ML070680160

LRPD-04, Revision 0; TLAA - Mechanical Fatigue, 01/25/2006

ML070680160
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	01/25/2006
From:	Finnin R Entergy Nuclear Operations, Entergy Nuclear
To:	Office of Nuclear Reactor Regulation
References
LRPD-04, Rev 0
Download: ML070680160 (35)

Text

VERIFICATION OF WNPS UCENSE RENEWAL PROJECT REPORT T i e of Report: -

TtAA Mechanicat1Fatigue RepoitNumber: LRpD-04 Revision: 0 This project report documents evahaths related b the WNPS license renewal pmj-Signatures certify that the &port was prepared, checked and reviewedby the License Renewal ProjectTeam in accurdance with the WNPS license renewal project guidelines and #at it was approved by the EN1 License Renewal Project Manger and the WNPS Manager, Engineering Projects.

redsion.

Other documnt(s) impacted by this revision: -Yes, See Revision Dexription Sheet xNo

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Prepared by Date: 97%-ok Reviewed by: Diite:@3/6 Approved by: mte: ,J,/~,/cJ 6 D. J. Lach ~ J % J--+r J

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 2 of 35 REVISION DESCRIPTION SHEET Revision Number Description Pages and/or Sections Revised 0 Initial Issue All

LRPD-04 VYNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Paae 3 of 35 Table of Contents 1.O Introduction..................................................................................................................... 4 2.0 VYNPS Evaluation of Metal Fatigue Time-Limited Aging Analyses ........................... 6 2.1 WNPS Class 1 Fatigue .................................................................................................... 6 2.1.1 ASME Code Requirements for Class 1 Fatigue 6 2.1.2 Cumulative Usage Factors 7 2.1.3 Reactor Pressure Vessel 10 2.1.4 Reactor lnternals 11 2.1.5 Class 1 Piping and Component Fatigue 12 2.2 WNPS Non-Class 1 Fatigue .......................................................................................... 14 2.2.1 Non-Class 1 Piping and In-Line Components 15 2.2.2 Non-Class 1 Pressure Vessels, Heat Exchangers, Storage Tanks, Pumps and Turbine Casings 17 2.3 WNPS Containment Liner Plate, Metal Containment, and Penetrations Fatigue Analyses ........................................................................................................ 18 2.3.1 Fatigue Analysis of the Torus 19 2.3.2 Fatigue Analysis of the Safety Relief Valve (SRV) Discharge Piping 19 2.3.3 Fatigue Analysis of Other Torus Attached Piping (TAP) 20 2.4 WNPS In-Service Inspection (ISI)- Fracture Mechanics Analyses ............................... 20 2.4.1 Recirculation (Jet Pump) Risers 21 2.4.2 Core Spray Piping in the Reactor Vessel. 21 2.4.3 Reactor Vessel Plate 1-15 21 2.4.4 Core Spray Nozzle to Safe End Weld Overlay 22 2.4.5 Containment corrosion 22 2.4.6 Primary Containment Localized Thinning 22 2.5 WNPS Response to Industry Issues on Fatigue............................................................ 22 2.5.1 VYNPS Response to NRC Bulletin 88-08 23 2.5.2 Effects of Reactor Water Environment on Fatigue Life 23 3.0 Summary and Conclusions.......................................................................................... 29 4.0 References...............................................,........................................................~............31

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 4 of 35 I.O Introduction This report supports the license renewal of Vermont Yankee Nuclear Power Station (VYNPS).

The purpose of this report is to evaluate time-limited aging analyses (TLAA) related to mechanical fatigue for WNPS components. In addition, the report discusses the VYNPS plant-specific responses to industry issues related to fatigue. Information on the overall license renewal project methodology and associated documentation can be found in LRPG-01, License Renewal Project Plan (Ref. 1).

Fatigue analyses are potential TLAA for Class 1 and selected non-Class 1 mechanical components. Fatigue is an age-related degradation mechanism caused by cyclic stressing of a component by either mechanical or thermal stresses that becomes evident by cracking of the component. Fatigue analyses are TLAA if they are based on a set of design transients that are based on the life of the plant.

When TLAA-metal fatigue is identified in the aging management program column, the TLAA associated with that fatigue is applicable. Review of the TLAA, per 10 CFR 54.21 (c)(l),

determines whether:

(i) the TLAA remains valid for the period of extended operation, (ii) the TLAA can be extrapolated to the end of the period of extend operation, or (iii) effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

For fatigue, if the TLAA remains valid (i) or is extrapolated to cover the period of extended operation (ii), then cracking due to fatigue is not an aging effect requiring management. If the TLAA does not remain valid for the period of extended operation, then cracking due to fatigue is an aging effect requiring management under 10 CFR 54.2l(c)(l)(iii). Cracking due to fatigue can be managed by a variety of plant programs including IS1 and BWRVIP.

Class 1 components (reactor vessel and recirculation system piping) received a fatigue analysis in accordance with ASME Section Ill, Subsection NB. ASME Section Ill requires evaluation of fatigue by considering design thermal and loading cycles. The design cycles for WNPS are listed in WNPS calculation WC-378 (Ref. 55). WNPS monitors transient cycles that contribute to fatigue usage in accordance with requirements of WNPS Administrative Procedure AP-0145 (Ref. 8). Reactor coolant system pressure boundary piping (with the exception of reactor recirculation piping) was designed to ANSI B31.Iand secondary stresses (e.g., stress due to thermal expansion and anchor movements) are analyzed for fatigue using stress intensification factors (SIFs) and stress range allowables. The stress range allowables are a function of thermal design cycles.

The non-Class 1 aging management reviews for WNPS identify non-Class 1 mechanical components that are within the scope of license renewal and are subject to aging management review. Based on exposure to mechanical and thermal cycling, specific components are subject to cracking by fatigue as identified in the non-Class 1 aging management reviews. Non-Class 1 component types subject to fatigue include pipe, tubing, fittings, tanks, vessels, heat exchangers, valve bodies and bonnets, pump casings, and miscellaneous process components.

LRPD-04 VYNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 5 of 35 Fatigue evaluations that meet the definition of TLAA for Class Iand non-Class 1 mechanical components at VYNPS are described and evaluated below. Cumulative usage factors have been documented and the actual numbers of design transient cycles have been projected to 60 years. Although some transients are projected to exceed the cycle limits before the end of 60 years, an adequate program is in place to track cycles and to provide corrective actions if limits are approached. The maximum cumulative usage factors (CUF) identified for WNPS components are summarized in Table 2.1-1.

In addition to metal fatigue, fracture mechanics analyses of flaws discovered during in-service inspection are TLAA for those analyses based on tirne-limited assumptions defined by the current operating term. When a flaw is detected during in-service inspections, either the flaw must be repaired or the component that contains the flaw can be evaluated for continued service in accordance with ASME Section XI. These evaluations may show that the component is acceptable to the end of the license term based on projected in-service flaw growth. Flaw growth is typically predicted based on the design thermal and loading cycles.

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 6 of 35 2.0 WNPS Evaluation of Metal Fatigue Time-Limited Aging Analyses The design basis for Class 1 nuclear power plant components subjected to cyclic service is in ASME Section 111. Transient events that the plant might credibly experience are evaluated to establish a design basis for plant equipment. The fatigue analyses rely on the definition of design basis transients that envelope the expected cyclic service and the calculation of a cumulative usage factor (CUF). In accordance with ASME Section 111, Subsection NB, the cumulative usage factor must be less than 1.O (Ref. 17).

IOCFR54.21(c) (Ref. 25) requires an evaluation of time-limited aging analyses to demonstrate that either the analyses remain valid for the period of extended operation, the analyses have been projected to the end of the period of extended operation, or the effects of aging on the intended function(s) will be adequately managed for the period of extended operation. This section will review the WNPS specific methods to meet the identified requirements and to ensure that the TLAA associated with fatigue are properly dispositioned for the period of extended operation. This review consists of the following sections.

0 Section 2.1 reviews the Class Ifatigue analysis documentation. The fatigue cumulative usage factors (CUFs) for Class 1 components are TLAA that are calculated by analyzing a number of transient cycles, The WNPS transient monitoring program tracks the actual transient cycles that the plant experiences to ensure they are maintained below the analyzed number of cycles.

0 Section 2.2 reviews the non-Class 1 fatigue analysis documentation. The design documentation is maintained to document the stress analysis determination of allowable cycles.

0 Section 2.3 summarizes the fatigue analyses associated with the WNPS containment.

Section 2.4 addresses specific analyses associated with flaws identified by inservice inspection. Any crack growth analyses justifying continued operation have been dispositioned for the period of extended operation.

0 Section 2.5 reviews the WNPS responses to industry experience. Appropriate actions are taken based on industry experience for significant fatigue concerns that were not considered during the original design. This includes environmental-assistedfatigue concerns as expressed in NUREGICR-6260.

2.1 WNPS Class 1 Fatique 2.1.1 ASME Code Requirements for Class 1 Fatigue WNPS Class 1 components evaluated for fatigue and flaw growth include the reactor pressure vessel (RPV) and appurtenances, certain reactor vessel internals, the reactor recirculation system (RRS), and the reactor coolant system (RCS) pressure boundary. The VYNPS Class 1 systems include components within the ASME Section XI, Subsection IWB inspection boundary and are described in aging management review reports AMRM-31, AMRM-32, and AMRM-33 (Refs. 5, 6 and 7).

VYNPS License Renewal Project '

Revision 0 TLAA - Mechanical Fatigue Fatigue evaluations were performed in the design of the WNPS Class 1 components designed in accordance with the requirements specified in ASME Section Ill (see Table 2.1-1). The fatigue evaluations are contained in analyses and stress reports, and because they are based on a number of transient cycles assumed for a 40-year plant life, these evaluations are TLAA.

Design cyclic loadings and thermal conditions for the Class 1 components are defined by the applicable design specifications for each component. The original design specifications provided the initial set of transients that were used in the design of the components and are included as part of each component analysis and stress reports, and because they are based on a number of transient cycles assumed for a 40-year plant life, these evaluations are TLAA.

