WM 07-0049, License Renewal Application, Amendment 1

License Renewal Application, Amendment 1

ML071580237
Person / Time
Site:	Wolf Creek
Issue date:	06/01/2007
From:	Matthew Sunseri Wolf Creek
To:	Document Control Desk, Office of New Reactors
References
WM 07-0049
Download: ML071580237 (64)

Text

W LEF CREEK NUCLEAR OPERATING CORPORATION June 1, 2007 Matthew W. Sunseri Vice President Oversight WM 07-0049 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Reference:

1) Letter ET 06-0038, dated September 27, 2006, from T. J. Garrett, WCNOC, to USNRC

Subject:

Docket No. 50-482: Wolf Creek Generating Station License Renewal Application, Amendment 1 Gentlemen:

Reference 1 provided Wolf Creek Nuclear Operating Corporation's (WCNOC) License Renewal Application (LRA) for the Wolf Creek Generating Station (WCGS). As part of the review for license renewal, the Nuclear Regulatory Commission (NRC) staff conducted two audits at WCGS. The LRA Aging Management Program (AMP) audit was conducted the week of March 26, 2007 and the LRA Aging Management Review (AMR) audit was conducted the week of May 7, 2007. During the course of these audits the NRC staff also audited Time Limited Aging Analyses (TLAA).Based on the results of the TLAA audits, it was determined that an Amendment to Sections 4.1 and 4.3 of Reference 1 would facilitate the NRC staff review. Enclosure 1 provides the amended sections to Reference 1.The question and answer database that was compiled during the TLAA audits will be submitted separately.

A2-P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET (UJ I J c PO WM 07-0049 Page 2 of 3 This letter contains no new commitments.

If you have any questions concerning this matter, please contact me at (620) 364-4008, or Mr. Kevin Moles at (620) 364-4126.Sincerely, Matthew W. Sunseri MWS/rlt Enclosure 1- LRA, Amendment 1 cc: J. N. Donohew (NRC), w/e V. G. Gaddy (NRC), w/e B. S. Mallett (NRC), w/e V. Rodriguez (NRC), w/e Senior Resident Inspector (NRC), w/e WM 07-0049 Page 3 of 3 STATE OF KANSAS )SS COUNTY OF COFFEY )Matthew W. Sunseri, of lawful age, being first duly sworn upon oath says that he is Vice President Oversight of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.By~r '/ A4J Matthew W. Sunseri Vice President Oversight SUBSCRIBED and sworn to before me this / day ofj,.fle, 2007.RHONDA L. TIEMEYER MY COMMISSION EXPIRES Notary Public -Exp.,irati January 11, 2010 Expiration Date l'/Z. / 0 Enclosure 1 to WM 07-0049 Wolf Creek Generating Station License Renewal Application Amendment I

Section 4 TIME-LIMITED AGING ANALYSES 4.0 TIME-LIMITED AGING ANALYSES

4.1 INTRODUCTION

Chapter 4 describes the Time-Limited Aging Analyses (TLAAs) for the WCGS in accordance with 10 CFR 54.3(a) and 54.21(c).

Subsequent sections describe TLAAs within these common general categories:

1. Neutron Embrittlement of the Reactor Vessel (Section 4.2)2. Metal Fatigue of Vessels and Piping (Section 4.3)3. Environmental Qualification (EQ) of Electrical Equipment (Section 4.4)4. Loss of Prestress in Concrete Containment Tendons (Section 4.5)5. Fatigue of the Containment Liner and Penetrations (Section 4.6)6. Other Plant-Specific TLAAs (Section 4.7)The information on each specific TLAA within these general categories is organized under three subheads: Summary Description A brief description of the TLAA topic and of the affected components.

Analysis A description of the current licensing basis analysis, that is, of the TLAA itself, including implications of the extended licensed operating period.Disposition The disposition of the TLAA for the extended licensed operating period, in accordance with 10 CFR 54.21(c)(1): " Validation

-10 CFR 54.21(c)(1)(i)

-The analysis remains valid for the period of extended operation, or" Revision -10 CFR 54.21(c)(1)(ii)

-The analysis has been projected to the end of the period of extended operation, or" Aging Management

-10 CFR 54.21(c)(1)(iii)

-The effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Wolf Creek Generating Station Page 4.1-1 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES 4.1.1 Identification of TLAAs Survey of Design and Licensing Bases An analysis, calculation, or evaluation is a "Time-Limited Aging Analysis" (TLAA) under the 10 CFR 54 License Renewal Rule (the Rule) only if it meets all six of the 10 CFR 54.3(a)criteria: Time-limited aging analyses are those licensee calculations and analyses that: (1) Involve systems, structures, and components within the scope of license renewal;(2) Consider the effects of aging;(3) Involve time-limited assumptions defined by the current operating term, for example, 40 years;(4) Were determined to be relevant by the licensee in making a safety determination; (5) Involve conclusions or provide the basis for conclusions related to the capability of the system, structure, and component to perform its intended functions; and (6) Are contained or incorporated by reference in the CLB [current licensing basis].The Rule requires that: (1) A list of time-limited aging analyses, as defined in §54.3, must be provided.The applicant shall demonstrate that -(i) The analyses remain valid for the period of extended operation;(ii) The analyses have been projected to the end of the period of extended operation; or (iii) The effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

(2) A list must be provided of plant-specific exemptions granted pursuant to 10 CFR 50.12 and in effect that are based on time-limited aging analyses as defined in§54.3. The applicant shall provide an evaluation that justifies the continuation of these exemptions for the period of extended operation.

A list of potential TLAAs was assembled from regulatory guidance and industry experience, including:

Wolf Creek Generating Station Page 4.1-2 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES" The NUREG-1800, "Standard Review Plan for License Renewal"* The NEI 95-10 "Industry Guideline for Implementing the Requirements of 10 CFR 54, the License Renewal Rule"* The 10 CFR 54 Final Rule "Statement of Considerations"* Westinghouse Engineering Owner's Group Topical Reports" Prior license renewal applications

• Plant-specific document reviews and interviews with plant personnel.

Keyword searches examined the current licensing basis (CLB) to determine whether the design or analysis feature of each potential TLAA exists in the WCGS licensing basis, and to identify additional potential unit-specific TLAAs. The CLB search included: " The Updated Safety Analysis Report (USAR)* Technical Specifications

• The NRC Safety Evaluation Report (SER) for the original operating license" All subsequent NRC Safety Evaluations (SEs)" Wolf Creek Nuclear Operating Corporation (WCNOC) and NRC docketed licensing correspondence.

Licensing basis program documents, such as the in-service inspection and equipment qualification programs (ISI and EQ programs), were reviewed separately.

Only those potential TLAAs meeting all six criteria of 10CFR54.3(a) are actual TLAAs requiring disposition in accordance with 54.21(c).

The list of potential TLAAs was reviewed (screened) against the six 10 CFR 54.3(a) criteria from information in the CLB source documents, such as:* Design calculations

• Code stress reports or code design reports" Environmental Qualification Work Packages* ISI reports (ASME Section XI Summaries of Reportable Indications)

• Design Basis Documents* Drawings Wolf Creek Generating Station Page 4.1-3 License Renewal Application

)

Section 4 TIME-LIMITED AGING ANALYSES Specifications These TLAA source documents confirmed the screening and provided the information and the basis for the dispositions in this chapter.4.1.2 Aging Management Review NUREG-1 801 identifies numerous aging effects that require evaluation as possible TLAAs in accordance with 10 CFR 54.21(c).

Each of these was reviewed in the appropriate aging management review, or in this chapter, and dispositioned as a TLAA if identified as such under the 10 CFR 54.3(a) criteria.

Tables 3.1.1, 3.2.1, 3.3.1, 3.4.1, 3.5.1 and 3.6.1, as discussed in Section 3.0, Aging Management Reviews, list the TLAA line items of the NUREG-1801 Volume 1 Summary Tables, and identify the specific sections relating to the required further evaluations.

4.1.3 Identification of Exemptions 10 CFR 54.21(c)(2) requires a list of plant-specific exemptions granted pursuant to 10 CFR 50.12 and in effect that are based on time-limited aging analyses as defined in§54.3. The applicant shall provide an evaluation that justifies the continuation of these exemptions for the period of extended operation.

A search of docketed correspondence, the operating license, and the Updated Safety Analysis Report (USAR) identified and listed exemptions in effect. Each exemption in effect was then evaluated to determine whether it involved a TLAA as defined in 10 CFR 54.3.The search found 12 exemptions that have been granted pursuant to 10CFR50.12.

One of the 10 CFR 50.12 exemptions from the original SER and Operating License is supported in part by a TLAA. This is an exemption from the then-current requirement of 10 CFR 50 Appendix A, General Design Criterion 4 to assume a break "...equivalent

... to the double-ended rupture of the largest pipe in the reactor coolant system." This exemption was granted with the Safety Evaluation Report for the original license. The supporting leak-before-break analysis is based in part upon a bounding-case evaluation of fatigue crack growth effects for a set of design transients consistent with those specified for WCGS for the original 40-year licensed operating period. This TLAA is described in Section 4.3.2.11,"Fatigue Crack Growth Assessment in Support of a Fracture Mechanics Analysis for the Leak-Before-Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures." One exemption granted in 1999 permitted use of ASME Code Case N-514 for determining the low-temperature overpressure protection (cold over pressurization mitigation system, COMS) pressure setpoint.

Although this permitted change in method did not depend on a TLAA -and the exemption therefore also did not depend on a TLAA -the calculation of the COMS setpoint itself is a TLAA, and is so described in Section 4.2.4, "Low Temperature Overpressure Protection (LTOP)." Wolf Creek Generating Station Page 4.1-4 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES 4.1.4 Summary of Results Sections 4.2 through 4.7 describe six general categories of TLAAs. They are listed in Table 4.1-1. They are presented in the order in which they appear in Sections 4.2 through 4.7 and following the order of NUREG-1800, "Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants." NUREG-1800, Tables 4.1-2 and 4.1-3, list examples of analyses that could be TLAAs, depending on the applicant's current licensing basis (CLB). Table 4.1-2 summarizes the results of the WCNOC review of the analyses identified in NUREG-1800 Tables 4.1-2 and 4.1-3.Table 4.1-1: List of TLAAs TLAA Disposition Report Category Descrption Categoryý'1 Section 1. Reactor Vessel Neutron Embrittlement 4.2 Neutron Fluence, Upper Shelf Energy and Adjusted ii, iii 4.2.1 Reference Temperature (Fluence, USE, and ART)Pressurized Thermal Shock (PTS) ii 4.2.2 Pressure-Temperature (P-T) Limits ii 4.2.3 Low Temperature Overpressure Protection (LTOP) ii 4.2.4 2. Metal Fatigue 4.3 Fatigue Management Program 4.3.1 ASME Section III Class 1 Fatigue Analysis of Vessels, 4.3.2 Piping, and Components Reactor Pressure Vessel, Nozzles, Head, and Studs iii 4.3.2.1 Control Rod Drive Mechanism (CRDM) Pressure Housings, Adapter Plugs, and Canopy Seals Reactor Coolant Pump Pressure Boundary iii 4.3.2.3 Components Pressurizer and Nozzles iii 4.3.2.4 Steam Generator ASME Section III Class 1, Class 2 Secondary Side, and Feedwater Nozzle Fatigue iii 4.3.2.5 Analyses ASME Section III Class 1 Valves iii 4.3.2.6 ASME Section III Class 1 Piping and Piping Nozzles iii 4.3.2.7 Wolf Creek Generating Station License Renewal Application Page 4.1-5 Section 4 TIME-LIMITED AGING ANALYSES Table 4.1-1: List of TLAAs Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and Stratification iii 4.3.2.8 Primary Coolant System Heatup Expansion Noise Events 4.3.2.9 High Energy Line Break Postulation Based on Fatigue iii 4.3.2.10 Cumulative Usage Factor Fatigue Crack Growth Assessment in Support of a Fracture Mechanics Analysis for the Leak-Before-Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures ASME Section III Subsection NG Fatigue Analysis of i 433 Reactor Pressure Vessel Internals Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety i, iii 4.3.4 Issue 190)Assumed Thermal Cycle Count for Allowable Secondary Stress Range Reduction Factor in B3 1.1 and ASME i, ii 4.3.5 Section III Class 2 and 3 Piping Fatigue Design of Spent Fuel Pool Liner and Racks for 4.3.6 Seismic Events Fatigue Design and Analysis of Class 1E.Electrical i 4.3.7 Raceway Support Angle Fittings for Seismic Events 3. Environmental Qualification of Electrical Equipment iii 4.4 4. Concrete Containment Tendon Prestress i, iii 4.5 5. Containment Liner Plate, Polar Crane Bracket, and Penetration 4.6 Load Cycles Absence of a TLAA for Containment Liner Plate, Polar Crane Bracket, and Containment Penetration Design -4.6.1 (Except Main Steam Penetrations)

Design Cycles for the Main Steam Line Penetrations i 4.6.2 6. Plant-Specific Time-Limited Aging Analyses 4.7 Wolf Creek Generating Station License Renewal Application Page 4.1-6 Section 4 TIME-LIMITED AGING ANALYSES Table 4.1-1: List of TLAAs Containment Polar Crane, Fuel Building Cask Handling Crane, Spent Fuel Pool Bridge Crane, and Fuel Handling Machine CMAA-70 Load Cycle Limits i 4.7.1 Absence of a TLAA for Reactor Vessel Underclad 4.7.2 Cracking Analyses Absence of a TLAA in a Reactor Coolant Pump Flywheel 473 Fatigue Crack Growth Analysis i 10 CFR 54.21(c)(1)(i)

-Validation:

Demonstration that "The analyses remain valid for the period of extended operation," ii 10 CFR 54.21 (c)(1)(ii)

-Revision:

Demonstration that "The analyses have been projected to the end of the period of extended operation," or iii 10 CFR 54.21(c)(1)(iii)

-Aging Management:

Demonstration that "The effects of aging on the intended function(s) will be adequately managed for the period of extended operation." Wolf Creek Generating Station Page 4.1-7 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Table 4.1-2 Review of Analyses Listed in NUREG-1800 Tables 4.1-2 and 4.1-3 NUREG-1 800 Examples Applicability to WCGS Section NUREG-1800, Table 4.1 Potential TLAAs Reactor vessel neutron Embrittlement Yes 4.2 Concrete containment tendon prestress Yes 4.5 Metal fatigue Yes 4.3 Environmental qualification of electrical Yes 4.4 equipment Metal corrosion allowance No N/A No explicit 40-year basis applies.Inservice flaw growth analyses that No demonstrate structure stability for 40 years No explicit 40-year basis applies.Inservice local metal containment corrosion No N/A analyses No explicit 40-year basis applies.High-energy line-break postulation based on Yes 4.3.2.10 fatigue cumulative usage factor UNREG-18O, TTable 4.1 Additional Examples of Plant-Specific TLAAs Intergranular separation in the heat-affected No zone (HAZ) of reactor vessel low-alloy steel No HAZ analyses were identified 4.7.2 under austenitic SS cladding within the CLB.Low-temperature overpressure (LTOP) Yes 4.2.4 analyses Fatigue analysis for the main steam supply Yes lines to the turbine-driven auxiliary feedwater 7000-cycle stress range reduction 4.3.5 pumps factor only.Fatigue analysis for the reactor coolant pump No flywheel No explicit 40-year basis applies.Wolf Creek Generating Station License Renewal Application Page 4.1-8 Section 4 TIME-LIMITED AGING ANALYSES Table 4.1-2 Review of Analyses Listed in NUREG-1800 Tables 4.1-2 and 4.1-3 NUREG-1800 Examples Applicability to WCGS Section Fatigue analysis of polar crane Yes 4.7.1 Design to CMVAA-70.Yes WCGS is designed to ASME Flow-induced vibration endurance limit,Section III Subsection NG, 1974 transient cycle count assumptions, and edition.4.3.3 ductility reduction of fracture toughness for No the reactor vessel internals WCGS is designed with no explicit 40-year embrittlement analysis for internals.

