ML15261A833
| ML15261A833 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 08/10/2015 |
| From: | Sutton K Entergy Nuclear Operations, Morgan, Morgan, Lewis & Bockius, LLP |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| RAS 28300, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR | |
| Download: ML15261A833 (172) | |
Text
ENT000679 Submitted: August 10, 2015 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and
) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. )
)
(Indian Point Nuclear Generating Units 2 and 3) )
) August 10, 2015 REVISED TESTIMONY OF ENTERGY WITNESSES NELSON F. AZEVEDO, ALAN B.
COX, JACK R. STROSNIDER, RANDY G. LOTT, MARK A. GRAY, AND BARRY M.
GORDON REGARDING CONTENTION NYS-26B/RK-TC-1B (METAL FATIGUE)
William B. Glew, Jr., Esq. Kathryn M. Sutton, Esq.
ENTERGY NUCLEAR OPERATIONS, INC. Paul M. Bessette, Esq.
440 Hamilton Avenue Raphael P. Kuyler, Esq.
White Plains, NY 10601 MORGAN, LEWIS & BOCKIUS LLP Phone: (914) 272-3360 1111 Pennsylvania Avenue, NW Fax: (914) 272-3242 Washington, DC 20004 E-mail: wglew@entergy.com Phone: (202) 739-3000 Fax: (202) 739-3001 E-mail: ksutton@morganlewis.com E-mail: pbessette@morganlewis.com E-mail: rkuyler@morganlewis.com COUNSEL FOR ENTERGY NUCLEAR OPERATIONS, INC.
TABLE OF CONTENTS Page I. WITNESS BACKGROUND ............................................................................................. 1 A. Nelson F. Azevedo (NFA).................................................................................. 1 B. Alan B. Cox (ABC) ............................................................................................ 5 C. Jack R. Strosnider, Jr. (JRS) ............................................................................... 8 D. Randy G. Lott (RGL) ....................................................................................... 11 E. Mark A. Gray (MAG) ...................................................................................... 14 F. Barry M. Gordon (BMG) ................................................................................. 17 II. OVERVIEW OF CONTENTION NYS-26B/RK-TC-1B ............................................... 20 III.
SUMMARY
OF DIRECT TESTIMONY AND CONCLUSIONS ................................ 25 IV. BACKGROUND ON METAL FATIGUE AND REGULATORY REQUIREMENTS AND GUIDANCE ........................................................................... 32 A. Technical Background on Metal Fatigue and Part 50 Requirements .................. 32
- 1. General Principles of Fatigue Analysis.................................................... 32
- 2. Design Margin and Other Conservatisms in Fatigue Analysis ................ 42
- 3. Fatigue and Other Aging Mechanisms .................................................... 48 B. Applicable 10 C.F.R. Part 54 Requirements and NRC Guidance ....................... 54 V. ENTERGYS LICENSE RENEWAL APPLICATION ADEQUATELY ADDRESSES METAL FATIGUE .................................................................................. 64 A. Overview of the License Renewal Application .................................................. 64 B. NRC Staff Review of the License Renewal Application ..................................... 67
- 1. The NRC Staff Approved the FMP in the Original SER ......................... 67
- 2. Issues Addressed in SSER 1 .................................................................... 68
- 3. Issues Addressed in SSER 2 .................................................................... 76 C. Overview of the EAF Evaluations Prepared for IPEC License Renewal ............ 78 D. The 2010 EAF Analyses for NUREG/CR-6260 Locations Conservatively Demonstrate that the CUFens for All NUREG/CR-6260 Locations at IPEC Do Not Exceed 1.0 ............................................................................................... 83
- 1. Summary of the 2010 EAF Evaluations .................................................. 83
- 2. Number of Transients .............................................................................. 91
- 3. Heat Transfer Coefficients ....................................................................... 97
- 4. Flow Rates and Bulk Liquid Temperatures ........................................... 109 i
Table of Contents (continued)
Page
- 5. Thermal Stratification and Thermal Striping in Pressurizer Surge Line System ........................................................................................... 111
- 6. Environmental Correction Factor .......................................................... 119
- 7. Dissolved Oxygen and Water Chemistry ............................................... 125
- 8. No Propagation of Error Analysis Is Required or Necessary ................ 136 E. Under Commitments 43 and 49 Entergy Has Evaluated the Limiting Locations for Fatigue at IPEC and Conservatively Demonstrated that the CUFens for Limiting Locations Do Not Exceed 1.0........................................... 145
- 1. Overview of the Limiting Locations Review......................................... 145
- 2. EAF Evaluations for RVI Components ................................................. 148
- 3. Other Criticisms of the Limiting Locations Review .............................. 152 F. The Balance of Entergys FMP Is Robust and Provides Reasonable Assurance that the Effects of Fatigue Will Be Adequately Managed ............... 157 VI. CONCLUSIONS............................................................................................................ 164 ii
TABLE OF ABBREVIATIONS T temperature gradient
- transformed strain rate ACRS Advisory Committee on Reactor Safeguards AMP aging management program AMR aging management review ANL Argonne National Laboratory ANO Arkansas Nuclear One ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials ATWS Anticipated Transient Without Scram B.S. Bachelor of Science BWR Boiling Water Reactor CFD computational fluid dynamics CLB current licensing basis CUF cumulative usage factor CUFen environmentally-adjusted CUF DBA design basis accident DO dissolved oxygen EAF environmentally-assisted fatigue EPRI Electric Power Research Institute FAC flow-accelerated corrosion Fen environmental correction factor FMP fatigue monitoring program GENE GE Nuclear Energy GSI generic safety issue HWC hydrogen water chemistry IGSCC intergranular stress corrosion cracking IPEC Indian Point Energy Center IP2 Indian Point Nuclear Generating Unit 2 IP3 Indian Point Nuclear Generating Unit 3 ISI inservice inspection LRA license renewal application LWR light water reactor MOP modified operating procedures MRP Materials Reliability Program M.S. Master of Science NACE National Association of Corrosion Engineers International NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NU Northeast Utilities NYS New York State O* transformed oxygen PEO period of extended operation
ppm parts per million PVRC Pressure Vessel Research Council PWR pressurized water reactor PWSCC primary water stress corrosion cracking QA quality assurance RAI request for additional information RCS reactor coolant system RHR residual heat removal RPI Rensselaer Polytechnic Institute RPV reactor pressure vessel RVI reactor vessel internals SCC stress corrosion cracking SSCs systems, structures and components SER Safety Evaluation Report SPU stretch power uprate SRM staff requirements memorandum T* transformed temperature TLAA time-limited aging analysis UFSAR updated final safety analysis report WOG Westinghouse Owners Group 2
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and
) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. )
)
(Indian Point Nuclear Generating Units 2 and 3) )
) August 10, 2015 REVISED TESTIMONY OF ENTERGY WITNESSES NELSON F. AZEVEDO, ALAN B.
COX, JACK R. STROSNIDER, RANDY G. LOTT, MARK A. GRAY, AND BARRY M.
GORDON REGARDING CONTENTION NYS-26B/RK-TC-1B (METAL FATIGUE)
I. WITNESS BACKGROUND A. Nelson F. Azevedo (NFA)
Q1. Please state your full name.
A1. (NFA) My name is Nelson F. Azevedo.
Q2. By whom are you employed and what is your position?
A2. (NFA) I am employed by Entergy Nuclear Operations, Inc. (Entergy), the applicant in this matter, as Supervisor of Code Programs at Indian Point Nuclear Generating Units 2 and 3 (IP2 and IP3, collectively Indian Point Energy Center or IPEC) in Buchanan, New York.
Q3. Please describe your role in this license renewal proceeding.
A3. I am involved in this proceeding as an Entergy witness in connection with the adjudication of this contention, the safety commitments contention (NYS-38/RK-TC-5), and the reactor vessel internals (RVI) contention (NYS-25). During the Track 1 hearings, I was an expert witness on the buried piping contention (NYS-5) and the flow-accelerated corrosion
contention (RK-TC-2). My role regarding NYS-26B/RK-TC-1B is to provide testimony based on my supervisory role at IPEC in the management of ASME Code programs at IPEC specifically the Code programs related to the management of fatigue.
Q4. Please describe your educational and professional qualifications, including relevant professional activities.
A4. (NFA) My professional and educational qualifications are summarized in the attached curriculum vitae (ENT000032). I hold a Bachelor of Science (B.S.) in Mechanical and Materials Engineering from the University of Connecticut and a Master of Science (M.S.)
in Mechanical Engineering from the Rensselaer Polytechnic Institute (RPI) in Troy, New York. In addition, I have received a Master of Business Administration from RPI.
I have more than 30 years of professional experience in the nuclear power industry.
During that time, I have held engineering, supervisory, and managerial positions with Northeast Utilities (NU) for nearly 19 years, and Entergy for more than 14 years. I became a Manager at NU in 1999, managing five engineering sections responsible for implementing numerous engineering programs at Millstone Station, including the fatigue monitoring programs, maintaining the Class 1 fatigue analyses, reactor pressure vessel (RPV) embrittlement and RVI programs. During that time, I performed several finite element analyses and fatigue analyses for both piping systems and for RPVs, pressurizers and steam generators. Prior to 1998, I was an Engineer for more than ten years and an Engineering Supervisor for another five years at NU.
Since 2001, I have managed the IPEC engineering section responsible for implementing American Society of Mechanical Engineers (ASME) Code programs, including the fatigue monitoring, inservice inspection, inservice testing, flow-accelerated corrosion, snubber testing, boric acid corrosion control, non-destructive examination, steam generators, buried piping, alloy 2
600 cracking, RPV embrittlement, RVIs, welding, and 10 C.F.R. Part 50, Appendix J containment leakrate programs. I also am responsible for ensuring compliance with the ASME Code,Section XI requirements for repair and replacement activities at IPEC. Further, I represent IPEC before industry organizations, including the pressurized water reactor (PWR) Owners Group Materials Subcommittee, and the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) committees.
Q5. Are you familiar with the sections of the IPEC License Renewal Application (Apr. 2007) (LRA) (ENT00015A-B), and its subsequent revisions that are relevant to the technical issues raised in NYS-26B/RK-TC-1B?
A5. (NFA) Yes. I am familiar with the technical issues related to the management of the effects of aging due to fatigue at IPEC, including the environmentally-assisted fatigue (EAF) evaluations prepared by Westinghouse. I have personal knowledge of the development, and subsequent revision, of the portions of the IPEC LRA that address such issues, including the relevant Entergy aging management programs (AMPs).
In particular, as relevant to NYS-26B/RK-TC-1B, I have personal knowledge about the development, and subsequent revision, of the portions of the IPEC LRA that address metal fatigue, including Sections 4.3.1 (Class 1 Fatigue); 4.3.2 (Non-Class 1 Fatigue); 4.3.3 (Effects of Reactor Water Environment on Fatigue Life); B.1.12 (the Entergy aging management program (AMP) that addresses metal fatigue, referred to as the fatigue monitoring program (FMP));
and Commitments 43 and 49. As a result, I am familiar with Entergys plans to manage the effects of metal fatigue at IPEC.
3
Q6. Please further describe the basis for your familiarity with these aspects of the LRA, and the associated technical issues raised in NYS-26B/RK-TC-1B.
A6. (NFA) In my capacity as Supervisor, Code Programs, at IPEC, I have been responsible for the IP2 FMP since 2001. I also supervise the IPEC engineering staff responsible for implementing the IP2 and IP3 FMPs. I reviewed draft versions of the Westinghouse Electric Company LLC (Westinghouse) EAF evaluations for IP2 and IP3 discussed below, and directly interfaced with Westinghouse personnel in resolving technical comments on those drafts before their final approval by Entergy. During my career, I have performed pipe stress analyses, finite element analysis of large components, ASME Code Section XI flaw evaluations, and ASME Code Section III, Class 1 fatigue analyses. Accordingly, I am very familiar with the IPEC FMP, including the description of that program in the LRA; relevant Nuclear Regulatory Commission (NRC) requirements and guidance; and applicable industry codes. I also have been directly involved in developing and reviewing Entergy responses to RAIs concerning fatigue issues addressed in the LRA and any necessary amendments or revisions to the application. I also supported Entergy at the Advisory Committee on Reactor Safeguards (ACRS) meetings for the IPEC LRA held in 2009 and 2015.
I have reviewed various materials in preparing this testimony, including those portions of Entergys LRA for IP2 and IP3 relating to Entergys evaluation of the effects of metal fatigue and EAF. In addition to the relevant sections of the LRA, I reviewed the parties pleadings, statements of position, and testimony on NYS-26B/RK-TC-1B, which are listed in response to Questions 41 and 42, below. I also reviewed the parties pleadings on NYS-26B/RK-TC-1B, the Atomic Safety and Licensing Boards (Boards) orders on this contention, including: (1)
Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 & 3), LBP-08-13, 68 4
NRC 43 (2008) (Order Admitting NYS-26/26A/RK-TC-1/1A); (2) Licensing Board Memorandum and Order (Ruling on Motion for Summary Disposition of NYS-26/26A/Riverkeeper TC-1/1A (Metal Fatigue of Reactor Components) and Motion for Leave to File New Contention NYS-26B/Riverkeeper TC-1B) (Nov. 4, 2010) (unpublished) (Order Admitting NYS-26B/RK-TC-1B); and the exhibits submitted by the State of New York (NYS or the State) and Riverkeeper (collectively, Intervenors) that are relevant to my testimony.
B. Alan B. Cox (ABC)
Q7. Please state your full name.
A7. (ABC) My name is Alan B. Cox.
Q8. By whom are you employed and what is your position?
A8. (ABC) I am now an independent consultant for Entergy; but, before my retirement from the company earlier this year, I was the Technical Manager of License Renewal with Entergy, the applicant in this matter. My office was located at Entergys Arkansas Nuclear One (ANO) facility in Russellville, Arkansas.
Q9. Please describe your role in this license renewal proceeding.
A9. (ABC) I am involved in this proceeding as an Entergy witness in connection with the adjudication of this contention, the safety commitments contention (NYS-38/RK-TC-5), and the RVI contention (NYS-25). During the Track 1 hearings, I was an expert witness on the buried piping, cables, and flow-accelerated corrosion contentions (NYS-5, NYS-6/7, and RK-TC-2, respectively). My role regarding NYS-26B/RK-TC-1B is to provide testimony based on my role, as part of Entergys license renewal services organization, in the development and review of the IPEC LRA.
5
Q10. Please describe your educational and professional qualifications, including relevant professional activities.
A10. (ABC) My professional and educational qualifications are summarized in the attached curriculum vitae (ENTR00031). Briefly summarized, I hold a B.S. degree in Nuclear Engineering from the University of Oklahoma and a Master of Business Administration degree from the University of Arkansas at Little Rock. I have over 38 years of experience in the nuclear power industry, having served in various positions related to engineering and operations of nuclear power plants. For example, I was licensed by the NRC as a reactor operator in 1981 and as a senior reactor operator in 1984 for ANO Unit 1. During operator training and while serving as a shift technical advisor for both ANO units, I was trained in reactor thermal hydraulics and in plant response to transients and accidents. From 1993 to 1996, I was employed by Entergy as a Senior Staff Engineer at ANO. From 1996 to 2001, I served as the Supervisor, Design Engineering, at ANO. I have previously held a professional engineers license in the State of Arkansas.
From 2001 to 2015, I worked for Entergys license renewal services organization, supporting the integrated plant assessment and LRA development for Entergy license renewal projects, as well as projects for other utilities. Specifically, as a member of the Entergy license renewal team, I participated in the development of LRAs for twelve Entergy plants and plants owned by other utilities. Since 2001, I have participated in peer reviews for numerous other LRAs for plants throughout the United States. For over ten years, I was a member of the Nuclear Energy Institute (NEI) License Renewal Task Force. During portions of that time, I served as Entergys representative on the NEI License Renewal Mechanical Working Group and the NEI License Renewal Electrical Working Group.
6
Q11. Are you familiar with the sections of the LRA, and its subsequent revisions that are relevant to the technical issues raised in NYS-26B/RK-TC-1B?
A11. (ABC) Yes. I am familiar with the technical issues related to the management of the effects of aging due to fatigue at IPEC, including the EAF evaluations prepared by Westinghouse. I have personal knowledge of the development, and subsequent revision, of the portions of the IPEC LRA that address such issues, including the relevant Entergy aging management programs (AMPs).
In particular, as relevant to NYS-26B/RK-TC-1B, I have personal knowledge about the development, and subsequent revision, of the portions of the IPEC LRA that address metal fatigue, including Sections 4.3.1 (Class 1 Fatigue); 4.3.2 (Non-Class 1 Fatigue); 4.3.3 (Effects of Reactor Water Environment on Fatigue Life); B.1.12 (the FMP); and Commitments 43 and 49.
As a result, I am familiar with Entergys plans to manage the effects of metal fatigue at IPEC.
Q12. Please further describe the basis for your familiarity with these aspects of the LRA, and the associated technical issues raised in NYS-26B/RK-TC-1B.
A12. (ABC) As Technical Manager, I was directly involved in preparing the LRA and developing AMPs, including the FMP for IPEC. I also have been directly involved in developing and reviewing Entergy responses to RAIs concerning the LRA and various amendments or revisions to the application (principally as they relate to aging management issues). I also supported Entergy at the related Advisory Committee on Reactor Safeguards (ACRS) License Renewal Subcommittee and Full Committee meetings for the IPEC LRA held in 2009 and 2015. Accordingly, I have personal knowledge of the development and subsequent revision of the LRA, including the FMP.
7
I have reviewed various materials in preparing this testimony, including those portions of Entergys LRA for IP2 and IP3 relating to Entergys evaluation of the effects of metal fatigue and EAF. In addition to the relevant sections of the LRA, I reviewed the parties pleadings, statements of position, and testimony on NYS-25, which are listed in response to Questions 41 and 42, below. I also reviewed the parties pleadings on NYS-26B/RK-TC-1B, the Boards orders on this contention, including the Order Admitting NYS-26/26A/RK-TC-1/1A, and the Order Admitting NYS-26B/RK-TC-1B, and the exhibits submitted by the Intervenors that are relevant to my testimony.
C. Jack R. Strosnider, Jr. (JRS)
Q13. Please state your full name.
A13. (JRS) My name is Jack R. Strosnider, Jr.
Q14. By whom are you employed and what is your position?
A14. (JRS) I am a Senior Nuclear Safety and Licensing Consultant with Talisman International, LLC. Since March 2007, I have provided consulting services to nuclear utilities and vendors on nuclear safety, performance issues, licensing and inspection activities.
Q15. Please describe your role in this license renewal proceeding.
A15. I have been retained by Entergy as an independent technical and regulatory expert in connection with the adjudication of this contention, the RVI contention (NYS-25), and the safety commitments contention (NYS-38/RK-TC-5). My role regarding NYS-26B/RK-TC-1B is to provide independent expert testimony based on my experience as a senior manager within the NRC, including supervising NRC Staff in engineering, inspection, research, and license renewal-related activities, and to provide technical testimony on the management of fatigue.
8
Q16. Please describe your educational and professional qualifications, including relevant professional activities.
A16. (JRS) My professional and educational qualifications are summarized in the attached curriculum vitae (ENTR00184). I hold a Bachelors and a Masters degree in Engineering Mechanics, both from the University of Missouri at Rolla. I also hold a Master of Business Administration from the University of Maryland. In brief, prior to April 2007, I was employed for 31 years by the NRC. I held numerous senior management positions at the NRC, including Director of the Office of Nuclear Material Safety and Safeguards, Deputy Director of the Office of Nuclear Regulatory Research, and Director of the Division of Engineering in the Office of Nuclear Reactor Regulation (NRR). I also was a supervisor for inspection activities in the NRCs Region I office from 1984 to 1990 and worked for two years at the Nuclear Energy Agency in Paris, France, which is an intergovernmental organization of industrialized countries that develops guidance and reports on issues that affect nuclear facilities around the world.
I have extensive experience in developing and applying NRC regulations and programs addressing the aging of nuclear power plant structures and components, including metal fatigue issues. In addition to my time as the supervisor of inspection activities in the late 1980s, as Director of the Division of Engineering in NRR from January 1999 to May 2001 I directed engineering reviews and preparation of safety evaluation reports (SERs) for license renewal.
This included developing technical resolutions for first-of-a-kind issues associated with license renewal. As it relates to this contention, I was responsible for research programs related to environmental effects on reactor component cracking, licensing reviews associated with resolution of Generic Safety Issue (GSI) 190, Fatigue Evaluation of Metal Components for 60-Year Plant Life, and the evaluation of the effects of fatigue on reactor components.
9
Q17. Are you familiar with the sections of the LRA, and its subsequent revisions that are relevant to the technical issues raised in NYS-26B/RK-TC-1B?
A17. (JRS) I am familiar with the technical issues related to the management of the effects of aging due to fatigue at IPEC, including the EAF evaluations prepared by Westinghouse. I have knowledge of the development, and subsequent revision, of the portions of the IPEC LRA that address such issues, including the relevant Entergy aging management programs (AMPs).
In particular, as relevant to NYS-26B/RK-TC-1B, I have knowledge of the portions of the IPEC LRA that address metal fatigue, including Sections 4.3.1 (Class 1 Fatigue); 4.3.2 (Non-Class 1 Fatigue); 4.3.3 (Effects of Reactor Water Environment on Fatigue Life); B.1.12 (the FMP); and Commitments 43 and 49. As a result, I am familiar with Entergys plans to manage the effects of metal fatigue at IPEC.
Q18. Please further describe the basis for your familiarity with these aspects of the LRA, and the associated technical issues raised in NYS-26B/RK-TC-1B.
A18. (JRS) I have reviewed various materials in preparing this testimony, including those portions of Entergys LRA for IP2 and IP3 relating to Entergys evaluation of the effects of metal fatigue and EAF. In addition to the relevant sections of the LRA, I reviewed the parties pleadings, statements of position, and testimony on NYS-25, which are listed in response to Questions 41 and 42, below. I also reviewed the parties pleadings on NYS-26B/RK-TC-1B, the Boards orders on this contention, including the Order Admitting NYS-26/26A/RK-TC-1/1A, and the Order Admitting NYS-26B/RK-TC-1B, and the exhibits submitted by the Intervenors that are relevant to my testimony.
10
D. Randy G. Lott (RGL)
Q19. Please state your full name.
A19. (RGL) My name is Randy G. Lott.
Q20. By whom are you employed and what is your position?
A20. (RGL) I am employed by Westinghouse as a Consulting Engineer.
Q21. Please describe your role in this license renewal proceeding.
A21. (RGL) I have been retained by Entergy as an independent technical expert in connection with the adjudication of this contention, the RVI contention (NYS-25), and the safety commitments contention (NYS-38/RK-TC-5). My role regarding NYS-26B/RK-TC-1B is to provide independent expert testimony based on my experience developing and implementing aging management strategies for PWR RPVs, RVIs, and other plant components, including work on developing the generic industry guidelines for managing the effects of aging on RVIs in MRP-227-A, and based on my experience with and knowledge of Westinghouses mechanical and structural evaluations of IPEC plant components.
Q22. Please describe your educational and professional qualifications, including relevant professional activities.
A22. (RGL) My professional and educational qualifications are summarized in my curriculum vitae (ENT000618). Briefly summarized, I hold a B.S. in Nuclear Engineering from the University of Michigan, and an M.S. and Doctor of Philosophy in Nuclear Engineering from the University of Wisconsin. I have over 35 years of experience in nuclear materials and radiation effects.
As noted above, I am employed by Westinghouse. Since joining Westinghouse in 1979, I have been the lead test engineer in the Remote Metallographic (Hot Cell) Facility. In this capacity, I have been responsible for numerous investigations of materials-related issues in 11
PWRs. I have supervised testing of RPV surveillance capsules and conducted research programs on irradiation embrittlement and annealing of RPV steels. In addition, I have pioneered the application of the Master Curve testing to characterize the ductile-to-brittle fracture toughness transition in RPV steels. My contributions have provided the basis for the reconsideration of Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2 (May 1988) (ENT000669), the development of Westinghouse RPV annealing technology, the safety analysis of reactor tanks at Savannah River, the determination of crack growth rates used in alternative plugging criteria for nuclear steam generators and the evaluation of RVI performance.
During my career at Westinghouse I have participated in the evaluation of aging degradation or failure of numerous reactor components including steam generator tubing, BMI flux thimbles, control rod guide tube split pins, baffle-former bolts and clevis insert bolts. I have also conducted numerous research programs on highly irradiated stainless steels, including tensile, fracture toughness and IASCC testing.
For the past eight years, I have been actively involved in the design and implementation of AMPs for RVIs. As a member of the MRP Reactor Internals Inspection and Evaluation Core Group, I was a contributor to the EPRI MRP Pressurized Water Reactor Internal Inspection and Evaluation Guidelines (MRP-227-A) (NRC000114A-F). My work on aging management strategies for the Westinghouse and Combustion Engineering plants provided the basis for the recommended guidelines. The same recommendations have been adopted in the most recent revision of the NRCs Generic Aging Lessons Leaned (GALL) Report (NUREG-1801)
(NYS00146A-C, NYS00147A-D).
12
Q23. Are you familiar with the sections of the LRA, and its subsequent revisions that are relevant to the technical issues raised in NYS-26B/RK-TC-1B?
A23. (RGL) Yes. I am familiar with the technical issues related to the management of the effects of aging due to fatigue at IPEC, including the EAF evaluations prepared by Westinghouse. I have personal knowledge of the development and subsequent revision of the portions of the IPEC LRA that address such issues, including the relevant Entergy AMPs.
In particular, as relevant to NYS-26B/RK-TC-1B, I have knowledge about the development, and subsequent revision, of the portions of the IPEC LRA that address metal fatigue, including fatigue of RVI components, including Sections 4.3.1 (Class 1 Fatigue); 4.3.2 (Non-Class 1 Fatigue); 4.3.3 (Effects of Reactor Water Environment on Fatigue Life); B.1.12 (the FMP); and Commitments 43 and 49. As a result, I am familiar with Entergys plans to manage the effects of metal fatigue at IPEC.
Q24. Please further describe the basis for your familiarity with these aspects of the LRA, and the associated technical issues raised in NYS-26B/RK-TC-1B.
A24. (RGL) I have reviewed various materials in preparing this testimony, including those portions of Entergys LRA for IP2 and IP3 relating to Entergys evaluation of the effects of metal fatigue and EAF. In addition to the relevant sections of the LRA, I reviewed the parties pleadings, statements of position, and testimony on NYS-25, which are listed in response to Questions 41 and 42, below. In preparing my testimony, I also reviewed the parties pleadings on NYS-26B/RK-TC-1B, the Boards orders on this contention, including the Order Admitting NYS-26/26A/RK-TC-1/1A, and the Order Admitting NYS-26B/RK-TC-1B, and the exhibits submitted by the Intervenors that are relevant to my testimony.
13
E. Mark A. Gray (MAG)
Q25. Please state your full name.
A25. (MAG) My name is Mark A. Gray.
Q26. By whom are you employed and what is your position?
A26. (MAG) I am employed by Westinghouse as a Principal Engineer, in the Primary Systems Design and Repair group.
Q27. Please describe your role in this license renewal proceeding.
A27. I have been retained by Entergy as an independent technical expert in connection with the adjudication of this contention, the RVI contention (NYS-25), and the safety commitments contention (NYS-38/RK-TC-5). My role regarding NYS-26B/RK-TC-1B is to provide independent expert testimony based on my experience with and knowledge of Westinghouses fatigue evaluations of IPEC plant components, as well as my experience in structural integrity issues in primary system piping and components, including ASME Code stress and fatigue analysis, and EAF evaluations.
Q28. Please describe your educational and professional qualifications, including relevant professional activities.
A28. (MAG) My professional and educational qualifications are summarized in the attached curriculum vitae (ENTR00186). Briefly summarized, I hold a B.S. in Mechanical Engineering, and an M.S. in Mechanical Engineering with a Nuclear Certificate, both from the University of Pittsburgh. I have over 34 years of experience in the nuclear power industry as an employee of Westinghouse. My principal activities at Westinghouse include the evaluation of structural integrity issues in primary system piping and components. This includes the development of plant life extension and monitoring programs and analysis. I have participated in the development and application of transient and fatigue monitoring algorithms and software for 14
the WESTEMS' Transient and Fatigue Monitoring System, and participated in cooperative efforts with vendors outside Westinghouse in the development of transient and fatigue monitoring systems. I am a member of ASME, the ASME Code Section III Working Group on Piping Design and Working Group on Environmental Fatigue Evaluation Methods, and the EPRI Environmentally-Assisted Fatigue Focus Group. I also was a member of the former EPRI/ASME Environmentally Assisted Fatigue Expert Panel. I am a registered professional engineer in the Commonwealth of Pennsylvania.
Q29. Please describe your specific mechanical and structural engineering experience, including experience with the analysis of fatigue in key reactor components.
