior data (discussed in Sections 3.1.1 and 3.3.2, respectively, of Reference 8.4) are calculated and formally transmitted to the fuel vendor. The interface between SCE and the fuel vendor with respect to reload specific data and computer files is achieved via formal transmittals. The fuel vendor provides a "data request" that specifies the data and computer files needed to perform the LOCA analyses. After completing the analyses described in Sections 3.1.1 and 3.3.2, respectively, of Reference 8.4, SCE prepares a "data transmittal" to the fuel vendor detailing the cycle specific data and computer files.

The final results of the fuel vendor's LOCA analysis are formally sent to SCE for incorporation into the Reload Analysis Report (RAR), which is implemented through the Nuclear Engineering Change Package (NECP) process. This process is described in Section 4.9.1 of Reference 8.4 for the interface with the current fuel vendor, ABB CE (now Westinghouse). The process continues to be applicable to both fuel vendors under consideration, with the only significant changes being the content and format of the data requests described above.

These minor variations in the relationships between SCE and the current or the alternate fuel vendor do not represent a significant change in the LOCA analysis interface described in the Reference 8.4 SCE Reload Analysis Methodology.

4.7 Fuel Mechanical Design Analysis - Interface with Fuel Vendor The fuel vendor will continue to perform all fuel assembly mechanical design analyses under contract to SCE. As is the case with the LOCA analyses, the fuel vendor requests certain physics data needed to verify the fuel assembly mechanical design for the new reactor core design. The fuel vendor provides a "data request" that specifies the data needed to verify the fuel assembly mechanical design. After completing the analyses described in Section 3.1.6, SCE prepares a "data transmittal" to the fuel vendor detailing the cycle specific data. The final results of the fuel vendor's mechanical design analyses are formally sent to SCE for incorporation into the Reload Analysis Report (RAR). This process is described in Section 4.9.2 of Reference 8.4 for Page 68 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel the interface with the current fuel vendor, ABB/CE (now Westinghouse). The process continues to be applicable to both fuel vendors under consideration, with the only significant changes being the content and format of the data requests described above.

Any minor variations in the relationships between SCE and the current or the alternate fuel vendor do not represent a significant change in the fuel mechanical design analysis interface described in the Reference 8.4, SCE Reload Analysis Methodology.

4.8 COLSS/CPCS Setpoints Analysis 4.8.1 SCE Reload Topical: Setpoint Methodology As discussed above, the SCE reload analysis methodology is described in the SCE Reload Topical, Reference 8.4. The Setpoints analyses in Reference 8.4 are described under the "Core Protection Calculator System (CPCS) Analysis," and "Core Operating Limiting Supervisory System (COLSS) Analysis" sections.

The CPCS is a set of digital computers and associated software which initiates the DNBR and Local Power Density (LPD) reactor trip signals. COLSS is a digital computer program in the Plant Monitoring System (PMS) which assists the plant operator in maintaining Technical Specification LCOs.

In the reload process, the COLSS and CPCS analyses receive input from upstream reload disciplines such as Physics, Thermal Hydraulics (TH), etc. An assessment of the impact of AREVA fuel on the COL-SS and CPCS input is a direct measure of the impact of AREVA fuel on the SCE Setpoints methods.

4.8.1.1 Physics Design Analyses Input to Setpoint Analyses The physics analyses at SCE explicitly model the fuel in the fuel management analyses and the generated input for the Setpoint analyses (e.g. neutronics model, physics related data base constants, and physics related reload data block (RDB), etc.

constants). In this manner, the impact of AREVA fuel on the physics related input to Setpoints analyses becomes transparent with regard to Setpoints analyses. Therefore, there is no change to the Setpoints methodology needed to account for the AREVA impact on the physics input to Setpoints analyses.

4.8.1.2 Core Thermal-Hydraulics Design Analyses Input to Setpoint Analyses As previously discussed in Section 4.2, the impact of the [

.] The CHF correlation [

Page 69 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

] A graphical representation of the relationship between the Thermal-Hydraulic Analysis process, the COLSS/CPCS Setpoints Analysis process and non-LOCA transients is shown in Figure 4.8.1. This existing process will continue.

Therefore, there is no change to the Setpoints methodology needed to account for the AREVA impact on the TH related input to Setpoints analyses.

4.8.1.3 Fuel Rod Behavior Analysis Input to CPCS Design The fuel behavior design provides the inputs which are used in the [

The fuel rod behavior analysis at SCE [

] In this manner, the impact of AREVA fuel on the physics related input to Setpoints analyses becomes transparent with regard to Setpoints analyses.

Therefore, there is no change to the Setpoints methodology needed to account for the AREVA impact on the fuel behavior analyses input to Setpoints analyses.

4.8.2 COLSS and CPCS Data Base Constants Verification Analyses As discussed earlier the impact of AREVA fuel on the COLSS and CPCS databases generated based on the upstream analyses (e.g., physics, safety design, etc.) is

[

.] Therefore, there is no change to the Setpoints methodology needed to account for the AREVA impact on the COLSS and CPCS database constant verification analyses.

4.8.3 COLSS and CPCS Addressable Constants Analyses As discussed above, the upstream input to Addressable Constants analyses are unchanged. Furthermore, the application of the Modified Statistical Combination Uncertainty (MSCU) topical, Reference 8.26 is unchanged. Therefore, there is no change to the Setpoints methodology needed to account for the AREVA impact on the Addressable Constants analyses.

4.8.3.1 Safety Design Analyses Input to CPCS Design The safety design analyses input into the Setpoints analyses are unchanged since the physics related changes are explicitly modeled in the physics input which is fed into the safety design analyses and the [

Page 70 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

.] In this manner, the impact of AREVA fuel on the safety design analyses related input to Setpoint analyses becomes transparent with regard to Setpoint analyses. Therefore, there is no change to the Setpoints methodology needed to account for the AREVA impact on the safety design analyses input to Setpoints analyses.

Page 71 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 4.8.1 Relationship Between T-H, COLSS/CPCS Setpoints and non-LOCA Transients I

Page 72 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.8.4 Related Industry Experience Reference 8.39 is the topical report for "ABB Critical Heat Flux Correlations for PWR Fuel." Reference 8.39 assessed the impact of the CHF correlations (NV and TV CHF correlations) on the generic COLSS/CPC Setpoints methodology. Reference 8.39 concluded that the Setpoints topical described in Reference 8.40 remains valid with the application of the new CHF correlation. It should be noted that the SCE approach with respect to application of the new AREVA BHTP CHF correlation is consistent with the approach described in Section 7.1.2 of the Reference 8.39 [

4.9 AST Dose Consequence The proposed fuel type change will have no effect on the current radiological consequence design basis accident (DBA) methodology. The current SCE reload analysis process generates bounding full core and average fuel rod source terms that are used in the radiological consequence DBA analyses with failed fuel. The isotopic distribution is calculated using the SAS2H/ORIGEN computer code. This bounding analysis was based on Westinghouse fuel and Alternative Source Term (AST)

Regulatory Guide 1.183 guidance. In the current reload analysis process, the bounding source terms and current radiological dose analyses are [

]

Regardless, the methodology to calculate and the process to verify the bounding source terms will remain unchanged during and after the transition to AREVA fuel. [

.1 Page 73 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 4.10 Fuel Centerline Melting Temperature 4.10.1 Proposed Change To utilize gadolinia burnable absorber fuel rods, SONGS Units 2 and 3 Technical Specification (Reactor Core Safety Limit) 2.1.1.2 and its bases are proposed to be modified to include Westinghouse temperature adjustments for gadolinia application to all fuel types resident in SONGS cores as described in References 8.46 and 8.47.

4.10.2 Generic Approval Topical Report CENPD-275-P (References 8.46 and 8.47) documents the NRC Staff review and approval of the Combustion Engineering (CE) analysis methods for gadolinium bearing uranium fuel. Section 2.2 in each of these Topical Reports explicitly discusses the treatment of gadolinia in the FATES3B computer code. Sections 2.2 and 2.2.2 explicitly note that "adverse" values for gadolinia melting temperatures are modeled for use in licensing analyses to conservatively account for uncertainties in these properties.

4.10.3 Previous Approvals for SONGS Although the two gadolinia Topical Reports are not currently addressed in Technical Specification 2.1.1.2, per the following documents SCE is currently licensed to use gadolinia as a burnable poison:

1) Technical Specification 4.2.1 states that integral or discrete burnable absorber rods may be used, including gadolinium oxide and erbium oxide.

2) SCE Reload Topical (Reference 8.4) addresses the use of gadolinia as a burnable absorber through its citation of CENPD-275-P-A.

3) The SCE CASMO4 and SIMULATE-3 Topical (Reference 8.5) addresses the modeling of gadolinia as a burnable absorber in the SCE reload design process.

4.10.4 Technical Basis

1) [.

Page 74 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

]

2) Basis for Approval is Not Dependent on the Manufacturer of the Fuel A review of the Westinghouse (References 8.46 and 8.47) and AREVA gadolinia topical reports (References 8.44 and 8.57) show that the sources of data are predominately laboratory-based. The use of Journal papers as references demonstrates the use of publicly available data that is not dependent of properties of commercially manufactured reactor fuel. Westinghouse and AREVA have manufactured fuel at a number of different factories including Windsor, Hematite, Columbia, Richland and Lynchburg. Fuel properties used in the analysis methods are not dependent on fuel manufacture locations. Therefore, fuel properties are not dependent on the manufacturer.

3)

I

4) Methodology is Independent of Fuel Rod Cladding The models, limits and application processes used for fuel pellets versus cladding with FATES3B w/M5TM are [

i 4.10.5 Conclusion The selection of the [

.1 Page 75 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 5.0 AREVA SCOPE OF RELOAD ANALYSIS 5.1 Fuel Mechanical Design Analysis In the SER accompanying the approved AREVA mechanical design topical report (Reference 8.17), the NRC stated:

"For each application of the mechanical design criteria, SPC (AREVA) must document the design evaluation process demonstrating conformance to these criteria and submit a summary of the evaluation to the NRC staff for possible use in an audit to confirm that SPC is in compliance with these design criteria."

There were no modifications made to the SCE scope of the SCE reload design process for evaluating fuel mechanical design analyses to evaluate the AREVA fuel for the Vendor Qualification Program fuel cycle.

The AREVA (formerly, Siemens Power Corporation, SPC) response to the above SER requirement for the scope of reload design analysis to be performed by AREVA is found in Section 5.1.4.

5.1.1 SONGS Mechanical Design Introduction SONGS plans unrestricted use of AREVA fuel starting with Unit 3, Cycle 17. The AREVA TM fuel design Tigm will be the CE 16x16 lattice HTP fuel assembly design consisting of M5 fuel rods, M5 corner and center guide tubes, Alloy 718 High Mechanical Performance (HMP) spacer at the lowermost axial elevation, M5TM HTP spacers in all other axial elevations, FUELGUARD lower tie plate (LTP), and the AREVA NP reconstitutable upper tie plate (UTP).

The AREVA fuel design is similar to the lead fuel assemblies that were introduced at San Onofre Unit 2 in Cycle 16 (Reference 8.3). The AREVA fuel assembly design offers two improvements relative to the lead fuel assembly design in order to improve seismic margin -

" Increased outer diameter (and wall thickness) of the corner guide tubes

" Increased thickness of the HTP spacer grid side plates The fuel assembly evaluations performed for the LFA designs are either applicable or more restrictive than for the reload design. The fuel rod performance for the LFA designs is representative of the reload design. The AREVA methodology requires the fuel performance to be evaluated for each cycle. Therefore, when the cycle design is established for the reload fuel for SONGS Unit 3 Cycle 17, the fuel performance will be explicitly demonstrated to satisfy the design criteria.

Page 76 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Section 5.1.2 of this report provides a detailed discussion of the design features of the AREVA fuel assembly. Section 5.1.3 provides the basis for verifying the mechanical compatibility of the AREVA reload fuel design to the SONGS reactor internals, control components, and handling and storage equipment. Section 5.1.4 of this report further performs a review of the NRC-approved mechanical design criteria that were utilized to license the lead fuel assemblies and which will also be used to license the CE16x16 lattice HTP batch fuel. The LFA results for the fuel assembly and structural performance are either applicable or more restrictive than for the reload fuel. Because the fuel rod design is unchanged, the only differences in the rod performance are due to the cycle to cycle variations in the power history. Therefore, the rod performance evaluations are representative of the reload fuel. The reload fuel will be explicitly evaluated when the cycle design is established.

5.1.2 SONGS Mechanical Design Details The AREVA fuel assembly for the San Onofre Unit 3 reactor is of a Combustion Engineering (CE) 16x16 lattice design. CE 16x16 lattice fuel designs contain 236 fuel rods, 4 corner guide tubes and 1 center guide tube. The corner and center guide tubes each occupy 4 fuel rod positions. The fuel rods are positioned within the fuel assembly by 11 spacer grids that are attached to the guide tubes.

The fuel bundle assembly design incorporates several proven design features to enhance performance. Figure 5-1 is a schematic of the AREVA fuel assembly for San Onofre. The fuel rod design in this assembly uses M5TM cladding and end caps. The M5TM material has very low corrosion and hydrogen pickup rates; providing substantial margin for end of life corrosion and hydrogen content. This material was developed in Europe and has been used extensively both in Europe and the United States for fuel rod cladding. The material has been generically reviewed and accepted by the United States Nuclear Regulatory Commission (USNRC) for use in CE fuel designs (Reference 8.10). Reloads with M5TM cladding have been provided in the United States since 2000.

Performance has been demonstrated to rod exposures in excess of 80 MWd/kgU. The fuel rod design includes uranium dioxide fuel rods and gadolinia bearing uranium dioxide fuel rods. Also, multiple uranium-235 enrichments are used within an assembly.

The lower tie plate design is a FUELGUARD structure. This structure uses curved vanes to provide non-line-of-sight flow paths for the incoming coolant to protect the fuel assembly from debris that may be present. This design is very efficient at preventing debris, including small pieces of wire, from reaching the fuel. The design uses the same vane configuration and spacing that has been used on CE 14x14, CE 15x15, Westinghouse 14x14, Westinghouse 15xl 5, Westinghouse 17xl 7, and Babcock &

Wilcox (B&W) 15x15 designs in the United States. This FUELGUARD design has been used in reloads in the United States since 1991. A schematic of the CE 16x16 FUELGUARD lower tie plate is provided in Figure 5-2.

Page 77 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The upper tie plate (UTP) design is the standard AREVA NP CE reconstitutable design.

The basic configuration is the same as that used for CE 14x14 plants supplied by AREVA NP, but modified to match the height of the co-resident fuel and to be compatible with the San Onofre core plate configuration. Figure 5-3 shows the San Onofre UTP configuration. This reconstitutable design uses the corner locking nuts to engage with the upper sleeves on the guide tube. The design allows the reaction plate to be depressed to a setting well beyond the end of life deflections, and the corner nuts rotated to disengage the upper tie plate from the locking nuts. The upper tie plate can then be removed. This design does not create any loose or disposable parts during the reconstitution. The reconstitution capabilities of the AREVA CE 16x16 fuel assemblies have already been successfully demonstrated in poolside examinations of CE16 lead fuel assemblies (LFA).

The cage or skeleton uses 4 M5TM corner guide tubes, 1 M5TM center guide tube, 10 M5TM HTP spacers, and 1 Alloy 718 High Mechanical Performance (HMP) spacer at the lowest spacer position. Figure 5-4 shows the cage configuration. The HTP spacers are welded directly to the five guide tubes; the HMP spacer is attached to the corner guide tubes by mechanically capturing the HMP between rings that are welded to the guide tubes. Because the guide tubes are of a zirconium alloy, they cannot be directly welded to the Alloy 718 material used in the HMP. The HTP spacer design was developed in the late 1980's and has been used in CE 14x14, CE 15x15, Westinghouse 14x14, Westinghouse 15x15, Westinghouse 17x17, and B&W 15x15 reloads in the United States. The initial reloads were in 1991. In addition to this reload experience, the HTP spacer design has been used in CE1 6x1 6 lead fuel assemblies at Palo Verde and San Onofre. The design provides 8 line contacts as the interface between the fuel rod and the spacer grid, and is therefore very resistant to fuel rod failures from flow induced vibration fretting.

The HTP design provides the line contact for the rods, but also is configured to improve heat transfer. As seen in Figure 5-5, the spring structure provides a flow path. This flow path is at an angle relative to the rod longitudinal direction, causing the water to swirl around the rod without creating a large pressure drop across the spacer. The HMP has the same line contact configuration but the channel is not angled. Because this spacer is at the lowermost position, the improved heat transfer is not necessary. As stated previously, the HMP material is Alloy 718. This material is very stable in irradiation environments, and provides additional assurance that the rod/spacer contact will be maintained throughout the design lifetime.

The assembly uses a MONOBLOC TM guide tube design for the corner guide tubes and a constant outer diameter and wall thickness design for the center guide tube. The MONOBLOCTM design maintains the same inner diameters in the dashpot and non-dashpot regions as the co-resident San Onofre fuel, but has a constant outer diameter the full length of the tube. Therefore, the wall thickness in the dashpot region (about the bottom 12 inches of the guide tube) is increased. The San Onofre co-resident fuel Page 78 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel maintains the same wall thickness instead of maintaining the same outer diameter as the MONOBLOCTM design. Therefore, the co-resident fuel has the same inner diameters in both the dashpot and non-dashpot region. The AREVA design has a larger outer diameter at all elevations. The MONOBLOCTM guide tube design has been used for fuel reload batches in Europe and in lead assemblies in the United States.

Southern California Edison (SCE) inserted lead assemblies in San Onofre Unit 2 in Cycle 16. The reload assemblies planned for San Onofre Unit 3 have 2 component changes to improve seismic margin. These component changes are:

" Increase in the OD of the MONOBLOCTM corner guide tubes from 0.980 inch to 1.023 inch.

This is the same OD as used for another CE16 lead assembly program. This increase in OD will increase the wall thickness of both the dashpot and non-dashpot regions by 0.0215 inch. The HMP and HTP spacers designed for the larger guide tube diameter will be used to accommodate the larger diameter of the guide tube. The rod/spacer grid structure/configuration is the same as the SONGS LFA spacer grids. The increased wall thickness increases the bundle stiffness thereby reducing the grid loading during a seismic event. The increased wall thickness also increases the stress margins for the guide tubes. The increase in the corner guide tube wall thickness also requires that the attachment to the guide tube upper sleeve and the guide tube lower end plug be modified.

The upper sleeve attachment to the upper tie plate and the lower end plug attachment to the lower tie plate are unchanged.

• Increase in the thickness of the HTP side plates from [

The rod/spacer grid interface structure is unchanged from the SONGS LFAs.

The thicker sideplate increases the crush strength of the spacer grid, thus improving the margin to plastic deformation.

The cladding, pellets, end caps, upper tie plate, lower tie plate, center guide tube are all unchanged from the SONGS LFAs.

Page 79 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-1 CE 16x16 Fuel Assembly for San Onofre Figure 5-2 CE 16x16 FUELGUARD Lower Tie Plate Page 80 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-3 CE 16x16 Reconstitutable Upper Tie Plate Page 81 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-4 CE 16x16 Cage Assembly Figure 5-5 CE16x16 HTP Spacer Page 82 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 5.1.3 Mechanical Compatibility AREVA NP and SCE have an on-going lead assembly program at San Onofre using AREVA fuel. Prior to insertion, the lead assemblies were shown to be compatible with the host reactor core internals, handling equipment, and storage racks as well as the co-resident fuel. The lead assembly operating experience has confirmed the results of the AREVA NP compatibility evaluations. The reload of fuel to be used at San Onofre Unit 3 has the same interface configurations as the lead assemblies (the upper and lower tie plate and guide tube ID dimensions are unchanged) and will continue to be mechanically compatible with the host reactor core internals, handling equipment, storage racks and co-resident fuel. A comparison of the mechanical design parameters of the reload assembly to the lead fuel assembly (LFA) is presented in Table 5-1. A summary of the lead assembly program mechanical compatibility evaluations is provided below.

5.1.3.1 Fuel Assembly For the LFA design, the fuel assembly overall length was confirmed to be compatible with the dimensions of the core internals (spacing between core support plate and fuel alignment plate) at beginning of life cold and hot conditions. Additionally, positive engagement of the center/locking nuts and fuel alignment plate was demonstrated. An axial growth analysis confirmed adequate assembly to core internals and differential fuel rod/fuel assembly growth margins up to the licensed fuel rod limit of 62 MWd/kgU.

The fuel assembly and fuel rod overall lengths are unchanged from the LFA design.

The array type, the number of fuel rods and guide tubes, and the fuel rod pitch dimensions are the same as the LFA design and the co-resident fuel.

The square and diagonal widths of the fuel assembly at the upper and lower tie plates and the spacer grids were confirmed to be compatible with the core internals, storage racks, fuel elevator, and co-resident fuel. The square envelope between the LFAs and the reload design is the same. There is a [ ] increase in the diagonal dimension of the reload fuel, relative to the LFA design; however, the diagonal dimension of the reload design is less than that of the co-resident fuel, and therefore within the plant operating experience. The axial elevations of the spacer grids for the LFAs were confirmed to have adequate overlap with the co-resident fuel. Except for the bottom spacer (the HMP), the LFA spacer elevations are unchanged for the reload design. The bottom spacer elevation has been changed slightly (about 0.250 inch) for manufacturing reasons. This revised elevation is within the co-resident design overlap and compatible with the LFA design.

These evaluations confirmed that the lead assemblies were compatible with the reactor components and co-resident fuel in the core. Additional evaluations of individual fuel assembly components were also performed.

Page 83 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 5.1.3.2 Upper Tie Plate The holes in the fuel alignment plate in the reactor core mate with the fuel assembly upper tie plate posts (both center and corner positions). There are three basic alignment plate patterns - locations with no control rods and no instrumentation, instrumentation locations, and control rod locations. In the three cases, the holes form a square array, which matches the locking nuts layout. The diameter of the corner posts and the center post (for instrumentation) are established to allow sufficient clearance with the fuel alignment plate holes. The length of the locking nuts are also set to allow engagement of the assembly and the fuel alignment plate at Beginning of Life (BOL) hot conditions and provide adequate clearance at end of life. The core plate pattern is the same throughout the core except for the outer four fuel assemblies on each side of the reactor core. These 16 fuel assemblies, 4 assemblies per face, are offset 1/2 bundle pitch.

The underside of the fuel alignment plate exhibits some protrusions in the form of socket head screws. The compatibility between the fuel alignment plate and the upper reaction plate of the fuel assembly was demonstrated by showing that the lock-bars and screws do not prevent the seating of the reaction plate against the fuel alignment plate.

These evaluations include the pattern change for the offset peripheral fuel assemblies.

The upper tie plate was also evaluated with respect to compatibility with the fuel grapples for fuel movement. Three types of grapples were evaluated: the spent fuel grapple, refueling machine grapple, and the new fuel grapple. It was shown that the grapples can fit over the center hole in the reaction plate, that the reaction plate arms fit within the grapples, and the reaction plate will not interfere with any part of the grapples.

Review of the grapple designs did not show any protrusions or unusual geometry that must be accommodated by the reaction plate. A prototype upper tie plate was tested at the reactor site prior to the LFAs.

These evaluations were demonstrated to be valid by the insertion of the LFAs and remain applicable for the reload design since the upper tie plate design is unchanged.

5.1.3.3 Lower Tie Plate The core plate and lower support assembly within the reactor vessel provide four alignment pins per assembly forming a square array. The mating holes in the fuel assembly lower tie plate are also on the same square array. The diameter dimensions of the mating holes are sized to provide adequate clearance with the alignment pins.

The lower support plate does not exhibit any protrusions within the confine of the core shroud other than the alignment pins.

Since the same lower tie plate design is maintained for the reload design, the lower tie plate compatibility is assured.

Page 84 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 5.1.3.4 Guide Tubes The radial locations of the guide tubes laterally within the assembly, the inner diameters of the guide tubes, and the weep hole diameters were chosen to be the same as the co-resident fuel. The axial locations of the guide tube dash pot and weep holes are also the same size as the co-resident fuel and at similar elevations (there are slight differences due to the differences in the spacer grid attachments). These critical dimensions assure that control element assembly drop times and guide tube cooling are not affected by the introduction of the AREVA assemblies.

The ID of the corner and center guide tubes is unchanged from the LFA design. The wall thickness is increased, but does not affect the interface with the control element.

Therefore, the compatibility evaluations performed for the lead assemblies remain applicable.

Table 5-1 Comparison of AREVA San Onofre LFA and Reload Fuel' Assembly Nominal Mechanical Design Features LFA Reload 1 Fuel Assembly Overall Length, 176.6 176.6 inch Fuel Rod Overall Length, inch 161.87 161.87 Fuel Rod Pitch, inch 0.506 0.506 Number of Fuel Rods/Assembly 236 236 Number of Corner Guide 4 4 Tubes/Assembly Number of Center Guide Tubes (Instrumentation 1 1 Tubes)/Assembly Fuel Rod Cladding Material M51M M5TM Fuel Rod Cladding Outer 0.382 0.382 Diameter (OD), inch Fuel Rod Cladding Thickness, 0.025 0.025 inch Fuel Pellet Diameter, inch 0.3255 0.3255 Fuel Stack Height (BOL, cold,), 150.00 150.00 inch Corner Guide Tube Material M5IM M5IM M

T M T

Corner Guide Tube Type MONOBLOC MONOBLOC Corner Guide Tube OD (upper), 0.980 1.023 inch Corner Guide Tube Wall 0.040 0.0615 Thickness (upper), inch 0.040 _0.0615 Page 85 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel LFA Reload1 Corner Guide Tube OD 0980 1.023 (Dashpot), inch Corner Guide Tube Wall Thickness (Dashpot), inch Center Guide Tube Material M5IM M51M Constant OD and wall Constant OD and wall Center Guide Tube Type thickness thickness Center Guide Tube OD (upper 0.980 0.980 and lower), inch Center Guide Tube Wall Thickness (upper and lower), 0.040 0.040 inch Number HTP Grids and Material 10, M5 10, M5

[ [ I [ ]

Number HMP Grids and Material 1, Alloy 718 1, Alloy 718 1 The detailed features of the reload fuel assembly design may change slightly. Each change will be explicitly reviewed for continued compliance with the design evaluations.

Overall, the mechanical compatibility evaluations performed for the LFA program, and the evaluations of the changes planned for the reload AREVA fuel have confirmed that the AREVA fuel assemblies are compatible with the SONGS reactor components and the co-resident fuel in the SONGS core.

5.1.4 Conformance with Mechanical Design Criteria The AREVA fuel design planned for introduction in a reload at San Onofre Unit 3 Cycle 17 is similar to the AREVA NP lead fuel assemblies that were introduced at San Onofre Unit 2 in Cycle 16 (Reference 8.3). The lead fuel assemblies were analyzed in accordance with the NRC-approved generic mechanical design criteria contained in EMF-92-116(P)(A) (Reference 8.17) in conjunction with NRC-approved topical report BAW-10240(P)(A) (Reference 8.10). This topical report incorporates the M5TM cladding material properties that were previously approved by the NRC in BAW-10227(P)(A)

(Reference 8.9) into the Reference 8.17 methodology. All the mechanical design criteria were shown to be met up to the licensed fuel rod burnup limit of 62 MWd/kgU in ANP-2839(P) (Reference 8.3). The design improvements that are mentioned in Section 5.1.2 relative to the lead fuel assembly design do not adversely influence the fuel assembly structural characteristics that were determined by prior mechanical testing of the lead assemblies. Therefore, the AREVA fuel design, with expected structural Page 86 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel behavior and projected performance, will meet design requirements throughout the life of the fuel.

The NRC-approved generic design criteria used to assess the performance of the lead fuel assemblies were developed to satisfy certain objectives (Reference 8.17). These objectives are used for designing fuel assemblies so as to provide the following assurances -

The fuel assembly (system) shall not fail as a result of normal operation and anticipated operational occurrences. The fuel assembly (system) dimensions shall be designed to remain within operational tolerances and the functional capabilities of the fuels shall be established to either meet, or exceed those assumed in the safety analysis.

" Fuel assembly (system) damage shall never prevent control rod insertion when it is required.

" The number of fuel rod failures shall be conservatively estimated for postulated accidents.

• Fuel coolability shall always be maintained.

" The mechanical design of fuel assemblies shall be compatible with co-resident fuel and the reactor core internals.

" Fuel assemblies shall be designed to withstand the loads from in-plant handling and shipping.

The generic criteria are applied to the fuel rod and fuel assembly designs. These criteria are listed in Table 5-2. Some of the criteria specified in EMF-92-116(P)(A)

(Reference 8.17) are for analyses other than the mechanical design evaluations. Only the criteria used for the mechanical design evaluations are presented below.

Table 5-2 Generic Mechanical Design Criteria Criteria io C Event Section Description Criteria Category 3.2 Fuel Rod Criteria 3.2.1 Internal Hydriding Hydrogen content in components controlled to a minimum Normal Operation level during manufacture to limit internal hydriding.

3.2.2 Cladding Sufficient plenum spring deflection and cold radial gap to Normal Operation Collapse prevent axial gap formation during densification.

Stress and Strain Limits Pellet / Cladding For M5' M cladding, strain [ Jand no centerline melting1 . Normal Operation Interaction I and AOOs Page 87 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Criteria Setin Description Criteria Event Category Section Ctgr ASME Section III, Appendix III Article 111-2000, Cladding Stress Normal Operation and AOOs Fuel Rod ACOs and 3.2.7 Mechanical ASME Section III, Appendix F. Postulated Fracturing Accidents Models included in NRC-approved fuel performance codes Normal Operation, 3.2.8 Fuel Densification and taken into account in analyses contained in Sections APOs, and 3.2.2, 3.2.4, 3.2.5, and 3.3.7 of EMF-92-116(P)(A) Accidents 2 3.3 Fuel System Criteria 3.3.1 Stress, strain, and loading limits on assembly components. (See 3.3.9 for handling and 3.4 for accident conditions.)