Leak before break Yes 4.3.2.11 Fatigue analysis for the containment liner No plate No fatigue evaluations were 4.6.1 performed.

Containment penetration pressurization Yes cycles Design cycles for main steam line 4.6.2 penetrations.

Reactor vessel circumferential weld No inspection relief (BWR) WCGS is a PWR.Wolf Creek Generating Station License Renewal Application Page 4.1-9 Section 4 TIME-LIMITED AGING ANALYSES 4.3 METAL FATIGUE ANALYSIS This section addresses design of mechanical system components supported by fatigue analyses; and also of components whose design depends on an assumed number of load cycles without a calculated fatigue usage factor.Section 4.6, "Containment Liner Plate, Polar Crane Bracket, and Penetration Load Cycles," describes fatigue in the containment vessel.Fatigue analyses are required for piping, vessels, and heat exchangers designed to American Society of Mechanical Engineers "Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components," Division 1, "Metal Components," Subsection NB, "Requirements for Class 1 Components" (ASME Section III Class 1).2 Fatigue analyses may also be invoked for Class 1 pump and valve pressure boundaries.

The design of piping and vessels to certain other codes and code sections, including ASME Section III Class 2 and 3, ANSI-ASME B31.1, and ASME Section VIII Division 2, may assume a stated number of full-range thermal and displacement cycles.Westinghouse license renewal topical report WCAP-14575-A, "Aging Management Evaluation for Class 1 Piping and Associated Pressure Boundary Components," and the NRC staff evaluation of it [Refs. 4.9.16 and 4.9.17], catalog aging effects and the Class 1 components affected by them, and propose dispositions for the period of extended operation consistent with 10 CFR 54. This Section, particularly Subsection 4.3.2, "ASME Section III Class 1 Fatigue Analysis of Vessels, Piping, and Components," supplies additional WCGS-specific information necessary for disposition of these TLAAs.This Section also describes fatigue analyses and evaluations of a limited number of other non-Class 1 components.

Basis of Fatigue Analyses ASME Section III Class 1 design specifications define a set of static and transient load conditions for which components are to be designed.

Although original design specifications commonly state that the transient conditions are for a 40-year design life, the fatigue analyses themselves are based on the specified number of occurrences of each transient rather than on this lifetime.

The number of occurrences of each transient for use in the fatigue analyses was specified to be somewhat larger than the number of occurrences expected during the 40-year licensed life of the plant, based on engineering experience and 2 Titles are from the 1971 edition of the code, as used for the reactor vessel and steam generators.

Later editions reorganized the Section III material and removed the Division 1 title, so that this subsection became"Division 1 -Subsection NB, Class 1 Components." Wolf Creek Generating Station Page 4.3-1 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES judgment.

This provides a margin of safety and an allowance for future changes in design or operation that may affect system design transients.

Operating experience at WCGS and at other similar units has demonstrated that the assumed frequencies of design transients, and therefore the number of transient cycles assumed for a 40-year life, were conservative.

4.3.1 Fatigue Management Program In accordance with WCGS Technical Specification program requirements for tracking component cyclic or transient occurrences, the present WCGS fatigue management program uses cycle counting and usage factor tracking to ensure that actual plant experience remains bounded by design assumptions and calculations reflected in the USAR. The program was prepared, verified, and validated under a 10 CFR 50 Appendix B quality assurance program for plant-specific implementation at WCGS.The present program will be enhanced in order to support safe operation of WCGS for the period of extended operation, as summarized in Section 4.3.1.3, "Program Enhancements for the Period of Extended Operation," and Appendix B3.1, "Metal Fatigue of Reactor Coolant Pressure Boundary." The enhanced WCGS fatigue management program will monitor plant transients and cumulative usage factors (CUFs) for a subset of ASME Section III Class 1 reactor coolant pressure boundary vessel and piping locations, to ensure that licensing basis limits on fatigue effects, in all locations, are not exceeded without appropriate corrective measures.4.3.1.1 Present WCGS Fatigue Monitoring Program Scope The present WCGS fatigue management program monitors the components and piping listed in Section 4.3.1.2, "Present and Projected Status of Monitored Locations." Corrective Action Corrective action is initiated whenever a cycle count or usage factor action limit is reached.See Section 4.3.1.3, "Program Enhancements for the Period of Extended Operation," for the action limits of the enhanced program.Margins Fatigue analyses incorporate several conservative assumptions and methods. The associated additional margins described below ensure that usage factors predicted by the design calculation will exceed (or "bound") the usage factors actually accumulated by the components.

Wolf Creek Generating Station Page 4.3-2 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Fatigue Desiqn Curve with Margin for Uncertainties and Moderate Environmental Effects: The ASME Section III fatigue S-N curves (allowable alternating stress intensity versus number of cycles) are based on regression analysis of a large number of fatigue data points for samples strain-cycled in air. The curves include adjustments for the elastic modulus and for departure from zero mean stress; and a design margin for uncertainties, including modest environmental effects (ASME Section III -1965, Par. N-415). The design margin is a factor of 2 on stress or a factor of 20 on cycles, whichever produced the lower, more conservative allowable for the data set.Boundinq Parameters for Transients:

Fatigue analyses assume a given number of cycles of each transient of a set of transient events, where each event is defined by limiting pressure and temperature transients and other load conditions.

Since actual event cycles are seldom as severe as those considered in the analysis, the resulting stress ranges are lower.Therefore, the contributions to cumulative usage factor are also lower.Actual Number of Event Cycles: The analytic limit for a fatigue analysis is a cumulative usage factor (CUF) of 1.0 at any location.

The CUF is calculated as the sum of all contributing partial usage factors for the design basis number of each of the design basis cyclic loading events. Even if the analysis showed a calculated usage factor at the 1.0 limit for a location, and even if the design basis number of events were reached for one or more events of a set, but not for the remainder of the assumed events, some margin will remain to the 1.0 limit because not as many design basis events will have occurred as assumed by the analysis.

Therefore they will not have contributed as much to the usage factor as the analysis assumed.4.3.1.2 Present Status of Monitored Locations The present WCGS fatigue management program was implemented in 1997. The usage factors calculated by the program include the effects of cycles incurred before the program was installed.

The cycle count input to the program was accumulated from two periods.Effects were counted or estimated from the WCGS operating history for the period between initial cold hydro in February 1982 to the installation of the automated transient data acquisition system in March 1992. Data from the data acquisition system and from operating records were used thereafter, up to the implementation of the fatigue management program.Cycle Counts Table 4.3-1, "Significant Transient Cycle Limits Tracked by the WCGS Fatigue Management Program," includes the cycle counts to December 31, 2005. The cycle accumulations shown in this table indicate that the original design basis number of events should not be reached in a 60-year operating life, nor should the code usage factor limit of 1.0 be exceeded.WCGS operating changes described in Section 4.3.2.4, "Pressurizer and Pressurizer Nozzles," have successfully mitigated surge line and pressurizer thermal stratification and Wolf Creek Generating Station Page 4.3-3 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES insurge-outsurge transient effects, limiting the remaining causes of fatigue usage to counted transients.

See also Section 4.3.2.8, "Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and Stratification." Cycle counting is therefore an effective means of determining the contribution of these effects to fatigue usage in these locations.

Table 4.3-1 Significant Transient Cycle Limits Tracked by the WCGS Fatigue Management Program Transient Design Cycles to Boldface are Technical Specification surveillance transients.

Limit 12/31/2005

"(M)" are manually recorded, all others are recorded automatically.

Normal Events 1. RCS Heatup at 5 100 °F/hour, 200 27(1)Tavg from 5 200 OF to a 550 OF 2. RCS Cooldown at 5 100 OFlhour, 200 25(1)Tavg from > 550 OF to 5 200 OF 3. Pressurizer Heatup at-< 100 'F/hour, 200 27 Tavg from -200 OF to > 650 OF 4. Pressurizer Cooldown at < 200 OF in any one hour, 200 25 Tavg from > 650 OF to 5 200 OF 5. Reactor Coolant Leak Test, 50(2) 6 Pressurized to > 2485 psig 6. Refueling (M) 80 14 7. Turbine Roll Test (M) 20 9 8. Loop Out of Service (M)a. Normal Loop Shutdown 80 NS(3)b. Normal Loop Startup 70 0 9. Reduced Temperature Return to Power (M) 2,000 0 Wolf Creek Generating Station License Renewal Application Page 4.3-4 Section 4 TIME-LIMITED AGING ANALYSES Table 4.3-1 Significant Transient Cycle Limits Tracked by the WCGS Fatigue Management Program Transient Design Cycles to Boldface are Technical Specification surveillance transients.

Limit 12/3112005

"(M)" are manually recorded, all others are recorded automatically.

Auxiliary Events 10. Accumulator Safety Injection 4 1 (Inadvertent Accumulator Blowdown)11. Inadvertent Safety Injection Actuation (High Head Safety Injection, 60 6 HHSI, By High-head Centrifugal Charging Pumps)12. Low Head Safety Injection (LHSI) Injection N/A(4) 1 13. COMS (LTOP) (M) 6,000(5) 0 14. Loss of Normal Charging [(Loop 1)] 120 34 15. Loss of Alternate Charging [(Loop 4)] 120 8 16. Loss of Letdown [and Return to Service] 200 18 Upset Events 17. Loss of Load Without Immediate Turbine or Reactor Trip, 80> 15% of Rated Thermal Power to 0% of Rated Thermal 18. Loss of Offsite AC Power (LOOP) to the ESF Busses 40 7 19. Loss of Flow in One Reactor Coolant Loop 80 2-Loss or Trip of Only One Reactor Coolant Pump 20. Reactor Trip, 100% to 0% of Rated Thermal Power 400 55 a. No Inadvertent Cooldown, or 230 55 b Cooldown with Safety Injection, or 10 0 c. Cooldown with No Safety Injection 160 0 d. No Inadvertent Cooldown, with Turbine Overspeed (Included (20 of NS(6)in the 230 with No Inadvertent Cooldown) 230)21. Auxiliary Spray Actuation, 10 0 Spray Water Differential

> 320 OF Wolf Creek Generating Station License Renewal Application Page 4.3-5 Section 4 TIME-LIMITED AGING ANALYSES Table 4.3-1 Significant Transient Cycle Limits Tracked by the WCGS Fatigue Management Program Tranisient Dsg ylst Boldface are Technical Specification surveillance transients.

Design Cycles to"(M)" are manually recorded, all others are recorded automatically.

Limit 12/3112005

22. Inadvertent RCS Depressurization (M) 20 0 23. Inadvertent Startup of Inactive Loop (M) 10 0 24. Excessive Feedwater Flow (M) 30 0 25. Operating Basis Earthquake (OBE) (M) 200(7) 0 Test Events 26. Reactor Coolant Hydrostatic Test, 5(8)Pressurized to a 3106 psig 27. Secondary System Hydrostatic Test, 5(9) 4(9)Pressurized to > 1350 psig Emergency Events 28. Small-Break Loss-of-Coolant Accident (LOCA) (M) 5 0 29. Small Steam Line Break (M) 5 0 30. Complete Loss of Flow (M) 5 0 Faulted Events 31. Large-Break LOCA (M) 1 0 32. Large Steam Line Break, 1 0> 6 inch Steam Line (M)33. Simultaneous Steam Line -Feedwater Line Break (M) 1 0 34. Post-LOCA Operation (M) (Not tracked by the fatigue management 1 0 program)Wolf Creek Generating Station License Renewal Application Page 4.3-6 Section 4 TIME-LIMITED AGING ANALYSES Notes to Table' The difference between the number of heatup and cooldown cycles occurs because the computer algorithm or manual additions to the cycle count. record can count additional heatups or additional cooldowns under special circumstances, such as prolonged holds at constant intermediate temperature.

In general, both the computer algorithm and manual additions will conservatively add these cycles and will not incorrectly omit these cycles.2 200 events documented in the original source and in USAR Table 3.9(N)-l.

However "In actual practice the primary side will be pressurized

[only] to the normal operating pressure..." [USAR § 3.9(N).1.1].

The allowed number of equivalent full-pressurization cycles is 50 for fatigue monitoring purposes.3 Not stated or not separately stated; no value reported.4 LHSI actuation as an independent transient is expected only following a faulted event such as a large-break LOCA, and is therefore not included in the design specifications for fatigue analysis.

LHSI actuation does not otherwise occur independent of other normal and upset transients.

Transients 20(b), "Reactor Trip and Cooldown with Safety Injection," and 22, "Inadvertent RCS Depressurization," both result only in HHSI actuation.

Shutdown residual heat removal (RHR) uses the LHSI pumps and may include connection to the primary system at temperatures as high as 350 TF, but these operations are included in the heatup, cooldown, and refueling transients.

5 The design specification includes 10 COMS actuation events of 600 relief valve operating cycles each.6 Not stated or not separately stated; no value reported in the periodic fatigue monitoring summary reports.7 20 earthquakes of 10 cycles each.8 10 events documented in the original design transient specification and in USAR Table 3.9(N)-I.

As a monitoring limit, five is conservative.

