ULNRC-06050, Response to RAI Set 27 (RAI B2.1.6-4c) and Amendment 27 to the Callaway LRA

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Response to RAI Set #27 (RAI B2.1.6-4c) and Amendment 27 to the Callaway LRA
ML13295A108
Person / Time
Site:  Ameren icon.png
Issue date: 10/17/2013
From: Kremer G
Ameren Missouri, Union Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME7708, ULNRC-06050
Download: ML13295A108 (19)
v • d • e


Text

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WAmeren MISSOURI Callaway Plant October 17, 2013 ULNRC-06050 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 10 CFR 2.101 10 CFR 2.109(b) 10 CFR 50.4 10 CFR 50.30 10 CFR 51.53(c) 10 CFR 54 Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.

FACILITY OPERATING LICENSE NPF -30 RESPONSE TO RAI SET #27 (RAI B2.1.6-4c) AND AMENDMENT 27 TO THE CALLAWAY LRA

References:

1) ULNRC-05830 dated December 15, 2011
2) NRC Letter, "Request for Additional Information for the Review of the Callaway Plant, Unit 1 License Renewal Application, Set 27 (TAC No. ME7708)," dated September 20, 2013
3) NRC Letter, "Safety Evaluation Report with Open Items Related to the License Renewal of Callaway Plant, Unit 1 (TAC NO. ME7708)," dated April23, 2013 By the Reference 1 letter, Union Electric Company (Ameren Missouri) submitted a license renewal application (LRA) for Callaway Plant Unit 1. Reference 2 dated September 20, 2013 transmitted the twenty-seventh Request for Additional Information (RAI) related to our application.

Enclosure 1 contains Ameren Missouri's response to request RAI B2.1.6-4c contained in the September 20, 2013 letter. Enclosure 2 contains LRA Amendment 27 to reflect changes made as a result of the RAI response. This response and LRA amendment provide information intended to assist the staff in partially closing Open Item B2.1.6-l identified in Reference 3.

~ ~~~~~ ~~~ ~~~~ ~~ ~~~~~ ~~ ~~ ~~~ ~~ ~~ ~~~~~~~ ~~~~ ~~ ~~~~~~~~~~~ ~~~~~~~~ ~~ ~~ ~~~~~ ~~ ~~~~ ~~ ~~~~ ~~~~~~ ~~~~ ~~~~~ ~~~ ~~~~~ ~~~~~ ~~ ~~~ ~~ ~~ PO Box 620 Fulton, MD 65251 AmerenMissouri.com

ULNRC-06050 October 17, 2013 Page 2 It should be noted that two (2) modified commitments are included in Amendment 27, Table A4-1, Item 4 and Item 31.

With the US Government shutdown that began on October 1, 2013, the suspension of normal NRC activities on October 10, 2013, and the resultant cancellation ofthe October 16, 2013 NRC Meeting on the Resolution of Plant-Specific Action Items Related to MRP-227-A, "Reactor Internals Aging Management Programs/Inspection Plans"; the response to RAI B2.1.6- 4b will not be submitted within 60 days as requested in Reference 2. We will communicate with NRC LRA project management to establish a revised submittal schedule.

If you have any questions with regard to this RAI response, or Amendment 27, please contact me at (314) 954-4319 or Ms. Sarah Kovaleski at (573) 489-9435.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on: OC'-ibhr /1 Zf)/ :J Director Engineering Programs DS/SGK/nls

Enclosures:

1) Request for Additional Information (RAI) Set #27 Response RAI B2.1.6-4c
2) Amendment 27, LRA Changes

ULNRC-06050 October 17, 2013 Page 3 cc: U.S. Nuclear Regulatory Commission (Original)

Attn: Document Control Desk Washington, DC 20555-0001 Steven Reynolds Acting Regional Administrator U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Samuel Cuadrado De Jesus Project Branch 1 Division of License Renewal Office ofNuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-11F1 Washington, DC 20555 Mr. Fred Lyon Project Manager, Callaway Plant Office ofNuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8B 1 Washington, DC 20555-2738 Mr. Gregory A. Pick U.S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511

ULNRC-06050 October 17, 2013 Page4 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 4150 International Plaza Suite 820 Fort Worth, TX 76109 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via Tech Spec ULNRC Distribution:

