ULNRC-05979, Enclosure 2 to ULNRC-05979, Amendment 23, LRA Changes from RAI Responses and Commitment Updates Enclosure 2 Summary Table

Enclosure 2 to ULNRC-05979, Amendment 23, LRA Changes from RAI Responses and Commitment Updates Enclosure 2 Summary Table

ML13119A137
Person / Time
Site:	Callaway
Issue date:	04/26/2013
From:	Ameren Missouri, Union Electric Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML131190229	List: ML13119A133 ML13119A136 ULNRC-05979, Enclosure 2 to ULNRC-05979, Amendment 23, LRA Changes from RAI Responses and Commitment Updates Enclosure 2 Summary Table
References
ULNRC-05979, TAC ME7708
Download: ML13119A137 (33)

Text

ULNRC-05979 April 26, 2013 Page 1 of 33 Amendment 23, LRA Changes from RAI Responses and Commitment Updates Enclosure 2 Summary Table Affected LRA Section LRA Page Section 4.3.4 4.3-33, 4.3-34, and 4.3-35 Table 4.3-6 4.3-35, 4.3-36, 4.3-37, and 4.3-38 Table 4.3-7 4.3-37 and 4.3-38 Section 4.8 4.8-1 and 4.8-2 Section A1 A-2 Table A4-1, item 1 A-36 Table A4-1, item 2 A-36 and A-37 Table A4-1, item 4 A-37 Table A4-1, item 31 A-46 and A-47 Table A4-1, item 37 A-49 Table A4-1, item 38 A-49 Table A4-1, item 39 A-49 Section B1.3 B-2 and B-3 Section B1.4 B-3 and B-4 Section B3.1 B-127, B-128, B-129, B-130, and B-131

ULNRC-05979 April 26, 2013 Page 2 of 33 Chapter 4 TIME-LIMITED AGING ANALYSES Evaluation of Sentinel Locations for Environmental Assisted Fatigue In order to assure that the limiting plant-specific EAF locations are identified as sentinel locations, Callaway performed a systematic review of all wetted, RCPB components with a Class 1 fatigue analysis [Ref. 17]. This was done either to show that the NUREG/CR-6260 locations are bounding or to incorporate EAF into the licensing basis for those more limiting components. The screening used EPRI Technical Report 1024995 Environmentally-Assisted Fatigue Screening, Process and Technical Basis for Identifying EAF Limiting Locations, [Ref. 18].

The CUF for wetted, RCPB locations were categorized based on the strain-rate of the dominant transient. The strain-rate classification was determined with a qualitative assessment based on experience and not a quantitative stress analysis. The estimated strain rate was used to calculate an estimated Fen. The estimated Fen value was also calculated assuming a low DO environment; the maximum fluid/metal temperature; and the maximum sulfur concentration, and is based on the methods in NUREG/CR-5704 for austenitic stainless steels and Ni-Cr-Fe steels, and NUREG/CR-6583 for carbon and low alloy steels. This estimated Fen was then averaged with the maximum Fen for that material type to calculate the average Fen. The average Fen and the design basis CUFs were used to calculate the estimated Uen.

These estimated Uen were then organized according to their system, thermal zone, and material type. A thermal zone is defined as a collection of piping and/or vessel components which undergo essentially the same group of thermal and pressure transients during plant operations. The maximum Uen for each thermal zone and material was selected as a sentinel location. In addition, if the next highest Uen with the same thermal zone and material is within 50% of the maximum, additional locations were identified as sentinel locations. This initial list was reduced further using EPRI Technical Report 1024995

[Ref. 18].

• One Thermal Zone can bound another Thermal Zone in a System:

Both the CUF and Fen values for one sentinel location in one thermal zone are each higher than the CUF and Fen values for the sentinel locations in other thermal zones.

• One material in a Thermal Zone can bound other materials in the same Thermal Zone:

This circumstance could be achieved if within the same thermal zone, both the CUF and Fen values for one sentinel location composed of one material are each higher than the CUF and Fen values for the sentinel locations composed for all other materials.

Callaway Plant Unit 1 Page 4.3-33 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 3 of 33 Chapter 4 TIME-LIMITED AGING ANALYSES

• One material in a Thermal Zone can bound other materials in another Thermal Zone:

This circumstance combines the guidelines of the two listed above and must satisfy both criteria listed.

• A location with Uen < 0.8 may be removed from the sentinel location list:

If the sentinel location Uen for the projected number of design cycles is low (e.g., (Uen)

< 0.8), that sentinel location may be removed from the final list due to the small likelihood that it will be the leading sentinel location in a system. If, however, the sentinel location Uen for the projected number of design cycles is fairly high (e.g., Uen >

0.8), the possibility exists that it could remain the sentinel location for its group and should be included in the monitoring program that ensures that it does not exceed a value of 1.0.