A29. (MAG) I have been involved in life extension and license renewal activities at Westinghouse since participating in the first Plant Life Extension pilot study for the Surry Unit 1 nuclear power plant in the mid-1980s. I co-authored the Westinghouse Owners Group (WOG)
Generic Technical Report on Aging Management for Pressurizers, contributed to a similar report covering Reactor Coolant System Piping, and represented Westinghouse before the NRC in their review of the generic reports. I have contributed to the development of transient and fatigue monitoring programs for over a dozen plants. These activities have included overall program development, as well as collection and interpretation of plant historical records and monitoring data for the establishment of baseline fatigue estimates, and identification of improvements to licensee fatigue management programs. I have performed and directed evaluations of the effects of reactor water environment on reactor component fatigue for a number of plants, including IPEC.
In addition, I have extensive experience performing ASME Code evaluations, and in evaluating actual plant transients, including pressurizer surge line stratification (NRC Bulletin 15
88-11), thermal stratification and cycling (NRC Bulletin 88-08), and pressurizer insurge/outsurge. From 1993 to 1998, I led the WOG program on Mitigation and Evaluation of Pressurizer Insurge and Outsurge Transients. I have led plant-specific activities for evaluation of pressurizer insurge/outsurge transients at a number of plants.
For approximately five years, I was lead engineer for fatigue analysis and fatigue-related issues affecting all Class 1 piping and related systems in U.S. Westinghouse plants. In that capacity, I was responsible for all design fatigue evaluations of Class 1 piping systems and components, as well as evaluation of reported non-design transients for their effects on design requirements. In sum, I have extensive experience in the application of finite element analysis, transfer function, and other techniques to evaluate heat transfer, stress and fatigue of components and structures subjected to complex thermal and mechanical loading conditions.
Q30. Please describe your role in the preparation of EAF analyses for IPEC license renewal.
A30. (MAG) During the preparation of the EAF analyses for IPEC license renewal, I provided general technical direction for the engineers performing the EAF analyses, and either co-authored or reviewed all of the resulting Westinghouse environmental fatigue reports relied upon in my testimony. These reports are referred to as WCAP reports or, in some cases, Calculation Notes.
Q31. Are you familiar with the sections of the LRA, and its subsequent revisions that are relevant to the technical issues raised in NYS-26B/RK-TC-1B?
A31. (MAG) Yes. I am familiar with the technical issues related to the management of the effects of aging due to fatigue at IPEC, including the EAF evaluations prepared by 16
Westinghouse. I have personal knowledge of the development, and subsequent revision, of the portions of the IPEC LRA that address such issues, including the relevant Entergy AMPs.
In particular, I have personal knowledge of the portions of the IPEC LRA that address metal fatigue, including Sections 4.3.1 (Class 1 Fatigue); 4.3.2 (Non-Class 1 Fatigue); 4.3.3 (Effects of Reactor Water Environment on Fatigue Life); B.1.12 (the FMP); and Commitments 43 and 49. As a result, I am familiar with Entergys plans to manage the effects of metal fatigue at IPEC.
Q32. Please further describe the basis for your familiarity with the IPEC license renewal project, including the associated LRA raised in NYS-26B/RK-TC-1B.
A32. (MAG) I have reviewed various materials in preparing this testimony, including those portions of Entergys LRA for IP2 and IP3 relating to Entergys evaluation of the effects of metal fatigue and EAF. In addition to the relevant sections of the LRA, I reviewed the parties pleadings, statements of position, and testimony on NYS-25, which are listed in response to Questions 41 and 42, below. In preparing my testimony, I also reviewed the parties pleadings on NYS-26B/RK-TC-1B, the Boards orders on this contention, including the Order Admitting NYS-26/26A/RK-TC-1/1A, the Order Admitting NYS-26B/RK-TC-1B, and the exhibits submitted by the Intervenors that are relevant to my testimony.
F. Barry M. Gordon (BMG)
Q33. Please state your full name.
A33. (BMG) My name is Barry M. Gordon.
Q34. By whom are you employed and what is your position?
A34. (BMG) I am an Associate at Structural Integrity Associates, Inc.
17
Q35. Please describe your role in this license renewal proceeding.
A35. (BMG) I have been retained by Entergy as an independent technical expert in connection with the adjudication of this contention and the safety commitments contention (NYS-38/RK-TC-5).
Q36. Please describe your educational and professional qualifications, including relevant professional activities.
A36. (BMG) My professional and educational qualifications are detailed in the attached curriculum vitae (ENT000680). Briefly summarized, I received an M.S. degree in Metallurgy and Material Science from Carnegie Mellon University. I have over 45 years of experience and expertise in materials corrosion behavior in nuclear power plant environments. Upon graduation I was employed by Westinghouse Bettis as a Materials Engineer, studying the corrosion and hydriding of zirconium fuel cladding followed by mitigation of steam generator corrosion in PWRs. I was subsequently hired by GE Nuclear Energy (GENE) to help resolve issues related to intergranular stress corrosion cracking (IGSCC) of austenitic stainless steels and nickel base alloys in Boiling Water Reactor (BWR) environments. During my 23 year career at GENE, I qualified hydrogen water chemistry (HWC) and patented zinc injection for water chemistry mitigation of IGSCC. In 1998, I became an Associate with Structural Integrity Associates, Inc.
and continue to work on a variety of materials corrosion issues in Light Water Reactors (LWRs) (both PWRs and BWRs) with continued emphasis on stress corrosion cracking (SCC). I am a Corrosion Specialist and Fellow in National Association of Corrosion Engineers (NACE) International and, as a consultant, have been teaching a class on Corrosion and Corrosion Control in LWRs at the NRC since 2004.
18
Q37. Are you familiar with the sections of the LRA, and its subsequent revisions that are relevant to the technical issues raised in NYS-38/RK-TC-5?
A37. (BMG) Yes. I am familiar with the technical issues related to the management of the effects of aging due to fatigue at IPEC, including chemistry issues related to EAF. I have knowledge of the portions of the IPEC LRA that address such issues, including the relevant AMPs.
In particular, I have personal knowledge about the development, and subsequent revision, of the portions of the IPEC LRA that are relevant to my areas of expertise, including Section B.1.41 (Water Chemistry).
Q38. Please further describe the basis for your familiarity with these aspects of the LRA, and the associated technical issues raised in NYS-26B/RK-TC-1B.
A38. (BMG) I have reviewed various materials in preparing this testimony, including those portions of Entergys LRA for IP2 and IP3 relating to Entergys evaluation of the effects of metal fatigue and EAF. In addition to the relevant sections of the LRA, I reviewed the parties pleadings, statements of position, and testimony on NYS-26B/RK-TC-1B, which are listed in response to Questions 41 and 42, below, as those documents are relevant to my testimony. In preparing my testimony, I also reviewed the parties pleadings on NYS-26B/RK-TC-1B, the Boards orders on this contention, including the Order Admitting NYS-26/26A/RK-TC-1/1A, the Order Admitting NYS-26B/RK-TC-1B, and the exhibits submitted by the Intervenors that are relevant to my testimony.
19
II. OVERVIEW OF CONTENTION NYS-26B/RK-TC-1B Q39. Are you familiar with Contention NYS-26B/RK-TC-1B, as originally proposed by the Intervenors?
A39. (NFA, ABC, JRS, RGL, MAG) Yes. We have reviewed the following: State of New Yorks and Riverkeepers Motion for Leave to File a New and Amended Contention Concerning the August 9, 2010 Entergy Reanalysis of Metal Fatigue, dated September 9, 2010 (Motion for Leave), the Petitioners State of New York and Riverkeeper, Inc. New and Amended Contention Concerning Metal Fatigue, dated September 9, 2010 (Amended Contention) and the associated Declarations of Dr. Richard T. Lahey, Jr. (Lahey Declaration),
dated September 9, 2010, and Dr. Joram Hopenfeld (Hopenfeld Declaration), dated September 9, 2010; the NRC Staffs Answer, dated October 4, 2010; Entergys Answer, dated October 4, 2010; and the State of New York and Riverkeepers Joint Reply, dated October 12, 2010.
The original contention alleges that Entergys LRA does not include an adequate plan to monitor and manage the effects of aging due to metal fatigue on key reactor components in violation of 10 C.F.R. § 54.21(c)(1)(iii). Order Admitting NYS-26B/RK-TC-1B at 7.
Specifically, NYS and Riverkeeper claim that Entergy has: (1) inappropriately limited the number of component locations for which EAF analyses must be performed; (2) failed to provide a propagation of error analysis; (3) improperly excluded RPV in-core structures and fittings from the scope of the EAF analyses; (4) not disclosed sufficient information about Westinghouses thermal hydraulic analysis used in the EAF analysis; (5) relied on incorrect or undisclosed assumptions regarding environmental correction (Fen) factors, dissolved oxygen (DO) levels, and numbers of plant transients; and (6) failed to provide a detailed, reliable, and prescriptive AMP. See Amended Contention at 6-13.
20
Q40. Have the Intervenors amended NYS-26B/RK-TC-1B since 2010?
A40. (NFA, ABC, JRS, RGL, MAG) No, they have not.
Q41. Did the Intervenors file statements of position, testimony, and exhibits on this contention in 2011 and 2012, and have you reviewed those materials?
A41. (NFA, ABC, JRS, RGL, MAG, BMG) Yes, we have reviewed the following documents filed by the Intervenors to the extent each is relevant to our testimony:
NYSR00343, State of New York and Riverkeeper, Inc. Initial Statement of Position [on] Consolidated Contention NYS-26B/RK-TC-1B (filed Dec. 27, 2011) (Position Statement);
NYS000439, State of New York and Riverkeeper, Inc.s Revised Statement of Position Regarding Consolidated Contention NYS-26B/RK-TC-1B (filed June 29, 2012) (Rebuttal Position Statement);
NYSR10344, Pre-Filed Written Testimony of Richard T. Lahey, Jr. Regarding Consolidated Contention NYS-26B/RK-TC-1B (filed Oct. 1, 2012) (Lahey Testimony);
NYS000295, Curriculum Vitae of Dr. Richard T. Lahey, Jr. (filed Dec. 22, 2011);
NYS000296, Report of Dr. Richard T. Lahey, Jr. in Support of Contentions NYS-25 and NYS-26B/RK-TC-1B (filed Dec. 22, 2011) (Lahey Report);
NYS000297, Supplemental Report of Dr. Richard T. Lahey, Jr. in Support of Contentions NYS-25 and NYS-26B/RK-TC-1B (filed Dec. 22, 2011)
(Supplemental Lahey Report);
NYS000440, Pre-Filed Written Reply Testimony of Richard T. Lahey, Jr.
Regarding Contention NYS-26B/RK-TC-1B (filed June 29, 2012) (Lahey Rebuttal Testimony);
RIV000034, Pre-Filed Written Testimony of Dr. Joram Hopenfeld Regarding Contention RK-TC-1B - Metal Fatigue (filed Dec. 22, 2011) (Hopenfeld Testimony);
RIV000004, Curriculum Vitae of Dr. Joram Hopenfeld (filed Dec. 22, 2011);
RIV000035, Report of Dr. Joram Hopenfeld in Support of Contention NYS-26B/RK-TC-1B - Metal Fatigue (filed Dec. 22, 2011) (Hopenfeld Report);
21
RIV000114, Pre-Filed Rebuttal Testimony of Dr. Joram Hopenfeld Regarding Contention NYS-26B/RK-TC-1B - Metal Fatigue (filed June 29, 2012)
(Hopenfeld Rebuttal Testimony); and Exhibits NYS00146A-C, NYS000147A-D, NYS000160, and NYS000161 (filed Dec. 15, 2011); NYS000195 (filed Dec. 16, 2011); NYS000298 through NYS000305, NYS000315 through NYS000317, NYS000323, NYS000325, NYS00326A-F, NYS000330, NYS000331, NYS000342, NYS000345 through NYS000369A-B (filed Dec. 22, 2011).
Q42. Did the Intervenors file additional statements of position, testimony, and exhibits on this contention in 2015, and have you reviewed those materials?
A42. (NFA, ABC, JRS, RGL, MAG, BMG) Yes, we have reviewed the following documents filed by the Intervenors to the extent each is relevant to our testimony:
NYS000529, State of New York and Riverkeeper, Inc. Revised Statement of Position [on] Consolidated Contention NYS-26B/RK-TC-1B (filed June 9, 2015) (Revised Position Statement);
NYS000530, Revised Pre-Filed Written Testimony of Dr. Richard T. Lahey, Jr. Regarding Consolidated Contention NYS-26B/RK-TC-1B (filed June 9, 2015) (Revised Lahey Testimony);
RIV000142, Supplemental Pre-Filed Written Testimony of Dr. Joram Hopenfeld Regarding Contention NYS-26B/RK-TC-1B (filed June 9, 2015)
(Supplemental Hopenfeld Testimony);
RIV000144, Supplemental Report of Dr. Joram Hopenfeld in Support of Contention NYS-26[B]/RK-TC-1B and Amended Contention NYS-38/RK-TC-5 (filed June 9, 2015) (Supplemental Hopenfeld Report); and Exhibits NYS000483 through NYS000528 (filed June 9, 2015); RIV000036 through RIV000058 (filed Dec. 22, 2011); RIV000103 through RIV000106 (filed June 19, 2012); RIV000115 through RIV000119 (filed June 29, 2012);
RIV000135 through RIV000141 (filed Nov. 9, 2012); RIV000143 and RIV000145 through RIV000160 (filed June 9, 2015).
In reviewing these statements of position, testimony, and reports, we note that, not only have Intervenors not amended this contention since 2010, the Intervenors also have not replaced their 2011 and 2012 submittals with updated materials in 2015, but merely added new information into the record in 2015. Thus, the claims of the Intervenors, and Drs. Lahey and 22
Hopenfeld, are cumulative and overlapping; redundant in some areas and contradictory in others.
Accordingly, we have focused our review on the most recent statement of position, testimony, and exhibits, filed on June 9, 2015. Nevertheless, we have reviewed all of these materials and our testimony represents our response to the totality of Intervenors and their experts claims, as they can best be understood.
Q43. What other materials have you reviewed in the preparation of your testimony to respond to this contention?
A43. (NFA, ABC, JRS, RGL, MAG, BMG) We have reviewed numerous documents in preparing this testimony, including, for example, those portions of Entergys LRA for IP2 and IP3 relating to Entergys evaluation of the effects of metal fatigue and EAF, and the pertinent portions of NRC regulations and guidance documents such as:
NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Rev. 1 (Sept. 2005) (SRP-LR) (NYS000195);
NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Rev. 2 (Dec. 2010) (SRP-LR, Rev. 2) (NYS000161);
NUREG-1801, Generic Aging Lessons Learned Report, Rev. 1 (Sept. 2005)
(NUREG-1801, Rev. 1) (NYS00146A-C);
NUREG-1801, Generic Aging Lessons Learned Report, Rev. 2 (Dec. 2010)
(NUREG-1801, Rev. 2) (NYS00147A-D);
NUREG/CR-5704, Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels (Apr. 1999) (NUREG/CR-5704)
(NYS000354);
NUREG/CR-6583, Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels (Mar. 1998) (NUREG/CR-6583)
(NYS000356); and NUREG/CR-6909, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials (Feb. 2007) (NUREG/CR-6909) (NYS000357);
NUREG-1930, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 (Nov. 2009) (SER) (NYS00326A-F);
23
NUREG-1930, Supp. 1, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 (Aug. 2011) (SSER 1)
(NYS000160); and NUREG-1930, Supp. 2, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 (Nov. 2014) (SSER 2)
(NYS000507);
We also have reviewed EPRI guidance documents, such as EPRI, Materials Reliability Program: Guidelines for Addressing Fatigue Environmental Effects in a License Renewal Application, Rev. 1 (Sept. 2005) (MRP-47) (NYS000350).
Q44. I show you what has been marked as Exhibits ENTR15001, ENT00015A-B, ENTR00031, ENT000032, ENTR00184 through ENT000231, ENT000369, ENT000616, ENT000618, ENT000627, ENT000631, ENT000636, ENT000646, ENT000659, ENT000665, ENT000669, and ENT000680 through ENT000697. Do you recognize these documents?
A44. (NFA, ABC, JRS, RGL, MAG, BMG) Yes. ENTR15001 is a list of Entergys exhibits, and includes those documents which we referred to, used, or relied upon in preparing this testimony. We have reviewed those documents, ENT00015A-B, ENTR00031, ENT000032, ENTR00184 through ENT000231, ENT000369, ENT000618, ENT000627, ENT000631, ENT000636, ENT000646, ENT000659, ENT000665, ENT000669, and ENT000680 through ENT000697, and these are true and accurate copies of the documents that we have referred to and/or relied upon in preparing this testimony. In those cases in which we have attached only an excerpt of a document as an exhibit, that is noted on Entergys exhibit list.
Q45. How do these documents relate to the work that you do as an expert in forming opinions such as those contained in this testimony?
A45. (NFA, ABC, JRS, RGL, MAG, BMG) These documents represent the type of information that persons within our fields of expertise reasonably rely upon in forming opinions of the type offered in this testimony. Many are documents prepared by government agencies, 24
peer reviewed articles, or documents prepared by Entergy or the utility industry. We note at the outset that we cannot offer legal opinions on the language of the NRC regulations or adjudicatory decisions discussed in our testimony. However, reading those regulations and decisions as technical statements, and relying on our expertise and experience, we can interpret the meaning of those documents as they relate to metal fatigue.
III.
SUMMARY
OF DIRECT TESTIMONY AND CONCLUSIONS Q46. What is the purpose of your testimony?
A46. (NFA, ABC, JRS, RGL, MAG, BMG) The purpose of our testimony is to demonstrate that NYS-26B/RK-TC-1B lacks merit and, accordingly, should be resolved in Entergys favor. The Entergy FMP (i.e., the AMP for managing the effects of aging due to metal fatigue), as set forth in Section B.1.12 of the LRA, as amended, is fully consistent with the program description and the associated Evaluation and Technical Bases provided in NUREG-1801, Revision 1. See NUREG-1801, Rev. 1, at X M-1 to X M-2 (NYS00146C). As we show in this testimony, the Intervenors have not demonstrated any discrepancy between the Entergy FMP and the AMP for Metal Fatigue of the Reactor Coolant Pressure Boundary specified in NUREG-1801, Revision 1.
In addition, although Entergys original aging management program referenced NUREG-1801, Revision 1 (NYS00146A-C), the program has been augmented in response to industry operating experience and also fully meets the intent of the guidance for the Fatigue Monitoring Program contained in NUREG-1801, Rev. 2 (NYS00147A-D). Our testimony shows that the FMP at IPEC provides reasonable assurance that, consistent with the current licensing basis (CLB) and considering environmental effects, the cumulative usage factors (CUFs) for components constituting the reactor coolant system (RCS) pressure boundary and for RVI components will not exceed a cumulative usage factor of 1.0 throughout the period of extended 25
operation (PEO). This, in turn, provides reasonable assurance that the effects of aging due to fatigue on RCS pressure boundary and RVI components will not prevent the components from performing their intended functions throughout the PEO. Our testimony refutes Intervenors criticisms of the FMP and claims to the contrary.
Q47. You mentioned changes in NUREG-1801, Revision 2. Can you briefly summarize the relevant changes related to metal fatigue?
A47. (NFA, ABC, JRS, RGL, MAG) NUREG-1801, Revision 2 has two significant changes that relate to the FMP and the EAF issues raised in this contention. The first is a change to the NRC Staffs guidance on the use of the critical components identified in NUREG/CR-6260. The second is a change to the set of approved formulae for evaluation of EAF. We explain these changes in greater detail in response to Question 91.
Q48. Please describe the scope of your testimony.
A48. (NFA, ABC, JRS, RGL, MAG, BMG) Our testimony identifies and describes the pertinent portions of the LRA for IP2 and IP3 relating to the FMP, which manage the effects of aging due to metal fatigue on the RCS pressure boundary and other components, including RVIs.
We show that the FMP described in the LRA is consistent with the NRC guidance for an acceptable FMP in NUREG-1801, and Entergys reliance on the FMP complies with 10 C.F.R.
Part 54, notwithstanding Intervenors claims to the contrary. Specifically, we describe how the FMP and supporting EAF analyses manage the effects of aging due to metal fatigue on key reactor components consistent with NUREG-1801 so that intended functions will be maintained as required by 10 C.F.R. §§ 54.21(a)(3) and (c)(1)(iii).
Ultimately, we conclude and explain why the projected CUFen values calculated for all three RPV locations (bottom head to shell; RPV inlet nozzle; and RPV outlet nozzle) identified 26
in NUREG/CR-6260 are acceptable through the end of the PEO. For those components where the preliminary scoping CUFen values in the LRA were projected to exceed 1.0 during the PEO, Westinghouse conducted additional fatigue analyses for Entergy, including EAF analyses of the:
(1) surge line hot leg and pressurizer nozzles; (2) RCS piping charging system nozzles; (3) RCS piping safety injection nozzles; and (4) residual heat removal Class 1 piping. Consistent with LRA Commitment 33, and consistent with established engineering methods, Entergy refined its fatigue analyses for NUREG/CR-6260 locations, as explained in Section V.D. These analyses demonstrate valid projected CUFen values that do not exceed 1.0 for all NUREG/CR-6260 locations.
We further describe the methods and inputs in detail, and explain that the CUFen calculations appropriately consider the number of transients, heat transfer coefficients, flow rates and bulk liquid temperatures, thermal stratification, environmental correction factors, and dissolved oxygen. We demonstrate that these methods and inputs are appropriately conservative, and that a propagation of error analysis for these calculations is not required by NRC regulations or recommended by NRC Staff guidance; nor is it necessary or appropriate.
Next, we explain that, to meet LRA Commitments 43 and 49, which also are explained further below, Westinghouse has conducted comprehensive new evaluations of all non-NUREG/CR-6260 IP2 and IP3 components with CLB CUF calculations, including RVIs, and confirmed that CUFen values for all limiting locations at IPEC are not projected to exceed 1.0 at the end of the PEO.
Finally, we explain that Entergy will continue to monitor the actual number of accumulated cycles and will take appropriate corrective actions, including more rigorous analyses, repairs or replacements prior to exceeding the CUF limit of 1.0 should the number of 27
accumulated cycles exceed the analyzed values during future plant operations. As a result, Entergy has demonstrated that it will adequately manage the effects of aging due to fatigue at the affected locations, and has committed to repair or replace the affected locations before their CUF values exceed 1.0, all of which is fully consistent with 10 C.F.R. §§ 54.21(a)(3) and (c)(1)(iii).
Q49. Please summarize the basis for your disagreement with the claims made by the Intervenors and their proffered experts, Drs. Lahey and Hopenfeld, in NYS-26B/RK-TC-1B.
A49. (NFA, ABC, JRS, RGL, MAG, BMG) As indicated in their Revised Position Statement, the Intervenors raise three basic criticisms of Entergys EAF analyses and FMP:
- 1. The methodology [relied upon by Entergy] to determine whether CUFen for any particular component is > 1 - i.e., the WESTEMS computer program - is technically deficient;
- 2. The input values chosen by Entergy for its use of WESTEMS are not technically defensible and understate the extent of metal fatigue; and
- 3. The range of components for which the CUFen calculations are proposed to be conducted is too narrow.
Revised Position Statement at 17 (NYS000529); see also Position Statement at 2-3 (NYSR00343). Our testimony shows that these claimsand the Intervenors numerous related and ancillary claimslack merit.
Q50. Please summarize why you disagree with Intervenors claim that the WESTEMSTM methodology is deficient.
A50. (NFA, ABC, JRS, RGL, MAG) Dr. Lahey and Dr. Hopenfelds critiques of the EAF analyses are based on apparent misunderstandings of the WESTEMS' program and the standard ASME Code methods used to perform fatigue analyses. The Intervenors speculate, for example, that WESTEMS may be nonconservative. Revised Lahey Testimony at 73 (NYS000530). The WESTEMS' software, however, uses standard ASME Code stress and 28
fatigue analysis methods, which contain considerable margin and conservatisms, as we will show. The NRC Staff has reviewed the use of the WESTEMS' code, as applied in the IPEC EAF analyses, and found it acceptable. While the Intervenors experts demand more conservatisms in the environmentally-adjusted CUF (CUFen) calculations, they overlook the significant design margin and conservatisms inherent throughout the fatigue analysis process.
Q51. Please summarize why you disagree with Intervenors claim that the inputs used in the EAF evaluations are not defensible.
A51. (NFA, ABC, JRS, RGL, MAG) The Intervenors criticisms of input values used in the WESTEMS' EAF evaluations are similarly based on a failure to acknowledge the substantial conservatisms in the selection of inputs to the analysis. This includes conservative assumptions regarding heat transfer coefficients, dissolved oxygen (DO) values, and the numbers of analyzed transients. In their testimony on input values, Dr. Lahey and Dr. Hopenfeld also overlook directly-relevant and readily-available information contained in the EAF analyses, and the supporting documentation that Entergy disclosed to the Intervenors in this proceeding.
For example, the LRA EAF evaluations, and their supporting documentation, demonstrate that Westinghouse used conservative assumptions regarding the numbers of transients, heat transfer coefficients, potential thermal stratification, and the environmental correction factor. And contrary to Intervenors claims, there is no valid technical basis, at this time, to conclude that an additional correction factor is needed to account for the potential effects of irradiation embrittlement on fatigue life for RVI components.
29
Q52. Please summarize why you disagree with Intervenors claim that the range of components subject to EAF evaluation is too narrow.
A52. (NFA, ABC, JRS, RGL, MAG) Consistent with the NRC guidance applicable to IPEC at the time the LRA was submitted, see NUREG-1801, Revision 1 at X-M1 (NYS00146C),
Entergy conducted EAF evaluations for all component locations identified in NUREG/CR-6260.
See generally Westinghouse, WCAP-17199-P, Rev. 1, Environmental Fatigue Evaluation for Indian Point Unit 2 (Dec. 2014) (WCAP-17199, Rev. 1) (ENT000681); Westinghouse, WCAP-17200-P, Rev. 1, Environmental Fatigue Evaluation for Indian Point Unit 3 (Dec. 2014)
(WCAP-17200, Rev. 1) (ENT000682). Following the NRCs issuance of NUREG-1801, Rev.
2, which calls for consideration of additional plant-specific component locations in the reactor coolant pressure boundary if they may be more limiting than those considered in NUREG/CR-6260, see NUREG-1801, Revision 2 at X M1-2 (NYS000147C), Entergy committed to conduct an additional review of potentially limiting locations, and has now completed that review for IP2 and IP3. See generally Westinghouse, Calculation Note CN-PAFM-12-35, Rev. 1, Indian Point Unit 2 and Unit 3 EAF Screening Evaluations (Nov. 26, 2012) (NYS000510) (Westinghouse Calculation Note CN-PAFM-12-35); Westinghouse, Calculation Note CN-PAFM-13-32, Rev. 3 Indian Point Unit 2 (IP2) and Unit 3 (IP3) Refined EAF Analyses and EAF Screening Evaluations (June 25, 2015) (Westinghouse Calculation Note CN-PAFM-13-32, Rev. 3)
(ENT000683). As we explain in further detail in Section V.E, this limiting location review included all components with CLB CUFs, including RVIs. Intervenors claim that Entergys review is too narrow lacks merit because Entergy has evaluated all CLB CUF locations for EAF.
30
Q53. In addressing the three basic claims presented by the Intervenors, how is your testimony presented below?
A53. (NFA, ABC, JRS, RGL, MAG) To address these three overarching claimsas well as the various additional ancillary issues raised in Intervenors testimonyour testimony presents a comprehensive discussion of the analytical inputs and methodology in Section V.D, which is further divided into eight parts, describing: (1) the EAF evaluations, generally, see Section V.D.1; (2) the number of transients, see Section V.D.2; (3) the heat transfer coefficients, see Section V.D.3; (4) flow rates and bulk liquid temperatures, see Section V.D.4; (5) thermal stratification and thermal striping in the pressurizer surge line, see Section V.D.5; (6) the environmental correction factor, see Section V.D.6; (7) dissolved oxygen and water chemistry, see Section V.D.7; and (8) that no propagation of error analysis is required or necessary, see Section V.D.8. In Section V.E, we explain that Entergy has completed its limiting locations review, pursuant to Commitments 43 and 49, which included RVIs, and that the scope of the review is appropriate. Finally, in Section V.F, we explain that the balance of Entergys FMP, in conjunction with the corrective action program, provides reasonable assurance that the effects of aging due to fatigue will be adequately managed through the PEO.
In summary, the FMP set forth in Entergys LRA for IP2 and IP3 is fully compliant with the applicable guidance in NUREG-1801 and with NRC license renewal regulations. As a result, contrary to the Intervenors contention, there is reasonable assurance that the aging effects of metal fatigue on the RCS pressure boundary and for RVI components will be adequately managed during the PEO, consistent with 10 C.F.R. §§ 54.21(a)(3), 54.21(c)(1)(iii), and 54.29(a).