Spacer Grid Lateral load < load limit. Normal Operation and AO0s Normal Operation, Upper and Lower Limiting loads occur during handling and postulated AOOs, and Tie Plates accidents. Postulated Accidents 3.3.2 Cladding Fatigue Cumulative usage factor for M5TM cladding (CUF) [ Normal Operation 3.3.2and AO0s 3.3.3 Fretting wear No fuel rod failures due to fretting wear. Normal Operation Oxidation, Acceptable maximum oxide thickness. For M5IM cladding, Ox3.4 Hridting, and best estimate oxide [ ]. Effects of oxidation and Normal Operation 3.3.4 Hydriding, and crud included in thermal and mechanical fuel rod analyses.

Crud Buildup Stress analysis to include metal loss due to oxidation.

3.3.6 Axial Irradiation Growth Fuel Rod Clearance remains between fuel rod and UTP/LTP at EOL. Normal Operation The fuel assembly length shall not exceed the minimum Fuel Assembly space between upper and lower core plates in the cold Normal Operation condition at EOL.

Acceptable maximum internal rod pressure. Allowable Rod Internal internal pressure not to exceed [ Normal Operation Pressure ]. When internal pressure exceeds system pressure, and AOOs pellet-to-clad gap does not open during steady-state or increasing power.

3.3.8 Assembly Liftoff No liftoff from core lower support. Normal Operation and AOOs 3.3.9 Fuel Assembly Assembly withstands 2 1/2 times the weight as a static force. Normal Operation Handling 3.4 Fuel Coolability Structural Maintain coolable geometry and ability to insert control rods. Postulated I Deformations ASME Section III, Appendix F. Accidents 1 Fuel centerline melt evaluations are addressed by SCE.

2 Fuel densification is included in the LOCA model.

The fuel design objectives stated earlier include assurance of fuel coolability and control rod insertability after postulated accident events. Criteria sections 3.2.7 and 3.4 in Table 5-2 pertain to these objectives. Seismic and Loss of Coolant Accident (LOCA)

Page 88 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel analyses were performed for the lead fuel assemblies in order to verify that these criteria were satisfied. Analyses reported in ANP-2839(P) (Reference 8.3) demonstrated that these criteria were met for the lead fuel assemblies. The lead fuel assembly analyses examined the various possible mixed row configurations for the lead assemblies along with the co-resident fuel. The maximum grid impact forces occurring from Safe Shutdown Earthquake (SSE) and LOCA events were combined using the SRSS method and compared to the allowable grid strength for the HTP spacer grid.

Results indicated that the combined maximum impact force was less than the allowable grid strength. The allowable grid strength is established at a 95 percent confidence level on the true mean from the distribution of experimentally determined grid crush data at the operating temperature. In addition, results also indicated that the stresses in the fuel rods, guide tubes, and other fuel assembly components resulting from seismic and LOCA-induced deformations are within acceptable limits. Therefore, fragmentation of the fuel rod will not occur and the reactor can be safely shutdown under faulted condition loading. These conclusions were also shown to be valid under an Operating Basis Earthquake (OBE) event.

Similar to the seismic and LOCA analyses performed for the lead fuel assemblies, AREVA will analyze accident loadings for the AREVA fuel assembly in the San Onofre core under mixed row and full row configurations. These analyses will serve to demonstrate fuel coolability and control rod insertability for the reload of AREVA fuel assemblies in mixed cores with the co-resident fuel as well as in a full core configuration.

AREVA intends to apply the generic mechanical design criteria contained in EMF 116 (Reference 8.17) and also listed in Table 5-2 to evaluate the design improvements to the lead fuel assembly design already operating at San Onofre. The fuel rod design used in the lead fuel assemblies that were introduced at San Onofre Unit 2 in Cycle 16 was evaluated against the fuel rod design criteria listed in Table 5-2 in accordance with the NRC-approved mechanical design methods described in References 8.10, 8.9, 8.12 and 8.14. Evaluations included cladding collapse, cladding steady-state and AOO (Condition II) strain, cladding stress, cladding fatigue, cladding oxidation, fuel rod growth, and rod internal pressure. Analyses reported in ANP-2839(P) (Reference 8.3) demonstrated that each of these criteria was met for the lead assembly fuel rod design.

This fuel rod design will be maintained for the reload design. It is expected that the reload fuel rod design will continue to meet the mechanical design criteria up to the licensed burnup limits for operation in any of the San Onofre units. AREVA NP will document the design evaluation process demonstrating compliance to the generic criteria and prepare a summary of the evaluation for possible use in an audit to confirm that AREVA NP is in compliance with these design criteria.

The mechanical design criteria for the reload fuel rod design will be evaluated using the RODEX2 fuel performance code (Reference 8.60). The RODEX2 code is benchmarked against measured fuel performance data that includes a wide range of design parameters and irradiation histories. The code is benchmarked against attributes such Page 89 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel as fuel centerline temperature, fission gas release, gas pressure, cladding diameter, rod internal free volume, and fuel rod axial elongation. Detailed benchmarking results are available in References 8.10, 8.60 and 8.61. Results indicate generally conservative predictions by the RODEX2 code of the fuel centerline temperature data available at the time of code approval. AREVA has reviewed the RODEX2 code in light of the information notice recently issued by the NRC regarding the lack of fuel thermal conductivity degradation with burnup in legacy codes and has determined that [

I I

Page 90 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Page 91 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-6 [

]

II

.11 Page 92 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-7 [

I

.I Page 93 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel In addition to the conservatism in the fuel temperature predictions used for strain calculations as discussed above, the approved RODEX2 methodology (Reference 8.60) has also established biases in certain input design parameters for calculation of cladding strain. The conservatisms include selection of the minimum cladding inner diameter along with the maximum pellet outer diameter (minimum pellet-to-clad gap) for earlier pellet-to-clad contact, maximum pellet density resulting in a harder pellet, and minimum cladding thickness for faster creepdown. The same conservatisms also apply to the cladding fatigue calculation methodology. In totality, the RODEX2 cladding strain and fatigue methodology provides conservative predictions of cladding strain and cyclic fatigue.

The cycle-specific cladding strain and cyclic fatigue analyses for the AREVA reload fuel assemblies will be performed in accordance with the approved RODEX2 methodology in Reference 8.60 using cycle-specific steady-state power histories. The analyses will confirm that the design criteria listed in Table 5-2 are met for the plant specific Condition II events and fatigue duty cycles.

Rod Internal Pressure RODEX2 was benchmarked against measured fission gas release data up to 62 GWd/MTU rod average exposure and its fission gas release model was calibrated to provide best-estimate fission gas release predictions up to this exposure (References 8.60 and 8.61). Figure 5-8 and Figure 5-9 present the RODEX2 fission gas release validation results. Figure 5-10 is a plot of calculated versus measured fission gas release. Figure 5-11 is a plot of fission gas release prediction deviation versus exposure. There is no bias of the prediction with exposure.

Page 94 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-8 [

I I

Figure 5-9 [

I Page 95 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-10 [

II co ARE VA considers its design analysis methodology for determining limiting rod internal Page 96 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel pressures, described and approved in Reference 8.12, to be appropriate and conservative.

The cycle-specific rod internal pressure analysis for the AREVA reload fuel assemblies will be performed in accordance with the approved RODEX2 methodology in Reference 8.12 using cycle-specific steady-state power histories. The analyses will confirm that the design criterion listed in Table 5-2 is met up to the fuel rod licensing limits.

5.1.5 Operating Experience Operational experience (OE) is an indispensable knowledge base to demonstrate the reliability and the performance of a fuel assembly design. The relevance of such OE increases all the more in the case of a design with technical features significantly different from all other designs.

HTP represents such a design. Whereas fuel assemblies equipped with traditional spacers employ springs and dimples to support each fuel rod in its spacer cell and have mixing vanes along the top edges of the spacer strips which significantly enhance thermal hydraulic performance, the HTP spacer represents an entirely different concept in spacer design for pressurized water reactor (PWR) fuel. The HTP spacer features strip doublets which are shaped such that they not only serve as spring elements to firmly hold the fuel rods in radial alignment but also produce curved internal flow channels to achieve the desired thermal hydraulic performance.

HTP is primarily the designation of a special type of spacer but is also used to denote a fuel assembly design in which this type of spacer is the major component. The first insertion was into a U.S. plant in 1988; the HTP design now has over 20 years of operational experience.

The advanced CE 16x16 HTP fuel assembly design currently in operation at San Onofre Unit 2 and intended for Unit 3 is an HTP-type fuel assembly design with M5TM fuel rod cladding, M5 TM MONOBLOC TM guide tubes, M 5 TM HTP intermediate and upper end grid, alloy-718 HMP lower end grid, and a Robust FUELGUARD lower tie plate. An overview of both the overall operating experience gained with the various components of the fuel assembly design as well as the specific operating experience in Westinghouse, B&W, and CE plants is provided below, including applicability to San Onofre.

5.1.5.1 Operating Experience with HTP Fuel Assemblies As of December 2009, the operational experience with HTPTM fuel assemblies comprises a total of 11,710 fuel assemblies irradiated in 47 nuclear power plants (NPP).

From these, 7,215 are in 27 European plants (Belgium, France, Germany, Spain, Page 97 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Sweden, Switzerland, UK, The Netherlands), 4,355 assemblies in 17 U.S. plants, 80 assemblies in 2 Japanese plants and 60 assemblies in a Brazilian plant.

This experience spans the entire range of fuel rod arrays from 14x14 to 18x18, as well as reactors supplied by various vendors, such as Combustion Engineering (CE),

Framatome, Westinghouse, Siemens and Babcock & Wilcox (B&W). The largest share, 4,765 assemblies have been loaded into 12 ft. Framatome/Westinghouse plants with a 17x17 array, followed by the 16x16 array for Siemens plants with 1,516 assemblies.

The operational experience gained with HTPTM fuel covers a variety of core formations.

Table 5-3 provides an overview or industry operating experience with HTPTM fuel.

HTPTM fuel assemblies have been loaded into reactors which are operated in significantly different strategies ranging from 6 to 24 month cycles. As of December 2009, more than 5,400 HTPTM fuel assemblies equipped with Gadolinia rods have been loaded worldwide into 29 NPPs. The number of Gadolinia rods within an assembly varied between 4 and 28 with Gd 2 0 3 concentrations from 2 up to 8 wt%. 15x15 and 17x17 HTPTM fuel assemblies with configurations ranging from 4 Gadolinia rods of 2 wt% to 24 Gadolinia rods of 8 wt% have been prepared for Westinghouse type plants.

A maximum fuel assembly average burnup of 67 MWd/kgU has been achieved with HTPTM assemblies containing Gadolinium poisoned rods.

Table 5-3 Operational Experience with HTP (Status December 2009)

1. of FAs Maximum # of defective Plant type pt o nrstin in irradiated FA burn-up rods plants Insertion operation (MWd/kgU) (accumulated)

CE-14x14 5 1988 556 1,163 60 [ ]

CE-15x15 1 1988 204 784 53 [ ]

CE-16x16 11 20081 81 81 8 [ ]

Westinghouse 3 1994 221 837 54 [1

-14x14 Westinghouse 1 1991 157 702 58 []

-15x15 Westinghouse 6 1994 726 1,971 57 17x17, 12ft 6_9472_,715__

Framatome 8 1993 467 2,794 67 []

17x17, 12ft B&W- 15x15 7 2003 774 839 50 Siemens- 3 2001 357 448 70 [ ]

15x1 5 Siemens- 9 1989 1,047 1,516 59 Page 98 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

1. of FAs Maximum # of defective Plant type plants Insertion in irofdFas FA burn-up rods operat irradiated (MWd/kgU) (accumulated) 16x1 6 Siemens- 3 1992 468 648 61 []

18x1 8 Total 47 4,985 11.710 70 [ ]

Note 1: An additional 8 CE-16x16 HTP lead assemblies were inserted for use at San Onofre Nuclear Generating Station in 2010.

Approximately 50% of the HTPTM experience is represented by fuel assemblies featuring a design with (non- HTPTM) bi-metallic ("bi-met", reflecting Zircaloy-4 strips with Alloy-718 springs) at the outermost positions, all Zircaloy-4 HTPTM spacers at intermediate positions, Zircaloy-4 cladding and structural material, and FUELGUARD debris filters. The bi-met upper and lower grids are being phased out in the U.S. and replaced with the Alloy-718 HMP grid in the lowest grid position and HTPTM grids in the uppermost grid position.

With 6,593 fuel assemblies, more than half of all inserted HTPTM fuel assemblies have achieved a burnup of higher than 40 MWd/kgU. The maximum assembly burnup is 70 MWd/kgU. The burnup distribution of the HTPTM fuel assemblies as of December 2009 is shown in Figure 5-11.

Page 99 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-11 Burnup Distribution of the HTP FA (Status December 2009)

Nwnbe of FaieAr :e?7ft~ T-4al Nunlt,ý--ý ý:,f 1ý -'-ý 3 O9, -

1'.00C -

'JOC-HTP Fuel Assemblies Equipped with an HMP Spacer at Lowermost Position The first insertion of the HTPTM fuel design with High Mechanical Performance (HMP)

Alloy-718 grids (straight flow channels) at the lower grid position was in 1998. Today, significant operational experience with the HTPTM fuel assembly featuring an HMP spacer is available. Altogether, 5,527 such HTPTM fuel assemblies have been loaded worldwide into 33 plants. Figure 5-12 shows the burnup distribution of HTPTM fuel assemblies featuring an HMP at the lowermost position as of December 2009. A maximum assembly burnup of 70 MWd/kgU has been achieved.

Page 100 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-12 Burnup Distribution of FA Featuring an HMP at Lowermost Position (Status December 2009)

Numbei of Fuel Assemblies Total Number of Fuel'Assembl' es: 5 527 900-500-500-400-200 ioc

-5 0,0 30 35 40 45 5: == 0 C-A~ssembiv4 mp, tdkL The first HTPTM fuel assemblies equipped with M5TM fuel rod cladding were inserted into four plants in 2003 - four LFAs into a South American plant, four LFAs into a US CE 14x14 plant (Ft. Calhoun), a reload consisting of 36 assemblies into a German plant with a 16x16 array, and one reload with 85 assemblies into a US plant of a 15x15 B&W design (Crystal River 3). As of December 2009, 3,574 HTPTM fuel assemblies with M5 T cladding have been irradiated in 28 plants in Brazil, Germany, the Netherlands, Sweden, Switzerland, South-America, the UK and in the US. The operational experience of the combination HTPTM fuel assembly and M5TM cladding covers all arrays from 14x14 up to 18x18. Up to now, a maximum assembly average burnup of 61 MWd/kgU has been achieved.

Page 101 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-13 shows the burnup distribution of HTPTM fuel assemblies equipped with M5TM cladding material as of December 2009.

Figure 5-13 Burnup Distribution of HTP FA having Fuel Rods with M5 Cladding Material (Status December 2009)

Number of Fuel Assembhes Total Number of Fuei Assembifes. 3-574

-150

-J00 350 300

]

200-

, 50.

100-50.

0-0 50 3r5 50 555055 -

Assembly Burnup [7Wd kgO]

5.1.5.2 Operating Experience with M5 Cladding The M5TM alloy is the reference alloy of AREVA NP for fuel rod cladding material. M5TM is the result of a vast program of optimization and industrial development which started at the end of the 1980's and reached completion at the beginning of this millennium.

Since 1993, more than three million fuel rods having M5TM cladding have completed their operation or are operating in 12,528 fuel assemblies in 79 commercial reactors worldwide. These include 53 reactors in Europe (Belgium, France, Germany, Netherlands, Spain, Sweden, Switzerland and UK), 17 in the US, 6 in China, 2 in South-Africa and 1 in Brazil (Table 5-4).

The irradiation experience covers all fuel assembly arrays ranging from 14x14 to 18x18, and different fuel assembly designs as AFA3G, HTP, Mark-B and Mark-BW. It includes Enriched Natural Uranium and Enriched Reprocessed Uranium fuel, both with and without Gadolinium. The range of enrichment extends at present from 3.2 to 4.95 w/o U235 . Mixed Oxide fuels are also included, particularly in Germany and in France.

Page 102 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 5-4 Operational Experience with M5 Cladding Material (Status December 2009)

Number Maximum Maximum Fuel First Number Status 12/2009 Array n oF/R Irradiation of FAs Burnup FA Burnup Reactors (MWd/kgU) (MWd/kgU) 14x14 1 1993 2 54 49 Belgium 15x1 5 1 1998 476 55 50 17xl 7 3 2000 436 59 53 Brazil 16x1 6 1 2003 60 49 44 China 17x17 6 1999 1704 54 49 France 900MWe 17x17 18 1993 378 80 57 France 17x17 8 1997 905 65 59 1300MWe France N4 17xl 7 4 2005 964 51 46 15xl 5 1 2004 200 65 59 Germany 16x16 7 1993 1497 65 59 18x18 3 1993 611 67 61 Netherlands 15x15 1 2004 144 59 54 South Africa 17x17 2 2002 416 63 57 Spain 17x17 1 1999 4 51 46 15xl5 1 2000 232 67 61 Sweden 17x17 2 1998 506 64 58 Switzerland 15x15 1 2005 5 64 58 UK 17x17 1 2008 168 31 28 14x14 2 2003 128 67 61 15xl5 8 1995 2205 68 56 USA 16x16 1 2008 8 17x17 6 1997 1479 72 68 TOTAL 79 12528 Page 103 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 5-14 shows the fuel assembly burnup distribution with status as of December 2009. 60% of the assemblies have achieved burnups in excess of 30 MWd/kgU, and 40 percent is in excess of 40 MWd/kgU. Thus far, the maximum fuel assembly average burnup achieved is 68 MWd/kgU while the maximum fuel rod burnup achieved is 80 MWd/kgU.

Figure 5-14 Burnup Distribution of AREVA NP FA Featuring M5 Fuel Rod Cladding Material (Status December 2009)

Number cf Fmel A&sembhe.

2500-1 Ttlal numbe! o~f Foe! Ac>-enbhefs . 12528 2000-

.5 500 "5 1*-10 16- 11 "-20 20-25 25-30 30-35 35-40 40-45 43-50 50-55 55-66 60-65 6570 4ssermolv Burnup fMWdrkgti_

5.1.5.3 Operating Experience with FUELGUARD Lower Tie Plate Table 5-5 summarizes the total number of fuel assemblies in the U.S. using the Robust FUELGUARD (FG) as an anti-debris filter, capturing significant debris, thereby reducing the potential for fretting failures. First introduced in 1993 in the U.S. at Robinson Unit 2 (Westinghouse 15x1 5 plant), the FUELGUARD MT debris filter design has now been used at sixteen U.S. plants in batch quantities, and at another six U.S.

plants as lead fuel assemblies. Over five-thousand assemblies with FUELGUARDTM anti-debris filters have been irradiated in the U.S. through the end of 2010 as shown in Table 5-5 below. Worldwide, 11,745 PWR fuel assemblies have been irradiated with the FUELGUARD TM anti-debris filter.

Page 104 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 5-5 US Operational Experience with FUELGUARD Lower Tie Plate (Status December 2010)

U.S. Power Plant Array # HTPTm FAs Kewaunee W14 172 Robinson 2 W15 662 Comanche Peak 1 173 Comanche Peak 2 266 Shearon Harris 1 W17 756 Braidwood 1 8 Sequoyah 1 4 Palisades CE15 536 Millstone 2 420 St. Lucie 1 508 Ft. Calhoun CE14 305 Calvert Cliffs 1 2 Calvert Cliffs 2 2 Palo Verde 1 8 SONGS 8 ANO1 237 Crystal River 3 242 Davis Besse 228 Oconee 1 B&W15 60 Oconee 2 136 Oconee 3 132 TMI1 161 Total 5026 5.1.5.4 Operating Experience with MONOBLOC Guide Tubes The MONOBLOC MT guide tube incorporates a solid tube design that features a constant outer diameter for the full length of the guide tube, and two inner diameters. Worldwide, as of December 2010, 22,623 fuel assemblies have been irradiated with MONOBLOCTM guide tubes made from Zircaloy-4 material, and an additional 3,209 fuel assemblies made from M5TM material. The MONOBLOC MT tube design has also been utilized for guide tubes in multiple lead assembly programs in the U.S., including SONGS, and is used for instrument tubes in all seven Babcock & Wilcox plants in the U.S.

5.1.5.5 Overall HTP Fuel Reliability Over the time period of more than 20 years, during which altogether approximately 2.7 million fuel rods have been irradiated worldwide, a total of [ ] fuel rod failures have been reported through December 2010. The defective fuel rods were found in [ ]

Page 105 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel separate fuel assemblies in [ ] different plants. Table 5-6 summarizes the fuel rod failures associated with HTP fuel designs.

[

.1 Table 5-6: [

I Baffle Contami Plant Type F/R Fretting Interaction Debris Handling nation PCI Unknown Page 106 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel I + I I I

]

Page 107 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Design features of the CE16 HTP fuel assembly, such as the lower HMP spacer grid and FUELGUARDTM lower tie plate, eliminates the majority of identified causes of fuel rods failures associated with fuel rod fretting at bi-met grid locations and spinning rods as shown in Table 5-6. AREVA's design control and fuel reliability program continues to evaluate all fuel rod failure mechanisms to eliminate such failures from reactor operation.

5.2 LOCA Analyses The loss-of-coolant accident is analyzed as required to assure that the design bases for the ECCS satisfy the requirements of 10 CFR 50.46 regarding ECCS acceptance criteria. The realistic large break LOCA (RLBLOCA) and the small break LOCA (SBLOCA) analyses for 10 CFR 50.46 and SRP Section 15.6.5 are discussed in Sections 5.2.1 and 5.2.2, respectively. Also required in SRP Section 6.3 is a review of the effects of pipe breaks, including containment response.

A summary discussion of the LOCA analyses is provided in this section. The complete AREVA report is provided in Enclosures 3 (LBLOCA) and 4 (SBLOCA) to this LAR.

5.2.1 Large Break LOCA The large break LOCA analysis is performed for SONGS by applying the S-RELAP5, RODEX3A, and ICECON computer codes. The large break LOCA approach applied for SONGS is based on the methodology documented in Reference 8.58, with specific deviations made in response to NRC reviews of other utility LARs. This altered methodology is referred to as the "transition program or transition package". The modifications to the methodology documented in EMF-2103(P)(A) are described in detail in Enclosure 3. This methodology follows the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach, which outlines an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifies the uncertainties for the large break LOCA analysis. The RLBLOCA methodology conforms to the SRP Section 6.3 acceptance criteria for realistic evaluation models as described in Regulatory Guide 1.157.

Page 108 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The large break LOCA event is characterized by a postulated large rupture in the reactor coolant system (RCS) cold leg. The RLBLOCA analysis considers a break range from [ ] times the reference break area of the cold leg pipe. Two scenarios are run, with loss of offsite power and with no loss of offsite power. The non-parametric statistical approach of the RLBLOCA analysis samples key plant parameters such as break size and pressurizer pressure through an operational range, or the Technical Specifications LCO range, as appropriate. A mixed core of AREVA HTP 16x16 fuel and Westinghouse 16x16 fuel is hydraulically taken into account in the analysis. The full list of sampled parameters and their range of values as well as more detailed large break LOCA event description may be found in Enclosure 3.

Results from the analysis show that the 10 CFR 50.46(b) acceptance criteria for PCT, maximum oxide thickness, and hydrogen generation are met with significant margin.

Detailed realistic large break LOCA analysis results are shown in Section 3.5 of .

5.2.2 Small Break Loss-of-Coolant Accident The AREVA S-RELAP5 SBLOCA evaluation model for event response of the primary and secondary systems and hot fuel rod used in this analysis consists of two computer codes, S-RELAP5 and RODEX2/2A. The appropriate conservatisms, as prescribed by Appendix K of 10 CFR 50, are incorporated. This methodology has been reviewed and approved by the NRC to perform SBLOCA analyses (Reference 8.59). The methodology documented in Reference 8.59 has been modified based on comments from NRC reviews of other utility LARs. The modifications are described in detail in .

The results of the SBLOCA analysis show that the 10 CFR 50.46(b) acceptance criteria for peak clad temperature (PCT), maximum oxide thickness, and hydrogen generation are met with significant margin. Detailed SBLOCA analysis results are shown in .

The SBLOCA is defined as a break in the RCS pressure boundary which has an area of up to approximately 10% of a cold leg pipe area. The most limiting break location is in the cold leg pipe on the discharge side of the reactor coolant pump (RCP), which results in the largest amount of inventory loss and the largest fraction of ECCS fluid being ejected out through the break. This behavior produces the greatest degree of core uncovery, the longest duration of fuel rod heatup, and consequently, the greatest challenge to the 10 CFR 50.46(b) criteria 1-4.

The SBLOCA event for SONGS is characterized by a slow depressurization of the RCS with a reactor trip occurring on a low pressurizer pressure signal. The Safety Injection Actuation Signal (SIAS) is initiated when the system has further depressurized. The Page 109 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel capacity and shutoff head of the high pressure safety injection (HPSI) pumps are important parameters in the SBLOCA analysis. For the limiting break size, the rate of inventory loss from the primary system is such that the HPSI pumps cannot preclude significant core uncovery. The primary system depressurization rate is slow, extending the time required to reach the safety injection tank pressure or to recover core liquid level with HPSI flow. This tends to maximize the heatup time of the hot rod, which therefore results in the maximum PCT and local cladding oxidation. Core recovery for the limiting break begins when the HPSI flow that is retained in the RCS exceeds the mass flow rate out the break, followed by injection of safety injection tank flow. For very small break sizes, the RCS pressure does not reach the safety injection tank pressure.

The full range of break sizes, as well as more detailed small break LOCA event descriptions may be found in Enclosure 4.

A Small Break Loss of Coolant Accident (SBLOCA) analysis was performed using the S-RELAP5 SBLOCA methodology for the San Onofre Nuclear Generating Station (SONGS) to support the implementation of AREVA 16x1 6 HTP fuel with M5TM cladding (see Enclosure 4). The analysis supports operation of SONGS Units 2 and 3 at a power level of 3458 MWt (including 20 MWt measurement uncertainty) and a steam generator tube plugging level of up to 8% in both steam generators.

In order to address requests from the U.S. NRC on recent AREVA SBLOCA submittals, the SBLOCA analysis exceeds the requirements of the approved Evaluation Model.

The analysis was performed for a range of break sizes and locations. The break spectrum included cold leg break sizes ranging from 1.0 inch diameter to 9.49 inch diameter, hot leg break sizes ranging from 1.0 inch diameter to 9.49 inch diameter, and safety injection tank (SIT) breaks. The entire break spectrum analyzed met criteria 1 through 3 set by 10 CFR 50.46(b). The results of the analysis indicate that criteria 1 through 4 of 10 CFR 50.46(b) are satisfied.

The break spectrum analysis assumed Reactor Coolant Pump (RCP) trip coincident with reactor trip. An evaluation of delayed RCP trip was also performed, since delayed RCP trip can potentially produce more limiting results.

5.2.3 Long Term Core Cooling Analysis Successful initial operation of the ECCS is shown by demonstrating that the core is quenched, and the cladding temperature is returned to near saturation. Thereafter, long-term cooling is achieved by the pumped injection systems. These systems are redundant and provide a continuous flow of cooling water to the core fuel assemblies so long as the coolant channels in the core remain open. For a cold leg break, the concentration of boric acid within the core can induce a crystalline precipitation that may Page 110 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel prevent coolant flow from reaching portions of the core. This section evaluates the initial operation of the ECCS, considers the long-term supply of water to the core, and discusses the procedures to prevent the build-up of boric acid in the core.

5.2.3.1 Initial Cladding Cooldown The LOCA calculations given in Sections 3.3 and 3.4 provide a simulation of the hottest fuel pins through core quench. After quenching, core heat transfer is through pool nucleate boiling or forced convection to liquid, depending on the break location (cold leg breaks are in pool nucleate boiling and hot leg breaks are in forced convection to liquid).

Either heat transfer mechanism is fully capable of maintaining the core within a few degrees of the coolant saturation temperature. Thus, within ten to fifteen minutes following a large break LOCA, the core is returned to an acceptably low temperature.

5.2.3.2 Extended Coolant Supply Once the core is cooled to low temperature, maintaining that condition relies upon the systems that are designed to provide a continuous supply of coolant to the core.

Detailed descriptions of the plant systems and functions are provided in UFSAR Section 6.3.3.4 and UFSAR Figure 6.3-1. Long-term core cooling with the ECCS is independent of the fuel design. Thus, the current licensing basis remains valid for AREVA 16x1 6 HTP fuel assemblies.

5.2.3.3 Boric Acid Concentration The long-term cooling mechanism for a hot leg break is forced convection to liquid.

Once cooling is established, and a positive core flow is assured, boron precipitation is not an issue, and no further consideration is necessary. For cold leg breaks, there is no forced flow through the core. The liquid head balance between the core and the downcomer prevents ECCS water from entering the core at a rate faster than core boil-off. Extra injection simply flows out the break and spills to the containment floor. With no core flow, core boiling acts to concentrate boric acid adding to the potential for precipitation and core blockage. To eliminate boron precipitation and any accompanying core blockage, operator action is required to establish hot and cold leg injection (positive core flow).

Hot and cold leg injection is initiated to provide long-term cooling; this induces a positive core flow, capable of controlling the concentration of boric acid. The timing and effectiveness of the hot leg injection is established by demonstrating that the in-vessel concentration of boric acid is below solubility limits. There is no dependency on the fuel Page 111 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel element design since concentrations depend on ECCS injection rate, RCS geometry, and core power level. Since the AREVA fuel does not alter these factors, the current evaluation remains valid and is equally applicable to AREVA 16x1 6 HTP fuel.