• A location that can be shown to be bounded by another location on a common basis stress evaluation may be removed from the sentinel location list:

This judgment relies upon the comparison of transients in terms of severity and/or number of occurrences between locations in the same or different thermal zones. For example, it may be possible to demonstrate that:

o The bounding location resides in two different thermal zones and experiences more numerous and more severe transients than the bounded location that only experiences transients from one of those thermal zones. An example of this is the charging nozzle which experiences transients in the CVCS and RCS thermal zones and will bound the fatigue behavior of the other CVCS locations. This is appropriate because all locations are designed to the same CVCS transients (e.g. loss of charging and loss of letdown). This is conservative because the charging nozzles CVCS transients have a reflood component where the nozzle temperature experiences a step increase from the CVCS transient temperature to the RCS Cold Leg. For example, during a Loss of Letdown event, the charging nozzle will cooldown from normal charging temperature (500°F) to the VCT temperature (70°F). This is followed by an abrupt increase in temperature from 70°F to RCS cold leg temperature (560°F) as the charging flow stops to maintain allowable pressurizer level and RCS cold leg flow refloods the nozzle. No other CVCS component experiences this reflood shock from the cold leg. Thus, the charging nozzles will bound the fatigue behavior of the other CVCS locations that flow to or Callaway Plant Unit 1 Page 4.3-34 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 4 of 33 Chapter 4 TIME-LIMITED AGING ANALYSES from the RCS cold legs and can eliminate these other locations from the list of sentinel locations. The auxiliary spray piping also participates in two thermal zones (CVCS transients plus the main spray piping transients) and is designated as a sentinel location.

o For two locations in the same thermal zone which are designed to identical transients (both in terms of number and severity), the location with the higher value of Uen can be used to manage both locations. An example is the pressurizer heater well (SS) (CUF = 0.562, Fen = 13.117, Uen = 7.372) and the pressurizer (lower) instrument nozzle (SS) (CUF = 0.439, Fen = 13.117, Uen =

5.758). The higher Uen of the former location identifies the bounding location.

This is appropriate because both components are designed to the same design basis transients including insurge/outsurge. This is conservative because the components will experience the same transients in terms of number and severity over the life of the plant and the relative ranking of Uen will not change as fatigue accumulates in each component.

Table 4.3-7 identifies the final sentinel locations used to manage the EAF aging mechanism during the period of extended operation. The plant specific sentinel locations are incorporated into Table 4.3-6 in order to demonstrate that the Uen will remain below 1.0.

Those plant specific locations with a Uen greater than 1.0 will be evaluated further using the same methods as those used to remove conservatisms for the NUREG/CR-6260 locations described above. Under the Fatigue Monitoring program 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.

The results of these final analyses will be incorporated into the Fatigue Monitoring program either by counting the transients assumed, or computing fatigue usage through the CBF/SBF capabilities of the program to ensure the component does not exceed an Uen of 1.0. Any use of SBF will be implemented in compliance with RIS 2008-30. Therefore, the effects of the reactor coolant environment on the plant specific locations will be managed for the period of extended operation. These TLAAs are dispositioned in accordance with 10 CFR 54.21(c)(1)(iii).

Disposition: Aging Management, 10 CFR 54.21(c)(1)(iii)

Callaway Plant Unit 1 Page 4.3-35 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 5 of 33 Table 4.3-6 Licensing Basis Uen for the Sentinel Locations Further Location Material CUF Fen Uen CUF-Fen Basis Eval.

Req.

NUREG/CR-6260 Locations SA 533, Design basis CUF No RPV Bottom Head to Shell Junction Grade B, Class 1, 0.0070 2.45 0.01715 NUREG/CR-6583 Low Alloy Steel maximum Fen Design basis CUF No SA 508, Class 2, RPV Inlet Nozzle 0.0795 2.45 0.195 NUREG/CR-6583 Low Alloy Steel maximum Fen Design basis CUF No SA 508, Class 2, RPV Outlet Nozzle 0.1078 2.45 0.264 NUREG/CR-6583 Low Alloy Steel maximum Fen CUF re-evaluated with NB- No 3200 elastic methods SA 182, Type 316, based on 60 year cycle Hot Leg Surge Line Nozzle 0.07572 10.097 0.7646 projections, Stainless Steel NUREG/CR-5704 strain-rate dependent Fen Callaway Plant Unit 1 Page 4.3-35 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 6 of 33 Table 4.3-6 Licensing Basis Uen for the Sentinel Locations Further Location Material CUF Fen Uen CUF-Fen Basis Eval.

Req.