31
IV. BACKGROUND ON METAL FATIGUE AND REGULATORY REQUIREMENTS AND GUIDANCE A. Technical Background on Metal Fatigue and Part 50 Requirements
- 1. General Principles of Fatigue Analysis Q54. Please explain the concept of metal fatigue.
A54. (NFA, ABC, JRS, RGL, MAG) Metal fatigue is a process of permanent structural change occurring in a material subjected to fluctuating stresses and strains that may culminate in cracks or fracture after a sufficient number of fluctuations. See ASTM Vol. 03.01, Metals -
Mechanical Testing; Elevated and Low Temperature Tests; Metallography, E1823 - 10a (2005)
(ENT000187). Metal components experience these stresses during transients, such as temperature changes during plant startup and shutdown. An excessive number of transients (or cycles), may result in cracking and, eventually, fracture of a component at stress levels below those for which the component was designed.
Q55. Please explain the difference between the terms transients and cycles, as those terms are used in fatigue analysis.
A55. (MAG, NFA) In general, these terms can be used somewhat interchangeably.
The term transient describes a component loading that varies with time. The transient loadings create stress cycles in a component. The stress cycles from the applicable transients are considered in the components fatigue evaluation. In this context, the term transients can typically be used interchangeably with stress cycles, fatigue cycles, load cycles, or simply, cycles.
Q56. What is meant by fatigue life?
A56. (NFA, ABC, JRS, RGL, MAG) For any material, there is a characteristic number of stress cycles that it can withstand at a particular applied stress level. The American Society 32
for Testing and Materials (ASTM) defines fatigue life (Nf) as the number of stress cycles that a material specimen can sustain before failure occurs. See ASTM Vol. 03.01, Metals -
Mechanical Testing; Elevated and Low Temperature Tests; Metallography, E1823 - 10a (2005)
(ENT000187). However, this definition applies to laboratory test specimens, for which the initiation of a crack and propagation of that crack are essentially contemporaneous.
Q57. What is meant by fatigue design life of a component?
A57. (NFA, ABC, JRS, RGL, MAG) Fatigue design life, is defined as the number of fatigue cycles that can be tolerated by a component based on an evaluation using the ASME Code Section III fatigue methodology and design fatigue curve. The fatigue design curve is based on measurements of fatigue life in air, offset by a factor of two on stress or twenty on cycles to account for additional factors such as specimen size, data scatter, environment or surface condition that might impact the application. See K. R. Rao, Companion Guide to the ASME Boiler & Pressure Vessel Code at 731 (3rd ed. 2009) (Companion Guide to the ASME Code) (ENT000191). The offset produces a curve that bounds all of the measured data by a large margin. The fatigue life measurements used to set the requirements in NUREG/CR-5704 and NUREG/CR-6909 defined fatigue life in terms of a 25% load drop in the test specimen, which was roughly equivalent to a 3 mm flaw. This corresponds to the onset of fatigue cracking.
The actual number of fatigue cycles that can be tolerated by a component before a crack will propagate sufficiently to cause loss of pressure boundary functionality or some other mode of failure is greater than the number of cycles that could lead to crack initiation. See ASME, Criteria of the ASME Boiler and Pressure Vessel Code for Design by Analysis in Sections III and VIII, Division 2 (1969) at 20 (Criteria for ASME Code Design) (ENT000188). Therefore, fatigue design life of a component is shorter than the fatigue life because the design life is 33
defined in terms of the onset of cracking and contains a large offset to accommodate factors that might impact the application.
Q58. What NRC regulations in 10 C.F.R. Part 50 and ASME Code specifications govern a licensees evaluation of metal fatigue?
A58. (NFA, ABC, JRS) 10 C.F.R. § 50.55a(c)(3) and (c)(4) require that the RCS pressure boundary meet the requirements of the original construction Code and supplemental Owners requirements, or a later version of ASME Code,Section III, as endorsed by the NRC in 10 C.F.R. § 50.55a with any applicable conditions or limitations. For license renewal, Westinghouse evaluated the RVIs for EAF in accordance with ASME Code,Section III, Subsection NG.
Q59. What codes and standards govern the number of fatigue cycles that safety-related components at IPEC can withstand?
A59. (NFA, ABC, MAG, RGL) The ASME and ANSI codes provide this information.
The design specification for a given safety-related component conservatively specifies the numbers of mechanical and thermal cycles that the component is expected to experience during its operation and defines the safety limits and applicable codes that must be satisfied. For components constituting the RCS pressure boundary, the specified requirements to account for cyclic loading and thermal conditions are contained in Section III of the ASME Code or ANSI B31.1. See ASME Code,Section III, Article NB-3000 § 3200 (1989) (ASME Code,Section III, Article NB-3000) (NYS000349); LRA at 4.3-18 (ENT00015B). Similar requirements govern RVI components, based on ASME Code Section III, Subpart NG. The ASME developed Subpart NG specifically to address RVI components.
34
Q60. What is the CUF?
A60. (NFA, ABC, MAG, JRS, RGL) CUF, or cumulative usage factor, quantifies the fatigue that a particular metal component experiences during plant operation, in comparison to its allowable design life. See Criteria for ASME Code Design at 18 (ENT000188).
Mathematically, it is the sum of the ratios of an actual or assumed number of design basis cycles (n), at various stress ranges, to the allowable number of cycles (N) from the ASME Code design fatigue curve, for each of these various alternating stress ranges. See id. Unless noted otherwise, CUF, in this testimony, refers to projected values that are based on the assumption that every analyzed transient has occurred the analyzed number of times.
Q61. Does the ASME Code define the allowable CUF value for components?
A61. (NFA, ABC, MAG, JRS, RGL) Yes. For component design purposes, ASME Code Section III requires that the CUF not exceed unity or 1.0; i.e., the total number of assumed cycles for design is not to exceed the allowable number of stress cycles, consistent with the fatigue design criteria. A CUF of less than one provides reasonable assurance that the component will not fail by fatigue cracking during its operation. But exceeding the criterion does not necessarily mean the component will exhibit fatigue cracking, given margins and conservatisms in the analytical process. See Criteria for ASME Code Design at 20 (ENT000188).
Importantly, the design CUF value is not indicative of the current condition of any component, or of any potential for fatigue cracking at the present time. See id. Instead, it represents a calculation of the condition at the end of life, assuming that every postulated transient included in the EAF analysis has taken place. A CUF value greater than 1.0 merely 35
indicates that, after all of the postulated transients have taken place, there is a potential for cracking at the affected location. See id.
Q62. Dr. Hopenfeld appears to take issue with these fundamental principles. Does he address ASME Code provisions in his testimony?
A62. (NFA, ABC, MAG, JRS, RGL) No. Dr. Hopenfeld suggests that existing reactor safety studies never addressed scenarios where many key safety components, such as those analyzed by Entergy, were allowed to remain in service without adequate useful fatigue life remaining. Supplemental Hopenfeld Report at 2 (RIV000144). However, there simply is no basis for his premise that components will be left in service at IPEC without adequate useful fatigue life. See id. The fatigue design life, as defined by the ASME Code and endorsed by NRC in 10 CFR 50.55a, is that associated with a CUF of 1.0. Moreover, as we will further explain below, the EAF evaluations prepared for IPEC contain considerable margin and conservatisms.
Q63. Dr. Lahey has suggested that significant cracking is expected of a component when it reaches a CUF = 1.0. Lahey Report at 24 (NYS000296). Do you agree?
A63. (MAG, NFA) No. This interpretation is inconsistent with the ASME Code principles we have described in this section. As we have shown, reaching a CUF of 1.0 does not translate to an expectation of significant cracking as posited by Dr. Lahey. See id. He ignores the ASME Code and the substantial design margins specified therein. In addition, beyond those design margins, even if a crack were to initiate when the CUF exceeds 1.0, additional load cycles would be required before component failure. See Criteria for ASME Code Design at 20 (ENT000188). Moreover, the relatively thin specimens used to obtain the test data that form the 36
basis for the ASME Code fatigue design curves have a very short cyclic life beyond detectable crack initiation, whereas actual, thicker, plant components that may be subject to fatigue crack initiation at a localized point on the surface of the component have much longer cyclic life after crack initiation. See id.
Q64. Do Dr. Lahey and Dr. Hopenfeld acknowledge the principle that the CUF evaluation is intended to assure there are no structurally significant cracks?
A64. (MAG, RGL) Dr. Lahey appears to, but Dr. Hopenfeld does not. Dr. Lahey has stated that [t]he maximum number of cycles which can be experienced by a structure or component before significant cracking is expected occurs when CUF = 1.0 . . . . Lahey Report at 24 (NYS000296). Dr. Hopenfeld, however, postulates that microscopic cracks can propagate and become a significant concern even when the CUFen < 1.0. See Supplemental Hopenfeld Report at 17-18 (RIV000144).
As recognized in ASME Code Section XI (endorsed in 10 C.F.R. § 50.55a(b)), the available evidence, however, suggests that such microcracking would have a negligible effect on structural strength.Section XI therefore distinguishes between structurally relevant defects and other defects which have no measurable effect on the load-carrying capability of the component.
See, e.g., ASME Boiler & Pressure Vessel Code,Section XI, Subsection IWB, Requirements for Class 1 Components of Light-Water Cooled Plants at tbl. IWB-3514-1; § IWB-3600 (2010)
(ENT000665).
For example, unirradiated austenitic stainless steels are expected to be highly flaw tolerant, and growth of a 3 mm flaw should not significantly reduce the load carrying capacity of the component. See EPRI, MRP-210, Materials Reliability Program: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Internal Components at 4-1 (Dec. 2007) 37
(MRP-210) (ENT000646). In fact, this study demonstrated that even after accounting for significant losses in fracture toughness, the critical flaw sizes for anticipated accident loads were significantly larger than 3 mm. See id.
Q65. Dr. Hopenfeld claims that merely because a CUF is less than 1.0 is not an adequate basis to conclude that fatigue is not a concern. See Supplemental Hopenfeld Report at 17-18 (RIV000144) (ANL [Argonne National Laboratory] and ORNL research invalidates the position that when CUFen values are less than one, fatigue initiation of cracks is not expected . . . .) (citing G. T. Yahr et al., Case Study of the Propagation of a Small Flaw Under PWR Loading Conditions and Comparison with the ASME Code Design Life, Comparison of the ASME Code Section III and XI (Nov. 1984) (RIV000118)) (Yahr Study). How do you respond?
A65. (NFA, ABC, JRS, MAG, RGL) As an initial matter, Dr. Hopenfelds argument is not with Entergys LRA or the Westinghouse fatigue analyses, but with the general principles of fatigue analysis set forth in the ASME Code and NRC regulations that incorporate the Code by reference. In any event, as to the Yahr Study referenced by Dr. Hopenfeld, that document is approximately 30 years old, and its recommendations have long ago been incorporated into the ASME Code, where appropriate. The Yahr Study was part of a larger effort at the time and was not intended to provide final recommendations. See id. at 2 (Therefore, the present study was initiated to determine whether these concerns are justified . . . . The present study was not designed to be a sufficient basis for recommending Code changes . . . .); see also id. at 24 (Conclusions, ¶ 4) (The current study is not a sufficient basis for defining any specific changes in either Sect. III or Sect. XI of the Code.). The Yahr Study itself suggests that recent 38
changes to the ASME Code (i.e., changes in the 1980s), may have alleviated some of the concerns that prompted the Yahr Study. See id. at 24 (Conclusions, ¶ 2).
Thus, the Yahr Study is outdated and identifies no deficiency in the current ASME Code or the IPEC EAF evaluations.
Q66. Dr. Hopenfeld also suggests that microscopic fatigue crack initiation and propagation can take place before the CUFen reaches 1.0. See Supplemental Hopenfeld Report at 17-18 (RIV000144) (citing Mohanty et al., A Review of Stress Corrosion Cracking/Fatigue Modeling for Light Water Reactor Cooling System Components (June 2012) (RIV000154)). How do you respond?
A66. (NFA, RGL, JRS) Although Dr. Hopenfeld asserts that laboratory data support his position, as mentioned in Question 65, he provides no evidence that the microscopic cracking caused by fatigue loading will result in a reduction in the ability of the component to withstand an accident load. On the contrary, as indicated in the response to Question 64, such small cracks are not considered structurally significant. The primary pressure boundary and RVI components included in the EAF evaluation are significantly larger than 3 mm. Such materials are highly flaw tolerant and, even under irradiated conditions, RVI components with flaws significantly larger than 3 mm would withstand accident loads. See MRP-210 at 4-1 (ENT000646).
As for the information in the Mohanty paper, it provides a survey of previously-published work on primary water stress corrosion cracking (PWSCC) in nickel alloy materials. A fatigue analysis is not intended to address PWSCCit is intended to provide reasonable assurance that a component will not experience fatigue cracking. See Question 61. The effects of aging due to PWSCC on susceptible primary plant components, including dissimilar metal welds, are managed through several inspection programs which address potential cracking (regardless of 39
the source mechanism), including the inservice inspection (ISI) Program, the Nickel Alloy Inspection Program, the Reactor Vessel Head Penetration Inspection Program, the Steam Generator Integrity Program, and the RVI AMP. See LRA App. B at B-63, B-74, B-109, B-118 (ENT00015B); NL-12-037, Letter from F. Dacimo, Entergy, to NRC Document Control Desk, License Renewal Application - Revised Reactor Vessel Internals Program and Inspection Plan Compliant with MRP-227-A, Attach. 1 (Feb.17, 2012) (NL-12-037) (NYS000496).
Q67. Returning to the discussion of general principles, please describe what is meant by EAF analysis.
A67. (NFA, ABC, MAG) As an initial matter, it is important to understand that the ASME Code design fatigue curves were developed on the basis of laboratory testing of smooth, polished specimens in air at a constant strain amplitude, with factors incorporated into the curves to account for certain variables such as atmosphere, specimen size, surface finish, etc. See NUREG/CR-6909 at xv, 1-5 (NYS000357); see also Memorandum from A. Thadani to W.
Travers, Closeout of Generic Safety Issue 190, Fatigue Evaluation of Metal Components for 60-year Plant Life, Attach. 1 (Dec. 26, 1999) (GSI-190 Closeout Memorandum)
(ENT000190).
EAF analysis is a fatigue analysis that adjusts for the potential reduction in fatigue life caused by the reactor coolant environment. See SRP-LR at 4.3-2 (NYS000195). For components exposed to reactor coolant water, the fatigue design life, as measured by the actual number of stress cycles tolerated prior to the onset of fatigue cracking, may be reduced compared to a components fatigue life in air. The cause of the lower fatigue life is the fact that corrosion of the material increases the effects of fatigue compared to the effects in an air environment. See NUREG/CR-6815, Review of the Margins for ASME Code Design Curve - Effects of Surface 40
Roughness and Material Variability at 10 (Sept. 2003) (NUREG/CR-6815) (ENT000225) (In LWR environments, the growth of small cracks in carbon and low-alloy steels occurs by a slip oxidation/dissolution process, and in austenitic SSs, most likely, by mechanisms such as H-enhanced crack growth.); see also id. at 9-21.
Therefore, a correction factor is applied to the calculated CUF to account for reactor coolant environmental conditions, such as DO and temperature. See NUREG/CR-6909 at 3 (NYS000357). A fatigue analysis that accounts for additional effects of operating in a reactor coolant environment is called an EAF analysis.
Q68. Has the NRC Staff determined whether applicants should address the effects of the reactor coolant environment on fatigue life?
A68. (NFA, ABC, JRS, MAG) Yes. In 1996, the NRC Staff issued GSI-190, Fatigue Evaluation of Metal Components for 60-Year Plant Life, which addresses the subject of fatigue evaluations for a 60-year plant life. The NRC Staff closed out GSI-190 in December 1999, without imposing additional requirements because it found only negligible calculated increases in core damage frequency between the 40-year and 60-year license periods. See GSI-190 Closeout Memorandum Attach. 1, at 5 (ENT000190). However, the NRC Staff concluded that applicants should address the effects of coolant environment on reactor coolant pressure boundary component fatigue life as part of license renewal, due to the potential for increased pressure boundary leakage from through-wall cracking as a result of fatigue. See id.; id. Attach.
2, at 1.
41
- 2. Design Margin and Other Conservatisms in Fatigue Analysis Q69. Turning to the issue of margin and conservatisms in fatigue analyses, please explain the difference between the design margins in the ASME Code and the conservatisms in fatigue calculations.
A69. (NFA, MAG, JRS, RGL) Fatigue calculations, including EAF calculations, include both Code-specified design margin and various conservatisms that are retained at the discretion of the analyst.
With respect to design margins, as we will explain in this section, the ASME Code fatigue design curves are best-fit curves based on experimental data, developed by applying the more conservative of a factor of two on stress or twenty on cycles to the best fit curves.
Companion Guide to the ASME Code at 731 (ENT000191). The Code also provides margins in stress allowables, worst case mean stresses and plasticity correction factor (i.e., Ke). These margins are inherent in the Code and cannot be changed by the analyst.
Conservatisms, in contrast, remain in the fatigue analysis at the discretion of the analyst.
We will further explain the various conservatisms normally present in fatigue analysis below.
Because these conservatisms are above and beyond the design margins required by the ASME Code and regulations, the licensee retains the option, consistent with established engineering practice, to perform more refined fatigue analyses to remove some of these conservatisms.
Q70. What are the design margins in the ASME Code for fatigue?
A70. (MAG, NFA, JRS) The key fatigue design margins prescribed in the ASME Code are the adjustment factors in the design fatigue curves and the design margin in the stress allowables.
Adjustment Factors in the Design Fatigue Curves: The ASME Code fatigue design curves used in the IPEC EAF evaluations have a built-in adjustment factor of 2 on stress or 20 on 42
cycles, whichever is more conservative, from the smooth, polished fatigue specimen data from laboratory testing. These adjustment factors were intended to cover such issues as scatter in the laboratory test data and size effect, and they contribute to the recognized overall fatigue design margin. See Criteria for ASME Code Design at 20 (ENT000188); see also Companion Guide to the ASME Code at 731 (ENT000191). Through this Code-mandated adjustment, all ASME Code fatigue analyses contain considerable inherent margin applied to the calculated design basis CUF. This margin between the Code-specified curves and the experimental data for carbon steel, low-alloy steel, and austenitic stainless steel, respectively, is illustrated in Figures 2-4, 2-14, and 2-22 from EPRI, Materials Reliability Program Evaluation of Fatigue Data Including Reactor Water Environmental Effects (Dec. 2001) (MRP-49) (ENT000214). In addition, as described in the Criteria for ASME Code Design (ENT000188), the margin with respect to significant fatigue cracking is at least a factor of three in air, as confirmed in experimental studies conducted in support of the ASME Code Class 1 fatigue design process. See Criteria for ASME Code Design at 20 (ENT000188). Thus, the adjustment factors in the fatigue design curves provide substantial margin.
Margin in the Design Stress Allowables: The design stress values used for allowable stress in ASME Class 1 fatigue analysis are developed with a prescribed design margin. The allowable stress is generally the lesser of 2/3 yield stress or 1/3 ultimate strength of the material.
This is further discussed in Section 4.2 of the Companion Guide to the ASME Code (ENT000191).
43
Q71. Beyond these design margins, what are the key conservatisms in the ASME Code fatigue analysis methodology?
A71. (MAG, NFA) Beyond the design margins, the key conservatisms include the following:
Stress Cycle Pairings: ASME Code,Section III, Article NB-3000 § 3222.4(e)(5)
(NYS000349) prescribes a conservative method for determining stress cycle pairings (stress cycle pairings are the calculated ranges of stress intensity between different states experienced by the component during various transients, and their associated cycles). The transients that result in the highest stress are paired with the transients that result in the lowest stress to produce the maximum stress intensity range. This approach introduces conservatism because these transients do not necessarily occur in the prescribed order, and are often separated by other transients with lower stress magnitudes. This approach leads to large calculated stress ranges, and is therefore conservative. For example, the graph and table on page A100 of CENC-1110 illustrates the process of stress cycle pairing. See Combustion Engineering, Inc, CENC-1110, Fatigue Evaluation of Head Flange Vessel Flange and Closure Studs at A100 (May. 1966)
(RIV00052C).
The Elastic-Plastic Penalty Factor (Ke): The elastic plastic penalty factor, Ke, is applied under ASME Code,Section III, Article NB-3000 § 3228.5, Simplified Elastic-Plastic Analysis (NYS000349) to account for fatigue cycling that may occur when the metal stress-strain cycling is in the plastic range. This penalty factor provides conservatism without requiring the more detailed and less conservative methods of a detailed inelastic analysis. A relatively recent ASME Code Case provides a more detailed, but less conservative, formulation for the Ke penalty for certain applications. See Cases of ASME Boiler and Pressure Vessel Code, Code Case N-44
779, Alternative Rules for Simplified Elastic-Plastic Analysis, Class 1 (Section III, Division 1)
(Jan. 26, 2009) (ENT000684). But Westinghouses EAF analyses applied the more conservative Ke provided in NB-3228.5. This factor, while it can be considered an ASME Code design margin, is categorized here as a conservatism because of the option to supersede this factor through a detailed inelastic analysis.
As explained in response to Question 69, conservatisms can be removed at the discretion of the analyst if necessary to show compliance with the CUF limit of 1.0, but this does not affect the design margins specified in the ASME Code. In addition, there are several other conservatisms in the EAF analyses.
Q72. What are the additional conservatisms that Westinghouse applied at IPEC, beyond those described above?
A72. (MAG, NFA) 45
46 Q73. Does NUREG/CR-6909 discuss the conservatisms in the EAF methodology?
A73. (MAG, NFA, JRS) Yes. NUREG/CR-6909 (NYS000357), a document on which Intervenors and their experts rely, see e.g., Lahey Report at 24 (NYS00296); Hopenfeld Report at 6 (RIV000035), explains that various entities, including the ASME, NRC, U.S. Department of Energy national laboratories, and EPRI, have evaluated the margin and conservatisms in the relevant ASME Code calculation inputs, analytical methods, and ASME Code fatigue design curves. See NUREG/CR-6909, § 7 (NYS000357). One report issued by Sandia and EPRI, and discussed in NUREG/CR-6909, states:
After review of numerous Class 1 stress reports, it is apparent that there is a substantial amount of conservatism present in many existing component fatigue evaluations. . . . It was concluded that the potential increase in predicted fatigue usage due to environmental effects should be more than offset by decreases in predicted fatigue usage if re-analysis were conducted to reduce the conservatisms that are present in existing component fatigue evaluations.
SAND94-0187, Evaluation of Conservatisms and Environmental Effects in ASME Code,Section III, Class 1 Fatigue Analysis at iii (Aug. 1994) (emphasis added) (SAND94-0187)
(ENT000189).
Q74. Dr. Lahey claims that Entergy is obligated to maintain its present day licensing basis . . . safety margins throughout the proposed 20 year PEO. Lahey Rebuttal Testimony at 11 (NYS000440). He has raised similar concerns about eroded . . .
safety margins and conservatisms, e.g., Revised Lahey Testimony at 73 (NYS000530), and CUFen values that are just below unity id. at 66. Do you agree?
A74. (ABC, JRS, NFA, MAG) Entergy intends to maintain the licensing basis margins associated with the CLB throughout the PEO under 10 C.F.R. Part 54. But for components with a CLB CUF, the licensing basis margins are embedded in the ASME Code 47
design limit of a CUF less than 1.0. Under 10 C.F.R. § 50.55a, the NRC has established that maintaining a CUF less than the ASME Code design limit of 1.0, in accordance with ASME Code design rules, provides reasonable assurance of public health and safety. Therefore, the notion suggested by Dr. Laheythat, in order to preserve design basis margin, the CUFen cannot be just below unity when projected to the end of the PEOis tantamount to changing the established design limit in the CLB to a lower value. This is not part of the license renewal process.
Nor is it logical. The components in a power plant begin their life with a CUF equal to zero. With each fatigue loading cycle during operation, the CUF increases as a normal and expected circumstance of operation, but the design margin (and the conservatisms) in the fatigue analysis remain in place. Under Dr. Laheys theory no plant could operate because even a single fatigue loading cycle would erode that margin. See Revised Lahey Testimony at 73, 75 (NYS000530). By analogy, if an automobile was required to permanently maintain the safety margin of its factory-condition tires, a motorist could never drive the car because even the slightest usage would reduce the tread. The appropriate question is how much wear can be safely tolerated. Here, the answer to that question is provided in the ASME Code, and incorporated in NRC regulations: as long as the CUF remains below 1.0, the licensing basis margins associated with the design remain intact and the CLB is maintained.
- 3. Fatigue and Other Aging Mechanisms Q75. Dr. Lahey generally criticizes Entergys fatigue calculations as using a silo approach that considered neither the effect of neutron-induced embrittlement nor the combined effects of fatigue damage and other degradation mechanisms, and argues that the most serious short-coming of this siloing approach is that synergistic interactions between radiation-induced embrittlement, corrosion-induced cracking, and fatigue-48
induced degradation mechanisms have not been considered. Revised Lahey Testimony at 14-15 (NYS000530). How do you respond?
A75. (NFA, JRS, RGL) Dr. Laheys concerns regarding an allegedly inappropriate silo approach are without foundation. As an initial matter, the NRCs license renewal regulations in 10 C.F.R. Part 54 focus on the evaluation and management of aging effects, rather than individual aging mechanisms. The Commission articulated this intent clearly in the Statements of Consideration for the current Part 54. See Final Rule, Nuclear Power Plant License Renewal; Revisions, 60 Fed. Reg. 22,461, 22,469 (May 8, 1995) (Part 54 SOC)
(NYS000016). Thus, rather than addressing the individual mechanismssuch as various types of fatigue, embrittlement, or corrosionthe NRCs license renewal process has long been focused on managing the effects of fatigue, embrittlement, and corrosion.
EAF analyses are not the only methods used to manage the effects of irradiation or fatigue on RVIs. The risk-prioritized inspections in the RVI AMPwhich consider the combined effects of multiple aging mechanismsprovide further assurance that the effects of aging will be adequately managed for RVI components throughout the PEO. See Entergys NYS-25 Testimony at Section VII.A (ENT000616); see also MRP-191, EPRI, Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Designs at 6-6 (Nov. 2006) (NYS000321).
For reactor coolant pressure boundary components, the ISI Program provides further assurance that the effects of aging will be adequately managed throughout the PEO. In addition, cracking due to PWSCC is managed through several other inspection programs, as we have previously explained. See Response to Question 66, above.
49
Q76. Do fatigue and embrittlement interact synergistically?
A76. (NFA, JRS, RGL) No. Both fatigue and irradiation embrittlement contribute to potential degradation, but they do not interact synergistically. Irradiation may have a positive or negative effect on the load carrying capability of the material, depending on the circumstances. One example of a positive effect on fatigue is provided in the work of P.
Shahinian et al, NRL Report 7446, Effect of Neutron Irradiation on Fatigue Crack Propagation in Types 304 and 316 Stainless Steels at High Temperature at 10-12 (July 21, 1972) (ENT000636),
which reported a reduction in fatigue crack growth rates in type 304 and 316 stainless steels irradiated under fast reactor conditions at temperatures up to 800°F.
Fatigue crack propagation depends on a number of factors; however, increased strength generally tends to increase the resistance to fatigue crack growth. See G. Was, FUNDAMENTALS OF RADIATION MATERIALS SCIENCE: METALS AND ALLOYS; PART III: MECHANICAL EFFECTS OF RADIATION DAMAGE at 689-90 (2007) (Was Text) (ENT000627). Similarly, irradiation effects also increase the material strength and fatigue resistance but decrease the ductility and fracture toughness (i.e., the ability of the material to resist fast fracture) of the material. See id. at 686 -
689. These mixed effects can be offsetting, and the results have been demonstrated experimentally. For example, as explained in MRP-175, [t]he work of several researchers suggest that neutron irradiation does not result in a further reduction in fatigue properties and in some cases suggests an improvement. MRP-175 at D-3 (ENT000631). While MRP-175 acknowledges that there is limited literature addressing this topic, see id., Draft NUREG/CR-6909 concludes that, although, the data in this area are inconclusive, the EAF methodology is appropriate for materials exposed to significant levels of irradiation. See Draft NUREG/CR-6909, Rev. 1 at 9 (NYS000490). Therefore, there is no basis, at this time, to conclude that an 50
additional correction factor is necessary to account for the effects of irradiation embrittlement on fatigue life. Moreover, as explained in Entergys testimony on NYS-25, the RVI AMP will manage the combined effects of these mechanisms through inspections, and provide reasonable assurance that the effects of aging (including the effects of fatigue and embrittlement) on RVIs will be adequately managed.