Emergency operating procedures provide guidance to address the boric acid precipitation issue and ensure that long-term cooling is maintained.

5.2.3.4 Adherence to Long Term Core Cooling Criterion Compliance with this criterion is demonstrated in UFSAR Section 6.3.3.4. It is independent of fuel design. The initial phase of core cooling results in low clad and fuel temperatures. A pumped injection system, capable of recirculation, is available and is actuated by the operators to provide extended coolant injection. The concentration of dissolved solids is limited to acceptable levels through the timely implementation of hot leg injection. Hence, long-term cooling is established and compliance to 10CFR50.46 demonstrated, and the existing analysis of record remains applicable to AREVA fuel.

5.2.4 Conclusions for LOCA Analysis The RLBLOCA analysis was performed in accordance with EMF-2103(P) (A) as modified by the description in Enclosure 3 to support application of the AREVA NP RLBLOCA analysis methodology to SONGS. The SBLOCA analysis was performed in accordance with EMF-2328(P)(A), as modified by the description in Enclosure 4. The existing Long Term Core Cooling analysis remains applicable.

Page 112 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 6.0 CHANGES TO REACTOR CORE DESIGN AND MONITORING PROGRAM The SCE Reactor Core Design and Monitoring Program is a comprehensive site wide program to ensure the thorough engineering and safety evaluation of all potential consequences of each reactor core design and operation. This program meets or exceeds all INPO recommendations in SOER 96-02, Design and Operating Considerations for Reactor Cores, and SOER 03-02, Managing Core Design Changes.

Details are provided in Section 4.0 of Reference 8.4.

6.1 Fuel Management Guidelines for SONGS Units 2 and 3 The SONGS Fuel Management Guidelines were utilized to generate the VQP cycles shown in Section 7.1. Changes made to the guidelines for AREVA fuel are described in Sections 4.1 and 4.4. No other changes to the design values or criteria were required.

6.2 Reactor Core Design Review Team The Reactor Core Design Review Team will continue to be composed of representatives from all potentially affected site groups and review core design and fuel design changes to determine potential site impact as early as possible. The Reactor Core Design Review Team is independent of the source of the core design. No changes are required to the Reactor Core Design Review Team process.

6.3 Reload Ground Rules The Reload Ground Rules are a compilation of plant design data, design changes, licensing changes and analysis assumptions that apply to the plant data. Changes to the data have been made to identify items used by AREVA and will be presented in the same fashion as Westinghouse values. While AREVA values will be added, SCE will continue the underlying process of engineering rigor of identification, validation and documentation of plant values for use in the safety analysis. No changes are required to the Reload Ground Rule process.

6.4 Vendor Fuel Design Change Interface The Vendor Fuel Design Change Interface will continue to evaluate improvements to the basic fuel design used at SONGS. Proposed changes will continue to be reviewed by the Nuclear Fuel Management (NFM) Division of SCE and then the affected site groups consistent with the review and approval process used for design change document review and approval controls. No changes are required to the Vendor Fuel Design Change Interface.

6.5 Vendor Reload Analysis Computer Codes and Methodology Interface The computer code and methodology interface will continue to be controlled by SCE.

Software changes continue to be controlled using SCE software V&V procedures. All changes to source code are required to be tested per an approved test plan and documented in a software installation report. Concerns, issues and error notices are Page 113 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel transmitted to code vendors and evaluated per SCE software control procedures.

Analysis methodology continues to be controlled by NFM's calculation documentation procedure (Reference 8.15). Significant deviations from established methodology that increase the conservatism in a methodology approved by the NRC will continue to be reviewed by SCE management. Significant deviations from established methodology that reduce known conservatism, but are still compliant with the methodology approved by the NRC will be reviewed by SCE management and/or the appropriate fuel vendor, as appropriate. Methodology changes that are not compliant with the NRC-approved methodology may be implemented only as permitted by 10 CFR 50.59. Fuel vendors are continually refining and updating their NRC-approved reload methodologies or developing new reload methodologies for NRC review and approval. These methodology improvements will be implemented at SCE discretion, if permitted by 10 CFR 50.59. All deviations are reviewed and approved by the analyst, NFM supervision and management. The NFM quality program will continue to be utilized to ensure high quality of analyses through the use of pre-job briefs, analysis review committee meetings and assessments of performance.

6.6 SCE Engineer Training Qualification Guide The pertinent qualification guides for core reload engineers are not impacted because SCE will continue to utilize the training process. Training needs analysis is performed in accordance with the SCE Reload Analysis Topical Report (Reference 8.4) and per INPO training guidelines (Reference 8.49).

6.7 Core Reload Analyses and Activities Checklist The Core Reload Analyses and Activities Checklist provides a list of the calculations performed during a core reload effort. The procedure will be updated for AREVA fuel supply, and will continue to be updated as needed each cycle with changes in calculation scope and computer codes.

6.8 Source Verification and Vendor Fuel Fabrication Interface The source verification and vendor fuel fabrication interface process will continue to ensure nuclear fuel and services supplied by either supplier are in compliance with Criterion VII of 10 CFR 50 Appendix B. Use of reactor core design information developed in the reload process will be used by the SCE quality assurance inspectors.

Verification activities of fuel and CEA manufacturing are credited in the Startup Activity Reduction Program to reduce required startup testing activities (Reference 8.47). No changes are required to the Source Verification and Vendor Fuel Fabrication Interface.

6.9 Fuel Vendor Engineering Interface The fuel vendor engineering interface will continue to operate utilizing a formal transmittal process for LOCA and fuel mechanical design. Formal data requests are made by the fuel vendor for information needed to perform vendor analyses. Formal data transmittals are utilized to provide the requested data to the fuel vendor. All Page 114 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel transmittals of data meet the SCE Nuclear Fuel Management (NFM) quality plan and 10 CFR Appendix B. No changes are required to the Fuel Vendor Engineering Interface.

6.10 Site Program Impact The Site Program Impact (SPI) process used at SONGS will continue to be used to identify and evaluate changes that impact other disciplines or departments at SONGS.

The SPI is a site wide process used in conjunction with the design change process, currently called Nuclear Engineering Change Package (NECP). For example, for the AREVA LFAs, the process of receiving, moving, and examining fuel was evaluated with the SPI process. The process of consultation with other SCE divisions will continue.

No changes are required to the SPI process.

6.11 Licensing and Design Basis Document Updates Updating of licensing and design basis documents due to core design changes will continue to be performed. The addition of AREVA data and processes has no impact on the requirement to update SONGS licensing basis and design basis documents.

SONGS will continue to use the design change process of evaluating and documenting SONGS core designs and impacts. No changes are required to the Licensing and Design Basis Documents Updates process.

6.12 Design Process Flow and Controls The design process flow and controls process will continue to be used to control design changes at SONGS including core design changes. Identification, evaluation, documentation and 10 CFR 50.59 review of changes is required for all core design changes as described in SONGS design change procedures. This activity is documented in a Nuclear Engineering Change Package (NECP) which meets 10 CFR 50 Appendix B requirements. No changes are required to the design process flow and controls process.

6.13 COLSS/CPCS Products The COLSS/CPCS products such as the data base, Reload Data Block and addressable constants will continue to be reviewed and updated as needed to address core or plant design changes. Change to inputs in the process of calculating the constants such as core thermal hydraulics overpower penalties, etc. will occur each cycle; however, the methodology and processes used to calculate the inputs and calculate constants is unchanged (See Section 4.8). No changes are required to the COLSS/CPCS products process.

6.14 Low Power and Power Ascension Testing Reactor startup testing scope for low power testing implements Startup Test Activity Reduction (STAR) Reference 8.43 criteria for selecting required tests and is not vendor dependent. Any impacts on the Low Power and Power Ascension Testing as a result of core design changes are captured in the supporting documents provided to Reactor Page 115 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Engineering. No changes are required to the Low Power and Power Ascension Testing process.

6.15 Core, SFP, and Dry Cask Storage Requirements Core: As described in Reference 8.4, as-built isotopic data and actual shutdown burnup will continue to be used to provide a balanced full core loading pattern. AREVA fuel weight is equivalent to Westinghouse fuel weight and fuel handling processes for AREVA fuel have been incorporated into the SCE processes through the AREVA LFA program.

Spent Fuel Pool (SFP): The AREVA fuel assembly weight is bounded by the weight assumed in the fuel rack design validating the SFP rack structural integrity analysis and fuel damage assumptions in the fuel handling accident analysis. [

TM]

Implementing AREVA fuel with M 5 TM cladding does not invalidate the current Technical Specifications 3.7.18 or 4.3.1, Licensee Controlled Specification (LCS) 4.0.100 requirements or allowed storage configurations.

Dry Cask Storage: The current SONGS dry storage design is not approved for storage of AREVA fuel design with M 5 TM cladding at this time. Dry cask storage for AREVA fuel is not requested in this submittal. AREVA fuel will be stored in the SONGS spent fuel pool until approved for dry storage at SONGS.

No changes will be made as part of this LAR to the Core, SFP, and Dry Cask Storage Requirements as a result of using AREVA fuel.

6.16 Core and Fuel Monitoring Core and Fuel Monitoring processes are controlled by SONGS procedures and monitor key COLSS and CPCS parameters and key plant parameters during the cycle. Data collected will continue to be evaluated to ensure the validity of the physics biases and uncertainties. No changes are required to the Core and Fuel Monitoring process.

Page 116 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 7.0 IMPACT OF CHANGES ON RELOAD ANALYSES SCE generated representative SONGS 3 Cycle 17 and Cycle 18 core designs (VQP cycles) that meet SCE fuel management guidelines to validate that the SCE Reload Analysis Methodology tools, computer codes and procedures were adequate to perform reloads for SONGS with AREVA fuel. Cycle 17 is a transition core and cycle 18 is a full core AREVA design (note the definition of full core as not including the center assembly per Section 4.1.3). These analyses demonstrate the acceptable performance of the AREVA fuel design in SONGS.

7.1 AREVA Fuel Reload Core Design Comparison of Key Physics Parameters Per the Reload Analysis topical report (Reference 8.4) SCE performs reload core physics calculations for downstream mechanical design, thermal hydraulic fuel behavior and UFSAR Chapter 15 transient analyses. This process was used to generate physics parameters for the AREVA VQP cycles.

Table 7.1.1 shows a comparison of key physics parameters for the current SONGS 3 Cycle 16 Westinghouse fuel design with the representative AREVA VQP Cycles 17 and

18. Placement of AREVA fuel and Westinghouse fuel for these cycles is shown in Figures 7.1.1 through 7.1.3. Key physics parameters for downstream use are shown in Tables 7.1.2 and 7.1.3 and Figures 7.1.4 through 7.1.6. These core designs will be utilized to describe the impact of AREVA fuel on the downstream reload analyses for SONGS Units 2 and 3.

Page 117 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 7.1.1 - Summary of VQP Fuel Cycle Design Verification SONGS3 SONGS3 SONGS 3 VQP ITEM PARAMETERS CONDITIONS Cycle 16 Cycle 17 Cycle 18 DESIGN DESIGN No 4.60w/o 3.50w/o 4.95w/o LIMIT TARGET (599 EFPD) (493 EFPD) (586 EFPD)

Cycle Maximum ARO HFP 1 Maximum 1.44 1.54 1.53 1.60 _<1.55 Fxy (Equilibrium) HFP, ARO Peak Linear Heat Cycle 2 Rate Maximum 9.27 9.52 9.18 12.8 < 9.74 (kw/ft) HFP, ARO 3 Peak (MWD/T) Rod Burnup Cycle Maximum 56,530 53,758 55,392 W:

A: 60,000 62,000 5S 60,000 62,000 HFP, ARO A:_62,00 ___62,000 Maximum HFP Cycle Boron Maximum 1663 1366 1430 1730 5 1674 Concentration HFP, ARO (ppm) HFP,_ARO Most Positive 4 -4 4 MTC at HZP (Ap/°F) BOC, HZP 0.06x10- 0.21x10- 0.15x10- 0.5x104 <0.25x104 Most Positive 6 MTC at 70%P BOC, 70%P -0.46 x 10-4 -0.43 x 10-4 -0.43 x 10. 4 0.0 x 10-4 5-0.31 x 10-4 (Ap/°F)

Most Positive MTC at HFP (AptF) BOC, HFP -0.68 x 10- -0.71 x 10.' -0.68 x 10.4 0.0 x 10-4 _<-0.55 x 10-4 Most Negative 4 4 4 4 MTC at HFP (ApvF) EOC, HFP -3.37 x 10-4 -3.18 x 10- -3.22 x 10- -3.7 x 10-4 > -3.45 x 10-Minimum HFP N-i, HFP, 6.99 7.07 6.96 6.00 > 6.00 Scram Worth (%Ap) BOC 10 Minimum HZP Scram SrmWrhBOC Worth N-i, BOGHZP, ZP 5.32 5.42 5.49 5.15 > 5.15

(%Ap) _____

Maximum Pellet Maximum in 4.60 3.50 4.95 495* <495*

11__ Enrichment (wt%) Feed Batch Gadolinia Loading:

12 - wt% of Gd 20 3 Maximum in [ ]

-No of Gad Feed Batch 8# 24 # 24 # 32 # 5 32 #

rods/assy Erbia Loading Maximum in 13 (wt% Er 20 3 in UO2) Feed Batch

• Maximum approved enrichment for spent fuel pool criticality is 4.80 w/o. 4.95 w/o enrichment was used in cycle 18 to evaluate core designs with this maximum enrichment. Increase in enrichment is not part of this request.

1. [

Page 118 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 7.1.2 - SONGS Unit 3 Nominal Physics Characteristics S3 S3 VQP Units S3C16 VQP C18 C17 Inverse Boron Worth Hot Full Power, BOC ppm/%Ap 149 138 150 Hot Full Power, EOC ppm/%Ap 115 111 112 Moderator Temperature Coefficients Hot Full Power, Equilibrium Xenon Beginning of Cycle 10-4 Ap/0 F -0.7 -0.7 -0.7 End of Cycle 10-4 ApI 0F -3.4 -3.2 -3.2 Doppler Coefficient Hot Full Power, BOC 10-5 Ap/0F -1.49 -1.49 -1.48 Hot Full Power, EOC 10-5 Ap/ 0F -1.59 -1.61 -1.61 Total Delayed Neutron Fraction, Peff BOC 0.0064 0.0064 0.0064 EOC 0.0052 0.0052 0.0052 CEA Ejected Worth Hot Full Power, BOC 0.2049 0.1968

• EOC %AP 0.2060 0.2197 *

%AP Hot Zero Power, BOC 0.4894 0.4067 EOC 0.6533 0.6010

• Maximum ARO HFP Fq (EqXe)

BOC 1.67 1.73 1.69 Maximum ARO HFP Fz (EqXe)

BOC 1.17 1.19 1.14 Average HFP ASI (EqXe)

BOC -0.014 -0.017 0.006 EOC 0.020 0.023 0.013

• Not calculated for VQP Cycle 18 since demonstration of SONGS CEA ejection analysis was only performed VQP cycle 17. CEA ejected worth is influenced by fuel management considerations (i.e. fuel enrichment, location, cycle length, etc.) and not fuel vendor.

Page 119 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 7.1.3 Comparison of HFP ARO Fxy and Fr for SONGS 3 S3-VQP S3-VQP S3-VQP S3-VQP C17 C18 C17 C18 Burnup Fxy Fxy Fxy Fr Fr Fr 0 1.446 1.488 1.529 1.423 1.460 1.498 0.25 1.441 1.484 1.529 1.419 1.457 1.503 0.5 1.441 1.487 1.529 1.418 1.461 1.506 1 1.441 1.496 1.530 1.419 1.475 1.508 2 1.440 1.523 1.526 1.418 1.500 1.505 3 1.434 1.539 1.517 1.415 1.501 1.496 4 1.430 1.515 1.504 1.410 1.460 1.482 5 1.425 1.460 1.490 1.403 1.404 1.462 6 1.424 1.436 1.494 1.396 1.385 1.452 7 1.422 1.434 1.508 1.390 1.386 1.457 8 1.419 1.442 1.521 1.389 1.406 1.467 9 1.416 1.448 1.522 1.387 1.410 1.470 10 1.414 1.439 1.520 1.387 1.400 1.469 11 1.414 1.421 1.511 1.386 1.385 1.463 12 1.411 1.398 1.498 1.385 1.368 1.455 13 1.407 1.381 1.485 1.381 1.355 1.448 14 1.402 1.367 1.470 1.376 1.343 1.444 15 1.394 1.351 1.465 1.371 1.330 1.439 16 1.386 1.334 1.456 1.365 1.317 1.431 17 1.378 1.318 1.444 1.358 1.304 1.422 18 1.371 1.306 1.434 1.351 1.293 1.412 19 1.361 N/A 1.424 1.343 N/A 1.402 20 1.352 N/A 1.413 1.335 N/A 1.393 21 1.342 N/A 1.402 1.328 N/A 1.384 21.75 1.335 N/A N/A 1.321 N/A N/A Page 120 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.1.1 Unit 3 Cycle 16 (Current) Fuel Placement Locations P4= Current C-C L~cabor, V = Fuel Vendor W W Z = E~tchvi 3

W W W W

-, -D F F

"...G W W W W W W W W W W W W W F F 2122 23 2-" 25 2-1 27 28 W W W W W W W W F F F F 22 33 34 35 W w w w W W w w F F ., F F --

3:-- ,3 43 w w w w w w w w W

w w w w W W w W 5

53 57 58 --,S *,i 52 F F F F -

W F W W W W W W W W

! F ".Eý F -= .E-* 1.'

=.

Key:

F: Fresh Fuel (108 Assemblies) 1B: Once Burned Fuel (108 Assemblies) 2B: Twice Burned Fuel (1 Assembly)

A: AREVA Fuel (0 Assemblies)

W: Westinghouse Fuel (217 Assemblies)

Page 121 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.1.2 Transition Unit 3 Cycle 17 VQP Fuel Placement Locations M- Curren,- -. C Location Fu IVe\~d r or W A Erazd- Tvp W W A A W

ý-B F 13 W A A A W A F F F 1B F

'i7 W A W A A F F lE'l F S2 2- 2 2 W A W A W A W A

-* F !z*F F 18F 2ý ýý 32 12 332 W A A W A W A W

-- F F E, F FLE 3*7 3 34 40 -2 44 A A W A W A W W FF F F W

4E. 5,617 £2 F4 A W A W A W A W F F F:2 FF A

55 5s. 5'5 52 F

W A W A W W W W IF Fd-Key:

F: Fresh Fuel (108 Assemblies) 1B: Once Burned Fuel (108 Assemblies) 2B: Twice Burned Fuel (1 Assembly)

A: AREVA Fuel (108 Assemblies)

W: Westinghouse Fuel (109 Assemblies)

Page 122 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.1.3 Full Core AREVA Unit 3 Cycle 18 VQP Fuel Placement Locations M = Cur-ent QC Lcocaric~r

.1=Fuel'-fendor A A A A A A A F F 4 + +

A A A A A A F

A A A A A A A F F 2-, 22 32- 2S 2a 2.7 28 A A A A A A A A F F F F 2;  : 0:*-2 -:-* -- 23 A A A A A A A A F F F.F 17 9 - 2 A A A A A A A A F F F F A

47 -52 A A A A A A A A F F F F A

F A A A A A A A W F F Key:

F: Fresh Fuel (108 Assemblies) 1B: Once Burned Fuel (108 Assemblies) 2B: Twice Burned Fuel (1 Assembly)

A: AREVA Fuel (216 Assemblies)

W: Westinghouse Fuel (1 Center Assembly - see Section 4.1.3)

Page 123 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.1.4 Comparison of HFP ARO Fxy for SONGS 3 1.580 1.530 1.480 x 1.430 -4c 16

-=G* C17 1.380 -- *C18 1.330 1.280 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 Burnup GWD/T Page 124 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.1.5 Comparison of HFP ARO Fr for SONGS 3 1.580 1.530 1.480 1.430 --- C16

--:=-ZC1 7

• C18 1.380 1.330 1.280 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 Burnup GWD/T Page 125 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.1.6 Comparison of Radial Power Distribution at BOC for SONGS 3 XXX Cycle 18 4223 0.9306 Y.YY Cycle 170.50 .60 Z.ZZ cycle 16. . - 0.3720 0.7830 0.3367 0.5400 1.0106 1.1226 0.9577 0.3320 0.5140 0.7790 0.8910 0.9600 03050 . -3 -:--r- 9730 0

.7-8* 9 10 i11 12 13

. ... ÷A- . 0.4371 1.0361 1 1448 0.9702 1.1951 0.4580 0.8690 1.0460 1.1630 1.2340 1.1260 0.4140 0.9240 1.0910 1.1980 1.1080 1.2970 14 15 16 17 18 19 20 0.4372 1.0876 1.0256 1.2066 0.9696 1.2288 1.0732 0.4560 1.0000 1.1180 1.1480 1.2280 1.1520 1.1320 0.4140 1.0020 1.1350 1.2910 1.1260 1.3190 1.0970 21 22 23 24 25 26 27 28 0.3380 1.0368 1.0260 1. 800 0.9791 1.1627 1.1437 1.0215 0.3320 0.8690 1.1190 1.2310 1.1450 1.2090 1.2360 1.2370 0.3060 0.9260 1.1360 1.2950 1.1100 1.3150 1.1860 1.3320 29 , 30 31 32 .33 '34 35 36 0.5409 1.1461 1.2080 0.9797 1.2119 1.0576 1.2493 1.0431 0.5130 1.0450 1.1480 1.1460 1.1900 1.2020 1.3380 1.2630 0.4830 1.0950 1.2930 1.1110 1.2840 1.0590 1.2570 1.0420 37 38 39 40 41 42 43 44 1.0121 1.1554 0.9721 1.1656 1.0602 1.2646 1.1202 1.2778 0.7760 1.1610 1.2270 1.2080 1.2030 1.1890 1.1580 1.1110 45 0.8590 1.2020 1.1290 1.3160 1.0600 1.2040 0.9490 0.8870 0.4231 0.3540 46 47 48 49 50 51 52 53 0.3780 1.1257 0.9736 1.2373 1.1487 1.2517 1.1203 1.2091 0.9410 0.8840 1.2320 1.1500 1.2330 1.3390 1.1570 1.1900 1.0400 54 1.0430 1.1130 1.3220 1.1860 1

.2570 0.9490 1.1440 0.8350 0.93 18 0.6160 55 56 57 58 , 59 , 60 61 62 0.7880 0.9577 1.1951 1.0732 1.0215 1.0431 1.2778 0.9410 0.8581 0.9600 1.1260 1.1320 1.2370 1.2630 1.1110 1.0400 0.8700 0.9730 1.2970 1.0970 1.3320 1.0420 0.8870 0.8350 0.7250 Page 126 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 7.2 AREVA Fuel Reload Core Thermal Hydraulic Analysis As described in Section 4.2, the [

The next step in the performance validation of the [

] was to perform analyses with the Thermal-Hydraulic codes. The following Thermal-Hydraulic analyses were performed in support of the AREVA fuel effort:

1. T-H Inputs: Evaluates the core design and plant design changes, and prepares the TORC input models for use in downstream analyses.

2. T-H Inlet Flow Distribution: A core inlet flow distribution was developed during initial plant design, based on a full core of the same fuel design. [

]

3. [

]

4. T-H MSCU Verification: The SONGS SAFDL Design Limit DNBR is based on the Modified Statistical Combination of Uncertainties (MCSU) method. [

Page 127 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 7.2.1 Thermal-Hydraulic Inputs Two-Stage TORC quarter-core models were developed for the different configurations to be considered:

[

For the purpose of this licensing effort of AREVA fuel, a [

] The TORC models were put through a series of test cases, and it was confirmed that the implemented changes provide expected results.

7.2.2 Thermal-Hydraulic Inlet Flow Distribution With the appropriate T-H models created and tested, the next effort was to determine the impact of the fuel on the [

Page 128 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.2.1 Inlet Loss Coefficient Comparison

[

.1 Page 129 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.2.2 Entire Spacer Region Loss Coefficient Comparison I

.1 Page 130 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.2.3 Pressure Drop Profile - Entire Assembly I:

I Page 131 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The results of the mixed core inlet flow distribution analysis for 4-RCP flow are shown in Figure 7.2.4.

Figure 7.2.4 Mixed Core Inlet Flow Distribution -4 Pump Flow

.1 The results of the full AREVA core inlet flow distribution analysis are shown in Figure 7.2.5.

Page 132 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.2.5 Full AREVA Core Inlet Flow Distribution - 4 Pump Flow I

Figure 7.2.5

.1 Page 133 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel The 3-RCP inlet flow distributions were also recalculated, and are shown in Figure 7.2.6 (mixed core) and 7.2.7 (full AREVA core).

Figure 7.2.6 Mixed Core Inlet Flow Distribution - 3 Pump Flow Page 134 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.2.7 Full AREVA Core Inlet Flow Distribution - 3 Pump Flow

[

I Page 135 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 7.2.2.1 Conclusion on Inlet Flow Distribution For future [

7.2.3 Thermal-Hydraulic Tuning and Benchmarking The purpose of the CETOP-D benchmarking analysis is to determine the [

Page 136 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 7.2.3.1 Mixed Core T-H Results Table 7.2.1 shows the overpower adjustment factors for [

Table 7.2.1

[

Page 137 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 7.2.3.2 Full AREVA Core T-H Results In order to evaluate the impact of the [

I Table 7.2.2

[

1 7.2.3.3 Conclusions on CETOP Benchmarking Analysis The discussion in this section shows that the CETOP benchmarking [

Page 138 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 7.3 FATES3B w/M5TM Fuel Behavior Predictions This section provides fuel behavior comparisons of [

] using the FATES3B w/M 5 TM code.

This comparison differs from that of Section 4.3.3.3, in that [

] on key performance parameters (Sections 7.3.1, 7.3.2 and 7.3.3). The comparison is based on the VQP Cycle 17 reload evaluation. The [

I Comparisons of FATES3B w/M 5 TM predicted [

7.3.1 Fuel Temperature The fuel average temperature comparison of the AREVA U0 2 M5TM clad fuel rod with the [

Page 139 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

]

7.3.2 Power-to-Centerline Melt (PTM)

The power-to-centerline meltcomparison of the [

7.3.3 Internal Hot Gas Pressure The rod internal pressure history comparison of the AREVA U0 2 M5TM clad fuel and the

[

Page 140 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel I

Page 141 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.3.1 Steady-State and Transient Rod Power - VQP Cycle S3C17

[

I Page 142 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.3.2 Axial Power Distributions - VQP Cycle S3C17 Page 143 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.3.3 Fuel Average Temperature Comparison

[

I Page 144 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.3.4 Fuel Average Temperature Comparison

[

I Page 145 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.3.5 Power-to-Centerline Melt (PTM) Comparison

[

I Page 146 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.3.6 Rod Internal Pressure Comparison

[

I Page 147 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.3.7 Rod Internal Pressure Comparison

[

I Page 148 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 7.4 VQP Comparison of Selected Non-LOCA Transient Analysis A summary of the impact of the AREVA fuel in the SONGS reactor for UFSAR Chapter 15 Non-LOCA events is provided in Attachment C. Based on this evaluation, the CEA ejection event is impacted by the change in fuel vendor and several other events are impacted through DNBR propagation and DNB statistical convolution fuel failure predictions. The Pre-trip Steam line break event (Section 7.4.1) was chosen to illustrate the application of DNB propagation and statistical convolution for fuel failures. The CEA ejection event (Section 7.4.2) is presented to illustrate the impact of the AREVA fuel change.

7.4.1 VQP Pre-trip MSLB Event Summary The Steam Line Break (SLB) Pre-Trip Power Excursion Event is analyzed to determine the response of the NSSS, prior to Reactor Trip, following a pipe break in the Main Steam Line. A rupture in the main steam piping system increases steam flow from the steam generators. This increase in steam flow increases the rate of RCS heat removal by the steam generators and leads to severe cooldown (temperature and pressure decrease) of the RCS. In the presence of large negative moderator temperature coefficient of reactivity (MTC), the decrease in temperature causes core power to increase. This may result in fuel failures and release of significant amounts of radioactive material.

The Reactor Protection System (RPS) trips available to mitigate this event are the Steam Generator Low Pressure (LSGP) and Low Level (LSGL) trips, the Core Protection Calculators (CPCS) Trips and Variable Overpower Trip (VOPT). The CPCS VOPT Trip produces a reactor trip when the reactor core power exceeds the VOPT Setpoint. The insertion of the CEAs upon reactor trip eventually terminates the DNB excursion. The reactor trip produces an automatic turbine trip.

Loss of Off-site AC Power (LOAC) was assumed to occur during the SLB Event, which results in a coast down of the reactor coolant pumps (RCPs). The RCP coastdown makes the event more adverse for DNBR. The core temperature decrease in the presence of a large negative MTC, results in an increase in the core power. The timing of LOAC is chosen to occur concurrent with Reactor Trip.

The CENTS, CPCS FORTRAN Simulation, and [

] such that, in Page 149 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel combination with the CPCS VOPT Trip and the predicted fuel failures, the radiological consequences of the Steam Line Break Pre-Trip Power Excursion Event will remain below the 10 CFR 50.67 and GDC 19 limits.

7.4.1.1 Analyses Performed for VQP Fuel Cycle with AREVA Fuels An analysis was performed for VQP Cycle 17 using AREVA fuel. The list of key input parameters used in that analysis is provided in Table 7.4.1. The resulting sequence of events is provided in Table 7.4.2. Figure 7.4.1 depicts steam flow for the duration of event prior and after MSIV closure. Figure 7.4.2 shows DNBR behavior for the event.