CUF re-evaluated with No benchmarked SBF and 60 Charging System Nozzle SA 182 Type 316, 0.0919 / 0.0782 6.22 / 6.75 0.5715 / 0.5273 year cycle projections,

[Normal and Alternate] Stainless Steel NUREG/CR-5704 strain-rate dependent Fen CUF re-evaluated with NB- No 3200 elastic methods Safety Injection Nozzle [Boron SA 182 Type 316, based on 60 year cycle 0.1135 6.495 0.7374 projections, Injection Header nozzles] Stainless Steel NUREG/CR-5704 strain-rate dependent Fen CUF re-evaluated with NB- No 3200 elastic methods Residual Heat Removal Inlet Nozzle SA 182 Type 316, based on design cycles, 0.0234 15.35 0.3591

[RHR nozzle-hot-leg] Stainless Steel NUREG/CR-5704 maximum Fen Plant Specific Limiting Locations Design basis CUF Yes CETNA Upper Nozzle Housing Stainless Steel 0.37 13.117 4.853 Estimated Fen based on NUREG/CR-5704 Design basis CUF Yes RPV Bottom Head Instrument Ni-Cr-Fe 0.3184 13.117 4.176 Estimated Fen based on Tubes NUREG/CR-5704 Callaway Plant Unit 1 Page 4.3-36 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 7 of 33 Table 4.3-6 Licensing Basis Uen for the Sentinel Locations Further Location Material CUF Fen Uen CUF-Fen Basis Eval.

Req.

Design basis CUF based No on NB-3200 elastic-plastic PZR Heater Penetration (Heater methods Stainless Steel 0.0103 13.117 0.135 Well)

Estimated Fen based on NUREG/CR-5704 Design basis CUF Yes Pressurizer Spray Nozzle Stainless Steel 0.411 9.013 3.704 Estimated Fen based on NUREG/CR-5704 CUF re-evaluated with CBF Yes and 60 year cycle Pressurizer Safety Valve Piping Stainless Steel 0.788 11.486 9.051 projections, Estimated Fen based on NUREG/CR-5704 Design basis CUF Yes Pressurizer Upper Instrument Stainless Steel 0.236 13.117 3.096 Estimated Fen based on Nozzle NUREG/CR-5704 Design basis CUF Yes Pressurizer Shell at Support Lug Low Alloy Steel 0.922 2.455 2.435 NUREG/CR-6583 maximum Fen Design basis CUF Yes Auxiliary Spray Piping Stainless Steel 0.72 10.350 7.452 Estimated Fen based on NUREG/CR-5704 Callaway Plant Unit 1 Page 4.3-37 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 8 of 33 Table 4.3-6 Licensing Basis Uen for the Sentinel Locations Further Location Material CUF Fen Uen CUF-Fen Basis Eval.

Req.

Design basis CUF Yes RCP Casing/Discharge Nozzle Stainless Steel 0.915 13.117 12.002 Estimated Fen based on Junction NUREG/CR-5704 Design basis CUF Yes RCS 2-inch Crossover Leg Loops 1 Stainless Steel 0.7 13.117 9.182 Estimated Fen based on

& 2 Drain Nozzles NUREG/CR-5704 CUF re-evaluated with CBF Yes and 60 year cycle Accumulator Nozzles Stainless Steel 0.54 10.350 5.589 projections, Estimated Fen based on NUREG/CR-5704 CUF re-evaluated with CBF No and 60 year cycle Hot Leg SIS Nozzles, projections, Stainless Steel 0.09 10.350 0.9315 Loops 2 and 3 Estimated Fen based on NUREG/CR-5704 Design basis CUF Yes RSG Tubesheet (Continuous Low Alloy Steel 0.428 2.455 1.051 NUREG/CR-6583 Region) maximum Fen Design basis CUF No RSG Tube-to-Tubesheet Ni-Cr-Fe 0.068 13.117 0.892 Estimated Fen based on Connection NUREG/CR-5704 Callaway Plant Unit 1 Page 4.3-38 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Enclosure 2 Page 9 of 33 Table 4.3-7: Sentinel Locations for EAF Monitoring Thermal NUREG Design Avg.

System Material Component Est. Uen Zone /CR-6260 CUF Fen

1. RPV Outlet Nozzle Y 0.1078 2.455 0.265 RPV Nozzles LAS

2. RPV Inlet Nozzle Y 0.0795 2.455 0.195 Reactor RPV Upper Pressure SS 3. CETNA Upper Nozzle Housing N 0.37 13.117 4.853 Head Vessel LAS 4. Bottom Head-to-Shell Junction Y 0.007 2.455 0.017 RPV Bottom Head Ni-Cr-Fe 5. Bottom Head Instrument Tubes N 0.3184 13.117 4.176 PZR Lower SS 6. Pressurizer Heater Penetration N 0.562 13.117 7.372 Head PZR Spray SS 7. Pressurizer Spray Nozzle N 0.411 9.013 3.704 PZR Pressurizer SS 8. Safety Valve Piping N 0.975 11.486 11.199 SRV/PORV SS 9. Pressurizer Upper Instrument Nozzle N 0.236 13.117 3.096 PZR Upper Head LAS 10. Pressurizer Shell at Support Lug N 0.922 2.455 2.435 Surge Piping Surge Line SS 11. Hot Leg Surge Nozzle Y 0.3 11.486 3.446 Callaway Plant Unit 1 Page 4.3-37 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Enclosure 2 Page 10 of 33 Table 4.3-7: Sentinel Locations for EAF Monitoring Thermal NUREG Design Avg.

