ML070680173

LRPD-03, Revision 0; TLAA and Exemption Evaluation Results, 01/25/2006

ML070680173
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	01/25/2006
From:	Finnin R J, Lach D J, Rinckel M A Entergy Nuclear
To:	Office of Nuclear Reactor Regulation
References
LRPD-03, Rev 0
Download: ML070680173 (98)

Text

VERIFEATION OF WNPS LICENSE RENEWAL PROJECT REPORT Ti& of Report: ReportNumber.

-43 7'iAA and Exemption Evaluation Rerrub Revision:

0 This report documents evaluations related to the WNPS license renewal project Si~res certify that the reportwas prepared, checked and reviewed by the License Renewal Project T' h accordance widh fhe WNPS license renewal pm@t guidelines and that it was approved by the EN1 License Renewal Project Manager and the WNPS Manager, Engineering prefects.

License RenewalProjectTeam signatures also certifythatareviewforbetemtinlngpotential impact to other ticense renewal documents, based on pmWs revisions, was conducted for this revisii. Other document&)

impacted by this revision: - Yes, See Attachment - x No Reviewed by: Date:

LRPD-03 Revision 0 Page 2 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations REVISION DESCRIPTION SHEET Revision Number Description Pages and/or Sections Revised LRPD-03 Revision 0 Page 3 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations Table of Contents I . 0 2.0 2.1 2.2 3.0 3 .I 3.2 3.3 3.4 3.5 3.6 3.7 3.8 4.0 4.1 4.2 5.0 6.0 Introduction

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4 Identification of TLAA and Exemptions

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6 Identification of TLAA .........................................................................................

6 Evaluation of TLAA ..........................................................................................

8 Reactor Vessel Neutron Embrittlement

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8 Metal Fatigue

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22 Environmental Qualification of Electrical Equipment

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22 Concrete Containment Tendon Prestress

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27 ASME Section XI Inservice Inspection

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29 TLAA in BWRVIP Documents

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30 Other Plant-Specific TLAA ...............................................................................

34 Fire Protection Requirement Exemptions

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37 Other 10 CFR Requirement Exemptions

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40 Identification of Exemptions

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7 Metal Corrosion Allowance

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27 Identification and Evaluation of Exemptions

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36 Summary and Conclusions

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42 References

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43 Attachment 1 . List of Potential TLAA and References

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51 Attachment 2 . List of Exemptions and References

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55 Attachment 3 . UFSAR TLAA Search Results ...........................................................

56 Attachment 4 . Estimate of Crane Cycles

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94 LRPD-03 Revision 0 Page 4 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations I .O Introduction This report is part of the integrated plant assessment (IPA) performed to extend the operating license of Vermont Yankee Nuclear Power Station (WNPS). This report reviews the time- limited aging analyses (TLAA), and exemptions to Part 10 of the Code of Federal Regulations (10 CFR), and evaluates them for the period of extended operation as required by 10 CFR 54. For additional information on the license renewal project and associated documentation, refer to the license renewal project plan. (Ref.

6.6.1) The Code of Federal Regulations 10 CFR 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants" governs the issuance of renewed operating licenses for nuclear power plants and includes requirements for the review of time-limited aging analyses (TLAA). The definition of time-limited aging analyses (TLAA) is in 10 CFR 54.3.

Time-limited aging analyses, for the purposes of this part, are those licensee calculations and analyses that:

I) involve systems, structures, and components within the scope of license renewal, as delineated in §54.4(a); 2) consider the effects of aging;

3) involve time-limited assumptions defined by the current term of operation, for example, 40 years;

4) were determined to be relevant 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, as delineated in §54.4(b); and

6) are contained or incorporated by reference in the current licensing basis.

An example of a TLAA is the reactor vessel neutron embrittlement analysis that is based on the neutron exposure for the current operating term and must be reevaluated for the period of extended operation.

Section 10 CFR 54.21(c) requires a list of time-limited aging analyses (TLAA) in the application for a renewed license.

Section 10 CFR 54.21 (c)(2) requires a list of current exemptions to 10 CFR 50 based on TLAA in the application for a renewed license.

954.21 Contents of application

-- technical information. (c) An evaluation of time-limited aging analyses.

(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; or ii) the analyses have been projected to the end of the period of extended operation; or LRPD-03 Revision 0 Page 5 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation In addition, IO CFR 54 states that a list must be provided of plant-specific exemptions granted (and still in effect) pursuant to 10 CFR 50.12 that are based on time-limited aging analyses as defined in 10 CFR 54.3. An applicant must provide an evaluation that justifies continuation of these exemptions for the period of extended operation.

TLAA and exemptions are discussed in this document and a reference is provided to supporting site documents. The methods used for identification and evaluation of TLAA and exemptions are described in Section 2.0 with the identified TLAA listed in Attachment

1. Identified exemptions are listed in Attachment

2. TLAA search results from the Updated Final Safety Analysis Report (UFSAR) and recommended UFSAR text changes are included in Attachment

3. The potential TLAA are evaluated in Section 3.0 while the WNPS exemptions based on TLAA are evaluated in Section 4.0. A summary description of the evaluation of TLAA for the period of extended operation will be provided in the UFSAR supplement.

VYNPS License Renewal Project TLAA and Exemption Evaluations LRPD-03 Revision 0 Page 6 of 98 2.0 Identification of TLAA and Exemptions

2.1 Identification

of TLAA The process used to identify the time-limited aging analyses is consistent with the guidance provided in NE1 95-1 0, lndustry Guidelines for lrnplernenfing the Requirernenfs of 70 CFR 54 - The License Renewal Rule, Revision 6, June 2005. Calculations and analyses that could potentially meet the definition of 1 OCFR 54.3 were identified by searching CLB documents including the following. Technical Specifications UFSAR docketed licensing correspondence fire protection program documents NRC safety evaluation reports BWRVIP documents Industry documents that list generic time-limited aging analyses were also reviewed to provide additional assurance of the completeness of the plant-specific list. These documents included NE1 95-10; NUREG-1800, Standard Review Plan (SRP) for Review for License Renewal Applications for Nuclear Power Plants, Revision I, September 2005; NUREG-1801 , Generic Aging Lessons Learned (GALL) Report, Revision 1, September 2005; and NRC safety evaluation reports related to license renewal applications by other BWR licensees.

Industry documentation, owners group reports, vendor reports, and site searches were utilized to identify TLAA that are applicable to WNPS. EPRl reports such as TR-105090 (Ref.

6.3.9) and other license renewal applications (Ref. 6.3.2, 6.3.3, 6.3.4) were used to identify generic TLAA. Site-specific evaluations ensured TLAA applicability. During preparation of the WNPS Class 1 and non-Class 1 aging management review reports, TLAA were identified in individual reports. The TLAA identified in individual reports are evaluated in this report or in LRPD-04, TLAA - Mechanical Fatigue.

A computer database search was performed to identify TLAA from the UFSAR (Ref.

6.5.1), the Technical Specifications (Ref. 6.5.4), the QA program (Ref. 6.5.6), the ASME Section XI lnservice Inspection Program including all Relief Requests (Ref. 6.5.58), the Fire Hazards Analysis (Ref.

6.5.7), the Fire Protection Program Commitment Reference Manual (Ref.

6.5.8), and available NRC correspondence (Ref.

6.5.10). The search criteria utilized key words and phrases such as age, aging, crack growth, corrosion allowance, cycles, cyclic, embrittlement, EFPY, fatigue, 40 years, life (design life, service life), RTND~, time limit, usage factor. The Vermont Yankee Fire Protection and Appendix R Program (Ref. 6.5.9) was reviewed manually.

The key word search of the aforementioned CLB documentation resulted in a list of potential TLAA. Attachment 1 lists the resulting potential TLAA and the documents referencing the potential TLAA. The TLAA identified by the various searches were consolidated. For example, the database search identified a number of reactor vessel neutron embrittlement analyses that were TLAA (e.g., RTNDT and CvUSE analyses). Section 3.1 of this report discusses the review of the reactor LRPD-03 Revision 0 Page 7 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations vessel neutron embrittlement TLAA. This is consistent with the license renewal application format and content guidelines presented in NE1 95-10. 2.2 Identification of ExemDtions A review of docketed correspondence identified VYNPS exemptions.

No WNPS exemptions depend on time-limited aging analyses.

To identify exemptions for WNPS, a keyword search was conducted on the UFSAR, Technical Specifications, and NRC correspondence.

This review involved a search of the database to identify exemptions that were granted pursuant to 10 CFR 50.1 2. The search criteria utilized key terms including "50.12and "exemption." Attachment 2 lists the identified exemptions and lists references for the exemptions. In accordance with 10 CFR 54.21 (c)(2), exemptions that are not in effect (Le., exemptions that were temporary or have been eliminated/withdrawn by later correspondence) are not discussed in the license renewal application and are not discussed further in this report.

LRPD-03 Revision 0 Page 8 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations I Page8of98 TLAA and Exemptioh rvaiuatioris I 3.0 Evaluation of TLAA Attachment 1 of this document summarizes potential TLAA applicable to WNPS. In the rest of Section 3, each potentia1 TLAA identified in Attachment 1 was examined to determine if it meets the definition of a TLAA in accordance with 10 CFR 54.3. Analyses and calculations that meet the TLAA definition are evaluated in accordance with the options provided in IO CFR 54.21 (c)(l). 3.1 Reactor Vessel Neutron Embrittlement The regulations governing reactor vessel integrity are in 10 CFR 50. Section 50.60 requires that all light-water reactors meet the fracture toughness, pressure-temperature limits, and material surveillance program requirements for the reactor coolant boundary as set forth in Appendices G and H of 10 CFR 50. The WNPS current licensing basis analyses evaluating reduction of fracture toughness of reactor vessel for 40 years are TLAA. The reactor vessel neutron embrittlement time-limited aging analyses were projected to the end of the period of extended operation (54 EFPY) in accordance with 10 CFR 54.21 (c)(l)(ii) as summarized below.

The WNPS current licensing basis contains calculations and analyses that address the effects of neutron irradiation embrittlement on the reactor vessel (Refs.

6.5.3, 6.5.1 1, 6.5.12, and 6.5.44). The analyses evaluating reduction of reactor vessel fracture toughness for 40-years are TLAA. The appropriate calculations have been updated based on a 60-year operating term assuming that licensed activities will continue to be conducted in accordance with the CLB. The Reactor Vessel Surveillance Program described in WNPS Report LRPD-02, "Aging Management Program Evaluation Report" will ensure that the time-dependent parameters used in the TLAA described below remain valid through the period of extended operation.

The reactor vessel neutron embrittlement TLAA was projected in Ref. 6.5.44 to approximately 51.6 EFPY. 51.6 EFPY was used in support of extended power uprate, based on actual EFPY before the uprate and an assumed capacity factor of 90% after the uprate. For license renewal, the fluence was extrapolated to 54 EFPY, as discussed in Section 3.1 .I below. Upper shelf energy (CVUSE) was calculated based on the 54 EFPY extrapolated fluence to demonstrate that IO CFR 50 Appendix G requirements are satisfied.

Section 3.1.2 below discusses the results.

Adjusted reference temperature has been calculated based on the 54 EFPY extrapolated fluence, and the results are presented in section 3.1.3 below. The currently licensed P-T limit curves remain bounding for the period of extended operation, including the extended power uprate (Ref.

6.5.61). See section 3.1.4 below for more detail.

3.1.1 Reactor

Vessel Fluence GE's Licensing Topical Report NEDC-32983P-A, which was approved by the NRC for licensing applications in Reference 6.2.31, documents the method used for the neutron flux calculation. The NRC found that, in general, this method adheres to the guidance in Regulatory Guide 1.190 for neutron flux evaluation.

The calculated reactor vessel ID fluence for 51.6 EFPY is 5.16~10" n/cm2 (E>1 MeV), assuming a power uprate from 1593 MWt to 1912 MWt (Ref. 6.5.44). The LRPD-03 VYNPS License Rer---' m--'--' TLAA and Exemption rvaiuarions I Page9of98 I Revision 0 neutron flux distribution was calculated based on the three-dimensional flux synthesis of two separate two-dimensional flux solution calculations performed in an (r,z) and an (r,O) model. These flux solution calculations use the two-dimensional discrete ordinates code DORTGOIV, which is a controlled version of DORT in the GE Engineering Computation Program (ECP) library. Extrapolated to 54 EFPY, the vessel surface (ID) fluence is 5.39 x10d7 n/cm2(E>l MeV). The fluence was extrapolated by simply extending the straight line between 33 EFPY and 51.6 EFPY to 54 EFPY. Using Re ulato Guide I .99, Revision 2, Equation (3), results in a 54 EFPY %T fluence of 3.98~10' n/cm . gY The beltline is defined by IO CFR 50 Appendix G, Fracture Toughness Requirements as the region of the reactor pressure vessel that directly surrounds the effective height of the active core and adjacent regions of the reactor pressure vessel that are predicted to experience sufficient neutron irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage. In addition, 10 CFR 50 Appendix H does not require material surveillance testing for ferritic materials unless the peak neutron fluence at the end of the design life exceeds I .OxIOd7 n/cm2. The beltline is thus considered the reactor pressure vessel ferritic materials with an end-of-life fluence that exceeds 1 .Ox1 Oq7 n/cm2. At VYNPS, the beltline for 40-years consists of four plates (1-14, 1-15, 1-16, 1-17) and their connecting welds, all adjacent to the active fuel zone. There are no nozzles in the beltline region (Ref. 6.2.1). The beltline has been re-evaluated for 60 years using the axial distribution of fast fluence at the RPV wall (Figure 3-7 of Ref. 6.5.44). Based on the additional fluence incurred during the period of extended operation, the vertical section of the reactor vessel ID that will receive greater than I x I Od7 n/cm2 extends from

3.5 inches

below the bottom of the active fuel to 10 inches above the top of the active fuel. There are no nozzles in this region. Based on drawing 5920-3773 (Ref. 6.5.63), this is equivalent to a vessel height of 204 inches to 361.5 inches. This same drawing shows that the centerline of the recirculation inlet nozzles is at 186 inches. The top of the nozzle weld is not specifically shown on the drawing, but can be approximated as 202 inches. Above the core, the nearest nozzles are the instrumentation nozzles (NI IA, N1 IB, N12A and N12B) at 422 inches. No nozzles will be added to the beltline region by additional fluence incurred during the period of extended operation at the uprated power. The limiting plate and weld material in the beltline for 40-years remain the limiting materials for the period of extended operation. Fluence, calculated based on the operating term, is a time-limited assumption for the TLAA that evaluate reactor vessel embrittlement.