A. C. Heflin F. M. Diya C. 0. Reasoner III D. W. Neterer B. L. Cox L. H. Graessle J. S. Geyer S. A. Maglio B. C. Daniels NSRB Secretary M. A. McLachlan G. S. Kremer S. G. Kovaleski T. B. Elwood G. G. Yates E. A. Blocher (STARS PAM COB)

Mr. Mike Westman (WCNOC)

Mr. Tim Hope (Luminant Power)

Mr. Ron Barnes (APS)

Mr. Tom Baldwin (PG&E)

Mr. Mike Murray (STPNOC)

Mr. Mark Morgan (SCE)

Mr. John O'Neill (Pillsbury Winthrop Shaw Pittman LLP)

Missouri Public Service Commission Ms. Leanne Tippett Mosby (DNR)

ULNRC-06050 October 17, 2013 Page 1 of 3 CALLAWAY PLANT UNIT 1 LICENSE RENEWAL APPLICATION REQUEST FOR ADDITIONAL INFORMATION (RAI) SET #27 RESPONSE RAI B2.1.6-4c

ULNRC-06050 October 17, 2013 Enclosure 1 Page 2 of 3 RAI B2.1.6-4c

Background:

By letter dated January 24, 2013 (Ameren Letter No. ULNRC-05950, dated January 24, 2013, as docketed in ADAMS ML13029A243 and ML13029A244), the applicant responded to RAI B2.1.6-4a. In that letter and response, the applicant used an LRA commitment (as given in Commitment No. 20 in license renewal application (LRA) UFSAR Supplement Table A4-1) as the basis for resolving the request in A/LAI No.8, Subitem 5 on the MRP-227-A report methodology.

LRA Table 4.3-5 provides those RVI components in the plant design that were analyzed in accordance with an applicable design basis fatigue analysis (i.e., cumulative usage factor [CUF]

analysis). In LRA Section 4.3.3 of the LRA, the applicant states that the CUF values for these RVI components are time-limited aging analysis (TLAAs) and that these TLAAs are acceptable in accordance with the criterion in 10 CFR 54.21 (c)(1 )(iii). The applicant currently credits its Fatigue Monitoring Program as the basis for managing fatigue induced cracking in the components during the period of extended operation.

Issue:

In the Ameren letter of April26, 2013 (Ameren Letter No. ULNRC-05979 as docketed in ADAMS ML13119A133, ML13119A136, and ML13119A137), the applicant amended the LRA to resolve the request in A/LAI No. 8, Subitem 5 by superseding the existing Commitment No.4 with the following amended commitment basis:

ApplicanULicensee Action Item (A/LAI) #8 Item #5 (in part- reactor coolant system water environment portion)

Enhance the Fatigue Monitoring program to evaluate the effects of the reactor coolant system water environment on the reactor vessel internal components with existing fatigue CUF analyses to satisfy the evaluation requirements of ASME Code, Section Ill, Subsection NG-2160 and NG-3121, Upon further review, the staff has determined that the neither UFSAR Supplement for the Fatigue Monitoring Program, as given in LRA UFSAR Supplement Section A2.1, "Fatigue Monitoring," nor Commitment No. 31 on the Fatigue Monitoring Program in UFSAR Table A4-1 have been amended to include the stated enhancement. Therefore, the staff seeks additional justifications on why UFSAR Supplement A2.1 and Commitment 31 in UFSAR Supplement Table A4-1 have not been enhanced consistent with the commitment change in UFSAR Supplement Table A4-1, Commitment No.4. The staff also needs clarification on how the program elements of the Fatigue Monitoring Program will be adjusted to evaluate (account for) the effects of the reactor coolant environment on the acceptability of the CUF analyses for the applicable RVI components.

ULNRC-06050 October 17, 2013 Page 3 of 3 Request:

Provide your basis why Commitment No. 31 in UFSAR Table A4-1 and UFSAR Section A2.1, "Fatigue Monitoring," has also not been amended to include this enhancement. Clarify and justify how the program elements of the Fatigue Monitoring Program will be adjusted to evaluate the effects of the reactor environment on the CUF analyses for the applicable RVI components.