System Material Component Est. Uen Zone /CR-6260 CUF Fen

12. Normal Charging Nozzles, Loop 1 Y 0.95 10.350 9.833 CVCS Charging SS

13. Alternate Charging Nozzles, Loop 4 Y 0.95 10.350 9.833 Auxiliary SS 14. Auxiliary Spray Piping N 0.72 10.350 7.452 Spray

15. RCP Casing/Discharge Nozzle SS N 0.915 13.117 12.002 Junction RCS RCS Cold Leg

16. RCS 2-inch Crossover Leg Loops 1 &

SS N 0.7 13.117 9.182 2 Drain Nozzles RHR Inlet RHR SS 17. RHR Nozzles, Hot Leg Loops 1 & 4 Y 0.81 10.350 8.384 (Suction)

BIT SS 18. BIT Nozzles (All Loops) Y 0.999 10.350 10.340 SI Accumulator SS 19. Accumulator Nozzles (All Loops) N 0.95 10.350 9.833 SIS SS 20. Hot Leg SIS Nozzles, Loops 2 & 3 N 0.1 10.350 1.035 LAS 21. RSG Tubesheet (Continuous Region) N 0.428 2.455 1.051 Steam Tubesheet Generator Ni-Cr-

22. RSG Tube-to-Tubesheet Connection N 0.068 13.117 0.892 Fe Callaway Plant Unit 1 Page 4.3-38 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 11 of 33 Chapter 4 TIME-LIMITED AGING ANALYSES

4.8 REFERENCES

1. Westinghouse Report WCAP-15400-NP. Analysis of Capsule X from the Ameren UE Callaway Unit 1 Reactor Vessel Surveillance Program. Rev. 0. June 2000.

Westinghouse Non-Proprietary Class 3.

2. Callaway PTLR. Callaway Plant Pressure and Temperature Limits Report. Rev. 5.

Released 11. December 2006.

3. Westinghouse Report. WCAP-17168-NP. Callaway Unit 1 Time-Limited Aging Analysis on Reactor Vessel Integrity. Rev. 0. September 2010. Westinghouse Non-Proprietary Class 3.

4. SIA Calculation 0900694.301. Environmentally-Assisted Fatigue (EAF) for Callaway.

Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 19 August 2010.

5. SIA Calculation 0901271.315. Residual Heat Removal (RHR) Inlet Nozzle Environmentally-Assisted Fatigue Analysis Calculation. Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 11 August 2010.

6. SIA Calculation 0901271.332. Charging Nozzle Environmentally-Assisted Fatigue (EAF) Analysis Using 60-Year of Operation Using Stress Based-Fatigue (SBF) Results from the Baseline Evaluation. Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 27 October 2011.

7. SIA Calculation 0901271.331. Safety Injection (BIT) Nozzle Environmentally-Assisted Fatigue (EAF) Analysis Using 60-Year Projected Numbers of Cycles. Rev 0. Structural Integrity Associates, Inc. San Jose, California. 16 September 2011. .

8. SIA Calculation 0901271.330. Hot Leg Surge Nozzle Environmentally-Assisted Fatigue (EAF) Analysis Using 60-Year Projected Numbers of Cycles. Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 15 September 2011.

9. Precision Surveillance Corporation Document No. CA-N1042-500. Final Report of the 25th Year IWL Inspection. Rev. 0. 16 September 2010. Supplemented by Callaway CAR 201009644.

10. Ameren Missouri Letter ULNRC-5100. Docket Number 50-483, Union Electric Company Callaway Plant, Transmittal of Inservice Inspection Summary Report for Refuel 13, and WCAP-16280-P, Flaw Evaluation Handbook For Callaway Unit 1 Reactor Vessel Inlet Nozzle Safe-End Weld Region, May 2004. 13 December 2004.

(ADAMS Accession No ML043650441).

Callaway Plant Unit 1 Page 4.8-1 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 12 of 33 Chapter 4 TIME-LIMITED AGING ANALYSES

11. Ameren Missouri Calculation BB-183. Evaluation of Reactor Vessel Cladding Indication Inside Bottom Head During Refuel 13. Rev. 1.

12. Westinghouse Topical Report WCAP-15666-A. Extension of Reactor Coolant Pump Motor Flywheel Examination. Rev. 1. October 2003.

13. Ameren Missouri Letter ULNRC-05553. Graessle, Luke H. Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility Operating License NPF-30 Follow-Up Information Regarding 10 CFR 50.55a Request: Proposed Alternative to ASME Section XI Requirements for Replacement of Class 3 Buried Piping (TAC No. MD6792). Fulton, MO. 9 October 2008. (ADAMS Accession No ML082900027).

14. Ameren Missouri Letter ULNRC-05542. Graessle, Luke H. Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility Operating License NPF-30 Additional Information Regarding 10 CFR 50.55a Request: Proposed Alternative to ASME Section XI Requirements for Replacement of Class 3 Buried Piping (TAC No MD6792). Fulton, MO. 15 September 2008. (ADAMS Accession No ML082630798).