Design cyclic loadings and thermal conditions for the Class 1 components are defined by the applicable design specifications for each component. The original design specifications provided the initial set of transients that were used in the design of the components and are included as part of each component analysis and/or stress report. The component analyses and stress reports contain the fatigue evaluations, which support the fatigue design basis for each component.

2.1.2 Cumulative Usage Factors A review of the fatigue evaluations reveals the maximum cumulative usage factors (CUFs) for applicable WNPS Class 1 components. The documents reviewed are current design basis fatigue evaluations that do not consider the effects of reactor water environment on fatigue life.

The maximum cumulative usage factors (CUF) for Class 1 components are summarized in Table 2.1-1 CUFs were found for the following Class 1 components.

Reactor pressure vessel 0 Reactor vessel internals The CUF for each component was determined by the following method.

1. Design documents (analytical reports, stress reports, design reports, evaluations, etc.)

were searched for documented usage factors.

2. The main source document for the reactor pressure vessel was WNPS calculation WC-378 (Ref. 9 and 55). This report was completed to replace the varied transients in the original analysis with transients that represented actual plant events and would make tracking of cycles more realistic. Secondary purposes were to remove unnecessary conservatism, and produce a more accurate calculation of vessel fatigue life using current analytical methods. This calculation was based on a licensed power level of 1593 MWt.

3. A search was performed to find relevant documentation that post-dated WC-0378 to determine if the usage factors documented in this analysis had been subsequently revised. A General Electric report (Ref. 61) prepared in support of the extended power uprate, was found. This report reviewed the existing CUFs, but did not refer to W C -

LRPD-04 VYNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 8 of 35 378. It then revised these CUFs to include the power uprate, and extended the CUFs to 60 years by multiplying them by 1.5 (60 years / 40 years). The feedwater nozzles and RR outlet nozzles exceeded CUFs of 1.O when extended by this method. However, unless the number of plant operational transients exceeds the analyzed numbers, these CUFs will not increase as predicted by this calculation.

No plant-specific CUFs were found for the core spray safe end, the feedwater piping, the RHR return piping, or the RR piping tee. As these iocations are reviewed for environmentally assisted fatigue in Section 2.5.2,the generic CUFs calculated in NUREG-6260 were used.

Table 2.1-1 Cumulative Usage Factors Location CUF Bottom Head and Support Skirt 0.40 1 Closure Flange I 0.57 Closure Studs 1 0.62 Core Spray Nozzle 0.63 Core Spray Safe End (I) 0.18 Core Support Structure 0.06 CRD Penetration 0.73 CRD Return Nozzle 0.39 Feedwater Nozzle 0.75 Feedwater Piping () 0.43 Refueling Bellows 0.67

~

RHR Return Piping () 0.03 RR inlet nozzle 0.61 RR outlet nozzle 0.81 RR Piping Tee (I) 0.40 Shroud Repair Rod Threaded Ends I 0.12 Shroud Support Plate Slotted Holes 1 0.23 Steam outlet nozzle I 0.17 (1) Plant-specific CUFs were not found. Generic CUFs from NUREG-6260were used.

Current design basis fatigue evaluations, including the CUFs, are based on design transients.

The design transients are listed in Table 2.1-2.

The VYNPS fatigue monitoring program tracks and evaluates the cycles and requires corrective actions if limits are approached. The VYNPS fatigue monitoring program (administrative procedure AP 0145, Ref. 8) ensures that the numbers of transient cycles experienced by the plant remain within the allowable numbers of cycles, and hence the component cumulative usage factors (CUFs) remain below the code allowable value of 1.O.

AP 0145 (Ref. 8)provides background information on the program and gives instructions for the acquisition, review, recording and maintenance of transient cycle data. The transient events to be monitored are listed in form 1 (APFO145.01) of the procedure and are taken from the revised reactor vessel cycle calculation (Ref. 55). The reporting frequency for cycle count data is once per fuel cycle. The most recent cycle monitoring program report (Ref. 57) provides the number

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Paae 9 of 35 of transients logged as of May 4, 2004. This report formed the basis for the projected number of cycles at sixty (60) years, Le., at the end of the license renewal period.

Through a review of Ref. 55,the transients currently being tracked at WNPS have been identified. In Revision 1 to WC-0378, WNPS reduced the number of transients monitored from the original 18 to 6. AP 0145 counts five transients and compares their actual number of cycles to the design allowed numbers of cycles. The other 13 of the original 18 transients are covered by the sixth transient labeled reactor startup/shutdowncycles, which consists of an engineering evaluation of plant data each fuel cycle. While this gives a more accurate evaluation of the fatigue to the plant from operation, there is no design allowable number of cycles for this transient. Actual events are analyzed by WNPS engineers to determine how many cycles need to be credited for these Startup/Shutdown events.

To ensure that cyclic fatigue limits will not be exceeded during vessel design life, as well as into the period of extended operation, it is necessary to count, evaluate and track the six cyclic fatigue inducing events.

Completion of the WNPS Equipment Cycle Record Keeping Procedure (Ref. 8) following each refueling alerts management if any transient is approaching its limiting number of cycles.

Therefore, the WNPS fatigue monitoring program provides an acceptable means for ensuring that TLAAs associated with fatigue for the RPV and associated Class 1 components will remain valid through the period of extended operation in accordance with I O CFR 54.21(c)(i). The fatigue monitoring program, is discussed more detail in LRPD-02.

The number of cycles accrued to date have been extrapolated to determine the numbers of cycles expected at the end of 60 years of operation. The results of the 60-year projections show that for the five WNPS transients with design allowable limits, the numbers will remain below the design allowable numbers during the period of extended operation. However, the projection for the sixth transient (engineering analysis of plant data for each fuel cycle) has no design allowable number against which to compare the projection.

First a projection factor for 60-years was calculated as the ratio of 60 years to the operating time through the last cycle update. For each transient, the 60-year projection is then calculated as the projection factor times the May 4, 2004, cycle count. The results are shown in Table 2.1-2.

The results of the 60-year projections show that for the five WNPS transients with design allowable limits, the numbers will remain below the design allowable numbers during the period of extended operation.

The WNPS fatigue monitoring program is adequate for monitoring plant transients and will assure that the allowed number of transients is not exceeded. Consequently, the TLAA (fatigue analyses and CUFs) based on those transients will remain valid for the period of extended operation in accordance with 10 CFR 54.2l(c)(l)(i). The fatigue monitoring program is discussed in more detail in LRPD-02, Aging Management Program Evaluation Results.

Design Transient Design Current Projection Projected  % used Basis Cycles Factor() Cycles at in 60 Cycles Logged as 60 Years of years of 05/04/04 Operation (Ref. 8) (Ref. 57)

Closure Flange 200 32 I.908 61 31%

Bolting Closure Flange 200 32 I.908 61 31%

Unbolting System Pressure 120 32 1.908 61 31%

Tests Heatup 300 90 1.908 172 57%

Cooldown 300 88 1.908 168 56%

Reactor 32 1.908 61 Startup/Shutdown Cycles The projection factor for 60-years was calculated as the ratio of 60 years to the operating time through the last cycle update.

Date of commercial operation for WNPS = November 20, 1972 Date after 60 years of operation = November 20,2032 Date of latest program report = May 04, 2004 (Ref. 57)

Projection factor = (November 20,2032 - November 20, 1972) /

(May 04,2004 - November 20, 1972

= 1.908 Reactor Pressure Vessel The reactor pressure vessel (and appurtenances) fatigue analyses were performed in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section 111, 1965 Edition, 1966 and 1967 addenda. (A complete listing of applicable codes is given in table 4.1-1 of the UFSAR.) The fatigue analyses based on a number of transient cycles assumed for a 40-year plant life are considered TLAA. The design transients are listed in Table 2.1 -2.

Design cyclic loadings and thermal conditions for the reactor pressure vessel were originally defined in the design specifications for the vessel. The original design specifications provided the set of transients that were used in the design of the components. Subsequent evaluations by WNPS modified the list of transients to one which more closely reflected actual plant transients that were easier to track, while still bounding the original design transients.

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 11 of 35 The WNPS Fatigue Monitoring program will assure that the allowed number of transient cycles is not exceeded. The program requires corrective action if transient cycle limits are approached.

Consequently, the TLAA (fatigue analyses) based on those transients will remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(l)(i) or the effects of aging on the intended function(s) will be adequately managed for the period of extended operation in accordance with 10 CFR 54.21(c)(l)(iii). Further details on cycle projections and cycle monitoring were provided in Section 2.1 2.

2.1.4 Reactor lnternals Although not mandatory, the design of the reactor vessel internals is in accordance with the intent of ASME Section 111 (Ref. 3, Section 3.3.4).

2.1.4.1 lnternals Fatigue UFSAR, Appendix C, section C.2.5.3 states that a fatigue analysis for the Dresden plant was performed that is a good approximation for WNPS, the implication being that there is no site specific fatigue analysis. UFSAR Section 3.3.6 states A vibration analysis of reactor vessel internals was performed in the design to reduce failures due to vibration. When necessary, vibration measurements were made during startup tests to determine the vibration characteristics of the reactor vessel internals and the recirculation loops under forced recirculation flow. Vibratory responses were recorded at various recirculationflow rates using strain gages on fuel channels and control rod guide tubes, accelerometers on the shroud support plate and recirculation loops, and linear differential transducers on the upper shroud and shroud head-steam separator assembly. The vibration analyses and tests were designed to determine any potential, hydraulically-inducedequipment vibrations and to check that the structures should not fail due to fatigue. The structures were analyzed for natural frequencies, mode shapes, and vibrational magnitudes that could lead to fatigue at these frequencies. With this analysis as a guide, the reactor internals were instrumented and tested to ascertain that there are no gross instabilities. The cyclic loadings were evaluated using as a guide the cyclic stress criteria of the ASME Code, Section 111. These field tests were only performed on reactor vessel internals that represent a significant departure from design configurations previously tested and found to be acceptable. Field test data were correlated with the analyses to ensure validity of the analytical techniques on a continuing basis.(3) Reference (3) is Reference 83 of this document. This analysis is not a TLAA as it is not based on any time-limited assumptions; vibration levels were analyzed and confirmed to be acceptable.