9 In each steam generator.

Monitored Locations Table 4.3-2 lists the locations monitored by the present WCGS fatigue management program. These locations were chosen to represent limiting usage factor locations in the Class 1 components and piping systems. With the exception of one location that has been validated for a 60-year life, the WCGS fatigue management program monitors all of the"NUREG\CR-6260" sample locations to be addressed at WCGS, as described in Section 4.3.4, "Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components." (The two charging nozzles, Numbers 5 and 6 in Table 4.3-2, count as a single NUREG\CR-6260 location in Section 4.3.4. They are identified as "Charging Nozzles" in Table 4.3-5 of Section 4.3.4.)Wolf Creek Generating Station Page 4.3-7 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Table 4.3-2 Limiting Locations Tracked by the WCGS Fatigue Management Program Locations Tracked with Cycle-Based Monitoring

1. Residual Heat Removal Cold Leg Return -Accumulator Safety Injection (ACCSI) Nozzles (4, 10-inch, 45-degree, tracked as one)(NUREG\CR-6260 location)2. High Head Safety Injection Boron Injection Tank (BIT HHSI) Cold Leg Nozzles (4, 3-inch, tracked as one)('1 (NUREG\CR-6260 location)3. Reactor Vessel Inlet Nozzles (NUREG\CR-6260 location)(4, tracked as one)4. Reactor Vessel Outlet Nozzles (NUREG\CR-6260 location)(4, tracked as one)Locations Tracked with Stress-Based Monitoring

5. Alternate Chemical and Volume Control System (CVCS) Charging Nozzle (Loop 4)(NUREG\CR-6260 location)6. Normal CVCS Charging Nozzle (Loop 1)(NUREG\CR-6260 location)7. Hot Leg Surge Line Nozzle (Including Thermal Stratification Effects) (NUREG\CR-6260 location)8. Pressurizer Lower Head 9. Pressurizer Heater Penetrations (all penetrations, tracked as one)1 OA. Stm Generator A FW Nozzle Loc. 1 Stm Generator A FW Nozzle Loc. 2 Stm Generator A FW Nozzle Loc. 3 Stm Generator A FW Nozzle Loc. 4 1OB. Stm Generator B FW Nozzle Loc. 1 Stm Generator B FW Nozzle Loc. 2 Stm Generator B FW Nozzle Loc. 3 Stm Generator B FW Nozzle Loc. 4 Wolf Creek Generating Station Page 4.3-8 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Table 4.3-2 Limiting Locations Tracked by the WCGS Fatigue Management Program 10C. Stm Generator C FW Nozzle Loc. 1 Stm Generator C FW Nozzle Loc. 2 Stm Generator C FW Nozzle Loc. 3 Stm Generator C FW Nozzle Loc. 4 10D. Stm Generator D FW Nozzle Loc. 1 Stm Generator D FW Nozzle Loc. 2 Stm Generator D FW Nozzle Loc. 3 Stm Generator D FW Nozzle Loc. 4 11. Pressurizer Spray Nozzle 12. Pressurizer Surge Line Nozzle (Including Thermal Stratification Effects)13. Pressurizer Surge Line (Including Thermal Stratification Effects)Notes to Table' From the Boron Injection Tank (BIT).4.3.1.3 Program Enhancements for the Period of Extended Operation The WCGS fatigue management program will be enhanced principally by including additional components to be monitored and by incorporating additional action limits and corrective action administrative controls.Scope The program monitors a representative set of locations within existing ASME Section III Class 1 vessel and piping fatigue analyses.

This set includes the NUREG/CR-6260 sample locations discussed in Section 4.3.4, "Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components," except the reactor vessel lower head to shell juncture, for which the CUF calculation including environmental effects has been successfully validated for a 60-year life. Monitoring this set of locations will ensure that fatigue in any other locations of concern, not included in the set, are within the same system and subject to the same transients, or within a system affected by the same transients.

Methods The fatigue usage factor at the monitored locations is tracked by one of two methods: Wolf Creek Generating Station Page 4.3-9 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES For the period of extended operation the WCGS fatigue monitoring program will use cycle-count-based monitoring for the first four locations listed in Table 4.3-2. These four locations are included among the six sample locations that will be monitored for the additional effect of the reactor coolant environment on fatigue usage, as discussed in Section 4.3.4, "Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components." Cycle-based monitoring assumes the alternating stress range of every cycle of a transient is equal to that of the design basis, worst-case events assumed by the code fatigue analysis.Accumulated fatigue usage is then the sum of the number of transient cycles times the per-cycle, design basis fatigue usage of each. The method uses event pairing methods similar to those of the ASME code to define transient stress ranges between events.For the period of extended operation the WCGS fatigue monitoring program will use stress-based monitoring for the remaining 12 locations in Table 4.3-2 (24 when the steam generator feedwater nozzle locations are counted separately).

These 12 locations include the two remaining locations monitored for the additional effect of the reactor coolant environment on fatigue usage (the hot leg nozzle connecting to the surge line, and the two charging nozzles).Stress-based monitoring uses actual plant transient profile data to determine the alternating stress range of monitored cycles between event pairs, from recorded pressure, temperature, flow, and rate-of-change data; using models based on the code fatigue analysis.

The stress range is determined from pairs of events as they actually occur. Fatigue usage accumulation is then calculated from this stress range, for each cycle.However, the WCGS fatigue monitoring program will use only cycle-based results for the locations monitored by the stress-based method, unless stress-based results are required to demonstrate acceptable fatigue usage.Enhanced Corrective Action Limits and Corrective Actions The WCGS fatigue management program provides for periodic evaluation (once per operating cycle) of fatigue usage and cycle count tracking of critical thermal and pressure transients to verify that the ASME Code CUF limit of 1.0 and other CUF design limits will not be exceeded.The program will be enhanced to specify corrective actions to be implemented to ensure that design limits are not exceeded.

These enhancements will include action limits for accrued transient cycles or CUF that require initiation of corrective actions, allowing sufficient time to effectively address the issues.Cycle Count Action Limit and Corrective Actions An action limit will be established that requires corrective action when the cycle count for any of the critical thermal and pressure transients is projected to reach a high percentage Wolf Creek Generating Station Page 4.3-10 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES (e.g., 90%) of the design-specified number of cycles before the end of the next operating cycle.If this action limit is reached, acceptable corrective actions include: 1. Review of fatigue usage calculations" To determine whether the transient in question contributes significantly to CUE." To identify the components and analyses affected by the transient in question.* To ensure that the analytical bases of the leak-before-break (LBB) fatigue crack propagation analysis and of the high-energy line break (HELB) locations are maintained.

2. Evaluation of remaining margins on CUF based on cycle-based or stress-based CUF calculations using the WCGS fatigue management program software.3. Redefinition of the specified number of cycles (e.g., by reducing specified numbers of cycles for other transients and using the margin to increase the allowed number of cycles for the transient that is approaching its specified number of cycles).Cumulative Fatigue Usage Action Limit and Corrective Actions An action limit will be established that requires corrective action when calculated CUF (from cycle based or stress based monitoring) for any monitored location is projected to reach 1.0 within the next 2 or 3 operating cycles.For WCGS locations identified in NUREG/CR-6260 and described in Section 4.3.4, "Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components," this action limit will be based on accrued fatigue usage calculated with the FEN factors required for including effects of the reactor coolant environment.

If this action limit is reached acceptable corrective actions include: 1. Determine whether the scope of the monitoring program must be enlarged to include additional affected reactor coolant pressure boundary locations.

This determination will ensure that other locations do not approach design limits without an appropriate action.2. Enhance fatigue monitoring to confirm continued conformance to the code limit.3. Repair the component.

4. Replace the component.

Wolf Creek Generating Station Page 4.3-11 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES 5. Perform a more rigorous analysis of the component to demonstrate that the design code limit will not be exceeded.6. Modify plant operating practices to reduce the fatigue usage accumulation rate.7. Perform a flaw tolerance evaluation and impose component-specific inspections.

These corrective actions are equally applicable to the WCGS NUREG/CR-6260 locations described in Section 4.3.4, "Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components," including consideration of the effects of the reactor coolant environment.

The enhancements described in this section will be required to address fatigue TLAAs in the period of extended operation.

Wolf Creek Nuclear Operating Corporation will complete these program enhancements before the end of the current licensed operating period.Changes in available monitoring technology or in the analyses themselves may by that time permit different action limits and action statements, or may otherwise change the program features and actions required to address the fatigue TLAAs.Wolf Creek Generating Station Page 4.3-12 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES 4.3.2 ASME Section III Class 1 Fatigue Analysis of Vessels, Piping, and Components Fatigue analyses are performed for ASME Section III Division 1 Class 1 piping, vessels, heat exchangers, pumps, and valves; and if applicable, their supports.

Table 4.3-3 lists all Class 1 vessels, heat exchangers, pumps, piping and subcomponents subject to Class 1 analyses, and the subsection of this chapter which addresses each component.

The reactor vessel internals are not designed to ASME Section III Class 1 but are analyzed to ASME Section III Subsection NG. See Section 4.3.3.Table 4.3-3 WCGS Class I Components and Piping Component WCGS Number(1) Subsection Reactor Pressure Vessel, Head, Studs, Shoes and Shims, and BBRBB01 4.3.2.1 Supports Control Rod Drive Motor (CRDM) Housings [BB-RBB01]

4.3.2.2 CRDM Head Adapter Plugs [BB-RBB01]

4.3.2.2 CRDM Seismic Support Platform, Spacer Plates, and Tie Rods No fatigue analysis Reactor Coolant Pump Casings, Supports, Main Flanges and BB-PBB01A, Thermal Barrier Heat Exchangers B, C, D Pressurizer BB-TBB03 4.3.2.4 Steam Generators (Primary or Tube Side and Shell Side)(2) BB-EBB01A, 4.3.2.5 B,C, D Pressure-Retaining Bolting (Included with the Reactor Vessel, Steam Generators, Reactor Coolant Pumps, Pressurizer, and -As noted Valves, as Applicable)

Valves See list in 4.3.2.6 4.3.2.6 Piping -4.3.2.7 Main Reactor Coolant Loop Piping Nozzles and Thermowells

-4.3.2.7 Wolf Creek Generating Station License Renewal Application Page 4.3-13 Section 4 TIME-LIMITED AGING ANALYSES Table 4.3-3 WCGS Class I Components and Piping ComonntWCGS CompOnent Number(1) Subsection Supports for Class 1 Piping and Valves No fatigue analysis Notes to Table 1 Brackets indicate a subcomponent.

2 The steam generator shell (steam) side is Class 2 but also received a Class 1 analysis.4.3.2.1 Reactor Pressure Vessel, Nozzles, Head, and Studs Summary Description The WCGS reactor pressure vessel (RPV) is designed to ASME Section III, Subsection NB (Class 1), 1971 Edition with addenda through Winter 1972.Pressure-retaining and support components of the reactor pressure vessel are subject to an ASME Boiler and Pressure Vessel Code, Division 1,Section III, fatigue analysis.

This analysis has been updated to incorporate redefinitions of loads and design basis events, operating changes, power rerate, and minor modifications.

The currently-applicable fatigue analyses of these components are TLAAs.See Section 4.3.2.9, "Primary Coolant System Heatup Expansion Noise Events," for the evaluation of certain noise events affecting the fatigue analyses of the primary coolant system and RPVI.Analysis Effects of Power Rerate, Thot Reduction, and up to 10 Percent Steam Generator Tube Plugging on the Vessel Fatigue Analysis The WCGS power rerate modification included evaluation of a proposed reduction in normal operating hot leg temperature (Thot reduction) and operation with up to 10 percent of steam generator tubes plugged. The code design report revision included review of operating conditions, including specified transient definitions, for the original power rating, rerated power, Thot reduction, and maximum and minimum tube plugging, to determine the most limiting parameters.

Stresses and fatigue usage were calculated for these most limiting parameters.

The calculated design stresses and fatigue usage factors in the revised design Wolf Creek Generating Station Page 4.3-14 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES report therefore bound all operating conditions, at up to rerated power, with or without Thot reduction, and for any level of steam generator tube plugging up to 10 percent.The maximum effect of power rerate on RPV fatigue was an increase in cumulative fatigue usage factor for the core support lugs of only 0.001.Effects of Cold Overpressure Mitigation System (COMS) Transients Evaluation of the COMS transients determined that an increase in design basis usage factor of 0.06 would bound fatigue effects of these cold overpressure events at all locations in the RPV.Effects on the Fatigue Analysis of the Expanded Stud Elongation Tolerance Band (Reduced-Pass Tensioning)

An amendment to the RPV code design report supported an increase in the RPV head stud elongation tolerance band. The analysis found that the calculated 40-year cumulative usage factor in the studs increased from 0.478 to 0.610, or to 0.670 with the conservative allowance for COMS transients.

Effects on the Fatigue Analysis of the Vessel, Head, and Studs of Operating with One Stud Detensioned An addendum to the RPV code design report supported operation for a single cycle with one RPV head stud detensioned.

This is a contingency because of difficulties encountered in removing some of the studs from the vessel flange during one refueling outage. The threads in some of the stud holes were subsequently repaired to alleviate the problem. The analysis shows operation for one cycle with a stud detensioned has a negligible effect on fatigue usage.WCGS has never operated with a stud detensioned.

However this analysis remains valid and could be invoked if needed.Effects on the Head Lifting Lug Fatigue Analysis from the Tensioner Clearance and Simplified Head Assembly (SHA) Modifications, COMS Transients, and Noise Events The outside corners of the lifting lugs were trimmed off to provide clearance for the head tensioner, and the SHA modification increased the mass of the lifting assembly attached to the lifting lugs. These modification increased lifting stresses and the design basis loads on the lifting lugs during the 400 OBE cycles assumed for the fatigue analysis.Evaluation of these modifications plus effects of the COMS and acoustic event transients (see Section 4.3.2.9, "Primary Coolant System Heatup Expansion Noise Events") found that the stresses remain well within allowables and that the design basis maximum usage factor in the lifting lug holes is 0.31.Wolf Creek Generating Station Page 4.3-15 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Effects on the Head and Lower Penetration Fatigue Analysis of the DMIMS Accelerometer Installation An evaluation of accelerometer installations on the lifting lugs and on lower head instrument penetrations for the Digital Metal Impact Monitoring System (DMIMS, also known as the loose parts monitoring system) reported margins in allowable stresses, and estimated increases in those stresses.

Since increases in fatigue usage factors were determined to be insignificant, numerical values were not calculated.

Summary of Analyses The limiting components for fatigue in the RPV pressure boundary and its supports are the bottom-mounted instrument tubes and the inlet and outlet nozzles. The design basis cumulative usage factors in these components do not exceed 0.4. Other components of the reactor pressure vessel pressure boundary and its supports are affected by similar loads and transients.

The closure flanges, studs, and nuts are also subject to boltup cycles.However the maximum 40-year usage factor in the head and flanges is only about 0.08, including the COMS transient allowance.

The maximum 40-year usage factor in the studs including effects of an increased elongation tolerance and the conservative COMS transient allowance, is 0.672.The RPV primary coolant inlet and outlet nozzles and lower-head-to-shell juncture are evaluated for effects of the reactor coolant environment on fatigue behavior of these materials, consistent with NUREG/CR-06260.