15. SIA Report FP-CALL-310. Benchmarking of Charging Nozzle Stress-Based Fatigue.

Rev. 0. San Jose, California: Structural Integrity Associates. 22 June 2011.

16. SIA Report FP-CALL-304. Baseline Analysis of Callaway Plant Cycles and Fatigue Usage - Startup through 1/31/2011. Rev. 2. San Jose, California: Structural Integrity Associates. 12 October 2012

17. SIA Report FP-CALL-307. Environmentally-Assisted Fatigue Screening. Rev. 4 5.

San Jose, California: Structural Integrity Associates. 28 November 2012 April 2013.

18. EPRI Technical Report 1024995. Environmentally-Assisted Fatigue Screening, Process and Technical Basis for Identifying EAF Limiting Locations.

19. WOG Topical Report WCAP-15338-A. A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants. Westinghouse Electric Company LLC.

October 2002.

Callaway Plant Unit 1 Page 4.8-2 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 13 of 33 Appendix A Final Safety Analysis Report Supplement A1

SUMMARY

DESCRIPTIONS OF AGING MANAGEMENT PROGRAMS The integrated plant assessment and evaluation of time-limited aging analyses (TLAA) identified existing and new aging management programs necessary to provide reasonable assurance that components within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis (CLB) for the period of extended operation. Sections A1 and A2 describe the programs and their implementation activities.

Quality Assurance for Aging Management Programs Three elements common to all aging management programs discussed in Sections A1 and A2 are corrective actions, confirmation process, and administrative controls. These elements are included in the Callaway Plant QA Program, which implements the requirements of 10 CFR 50, Appendix B. The Callaway Plant QA Program is applicable to safety-related systems, structures and components that are subject to aging management review activities for license renewal. These three elements will are also be applied to the nonsafety-related systems, structures and components subject to aging management activities after enhancement to existing Callaway procedures.

Consideration of Operating Experience in Aging Management Programs (AMPs)

Operating Experience from plant-specific and industry sources is captured and systematically reviewed on an on-going basis in accordance with the quality assurance program which meets the requirements of 10 CFR Part 50, Appendix B and the operating experience program, which meets the requirements of NUREG-0737, Clarification of TMI Action Plan Requirements, Item I.C.5, Procedures for Feedback of Operating Experience to Plant Staff. The operating experience program interfaces with and relies on active participation in the Institute of Nuclear Power Operations operating experience program, as endorsed by the NRC. In accordance with these programs, all incoming operating experience items are screened to determine whether they may involve age-related degradation or aging management impacts. Items so identified are further evaluated and applicable AMPs may be enhanced or new AMPs may be developed, as appropriate, if it is determined that the effects of aging may not be adequately managed. Training on age-related degradation and aging management is provided to those personnel responsible for implementing the AMPs and who are likely to submit, screen, assign, evaluate, or otherwise process plant-specific and industry operating experience. Plant-specific operating experience associated with aging management and age-related degradation is reported to the industry in accordance with guidelines established in the operating experience program.

Callaway Plant Unit 1 Page A-2 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 14 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 1 Procedures will be enhanced to apply the elements of corrective actions, confirmation process, B1.3 Completed no later and administrative controls of the Callaway Plant Quality Assurance program to those than six months prior nonsafety-related SSCs requiring aging management. (Completed Amendment 23) to the PEO Completed Callaway Plant Unit 1 Page A-36 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 15 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 2 Upon receipt of the renewed operating license, the station operating experience review B1.4 Upon receipt of the process and Corrective Action Program will perform reviews of plant-specific and industry renewed operating operating experience to confirm the effectiveness of the license renewal aging management license Completed programs, to determine the need for aging management programs to be enhanced, or indicate the need to develop a new aging management program.

In order to provide additional assurance that internal and external operating experience related to aging management continues to be used effectively in the aging management programs, Callaway will enhance its operating experience program to:

• Explicitly require the review of operating experience for age-related degradation.

(Completed Amendment 18)

• Establish criteria to define age-related degradation. In general, the criteria will be used to identify aging that is considered excessive relative to design, previous inspection experience, and inspection intervals. (Completed Amendment 18)

• Establish coding for use in identification, trending and communications of age-related degradation. This coding will assist plant personnel in ensuring that, in addition to addressing the specific issue, the adequacy of existing aging management programs is assessed. This could lead to AMP revisions or the establishment of new AMPs, as appropriate. (Completed Amendment 18)

• Require communication of significant internal age-related degradation, associated with SSCs in the scope of license renewal, to the industry. Criteria will be established for determining when aging-related degradation is significant. (Completed Amendment 18)

Callaway Plant Unit 1 Page A-36 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 16 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule

• Require review of external operating experience for information related to aging management, and evaluation of such information for potential improvements to Callaway aging management activities. License Renewal Interim Staff Guidance (LR-ISG) documents will be reviewed as part of this external operating experience information as they are issued on an ongoing basis, capturing new insights or addressing issues that emerge from license renewal reviews. (Completed Amendment 21)

• Provide training to those responsible for screening, evaluating and communicating operating experience items related to aging-related degradation. This training will be commensurate with their role in the process. (Completed Amendment 23)

• Explicitly require AMP activities, criteria, and evaluations integral to the elements of the AMPs be included in the operating experience evaluation. (Completed Amendment 21)

Callaway Plant Unit 1 Page A-37 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 17 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 4 Implement the PWR Vessel Internals program as described in LRA Section B2.1.6. As part of B2.1.6 PWR Vessel 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 Each applicant/licensee is responsible for assessing its plants design and operating after the issuance of history and demonstrating that the approved version of MRP-227 is applicable to the MRP-227-A, PWR facility. Each applicant/licensee shall refer, in particular, to the assumptions regarding Internals Inspection plant design and operating history made in the FMECA and functionality analyses for and Evaluation reactors of their design (i.e., Westinghouse, CE, or B&W) which support MRP-227 and Guideline 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 Applicant/Licensee Action Item (A/LAI) #8 Item #5 A/LAI #8 Item 5:

(in part - reactor coolant system water environment portion) Completed no later For those cumulative usage factor (CUF) analyses that are TLAAs, the applicant may than six months prior use the PWR Vessel Internals Program as the basis for accepting these CUF analyses to the PEO in accordance with 10 CFR 54.21(c)(1)(iii) only if the RVI components in the CUF analyses are periodically inspected for fatigue-induced cracking in the components during the period of extended operation. The periodicity of the inspections of these components shall be justified to be adequate to resolve the TLAA. Otherwise, acceptance of these TLAAs shall be done in accordance with either 10 CFR 54.21(c)(1)(i) or (ii), or in accordance with 10 CFR 54.21(c)(1)(iii) using the applicants program that corresponds to NUREG-1801, Revision 2, AMP X.M1, Metal Fatigue of Callaway Plant Unit 1 Page A-37 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 18 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule Reactor Coolant Pressure Boundary Program. To 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 III, Subsection NG-2160 and NG-3121, the existing fatigue CUF analyses should include the effects of the reactor coolant system water environment.

Callaway Plant Unit 1 Page A-37 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 19 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation 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 system locations. The set includes the 4.3.4 to the PEO NUREG/CR-6260 sample locations for a newer-vintage Westinghouse Plant and plant- B3.1 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 III fatigue and fatigue crack growth analyses.

• include additional locations which receive more detailed monitoring. These locations were identified by evaluation of ASME Section III fatigue analyses and the locations evaluated for effects of the reactor coolant 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 Page A-46 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 20 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation 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, 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.

• 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.

• 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 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 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 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 Callaway Plant Unit 1 Page A-47 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 21 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule sentinel locations are updated appropriately and remain bounded consistent with the refined analysis.

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ULNRC-05979 April 26, 2013 Page 22 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 37 Complete an evaluation to determine if there are any additional plant-specific bounding EAF 4.3.2.2 Completed no later locations. The supporting environmental factors, F(en), calculations will be performed with 4.3.4 than six months prior NUREG/CR-6909 or NUREG/CR-6583 for carbon and low alloy steels, NUREG/CR-6909 or to the PEO NUREG/CR-5704 for austenitic stainless steels, and NUREG/CR-6909 for nickel alloys. Completed (Completed Amendment 2)

In order to determine if the pressurizer contains a limiting EAF location, the fatigue analyses will be revised to incorporate the affect effect of insurge-outsurge transients on the pressurizer lower head, surge nozzle, and heater well nozzles at plant specific conditions.

(Completed Amendment 2)

Those non-NUREG/CR-6260 locations with an EAF CUF greater than 1.0 will be further evaluated using same methods as those used for NUREG/CR-6260 locations to remove conservatisms from the preliminary EAF CUF. 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 alternative, the Fatigue Monitoring program will implement SBFs of certain locations in order to ensure the component does not exceed an EAF CUF of 1.0. Any use of SBF will be implemented in compliance with RIS 2008-30. (Moved to Item #31.)

The pressurizer contains a limiting EAF location. The fatigue analyses will be revised to incorporate the effect of insurge-outsurge transients in the pressurizer lower head.

(Completed Amendment 11)

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ULNRC-05979 April 26, 2013 Page 23 of 33 Appendix A Final Safety Analysis Report Supplement Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 38 The number of the most severe RCP component cooling water transient, elevated CCW inlet 4.3.2.1 Completed no later temperature transients, will be limited to 75 percent of its design value, i.e. limited to 150, in than six months prior order to accommodate the seasonal temperature change transient in the RCP thermal barrier to the PEO flange fatigue analysis. (Moved to Item #31.)