Section 4.3.2.2 of the Dresden License Renewal Application (Ref. 85) states that this analysis does not contain any time-limited assumptions. The NRC SER for the Dresden Application (Ref. 86) does not dispute this statement but notes that the analysis has been redone for the Extended Power Uprate at Dresden. Section 2.2.3.2 of the NRC draft SER for the WNPS extended power uprate (Ref. 87) states For components other than the steam separators and dryers, the evaluation of FIV for the reactor internal components was performed based on the vibration data recorded during startup testing at the GE prototype BWW4 plant (Monticello) and WNPS. The vibration levels were calculated by extrapolating the recorded vibration data to EPU conditions and compared to the plant allowable limits. The stresses at critical locations were calculated based on the extrapolated vibration peak response displacements and found to be within the

I LRPD-04 VYNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 12 of 35 GE allowable design criteria of 10 ksi (where 1 ksi = 1000 pounds per square inch). Stress values less than 10 ksi for stainless steel are within the endurance limit under which sustained operation is allowed without incurring any cumulative fatigue usage. The vibration evaluation methodology, as described in Section 3.4.2 of the PUSAR, is conservative based upon the absolute sum combination of the various modes of vibration, including the absolute sum of the maximum vibration amplitude occurring in each mode. The licensee concluded that vibration levels of all safety-related reactor internal components are within the acceptance criteria. The NRC staff finds the licensees specified stress limit of 10 ksi for the reactor internal components to be reasonably conservative in comparison to the ASME Code limit of 13.6 ksi for the peak vibration stress and is, therefore, acceptable.

Again, the described analysis is not based on any time-limited assumptions and is thus not a TLAA.

2.1.4.2 Shroud Repair Fatigue UFSAR Section K.3.1 states that the core shroud repair was designed for a 40 year life. Refs.

14 and 15, the shroud repair stress reports, provide the fatigue design basis and were the source documents for the,core shroud stabilizer CUFs.

In response (BVY 96-48 Ref. 14) to the NRCs Request for Additional Information on the core shroud repair, WNPS stated that the fatigue analysis had been performed for the shroud repair hardware in calculation 2499502-601 (Ref. 15). This calculation included a fatigue analysis of the slotted hole in the shroud support plate where the shroud repair ligaments attach. This analysis is a TLAA. The resulting CUF was 0.23 (Table 5-6 of Ref. 15). This CUF is based on the design transients in the original reactor vessel design report. As such, the WNPS fatigue monitoring program will assure that those transients are not exceeded and the TLAA will remain valid for the period of extended operation in accordance with 10CFR54.21(c)(l)(i). The fatigue monitoring program is discussed in LRPD-02.

The shroud and shroud repair was re-evaluated for the constant pressure power uprate (CPPU)

(Ref. 28) as discussed above. The results of the CPPU evaluations showed that the loads and stresses for all shroud repair components remained within design limits, and thus the results of calculation 2499502-601 remain valid. No new fatigue analysis was performed.

The WNPS Fatigue Monitoring program will assure that the allowed number of transient cycles is not exceeded. The program requires corrective action if transient cycle limits are approached.

Consequently, the TLAA (fatigue analyses) based on those transients will remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(l)(i) or the effects of aging on the intended function(s) will be adequately managed for the period of extended operation in accordance with 10 CFR 54.21 (c)(l)(iii).

2.1.5 Class 1 Piping and Component Fatigue The VYNPS Class Iboundary corresponds to all reactor coolant system (RCS) pressure boundary components within the ASME Section XI, IWB inspection boundary and is described in aging management review reports AMRM-31, AMRM-32 and AMRM-33 (Refs. 5, 6, and 7).

WNPS Class 1 RCS pressure boundary piping was designed and analyzed in accordance with ANSI B31.I. (UFSAR Table 4.1-1 gives a complete listing of applicable codes.) As such, fatigue analyses that calculate cumulative usage factors are not required. Rather stress range

W N P S License Renewal Project

. . I LRPD-04 Revision 0 TLAA - Mechanical Fatigue rarigue Page 13 of 35 reduction factors are used to account for anticipated transients (normally 7000 cycles).

Applicable design specifications define the design cyclic loadings and thermal conditions for the Class 1 components.

Portions of certain systems have been replaced or modified over the course of operation of WNPS. Those modifications may be done to more recent code requirements. Review of Current Licensing Basis documentation has found the following additional Class 1 fatigue topics.

2.1 -5.1 Reactor Recirculation System WNPS replaced the reactor recirculation (RR) system piping in 1986. (See Section 4.3.6 of the UFSAR). Also replaced were connecting portions of the residual heat removal (RHR) system piping. The new piping was designed and analyzed to ANSI 631. I , but was inspected and tested to ASME Section Ill requirements. Stress analyses for the reactor recirculation system were performed to B31.1 requirements in Refs. 10 and 11. Even though B31.1 does not require a fatigue analysis, such an analysis was done for the highest anticipated usage factor location -

the RHR to RR tee. These analyses were based on a number of cycles not expected to be exceeded in 40 years and as such are treated as TLAA.

The VYNPS Fatigue Monitoring program will assure that the allowable number of transient is not exceeded. The program requires corrective action if transient cycle limits are approached.

Consequently, the TLAA (fatigue analyses) based on those transients will remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(l)(i) or the effects of aging on the intended function(s) will be adequately managed for the period of extended operation in accordance with 10 CFR 54.2l(c)(l)(iii).

Section 4.3 of Ref. I O makes the following two points about this calculation.

1. The initiation of RHR has the highest delta-temperature (up to 200 degrees F) between RHR and RR. This relatively short transient is permissible for 4000 cycles. As it occurs only during plant cooldown, there will be far less than 4000 cycles in 60 years (300 cooldowns per Table 2.1-2 of this report). Consequently this portion of the analysis remains valid for the period of extended operation per 10 CFR 5421(c)(l)(i).

2. A detailed ASME Class 1 Section 111 1983 edition fatigue analysis for the Tee has been performed. The duty cycles used consisted of 27 load sets. Each of the load sets includes pressure, thermal expansion effect, earthquake and the temperature gradient through wall at the Tee. The total number of cycles for each load set is based on 40 years of plant operation. The results of this study indicate that the RHR Tee meets the Class 1 fatigue criteria. Again, as noted in Section 2.1.2 W N P S will not exceed the allowable number of cycles for 40 years even after 60 years of operation, this portion of the calculation will also remain valid for the period of extended operation per 10 CFR 54.2l(c)(l)(i).

2.1 5.2 Main Steam Isolation Valve Cycles UFSAR Section 4.6.3 states that the main steam isolation valves (MSIV) are designed for 40 years based on 100 cycles during the first year and 50 cycles per year thereafter. This statement may be interpreted to imply a TLAA. The UFSAR statement refers to mechanical cycles of the valve (in the UFSAR, 2,050 cycles are assumed for 40 years).

LRPD-04 VYNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 14 of 35 Cycling of these valves will lead to wear of active subcomponents (disc and seat) with no license renewal function and fatigue of the passive components (body and bonnet) that have a license renewal intended function (pressure boundary).

Fatigue of the valve body is not an issue due to the large allowable number of cycles on the valves. Projected cycles of the MSlVs are far below the 50 cycles per year allowed. This TLAA will remain valid for the period of extended operation in accordance with I O CFR 54.21(c)(I)(i).

The MSlVs will not exceed 2050 cycles in 60 years (34 cycles per year).

The MSlVs are monitored for cracking by the ASME Inservice Inspection program, which would also detect fatigue cracking if it were to occur.

2.1.5.3 Flow Induced Vibration (FIV) of the Main Steam (MS) and Feedwater (FW) Piping The constant pressure power uprate (CPPU) increases the main steam and feedwater flowrates by approximately 20%. The GE CPPU report (Ref. 28) evaluated this increase in flow as it affects FIV of the MS and FW piping. By reference 28, VYNPS committed to perform a piping vibration startup test program, at the uprated power level, consistent with the ASME code and regulatory requirements. This testing is expected to confirm that FIV of this piping is not a concern for the period of extended operation. No TLAA is associated with these components.

2.2 WNPS Non-Class 1 Fatisue The non-Class 1 aging management reviews (AMRs) for WNPS identify the non-Class 1 mechanical components susceptible to cracking by fatigue including pipe, tubing, fittings, tanks, vessels, heat exchangers, valve bodies and bonnets, pump casings, and miscellaneous process components. These components may experience thermal cycles as a result of either flow transients or cyclic thermal stratification.. These thermal cycles are not tracked as part of the fatigue monitoring program. The impact of thermal cycles on non-Class 1 components is reflected in the calculation of the allowable stress range. The allowable stress range is reduced by the stress range reduction factor if the number of thermal cycles exceeds 7000. (1995 ASME Boiler and Pressure Vessel Code, Division I,Subsection NC, Class 2 Components, Page 158.)

Only those systems that may experience a high cycle frequency are of concern. 7000 cycles equates to approximately one cycle every 2 days for forty (40) years of operation, or approximately one cycle every 3 days for 60 years of operation.

The non-Class 1 fatigue screening document in Appendix H of the Mechanical Tools (Ref. 18) was used to determine locations susceptible to fatigue cracking in non-Class 1 systems at VYNPS. The first step in the screening process was to identify non-Class 1 components that may have normalhpset condition operating temperature in excess of 220°F for carbon steel or 27OoFfor stainless steel. These values are based on recommendations in the EPRl Fatigue Management Handbook (TR-104534), as summarized in the Mechanical Tools. Although most WNPS non-Class 1 components do not exceed the temperature thresholds, some components, identified in the appropriate AMRs are further evaluated for fatigue below.

I LRPD-04 VYNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue 2.2.1 Non-Class 1 Piping and In-Line Components The design of ASME 111 Code Class 2 and 3 piping systems incorporates the Code stress reduction factor for determining acceptability of piping design with respect to thermal stresses.