The reactor vessel fluence calculation has been projected to the end of the period of extended operation and that result is used throughout the remainder of Section 3.1 of this report. 3.1.2 Pressureme m pe rat u re Limits Appendix G of 10 CFR 50 requires the reactor vessel to remain within established pressure-temperature (P-T) limits during reactor vessel boltup, hydrotest, pressure tests, normal operation, and anticipated operational occurrences. These limits are from calculations that use the materials and fluence data obtained through the reactor vessel surveillance program. Normally, the pressure-temperature limits are calculated for several years into the future.

LRPD-03 Revision 0 Page IO of 98 WNPS License Renewal Project TLAA and Exemption Evaluations In March 2003 (Ref. 6.5.1 6), WNPS submitted a license amendment request to change the P-T limits to incorporate data from analysis of the first WNPS surveillance capsule and to extend the curves to 32 EFPY. The NRC approved this submittal as Amendment 21 8 to the WNPS license (Ref. 6.2.1 1). As stated in that SER, VYNPS used conservative values for determining the P-T limits. Those values were peak vessel fluence of 1

.24x101* n/cm2, '/4 T ART of 89°F and a % T ART of 73°F. Table 3-5 of this report compares the bases for the present curves with the projected fluence and ARTS for 54 EFPY and shows that the projected values at 54 EFPY (fluence of 5.39 xlOI7 n/cm2, '/4 T ART of 685°F and a % T ART of 56.9"F) are still less that those used for the P-T curves. As such the TLAA for Pressure Temperature limits remains valid in accordance with lOCFR54.21 (c)(l)(i).

WNPS will submit a technical specification change request prior to 32 EFPY to officially update the curves in the Technical Specifications.

Even though the curves may be the same, the applicable EFPY will be changed. 3.1.3 Charpy Upper Shelf Energy (CVUSE)

Appendix G of 10 CFR 50 requires that reactor vessel beltline materials "have Charpy upper- shelf energy . . . of no less than 75 ft-lb initially and must maintain Charpy upper-shelf energy throughout the life of the vessel of no less than 50 ft-lb...." Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"

provides two methods for estimating Charpy upper-shelf energy (CVUSE) at end of life. Position 1 applies for material that does not have surveillance data. Position 2 applies for material with surveillance data. Position 2 requires a minimum of two sets of credible surveillance data. Since WNPS has data from only one material surveillance capsule, Position 2 does not apply.

For Position 1, the percent drop in CvUSE for a stated copper content and neutron fluence is determined by reference to Figure 2 of Regulatory Guide 1.99, Revision 2. This percentage drop is applied to the initial CvUSE to obtain the adjusted CVUSE. Table 3.1 calculates the end of life CvUSE by this method. Safety analysis report NEDC-33090P (Ref.

6.5.3) documents the most recent calculations of CVUSE. NEDC-33090P was submitted to the NRC as part of the WNPS power uprate request (Ref. 6.5.2). Analyses were done for 32 EFPY at the previously licensed power level of 1598 MWt, and for 33 and 51.6 EFPY with a power uprate to 1912 MWt at 25 EFPY. Results of NEDC-33090P are extrapolated to 54 EFPY in this report. The WNPS unirradiated surveillance specimens were from plate 1-14 with a CvUSE of 89 ft-lb (137 ft-lb times 0.65) (Ref. 6.5.14). The 54 EFPY CvUSE value for plate 1-14 was calculated using Regulatory Guide 1.99, Position I, Figure 2. Specifically, the formulae for the lines were used to calculate the percent drop in &USE (Ref.

6.2.9). The calculation used the fluence determined in Section 3.1.1 above. For 54 EFPY, Table 3-1 shows the minimum projected CvUSE for platel-14 remains above the 50 ft-lb requirement of Appendix G of 10 CFR 50. As such, this TLAA has been extrapolated for the period of extended operation in accordance with 10CFR54.21 (c)(l)(ii). Initial (un-irradiated) upper shelf energy data for the weld materials and for plates 1-15, 1-16, and 1-17 do not exist. (Ref.

6.5.14) The BWR Owners Group prepared an equivalent margins analysis for plants without this data. The analysis (NEDO-32205-A, -Ref. 6.5.13) used Code case N-512. The NRC reviewed and accepted the evaluation, as documented in the SER in LRPD-03 Revision 0 Page I1 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations Ref. 6.2.32. Rather than calculating an end of life CvUSE (impossible without an initial CvUSE) a plant may calculate the percent drop in CVUSE, and show that the percent drop is less than the percent drop in the equivalent margins analysis.

Appendix B of BWRVIP-74 provides a method to evaluate USE at 54 EFPY using plant-specific surveillance data. BWRVIP-74 gives allowable percent drops in CvUSE of 23.5% for BWR 3-6 plate and from 39% for welds. The NRC approved the use of these new values in their SER (Ref. 6.2.24). Table 3-4 uses the BWRVIP-74 method to verify that the WNPS reductions in USE remain less than the reduction calculated in the BWRVIP-74 equivalent margins analyses at 54 EFPY for beltline welds and plates 1-15, 1-16, 1-17. As such, this TLAA has been projected to the end of the period of extended operation in accordance with 10CFR54.21 (c)(l)(ii).

3.1.4 Adjusted

Reference Temperature Irradiation by high-energy neutrons raises the value of adjusted reference temperature (ART) for the reactor vessel.

The initial RTNDT is determined through testing un-irradiated material specimens. The shift in reference temperature, ARTNDT, is the difference in the 30 ft-lb index temperatures from the average Charpy curves measured before and after irradiation. (ART)

= RTNDT + ARTNDT + margin. (Regulatory Guide 1.99, Revision 2) Safety analysis report NEDC-33090P (Ref.

6.5.3) includes the most recent calculations of RTNDT. NECD-33090P was submitted to the NRC as part of the WNPS power uprate request (Ref. 6.5.2). The report calculated the adjusted RTNDT for the welds and plates. Analyses were completed for both 32 EFPY at the previously licensed power level of 1598 MWt and for 33 EFPY with a power uprate to 1912 MWt at 25 EFPY. In addition to new fluence values, this report provided initial RTNDT for each plate, rather than the maximum plate value found in the Reactor Vessel Integrity Database. Results of NEDC-33090P are extrapolated to 54 EFPY in this report. Regulatory Guide 1.99 Revision 2, Regulatory Position 1 , defines the calculation methods used for ARTNDT and ART. WNPS response to GL 92-01 (Ref. 6.5.15) included chemistry data. Chemistry factors (CF) were interpolated from Table 1 in RG 1.99. Initial RTNDT values and standard deviations were taken from VYNPS NEDC-33090P, Table 3-2a. Standard deviations for ARTNDT, oA, were calculated as one-half the ARTNDT since in all instances

0.5 times

the ARTNDT was less than 28 OF for welds and 17 OF for plates. Margins were calculated as twice the square root of the sum of the squares of the two standard deviations. Note that adjusted reference temperatures use %T fluence. Section 3. I .I discussed calculation of fluence. Fluence factors (FF) were calculated using Equation 2 in Regulatory Guide 1.99, Revision 2. Extrapolated ARTNDT values were calculated by multiplying the CF and the FF for each plate and weld. The initial RTNDT, the calculated ARTN~T and the calculated margins were then added to get the new value of ART. Table 3-3 shows the 54 EFPY values of ART. As indicated in the table, the plates remain the limiting subcomponents rather than the welds; and Plate 1-14 remains the limiting plate. All calculated values are well below the 200 OF suggested in Section 3 of Regulatory Guide 1.99 and are thus acceptable for the period of extended operation.

The TLAA for RTNDT is thus projected through the period of extended operation in accordance with 1 OC FR54.2 1 (c)( 1 ) (i i).

WNPS License Renewal Project TLAA and Exemption Evaluations

3.1.5 Reactor

Vessel Circumferential Welds LRPD-03 Revision 0 Page 12 of 98 BWRVIP-74 reiterated the recommendation of BWRVIP-05 that RPV circumferential welds could be exempted from examination.

The NRC SER for BWRVIP-74 agreed, but required that plants apply for this relief request individually. The relief request should demonstrate that at the expiration of the current license, the circumferential welds satisfy the limiting conditional failure probability for circumferential welds in the (BWRVIP-05) evaluation.

WNPS has applied for the relief request (Ref. 6.5.22), but has only evaluated the welds to the end of the current operating license. The changes in metallurgical conditions expected over the period of extended operation require additional analysis for 54 EFPY to extend the reactor vessel circumferential weld inspection relief request. The evaluations have been extended to 54 EFPY and the results are presented here. WNPS requested relief from the inspection of reactor vessel circumferential welds (Ref. 6.5.22). The WNPS submittal included an analysis that showed that the reactor vessel parameters after 32 EFPY were within the NRC's 32 EFPY bounding Chicago Bridge & Iron (CBI) vessel parameters from the BWRVIP-05 SER. As such, there is a lower conditional probability of failure for circumferential welds at VYNPS than that stated in the NRC's Final Safety Evaluation Report of BWRVIP-05.

Table 3-6 reproduces the table from the submittal, with an added column providing the values for 54 EFPY. Consistent with earlier submittals, this table conservatively uses surface fluence rather than

%T fluence, so the resulting change in RTNDT is slightly higher than shown in Section 3.1.4 of this report. The WNPS reactor pressure vessel circumferential weld parameters at 54 EFPY will remain within the NRCs (64 EFPY) bounding CBI vessel parameters from the BWRVIP-05 SER. As such, the conditional probability of failure for circumferential welds remains below that stated in the NCR's Final Safety Evaluation of BWRVIP-05.

Therefore, this analysis has been projected for the period of extended operation per 10 CFR 54.21 (c)(l)(ii).

VYNPS will officially request this relief request for the period of extended operation.

3.1.6 Reactor

Vessel Axial Welds Applicants must evaluate axially oriented RPV welds to show that their failure frequency remains below the 5x1 0-6 calculated in the BWRVIP-74 SER. The SER states that an acceptable way to do this is to show that the mean RTNDT of the limiting axial beltline weld at the end of the period of extended operation is less than the values specified in Table 1 of that SER. Table 3-7 of this report reproduces the 32 EFPY and 64 EFPY data from the SER, and adds the WNPS data for 32 and 54 EFPY. The table shows that the WNPS mean RTNDT is well below that in the SER, and thus the WNPS axial weld failure frequency is well below the acceptable limit of 5x1 O-6. Therefore, this analysis has been projected for the period of extended operation per 10 CFR 54.21 (c)(l)(ii).

3.1.7 Surveillance

Specimen Testing IOCFRSO, Appendix H, requires a reactor vessel materials surveillance program that can verify the TLAAs for vessel embrittlement discussed above, and modify the projections if needed, based on measured embrittlement of actual material samples. The first WNPS surveillance capsule was withdrawn from the vessel after approximately 4.3~1 0l6 n/cm2 and tested.

The LRPD-03 Revision 0 Page 13 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations results are presented in Battelle Columbus Laboratories report BCL-585-84-3 (Ref.

6.5.1 8). The data agree with the data in the NRC RVID2 database (Ref. 6.2.1). The data in the capsule report showed the decrease in plate CvUSE to be 2.5 times that predicted by RG 1.99. W re-evaluated the raw data points determined by Battelle using an EPRl hyperbolic tangent curve fitting routine. This resulted in revised unirradiated and irradiated CvUSE results for both the plate and weld specimen. The new analyses still resulted in larger CvUSE reductions than predicted by RG I .99 for the plate, but not as large as predicted by the original Battelle report. The revised analyses resulted in a decrease in weld CvUSE very close to the RG 1.99 predicted decrease opposed to the increase in the original analyses.

The revised analyses were submitted to the NRC in WNPS letter BW 93-146. (Ref.

6.5.14) Table 3-2 summarizes both the Battelle report and BW 93-146 for comparison.

In March 2003 (Ref. 6.5.16), WNPS submitted a license amendment request to remove the plant-specific reactor vessel surveillance requirements from the Technical Specifications and replace them with the BWRVIP Integrated Surveillance Program (ISP). The NRC approved this submittal as Amendment 218 to the WNPS license (Ref. 6.2.1 1). This amendment removed the plant-specific surveillance capsule requirements.

For the period of extended operation, WNPS will continue to participate in the BWRVIP Integrated Surveillance Program (BWRVIP-74, 86 and 116).

WNPS will periodically adjust the projected values of fluence, CvUSE and RTNDT as additional surveillance capsule results are collected by the BWRVIP Integrated Surveillance Program (BWRVIP Reports 78, 86, and 116). See the Reactor Pressure Vessel Monitoring Program in LRPD-02, Aging Management Program Evaluation Report, for additional details. Surveillance capsule data is not a TLAA.

LRPD-03 Revision 0 Page 14 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations Heat# C2640-1 C2653-3 C3116-2 C3017-2 Table 3.1 WNPS Charpy Upper Shelf Energy Data for 54 Effective Full-Power Years (EFPY)

I I I Material Description 32 EFPY Projection 54 EFPY Projection

%Cu 0.12 0.13 0.14 0.11 t i I Initial USE EMA 1/4 T 1/4 T fluence fluence (10'~ %Drop USE (10'~ %Drop USE n/cm2) in USE (1/4T) n/cm2) in USE (114T) 0.017 8.00% EMA 0.0398 9.79% EMA Plate 1-17 Plate 1-16 Reactor Vessel Beltline Region Plates I Plate 1-15 I A533B I 328 Material Plate Type ID EMA EMA 89 I Welds 1 SMAW I 955 0.017 8.38% EMA 0.0398 10.3% EMA 0.017 8.76% EMA 0.0398 10.7% EMA 0.01 7 7.62% 82.2 0.0398 9.32% 67.7

References:

Plate 1-14 Reactor Vessel Beltline Region Welds A533B 327 Weld Plate Type ID Heat# NANV-A 1/4 T 1/4 T fluence fluence Initial (1 of9 %Drop USE (ioi9 %Drop USE %Cu USE2 n/cm2) in USE (1/4T) n/cm2) in USE (1/4T) 0.04 EMA 0.017 6.86% EMA 0.0398 8.39% EMA 1 2 3 The material description and 32 EFPY projections came from the NRC Reactor Vessel Integrity Database (RVID2), Ref.

6.2.1. The 54 EFPY projection uses the vessel ID fluence given in Ref. 6.5.44 converted to 1/4T fluence using the RG 1.99 formula. Vessel thickness

= 5.064 inches (Ref. 6.5.3) The 54 EFPY % drop in use is calculated from the fluence and the formulae for the curves in RG 1.99.