Q77. Dr. Lahey also suggests that ASME Code fatigue evaluations, even when adjusted to address environmental effects, do not include evaluations of the potential failure of highly fatigued structures and fittings external to the RPV due to a secondary side LOCA or any other thermal shock loads. Unfortunately, these transient loads can be much larger than what W has included in their fatigue failure evaluations for IP-2 & 3 and they can lead to the early failure of fatigue-weakened components. If so, this can lead to a primary side LOCA which, in turn, will challenge core cooling. Supplemental Lahey Report at 4 (NYS000297)); see also Revised Lahey Testimony at 15 (NYS000530)
([V]arious operational and accident-induced shock loads could cause failures well before the fatigue limit is reached (i.e., when CUFen < 1.0).). How do you respond?
A77. (NFA, ABC, JRS) The ASME Code fatigue analyses prepared for IPEC components are intended to ensure that components will perform their intended function under design basis loads. Because the updated CUFen values (considering fatigue from normal operation and transient conditions) of the relevant IPEC components are less than 1.0 (i.e.,
cracking is not predicted), there is no basis for Dr. Laheys suggestion that such components are more susceptible to failure when subjected to accident loads. In sum, as long as the Class 1 components continue to operate with a CUFen that does not exceed 1.0as has been demonstrated to be the case at IPECthen cyclic operation has not compromised the load-51
carrying ability of the material, and the Class 1 components remain within their design basis and fully capable of withstanding design basis loadings of all types. The NRCs license renewal process recognizes this principle, as the Staff explains in the SRP-LR:
The applicable aging effects to be considered for license renewal include those that could result from normal plant operation, including plant/system operating transients and plant shutdown.
Specific aging effects from abnormal events need not be postulated for license renewal. . . .
DBEs are abnormal events; they include: design basis pipe break, LOCA, and safe shutdown earthquake (SSE). Potential degradations resulting from DBEs are addressed, as appropriate, as part of the plants CLB. There are other abnormal events which should be considered on a case-by-case basis. For example, abuse due to human activity is an abnormal event; aging effects from such abuse need not be postulated for license renewal. When a safety-significant piece of equipment is accidentally damaged by a licensee, the licensee is required to take immediate corrective action under existing procedures (see 10 CFR Part 50 Appendix B) to ensure functionality of the equipment. The equipment degradation is not due to aging; corrective action is not necessary solely for the period of extended operation.
SRP-LR, App. A, § A.1.2.1 (NYS000195).
Q78. Dr. Lahey notes that the EAF evaluations do not take into account that the structures, components and fittings being analyzed may have experienced significant corrosion and irradiation-induced embrittlement, and thus can experience early fatigue-induced failures. Lahey Report at 25 (NYS000296). Dr. Hopenfeld makes similar claims in his 2015 testimony. Supplemental Hopenfeld Report at 28 (RIV000144) (discussing impact of potential PWSCC on fatigue evaluations). How do you respond?
A78. (NFA, JRS, RGL, ABC) We disagree with Drs. Lahey and Hopenfeld. First, none of the reactor coolant pressure boundary components with CLB CUF analyses exceed the threshold of 1.0 x 1017 neutrons/cm2 for irradiation damage set forth in 10 C.F.R. Part 50, Appendix H, including the reactor vessel inlet and outlet nozzles, which are the limiting 52
locations for fatigue. See NL-08-143, Letter from F. Dacimo, Entergy, to NRC, Additional Information Regarding License Renewal Application - Reactor Vessel Fluence Clarification, Attach. 1 at 1 (Sept. 24, 2008) (ENT0000231).
The portion of the reactor coolant pressure boundary that exceeds the threshold is the reactor vessel belt line region directly adjacent to the core which could experience neutron fluence greater than the 1.0 x 1017 neutrons/cm2 value set forth in Part 50, Appendix H. Any potential effects of neutron irradiation embrittlement on EAF evaluations would be limited to the RPV beltline and RVI components. The limiting locations for fatigue, however, are the reactor vessel inlet and outlet nozzles. See LRA at 4.3-9, 4.3-10 (ENT00015B). This is because the beltline region of the reactor vessel has no nozzles that create structural discontinuities, thereby eliminating stress concentrations and resulting in very low CUFs. Therefore, the interaction between fatigue and irradiation damage is not a new or significant concern for the pressure boundary components.
The impact of irradiation on the fatigue life evaluations of the RVIs was previously summarized in our response to Question 75, and we will further discuss this issue in Section V.E.2, below. As to the interaction between corrosion and fatigue, this potential has been recognized and addressed. This is precisely the focus of the environmental correction factors, which account for the fact that cyclical loads applied in the reactor coolant environment will result in a decrease in the fatigue life of the component compared to the fatigue life for the same cyclical loads applied in air environment. Furthermore, with respect to the potential for SCC to affect primary plant components, such effects on susceptible components are managed through other inspection programs, as we discussed in response to Question 66, above.
53
Q79. Dr. Hopenfeld asserts that flow-accelerated corrosion (FAC) can lead to non-uniform stress distributions in low-alloy components. Hopenfeld Report at 15 (RIV000035). He also asserts that the CLB CUFs must be revised because [t]he geometry of many components at IP2 and IP3 have changed over the past 40 years due to flow-accelerated corrosion (FAC). Supplemental Hopenfeld Report at 20 (RIV000144). Is this issue a concern for the Westinghouse EAF evaluations?
A79. (NFA, RGL) No. As its name implies, the phenomenon of EAF relates to the reactor coolant environment. The EAF evaluations cover reactor coolant pressure boundary components, all of which are either constructed of or clad with stainless steel. At IPEC, there are no carbon or low-alloy steel surfaces in contact with primary coolant. Since stainless steels are not susceptible to FAC, this issue is not relevant to the IPEC EAF analyses. See Riverkeeper Counter-Statement of Material Facts at 8 (Aug. 16, 2010) (Attachment 1 to Riverkeeper Opposition to Entergys Motion for Summary Disposition of Riverkeeper Technical Contention 2 (Flow-Accelerated Corrosion) (Riverkeeper Counter-Statement of Material Facts), available at ADAMS Accession No. ML102371214 (acknowledging that it is [u]ndisputed that stainless steel is a FAC-resistant material).
B. Applicable 10 C.F.R. Part 54 Requirements and NRC Guidance Q80. Please identify and briefly describe the NRC aging management review (AMR) requirements applicable to IPEC systems, structures, and components (SSCs).
A80. (ABC, JRS) 10 C.F.R. Part 54 governs the matters that must be considered for purposes of operating license renewal, and in this adjudicatory proceeding. Section 54.4 defines the plant SSCs that are within the scope of the license renewal rule based on their intended functions. Part 54 also requires an aging management review of in-scope SSCs that are subject to AMR and evaluation of time-limited aging analyses (TLAAs).
54
Q81. How do the NRC regulations define TLAAs?
A81. (ABC, JRS) TLAAs are calculations and analyses that: (1) involve SSCs as delineated in § 54.4(a); (2) consider the effects of aging; (3) involve time-limited assumptions defined by the current (e.g., 40-year) operating term; (4) were determined to be relevant by the licensee in making a safety determination; (5) involve conclusions or provide the basis for conclusions related to the capability of the SSC to perform its intended functions; and are (6) contained or incorporated by reference in the CLB. 10 C.F.R. § 54.3(a).
Q82. How do license renewal applicants evaluate TLAAs for the PEO?
A82. (ABC, JRS) NRC regulations require applicants to either: (1) demonstrate that a TLAA remains valid for the PEO (10 C.F.R. § 54.21(c)(1)(i)); (2) revise the analyses to remain valid for the PEO (10 C.F.R. § 54.21(c)(1)(ii)); or (3) demonstrate that the effects of aging on the intended functions of the SSC will be adequately managed for the PEO (10 C.F.R.
§ 54.21(c)(1)(iii)). The third option does not rely on demonstrating the validity of the TLAA throughout the PEO prior to issuance of a renewed license; instead, the third option relies upon an AMP for managing the effects of aging during the PEO.
Q83. Can any of the above three options for evaluating a TLAA be applied to managing the effects of fatigue?
A83. (ABC, JRS) Yes. But in this case, Entergy is managing the effects of fatigue through the elements of an AMP pursuant to 10 C.F.R. § 54.21(c)(1)(iii). This involves determining the number of cycles at which the CUF would be expected to reach its allowable limit, and then tracking the number of cycles through an appropriate AMP during the PEO in accordance with 10 C.F.R. § 54.21(c)(1)(iii). If, based on the actual accrued numbers of cycles, a components CUF value is expected to exceed 1.0 during the PEO, then corrective actions 55
including more rigorous analyses, repair or replacement would be taken before the CUF exceeds 1.0. In the remainder of this testimony, we explain how, at IPEC, the effects of fatigue will be adequately managed in accordance with 10 C.F.R. § 54.21(c)(1)(iii).
Q84. What findings must the NRC make to issue a renewed operating license?
A84. (ABC, JRS) 10 C.F.R. § 54.29(a) requires a finding that the applicant has identified and has taken, or will take, actions to address the TLAAs that have been identified for review under 10 C.F.R. § 54.21(c). Specifically, the NRC must find that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the plants CLB during extended operation. 10 C.F.R. § 54.29(a); see also id.
§ 54.21(c)(1)(iii). Importantly, the standard for this demonstration is one of reasonable assurance. See 10 C.F.R. § 54.29(a); Part 54 SOC, 60 Fed. Reg. at 22,479 (NYS000016) (the
[license renewal] process is not intended to demonstrate absolute assurance that structures or components will not fail, but rather that there is reasonable assurance that they will perform such that the intended functions . . . are maintained consistent with the CLB). The Commission has recognized that adverse aging effects generally are gradual and thus can be detected by programs that ensure sufficient inspections and testing. See id. at 22,475.
Q85. What guidance documents has the NRC Staff issued to assist in applicants in implementing the requirements of 10 C.F.R. Part 54?
A85. (ABC, JRS) The two primary guidance documents issued by the NRC Staff are NUREG-1801 (NYS000146) and the SRP-LR (NYS000195).
Q86. Please describe the function of SRP-LR as it relates to AMPs.
A86. (ABC, JRS) The SRP-LR provides guidance to NRC staff for conducting their review of LRAs. It provides acceptance criteria for determining whether the applicant has met 56
the requirements of the NRCs regulations in 10 C.F.R. § 54.21. See SRP-LR § 3.1.2 (NYS000195). For each of the SSCs identified as subject to aging management, one acceptable way to manage aging effects for license renewal is to use an AMP that is consistent with NUREG-1801. See id. § 3.0.1.
Q87. Please describe the origin and purpose of NUREG-1801.
A87. (ABC, JRS) In an NRC Staff paper, SECY-99-148, Credit for Existing Programs for License Renewal, dated June 3, 1999, the Staff described options for crediting existing licensee AMPs to satisfy the requirements of Part 54. By a Staff Requirements Memorandum (SRM) dated August 27, 1999, the Commission directed the Staff to develop NUREG-1801, which the Staff first issued in 2001, to document its evaluation of existing aging management programs. NUREG-1801 is referenced as a technical basis document in the SRP-LR.
The purpose of NUREG-1801, also referred to as the GALL Report, is to provide generic aging management review results of SSCs in the scope of license renewal. It also identifies and describes generic AMPs that the NRC Staff has found acceptable for managing the effects of aging on SSCs, based in part on the experience with evaluations of existing programs at operating plants during the initial license period. See NUREG-1801, Rev. 1, at 1-2 (NYS00146A). An applicant may reference NUREG-1801 in an LRA to show that its AMPs are consistent with those reviewed and approved in NUREG-1801. See id. at 3 (NYS00146A).
The original GALL Report was issued in July 2001. See NUREG-1800, Rev. 1 at 4.4-6 (NYS000195). Revision 1 was issued in September 2005. See NUREG-1801 at i (NYS00146A). Revision 2 was issued in December 2010. See NUREG-1801, Rev. 2 at i (NYS00147A). The revisions reflect further lessons learned from the reviews of LRAs, 57
operating experience, and other public input including industry comments. See NUREG-1801, Rev. 2 at 3 (NYS00147A).
Q88. Please describe the basic format and content of NUREG-1801, Revision 1.
A88. (ABC, JRS) NUREG-1801, Revision 1 includes tables summarizing various structures and components, the materials from which they are made, the environment to which they are exposed, the relevant aging effects (e.g., cracking due to fatigue, loss of material through pitting, leaching or corrosion), the AMP found to effectively manage the particular aging effect in that component, and whether a further evaluation is necessary. NUREG-1801, Rev.
1, at 5 (NYS00146A). The evaluation results documented in NUREG-1801, Revision 1 indicate that many existing programs are adequate without change to manage aging effects on particular structures or components for purposes of license renewal. Id. at 4. NUREG-1801, Revision 1 also contains recommendations concerning specific areas for which existing programs should be augmented for purposes of license renewal.
Q89. How does the NRC evaluate AMPs in light of NUREG-1801 A89. (ABC, JRS) The NRC Staff reviews AMPs listed in a license renewal application for consistency with NUREG-1801. See SER at 3-1 to 3-5 (NYS00326B). For plant-specific AMPs, the NRC Staff will review whether the applicants new AMP addresses the ten elements of an AMP, as specified in SRP-LR, Appendix A. See SRP-LR, Rev. 2 at A.1-3 (NYS000161).
Q90. Was NUREG-1801 revised following the preparation, submittal and NRC Staff review of the IPEC LRA?
A90. (ABC, JRS) Yes. In December 2010, the NRC Staff issued NUREG-1801, Revision 2. This revision was issued more than three years after the IPEC LRA was submitted, and more than a year after the NRC staff issued its original SER on the IPEC LRA in August 58
2009. Therefore, Entergy prepared the IPEC LRA using the guidance in NUREG-1801, Revision 1.
Q91. With respect to the issues raised in this contention, are there significant changes in NUREG-1801, Revision 2? If so, please explain.
A91. (ABC, JRS) NUREG-1801, Revision 2 has two significant changes that relate to the FMP and the EAF issues raised in this contention. The first is a change to the NRC Staffs discussion of the critical components identified in NUREG/CR-6260. See NRC Regulatory Issue Summary 2011-05, Information on Revision 2 to the Generic Aging Lessons Learned Report for License Renewal of Nuclear Power Plants at 4 (July 1, 2011) (RIS 2011-05)
(ENT000192).
As explained further in response to Question 95, NUREG-6260 identified a set of representative components for environmental fatigue analysis purposes based on high fatigue usage and risk importance. See NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components at xx, xxi (Feb. 1995)
(NUREG/CR-6260) (NYS000355). In NUREG-1801, Revision 1, the Staff specified that the sample of components to be evaluated for EAF is to include the locations identified in NUREG/CR-6260, as minimum, or [the applicant should] propose alternatives based on plant configuration. NUREG-1801, Rev. 1, at X M-1 (NYS00146C). The new guidance in NUREG-1801, Revision 2 states that [t]his sample set should include the locations identified in NUREG/CR-6260 and additional plant-specific component locations in the [RCS] pressure boundary if they may be more limiting than those considered in NUREG/CR-6260. NUREG-1801, Rev. 2, at X M1-2 (NYS00147C).
59
The second significant change to the discussion of fatigue monitoring in NUREG-1801, Revision 2 is a change to the set of approved formulae for evaluation of environmental effects on fatigue. In NUREG-1801, Revision 1, the NRC Staff approved the use of the formulae in NUREG/CR-6583 at pages 65 to 66 (NYS000356) for carbon and low-alloy steels, and in NUREG/CR-5704 at page 31 (NYS000354) for austenitic stainless steels. In NUREG-1801, Revision 2 (at X M1-1), the NRC approved the use of any one of the following options:
- 1. For carbon and low-alloy steels, either the formulae set forth NUREG/CR-6583 (NYS000356), or in Appendix A of NUREG/CR-6909 (NYS000357), or a staff-approved alternative.
- 2. For austenitic stainless steels, either the formulae set forth in and NUREG/CR-5704 (NYS000354), or in NUREG/CR-6909 (NYS000357),
or a staff-approved alternative.
- 3. For nickel alloys, either the formulae set forth in NUREG/CR-6909 (NYS000357), or a staff-approved alternative.
Q92. Did Entergy revise its LRA to account for the changes in NUREG-1801, Revision 2?
A92. (ABC, JRS, MAG, NFA) Yes, as we will explain in Section V.E, below, since the issuance of NUREG-1801, Revision 2, Entergy has revised its FMP to meet the intent of the new guidance.
Q93. Has the NRC Staff recently issued a draft revision to NUREG/CR-6909?
A93. (MAG, RGL, NFA) Yes. In March 2014, the NRC Staff issued a draft proposed update to the original NUREG/CR-6909 (i.e., Revision 0). See Draft NUREG/CR-6909, Rev. 1 (NYS000490A-B). The Staff also issued a draft revision to Regulatory Guide 1.207. See Draft Regulatory Guide DG-1309 (Proposed Revision 1 of Regulatory Guide 1.207, dated March 2007), Guidelines for Evaluating the Effects of Light-Water Reactor Coolant Environments in Fatigue Analyses of Metal Components (Nov. 2014) (DG-1309) (NYS000523).
60
Q94. What are the most significant proposed changes to the guidance in the draft revision to NUREG/CR-6909?
A94. (MAG, RGL, NFA) The most significant change between Revision 0 and draft Revision 1 is that the Fen formulas were revised for each material exposed to the reactor coolant environment. In general, the Fen values from draft Revision 1 would be lower than in Revision 0.
For example, for carbon/low-alloy steel in PWRs at temperatures below the threshold and typical maximum DO of 0.05 ppm, Fen is 1.89 which is lower than Revision 0 for the same conditions.
At high temperatures, when the DO is within plant specifications (i.e., DO < 0.005 ppm), the proposed new Fen is 1.0. For stainless steels, the minimum Fen is 1.0 when temperature or strain rate are at their respective thresholds. The maximum Fen for stainless steel material is 14.1, which is slightly less than the Revision 0 maximum of 15.4.
Additionally, as explained further below, draft Revision 1 considers the potential effects of irradiation on Fen, and generally concludes that current available information does not warrant any additional accounting in the Fen values. In any event, the NRC has received comments on the draft Revision 1 to NUREG/CR-6909, and it remains under review.
Q95. Based on these regulations and guidance, how can an applicant address the effects of the reactor coolant environment on metal fatigue?
A95. (ABC, JRS, NFA, MAG) As introduced in Question 91, the applicant can address the effects of the coolant environment on component fatigue life by assessing the impact of the reactor coolant environment on a sample of critical components for the plant, using the formulae specified in NRC guidance. See NUREG-1801, Rev. 1, at X M-1 (NYS00146C).
61
Q96. You mentioned that NUREG/CR-6260 identifies a set of critical locations for fatigue usage. What are those critical locations?
A96. (ABC, JRS, NFA, MAG) NUREG/CR-6260 identifies six generic critical locations for Westinghouse Plants. See NUREG/CR-6260 at xx-xxi (NYS000355). Those generic locations are: (1) the reactor vessel shell and lower head; (2) the reactor vessel inlet and outlet nozzles; (3) the pressurizer surge line (including hot leg and pressurizer nozzles); (4) RCS piping charging system nozzle, (5) RCS piping safety injection nozzle; and (6) residual heat removal (RHR) Class 1 piping. Id. The corresponding IPEC plant-specific locations are listed in Tables 4.3-13 and 4.3-14 in the LRA. See LRA at 4.3-24 to 4.3-25 (ENT00015B). As explained in NUREG/CR-6260, these components are not necessarily the locations with the highest design CUFs in the plant, but were chosen to give a representative overview of components that had higher CUFs and/or were important from a risk perspective. For example, the reactor vessel shell (and lower head) was chosen for its risk importance. NUREG/CR-6260 at xxi (NYS000355).
A97. (JRS, ABC) No. Because EAF (CUFen) analyses are not contained within an applicants CLB, they are not TLAAs themselves. See 10 C.F.R. § 54.3. Entergy prepared its EAF analyses as part of the FMP, which is subject to 10 C.F.R. §§ 54.21(a) (3) and (c)(1)(iii)
(i.e., they are part of the AMP used to address the CUF TLAA). The environmental effects of the reactor coolant water are applied to the analyses that are used in the FMP as the trigger for further actions before the CUFen of any component exceeds 1.0.
62
Q98. Alongside the FMP and the EAF evaluations conducted under it, are there other programs that provide further assurance that plant components will not fail due to fatigue damage?
A98. (ABC, NFA, JRS) Yes. Other AMPs provide defense-in-depth.
For example, the design basis fatigue analyses provide reasonable assurance that the component can be placed into service. Once placed in service, the ASME Section XI ISI Program is designed to detect and size any defects that might threaten the components ability to continue to perform its functions. See W. E. Cooper, The Initial Scope and Intent of the Section III Fatigue Design Procedures, in PRESSURE VESSEL RESEARCH COUNCIL, TECHNICAL INFORMATION FROM WORKSHOP ON CYCLIC LIFE AND ENVIRONMENTAL EVENTS IN NUCLEAR APPLICATIONS, Vol. 1, at 1, 7 (Jan. 20, 1992) (Cooper Paper) (ENT000215). The IPEC ASME Section XI ISI Program, therefore, also manages the effects of fatigue by inspecting reactor coolant system components for cracking caused by any mechanism. See LRA App. B at B-63 to
-68 (ENT00015B). Entergy periodically updates the ISI Program to address both IPEC-specific and industry operating experience. See id. at B-63, B-67 (ENT00015B). For example, as new operating experience is identified, the EPRI MRP disseminates that information throughout the industry for plants to take appropriate action under the ISI Program. The program reinforces our conclusion that there is reasonable assurance that the effects of fatigue will be adequately managed throughout the PEO. The Intervenors do not challenge the adequacy of the IPEC ISI Program.
For the RVIs, the RVI AMP, which is consistent with the NRC Staff-approved guidance in MRP-227-A, provides further assurance that components within the scope of that program will not fail due to fatigue damage. Specifically, even though none of the CUFen values for IP2 and 63
IP3 exceed 1.0, Entergy will conduct risk-prioritized inspections of components for cracking due to IASCC or fatigue, including areas that are susceptible to fatigue, such as the lower core plate.
See Entergys NYS-25 Testimony at Section VII.A (ENT000616). Together, the RVI AMP and the FMP provide reasonable assurance that the effects of aging will be adequately managed for RVI components throughout the PEO.
V. ENTERGYS LICENSE RENEWAL APPLICATION ADEQUATELY ADDRESSES METAL FATIGUE A. Overview of the License Renewal Application Q99. What section(s) of the LRA evaluate metal fatigue?
A99. (NFA, ABC) Chapter 4 of the IPEC LRA summarizes Entergys evaluation of metal fatigue and EAF on in-scope piping and components for the PEO. Section 4.3.1 (Class 1 Fatigue) addresses components and subcomponents that were designed in accordance with ASME Code,Section III. These components include the RPV, RVIs, pressurizer, steam generators, reactor coolant pumps, control rod drive mechanisms, Class 1 heat exchangers, and Class 1 piping and components. LRA Section 4.3.2 (Non-Class 1 Fatigue) addressesSection III Class 2 and 3 piping systems. LRA Section 4.3.3 (Effects of Reactor Water Environment on Fatigue Life) addresses EAF with respect to the critical component locations identified in NUREG/CR-6260.
Q100. What IPEC AMP addresses metal fatigue?
A100. (NFA, ABC) As previously noted, the AMP for fatigue described in the appendices to the LRA is the FMP. Appendix A to the LRA presents the information required by 10 C.F.R. § 54.21(d) relating to the AMP for fatigue monitoring that supplements the updated final safety analysis report (UFSAR) for IPEC. The supplements to the UFSAR, presented in Sections A.2 and A.3 of Appendix A for IP2 and IP3, respectively, contain summary descriptions 64
of the program for managing the effects of metal fatigue during the PEO. Appendix A indicates that the FMP will be enhanced to meet commitments in the LRA prior to the PEO. See LRA App. A at A-1, A-22, A-49 (ENT00015B); see also SSER 2 at A-3, A-14, A-15 (NYS000507).
In Appendix B to the LRA, Section B.1.12 describes the IPEC FMP, which is designed to track the number of transients for selected RCS components. LRA, App. B at B-1, B-44 (ENT00015B). Section B.1.12 indicates that it is consistent with the program described in NUREG-1801, Revision 1,Section X.M1 (Metal Fatigue of Reactor Coolant Pressure Boundary), but took one exception to the detection of aging effects program element. Id.
Entergy, however, later removed this exception, and it is not at issue in this contention. See NL-08-092, Letter from Fred R. Dacimo, Entergy, to NRC, Amendment 5 to License Renewal Application (LRA), Attach. 1 at 5 (June 11, 2008) (NL-08-092) (ENT000193).
Q101. Please explain how Entergy originally calculated CUFen values for Class 1 components in the LRA.
A101. (NFA, ABC) As discussed in LRA Section 4.3.1, components designed in accordance with ASME Code,Section III have fatigue analyses with resultant CUF values. LRA Tables 4.3-3 to 4.3-12 list the CUF values developed in the initial design calculations (based on assumed numbers of cycles for 40 years of operation) for the various Class 1 components and subcomponents. To calculate the projected CUFen values for NUREG/CR-6260 components in the LRA for 60 years of operation:
First, Entergy evaluated the numbers of cycles experienced through the time of LRA preparation to determine a rate of occurrence, which was then used to project the number of cycles expected to occur by the end of 60 years of plant operation, as shown in LRA Tables 4.3-1 and 4.3-2. For both plants, the number of cycles expected at the end of 60 years of plant operation is less than the number of cycles considered in the original design analyses for all but a few of the transients. See LRA at 4.3-2 (ENT00015B). Accordingly, Entergy determined that the CUF values for the NUREG/CR-6260 components with fatigue calculations were valid for the PEO.
65
Consistent with NUREG-1801, Revision 1, Entergy applied Fens calculated as described in NUREG/CR-6583 and NUREG/CR-5704 to the CUF values to determine CUFen values, as applicable. See LRA at 4.3-21, tbls. 4.3.13 (IP2), 4.3-14 (IP3) (ENT00015B).
Q102. What NUREG/CR-6260-specified component locations had projected CUFen values less than one?
A102. (NFA, ABC, JRS) As indicated in Tables 4.3.13 and 4.3.14, the CUFen values for all three reactor vessel locations (lower head to shell; reactor vessel inlet nozzle; and reactor vessel outlet nozzle) identified in NUREG/CR-6260 were less than 1.0 for both IP2 and IP3. In addition, the IP2 (but not IP3) pressurizer surge line nozzle had a CUFen less than 1.0.
Q103. What NUREG/CR-6260-specified component locations did not have projected CUFen values less than one?
A103. (NFA, ABC, JRS) There were four NUREG/CR-6260-specified component locations in LRA Tables 4.3-13 and 4.3-14 with projected CUFen values greater than 1.0, including the pressurizer surge line piping for IP2 and IP3; the RCS piping charging system nozzles for IP2; and the pressurizer surge line nozzle for IP3. See LRA at 4.3-24 to -25 (ENT00015B). No CUFen values were calculated for five NUREG/CR-6260-specified componentsthe RCS piping safety injection nozzles for IP2 and IP3, the RHR Class 1 piping for IP2 and IP3, and the RCS piping charging system nozzle for IP3because they are designed to ANSI B31.1 and, thus, no design fatigue analyses were required. See id.
To address these nine locations (the four components with CUFen values that exceeded 1.0 and the five components without CUFen values), Entergy committed to take one of the following actions: (1) refine the fatigue analyses, at least two years before entering the PEO, to determine valid CUFen values below the limit; (2) manage the effects of aging due to fatigue at the affected locations by an inspection program that has been reviewed and approved by the 66
NRC; or (3) repair or replace the affected locations before exceeding CUF of 1.0 (collectively referred to as Commitment 33). LRA at 4.3-22 to -23 (ENT00015B).
B. NRC Staff Review of the License Renewal Application
- 1. The NRC Staff Approved the FMP in the Original SER Q104. Please describe the NRC Staffs conclusions in its SER with respect to the FMP and the EAF evaluations.
A104. (ABC, NFA, JRS) As documented in its original SER in 2009, the NRC Staff conducted a detailed technical review of the LRA, which included a related onsite audit. It independently determined that IPECs FMP includes acceptable program elements that are consistent with recommendations in NUREG-1801, Revision 1,Section X.M1. See SER at 3-78 to 3-81 (NYS00326B). The Staff further concluded that Entergy has demonstrated that the effects of aging will be adequately managed so that the intended component functions will be maintained consistent with the CLB throughout the PEO, as required by 10 C.F.R. § 54.21(a)(3).
See id. at 3-79.
In addition to evaluating IPECs FMP, the Staff reviewed LRA Section 4.3.3 regarding EAF. As part of its review, the NRC Staff confirmed that Entergy had correctly accounted for the environmental factors used as inputs for calculating Fen factors. It found that Commitment 33 is consistent with 10 C.F.R. § 54.21(c)(1)(iii). Id. at 4-41, 4-44 (NYS00326E). The Staff also found the corrective action program element, see id. at 3-79 (NYS00326B), and the operating experience program element satisfy the recommendations in NUREG-1801, Revision 1 and the guidance in the SRP-LR, and therefore these program elements were acceptable. See id. at 3-81.