Table 7.4.1 Pre-Trip Steam Line Break Key Parameters Parameters Units Unit

_J VQP - Cycle 17 Core Power MWt 3458 (100.582% of 3438 MWt)

Initial Core Inlet Temperature OF 560 (a) 443,520 Initial RCS Flow Rate gpm (45,695 Ibm/sec)

Initial RCS Pressure psia 2,300 (b)

Moderator Temperature Coefficient pp/0 F -3.7x1 0- (c)

(MTC) __/_F_-3.7 __10 _4 __c_

CEA Worth at Trip %pp -6.0 Steam line Area ft2 7.41 Steam Line Break Area ft2 Double Ended Guillotine (modeled as 100)

Delayed Neutron Fraction --- 0.00412 (a) Technical Specification 3.4.1 (referencing LCS 3.4.100) upper limit is 558 °F (b) Technical Specification 3.4.1 (referencing LCS 3.4.100) upper limit is 2,275 psia (c) Technical Specification 3.1.4 and LCS 3.1.100 limit Page 150 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 7.4.2 Sequence of Events for the Outside Containment Pre-Trip DEG Steam Line Break Event Time (see) Event Setpoint or Value 0.0 Outside Containment Break in the Main Steam Line 7.41 ft2 5.7 CPCS VOPT Trip Setpoint Reached ---

Reactor Trip Generated ---

6.1 Loss of Normal AC Power Occurs ---

Reactor Coolant Pumps Begin to Coastdown ---

6.3 Main Steam Isolation Signal (MSIS) Setpoint 675 psia Reached 6.3 Maximum Core Power 115% of 3438 MWt 6.9 CEAs Begin to Drop ---

7.2 (b) Main Steam Isolation Valves (MSIVs) Begin to ---

Close 7.4 Minimum DNBR Occurs <1.31 15.2 (b) MSIVs Full Closed ---

26.9 Safety Injection Actuation Signal (SIAS) (a) 1560 psia 1800 Plant Cooldown Initiated by Manual Operation of ---

ADV on the Intact Steam Generator (a) SIAS not credited in analysis.

(b) See Figure 7.4.1 for steam flow before and after MSIV closure 7.4.1.2 Pre-Trip MSLB Event DNBR Propagation A calculation was performed to compare the Pre-Trip Main Steam Line Break cladding strain during DNB conditions for [

] relative to AREVA M5TM cladding. Results are presented in Table 7.4.3.

Page 151 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 7.4.3

[

+ 4 4 4 +

4 4

.] Therefore, DNB propagation will not occur for M 5 TM fuel during this event and coolable geometry will continue to be maintained.

7.4.1.3 Pre-Trip MSLB Event Analysis Dose Results The Pre-Trip MSLB analysis is also performed to determine the dose consequences as a result of the implementation of AREVA fuel. For the VQP cycle 17 core described in Table 7.1.1, the Pre-Trip MSLB event predicted fuel failure values with [

].1 Page 152 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 7.4.4 Comparison of Results for Pre-Trip Main Steam Line Break Parameter Acceptance Criterion Analysis Results AREVA Fuel Westinghouse Fuel Coolable Geometry Maintain Maintained Maintained Exclusion Area 5100% of 10 [ Bounded by doses Boundary Doses CFR50.67 reported in UFSAR Low Population Zone _<100% of 10 [ Bounded by doses Dose CFR50.67 reported in UFSAR Control Room Dose _< GDC 19 [ Bounded by doses reported in UFSAR Page 153 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.4.1 Steam Generator Steam Flow 6000 5000 U

4000 0

r-4 3000 14J V) r4j 2000 0

E-1000 0

0 5 15 20 25 30 Time, Seconds

-- Affected SG Total Steam Flow (lbm/sec)

Intact SG Total Steam Flow (lbm/sec)

Page 154 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Figure 7.4.2 DNBR vs. Time 7.4.1.4 VOP Pre-Trip MSLB Event Analysis - Conclusion As shown in Table 7.4.1.4, coolable geometry will be maintained and all dose criteria for this event will be met with the use of AREVA fuel.

Page 155 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 7.4.2 VQP CEA Ejection Event Summary 7.4.2.1 CEA Ejection Analysis for VQP Fuel Cycle with AREVA Fuels A CEA Ejection analysis was performed with the current UFSAR Chapter 15 analysis methodology for the VQP Cycle 17 with both Westinghouse fuel and AREVA fuel, as described in Section 7.1. This analysis incorporated the process and input changes described in Section 4.5.4.

The time in life considered in the analysis was not limited to beginning of cycle (BOC) or end of cycle (EOC). Burnup-dependent parameters were evaluated at different burnup points to determine the limiting combination of ejected CEA worth and power peaking factor.

The CPCS Variable Over Power Trip (VOPT) is credited in this analysis. [

.1 The power rise caused by CEA ejection is terminated by Doppler effect well before the scram CEAs drop into the core. Any changes to the VOPT setpoints due to harsh environment will not have a significant impact to the analysis results.

As described in the SCE reload analysis methodology (Reference 8.4, Section 3.4.2.1.4), the CEA Ejection event performed during each reload cycle is evaluated at various power levels corresponding to the LCS Power Dependent Insertion Limit (PDIL) breakpoints. This PDIL verification methodology will remain unchanged for AREVA fuel. In this section the standard HFP and HZP cases are presented to illustrate the impact of AREVA fuel on this event.

7.4.2.2 CEA Ejection Analysis Results The analysis results for [

] The key input data used in the analysis are presented in Tables 7.4.5 and 7.4.6. The sequences of events are listed in Tables 7.4.7a/7b and 7.4.8a18b.

Page 156 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel I

It was verified that DNB propagation did not occur during the event; therefore, the core coolable geometry is maintained.

The CEA ejection analysis results were compared against the acceptance criteria as shown in Table 7.4.9.

7.4.2.3 VQP CEA Ejection Event Analysis - Conclusion The results of the VQP CEA ejection analysis show that all event acceptance criteria will continue to be met. [

1.

Table 7.4.5 Key Parameters Assumed in the HFP CEA Ejection Analysis Parameter Units Value Power Level MWt 3458 RPS Trip Value  % full power 125 MTC 10-4 Ap/OF 0.0 Ejected CEA Worth %Ap 0.2319 Doppler Weighting Factor (Wr) (Wr) 1.0 Delayed Neutron Fraction 13 0.00517 Post-Ejected 3-D Power Peak, --- 4.2359 Fq CEA Bank Worth at Trip %Ap -6.0 CEA Drop Time Seconds 3.4 Page 157 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 7.4.6 Key Parameters Assumed in the HZP CEA Ejection Analysis Parameter Units Value Core Power Level MWt 1.0 RPS Trip Value  % full power 70 MTC 10-4 Ap/OF +0.5 Ejected CEA Worth %Ap 0.6010 Doppler Weighting Factor (Wr) 2.0272 Delayed Neutron Fraction P 0.00412 Post-ejected 3-D Power Peak, Fq --- 15.6956 CEA Bank Worth at Trip %Ap -4.0 CEA Drop Time Seconds 3.4 Page 158 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel I

]

.4 4 II Page 159 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

[

I I

Page 160 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel Table 7.4.9 Analysis Results for the CEA Ejection Analysis Acceptance Analysis Results Parameter Criteria AREVA Westinghouse Fuel Fuel Control Room Dose, Rem TEDE <5 bounded by USFAR value Offsite Dose, Rem TEDE <6.3 bounded by UFSAR value Page 161 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel

8.0 REFERENCES

8.1 Letter From SCE (Short) to NRC dated January 30, 2009, "Request for Temporary Exemption from the Provisions of 10 CFR 50.46 and 10 CFR 50, Appendix K for Lead Fuel Assemblies... for San Onofre Nuclear Generating Station, Units 2 and 3," (ADAMS Accession Number ML090360738).

8.2 Letter from NRC (Hall) to SCE (Ridenoure) dated December 17, 2009, "SONGS, Units 2 and 3 - Temporary Exemption from the Requirements of 10 CFR Part 50, Section 50.46 and Appendix K for Lead Fuel Assemblies (TAC No. ME0602 and ME0603)," (ADAMS Accession number ML090860415).

8.3 ANP-2839(P), Revision 1, "San Onofre Nuclear Generating Station Lead Fuel Assemblies Fuel Design Criteria Review," September 2009 8.4 Southern California Edison Company (SCE), "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3," Topical Report SCE-9801 -P, November 1998 and approved version Topical Report SCE-9801 -P-A, June 1999.

8.5 Southern California Edison Company (SCE), "PWR Reactor Physics Methodology Using Studsvik Design Codes," Topical Report SCE-0901-A, December 2009.

8.6 Letter from NRC (Hall) to SCE (Ridenoure) dated December 15, 2009, "San Onofre Nuclear Generating Station Units 2 and 3 - Issuance of Amendments Revising Technical Specification 5.7.1.5, Core Operating Limits Report (COLR)"

(TAC Nos. ME0604 and ME0605)" (ADAMS Accession Number ML093220105) 8.7 CENPD-404-P-A,"lmplementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," November 2001.

8.8 BAW-1 0227r(P-A, Rev 0, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," approved by the NRC in February 2000.

This Reference contains the NRC Safety Evaluation and approval letter identified in Reference 8.13.

8.9 BAW-10227(P)-A, Rev 1, "Evaluation of Advanced Cladding and Structural Material (MS ) in PWR Reactor Fuel," approved by the NRC in June 2003. This Reference contains the NRC Safety Evaluation and approval letter identified in Reference 8.13.

8.10 BAW-10240(P)-A, "Incorporation of M5TM Properties in Framatome ANP Approved Methods," approved by the NRC in May 2004.

8.11 BAW-10241(P)(A), Revision 1, "BHTP DNB Correlation Applied with LYNXT,"

approved by the NRC in July 2005.

8.12 XN-NF-82-06 (P)(A), Revision 1 and Supplements 2, 4 and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup."

Page 162 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 8.13 Letter, S. A. Richards (NRC) to T. A. Coleman (Framatome Cogema Fuels),

Revised Safety Evaluation (SE) for Topical Report BAW-1 0227P: "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," February 4, 2000.

8.14 ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," December 1991.

8.15 SONGS Procedure S023-XXXVI-1.4.1, Revision 0, "Documentation of NFM Vendor Qualification Program Analyses."

8.16 CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983 (PROPRIETARY) 8.17 EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs,"

Revision 0, February 1999.

8.18 XN-75-32(P)(A) and Supplements 1, 2, 3 and 4, "Computational Procedure for Evaluating Fuel Rod Bow," February 1983.

8.19 CENPD-161(P)(A), "TORC Code: A Computer Code for Determining Thermal Margin of a Reactor Core," April 1986 8.20 Letter, N. Kalyanum (NRC) to R. M. Rosenblum (SCE), "San Onofre Nuclear Generating Station, Units 2 and 3 - Issuance of Amendments Re: Full-Scope Implementation of an Alternative Source Term (TAC Nos. MC5495 and MC5496)," December 29, 2006 8.21 Letter, N. Kalyanum (NRC) to R. T Ridenoure (SCE), "San Onofre Nuclear Generating Station, Unit 2 And 3 - Request To Revise Safety Evaluation To License Amendment Nos. 210 And 202 (TAC Nos. MD6759 and MD6760),"

September 16, 2008 8.22 Letter, D. Nunn (SCE) to NRC, "San Onofre Nuclear Generating Station, Units 2 and 3 Docket Nos. 50-361 and 50-362 Proposed Change Number (PCN) 555, Alternative Source Term," December 27, 2004 8.23 CENPD-183-A, "CE Methods for Loss of Flow Analysis," June 1984 8.24 CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990 (including SER and Response to NRC Questions on CEN-372-P) 8.25 Letter, NRC-SCE, "Plant-Specific Use Of Topical Report CEN-372-P, Fuel Rod Maximum Allowable Gas Pressure," San Onofre Nuclear Generating Station, Unit Nos. 2 and 3 (TAC Nos. 77123 and 77124)," August 3, 1990 8.26 CEN-356(V)-P-A, Revision 01-P-A, "Modified Statistical Combination of Uncertainties," ABB Combustion Engineering Nuclear Power, May, 1988 8.27 XN-NF-621 (P)(A), Revision 1, "Exxon Nuclear DNB Correlation for PWR Fuel Designs" Page 163 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 8.28 CENPD-1 62-A, Rev. 000, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids: Part 1 Uniform Axial Power Distributions" 8.29 CENPD-207-NP-A, Rev. 000, "Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids: Part 2 Nonuniform Axial Power Distributions" 8.30 CENPD-135-P, "STRIKIN-Il Cylindrical Geometry Fuel Rod Heat Transfer Program," August 1974.

8.31 CENPD-1 90-A, "C-E Method for Control Element Assembly Ejection Analysis,"

7/30/1976.

8.32 CENPD-382-P-A, "Methodology for Core Design Containing Erbium Burnable Absorbers," August 1993.

8.33 WCAP-1 5996-P-A, Rev. 1, "Technical Description Manual for the CENTS Code",

Volumes 1 through 4, dated March 2005 8.34 CEN-193(B), Supplement 2-P, Partial Response to NRC Questions on CEN-161 (B)-P, 'Improvements to Fuel Evaluation Model', dated March 21, 1982 8.35 CENPD-139-P-A, "C-E Fuel Evaluation Model Topical Report," Combustion Engineering Inc., July 1974 8.36 CENPD-139, Supplement 1, Revision 01, "C-E Fuel Evaluation Model Topical Report," Combustion Engineering Inc., September 1974 8.37 CEN-1 61 (B)-P-A, "Improvements to Fuel Evaluation Model," ABB Combustion Engineering Nuclear Fuel, August 1989 8.38 CEN-1 61 (B)-P, Supplement 1-P-A, "Improvements to Fuel Evaluation Model,"

ABB Combustion Engineering Nuclear Fuel, January 1992 8.39 CENPD-387-P-A, Rev. 000, "ABB Critical Heat Flux Correlations for PWR Fuel,"

2000.

8.40 CENPD-199-P, Supplement 2-P, "CE Setpoint Methodology," September 1997.

8.41 NRC Letter from A.C. Thadani to R. Copeland of Advance Nuclear Fuels Corporation, "Acceptance for Referencing of Licensing Topical Report ANF 133(P), Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU (TAC No. M69719)", September 9,1991.

8.42 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-0800. 1 st ed. November 1975, 2 nd ed. March 1980, 3 rd ed. July 1981, U. S. Nuclear Regulatory Commission.

8.43 WCAP-1 6011-P-A, Rev. 0, "Startup Test Activity Reduction Program", February 2005.

Page 164 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 8.44 XN-NF-79-56(P)(A), Revision 1, Supplement 1," "Gadolinia Fuel Properties for LWR Fuel Safety Evaluation" 8.45 Not Used 8.46 CENPD-275-P, Revision 1-P-A,"C-E Methodology for Core Designs Containing Gadolinia-Urania Burnable Absorbers," May 1988.

8.47 CENPD-275-P, Revision 1-P, Supplement 1-P-A, "C-E Methodology for PWR Core Designs Containing Gadolinia-Urania Burnable Absorbers," April 1999.

8.48 AREVA NP Inc. Letter to NRC, "Typographical Error in BAW-1 0240(P)(A),

Revision 0, 'Incorporation of M5TM Properties in Framatome ANP Approved Methods', May 2004," January 31, 2011 (non-proprietary version available in NRC ADAMS Database as Accession Number ML110320443) 8.49 INPO ACAD 98-004, "Guidelines for Training and Qualification of Engineering Personnel", June 2003.

8.50 LD-82-001 (dated 1/6/82) "CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System", Enclosure 1-P to letter from A. E.

Scherer to D. G Eisenhut, December 1981.

8.51 CENPD-188-A. "A Multi-Dimensional Space-Time Kinetics Code for PWR Transients", July 1976.

8.52 NRC Information Notice (IN) 2009-23, "Nuclear Fuel Thermal Conductivity Degradation," October 8, 2009.

8.53 Topical Report WCAP-1 6500-NP-A, "CE 16x16 Next Generation Fuel Core Reference Report," August 2007 8.54 NRC Letter, "Final Safety Evaluation for Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-16500-P, Revision 0, "CE (Combustion Engineering) 16x16 Next Generation Fuel (NGF) Core Reference Report," July 30, 2007.

8.55 Westinghouse Letter LTR-NRC-06-66 (non-Proprietary) to U.S. NRC, "Response to NRC's Request for Additional Information By the Office of Nuclear Reactor Regulation Topical Report WCAP-1 6500-P, "CE 16x1 6 Next Generation Fuel Core Reference Report," November 29, 2006.

8.56 CEN-386-P-A, Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16x1 6 PWR Fuel, August 1992.

8.57 XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," November 1986 8.58 EMF-2103(P)(A), Revision 0, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors, April 2003."

Page 165 of 166 Enclosure 5 to SONGS PCN 600

LICENSEE'S EVALUATION SONGS PCN 600 Request for Unrestricted Use of AREVA Fuel 8.59 EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based."

8.60 Document No. XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," March 1984.

8.61 Document No. XN-NF-81-58(P)(A) Revision 2 and Supplements 3 and 4, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," April 1990.

8.62 Letter from R.A.Copeland (Siemens) to R.C.Jones (NRC), "Gadolinia Bearing Fuel Rod Design Methodology" dated May 13, 1992.

Page 166 of 166 Enclosure 5 to SONGS PCN 600

ATTACHMENTS A. Listing of Acronyms

Attachment 7.A Listing of Acronyms LISTING OF ACRONYMS Acronym Meaning 2-D Two-Dimensional 3-D Three-Dimensional ASEA Brown-Boveri/Combustion Engineering, now ABB/CE Westinghouse ADAMS NRC Agencywide Document Access Management System ARO All Rods Out ANF Advanced Nuclear Fuels, now AREVA AREVA, a nuclear fuel and services provider. AREVA has combined with other corporate entities to form the current company. Included in these combinations and acquisitions, AREVA companies formerly known as:

Exxon Nuclear Corporation - ENC Framatome-ANP - FANP Advanced Nuclear Fuels - ANF Siemens Power Corporation - SPC AST Alternative Source Term BOC Beginning Of Cycle BWR Boiling Water Reactor C-3 CASMO-3 C-3/S-3 CASMO-3 / SIMULATE3 C-4 CASMO-4 C-4/S-3 CASMO-4 / SIMULATE-3 cal/gm Calories per gram.

CBC Critical Boron Concentration CCFL Counter-Current Flow

_CE Combustion Engineering, now Westinghouse CEA Control Element Assembly (Control Rod)

CFR US Code of Federal Regulations CHF Critical Heat Flux COLR Core Operating Limits Report COLSS Core Operating Limit Supervisory System CPC(S) Core Protection Calculator (System)

CPS Core Protection System DBA Design Basis Accident DC Downcomer - Down flow coolant pathway outside core barrel DEGB Double Ended Guillotine Break DNB Departure from Nucleate Boiling DNBR DNB Ratio EARO Essentially All Rods Out ECCS Emergency Core Cooling System EFPH Effective Full Power Hours Page 1 of 3 Attachment 7.A to Enclosure 5 of SONGS PCN 600

Attachment 7.A Listing of Acronyms Acronym Meaning ENC Exxon Nuclear Corporation, Now AREVA EOC End Of Cycle FAH Nuclear Enthalpy Rise Hot Channel Factor Fq, Fq Power Distribution Total Peaking Factor Fr Power Distribution Integrated Peaking Factor Fxy Power Distribution Planar Peaking Factor FCE Facility Change Evaluation FP Fuel Performance FANP Framatome - ANP, Now AREVA GDC General Design Criterion (or Criteria)

GWD/MTU, GWd/MTU Gigawatt-Day(s) per Metric Ton Uranium HFP Hot Full Power HMP High Mechanical Performance HPSI High Pressure Safety Injection HTP Hi h Thermal Performance HZP Hot Zero Power kw/ft Kilowatt per foot LAR License Amendment Request LBLOCA Large Break LOCA LCS Licensee Controlled Specifications LCO Limiting Condition of Operation LFA Lead Fuel Assembly LEF Lower End Fitting LHR Linear Heat Rate LOCA Loss of Coolant Accident LOOP Loss of Offsite Power LPD Local Power Density, as in LPD Trip MSCU Modified Statistical Combination of Uncertainties MSLB Main Steam Line Break MWD/MTU, MWd/MTU Megawatt-Day(s) per Metric Ton Uranium MWt, MWth Megawatt(s) Thermal NECP Nuclear Engineering Change Package NCLO No-Clad-Lift-Off NFM SCE Nuclear Fuel Management NRC, USNRC United States Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OD Outer Diameter PCM Percent Mil (Unit of Reactivity = 10`5 Ak/k)

PCN Proposed Change Number PCT Peak Cladding Temperature pdf or p.d.f. fuel failure Probability Distribution Function Page 2 of 3 Attachment 7.A to Enclosure 5 of SONGS PCN 600

Attachment 7.A Listing of Acronyms Acronym Meaning PDIL Power Dependent Insertion Limit PLHGR Planar Linear Heat Generation Rate PLHR Peak Linear Heat Rate PMS Plant Monitoring System PSV Pressurizer Safety Valve PTM power-to-centerline melt PWR Pressurized Water Reactor RAI, RAIs Request(s) for Additional Information RCP Reactor Coolant Pump RCS Reactor Coolant System RDB Reload Data Block RLBLOCA Realistic LBLOCA RLBLOCA EM RLBLOCA Evaluation Model ROPM Required Over Power Margin RPS Reactor Protection System RWST Refueling Water Storage Tank SAFDL Specified Acceptable Fuel Design Limits SBLOCA Small Break LOCA SCE Southern California Edison Company SCU Statistical Combination of Uncertainties SE, SER US NRC Safety Evaluation, Safety Evaluation Report SFP Spent Fuel Pool SIAS Safety Injection Actuation Signal SL Safety Limit SP, SP's Set Point(s)

SPC Siemens Power Corporation, now AREVA SONGS San Onofre Nuclear Generating Station STAR Startup Test Activity Reduction TBD To Be Determined TEDE Total Effective Dose Equivalent TER Technical Evaluation Report TH Thermal Hydraulic TS Technical Specifications UFSAR Updated Final Safety Analysis Report VOPT Variable Over Power Trip VQP Vendor Qualification Program V&V Verification and Validation Page 3 of 3 Attachment 7.A to Enclosure 5 of SONGS PCN 600

ATTACHMENTS B. Technical Specification and Bases Changes

Attachment B.1 (Existing Pages)

SONGS Unit 2

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at > 1.31.

2.1.1.2 In MODES I and 2, peak fuel centerline temperature shall be maintained at < 5080'F, decreasing by 58°F per 10,000 MWD/MTU and adjusted for burnable poison per CENPD-382-P-A.

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at < 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE I or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

(continued)

SAN ONOFRE--UNIT 2 2.0-1 Amendment No. 207

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Exclusion Area Boundary The exclusion area boundary shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone (LPZ)

The LPZ shall be as shown in Figure 4.1-2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 217 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLOTM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Integral or Discrete Burnable Absorber Rods may be used. They may include:

borosilicate glass - Na20-B 20 3-Si0 2 components, boron carbide -

B4C, zirconium boride - ZrB2, gadolinium oxide - Gd203 , erbium oxide - Er 20 3 . Limited substitutions of zirconium alloy (such as ZIRLOTM or Zircaloy) or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 83 full length and eight part length control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, and inconel as approved by the NRC.

(continued)

SAN ONOFRE--UNIT 2 4.0-1 Amendment No. +H--199

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2. CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3. CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4. SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5. CEN-635(S), "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6. Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-1O and Amendment No. 36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3 SER)

7. Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-10 and Amendment No. 19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8. "Imýlementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs,"' CENPD-404-P-A

9. SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes"

c. The core operating limits shall be determined assuming operation at RATED THERMAL POWER such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(continued)

SAN ONOFRE--UNIT 2 5.0-27 Amendment No. 222

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.6 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

Technical Specification 3.4.3 RCS Pressure and Temperature (P/T) Limits, Technical Specification 3.4.6 RCS Loops - MODE 4, Technical Specification 3.4.7 RCS Loops - MODE 5, Loops Filled, Technical Specification 3.4.12.1 Low Temperature Overpressure Protection (LTOP) System RCS Temperature PTLR Limit, and Technical Specification 3.4.12.2 Low Temperature Overpressure Protection (LTOP) System RCS Temperature > PTLR Limit.

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

CE NPSD-683-A, The Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Setpoints from the Technical Specifications.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (1-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

(continued)

SAN ONOFRE--UNIT 2 5. 0-27a Amendment No. 203 1

Attachment B.2 (Existing Pages)

SONGS Unit 3

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES I and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at Ž 1.31.

2.1.1.2 In MODES I and 2, peak fuel centerline temperature shall be maintained at < 5080 F, decreasing by 58 F per 10,000.

MWD/MTU and adjusted for burnable poison per CENPD-382-P-A.

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

(continued)

SAN ONOFRE--UNIT 3 2.0-1 Amendment No. 199 1

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Exclusion Area Boundary The exclusion area boundary shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone (LPZ)

The LPZ shall be as shown in Figure 4.1-2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 217 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLOTM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Integral or Discrete Burnable Absorber Rods may be used. They may include:

borosilicate glass - Na20-B 2 03-Si0 2 components, boron carbide -

B4C, zirconium boride - ZrB2, gadolinium oxide - Gd20 3 , erbium oxide - Er 20 3 . Limited substitutions of zirconium alloy (such as ZIRLOTM or Zircaloy) or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 83 full length and eight part length control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, and inconel as approved by the NRC.

(continued)

SAN ONOFRE--UNIT 3 4.0-1 Amendment No. 1+6-,190

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2. CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3. CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4. SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5. CEN-635(S), "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6. Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-1O and Amendment No. 36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3 SER)

7. Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-1O and Amendment No. ig to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8. "Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A

9. SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes"

c. The core operating limits shall be determined assuming operation at RATED THERMAL POWER such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuc'lear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

(continued)

SAN ONOFRE--UNIT 3 5.0-27 Amendment No. 215

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.6 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

Technical Specification 3.4.3 RCS Pressure and Temperature (P/T) Limits, Technical Specification 3.4.6 RCS Loops - MODE 4, Technical Specification 3.4.7 RCS Loops - MODE 5, Loops Filled, Technical Specification 3.4.12.1 Low Temperature Overpressure Protection (LTOP) System RCS Temperature PTLR Limit, and Technical Specification 3.4.12.2 Low Temperature Overpressure Protection (LTOP) System RCS Temperature > PTLR Limit.

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

CE NPSD-683-A, The Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Setpoints from the Technical Specifications.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (1-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

(continued)

SAN ONOFRE--UNIT 3 5.0-27a Amendment No. 195 1

Attachment B.3 (Proposed Pages)

(Redline and Strikeout)

SONGS Unit 2

SLs 2.C 2.0 SAFETY LIMITS (51) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at 1.31.

2.1.1.2 In MODES 1 and 2, peak fuel centerline temperature shall be maintained at < 5080 F, decreasing by 58 F per 10,000 MWD/MTU and adjusted for burnable poison per CENPD-382-P-AM or per CENPD-275-P-A including Supplement 1-P-A.

2.1.2 Reactor Coolant System (RCS) Pressure SL in MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within I hour.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

(continued)

SAN ONOFRE-UNIT 2 2.o0-1 S~OOFR--IITNo. XxX.

22.-1A~r-endmý,ent

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Exclusion Area Boundary The exclusion area bcundary shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone (LPZ)

The LPZ shall be as shown in Figure 4.1-2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 2117 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, or ZIRLOTm or M5-' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UW) as fuel material. Integral or Discrete Burnable Absorber Rods may be used. They may include:

borosilicate glass - Na2O-B*O 3 -SiO 2 components, boron carbide - B4C, zirconium boride - ZrB2 , gadolinium oxide - Gd203 , erbium oxide -

Er 203 . Limited substitutions of zirconium alloy (such as ZIRLOT4 5r Zircaloy or M57) or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 83 full length and eight part length control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, and inconel as approved by the NRC.

(continued)

SAN ONOFRE-UNIT 2 4 .0-i Amendment No. -- 1-5 X0,.

I

Reporting Requirements

..7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2. CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3. CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4. SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5. CEN-635(S), "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodol ogy"

6. Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-1O and Amendment No. 36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3 SER)

7. Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-1O and Amendment No. 19 to Facility Operating License NPF-15 " San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8. "Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A

9. SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes"

10. EMF-2103(P)(A), "Realistic Larqe Break LOCA Methodology."

i. EMF-2328(P)(A)* "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based."

12. Letter dated , , , (NRC)to (SCE),

"Issuance of Amendments for Unit 2 and for Unit 3, Request for Unrestricted Use of AREVA fuel, Sar Onofre Nuclear Generatijn Statiýn."

13. BAW-10240(P)-3/4, 'Incorpora[ion o1f N.5" Prooerýties in Framatome ANP Approved Methods."

(continued)

SAN ONOFRE--UNIT 2 5.0-27 Amendment No. 4-+ XXX

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7*1.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined assuming operation at RATED THERMAL POWER such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Technical Specification 3.4.3 RCS Pressure and Temperature (P/T) Limits, Technical Specification 3.4.6 RCS Loops - MODE 4, Technical Specification 3.4.7 RCS Loops - MODE 5, Loops Filled, Technical Specification 3.4.12.1 Low Temperature Overpressure Protection (LTOP) System RCS Temperature PTLR Limit, and Technical Specification 3.4.12.2 Low Temperature Overpressure Protection (LTOP) System RCS Temperature > PTLR Limit.

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

CE NPSD-683-A, The Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Setpoints from the Technical Specifications.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (1-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

(continued)

SAN ONOFRE- _UNTIT 2 5.0-2 7a AmendmeNt No. £2.&3 XXX I

Attachment B.4 (Proposed Pages)

(Redline and Strikeout)

SONGS Unit 3

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at Ž3 1.31.