WNPS License Renewal Project TLAA and Exemption Evaluations Table 3-2 VYNPS Surveillance Capsule #I Test Data (Discussed in Section 3.1.4) LRPD-03 Revision 0 Page 45 of 98 Material Description Plate, Longitudinal Weld, NA Heat Number' Capsule No. Lead Factor Copper % Neutron fluence (IO" n/cm2) fluence factor Measured Initial USE Measured Radiated USE Drop in Use % Drop in USE RG 1.99 Predicted

% drop in USE USE correction factor Capsule Report BVY 93-146 C3017-2 C3017-2 30 deg 30 deg2 0.83 0.83*

0.106 0.1 1 0.0043 0.0043 0.063 0.063 148 137 128 126 20 11 1 3.5%4 8.0% 5.39%4 5.50% 2.514 1.45 Capsule Report BVY 93-146 3P4966 3P4966 30 deg 30 deg2 0.83 0.832 0.030 0.030 0.0043 0.0043 0.063 0.063 107 125 122 119 -1 5 6 -14.02%4 4.80% 4.68%4 4.68%

1.00~,~ 1.03 1 2 3 4 5 The heat number is from RVID2 (Ref. 6.2.1), it is not used in any calculation.

The capsule number and lead factor are from the capsule report, Ref. 6.5.18. They are not used in any calculation.

The fluence factor is not given in either report.

It is calculated here using the fluence and the formula in RG 1.99. The % drop in USE and the RG 1.99 predicted

% drop in USE, and USE correction factor were not in the capsule report. They have been calculated here using the data above and the formulae for the curves in RG 1.99. USE correction factor was set =I as measured data showed an increase in USE.

LRPD-03 Revision 0 Paae 16 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations Material Description Initial Base Plate RTNDT Chemism Metal ID Heat# %Cu %Ni [DegF) 0, Factor 45336 330 C2640-1 0.12 0.61 0.0 0 83.2 45338 329 C2653-3 0.13 0.59 0.0 0 90.7 95336 328 C3116-2 0.14 0.66 -10.0 0 101.5 45336 327 C3017-2 0.11 0.63 30.0 0 74.5 Reactor Vessel Beltline Region Location Beltline ID Location Unknowr Location Unknowr Location Unknowr Location Unknowr 1-15 1-14 32 EFPY 54 EFPY Adjusted Adjusted 1/4T fluence Fluence ARTNDT Margin RTNDT 1/4T fluence Fluence ARTNDT Margin RTNDT (10'gn/cm2) Factor (Deg F)

GA (Deg F) (Deg F) {1019n/cm2) Factor (Deg F: b~ :Deg F) (Deg F) 0.0170 0.155 12.9 6.5 12.9 25.8 0.0398 0.258 21.5 10.7 21.5 42.9 0.0170 0.155 14.1 7.0 14.1 20.2 0.0398 0.258 23.4 11.7 22.8 46.8 0.0170 0.155 15.8 7.9 15.8 21.5 0.0398 0.258 26.2 13.1 25.6 42.4 0.0170 0.155 11.6 5.8 11.6 53.1 0.0398 0.258 19.2 9.6 18.8 68.4 Table 3-3 VYNPS RTNDT for 32 and 54 Effective Full-Power Years (EFPY) Reactor Vessel Beltline Region Initial Location Flux Weld RTNDT Chemistb (Beltline ID) type ID Heat# %Cu %Ni lDeg~I bu Factor Welds' SMAW 955 NAN-A 0.04 1.00 0 13 54 Welds' SMAW 955 NAN-A 0.04 1.00 0 0 54 Adjusted Adjusted 1/4T fluence Fluence ARTNDT Margin RTNDT 1/4T fluence Fluence ARTNDT Margin RTNDT (10'9n/crn2)

Factor (Deg F) CA (Deg F) (Deg F) (10'9n/cm2)

Factor :DegF: oA (Deg F) (Deg F) 0.0170 0.155 8.4 10.0 27.3 35.7 0.0398 0.258 13.9 10.0 29.3 43.4 0.0170 0.155 8.4 4.2 8.4 16.8 0.0398 0.258 13.9 7.0 13.6 27.9 , 1 This line mimics RVID2 and uses override values for 0, and oA. Results in conservative margin. 2 This line mimics NEDC-33090P and uses 0 for crU and calculates bA per RG 1.99, consistent with the way RVID2 calculates the plates. 3 The 54 EFPY projection uses the vessel fluence as determined in Section 3.1.1 of this report. 4 The initial RTN~T values are from the Structural Integrity Associates report attached to BVY-OO- 113. These values supersede RVID2, as agreed to by the NRC in their SER (Ref. 6.2.33).

LRPD-03 Revision 0 Page 17 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations Table 3-4 Equivalent Margin Analysis for WNPS Plate Material USE 32 EFPY 33 EFPY 54 EFPY Surveillance Plate % Cu 0.11% 0.11% 0.11% Surveillance Plate Fluence (10" n/cm2) 6 6 4.49E+16 CLTP' CPPU* CPPU3 4.49E+1 4.49E+1 Surveillance Plate Measured Decrease 8.03% 8.03% 8.03%

RG 1.99 Predicted Decrease 5.55% 5.55% 5.55% Ratio of Measured to Predicted 1.448 1.448 1.448 Beltline Plate

% Cu 0.14% 0.14% 0.14% 32 EFPY 1/4T fluence (IO" n/cm2) 0.0221 0.0235 0.03984 RG 1.99 Predicted Decrease 9.4% 9.5% 1 0.7%5 Limiting % Decrease 21 .O% 21 -0% 23.5%? Plate Acceptable Yes Yes Yes Adjusted % Decrease 13.5%

13.8% 1 5.5%6 All of the above decreases are less than the 23.5% decrease in the bounding equivalent margin analysis, so the analysis conclusions apply to the vessel plates. 1 The 32 EFPY, Current Licensed Thermal Power (CLTP) column is from NEDC-33090P, Table 3-la. (Ref. 6.5.2) Note that as part of the power uprate, the capsule fluence and the 32 EFPY fluence were recalculated using neutron transport theory consistent with RG1.190. Hence the numbers vary slightly from those given in table 3-2. The 33 EFPY, Constant Pressure Power Uprate (CPPU) column is from NEDC-33090P, Table 3-lc (Ref. 6.5.2) The 54 EFPY, CPPU column is created here. The surveillance capsule data is the same as the first two columns, other values are discussed in the following footnotes.

The 54 EFPY 1/4T fluence was calculated in Section 3.1.1 of this report. The RG 1.99 predicted value was calculated using the formula for the curves in RG 1.99. The adjusted decrease equals the product of the RG 1.99 prediction (10.7%) and the surveillance capsule ratio of measured to predicted (I .448). The limiting percent decrease for 54 EFPY is 23.5% per BWRVIP-74 (Ref. 6.4.1 I) as approved by the NRC in their SER (Ref. 6.2.24). 2 3 4 5 6 7 LRPD-03 VYNPS License Ren-----'

n--'--' TLAA and Exemption waiuarions I Page18of98 I Revision 0 Table 3-4 (continued)

Equivalent Margin Analysis for VYNPS Weld Material USE Surveillance Weld

% Cu Surveillance Weld Fluence (1 0" n/cm2) Surveillance Weld Measured Decrease RG 1.99 Predicted Decrease Ratio of Measured to Predicted Beltline Weld % Cu 32 EFPY 1/4T fluence (IO" n/cm2) RG 1.99 Predicted Decrease Adjusted % Decrease Limiting % Decrease Weld Acceptable 32 EFPY 0.03% 4.49E+1 6 4.80% 4.77% 1.005 0.04% 0.0221 7.32% 7.36% 34.0% Yes CLTP' 33 EFPY 54 EFPY 0.03% 0.03% CPPU2 CPPU3 4.49E+1 6 4.49E+16 4.80% 4.77% 1.005 0.04% 0.0235 7.43% 7.47% 34.0% Yes 4.80% 4.77% 1.005 1 1.1 9YO6 11 .24%7 39.0%* Yes All of the above decreases are less than the decrease in the bounding equivalent margins analysis, so the analysis conclusions apply to the vessel welds. The 32 EFPY, CLTP column is from NEDC-33090P, Table 3-1 b. (Ref. 6.5.2). Note that as part of the power uprate, the capsule fluence and the 32 EFPY fluence were recalculated using neutron transport theory consistent with RG1.190. Hence the numbers vary slightly from those given in table 3-2. The 33 EFPY, CCPU column is from NEDC-33090P, Table 3-Id (Ref. 6.5.2) The 54 EFPY, CPPU column is created here. The surveillance capsule data is the same, other values are discussed in the following footnotes.

A maximum weld copper content of 0.1% was conservatively used, consistent with the original equivalent margin evaluations in BVY 93-146 The 1/4T fluence was calculated in Section 3.1.1 of this report. The RG I .99 predicted value was calculated using the formula for the curves in RB 1.99. The adjusted decrease equals the product of the RG 1.99 prediction (1 1.1 9%) and the surveillance capsule ratio of measured to predicted (1.005). The limiting percent decrease for 54 EFPY for welds is 39% per BWRVIP-74 (Ref.

6.4.1 1) as approved by the NRC in their SER (Ref. 6.2.24).

WNPS License Renewal Project TLAA and Exemption Evaluations Vessel Beltline Region Location Beltline ID Unknown 1-14 LRPD-03 Revision 0 Page 19 of 98 Table 3-5 VYNPS P-T Curve Bases, (Current 32 EFPY and 54 EFPY) Material Description 32 EFPY P-T Curve Bases Initial fluence Adjusted RTNDT Chemistry Thickness (IOi9 Fluence ARTNDT Margin RTNDT ,DeqF, (T" Factor Location (inches) n/cm2) Factor (Deg F) (JA (Deg F) (Deg F) 30.0 0 74.5 ID 0 (5.06) 1.24E+18 0.461 34.3 17.0 34.0 98.3 30.0 0 74.5 %T 1.3 9.15E+17 0.399 29.7 14.9 29.7 89.5 30.0 0 74.5 %T 3.8 4.99E+17 0.292 21.8 10.9 21.8 73.5 54 EFPY fluence ARTND Adjusted (IOi9 FluenceT (Deg Margin RTNDT n/cm2) Factor F) (JA (Deg F) (Deg F) 5.39E+17 0.305 22.7 11.3 22.7 75.4 3.98E+17 0.258 19.2 9.6 19.2 68.5 2.17E+17 0.181 13.5 6.7 13.5 56.9 2 The basis for the current P-T limits (32 EFPY) are found in Reference 6.2.1 1. 3 The 54 EFPY projection uses the vessel fluence as determined in Section 3.1.1 of this report.

Table 3-6 32 EFPY Bounding Parameters

-65 I WNPS RPV C Beltline 64 EFPY Beltline Circ Weld Bounding Circ Weld 32 EFPY Parameters 54 EFPY 65 0 Parameter Description 44.5 Initial (unirradiated) reference temperature (RTNDT), OF -58.2 70.6 32.9 Neutron fluence at the end of the 1 requested relief period (Peak Surface Fluence Entire Beltline) Fluence Factor (calculated Der RG 1.99 based on fluence in previous line.) Weld Copper content, % Weld Nickel Content Weld Chemistry Factor (CF) Chemistry Factor times Fluence Factor Margin (Implied), OF Increase.in reference temperature (ARTNDTL OF Mean- adjusted reference temperature (ART), OF = RTNDT + ARTNDT 1 The first column is from Table

2.1 cumferential

Shell Welds USNRC I VYNPS I USNRC I VYNPS I I 1 109.5 109.5 2 3 4 The second column is from BVY 03-83 (Ref. 6.5.22) The third column is from Table 2.6-5 of the NRC SER for BWRVIP-05.

The fourth column is new material for this report.

WNPS License Renewal Project TLAA and Exemption Evaluations LRPD-03 Revision 0 Page 21 of 98 I I I I Column 1 is from Table 2.6-4 of BWRVIP-05 SER. Column 2 is generated from data in this report with some calculations using that data. WNPS RPV Axial Shell Welds Parameter Description USNRC VYNPS USNRC Limiting Data for Limiting Plant- Plant- axial weld Specific Data Specific Data 32 32 64 -30 0 -30 EFPY Initial (unirradiated) reference temperature (RTNDT), OF Neutron Fluence 6.90E+18 2.99E+17 1.38E+18 0.896 0.21 9 I .089 FF = Fluence Factor Weld Copper content, % 0.10% 0.04% 0.10% 135.0 54 135.0 CF = Chemistry Factor 121 .o 11.8 147.1 Increase in reference temperature (ARTNDT), OF =FF*CF temperature (ART), OF Weld Nickel Content, % I .08% 1 .OO% 1.08% 91 .o 11.8 117.1 Mean adjusted reference

= RTNDT + ARTNDT No previously submitted data was located. Column 3 is from Table 2.6-5 of BWRVIP-05 SER. Column 4 is generated from data in this report with some calculations using that data VYNPS Data for axial weld 54 0 5.39E+17 0.305 0.04% I .OO% 54.00 16.5 16.5 LRP D -03 Pane 27 nf 98 Revision 0 WNPS License Renewal Project TLAA and Exemption Evaluations

3.2 Metal

Fatique LRPD-04 describes and evaluates fatigue evaluations that meet the definition of TLAA for Class I and non-Class 1 mechanical components. Cumulative usage factors have been documented and the actual numbers of design transient cycles have been projected to 60 years. An adequate program is in place to track cycles and to provide corrective actions if limits are approached.

For details of fatigue TLAA evaluations, see LRPD-04, TLAA - Mechanical Fatigue. 3.3 Environmental Qualification of Electrical Equioment This section provides the evaluation of TLAA for EQ components.

3.3.1 Background

For certain important-to-safety electrical components, operating plants must meet the requirements of 10 CFR 50.49 (Ref. 6.1.7) which defines the scope of components to be included, requires the preparation and maintenance of a list of in-scope components and requires the preparation and maintenance of a qualification file. Environmental Qualification Program Manuals, Volume I & II (Ref. 6.5.47), document the WNPS EQ program basis and history. Volumes I & II include the general engineering documentation that provides the environmental qualification specifications for each component.

Environmental Qualification Program Manual, Volume I, identifies and summarizes the EQ program activities and processes for implementing EQ requirements, which exist to assure that EQ components and EQ-related activities satisfy applicable industry and regulatory requirements. Volume I of the program manual documents the philosophy and methodology for meeting the requirements outlined in IO CFR 50.49.

Also, Volume I (Section 4.0) addresses how the industry and WNPS have dealt with equipment qualification as it has evolved since licensing of the first plants. Volume I directs the user to appropriate documents for detailed processes and implementation requirements as necessary.

In addition, the EQ Program includes enhanced Maintenance and Surveillance (M&S) Program requirements for electrical components that require environmental qualification.

The EQ M&S Program is not included in the EQ Program Manual but utilizes input from the Program Manual. The EQ M&S Program records are controlled and maintained at the plant per plant procedure AP0305 (Ref. 6.5.50). The EQ M&S Program is to assure that the environmental qualification of specific plant equipment remains valid through the expected life of that equipment and that any change to the expected life is recognized and reanalyzed. The EQ M&S Program provides controls for scheduling, performing, and documenting maintenance performed on EQ equipment.