67
- 2. Issues Addressed in SSER 1 Q105. After issuance of the NRC Staffs SER, did the Staff issue additional metal fatigue-related requests for additional information (RAIs)?
A105. (NFA, ABC, JRS) Yes. The NRC Staff issued RAIs related to metal fatigue. See NL-11-032, Letter from F. Dacimo, Entergy, to NRC, Response to Request for Additional Information (RAI) Aging Management Programs, Attach. 1 at 25-26 (Mar. 28, 2011) (NL 032) (NYS000151). Specifically, the NRC Staff asked Entergy to confirm and justify that the IPEC plant-specific locations listed in LRA Tables 4.3-13 and 4.3-14 are equivalent to the generic NUREG/CR-6260 components, and to confirm that the plant-specific locations analyzed for EAF are the most limiting locations for the plant, beyond the generic NUREG/CR-6260 components. See id. at 26. The NRC Staff issued similar RAIs to other license renewal applicants following the issuance of NUREG-1801, Rev. 2. See, e.g., Letter from A. Cunanan, NRC, to S. Gambhir, Energy Northwest, Request for Additional Information for the Review of the Columbia Generating Station License Renewal Application for Metal Fatigue (TAC No.
ME3058) encl. at 1 (Feb. 3, 2011) (ENT000550).
The Staff also issued RAIs regarding the use of the Design CUF module in the WESTEMS' software, which referred to the modeling software used by Westinghouse in preparing the ASME CUF calculations in its EAF evaluations. First, the NRC Staff asked Entergy to commit to include a written explanation and justification of any user intervention in the analyses conducted using the WESTEMS' software, Design CUF module. See NL 032, Attach. 1 at 27 (NYS000151). Second, the NRC Staff asked Entergy for a commitment that the NB-3600 option of the WESTEMS' Design CUF module will not be used or implemented in the future at IPEC. See id. The Staff issued similar RAIs to other license renewal applicants as well.
68
These issues are further addressed in this Section, in the questions that follow, as well asSection V.D, below.
Q106. What is the WESTEMS' software?
A106. (MAG) WESTEMS' is a software code used by Westinghouse to conduct fatigue evaluations in support of license renewal and other activities at nuclear plants, including IPEC. WESTEMS' is a computer program that implements the requirements of the ASME,Section III Code.
Q107. Dr. Lahey has suggested that a WESTEMS' fatigue evaluation is not a standard ASME code evaluation[]. See, e.g., Lahey Report at 26 (NYS000296). Do you agree?
A107. (MAG) No. Dr. Lahey is incorrect. WESTEMS' performs a standard ASME Code fatigue evaluation. See, e.g., NUREG-2101, Safety Evaluation Report Related to the License Renewal of Salem Nuclear Generating Station at 3-162 to -188 (June 2011) (Salem SER) (ENTR00195) (showing the NRC Staff evaluation of the Salem WESTEMS' fatigue analyses and concluding that WESTEMS' performs a standard ASME Code fatigue evaluation); WCAP-17199, Rev. 1 at 1-1 (ENT000681) (The stress and fatigue evaluations . . .
were performed using the standard methods of the ASME Code,Section III).
69
Q108. In its February 2011 RAI, the NRC Staff referred to the Design CUF module in the WESTEMS' software. See NL-11-032, Attach. 1 at 27 (NYS000151). What is this module?
A108. (MAG)
Q109. In the same RAI, the NRC Staff used the term user intervention in the context of the WESTEMS' Module. See NL-11-032, Attach. 1 at 27 (NYS000151).
Please explain what that term means.
A109. (MAG) As part of any fatigue analysis, whether prepared by hand or using computer software, a qualified analyst must use accepted engineering methods and apply his or her expertise to determine the stress peaks and valleys to be used as input. The ASME Code provides guidelines to be used by the analyst for defining the transient stress cycle extremes (peaks and valleys) used in the stress and fatigue calculations. See ASME Code,Section III, Article NB-3000 at 67-88 (NYS000349). ASME Code Section NB-3222.4(e) also describes the process to be used in calculating fatigue usage factors from the stress peaks and valleys. See id.
at 80.
70
The NRC has now approved this practice. See NRC, Safety Evaluation Report, Topical Report on ASME Section III Piping and Component Fatigue Analysis Utilizing the WESTEMSTM Computer Code at 16 (WCAP-17577, Revision 2)
(undated) (SER for WCAP-17577) (ENT000687).
In summary, WESTEMS' simply uses an automated approach to assist the analyst in selecting the stress peak and valley times in each transienta process that, under traditional methods would be accomplished entirely by the analyst. See Letter from P. Davison, PSEG Nuclear, LLC, to NRC, Close-out of the NRC Audit Associated with Use of WESTEMS' Related to the Salem Nuclear Generating Station, Units 1 and 2 License Renewal Application Encl. A at 6-8 (Feb. 24, 2011) (PSEG Letter) (ENT000197); see also generally id. Encl. C (PVP2010-25891, Method for Selecting Stress States for Use in an NB-3200 Fatigue Analysis);
WESTEMS User Manual, Rev. 6, at 299-301 (ENT000686).
Q110. Please summarize Entergys response to the 2011 RAIs.
A110. (NFA, ABC, JRS) First, Entergy confirmed that the locations selected for EAF analysis in the IPEC LRA were equivalent to the generic components identified in the NUREG/CR-6260 guidance. See NL-11-032, Attach. 1 at 26 (NYS000151). Second, Entergy 71
committed to review the IPEC design basis ASME Code fatigue evaluations to determine whether the NUREG/CR-6260 locations that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting locations for IPEC (Commitment 43). See id. If more limiting locations were identified, then Entergy would evaluate the most limiting locations for the effects of the reactor coolant environment on fatigue usage. See id. This evaluation has now been completed for both units. See Westinghouse Calculation Note CN-PAFM-12-35 (NYS000510); Westinghouse Calculation Note CN-PAFM-13-32, Rev. 3 (ENT000683).
Entergy also made two additional commitments: to include a written explanation and justification of any user intervention in the analyses conducted using the WESTEMS' software (Commitment 44), and to not use the ASME Code,Section III, NB-3600 option of the WESTEMS' Design CUF module (Commitment 45). See NL-11-032, Attach. 1 at 27 (NYS000151). Commitment 44 is consistent with standard engineering practice to document inputs in the preparation of calculations. And as explained in response to Question 113, Westinghouse did not use the ASME Code,Section III, NB-3600 option of the WESTEMS' Design CUF module for any EAF evaluations for IPEC license renewal.
Q111. Did Westinghouse undertake the elimination of redundant stress peaks or valleys in the EAF analyses conducted for IPEC license renewal?
A111. (MAG) 72
Q112.
A112. (MAG)
Q113. The NRC Staffs RAIs also refer to the NB-3600 option in the WESTEMS' Design CUF module. See NL-11-032, Attach. 1 at 27 (NYS000151). What is that option?
A113. (MAG) 73
Q114. In his 2011 reports, Dr. Lahey applies the term user intervention to: (1) to the assumption of a constant heat transfer coefficient in the pressurizer surge line nozzle evaluations, see Supplemental Lahey Report at 7 (NYS000297); and (2) the thermal stress results for the CUFen values. Lahey Report at 28 (NYS000296). Do you agree with this use of the term?
A114. (MAG) No. Dr. Laheys use of the term user intervention is inconsistent with the NRC Staffs use and interpretation of that term in SSER 1. The term user intervention, as used by the NRC Staff, did not refer to the broader issue of user definition of inputs, such as heat transfer coefficients and loads. Further, it is not possible to eliminate the need for the exercise of specialized engineering expertise in fatigue analyses. Dr. Lahey does not explain why the practice of using engineering judgment to conservatively define input loadsas is done in all ASME Code stress and fatigue analyses, including the Westinghouse EAF evaluations for IPECis in any way deficient. The issue of user intervention is further discussed in Entergys testimony on the safety commitments contention, with reference to the specific claims raised in testimony on that contention. See Entergys NYS-38/RK-TC-5 Testimony at Section V.B.2 (ENT000699).
Q115. More recently, Dr. Lahey suggested that Westinghouse has disclose[d] the use of engineering judgment and user intervention . . . for some, but not all of the fatigue evaluations performed to date. Revised Lahey Testimony at 72 (NYS000530). Is this an accurate statement?
A115. (MAG) No. As we explained in response to Question 112, above, in accordance with Entergys Commitment 44, Westinghouse has documented the one instance where peak editing was used in a fatigue evaluation for Entergy in support of 74
the Indian Point LRA. See Westinghouse Calculation Note CN-PAFM-13-40, at App. D (ENT000688). To the extent Dr. Laheys concerns are on the broader issue of engineering judgment, each of the fatigue calculations include documentation of the assumptions used.
Q116. Dr. Hopenfeld similarly argues that Entergy has not disclosed how it will control the modifications of the WESTEMS' program by the analyst during future evaluations. See Hopenfeld Rebuttal at 48-49 (RIV000114). How do you respond to this statement?
A116. (MAG, NFA) It is not clear what Dr. Hopenfeld means by modifications or manipulations. Neither Entergy nor Westinghouse have modified or manipulated any fatigue analyses performed using the WESTEMS' computer program, nor will they do so in the future.
Entergy and Westinghouse retain the option to use peak editing to eliminate redundant peaks and valleys, consistent with Commitment 44 in the LRA, which the NRC Staff approved in SSER 1.
Q117. Dr. Hopenfeld criticizes Entergys Commitment 43 to review its design basis ASME Code fatigue evaluations to confirm that the locations specified in NUREG/CR-6260 are limiting locations for IPEC. See Hopenfeld Report at 23-24 (RIV000035). Specifically, he describes this as a vague commitment to perform necessary metal fatigue investigation and analyses in the future. Id. at 24. How do you respond?
A117. (NFA, ABC, JRS, MAG) Contrary to Dr. Hopenfelds characterization of this commitment as vague, the commitment is very specific with regard to the timing, nature, and scope of the evaluation to be performedall design basis ASME Code fatigue evaluations would be reviewed for environmental effects prior to the PEO. In addition, in making Commitment 49 in 2013 (discussed further below in response to Question 120), Entergy later confirmed that its comprehensive review of CUFs for IP2 and IP3 will include all RVI components with CLB 75
CUFs. See SSER 2 at A-15 (NYS000507) (committing to [r]ecalculate each of the limiting CUFs provided in Section 4.3 of the LRA for the reactor vessel internals); LRA at 4.3-11 to 4.3-12, tbls. 4.3-5 & 4.3-6 (ENT00015B) (listing the CLB RVIs CUFs for IP2 & IP3). Therefore, we disagree with Dr. Hopenfeld that this commitment is vague or otherwise insufficient in scope.
Q118. What conclusions did the NRC Staff reach following its review of Entergys RAI responses?
A118. (ABC, NFA, JRS) After Entergy responded to these and other RAIs unrelated to fatigue, the NRC Staff issued a Supplemental Safety Evaluation Report. See generally SSER 1 (NYS000160). SSER 1 concludes that LRA Commitment 43 is acceptable because it is consistent with the recommendations in SRP-LR Sections 4.3.2.2 and 4.3.3.2, and NUREG-1801, Revision 1 AMP X.M1, to consider environmental effects for the NUREG/CR-6260 locations, at a minimum. See id. at 4-2 (NYS000160). SSER 1 further explains that new Commitment 44 also is acceptable because it specifies that records of WESTEMS' Design CUF module fatigue evaluations will contain sufficient information to document and justify any assumptions and engineering judgment, and provides that the bases for fatigue calculation conclusions are auditable and retrievable. See id. Finally, the NRC Staff concluded that Commitment 45 was acceptable because Entergy committed not to use the NB-3600 option of WESTEMS' pending resolution of the NRC Staffs concerns with that option. See id. at 4-3.
- 3. Issues Addressed in SSER 2 Q119. After SSER 1, did the NRC Staff issue any metal-fatigue related RAIs pertaining to RVIs?
A119. (NFA, ABC, RGL) Yes. Related to the current FMP, the Staff requested clarification of whether Entergy would use its RVI AMP, its FMP, or a combination of both to manage RVI fatigue during the PEO. See Letter from R. Kuntz, NRC, to Vice President, 76
Operations, Entergy, Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Nos. 2 and 3, License Renewal Application, Encl. at 7 (May 15, 2012)
(ENT000659).
Entergy responded indicating that for locations with a fatigue TLAA, Entergy would manage the effects of aging due to fatigue through the FMP, but for locations without a CLB fatigue analysis, Entergy would rely on the inspections in the MRP-227-A-based RVI AMP. See NL-12-089, Letter from F. Dacimo, Entergy, to NRC Document Control Desk, Reply to Request for Additional Information Regarding the License Renewal Application, Attach. 1 at 17-18 (June 14, 2012) (NL-12-089) (ENT000554). This is consistent with the NRC Staffs Safety Evaluation for MRP-227-A. See Letter from R. Nelson, NRC, to N. Wilmshurst, EPRI, Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Prog[ra]m (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (TAC No. ME0680),
encl. 1 at 26 (Dec. 16, 2011) (SE for MRP-227-A) (ENT000230).
Entergy also noted that under Commitment 43 the existing RVI fatigue calculations for each unit would be reviewed prior to entering the PEO to determine whether the locations specified in NUREG/CR-6260 were the limiting locations for IP2 and IP3. See NL-12-089, Attach. 1 at 18 (ENT000554); see also generally NUREG/CR-6260 (NYS000355). As noted later in response to Question 122, this review has now been completed.
Q120. Did the NRC Staff seek further information after this response?
A120. (NFA, ABC, RGL) Yes. In RAI 15, issued February 6, 2013, the Staff asked Entergy to provide a new commitment, in addition to Commitment 43, to specifically address the review of RVI components for EAF as part of the FMP. See NL-13-052, Letter from F. Dacimo, 77
Entergy, to NRC Document Control Desk, Reply to Request for Additional Information Regarding the License Renewal Application, Attach. 1 at 8 (May 7, 2013) (NL-13-052)
(NYS000501). In response, Entergy provided Commitment 49, committing to [r]ecalculate each of the limiting CUFs provided in Section 4.3 of the LRA for the reactor vessel internals.
See id. at 9. Again, as explained below, this review has now been completed by Westinghouse.
Q121. What conclusions did the NRC Staff reach following its review of Entergys RAI responses regarding the effects of fatigue on RVIs?
A121. (NFA, ABC, RGL) The NRC Staff issued its second supplemental safety evaluation report in November 2014. See generally SSER 2 (NYS000507). In SSER 2, the Staff concluded that its concerns were resolved, and that the responses were technically acceptable because the methods proposed for managing fatigue of the RVI are consistent with those allowed by A/LAI 8 from MRP-227-A, and consistent with NUREG-1801,Section X.M.1, Fatigue Monitoring. SSER 2 at 3-53 (NYS000507).
C. Overview of the EAF Evaluations Prepared for IPEC License Renewal Q122. Please provide a general overview of the sequence and chronology of the major EAF evaluations that have been prepared in support of the IPEC LRA.
A122. (NFA, ABC, MAG) To summarize the EAF evaluations performed under the FMP to date, Entergy first prepared an initial fatigue screening evaluation in its 2007 LRA.
Second, to satisfy Commitment 33, Entergy retained Westinghouse to perform comprehensive refined EAF analyses, completed in 2010, for all locations identified in NUREG/CR-6260. See generally, WCAP-17199, Rev. 1 (ENT000681); WCAP-17200, Rev. 1 (ENT000682).
Later, to address the subsequent Commitments 43 and 49, Entergy retained Westinghouse to perform additional EAF reviews to determine whether there are any other potentially leading locations beyond the NUREG/CR-6260 locations, including RVIs. In 2012, Westinghouse 78
completed the screening assessment for non-NUREG/CR-6260 locations and RVIs as part of this process. See generally Westinghouse Calculation Note CN-PAFM-12-35 (NYS000510). This screening review included all ASME Class 1 design basis fatigue evaluations, and included all RVI components with CLB CUF fatigue evaluations, consistent with Commitment 43, as clarified in Commitment 49. See id. at 9-11.
In 2013 for IP2 and in 2015 for IP3, Westinghouse then completed refined evaluations of the non-NUREG/CR-6260 locations and RVIs that were identified as potentially leading locations in the CN-PAFM-12-35 screening analysis. See generally CN-PAFM-13-32, Rev. 3 (ENT000683). As we will discuss further, there are several additional supporting analyses that support the major documents cited above. The sequence and supporting analyses of the major evaluations are represented in the chart below:
2007: 6260 Locations Screening LRA For original application 2010: 6260 Locations Refined WCAP-17199 & 17200 For Commitment 33 2012: Additional Locations Screening CN-PAFM-12-35 For Commitments 43 & 49 2013: Additional Locations Refined (IP2)
CN-PAFM-13-32, Rev. 1 For Commitments 43 & 49 2015: Additional Locations Refined (IP3)
CN-PAFM-13-32, Rev. 3 For Commitments 43 & 49 79
Q123. Was Entergy permitted to refine the EAF analyses presented in its initial LRA through the further evaluations listed above?
A123. (NFA, ABC, JRS) Yes. An applicant may perform refined fatigue analyses that supersede prior analyses. See NUREG-1801, Revision 1 at X M-2 (NYS00146C); see also NUREG-1801, Revision 2 at X M1-2 (NYS00147C) (allowing a more rigorous analysis of the component to demonstrate that the design code limit will not be exceeded during the period of extended operation); see also MRP-47 at 3-7 (NYS000350) (Possible reasons for updating the fatigue analysis could include . . . [e]xcess conservatism in original fatigue analysis with respect to modeling, transient definition, transient grouping and/or use of an early edition of the ASME Code.).
The elimination of unnecessary conservatisms through re-analysis yields a new CUF value (to which the Fen is then applied). See id. at 4-4 (stating that techniques for removing excess conservatisms from the input (stress) values of CUF calculations are generally well understood by engineers performing these assessments throughout the industry). If a CUFen still exceeds 1.0 after more detailed fatigue analysis, then NRC guidance recommends reviewing additional RCS pressure boundary locations (beyond the NUREG/CR-6260 locations) where high usage factors might be a concern. See NUREG-1801, Rev. 1, at X M-2 (NYS00146C);
MRP-47 at 3-4 to 3-5 (NYS000350).
As previously explained, Entergy has prepared detailed EAF evaluations for all NUREG/CR-6260 locations, and reviewed all CLB CUF analyses for reactor coolant pressure boundary and RVI locations, and evaluated the limiting locations for EAF. Consistent with the regulations, the refined EAF analyses conducted by Westinghouse showed the CUFen to be less 80
than or equal to 1.0, and these new evaluations supersede any corresponding prior initial screening evaluations.
Q124. Drs. Hopenfeld and Lahey assert that once the LRA showed a CUFen value greater than 1.0, then Entergy was required to expand the scope of components reviewed for EAF. See, e.g., Lahey Report at 25-26 (NYS000296), Hopenfeld Report at 24 (RIV000035). How do you respond?
A124. (NFA, ABC, JRS, MAG) This issue is moot. Consistent with Commitments 43 and 49, Entergy expanded the scope of its EAF evaluations to cover all design basis ASME Code Class 1 fatigue evaluations and all RVI components with CLB CUF analyses. See generally Westinghouse Calculation Notes CN-PAFM-12-35 (NYS000510) and CN-PAFM-13-32, Rev. 3 (ENT000683).
To the extent Dr. Hopenfeld is seeking EAF evaluations of secondary plant componentssuch components are not part of the reactor coolant pressure boundary and are not exposed to the reactor water environment. Therefore, an EAF evaluation is not necessary. See GSI-190 Closeout Memorandum Attach. 2, at 1 (ENT000190); see also NUREG-1801, Revision 1, at X M-1 (Thus, no further evaluation is recommended for license renewal if the applicant selects this option under 10 CFR 54.21(c)(1)(iii) to evaluate metal fatigue for the reactor coolant pressure boundary) (NYS00146C).
Further, to the extent he is requesting EAF analyses of primary plant components beyond those with CLB CUF evaluations, such claims are a challenge to the CLB and beyond the scope of this proceeding. Under 10 C.F.R. § 54.21(c)(1)(iii), the FMP is intended to manage the effects of aging addressed by fatigue TLAAs that are part of the CLB for IP2 and IP3. See NUREG-81
1801, Rev. 1 at X M-1 (NYS00146C) (In order not to exceed the design limit on fatigue usage
. . . .) (emphasis added).
Q125. Dr. Lahey has characterized the subsequent EAF analyses performed after submission of the original LRA as an attempt to selectively remove conservatisms to reach a manipulated and predetermined result. Declaration of Dr. Richard T. Lahey, Jr. in Support of the State of New Yorks Supplemental Contention 26-A ¶ 5 (Apr. 7, 2008)
(NYS000299). Similarly, in his Supplemental Report, he states that the thermal stress results for CUFen are strongly influenced by the code users assumptions, manipulations and interventions. There is a lot of engineering judgment implicit in the CUFen results, and, since an error analysis has not been done to bound the uncertainty, and many results are disturbingly close to the CUFen = 1.0 limit, I do not believe that one can trust these results . . . . Supplemental Lahey Report at 8 (NYS000297). How do you respond?
A125. (NFA, ABC, MAG) We disagree with Dr. Laheys characterizations of the EAF analyses. Entergy and Westinghouse followed standard engineering practice in preparing refined EAF analyses for IPEC and did not make unsupported assumptions to reach a manipulated and predetermined result. As we demonstrate in response to his specific allegations throughout this testimony, the refined IPEC EAF analyses are sufficiently conservative to demonstrate that each analyzed component is not expected to experience fatigue cracking during the PEO. The refined analyses and the underlying Westinghouse documents are transparent with regard to the assumptions and methods used. Furthermore, the margin and conservatism in design basis fatigue calculations is well-documented, as we have described in detail in Section IV.A.2 of our testimony.
82
Q126. Dr. Lahey argues that Entergys general defense of the EAF evaluations as standard industry practice is inadequate. See Lahey Rebuttal Testimony at 6 (NYS000440). What is your response to this statement?
A126. (NFA, ABC, MAG) To the extent Dr. Lahey is criticizing the ASME Code fatigue design process, he is attacking the NRCs regulations in 10 C.F.R. § 50.55a(c)(1), which specifies that the metal fatigue standards in the ASME Code,Section III, must be met. In any event, Entergy does not rest merely on a defense of standard industry practice. That is Dr.
Laheys phrase, not ours. On the contrary, Entergy has fully explained how the ASME Code fatigue analyses Westinghouse performed contain considerable margin and conservatism. See Section IV.A.2, above.
D. The 2010 EAF Analyses for NUREG/CR-6260 Locations Conservatively Demonstrate that the CUFens for All NUREG/CR-6260 Locations at IPEC Do Not Exceed 1.0
- 1. Summary of the 2010 EAF Evaluations Q127. Did Entergy perform refined fatigue analyses to determine environmentally-adjusted CUF values for NUREG/CR-6260 locations?
A127. (NFA, ABC) Yes. To meet Commitment 33, Entergy retained Westinghouse in 2008 to perform refined fatigue analyses to determine environmentally-adjusted CUFs for those locations listed in LRA Tables 4.3-13 and 4.3-14 that did not have initial screening CUF values less than or equal to 1.0. Westinghouse completed the refined fatigue analyses in June 2010.
See Westinghouse, WCAP-17199, Environmental Fatigue Evaluation for Indian Point Unit 2, Rev. 0 (June 2010) (NYS000361); Westinghouse, WCAP-17200, Environmental Fatigue Evaluation for Indian Point Unit 3, Rev.0 (June 2010) (NYS000362). Entergy approved the analyses on July 29, 2010, and shortly thereafter, notified the Staff of the results of the refined EAF analyses. See generally NL-10-082, Letter from Fred R. Dacimo, Entergy, to NRC 83
Document Control Desk, License Renewal Application - Completion of Commitment #33 Regarding the Fatigue Monitoring Program (Aug. 9, 2010) (NL-10-082) (NYS000352).
Q128. Has Westinghouse since revised these two reports?
A128. (MAG, NFA, ABC) Yes. In December 2014, Westinghouse issued revised WCAP-17199, Rev. 1 (ENT000681) and WCAP-17200, Rev. 1 (ENT000682).
Q129. Please summarize the changes Westinghouse made in the revised reports.
A129. (MAG)
Q130. Please summarize the methodology used by Westinghouse to conduct the 2010 fatigue evaluations for NUREG/CR-6260 locations.
A130. (MAG, NFA)
Q131. Please provide a general overview of the ASME Code Section III stress and fatigue evaluations including a summary of the thermal-hydraulic modeling used by Westinghouse for the IPEC EAF analyses.
A131. (MAG, NFA) In general, the stress and fatigue analyses were performed as follows.
84
Q132. Please identify the major documents which are part of the 2010 EAF evaluations for IPEC.
A132. (MAG) The major documents supporting the 2010 Westinghouse EAF analyses for IPEC components are summarized in the following tables. The location column shows both the NUREG/CR-6260-specified generic location and the plant-specific limiting locations for IP2 and IP3.
Table 1: IPEC Unit 2 2010 Westinghouse EAF Calculations Location Document Exhibit Pressurizer surge Westinghouse Calculation Note ENT000199 85
line nozzles - Number CN-PAFM-09-100, Rev. 0, Pressurizer surge Indian Point Unit 2 Insurge/Outsurge nozzle and Environmental Fatigue Evaluations (June 10, 2010)
(Westinghouse Calculation Note CN-PAFM-09-100);
Westinghouse Calculation Note ENT000200 Number CN-PAFM-09-63, Rev 0, Indian Point Unit 2 Pressurizer Insurge/Outsurge Transient Development (June 18, 2010)
(Westinghouse Calculation Note CN-PAFM-09-63); and NYS000365 Westinghouse Calculation Note Number CN-PAFM-09-67, Rev. 0, Pressurizer Surge Nozzle and Lower Head Transfer Functions for Indian Point Units 2 and 3 (June 18, 2010)
(Westinghouse Calculation Note CN-PAFM-09-67).
Pressurizer surge Westinghouse Calculation Note NYS000368 line nozzles - Surge Number CN-PAFM-09-117, Rev. 0, line hot leg nozzle Indian Point Units 2 and 3 Hot Leg (RCS piping surge Surge Nozzle Environmental Fatigue nozzle) Evaluations (June 18, 2010)
(Westinghouse Calculation Note CN-PAFM-09-117);
Westinghouse Calculation Note CN- ENT000200 PAFM-09-63; and Westinghouse Calculation Note ENT000201 Number CN-PAFM-08-109, Rev. 0, Indian Point Units 2 and 3 Transfer Function Database Development for Hot Leg Surge Nozzle (June 18, 2010)
(Westinghouse Calculation Note CN-PAFM-08-109).
RCS piping charging Westinghouse Calculation Note NYS000364 system nozzle - Number CN-PAFM-09-21, Rev. 0, Charging system Indian Point Units 2 and 3 Charging nozzle Nozzles Environmental Fatigue Evaluation (June 18, 2010)
(Westinghouse Calculation Note CN-86
PAFM-09-21); and Westinghouse Calculation Note ENT000202 Number CN-PAFM-08-78, Rev. 0, Indian Point Unit 2 and Unit 3:
Transfer Function Database Development for Hot Leg Charging Nozzle (June 18, 2010)
(Westinghouse Calculation Note CN-PAFM-08-78).
RCS piping safety Westinghouse Calculation Note NYS000367 injection nozzle - Number CN-PAFM-09-79, Rev. 0, Boron injection tank Indian Point Unit 2 Boron Injection nozzle Tank Nozzle Environmental Fatigue Evaluations (June 14, 2010)
(Westinghouse Calculation Note CN-PAFM-09-79); and Westinghouse Calculation Note ENT000203 Number CN-PAFM-08-116, Rev. 0, Indian Point Unit 2 BIT Nozzle Transfer Function Database Development (June 18, 2010)
(Westinghouse Calculation Note CN-PAFM-08-116).
RHR class 1 piping - Westinghouse Calculation Note NYS000366 Accumulator nozzle Number CN-PAFM-09-77, Rev. 0, Indian Point Units 2 & 3 Accumulator Nozzle Environmental Fatigue Evaluation (June 18, 2010)
(Westinghouse Calculation Note CN-PAFM-09-77); and Westinghouse Calculation Note ENT000204 Number CN-PAFM-08-103, Rev. 0, Indian Point Units 2 and 3 Transfer Function Database Development for Accumulator Nozzles (June 17, 2010)
(Westinghouse Calculation Note CN-PAFM-08-103).