The design of ASME B31 .ICode components also incorporates stress reduction factors based upon an assumed number of thermal cycles. In general, 7000 thermal cycles are assumed, leading to a stress reduction factor of 1.O in the stress analyses. VYNPS evaluated the validity of this assumption for 60 years of plant operation. The results of this evaluation indicate that the 7000 thermal cycle assumption is valid and bounding for 60 years of operation. Therefore, the pipe stress calculations are valid for the period of extended operation in accordance with 10 CFR 54.2l(c)(l)(i).

The design code for most non-Class 1 piping and in-line components (e.g., fittings and valves) is ANSI B31 .Ior ASME Ill Subsections NC and ND. These Codes specify evaluation of cyclic secondary stresses (Le., stresses due to thermal expansion and anchor movements) using stress intensification factors (SIFs) and allowable stresses (SA). The allowable secondary stress range is 1.O SAfor 7000 cycles or less and is reduced in steps to 0.5 SAfor greater than 100,000 cycles. No increase is allowed for less than 7000 cycles.

For all those non-Class Icomponents identified in the AMR reports described above as subject to cracking due to fatigue, a review of system operating characteristics was conducted to determine the approximate frequency of significant thermal cycling. If the number of equivalent full temperature cycles for 60 years of operation is below the limit used for the original design (usually 7000 cycles, as described in Section 2.2 above), the component is suitable for the period of extended operation. If the number of equivalent full temperature cycles exceeds the limit, evaluation of the individual stress calculations will be required.

The following non-Class 1 systems include piping and in-line components subject to fatigue.

The specific component types for each system are identified in the aging management review results table for the associated AMR..

0 AMRM-02: The residual heat removal system (RHR) system contains filters, orifices, piping, thermowells, and valve bodies subject to fatigue. RHR is used during plant shutdowns to cold conditions. Even though cycling of the RHR system may occur during outages, significant thermal transients occur only early in the shutdown when reactor coolant is still at temperatures above 220 O F . Thus significant cycles of the RHR system are coincident with plant cooldowns, which are limited to 300 per Table 2.1-2. The RHR system fatigue analysis remains valid for the period of extended operation in accordance with 10 CFR 542I(c)(l)(i).

0 AMRM-04: The automatic depressurization system (ADS) contains piping, tubing, and valve bodies subject to fatigue. ADS relieves pressure following a plant upset, not during normal plant operation. The number of cycles is expected to be orders of magnitude below 7000. The safetyhlief valves are tested when removed, consequently testing does not fatigue the discharge piping. The ADS fatigue analysis remains valid for the period of extended operation in accordance with 10 CFR 54.2l(c)(l)(i). Relief valve discharge piping and torus penetrations have unique fatigue analyses based on a much lower number of cycles. These components are discussed in Section 2.3 of this report.

I Revision 0 I 0 AMRM-05: The high pressure coolant injection (HPCI) system contains drain pots, orifices, piping, steam traps, strainer housings, thermowells, tubing, and valve bodies subject to fatigue. The steam supply to the HPCl pump turbine is exercised during testing and is used to mitigate design basis events. The pump is tested quarterly (when steam pressure is available), so the number of tests in sixty years will be no more than 240. The plant is restricted to less than 400 reactor trips which limits the number of HPCl actuations. The number of total cycles is expected to be significantly below 7,000 equivalent full temperature cycles (less than 640 cycles) during the period of extended operation. The HPCl piping fatigue analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(l)(i).

0 AMRM-06: The reactor core isolation cooling (RCIC) system contains drain pots, orifices, piping, steam traps, strainers, tubing, and valve bodies subject to fatigue. The steam supply to the RClC pump turbine and turbine exhaust are exercised during testing and to mitigate design basis events. The pump is tested quarterly (when steam pressure is available), so the number of tests in sixty years will be no more than 240.

The plant is restricted to less than 400 reactor trips which limits the number of RClC actuations. The number of total cycles is expected to be significantly below 7,000 equivalent full temperature cycles (less than 640 cycles) during the period of extended operation. The RClC pump turbine piping fatigue analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21 (c)(l)(i).

0 AMRM-13: The emergency diesel generators (EDG) contain expansion joints, piping, silencers, and turbocharger housings subject to fatigue. The diesels are tested monthly giving 720 full temperature cycles in 60 years. EDG actuations occur less frequently than the testing. Thus the total equivalent full-temperature cycles in 60 years will not exceed 7000. The EDG fatigue analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(l)(i).

0 AMRM-17: The fire protection system (FP) contains diesel exhaust lines and lube oil return lines subject to fatigue. As these components are only thermally cycled when the fire protection diesel is loaded, they are expected to experience far less than 7000 cycles in 60 years similar to the emergency diesels above. The fire protection system fatigue analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(?)(i).

0 AMRM-21: The John Deere Diesel contains a silencer, turbocharger, and exhaust piping subject to fatigue. As these components are only thermally cycled when the John Deere diesel is loaded, they are expected to experience far less than 7000 cycles in 60 years, as explained for the emergency diesels above. The John Deere diesel fatigue analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(l)(i).

0 AMRM-26: The main condenser and MSlV leakage pathway contain orifices, heat exchanger tubes, piping, strainer housings, thermowells, steam traps, tubing, and valve bodies subject to fatigue. The main condenser and MSlV leakage pathway are normally in service during plant operation. However, they only see significant temperature transients during plant startup and shutdown. Again, the number of cycles expected on

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 17 of 35 these systems is well below the 7000 cycles allowable. The main condenser and MSlV leakage pathway fatigue analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(l)(i).

0 AMRM-30: Non-safety systems affecting safety related systems contain filter housing, flow elements, rupture disks, piping, and valve bodies subject to fatigue. The bulk of the components that experience temperatures above 220°F are in the steam/feedwater cycle. As such these systems are limited to the number of transients discussed above for the main condenser and main steam isolation valve leakage pathway and will not exceed 7000 cycles during the period of extended operation.

The heating boiler (HB) system piping and valves carry steam to local space heaters, and HB piping and valves are in several rooms with safety related equipment. This system is designed to B31.Iand is thus inherently acceptable for 7000 cycles of operation. As the system is normally started and stopped (thermally cycled) several times per year based on prevailing weather, the total cycles for 60 years are expected to be in the hundreds of cycles, not thousands of cycles. The heating boiler fatigue analysis remains valid for the period of extended operation in accordance with I O CFR 54.21(c)(l)(i).

The post accident sampling system (PASS) system contains tubing and valve bodies subject to fatigue. PASS is an accident mitigation systems that is not used during normal plant operation. The number of cycles to the PASS system is expected to be orders of magnitude below 7000. The PASS system fatigue analysis remains valid for the period of extended operation in accordance with 10 CFR 54.21(c)(l)(i).

Some applicants for license renewal have estimated that piping in the primary sampling system will have more than 7000 thermal cycles before the end of the period of extended operation. At WNPS, the sampling system is used to take reactor coolant samples every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> during normal operation. However, the normal samples are taken from the RWCU filter influent, where the water has already been cooled. Thus normal sampling does not cause a thermal cycle. Alternate samples may be taken directly from the B discharge header of the reactor recirculation system via containment penetration X-41; however, this is an infrequently performed procedure and this piping, designed to ASME B31.Iwill , not exceed 7000 cycles prior to 60 years of operation.

The non-safety systems affecting safety related systems fatigue analyses remain valid for the period of extended operation in accordance with 10 CFR 54.2l(c)(l)(i).

Thus, the TLAA for the all non-Class Ipiping and piping components remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(i).

2.2.2 Non-Class IPressure Vessels, Heat Exchangers, Storage Tanks, Pumps and Turbine Casings Non-Class 1 pressure vessels, heat exchangers, storage tanks and pumps are designed in accordance with ASME Vlll or ASME Ill Subsection NC or ND (e.g., Class 2 or 3). Some tanks and pumps are designed to other industry codes and standards, reactor designer specifications,

.- . , LRPD-04 I Revision 0 I

and architect engineer specifications. ASME Section Vlll Division 2 and ASME Section I l l Subsection NC-3200 include fatigue design requirements. Due to conservatism in ASME Section Vlll Division 1 and ASME Section Ill NC-3100/ND-3000, detailed fatigue analysis is not required. Fatigue analyses are also not required for NClND pumps and storage tanks (e 15 psig). Also, the component designer should have specified ASME Sectiqn Vlll Division 2 or NC-3200 if cyclic loading and fatigue usage could be significant. The design specification identifies the applicable design code for each component.

Both ASME Section Vlll Division 2 and ASME Section Ill NC-3200 include provisions for "exemption from fatigue", which is actually a simplified fatigue evaluation based on materials, configuration, temperature and cycles.

Fatigue analysis is not required for other design codes (e.g. ASME Section Vlll Division 1, A W A , MSS, NEMA), and components designed and fabricated with these codes are suitable for the period of extended operation without further evaluation.

The non-Class 1 system heat exchangers, pumps and turbine casings identified in the aging management review reports as subject to fatigue are evaluated below.

AMRM-02 The residual heat removal system (RHR) system contains heat exchangers and pump casings subject to fatigue. The RHR pumps are designed to ASME Section 111, Class C. The heat exchanger shell side is designed to ASME Section Ill, Class C while the heat exchanger tube side is designed to ASME Section Vlll/TEMA Class 2. (Section 2.2.5.2 of Design Basis Document RHR, Revision 1, January, 1999) These codes do not require a fatigue analysis. As such, there is no TLAA associated with fatigue of the RHR components. The RHR pump and heat exchanger are acceptable for the period of extended operation.

AMRM-05: The HPCl system contains a turbine casing subject to fatigue. This turbine is designed to the NEMA code. This code does not require a fatigue analysis. As such, there is no TLAA associated with fatigue of the HPCl turbine. The HPCl turbine is acceptable for the period of extended operation.

AMRM-06: The RClC system contains a turbine casing subject to fatigue. This turbine is designed to the NEMA code. This code does not require a fatigue analysis. As such, there is no TLAA associated with fatigue of the HPCl turbine. The HPCl turbine is acceptable for the period of extended operation.