The WNPS EQ Master Equipment List (EQMEL) (Ref. 6.5.45) is a hard copy of the EQ Program Volume I document. The EQMEL is based on plant procedure AP 0092 (Ref. 6.5.46).

Volume I, Section 7.0 and calculation WC-193 (Ref. 6.5.47 and 6.5.48) define environmental service parameters for the environmental qualification of equipment.

EQ documentation for equipment is maintained in qualification documentation review (QDR) packages.

The index in Volume I, Section 1 provides the complete list of QDRs and Section 6 identifies the applicable QDR number for each component, which includes the analysis that supports the equipment qualification (Ref. 6.5.45). The system component evaluation worksheet (SCEW) is a form, LRPD-03 Revision 0 Page 23 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations which summarizes the environmental qualification data and qualification status for components or component types in the format requested by the NRC in IEB 79-01 B (Ref. 6.2.5). The SCEW is part of the QDR (Tab B). EQ-related maintenance requirements (if necessary to maintain qualification) are in the (QDR) for the specific EQMEL item (Ref. 6.5.47 and 6.5.50). In order to identify EQ commitments and exemptions to 10 CFR 50.49 for VYNPS, a review was conducted of the searchable plant databases that include the UFSAR, the operating license, and NRC correspondence.

This review involved a search of this database to identify exemptions based on EQ TLAA that were granted pursuant to 10 CFR 50.1 2. The search criteria included key terms including "50.12", 'cexemption*, "EQ", "environmental qualification", "40 years", "forty years", "plant life", "design life", "qualified life", "life of the plant", "service period" and "operating term". In addition to the database searches, sections of the Technical Specifications, the original Safety Evaluation Report, and selected correspondence, such as NRC generic letters and bulletins were reviewed.

The review at WNPS identified no exemptions based on TLAAs for EQ electrical components.

3.3.2 Environmental

Qualification All operating plants must meet the requirements of 10 CFR 50.49 for certain important-to-safety electrical components.

10 CFR 50.49 defines the scope of components to be included, requires the preparation and maintenance of a list of in-scope components, and requires that preparation and maintenance of a qualification file that includes component performance specifications, electrical characteristics and environmental conditions.

The EQ master equipment list (EQMEL) meets the 10 CFR 50.49(d) and IE Bulletin 79-01 B requirements for the list of EQ components.

The listing of electrical equipment in the EQMEL (Ref. 6.5.45) is controlled under the M&S Program Manual in accordance with plant procedure AP 0092 (Ref. 6.5.46). 10 CFR 50.49(e)(5) requires component replacement or refurbishment prior to the end of the designated life, unless additional life is established through ongoing qualification. The equipment included in the scope of EQ is based on specific screening criteria in 10 CFR 50.49(b).

As part of the EQ program, when EQ equipment or parts thereof have a limited life, the preventive maintenance process ensures the equipment or parts are replaced prior to expiration of the qualified life. If excess conservatism exists in the original qualified life determination, then reanalysis, which meets the requirements of 10 CFR 50.49, can extend the qualified life.

The reanalysis utilizes standard EQ techniques (such as the Arrhenius method), and becomes part of the EQ documentation. Conservatism may exist in the ambient temperature of the equipment, in unrealistically low activation energy, and in the application of the equipment. The primary method used for reanalysis is reducing excess conservatism in equipment service temperatures by using temperature values closer to actual temperature in the area around the applicable equipment. These reanalysis methods for EQ components are discussed in NUREG-1801,Section X.E1 (Ref. 6.1.3). EQ equipment will be reevaluated for the environmental service conditions that are applicable the equipment (i.e., 60 years of exposure versus 40 years). The environmental service conditions considered are normal and accident. For electrical equipment exposed to a harsh environment, IO CFR 50.49 requires consideration of all significant aging effects from normal service conditions. This includes the expected thermal aging effects from the temperature to which the device is normally exposed, wearkycle aging (applicable to limited types of EQ to LRPD-03 Revision 0 Page 24 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations equipment), and radiation aging effects during normal plant operation.

10 CFR 50.49 also requires evaluation of the effects from harsh environments to which the equipment could be exposed under accident conditions.

In general, the harsh environments analyzed as part of the EQ Program are those caused by loss of coolant accidents (LOCA), high energy line breaks (HELBs) inside the reactor building, and HELBs outside the reactor building.

For EQ equipment with a qualified life less than the design-life of the plant, "ongoing qualification" is a method of long-term qualification involving additional testing.

Ongoing qualification or retesting, as described in IEEE Std. 323-1974 (Ref.

6.1.4), Section 6.6(1) or (2), is not considered a viable option, and there are no immediate plans to implement such an option. If this option becomes viable, ongoing qualification or re-testing would be performed in accordance with accepted industry and regulatory standards.

The evaluation of the environmental service conditions for the license renewal period requires a reevaluation of only the normal aging effects. Therefore, the normal effects from operation for a 60-year period instead of a 40-year period will be evaluated. Radiation aging effects, thermal aging effects, and wear/cycle aging effects, as applicable, are analyzed for the period of extended operation. The following sections describe each of these considerations in more detail. 3.3.3 Radiation Considerations Before entering the period of extended operation, the additional normal dose for the license renewal period will be evaluated. Typically a normal does that is 1.5 times (i.e., 60 years/40 years) the dose for the 40-year period is evaluated for the extended period.

The dose values used are based on equipment locations, applying either an inside reactor building value, outside reactor building value or application-specific value. The total integrated dose for the 60-year period is determined by adding the established accident dose to the newly determined 60-year normal dose for the device. If the device is qualified for this total integrated dose, no additional review is required.

If the increased normal dose results in a total integrated dose above the qualified dose for the component, a location-specific review may be required to determine a lower dose for the specific component. Other options include requiring component or part replacement prior to exceeding the qualified total dose or performing radiation surveys to determine actual operating dose and evaluating against this value. WNPS EQ Manual Volume I, Section 7.0 and WC-193 (Ref. 6.5.47 and 6.5.48) provide normal and accident radiation environmental data used for evaluating the environmental qualification of electrical equipment subject to 10 CFR 50.49.

In addition WC2339, which is referenced in the QDRs, includes more detailed analysis of specific components.

Some components have been installed under a plant modification, and will not experience 60 years of radiation aging by the end of the license renewal period. In these cases, credit may be taken for less than 60 years of aging.

Also, plant modifications that affect the normal dose, such as power uprate, will be addressed for impact on the EQ TLAAs.

LRPD-03 Revision 0 Page 25 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations

3.3.4 Thermal

Considerations The average (ambient) temperatures inside the reactor building in occupied areas are 84°F and non-occupied areas are1 00°F with peak allowable temperatures 106°F and 120°F respectively.

The average temperature inside the drywell is dependent upon elevation with a range of 150°F to 270°F. The peak allowable temperatures are in a range of 160°F to 280°F. The average (ambient) temperatures for the auxiliarykurbine building in occupied areas is 100°F with peak allowable temperature 105"F, except where equipment walkdowns or temperature monitoring have identified localized elevated temperatures. Section 7.0 of the EQ Program Manual provides an extensive plant area listing of (ambient) temperatures (Ref. 6.5.47).

For components exposed to (ambient) temperatures lower than the given averages, temperature monitoring can be utilized to confirm lower than design temperatures exist in these areas, and on that basis, extend qualified lives through the license renewal period. If extension of the 40-year qualified life is chosen rather than component replacement, it would be based upon re-evaluation of an aging analysis as defined by 10 CFR 50.49 and discussed in NUREG-I 801 Section X.EI (Ref. 6.1.3). The aging analysis will be revised, as applicable, to identify the maximum service life based on traditional EQ techniques such as the Arrhenius method. Some components have been installed under a plant modification, and will not experience 60 years of thermal aging by the end of the license renewal period.

In these cases, credit may be taken for less than 60 years of aging. 3.3.5 Wear/Cycles Considerations EQ evaluations for the license renewal period will address wear/cycle aging qualification prior to entering the period of extended operation. For most components, this aging does not apply; however, for electromechanical equipment like solenoid valves there would be an associated number of cycles over 40 years. The number of cycles assumed for the license renewal period is 1.5 times (Le., 60 years/40 years) the number established for the 40-year period.

Some components have been installed under a plant modification, and will not experience 60 years of cycling by the end of the license renewal period. In these cases, credit may be taken for less than 60 years of aging. Credit may also be taken for lower actual frequency of cycles. 3.3.6 GSI-168, EQ of Electrical Components As discussed in SECY-93-049 (Ref.

6.1.6), the staff reviewed significant license renewal issues and found that several were related to environmental qualification.

A key aspect of these issues was whether the licensing basis should be reassessed or enhanced in connection with license renewal, and whether this reassessment should be extended to the current license term. In late 1993, the Commissioners instructed the Staff that the current EQ licensing basis must be used in the license renewal period and that any EQ concerns identified by the staff during the review of EQ for license renewal should be evaluated for the effect on current licenses, independent of license renewal.

The NRC Staffs EQ Task Action Plan (EQ-TAP) was initiated to address the adequacy of current EQ practices. (Ref. 6.2.4) Upon completion of the EQ-TAP review, the Staff concerns focused on issues related to the adequacy of accelerated aging practices in existing LRPD-03 WNPS License Rer-*-'

TLAA and Exemption cvaiuarions 1 Page26of98 I I qualifications, and the lack of a "feedback mechanism" in EQ programs (i.e., programmatic requirements to determine the current condition of EQ equipment so that it can be evaluated against the assumptions and parameters for qualification).

The EQ-TAP was subsequently closed and the remaining open issues were incorporated into GSI-168 for management tracking purposes. The EQ-TAP review did not identify any generic safety issues related to these open issues. NRC guidance for addressing GSI-168 for license renewal is contained in a June 1998 letter to NE1 (Ref. 6.2.29). In this letter, the NRC states: LRPD-03 "With respect to addressing GSI-168 for license renewal, until completion of an ongoing research program and staff evaluations, the potential issues associated with GSI-168 and their scope have not been defined to the point that a license renewal applicant can reasonably be expected to address them at this time. Therefore] an acceptable approach described in the SOC is to provide a technical rationale demonstrating that the current licensing basis for EQ pursuant to IO CFR 50.49 will be maintained in the period of extended operation. Although the SOC also indicates that an applicant should provide a brief description of one or more reasonable options that would be available to adequately manage the effects of aging, the staff does not expect an applicant to provide the options at this time." WNPS License Renewal Project TLAA and Exemption Evaluations Consistent with previous NRC guidance, and RIS-2003-09 (Ref.

6.2.2) no additional information is required to address GSI-168 in a license renewal application (Ref. 6.1.2). Revision 0 Page 26 of 98 3.3.7 Summary The VYNPS environmental qualification (EQ) of electrical components program (Ref. 6.5.46) manages component thermal, radiation and cyclical aging, as applicable, through the use of aging evaluations based on IO CFR 50.49(f) qualification methods. As required by 10 CFR 50.49, EQ components not qualified for the current license term will be refurbished, replaced, or have their qualification extended prior to reaching the aging limits established in the evaluation.

Aging evaluations for EQ components that specify a qualification of at least 40 years are considered TLAA for license renewal. The EQ program ensures that these analyses for EQ components are maintained in accordance with their qualification bases. The WNPS program is an existing program established to meet WNPS commitments for 10 CFR 50.49. It is consistent with NUREG-1801,Section X.EI, "Environmental Qualification (Ea) of Electric Components."

The VYNPS program includes consideration of operating experience to modify qualification bases and conclusions, including qualified life. Compliance with IO CFR 50.49 provides reasonable assurance that components can perform their intended function(s) during accident conditions after experiencing the effects of inservice aging. Consistent with NRC guidance provided in RIS 2003-09, no additional information is required to address GSI-168, "EQ of Electrical Components."

Based upon a review of the existing program and associated operating experience, continued implementation of the WNPS environmental qualification of electrical components program provides reasonable assurance that the aging effects will be managed and that the in-scope EQ components will continue to perform their intended function(s) for the period of extended operation.

The effects of aging will be managed by the WNPS program in accordance with the requirements of 10 CFR 542I(c)(l)(iii). (Ref. 6.1.5 and 6.5.47)

VYNPS License Renewal Project TLAA and Exemption Evaluations

3.4 Concrete

Containment Tendon Prestress LRPD-03 Revision 0 Page 27 of 98 This section is not applicable as WNPS does not have pre-stressed tendons in the containment building.

Containment (Torus) and Torus Attached Piping The torus and torus attached piping systems were analyzed for fatigue due to mechanical loadings as well as thermal and anchor motion. LRPD-04, Mechanical Fatigue addresses these analyses.

3.5 Metal

Corrosion Allowance Most pressure retaining components are constructed with a wall thickness in excess of minimum required wall thickness for that component.

This excess wall thickness provides a corrosion allowance over the life of the component to assure that minimum wall thickness requirements are still met at end of life. If these corrosion allowances were meant to cover the original 40 year design life of the component, they could be considered TLAA. Individual corrosion allowances are discussed below. The results show that there are no analyses for corrosion allowances based on time-limited assumptions and hence no TLAA. Loss of material caused by corrosion of metal components will be managed for the period of extended operation.

3.5.1 Reactor

Pressure Vessel UFSAR Section 4.2.4.1 states "Although little corrosion of plain carbon or low alloy steels occurs at temperatures of 500°F to 600"F, higher corrosion rates occur at temperatures around 140°F. The 0.1 25-inch minimum thickness stainless steel cladding provides the necessary corrosion resistance during reactor shutdown and also helps maintain water clarity during refueling operations.

Exterior exposed ferritic surfaces of pressure-containing parts have a minimum corrosion allowance of 1/32-inch.

All carbon and low alloy steel nozzles exposed to the reactor coolant have a corrosion allowance of 1/16-inch.

The vessel shape is designed to limit coolant retention pockets and crevices". Although the original reactor vessel corrosion allowances were conservative values intended to encompass 40 years of operation, no specific corrosion rate, and no specific analysis associated with these values has been identified. As such, there are no TLAA associated with these corrosion allowances.

Loss of material from carbon steel and low alloy steel is an aging effect requiring management as identified in WNPS Report AMRM-31, Aging Management Review of the Reactor Pressure Vessel. The lnservice Inspection Program and the Water Chemistry Control Program manage this effect as detailed in AMRM-31, Aging Management Review of the Reactor Pressure Vessel.