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ULNRC-05979 April 26, 2013 Page 24 of 33 Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 39 NFPA 805 and LRA GAP analysis: B2.1.13 Prior to March 25, A gap analysis of LRA Tables 2.3.3-20 and 3.3.2-20 will be provided to identify B2.1.14 2013.

differences between the existing and NFPA 805 post-transition changes. The results If the draft NFPA 805 and the impacts of these gaps on the fire protection program described in LRA Tables Safety Evaluation 2.3.3-20 and 3.3.2-20 will be summarized, as the basis for transitioning to the NFPA Report is not 805 nuclear safety capabilities. The summary will also list the fire protection systems available in February and components including structural fire barriers, (e.g., fire walls and slabs, fire doors, 2013, Ameren will fire barrier penetration seals, fire dampers, fire barrier coatings/wraps, provide an alternate equipment/personnel hatchways and plugs, metal siding), that will be added or schedule to address removed based on the NFPA 805 transition in the scope of license renewal in this commitment.

accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in (Revised Amendment accordance with 10 CFR 54.21(a)(1). 19)Within thirty days after the final NFPA 805 Safety Evaluation Report is issued. (Revised Amendment 23)

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ULNRC-05979 April 26, 2013 Page 25 of 33 Appendix B AGING MANAGEMENT PROGRAMS B1.3 QUALITY ASSURANCE PROGRAM AND ADMINISTRATIVE CONTROLS The Callaway Plant Quality Assurance (QA) Program implements the requirements of 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants and with the guidance of Regulatory Guide 1.33, Quality Assurance Program Requirement Requirements (Operation), as clarified by Appendix A of the Callaway Plant Operating Quality Assurance Manual (OQAM). This QA Program is consistent with the summary provided in Appendix A.2 of NUREG-1800 and the Appendix, Quality Assurance for Aging Management Programs, of NUREG-1801. The corrective action, confirmation process, and administrative controls of the Callaway (10 CFR 50 Appendix B)

Quality Assurance Program are applicable to all systems, structures and components (SSCs) subject to aging management programs and activities required during the period of extended operation.

The program elements of Corrective Action, Confirmation Process, and Administrative Controls are applicable as follows:

Corrective Action Callaway Plant applies its Corrective Action Program to safety-related and nonsafety-related SSCs that are subject to aging management. Corrective Action Program procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.

The Corrective Action procedures specify the methods to promptly report, evaluate, and correct conditions adverse to quality and significant conditions adverse to quality commensurate with the significance of the SSC or activity. Consistent with the significance of the adverse condition, these methods include: (1) problem identification, (2) problem reporting, (3) immediate response, (4) investigative action to determine the cause, (5) evaluation of the extent of condition and extent of cause, (6) assessment of impact on operability and assessment for reportability, (7) determination of corrective action to prevent recurrence or minimize the consequences, and (8) the performance and verification of corrective actions.

In the case of significant conditions adverse to quality, measures are implemented to ensure that the cause is determined and that corrective action is taken to preclude recurrence.

Significant conditions adverse to quality receive independent review and approval and are reported to appropriate levels of management.

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ULNRC-05979 April 26, 2013 Page 26 of 33 Appendix B AGING MANAGEMENT PROGRAMS Confirmation Process The Callaway Corrective Action Program requires that measures be taken to preclude repetition of significant conditions adverse to quality. These measures include documented actions to verify effective implementation of corrective actions.

Plant procedures include provisions for timely evaluation of adverse conditions and implementation of corrective actions required, including root cause evaluations and prevention of recurrence where appropriate (e.g.; significant conditions adverse to quality).

These procedures provide for tracking, coordinating, monitoring, reviewing, verifying, validating and approving corrective actions, and ensure that corrective actions have been effectively implemented.

The corrective action process is also monitored for potentially adverse trends. Identification of a potentially adverse trend due to recurring or repetitive unacceptable conditions will result in the initiation of a corrective action document.

Administrative Controls Callaway Plant organizational structure, responsibilities and authorities and personnel qualification requirements conform to Appendix B of 10 CFR 50. Formal review and approval processes exist for procedures and other forms of written instruction utilized for the activities performed under the programs credited for aging management. These procedures contain objectives, program scope, responsibilities, methods for implementation, and acceptance criteria.

Enhancement Procedures will be enhanced apply the elements of corrective actions, confirmation process, and administrative controls of the Callaway Plant Quality Assurance program to those nonsafety-related SSCs requiring aging management.

None.

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ULNRC-05979 April 26, 2013 Page 27 of 33 Appendix B AGING MANAGEMENT PROGRAMS B1.4 OPERATING EXPERIENCE Operating experience is used at Callaway to enhance plant programs, prevent repeat events, and prevent events that have occurred at other plants from occurring at Callaway. The operating experience process screens, evaluates, and acts on operating experience documents and information to prevent or mitigate the consequences of similar events. The operating experience process reviews operating experience from external (also referred to as industry operating experience) and internal (also referred to as plant-specific operating experience) sources. External operating experience includes INPO documents, NRC generic communications (e.g., NRC Generic Letters, Bulletins, Information Notices, Regulatory Issue Summaries), License Renewal Interim Staff Guidance, and other documents (e.g., 10 CFR 21 Reports, Licensee Event Reports, Nonconformance Reports). Recognizing that industry operating experience may be derived from other sources, the Corrective Action Program procedure requires that the identification of industry operating experience applicable to Callaway be documented in the Corrective Action Program for further evaluation. Internal operating experience includes event investigations, trending reports, lessons learned from in-house events, self-assessments, and the 10 CFR 50, Appendix B, corrective action process.