Table 2: IPEC 2010 Unit 3 Westinghouse EAF Calculations Location Document Exhibit Pressurizer surge Westinghouse Calculation Note ENT000205 line nozzles - Number CN-PAFM-09-105, Rev. 0, 87
Pressurizer surge Indian Point Unit 3 Insurge/Outsurge nozzle and Environmental Fatigue Evaluations (June 22, 2010)
(Westinghouse Calculation Note CN-PAFM-09-105);
Westinghouse Calculation Note RIV000055 Number CN-PAFM-09-64, Rev. 0, Indian Point Unit 3 Pressurizer Insurge/Outsurge Transient Development (June 18, 2010)
(Westinghouse Calculation Note CN-PAFM-09-64); and NYS000365 Westinghouse Calculation Note CN-PAFM-09-67.
Pressurizer surge Westinghouse Calculation Note CN- NYS000368 line nozzles - Surge PAFM-09-117; line hot leg nozzle (RCS piping surge Westinghouse Calculation Note CN- RIV000055 nozzle) PAFM-09-64; and Westinghouse Calculation Note CN- ENT000201 PAFM-08-109.
RCS piping charging Westinghouse Calculation Note CN- NYS000364 system nozzle PAFM-09-21; and Westinghouse Calculation No. CN- ENT000202 PAFM-08-78.
RCS piping safety Westinghouse Calculation Note ENT000206 injection nozzle - Number CN-PAFM-09-74, Rev. 0, Boron injection tank Indian Point Unit 3 Boron Injection nozzle Tank Nozzle Environmental Fatigue Evaluations (June 14, 2010)
(Westinghouse Calculation Note CN-PAFM-09-74); and Westinghouse Calculation Note ENT000207 Number CN-PAFM-09-13, Rev. 0, Indian Point Unit 3 BIT Nozzle Transfer Function Database Development (June 18, 2010)
(Westinghouse Calculation Note CN-PAFM-09-13).
88
RHR class 1 piping Westinghouse Calculation Note NYS000366
- Accumulator Number CN-PAFM-09-77; and nozzle Westinghouse Calculation No. CN- ENT000204 PAFM-08-103.
Q133. What were the results of Westinghouses 60-year EAF evaluations for NUREG/CR-6260 locations?
A133. (MAG, NFA) The 60-year fatigue results for the critical component locations are provided in Tables 5-8 through 5-14 of WCAP-17199 and WCAP-17200. See WCAP-17199, Rev. 1 at 5-26, 5-28, 5-32, 5-35, 5-39, 5-41, 5-45 (ENT000681); WCAP-17200, Rev. 1 at 5-26, 5-28, 5-32, 5-35, 5-39, 5-41, 5-45 (ENT000682). Westinghouse determined that, for IP2 and IP3, the refined CUFen values for pressurizer surge line piping, RCS piping charging system nozzle, RCS piping safety injection nozzle, and RHR Class 1 piping all are below 1.0 when projected to the end of the PEO. See NL-10-082, Attach. 1 at 2-4 (NYS000352). The refined CUFen values supersede the screening values contained in the April 2007 LRA.
Q134. Dr. Lahey observes that WESTEMS' is based on rather simple models (particularly the thermal-hydraulic models). Lahey Report at 28 (NYS000296). Does the use of a simplified model necessarily mean that the results will be unreliable or less conservative?
A134. (MAG, NFA) No. The development of the input transient temperature loads is consistent with accepted engineering practice. Dr. Lahey appears to imply, erroneously, that simplified models are non-conservative. The use of simplified models inherently introduces more conservatism, because to simplify a model, a fatigue analyst must use more conservative values for it to be a valid simplification.
89
Q135. Relatedly, Dr. Lahey expresses the desire to see how the EAF analyses were bench-marked against representative experimental data and/or analytical solutions.
Lahey Report at 27 (NYS000296). Have the WESTEMS' EAF results been benchmarked as Dr. Lahey proposes?
A135. (MAG) Yes. In response to NRC Staff RAIs, WESTEMS' results have at least twice been benchmarked against other methods, and each time were shown to be valid. See Salem SER at 1-9 (ENTR00195); NUREG-1916, Safety Evaluation Report Related to the License Renewal of Shearon Harris Nuclear Power Plant, Unit 1 at 4-28 (Nov. 2008) (Shearon Harris SER) (ENT000223). The NRC Staffs review of the Salem License Renewal application included an audit of WESTEMS' results against the ASME Code methodology, and found the results acceptable. See Salem SER at 1-9 (ENTR00195) (The audit confirmed that for the two monitored locations, Salems use of WESTEMS' NB-3200 module produced results that were consistent with those using the methodology in ASME Code Section III, NB-3200.).
Similarly, in the Shearon Harris license renewal SER, the Staff explained that it reviewed the applicants benchmark verification results . . . all indicating that the stress results generated from fatigue analysis software [i.e., WESTEMS'] and those generated from traditional finite element ANSYS analysis have negligible differences. Shearon Harris SER at 4-28 (ENT000223). On that basis, the Staff conclude[d] that stress evaluation by fatigue analysis software is acceptable. Id. The methods that were reviewed and approved by the NRC Staff in the Salem and Shearon Harris license renewals are the same methods used to prepare the IPEC EAF analyses.
90
- 2. Number of Transients Q136. Turning to a more detailed consideration of the 2010 EAF analyses and the Intervenors critique of specific aspects of them, please summarize how the transient cycles for each component were projected.
A136. (MAG, NFA)
Q137. Are actual cycles tracked as part of the FMP?
A137. (ABC, NFA) Yes. Entergy is tracking cycles as part of its FMP. See LRA, App.
B at B-44 (ENT00015B). If, for some reason, the number of actual occurrences of a transient begins to approach the analyzed number of occurrences of that transient, then the FMP will require corrective action under the FMP, including more rigorous analyses, or repair or replacement of components, as appropriate, before the analyzed number of transients is 91
exceeded. See id. NUREG-1801 specifies this approach. See generally NUREG-1801, Revision. 2 (NYS00147A-D).
Q138. What transient loads are considered as inputs into the fatigue analysis?
A138. (NFA, ABC, JRS, MAG) Under ASME Code, Section NB-3200, the fatigue analysis considers normal operation, upset condition, and test transient conditions (NB-3222, NB-3223, NB-3226). See ASME Code,Section III, Article NB-3000 §§ NB-3222.4(b), NB-3222.4(d) & n.8 (stating that peak stress is derived from normal service conditions)
(NYS000349). The cyclic service evaluation does not require consideration of postulated accident and post-accident conditions, which are addressed in other sections of the ASME Code.
Specifically, accident loads are addressed by a separate analysis as described in NB-3224 and NB-3225. LOCA and Anticipated Transient Without Scram (ATWS) loads are not normal service conditions as defined in NB-3222.4 and, therefore, are not included as transients leading to fatigue of a component in an NB-3200 fatigue analysis of RCS Class 1 components. See 10 C.F.R. Pt. 50, App. A (defining LOCA); 10 C.F.R. § 50.62(b) (defining ATWS).
Q139. How were past transients used as inputs into the EAF analyses?
A139. (NFA, MAG) Contrary to Dr. Hopenfelds statements, the Westinghouse EAF evaluations provide ample documentation on the past transients used in the EAF analyses. In general, Westinghouse reviewed the IP2 and IP3 plant operating records to determine when the plant was at power operation and when the plant was shut down. See generally Westinghouse, WCAP-12191, Rev. 4, Transient and Fatigue Cycle Monitoring Program Transient History Evaluation Report for Indian Point Unit 2 (Dec. 2014) (ENT000689); Westinghouse, WCAP-16898-P, Rev. 1, Indian Point Unit 3 Transient and Fatigue Cycle Monitoring Program Transient History Evaluation (May 2015) (ENT000690). Available plant computer data were used to 92
characterize plant cycles. See id. When sufficient data were not available, appropriate alternatives were used, based on a review of plant history and operating procedures. See, e.g.,
Westinghouse Calculation Note CN-PAFM-09-64 at 6 (RIV000055).
Q140.
A140. (NFA, MAG) 93
94 Q141. Dr. Hopenfeld also takes issue with the straight-line extrapolation of the number of plant transients from 40 to 60 years. See Hopenfeld Report at 19-20 (RIV000035). Rather than a straight-line extrapolation, Dr. Hopenfeld believes the bathtub curve better represents the number of transients that will be experienced during the PEO. See id. at 20. How do you respond?
A141. (ABC, NFA) As a threshold matter, we do not agree that there is any logical basis to conclude that IP2 or IP3 would be subjected to an increasing number of cycles as the units approach 60 years of operation.
In any event, Entergys FMP for IPEC does not simply rely on straight-line extrapolation of transients. Entergy tracks all operating cycles used to calculate the CUFen, and therefore will ensure that the numbers of actual cycles through 60 years do not exceed the numbers of cycles assumed in the fatigue analysis. If future plant operating changes are implemented which result in increased numbers of cycles (as Dr. Hopenfeld postulates), or if the analyzed number of cycles is approached for some other reason, such that actual cycles are expected to exceed the number analyzed, then Entergy will evaluate the fatigue analysis for the affected components to ensure that the CUFen does not exceed 1.0. See SER at 3-79, 4-44 (NYS00326B, NYS00326E); NL-08-084, Letter from Fred R. Dacimo, Entergy, to NRC Document Control Desk, Reply to Request for Additional Information Regarding License Renewal Application - Time-Limited Aging Analyses and Boraflex, Attach. 1 at 4 (May 16, 2008) (NL-08-084) (ENT000194). Consistent with 10 C.F.R. §§ 54.21(a)(3) and (c)(1)(iii),
those components which cannot be demonstrated to comply with a CUF of 1.0 based on such a re-analysis will be repaired or replaced to ensure they meet required structural capabilities. Id.
95
Thus, Dr. Hopenfelds general comments about the bathtub curve are not relevant. If, for some reason, the number of occurrences of a transient approaches the number analyzed, then the FMP will require corrective action before the analyzed number of transients is reached.
Q142.
A142. (MAG, NFA)
Under NRC guidance, this is an acceptableand indeed, expectedremoval of excess conservatism. See NUREG-1801, Revision 2 at X M1-2 (NYS00147C); MRP-47 at 3-7 (NYS000350) (Possible reasons for updating the fatigue analysis could include . . . [e]xcess conservatism in original fatigue analysis with respect to modeling, transient definition, transient grouping and/or use of an early edition of the ASME Code.). In any event, we would not describe this process as the modification of assumed plant-specific thermal transients. As to the proximity of any calculated CUFen to 1.0, this is irrelevant, as we explained in response to Question 74.
96
- 3. Heat Transfer Coefficients Q143. Dr. Hopenfeld states that Hopenfeld Report at 13 (RIV000035); see also id. at 14. Do you agree?
A143. (MAG, NFA) Heat transfer is a factor in the calculation of transient thermal stress, but given the conservative heat transfer coefficients used in the IPEC analysis (which we will explain further below), the major factor controlling thermal fatigue damage is the magnitude of the variation in temperature. While the heat transfer coefficient does impact the calculation of stresses (i.e., a higher heat transfer coefficient can lead to increased calculated stresses), unlike the variation in temperature, it is not the main contributor to the thermal or stress gradient through the component wall.
Q144.
A144. (MAG) 97
98 F. Kreith, PRINCIPLES OF HEAT TRANSFER at 396 (3rd ed. 1973) (Principles of Heat Transfer) (ENT000208).
99
. Principles of Heat Transfer at 396 (ENT000208).
Q145. In the same section, Dr. Hopenfeld asserts that to assess the uncertainty of the heat transfer coefficients used, it is imperative to know the component geometry, the piping geometry upstream of the component, [and] the flow velocities as well as the heat transfer coefficients, but that this information was not specified. Hopenfeld Report at 18 (RIV000035). Do you agree with Dr. Hopenfeld?
A145. (MAG) No.
100
Q146.
A146. (MAG) 101
Q147.
A147. (MAG)
Q148.
A148. (MAG)
Q149. What is a thermal stress ratchet?
A149. (MAG) The term thermal stress ratchet refers to the potential for permanent cumulative strain to cause additional reduction in service life of a component.
Q150.
A150. (MAG) 102
Q151. Please respond to Dr. Laheys statements regarding alleged errors from modifying the heat transfer coefficient to account for the effect of the thermal sleeves.
Supplemental Lahey Report at 3 (NYS000297).
A151. (MAG, NFA) Dr. Lahey argues that modifying the heat transfer coefficient to account for the effects of a thermal sleeve is a crude analytical approach. Supplemental Lahey Report at 3 (NYS000297). We disagree. A thermal sleeve is a thin cylindrical part welded to the inside of a nozzle to shield the nozzle surface from the effects of thermal transients. In all cases, as described above, the film coefficients are calculated in a conservative manner, and therefore yield conservative heat transfer coefficients. Accounting for thermal sleeve heat transfer effects with an effective heat transfer coefficient is a standard analytical practice, and is supported by industry papers and standard textbooks in the field. See, e.g., M. E. Nitzel et al., Comparison of ASME Code NB-3200 and NB-3600: Results for Fatigue Analysis of B31.1 Branch Nozzles (Apr.
1996) (ENT000221); J. Holman, HEAT TRANSFER (4th ed. 1976) (ENT000209). The acceptability of the use of an effective heat transfer coefficient is further described in the NRC-endorsed NUREG/CR-6260. See NUREG/CR-6260 at 5-67, 5-72 (NYS000355).
103
Q152. How were thermal sleeves evaluated in the IPEC EAF analyses?
A152. (MAG, NFA)
Q153. Dr. Lahey suggests that a more accurate 3-D thermal analysis is feasible. See Supplemental Lahey Report at 3 (NYS000297) Do you agree?
A153. (MAG, NFA) It is theoretically possible to estimate the effects of thermal sleeves through a detailed finite element analysis, including the thermal sleeve. Dr. Lahey asserts that such modeling would be more accurate, which may be true. But this assertion does not show that Westinghouses method is deficient, or not conservative.
In any event, such a detailed analysis is not required or necessary, as the methodology used by Westinghouse is used throughout the nuclear industry and other industries to evaluate the effects of a thermal sleeve on a nozzle, and has been endorsed in NRC guidance. See NUREG/CR-6260 at 5-67, 5-72 (NYS000355). As previously noted, the ASME Code requires conservatism in fatigue evaluations, not an exact calculation of fatigue usage.
104
Q154. Dr. Lahey also states that the inherent uncertainty in single-phase heat transfer coefficients, which are allegedly at least +/- 25%, has not been accounted for in the IPEC EAF evaluations. Supplemental Lahey Report at 4 (NYS000297). How did Westinghouse account for uncertainty in the heat transfer coefficients it used?
A154. (MAG) It is well known to experts in the field that single-phase heat transfer coefficients are approximate and empirical. See generally, e.g., Principles of Heat Transfer, Ch.
6 (ENT000208).
Q155.
A155. (MAG) 105
Q156.
A156. (MAG) 106
Q157.
A157. (MAG, NFA) If the objective of the analysis was to evaluate transient heat transfer phenomena as accurately as possible, then Dr. Lahey would be correct. But as we have previously explained, that is not the objective of an EAF analysis. An EAF analysis is intended to demonstrate, with conservatism, whether the CUFen of each analyzed component exceeds 1.0.
Q158. In support of his point regarding the desirability of more detailed, three-dimensional thermal-hydraulic analysis, Dr. Lahey relies on Cengel & Turner, 107
Fundamentals of Thermal-Fluid Sciences at 759-760 (1st ed. 2001) (Fundamentals of Thermal-Fluid Sciences) (NYS000345) to assert that one-dimensional evaluations will under-predict fluid-to-wall heat transfer and therefore lead to non-conservative overall results. See Supplemental Lahey Report at 3 (NYS000297). Please respond.
A158. (MAG, NFA) This reference does not support Dr. Laheys point that only a 3-D model can accurately define the transient loadings at locations such as pipes and nozzles where fully-developed flow conditions are lacking. It simply suggests that this issue must be considered when analyzing such components. See Fundamentals of Thermal-Fluid Sciences at 760 (NYS000345). In fact, the reference states that [p]recise correlations for the heat transfer coefficient for the entry regions are available in the literature. Id. This clearly implies that this is an issue which can be solved using available formulaswhich is what Westinghouse did without the need for complex 3-D models. The reference goes on to state that [t]his approach, which we will use for simplicity, gives reasonable results for long tubes and conservative results for short ones. Id. (emphasis in original). Thus, Fundamentals of Thermal-Fluid Sciences clearly endorses the approach of using simplified solutions, rather than the use of complex 3-D models. This document does not contradict our opinion that it is possible to develop a simplified correlation for the heat transfer coefficient that is sufficiently conservative without using a 3-D model.
Q159. In discussing the accumulator nozzles, Dr. Lahey observes that WESTEMS does not allow for the onset of nucleate boiling during depressurization transients . . . .
Supplemental Lahey Report at 6 n.1 (NYS000297). How do you respond?
A159. (MAG, NFA) Under normal operating conditions, nucleate boiling is a phenomenon which occurs at the fuel cladding. It does not occur at the accumulator nozzles or 108
anywhere else in the reactor coolant system because the system is maintained in a subcooled condition throughout all modes of operation. Pressure decreases during the cyclic transients analyzed in fatigue evaluations do not result in water flashing to steam at these locations. In the event of a large LOCA, it is possible for steam to be present, but this is a one-time accident condition, quite distinct from the cyclic transient loads evaluated in the fatigue analysis.
Q160.
A160. (MAG)
- 4. Flow Rates and Bulk Liquid Temperatures Q161. Dr. Hopenfeld asserts that information on flow velocities is also necessary to assess the uncertainty of the heat transfer coefficients used, but that this information was not specified. Hopenfeld Report at 18 (RIV000035). What flow rates and bulk liquid temperatures did Westinghouse use in its 2010 analyses for each component, and why are they appropriate?
A161. (MAG) 109
110
- 5. Thermal Stratification and Thermal Striping in Pressurizer Surge Line System Q162. Dr. Hopenfelds testimony discusses the phenomena of thermal stratification and thermal striping at length. By way of background, did Westinghouse prepare a specialized thermal hydraulic model to represent the pressurizer and surge line system?
A162. (MAG)
Q163. Dr. Hopenfeld suggests that stratified flow in the pressurizer surge line is another non-uniform heat load that must be addressed in the fatigue evaluation. See Hopenfeld Report at 15 (RIV000035). How do you respond?
A163. (MAG) The Westinghouse EAF evaluations, including the thermal hydraulic model we just described, fully account for thermal stratification in the pressurizer surge line.
Thermal stratification refers to transient fluid temperature differences across the piping, such as a layer of warmer water lying above a layer of colder water. Dr. Hopenfeld does not address or refute any of the information in the Westinghouse calculation notes on this point. See 111
Q164. Dr. Hopenfeld also raises the issue of high frequency temperature fluctuations on the surface of the component. See Hopenfeld Report at 15 (RIV000035).
Are you familiar with this potential phenomenon?
A164. (MAG) Dr. Hopenfeld appears to be referring to the phenomenon of thermal striping in feedwater nozzles.
112
Q165. Can thermal striping (as distinct from thermal stratification) affect the pressurizer surge line?
A165. (MAG) 113
Dr. Hopenfeld, however, does not appear to distinguish between these two separate phenomena (thermal striping and thermal stratification) in his testimony. See, e.g., Supplemental Hopenfeld Report at 22 (RIV000144) (citing to studies of thermal stratification in the pressurizer surge line as the basis an assertion that [t]he pressurizer surge line is most vulnerable to fatigue failure from thermal striping.); see also id. 22 n.68.
Q166.
A166. (MAG) 114
Q167.
What is your response?
A167. (MAG)
For this reason, Dr. Hopenfeld presents no valid critique of the IPEC EAF evaluations for either the surge line hot leg nozzles or the pressurizer surge nozzles.
Q168. As an example of his concerns regarding thermal stratification, Dr.
Hopenfeld points to variations in heat transfer along nozzles and bends, which he says can introduce larger uncertainty in fatigue calculations. Hopenfeld Report at 15 (RIV000035). How do you respond?
A168. (MAG, NFA) 115
Q169. In rebuttal, Dr. Hopenfeld suggests that Entergy relied on heat transfer coefficients for the feedwater line in its evaluations of the pressurizer surge line. See Hopenfeld Rebuttal at 62 (RIV000114). How do you respond?
A169. (MAG) Dr. Hopenfeld is incorrect.
Q170. Dr. Hopenfeld also claims that the IPEC EAF evaluations for the pressurizer surge line do not reflect the most up to date work on thermal stratification. Hopenfeld Report at 16 (RIV000035). How do you respond?
A170. (MAG) Dr. Hopenfelds reference to the most up to date work is based on two documents: Institute for Energy, EUR 22763, Development of a European Procedure for Assessment of High Cycle Thermal Fatigue in Light Water Reactors: Final Report of the NESC-116
Thermal Fatigue Project (2007) (Institute for Energy Paper) (RIV000048) and Kwang-Chu Kim et al., Thermal Fatigue Estimation Due to Thermal Stratification in the RCS Branch Line Using One-Way FSI Scheme, 22 J. of Mech. Sci. & Tech. 2218 (2008) (Kim Paper)
(RIV000051).
The Institute for Energy Paper focuses on thermal stratification and cycling in mixing tees that are connected to the reactor coolant loop piping. Pressurizer surge line stratification is mentioned in this report only briefly when describing NRC Bulletin 88-11 (RIV000015).
Institute for Energy Paper at 33 (RIV000048). The focus of the paper is on thermal stratification cycling due to valve leakage and turbulent penetration in normally stagnant unisolable piping, which is not applicable to the pressurizer surge line.
As to the Kim Paper, it analyzed a shutdown cooling branch line connected to the reactor coolant system hot leg, which can experience thermal cycling and stratification, but the nature of the loading is different from the pressurizer surge line. See Kim Paper at 2218 (RIV000051).
The analysis in the Kim Paper applied thermal loading based on computational fluid dynamics (CFD). See id.
117
For example, removal of such conservatism by defining a less conservative profile between the hot and cold fluids was the subject of PVP2011-57700. See generally J. D. Burr et al., PVP2011-57700, Simulation and Evaluation of Thermal Stratification in a Sloped Surge Nozzle Correlated with Plant Measurements (July 2011) (ENT000220).
Q171. In his 2015 testimony, Dr. Hopenfeld argues that fatigue calculations that utilize an environmental factor (Fen) in accordance with MRP-47 (NYS000350) are meaningless when the component is subjected to thermal striping. Supplemental Hopenfeld Report at 26-27 (RIV000144). Do you agree?
A171. (MAG, NFA) No.
118
Q172. Dr. Hopenfeld concludes his discussion of the pressurizer surge line evaluations by claiming that more realistic heat transfer calculations alone would have increased the CUFen values calculated by Westinghouse by as much as a factor of two.
See Hopenfeld Report at 17 (RIV000035). Do you agree?
A172. (MAG, NFA) No. Dr. Hopenfeld presents no basis for the factor of two figure he asserts. For the reasons we have stated in response to Question 166, Dr. Hopenfelds critique of the pressurizer surge line nozzle evaluations lacks merit.
- 6. Environmental Correction Factor Q173. How did Westinghouse determine the Fen values to use for each component in the 2010 evaluations of NUREG/CR-6260 locations?
A173. (MAG, NFA) 119
Q174. Dr. Hopenfeld identifies numerous uncertainties inherent in the determination of CUFen, and in particular, in accounting for the reactor coolant environment, citing, among other documents, NUREG/CR-6909 (NYS000357). See Hopenfeld Report at 4-9 (RIV000035). He concludes that the EAF analyses should use appropriate bounding Fen values of 12 and 17 for stainless steel and carbon and low alloy steel, respectively. Id. at 7. How do you respond?
A174. (MAG, NFA) The methodologies and formulae set forth in NUREG/CR-6583 and NUREG/CR-5704 appropriately account for the uncertainties identified in NUREG/CR-6909 and recited by Dr. Hopenfeld. While NUREG/CR-6909 accounts for the effects of the reactor coolant environment through changes to the ASME Code design fatigue curves and through the use of Fen factors, NUREG/CR-6583 and NUREG/CR-5704 account for the same effects through defined Fen formulae.
There also is no technical basis for the claim that Entergy must use only the bounding Fen values identified in NUREG/CR-6909. The NRC Staff has found the use of NUREG/CR-6909, NUREG/CR-6583, or NUREG/CR-5704 to be acceptable. See NUREG-1801, Rev. 1, at X M-1 (NYS00146C); NUREG-1801, Rev. 2, at X M1-1 (NYS00147C).
Moreover, Dr. Hopenfeld misreads NUREG/CR-6909 itself. Nothing in that document requires or recommends the use of the bounding values of 12 and 17 as the Fen in all cases.
Thus, while Dr. Hopenfeld portrays the field of fatigue analysis as facing insurmountable 120
obstacles that demand the use of only the most conservative Fen factors, see, e.g., Hopenfeld Report at 5 (RIV000035) (The Electrical [sic - Electric] Power Research Institute has explained that [i]n many respects, the current state of the technology with respect to the Fen methodology is incomplete or lacking in detail and specificity.) (quoting MRP-47 at 4-25 (NYS000350)),
this is not the case. On the contrary, the very next sentence of MRP-47, omitted by Dr.
Hopenfeld, states that [r]ecommendations are made in this guideline where needed to fill in these missing details. MRP-47 at 4-25 (NYS000350). That is what Westinghouse did.
Q175. Dr. Hopenfeld identifies what he describes as uncertainties inherent in the determination of CUFen purportedly identified in NUREG/CR-6909. See Hopenfeld Report at 5 (RIV000035). How did the Westinghouse EAF analyses account for uncertainties in applying the formulae in NUREG/CR-6583 and NUREG/CR-5704?
A175. (MAG, NFA) For stainless steels, NUREG/CR-5704, Section 7 defines the Fen formulas to be used with the design fatigue curves, and, in Section 9 describes the approach to evaluate EAF by applying the Fen factors to the ASME fatigue usage. For carbon and low-alloy steels, NUREG/CR-6583, Section 6 defines the Fen formulas to be used with the design fatigue curves, and, in Section 8 describes the approach to evaluate EAF by applying the Fen factors to the ASME fatigue usage. These factors are designed to account for uncertainties in applying the ASME Code fatigue curves to the reactor coolant environment, including the uncertainties in materials, loading, and environment Dr. Hopenfeld has identified in his quotation from NUREG/CR-6909. See generally NUREG/CR-5704 (NYS000354); NUREG/CR-6583 (NYS000356). The NRC Staff has approved both NUREG/CR-5704 and NUREG/CR-6583 for this purpose. See NUREG-1801, Rev. 1, at X M-1 (NYS00146C); NUREG-1801, Rev. 2, at X M1-1 (NYS00147C).
121
Q176. On page 8 of his Report (RIV000035), Dr. Hopenfeld presents a chart applying the bounding Fen factors identified in NUREG/CR-6909 to the Westinghouse EAF evaluations, to derive new CUFen values. Do you agree with this approach?
A176. (MAG, NFA) No. Dr. Hopenfelds combination of the bounding Fen factors mentioned in NUREG/CR-6909 (i.e., disregarding the methodology specified in NUREG/CR-6909 to calculate more specific Fens) with values derived from the ASME Code design air curves for carbon steel and low-alloy steels contained in NUREG/CR-6583 and NUREG/CR-5704, is inappropriate. The values derived from this approach are unrealistically high, and are not consistent with the guidance in those three documents. Dr. Hopenfelds CUFen calculations certainly reveal no deficiency in the Westinghouse EAF evaluations, which relied on the guidance in NUREG/CR-6583 and NUREG/CR-5704.
Q177. Further, Dr. Hopenfelds recommendations are inconsistent with the NRC Staffs guidance in NUREG-1801. See NUREG-1801, Rev. 1, at X M-1 (NYS00146C)
(Formulae for calculating the environmental life correction factors are contained in NUREG/CR-6583 for carbon and low-alloy steels and in NUREG/CR-5704 for austenitic stainless steels.); NUREG-1801, Rev. 2, at X M1-1 (NYS00147C) (allowing licensees to use the formulae provided in NUREG/CR-6583 or NUREG/CR-6909 for carbon and low-alloy steels, and those provided in NUREG/CR-5704 or NUREG/CR-6909 for stainless steels).
Within his discussion of NUREG/CR-6909, Dr. Hopenfeld states that Fen values of 12 and 17 were bounding in laboratory environments, it is reasonable to expect even higher Fen values in the actual reactor environment, especially for those components that experience stratified flows and thermal striping. Hopenfeld Report at 7 (RIV000035). Do you agree?