2.1.1.2 In MODES 1 and 2, peak fuel centerline temperature shall be maintained at < 5080 F, decreasing by 58 F per 10,000 MWD/MTU and adjusted for burnable poison per CENPD-382-P-AT or cer CLEPD-275-P-A including Supplement I-P-A.

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at 2750 psia.

2.2 SL Violations 2.2.1 if SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

(continrued)

SAN ONOFRE--UNIT 3 2.0-1 Amendment No. 0i XXX

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Exclusion Area Boundary The exclusion area boundary shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone (LPZ)

The LPZ shall be as shown in Figure 4.1-2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 217 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, &r ZIRLOTM or M5-' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02 ) as fuel material. Integral or Discrete Burnable Absorber Rods may be used. They may include:

borosilicate glass - Na2O-B2 0 3-SiO2 components, boron carbide - B4C, zirconium boride - ZrB2, gadolinium oxide - Gd'0 3 , erbium oxide -

Er2 0 3 . Limited substitutions of zirconium alloy (such as ZIRLOTML

" Zircaloy or M5 7) or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 83 full length and eight part length control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, and inconel as approved by the NRC.

(conti nued)

SAN ONOFFRE--UNIT 3 4.0-! Amendment No. iý-4 XXX

Reporting Requirements 5.7 7.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 1i CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2. CENPD-137P, "'Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3. CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4. SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5. CEN-635(S), "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6. Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-1O and Amendment No. 36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3 SER)

7. Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-1O and Amendment No. 19 to Facility Operating License NPF-15 " San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8. "Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A

9. SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes"

10. EMF-2103(P)(A). "Realistic Large Break LOCA Methodology."

i1. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based."

12. Letter dated , , , (NRC)to _SCE),

"Issuance of Amendments for Unit 2 and for Unit 3, Request for Unrestricted Use of AREVA fuel. San Onofre Nuclear Generating Station."

Is. BAW-10240(P)-A. "Incorporation of M5 Properties in Framatome ANP Anproved Methods."

conti nued)

SAN ONOFREE-UNIT 5.0-27 SAN Oi0FR--UNI 3 50-27imendment No. -5XAXX

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

Co The core operating limits shall be determined assuming operation at RATED THERMAL POWER such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Technical Specification 3.4.3 RCS Pressure and Temperature (P/T) Limits, Technical Specification 3.4.6 RCS Loops - MODE 4, Technical Specification 3.4.7 RCS Loops - MODE 5, Loops Filled, Technical Specification 3.4.12.1 Low Temperature Overpressure Protection (LTOP) System RCS Temperature PTLR Limit, and Technical Specification 3.4.12.2 Low Temperature Overpressure Protection (LTOP) System RCS Temperature > PTLR Limit.

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

CE NPSD-683-A, The Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Setpoints from the Technical Specifications.

C. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (1-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional (continued)

SAN ONOFRE--UNIT 3 5.0-27a No. +q- XXX SANONORE-U~i 3 .O-aAimer-ndment

Attachment B.5 (Proposed Pages)

SONGS Unit 2

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at 1.31.

2.1.1.2 In MODES I and 2, peak fuel centerline temperature shall be maintained at < 5080 F, decreasing by 58 F per 10,000.

MWD/MTU and adjusted for burnable poison per CENPD-382-P-A or per CENPD-275-P-A including Supplement I-P-A.

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within I hour.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

(continued)

SAN ONOFRE--UNIT 2 2.0-1 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Exclusion Area Boundary The exclusion area boundary shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone (LPZ)

The LPZ shall be as shown in Figure 4.1-2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 217 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLOTM or M5TM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02 ) as fuel material. Integral or Discrete Burnable Absorber Rods may be used. They may include:

borosilicate glass - Na20-B 203 -Si0 2 components, boron carbide -

B4C, zirconium boride - ZrB2, gadolinium oxide - Gd20 3 , erbium oxide - Er 203 . Limited substitutions of zirconium alloy (such as ZIRLOTM , Zircaloy or M5TM) or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 83 full length and eight part length control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, and inconel as approved by the NRC.

(continued)

SAN ONOFRE--UNiT 2 4.0-1 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2. CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3. CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4. SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units and 3"

5. CEN-635(S), "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6. Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-1O and Amendment No. 36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3 SER)

7. Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-10 and Amendment No. 19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8. "Implementation of ZIRLOTM Cladding Material in CE Nuc ear Power Fuel Assembly Designs," CENPD-404-P-A

9. SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes"

10. EMF-2103(P)(A), "Realistic Large Break LOCA Methodology."

11. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based."

12. Letter dated , , , (NRC)to (SCE),

"Issuance of Amendments for Unit 2 and for Unit 3, Request for Unrestricted Use of AREVA fuel, San Onofre Nuclear Generating Station."

13. BAW-10240(P)-A, "Incorporation of M5TM Properties in Framatome ANP Approved Methods."

(continued)

SAN ONOFRE--UNIT 2 5.0-27 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined assuming operation at RATED THERMAL POWER such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

a. RCS pressure and temperature limits for heatup, cooldown, low tem erature operation, criticality, and hydrostatic testing as wel as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Technical Specification 3.4.3 RCS Pressure and Temperature (P/T) Limits, Technical Specification 3.4.6 RCS Loops - MODE 4, Technical Specification 3.4.7 RCS Loops - MODE 5, Loops Filled, Technical Specification 3.4.12.1 Low Temperature Overpressure Protection (LTOP) System RCS Temperature PTLR Limit, and Technical Specification 3.4.12.2 Low Temperature Overpressure Protection (LTOP) System RCS Temperature > PTLR Limit.

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

CE NPSD-683-A, The Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Setpoints from the Technical Specifications.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (1-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

(continued)

SAN ONOFRE--UNIT 2 5.0-27a Amendment No.

Attachment B.6 (Proposed Pages)

SONGS Unit 3

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, departure from nucleate boiling ratio (DNBR) shall be maintained at Ž 1.31.

2.1.1.2 In MODES 1 and 2, peak fuel centerline temperature shall be maintained at < 5080 F, decreasing by 58 F per 10,000 MWD/MTU and adjusted for burnable poison per CENPD-382-P-A or per CENPD-275-P-A including Supplement 1-P-A.

2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained at 2750 psia.

2.2 SL Violations 2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

(continued)

SAN ONOFRE--UNIT 3 2.0-1 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site 4.1.1 Exclusion Area Boundary The exclusion area boundary shall be as shown in Figure 4.1-1.

4.1.2 Low Population Zone (LPZ)

The LPZ shall be as shown in Figure 4.1-2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 217 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLOTM or M5TM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02 ) as fuel material. Integral or Discrete Burnable Absorber Rods may be used. They may include:

borosilicate glass - Na20-B 20 3-Si0 2 components, boron carbide -

B4C, zirconium boride - ZrB2, gadolinium oxide - Gd203 , erbium oxide - Er 20 3 . Limited substitutions of zirconium alloy (such as ZIRLOTM, Zircaloy or M5TM) or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 83 full length and eight part length control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, and inconel as approved by the NRC.

(continued)

SAN ONOFRE--UNIT 3 4.0-1 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model"

2. CENPD-137P, "Calculative Methods for the C-E Small Break LOCA Evaluation Model"

3. CEN-356(V)-P-A, "Modified Statistical Combination of Uncertainties"

4. SCE-9801-P-A, "Reload Analysis Methodology for the San Onofre Nuclear Generating Station Units 2 and 3"

5. CEN-635(S), "Identification of NRC Safety Evaluation Report Limitations and/or Constraints on Reload Analysis Methodology"

6. Letter, dated May 16, 1986, G. W. Knighton (NRC) to K.

P. Baskin (SCE), "Issuance of Amendment No. 47 to Facility Operating License NPF-1O and Amendment No. 36 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 3 SER)

7. Letter, dated January 9, 1985, G. W. Knighton (NRC) to K. P. Baskin, "Issuance of Amendment No. 30 to Facility Operating License NPF-1O and Amendment No. 19 to Facility Operating License NPF-15," San Onofre Nuclear Generating Station Units 2 and 3 (Cycle 2 SER)

8. "Implementation of ZIRLOTM Cladding Material in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A

9. SCE-0901, "PWR Reactor Physics Methodology Using Studsvik Design Codes"

10. EMF-2103(P)(A), "Realistic Large Break LOCA Methodology."

11. EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based."

12. Letter dated , , , (NRC)to (SCE),

"Issuance of Amendments for Unit 2 and for Unit 3, Request for Unrestricted Use of AREVA fuel, San Onofre Nuclear Generating Station."

13. BAW-10240(P)-A, "Incorporation of M5TM Properties in Framatome ANP Approved Methods."

(continued)

SAN ONOFRE--UNIT 3 5.0-27 Amendment No.

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined assuming operation at RATED THERMAL POWER such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.7.1.6 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

Technical Specification 3.4.3 RCS Pressure and Temperature (P/T) Limits, Technical Specification 3.4.6 RCS Loops - MODE 4, Technical Specification 3.4.7 RCS Loops - MODE 5, Loops Filled, Technical Specification 3.4.12.1 Low Temperature Overpressure Protection (LTOP) System RCS Temperature PTLR Limit, and Technical Specification 3.4.12.2 Low Temperature Overpressure Protection (LTOP) System RCS Temperature > PTLR Limit.

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

CE NPSD-683-A, The Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Setpoints from the Technical Specifications.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.7.1.7 Hazardous Cargo Traffic Report Hazardous cargo traffic on Interstate 5 (1-5) and the AT&SF railway shall be monitored and the results submitted to the NRC Regional Administrator once every three years.

(continued)

SAN ONOFRE--UNIT 3 5.0-27a Amendment No.

Attachment B.7 (Proposed Bases Pages - for information only)

SONGS Unit 2

Reactor Core SLs B2.1.1 BASES APPLICABLE h Local Power Density-High trip; SAFETY ANALYSES (continued) i. DNBR-Low trip;

j. Reactor Coolant Flow-Low trip; and

k. Steam Generator Safety Valves.

The SL represents a design requirement for establishing the protection system trip setpoint allowable values identified previously. LCO 3.2.1, "Linear Heat Rate (LHR)," and LCO 3.2.4, "Departure From Nucleate Boiling Ratio (DNBR),"

or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

SAFETY LIMITS SL 2.1.1.1 and SL 2.1.1.2 ensure that the minimum DNBR is not less than the safety analyses limit and that fuel centerline temperature remains below melting.

The minimum value of the DNBR during normal operation and design basis AO0s is limited to 1.31, based on a statistical combination of CE-i CHF correlation and engineering factor uncertainties, and is established as an SL. Additional factors such as rod bow and spacer grid size and placement will determine the limiting safety system settings required to ensure that the SL is maintained.

A steady state peak linear heat rate of 21 KW/ft has been established as the Limiting Safety System Setting to prevent fuel centerline melting during normal steady state operation. Following design basis anticipated operational occurrences, the transient linear heat rate may exceed 21 KW/ft provided the fuel centerline melt temperature is not exceeded.

The design melting point of new fuel with no burnable poison is 5080 F. The melting point is adjusted downward from this temperature depending on the amount of burnup and amount and type of burnable poison in the fuel. The 58 F per 10,000 MWD/MTU adjustment for burnup was accepted by the NRC in Topical Report CEN-386-P-A, Reference 3. Adjustments for burnable poisons are established based on NRC approved Topical Report CENPD-382-P-A, (Reference 4) or per CENPD-275-P-A includina Suo)Ieent 1-P-A (Peferences 5 and 6).

(continued)

XIXX "XX SAN ONOFRE--UNIT 2 B 2.0-3 Amendment No. 207 1-2e%

Reactor Core SLs B 212 BASES (continued)

SAFETY LIMIT The following violation responses are applicable to the VIOLATIONS reactor core SLs.

2.2.1 If SL 2.1.1.1 or SL 2.1.1.2 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing the unit to a MODE where this SL is not applicable and reduces the probability of fuel damage.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. UFSAR, Section 15.0.3.2, "Initial Conditions,"

3. CEN-386-P-A, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/MTU for Combustion Engineering 16x06 PWR Fuel," August 1992.

4. CENPD-382-P-A, "Methodology for Core Designs Containing Erbium Burnable Absorbers," August 1993.

CENPD-275-P. Revision I-P-A, "C-E Methodoloay for Core Desicns Containinq Gadolinia-Urania Burnable Absorbers," May 1998.

CENPD-275-P, Revision I-P, Suppleyent I-P-A, "C-E Methodology for PIR Core designs Containing Gadolinia-Urania Burnable Absorbers." Anril 1999.

(continued)

\XX,xlx SAN ONOFRE--UNIT 2 B 2.0-4 Amendment No. 207 I4244+je -

ATTACHMENTS C. SONGS - Summary of Impact on UFSAR Chapter 15 Events

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events Summary of Impacts on Other UFSAR Chapter 15 Non-LOCA Events Discussion of Chapter 15 Events The Attachment summarizes the impact of the change to AREVA fuel on the SONGS accident analyses presented in UFSAR Chapter 15.

Table C.1 shows the impact of the changes associated with the use of AREVA fuel on all UFSAR Chapter 15 events.

An assessment was performed of all UFSAR safety analysis events to determine the impact of the proposed use of AREVA fuel on the event and its acceptance criteria.

Table C.1 shows the acceptance criteria for all relevant UFSAR Chapter 15 events included in the assessment, and the results of the assessment.

The Table C.1 assessment was based on the following consideration:

1) Impact of AREVA fuel on non-LOCA events input

a. Impact of AREVA fuel on Physics Inputs:

Since non-LOCA simulation of transients is based on a point-kinetics model, physics provides input that represents a given core based on some global physics parameters such as power distributions and various reactivity rates.

Since all of these input parameters are based on the fuel management and the fuel management is controlled though the fuel management guidelines, there is no direct impact from AREVA fuel on these parameters and they become transparent to non-LOCA safety analyses.

b. Impact of AREVA fuel on TH input The TH section of the LAR has considered the impact of the AREVA fuel on the TH parameters. From a non-LOCA perspective this impact comes through

[ ]. Explicit analysis has been performed and discussed in Table C.1 to address this impact.

c. Impact of AREVA fuel on Core Operating Limits Supervisory System/Core Protection Calculator (COLSS/CPCS) setpoints The use of AREVA fuel has not necessitated any change on the COLSS/CPCS setpoints (e.g., VOPT parameters). The COLSS/CPCS and reactor protection system setpoints, as discussed in the SONGS Reload Topical (Reference 8.4),

ensure that DNBR and linear heat rate are maintained within allowable limits as approved for SONGS by the NRC.

Page 1 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events

2) Impact of AREVA fuel on the transient dynamics Again since the non-LOCA transient simulations are based on a point-kinetic model, the fuel type does not impact the transient dynamics (i.e., for a loss of flow event, the flow degradation is not a function of the fuel type).

3) Impact of AREVA fuel on plant systems that mitigate the non-LOCA events The use of AREVA fuel has not necessitated any change to plant systems that mitigate the non-LOCA events If the nature of the transient and its consequences are such that it cannot be assessed solely based on the above consideration, an explicit analysis is performed to assess the impact of the AREVA fuel on the transient (e.g., CEA Ejection).

Page 2 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events Chapter 15 Event Impact Evaluation Table Table C.1 IMPACT OF THE USE OF AREVA FUEL ON UFSAR CHAPTER 15 ACCIDENT ANALYSES UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION 15.1 Increase in Heat Removal by the Secondary System 15.1.1.1 Decrease in Peak RCS Pressure <110% of Peak pressures and fuel Feedwater Design performance are bounded by Temperature Peak Secondary Pressure <110% Increased Main Steam Flow (DFWT) of Design (Section 15.1.1.3).

No Fuel Failure (Minimum DNBR Doses bounded by

>1.31 and Peak LHR <21 kw/ft (a)) IOSGADV.

Offsite Doses < 0.5 Rem TEDE; Control Room Dose < GDC 19 15.1.1.2 Increase in Peak RCS Pressure <110% of Peak pressures and fuel Feedwater Design performance are bounded by Flow(IFF) Peak Secondary- Pressure 110% Increased Main Steam Flow of Design (Section 15.1.1.3).

No Fuel Failure (Minimum DNBR Doses bounded by

>1.31 and Peak LHR 521 kw/ft (a)) IOSGADV Offsite Doses < 0.5 Rem TEDE; Control Room Dose < GDC 19 15.1.1.3 Increased Main Peak RCS Pressure _110% of Peak pressure criteria are Steam Flow (IMSF) Design not challenged for this event.

Peak Secondary Pressure *5110% Appropriate Core Operating of Design Limits Supervisory System and Core Protection No Fuel Failure (Minimum DNBR Calculator setpoints ensure

_>1.31 and Peak LHR *21 kw/ft (a)) fuel failure does not occur.

Offsite Doses < 0.5 Rem TEDE; Control Room Dose < GDC 19 Doses bounded by IOSGADV 15.1.1.4 Inadvertent Opening Peak RCS Pressure _110% of Peak pressures and fuel of a Steam Design performance are bounded by Generator Peak Secondary Pressure <110% Increased Main Steam Flow Atmospheric Dump of Design (Section 15.1.1.3)

Valve (IOSGADV) ofDsg No Fuel Failure (Minimum DNBR Since this event does not fail

-1.31 and Peak LHR <21 kw/ft (a))

Page 3 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION Offsite Doses < 0.5 Rem TEDE; fuel, the dose is based on Control Room Dose < GDC 19 Technical Specification activity limits, and is independent of the fuel in the core.

15.1.2.1 Decrease in Peak RCS Pressure _1 10% of Bounded by Increased Main Feedwater Design Steam Flow with Single Temperature with a Peak Secondary Pressure _<110% Failure (Section 15.1.2.3).

Concurrent Single of Design Failure of an Active Doses bounded by Component Maintain coolable geometry IMSF+SF.

(DFWT+SF) Offsite Doses _25% of 10CFR50.67; Control Room Doses - GDC 19 15.1.2.2 Increase in Peak RCS Pressure <1 10% of Bounded by Increased Main Feedwater Flow Design Steam Flow with Single with a Concurrent Peak Secondary Pressure _1 10% Failure (Section 15.1.2.3).

Single Failure of an of Design Active Component Doses bounded by (IFF+SF) Maintain coolable geometry IMSF+SF.

Offsite Doses _25% of 10CFR50.67; Control Room Doses - GDC 19 15.1.2.3 Increased Main Peak RCS Pressure -1 10% of Peak pressure and peak Steam Flow with a Design LHR criteria are not Concurrent Single Peak Secondary Pressure <1 10% challenged for this event.

Failure of an Active of Design Component For purposes of DNB IMSF+SF) Maintain coolable geometry propagation analysis Maintain Peak LHR <21 kw/ft(a) (Section 7.4.1), [

Page 4 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION Offsite Doses <25% of 10CFR50.67; Control Room Doses < GDC 19 ]. As such, introduction of AREVA fuel does not challenge maintaining coolable geometry.

As shown in the Pre-Trip Steam Line Break (Section 7.4.1), [

]. As such, doses currently reported in UFSAR remain bounding.

15.1.2.4 Inadvertent Opening Peak RCS Pressure _110% of Peak pressures and fuel of a Steam Design performance bounded by Generator Peak Secondary Pressure <110% Increased Main Steam Flow Atmospheric Dump of Design with Single Failure (Section Valve with a 15.1.2.3).

Concurrent Single Maintain coolable geometry Failure of an Active Offsite Doses <25% of Since this event does not fail Component 10CFR50.67; fuel, the dose is based on (IOSGADV+SF) Control Room Doses - GDC 19 Technical Specification activity limits, and is independent of the fuel in the core.

15.1.3.1.1 Steam System Maintain coolable geometry See SONGS re-analysis of Piping Failures (Pre- Offsite Doses -100% of pre-trip steam line break trip power 10CFR50.67; assuming AREVA fuel excursion) Control Room Doses _ GDC 19 (Section 7.4.1).

15.1.3.1.2 Steam System Maintain coolable geometry For purposes of DNB Page 5 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION Piping Failures Offsite Doses < 10% of propagation analysis, [

(Post-trip return to 10CFR50.67 (With pre-existing power) Iodine spike the limit is 100% of 10CFR50.67);

Control Room Doses < GDC 19 As such, introduction of AREVA fuel does not challenge maintaining coolable geometry.

Since this event does not fail fuel, the dose is based on Technical Specification activity limits, and is independent of the fuel in the core.

15.2 Decrease in Heat Removal by the Secondary System 15.2.1.1 Loss of External Peak RCS Pressure <110% of Bounded by Loss of Load Design Condenser Vacuum (Section (LOL) Peak Secondary Pressure *110% 15.2.1.3).

of Design 15.2.1.2 Turbine Trip (TT) Peak RCS Pressure _110% of Bounded by Loss of Design Condenser Vacuum Peak Secondary Pressure *110% (Section 15.2.1.3).

of Design 15.2.1.3 Loss of Condenser Peak RCS Pressure _110% of No impact with use of Vacuum (LOCV) Design AREVA fuel, as this is a Peak Secondary Pressure is system response event not

<110% of Design impacted by fuel type.

Offsite Doses < 0.5 Rem TEDE; Doses bounded by Control Room Dose - GDC 19 IOSGADV 15.2.1.4 Loss of Normal AC Peak RCS Pressure _110% of Peak pressures bounded by Power (LONAC) Design Loss of Condenser Vacuum Peak Secondary Pressure _<110% (15.2.1.3) and Fuel Failure of Design criteria are not challenged by this event.

No Fuel Failure (Minimum DNBR

>1.31 and Peak LHR <21 kw/ft(a)) Adequate heat sink analysis Adequate SG inventory to is not impacted by fuel type.

maintain adequate heat sink Doses bounded by Offsite Doses < 0.5 Rem TEDE; IOSGADV.

Control Room Dose - GDC 19 Page 6 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION 15.2.2.1 Loss of External Peak RCS Pressure <110% of Bounded by Loss of Load with a Design Condenser Vacuum with Concurrent Single Peak Secondary Pressure *110% single failure (Section Failure of an Active of Design 15.2.2.3).

Component (LOL+SF) 15.2.2.2 Turbine Trip with a Peak RCS Pressure *5110% of Bounded by Loss of Concurrent Single Design Condenser Vacuum with Failure of an Active Peak Secondary Pressure *110% Concurrent Single Failure of Component of Design an Active Component (TT+SF) (Section 15.2.2.3).

15.2.2.3 Loss of Condenser Peak RCS Pressure -110% of No impact with use of Vacuum with a Design AREVA fuel, as this is a Concurrent Single Peak Secondary Pressure -5110% system response event not Failure of an Active of Design impacted by fuel type.

Component (LOCV+SF) 15.2.2.4 Loss of all Normal Peak RCS Pressure <110% of Peak pressures bounded by AC Power with a Design Loss of Condenser Vacuum Concurrent Single ak Secondary Pressure *110% with a Concurrent Single Failure of an Active of Design Failure of an Active Component Component (Section (LONAC+SF) Adequate SG inventory to 15.2.2.3). Adequate heat maintain adequate heat sink sink analysis is not impacted No Fuel Failure (Minimum DNBR by fuel type.

Ž1.31 and Peak LHR:*21 kw/ft(a)) Fuel performance criteria are Offsite Doses *25% of not challenged by this event.

10CFR50.67; Control Room Doses - GDC 19 Doses bounded by IOSGADV+SF 15.2.2.5 Loss of Normal Peak RCS Pressure <110% of Peak pressure criteria are Feedwater Flow Design not challenged by this event.

(LOFW) Peak Secondary Pressure *110% No impact with use of of Design AREVA fuel, as this is a system response event not Adequate SG inventory to impacted by fuel type.

maintain adequate heat sink Adequate heat sink analysis is not impacted by fuel type.

Page 7 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION 15.2.3.1 Feedwater System Peak RCS Pressure <120% of No impact with use of Pipe Breaks (FSPB Design AREVA fuel, as this is a or FWLB) Peak Secondary Pressure <110% system response event not of Design impacted by fuel type.

No Liquid release through the PSV for peak RCS pressure case Adequate heat sink.

15.2.3.2 Loss of Normal Peak RCS Pressure <110% of Peak Pressure criteria are Feedwater Flow Design not challenged by this event.

with a concurrent Peak Secondary Pressure <1 10% No impact with use of single failure of an of Design AREVA fuel, as this is a active component system response event not (LOFW+SF) Adequate SG inventory to impacted by fuel type.

maintain adequate heat sink Adequate heat sink analysis is not impacted by fuel type.

15.3 Decrease in Reactor Coolant Flow Rate 15.3.1.1 Partial Loss of Peak RCS Pressure <110% of The Partial Loss of Forced Forced Reactor Design Flow is bounded by the Total Coolant Flow Peak Secondary Pressure <110% Loss of Flow (Section (PLOF) of Design 15.3.2.1).

No Fuel Failure (Minimum DNBR Doses Bounded by

>1.31 and Peak LHR 521 kw/ft(a)) IOSGADV Offsite Doses _25% of 10CFR50.67; Control Room Doses _ GDC 19 Page 8 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION 15.3.2.1 Total Loss of Forced Peak RCS Pressure 5110% of Peak Pressure Criteria are Reactor Coolant Design not challenged for this event.

Flow (TLOF) Peak Secondary Pressure <110% Appropriate Core Operating of Design Limits Supervisory System and Core Protection No Fuel Failure (Minimum DNBR Calculator setpoints ensure

>1.31 and Peak LHR _21 kw/ft(a)) fuel failure does not occur.

Offsite Doses _25% of 10CFR50.67; Doses bounded by Control Room Doses < GDC 19 IOSGADV+SF 15.3.2.2 Partial Loss of Peak RCS Pressure -1 10% of Bounded by Single Reactor Forced Reactor Design Coolant Pump Sheared Coolant Flow with Peak Secondary Pressure 5110% Shaft (RCPSS) (Section Concurrent single of Design 15.3.3.2).

failure of an active component Maintain coolable geometry Doses bounded by RCPSS (PLOF+SF) Offsite Doses -<10% of 10CFR50.67; Control Room Doses _ GDC 19 15.3.3.1 Single Reactor Peak RCS Pressure _1 10% of Bounded by Single Reactor Coolant Pump Shaft Design Coolant Pump Sheared Seizure Peak Secondary Pressure <1 10% Shaft (Section 15.3.3.2).

of Design Doses bounded by RCPSS Maintain coolable geometry Offsite Doses <10% of 10CFR50.67; -<

Control Room Doses _ GDC 19 Page 9 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION 15.3.3.2 Single Reactor Peak RCS Pressure <1 10% of Peak Pressure Criteria are Coolant Pump Design not challenged for this event.

Sheared Shaft Peak Secondary Pressure <1 10%

(RCPSS) of Design For purposes of DNB (RCPS)f Deignpropagation analysis, [

geometry Maintain coolable Offsite Doses 510% of 10CFR50.67; Control Room Doses < GDC 19 As such, introduction of AREVA fuel does not challenge maintaining coolable geometry.

As shown in the Pre-Trip Steam Line Break (Section 7.4.1), [

.] As such, doses currently reported in UFSAR remain bounding.

15.3.3.3 Total Loss of Forced Peak RCS Pressure <1 10% of Bounded by Single Reactor Reactor Coolant Design Coolant Pump Sheared Flow with Peak Secondary Pressure !5110% Shaft (Section 15.3.3.2).

Concurrent Single of Design Failure of an Active Doses bounded by RCPSS Component Maintain coolable geometry (TLOF+SF) Offsite Doses -10% of 10CFR50.67; Control Room Doses < GDC 19 Page 10 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION 1 1 15.4 Reactivity and Power Distribution Anomalies 15.4.1.1 Uncontrolled CEA Peak RCS Pressure 5110% of The only key input Withdrawal Design parameter impacted by the (CEAW) from a No Fuel Failure (Minimum DNBR choice of fuel vendor is [

Subcritical or Low >1.31)

Power Condition Fuel Centerline Temperature

!_4706°F Core Operating Limits Supervisory System, Core Protection Calculator, and reactor protection system setpoints are adequate to ensure that fuel failure and fuel centerline melting do not occur.

15.4.1.2 Uncontrolled CEA Peak RCS Pressure _110% of The only key input Withdrawal at Design parameter impacted by the Power (CEAW) No Fuel Failure (Minimum DNBR choice of [

Ž1.31 and Peak LHR 521 kw/ft(a))

.]

Appropriate Core Operating Limits Supervisory System and Core Protection Calculator setpoints ensure fuel failure does not occur.

Control Element Peak RCS Pressure <110% of Peak Pressure Criteria are Assembly (CEA) Design not challenged for this event.

Misoperation No Fuel Failure (Minimum DNBR

>1.31 and Peak LHR <21 kw/ft(a)) Appropriate Core Operating Limits Supervisory System and Core Protection Calculator setpoints ensure fuel failure does not occur.

CVCS Malfunction Time after boron dilution alarm for No impact due to the use of (Inadvertent Boron operator action before shutdown AREVA fuel, as this is a Dilution) margin is lost (Modes 2 to 5) >15 system response event not minutes impacted by fuel type.

Page 11 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION Time after boron dilution alarm for operator action before shutdown margin is lost (Mode 6) >30 minutes 15.4.1.5 Startup of an Reactor remains subcritical No impact due to the use of Inactive Reactor (shutdown margin > 0) AREVA fuel, as this is a Coolant System system response event not Pump impacted by fuel type.

15.4.3.1 Inadvertent Loading Fuel loading errors large enough No impact due to the use of of a Fuel Assembly to adversely affect core AREVA fuel, as this event is into the Improper performance under normal or prevented by administrative Position accident conditions would be controls.

detectable during startup testing.

15.4.3.2 Control Element Peak RCS Pressure _1 10% of No impact to peak pressure Assembly (CEA) Design analysis with use of AREVA Ejection Total centerline enthalpy of hottest fuel; since the peak pressure fuel pellet _250 cal/gm analysis is not impacted by fuel type.