Each aging management program summary in this appendix contains a discussion of operating experience relevant to the program. This information was obtained through the review of in-house operating experience in the Corrective Action Program and the review of industry operating experience. Plant-specific operating experience was obtained by a review of the Callaway corrective action program records for the period January 1999 through June 2011 and applicable industry operating experience was reviewed based on plant responses to specific NRC Generic Letters, Generic Safety Issues, Information Circulars, IE Bulletins, Information Notices, and Regulatory Issue Summaries. This population of industry experience was supported by plant documentation available since the beginning of the project and includes the operating experience associated to the NUREG-1801, Revision 2 (January 2004 to approximately April 2009). These reviews ensured that there was no unique, plant-specific operating experience in addition to that provided in NUREG-1801. This review was augmented with information from the Callaway staff.

The applicable operating experience for each aging management program was reviewed and summarized in the Appendix B program summaries. Detailed records on the performance and effectiveness of each program are maintained in the Callaway records management system (including the Corrective Action Program). New programs utilized plant and/or industry operating experience as applicable, and discussed the operating experience and associated corrective actions as they relate to the implementation of the new program. The operating experience summary in each aging management program identifies past corrective actions and provides objective evidence that the effects of aging have been, and will continue to be, adequately managed so that the intended functions of Callaway Plant Unit 1 Page B-3 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 28 of 33 Appendix B AGING MANAGEMENT PROGRAMS the structures and components within the scope of each program will be maintained during the period of extended operation.

Upon receipt of the renewed operating license, the station operating experience review process and Corrective Action Program will perform reviews of plant-specific and industry operating experience to confirm the effectiveness of the license renewal aging management programs, to determine the need for aging management programs to be enhanced, or indicate the need to develop a new aging management program.

Evaluation of operating experience that relates to aging management will consider and document as appropriate:

• Systems, structures or components that are similar or identical to those involved with the identified operating experience issue, to gain relevant lessons learned.

• Material of construction, operating environment and aging effects associated with the identified aging issue so that lessons learned can be applied to susceptible SSCs within the scope of license renewal.

• Aging mechanisms associated with the operating experience to confirm that Callaway has appropriate AMPs in place to manage aging that could be caused by these mechanisms.

• AMPs associated with this operating experience so that if the AMPs have been demonstrated to be ineffective, similar AMPs in place at Callaway can be evaluated to determine if AMP changes are appropriate, or a new AMP is needed. Included in this review is consideration of activities, criteria, and evaluations integral to the elements of the plant AMPs.

Training on age-related degradation and aging management is provided to those personnel responsible for implementing the AMPs and who are likely to submit, screen, assign, evaluate, or otherwise process plant-specific and industry operating experience. Plant-specific operating experience associated with aging management and age-related degradation is reported to the industry in accordance with guidelines established in the operating experience program.

In order to provide additional assurance that internal and external operating experience related to aging management continues to be used effectively in the aging management programs, Callaway will enhance its operating experience program to:

1. Provide training to those responsible for screening, evaluating and communicating operating experience items related to aging-related degradation. This training will be commensurate with their role in the process.

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ULNRC-05979 April 26, 2013 Page 29 of 33 Appendix B AGING MANAGEMENT PROGRAMS B3.1 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 III 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 III NB-3200 requirements, or a method will be benchmarked 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 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 and plant-specific bounding environmentally assisted fatigue (EAF) locations. 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 Callaway Plant Unit 1 Page B-127 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 30 of 33 Appendix B AGING MANAGEMENT PROGRAMS for nickel alloys.

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 locations. The 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 III fatigue and fatigue crack growth analyses.

Procedures will be enhanced to include additional locations which receive more detailed monitoring. These locations were identified by evaluation of ASME Section III fatigue analyses and the locations evaluated for effects of the reactor coolant environment. The monitoring methods will be benchmarked consistent with the NRC RIS 2008-30.

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ULNRC-05979 April 26, 2013 Page 31 of 33 Appendix B AGING MANAGEMENT PROGRAMS 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.

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.

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ULNRC-05979 April 26, 2013 Page 32 of 33 Appendix B AGING MANAGEMENT PROGRAMS 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 (Greens 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 III 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 III 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.

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 Callaway Plant Unit 1 Page B-130 License Renewal Application Amendment 23

ULNRC-05979 April 26, 2013 Page 33 of 33 Appendix B AGING MANAGEMENT PROGRAMS 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.

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