122
A177. (MAG, NFA) No. Dr. Hopenfeld provides no support for his assertion and we disagree with it. First, there is no evidence that the bounding values derived from laboratory testing are non-conservative relative to actual components in typical plant operating environments. In our opinion, the opposite is truethe effects of reactor water environments seen in the laboratory have yet to be seen in operating environments. See Cooper Paper at 7 (ENT000215); SECY-95-245, Completion of the Fatigue Action Plan (Sept. 25, 1995)
(ENT000224) (documenting the completion of the NRCs fatigue action plan, including evaluations based on laboratory data, not operating experience); SRP-LR Rev. 2, at 4.3-3 (NYS000161) (the calculations supporting resolution of this issue, which included consideration of environmental effects, indicate the potential for an increase in the frequency of pipe leaks as plants continue to operate) (emphasis added).
The other part of Dr. Hopenfelds assertion appears to claim that the assessment of stratified flow and thermal striping is inaccurate when reactor water environmental effects are taken into account. However, he provides no justification for his assertion. If anything, the opposite is true, because the high-cycle nature of the loading in those cases has a high strain rate.
Under those conditions, the effects of the reactor water environment are insignificant. See MRP-47 at 4-23 (NYS000350).
Q178. In his 2015 supplemental report, Dr. Hopenfeld relies on the alleged statements of Dr. Omesh Chopra of the Argonne National Laboratory before the ACRS for the propositions that it is the responsibility of the operator to account for the differences between the lab and plant environments when applying the results, and that the ANL results may not be conservative. See Supplemental Hopenfeld Report at 7 (RIV000144)
(citing Transcript, Advisory Committee on Reactor Safeguards, Subcommittee on 123
Materials, Metallurgy and Reactor Fuels at 22 and generally (Dec. 6, 2006) (RIV000037)).
Do you agree with his discussion of this issue?
A178. (MAG, NFA, RGL) No. Dr. Hopenfeld attributes these statements to Dr.
Chopra, the principal investigator of the ANL research. However, these are not quotations from Dr. Choprathey merely represent Dr. Hopenfelds selective interpretation, and do not reflect Dr. Chopras statements before the ACRS.
Q179. Relying on the ACRS transcript, Dr. Hopenfeld goes on to assert that the IPEC EAF evaluations have not accounted for the known differences between the laboratory and plant environments. Supplemental Hopenfeld Report at 7 (RIV000144)
How do you respond?
A179. (MAG, NFA, RGL) Based on his incorrect characterization of Dr. Chopras statements, Dr. Hopenfeld asserts that Entergy must use the general form of the Fen equation presented early in NUREG/CR-6909 (NYS000357) and asserts, without support, that the designer must use this general equation for each location analyzed. See Supplemental Hopenfeld Report at 7 (RIV000144). However, Dr. Hopenfeld fails to acknowledge the rest of NUREG/CR-6909 which: (1) develops applications of test data for different materials; (2) develops methods and margins to account for the various factors to be considered in evaluations; and (3) presents final equations for specific material types, with the ranges and limits specified for each input variable. See NUREG/CR-6909, App. A (NYS000357). In other words, the ANL results discussed in NUREG/CR-6909 (and the correction factors specified in other NRC guidance documents) already account for the differences between the lab and plant environments. Dr. Hopenfelds statements disregard these facts.
124
Q180. Dr. Hopenfeld has also stated that Entergys witnesses fail[ed] to address the twelve factors that are allegedly prerequisites for the use of NUREG/CR-5704 and NUREG/CR-6583 rather than NUREG/CR-6909. See Hopenfeld Rebuttal at 11-13 (RIV000114). Do you agree?
A180. (MAG) No. NUREG/CR-5704 and NUREG/CR-6583 in fact address the factors required in their determination of the applicable fatigue curves and corresponding Fen equations to be used with respect to the data. See NUREG/CR-5704 §§ 3-5 (Fen), 8 (design fatigue curves)
(NYS000354); NUREG/CR-6583 §§ 4-6 (Fen), 7 (design fatigue curves) (NYS000356).
- 7. Dissolved Oxygen and Water Chemistry Q181. Dr. Hopenfeld raises several questions about the consideration of dissolved oxygen (DO) in the EAF evaluations. For background purposes, how did the Westinghouse EAF analyses consider the concentration of DO in the reactor coolant system, and its potential effect on the EAF correction factor?
A181. (MAG) 125
Under the approach in both NUREG/CR-5704 (stainless steels) and NUREG/CR-6583 (carbon steels), the Fen factor to be used is in part dependent on the product of three values:
transformed oxygen (O*), which represents the impact of DO concentration on the Fen, transformed temperature (T*), which represents the impact of fluid temperature on the Fen, and transformed strain rate (*), which represents the impact of the rate of change of the strain in the material during the transient. If either O* or T* (or *) equals zero, then the product of this portion of either Fen formula also equals zero and the Fen value is determined by other empirically-derived constants. These formulas were developed by ANL and are approved by the NRC in NUREG-1801, Revisions 1 and 2. See NUREG-1801, Rev. 1, at X M-1 (NYS00146C);
NUREG-1801, Rev. 2, at XM1-1 (NYS00147C).
Q182. How does DO concentration impact the correction factors for stainless steels and carbon steels?
A182. (MAG) For stainless steels, lower DO generally will actually produce larger Fen values. The most conservative NUREG/CR-5704 Fen is based on DO results from DO < 0.05 ppm, yielding an O* term of 0.260. See NUREG/CR-5704, § 3.2.3 (NYS000354).
For carbon steels, the NUREG/CR-6583 Fen equation specifies that the O* parameter, which is affected by the input for DO, will be zero when DO < 0.05 ppm. See NUREG/CR-6583, § 3.2.3 at 60 (NYS000356).
Q183. Did the Westinghouse EAF evaluations for IPEC follow this guidance in considering DO?
A183. (MAG) Yes. The 2010 EAF evaluations followed this guidance.
126
Q184. How do temperature and DO concentration interact in the calculation of the Fen?
A184. (MAG) As previously noted, the NUREG/CR-6583 Fen equation for carbon steels contains a transformed temperature (T*) parameter, which is affected by input for temperature.
NUREG/CR-6583 at 60 (NYS000356).
The use of the product of the transformed strain rate, oxygen, and temperature values is based on experimental data underlying the environmental correction factors, which shows that the decrease in fatigue life due to environmental factors is significant only when four conditions are satisfied simultaneously, viz., when the strain amplitude, temperature, and DO in water are above certain threshold values, and the strain rate is below a threshold value. NUREG/CR-6815 at 10 (emphasis added) (ENT000225) 127
Q185. Dr. Hopenfeld states that the Westinghouse EAF evaluations do not correct the Fen value to account for the fact that when temperature is below 150º C, fatigue life could be reduced by a factor of two. Hopenfeld Report at 7 (RIV000035) (citing NUREG/CR-6909 at 26-28 (NYS000357)). How do you respond?
A185. (MAG, NFA) The referenced section of NUREG/CR-6909 states the decrease in life is greater at high temperatures and DO levels. [However,] a moderate decrease in fatigue life is observed in water at temperatures below the threshold value of 150º C or at DO levels 0.05 ppm . . . .
NUREG/CR-6909 at 26 (NYS000357) (emphasis added).
Q186. Why are the assumptions regarding DO content in the reactor coolant system valid?
A186. (NFA, BMG) These assumptions are valid because hydrazine, which acts as an oxygen scavenger, reacting with oxygen to form nitrogen and water, is added to the IP2 and IP3 reactor coolant prior to heatup above 180 ºF (82 ºC). See EPRI, BWRVIP-187, BWR Vessel and Internals Project: Controlling Intergranular Stress Corrosion Cracking During a BWR Startup-Addition of Hydrazine or Carbohydrazide at 4-1 (Mar. 2008) (ENT000691) (showing hyrdazine-oxygen reaction); Entergy, 0-CY-2310, Rev. 24, Reactor Coolant System Specification and Frequencies at 11 (Jan. 16, 2015) (IPEC RCS Specifications) (ENT000692) (showing IPEC specifications). At higher temperatures, including during power operations, hydrogen is added to 128
the system to scavenge oxygen by combining with it to form water. See id. This is illustrated in plant chemistry records, which show that during the startup following the most recent refueling outage, DO was measured to be below 0.05 ppm (50 ppb) before the plants heated up above 200ºF (93ºC), which is consistent with the assumptions made by Westinghouse. See IPEC, Unit 3 Chemistry Data at 2 (Mar. 21-22, 2015) (ENT000693) (showing DO < 2.5 ppb prior to heatup above 188 ºF).
Q187. How do you respond to Dr. Hopenfelds claim that oxygen dissolved in the coolant will increase significantly during shutdown transients? See Hopenfeld Report at 10-11 (RIV000035).
A187. (NFA, ABC, BMG) To the extent Dr. Hopenfeld is referring to a transient involving the shutdown and cooldown of the plant, this is not an issue of concern for PWRs like IP2 and IP3 because the temperature term in the Fen equation is zero at temperatures less than 150°C. To the extent he is referring to transients that can take place while the plant temperature is above 150ºC, his issue also is not of concern because the IPEC units are operated in a manner that precludes a ready source of DO.
During normal PWR operation, DO is limited to less than 0.005 ppm. See IPEC RCS Specifications at 11 (ENT000692). Dr. Hopenfelds reliance on his fundamental law of physics regarding the negative solubility coefficient of oxygen in water is misplaced.
Hopenfeld Report at 10 (RIV000035). In a transient accompanied by significant temperature reduction, the ability of the water to hold DO will increase. However, the DO concentration will not spontaneously increase as temperature is lowered if there is no available source of oxygen.
The reactor coolant system operates well above atmospheric pressure, and is treated with hydrogen to scavenge oxygen. Accordingly, the DO will not increase above specifications until 129
the system is depressurized, which takes place after the temperature is below the T* threshold described above.
Q188. Dr. Hopenfeld further asserts that the heating period below 150°C (300°F) does not represent or bound all transients and that for transients where the temperature is between 150º F and 600º F, the EAF analyses assume that T* = 0, improperly discount[ing] the presence of oxygen and the effect it will have on fatigue life. Hopenfeld Report at 12 (RIV000035); see also Supplemental Hopenfeld Report at 9-11 (RIV000144).
How do you respond?
A188. (MAG, NFA)
Under the equations in NUREG/CR-6583, because of how the IPEC PWRs are operated and how water chemistry is controlled, the product of O* and T* always equals zero.
Q189. Dr. Hopenfeld cites to a diagram in MRP-47 which, he states, calculates Fen values as high as 130 at high DO levels, which is two orders of magnitude higher than those 130
calculated by Entergy. Hopenfeld Report at 13 (RIV000035) (citing MRP-47 at 4-22 (NYS000350)). How do you respond?
A189. (MAG, NFA) We disagree. MRP-47 actually shows that the high DO curve cited in the assertion applies only to BWRs that do not use hydrogen water chemistry (HWC) and has no relationship to the operating conditions of PWRs, since PWRs such as IP2 and IP3 operate to chemistry specifications requiring DO below 0.05 ppm at high temperatures. If one looks at the portions of the curves on the figures Dr. Hopenfeld cites that are applicable to PWRs, then one can see that the Fen multipliers are considerably lower than 130.
Q190. Does Dr. Hopenfeld base his concerns about DO on other documentation regarding BWRs?
A190. (NFA, ABC, BMG) Yes. Dr. Hopenfeld fails to distinguish between PWRs and BWRsthe latter operating at higher DO concentrationswhen he cites an EPRI R&D Status Report from over thirty years ago that applies to BWRs, claiming that EPRI data on actual oxygen concentrations vary with the change in temperature by more than an order of magnitude in comparison to oxygen levels during normal operating conditions. Hopenfeld Report at 10 &
- n. 36 (RIV000035) (citing John J. Taylor, R&D Status Report, Nuclear Power Division, EPRI J.
52 (Jan./Feb. 1983) (RIV000040)). He goes on to assert, without any basis, that [s]imilar oxygen dependence on temperature can be expected in PWRs. Id. at 10. As we have just shown, however, that is not the case at the temperatures of concern because the PWR environment is deoxygenated.
Q191. Dr. Hopenfeld criticizes your testimony on DO issues, claiming that Entergy has ignored his argument that it has presented no plant data on DO. See Hopenfeld Rebuttal Testimony at 38-40 (RIV000114); Supplemental Hopenfeld Report at 14 131
(RIV000144). He also asserts that [t]he Fen equation requires that the oxygen input be a real value based on actual measurements not on assumptions. Id. How do you respond?
A191. ( NFA) DO levels in the RCS at IP2 and IP3 are measured approximately three times per week and the normal values are < 0.0025 PPM. See IPEC, Unit 3 Chemistry Data at 2 (Mar. 21-22, 2015) (ENT000693) (showing DO < 2.5 ppb prior to heatup above 188 ºF). This value is 20 times lower than the conservative lower bound of 0.05 ppm used in the EAF evaluations. Thus, Dr. Hopenfelds claim that actual plant measurements must be used instead of Westinghouses conservative assumptions appears to be based on BWR practices and lacks basis for IPEC.
Q192. Relatedly, Dr. Hopenfeld asserts that Entergy and Westinghouse completely ignored ANL guidance on DO values, which recommend[s] using 0.4 ppm, and EPRI guidance on DO values, which requires the use of maximum DO values for carbon steel.
Supplemental Hopenfeld Report at 9 (RIV000144) (citing NUREG/CR-6909 at A.5 (NYS000357)). Do you agree?
A192. (MAG, NFA) No. NUREG/CR-6909 says that a value of 0.4 ppm can be used for the DO content to perform a conservative evaluationit does not say that it recommends these values for all transients. NUREG/CR-6909 at A.5 (NYS000357). This is simply a worst-case scenario suggestion if no information is available or the analyst is satisfied with using excessive conservatism. Because we know the DO content at IP2 and IP3 is considerably lower when the plants are in operation, there is no reason to use this overly-conservative value.
Q193. Dr. Hopenfeld claims that, in 2011, ANL modified the definitions of T* and O* . . . such that now the second term in the exponent of the Fen equation is not zero any more below 150°C. Supplemental Hopenfeld Report at 11 (RIV000144) (citing ANL-132
LWRS-47, Report on Assessment of Environmentally-Assisted Fatigue for LWR Extended Service Conditions (Sept. 2011) (RIV000150)). Is Entergy required to use this revised equation?
A193. (MAG, NFA) No. The subsequent ANL change to the equation was published in the Draft NUREG/CR-6909, Rev. 1 (NYS000490A-B), which was made available for public comment in 2014. Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, 79 Fed. Reg. 21,811 (Apr. 17, 2014). The draft has not been finalized, and it is certainly not required, or even endorsed, by NUREG-1801. In any event, as we have previously explained in response to Question 94, in general, under the proposed new guidance, the Fen values would be lower than those used in the IPEC EAF analyses.
Q194.
A194. (MAG, NFA)
. ANL reports repeatedly state that all of the parameters (temperature, oxygen, and strain rate) must be above their respective threshold values for there to be a significant environmental effect. See ANL-LWRS-47 at 27 ([E]nvironmental effects on fatigue life are significant only when critical parameters (temperature, strain rate, DO level, and strain amplitude) meet certain threshold 133
values. Environmental effects are moderate, e.g., less than a factor of 2 decrease in life, when any one of the threshold conditions is not satisfied.) (RIV000150) (emphasis added).
Q195. Dr. Hopenfeld states that [h]ydrazine does not remove oxygen from the reactor system; it only changes its form, which varies with the temperature. Thus, the oxygen could be bound to metal surfaces or be floating crud in the system, and thereby presumably contribute to environmental effects on fatigue. Supplemental Hopenfeld Report at 8 (RIV000144). Do you agree?
A195. (NFA, BMG, ABC) No. Dr. Hopenfeld cites no references or evidence for his speculative concerns about additional DO that might be bound to metal surfaces or in floating crud in the system. On the contrary, hydrazine combines with oxygen to form stable compounds that do not yield DO when system conditions change. See BWRVIP-187 at 4-1 (ENT000691). The DO used in the development of the Fen equations is the bulk concentration in the fluid used in the laboratory tests from which the equations were developed. Using bulk concentration of DO measured in the plant would be consistent with the measurement of DO in the lab tests. Dr. Hopenfeld has provided no basis to support a theory that the DO concentration at the surface of a component could be a factor of 10 above the bulk DO concentration of the reactor coolant water.
134
Q196. Dr. Hopenfeld asserts that there is no evidence that Entergy considered the presence of trace impurities on water conductivity, which reduces fatigue life and that the EAF evaluations did not consider the potential synergistic interaction between fatigue and stress corrosion cracking caused by chlorides. Hopenfeld Report at 7 (RIV000035)
(citing NUREG/CR-6909 at 30-31) (NYS000357). How do you respond?
A196. (NFA, BMG) The potential for trace impurities in the reactor coolant to contribute to stress corrosion cracking is addressed through the Water Chemistry Control -
Primary and Secondary Program. See NUREG-1801, Rev. 1, at XI M-10 (NYS00146C); LRA App. B at B-137 to -39 (ENT00015B). This is consistent with the approach in NUREG/CR-6909, which, on the very pages cited by Dr. Hopenfeld, states: Normally, plants are unlikely to accumulate many fatigue cycles under off-normal conditions. Thus, effects of water conductivity on fatigue life have not been considered in the determination of Fen. NUREG/CR-6909 at 30 (NYS000357). Moreover, as previously noted in response to Question 66, above, Entergy relies on several other inspection programs to manage the effects of aging due to cracking caused by SCC or other mechanisms.
Q197. Dr. Hopenfeld also claims that since oxygen plays a part in causing stress corrosion cracking (SCC) and SCC has been observed on numerous occasions in PWRs, oxygen simply cannot be zero. Supplemental Hopenfeld Report at 14 (RIV000144). Is this assertion valid?
A197. (NFA, BMG) As a threshold matter, we have never stated that DO is zero; only that DO concentration is very low, below 0.005 ppm (5 ppb) during normal operation. Dr.
Hopenfeld provides no support for the idea that SCC requires the presence of levels of oxygen higher than plant chemistry specifications. In fact, SCC mechanisms can occur, and have 135
occurred, in deaerated PWR environments with <5 ppb DO content. See Z. Szklarska-Smialowska et al., Mechanism of Crack Growth in Alloy 600 in High-Temperature Deaerated Water, 50 CORROSION 676 (Sept. 1994) (ENT000694). Therefore, there is no basis for Dr.
Hopenfelds assertion that the presence of SCC in PWRs somehow indicates that there are higher DO levels.
- 8. No Propagation of Error Analysis Is Required or Necessary Q198. Dr. Lahey raises a number of issues regarding the use of engineering judgment in the EAF analyses. For example, he states that [t]here was obviously a lot of engineering judgment used in obtaining some of the fatigue analysis results in the EAF analyses. Supplemental Lahey Report at 3; see also id. at 4, 7-8 (NYS000297); Lahey Report at 28 (NYS000296). Please respond.
A198. (MAG, NFA) As we explain throughout our testimony, the assumptions used in Westinghouses EAF evaluations are consistent with accepted engineering practice. The practice has been developed over time, contains considerable conservatisms, and has been used to effectively evaluate the effects of fatigue. As we also show throughout our testimony, Dr. Lahey has not cited any valid example of where the Westinghouse evaluations are non-conservative, or how the calculations consistent with these practices are deficient for their intended purposes.
The completed EAF calculations are conducted by qualified analysts in accordance with Westinghouses NRC-approved quality assurance program and are reviewed by other experienced and qualified analysts. See Westinghouse Level 2 Policy/Procedures, NSNP 3.2.6, Design Analysis at 5-6 (Mar. 2011) (ENT000196). This is consistent with traditional ASME Code stress analysis. Engineering expertise is necessary in performing fatigue analyses, as it is in other types of complex engineering analysis. See 10 C.F.R. Pt. 50, App. B, § II (Quality Assurance Program) (The program shall take into account the need for special controls, 136
processes, test equipment, tools, and skills to attain the required quality . . . . The program shall provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained.). Fatigue analysis demands training and experience, and is not simply a mathematical exercise that can be undertaken without the appropriate background.
Dr. Laheys observation about the obvious use of a lot of engineering judgment may stem from his apparent unfamiliarity with fatigue analyses of this type.
137
In any case, there is no need to precisely quantify uncertainties arising from the use of engineering judgment because the EAF analyses are conservative, bounding analyses. While Dr.
Lahey argues that modeling and input assumptions lead to results that are highly uncertain and unreliable, see Supplemental Lahey Report at 8 (NYS000297), most engineering analysis requires assumptions and inputs. As we have shown in throughout our testimony, conservative modeling and input assumptions have been used at each step of the fatigue analyses, thereby providing confidence that the results are reliable for managing the effects of fatigue throughout the PEO.
Q199. On the issue of engineering judgment, Dr. Lahey identifies a series of possible sources of error in the EAF analyses.
A199. (MAG, NFA) 138
Q200. Dr. Lahey has suggested that a CUFen that is close to, but does not exceed 1.0, means that virtually any error would put some of the calculated values of CUFen over the CUFen = 1.0 fatigue failure limit, and that, accordingly, Entergy must conduct a propagation of error analysis. Revised Lahey Report at 67 (NYS000530); see also Lahey Report at 27 (NYS000296); Hopenfeld Report at 21 (RIV000035). Do you agree?
A200. (MAG, NFA, RGL) No. As a threshold matter, the EAF calculations determine projected CUFen values that would result if all of the transients actually occurred the number of times assumed in the calculation. The ASME Code has long recognized that there are uncertainties associated with both analytical inputs and modeling techniques. These uncertainties are addressed through the margin factors we discussed in response to Question 70, above, rather than through error analyses, as Dr. Lahey suggests. As we explain throughout our testimony, the IPEC EAF evaluations have been prepared with variables purposefully chosen to reasonably bound expected values. Because the inputs are not best-estimate values of a normal distribution, a propagation of error analysis is inappropriate.
For this reason, NUREG-1801, Revision 1 (Section X.M1) (NYS00146C) and the acceptance criteria for fatigue analysis in the SRP-LR (Section 4.3) (NYS000195) do not specify any need for uncertainty analyses to validate ASME Code or ANSI B31.1 fatigue analyses. In addition, the ASME Code fatigue analysis methods endorsed by NRC in 10 C.F.R. § 50.55a do not establish any requirements for propagation of error analyses. See generally ASME Code,Section III, Article NB-3000 (NYS000349). Dr. Lahey has provided no regulatory or technical 139
basis to demonstrate the need to perform uncertainty analyses for ASME Code Section III or ANSI B31.1 fatigue analyses.
Finally, Drs. Lahey and Hopenfeld only speculate that there are many possible sources of error in the EAF analyses, Supplemental Lahey Report at 2 (NYS000297), that could lead to a violation of the 1.0 limit, Lahey Report at 27 (NYS000296) (emphasis added), they do not substantiate or quantify the postulated errors or uncertainties. See also Hopenfeld Report at 21 (RIV000035) (Given the large uncertainties . . . the detrimental effects of the environment on fatigue strength, and resulting predicted fatigue life, of the components evaluated are likely grossly underestimated.) (emphasis added)). To the contrary, the detrimental effects of the environment are likely overestimated because of the conservative bias applied to the analyses.
Therefore, Dr. Laheys and Dr. Hopenfelds assertions are unsupported.
Q201. Dr. Lahey cites to an engineering textbook in support of his theory that one would normally expect to see a detailed propagation-of-error analysis for the EAF evaluations. See Lahey Report at 27 (NYS000296) (citing S. Vardeman and J.M. Jobe, Basic Engineering Data Collection and Analysis, at 310-11 (2001) (NYS000347)); see also Revised Lahey Testimony at 70 (NYS000530). Do you agree with Dr. Lahey?
A201. (MAG, NFA, RGL) No. The Vardeman & Jobe textbook (NYS000347) Dr.
Lahey references deals with the interpretation of randomly collected test dataan issue that is not applicable to the IPEC EAF analyses, which considered the entire relevant data population.
Q202.
140
A202. (MAG, NFA)
Q203. Relatedly, Dr. Lahey asserts that the NRC should insist that an error analysis be performed because the NRC recently issued an inspection report with notices of non-conformance of Westinghouses QA program. Revised Lahey Testimony at 70-71 (NYS000530) (citing Letter from E. Roach to S. Hamilton, Nuclear Regulatory 141
Commission Vendor Inspection of Westinghouse Electric Company LLC, Cranberry Township, Report No. 99900404/2015-202 and Notices of Nonconformance (Apr. 24, 2015)
(NYS000515) (NRC Inspection Letter)). How do you respond?
A203. (MAG, RGL) The NRC Inspection Letter has nothing to do with WESTEMSTM, the EAF evaluations prepared for IPEC, or any other Westinghouse work related to the license renewal proceeding for IPEC. As Dr. Lahey himself states, the NRC Inspection Letter documented certain nonconformances in Westinghouses corrective actions, oversight of suppliers, and audits. Revised Lahey Testimony at 70 (NYS000530); see also generally NRC Inspection Letter (NYS000515). Under these circumstances, the NRC Inspection Letter provides no basis for the NRC to insist that an error analysis be performed to ensure the validity of Westinghouses fatigue evaluations for IP2 and IP3.
Q204. Dr. Hopenfeld alleges that the EAF analyses lack conservatism because of particular uncertainties that he has identified. E.g., Hopenfeld Report at 21 (RIV000035). For example, Dr. Hopenfeld states that Entergy repeatedly applies the term bounding to its analyses and results, implying that such results are conservative, and that no error analysis is necessary. Id. How do you respond to this point?
A204. (MAG, NFA) Dr. Hopenfelds statements conflate uncertainty with non-conservatism. As we explained in our response to Question 71, there are significant margins and conservatism inherent in ASME Code requirements and, thus, the Westinghouse EAF analyses.
Entergy and Westinghouse followed accepted engineering practice, where analysis inputs are selected conservatively with respect to expected loading parameters, in order to ensure the results are conservative. Best-estimate analyses for which uncertainty analyses are typically and more appropriately performed would result in less conservative results than the Westinghouse 142
analyses. We address the specific uncertainties that Dr. Hopenfeld identifies individually and in more detail throughout this Section (Section V.D.8). The States expert, Dr. Lahey, makes similar claims when he identifies various possible sources of error in the Westinghouse EAF analyses, Supplemental Lahey Report at 2 (NYS000297), without acknowledging that the uncertainties he identifies were recognized by the analysts and addressed with the selection of conservative parameter values.
Q205.
A205. (MAG, NFA, RGL)
In fact, to the extent that the environmental adjustment introduces additional conservatism, the conservatisms in the analysis are increased.
143
Q206. In your 2012 testimony, you stated that the purpose of a CUFen calculation is to determine whether or not the CUFen exceeds 1.0, not to calculate a precise CUFen value.
In rebuttal, Dr. Lahey labeled this statement as absurd and states that this reflects the lack of rigor in the refined EAF analyses prepared by Westinghouse. Lahey Rebuttal at 11 (NYS000440). How do you respond?
A206. (NFA, ABC, JRS, MAG) Our statement is consistent with 10 C.F.R.
§ 50.55a(c)(1), which specifies that components must meet the requirements of the ASME Code,Section III. To satisfy this regulatory requirement, it is acceptable if the analyst can show that the CUFen remains below 1.0.
Dr. Laheys labeling this process as absurd also is contrary to the license renewal process under 10 C.F.R. Part 54. Under the GALL Report, the acceptance criterion for the Fatigue Monitoring AMP is that the CUFen values, calculated using an acceptable methods provided in the GALL Report, remain below the fatigue design limit of 1.0 specified in the ASME Code and the regulations. See NUREG-1801, Revision 1 at X M-1 (NYS00146C); see also NUREG-1801, Revision 2 at X M1-2 (NYS00147C); 10 C.F.R. § 50.55a(c)(1). If a fatigue evaluation shows that the 1.0 limit is exceeded, then it is acceptable to prepare a more rigorous analysis of the component to demonstrate that the design code limit will not be exceeded during the extended period of operation. NUREG-1801, Revision 1 at X M-2 (NYS00146C)
(emphasis added); see also NUREG-1801, Revision 2 at X M1-2 (NYS00147C). If the applicant can show that CUFs, considering environmental effects, remain below 1.0, then the applicant has shown that there is reasonable assurance that plant activities will continue to be conducted in accordance with the CLB throughout the PEO, as required under 10 C.F.R. §§ 54.21(c)(iii) and 54.29.
144
E. Under Commitments 43 and 49 Entergy Has Evaluated the Limiting Locations for Fatigue at IPEC and Conservatively Demonstrated that the CUFens for Limiting Locations Do Not Exceed 1.0
- 1. Overview of the Limiting Locations Review Q207. Did Westinghouse conduct additional analyses to determine whether there were any locations that may be more limiting than those identified in NUREG/CR-6260?
A207. (MAG, NFA) Yes. Westinghouse completed a screening evaluation of CLB ASME Code Class 1 components and RVI components to identify leading locations at IP2 and IP3 in November 2012. See generally Westinghouse Calculation Note CN-PAFM-12-35 (NYS000510). This screening was a first-pass evaluation designed to identify potential locations for further analysis, the latter of which is discussed below.