3.5.2 Recirculation

Pump Casing Section 4.3.4 of the UFSAR states: "The design objective for the recirculation pump casing is a useful life of 40 years, accounting for corrosion, erosion, and material fatigue." Although the original corrosion allowances were conservative values intended to encompass 40 years of operation, no specific corrosion rate, and no specific analysis associated with these values has been identified. As such, there are no TLAA associated with these corrosion allowances.

Corrosion is an aging mechanism leading to loss of material. Loss of material is an aging effect evaluated in AMRM-33, Aging Management Review of the Reactor Coolant System. The WNPS License Renewal Project TLAA and Exemption Evaluations Lnru-ua Revision 0 recirculation pump casing is cast of austenitic stainless steel, which is inherently resistant to corrosion; consequently significant corrosion of this pump casing is not expected. The lnservice Inspection Program and the Water Chemistry Control Program manage this effect. 3.5.3 Main Steam Isolation Valve UFSAR section 4.6.3, Description (of the MSIVs), states "The design objective for the valve is a minimum of 40 years' service at the specified operating conditions.

The estimated operating cycles per year is 100 cycles during the first year and 50 cycles per year thereafter.

In addition to minimum wall thickness required by applicable codes, a corrosion allowance of 0.120 inch minimum is added to provide for 40 years' services." Although the original corrosion allowances were conservative values intended to encompass 40 years of operation, no specific corrosion rate, and no specific analysis associated with these values has been identified. As such, there are no TLAA associated with these corrosion allowances.

Corrosion is an aging mechanism leading to loss of material.

Loss of material is an aging effect evaluated in AMRM-33, "Aging Management Review of the Reactor Coolant System."

The Water Chemistry Control - BWR program, the ASME Section XI Inservice Inspection Program and the System Walkdown program manage loss of material from these valves. . 3.5.4 HPCI System Section 6.4.1 of the UFSAR states "The system is designed for a service life of 40 years, accounting for corrosion, erosion, and material fatigue". Although the original corrosion allowances were conservative values intended to encompass 40 years of operation, no specific corrosion rate, and no specific analysis associated with these values has been identified.

As such, there are no TLAA associated with these corrosion allowances. Corrosion and erosion are aging mechanisms that lead to loss of material.

Loss of material for the HPCl system is identified as an aging effect requiring management in AMRM-05, "Aging Management Review of the High Pressure Coolant Injection System". The Water Chemistry Control - BWR program and the Flow Accelerated Corrosion Program manage loss of material due to corrosion for the HPCl system. 3.5.5 The table on page C.2-52 of the UFSAR states RClC and HPCl Turbine Casings "2. The minimum wall thickness of the turbine casing shall be based on that to limit stress to the allowable working stress when subjected to design pressure plus corrosion allowance. Allowable stresses shall be in accordance with ASME B&PV Code,Section VI 11." Although the original corrosion allowances were conservative values intended to encompass 40 years of operation, no specific corrosion rate, and no specific analysis associated with these values has been identified. As such, there is no TLAA associated with these corrosion allowances.

Corrosion is an aging mechanism leading to loss of material.

Loss of material from the RCIC turbine casing is addressed in AMRM-06, Aging Management Review of Reactor Core Isolation Cooling System. The Water Chemistry Control - BWR Program, the Periodic Surveillance and LRPD-03 Revision 0 Page 29 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations Preventive Maintenance Program manage loss of material for the RClC and HPCl turbine casings. 3.5.6 RClC and HPCl Pumps The table on page C.2-49 of the UFSAR states "2. The minimum wall thickness of the pump shall be based on that to limit stress to the allowable working stress when subjected to design pressure plus corrosion allowance. Allowable stresses shall be in accordance with ASME B&PV Code, Section 111." Although the original corrosion allowances were conservative values intended to encompass 40 years of operation, no specific corrosion rate, and no specific analysis associated with these values has been identified. As such, there are no TLAA associated with these corrosion allowances. Corrosion is an aging mechanism leading to loss of material.

Loss of material from the RClC pump casing is addressed in AMRM-06, Aging Management Review of Reactor Core Isolation Cooling System. Loss of material from the HPCl pump casing is addressed in AMRM-05, Aging Management Review of the High Pressure Coolant Injection System. The Water Chemistry Control - BWR Program, the Flow Accelerated Corrosion Program, and the Periodic Surveillance and Preventive Maintenance Program manage this aging effect. 3.5.7 RHR Heat Exchanger The table on page C.2-55 of the UFSAR states "The minimum thickness of the following components shall be designed to contain the design pressure plus corrosion allowance.

A. Shell B. Shell Cover C. Channel Ring D. Tubes." Although the original corrosion allowances were conservative values intended to encompass 40 years of operation, no specific corrosion rate, and no specific analysis associated with these values has been identified.

As such, there is no TLAA associated with these corrosion allowances.

Corrosion is an aging mechanism leading to loss of material. Loss of material from the RHR heat exchanger components is addressed in AMRM-02, Aging Management Review of Residual Heat Removal System. The Water Chemistry Control - Closed Cooling Water Program, the Systems Walkdown Program (external surfaces), and the Service Water Integrity Program(a1l surfaces) manage this aging effect. 3.6 ASME Section XI Inservice Inspection All currently active Section XI Code relief requests submitted to the NRC by VYNPS are specific to the third IS1 inspection interval, and thus by definition do not involve TLAA. Descriptions of each of these relief requests may be found in the IS1 Program Procedure. (Ref. 6.5.58)

LRPD-03 Revision 0 Page 30 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations 3.7 TLAA in BWRVIP Documents BWR Vessel and lnternals Project (BWRVIP) documents identify various potential TLAA. The TLAA applicable to WNPS are described below.

3.7.1 BWRVIP-05 Reactor Vessel Axial Welds BWRVIP-05 justified the elimination of reactor vessel circumferential welds from examination.

BWRVIP-74 extended this justification to cover the period of license renewal. See BWRVIP-74 below and Section 3.1.6 above for review of the TLAA associated with this issue. 3.7.2 BWRVIP-18 Core Spray lnternals There are no TLAA identified in BWRVIP-18.

3.7.3 BWRVIP-25 Core Plate This document concerns two aging effects for core plate rim hold-down bolts:

loss of preload and cracking.

The calculation of loss of preload on the core plate rim hold-down bolts is a TLAA (Ref. 6.2.25). BWRVIP-25 calculated the loss of preload for these bolts for forty years. Appendix B to BWRVIP-25 projected this calculation to 60 years, showing that the WNPS bolts would experience only 5 to 19 percent loss of preload. This TLAA is thus projected to the end of the period of extended operation in accordance with 10 CFR 54.21 (c)(l)(ii).

There is no TLAA associated with cracking of the core plate bolts. The inspection recommendations of BWRVIP-25 are intended to manage cracking of the core plate bolts for the period of extended operation.

WNPS has implemented (Ref. 6.5.43) the inspection requirements of BWRVIP-25 in the WNPS BWR Vessel lnternals Program, which will adequately manage cracking of the core plate rim hold down bolts for the period of extended operation.

3.7.4 BWRVIP-26 Top Guide BWRVIP-26 calculated the minimum top guide fluence for 32 EFPY (40 years) as 4 x102' n/cm2. Appendix C to BWRVIP-26 projected the calculation of the top guide fluence to 6 x102' n/cm2 for 48 EFPY (60 years). (Ref. 6.2.26) BWRVIP-26 and the NRC SER for BWRVIP-26 consider this a TLAA. This calculation confirms that every BWR exceeded the IASCC threshold after approximately 4 EFPY and must therefore inspect for IASCC. This analysis does not meet the criteria for a TLAA as there is no safety determination'based on this analysis (the analysis does not justify performing less inspections).

The threshold for IASCC is 5~10~~ n/cm2 (Refs.

6.4.4 and 6.2.26). The WNPS top guide fluence will exceed this threshold. Therefore WNPS must manage IASCC of the top guide assembly.

WNPS has implemented the inspection recommendations in BWRVIP-26 through the BWR Vessel lnternals Program (Ref. 6.5.43). The BWR Vessel lnternals Program will adequately manage the effects of aging on the top guide for the period of extended operation.

WNPS License Renewal Project TLAA and Exemption Evaluations 3.7.5 BWRVIP-27 SLClCore AP LRPD-03 Revision 0 Page 31 of 98 The BWRVIP-27 fatigue analysis of the SLC/core AP line for 60 years of operation is a TLAA. The NRC SER (Ref. 6.2.30) states that fatigue and the projected cumulative usage factors (CUF) should be addressed by each applicant who applies for license renewal. The WNPS SLC/AP nozzle is a low alloy steel (A508 C12) nozzle. WNPS reviewed the CLB fatigue analyses and CUFs of all components in WNPS Report LRPD-04, "TLAA - Mechanical Fatigue".

No fatigue analysis for the SLC/AP nozzle was found and no CUF for the nozzle was identified.

As such, WNPS has no TLAA associated with the SLC/AP nozzle. Cracking is an aging effect for this nozzle that is managed for the period of extended operation per WNPS Aging Management Review Report MARM-32, Aging Management Review of the Reactor Vessel Internals. 3.7.6 BWRVIP-38 Shroud Support The BWRVIP-38 fatigue analysis of the shroud support is a TLAA. Fatigue of the reactor vessel internals, including the shroud, are discussed in WNPS Report LRPD-04, TLAA - Mechanical Fatigue. The CUFs for the shroud are based on the design basis transients and remain valid for the period of extended operation in accordance with 10CFR54.21 (c)(l)(i).

BWRVIP-38 also identifies that the original stress analysis contains a bounding crack growth rate. It further points out that individual plants may seek approval to use a Jower crack growth rate based on operating experience.

If a plant has received approval to use a different crack growth rate, it may involve a TLAA. VYNPS has not sought nor received such approval and therefore has no TLAA relating to crack growth for the shroud support.

3.7.7 BWRVIP-41 Jet Pump Assembly The NRC SER for BWRVlP-41 requires plant-specific evaluation of jet pump fatigue by each applicant.

WNPS addresses fatigue in Report LRPD-04, TLAA - Mechanical Fatigue.

LRPD- 04 found no fatigue analysis and no CUF for the VYNPS jet pumps. As such, WNPS has no TLAA associated with the jet pumps. Cracking is an aging effect for these pumps that is managed for the period of extended operation per WNPS Aging Management Review Report MARM-32, Aging Management Review of the Reactor Vessel Internals.

The SER for BWRVIP-41 identifies evaluation of thermallradiation embrittlement of the jet pump cast austenitic stainless steel components as a TLAA if cracks exist in the components.

If the applicant can show that cracks do not exist, "...loss of fracture toughness resulting from thermal and/or neutron embrittlement will not be a significant aging effect."

WNPS has observed no cracking in the cast components of the jet pump assemblies. The BWRVIP has not reported cracks in these components.

In fact, BWRVIP-41 states that cracks are not expected in these components. Therefore, this is not a TLAA. If cracking appears, VYNPS will follow the recommendations of the BWRVIP to manage that cracking.

3.7.8 BWRVIP-47 Lower Plenum BWRVIP-47 identified fatigue analyses, especially of lower plenum pressure boundary components, as a TLAA. (Some plants have components whose CUF will exceed 1 .O during the period of extended operation).

WNPS addresses fatigue in Report LRPD-04, "TLAA -

LRPD-03 Revision 0 Page 32 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations Mechanical Fatigue". The only lower plenum CUF identified by LRPD-04 for VYNPS was a CUF for the CRD penetrations equal to 0.13. This CUF is determined by the allowed number of transients and as such remains valid for the period of extended operation per lOCFR54.2l(c)(l)(i).

3.7.9 BWRVlP-48 Vessel ID Attachment Welds The BWRVIP-48 fatigue analyses for various configurations of different vessel ID bracket attachments are TLAA. The analyses addressed VYNPS bracket configurations. (VYNPS has no unique bracket configurations.)

Analysis of fatigue for 60 years showed that no CUFs are above 0.4. This analysis remains valid for the period of extended operation in accordance with IOCFR54.21(c)(l)(i).

3.7.10 BWRVIP-49, Instrument Penetrations The BWRVIP-49 fatigue analysis for several configurations of instrumentation penetrations, including the WNPS configuration, is a TLAA. Analysis of fatigue for 60 years showed that all CUFs are below 0.4. This analysis remains valid for the period of extended operation in accordance with 10CFR54.21 (c)(l)(i).

3.7.1 1 BWRVIP-74, Reactor Pressure Vessel BWRVIP-74 and the NRC SER for BWRVIP-74 (Ref. 6.2.24) discuss the following four TLAA. I. Pressurenemperature Curves The SER concludes "a set of P-T curves should be developed for the heatup and cooldown operating conditions in the plant at a given EFPY in the LR period." Section 3.1.2 and Table 3-5 address the WNPS P-T curves.

2. Fatigue The SER states that the license renewal applicant should not rely solely on the analysis in BWRVIP-74, but should also verijr that the number of cycles assumed in the original fatigue design is conservative.

VYNPS Report LRPD-04, TLAA - Mechanical Fatigue addresses fatigue of the reactor pressure vessel. The SER also states that NRC staff concerns on environmental fatigue were not resolved and that each applicant should address environmental fatigue for the components covered by BWRVIP-74.

WNPS Report LRPD-04, TLAA - Mechanical Fatigue addresses environmentally assisted fatigue.

3. Equivalent Margins Analysis for RPV Materials with Charpy USE Less than 50 ft-lbs BWRVIP-74, addresses the percent reductions in Charpy USE for limiting BWW3-6 plates and BWR non-Linde 80 submerged arc welds (23.5 percent and 39 percent, respectively).

The NRC SER for BWRVIP-74 (Ref. 6.2.24) states that the applicant shall demonstrate that the percent reduction in USE for their beltline materials is less than the BWRVIP-74 values. Further, the SER states that the applicant shall demonstrate that LRPD-03 Revision 0 Page 33 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations the percent reduction of their surveillance weld and plate material is less than or equal to the values predicted by RG 1.99, Revision 2. Section 3.1.3, Table 3-1 and Table 3-4 address Charpy USE for reactor pressure vessel materials.

For WNPS, the percent reduction in CvUSE for the beltline materials remains well below the percent reduction in the equivalent margins analysis.

4. Material Evaluation for Exempting RPV Circumferential Welds from Inspection See Section 3.1.5, Section 3.1.6, Table 3-6 and Table 3-7 for a discussion of the reactor vessel welds.