2.3 WNPS Containment Liner Plate, Metal Containment, and Penetrations Fatiaue Analyses The torus and torus attached piping systems were analyzed as part of the Mark 1 containment long-term program, using methods and assumptions consistent with NUREG-0661 (Ref. 80).

The fatigue analyses performed included the torus, SRV piping and penetrations, and other torus attached piping.

The VYNPS torus analysis is Technical Report TR-5319-1 (Ref. 77), which was transmitted to the NRC via Ref. 78. The VYNPS torus attached piping analysis is Technical Report TR-5319-2 (Ref. 79). The piping analysis in turn references GE report MPR-751 (Ref. 16).

VYNPS License Rene--- - - I - -I Renewal Project TLAA - Mechanical Fatigue rauyue I LRPD-04 Revision 0 Page I 9 of 35 2.3.1 Fatigue Analysis of the Torus The fatigue analyses of the torus looked at both the torus shell and attached piping systems.

The WNPS plant specific fatigue usage factor (TR-5319-1, Section 3.3.1, Ref. 77) for the torus shell is 0.001 for normal operation and 0.01 1 for upset conditions. These values are so small that when multiplied by 1.5 to account for 60 years rather than 40 years, they are still insignificant usage factors. The fatigue analysis of the torus during normal operation and upset conditions has thus been projected through the period of extended operation in accordance with IOCFR54.21(c)(l)(ii).

The WNPS fatigue usage factor (TR-5319-1, Section 3.3.1, Ref. 77) for the torus shell is 0.078 for the design basis accident. As there will still be only one design basis LOCA for the life of the plant, including the period of extended operation, this analysis is not based on a time-limited assumption and is not a TLAA.

The vent system was conservatively analyzed by assuming that all maximum stresses occur simultaneously, and that all cycles reach these maximum values. This produced a conservative CUF for the vent system of 0.76. The significant contributor to this CUFs is post-LOCA chugging, a once in plant-life event. As there will still be only one design basis LOCA for the life of the plant, including the period of extended operation, this analysis is not based on a time-limited assumption and is not a TLAA.

2.3.2 Fatigue Analysis of the Safety Relief Valve (SRV) Discharge Piping TR-5319-2 (Ref. 79) documents stress evaluations for the SRV piping for various load combinations, but does not include a fatigue analysis. The fatigue analysis of the SRV piping, along with all the other torus attached piping (TAP) is bounded by MPR-751, the GE Mark I containment program (Ref. 16). MPR-751 was designed to bound all BWR plants which utilize the Mark I containment design. The analysis concluded that for all plants and piping systems considered, in all cases the fatigue usage factors for an assumed 40-year plant life was less than 0.5. In a worst-case scenario, extending plant life by an additional 20 years would produce usage factors below 0.75. Since this is less than 1.O, the fatigue criteria are satisfied. The MPR-751 generic fatigue analysis is thus projected for the period of extended operation in accordance with 10CFR54.21(c)(l)(ii).

A WNPS plant specific analysis (TR-3519-2) addresses the torus SRV penetration sleeves and bellows. This analysis states that the SRV penetrations are qualified for 7500 cycles of maximum load while the SRVs are expected to see less than 50 cycles at maximum load and less than 4500 cycles a partial load. The report concludes Since the 7500 cycles of maximum load bounds both of these by such a large margin and since no other significant loads are imposed on the line, the penetration was assumed acceptable for fatigue without further evaluation. (Section 2.4.5 of Ref. 79) Increasing the 40 year cycles by 1.5 for the period of extended operation would still be only 75 maximum load cycles and 6750 low load cycles for a total of 6850 mixed load cycles, less than the 7500 maximum load cycles permitted. The fatigue analysis for torus penetrations thus remains valid for the period of extended operation in accordance with 10CFR54.21(c)(l)(i).

LRPD-04 WNPS License Rene-' D--:--L TLAA - Mechanical rarigue I Revision 0 1 Page 20 of 35 I 2.3.3 Fatigue Analysis of Other Torus Attached Piping (TAP)

The WNPS plant-specific analysis (TR-5319-2) references the generic GE Mark Icontainment program (Ref. 16), for other torus attached piping. The results of the generic GE Mark 1 containment program (based on 40 years of operation) were that 92% of the TAP would have cumulative usage factors of less than 0.3, and that 100% would have usage factors less than 0.5. Conservatively multiplying the CUFs by 1.5 shows that for 60 years of operation, 92% of the TAP would have CUFs below 0.45, and 100% would have CUFs below 0.75. These calculations have thus been projected through the period of extended operation in accordance with I O CFR 50.2I(c)(ii).

2.4 WNPS In-Service InsDection (ISI) - Fracture Mechanics Analvses Flaws in Class 1, 2, or 3 components discovered during inservice inspections (ISI) must be evaluated in accordance with ASME Section XI, Subsections IWB, IWC, or IWD, respectively.

One option is to show the indication is acceptable without repair based on a fracture mechanics analyses (FMA). FMA evaluation requires a prediction of flaw growth considering a chosen evaluation period, usually either the time until the next inspection or the remaining licensed life of the plant. FMA evaluations performed for the licensed life of the plant may be TLAA that must be addressed for license renewal.

Plant Technical Specifications (Ref. 4, Section 4.6.E) require an in-service inspectionltesting program to verify the integrity of the reactor coolant pressure boundary, Specifically, 10 CFR 50.55a(g) (Ref. 22) requires IS1 per ASME Section XI, and I O CFR 50.36(~)(3)(Ref.23) provides general surveillance requirements. In accordance with 10 CFR 50.55a, the IS1 Program Plan is reviewed every 120 months and revised, as necessary, to meet the latest NRC authorized edition of ASME Section XI. This revision is submitted to the NRC for approval.

The examination categories defined in Table IWB-2500-1 require the use of nondestructive examination (NDE) techniques to detect and characterize flaws. The flaws may be service-induced (e.g., fatigue) or may be fabrication flaws that have grown due to service loads. T able IWB-2500-1specifies the extent and frequency of inspection . The inspection intervals are valid for any period of extended operation.

Flaws detected during examination are evaluated by comparing the examination results to acceptance standards established in ASME Section XI. Unacceptable indications require detailed analysis (e.g., ASME Section XI, Appendix A), repair, or replacement.

This section reviews analyses of flaws discovered during inservice inspections (El)at WNPS.

Class 1, 2, or 3 components require evaluation in accordance with ASME Section XI, Subsections IWB, IWC, or IWD, respectively. For any indication discovered during IS1 that exceeds acceptance standards,Section XI requires that (I) repairs be made, (2) affected portions of the item be replaced, or (3) the indication be shown acceptable through fracture mechanics analysis (FMA).

Acceptance through FMA requires a prediction of flaw growth considering either a chosen evaluation period (i.e., no shorter than the time until the next inspection following discovery of the flaw), or the remaining service life of the component. Flaw indications that are determined not to grow beyond acceptance limits during the evaluation period are justified for continued

- I LRPD-04 VYNPS License TLAA - Mechanicai rarigue I Revision 0 Page 21 of 35 operation. FMA evaluations performed for the current operating term may be TLAA. This review of the VYNPS inservice inspection records found six fracture mechanics analysis that were evaluated as potential TLAA; it was determined that none of them were actual TLAA.

Results of the review are given below.

2.4.1 Recirculation (Jet Pump) Risers Indications on four jet pump riser welds were found in the spring of 1998. These welds were originally evaluated for one fuel cycle (Ref. 70). However, Ref 71 was submitted in 1999 to justify extending the time between inspections to two fuel cycles, and this was accepted by the NRC in Reference 72.

The follow up inspection in 2001, and reanalysis of the 1998 data, showed that two of the 1998 indications were false. Indications on jet pump risers N2H and N2K were confirmed and are documented on IR# 01-10 (Ref. 26). The indications were analyzed by General Electric in Report JXOAL (Ref. 73) and determined to be acceptable for two cycles of operation.

Additional evaluations performed by Entergy (Ref. 88) determined that these welds are acceptable until the refueling outage in 2007. The analysis in Reference 73 is still valid and the welds were scheduled for re-inspection in RFO 24 (2004).

The fatigue growth analysis does not justify leaving the flaws in service for the design life of the plant, rather it justifies leaving them in service for four cycles of operation (until RFO 26) after which they are inspected and sized. If the flaws are still below the size analyzed, they may stay in service two additional cycles and then be inspected again. This analysis is not a TLAA since it is not based on a time-limited assumption defined by the current operating term.

2.4.2 Core Spray Piping in the Reactor Vessel.

Indications were identified in core spray piping welds in 1996. Results and evaluations were submitted to the USNRC by Ref. 27. The flaw growth calculations (Ref. 76) for these indications are acceptable for 2 cycles of operation. Additional evaluations performed by Entergy (Ref. 89) determined that these indications are acceptable until RFO 26 (2007).WNPS is continuing to inspect the core spray piping in accordance with the guidelines of BWRVIP-18.

This analysis is not a TLAA since it is not based on a time-limited assumption defined by the current operating term.

2.4.3 Reactor Vessel Plate 1-15 Reactor vessel plate 1-15 had one indication in the 1995 inspection. The indication is located in the plate below weld EF which joins plates 1-12 and 1-15. The weld is outside the core region.

The indication is acceptable for continued service per calculation package YAEC-25Q-301, which was submitted to the NRC by VYNPS letter B W 96-119 (Ref. 74), and approved by the NRC in their letter of I 1 Oct 96 (Ref. 75).

The analysis is not a TLAA since it does not involve a time-limited assumption defined by the current operating term. Allowable flaw size was calculated based on flaw growth, assuming a fixed number of cycles and worst case stresses. The predicted flaw size after the assumed number of cycles is significantly less than the allowable flaw size.

LRPD-04 VYNPS license Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 22 of 35 The flaw growth projection assumed 105 cycles (heatups). The flaw was re-inspected in 2004 (Ref. 94) and no flaw growth was observed. Conservatively assuming two heatups per year from 2004 through the period of extended operation (2004-2032, inclusive) results in only 58 cycles. Since the projected number of cycles is much less than the analyzed value, the analysis remains valid well beyond the period of extended operation. Note that the plant is expected to periodically re-inspect this weld each 10 year interval, thereby verifying there is no crack growth and essentially zeroing the count on allowable cycles.