Q208.
A208. (MAG) 145
Q209. What were the results of the screening analyses?
A209. (MAG, NFA)
Q210.
A210. (MAG) 146
Q211. What were the results of the refined EAF evaluations for non-NUREG/CR-6260 locations?
A211. (MAG, NFA) The Westinghouse calculations found that all CUFens for RVI locations and potentially limiting equipment locations were less than 1.0. See Westinghouse Calculation Note CN-PAFM-13-32, Rev. 3 at 7-8 (ENT000683); Westinghouse Calculation Note CN-PAFM-13-40 at 11 (ENT000688). These results indicate that further refined analysis (such as a WESTEMSTM analysis, as performed for the 6260 locations) would result in even lower CUFen values; therefore, the analyses demonstrated that the NUREG/CR-6260 locations originally evaluated were in fact limiting locations for fatigue at IP2 and IP3, and that the CUFen does not exceed 1.0 for all RVI components with CLB CUFs.
Q212. Do the evaluations documented in Westinghouse Calculation Notes CN-PAFM-12-35 (NYS000510), CN-PAFM-13-32, Rev. 3 (ENT000683), and CN-PAFM-13-40 (ENT000688) address Entergys LRA Commitments 43 and 49?
A212. (MAG, NFA, JRS) Yes. In Westinghouse Calculation Note CN-PAFM-12-35 (NYS000510), Westinghouse review[ed] design basis ASME Code Class 1 fatigue evaluations to determine whether the NUREG/CR-6260 locations that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting locations for the IP2 and IP3 configurations, as Entergy committed to do in the first part of Commitment 43. See NL-11-032, Attach. 2 at 17 (NYS000151). Since more potential limiting locations were identified, Westinghouse, in Calculation Notes CN-PAFM-13-32, Rev. 3 (ENT000683) and CN-PAFM 147
40 (ENT000688), evaluated the most limiting locations for the effects of the reactor coolant environment on fatigue usage, as Entergy committed to do in the second part of Commitment 43.
See generally Westinghouse Calculation Note CN-PAFM-13-40 (ENT000688). Additionally, Westinghouse use[d] the NUREG/CR-6909 methodology in the evaluation of the limiting locations consisting of nickel alloy. NL-11-032, Attach. 2 at 17; Westinghouse Calculation Note CN-PAFM-12-35 at 25 (NYS000510) (For the IP2/IP3 EAF screening . . . NUREG/CR-6909 is used for nickel alloy steels); see also generally Westinghouse Calculation Note CN-PAFM-13-32, Rev. 3 (ENT000683). These evaluations included recalculations of limiting RVI locations as well, so they also address Commitment 49 for both units.
Q213. Did the NRC Staff conduct an inspection of license renewal commitment implementation at IP2?
A213. (NFA, ABC) Yes. In September 2013, the NRC Staff examined Westinghouse Calculation Notes CN-PAFM-13-32, Rev. 0, and CN-PAFM-12-35, Rev. 1, and determined that Commitment 43 for IP2 had been appropriately implemented, and therefore closed the commitment. See Letter from J. Trapp, NRC, to J. Ventosa, Entergy, Indian Point Nuclear Generating Unit 2 - NRC License Renewal Team Inspection Report 05000247/2013010 at 6-7, A-1 (Sept. 19, 2013) (ENT000695). Entergy anticipates a similar inspection at IP3 prior to the PEO.
- 2. EAF Evaluations for RVI Components Q214. Did the refined EAF evaluations documented in Westinghouse Calculation Note CN-PAFM-13-32, Rev. 3 (ENT000683) include consideration of RVI locations?
A214. (MAG, NFA, RGL) Yes. Consistent with Entergys LRA Commitment 49, Westinghouse recalculate[d] each of the limiting CUFs provided in Section 4.3 of the LRA for the reactor vessel internals to include the reactor coolant environment effects (Fen) as provided in 148
the IPEC Fatigue Monitoring Program using NUREG/CR-5704 or NUREG/CR-6909. NL 052, Attach. 1 at 9 (NYS000501); see also Westinghouse Calculation Note CN-PAFM-13-32, Rev. 3 (ENT000683).
Q215. In evaluating environmental effects on RVI components, is it necessary to apply an additional correction factor for the effects of irradiation embrittlement, as Drs.
Lahey and Hopenfeld suggest? See, e.g., Revised Lahey Testimony at 15 (NYS000530)
(claiming that synergistic interactions have not been considered for RVIs); see also Supplemental Hopenfeld Report at 23-25 (RIV000144).
A215. (NFA, RGL) No. As we have previously explained in response to Question 76, fatigue and irradiation embrittlement are separate effects and there is no basis to apply an additional fatigue correction factor to address potential embrittlement. Data on the effects of irradiation on fatigue suggest that further correction factors for fatigue are not warranted. See MRP-175 at D-3 ([t]he work of several researchers suggest that neutron irradiation does not result in a further reduction in fatigue properties and in some cases suggests an improvement.)
(ENT000631).
Q216. Is this issue considered in the draft revision to NUREG/CR-6909?
A216. (MAG, NFA, RGL) Yes. The recently-issued Draft NUREG/CR-6909, Rev. 1 (NYS000490A-B) includes a discussion of irradiation effects, and concludes that no additional augmentation of the calculated Fen is required to account for irradiation effects.
Specifically, the authors of the draft report, Dr. Omesh Chopra of ANL and Mr. Gary Stevens of the NRC, considered the potential effects of neutron irradiation on fatigue life. Based on their review of the available data:
Additional fatigue data on reactor structural materials irradiated under LWR operating conditions are needed to determine whether 149
there are measurable effects of neutron irradiation on the fatigue lives of these materials and, if so, to better define how those impacts may be quantified. In the absence of such data, the methods described in this report are considered appropriate for application to materials exposed to significant levels of irradiation, such as SS reactor internals components, when mandated by regulation or required by the current licensing basis.
Id. at 11 (NYS00490A). Thus, there is no basis, at this time, to conclude that an additional correction factor is necessary to account for the effects of irradiation embrittlement on fatigue life.
It also is important to emphasize again that fatigue analyses are not the only methods used to manage the effects of irradiation or fatigue on RVIs. The risk-prioritized inspections in the RVI AMP provide further assurance that the effects of aging will be adequately managed for RVI components throughout the PEO. See Entergys NYS-25 Testimony at Section VII.A (ENT000616).
Q217. Dr. Hopenfeld disputes the conclusion, in Draft NUREG/CR-6909, Rev. 1 (NYS000490A-B), that no additional correction factor is necessary to address the effects of irradiation on fatigue life. He labels the conclusion entirely arbitrary, non-scientific, and inconsistent with the ASME Code, and claims that irradiation can observed factor of 5 in the crack growth rate is the appropriate bounding multiplier. Supplemental Hopenfeld Report at 14-16 (RIV000144). How do you respond to these claims?
A217. (RGL, JRS, ABC) We disagree. For all of the reasons previously stated in response to Questions 76, 215, and 216, there is no basis, at this time, to conclude that an additional correction factor is necessary to account for the effects of irradiation embrittlement on fatigue life. As for Dr. Hopenfelds claim that a factor of 5 multiplier is appropriate, he appears to have misread his reference document. His testimony states, Tests in the U.S. and Japan show that irradiation can increase crack growth rates by more than a factor of 5 in low oxygen boiling 150
water reactor (BWR) environments. Supplemental Hopenfeld Report at 15 (RIV000144). But page 2 of Exhibit RIV000153, the ANL web page Dr. Hopenfeld relies upon, states, In low-DO BWR environments, the CGRs of the irradiated steels decreased by an order of magnitude.
Emphasis added. His claim is irrelevant for the additional reason that the EAF analysis evaluates when there is the potential for crack initiation. It does not provide any estimate of a crack growth rate, which is Dr. Hopenfelds stated concern.
Q218. Dr. Hopenfeld argues that [t]he synergistic effect of metal fatigue and irradiation embrittlement are of particular concern during loss of coolant accident (LOCA) transients where very small cracks can propagate to failure under the high intensity LOCA loads. Supplemental Hopenfeld Report at 18 n. 56 (RIV000144). Do you agree with these statements?
A218. (RGL, JRS, NFA) As explained in greater detail in our testimony on contention NYS-25, the conditions addressed in MRP-227-A include significant transients, design basis accidents, and seismic loads; the very purpose of the program is to provide reasonable assurance that components will continue to perform their intended functions, consistent with the CLB including the consideration of accident loadsthrough the end of the PEO. See Entergys NYS-25 Testimony at Q180 (ENT000616). Additionally, as explained in our response to Question 64, Dr. Hopenfelds hypothetical very small cracks are not of engineering significance, according to the ASME Code.
Q219. Dr. Hopenfeld also claims that Entergy employs circular logic to assert that synergistic effects can be ignored. Supplemental Hopenfeld Report at 18 (RIV000144). He claims that the EAF calculations found CUFen < 1.0 without accounting 151
for radiation, and then used the CUFen < 1.0 result to claim that radiation need not be considered. Id. Is this an accurate assertion?
A219. (RGL, JRS) There is no circular logic. Entergy is not relying on any EAF evaluation to claim that synergistic effects from irradiation do not exist. On the contrary, as we have shown in this section, fatigue and irradiation embrittlement contribute to potential aging effects in very different ways, and there is no basis to apply an additional fatigue correction factor to address potential embrittlement. See MRP-175 at D-3 (ENT000631).
In addition, the RVI AMP is a risk-prioritized inspection program which inspects high-susceptibility RVI components for cracking and other aging effects, regardless of the underlying aging mechanisms. The RVI and FMP together provide reasonable assurance that the effects of aging on RVIs will be adequately managed throughout the PEO.
- 3. Other Criticisms of the Limiting Locations Review Q220. In his 2015 testimony, Dr. Lahey suggests that an article (NYS000513) shows the iterative process used by Westinghouse in which safety margin is removed in its
[EAF] calculations in an effort to reduce the output or result below CUFen = 1.0. Revised Lahey Testimony at 67-68 (NYS000530). Is this an accurate characterization?
A220. (MAG) No. The paper, of which I was one of the co-authors, describes an initial process to compare plant components on a consistent basis for the purpose of determining the leading locations with respect to EAF. C. Kupper and M. Gray, PVP2014-29093, License Renewal Environmental Fatigue Screening Application at 1 (July 2014) (NYS000513). It supports the guidance in NUREG-1801, Rev. 2 to investigate if there can be other leading EAF locations in addition to the NUREG/CR-6260 locations. The general process presented in the paper is consistent with that followed in CN-PAFM-12-35 (NYS000510). The paper, however, does not describe any process for removal of safety margin. There is a brief summary near the 152
end of the paper of possible ways to reduce conservatism in a subsequent phase of screening, i.e.,
similar to what was done in CN-PAFM-13-32, but no details of that process are described in exhibit NYS000530. Regardless, as we have previously shown, reduction of conservatism in a fatigue analysis does not constitute removal of required safety margin.
Q221. Dr. Hopenfeld claims that neither the screening nor refined limiting locations analyses gave any indication that thermal striping was considered in the identification of the most limiting locations. Supplemental Hopenfeld Report at 22 (RIV000144). Is Dr.
Hopenfeld correct?
A221. (MAG, NFA) We have explained the distinctions between thermal striping and thermal stratification in response to earlier questions in Section V.D.5, supra.
153
Q222.
A222. (MAG, NFA) 154
Finally, as far as the influence of potential support plate crud buildup on the CLB loads, the Steam Generator Integrity Program addresses this possibility through inspections of the steam generators. See LRA at B-118 (ENT00015B).
Q223.
A223. (NFA, MAG) No. Again, as we explained in response to Dr. Laheys claims, a projected CUF of less than 1.0 indicates that no fatigue cracking is expected at the end of 60 years of operation, assuming that all of the transients assumed in the EAF analysis have taken place. In addition, we must note that the CUFen values calculated in Westinghouse Calculation Note CN-PAFM-13-32 are not minimum values. They represent maximum values that retain design margin and are still based on conservative assumptions, even though some of the many 155
conservatisms of earlier evaluations may have been reduced through more refined engineering techniques.
Q224. Dr. Hopenfeld also suggests that the Westinghouse CUFen values are non-conservative because, [w]hen the original calculations of the CUFs of record were made, it was unknown that the maximum reduction in fatigue life would occurs [sic] at low strain rates. Supplemental Hopenfeld Report at 23 (RIV000144). Is this a valid critique?
A224. (MAG) No. Dr. Hopenfeld presents no support for his claim, and it is incorrect.
Thus, Dr. Hopenfelds generalized statement does not identify any error in the EAF calculations.
Q225. Dr. Hopenfeld asserts that Entergy should have revised the CLB CUF values because, [a]fter 40 years of exposure to a hostile LWR environment most of the components have undergone a change in geometry and surface structure due to erosion/corrosion, stress corrosion, swelling, pitting, and cavitation, and because [t]he ASME fatigue curves are based on average stresses only . . . [whereas] sharp surface 156
discontinuities introduce high local stress concentrations. Supplemental Hopenfeld Report at 19-20 (RIV000144). How do you respond?
A225. (NFA, MAG) First, the NUREG reports that provide guidance on EAF evaluations account for surface finish effects. See, e.g., NUREG/CR-6909 at 76 (NYS000357).
Thus, Dr. Hopenfelds recommendation to multiply the IP2 and IP3 CUFs of record by a factor of 10 . . . to account [for] surface roughness, Supplemental Hopenfeld Report at 21 (RIV000144), is entirely unsupported.
Dr. Hopenfeld, moreover, provides no supporting evidence for his speculative assertion that IP2 and IP3 operating conditions cause surface discontinuities in primary plant components subject to EAF evaluations. In fact, visual examinations of IP2 and IP3 prove otherwise. These visual examinations, performed as part of the ASME Section XI ISI program, confirm that the internal surfaces of primary components are not susceptible to increased surface roughness during plant operations. See LRA at B-64 (ENT00015B) (showing that ISI program visual examinations look for cracks and symptoms of wear, corrosion, physical damage, evidence of leakage, and general mechanical and structural condition); see also Entergy, Visual Examination of Reactor Vessel and Internals (VT-3), Summary No. 206913-RVVINT for IP2 (Mar. 10, 2014), and Summary No. 1-1200-VESS. INT. for IP3 (Mar. 13, 2015) (ENT000697).
F. The Balance of Entergys FMP Is Robust and Provides Reasonable Assurance that the Effects of Fatigue Will Be Adequately Managed Q226. To the extent the Intervenors argue that the overall FMP is insufficient to satisfy NUREG-1801, Revision 1, see Hopenfeld Report at 21-25 (RIV000035), how do you respond?
A226. (NFA, ABC, JRS) As explained above, the IPEC FMP fully complies with NRC regulations and NUREG-1801, Revision 1 recommendations, and provides the level of detail 157
necessary for an AMP. See SER at 3-78 to 3-81 (NYS00326B) (finding the IPEC FMP to be consistent with NUREG-1801, Revision 1 AMP). Entergy has committed to manage the effects of fatigue throughout the PEO by monitoring cycles incurred and ensuring they do not exceed the analyzed numbers of cycles, such that the CUFen analyses remain valid. See SER at 4-44 to -
45 (NYS00326); NL-08-084, Attach. 1 at 3-4 (ENT000194). Under this AMP, Entergy will track the numbers of actual plant transients and evaluate those numbers against the numbers of such transients assumed in the fatigue analyses. See NL-08-084, Attach. 1 at 4 (ENT000194);
SER at 4-45 (NYS00326E). The plant transient counts will be updated periodically to ensure that the analyzed number of transients remains valid and to ensure that appropriate corrective actions are implemented prior to reaching the CUFen limit of 1.0. See NL-08-084, Attach. 1 at 4 (ENT000194); SER at 3-79, 4-44 (NYS00326B, NYS00326E). The Intervenors criticisms of the FMP focus almost exclusively on alleged deficiencies in the EAF analysis prepared in support of it. We have shown that those criticisms lack merit.
Q227. With respect to the consequences of fatigue failure, Dr. Lahey also observes that fatigue can result in pipe ruptures, physical failures, and the relocation of loose pieces of metal throughout the reactor system, which, in turn, may result in core blockages and interfere with the safe operation of a nuclear power plant. Lahey Testimony at 37 (NYSR00344). Do you agree with this assessment?
A227. (JRS, RGL) No. To the extent Dr. Laheys assertions regarding the consequences of fatigue failure refer to reactor coolant pressure boundary components, his basis is unclear. Structures, component, or fittings for which the CUFen is maintained less than or equal to 1.0 (including consideration of environmental effects) are not expected to experience cracking, so his assertion also is premised on the validity of his criticisms of the Westinghouse 158
EAF evaluations; such criticisms have already been addressed. To the extent Dr. Laheys assertions are referring to RVI components, they are addressed in Entergys testimony on NYS-25.
Q228. On page 17 of his Report, Dr. Hopenfeld provides a critique of the original design CUF calculations for the reactor vessel inlet and outlet nozzles at IPEC, prepared over 45 years ago. Hopenfeld Report (RIV000035). See C.R. Crockrell and J. C. Lowry, Combustion Engineering, Inc., C.E. CENC-1110, Analytical Report for Indian Point Reactor Vessel Unit No. 2, (Apr. 22, 1968) (CENC-1110) (RIV000052A-D); C.R.
Crockrell and J. C. Lowry, Combustion Engineering, Inc., CENC-1122, Analytical Report for Indian Point Reactor Vessel Unit No. 3 (June 1969) (CENC-1122) (RIV000053A-O).
Are these calculations part of the IP2 and IP3 CLBs?
A228. (NFA, ABC) Yes. The design basis CUF calculations in CENC-1110 and CENC-1122 for reactor vessel inlet and outlet nozzles are part of the CLB of IP2 and IP3 and Westinghouse did not perform a detailed EAF evaluation for these components. See Calculation Note RCDA-03-75, Indian Point Unit 3 Stretch Power Uprate Reactor Vessel Structural Evaluation at 23 (May 2005) (ENT000228); Calculation Note RCDA-03-64, Indian Point Unit 2 Stretch Power Uprate Reactor Vessel Evaluation at 15 (Feb. 2005) (ENT000229) (including these original calculations as part of the post-SPU fatigue evaluation). In the LRA, the CLB CUFs of record for the reactor vessel inlet and outlet nozzles, when corrected for environmental effects and projected through the PEO, continued to show CUFen values that did not exceed 1.0.
See LRA at 4.3-24 to -25 (Tables 4.3-13 and 4.3-14) (ENT00015B). Thus, there has been no need for further refined analysis of these components. We therefore disagree with Riverkeeper when it stated, in a discussion of CENC-1110 and CENC-1122, that Dr. Hopenfeld provided 159
testimony about the deficiencies with Entergys refined fatigue analyses. Riverkeeper, Inc.
Opposition to Entergys Motion in Limine to Exclude Portions of Pre-filed Testimony, Expert Report, Exhibits, and Statement of Position for Contention NYS-26B/RK-TC-1B (Metal Fatigue) at 11 (Feb. 17, 2012).
Thus, any question of the adequacy of CENC-1110 and CENC-1122 is not relevant to any future CUFen calculations that Entergy might perform under its FMP, but is only relevant to the adequacy of the plants design basis and CLB.
Q229. Although part of the CLB, how do you respond to Dr. Hopenfelds critique of these calculations?
A229. (NFA, JRS) Dr. Hopenfeld presents no valid critique of these design calculations.
We recognize that these components were evaluated 40 years ago using simplified 2-D models.
But as we have previously explained in response to Question 134, simplified models require the use of conservative inputs and assumptions, which result in significant conservatism. Dr.
Hopenfelds observation that these calculations did not use a finite element analysis and were based on a simplified 2-D model (Hopenfeld Report at 17 (RIV000035), Supplemental Hopenfeld Report at 21-22 (RIV000144)), does not reveal any deficiency in the calculations. As to the potential variability in heat transfer coefficients, see id., Dr. Hopenfeld does not explain why the conservative values used in these analyses do not account for the variability he assumes.
See CENC-1110, App. B (RIV000052B-C); CENC-1122, App. B (RIV000053L-O).
160
Q230. Dr. Lahey suggests that Entergy is required to preemptively declare the CUFen value that would trigger component replacement. Lahey Rebuttal at 13 (NYS000440). How do you respond?
A230. (ABC, JRS, NFA, MAG) Entergy has, in fact, established the value of CUFen at which they would repair or replace components, if necessary. Pursuant to Commitment 33, if Entergy does not demonstrate valid projected CUFen values below 1.0 via refined CUFen analyses (Option 1), then Entergy must repair or replace the affected locations before exceeding a CUF of 1.0.
Q231. Relatedly, the Intervenors 2011 Position Statement (but not in their Revised Position Statement (NYS000529)) is that Entergys FMP does not include specific criteria for determining when corrective actions should be taken. Position Statement at 32 (NYSR00343). Please describe the corrective action components of the FMP.
A231. (NFA, ABC) There is no ambiguity or uncertainty about the timing or scope of corrective actions, including repair and replacement activities under the FMP. The program requires that corrective action be implemented before the plant exceeds the analyzed number of transient cycles. See NL-08-084, Attach. 1 at 4 (ENT000194); SER at 4-44, 4-45 (NYS00326E).
IPEC procedures contain specific alert levels that trigger the initiation of corrective actions under the Fatigue Monitoring Program. SER at 4-44 (NYS00326E). Any necessary future analysis updates would be governed by Entergys quality assurance (QA) program. See NL-08-084, Attach. 1 at 4 (ENT000194).
Repair or replacement of a component, if necessary, also would be accomplished in accordance with established plant procedures that are governed by Entergys QA program, as credited in the SER. See SER at 3-216 (NYS00326C). As required by 10 C.F.R. § 50.55a, 161
repair and replacement will be accomplished in accordance with the applicable requirements of ASME Code Section XI, Inservice Inspection of Nuclear Power Plant Components. See NL-08-084, Attach. 1 at 4 (ENT000194); SER at 3-173 to -189 (NYS000326C) (discussing ASME Code Section XI ISI and repair and replacement requirements and their implementation at IPEC during the PEO). Intervenors allegations of vagueness thus are entirely unfounded.
Q232.
A232. (NFA, ABC)
The main feedwater nozzle provides water to the secondary side of the steam generator, so, as we have explained in response to Question 124, above, they are not subject to EAF.
162
Q233. Dr. Hopenfeld also raises a concern regarding that [f]atigue may also create small cracks that propagate and cause a given component to malfunction and/or break up and form loose parts which would interfere with the safe operation of the plant.
Hopenfeld Report at 3 (RIV000035). As an example he posits if one of the feed water distribution nozzles (J tubes) were to fail from fatigue, pieces from the broken nozzle could be lodged between steam generator tubes, causing the tubes to rupture and leading to a potential core melt. Id. How do you respond to this point?
A233. (ABC, NFA) As we explained in response to Dr. Laheys claims, the adjustment of the CUF to account for the effects of the reactor coolant environment is not applicable to secondary side components that are not exposed to the reactor coolant environment. This applies to Dr. Hopenfelds concerns regarding cracking and loose parts on the secondary side of the steam generator. Aging effects applicable to those componentsincluding those creating the potential for loose parts in the steam generatorare managed under the Water Chemistry Control - Primary and Secondary Program and the Steam Generator Integrity Program. See LRA Tbls. 3.1.2-4-IP2, 3.1.2-4-IP3 (ENT00015A); id. App. B at B-118, B-137 (ENT00015B).
In particular, the Steam Generator Integrity Program includes processes for monitoring and maintaining secondary side components, through visual inspections of feedwater rings, performed by qualified personnel using approved non-destructive examination processes and procedures. See id. App. B at B-118 (ENT00015B). The adequacy of these programs is unchallenged in this contention. Intervenors do appear to raise issues with this program in a 163
different contention, NYS-38/RK-TC-5. However, those issues are unrelated to fatigue or feedwater rings. Moreover, historic issues with failures of feedwater distribution components in steam generators are due to erosionnot fatigue, as Dr. Hopenfeld suggestsand have been addressed in response to generic industry communications. See NRC Information Notice 91-019, Steam Generator Feed Water Distribution Piping Damage (Mar. 12, 1991) (RIV000008);
Morning Report 5-93-0042, Steam Generator Feedring Nozzle Through Wall Erosion (June 16, 1993) (RIV000009).
VI. CONCLUSIONS Q234. Please summarize your testimony and the bases for your conclusion that NYS-26B/RK-TC-1B lacks factual and technical merit.
A234. (NFA, ABC, JRS, MAG, RGL, BMG) NYS-26B/RK-TC-1B lacks merit for the following principal reasons.
The LRA addresses in detail the aging effects due to metal fatigue on RCS components. Entergys FMP is consistent with NUREG-1801, Revision 1, and therefore, in conjunction with the further evaluations under Commitments 33, 43 and 49, provides reasonable assurance that the effects of aging will be adequately managed throughout the PEO, consistent with 10 C.F.R. §§ 54.21(a)(3), (c)(1)(iii), and 54.29(a).
Consistent with accepted NRC guidance and industry methods of analysis, Entergy has demonstrated that the effects of fatigue, including the effects of the reactor water environment, will be adequately managed such that affected components will remain capable of performing their intended function throughout the PEO, consistent with 10 C.F.R. §§ 54.21(a)(3),
(c)(1)(iii), and 54.29(a).
Under NRC regulations and guidance, Entergy is permitted to refine its EAF analyses to demonstrate that the CUFen values for the limiting plant locations are less than or equal to 1.0. In accordance with NRC regulations and established engineering procedures, Entergy appropriately refined its EAF analyses.
Entergy and Westinghouse have fully documented and detailed the conservative methods used to determine refined CUFen values. The various criticisms leveled by Intervenors against these analyses are 164
unfounded and lack technical merit. In particular, the analyses used appropriate heat transfer coefficients, environmental correction factors, and DO values, and conservatively estimated the number of transients for each analyzed component. These analyses are conservative analyses, so no propagation of error evaluation is appropriate or necessary, as might be the case with a best-estimate analysis.
Entergy analyzed the effects of EAF for the NUREG/CR-6260 locations at IPEC, consistent with the guidance in NUREG-1801, Revision 1 and NUREG/CR-6583 and 5704. Entergy additionally reviewed its design basis ASME Code fatigue evaluations to determine whether the NUREG/CR-6260 locations that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting locations for IPEC. This limiting locations review included consideration of limiting reactor coolant pressure boundary and RVI locations. This review further supports the adequacy of the Entergy FMP, by providing additional assurance that the CLB will be maintained throughout the PEO.
The evaluation of limiting RVI components appropriately incorporated the environmental correction factors specified in NRC guidance. There is no technical basis to require any additional correction factor to account for the effects of irradiation embrittlement on fatigue life. Moreover, in addition to the FMP, under the IPEC RVI AMP, Entergy will conduct risk-prioritized inspections of irradiated components for cracking due to fatigue or other mechanisms. Together, the RVI AMP and the FMP provide reasonable assurance that the effects of aging will be adequately managed for RVI components throughout the PEO.
Q235. Does this conclude your testimony?
A235. (NFA, ABC, JRS, MAG) Yes.
Q236. In accordance with 28 U.S.C. § 1746, do you state under penalty of perjury that the foregoing testimony is true and correct?
A236. (NFA, ABC, JRS, MAG, RGL, BMG) Yes.
Executed in accord with 10 C.F.R. § 2.304(d)
Nelson F. Azevedo Supervisor of Code Programs Entergy Nuclear Generation Co.
295 Broadway, Suite 1 Buchanan, NY 10511 914-734-6775 nazeved@entergy.com 165
Executed in accord with 10 C.F.R. § 2.304(d)
Alan B. Cox Independent Consultant Entergy License Renewal Services 1448 SR 333 N-GSB-45 Russellville, AR 72802 479-858-3173 acox@entergy.com Executed in accord with 10 C.F.R. § 2.304(d)
Jack. R. Strosnider Senior Nuclear Safety Consultant Talisman International, LLC 9712 Breckenridge Pl.
Montgomery Village, MD 20886 202-471-4244 jstrosnider@talisman-intl.com Executed in accord with 10 C.F.R. § 2.304(d)
Mark A. Gray Principal Engineer Westinghouse Electric Company LLC Nuclear Services 1000 Westinghouse Drive Cranberry Township, PA 16066 412-374-4602 GrayMA@westinghouse.com Executed in accord with 10 C.F.R. § 2.304(d)
Randy G. Lott Consulting Engineer Westinghouse Electric Company LLC Nuclear Services 1000 Westinghouse Drive Cranberry Township, PA 16066 (412) 374-4157 LottRG@westinghouse.com Executed in accord with 10 C.F.R. § 2.304(d)
Barry M. Gordon Associate Structural Integrity Associates, Inc.
5215 Hellyer Ave., Suite 210 San Jose, CA 95138 166
(408) 978-8200 BGordon@structint.com August 10, 2015 DB1/ 84089997 167