3.7.12 BWRVIP-76 Core Shroud BWRVIP-76, Appendix K, states that plant-specific analyses for shroud fatigue will be reviewed to determine if there is a TLAA. A review of the WNPS plant-specific shroud analyses (Refs. 6.5.55, 6.5.56, and 6.5.57) identified one TLAA. WC-1362 (Ref. 6.5.55) contains GE report GE-NE-523-A005-0195, Vermont Yankee Core Shroud Primary Stresses". This report is a stress calculation and does not involve a TLAA. VYC-I 363 (Ref. 6.5.56) contains GE proprietary report GE-NE-523-A194-I 294, which develops the seismic model and does the seismic analysis for the core shroud. This report is structural model calculation and does not involve a TLAA. The WNPS calculation (WC-1364, Ref. 6.5.57) of the allowable interval between inspections for various core shroud welds (2 cycles for some and 1 cycle for others) uses the limit load analysis techniques described in ASME Code,Section XI to calculate crack growth, which is valid as long as total neutron fluence remains below 3~10~' n/cm2. (Ref. 6.5.57) Extrapolation of neutron fluence data from References 6.5.1 1 and 6.5.44 show that shroud fluence will be approximately 15x1 O2' n/cm2 at the end of the period of extended operation (54 EFPY). The extrapolation was based on 3.518~10~ MWH (Ref.

6.5.44) at 1583 MWt with a peak flux of 8.17~10'~ n/cm2/sec (Ref.

6.5.11) and 7.943E8 MWH (Ref. 6.5.44) at 1912 MWt with a peak flux of 9.67~10'~

n/cm2/sec (Ref. 6.5.44). This compares reasonably to 1.39~10" n/cm2 for 54 EFPY at 1583 MWt as calculated in Ref 6.5.1 1. Therefore, this calculation remains valid for the period of extended operation per 10 CFR 54.21 (c)('l)(i).

LRPD-03 Revision 0 Page 34 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations Best Estimate k Reactor Building Crane 3007 1 0.40 3.8 Other Plant-Specific TLAA Conservative Cycles 1 o.50 7590 3.8.1 Crane Load Cycles In the late 1970s, the NRC requested all licensees of operating reactors to review their controls for handling heavy loads to determine the extent to which the guidelines of NUREG-0612 were satisfied, and to identify the changes and modifications that would be required to fully satisfy these requirements (Ref.

6.2.22). Licensee responses required verification that crane designs complied with the guidelines of Crane Manufacturer's Association of America (CMAA)

Specification 70 and Chapters 2-1 and 2-2 of ANSI B30.2-1976, including the demonstration of equivalency of actual design requirements for instances where specific compliance with these standards is not provided.

WNPS's response (Ref. 6.5.42) identified that NUREG-0612 applied only to the reactor building crane at WNPS. The reactor building crane was designed and built by Whiting Corporation and has a main hook load rating of 110 tons, and an auxiliary hook load rating of 7.73 tons. The reactor building crane was modified in 1976 by replacing the original trolley with one that has a dual load path on the main hoist when used for shipping cask operation. The modification satisfies the intent of APCSB BTP 9-1 which called for the crane to be designed and fabricated to a number of industry standards, including ANSI B30.2 and CMAA-70. The staffs safety evaluation of this modification, as transmitted by letter from R. Reid (NRC) to R. Groce (Yankee Atomic) on January 28, 1977, approved the modifications, implicitly confirming compliance with CMAA-70. A subsequent review was deemed unnecessary.

CMAA-70 calculates allowable stress range based on joint category and service class, which in turn assumes a number of cycles. However, this is not a TLAA as the calculation does not specifically use the number of cycles. The minimum number of load cycles in CMAA-70 is 20,000, for Class A cranes, with a mean effective load factor range of 0.35-0.53. The reactor building crane at WNPS has been reviewed in accordance with CMAA-70, and is conservatively classified as a Class A crane for this review. The total load cycles and mean effective load factors for this crane have been estimated for the period of extended operation.

Even using conservative estimates, total load cycles are well below 20,000 and effective load factors are well below 0.53 (Table 3.9-1). Therefore, the crane allowable stress range remains valid through the period of extended operation.

3.8.2 Reflood

Thermal Shock of the Reactor Vessel lnternals UFSAR Section 3.3.5.4 addresses reflood thermal shock of the reactor vessel internals (core shroud). This evaluation of thermal shock is a TLAA as it is based on the shroud receiving a maximum integrated neutron fl uence of 2.7~1 O2' n/cm2 (greater than 1 MeV) by the end of plant life. The value of 2.7~10~~ n/cm2 is a generic value that bounds all BWRs. To show that I Page35of98 TLAA and Exemption Evaluations I WNPS License Renewal Project TLAA and Exemption Evaluations VYNPS remains bounded for the period of extended operation, it is adequate to show that shroud fluence for 54 EFPY remains below 2.7~10~~ n/cm2. The peak shroud flux was calculated (Ref. 6.5.44) for the extended power uprate at 9.67~10'~

n/cm2-sec. Integrating this and the pre-uprate flux from Ref. 6.5.1 1 gives an end of life shroud fluence of 1 .5x1020 n/cm2. This value remains below the 2.7 x102' n/cm2 value used in the evaluation discussed in the UFSAR. As such, this TLAA remains valid for the period of extended operation in accordance with 1 OCFR54.21 (c)(l)(i).

LRPD-03 Revision 0 Page 35 of 98 3.8.3 Reflood Thermal Shock of the Reactor Vessel I WNPS License Rer Section B.4.9 of the WNPS UFSAR refers to a thermal shock analysis performed on a representative GE BWR reactor vessel. This analysis is in GE topical report NEDO-10029, An Analytical Study in Brittle Fracture of GE-BWR Vessel subject to the Design Basis Accident (LOCA). NEDO-10029 also appears in Table 1 .IO of the WNPS UFSAR. WNPS has reviewed this analysis and determined that it is not a TLAA. It does not satisfy 10 CFR 54.3 (4) in that this analysis was not used by the licensee in making any safety determination. Rather, this analysis was prepared by GE in 1969 to answer concerns of the ACRS. In the 1969 time frame, thermal shock was a concern for all commercial reactor vessels. Regulatory Guide 1.2, Thermal Shock to Reactor Pressure Vessels, addressed these concerns. As more information was developed, it became clear that this was a concern for Pressurized Water Reactors and not for Boiling Water Reactors. Regulatory Guide 1.2 was withdrawn. The withdrawal notice says Regulatory Guide 1.2 was superceded by 1 OCFR50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, and by Regulatory Guide I .54, Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors. These two replacement documents apply only to PWRs. There are currently no requirements for analysis of thermal shock in BWRs. Brittle fracture of BWR vessels is addressed in BWRVIP-05. These analyses demonstrate that the probability of brittle fracture is acceptably low for BWR vessels, low enough to justify reduced inspection of the circumferential welds. This analysis included Level C and Level D events (including LOCA) as well as cold overpressure events. The analyses demonstrated that LOCAs were not limiting events because conditions that cause rapid vessel cooling also cause rapid depressurization. The NRC SER and SER supplement for BWRVIP-05 agrees that the probability of failure of a BWR vessel is sufficiently low that 100% inspection of the axial and circumferential welds is not required. Reactor vessel neutron embrittlement (CVUSE, RTNDT, P-T limits, circumferential and axial welds) have been addressed as TLAA based on accumulated fluence over the life of the plant. These TLAA address all required aspects of the BWR vessel. As such there is no additional significant information in NEDO-10029 requiring evaluation and it is not a TLAA. 3.8.4 Feedwater Nozzle Crack Growth Cracks were identified in the reactor vessel main feedwater nozzles in the mid 1970s. Modifications to the plant, including removing the cracks by grinding and installing re-designed WNPS License Renewal Project TLAA and Exemption Evaluations feedwater nozzle thermal sleeves, were completed in 1976 (Ref.

6.5.51). The maximum growth rate of feedwater nozzle cracks was evaluated by calculation WC-1005 (Ref. 6.5.51). This calculation is not a TLAA as it does not justify the use of the feedwater nozzles for the life of the plant. Rather, it calculates an acceptable interval between ultrasonic inspections to measure actual crack growth. As committed to the NRC in BW 01-02 (Ref. 6.5.52), the fatigue monitoring program as implemented in OP-4172 (Ref.

6.5.53) and AP-0145 (Ref.

6.5.54). counts the thermal cycles on the feedwater nozzles and requires re-inspection of the nozzles before the allowed number of cycles is reached.. Mitigation efforts (via chemistry control) to date have been very successful with little or no flaw growth noted (Ref. 6.5.51). W will continue to mitigate cracking, monitoring cycles, and re-inspect as necessary throughout the period of extended operation.

LRPD-03 Revision 0 Page 36 of 98 3.8.5 Section C2.2.2 of Appendix C to the UFSAR, discusses the probability of upset, emergency, and faulted conditions occurring in 40 years. The definitions support the quantitative event classifications, and were never meant to be precise quantitative values.

The probability of an event occurring in 40 years, P4*, was approximated only within an order of magnitude rather than calculated. Given the lack of any specific analysis, these probabilities are not TLAA. Upset, Emergency and Faulted Conditions

3.8.6 Section

1.4 of Appendix I of the UFSAR calculates the steam line break probability for a 40 year plant life. The UFSAR (Section 1.1) says these probabilities are being calculated to "point out the more critical components and locations within the reactor piping system, so that attention can be directed to those areas." The small changes in system reliability due to the period of extended operation will affect all systems and will not change the relative system reliabilities. This is not a TLAA as no safety related decisions are made based on the results of this analysis.

Probability of a Steam Line Break 4.0 Identification and Evaluation of Exemptions Pursuant to 10 CFR 54.21(~)(2), an applicant for license renewal must provide (1) a listing of plant-specific exemptions granted pursuant to 10 CFR 50.12 that are in effect and based on TLAA, and (2) an evaluation of these exemptions to justify their continuation for the period of extended operation.

This section identifies exemptions for VYNPS and concludes that no exemptions that remain in effect are based on TLAA. As discussed in Section 2.0, the searchable computer database that includes the UFSAR and NRC correspondence was reviewed. Attachment 2 provides a listing of the exemptions that were identified and lists the identified references for the exemptions.

In accordance with 10 CFR 54.21(~)(2), exemptions that are not in effect are not required to be discussed in the license renewal application. For evaluation purposes, exemptions that were found to remain in effect were grouped into the following categories:

0 Fire Protection Requirement Exemptions Other IO CFR Requirement Exemptions The following sections discuss the exemptions in each category and identify whether the exemptions are based on TLAA.

LRPD-03 Revision 0 Paae 37 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations 4.1 Fire Protection Requirement Exemptions The WNPS implementation of Appendix R resulted in eleven exemption requests to Sections III.G, 1II.J and 1II.L of 10 CFR 50 Appendix R. These exemptions are discussed in the WNPS Safe Shutdown Capabilities Analysis (Ref. 6.5.19) with additional references provided therein. None of these exemptions are based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore are not based on TLAA. 4.1 .I Exemption from Section lll.G.3.bJ relief from installation of automatic fire suppression through the control room, (Section 6.1 of Ref. 6.5.19) WNPS requested an exemption from the 10 CFR 50, Appendix R, Section III.G.1 requirements to have a fixed fire suppression system in the control room on the basis that the existing fire protection features in the control room are equal in effectiveness to a fixed fire suppression system. This exemption was granted because the control room is a unique area of the plant that is required to be continually occupied by the operators.

In the event of a fire, manual fire suppression would be effective and prompt. Because the operators provide a continuous fire watch in the control room, a fixed fire suppression system is not necessary to achieve adequate fire protection.

This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.2 Exemption from Section lll.G.2.a, installation of 3-hour rated fire barriers for the RCIC room, (Section 6.2 of Ref. 6.5.1 9) WNPS requested an exemption from the 10 CFR 50, Appendix R, Section lll.G.2.a because the door, stairwell and equipment hatch that provide RClC fire separation are not 3-hour fire rated fire barriers. This exemption was granted because the majority of the room is 3-hour fire rated, and the un- rated door, hatch, and stairs are heavy steel designed to withstand a high energy line break.

The fire load is low, and the fire detection system is alarmed in the control room, allowing early dispatch of the fire brigade. The staff concluded that completing the 3-hour fire rating of all room components would not significantly increase the level of fire protection in this zone. This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.3 Exemption from Section lll.G.2.bJ various reactor building area fire suppression (Section 6.3 of Ref. 6.5.19)

VYNPS requested an exemption from the 10 CFR 50, Appendix R, Section lll.G.2.b to the extent that the automatic fire suppression systems required by the code were not installed in several areas of the reactor building.

LRPD-03 Revision 0 Paae 38 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations The exemption was granted because the staff concluded that the existing fire protection, combined with the proposed fire protection measures in the subject zones, would provide a level of fire protection equivalent to the technical requirements of Section lll.G.2.b of Appendix R. This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.4 Exemption from Section lll.G.2.b, 20 feet of separation in Reactor Building elevation 252 East and West (RB3 and RB4), (Section 6.4 of Ref. 6.5.19)

WNPS requested an exemption from the 10 CFR 50, Appendix R, Section lll.G.2.b to the extent that it requires the installation of an automatic fire suppression system in the area and to the extent that it requires 20 feet of separation free of intervening combustibles.

This exemption was granted because W committed to install an early fire detection system, and several fire barriers. The NRC concluded that the fire load in the area was low and that a fire would develop slowly. Given the early warning detection system, and the new barriers, the fire could be extinguished manually before it damaged redundant equipment.

This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.5 Exemption from Section lll.G.2.b, 20 feet of separation in Reactor Building elevation 252, NW corner, (Section 6.5 of Ref. 6.5.19) WNPS requested an exemption for the northwest corner of the reactor building, RB-3 and RB- 4, from the 10 CFR 50, Appendix R, Section lll.G.2.b.

to the extent that it requires 20 feet of separation free of intervening combustibles.

This exemption was granted because the low fire loading, early warning detection system, manual fire suppression systems, other fire barriers and the 18 feet of existing separation provided adequate fire protection.

This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.6 Exemption from Section III.G.l.a, electrical repair (Section 6.6 of Ref. 6.5.19) WNPS requested an exemption from the 10 CFR 50, Appendix R, Section II1.G.l.a to allow connection of a batter charger, and replacement of fuses following a fire in the cable vault/cable spreading area. This exemption was granted based on the fact that the repairs were simple and quick, involving equipment staged in appropriate locations, with approved procedures in place. The repairs can be completed well before the systems are needed following a reactor scram.

LRPD-03 WNPS License Ren-----I n--'--A I Revision 0 Page 39 of 98 ewai rrqt?r;t TLAA and Exemption Evaluations This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.7 Exemption from Section lll.G.2.a, fire sealing on main steam and main feedwater line penetrations, (Section 6.7 of Ref. 6.5.19) WNPS requested an exemption from the 10 CFR 50, Appendix R, Section lll.G.2 because the penetration seals on the main steam and feedwater lines were un-rated versus the required 3- hour rated seals. This exemption was granted based on the subject penetration area being a high radiation area that is inaccessible to personnel during plant operation and hence free of transient combustibles. There are no other possible sources of fire. The staff concluded that the subject penetration provides adequate fire protection even with the unqualified seal, and that no particular enhancement would be gained if the existing seal were replaced with a qualified seal. This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.8 Exemption from Section lll.L.3, use of Vernon Tie for alternative shutdown equipment, (Section 6.8 of Ref. 6.5.19) WNPS requested an exemption from the 10 CFR 50, Appendix R, Section lll.L.3 to allow use of the Vernon tie-line as an alternative to the onsite emergency diesel generator for fire events involving the control room, the cable spreading room, and fire zones RB-1, RB-2, RB-3 and RB- 4 when offsite power is not available.