Total average enthalpy of hottest fuel pellet <200 Cal/gm) See SONGS Re-analysis of Maintain coolable geometry CEA Ejection event

-assuming AREVA fuel.

Offsite Doses:525% of 10CFR50.67; Control Room Doses - GDC 19 15.5 Increase in Reactor Coolant Inventory 15.5.1.1 Chemical and Peak RCS Pressure -1 10% of No impact due to the use of Volume Control Design AREVA fuel, as this is a System (CVCS) Peak Secondary Pressure _1 10% system response event not Malfunction of Design impacted by fuel type.

No liquid flow through PSVs for Doses bounded by peak RCS pressure case IOSGADV.

Offsite Doses < 0.5 Rem TEDE; Control Room Dose _ GDC 19 15.5.1.2 Inadvertent Peak RCS Pressure <110% of Bounded by CVCS Operation of the Design malfunction (Section ECCS During Power No liquid flow through PSVs for 15.5.1.1).

O peration No RCS p r ou ghca se .

(IOECCS) peak RCS pressure case.

15.5.2.1 Chemical and Peak RCS Pressure 5110% of No impact due to the use of Volume Control Design AREVA fuel, as this is a System Malfunction Peak Secondary Pressure *110% system response event not with a Concurrent of Design impacted by fuel type.

Page 12 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION Single Failure No liquid flow through PSVs for (CVCS+SF) peak RCS pressure case Doses bounded by Offsite Doses < 0.5 Rem TEDE; IOSGADV.

Control Room Dose s GDC 19 15.5.2.2 Inadvertent Peak RCS Pressure < 110% of Bounded by CVCS Operation of the Design malfunction with concurrent ECCS During Power SF of an active component operation With a No liquid flow through PSVs for (Section 15.5.2.1).

Concurrent Single peak RCS pressure Failure of an Active Component 15.6 Decrease in Reactor Coolant Inventory 15.6.3.1 Primary Sample or Doses _<10% of 10CFR100 (With No impact due to the use of Instrument Line pre-existing Iodine spike the limit AREVA fuel, as this is a Break (PSILB) is 100% of 10CFR100); system response event not Control Room Doses < GDC 19 impacted by fuel type.

15.6.3.2 Steam Generator Offsite Doses _510% of 10CFR100 No impact due to the use of Tube Rupture (With pre-existing Iodine spike the AREVA fuel, as this is a (SGTR) limit is 100% of 10CFR100); system response event not Control Room Doses < GDC 19 impacted by fuel type.

Appropriate Core Operating Limits Supervisory System and Core Protection Calculator setpoints ensure fuel failure does not occur.

15.6.3.4 Inadvertent Opening Bounded by Small Break LOCA See Section 5.2.2, "SBLOCA of a Pressurizer Analysis," and Enclosure 4.

Safety Valve (IOPSV) 15.7 Radioactive Release from a Subsystem or Component 15.7.3.1 Radioactive Waste Offsite Doses 5 0.5 Rem Whole No impact due to the use of Gas System Leak or Body AREVA fuel, as this is a Failure system response event not impacted by fuel type.

15.7.3.2 Radioactive Waste Offsite Doses < 0.5 Rem Whole No impact due to the use of System Leak or Body and < 1.5 Rem Thyroid AREVA fuel, as this is a Failure (Release to system response event not Atmosphere) impacted by fuel type.

Page 13 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION 15.7.3.3 Postulated N/A. No credible accident. No credible accident can Radioactive expose the public to Release Due to radioactive material released Liquid Tank Failure to surface water or groundwater in the event of liquid tank failures.

15.7.3.4 Design Basis Fuel Offsite Doses < 25% of As all pins in both the Handling Accident 10CFR50.67; dropped and impacted Inside Fuel Building Control Room Doses < GDC 19 assemblies are assumed to fail, there is no difference with use of AREVA fuel.

15.7.3.5 Spent Fuel Cask Offsite Doses < 25% of A spent fuel cask drop into Drop Accidents 10CFR50.67; the spent fuel pool is not Control Room Doses _ GDC 19 credible and therefore has no radiological consequences.

A spent fuel cask dropped from the Cask Handling Crane onto a flat surface is not credible and therefore has no radiological consequences.

The radiological consequences of a spent fuel transfer cask drop from the upper shelf in the cask pool back into the lower portion of the cask pool are bounded by the results of a fuel handling accident in the fuel handling building.

15.7.3.6 Spent Fuel Pool Offsite Doses < 25% of To eliminate the potential for Gate Drop Accident 10CFR100; adverse radiological Control Room Doses _ GDC 19. consequences, administrative controls provide assurance that the cask pool and transfer pool gates will not impact fuel assemblies.

15.7.3.7 Test Equipment N/A. No credible accident. Administrative controls have Skid Drop been implemented to preclude fuel failure in the event of a test equipment skid drop.

Page 14 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

Attachment C SONGS - Summary of Impact on UFSAR Chapter 15 Events UFSAR EVENT ACCEPTANCE CRITERIA IMPACT OF AREVA FUEL SECTION 15.7.3.8 Spent Fuel Pool Offsite Doses <25% of 10CFR100 No impact due to the use of Boiling Accident AREVA fuel, as this is a system response event not impacted by fuel type.

15.7.3.9 Design Basis Fuel Offsite Doses < 25% of As all pins in both the Handling Accident 10CFR50.67; dropped and impacted Inside Containment Control Room Doses < GDC 19 assemblies are assumed to fail, there is no difference with use of AREVA fuel.

15.7.3.10 Spent Fuel Offsite Doses <25% of Administrative controls will Assembly Drop onto 10CFR50.67; preclude damage to spent Reconstitution Control Room Doses < GDC 19 fuel assemblies stored atop Station spacers in the reconstitution station. Potential damage to the falling assembly is bounded by the design basis fuel handling accident in the Fuel Handling Building.

15.7.3.11 Use of Offsite Doses < 25% of Doses consequences Miscellaneous 10CFR50.67; bounded by the Equipment Under Control Room Doses < GDC 19 consequences of a fuel 2000 lbs handling accident in the Fuel 1_ _1 _ Handling Building.

15.8 Anticipated Transient Without Scram 15.8 Anticipated Event mitigated by design The requirements of 10 CFR Transient Without features. 50.62 were met by Scram (ATWS) installation of a Diverse SCRAM System and Diverse Emergency Feed Water Actuation System, together with the existing turbine trip function.

15.9 Miscellaneous 15.9.1.1 Asymmetric Steam No Fuel Failure (Minimum DNBR Appropriate Core Operating Generator Transient _>1.31 and Peak LHR <21 kw/ft (a) Limits Supervisory System (ASGT) ) and Core Protection Calculator setpoints ensure fuel failure does not occur.

(a) Technical Specification 2.1.1.2 is stated in terms of peak fuel centerline temperature, not to exceed the design melting point of new fuel, adjusted for burnup and burnable poison content. Maintaining PLHR _<21 kw/ft for this event ensures that fuel centerline melt will not occur.

Page 15 of 15 Attachment C to Enclosure 5 of SONGS PCN 600

ATTACHMENTS D. Compilation of Calvert Cliffs RAI's - Application to SONGS and Responses

Attachment D - Response to Calvert Cliffs RAIs SCE reviewed the Calvert Cliffs regulatory record regarding their request to implement AREVA fuel and reload methodology at their site. Their request differs from the request that SCE is making in this LAR. However, SCE has provided the following summary of our review of a composite list of the NRC Requests for Additional Information (RAIs) that were identified in docketed correspondence between Calvert Cliffs and the NRC.

The text of the RAI Questions has been extracted from the referenced ADAMS documents, which are generally Calvert Cliffs' response documents. The RAI responses in this document are specific to the SONGS Units 2 and 3 application supporting HTP fuel. As a result, the RAI text is left exactly as written by the NRC with reference to either Calvert Cliffs or CC or CCNPP, but the responses in this document refer only to SONGS.

D.1 CC Letter dated 1-26-2010 Question: The staff requested that a listing of the safety evaluation conditions, limitations and restrictions be provided, and document how compliance with each is attained and confirmed for each new methodology proposed for reference in the Core Operating Limits Report.

Response: Attachment E is a listing of the SCE evaluation of AREVA fuel safety evaluation conditions, limitations and restrictions.

4 D.2 CC Letter Dated 4-22-10 Question: The Nuclear Regulatory Commission staff intends to run FRAPCON-3 benchmark calculations of the AREVA CE-1 4 HTP fuel rod design. Please provide (the list of inputs tabulated in the NRC request - list not reproduced here).

Response

This is a Calvert Cliffs specific request, not applicable to SONGS. If the NRC would like to benchmark FRAPCON-3 for SONGS, SCE will provide the requested inputs.

D.3 CC Letter Dated 8-9-2010 Question 1 Question: Please provide more information about the management of the fuel thermal conductivity degradation issue identified in NRC Information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation. Specifically:

D.3.a. ANP-2834(P), Page 1-3, states, "For each specific time in cycle, the fuel conditions are computed using RODEX3A prior to starting the S-RELAP5 portion of the analysis. A steady state condition for the given time in cycle using S-RELAP5 is established. A base fuel centerline temperature is established in this process. Then two-transformation adjustment to the base fuel centerline temperature is computed. The first transformation is a linear adjustment for an exposure of 10 MWd/MTU or higher. In the new process, a polynomial transformation is used in the first transformation instead of a linear transformation." Please clarify the following:

i. Explain how the fuel pellet radial temperature profile is computed ii. Explain which code is used to calculate this profile, both for initial conditions and through the postulated accident.

iii. Explain whether the polynomial transformation is applied merely to the centerline temperature, or to the entire pellet temperature.

Page 1 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls D.3.b. Provide additional information to describe the polynomial transformation.

Summarize data used to develop the polynomial transformation and discuss consideration of applicable uncertainties.

Response

The RODEX3 topical report, ANF-90-145(P)(A), Appendix B provides details of the calculation of the radial temperature distribution.

A portion of the RODEX3A fuel model was incorporated into the S-RELAP5 code to calculate fuel response for transient analyses. This coding, referred to as the S-RELAP5/RODEX3A model, deals only with transient predictions and does not calculate the burnup response of the fuel. Instead, fuel conditions at the burnup of interest are transferred via a binary data file from RODEX3A to S-RELAP5/RODEX3A, establishing the initial state of the fuel prior to the transient. The data transferred from RODEX3A describes the fuel at zero power. A steady--state S-RELAP5/RODEX3A calculation is required to establish the fuel state at power. The transient fuel pellet radial temperature profile is computed by solving the conduction equation in S-RELAP5. Material properties are calculated in S-RELAP5/RODEX3A.

The adjustment is applied to the entire fuel pellet. The polynomial transformation provides a bias adjustment to the fuel centerline temperature. A sampled parameter provides a random assessment and adjustment of the centerline temperature uncertainty. These are combined and the total adjustment is achieved by iterating a multiplicative adjustment to the fuel thermal conductivity until the desired fuel centerline temperature is reached.

Section 6.1 of ANP-2975(P) provides further details.

D.4 CC Letter Dated 8-9-2010 Question 2 Question: The current licensing basis, deterministic loss of coolant accident (LOCA) analysis concluded that the limiting condition did not involve a worst-case single failure, but rather that it depended on injected coolant delivered in such a condition that the resultant containment environment, specifically the lower containment pressure, contributed to the limiting peak cladding temperature (PCT). Please provide information describing how this potentially limiting scenario was evaluated using the proposed best-estimate methodology.

Response

AREVA NP EMF-2103(P)(A) Revision 0 (Reference 8.58) conservatively prescribes:

• The use of full containment sprays without a time delay at the minimum technical specification temperature;

" Pumped ECCS injection at the maximum technical specification temperature; and

• Sampling of the containment volume (indirectly sampling containment pressure) from its nominal volume to its empty volume.

Page 2 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls Studies, comparing several failure assumptions, including a no-failure assumption (see Reference 8.58, EMF-2103(P)(A) Revision 0, RAI response Numbers 26 and 111) validate that the ECCS and containment modeling of the AREVA NP methodology trends to the conservative. The containment pressure response is indirectly ranged by sampling the containment volume. The possible range to be sampled from was 2.305E+6 ft3 to 2.335E+6 ft3 for SONGS Units 2 and 3 containment volume. Reference 8.58, Figure 4-21, shows that there is little sensitivity between containment volume (indirectly pressure) and PCT for a statistical application. Thus, the methodology is responsive to the goal of a realistic evaluation, yet slightly conservative.

D.5 CC Letter Dated 8-9-2010 Question 3 Question: Please provide additional information summarizing the single-failure evaluation performed to establish compliance with General Design Criterion (GDC) 35 requirements. Identify which single failures were considered, discuss whether each failure was evaluated or explicitly analyzed, and for those failures which were explicitly analyzed, explain whether they were analyzed in a reference case or explicitly as a part of the statistical methodology. Also discuss the basis for the single failure evaluation.

For example, were single failures considered as a matter of experience with CCNPP specifically, or with a generic Combustion Engineering nuclear steam supply system design?

Response

The single failures considered in UFSAR Chapter 15 non-LOCA events will remain unchanged for AREVA fuel. SCE is retaining the current SONGS licensing basis event methods and not transitioning to AREVA Non-LOCA chapter 15 methodology.

The single failure assumed in the SONGS RLBLOCA analysis is a loss of one train of ECCS. This single failure assumption is neither the loss of a single LPSI nor the loss of a diesel. It is modeled as a loss of one LPSI pump and one HPSI pump. All containment pressure-reducing systems such as containment fans and containment sprays are modeled as fully functional. This approach conservatively reduces containment pressure and increases break flow. The results of sensitivity study demonstrate that this configuration is PCT-limiting and oxidation-limiting. Section 6.5 of ANP-2975(P) provides further details.

D.6 CC Letter Dated 8-9-2010 Question 4 Question: Page 3-6 states, "the RLBLOCA transients are of sufficiently short duration that the switchover to sump cooling water (i.e., RAS) for ECCS pumped injection need not be considered." For the limiting transient, the collapsed core liquid level from 200-350 seconds appears to trend downward (Figure 3-20). An indication of stable and increasing collapsed liquid level would substantiate the statement quoted above, but this is not the case for Figure 3-20. Is the SRELAP-5 model of the limiting case capable of generating credible results after 350s? If so, please provide results for a period of the transient sufficient to demonstrate that the core collapsed liquid levels are stable or increasing.

Page 3 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls *

Response

The "Collapsed Liquid Level in the Core for the Limiting Case" shows an increase in core collapsed liquid level from 100 seconds to about 150 seconds. After 150 seconds, it shows the core collapsed level decreasing slightly through transient termination. At this time the inner-vessel two-phase mixture extends into the upper plenum. The factors that govern the core liquid content are the steam generation rate, the steam rise velocity and the steam specific volume. After 150 seconds the steam generation rate and the steam velocity change only slightly. However, as the containment pressure falls and as the various steam generation sources outside of the core cool, reducing the differential pressure across the break, the RCS pressure decreases. This decrease in RCS pressure increases the specific volume of the steam in the core, allowing less room for water and thus, decreasing the core collapsed water level. A better measure of a stable cooling inventory is the vessel mass, which shows that water is being supplied to the vessel at the rate that boiling is occurring. The reactor vessel mass shown in ANP-2975(P), Figure 3.22 (Enclosure 3) confirms a stable reactor vessel liquid mass from about 150 seconds until the end of the transient.

D.7 CC Letter Dated 8-9-2010 Question 5 Question: Please provide information to enable comparison between Technical Specifications (TS) requirements and analytic input parameters for Pressurizer Level.

The TS requirement is given in inches and the input parameters are specified in percent span.

Response

Technical Specification LCO 3.4.9 states that the pressurizer shall be OPERABLE with pressurizer water level of less than or equal to 57% and at least two groups of pressurizer heaters OPERABLE each having a capacity greater than or equal to 150 kW. The Technical Specifications for SONGS do not have the requirement in inches, just percent span. The sampled range for the liquid level uncertainty in the pressurizer was 22 to 61 percent of span.

D.8 CC Letter Dated 8-9-2010 Question 6 Question: Please provide discussion to confirm that the assumed 60'F containment temperature is an acceptable minimum without a TS requirement.

Response

The current SONGS Unit 2 and 3 analysis value is 50°F. This value will remain unchanged for AREVA fuel analyses. Multiple Containment Initial Temperature indications inside containment are averaged and logged by the Control Room once-a-day. Selection of 50 OF for the minimum average temperature expected inside containment in the ECCS Performance/Containment Minimum Pressure analyses bounds any credible minimum temperature expectation because of the large amount of energy released by the 3438 MWth reactor, reactor coolant system piping, Pressurizer, and RCP motor operation.

Page 4 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls Containment temperature has a minor effect on the peak clad temperature and clad oxidation that occurs in the ECCS LOCA evaluations. The minimum containment pressure is calculated using the containment air temperature as an input. The minimum containment pressure is used to allow more flow out the break after the blow down phase of the LOCA. The minimum containment air temperature is not a direct impact on LOCA criteria and thus is not a limiting condition for operation. Therefore, the minimum containment temperature does not need to be a technical specification LCO.

SCE and AREVA use a bounding value and this input has a secondary impact on the LOCA ECCS performance.

D.9 CC Letter Dated 8-9-2010 Question 7 Question: The TS minimum for the refueling water storage tank (RWST) temperature is 45 0 F. Previous, deterministic analyses demonstrated that minimum safety injection temperatures resulted in a limiting PCT In light of this information, please explain why a minimum RWST temperature case was not evaluated, or if a minimum RWST temperature case was evaluated, please summarize the evaluation and discuss its conclusions.

Response

The RWST borated water temperature Technical Specifications requirement for SONGS Units 2 and 3 is > 40°F and < 100°F (SR 3.5.4.1). The NRC-approved RLBLOCA EM, EMF-2103(P)(A), prescribes use of the maximum temperature for the ECCS pumped injection and use of the minimum temperature for containment sprays.

The temperatures for the SONGS analysis are 11 0°F for pumped injection and 35 0 F for the containment sprays. The low containment spray temperature was selected to minimize containment pressure and maximize break flow thus increasing cladding temperature and oxidation. High safety injection flow temperature was selected to reduce core cooling thus increasing cladding temperature and oxidation. Therefore conservative input values were used for application of RWST inventory in the LOCA analyses.

D.10 CC Letter Dated 8-9-2010 Question 8 Question: As noted in Section 1 of ANP-2834(P), deviations from the approved RLBLOCA evaluation model (EMF-2103(P)(A), Revision 0) are necessary to demonstrate compliance with 10 CFR 50.46 requirements. Please provide a commitment to adhere to the deviations noted in Section 1 of ANP-2834(P)(A) until such time as:

a. AREVA develops a new revision of EMF-2103,

b. The NRC approves the new revision of EMF-2103, and

c. CCNPP implements the new, NRC-approved revision of EMF-2103.

The commitment should include language to indicate that meeting Conditions a, b, and c, above, or submitting a license action request to implement a different evaluation method, will obviate the need for this commitment.

Page 5 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs

Response

ANP-2975(P) is SCE's equivalent of Calvert Cliffs document ANP-2834(P). In Section 3.2.1.3 of this LAR, SCE proposes adding EMF-2103(P)(A),Revision 0 and the Safety Evaluation Request for this LAR as COLR references. As such the methodology discussed in EMF-2103(P)(A) together with the deviations demonstrated in Section 1 of ANP-2975(P) become part of the SCE methodology for performing RLBLOCA to demonstrate compliance with 10CFR 50.46. Therefore, since adherence to the deviations now become part of SCE's base RLBLOCA methodology for SONGS 2 and 3, no additional commitments are necessary for this purpose.

In the future if AREVA develops, NRC approves and SCE decides to implement a new revision of EMF-2103, SCE will need to get NRC approval for adding this new revision as a COLR reference. Thus the future removal of the methodology deviations described in Section 1 of ANP-2975(P) will only occur if NRC approval of adding the new revision is obtained For these reasons no additional SCE commitment is needed for this item.

D.11 CC Letter Dated 7-23-10 Question 1 Question: A modification to the licensing basis fuel type can have the potential to change the core isotopic distribution assumed in post-accident conditions. Based upon this, please provide additional information regarding the effect the proposed fuel type change has on the current radiological consequence design basis analyses. Please provide any changes to the parameters, assumptions, or methodologies in the radiological design-basis accident (DBA) analyses as a result of the proposed fuel type change and justification for those changes. If there are changes to the radiological DBA analyses, please provide the resulting change to the calculated radiological consequence of the DBAs.

Response

The current SCE analysis generates bounding full core and average fuel rod source terms that are used in the radiological analysis Design Basis Accidents with failed fuel.

The isotopic distribution is calculated using the SAS2H/ORIGEN computer code. This bounding analysis of record was based on Westinghouse fuel and Regulatory Guide 1.183 guidance. In the current reload analysis process, the bounding source terms and current radiological dose analyses are verified as applicable to the new fuel cycle, or new cycle-specific source terms are generated for use in the accident radiological dose analyses. The key parameters that influence the source terms are [

Regardless, the methodology to calculate and the process to verify the bounding source terms will remain unchanged during and after the transition to AREVA fuel. As described in Section 4.9, [

Page 6 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs

] the current bounding analysis described in the UFSAR.

There are no expected changes to the radiological consequences of any UFSAR Chapter 15 described accident.

D.12 CC Letter Dated 10-29-2010 Question 1 Question: Modeling assumptions for flow mixing in the lower plenum of the reactor vessel and non-uniform fuel assembly inlet flow distribution have a 1 st order impact on calculated core parameters (e.g., power distribution, minimum DNBR) during anticipated operational occurrences (AOOs) and accidents.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic methodology, which used uniform inlet flow in thermal hydraulic analyses. SCE will retain responsibility for thermal hydraulic analysis, rather than adopt the AREVA analysis methodology.

The SCE Reload Topical (Reference 8.4) methodology explicitly determines non-uniform inlet flow across the core to be used as input to downstream thermal hydraulic analyses. SCE explicitly calculated [

] Detailed description of the methods is shown in Section 4.2 and the application of the methods is in Section 7.2 Impact of the thermal hydraulic changes on a typical Non-LOCA transient is shown in Section 7.4.1.

D.13 CC Letter Dated 10-29-2010 Question 1 Question: 1.a. The current UFSAR methodology for calculating minimum DNBR consists of a detailed 3D open channel core thermal hydraulics model (i.e., TORC) which specifically models the core inlet flow distribution (mapping of fuel assembly flow factors). This current methodology accounts for flow mixing and non-uniform flow distribution in the lower plenum of the reactor vessel. Separate core inlet flow distributions exist for 4-pump and 3-pump configurations. Please identify and discuss differences in the treatment of core inlet flow distribution in all current and new UFSAR

.Chapter 15 analysis of records (AORs). Include a description of the basis of each model and whether empirical data (e.g., plant flow testing measurements, scale models) were used in their development.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic methodology, which used uniform inlet flow in thermal hydraulic analyses. SCE will retain responsibility for thermal hydraulic analysis, rather than adopt the AREVA analysis methodology.

The SCE Reload Topical (Reference 8.4) methodology [

.] Detailed description of the methods is shown in Section 4.2 and the application of the methods is in Section 7.2. Impact of the thermal hydraulic changes on a typical Non-LOCA transient is shown in Section 7.4.1. The [

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Attachment D - Response to Calvert Cliffs RAls

.] Westinghouse fuel will continue to be modeled with the current NRC-approved SCE methodology and modeling inputs.

D.14 CC Letter Dated 11-19-2010 Question 1 Question: 1.b. The new Asymmetric Steam Generator Transient analysis does not model an asymmetric core inlet temperature distribution and its impact on power distribution. Please identify and discuss differences in the treatment of core inlet temperature distribution in all current and new UFSAR Chapter 15 AORs. Include a description of the basis of each model and whether empirical data (e.g., plant flow testing measurements, scale models) were used in their development. Provide information to justify that any analytic penalties are appropriately conservative.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic methodology, which did not model the asymmetric core inlet temperature distribution and its impact on power distribution. SCE will retain responsibility for thermal hydraulic analysis, rather than adopt the AREVA analysis methodology.

In the SCE submittal, the SCE Reload Topical (Reference 8.4) approved methodology is maintained. As described in Reference 8.4, Section 3.4.2.2.2 and UFSAR Section 15.9.1, the SCE methodology explicitly accounts for the asymmetrical temperature tilt across the core. As the approved SCE methodology will continue to be used for these analyses, there is no change in modeling assumptions.

D.15 CC Letter Dated 10-29-2010 Question 2, CC Letter Dated 11-19-2010 Question 2

Question: The strategy for addressing the presence of both Westinghouse TURBO fuel assemblies and AREVA CE14 HTP fuel assemblies relies on limiting the relative power in the TURBO fuel bundles. During transition cores, fuel management schemes will ensure that resident TURBO fuel assemblies operate at reduced power levels relative to the AREVA CE1 4 HTP fuel assemblies. It is the staffs understanding that peak fuel rod radial peaking factors (Fr) within any TURBO fuel assembly will remain 9% lower than the leading Fr within any AREVA CE14 HTP fuel assembly. In theory, this additional thermal margin will ensure that resident TURBO fuel assemblies will never be limiting during any AOO and accident condition.

Response

The presence of both Westinghouse and AREVA fuel assemblies in the transition core is explicitly modeled in the SONGS Core Physics, Thermal-Hydraulics and Fuel Behavior analyses. As described in the SCE Reload Topical (Reference 8.4), the most limiting parameters from the above analyses are used in the UFSAR Chapter 15 Non-LOCA Transient Analyses. Therefore, no relative radial peaking limits/penalties are needed for either the Westinghouse or the AREVA fuel assemblies (see Sections 4.1 and 7.1 of this document).

Page 8 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs D.16 D16 CC Letter Dated 11-19-2010 Question 2 Question: The staff requests further information to assess this strategy:

a. For lower power events which do not rely upon initial HFP thermal margin (e.g., Post-Trip MSLB, CEA ejection, bank withdrawal, excess load), neither approach to DNBR or fuel centerline melt SAFDLs will be quantified for Westinghouse TURBO fuel rods. How do the transition core reload methods ensure that Westinghouse fuel does not violate its own SAFDLs during these events?

Response

As described in Section 4.5, SCE will continue to validate that the SONGS transition and full cores meets the relevant acceptance criteria as described in the SCE reload topical (Reference 8.4). Per Section 3.4 of Reference 8.4, the SCE analysis of various reactivity and power distribution events are evaluated at intermediate power levels as appropriate and therefore there is no need to demonstrate that a HFP case or margin bounds the results of part power initial condition cases. Both the Westinghouse and AREVA fuel will be analyzed to meet the acceptance criteria for each transient.

The Fuel Centerline Melt Temperature is defined in Technical Specification 2.1.1.2 and described in Section 3.2.1.1 of this submittal. The fuel centerline melt temperature will be applied to Non-LOCA transients for both Westinghouse and AREVA fuel types in the SONGS reactor.

SCE has elected to retain the current DNBR SAFDL (Technical Specification 2.1.1.1) to be used as the acceptance criteria for both AREVA and Westinghouse fuel DNBR calculations, as discussed in Sections 4.2 and 7.2. SCE will [

]for Non-LOCA Transients.

D.17 CC Letter Dated 10-29-2010 Question 2 Question: The staff requests further information to assess this strategy:

b. For CCNPP-2 Cycle 19 and future transition cores, will the 9% thermal margin be preserved under all rodded conditions allowed by the COLR PDIL?

Response

SCE does not preserve any thermal margin between Westinghouse and AREVA fuel.

Explicit analysis is performed for both Westinghouse and AREVA fuel. No changes were made to the current Reference 8.4 NRC-approved SCE methodology which analyzes both multiple times-in-life and partial powers with rodded configurations as allowed by the PDIL. Therefore, no relative radial peaking limits/penalties are credited.

Page 9 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs D.18 CC Letter Dated 11-19-2010 Question 2 Question: The staff requests further information to assess this strategy:

c. For CCNPP-2 Cycle 19 nominal HFP conditions, provide the Fr, calculated DNBR, and overpower DNB margin for the limiting Westinghouse and AREVA fuel rods.

Response

The CC LAR proposed adoption of AREVA methodology for evaluating transition core thermal hydraulic analyses.

In the SCE submittal, the current Reference 8.4 NRC-approved SCE methodology will continue to be used. For Westinghouse and AREVA fuel, the thermal margins are evaluated explicitly. [

.1 SONGS utilizes COLSS and CPCS to monitor and protect the plant. The required thermal margin [

] is then preserved in COLSS constants.

D.19 CC Letter Dated 11-19-2010 Question 2 Question: The staff requests further information to assess this strategy:

d. At different exposure levels, compare the calculated AOO and accident overpower required to achieve the Westinghouse and AREVA cladding strain SAFDL and compare to the predicted overpower for all Chapter 15 AOO and accidents.

Response

[

Page 10 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls D.20 CC Letter Dated 11-19-2010 Question 2 Question: The staff requests further information to assess this strategy:

e. NUREG-0800, SRP-4.2 requires that the number of failed fuel rods not be under- predicted. How do the transition core reload methods quantify the number of Westinghouse fuel rods which violate any SAFDLs during any accident conditions?

Response

The presence of both Westinghouse and AREVA fuel assemblies in the transition core is explicitly modeled in the SONGS Core Physics, Thermal-Hydraulics and Fuel Behavior analyses. The method used in the evaluation of the number of failed fuel rods is [ ] from that described in Attachment C and Section 4.5.2. A Page 11 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls demonstration of the application of the methods for both Westinghouse and AREVA fuel is shown in Section 7.4.1.

D.21 CC Letter Dated 10-29-2010 Question 2 Question: The staff requests further information to assess this strategy:

f. For CCNPP-2 Cycle 19, provide a plot of minimum DNBR versus time (current UFSAR analysis (TURBO) versus new AREVA analysis (CE14 HTP) for several AOO and accident analyses.