Callaway Response LRA Table A4-1 item 4 has been revised to delete the A/LAI #8 Item #5 commitment on effects of the reactor water environment on the reactor vessel internals locations with fatigue usage calculations and incorporate implementation of the commitment into LRA Table A4-1 item 31.

Implementation of the commitment is described in LRA Appendix A2.1, LRA Section A3.2.3, and LRA Section 83.1. Callaway will recalculate each of the reactor vessel internals CUFs identified in LRA Table 4.3-5, Reactor Internals Design Basis Fatigue Analysis Results, to consider the reactor water environment effects (F.n) using NUREG/CR-5704 or NUREG/CR-6909. Consistent with the corrective actions specified in the Fatigue Monitoring program (83.1 ), corrective actions include fatigue re-analysis, repair, or replacement of the affected components prior to the U."

reaching 1.0.

LRA Table A4-1 item 4, LRA Table A4-1 item 31, LRA Section A2.1, LRA Section A3.2.3, and LRA Section 83.1 have been revised as shown in LRA Amendment 27 in Enclosure 2 to consider the effects of the reactor water environment on the reactor vessel internals locations with fatigue usage calculations.

Corresponding Amendment Changes Refer to the Enclosure 2 Summary Table "Amendment 27, LRA Changes" for a description of LRA changes with this response.

ULNRC-06050 October 17, 2013 Page 1 of 12 Amendment 27, LRA Changes Enclosure 2 Summary Table Affected LRA Section LRA Paae(s\

Section A2.1 A-21 Section A3.2.3 A-27 Table A4-1 item 4 A-37 Table A4-1 item 31 A-46, A-47 and A-48 Section 83.1 B-127, B-128, B-129, B-130, and B-131

ULNRC-06050 October 17, 2013 Enclosure 2 Page 2 of 12 Appendix A Final Safety Analysis Report Supplement A2.1 FATIGUE MONITORING The Fatigue Monitoring program manages fatigue cracking caused by anticipated cyclic strains in metal components of the reactor coolant pressure boundary. The program ensures that actual plant experience remains bounded by the transients analyzed in the design calculations and fatigue crack growth analyses, or that corrective actions maintain the design and licensing basis. The Fatigue Monitoring program tracks the number of transient cycles and will track cumulative fatigue usage at monitored locations. The Fatigue Monitoring program tracks fatigue by one of the following methods:

1) The Cycle Counting (CC) monitoring method tracks transient event cycles affecting the location to ensure that the numbers of transient events analyzed by the fatigue analyses are not exceeded. This method does not calculate cumulative usage factors (CUFs).
2) The Cycle-Based Fatigue (CBF) monitoring method utilizes the CC results and stress intensity ranges generated with the ASME Ill methods that use three dimensional six component stress-tensor methods to perform CUF calculations for a given location. The fatigue accumulation is tracked to determine approach to the ASME allowable fatigue limit of 1.0.
3) The Stress-Based Fatigue (SBF) monitoring method computes a "real time" stress history for a given component from data collected from plant instruments to calculate transient pressure and temperature, and the corresponding stress history at the critical location in the component. The stress history is analyzed to identify stress cycles, and then a CUF is computed. The CUF will be calculated using a three dimensional, six component stress tensor method meeting ASME Ill NB-3200 requirements, or a method will be benchmarked consistent with the NRC Regulatory Issue Summary RIS 2008-30.

The program will also consider the effects of the reactor water environment for a set that includes the NUREG/CR-6260 sample locations for a newer-vintage Westinghouse Plant, aA1 plant-specific bounding EAF locations in the reactor coolant pressure boundary, and reactor vessel internals locations with fatigue usage calculations. F!ill factors will be determined as described in Section A3.2.3.

If a cycle count or cumulative usage factor value increases to a program action limit, corrective actions include fatigue reanalysis, repair, or replacement. Action limits permit completion of corrective actions before the design limit is exceeded. The sentinel location analysis, when refined , will be revisited to confirm bounding Reactor Coolant Pressure Boundary Environmentally Assisted Fatigue susceptible sentinel locations are updated appropriately and remain consistent with the refined analysis.