This exemption was granted based on the staffs conclusion that the Vernon tie line provides an acceptable alternative to power from an onsite emergency diesel generator when normal sources of offsite power are not available for a fire in the control room, cable spreading room, or reactor building fire zones RB-1/2/3/4. This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.9 Exemption from Section lll.L.2 and G.l .a, ADS and Low Pressure Injection use in fire areas where a offsite power is not available, (Section 6.9 of Ref. 6.5.19) WNPS requested an exemption from the 10 CFR 50, Appendix R, Section lll.L.2 and III.G.l to allow use of the automatic depressurization system (ADS) and low pressure injection (either core spray of low pressure coolant injection) as a means of achieving post-fire safe-shutdown conditions in fire zones RB-1, RB-2, RB-3, and RB-4 when offsite power is not available; Le. high pressure injection is not available.

This exemption was granted based on the staffs conclusion that the detection and suppression capabilities in fire zones RB-1/2/3/4 would be adequate to protect against fire hazards in the zones contingent on WNPS installing additional fire detection capability as committed.

The I Page40of98 TLAA and Exemption EvaiuaIions I - LRPD-03 Revision 0 Page 40 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations staff further concluded that the revised shutdown strategy for fire zones RB-1/2/3/4 (use of ADS with either CS or LPCI) and the re-designation of these fire zones as areas requiring an alternative shutdown capability provide and acceptable level of safe-shutdown protection.

i WNPS License Re1 ' -. -I This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.10 Exemption from Section IILJ, emergency yard lighting, (Section 6.10 of Ref. 6.5.19) VYNPS requested an exemption from the 10 CFR 50, Appendix R, Section 1II.J requirement to have 8-hour battery backup for lighting in the general yard and nitrogen storage area and instead use the security lighting as access lighting for these areas. This exemption was granted based on the security lighting being powered from a separate power source and therefore not being subject to fire loss. This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.1.1 1 Exemption from Section lll.G.2.c, Rockbestos cable in fire zone R, (Section 6.11 of Ref. 6.5.19) WNPS requested an exemption from the 10 CFR 50, Appendix R, Sectionlll.G.2.c to use fire resistant cables with Rockbestos insulation instead of the code requirement to enclose the cables in a l-hour rated fire barrier. This exemption was granted when the NRC staff concluded that the fire resistant cables provided essentially the same protection that a l-hour fire barrier would provide.

This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.2 Other 10 CFR Requirement Exemptions Two additional exemptions to the Code of Federal Regulations were identified. These exemptions involve the calculation of P-T limits using Code Case N-640 and the use of the alternate source term (AST) in plant analyses.

None of these exemptions are based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore none of these exemptions are based on TLAA.

LRPD-03 Revision 0 Page 41 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations

4.2.1 Exemption

from 10 CFR 50 Appendix G, Use of Code Cases N-588 and N-640 for calculating PTT limits In December, 2000, WNPS requested an exemption to use Code Case N-588 and N-640 to calculate the pressureltemperature curves for the technical specifications.

In February of 2001, WNPS amended the request to use just code case N-640. In April, 2001 the NRC (Ref.

6.2.20) acknowledged the withdrawal of Code Case N-588 and approved the use of code case N-640 for determining the P/l limits. Case N-640 allows use of the Klc equation in place of the KI, equation to calculate the P/T curves. The NRC affirmed that knowledge gained since issuance of the code demonstrates the margin of safety to protect the public health and safety from potential reactor vessel failure is conservative using the kl, equation and still sufficient to ensure the structural integrity of the reactor vessel using the k,, equation. This exemption is not based on calculations or analyses that consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore is not based on a TLAA. 4.2.2 Alternative Source Term (AST) Methodology W requested (Ref. 6.5.40) an exemption from lOCFR50.54(0) and 10CFR50, Appendix J, Option 8, Sections 1II.A and Sections 1II.B. The request was to use an alternative source term as discussed in 1 OCFR50.67.

Since that submittal, eleven supplemental submittals have been made and approval of this exemption is still anticipated by WNPS. The alternative source term calculation anticipated the requested power uprate to 1912 megawatts thermal, and operation at the maximum extended load line limit (MELLA) power-flow condition thus ensuring a bounding core isotopic inventory. This AST calculation, and associate accident re-analyses do not consider the effects of aging or involve time-limited assumptions defined by the current operating term and therefore the associated exemption is not based on TLAA.

LRPD-03 Revision 0 Page 42 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations

5.0 Summary

and Conclusions This report identified and listed TLAA and exemptions that were potentially applicable to the WNPS. The TLAA were evaluated using one or more of the three options in 10 CFR 54.21 (c)(l)* No exemptions that will remain in effect for the period of extended operation are based on TLAA.

VYNPS License Renewal Project TLAA and Exemption Evaluations LRPD-03 Revision 0 Page 43 of 98 6.0 References

6.1 Codes

and Standards VYNPS License Rer TLAA and Exemption Evaluations 6.1 .I 6.1.2 6.1.3 6.1.4 ' 6.1.5 6.1.6 6.1.7 I Page43of98 10 CFR Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants NUREG-1800, Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants, July 2001. NUREG-1801, Rev. 0, Volumes I and 2, Generic Aging Lessons Learned (GALL)

Report IEEE Std. 323-1974, Quali&ing Class ?E Equipment for Nuclear Power Generating Stations, The Institute of Electrical and Electronics Engineers, Inc., 1974. Requirements For Renewal of Operating Licenses For Nuclear Power Plants, IO CFR Part 54, Federal Register, Vol. 60 No. 88, Monday, May 8, 1995, (60 FR 22461 ), Final Rules, Includes the Statement of Considerations (SOC) for the Final Rule. SECY-93-049, Implementation of 70 CFR Part 54, ' Requirements for Renewal of Operating Licenses for Nuclear Power Plants,' March 1 , 1993 Environmental Qualification of Electrical Equipment Important to Safety, 10 CFR 50.49, Federal Register, Volume 48, No. 15, January 21, 1983.

6.2 NRC Documents 6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.2.6 6.2.7 6.2.8 6.2.9 US NRC, Reactor Vessel Integrity Database (RVID), Version 2.0.1, July, 2000 NRC Regulatory Issue Summary 2003-09 Environmental Qualification of Low- Voltage Instrumentation and Control Cables, May 2, 2003 Reg. Guide 1.89, Rev. I, Environmental Qualification of Certain Electrical Equipment Important to Safety for Nuclear Power Plants Memo from L. Joseph Callan, Executive Director for Operations, NRC, to Chairman and Commissioners, USNRC,

Subject:

Report on the Status of the Environmental Qualification Task Action Plan, November 15, 1996 NRC IE Bulletin 79-01 B, Environmental Qualification of Class IE Electrical Equipment, January 14, 1980, Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission NUREG 0588, Rev. 1, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors (DOR Guidelines), Enclosure 4 to IE Bulletin 79- 016, U.S. Nuclear Regulatory Commission, Washington, DC, January 14, 1980 NRC Regulatory Issue Summary (RE) 2003-09, May 2, 2003, Environmental Qualification of Low-Voltage Instrumentation and Control Cables NUREG/CR-5799, March, 1992, Review of Reactor Pressure Vessel evaluation for Yankee Rowe Nuclear Power Station LRPD-03 Revision 0 Page 44 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations 6.2.10 NW 04-016, Safety Evaluation of Relief Requests for the fourth 10-year interval of the Inservice Inspection program - Vermont Yankee Nuclear Power Station (TAC NOS. MB8349 through MB 8358) 05 March 2004 6.2.1 1 NW 04-027, R. B. Ennis to M. Kansler, Vermont Yankee Nuclear Power Station - Issuance of Amendment re: Reactor Pressure Vessel Fracture Toughness and Material Surveillance Requirements (TAC NOS. MB8119 and MB8379), 29 Mar 2004 6.2.12 NW 03-78, Vermont Yankee Nuclear Power Station - Relief Request NOS. RR-POI, RR-P02, RR-P03, RR-P04, RR-VOI, RR-V02 (TAC NOS. MB7489 through MB7494), 6 October 2003 6.2.13 NW 82-66, Exemption pertaining to requirement for fixed fire suppression in the Control Room, 10 May 1982 6.2.14 NW 86-240, Exemption from Appendix R to 10CFR50 Concerning Automatic Fire Suppression, Separation, and Repairs, 1 December 1986 6.2.15 NW 89-137, Issuance of Exemptions to IOCFRSO, Appendix R, Section 1II.J. Emergency Lighting, and Section lll.G.Z.a, Separation, 26 June 1989 6.2.16 NW 97-128, Vermont Yankee Nuclear Power Station (TAC NOS. M95442 and M95149), 12 August 1997 6.2.17 NW 97-42, Vermont Yankee Nuclear Power Station (TAC NO. M95760), 23 March 1997 6.2.18 NW 97-81, Vermont Yankee Nuclear Power Station (TAC NO. M95482), 9 June 1997 6.2.19 NRC letter, T. A. Alexion to M. R. Kansler (Entergy Nuclear Operations), . . . Exemption from the Requirements of IOCFR Part 20, Section 20.1003 Definition of Total Dose Equivalent . . . , 12 September 2002 requirements of 1 OCFR Part 50, Appendix G (TAC NO. MB0763), 16 April 2001 6.2.20 NW 01-39, Vermont Yankee Nuclear Power Station - Exemption from the 6.2.21 NW 84-139, Control of Heavy Loads (Phase I), 27 June 1984 6.2.22 Letter, USNRC to Vermont Yankee, Control of Heavy Loads at Nuclear Power Plants, 22 December 1980 6.2.23 NW 00-49, Safety evaluation of the request for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements for the containment inservice inspection program, Vermont Yankee Nuclear Power Station (TAC NO. MA8658), 19 May 2000 6.2.24 NRC letter from C.I. Grimes (NRC) to C. Terry, (BWRVIP Chairman), Acceptance for referencing of EPRl Proprietary Report TR-I 13596, "BWR Vessel and lnternals Project, BWR Reactor Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP- 74) and Appendix A, 'Demonstration of Compliance with the Technical Information requirements of the License Renewal Rule (10CRF54.21)', 18 October 2001 6.2.25 NRC letter from C.I. Grimes (NRC) to C. Terry (BWRVIP Chairman), Safety Evaluation for Referencing of BWR Vessel and lnternals Project, BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25) Report for Compliance with the License Renewal Rule (IOCFR Part 54), 7 December 2000 LRPD-03 Revision 0 Page 45 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations 6.2.26 NRC letter from C.I. Grimes (NRC) to C. Terry (BWRVIP Chairman), Acceptance for Referencing of BWR Vessel and lnternals Project, BWR Top Guide Inspection and Flaw Evaluation Guidelines (BWRVIP-26) Report for Compliance with the License Renewal Rule (10CFR50 Part 54), 7 December 2000 6.2.27 USNRC letter from Gus C. Lainas to Car Terry, Niagara Mohawk Power Company, BWRVIP Chairman, Final Safety Evaluation of the BWRVIP Vessel and lnternals Project BWRVIP-05 Report, (TAC No. M93925), July 28, 1998 Supplement to Final Safety Evaluation of the BWRVIP Vessel and lnternals Project BWRVIP-05 Report, (TAC No, MA3395), March 7,2000 GSI 168 for License Renewal, Project 690, dated June 2, 1998, ML031500232.

BWRVIP Vessel and lnternals Project, 'BWR Standby Liquid Control SystemKOre Plate DP Inspection and Flaw Evaluation Guidelines (BWRVIP-27)'

EPRl Report TR-107236 (TAC NO. M98708), 27 April 1999 6.2.31 USNRC letter S.A. Richards to J.F. Klapproth, Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891), MFN 01-050, September 14,2001 6.2.32 USNRC letter, J. T. Wiggins to L. A. England, Acceptance for referencing of Topical Report NEDO-32205, Revision 1, '10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWW2 through BWRl6 Vessels, 8 Dec 1993 6.2.33 NVY 01-46, NRC Letter with SER, R.M. Pulsifer to M.A. Balduzzi, Vermont Yankee Power Station - Issuance of Amendment RE: PIT Limit Curves (TAC NO. MB0764), May 4,2001 6.2.28 USNRC letter from Jack R. Strosnider, Jr., to Car Terry, BWRVIP Chairman, 6.2.29 USNRC letter from C. I. Grimes (NRC) to D. Walters (NEI), Guidance on Addressing 6.2.30 USNRC letter from J.R. Strosnider to Carl Terry (BWRVIP), Safety Evaluation of the 6.3.1 6.3.2 6.3.3 6.3.4 6.3.5 6.3.6 6.3.7 6.3.8 6.3 lndustry Documents NE1 95-10, Revision 3, lndustry Guideline for lmplernenting the Requirements of IO CfR Part 54- the License Renewal Rule License Renewal Application, Dresden Nuclear Power Station and Quad Cities Nuclear Power Station, January 2003 License Renewal Application, Edwin I. Hatch Nuclear Power Plant Units I and 2 License Renewal Application, Peach Bottom Atomic Power Station Units 2 and 3, July 2001 EPRl 1000174, Rev. 1, Oconee Electrical Component Integrated Plant Assessment and Time-Limited Aging Analyses, November 1995 EPRl 1003458, License Renewal Electrical Template EPRl 1003057, License Renewal Electrical Handbook EPRl 1000866, Rev. 1, Summary of Generic License Renewal Technical Issues, June 2001 LRPD-03 Revision 0 Page 46 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations 6.3.9 EPRl TR-105090, Guidelines to Implement the License Renewal Technical Requirements of 10 CFR 54 for Integrated Plant Assessments and Time-Limited Aging Analyses, November 1995 6.4.1 6.4.2 6.4.3 6.4.4 6.4.5 6.4.6 6.4.7 6.4.8 6.4.9 6.4 BWRVIP Documents BWRVIP-05, EPRl Report TR-105697, BWR Vessel and lnternals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), For the Boiling Water Reactor Owners Group (Proprietary), September 28, 1995, with supplementing letters of June 24 and October 29, 1996, May 16, June 4, June 13, and December 18,1997, and January 13,1998 BWRVIP-18, EPRl Report TR-106740, BWR Core Spray lnternals Inspection and Flaw Evaluation Guidelines (BWRVIP-18), July 1996 BWRVIP-25, EPRl Report TR-107284, BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25), December 1996 BWRVIP-26, EPRl Report TR-107285, BWR Top Guide Inspection and Flaw Evaluation Guidelines (BWRVIP-26), December 1996 BWRVIP-27, EPRl Report TR-107286, BWR Standby Liquid Control SystemKOre Plate UP Inspection and Flaw Evaluation Guidelines (BWRVIP-27), April 1997 BWRVIP-38, EPRI Report TR-108823, BWR Shroud Support Inspection and Flaw Evaluation Guidelines (BWRVIP-38), September 1997 BWRVIP-41, EPRl Report TR-108728, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines (BWRVIP-41), October 1997 BWRVIP-47, EPRl Report TR-108727, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines (BWRVIP-47), December 1997 BWRVIP-48, EPRl Report TR-108724, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines (BWRVIP-48), February, 1998 6.4.10 BWRVlP-49A, EPRl Report 1006602, BWR Vessel and lnternals Project Instrument 6.4.1 I BWRVIP-74, EPRl Report TR-113596, BWR Reactor Pressure Vessel Inspection 6.4.12 BWRVIP-76, EPRl Report TR-I 14232, BWR Core Shroud Inspection and Flaw 6.4.13 BWRVIP-86, EPRl Report 1000888, BWR Integrated Surveillance Program 6.4.14 BWRVIP-I 16, EPRl Report 1007824, Integrated Surveillance Program (ISP)