See Section 4.3.4, "Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components." Disposition:

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

Fatigue usage factors in the reactor vessel pressure boundary do not depend on effects that are time-dependent at steady-state conditions, but depend only on effects of operational, abnormal, and upset transient events, principally on startup and shutdown transients and on RPV head flange boltup.The WCGS fatigue management program will ensure that the fatigue analyses remain valid, or that appropriate reevaluation or other corrective measures maintain the design and licensing basis. Therefore, effects of fatigue in the RPV pressure boundary and its supports will be managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii).

4.3.2.2 Control Rod Drive Mechanism (CRDM) Pressure Housings, Adapter Plugs, and Canopy Seals Summary Description The WCGS CRDM housings are designed to ASME Section III, Subsection NB (Class 1), 1974 Edition with addenda through Winter 1974.Wolf Creek Generating Station Page 4.3-16 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Pressure-retaining components of the reactor control rod drive mechanisms

-the CRDM housings and adapter plugs, as well as the added canopy seal clamp assemblies

-are ASME Section III Class 1 components and have a Class 1 fatigue analysis.

The analysis was reexamined for the power rerate, Thot reduction modification, for 10 percent of steam generator tubes plugging and for addition of canopy seal clamp assemblies.

Analysis The highest calculated fatigue usage factor in the CRDM housings is 0.1093, significantly less than 1.0.Effects of Power Rerate, Thot Reduction, and up to 10 Percent Steam Generator Tube Plugging on the CRDM Adaptor Plug Analysis and CRDM Nozzle and Housing Analysis The WCGS power rerate modification included evaluation of a proposed reduction in normal operating hot leg temperature and operation with up to 10 percent of steam generator tubes plugged. Power rerate had no effect on the number of transient cycles nor on the resulting stress ranges in the CRDM adapter plugs, and therefore no effect on the associated fatigue analysis.

It similarly had only a diminishing effect on the thermal analysis of other CRDM pressure boundary components.

Therefore the original fatigue analysis of these components remained valid.Canopy Seal Clamp Assembly Modifications WCGS experienced leakage at canopy seal welds on spare CRDM penetrations.

The leaking welds were encapsulated by canopy seal clamp assemblies.

The modification also included evaluation of similar seal weld clamp assemblies for spare core exit thermocouple penetrations.

The clamp assemblies are qualified for a 40-year design life.The code analysis for this modification examined the worst-case loads and fatigue usage factors for core exit thermocouple, active CRDM, and spare nozzles. The fatigue evaluation for this modification confirmed the previously-calculated maximum usage factor of 0.1093, at the worst-case CRDM housing to head weld.Disposition:

Validation, 10 CFR 54.21 (c)(1)(i)The maximum calculated usage factor in the CRDM pressure housings indicates that the design is adequate for nine times the number of specified design transient events. The evaluation of fatigue effects in the CRDM pressure housings will therefore remain valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i).

Wolf Creek Generating Station Page 4.3-17 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES 4.3.2.3 Reactor Coolant Pump Pressure Boundary Components Summary Description The pump pressure boundary was designed to ASME Section III, 1971 edition with addenda through Summer 1973. Subarticle NB-3400 of this edition and addenda, "Design of Class 1 Pumps," does not require a fatigue analysis, but the vendor specified a fatigue analysis.Subparagraph NB-3222.4(d) includes provision for waiver of the fatigue analysis if limiting criteria are met. The original analysis included a Subparagraph NB-3222.4(d) fatigue waiver analysis, which demonstrated that no pump pressure boundary components required a fatigue analysis.

However the vendor elected to perform conservative, simplified fatigue analyses for a number of components, and subsequent load definition changes eventually resulted in fatigue waivers or analyses applicable to a number of additional locations in the pump pressure boundary.

These fatigue and fatigue waiver analyses have been updated to incorporate redefinitions of loads and design basis events, operating changes, power rerate, and minor modifications.

The fatigue analyses are TLAAs. The fatigue waiver analyses are also TLAAs because they depend in part on the assumed numbers of design basis normal and upset transient cycles.The WCGS reactor coolant pump casings and their support feet are one-piece castings.The casing therefore requires no internal weld inspections, and Code Case N-481 inspection relief is not applicable.

Analysis A review of the code design reports finds that a Subparagraph NB-3222.4(d) fatigue waiver still applies to many components.

Confirming conservative fatigue analyses were applied to some of the components for which a fatigue waiver applies. The maximum design basis calculated cumulative usage factor for all components is less than 0.82. Many of the analyses are conservative.

Since few of the components experience significant alternating stress, the alternating stress ranges used for the fatigue analyses were taken from zero to the limit stress for the particular transient, resulting in overly-conservative stress ranges.The most significant contributors to usage factor in all locations are startup and shutdown cycles.Disposition:

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

Fatigue usage factors in the reactor coolant pumps do not depend on effects that are time-dependent at steady-state conditions, but depend only on effects of operational, abnormal, and upset transient events, principally on startup and shutdown transients.

The WCGS fatigue management program will track events to ensure either that the code design analyses remain valid, or that appropriate reevaluation or other corrective action is taken if a design basis number of events is exceeded, or if usage factors approach the limit Wolf Creek Generating Station Page 4.3-18 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES of 1.0. Therefore, effects of fatigue in the reactor coolant pump pressure boundaries will be managed for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(iii).

4.3.2.4 Pressurizer and Pressurizer Nozzles Summary Description The WCGS pressurizer is designed to ASME Section III, Subsection NB (Class 1), 1974 Edition.Pressure-retaining and support components of the pressurizer are subject to an ASME Boiler and Pressure Vessel Code, Division 1,Section III, fatigue analysis.

This analysis has been updated from time to time to incorporate redefinitions of loads and design basis events, operating changes, power rerate, and minor modifications; including the effects of thermal stratification and insurge-outsurge transients not included in the original analyses.The currently-applicable fatigue analyses of the pressurizer components are TLAAs.The pressurizer and its nozzles are the subject of a Westinghouse Owners Group (WOG)License Renewal Topical Report, WCAP-14574-A, "License Renewal Evaluation:

Aging Management Evaluation for Pressurizers." Analysis Effects of Power Rerate, Thot Reduction, and up to 10 Percent Steam Generator Tube Plugging on the Pressurizer Fatigue Analysis The WCGS power rerate modification included evaluation of a proposed reduction in normal operating hot leg temperature (Thot reduction) and operation with up to 10 percent of steam generator tubes plugged. The code design report revision included review of operating conditions, including specified transient definitions, for the original power rating, rerated power, Thot reduction, and maximum and minimum steam generator tube plugging, to determine the most limiting parameters.

Stresses and fatigue usage were calculated for these limiting parameters.

Therefore the calculated design stresses and fatigue usage factors in the revised design report bound all operating conditions, at up to rerated power, with or without Thot reduction, and for any level of steam generator tube plugging up to 10 percent. The evaluation required no revision to the pressurizer fatigue analysis.Effects of NRC Bulletin 88-11 Thermal Stratification Transients The current code design report includes a revision to the maximum usage factor in the surge nozzle, including effects of NRC Bulletin 88-11 thermal stratification.

For related thermal stratification effects in the surge line, see Section 4.3.2.8, "Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and Stratification." Wolf Creek Generating Station Page 4.3-19 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Insurge-Outsurge Transients Insurge-outsurge events during startup or shutdown can introduce cooler water through the pressurizer surge line into the pressurizer, against the wall previously heated by hot pressurizer water. This causes a significant thermal gradient in the pressurizer wall.Surge effects in the pressurizer are mitigated by Technical Specification heatup and cooldown rate limits; and by the use of continuous spray during heatup and cooldown transients.

Continuous spray maintains a small flow from the pressurizer to the hot leg during heatup and cooldown, which maintains a uniform fluid temperature below the pressurizer heaters and in the upper portion of the surge line to minimize thermal stratification.

The heatup and cooldown rate operating limits and the use of continuous spray flow were instituted in 1993, and have been very effective in reducing fatigue usage accumulation in the lower pressurizer components since that time. Based on this experience, WCGS has concluded that a generic Westinghouse analysis of fatigue usage, including insurge/outsurge transient effects (Reference 4.9.18), is conservative for WCGS for 60 years of operation.

Effect of a Pinned Support on the Relief Line to BB-V-8010C A pinned support in the discharge line from this pressurizer code relief valve had been recognized in 1987, and the effect of thermal cycles under this pinned condition on the pressurizer nozzle usage factor was calculated and included in the fatigue analysis and in the pressurizer design report. The pinned support also produced a plastic displacement in the relief valve line. An evaluation of the effect of the plastic displacement on the code fatigue analysis of the line found a small effect on usage factor in the line. The evaluation determined, by comparison that the additional effect on the nozzle was negligible

(<0.001).Therefore no change to the pressurizer fatigue analysis was initiated.

Effect of a Pressurizer-Surge-Nozzle-to-Safe-End Weld, Safe End, and Safe-End-to-Surge-Line Weld Overlay A weld overlay was installed over the surge-nozzle-to-safe-end weld, safe end, and safe end to pipe weld during Refuel 15. The maximum fatigue usage in the surge nozzle is at a location remote from this overlay and is unaffected by the increased thickness of the overlay.Summary of Analyses With the design basis set of transients, including the power rerate and Thot modification and other effects described above, worst-case fatigue usage factors for the present design exceed 0.9 at three pressurizer locations.

Although the pressurizer surge and spray nozzles and pressurizer lower head are subject to significant operating thermal cycles from thermal stratification and insurge-outsurge transients not considered in the original code analysis, operating procedure changes have Wolf Creek Generating Station Page 4.3-20 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES minimized these transients, and updated analyses confirm that fatigue usage factors in the affected pressurizer components will be within acceptable limits for the originally-specified design transient events, plus the number of these additional transient events expected for an operating life of 60 years.Disposition:

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

The WCGS fatigue management program will track events to ensure either that the number of assumed events are sufficient and the usage factors are not exceeded, or that appropriate reevaluation or other corrective measures maintain the design and licensing basis.The pressurizer surge nozzle, spray nozzle, and lower head may be subject to significant operating thermal stress cycles due to thermal stratification and insurge-outsurge cycles, and are therefore expected to be the limiting pressurizer components for fatigue. As a result the fatigue usage factors of these locations are specifically monitored.

Therefore the effects of fatigue in the pressurizer Class 1 pressure boundary and supports will be managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii).

4.3.2.5 Steam Generator ASME Section III Class 1, Class 2 Secondary Side, and Feedwater Nozzle Fatigue Analyses Summary Description The steam generators are designed to ASME Section III, Subsection NB (Class 1) and Subsection NC (Class 2), 1971 Edition with addenda through Summer 1973.Pressure-retaining and support components of the primary coolant side of the steam generators are subject to an ASME Boiler and Pressure Vessel Code, Division 1,Section III fatigue analysis.

Although the secondary side is Class 2, pressure retaining parts of the steam generator satisfy the Class 1 criteria, including the Class 2 secondary side boundaries.

These analyses have been updated to incorporate redefinitions of loads and design basis events, operating changes, power rerate, primary loop Thot reduction, and minor modifications.

The currently-applicable fatigue analyses of these components are TLAAs.There are also some fatigue evaluations of feedwater lines and of the auxiliary feedwater tees. However, none of these evaluations have produced licensing basis commitments or safety determinations supported by fatigue analyses.

Therefore the fatigue evaluations associated with feedwater lines and auxiliary feedwater tees are not TLAAs.Wolf Creek Generating Station Page 4.3-21 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Analysis Steam Generator Tube Code Fatigue Analysis Not a TLAA The design of the steam generators includes a code fatigue analysis of the steam generator tubes. This analysis would be a TLAA if the safety determination depended upon it.However the code fatigue analysis has not proved sufficient to support the safety determination.

The various tube degradation mechanisms not anticipated in the original design have required stringent periodic inspection programs in order to ensure adequate steam generator tube integrity.

The fatigue analysis is therefore no longer the basis of the safety determination that the tubes will maintain their pressure boundary function (Criterion 5).Therefore, even in installations (such as WCGS) with excellent material and chemistry control, the safety determination for integrity of steam generator tubes now depends on managing aging effects by a periodic inspection program rather than on the fatigue analysis.Therefore the code fatigue analysis of the tubes is not a TLAA.Steam Generator Fatigue Analysis Including Effects of Power Rerate, Thot Reduction, and up to 15 Percent Steam Generator Tube Plugging The WCGS power rerate modification included evaluation of a proposed reduction in normal operating hot leg temperature (Thot reduction) and operation with up to 10 percent of steam generator tubes plugged. The current analysis encompasses these design limits, including up to 15 percent plugged tubes.The current steam generator code design report reflects refinements in the analyses of some components and resulting reductions in calculated usage factors, including qualification of the primary manway studs and secondary closure bolting by test. It also includes effects of up to 15 percent steam generator tube plugging, revised design basis transients, revised seismic spectra, and feed line acoustic pressure pulse transients.

Therefore the calculated design stresses and fatigue usage factors in the revised design report bound known operating conditions, at up to rerated power, with or without Thot reduction, and for any level of tube plugging up to 15 percent.With power rerate and the Thot modification the worst-case usage factors calculated for the specified set of design basis transients exceed 0.9 in two steam generator locations.

However, excepting the tubes (for which the safety determination depends on managing aging effects by a periodic inspection program rather than on the fatigue analysis), fatigue usage factors in the steam generator components do not depend on flow-induced vibration or other effects that are time-dependent at steady-state conditions, but depend only on effects of operational and upset transient events. The WCGS fatigue management program tracks these operational and upset events to ensure that the design basis number of them is not exceeded without an appropriate evaluation and any necessary mitigating actions.Wolf Creek Generating Station Page 4.3-22 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Primary Manway Studs, with Power Rerate and Thot Reduction The primary manway bolts were replaced by bored studs to permit hydraulic tensioning and measurement of preload. The replacement studs met code stress criteria, but high calculated usage factors would have required their periodic replacement, at the rate of transient cycle accumulation implied by the original 40-year design life. The studs and nuts were qualified by test, with a sufficient number of test cycles to envelope the entire set of design basis transients.

Secondary Manway Bolts, Handhole (Inspection Port) Bolts, and Instrument Opening Bolts, with Power Rerate and Thot Reduction The secondary bolting is qualified for fatigue effects by analyses that apply the results of the primary manway stud tests.The high calculated usage factor in bolting for these openings originally required their periodic replacement, as determined by the rate of transient cycle accumulation implied by the original 40-year design life. The increased rate of accumulation of fatigue usage factor with rerate and Thot reduction reduced the secondary manway bolt replacement interval from 20 to 18 years. Since the bolting replacement interval was less than the design life, its basis was not, at that point, a TLAA.However, the current code design report extends the primary manway bolting fatigue qualification tests to the secondary side bolting, and the basis of the safety determination for this bolting is qualification by test. The secondary-side steam generator pressure boundary bolts are no longer periodically replaced, therefore the application of the primary bolting fatigue test as the basis for secondary side bolting qualification for fatigue effects is now a TLAA. If the number of load cycles assumed by the evaluation, used in the fatigue test, is not exceeded, the qualification basis will remain valid.Stub Barrels and Channel Heads Drilled and Tapped for DMIMS-DX Loose Parts Monitor Accelerometer Mountings A 2004 addendum to the code stress report includes effects of tapping 1/4" -28 UNF2B holes in the stub barrel and channel head of each steam generator for mounting digital metal impact monitoring system (DMIMS) accelerometers.