Response

The presence of both Westinghouse and AREVA fuel assemblies in the transition core is explicitly modeled in the SONGS Core Physics, Thermal-Hydraulic, and Fuel Behavior analyses. The SCE LAR is not requesting any change to the evaluation models applied to Westinghouse fuel. Consistent with the Reference 8.4 NRC-approved SCE Reload Design Methodology, the most limiting parameters from the above analyses are used in the UFSAR Chapter 15 Non-LOCA Transient Analyses. As shown in Section 7.4.1 for the pre-trip MSLB event, the mixed core will be treated [

] to compare to the acceptance criteria.

D.22 CC Letter Dated 10-29-2010 Question 3, CC Letter Dated 11-19-2010 Question 3

Question: Some of the postulated accidents and transients that are analyzed and described in the Calvert Cliffs Updated Final Safety Analysis Report are sensitive to the initial power level. This is of concern, but may not be limited to, reactivity and power distribution anomalies.

While the current licensing basis and the proposed safety analysis methodology include prospects for analyzing these events at zero- and full-power conditions, the NRC staff has not located documentation describing further analyses, data, and/or sensitivity studies to indicate that the consequences of these events, if initiated at a power level between zero- and full-power, would be less severe than the two power levels analyzed.

Further, allowable operating ranges in the COLR LCOs often vary as a function of power level (e.g., ASI, peaking factor, control rod insertion). The basis for these power-dependent breakpoints must be grounded in safety analysis.

Please identify the limiting set of initial conditions for those transients that are sensitive to the initial core power level and demonstrate that, when initiated at those initial conditions, the analytic results remain within the applicable acceptance criteria. In particular, provide information to demonstrate appropriate consideration of the following:

a. Combinations of initial power level and instrument uncertainty that will provide for i) the greatest challenge to reactor protection system effectiveness, and ii) the greatest rise in power between event initiation and trip completion Page 12 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls

b. The basis for allowable control rod insertion as a function of core power, and the CEA worth and core design parameters that correspond to those limits

c. Initial thermal margin available at the transient onset and the reduction in that margin throughout the transient

d. Core conditions at varying exposures, including mid-cycle cases

e. Assumption of more severe axial power shapes and radial power distributions reflective of operation at lower power levels

f. For CCNPP-2 Cycle 19, provide a plot of minimum DNBR versus time (current UFSAR analysis (TURBO) versus new AREVA analysis (CE1 4 HTP)) for several AOO and accident analyses

Response

The CC LAR proposed adoption of the AREVA non-LOCA transient methodology, which used a different set of assumptions and cases for these events relative to the CC UFSAR description of some events.

In the SCE submittal, the current Reference 8.4 NRC-approved SCE methodology for non-LOCA transients will continue to be used for these analyses. As described in Section 3.4 of Reference 8.4, the SCE analysis of various reactivity and power distribution events are evaluated at intermediate power levels as appropriate, and therefore there is no need to demonstrate that a HFP case bounds the results of part power initial condition cases. See Attachment C for a summary of the impact on SONGS Chapter 15 transients.

D.23 CC Letter Dated 11-19-2010 Question 4 Question: Per the EMF-2310 methodology, the S-RELAP5 analysis for any given transient typically assumes a significant number of initial conditions are taken at nominal values. The licensing basis transient analysis, however, must demonstrate acceptable results with respect to both specified acceptable fuel design limits and reactor coolant pressure boundary integrity. The analytic assumptions that deliver a conservative result for one will, at times, deliver a non-conservative result with respect to the other.

While the EMF-2310 methodology describes detailed thermal-hydraulic analysis, which relies on parametric biasing to provide conservative results with respect to fuel thermal margin, similar parametric biasing to provide conservative results with respect to peak RCS pressure is not always performed.

For transients and accidents that challenge both fuel thermal and RCS pressure margins, provide plant analyses to demonstrate the effects of initiating the selected transients at pressure-limiting initial conditions, including, for example, RCS pressure, main steam system initial pressure, and steam generator initial level.

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Response

The CC LAR proposed adoption of the AREVA non-LOCA transient methodology, which used a different computer code (S-RELAP5) and initial conditions relative to the computer code and methods described in the CC UFSAR description of the determination of this parameter for NSSS pressurization events.

In the SCE submittal, the current Reference 8.4 NRC-approved SCE methodology and computer codes will continue to be used for these analyses. That reference methodology uses CENTS NSSS simulation (Reference 8.33) code to evaluate NSSS events. SCE is not adopting the AREVA S-RELAP5 analysis non-LOCA models and methodology.

SCE will continue to use Reference 8.4 NRC-approved methodology for non-LOCA transient analysis. For the events that using one set of assumptions will deliver conservative result for one criterion but deliver non-conservative result to another criterion, the acceptance criteria are analyzed in separate analyses. For example, for the CEA Ejection event, the RCS peak pressure criterion is verified in a separate analysis using assumptions different from the analysis for the energy deposition criteria.

In the CEA Ejection RCS peak pressure analysis, assumptions are made to maximize the fuel to coolant heat transfer rate, therefore conservatively increase the RCS coolant energy content and challenge the RCS peak pressure criterion.

D.24 CC Letter Dated 10-29-2010 Question 5 Question: Provide a detailed summary describing the process for transient-specific verification of analog instrument setpoints, delays, and uncertainties, and the evaluation of the resultant impact on transient and accident analysis results

Response

The CC LAR proposed a change from the Westinghouse methodology to the AREVA methodology for evaluation of non-LOCA transients including the treatment of analog instrument setpoints, delays, and uncertainties.

The scope of the proposed change in Reference 8.4 SCE non-LOCA transient methodology is confined to Thermal Hydraulic (TH) and Fuel Performance (FP) areas.

Once generated, the results of the TH and FP analyses will be used in the same manner as was applied in the Reference 8.4 NRC-approved SCE reload analysis methodology topical. Therefore, no changes are necessary to the methodology utilizing analog instrument setpoints, delays, and uncertainties.

D.25 CC Letter Dated 10-29-2010 Question 6 Question: Provide recent data concerning fuel rod bowing to demonstrate that (1) legacy analyses for fuel rod bowing remain applicable to modern fuel designs and operating strategies, (2) that thermal-hydraulic testing accounts for fuel rod bowing, and Page 14 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs (3) thermal-hydraulic analysis includes appropriate treatment of fuel rod bowing in light of recently observed data.

Response

The CC LAR proposed a change in licensing basis from Westinghouse rod bow methodology to AREVA rod bow methodology.

For Westinghouse fuel, Rod Bow Penalty is determined via the Rod Bow Topical Report (Reference 8.16). For AREVA fuel, calculation of Rod Bow Penalty is described in Section 4.2.2 and AREVA's topical report (Reference 8.18).

D.26 CC Letter Dated 11-19-2010 Question 7 Question: (for Locked Rotor Event) Section 5.1, Assumption #1, [

]. In light of this assumption, describe the assembly inlet flow factors and flow coast down characteristics in each region of the core. Provide a justification for this assumption. As part of this justification, identify any differences between the new core inlet flow distribution and the current UFSAR AOR.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used uniform inlet flow.

The approved SCE Reference 8.4 analysis methodology will continue to be used for these Locked Rotor analyses. The 3-pump flow distribution methods are described in Section 4.2 and applied in section 7.2 for both the mixed core and full core configurations. Since the RCP pumps remain unchanged and the flow coastdown methods are unchanged from the SCE methods described in Reference 8.4, the flow coastdown curve will not be adversely impacted by the change to AREVA fuel.

D.27 CC Letter Dated 10-29-2010 Question 8 Question: (for Pre-Trip MSLB Event) Item 11 on Page 26 indicates that RCP coastdown begins at reactor scram and not concurrent with reactor trip signal. Justify this change relative to UFSAR.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and trip assumptions for this event relative to the CC UFSAR description of the event.

The SCE Reload Topical (Reference 8.4) methodology will continue to be used for these Pre-Trip MSLB event analyses (See application in Section 7.4.1), and therefore there is no change to initial conditions or trip assumptions for this event.

Page 15 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs Consistent with Table 15.7 of the SONGS Safety Evaluation Report (NUREG-0712, Safety Evaluation Report Related to the Operation of San Onofre Nuclear Generating Station, Units 2 and 3, February 1981) a loss of normal AC power (which powers the RCP's) concurrent with time of reactor trip will be maintained as the SONGS licensing basis.

D.28 CC Letter Dated 11-19-2010 Question 9 Question: (for Pre-Trip MSLB - Continued) The S-RELAP5 scenarios describe symmetric and asymmetric cases. Prior to MSIV closure, steam flow should increase from both SG's. Describe the asymmetric steam flow cases. Include in your description plots of steam flow versus time for all of the cases. In addition, discuss the scenarios which credit the asymmetric SG trip.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which uses different computer codes and modeling assumptions.

For SCE, the LAR does not propose any different modeling assumptions or RPS trips to be credited for this event. SONGS does not credit an asymmetric SG trip for Pre-Trip steam line break. Low Steam Generator Pressure (LSGP)/Core Protection Calculator (CPCS) Variable OverPower Trip (VOPT) are credited as shown in SONGS UFSAR 2/3 Section 15.1.3.1. The event details with steam flow figures are demonstrated in Section 7.4.1 of the LAR.

D.29 CC Letter Dated 10-29-2010 Question 10 Question: (for Pre-Trip MSLB - Continued) New reactor trips are credited (i.e., Thermal Margin/Low Power, Low Steam Generator Pressure, SGAP) relative to trip functions cited in the UFSAR for the pre-trip scenario (i.e., HCPT and Variable High Power Trip).

Describe how initial conditions and assumptions were manipulated to delay these trips.

Response

The CC LAR proposed to credit new trips to the Pre-Trip Steam Line Break.

For SCE, the LAR[ ]

or RPS trip timing to be credited for this event. Low Steam Generator Pressure (LSGP)

/ Core Protection Calculator (CPCS) Variable OverPower Trip (VOPT) are credited as shown in SONGS UFSAR 2/3 Tablel5.10.1.3.1.1-1.

D.30 CC Letter Dated 10-29-2010 Question 11 Question: (for Post-trip MSLB) [

] Describe whether the inclusion of these wider ranges would influence the timing of the transient scenario. Specifically discuss:

a. Higher initial pressurizer pressure may delay timing of LPP SIAS.

b. Higher initial pressurizer pressure may delay delivery of HPSI.

Page 16 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and trip assumptions for this event relative to the CC UFSAR description of the event.

i In the SCE submittal, the current Reference 8.4 NRC-approved SCE methodology will continue to be used for these analyses, and therefore there is no change to methodology for determining initial conditions or trip assumptions to evaluate or justify.

The SCE methodology accounts for measurement uncertainty for both the initial pressurizer pressure and the core inlet temperature.

D.31 CC Letter Dated 11-19-2010 Question 12 Question: (for Post-trip MSLB) The MSLB analysis supporting the migration to AREVA fuel and methods does not include scenarios initiated from lower plant operating modes (as defined in the plant Tech Specs). In lower modes, certain trip functions and ESFAS equipment important in the mitigation of the event may be unavailable. Please discuss the availability of safety related equipment and demonstrate that the HZP case bounds scenarios initiated from lower modes.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions for this event relative to the CC UFSAR description of the event.

For SCE, no changes are proposed to be implemented regarding Post-Trip MSLB analysis methods. Since there are no changes being proposed for the MSLB analysis methods, the Reference 8.4 approved SCE analysis methodology remains applicable.

D.32 CC Letter Dated 10-29-2010 Question 13 Question: (for Post-trip MSLB) Discuss differences in the moderator reactivity versus moderator density curve used in the current S-RELAP5 calculations relative to the current UFSAR AOR. Include a discussion of the effects of stuck rod core location and how cycle-specific differences will be addressed for future reloads.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions for this event relative to the CC UFSAR description of the event.

For SCE, no changes in physics methods are proposed to be implemented regarding the treatment of moderator reactivity or stuck rod worths for this non-LOCA transient event.

Page 17 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls D.33 CC Letter Dated 11-19-2010 Question 14 Question: (for CEA Ejection event) The current UFSAR AOR includes a single case, bounding each input parameter based on conservative selection throughout burnup (BOC to EOC). The new analysis documents a single BOC and EOC case based on predicted physics parameters at the 2 exposure points. More cases may be necessary to ensure that the limiting combination of burnup-dependent parameters has been identified. Demonstrate that the limiting combination of initial conditions and core physics parameters has been captured by these 2 exposure points.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions for this event relative to the CC UFSAR description of the event.

In the SCE submittal, the Reference 8.4 current NRC-approved SCE methodology will continue to be used for this analysis without change, and therefore there is no change to evaluate or justify. The SCE methodology identifies the limiting time in cycle life using a conservative combination of initial conditions and core operating parameters, not just the two cycle end points, to evaluate the event consequences as described in Section 4.5.4 and Section 7.4.2.

D.34 CC Letter Dated 10-29-2010 Question 15 Question: (for CEA Ejection Event - continued) The current UFSAR AOR cites a 0.05 sec time for the control rod to fully eject from the core. The new analysis assumes an ejection time of [ ]. As a result of this change, the reactor trip signal setpoint is reached prior to full withdrawal. Please justify the change in assumed CEA ejection time.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions for this event relative to the CC UFSAR description of the event.

In the SCE submittal, the Reference 8.4 current NRC-approved SCE methodology will continue to be used for this analysis without change. The 0.05 sec time for the control rod to fully eject from the core shown in Section 7.4.2.2 is unchanged from the value reported in UFSAR Section 15.4.3.2. Therefore, there is no change to evaluate or justify.

D.35 CC Letter Dated 11-19-2010 Question 16 Question: (for CEA Ejection Event - continued) In the new analysis, Table 6.1 scram reactivity refers to N-1 values. CCNPP has traditionally used an N-2 scram curve (1 control rod sticks and 1 control rod ejected). Please justify the use of an N-1 scram curve.

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Attachment D - Response to Calvert Cliffs RAls

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions for this event relative to the CC UFSAR description of the event.

In the SCE submittal, the Reference 8.4 current NRC-approved SCE methodology will continue to be used for this analysis without change. This existing analysis is based on an N-2 scram curve. Therefore, there is no change to the existing process.

D.36 CC Letter Dated 10-29-2010 Question 17 Question: (for CEA Ejection Event - continued) Calvert Cliffs reactor and protection system design criteria (UFSAR Chapter 1) dictate that the RPS be capable of performing its function in the event of a single failure. In addition, CCNPP Technical Specifications allow a single excore safety channel to be inoperable. The CEA ejection event exhibits a rapid, localized power excursion. The neutron flux levels measured and timing to reach the VHPT analytical setpoint at each of the four excore safety channels will be influenced by their proximity to the ejected rod (as well as other factors including initial control rod configuration). Furthermore, a harsh environment may exist in containment and must be considered in the instrument response. Please describe how these factors were accounted for in the new analysis

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions for this event relative to the CC UFSAR description of the event.

In the SCE submittal, the Reference 8.4 current NRC-approved SCE methodology will continue to be used for this analysis without change. As described in Section 7.4.2.1, the impact of inoperable excore detectors and the potential harsh environment in containment are explicitly considered.

D.37 CC Letter Dated 10-29-2010 Question 18 Question: (for CEA Ejection Event - Continued) Please discuss differences in analytical methodology and assumptions which prompted the significant change in predicted ejected rod worth in the new analysis relative to the current UFSAR AOR.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions for this event relative to the CC UFSAR description of the event.

In the SCE submittal, the Reference 8.4 current NRC-approved SCE methodology will continue to be used for this analysis without change. Both AREVA and Westinghouse fuel is explicitly modeled at SONGS on a cycle specific basis. CEA ejected worth is influenced by fuel management considerations (i.e. fuel enrichment, location, cycle length, etc.) and not fuel vendor. Therefore significant changes in ejected rod worth at SONGS are not expected due to the change in fuel vendor.

Page 19 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs D.38 CC Letter Dated 10-29-2010 Question 19 Question: (for CEA Ejection Event - Continued) Please discuss the selection of the initial and final AXPD for each case. For example, the DNBR calculation for the BOC HZP case used a bottom peaked AXPD with a peak Fz of 1.3858. This benign AXPD does not appear to be limiting with respect to DNBR.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions for this event relative to the CC UFSAR description of the event.

In the SCE submittal, the Reference 8.4 current NRC-approved SCE methodology will continue to be used. The SCE methodology references the approved CEA Ejection methodology, Reference 8.31, which will continue to be used for this analysis without change including determination of limiting core power distributions. Implementation of the CEA ejection methodology at SONGS is described in Section 7.4.2.

D.39 CC Letter Dated 10-29-2010 Question 20 Question: (for Excess Load) Please provide a plot of the AXPDs (current UFSAR AOR versus new analysis) used in all of the lower power AOO and accident calculations.

Discuss any significant differences.

Response

The CC LAR proposed adoption of the AREVA non-LOCA transient methodology, which used a different set of initial conditions for this event relative to the CC UFSAR description of the Excess Load event.

In the SCE submittal the Reference 8.4 current NRC-approved methodology will continue to be used for these analyses, and therefore there are no changes to the process used to develop and use axial shapes for this event.

Page 20 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs D.40 CC Letter Dated 11-19-2010 Question 22 Question: Section 7.0 identifies penalties used to compensate for potential non-conservative impacts related to the lack of a fuel thermal conductivity model which accurately captures its degradation at higher exposures. The application of these penalties is outside the approved methodology listed in the proposed CCNPP technical specifications. Please provide a detailed description of the augmented methodology. In your description, identify the applicability of these penalties up to a peak rod power of 15 KW/ft.

Additionally, because the augmented methodology is not described in documents listed in the TS COLR - References section, please ensure that the augmentation is summarized or described in an NRC-tracked manner (i.e., the NRC staff recommends adding a reference to TS COLR-References section or documenting the methodology augmentation in a Regulatory Commitment).

Response

NRC information Notice 2009-23, "Nuclear Fuel Thermal Conductivity Degradation" (Reference 8.52) addresses a decrease in fuel pellet thermal conductivity versus burnup based on measured data from 1990s. The NRC asked licensees to review their fuel rod behavior models for pre-1999 fuel thermal model codes. Fuel conductivity affects the heat transfer within the fuel rod which affects fuel rod behavior (e.g., fuel temperature and pressure). The results of the fuel behavior analysis are used as input to the non-LOCA safety analyses. The discussion in IN 2009-23 characterizes the issue as not the specific value of fuel thermal conductivity but rather that the results used in the downstream analyses may be nonconservative. Specifically, IN 2009-23 notes that if the pre-1999 methods misrepresent fuel thermal conductivity, then calculated margins to SAFDLs and other limits may be less conservative than previously understood.

Use of RODEX Section 5.1.4 provides a detailed description of the applications (that are within AREVA's scope for this licensing effort) using the RODEX2 fuel performance code that are [

Use of FATES In response to IN 2009-23 (Reference 8.52), Westinghouse [

Page 21 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs D.41 CC Letter Dated 10-29-2010 Question 23 Question: [

Response

]

[

Page 22 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls Page 23 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs Page 24 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls Page 25 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs I

D.42 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

a. Axial power shapes for the hot rod, hot bundle, and average core region

Response

The plot requested is provided in Enclosure 4, ANP-2974(P) Rev. 0, Figure 3-5.

D.43 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

b. Were the upper core barrel and hot leg nozzle gap leakage paths included in the RELAP5 model?

Response

D.44 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

c. Moderator temperature coefficient

Response

The SBLOCA cold leg pump discharge (CLPD) break cases in Enclosure 4, ANP-2974(P) Rev. 0 were executed [

• ]

Page 26 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs D.45 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

d. Moderator reactivity curve of reactivity vs. moderator density

Response

See response to RAI D.44 above.

D.46 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

e. Were the charging pumps credited in the analysis?

Response

[ .]

D.47 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

f. Decay heat power (fraction) vs. time curve

Response

[ ] are used to specify fission product decay heat plus actinide decay, with a fission product yield factor of 1.2 and a conservative EOC core average 239U yield factor. This results in a decay heat power fraction vs. time curve found in ANP-2974(P) Rev. 0 Figure 4-26.

D.48 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

g. Void distribution in the hot bundle at time of PCT for the 0.09 ft2 and 0.15 ft2 CLBs

Response

The void distribution in the hot bundle for the limiting CLPD [ ] break is shown in ANP-2974(P) Rev. 0, Figure 4-22 and Figure 4-23.

D.49 CC Letter Dated 11-19-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

h. Please also explain the reasons for [

I

Response

I

.1 Page 27 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs A detailed explanation of the differences is provided in the response to D.41.

D.50 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

i. [

.] Assure that the break spectrum identifies the largest break that results in RCS pressure hangup just above the SIT actuation pressure.

Response

[

.] Data for these cases is provided in ANP- 2974(P) Rev. 0, Table 4-1 and 4-2.

D.51 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

j. The axial power shape for the hot rod appears to be a mid-peaked shape with the peak axial power just above the 6 foot elevation (node 13) vs. node 20 for the remainder of the core (see Doc no. [

.J Please verify that the most top peaked axial distribution was used in the analysis, if not, please correct the shape in the re analysis for the hot rod

Response

[ .] Figure 3-4 and Figure 3-5 in ANP-2974(P) Rev. 0 provide the plots for the axial shape and axial distribution.

D.52 CC Letter Dated 11-19-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

k. Please also include moderator reactivity feedback effects in the SBLOCA analyses (moderator reactivity vs. core density) basing the feedback curve on the most positive MTC.

Response

See response to RAI D.44 above.

D.53 CC Letter Dated 11-19-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

I. The HPSI delivery flow rates are higher than those used in the last CE analysis, please explain the differences and verify the HPSI flow curves used in the re-analysis Page 28 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs

Response

The HPSI flow curve used in the SBLOCA analysis documented in ANP-2974(P) is the same curve that was used in SONGS SBLOCA analysis of record as presented in the current FSAR. The HPSI curve data is provided in Table 3-3 in ANP-2974(P) Rev. 0.

D.54 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

m. Please explain and justify the SIT maximum temperature of 90 F compared to 120 F used in the CE analysis. Please also explain and justify the 100°F RWST maximum temperature assumed in the analysis.

Response

The SIT temperature analyzed for SONGS was 130 OF in ANP-2974(P) Rev. 0, which represents the maximum temperature plus 10 OF measurement uncertainty. The RWST (HPSI) temperature analyzed was 110 °F in ANP-2974(P) Rev. 0, which is the maximum Technical Specification temperature including uncertainties.

D.55 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

n. Please provide the results of a severed injection line and provide the values of the degraded HPSI flows to each cold leg.

Response

.1 D.56 CC Letter Dated 10-29-2010 Question 23 Question: (for SBLOCA Event) The staff requests the following information:

o. Provide additional information regarding the 7-minute operator action that is credited to secure the reactor coolant pumps:

i. How does operator training assure that this 7-minute action time will be executed successfully?

ii. Provide EOP revisions that incorporate this action.

Response

The RCP trip criteria are unchanged from the existing step 6 of the S023-12-1 Standard Post Trip Actions Emergency Operating Instruction (EOI). As the step in question exists (without a specific time limit), the operators are already trained on and familiar with the step in question, and the SONGS simulator is capable of modeling the conditions under which the action would be required. The operators are trained on this EOI as a part of both initial and continuing licensed operator training. Therefore, actual performance of the step is feasible and reliable. The additional procedure revisions and operator training to support the 7-minute time requirement, including the appropriate validations, will be tracked as a requirement for implementation of this PCN.

Page 29 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs D.57 CC Letter Dated 11-19-2010 Question 24 Question: The proposed removal of TS 3.2.2, Total Planar Radial Peaking Factor (Fxy), appears to adversely affect the surveillance requirements for TS 3.2.1, Linear Heat Rate (LHR). Specifically, SR 3.2.1.1 stipulates that when monitoring the COLR LHR limit using the excore detector monitoring system, Fxy must be verified to be within specified limits every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In addition to removing TS 3.2.2, the proposed TS change package includes the elimination of SR 3.2.1.1. No alternate means of surveillance for the LHR limit is proposed (when using the excore detector monitoring system). Please discuss the impact of removing this surveillance requirement or propose an alternative.

Response

The SONGS Units 2 and 3 LAR is not proposing any changes to the Technical Specifications or surveillances relative to the radial peaking factors.

D.58 CC Letter Dated 1-14-2011 Question 1 Question: In response to Question (D.14), the licensee states that a reactor trip terminates the asymmetric steam generator transient (ASGT) event prior to the development of any asymmetry at the core inlet. Based on this position, the ASGT is modeled using a uniform core inlet flow and temperature distribution. This modeling assumption essentially removes the unique aspects of the ASGT including the asymmetric core inlet temperature distribution and resulting core power tilt. The existing CCNPP licensing basis for the ASGT specifically captures the asymmetric core inlet flow distribution. Without new information, the staff is unable to accept this change to the CCNPP licensing basis. The staff requests that the ASGT be re-analyzed using a justified asymmetric core inlet temperature distribution.

Response

The CC LAR proposed adoption of the AREVA non-LOCA transient methodology, which used a different set of initial conditions for this event relative to the CC UFSAR description of the event.

The SCE LAR does not propose to make any changes to the approved SCE methodology for the ASGT events as described in Reference 8.4, Section 3.4.2.2.2.

The methodology description in this section explicitly considers an asymmetric core inlet temperature distribution for these events.

D.59 CC Letter Dated 12-30-2010 Question 2 Question: In response to Questions (D.1 3) and (D.26), the licensee states that "modeling assumptions for flow mixing in the lower plenum do not have a first order effect on the minimum DNBR." The response also states that inlet flow distributions are "washed out quickly" in an open lattice PWR core. Based on this position, the single Reactor Coolant Pump Locked Rotor event is modeled using a uniform core inlet flow and temperature distribution. This modeling assumption essentially removes the unique aspects of the asymmetric core inlet flow distribution resulting from a coast down from 4-pump to 3-pump conditions. The existing CCNPP licensing basis for the LR event Page 30 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs specifically captures the symmetric core inlet flow distribution. Without new information, the staff is unable to accept this change to the CCNPP licensing basis. The staff requests that the Locked Rotor minimum DNBR be re-calculated using the existing 3-pump limiting assembly inlet flow factor.

Response

The CC LAR proposed adoption of the AREVA non-LOCA transient methodology, which used a different set of initial conditions for this event relative to the CC UFSAR description of the event.

No changes to the SCE Reference 8.4 NRC-approved methodologies for these events are being requested. At SONGS, the limiting assembly for Locked Rotor is determined based on a wide range of initial operating conditions as well as an inlet flow distribution representing a 3 pump flow model.

The inlet flow distributions for both the 4 pump and 3 pump initial conditions were explicitly calculated (see Section 7.2.2) for the mixed core transition cycle and for the full core of AREVA fuel.

D.60 CC Letter Dated 12-30-2010 Question 3 Question: Based on the licensee's letter dated November 19, 2010, the staff has identified the need for additional information. In response to Question (D.19), AREVA appears to have run some RODEX-2 cases for the CEA drop transient. The maximum cladding strain is calculated to be 0.78% strain at 16.4 GWd/MTU. The CEA drop case was selected because it exhibited the "peak attainable linear heat rate." A peak LHGR is limiting with respect to fuel centerline temperature. However, the maximum change in LHGR is limiting with respect to calculated cladding strain. For each AOO and accident, please provide the peak calculated cladding strain along with the pre- and post-LHGR. Specify radial (Fr) and axial power peaking (Fz) components.

Response

This RAI is a continuation of D.20 that was asked by the NRC during the first round of RAls. The response provided for D.20 applies. Section 5.1.4 describes AREVA's cladding strain analysis methodology using the RODEX2 fuel performance code as well as the augmentation factors developed by AREVA to account for degradation of fuel thermal conductivity with burnup within RODEX2. The NRC-approved RODEX2 code and methodology (Reference 8.14) along with the appropriate penalty will be used by AREVA to demonstrate that occurrence of the most severe plant specific Condition II event in terms of incremental power during the transient will not result in violation of the cladding transient strain criterion for the fuel design intended for operation at the San Onofre units.

D.61 CC Letter Dated 11-19-2010 Question 4 Question: Per the EMF-231 0 methodology, the S-RELAP5 analysis for any given transient typically assumes a significant number of initial conditions are taken at nominal values. The licensing basis transient analysis, however, must demonstrate acceptable results with respect to both specified acceptable fuel design limits and Page 31 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs reactor coolant pressure boundary integrity. The analytic assumptions that deliver a conservative result for one will, at times, deliver a non-conservative result with respect to the other.

While the EMF-231 0 methodology describes detailed thermal-hydraulic analysis, which relies on parametric biasing to provide conservative results with respect to fuel thermal margin, similar parametric biasing to provide conservative results with respect to peak RCS pressure is not always performed.

For transients and accidents that challenge both fuel thermal and RCS pressure margins, provide plant analyses to demonstrate the effects of initiating the selected transients at pressure-limiting initial conditions, including, for example, RCS pressure, main steam system initial pressure, and steam generator initial level. For transients that challenge both fuel thermal and RCS pressure margins, please see an example of SONGS analysis described in the response to D.23.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions relative to the CC UFSAR description.

In the SCE submittal, the current Reference 8.4 NRC-approved SCE methodology and computer codes will continue to be used for these analyses. That reference methodology uses CENTS NSSS simulation (Reference 8.33) code to evaluate NSSS events. SCE will not adopt the AREVA S-RELAP5 analysis models and methodology.

For transients and accidents that challenge both fuel thermal and RCS pressure margins, separate analyses with the appropriate limiting initial conditions are used for each criterion as necessary.