2.4.4 Core Spray Nozzle to Safe End Weld Overlay In 1986 cracks were found on the core spray nozzle to safe end welds. The repair, described to the NRC in FVY 86-036 (Ref. 67),was accomplished using a weld overlay, leaving the cracks in service. The original intent was to use the weld overlay for up to 5 years before doing a replacement of the nozzle safe ends. An analysis of the overlay acceptability for 5 years was provided in Appendix C to Fvy 86-036. Subsequently, due primarily to the high radiation exposure that would be required for the repair, WNPS decided to use the weld overlay as a permanent repair. This was presented to the NRC in WNPS letter FVY 88-19 (Ref. 68). That letter included a report that reviewed the expected flaw growth for the as-found cracks. The conclusion was that flaw growth would be arrested by the compressive stress in the component and no growth above 75% thru-wall was expected. The NRC accepted use of the weld overlay as a permanent repair in letter NVY 88-080 (Ref. 69). The acceptance was based on the analysis provided and VYNPSs commitment to continue periodic inspection of these weld overlays in accordance with NUREG-0313 and Generic Letter 88-01. This analysis is not a TLAA as it is not based on time-limited assumptions for the current operating term.

2.4.5 Containment corrosion In 2001, the Inspection of the containment structure (per ASME Section XI - IWE) found corrosion of the vent header and vent pipe bowls in the W primary containment. This corrosion was analyzed by VYNPS Technical Evaluation TE 2001-25 (Ref. 90) and found to be acceptable for use as is. TE 2001-25 is a comparison of as-found wall thickness to minimum required wall thickness. This analysis does not include any time limited assumptions and is not based on any time period. As such this analysis is not a TLAA.

2.4.6 Primary Containment Localized Thinning In 1999, corrosion was found on approximately 20 square inches of the containment. The condition was evaluated by WNPS calculation WC-2043 (Ref. 91) and found acceptable for the life of the plant. This analysis is not a TLAA since it does not involve a time-limited assumption defined by the current operating term. The calculation is based on 100 test cycles (design basis accident cycles) of containment pressure. At an expected test frequency of less than once every two years, the analysis remains valid for over 200 years, which is well beyond the period of extended operation.

2.5 VYNPS Response to lndustrv Issues on Fatique Industry experience and research efforts have revealed fatigue issues not considered as part of the original design basis. Some of these issues impact fatigue analyses and resulted in the

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 23 of 35 issuance of NRC generic communications. The following subsections discuss concerns directly related to fatigue.

Although aging management review reports AMRM-31, AMRM-32 and AMRM-33 (Refs. 5, 6 and 7) review the industry experience and NRC identified concerns for Class Icomponents, the reports do not review all of the fatigue-related concerns and do not contain site-specific responses. Industry events have identified some specific fatigue situations that were not a part of the original design evaluations. This section provides the WNPS specific responses for fatigue related issues.

This review has determined that WNPS has properly evaluated the existing industry experience on fatigue. WNPS will continue to evaluate any future industry experience in accordance with plant procedures.

2.5.1 WNPS Response to NRC Bulletin 88-08 NRC Bulletin 88-08 (Ref. ZI), its supplements and associated information notices identified concerns with thermal stresses due to unanalyzed temperature distributions in piping connected to the reactor coolant system (RCS). The bulletin required a review of the systems connected to the RCS and a reporting of the results and necessary corrective actions. The supplements provided additional information on events and examinations to detect crack locations.

WNPS provided responses to the NRC for NRC Bulletin 88-08 and its supplements. Based on these responses, the NRC staff found that WNPS met the requirements of NRC Bulletin 88-08 (Ref. 66).

Subsequently, commitments regarding inspections in response to NRC Bulletin 88-08 have been superseded by the WNPS risk-informed in-service inspection (RI-ISI) of ASME Class 1 piping as approved by the NRC (Ref. 64). Technical Specification 4.6.E.1 allows this alternate method as approved by the NRC staff (Ref. 65). Aging effects due to thermal stratification as described in Bulletin 88-08 will be managed by augmented inspections (as part of the IS1 program) through the period of extended operation. There is no TLAA associated with Bulletin 88-08 as there are no analyses based on time-limited assumptions.

2.5.2 Effects of Reactor Water Environment on Fatigue Life Test data indicate that certain environmental effects (such as temperature, oxygen content, and strain rate) in the primary systems of light water reactors could result in greater susceptibility to fatigue than would be predicted by fatigue analyses based on the ASME Section Ill design fatigue curves. The ASME design fatigue curves were based on laboratory tests in air and at low temperatures. Although the failure curves derived from laboratory tests were adjusted to account for effects such as data scatter, size effect, and surface finish, these adjustments may not be sufficient to account for actual plant operating environments (Ref. 19).

As reported in SECY-95-245, the NRC believes that no immediate staff or licensee action is necessary to deal with the environmentally assisted fatigue issue. In addition, the staff concluded that it could not justify requiring a back fit of the environmental fatigue data to operating plants. However, the NRC concluded that, because metal fatigue effects increase

with service life, environmentally assisted fatigue should be evaluated for any proposed extended period of operation for license renewal.

NUREG/CR-6260 (Ref. 50), applied the fatigue design curves that incorporated environmental effects to several plants and identified locations of interest for consideration of environmental effects. Section 5.7 of NUREG/CR-6260 identified the following component locations to be most sensitive to environmental effects for WNPS vintage General Electric plants. These locations and the subsequent calculations are directly relevant to WNPS.

1. Reactor vessel shell and lower head

2. Reactor vessel feedwater nozzle

3. Reactor recirculation piping (including inlet and outlet nozzles)

4. Core spray line reactor vessel nozzle and associated Class 1 piping

5. Residual heat removal (RHR) return line Class 1 piping

6. Feedwater line Class 1 piping Entergy evaluated the limiting locations (a total of nine components corresponding with the above six locations) using the guidance provided in NUREG-I801 (Ref. 51, Volume 2,Section X.M.l). Seven of the nine components reviewed have an environmentally adjusted CUF of greater than 1.O (see Table 2.5-1). The ASME Code does not require environmental adjustment to fatigue analyses, let alone the extremely conservative evaluations done in this report. (The method used in this report essentially applies the worst case temperature and worst case strain rate to the full range of all transients. More refined finite element analyses could analyze each transient in sections, using realistic values for temperature and strain, and greatly reduce the predicted CUFs.)

However, there is an increased potential for fatigue cracking during the period of extended operation at locations having CUFs exceeding 1.O. Prior to entering the period of extended operation, for each location that may exceed a CUF of 1.O when considering environmental effects, WNPS will implement one or more of the following:

(1) further refinement of the fatigue analyses to lower the predicted CUFs to less than 1.O.

Such analyses would include calculation of site specific environmentally adjusted CUFs for all of the NUREG-6260 locations. (Projection of the TLAA per 10CFR54.21(c)(l)(ii).)

(2) management of fatigue at the affected locations by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method accepted by the NRC). (Aging effects managed per 10CFR54.21(c)(l)(iii).)

(3) Repair or replacement of the affected locations. (Aging effects managed per 10CFR54.21(c)(l)(iii).)

Should WNPS select the option to manage environmental-assisted fatigue during the period of extended operation, details of the program to be implemented such as scope, qualification, method, and frequency will be provided to the NRC prior to the period of extended operation.

The effects of environmental-assisted thermal fatigue for the limiting locations identified in

LRPD-04 VYNPS License Renewal Project Revision 0 TLAA Mechanical Fatigue Page 25 of 35 NUREG-6260 have been evaluated. Cracking by environmentally-assistedfatigue of these locations is addressed using one of the above three approaches in accordance with I O CFR 54.21(c)(l).

Table 2.5-1 VYNPS Cumulative Usage Factors for NUREGICR-6260 Limiting Locations Environmentally Adjusted CUF 0.98 2.86 1.50 1.99 6.09 4 Core Spray Nozzles 0.625 LAS 2.45 1.55 4 Core Spray Safe End 0.182' ss 15.35 2.79 5 RHR Return Piping 0.032' ss 15.35 0.49 6 Feedwater Piping 0.427' cs 3.01 1.29 1 No WNPS specific CUF is available; used value documented in NUREGCR-6260.

Carbon Steel The environmentally assisted fatigue correction factor (Fen) for carbon steel is calculated as follows.

Fen = exp(0.585-0.00124T-0.101 S'T*O*E*) (NUREGER-6583, Equation 6.5a)

T = 25 "C Ambient Temperature (Ref. 82) s'= s (0 < S (Sulfur) I 0.015 wt%)

S' = 0.015 ( S 2 O.O15wt%)

T*= 0 (Temperature (T) c 150°C)

T' = T-I 50 (T = 150-350°C) o*=0 (DO I 0.05 ppm)

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Paae 26 of 35 O'= ln(D0/0.04) (0.05 ppm DO I 0.5 ppm) 0*=ln(12.5) = 2.53 (DO > 0.5 ppm)

E*= 0 (strain rate ( E ) > l%/s)

E* = In(&) (0.001 I E I 1%/s)

E* = ln(O.001) & 0.001%/s)

For WNPS, the only carbon steel location in NUREGICR-6260 is the feedwater piping.

Therefore the carbon steel correction factor is calculated using FW temperature and dissolved oxygen. Sulfur content and strain rate are assumed at the values that yield the highest Fen.

S* = 0.015 - Assume bounding Sulfur > 0.015 wt%

T* = 49.72 - FW Temp 391.5OF(Ref., 28) 0' = 1.056 - 0.03 ppm <FW DO 0.2 ppm (Ref. 59)

E* = -6.91 - Assume bounding strain rate (0.001%/s)

Fen (Feedwater) = exp(0.55-(0.00124)(25)-(0.101)(0.015)(49.72)(1.056)(-6.91)) = 3.01 The adjusted CUF for the WNPS feedwater piping is shown in Table 2.5-1.