If the number of load cycles assumed by the fatigue analysis is not exceeded, the predicted usage factor will remain within the allowable of 1.0.No TLAA in the Finite Element Analysis in Support of Feedwater Nozzle Thermal Stratification Transfer Functions for Stress-Based Fatigue Monitoring NRC Information Notices 91-38 "Thermal Stratification in Feedwater System Piping" and 93-20 "Thermal Fatigue Cracking of Feedwater Piping to Steam Generators" identified concerns with thermal stratification in feedwater piping and nozzles. The WCNOC resolution of this problem included thermal monitoring over several operating cycles to Wolf Creek Generating Station Page 4.3-23 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES assess the severity of the concern. Analysis of these effects indicated more-rapid-than-design accumulation of fatigue usage factor, which then prompted plant operational changes and addition of a startup feedwater heating system.A finite element analysis identified the limiting locations, and provided scaling for global-to-local transfer functions to permit stress-based fatigue monitoring at the limiting locations.

However, thermal monitoring following the startup feedwater heating addition and operational changes did not indicate any significant alteration of the code analysis results.Therefore the code analysis does not reflect these effects. Since the finite element analysis used to develop the fatigue transfer functions has not been incorporated into the code stress analysis, the finite element analysis is not a TLAA.Repair of Primary Chamber Drains The 2005 refueling outage inspections found cracked welds at the connection of the C and D steam generator primary drains to the lower heads. The welds and the drain couplings in all four steam generators were excavated and replaced, and the steam generator analysis was amended. However the limiting usage factor at these drains is inside the heads (locations unaffected by the repair), and remains limiting at this location in the steam generator heads.Disposition:

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

The fatigue analysis of the steam generator tubes does not support the safety determination and is therefore not a TLAA. Fatigue usage factors in other steam generator pressure boundary and Class 1 support components, and qualification of the primary manway studs by test, do not depend on effects that are time-dependent at steady-state conditions, but depend only on effects of operational, abnormal, and upset transient events.Manway, Handhole, and Instrument Opening Bolting -Possible Requalification or Replacement Appropriate corrective measures, which may include requalification or replacement, will ensure that the design basis of the bolting is maintained if fatigue monitoring indicates that the numbers of load cycles assumed by the qualification by test may be exceeded.All Components The WCGS fatigue management program will track events to ensure that appropriate reevaluation or other corrective action is taken if a design basis number of events is exceeded, and will maintain a current record of cumulative usage factor for each monitored location.Therefore, effects of fatigue in the steam generator pressure boundaries and their supports will be managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii).

Wolf Creek Generating Station Page 4.3-24 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES 4.3.2.6 ASME Section III Class I Valves Summary Description WCGS Class 1 valves (power-operated relief valves, pressurizer safety valves, control valves, motor- and air-operated valves, manual valves, and check valves) are designed to ASME Section III, Subsection NB, 1974 Edition and later addenda.A review of WCGS Class 1 valve analyses and specifications found that fatigue analyses and possible TLAAs were performed only for the 6-inch pressurizer safety valves, for Class 1 check and gate valves over 4 inches nominal, and for one model of 1 1/2 inch angle globe valves.However, the fatigue analysis for the 1 1/2 inch angle globe valves was not required by code or specification, is not discussed in any licensing basis document, and was therefore not the basis for a safety determination.

Therefore the the 1 1/2 inch angle globe valve fatigue analysis is not a TLAA. With that exception, no fatigue analyses were applied to any valves of four inches nominal inlet or less. Therefore no Class 1 fatigue analysis TLAAs support design of valves with inlets four inches or less. Conversely, fatigue analyses were applied to Class 1 valves with inlets greater than four inches, and all fatigue analyses of Class 1 valves greater than four inches are TLAAs.Analysis TLAA fatigue analyses or evaluations were performed for the following Class 1 valves: Table 4.3-4 Class I Valves With TLAA Fatigue Analyses Tag Number Description Normal Duty CUF 1,.....________ Ops NA BB8010A, B, C 6" x 6" Pressurizer Safety Valves >106 <0.4 BBPV8702A, B 12" RHR Suction Gate Valves >106 <1.0 EJHV8701A, B BB8949A,B,C,D 6" Swing Check Valves >106 <0.4 EJ8841A,B EP8818A,B,C,D BB8948A,B,C,D 10" Swing Check Valves 820,000 <1.0 EP8956A,B,C,D The allowed NB-3545.3 NA normal duty operations far exceed those expected to occur. The calculated cumulative usage factors It for NB-3550 cyclic loads are less than the code limit of 1.0, and in all but the 12 inch RHR gate valves and 10 inch swing checks It is less than 0.4.Wolf Creek Generating Station Page 4.3-25 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Disposition:

Validation, 10 CFR 54.21(c)(1)(i);

and Aging Management, 10 CFR 54.21(c)(1)(iii)

Validation The calculated worst-case usage factors for Class 1 pressurizer safety valves and for Class 1, 6 inch swing check valves indicate that the designs have large margins, and the pressure boundaries would withstand fatigue effects for at least two of the original design lifetimes.

Therefore the design of these valves for fatigue effects is valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i).

Aging Management The calculated worst-case usage factors for the Class 1, 12 inch RHR suction gate valves and the Class 1, 10 inch check valves exceed 0.4. However, fatigue usage factors in these valves do not depend on effects that are time-dependent at steady-state conditions, but depend only on effects of operational, abnormal, and upset transient events. As discussed in Section 4.3.2.7, "ASME Section III Class 1 Piping and Piping Nozzles," the 40-year design basis number of events should be sufficient for 60 years of operation, of Class 1 piping systems containing valve. Therefore the calculated usage factors should not be exceeded.(The exceptions discussed in Section 4.3.2.7 are the surge line and surge line nozzle, which contain no valves).The WCGS fatigue management program will ensure that calculated usage factors will not be exceeded, or that appropriate corrective action is taken if a design basis number of events is exceeded.

Therefore, effects of fatigue in Class 1 valve pressure boundaries will be managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii).

4.3.2.7 ASME Section III Class 1 Piping and Piping Nozzles Summary Description Class 1 reactor coolant main-loop piping designed and supplied by Westinghouse is designed to ASME Section III, Subsection NB, 1974 edition with addenda through Winter 1975. The main loop piping fatigue analysis was performed to the 1974 edition with addenda through Winter 1975. The fatigue analyses of piping outside the main loop used code addenda through Summer 1979.These analyses have been updated from time to time to incorporate redefinitions of loads and design basis events, operating changes, power rerate, and minor modifications.

The currently-applicable fatigue analyses of these components are TLAAs.For fatigue in the pressurizer surge line see Section 4.3.2.8, "Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and Stratification." Wolf Creek Generating Station Page 4.3-26 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES For the evaluation of certain noise events affecting the fatigue analyses of the primary coolant system and reactor pressure vessel see Section 4.3.2.9, "Primary Coolant System Heatup Expansion Noise Events." The evaluation of these noise events found no effect on the primary coolant piping fatigue analysis.Analysis In the primary coolant system and large-bore emergency core cooling (EGOS) lines the attachment welds to the reactor vessel inlet and outlet nozzles, and the primary coolant system EGGS injection nozzles (Loop 1 and 4 GVCS charging nozzles, BIT (HHSI) nozzles, and accumulator safety injection (ACCSI) nozzles) have the most limiting calculated design basis usage factors. In a number of these nozzles, the analysis of record was refined only to the level necessary to demonstrate a usage factor just under 1.0. The high usage factors in the EGGS injection nozzles are primarily due to transient thermal stresses from normal operating and upset injection events.Calculated design basis usage factors in smaller Class 1 lines also exceed 0.9 at a number of locations, in many cases due to operating transients that specifically affect the location.With the exception of the thermowells and pressurizer surge line nozzle discussed in this section and the pressurizer surge line discussed in Section 4.3.2.8, "Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and Stratification," fatigue usage factors in these components do not depend on effects that are time-dependent at steady-state conditions, but depend only on effects of operational, abnormal, and upset transient events. Since the WGGS fatigue management program will track these events, the design basis fatigue usage factor limit (1.0) will not be exceeded in these locations without an appropriate evaluation and any necessary mitigating actions.Analysis of Supports Only the pressurizer surge line was re-evaluated to a code edition and addendum (1986, no addenda) which in some cases would have required design of the supports for stress limits based on a finite number.of lifetime load cycles. However, the original code of record (1974 W'75, fatigue analysis to 1977 S'79) was the same as that for other Class 1 lines and did not invoke this requirement, and as permitted by code rules, the later edition was not invoked for the support reanalysis.

The supports were analyzed to the 1974 W'75 code of record.See Section 4.3.2.8, "Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and Stratification." Effects of NRC Bulletin 88-11 Thermal Stratification on the Hot Leg Pressurizer Surge Line Nozzle The current code analysis includes this effect. See also Section 4.3.2.8. The WCGS fatigue management program calculates fatigue usage factor in this nozzle.Wolf Creek Generating Station Page 4.3-27 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Effects of Power Rerate, Thot Reduction, and Allowance for Increased Steam Generator Plugging on the Piping Fatigue Analyses Evaluations performed to incorporate the effects of power rerate, Thot reduction (hot leg normal operating temperature reduction), and steam generator tube plugging selected parameters and transient descriptions that envelope worst case conditions for original power rating, rerated power, original Thot, proposed (but not implemented) reduced That, and steam generator tube plugging up to 10 percent. Therefore, the analysis results are conservative for any combination of these conditions.

Charging Lines and Nozzles In 1990 Westinghouse identified concerns with CVCS injection path switching and containment isolation testing practices that might introduce a larger-than-design number of significant thermal transients in these nozzles. WCGS therefore revised operating procedures to ensure that significant injection nozzle thermal cycles are minimized.

RTD Nozzles The RTD piping has been removed and these nozzles will therefore accumulate no significant additional fatigue usage factor. Thermowells have been added at some of these nozzles, as described below.Thermowells added at RTD Nozzles The modification that removed RTD bypass piping added thermowells at the 12 primary loop hot leg RTD scoop nozzles (3 nozzles per hot leg) and at the 4 cold leg RTD nozzles, and capped the 4 return nozzles to the crossover legs. The thermowells were analyzed for fatigue due to flow-induced vibration.

The maximum calculated usage factor for a 40-year life at 75 per cent availability is 0.025, in the cold leg thermowells Fatigue due to these loads is proportional to operating time. The worst-case usage factor can therefore be projected to and validated for a 60-year life. The 90 percent capacity factor now expected requires no more than about 98 percent availability.

Hence the worst usage factor in any of the RTID thermowell locations would be CUF 6 0 = CUF 4 0 x 0.98/0.75 x 60/40 CUF 6 0 = 0.025 x 0.98/0.75 x 60/40 = 0.049, and therefore remains negligible.

Effects of NSAL-94-025 Reactor Coolant Pump Support Column Tilt on Main Loop Piping and Supports Westinghouse identified a concern that reactor coolant pump support column tilt may have an adverse effect on main loop piping thermal stresses during heatup and cooldown Wolf Creek Generating Station Page 4.3-28 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES transients.

The Westinghouse evaluation found a large increase in the crossover and cold leg stresses at the reactor coolant pump, and a significant change in the load of the tilted column; but since original analysis stresses were low the effects on stresses and usage factors would not affect code compliance or the conclusions of the leak-before-break analysis.

See also Section 4.3.2.11, "Fatigue Crack Growth Assessment in Support of a Fracture Mechanics Analysis for the Leak-Before-Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures".

Disposition:

Validation, 10 CFR 54.21(c)(1)(i);

and Aging Management, 10 CFR 54.21(c)(1)(iii)

Validation The fatigue analysis of primary loop thermowells has been validated for the period of extended operation as described above, in accordance with 10 CFR 54.21(c)(1)(i).

Aging Management With the exception of the thermowells and surge line nozzle discussed above and the pressurizer surge line discussed in Section 4.3.2.8, "Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and Stratification," usage factors in Class 1 piping pressure boundaries do not depend on effects that are time-dependent at steady-state conditions, but depend only on effects of operational, abnormal, and upset transient events.The WCGS fatigue management program will ensure either that the original design basis number of events or usage factor is not exceeded, or that appropriate reevaluation or other corrective action is taken.Therefore, effects of fatigue in the Class 1 piping pressure boundary will be managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii).

4.3.2.8 Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and Stratification NRC Bulletin 88-11 "Pressurizer Surge Line Thermal Stratification" requested that licensees"establish and implement a program to confirm pressurizer surge line integrity in view of the occurrence of thermal stratification" and required licensees to "inform the staff of the actions taken to resolve this issue." A similar earlier Bulletin 88-08 "Thermal Stresses in Piping Connected to Reactor Coolant System" requested that licensees review the primary coolant pressure boundary and connected interfaces for possible effects of thermal cycles due to leaking interface valves.WCGS installed temperature monitoring to detect leakage past the auxiliary spray valve.Monitoring has not prompted a revision to the piping analysis.Wolf Creek Generating Station Page 4.3-29 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES See Section 4.3.2.4, "Pressurizer and Nozzles," for effects on the pressurizer surge nozzle.See Section 4.3.2.7, "ASME Section III Class 1 Piping and Piping Nozzles," for effects on the hot leg surge nozzle.Summary Description The original surge line fatigue analysis used code addenda through summer 1979. The surge line design was re-analyzed to the 1986 code in response to the NRC Bulletin 88-11 thermal stratification concerns.

This analysis was later reevaluated for effects of snubber removals.

The results of these analyses have been incorporated into the piping and main-loop nozzle code design reports.Winter 1982 and later code addenda provide stress limits for high-cycle fatigue of Class 1 supports, under Subsubarticle NF-3330. However the re-evaluation of the surge line for NRC Bulletin 88-11 did not retroactively impose these requirements.

Therefore no TLAA arises for design of the supports.Analysis Effects of Thermal Stratification on the Surge Line Piping Fatigue Analysis The maximum calculated CUF at any location in the surge lines, under the current analysis of record, including thermal stratification effects, is less than 0.1.Effects of Power Rerate and Thot Reduction on the Surge Line Piping Fatigue Analysis The evaluation of these modifications found that the resulting changes in temperature ranges have negligible effect on the surge line analysis.Effect of a Pressurizer-Surge-Nozzle-to-Safe-End Weld, Safe End, and Safe-End-to-Surge-Line Weld Overlay A weld overlay was installed over the surge-nozzle-to-safe-end weld, safe end, and safe end to pipe weld during Refuel 15.The fatigue usage of the nozzle-to-safe-end and safe-end to pipe welds are no longer the'basis of a safety determination, because the reliability of these welds will be verified by periodic inspections and by flaw propagation analyses.