D.62 CC Letter Dated 1-14-2011 Question 4.a Question: The response to RAI Question D.22 does not provide sufficient information to address the staffs concerns. The staff has prepared this request for follow-up information using two examples: the quasi-steady state control element assembly (CEA) withdrawal error at power (CWAP) and the transient CEA drop events.

a. CWAP The CWAP transient is analytically terminated by the variable high power trip (VHPT). It is asserted that the limiting power ascension occurs with the zero-power transient because the trip ceiling at that point provides for the longest trip delay, and with reactor power greater than 30-percent, the trip setpoint is 10-percent greater than the reactor power level.

The response provided considers various phenomena as separate effects, including perturbations in peaking factor, core response characteristics, fractional power level, and setpoint methodology conservatisms. The response asserts that each of these effects provides sufficient conservatism in the analysis to assure that a HZP and a HFP analytic case are bounding of power levels in Page 32 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls between. The conservatism is un-quantified and is not supported with analytic examples.

Please perform a sensitivity analysis of the CWAP transient to demonstrate the effectiveness of the VHPT as a function of reactor power..

It is stated, "The (linear heat rate) is proportional to the fraction of power, so that at powers below 60% this effect overshadows the increases in local peaking from the axial shape index and CEAs at lower powers." Demonstrate that this is true.

At each power level, consider limiting achievable initial axial shape indices (ASIs) and address the attendant DNBR effects. Also include appropriate consideration of the transient radial power redistribution.

Response

The CC LAR proposed adoption of the AREVA non-LOCA transient methodology, which used a different set of assumptions and cases for these events relative to the CC UFSAR description of the event.

In the SCE submittal, the current NRC-approved methodology in Reference 8.4 will continue to be used for these analyses. This approved methodology includes performance of part power calculations, including corresponding initial conditions, in addition to HZP and HFP calculations and conservatively adjusts the overpower trip for power uncertainties.

D.63 CC Letter Dated 12-30-2010 Question 4b, CC Letter Dated 1-14-2011 Question 4.b Question: The response to RAI Question D.22 does not provide sufficient information to address the staff's concerns. The staff has prepared this request for follow-up information using two examples: the quasi-steady state control element assembly (CEA) withdrawal error at power (CWAP) and the transient CEA drop events.

b. CEA Drop The CEA drop event is evaluated for both minimum departure from nucleate boiling ratio and for fuel centerline melt. The peak linear heat generation rate is predicted based on a steady-state evaluation of the end state power level, and a linear heat rate calculation factoring in the maximum allowable Technical Specification peaking factor values. The CEA drop event, however, causes a transient change to the peaking factors.

While the response to RAI (D.22) states that setpoint verification calculations account for transient variations in local power distribution, this transient is unmitigated by a reactor trip. The response to RAI (D.22) also asserts that proportionality in linear heat rate provides adequate margin to SAFDLs to assure that lower-power transients remain bounded by the HFP analyses.

Demonstrate that the above discussed phenomenology holds true for the CEA Drop transient. Analyze the hot full-power sensitivities to dropped rod worth, initial and final power level, and limiting power distributions and redistributions at Page 33 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls lower power levels. Confirm that the limiting results from these studies are bounded by the return to power and radial peaking augmentation factors obtained from the full-power neutronic analyses [

1.

Response

The CC LAR proposed adoption of the AREVA non-LOCA transient methodology, which used a different set of assumptions and cases for these events relative to the CC UFSAR description of the event.

In the SCE LAR submittal, the current Reference 8.4 NRC-approved SCE methodology will continue to be used for these analyses. The SCE reload analysis methodology evaluates various CEA Drop events at intermediate power levels corresponding to the Power Dependent Insertion Limit (PDIL) breakpoints, as appropriate (e.g. Reference 8.4, Section 5.2, physics inputs, Section 5.5.7 Table 5.5.7.-2 for CEA Ejection, and Section 5.5.9 Table 5.5.9-1 for Part Length CEA Drop, Section 5.5.10 Table 5.5.10-1 for DEA Withdrawal Within Deadband, Section 5.5.12 Table 5.5.12-1 for CEA Misalignment, and Section 5.5.13 for CEA Withdrawal). As intermediate power levels are explicitly analyzed as appropriate, there is no need to demonstrate that a HFP case bounds the results of part power initial condition cases.

D.64 CC Letter Dated 12-30-2010 Question 4b, CC Letter Dated 1-14-2011 Question 4.b Question: The staff will request a license condition from CCNPP to limit itself from changing Technical Specification COLR Figures 3.1.6, 3.2.1-2, 3.2.3, or 3.2.5 without prior NRC review and approval until an NRC-accepted, generic or CCNPP-specific basis is developed for analyzing power level-sensitive transients (Control Rod Bank Withdrawal, CEA Drop, and CEA Ejection) at full power conditions only.

Response

The CC LAR proposed adoption of the AREVA reload analysis methodology in place of the Westinghouse analysis methodology of record. This imposed NRC restriction on CC was generated by the issues involved with CC proposing to change their scope of licensing basis calculations (eliminated some confirmatory analyses, changing some analyses to evaluate different criteria).

In the SCE submittal, the scope of analysis in the Reference 8.4 NRC-approved SCE reload methodology will be unchanged. Specifically, the power dependent analyses that are performed for events that are sensitive to initial power levels will continue to be performed at several initial power levels and PDIL breakpoints. Therefore, imposition of a license condition regarding the overall topic of power level sensitive transient methodology is not required.

Page 34 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls D.65 CC Letter Dated 12-30-2010 Question 8 Question: The response to Question (D.22) provides no evidence that CEA ejection events initiated at mid-power conditions along the COL PDIL are bounded by HZP and HFP conditions. Please provide CEA ejection cases at several at-power conditions which capture power-dependent parameters such as PDIL LCOs, ASI LCOs, power peaking LCOs, LSSSs, and power measurement uncertainty. For example, one scenario might be allowable initial conditions at 19.9% power (e.g., Ejected rod worth spanning up to Bank 3 60% inserted, most severe AXPD, etc.) while initiated at a higher power corresponding to 19.9% plus power measurement uncertainty (e.g., secondary calorimetric uncertainty) along with a similarly de-calibrated high power trip.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which used a different set of initial conditions and modeling assumptions for this event relative to the CC UFSAR description of the event As described in the SCE reload analysis methodology Reference 8.4, Section 3.4.2.1.4, for every reload cycle the CEA Ejection event is evaluated at various power levels corresponding to the PDIL breakpoints. Therefore, there is no need to demonstrate that a HFP or HZP case bounds the results of part power initial condition cases. The current NRC-approved SONGS reload methodology will continue to be used for these analyses.

D.66 CC Letter Dated 11-19-2010 Question 9, CC Letter Dated 12-30-2010 Question 9

Question: The response to Question (D.23) appears to be non-responsive to the staffs technical concern.

In Question (D.23) the staff requested that the licensee address pressurization effects of allowable initial conditions other than nominal. In response, the licensee stated that the S-RELAP5 code is being used to analyze for conformance to specified acceptable fuel design limits, and not to analyze reactor coolant pressure boundary integrity.

Because this response does not address the staffs technical concern, we would need to identify a way to restrict our approval of S-RELAP5 to only those safety analyses that confirm acceptable transient performance relative to the specified acceptable fuel design limits, and require prior, transient-specific NRC review and approval of any use of S-RELAP5 to demonstrate reactor coolant pressure boundary integrity.

If the licensee is amenable to proceeding in this fashion, we would close Question (D.23) as resolved, pending the development of the appropriate license condition. If not, we will need to identify an alternative path forward.

Response

The CC LAR proposed adoption of the AREVA non-LOCA transient SCE reload methodology, which used a different computer code (S-RELAP5) relative to the computer code described in the CC UFSAR description of the determination of this parameter for NSSS pressurization events.

Page 35 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls In the SCE submittal, the current Reference 8.4 NRC-approved methodology and computer codes will continue to be used for these analyses. That reference methodology uses CENTS NSSS simulation (Reference 8.33) code to evaluate NSSS pressurization events. The initial conditions used in the CENTS NSSS simulations are unchanged, relative to the SCE approved methodology (Reference 8.4). SCE will not adopt the AREVA S-RELAP5 analysis non-LOCA models and methodology. For transients that challenge both fuel thermal and RCS pressure margins, please see an example of SONGS analysis described in the response to D.23.

D.67 CC Letter Dated 12-30-2010 Question 10 Question: For the SBLOCA please provide the plots of the key system parameters for the breaks re-analyzed that are provided in the UFSAR. No plots were provided nor were the tables summarizing the timing for the key events provided.

Response

These plots are provided in Enclosure 4, ANP-2974(P) Rev. 0, all figures in Section 4.0.

D.68 CC Letter Dated 12-30-2010 Question 11 Question: An HPSI flow delivery curve with 5% more flow was used in the AREVA analysis compared to the previous SBLOCA submittal by CE. Please justify this new HPSI delivery curve and demonstrate that it meets the latest surveillance measurement for HPSI pressure and flow. The HPSI curve is adjusted to account for measurement error (approximately 5%) for pressure and flow when the surveillance pressure/flow measurements are taken. Please demonstrate that these errors are accounted for in the HPSI pressure and flows used in the re-analysis. Please provide the HPSI pressure vs. flow curve used in the analysis.

Response

RAI not applicable to SONGS analysis. The HPSI flow curve used in the SBLOCA analysis documented in ANP-2974(P) is the same curve that was used in the SONGS SBLOCA as reported in the current UFSAR. The HPSI curve data is provided in Table 3-3 in ANP-2974(P) Rev. 0.

D.69 CC Letter Dated 12-30-2010 Question 12, CC Letter Dated 1-14-2011 Question 12 Question: It was stated that the AREVA SBLOCA analysis results in multiple loop seals clearing relative to the previously approved CE analysis where only one loop seal clears. Please provide the plots of the liquid levels in the loop seals for the limiting break and the steam mass flow and velocity entering the loop seal from the horizontal portions of the suction legs. Please show that the conditions in the loop seals in the unbroken loop support clearing of this additional loop. Loop seal clearing phenomena following SBLOCA is very difficult to predict correctly and has historically been poorly predicted by all T/H codes, including the RELAP5 series of codes, As such, loop seal clearing only in the broken loop has been the accepted approach by the NRC staff during the review of evaluation models. Please provide benchmarking of the Page 36 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls S-RELAP5 model against loop seal clearing separate effects tests as well as integral experimental data. While it is recognized that AREVA is using an approved RELAP5 model, it is still necessary to demonstrate that the model is performing correctly in all plant specific calculations, with a physically based thermal hydraulic behavior that supports the loop seal clearing behavior. It is not clear that additional loop seals will clear once the broken loop seal has cleared for such small break sizes. Please provide justification for the RELAP5 multiple loop seal clearing following a small break in the discharge leg. As a comparison, please provide the results of the limiting SBLOCA with only the single broken cold leg loop seal cleared.

Response

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.1 D.70 CC Letter Dated 1-14-2011 Question 13 Question: An analysis of the (SBLOCA) severed injection leg is needed. Please provide the results of an analysis of the severed injection line with the degraded injection into the RCS since one of the line spills to containment while the others inject 2

at the much higher RCS pressures. Breaks up to and including approximately 1.0 ft are considered in the small break spectrum even though they are in the transition region.

Response

.]

D.71 CC Letter Dated 1-14-2011 Question 14 Question: Please perform an analysis of hot leg breaks to demonstrate that the limiting break location for the RCP trip timing criteria has been identified

Response

.1 D.72 CC Letter Dated 12-30-2010 Question 16 Question: In response to RAI (D.57), Constellation described the regulatory basis for removing Surveillance Requirement 3.2.1.1, but did not address the technical basis.

During the audit, please be prepared to describe (1) how "peripheral" ASI will be used to confirm "interior" linear heat rate limits (TS 3.2.1), (2) will peripheral ASI (Fz) be Page 37 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls combined with another local power surveillance (e.g., Fr) to validate peak LHGR (Fq),

and (3) how does the removal of Fxy surveillance guarantee the same level of protection?

Response

The SONGS Units 2 and 3 LAR is not proposing any changes to the radial peaking factor Technical Specification (3.2.2) or the associated surveillance requirements.

D.73 CC Letter Dated 12-30-2010 Question 17 Question: During the audit, we'll need to speak to individuals familiar with the RLB LOCA analysis. The staff doesn't fully understand CCNPP's disposition and RAI response concerning the single failure selection. ANP-2834(P), Section 1, states that "A conservative loss of a diesel assumption is applied in which LPSI inject into the broken loop and one intact loop and HPSI inject into all four loops." CCNPP response to RAI (D.59) (letter dated August 9, 2010), as well as Section 3.1 of ANP-2834(P) state that the limiting single failure has been determined to be the loss of one ECCS pumped injection train. The staff needs to confirm that these failures are one in the same.

The staff also needs to understand how the limiting single failure for the CE NSSS was determined, since the basis for the RAI response defers to NRC-approved methodology. Poring through EMF-2103, the staff only located sensitivity results on 3-loop W systems. In some cases, the limiting failure would be a single LPSI and in others it was a diesel. The staff could not locate a clear, generic disposition for the single failure at any place in EMF-2103.

What was done under the auspices of EMF-2103 development to ensure that the containment analysis produced a sufficiently conservative prediction that a no failure, max S1 spillage case, for a CE NSSS, is bounded by the chosen single failure? The staff will need to see that work.

Response

The single failure assumed in the SONGS RLBLOCA analysis is a loss of one train of ECCS. This single failure assumption is neither the loss of a single LPSI nor the loss of a diesel. It is modeled as a loss of one LPSI pump and one HPSI pump. All containment pressure-reducing systems such as containment fans and containment sprays are modeled as fully functional. This approach conservatively reduces containment pressure and increases break flow. The results of sensitivity study demonstrate that this configuration is PCT-limiting and oxidation-limiting. For a detailed response, see Enclosure 3, ANP-2975(P) Rev. 0 Section 6.5.

D.74 CC Letter Dated 12-30-2010 Question 18, CC Letter Dated 1-14-2011 Question 18 Question: The response to RAI (D.56) discussed the 7-minute operator action that is credited to secure the reactor coolant pumps. Please address the following:

a. Will operators need to know that there is a time-constraint of 7 minutes associated with this action?

b. How have these actions been validated to be feasible and reliable?

c. Who was, or will be, involved in the validation?

Page 38 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAts

d. Describe the changes, if any, to the plant-reference simulator and training that are planned to support these actions.

e. The response to the RAI states, "The new step is in the Pressure and Inventory Safety Function, normally the second safety function performed by the Reactor Operator, and normally within two minutes of a reactor trip." Where did the two minute time frame originate and what is its basis? Is this from another procedure that has previously been validated?

Response

See the prior response to D.56.

D.75 CC Letter Dated 12-30-2010 Question 19 Question: In response to RAI (D.28), the licensee provided Figures 9-4 and 9-6 illustrating steam flow versus time for a 1.0 ft2 break outside containment and inside containment respectively. Both breaks are located upstream of the MSIV.

a. Explain the asymmetric steam flow prior to MSIV closure which is not exhibited in larger breaks.

Response

This RAI is similar to RAI D.28. The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which uses different computer codes and modeling assumptions.

For SCE, the LAR does not propose any changes to the non-LOCA transient computer codes or modeling assumptions, other than those explicitly described in this LAR related to DNB propagation and statistical convolution. The Pre-Trip Steam Line Break event details, with steam flow figures, are demonstrated in Section 7.4.1 of the LAR.

D.76 CC Letter Dated 1-14-2011 Question 20 Question: In response to RAI (D.29), the licensee described various credited trip functions. Please discuss the harsh environment uncertainty for the ASGT and Containment Pressure High trip functions.

Response

The CC LAR proposed adoption of the AREVA thermal hydraulic and non-LOCA transient methodology, which uses different computer codes and modeling assumptions.

For SCE, the LAR does not propose any different modeling assumptions or RPS trips to be credited for this event. SONGS does not credit an asymmetric SG trip for Pre-Trip steam line break. Low Steam Generator Pressure (LSGP) / Core Protection Calculator (CPCS) Variable OverPower Trip (VOPT) are credited as shown in SONGS UFSAR 2/3 Tablel 5.10.1.3.1.1 -1.

The event details are included in Section 7.4.1 of the LAR. For events which create a harsh environment at a credited instrument, the appropriate harsh environment adjustments are made.

Page 39 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs D.77 CC Letter Dated 12-30-2010 Question 22 Question: For the PCT-limiting RLBLOCA case, please provide:

a. Corrected and uncorrected radial temperature profile of the hot rod at the time and location of peak cladding temperature.

b. Temperature vs. time for the limiting PCT case at the limiting location, including the fuel centerline, fuel average, and clad surface temperatures, Indicate the end of blowdown, start of refill, and start of reflood on this graph.

c. Burnup for the limiting rod.

Response: , ANP-2975(P), Figure 6-2 provides the uncorrected and corrected temperature profile for the hot rod. Figure 6-3 provides temperature vs. time for the limiting PCT case at the limiting location. All of the fresh rods have higher PCTs than the once-burnt rods. The most limiting once-burnt rod is the U0 2 rod. With a cycle burnup of approximately 9664 EFPH, the fresh 6% Gad rod has an assembly burnup of 18.2 GWd/MTU while the once-burnt U0 2 has an assembly burnup of 33.2 GWd/MTU.

For a detailed response, see Enclosure 3, ANP-2975(P) Rev. 0 Section 6.1.

D.78 CC Letter Dated 12-30-2010 Question 23 Question: The issue described in IN 2009-23 invalidates AREVA's generic disposition for analyzing fresh fuel only, which is based on sensitivity studies indicating that mid-second-cycle fuel had a PCT of 80'F lower than the limiting PCT. This work needs to be repeated accounting for fuel thermal conductivity degradation. Please provide several cases run at various times-in-life for once-burnt fuel, with information similar to the list provided below; Item ".c" is only necessary for the most limiting second cycle case analyzed.

a. Corrected and uncorrected radial temperature profile of the hot rod at the time and location of peak cladding temperature.

b. Temperature vs. time for the limiting PCT case at the limiting location, including the fuel centerline, fuel average, and clad surface temperatures, Indicate the end of blowdown, start of refill, and start of reflood on this graph.

c. Burnup for the limiting rod.

Response

The SONGS RBLOCA analysis explicitly accounts for the impact of once-burnt fuel on the degradation of thermal conductivity which is burnup dependent. Thermal conductivity degradation is accounted for by applying a bias determined by benchmarking the fuel performance code, RODEX3A, to a set of data that extends past the licensed burnup. The highest burnup for which case calculations are performed is the maximum burnup anticipated for once-burnt fuel in the design. This approach provides bounding LOCA licensing for fuel up to the current AREVA PWR fuel licensed burnup limit of 62 GWd/MTU. For a detailed response, see Enclosure 3, ANP-2975(P)

Rev. 0 Section 6.1.

D.79 CC Letter Dated 1-14-2011 Question 24 Question: In RAI (D.37), the staff requested an explanation of apparent differences between the UFSAR AOR and the AREVA analysis. The response was insufficient to Page 40 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls understand differences in analytical techniques, assumptions, and initial conditions.

Please expand this response.

(a) Provide a comparison of Cycle 18 versus Cycle 19 calculated values for initial ASI, initial Fq, ejected rod worth, post ASI, and post Fq for the limiting rod configuration at HZP and HFP, and (b) Using prior reload cycle calculated parameters (ejected worth, post-Fq, fuel enthalpy), populate the plots of Deposited Enthalpy vs. Rod Worth in XN-NF 44.

Response

This CC RAI was asked as a result of the CC proposed change in licensing basis from Westinghouse methodology to AREVA methodology.

In the SCE submittal, the Reference 8.4 current NRC-approved SCE methodology will continue to be used for this analysis without change. A description of the CEA ejection event showing the impact of Westinghouse and AREVA fuel caused by the changes for M5TM cladding and HTP grids is shown in Section 4.5.4. Section 7.1 also shows the details of the SONGS 3 VQP Cycle 17 core used in the Section 7.4.2 CEA ejection analysis.

D.80 CC Letter Dated 1-14-2011 Question 26 Question: In RAI (D.33), the staff requested an explanation of the use of BOC and EOC physics parameters. The response (provided in RAI (D.22)) was insufficient to justify analyzing only these two extreme cases. At EOC conditions, FTC is most negative (turns event around) and Beff is smallest (promotes power excursion). These parameters act to offset each other, so the limiting scenario may be at another BU point where FTC is not so negative. This is probably the reason why the UFSAR AOR combines limiting parameters, regardless of exposure dependence.

Response

These CC RAI's were asked as a result of the CC proposed change in licensing basis from CE/Westinghouse methodology to AREVA methodology.

In the SCE submittal, the Reference 8.4 current NRC-approved SCE methodology will continue to be used for this analysis without change. The SCE methodology identifies the limiting time in cycle life, not just the two cycle end points, to evaluate the event consequences. Physics input parameters for the cycle 17 VQP core is shown in Section 7.1.1 of this LAR. The CEA ejection event showing the impact of the AREVA fuel is presented in Section 7.4.2 of this LAR.

D.81 CC Letter Dated 12-30-2010 Question 27a Question: With respect to the performance of co-resident Westinghouse fuel assemblies:

a. Please explain how the rod power histories considered in the Westinghouse fuel rod thermal-mechanical performance calculations will be verified for future operating cycles.

Page 41 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs

Response

The SCE cycle specific Physics Fuel Mechanical Analysis calculation (discussed in Reference 8.4, Section 3.1.2) is performed to validate the bounding physics inputs used by Westinghouse for their thermal-mechanical performance verification calculations.

This will remain unchanged with a transition to AREVA fuel.

D.82 CC Letter Dated 12-30-2010 Question 27b Question: With respect to the performance of co-resident Westinghouse fuel assemblies:

b. Did the Post-Trip MSLB analysis consider the performance of Westinghouse fuel should the stuck CEA be located above a Westinghouse assembly?

Response

This CC RAI was asked as a result of the CC proposed change in licensing basis from CE/Westinghouse methodology to AREVA methodology.

In the SCE submittal, the current Reference 8.4 NRC-approved SCE methodology will continue to be used for performing this analysis for Westinghouse fuel, and therefore there is no change to evaluate or justify. The SCE methodology explicitly models all fuel in the core.

D.83 CC Letter Dated 1-14-2011 Question 28 Question: The staff has completed FRAPCON-3 benchmark calculations based upon the limiting CC2CY1 9 rod internal pressure power histories specified in 32-9135500-001. Code inputs and power histories have been coordinated, such that differences are most likely due to code-to-code variations and methodology (e.g., application of uncertainties). FRAPCON-3 benchmark calculations predict a nominal U0 2 fuel rod internal rod pressure of 1714 psia (no AOO power ramps). A statistical evaluation simulating 500 cases; each randomly sampling the manufacturing tolerances, yields a 95/95 UTL internal pressure of 1820 psia. The addition of the fission gas release model uncertainty to this statistical evaluation yields a 95/95 UTL internal pressure of 2260 psia.

Discuss the conservatism of the RODEX2 calculated rod internal pressure, relative to the staffs calculations for Calvert Cliffs.

Response

The FRAPCON-3 benchmark calculations completed by the NRC, the results of which are detailed in this RAI, was specific to the Calvert Cliffs fuel design and operating parameters. AREVA will supply any specific information for the design intended for operation at the San Onofre units that the NRC requires for similar benchmarking upon the NRC's request during the review process.

Page 42 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls

] Section 5.1.4 of this LAR describes the conservatisms applied by the RODEX2 code in the treatment of the power histories as well as in the fuel design parameters. In addition, Section 5.1.4 also describes why the rod internal pressure design analyses performed by the RODEX2 code using the design methodology described in Reference 8.12 are appropriate and conservative.

D.84 CC Letter Dated 12-30-2010 Question 29 Question: As part of a recent Fitzpatrick fuel transition, the staff developed penalties on calculated rod internal pressure and limitations on fuel rod burnup to address outdated fuel thermal-mechanical methods. The use of cycle-specific rod power histories in the AREVA RODEX-2 methodology makes this difficult.

Response

[

.j Section 5.1.4 of this LAR describes the conservatisms applied by the RODEX2 code in the treatment of the power histories as well as in the fuel design parameters. In addition, Section 5.1.4 also describes why the rod internal pressure design analyses performed by the RODEX2 code using the design methodology described in Reference B-6 are appropriate and conservative.

Page 43 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs D.85 CC Letter Dated 12-30-2010 Question 31 Question: Realistic Large Break LOCA Methodology Question:

a. AREVA postulates that clad swelling and rupture produces a benefit to PCT, and because of this, the realistic large break loss of coolant accident (LBLOCA) model does not include a clad swelling and rupture model. Does this conjecture include consideration of test data, which has shown that following fuel rupture, the ballooned region fills with fuel fragments? What analytic studies support this conclusion? How are they applicable to CCNPP? Please also address the potential for co-planar blockage with the fuel relocation evaluation.

b. Since blowdown ruptures can occur at end of life conditions, show that blowdown ruptures do not occur at the end of life for the postulated CCNPP large break LOCA.

Response

Under a condition of fuel relocation, wherein the fuel above the ballooned region drops into the ballooned region, it has been postulated that increased decay heat generation will lead to an increase in cladding heat flux resulting in higher cladding temperatures.

Various presentations purport to show the effect. However, these studies have uniformly incorporated extreme assumptions on the conditions of relocation and the resultant heat transfer processes. Few include provisions for rupture-induced cooling mechanisms. Most assume that the cladding expands circularly without being encumbered by the surrounding pins in the fuel assembly. In fact, a free expansion of the fuel rod is only possible up to pin strains in the mid-30 percents. For higher strains the local gap volume no longer increases faster than the clad surface area. Finally, the packing factor of the rubble filling the ballooned region is over-predicted. If reasonable, yet conservative, assumptions are made, study results would lead to the expectation that fuel relocation, which is real, does not pose a condition by which the ruptured or ballooned regions will exceed the consequence of the non-ballooned regions of the hot pin.

The above conclusion was observed experimentally in the KfK experiments. In the KfK In-Pile Tests, fuel relocation into the ballooned area of the fuel rod occurred but did not adversely affect the subsequent clad temperature behavior. Thus, the KfK experiments demonstrate that analyses which ignore the beneficial effects of swelling and rupture provide conservatively high clad temperature estimates for the ruptured region during reflood even when fuel relocation occurs.

Page 44 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAls

.1I In conclusion, the AREVA NP RLBLOCA EM does not incorporate a clad ballooning, rupture and fuel relocation model. To support this modeling, the cladding temperature and pin stress evolution has been assessed against rupture criteria appropriate for the cladding being evaluated. No rupture occurred during blowdown or refill for fresh or once- and twice-burnt fuel. The assessment was performed for the limiting PCT case.

The limiting case is approximately 60% through the cycle. The PCT reduction between the limiting PCT case and the end-of-life cases is such that blowdown rupture would not be challenged. For rupture during reflood, the cladding temperature for the most severe location on the un-ruptured rod has been demonstrated to conservatively bound the result for any possible rupture location.

For a detailed response, see Enclosure 3, ANP-2975(P) Rev. 0 Section 6.3.

D.86 CC Letter Dated 12-30-2010 Question 32 Question: Provide information to illustrate the conservative nature of the single-side only oxidation model and its application to the CCNPP RLBLOCA analysis.

Response

AREVA NP's NRC-approved RLBLOCA EM uses the [

.]

Thus, the RLBLOCA Revision 0 EM approach of determining local transient oxidation is clearly appropriate to demonstrate compliance with the local oxidation criterion of 10CFR50.46, when combined with the pre-transient oxidation. For a detailed response, see Enclosure 3, ANP-2975(P) Rev. 0 Section 6.4.

D.87 CC Letter Dated 12-30-2010 Question 33 Question: Provide additional information to justify the use of the selected analytic treatment for decay heat uncertainty in the RLBLOCA model.

Response

The SONGS RLBLOCA analysis utilizes the U235 decay curve from the 1979 ANSI/ANS standard for fully saturated decay chains from infinite operation as the decay for all fission products. No bias or uncertainty certainty is assigned to the fission product decay heat. Differing from the base EMF-2103 evaluation model approach, the uncertainty for the decay heat parameter is set to zero and no sampling is done on this parameter, resulting in the decay heat being used with a 1.0 multiplier. The decay heat Page 45 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment D - Response to Calvert Cliffs RAIs in the analysis is always the 1979 ANS standard for decay heat from U235 with fully saturated decay chains, corresponding to infinite operation, assuming 200 MeV per fission. The choice of infinite operation with pure U235 fission product decay heat provides a base model that is conservative for the RLBLOCA time frame of interest, relative to the decay heat for finite operation with uncertainties. For a detailed response, see Enclosure 3, ANP-2975(P) Rev. 0 Section 6.2.

D.88 CC Letter Dated 2-11-2011 Supplemental Info 1 Question: Verify that the most limiting SB LOCA peak cladding temperature was appropriately determined based on a 0.01 ft2 break size increment in a sensitivity study.

Response

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.1 D.89 CC Letter Dated 2-11-2011 Supplemental Info 2 Question: The figures below are provided in connection with discussions regarding the Response to Question 33 (RAI D.87)-provided-in-CC-Letter-dated-1-2-30-201 0-.---...... ..

Response

The response is provided D.87. For a detailed response, including the figures supporting the D.87 response, see Enclosure 3, ANP-2975(P) Rev. 0 Section 6.2.

D.90 CC Letter Dated 2-11-2011 Supplemental Info 3 Question: Provide additional discussions of the Response to Question (D.69) provided in CC Letter dated 1-14-2011, including RELAP5 results for SBLOCA calculations involving loop seal clearing, and the resulting core mixture level.

Response

This request is addressed in the response to D.69.

Page 46 of 46 Attachment D to Enclosure 5 of SONGS PCN 600

Attachment E - Limitations and Constraints ATTACHMENTS E. Evaluation of AREVA SER and TER Methodology Limitations

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