The feedwater line has an environmentally adjusted CUF greater than 1.O and requires a more rigorous management strategy as discussed above. Conservative assumptions were made regarding the sulfur content and strain rate of the carbon steel feedwater line. If less conservative values can be justified for these parameters, or if more detailed calculations are performed, the adjusted CUF for the feedwater line could be less than 1.O.

Low Alloy Steel The environmentally assisted fatigue correction factor (Fen) for low alloy steel is calculated as follows:

Fen = exp(0.929-0.00124T-0.101 S*T'O*E*) (NUREG/6583, Eq. 6.5b)

T=25 "C Ambient Temperature (Ref. 82) s'= s (0 c S (Sulfur) I 0.015 wt%)

S'= 0.015 (S1 O.O15wt%)

T'= 0 (Temperature (T) 150°C)

T' = T-I 50 (T = 150-350°C) o*=0 (DO I 0.05 ppm)

O'= ln(D0/0.04) 0.05 ppm DO I 0.5 ppm) 0'= ln(12.5) = 2.53 (DO> 0.5 ppm)

E*= 0 (strain ( E ) > 1%/s)

E*= In(&) (0.001 5 E I 1%/s)

Et= In(O.001) E c O.OOl%/s)

There are five subcomponents of low alloy steel in the 6 limiting locations at WNPS. A separate Fenwill be calculated for each location to account for materiaVenvironment differences.

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 27 of 35 RV material certs for shell plates typically > 0.01 5 ppm Sulfur RV material certifications (Ref. 60)

RV material certifications (Ref. 60)

RV material certifications (Ref. 60)

RV material certifications (Ref. 60)

Averaged FW inlet temp and saturation temp.

FW Temp 3915°F (Ref., 28)

Averaged FW inlet temp and saturation temp.

Averaged FW inlet temp and saturation temp.

Saturation temp = 547.9 O F .

- 0.03 ppm FW DO <0.2 ppm (Ref. 59)'

Oxygen target value is 0.015 ppm (Ref. 92)

- Assume bounding strain rate (0.001%)

Fen (lower head) = ~X~(O.~~~-(O.OOI~~)(~~)-(O.IOI)(0.015)(93.17)(0.000)(-6.91)) = 2.45 Fen (FWnonles) = exp(0.929-(0.00124)(25)-(0.101)(0.012)(49.72)(1.056)(-6.91)) = 3.81 Fen (RRoutnonles) = ~ X ~ ( ~ . ~ ~ ~ - ( ~ . ~ ~ ~ ~ ~ ) ( ~ ~ ) - ( ~ . ~ ~ ~ ) ( 0 . 0 1 5 ) ( =

9 2.45 3.17)(0.000)(-

Fen (cs nozzles) = exp(0.929-(0.00124)(25)-(0.101)(0.012)(93.17)(0.000)(-6.91)) = 2.45 The adjusted CUFs for the WNPS low alloy steel locations when considering environmental effects are shown in Table 2.5-1. Four of the five low alloy steel locations have environmentally adjusted CUFs greater than 1.O and require a more rigorous management strategy as discussed above. Conservative assumptions were made regarding the strain rate of the low alloy steel components. If a less conservative value can be justified, or if more detailed calculations are performed, the adjusted CUF for these components could be less than 1.O.

Austenitic Stainless Steel The environmentally assisted fatigue correction factor (Fen) for stainless steel is calculated as follows:

Fen = exp(0.935-T'O~') (NUREGiCR-5704, Equation 13)

T'=O (T 200OC)

T'=1 (T > 200OC) 0' = 0.26 (DO 0.05 ppm) 0' = 0.172 (DO 2 0.05 ppm)

E'=o (E > 0.4%/s)

E' = ln(~/0.4) (0.0004 I E 0.4%/s)

I E' = ln(0.0004/0.4) (E< 0.0004%/s)

For WNPS there are three stainless steel components in the locations of interest, the RR piping tee, the RHR return piping, and the core spray nozzle safe ends. Fen is calculated for all three locations as follows.

LRPD-04 VYNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 28 of 35 T(all) = 1 Temp = 547.9 O F or 469.7OF (Ref. 28)

O(a11)= 0.26 Oxygen target value is 0.015 ppm (Ref. 92)

E (all)= -6.91 Assume bounding strain rate Fen=exp(0.935-(1.0)(0.172)(-6.91)) = 15.35 The target DO of 15 ppb was used for all components. The fatigue factor for stainless steel actually goes up as DO decreases, so justifying a lower DO would make these factors larger.

The adjusted CUFs for the limiting VYNPS stainless steel locations when considering environmental effects are shown in Table 2.5-1. All stainless steel locations (the RHR tee, the RHR piping and the Core Spray safe end) have environmentally adjusted CUFs greater than 1.O and require a more rigorous management strategy as discussed above. Conservative assumptions were made regarding the strain rate of the stainless steel components. If a less conservative value can be justified, or if more detailed calculations are performed, the adjusted CUF for these components could be less than 1.O.

LRPD-04 VYNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 29 of 35 3.0 Summary and Conclusions This report reviewed the time-limited aging analyses associated with metal fatigue of Class 1 and non-Class 1 components and fracture mechanics evaluations for flaws discovered in mechanical components. This report documents that WNPS time-limited aging analyses related to fatigue on Class 1 and non-Class Icomponents have been appropriately evaluated for the period of extended operation in accordance with the requirements of 10CFR54.21(c).

Two specific items of interest are noted below.

None of the monitored transients are projected to exceed the allowable cycle limits before the end of 60 years. The Fatigue Monitoring Program, once enhanced as discussed in LRPD-02 (Ref. 24), will provide an adequate program to count, evaluate, track and trend cycles and to provide corrective actions if limits are approached. For a more detailed discussion of this program see LRPD-02.

The effects of reactor water environment on fatigue life were identified and adjusted CUFs were calculated. Several components will exceed a CUF of 1.O when environmental effects are considered. This requires enhanced fatigue management that must be resolved before the period of extended operation.

The time-limited aging analyses that are associated with fatigue have been reviewed to ensure acceptability for the period of extended operation. All three of the methods listed in 10 CFR 54.21(c)(l)(i) through (iii) for demonstrating acceptability of TLAA for license renewal have been utilized for different fatigue evaluations. The combination of these efforts show that all fatigue related TLAA are adequately addressed for the period of extended operation. A summary of the results is provided in Table 3-1.

Table 3-1: Summary of Fatigue TLAA Section of this Report Component(s) Result 2.1.2 Cumulative Usage Factors (CUFs) 10 CFR 54.21(c)(i)

I 2.1.3 I Reactor pressure vessel I 10 CFR 54.21(c)(i) 1 2.1.4.1 I Reactor vessel internals I NOTM 2.1.4.2 Shroud Repair 10 CFR 54.21(c)(i) 2.1 5.1 Reactor recirculation system 10 CFR 54.21(c)(i)

I 10 CFR 54.21(c)(i)

I 2.1 5.3 I FIV for MS and FW components I NoTLAA 12.2.1 I Non-Class Ipiping and in line components I 10 CFR 54.21(c)(i)

I 2.2.2 I Non-Class 1 vessels, heat exchangers, tanks and pumps( No TLAA

LRPD-04 WNPS License Renewal Project Revision 0 TLAA Mechanical Fatigue Page 30 of 35 2.3.1 Torus 10 CFR 54.21(c)(ii) and No TLAA 2.3.2 SRV discharge lines 10 CFR 54.21(c)(i) and 10 CFR 54.21(c)(ii) 2.3.3 Torus Attached Piping (TAP) 10 CFR 54.21(c)(ii)

, 2.4.1 Jet Pump Riser flaws No TLAA 1 2.4.2 Core Spray piping No TLAA F V p l a t e 1-15 flaw I 10 CFR 54.21(c)(i) 2.4.4 Core Spray weld overlay No TLAA 2.4.5 Containment Corrosion Not a TLAA 2.4.6 Containment Localized Thinning 10 CFR 54.21(c)(i) 2.5.1 Bulletin 88-08 response No TLAA

~

2.5.2 EnvironmentallyAssisted Fatigue 10CFR 54.21(c)(ii) or 10CFR 54.21(c)(iii)

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical1 r-&:-.-.-

rariyut:

Page 31 of 35 1 4.0 References

1. LRPG-01, License Renewal Project Plan.

2. Project Report LRPD-03, TLAA and Exemption Evaluations.

3. Vermont Yankee Nuclear Power Station, Final Safety Analysis Report, Revision 17.

4. Vermont Yankee Nuclear Power Station, Technical Specifications.

5. AMRM-31, Aging Management Review of the Reactor Pressure Vessel

6. AMRM-32, Aging Management Review of the Reactor Vessel Internals

7. AMRM-33, Aging Management Review of the Reactor Coolant System

8. VYNPS Administrative Procedure AP-0145, Equipment Cycle Record Keeping, Revision 8, Ipc 3, 11/29/02

9. WNPS Calculation WC-0378, Vermont Yankee Reactor Cyclic Limits for Transient Events, IO. WNPS Calculation W C 23A5569, Recirculation System Stress Analysis Loop A, Rev.0, ccn 2, 11/07/2002

11. WNPS Calculation W C 23A5570, Recirculation System Stress Analysis Loop B, Rev. 0, ccn I,11/07/2002

12. Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, January 25, 1988.

13. NUREG-0313, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, Revision 2, January 1988.

14. B W 96-48,7 August 1996, R. E. Sojka to USNRC Document Control Desk, Response to Request for Additional Information Regarding Vermont Yankee Core Shroud Modification

15. MPR Calculation 2499502-601 Stress Analysis of Shroud Support Plate, Revision 1, 28 March 1996

16. GE calculation MPR-751, Mark 1 Containment Program Augmented Class 2/3 Fatigue Evaluation Method and Results for Typical Torus Attached and SRV Piping Systems, November 1982

17. 1982 ASME Boiler and Pressure Vessel Code, Division 1, Subsection NB, Class 1 Components.