The flaw propagation analyses are not TLAAs.Disposition:

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

The surge line is subject to fatigue monitoring.

The WCGS fatigue management program will ensure either that the usage factor remains valid for the period of extended operation or that appropriate corrective measures maintain the design and licensing basis. Therefore, effects of fatigue in the Class 1 surge line will be managed for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(iii).

Wolf Creek Generating Station Page 4.3-30 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES 4.3.2.9 Primary Coolant System Heatup Expansion Noise Events Summary Description Since 1990, abrupt audible events have been heard inside containment at WCGS toward the end of primary system heatups. They have been attributed to an abrupt release of differential expansion energy, originally believed to be at the crossover piping support saddle shims, later found to have probably also occurred between the reactor vessel support pads and shoes, under the vessel main loop nozzles. The evaluation of these effects on the vessel, piping, nozzle, and component fatigue analyses is a TLAA, as is the projection of shakedown effects in a reactor vessel support element.Analysis The driver for these events was modeled as the simultaneous sudden release of compressive loads on the reactor vessel nozzles. The resulting piping, support, and nozzle loads are within allowable limits, with the exception of a local region of the reactor vessel support cooling box, described below.For purposes of evaluating fatigue effects the analysis assumed 330 such heatup noise events would occur in a 60-year design life.Monitoring for the noise occurrence from Refuel 5 through Refuel 14 has detected 16 noise events, or about 1.6 per refueling cycle. Assuming a total of 30 cycles have already been expended, 300 cycles remain for the remaining life of the plant. Plant life from the present to the end of the extended operating period is an additional 40 years, or about an additional 27 refuelings at the present 18-month cycle. Up to 11 noise events could therefore occur per refueling cycle, at the magnitude assumed in the fatigue calculation, and still remain within the limits assumed by the fatigue calculation.

Reactor Pressure Vessel Structural Analysis The increase in cumulative usage factor (CUF) in the affected inlet and outlet nozzle-to-shell junctures was calculated by combining the effects of 330 such peak stress range events with the appropriate ranges of related events from the original vessel fatigue analysis, under the Code stress range combination rules for fatigue. The resulting total CUFs are nominal, about 0.11 for the inlet nozzles, 0.18 for the outlet nozzles.Reactor Coolant Loop Piping Analysis Resulting stresses in piping are much less than the endurance limit and the resulting moment stress ranges at piping nozzle welds are less than the T-Z limit. Therefore, the events have no effect on fatigue usage nor on the conclusions of the piping analysis.Wolf Creek Generating Station Page 4.3-31 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Reactor Coolant Loop Leak-Before-Break (LBB) Analysis The recommended LBB margins are maintained.

Therefore the conclusions of the LBB analysis remain valid.Reactor Coolant Loop Primary Component Supports Evaluation All of the primary equipment supports were qualified for normal and upset allowables for the sudden-release loads of the noise event.Reactor Vessel Support Cooling Box Evaluation The support cooling box is not a pressure boundary component.

A local region of the cooling box, bearing on the imbedded steel, may have exceeded yield, but shakedown is occurring or has occurred and the support is and will remain stable. This was confirmed by an elastic-plastic shakedown analysis, and is indicated by the fact that the event occurred at increasingly higher temperatures with successive heatups, and finally at the same temperature for the last three heatups prior to the May 2001 report date.Steam Generator Primary Nozzle Evaluation The effect of these noise events on the steam generator primary nozzle fatigue analysis was evaluated assuming 330 noise events might occur through the end of an extended 60-year licensed operating period. The sum of the added fatigue usage factor due to the noise events, plus the originally-calculated 40-year usage factor, is only 0.063 at the worst location in the primary nozzles. If the originally-calculated 40-year usage factor were multiplied by 1.5 to account for the increased life, the sum of these two values would be only 0.087.Results of the Noise Event Monitoring Program The noise event was first monitored to fulfill a commitment to the NRC, and subsequently for tracking and trending purposes.

The commitment to the NRC has been met. The analysis of data to date indicates no effects on the vessel, piping, or components sufficient to cause a loss of safety function or to invalidate the design basis of a component, and no increase in event severity.Since the original WCGS LRA was filed, WCNOC has conducted a preliminary examination of Refuel 15 monitoring data. These results introduced some uncertainty in the statement, that previously appeared under this subheading in this section, that analysis of data to date indicates "apparent declines" in event severity.

However, the additional data continue to indicate that the event severity remains bounded by earlier instances.

This noise event has been observed since Refuel 5. Indicated severity has not been uniform between occurrences.

This variation is expected due to several factors:* The system operating sequence varies prior to each occurrence.

• Monitoring equipment and methods have changed due to upgrades.Wolf Creek Generating Station Page 4.3-32 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES* Data from some events has been partially lost due to monitoring equipment failures.* Equipment has been modified, notably primary loop restraint changes and snubber removal, and reactor vessel head modification.

All of these factors have contributed to and will continue to contribute to variability in the measured results; and effects of particular changes are not clearly discernable from the data. Thus, correlation of data from the various occurrences has involved considerable uncertainty.

Raw data from Refuel 15 indicate somewhat higher responses than those observed during Refuels 13 and 14, and with these uncertainties, WCGS therefore no longer concludes that there have been "apparent declines" in event severity.

However, even with these uncertainties and the Refuel 15 data, the measured magnitudes and characteristics of these events collected over the period from Refuel 5 through Refuel 15 continue to indicate that effects are very limited and that the characteristics remain consistent, WCNOC therefore concludes that results of previous evaluations remain valid and will continue to remain valid.Effects of Power Rerate and Thot Reduction on the Analysis of Effects of the Noise Events The rerate report found that revised fatigue results "...accounted for both the noise program loadings and the modified design thermal transient conditions of the rerating program," and that other effects of rerate on the analysis of these events are negligible.

Disposition:

Validation, 10 CFR 54.21(c)(1)(i)

Reactor Pressure Vessel, Reactor Coolant Loop, Primary Loop Component Supports, and LBB Analyses The evaluation found that the effect of these events on the reactor pressure vessel and reactor coolant loop and support fatigue analyses, and on the reactor coolant loop LBB analysis, is zero or insignificant for the period of extended operation.

The effect of these events has therefore been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

Reactor Vessel Supports The effect of this event on vessel supports is within normal and upset allowables, with the exception of a local region of the reactor vessel support cooling box. This region is shaking down or has shaken down to a stable response to these events, and will therefore be suitable for the period of extended operation.

The effect of these events has therefore been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).

Wolf Creek Generating Station Page 4.3-33 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Steam Generator Primary Nozzles The effect of these events on the steam generator primary nozzles has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(i).

4.3.2.10 High Energy Line Break Postulation Based on Fatigue Cumulative Usage Factor Summary Description Selection of pipe failure locations for evaluation of the consequences of a high energy line break on nearby essential systems, components, and structures, except for the reactor coolant loop, is in accordance with Regulatory Guide 1.46, and NRC Branch Technical Positions ASB 3-1 and MEB 3-1.A revised stress analysis also permitted omission of the pressurizer surge line intermediate breaks.A leak-before-break analysis (LBB) eliminated large breaks in the main reactor coolant loops. See Section 4.3.2.11, "Fatigue Crack Growth Assessment in Support of a Fracture Mechanics Analysis for the Leak-Before-Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures." Analysis With the stated exception of the reactor coolant system primary loops, the citation of MEB 3-1 means that breaks in piping with ASME Section III Class 1 fatigue analyses are identified based on a limiting stress criterion (a break is assumed if the ASME Section III NB-3653 Equation (10), (12), or (13) stress range between any two load sets is greater than 2.4 SM); and on a cumulative usage factor criterion (a break is assumed if the cumulative usage factor exceeds 0.1). Therefore, the location determinations that depend on usage factor are time-dependent and are TLAAs.The surge line intermediate break locations were eliminated based on usage factor. The most recent piping analysis confirmed the elimination of these break locations.

Therefore, the analysis that justified the elimination of these intermediate locations in the surge line is a TLAA.The same would be true of other line sections with no intermediate locations with fatigue usage factors above 0.1, if this analysis result were used to eliminate intermediate breaks -that is, the determination that there are no intermediate breaks in these sections based on a low usage factor would, for the same reason, be a TLAA. However, no additional cases similar to the surge line occur in the WCGS licensing basis. Therefore, the scope of these HELB-location TLAAs is limited to ASME Section III Class 1 piping analyses of other than the RCS primary coolant loops, including the surge line.Wolf Creek Generating Station Page 4.3-34 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES WCGS has containment penetration break exclusion regions ("no break zones"). However, these contain no ASME Section III Class 1 piping with fatigue analyses.

Therefore, their qualification is based only on calculated stress and the break locations in these no break zones are independent of time and are not supported by a TLAA.Disposition:

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

Break locations which depend on usage factor, and their absence in the surge line, will remain valid as long as the calculated usage factors are not exceeded.The WCGS fatigue aging management program will ensure that the calculated fatigue usage factors upon which the HELB break locations are based, and the HELB locations, will remain valid for the period of extended operation, or that appropriate corrective measures maintain the design and licensing basis.4.3.2.11 Fatigue Crack Growth Assessment in Support of a Fracture Mechanics Analysis for the Leak-Before-Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures Summary Description A leak-before-break analysis eliminated the large breaks in the main reactor coolant loops, which permitted omission of evaluations of their jet and pipe whip effects. This permitted omission of large jet barriers and whip restraints.

The containment pressurization and equipment qualification analyses retained the large-break assumptions.

The dynamic effects from postulated pipe breaks have been eliminated from the structural design basis of the reactor coolant system primary loop piping, as allowed by revised General Design Criterion

4. The elimination of these breaks is the result of the application of leak-before-break (LBB) technology which has been approved for WCGS by the NRC.The final licensing basis LBB submittal for WCGS is the proprietary WCAP-10691,"Technical Basis for Eliminating Large Primary Loop Pipe Rupture as a Structural Design Basis for Callaway and Wolf Creek Plants." The NRC approval of this use of LBB at WCGS was granted with the Safety Evaluation Report for the original license as a 10 CFR 50.12 exemption from parts of General Design Criterion 4 (GDC-4). See Supplement 5 Section 3.6.1.1 of the NUREG-0881 SER for the original WCGS operating license.The fracture mechanics analysis is not time-dependent and therefore is not a TLAA.However, the final LBB submittal is also supported by a fatigue crack growth assessment for a 40-year design life (WCAP-10691 Section 6.0), which is a TLAA.There is no licensing basis evaluation of embrittlement of the cast reactor coolant piping or other cast austenitic stainless steel (CASS) at WCGS, apart from the LBB question.Wolf Creek Generating Station Page 4.3-35 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Analysis Fracture Mechanics Analysis (not a TLAA)The fracture mechanics analysis depends in part on a material property, the crack initiation energy integral, JIN. The primary coolant loops at WCGS are SA 351 Grade CF8A cast stainless steel, which at PWR operating temperatures is subject to time-dependent thermal embrittlement.

Embrittlement reduces the JIN integral.Thermal embrittlement effects depend approximately logarithmically on time (more rapid initial change, achieving a saturation value after a long time). The available margins permitted use of a saturation value of JIN for this analysis.

Since a saturated JIN is not time-dependent, this fracture mechanics analysis is not a TLAA.Other supporting parts of the fracture mechanics analysis are supported in part by calculation of JIN values for WCGS for a 40-year life. These supporting analyses are therefore time-dependent.

However, the conclusion and safety determination of the LBB analysis does not depend on these supporting time-dependent analyses.

Therefore the supported fracture mechanics analysis is not a TLAA. Additionally, since the supporting analyses do not determine the result of the safety determination, the supporting analyses are also not TLAAs.Fatigue Crack Growth Assessment The final LBB submittal for WCGS includes a fatigue crack growth assessment for a typical-plant case representative of WCGS, for a range of materials at a typical location.

The analysis includes estimates of effects of the reactor coolant environment, and concludes that fatigue crack growth effects will be negligible.

The evaluation of these typical 40-year LBB fatigue crack growth effects assumed load transients, stresses, and numbers of transient events which are representative of the WCGS reactor coolant system primary loop design.Effects of Power Rerate and Tht Reduction on the LBB Analysis The power rerate and Thot reduction modifications had no effects on the LBB analysis.Effects of the Primary System Heatup Expansion Noise Events on the LBB Analysis The evaluation of effects of the primary system noise events considered possible effects on the LBB analysis and found that recommended LBB margins are maintained, and that the conclusions of the LBB analysis remain valid. See Section 4.3.2.9, "Primary Coolant System Heatup Expansion Noise Events." Effects of NSAL-94-025 Reactor Coolant Pump Support Column Tilt Westinghouse identified a concern that reactor coolant pump support column tilt may have an adverse effect on main loop piping thermal stresses during heatup and cooldown transients.

As described in Section 4.3.2.7, "ASME Section III Class 1 Piping and Piping Wolf Creek Generating Station Page 4.3-36 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Nozzles," the Westinghouse evaluation found a large increase in the crossover and cold leg stresses at the reactor coolant pump, and a significant change in the load of the tilted column; but since original analysis stresses were low the effects on stresses and usage factors would not affect code compliance or the conclusions of the LBB analysis.Disposition:

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

The LBB analysis found that fatigue crack growth effects will be negligible.

The basis for evaluation of fatigue crack growth effects in the LBB analysis will remain unchanged so long as the number of occurrences of each transient remains below the number assumed for the existing analysis of fatigue crack growth effects.The WCGS fatigue aging management program will ensure that the number of occurrences of each transient cycle in the primary loop piping remains below the number specified by the design specifications during the period of extended operation, and therefore below the number assumed for the existing analysis of fatigue crack growth effects; or that appropriate corrective measures maintain the design and licensing basis. Therefore, the effects will be managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).

4.3.3 ASME Section III Subsection NG Fatigue Analysis of Reactor Pressure Vessel Internals Summary Description The WCGS reactor vessel internals were designed after the incorporation of Subsection NG into the 1974 Edition of Section III of the ASME Boiler and Pressure Vessel Code. The design meets the intent of paragraph NG-331 1(c); that is, design and construction of core support structures meet Subsection NG in full, and other internals are designed and constructed to ensure that their effects on the core support structures remain within the core support structure code limits.Wolf Creek Generating Station License Renewal Application Page 4.3-37 Section 4 TIME-LIMITED AGING ANALYSES Analysis Westinghouse topical report WCAP-14577, "License Renewal Evaluation:

Aging Management for Reactor Internals" found that the only TLAAs supporting design of the reactor vessel internals are the fatigue analyses.