Callaway Plant Unit 1 Page A-21 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Page 3 of 12 A3.2.3 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)

All of the locations specified in NUREG/CR-6260 for newer vintage Westinghouse plants will be monitored by the Fatigue Monitoring program, described in Section A2 .1. If any of the analyzed CUF values for these locations exceeds the fatigue design limit, the analyses may be revised using actual plant transients experienced. Callaway has completed an evaluation to identify any additional plant-specific bounding EAF locations.

The effects of the reactor coolant environment on fatigue usage factors in the NUREG/CR-6260 and plant-specific bounding EAF locations in the reactor coolant pressure boundary will be managed for the period of extended operation. The supporting environmental factors, Fen. calculations will be performed with NUREG/CR-6909 or NUREG/CR-6583 for carbon and low alloy steels, NUREG/CR-6909 or NUREG/CR-5704 for austenitic stainless steels, and NUREG/CR-6909 for nickel alloys.

In addition. reactor vessel internals locations with fatigue usage calculations will be evaluated for the effects of the reactor water environment. These TLAAs are dispositioned in accordance with 10 CFR 54.21(c)(1)(iii).

Callaway Plant Unit 1 Page A-27 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Page 4 of12 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments LRA Implementation Item# Commitment Section Schedule 4 Implement the PWR Vessel Internals program as described in LRA Section 82. 1.6. As part of 82.1.6 PWRVessel the implementation activities address the following Applicant/Licensee Action Items (A/LAI) of Internals program NRC MRP-227-A Safety Evaluation dated December 16, 2011. implementation and A/LAI #1:

Applicant/Licensee Action Item (A/LAI) #1 Within 24 months after the issuance of Each applicant/licensee is responsible for assessing its plant's design and operating MRP-227-A, PWR history and demonstrating that the approved version of MRP-227 is applicable to the Internals Inspection facility. Each applicant/licensee shall refer, in particular, to the assumptions regarding and Evaluation plant design and operating history made in the FMECA and functionality analyses for Guideline reactors of their design (i.e., Westinghouse, CE, or 8&W) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RVI components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MRP-227.

,A,~~IisaAt.lbiseAsee AstieA Item (AibAI) #6 Item #a AtbAI #8 Item a:

(iA ~art reaster seelaAt system water eAvireAmeAt ~ertieA) Gem~leteEI Ae later thaA six meAths ~rier eAhaAse the ~atiJI:le MeAiteriA1 ~reJram te evall:late the effests ef the reaster seelaAt te the PeO.

system ,.,,~ater eA 11ireAmeAt eA the reaster 1/essel iAterAal sem~eAeAts 'llith existiA1 fatiJI:le GUF aAalyses te satisfy the evall:latieA reql:liremeAts ef ASMe GeEie, SestieA Ill , Sl:lesestieA NG 2HIQ aAEI NG 3~2~ . (moved to item 31 )

Callaway Plant Unit 1 PageA-37 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Page 5 of 12 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments LRA Implementation Item# Commitment Section Schedule 31 Enhance the Fatigue Monitoring program procedures to: 4.3.2.1 Completed no later

  • include fatigue usage calculations that consider the effects of the reactor water 4.3.2.2 than six months prior environment for a set of sample reactor coolant systeFR pressure boundarv locations 4.3.4 to the PEO and reactor vessel internals locations with fatigue usage calculations. The reactor 83.1 coolant pressure boundary set includes the NUREG/CR-6260 sample locations for a newer-vintage Westinghouse Plant and plant-specific bounding EAF locations.
  • ensure the scope includes the fatigue crack growth analyses, which support the leak-before-break analyses, ASME Section XI evaluations, and the HELB break selection criterion remain valid by counting the transients used in the analyses.
  • require the review of the temperature and pressure transient data from the operator logs and plant instrumentation to ensure actual transient severity is bounded by the design and to include environmental effects where applicable. If a transient occurs which exceeds the design transient definition the event is documented in the Corrective Action Program and corrective actions are taken.
  • include additional transients that contribute significantly to fatigue usage. These additional transients were identified by evaluation of ASME Section Ill fatigue and fatigue crack growth analyses.
  • include additional locations which receive more detailed monitoring. These locations were identified by evaluation of ASME Section Ill fatigue analyses and the locations evaluated for effects of the reactor coolant environment. In addition. reactor vessel internals locations with fatigue usage calculations will be evaluated for the effects of the reactor water environment. The monitoring methods will be benchmarked consistent with the NRC RIS 2008-30.
  • project the transient count and fatigue accumulation of monitored components into the future. - -- --- - -