18. EPRI Report 1003056, Revision 3, Non-Class 1 Mechanical Implementation Guideline and Mechanical Tools.

19. GSI 190, Fatigue Evaluation of Metal Components for 60-year Plant Life.

20. SECY-95-245, Completion of the Fatigue Action Plan.

21. NRC Bulletin No. 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems, including Supplements 1, 2, and 3.

LRPD-04 WNPS License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 32 of 35

22. 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Part 55a, Codes and Standards, (9) lnservice Inspection Requirements.

23. I O CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Part 36, Technical Specifications.

24. Project Report LRPD-02, Aging Management Program Evaluation Results, Rev. 0

25. 10 CFR Part 54, Requirementsfor Renewal of Operating Licenses for Nuclear Power Plants, Part 21, Contents of Application - Technical Information.

26. lnservice Discrepancy Report, IDR# 01-10, 05/08/01, RS-1 Recirc Riser Welds K, B, and C

27. B W 96-118, J. J. Duffy to USNRC Document Control Desk, Core Spray System Inspection at Vermont Yankee, 9 October 96

28. WNPS letter B W 03-80, J. K. Thayer to USNRC Document Control Desk, Sept I O ,

2003, Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No.

50-271) Technical Specification Proposed Change No. 263 Extended Power Uprate

29. AMRM-01 , Standby Liquid Control System.

30. AMRM-02, Residual Heat Removal System.

31. AMRM-03, Core Spray System.

32. AMRM-04, Automatic Depressurization System.

33. AMRM-05, High Pressure Coolant Injection System.

34. AMRM-06, Reactor Core Isolation Cooling System.

35. AMRM-07, Standby Gas Treatment System.

36. AMRM-08, Primary Containment Atmosphere Control System.

37. deleted

38. AMRM-11, Service Water System.

39. AMRM-12, Reactor Building Closed Cooling Water Systems.

40. AMRM-13, Emergency Diesel Generator System.

41. AMRM-14, Standby Fuel Pool Cooling System.

42. AMRM-15, Fuel Oil System.

43. AMRM-16, Instrument Air System.

44. AMRM-17, Fire Protection - Water System.

45. AMRM-18, Fire Protection - C02 Systems.

46. AMRM-19, Heating, Ventilation and Air Conditioning Subsystems.

47. AMRM-20, Primary Containment Penetrations.

48. AMRM-26, Main Condenser and MSlV Leakage Pathway.

49. deleted

WNPS License Renewal Project LRPD-O~ --

Revision 0 TLAA - Mechanical Fatigue Page 33 of 35

50. NUREG/CR-6260, Application of NUREGKR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components.

51. NUREG-1801, Generic Aging Lessons Learned, Revision 1, September, 2005

52. NUREGKR-5999, Interim Fatigue Design Curves for Carbon, Low-Alloy, and Austenitic Stainless Steels in LWR Environments, 1993.

53. NUREG/CR-6717, Environmental Effects on Fatigue Crack Initiation in Piping and Pressure Vessel Steels, May 2001.

54. Reactor Vessel Record, Exhibit C, GE Purchase Specification 21A1115, Rev 3, 7/14/69

55. WNPS Calculation WC-378, Rev. 1, 6 April 88, Vermont Yankee Reactor Cyclic Limits for Transient Events

56. AMRM-21, John Deere Diesel

57. Entergy Condition Report LO-VTYLO-2004-00253, Complete the end of cycle report per AP 0145, Closed 11/29/2004

58. WNPS report LRPD-02, Aging Management Program Evaluation Results, Rev. 0

59. W Procedure OP4612, Reactor Water System Sampling and Treatment, Rev 25, July 21,2005

60. Pressure Vessel Record, Exhibit D, Certified Test Reports Boiling Water Nuclear Reactor Vessel 17.167 x 63.167 Ins. Hds.

Manufacturers Serial No. B-4698 Vermont Yankee Project, Vernon, VT.

G.E. CO. P.O. 205-55565-1 CB&I contract 9-2601

61. GE-NE-0000-0009-9951-01, Project Task Report, Entergy Nuclear Opefations Incorporated Vermont Yankee Nuclear Power Station Extended Power Uprate Task 302: Reactor Vessel Integrity - Stress Evaluation, September, 2003

62. WVY 78-0,R. H. Groce to USNRC Office of Nuclear Reactor Regulation, Results of the Feedwater and Control Rod Drive hydraulic return line nozzle inspection 20 January 1978

63. WNPS Calculation WC-216, Fatigue Analysis of RWCU Non-standard Tee, Rev 0, 01/19/1984

64. N W 98-155, C. 0. Thomas to G. A. Maret, 9 Nov 98, Request to use Code Case N560 as an alternative to the requirements of ASME Code,Section XI, Table IWB-2500-1 at Vermont Yankee Nuclear Power Station (TAC No. M99389)

65. N W 99-76, R. P. Croteau to R. J. Wanczyk, 13 August 99, Vermont Yankee Nuclear Power Station - Issuance of Amendment RE: Clarification of lnservice Inspection requirements (TAC No. MA5281)

66. N W 91-167, 19 September 1991, M. B. Fairtile to L. A. Tremblay, NRC Bulletin 88-08 Thermal Stresses in Piping connected to Reactor Coolant systems (TAC No.

69702)

LRPD-04 W N P S License Renewal Project Revision 0 TLAA - Mechanical Fatigue Page 34 of 35

67. F W 86-036, W. P. Murphy to D. R. Muller (NRR), 5 May 1986, Core Spray Nozzle Weld Overlay

68. FVY 88-19, W. P. Murphy to USNRC Document Control Desk, 1 March 1988, Long-term operation with Core Spray Safe End Nozzle Weld Overlays

69. N W 88-080, Vernon L. Rooney to R. W. Capstick, 9 May 1988, Core Spray Safe-End Inspection (TAC No. 67522)

70. BVY 98-67, 4 May 1998, Jet Pump Riser Circumferential Weld Inspections

71. B W 99-43, D. M. Leach to USNRC Document Control Desk, Vermont Yankee Nuclear Power Corporation, License No. DPR-28 (Docket No. 50-271), Jet Pump Riser Circumferential Weld Inspections and Flaw Evaluation, 29 March 1999

72. N W 99-46, R. P. Croteau to G. A. Maret, 29 April 1999, Jet pump riser circumferenctial weld inspections at Vermont Yankee Nuclear Power Station (TAC No. MA5109)

73. Curator document RFO 22, Memorandum from C. B. Larsen to D. C. Girroir, 9 May 2001, Jet pump NDE Uncertainty, includes copy of GE report JXOAL, May 2001.

74. B W 96-119, 9 October 1996, J. J. Duffy to USNRC Document Control Desk, Reactor Pressure Vessel Inspection at Vermont Yankee

75. NRC Letter dated 1IOctober 1996, C. C. Harbuck to D. A. Reid, Evaluation of Flaw Indicationfound during reactor pressure vessel inspections at Vermont Yankee Nuclear Power Station (TAC No. M96670)

76. Technical Evaluation No. 2001-029, Revision 1, 1/30/2003, Evaluation of Internal Core Spray Piping Flaws

77. TR-5319-1, Technical Report (Teledyne) Mark 1 Containment Program Plant Unique Analysis of the Torus Suppression Chamber for Vermont Yankee Nuclear Power Station, Revision 2, 30 November 1983

78. F W 84-38, J. B. Sinclair to D. B. Vassalo, Mark I Containment Program, 27 April 1984

79. Technical Report TR-5319-2, Mark I Containment Program Plant Unique Analysis Report of the Torus Attached Piping for Vermont Yankee Nuclear Power Plant, 3 October 1993

80. NUREG-0661, Safety Evaluation Report, Mark 1 Containment Long-Term Program.

November, 1981.

81. Project Aging Management Review AMRM-30, Non-safety related components affecting safety-related components

82. BWRVIP letter (EPRI letter 2005-271), Potential Error in Existing Fatigue Reactor Water Environmental Effects Analyses, July 1 2005

83. APED-5453, Wetzel, V. R., Duckwald, C. S., and Head, M. A., Vibration Analysis and Testing of Reactor Internals, General Electric Company, Atomic Power Equipment Department, April 1967

84. LRPD-02, Aging Management Program Evaluation Results

I I I n n n nr I TL

85. License Renewal Application, Dresden Nuclear Power Station, Docket Nos. 50-237 AND 50-249, Facility Operating License Nos. DPR-19 and DPR-25, Quad Cities Nuclear Power Station, DOCKET Nos. 50-254 AND 50-265, Facility Operating License Nos. DPR-29 and DPR-30, January 2003

86. NRC Letter, Pao-Tsin Kuo to John L. Skolds, License Renewal Safety Evaluation Report for the Dresden and Quad Cities Nuclear Power Stations, February 12, 2004

. 87. Safety evaluation by the office of Nuclear Reactor Regulation related to Amendment No. -to facility operating license no. DPR-28, Entergy Nuclear Vermont Yankee, LLC And Entergy Nuclear Operations, Inc., Vermont Yankee Nuclear Power Station, Docket NO. 50-271, DRAFT, Revision 1November 2,2005

88. WNPS Calculation WC-2400, Evaluation of Jet Pump Riser Flaws at weld N2K-RS-I,Rev. 0, 03/09/2005

89. WNPS Engineering Report W-RPT-05-00? 5, Core Spray Piping Weld P8.P9 Evaluation, Rev. 0, 06/20/2005

90. WNPS Technical Evaluation TE 2001-25, Evaluation of ASME Section XI - IWE Findings for IDR 01-07 and IDR 01-08, 2001

91. WNPS Calculation VYC-2043, Evaluation of Primary Containment Localized Thinning and Screening Criteria for ASME XI-IWE Inspections

92. GE Service Information Letter, SIL No. 408, Implementation of Hydrogen Water Chemistry is Recommended in all BWRs, July 9,1984.

93. email from Jim Fitzpatrick to Dave Lach, FW: LRPD Metal Fatigue, 19 December 2005

94. VY Owners Activity Report (OAR-1) for Inservice Inspections October 25,2002 through May 4,2004