The NRC staff safety evaluation noted this finding, and concluded that (1) the aging effects of fatigue will be adequately managed, and (2) although the fatigue calculations needed for the TLAA have not been performed and/or have not been updated by the Westinghouse Owners Group to reflect operations during the license renewal period, the screening process and methodology presented are acceptable for licensees' use in preparing plant-specific fatigue TLAA evaluations to support license renewal applications.

The plant-specific requirements of 10 CFR 54.3 and 10 CFR 54.21 (c)(1) for fatigue TLAAs must be addressed by the license renewal applicant.

Code Fatigue Analyses of Record of the WCGS Reactor Vessel Internals The current reactor vessel internals analyses of record for WCGS are summarized in a Westinghouse supplementary design report that incorporates effects of power rerate and of the hot leg temperature reduction (Thor). The supplementary design report identifies a number of usage factors above 0.66 for the specified set of design basis transient events and for 40 years of high-cycle effects, and several above 0.9.The greater part of each calculated fatigue usage factor is due to effects of significant transients.

However, some part of fatigue usage in internals is due to the high-cycle effects, and therefore depends on steady-state operating time rather than on the number of transient events. High-cycle fatigue must therefore be evaluated separately in order to extend the conclusion of the supplementary design report to the end of the 60-year licensed operating period.Fatigue Analyses of Barrel-to-Former and Baffle-to-Former Bolts Cracked baffle-to-former bolts were found in a few European reactors with designs and materials similar to Westinghouse units, multiple failures have occurred in Alloy A-286 internals bolting in B&W reactors; and stress corrosion cracking has occurred in tack-welded Alloy X-750 bolts in German Siemens and Kraftwerk Union internals.

The failures have been attributed to a combination of fatigue, neutron embrittlement, and irradiation-assisted stress corrosion cracking (IASCC). With extended operation the stresses induced by differential void swelling between the bolts and bolted members, and between the bolted members, may also become significant, although the stresses will probably be somewhat mitigated by irradiation and thermal creep relaxation.

All of these effects are time-dependent and all except fatigue and thermal creep depend on neutron fluence.Fatigue in these bolts is the subject of an ASME code analysis, which is a TLAA. No other evaluations of these other effects have been introduced into the licensing basis at WCGS.Therefore, the baffle-to-former and barrel-to-former bolted connection designs are supported Wolf Creek Generating Station Page 4.3-38 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES by no TLAAs addressing effects other than fatigue. This finding agrees with the conclusion of Westinghouse topical report WCAP-14577-A.

The Westinghouse topical report observed that the barrel-to-former and baffle-to-former bolts are in the category of components

"...where the cyclic loadings are sufficiently uncertain to preclude the effective use of detailed fatigue design analysis, ..." and therefore for which "...alternatives for managing the effects of the age-related degradation are described...." The 40-year predicted usage factor in at least some of at least the baffle-to-former bolts is a significant fraction of the limit of 1.0. That is, the high predicted usage factor, the additional aging effects requiring mitigation, and the fact that some of these are synergistic (e.g., fatigue and the other cracking mechanisms) dictate that management of the fatigue usage factor in these bolts will be insufficient by itself, and that an aging management program must be constructed for the bolts which either adequately address all of these effects, or which will ensure their safety function despite these effects.WCNOC reviewed Westinghouse Technical Bulletin TB-03-2, "Reactor Internals Baffle-to-Former Bolt/Core Design Interface," and concluded that fatigue failures would not be expected during the original 40-year licensed operating period, but must be addressed for license renewal.Disposition:

Revision, 10 CFR 54.21(c)(1)(ii);

and Aging Management, 10 CFR 54.21(c)(1)(iii)

Confirmation of Negligible Effects of High-Cycle Fatigue Since fatigue usage factor does not depend strongly on flow-induced vibration or other high-cycle effects that are time-dependent at steady-state conditions, but depends more strongly on effects of operational, upset, and emergency transient events, the increase in operating life to 60 years should not have a significant effect on fatigue usage factor so long as the number of design basis transient cycles remains within the number assumed by the original analysis.

WCNOC will obtain a design report amendment to either quantify the increase in high-cycle fatigue effects, or to confirm that the increase will be negligible.

WCNOC will complete this action before the end of the current licensed operating period. The analysis of high-cycle fatigue effects in reactor internals will be revised for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(ii).

Cycle Count and Usage Factor Tracking for the Balance of Reactor Internals with Fatigue Analyses Transient cycle counting under the WCGS fatigue management program will ensure that the design basis fatigue usage factor limit (1.0) will not be exceeded in any analyzed location in the reactor internals without being identified and evaluated, including taking any necessary mitigating actions. Therefore, fatigue in the reactor vessel internals (other than the barrel-former and baffle-former bolts, below) will be adequately managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii).

Wolf Creek Generating Station Page 4.3-39 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES This disposition conforms to the "primary step" of an acceptable aging management program for fatigue effects in reactor internals described in WCAP-14577-A.

Aging Management for Barrel-to-Former and Baffle-to-Former Bolts The WCGS aging management program for reactor vessel internals for the license renewal period is identified in Section B2.1.35, Reactor Coolant System Supplement.

This program or programs will adequately manage combined effects of fatigue and other aging mechanisms in the barrel-to-former and baffle-to-former bolts for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii).

Wolf Creek Generating Station License Renewal Application Page 4.3-40 Section 4 TIME-LIMITED AGING ANALYSES 4.3.4 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)Summary Description The fatigue data upon which the ASME Section III fatigue curves are based are the result of tests in air at room temperature and constant strain rate. Concerns with possible effects of elevated temperature, reactor coolant chemistry environments, and different strain rates prompted NRC-sponsored research to assess these effects, first presented in the 1993 NUREG/CR-5999, "Interim Fatigue Design Curves for Carbon, Low-Alloy, and Austenitic Stainless Steels in LWR Environments." Subsequent research and studies, including NUREG/CR-6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," refined the methods.The NRC concluded that effects of the reactor coolant environment might need to be included in the calculated fatigue life of components, and opened three generic safety issues to address this question, all finally closed to a single Generic Safety Issue, GSI 190.Although GSI 190 has been closed for plants with 40-year initial licenses, NUREG-1800 states that "The applicant's consideration of the effects of coolant environment on component fatigue life for license renewal is an area of review," noting the staff recommendation

"...that the samples in NUREG/CR-6260 should be evaluated considering environmental effects for license renewal." Analysis Sample Locations NUREG/CR-6260 identifies seven sample locations for newer Westinghouse plants such as WCGS:* Reactor Vessel Lower Head to Shell Juncture* Reactor Vessel Primary Coolant Inlet Nozzles* Reactor Vessel Primary Coolant Outlet Nozzles* Surge Line Hot Leg Nozzle* Charging Nozzles* Safety Injection Nozzles [BIT Nozzles]* Residual Heat Removal Line Inlet Transition.

NUREG/CR-6260 does not distinguish between the "normal" (loop 1 cold leg) and"alternate" (loop 4 cold leg) charging nozzles. The two WCGS charging nozzles have equal calculated usage factors, but have had different operating histories.

Wolf Creek Generating Station Page 4.3-41 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES The four WCGS 10-inch accumulator and RHR cold leg safety injection nozzles (ACCSI nozzles) correspond to the NUREG/CR-6260 "inlet transition." Therefore, the WCGS evaluation includes these four nozzles and both charging nozzles. See Table 4.3-5.Analysis of Sample Locations WCGS performed plant-specific calculations for the seven sample locations applicable to WCGS identified in NUREG/CR-6260 for newer Westinghouse plants. WCGS evaluated effects of the reactor coolant environment on fatigue calculations using the appropriate Fen factors from NUREG/CR-6583 for carbon and low-alloy steels and from NUREG/CR-5704 for stainless steels, as appropriate for the material at each of these seven locations.

See the notes to Table 4.3-5 for the method used for each FEN multiplier.

At the first location, the vessel lower head to shell juncture, the expected 60-year fatigue usage factor was determined by multiplying the design basis 40-year usage factor times 1.5.All others were projected from historical and current rates of accumulation of transient cycles and usage factors, using either the cycle-based method or the stress-based method of the fatigue monitoring program described in Section 4.3.1, "Fatigue Management Program." The inlet, outlet, and hot leg nozzle predictions used the cycle-based method. The remaining charging, safety injection (BIT), and accumulator-RHR nozzle predictions used the stress-based method.Table 4.3-5 Summary of Fatigue Usage Factors at NUREG/CR-6260 Sample Locations, Adapted to WCGS CUF Expected Location material Expected FEN 60-year LaoMeaat 60 CUF with Years FEN Reactor Vessel Lower Head to SA-533 Gr. B Cl. 1 0.1005 2.45(1) 0.2462 Shell Juncture (Not Monitored)

Reactor Vessel Primary Coolant SA-508 Cl. 3 0.13467 2.45(1) 0.3299 Inlet Nozzle Reactor Vessel Primary Coolant SA-508 Cl. 2 0.21597 2.45(1) 0.5291 Outlet Nozzle Surge Line Highest-CUF Location, SA-182 F316N 0.05849 8.593(2) 0.50257 Hot Leg Nozzle-Normal, Loop 1 0.15863 0.87028 Charging Nozzles SA-182 F316N 5.486(2)-Alt., Loop 4 0.09847 0.54024 Wolf Creek Generating Station License Renewal Application Page 4.3-42 Section 4 TIME-LIMITED AGING ANALYSES Table 4.3-5 Summary of Fatigue Usage Factors at NUREG/CR-6260 Sample Locations, Adapted to WCGS Expected Expected<, :Location Material atp cted FEN CUF with..... ~~~~ ~a ..........

0.. .... ... CUF= Years FEN Safety Injection (BIT) Nozzles SA-182 F316N 0.16351 5.535(3) 0.9050 Accumulator and RHR Cold Leg Safety Injection Nozzles, "RHR SA-351 Gr. CF8A 0.06328 15.35(4) 0.9713 Line Inlet Transition" Notes to Table Maximum FEN for low dissolved oxygen.2 Computed FEN for low dissolved oxygen.3 Computed FEN for low dissolved oxygen, similar plant.4 Maximum FEN for low dissolved oxygen and slow strain rate.Disposition:

Validation, 10 CFR 54.21(c)(1)(i);

and Aging Management, 10 CFR 54.21(c)(1)(iii)

Validation

-Reactor Vessel Lower Head to Shell Juncture The low design basis usage factor for this location permits a projection of the usage for a 60-year life, equal to 1.5 times the design basis usage factor and times a conservative FEN for carbon steel, with considerable margin to the code allowable of 1.0. The evaluation of fatigue effects in this location, and effects of the reactor coolant on them, have thereby been validated and projected to the end of the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i).

Aging Management The analysis showed that the fatigue usage factors in the NUREG/CR-6260 sample locations should remain less than 1.0 for the period of extended operation, including the effects of the reactor coolant environment, and that the safety determination supported by the code fatigue analyses should therefore remain valid. The WCGS fatigue management program described in Section 4.3.1, "Fatigue Management Program," and in Appendix B3.1,"Metal Fatigue of Reactor Coolant Pressure Boundary," will include appropriate usage factor action limits to ensure that the usage factor at the remainder of these locations, including FEN, does not exceed 1.0 before an evaluation is completed and appropriate actions have been identified.

See Section 4.3.1.3, "Program Enhancements for the Period of Extended Operation," for a description of these action limits and corrective actions.Wolf Creek Generating Station Page 4.3-43 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Therefore, the effects of the reactor coolant environment on fatigue usage factors in these locations will be managed for the period of extended operation in accordance with 10 CFR 54.21 (c)(1)(iii).

4.3.5 Assumed Thermal Cycle Count for Allowable Secondary Stress Range Reduction Factor in B31.1 and ASME Section III Class 2 and 3 Piping Summary Description Piping in the scope of license renewal that is designed to B31.1 or ASME Section III Class 2 and 3 requires the application of a stress range reduction factor to the allowable stress range for secondary stresses (expansion and displacement) to account for thermal cyclic conditions.

If the number of equivalent full temperature cycles exceeds 7,000, a factor less than I must be used. These piping analyses are TLAAs because they are part of the current licensing basis, are used to support safety determinations, and depend on an assumed number of thermal cycles that can be linked to plant life.Analysis None of ANSI B31.1 or the ASME Section III Subsections NC and ND for Class 2 and 3 piping invokes fatigue analyses.

If the number of full-range thermal cycles is expected to exceed 7,000, these codes require the application of a stress range reduction factor to the allowable stress range for expansion stresses (secondary stresses).

The allowable secondary stress range is 1.0 SA for 7000 equivalent full-temperature thermal cycles or less and is reduced in steps to 0.5 SA for greater than 100,000 cycles. Partial cycles are counted proportional to their temperature range.A review of ASME Section III Class 2 and 3 and B31.1 piping specifications found no indication of a number of expected lifetime full-range or equivalent full-range thermal cycles greater than 7,000 during the original 40-year plant life.The EPRI license renewal "Non-Class 1 Mechanical Implementation Guideline and Mechanical Tools" includes .temperature screening criteria to identify components that might be subject to significant thermal fatigue effects. Normal and upset operating temperatures less than 220 OF in carbon steel components, or 270 OF in stainless steel, will not produce significant thermal stresses, and will not therefore produce significant fatigue effects. A systematic survey of plant piping systems in the scope of license renewal found that with the exception of reactor coolant sampling lines, described below, the piping and components in the scope of license renewal: " Do not meet the operating temperature screening criteria of the EPRI Mechanical Tools, and therefore do not experience significant thermal cycle stresses; or" Clearly do not operate in a cycling mode that would expose the piping to more than three thermal cycles per week, i.e., to more than 7,000 cycles in 60 years; or Wolf Creek Generating Station Page 4.3-44 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES* The assumed thermal cycle count for the analyses depends closely on reactor operating cycles, and. can therefore be conservatively estimated by the thermal cycles used in the ASME Section III Class 1 vessel and piping fatigue analyses.For this last case, see the reactor coolant system thermal cycles discussed in USAR Section 3.9(N) and listed in its Table 3.9(N)-i, and in LRA Table 4.3-1, "Significant Transient Cycle Limits Tracked by the WCGS Fatigue Management Program." Of these, those likely to produce full-range thermal cycles in balance-of-plant Class 2, 3, and B31.1 piping, in a 40-year plant lifetime, are the 200 heatup-cooldown cycles plus 400 reactor trips (Table 4.3-1).Other events may contribute a few full-range or a number of part-range cycles. However the total count of all design basis events (second column of Table 4.3-1) is only about 2000, discounting the reduced temperature return to power cycles, which are over-estimated and will contribute at most some part-range cycles to non-Class 1 systems, and noting that the RCS and pressurizer heatup-cooldown cycles are common.Therefore, the total count of expected full-range thermal cycles for most of these systems is well under 1000 for a 40-year plant life. For the 60-year extended operating period the number of thermal cycles for piping analyses would be proportionally increased to less than 1500, which is only a fraction of the 7000-cycle threshold for which a stress range reduction factor is required in the applicable piping codes.The WCNOC review determined that piping calculations included appropriate stress ranges for. any temperatures other than normal ambient, with one exception that has been satisfactorily addressed.