Callaway Plant Unit 1 PageA-46 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Page 6 of 12 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments LRA Implementation Item# Commitment Section Schedule

  • include additional cycle count and fatigue usage action limits, which permit completion of corrective actions if the design limits are expected to be exceeded within the next 3 fuel cycles. The fatigue results associated with the NUREG/CR-6260 sample locations for a newer vintage Westinghouse plant and plant-specific bounding environmental-assisted fatigue locations will account for environmental effects on fatigue. The cycle count action limits for the hot leg surge nozzle will incorporate the 60-year cycle projections use in the hot leg surge nozzle EAF analysis.
  • include appropriate corrective actions to be invoked if a component approaches a cycle count or CUF action limit or if an experienced transient exceeds the design transient definition. If an action limit is reached, corrective actions include fatigue reanalysis, repair, or replacement. When a cycle counting action limit is reached, I action will be taken to ensure that the analytical bases of the HELB locations are maintained. Re-analysis of a fatigue crack growth analysis must be consistent with or I

reconciled to the originally submitted analysis and receive the same level of regulatory review as the original analysis.

  • limit the number of the most severe RCP component cooling water transient, elevated I CCW inlet temperature transients, to 75 percent of its design value, i.e. limited to 150, in order to accommodate the seasonal temperature change transient in the RCP thermal barrier flange fatigue analysis. I
  • include non-NUREG/CR-6260 locations with an Uen greater than 1.0 for further evaluation using the same methods as those used for NUREG/CR-6260 locations to I remove conservatisms from the preliminary Uen* The results of these final analyses will be incorporated into the Fatigue Monitoring program by either counting the transients assumed or incorporate the stress intensities into a CBF ability of the program. As an I alternative, the Fatigue Monitoring program will implement SBFs of certain locations in order to ensure the component does not exceed an Uen of 1.0. Any use of SBF will be imQ(emented consistent with RIS 2008-30. I Callaway Plant Unit 1 PageA-47 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Page 7 of 12 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments LRA Implementation Item# Commitment Section Schedule

  • the sentinel location analysis, when refined, will be revisited to confirm bounding Reactor Coolant Pressure Boundary Environmentally Assisted Fatigue susceptible sentinel locations are updated appropriately and remain bounded consistent with the refined analysis. I Callaway Plant Unit 1 PageA-48 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Enclosure 2 Page 8 of 12 Appendix B AGING MANAGEMENT PROGRAMS 83.1 FATIGUE MONITORING Program Description The Fatigue Monitoring program manages fatigue cracking caused by anticipated cyclic strains in metal components of the reactor coolant pressure boundary. The program ensures that actual plant experience remains bounded by the thermal and pressure transient numbers and severities analyzed in the design calculations, or that corrective actions maintain the design and licensing basis.

The Fatigue Monitoring program tracks fatigue by one of the following methods:

1) The Cycle Counting (CC) monitoring method tracks transient event cycles affecting the location to ensure that the numbers of transient events analyzed by the fatigue analyses are not exceeded. This method does not calculate cumulative usage factors (CUFs).
2) The Cycle-Based Fatigue (CBF) monitoring method utilizes the CC results and stress intensity ranges generated with the ASME Ill methods that use three dimensional six component stress-tensor methods to perform CUF calculations for a given location. The fatigue accumulation is tracked to determine approach to the ASME allowable fatigue limit of 1.0.
3) The Stress-Based Fatigue (SBF) monitoring method computes a "real time" stress history for a given component from data collected from plant instruments to calculate transient pressure and temperature, and the corresponding stress history at the critical location in the component. The stress history is analyzed to identify stress cycles, and then a CUF is computed. The CUF will be calculated using a three dimensional, six component stress tensor method meeting ASME Ill NB-3200 requirements, or a method will be bench marked consistent with the NRC Regulatory Issue Summary RIS 2008-30.

The Fatigue Monitoring program requires periodic reviews of the plant instrumentation and operator logs to ensure that the fatigue critical thermal and pressure transients have not exceeded design transient severity or analyzed number, and to ensure that usage factors will not exceed the allowable value of 1.0 without corrective actions.

The Fatigue Monitoring program will be enhanced to include the effects of the reactor coolant water environment on component fatigue life for a set of sample reactor coolant system locations. The set includes fatigue monitoring of the NUREG/CR-6260 sample locations for a newer-vintage Westinghouse Plant.. arui plant-specific bounding environmentally assisted fatigue (EAF) locations in the reactor coolant pressure boundary, and reactor vessel internals locations with fatigue usage calculations . The supporting environmental factors, F(en), calculations will be performed with NUREG/CR-6909 or NUREG/CR-6583 for carbon and low alloy steels, NUREG/CR-6909 or NUREG/CR-5704 for austenitic stainless steels, and NUREG/CR-6909 for nickel alloys.

Callaway Plant Unit 1 Page B-127 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Page 9 of 12 Appendix B AGING MANAGEMENT PROGRAMS NUREG-1801 Consistency The Fatigue Monitoring program is an existing program that, following enhancement, will be consistent with NUREG-1801,Section X.M1, Fatigue Monitoring.

Exceptions to NUREG-1801 None Enhancements Prior to the period of extended operation, the following enhancements will be implemented in the following program elements:

Scope of the Program - Element 1 Procedures will be enhanced to include fatigue usage calculations that consider the effects of the reactor water environment for a set of sample reactor coolant system pressure boundary locations and reactor vessel internals locations with fatigue usage calculations.

The reactor coolant pressure boundary set includes the NUREG/CR-6260 sample locations for a newer-vintage Westinghouse Plant, and plant-specific bounding EAF locations.

Procedures will be enhanced to ensure the fatigue crack growth analyses, which support the leak-before-break analyses, ASME Section XI evaluations, and the HELB break selection criterion remain valid by counting the transients used in the analyses.

Preventive Actions - Element 2 Procedures will be enhanced to require the review of the temperature and pressure transient data from the operator logs and plant instrumentation to ensure actual transient severity is bounded by the design and to include environmental effects where applicable. If a transient occurs which exceeds the design transient definition the event is documented in the Corrective Action Program and corrective actions are taken .

Parameters Monitored or Inspected- Element 3 Procedures will be enhanced to include additional transients that contribute significantly to fatigue usage. These additional transients were identified by evaluation of ASME Section Ill fatigue and fatigue crack growth analyses.

Callaway Plant Unit 1 Page B-128 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Page 10 of 12 Appendix 8 AGING MANAGEMENT PROGRAMS Procedures will be enhanced to include additional locations which receive more detailed monitoring. These locations were identified by evaluation of ASME Section Ill fatigue analyses and the locations evaluated for effects of the reactor coolant water environment. In addition. reactor vessel internals locations with fatigue usage calculations will be evaluated for the effects of the reactor water environment. The monitoring methods will be benchmarked consistent with the NRC RIS 2008-30.

Procedures will be enhanced to limit the number of the most severe RCP component cooling water transient, elevated CCW inlet temperature transients, to 75 percent of its design value, i.e. limited to 150, in order to accommodate the seasonal temperature change transient in the RCP thermal barrier flange fatigue analysis.

Monitoring and Trending - Element 5 Procedures will be enhanced to project the transient count and fatigue accumulation of monitored components into the future.

Acceptance Criteria - Element 6 Procedures will be enhanced to include additional cycle count and fatigue usage action limits, which permit completion of corrective actions if the design limits are expected to be exceeded within the next three fuel cycles. The fatigue results associated with the NUREG/CR-6260 sample locations for a newer vintage Westinghouse plant and plant-specific bounding EAF locations will account for environmental effects on fatigue. The cycle count action limits for the hot leg surge nozzle will incorporate the 60 year cycle projections used in the hot leg surge nozzle EAF analysis.

Corrective Actions - Element 7 Procedures will be enhanced to include appropriate corrective actions to be invoked if a component approaches a cycle count or CUF action limit or if an experienced transient exceeds the design transient definition. If an action limit is reached, corrective actions include fatigue reanalysis, repair, or replacement. When a cycle counting action limit is reached, action will be taken to ensure that the analytical bases of the HELB locations are maintained. Re-analysis of a fatigue crack growth analysis must be consistent with or reconciled to the originally submitted analysis and receive the same level of regulatory review as the original analysis.