The piping calculations included appropriate stress intensification factors, which do not depend on the number of cycles.. Therefore changes to allowable stress as the result of additional thermal cycles is correctly addressed by the required changes to stress range reduction factor.Reactor Coolant Sample Lines The survey of plant piping systems found that the reactor coolant sample lines may be subject to more than 7000 thermal cycles. Review of operating practice determined that lines used for daily samples remain hot except for a Technical Specification leakage test about every two days and are therefore subject to less than about 11,000 cycles in 60 years, permitting a stress range reduction factor (SRRF) of 0.9. WCNOC reviewed the design of this piping and identified three segments whose design calculations require revision to meet this SRRF.Wolf Creek Generating Station Page 4.3-45 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES Disposition:

Validation, 10 CFR 54.21(c)(1)(i);

and Revision, 10 CFR 54.21(c)(1)(ii)

For less than 7000 equivalent full-temperature thermal cycles the stress range reduction factor is 1. Therefore, so long as the estimated number of cycles remains less than 7000 for a 60-year life, the stress range reduction factor remains at 1 and the stress range reduction factor used in the piping analysis will not be affected by extending the operation period to 60 years.Validation for Piping Other than Sample Lines and Other High-Thermal-Cycle Lines The expected number of equivalent full-range thermal cycles for other than reactor coolant sample piping should be less than 1500 in 60 years, which is only a fraction of the 7000-cycle threshold for which a stress range reduction factor is required in the applicable piping codes. Therefore, the existing analyses of piping for which the allowable range of secondary stresses depends on the number of assumed thermal cycles and that are within the scope of license renewal, other than ASME Class 1 analyses and sample lines, are valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).

Validation or Revision for Reactor Coolant Sample Lines Reactor coolant sample lines are subject to a thermal cycle about every two days in modes 1 through 4, and may therefore be subject to something less than 11,000 significant thermal cycles in a 60-year plant lifetime.

For this case the design codes impose an SRRF of 0.9.WCNOC reviewed the design analyses of these lines and has determined that the secondary stress ranges are within the limits imposed by the 0.9 SRRF in all but three line segments.

Therefore, with the exception of three line segments, the design of these sample lines has been validated for the duration of the period of extended operation in accordance with10 CFR 54.21 (c)(1)(i).

For the three line segments where secondary stress ranges were not within 0.9 of the allowable, WCNOC has determined that reanalysis should be able to demonstrate secondary stress ranges below this limit. WCNOC will complete these reanalyses, and any additional corrective actions or modifications indicated by them, before the end of the current licensed operating period. These analyses will be revised in accordance with 10 CFR 54.21(c)(1)(ii), for the duration of the period of extended operation.

4.3.6 Fatigue Design of Spent Fuel Pool Liner and Racks for Seismic Events Summary Description The spent fuel pool racks were replaced in order to accommodate a larger inventory.

The design of the replacement racks included a fatigue analysis of the racks and of high-stress locations in the pool liner. These analyses are described in USAR Section 9.1A.4.3.5.4.

Wolf Creek Generating Station Page 4.3-46 License Renewal Application 11 Section 4 TIME-LIMITED AGING ANALYSES Analysis Fuel Racks The specification for the replacement spent fuel storage racks requires them to be designed, fabricated, and installed to ensure operation for an intended period of 60 years.However, examination of the design criteria document and the calculation found that the fatigue analysis did not increase the assumed number of safe-shutdown earthquakes (SSE)and operating basis earthquakes (OBE) events above the 20 SSEs and 1 OBE assumed for a 40-year license.The replacement racks in the spent fuel pool were analyzed for fatigue effects of the SSE and OBE events using methods similar to those for ASME Section III Class 1 analyses.

The analysis calculated a cumulative usage factor of 0.404 for these events at the maximum-stress location, the pedestal rack to baseplate junction.No detectable seismic events have occurred in the 20-year operating history of the plant to date.3 The design basis number of events therefore remains sufficient for the remainder of the original licensed operating period, plus the 20-year licensed operating period extension, and the replacement racks are therefore presently qualified for the number of these events now expected for the remainder of a 60-year life..Spent Fuel Pool Liner The docketed reracking licensing report also describes a fatigue evaluation of locations of the pool liner most affected by seismic loads imposed by the racks, and for the same 20 OBE plus 1 SSE events described above for the racks. The report does not state a calculated usage factor for the liner, but states that the result was acceptable for this set of cyclic loading events.The question on fatigue in the liner was closed on an NRC staff conclusion that "Based on the maximum stress level in the liner, the cumulative usage factor was shown to be well below 1.0." Summary The analysis for both the racks and liner depend only on the assumed number of OBE and SSE events. Although an ASME Section III Class 1 pressure boundary fatigue analysis would omit the faulted, SSE loads, this analysis included them because spent fuel storage must continue to function following these events. The analysis remains valid for any period for which the number of OBE events has not been and is not expected to be exceeded, 3 The free-field Strong Motion Accelerometer (SMA) trigger level is adjustable over a minimum range of 0.01 g to 0.03 g (USAR 3.7(B).4.1b).

Exceeding the trigger level actuates an annunciator in the control room to indicate a possible seismic event. No SMA trigger attributable to an earthquake has occurred at Wolf Creek to date.Wolf Creek Generating Station Page 4.3-47 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES assuming an additional SSE event then occurs. Since the remaining plant life from the present to the end of the period of extended operation (2006 to 2045) is less than that of the original license to which the numbers of OBE and SSE events apply, and since no detectable SSE or OBE has occurred, these analyses remain valid for the period of extended operation.

Disposition:

Validation, 10 CFR 54.21(c)(1)(i)

Operating basis and safe-shutdown earthquakes (OBE and SSE) are the only design basis events for cyclic loading on the fuel pool liner and racks. No detectable seismic events have occurred in the 20-year operating history of the plant to date, so that the design basis number of events remains sufficient for the remainder of the original licensed operating period, plus the 20-year licensed operating period extension, and the replacement racks are therefore presently qualified for the number of these events now expected for the remainder of a 60-year life. Therefore the analyses are valid for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(i).

4.3.7 Fatigue Design and Analysis of Class IE Electrical Raceway Support Angle Fittings for Seismic Events Summary Description The design of Class IE electrical raceway included a fatigue evaluation of the effects of operating basis and safe shutdown earthquake loads (OBE and SSE loads) on angle fittings used at the connections of strut hangers to overhead supports, or at interhanger locations.

Analysis A cumulative usage factor was calculated and compared to a fatigue curve. The usage factor was based on tests of typical designs to failure, and the number of fatigue cycles to failure was divided by a factor of safety of 1.5 in order to establish an allowable number of fatigue cycles for design. The design assumed the number of OBE and SSE events recommended by IEEE 344-1975, which states that the maximum number of OBE and SSE events plausible during a plant lifetime is five and one, respectively.

The fatigue analysis was extremely conservative.

Although an ASME Section III Class 1 pressure boundary fatigue analysis would omit the faulted, SSE loads, this analysis included them. The analysis assumes 150 alternating stress cycles per OBE or SSE seismic event, based on a design basis maximum acceleration period "in the neighborhood of 15 seconds," and a conservative support first-mode resonance of 10 Hz, or a total of 750 cycles for the 5 OBE events assumed plus 150 for the single SSE. No detectable OBE or SSE has occurred in the first 20 years of operation, 4 so that qualification for the original design basis number of OBE and SSE events is sufficient for both the 20 years remaining in the original license from the time of this application, and to the end of the period of extended operation.

4 See Footnote 3 in Section 4.3.6.Wolf Creek Generating Station Page 4.3-48 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES The analysis used a fatigue curve of angular rotation vs. cycles to failure for the specific Power Strut Welded-Fillet PS-608 angle fittings.

This fatigue failure curve is asymptotic to infinity beyond 1000 cycles; that is, angular deflections below that for which 1000 cycles are permitted are below the endurance limit. A limiting angular deflection was used, calculated from the manufacturer's allowed ultimate moment. Although this calculated angular deflection at this ultimate moment was less than 60 percent of the angular deflection at 1000 cycles-to-failure, and therefore also less than 60 percent of the angular deflection that would permit an infinite number before failure, 1000 cycles was assumed as the allowable for the usage factor evaluation.

This is extremely conservative since neither the design fatigue curve nor the failure curve provides any finite limits in this region.A realistic usage factor would therefore be significantly less than the 0.9 reported (900 assumed/1 000 allowed cycles). The calculation could have concluded, with equal validity, that the resulting usage factor is negligible.

Disposition:

Validation, 10 CFR 54.21(c)(1)(i)

The seismic fatigue analysis of Class IE electrical support angle fittings was extremely conservative, assuming 1000 allowable cycles for a deflection considerably less than the endurance limit. Operating basis and safe-shutdown earthquakes (OBE and SSE) are the only design basis events for cyclic loading on these support angle fittings.

Furthermore, no detectable seismic events have occurred in the 20-year operating history of the plant to date, so that the design basis number of events remains sufficient for the remainder of the original licensed operating period, plus the period of extended operation.

Therefore, the analysis is valid for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(i).

Wolf Creek Generating Station License Renewal Application Page 4.3-49 Section 4 TIME-LIMITED AGING ANALYSES

4.9 REFERENCES

4.9.1 US NRC NUREG 0881. Safety Evaluation Report Related to the Operation of Wolf Creek Generating Station, Unit No. 1. Washington:

USNRC, April 1982[WFCKNUREG0881V01.pdf].

4.9.2 Federal Register Notice at 68 FR 60422. "Notice of Availability of Model Application Concerning Technical Specification Improvement Regarding Extension of Reactor Coolant Pump Motor Flywheel Examination for Westinghouse Plants Using the Consolidated Line Item Improvement Process." 22 October 2003.4.9.3 WCNOC letter W004-0001, Britt T. McKinney, Site Vice President; to US NRC Document Control Desk. "Docket No. 50-482: Application for Technical Specification Improvements to Extend the Inspection Interval for Reactor Coolant Pump Flywheels Using the Consolidated Line Item Improvement Process." 9 February 2004.4.9.4 US NRC letter, Jack N. Donohew, Senior Project Manager, Section 2, Project Directorate IV, Division of Licensing Project Management, Office of Nuclear Reactor Regulation; to Rick A. Muench, President and Chief Executive Officer, WCNOC."Wolf Creek Generating Station -Issuance of Amendment Re: Extending the Inspection Interval for Reactor Coolant Pump Flywheels." 16 June 2004.With attached Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 153 to Facility Operating License No. NPF-42. Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Docket No. 50-482.16 June 2004.4.9.5 WCAP-8163, Topical Report, Reactor Coolant Pump Integrity in LOCA. Pittsburgh:

Westinghouse, September 1973 4.9.6 WCAP-15338-A.

Warren Bamford and R. D. Rishel. Westinghouse Owners Group Generic License Renewal Program Topical Report. A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants. Pittsburgh:

Westinghouse Electric Company LLC, October 2002.4.9.7 USNRC Letter, Grimes (NRC) to Newton (WOG). "Safety Evaluation of WCAP-15338, 'A Review of Cracking Associated With Weld Deposited Cladding in Operating PWR Plants."'

15 October 2001.4.9.8 WCAP-15666-A.

Westinghouse Topical Report. P. L. Strauch et al. Extension of Reactor Coolant Pump Motor Flywheel Examination.

Rev. 1. For the Westinghouse Owners Group (WOG). Pittsburgh:

Westinghouse, October 2003.4.9.9 US NRC Letter, Herbert N. Berkow, Director, Project Directorate IV, Division of Licensing Project Management, Office of Nuclear Reactor Regulation; to Robert H.Bryan, Chairman, Westinghouse Owners Group. "Safety Evaluation of Topical Report WCAP-15666, 'Extension of Reactor Coolant Pump Motor Flywheel Examination'." 5 May 2003.Wolf Creek Generating Station Page 4.9-1 License Renewal Application Section 4 TIME-LIMITED AGING ANALYSES 4.9.10 WCAP-16028.

Westinghouse Report. T. J. Laubham, J. Conermann, and R. J.Hagler. Analysis of Capsule X from Wolf Creek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation Surveillance Program. Rev. 0, March 2003.4.9.11 WCAP-16030.

Westinghouse Report. T. J. Laubham. Evaluation of Pressurized Thermal Shock for Wolf Creek. Rev. 0, May 2003.4.9.12 US NRC Letter, Jack Donohew, Senior Project Manager, Section 2, Project Directorate IV, Division of Licensing Project Management, Office of Nuclear Reactor Regulation; to Rick A. Muench, President and CEO, WCNOC. "Wolf Creek Generating Station -Test Results for the Withdrawal of Surveillance Capsule X." 21 January 2004.With Attached Evaluation by the Office of Nuclear Reactor Regulation Related to Capsule X Surveillance Capsule Program Report, Summary of Findings (Reactor Vessel Integrity Neutron Irradiation), Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station, Docket No. 50-482. 21 January 2004.4.9.13 BC-TOP-I.

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Containment Building Liner Plate Design Report. Rev. 1. Bechtel Corporation, December 1972.4.9.14 BC-TOP-5-A.

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Rev. 3. Los Angeles: Bechtel Power Corporation, February 1975.4.9.15 NUREG-1774.

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Frank Klanica and Charlie Gay. Westinghouse -Owners Group Generic License Renewal Program Topical Report. Aging Management for Class I Piping and Associated Pressure Boundary Components.

Pittsburgh:

Westinghouse Electric Company LLC, December 2000. (Accepted version of WCAP-14575, Rev. 1, August 1996.)4.9.17 US NRC Letter, Grimes (NRC) to Newton (WOG). "Acceptance for Referencing of Generic License Renewal Program Topical Report Entitle, "License Renewal Evaluation:

Aging Management for Class 1 Piping and Associated Pressure Boundary Components," WCAP-14575, Rev. 1, August 1996." 8 November 2000.With attached Final Safety Evaluation by the Office of Nuclear Reactor Regulation Concerning Westinghouse Owners Group Report, WCAP-14575, Revision 1,"License Renewal Evaluation:

Aging Management for Class I Piping and Associated Pressure Boundary Components," Project No. 686.4.9.18 WCAP-14950.

M. A. Gray et al. Westinghouse Report. Mitigation and Evaluation of Pressurizer Insurge-Outsurge Transients.

Westinghouse Proprietary Class 2C.February 1998.Wolf Creek Generating Station Page 4.9-2 License Renewal Application