Procedures will be enhanced to include non-NUREG/CR-6260 locations with a Uen greater than 1.0 for further evaluation using the same methods as those used for NUREG/CR-6260 locations to remove conservatisms from the preliminary Uen* The results of these final analyses will be incorporated into the Fatigue Monitoring program by either counting the transients assumed or incorporate the stress intensities into a CBF ability of the program.

Callaway Plant Unit 1 Page B-129 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Page 11 of 12 Appendix B AGING MANAGEMENT PROGRAMS As an alternative, the Fatigue Monitoring program will implement SBFs of certain locations in order to ensure the component does not exceed a Uen of 1.0. Any use of SBF will be implemented consistent with RIS 2008-30.

The sentinel location analysis, when refined, will be revisited to confirm bounding Reactor Coolant Pressure Boundary Environmentally Assisted Fatigue susceptible sentinel locations are updated appropriately and remain bounded consistent with the refined analysis.

Operating Experience The following discussion of operating experience provides objective evidence that the Fatigue Monitoring program will be effective in ensuring that intended functions are maintained consistent with the current licensing basis for the period of extended operation.

1. In response to NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification, Westinghouse performed a plant-specific evaluation of Callaway pressurizer surge line.

It was concluded that thermal stratification does not affect the integrity of the pressurizer surge line. Callaway responses to NRC Bulletin 88-11 describe the inspections, analyses, and procedural revisions made to ensure that thermal stratification does not affect the integrity of the pressurizer surge line. There have been no signs of damage from surge line movement.

2. NRC Regulatory Issue Summary RIS 2008-30, Fatigue Analysis Of Nuclear Power Plant Components informed licensees of analysis methodology (Green's function) used to demonstrate compliance with the ASME Code fatigue acceptance criteria could be non-conservative if not correctly applied. Ameren Missouri is committed to using a three dimensional, six component stress tensor method meeting ASME Ill NB-3200 requirements, or benchmarking the chosen method. This benchmarking has been performed for the normal and alternate charging nozzle in order to implement SBF at that location. Any additional locations which will be monitored with SBF must meet the ASME Ill NB-3200 requirements or be benchmarked.
3. An error was identified in the previous SBF transfer function for the normal and alternate charging nozzles. The SBF transfer function incorrectly included thermal sleeves for the nozzles and therefore would calculate less fatigue than the nozzles would actually accumulate. The extent of this condition is limited only to the normal and alternate charging nozzle SBF models. The transfer function has been updated to exclude thermal sleeves. The SBF transfer functions for the normal and alternate charging nozzles were also benchmarked in accordance with NRC RIS 2008-30.

Callaway Plant Unit 1 Page B-130 License Renewal Application Amendment 27

ULNRC-06050 October 17, 2013 Page 12 of 12 Appendix B AGING MANAGEMENT PROGRAMS

4. The CVCS design specification identifies the nominal letdown flow of 75 gpm with maximum flow of 120 gpm. Callaway operated from 1993 to 2011 at the maximum letdown flow, but has returned to the nominal value of 75 gpm. The effects of this increased flow rate have been evaluated. To account for the increase in fatigue, Callaway reduced the assumed number of load following transients to be more consistent with its operation as a base load plant. Also, starting in Refuel Outage 17, Callaway has switched from the normal charging flow path to the alternate charging flow path in order to spread fatigue over the two paths.

The operating experience of the Fatigue Monitoring program did not identify an adverse trend in performance. Occurrences that would be identified under the Fatigue Monitoring program will be evaluated to ensure there is no significant impact to safe operation of the plant, and corrective actions will be taken to prevent recurrence.

Guidance for re-evaluation, repair, or replacement is provided for locations where aging is found. There is confidence that the continued implementation of the Fatigue Monitoring program will effectively identify aging prior to loss of intended function.

Conclusion The continued implementation of the Fatigue Monitoring program, following enhancement, provides reasonable assurance that aging effects will be managed such that the systems and components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

Callaway Plant Unit 1 Page B-131 License Renewal Application Amendment 27

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