Penetration Inspection and Flaw Evaluation Guidelines, March 2002 and Flaw Evaluation Guidelines (BWRVIP-74), September 1999 Evaluation Guidelines November 1999 Implementation Plan, December 2000 Implementation for License Renewal, July 2003 6.5 WNPS Documents 6.5.1 WNPS Updated Final Safety Analysis Report, Revision

17. 6.5.2 BW 03-80, 9/10/2003, J.K.Thayer to USNRC Document Control Desk, Technical Specification Proposed Change No. 263, Extended Power Uprate LRPD-03 Revision 0 Page 47 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations 6.5.3 6.5.4 6.5.5 6.5.6 6.5.7 6.5.8 6.5.9 6.5.10 6.5.1 1 6.5.12 6.5.13 6.5.14 6.5.15 6.5.16 6.5.17 6.5.18 6.5.19 6.5.20 6.5.21 NEDC-33090P Rev 0, (Attachment 4 to Reference 6.5.2), September 2003, Safety Analysis Report for Vermont Yankee Nuclear Power Station Constant Pressure Power Uprate Appendix A to Operating License DPR-28, Technical Specifications and Bases for Vermont Yankee Nuclear Power Station Vernon, Vermont Docket No. 50-271, Amendment

1. 208 Vermont Yankee Technical Requirements Manual (TRM), Revision 19 QAPM, Entergy Quality Assurance Program Manual, Effective Oct. 27,2003 Vermont Yankee Nuclear Power Corporation Fire Hazards Analysis, Revision 5, 8/20/02 Vermont Yankee Fire Protection Commitment Reference Manual, Revision 1, 9/5/2000 PP 701 1, Vermont Yankee Fire Protection and Appendix R Program, Rev.

1, 07/05/01 CURATOR search engine - includes plant correspondence GE-NE-0000-0007-2342-R1 -NP, Rev. 1, July 2003, Entergy Northeast Vermont Yankee Neutron Flux Evaluation Operation and Maintenance Instructions:

Reactor Assembly for Vermont Yankee Nuclear Power Station, General Electric Report GEK-9608, December 1970 NEDO-32205-A, Rev.

1 1 OCFR50 Appendix G, Evaluation of Equivalent Margin Analysis for Low Upper Shelf Energy in BWW2-6 Vessel, February 1994 BW 93-146, L.A. Tremblay, Jr. to USNRC Document Control Desk, Additional Information Regarding Generic Letter 92-01

Reactor Pressure Vessel Structural Integrity, 21 Dec 1993 BW 93-107, L. A. Tremblay, Jr.

to NRC Document Control Desk, Response to Request for Additional Information, GL 92 Reactor Vessel Structural Integrity, September 24, 1993 BW 03-29, M. A. Balduzzi to USNRC Document Control Desk, Technical Specifications Proposed Change No. 258, RPV Fracture Toughness and Material Surveillance Requirements, 26 March 2003 BW 03-28, 1 April 2003, Fourth-Interval lnservice Inspection Program Plan, Fourth-Interval Inservice Inspection Pressure Test Program, and Request for Approval of IS1 Relief Requests Battelle-Columbus Report BCL-585-84-3, Examination, Testing and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Vermont Yankee Nuclear Power Station, 8/15/84 Vermont Yankee Nuclear Power Plant Safe Shutdown Capability Analysis (SSCA), Revision 6, 12/22/99 BW 03-63, Relief Request to use ASME Code Case N-600, August 11,2003 BW 03-87, Supplement to Third Interval Inservice Inspection (ISI) Program - Submittal of Relief Request B-5 'Limited Examinations', 1 October 2003 WNPS License Renewal Project TLAA and Exemption Evaluations 6.5.22 6.5.23 6.5.24 6.5.25 6.5.26 6.5.27 6.5.28 6.5.29 6.5.30 6.5.31 6.5.32 6.5.33 6.5.34 LRPD-03 Revision 0 Page 48 of 98 BW 03-83, Supplement to Fourth-Interval Inservice Inspection (ISI) Program Plan - Submittal of Relief Request ISI-O6,25 September 2003 BW 03-89, Supplement 2 to Fourth-Interval Inservice Inspection (El) Program Plan - Submittal of Relief Request Rl-01, 1 October 2003 BW 03-1 20, Supplement to Relief Request RI-01, 23 December 2003 BVY 04-04, Supplement to Fourth-Interval Inservice Inspection (ISI) Program Plan - Submittal of Revised Relief Request ISI-09, 12 Jan 2004 BW 04-07, Supplement to Relief Request RI-01, 22 January 2004 BW 04-20, Supplement to Fourth-Interval Inservice Inspection (ISI) Program Plan - Withdrawal of Relief Request PT-1, 18 February 2004 BW 04-22, Response to RAI on Relief Request to use ASME Code Case N-600,4 March 2004 BW 04-027, Response to RAI on Relief Request to use ASME Code Case N-600 - Supplement I, 18 March 2004 FW 82-13, Request for Exemption from Appendix R, Section lll.G.3.6, relating to fixed suppression in the Control Room, 17 February 1982 FW 85-38, letter WNPC to USNRC requesting exemptions from the provisions of Appendix R 24 April 1985 BW 89-13, Request for Exemption to Section lll.G.2.a for Steam Tunnel to Turbine Building Unqualified Penetration, 2 February 1989 BW 96-43, Request for Exemption from 10CFR Part 50, Appendix R, Section III.L, 'Alternative and Dedicated Shutdown Capability', 04 April 1996 BW 96-67, Request for Exemption from IOCFR Part 50, Appendix R,Section III.G, 'Fire protection of safe shutdown capability' and Section III.L, 'Alternative and Dedicated Shutdown Capability', 21 May 1996 6.5.35 BW 94-128, Request for Exemption, IOCFR, Part 50, Appendix R,Section III.J, 6.5.36 BW 96-58, Request for Exemption from IOCFR Part 50, Appendix R,Section III.G, 6.5.37 BW 02-60, Amendment to 'Request for Exemption from 1 OCFR20.1003 Definition of 'Emergency Lighting', 28 December 1994 'Fire protection of safe shutdown capability', 28 May 1996 "Deep-Dose Equivalent" and Permission to use External Whole Body "Weighing Factors" other than 1 .O', 07 August 2002 and Exemption Request to use Code Case N-588 and N-640,19 Dec 2000 Withdrawal of Exemption Request to use Code Case N-588 Term, 31 July 2003 Alternative Source Term - Meteorological Database for Ground-Level Releases, 17 March 2004 6.5.38 BVY 00-1 13, Technical Specification Proposed Change No. 244 Revised P/T Curves 6.5.39 BW 01-13, Supplement to Technical Specification Proposed Change No. 244 6.5.40 BVY 03-70, Technical Specification Proposed Change No. 262 Alternative Source 6.5.41 BW 04-032, Technical Specification Proposed Change No. 262 - Supplement 11 WNPS License Renewal Project TLAA and Exemption Evaluations 6.5.42 6.5.43 6.5.44 6.5.45 6.5.46 6.5.47 6.5.48 6.5.49 6.5.50 6.5.51 6.5.52 6.5.53 6.5.54 6.5.55 6.5.56 6.5.57 6.5.58 6.5.59 6.5.60 6.5.61 LRPD-03 Revision 0 Page 49 of 98 FW 81-134, Control of Heavy Loads, 11 September 1981 WNPS Reactor Vessel lnternals Management Program, PP 7027, Revision 1, 27 September 2002 GE-NE-0014-0292-01, Entergy Nuclear Operations Incorporated Vermont Yankee Nuclear Power Station Extended Power Uprate, Task TO31 3: RPV Flux Evaluation, May 2003 WNPS Document, Environmental Qualification Master Equipment List (EQMEL) - EQ Program Manual, Volume I, Section 6.0. WNPS Document, AP 0092, Environmental Qualification (EQ) Document Change Notification WNPS Document, Environmental Qualification Program Manuals, Volume I (Rev. 47) & II (Rev. 13)

WC-193, Main (W Design Basis Radiation Dose Calculation Specifications)

WNPS EQ Database (information tool only at this time)

VYNPS Document, AP 0305 Rev 1 I, EQ Maintenance and Surveillance (ME) Program VYNPS Calculation, VYC-1005, Revision 2, 711 1/02, Crack Growth Calculation for the Vermont Yankee FW Nozzles VYNPS letter BW 01-02,22 January 2001, D. M. Leach to USNRC Document Control Desk, Vermont Yankee Nuclear Power Station License No.

DPR-28 (Docket NO. 50-271 ) Alternative Feedwater Nozzle Inspection VYNPS Operating Procedure OP-4172, Revision 25, 04/27/2004, Feedwater System Surveillance VYNPS Administrative Procedure AP-0145, Revision 8, 02/26/1999, Equipment Cycle Record Keeping VYNPS Calculation WC-1362, Vermont Yankee Core Shroud Primary Stresses, Revision 1, 3/16/95 VYNPPS Calculation WC-1363, Vermont Yankee Core Shroud Primary Stresses VYNPS Calculation WC-1364, Vermont Yankee Core Shroud Flaw Evaluation, Revision 4, 11/13/96 VYNPS Procedure PP-7015, Vermont Yankee lnservice Inspection Program; Revision 3, 09/01/2003 Technical Report TR-5319-1 (Teledyne Engineering Services), Plant Unique Analysis Report of the Torus Suppression Chamber for Vermont Yankee Nuclear Power Station, Revision 2, 30 November 1983 Technical Report TR-5319-2 (Teledyne Engineering Services), Plant Unique Analysis Report of the Torus Attached Piping for Vermont Yankee Nuclear Power Station, 30 September 1983 GE-NE-0000-0010-6295-01, Task Report 0301 WNPS License Rer-----'

TLAA and Exemption waiuarions I Page50of98 I 6.5.62 MPR-751, Mark 1 Containment Program Augmented Class 2/3 Fatigue Evaluation Methods and Results for Typical Torus Attached and SRV Piping Systems, November, 1982 6.5.63 WNPS Drawing 5920-3773, Assembly, Reactor, (GE Drawing 104R940) 6.5.64 Minor modification MM 2003-040, Reactor Building Auxiliary Hoist Upgrade 6.5.65 WNPS Procedure, OP-2200, Operation of the Reactor and Turbine Bridge Cranes, Revision 17, LPC #I 1 , 1211 6/2003 6.6 WNPS License Renewal Documents 6.6.1 LRPG-01 , License Renewal Project Plan 6.6.2 WNPS Report LRPD-02, Aging Management Program Evaluation Results 6.6.3 WNPS Report LRPD-04, TLAA - Mechanical Fatigue 6.6.4 LRPG-08, TLAA and Exemption Evaluations Attachment 1 - List of Potential TLAA and References 6.5.14 Analysis projected per 3.1.4 6.5.3 10CFR54.21 (c)(l)(ii) Adjusted Reference Temperature Analysis projected per 3.1.5 6.4.1 1 10CFR54.21 (c)(l)(ii)

Reactor vessel circumferential welds ~~ I Analysisprojectedper 6.4.1 1 10CFR54.21 (c)(l Mil Reactor vessel axial welds . ,. ,. , , Not a TLAA 3.1.7 6.5.18 Surveillance Specimen testing I Metalfati ue NA LRPD-04 Class 1 fatigue Analysis (CUFs) remain NA LRPD-04, Section 2.1.3 .- weld overlay Containment Corrosion I NA LRPD-04, Section 2.2.1 valid for'the period of extended operation per 10CFR54.21 (c)( l)(i) OR Aging effect managed per

?OCFR54.21(c)(?)(iii) Analyses remains valid through the period of extended operation per 'lOCFR!X21(c)(-l)(i). Analyses are not TLAA NA LRPD-04, Section 2.2.2 Analysis is not a TLAA. NA LRPD-04 I Section 2.4.1 Analysis is not a TLAA. 1 NA LRPD-04 Section 2.4.2 Section 2.4.3 Analysis remains valid NA LR PD-04 through the period of extended operation per 1OCFR54.21 (c)(l)(i).

Analysis is not a TLAA. NA LRPD-04 Section 2.4.4 ~ Analysis is not aTLAA. 1 NA I LRPD-04 I I Section 2.4.5 LRPD-03 Revision 0 Page 52 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations Attachment 1 - List of Potential TLAA and References TLAA Description Primary Containment Localized Thinning VYNPS Response to Bulletin 88- 08 Effects of Reactor Water Environment on Fatigue Life Environmental Qualification Analyses for Electrical Components Concrete Containment Tendon Prestress Analysis Containment Liner Plate, M Penetrations Fatis Fatigue analysis of the torus Fatigue analysis of the SRV discharge piping Fatigue analysis of other torus attached piping Reference Analysis projected per 1 OCFR54.21 (c)(l)(ii)

Aging effect managed lOCFR54.21 (c)(l)(iii)

Not applicable for BWRs 3.4 Section 2.5.2 6.2.3 6.2.5 6.5.45 6.5.46 I I al Containment, and I I LRPD-04 e Analyses I I Analyses are proiected I I LRPD-04, Section

2.3.1 through

the period of extended operation per 1 OCFR54.21 (c)(l )(ii). Analvsis remains valid I LRPD-04, Section 2.3.2 per 10CFR54.21 (c)(l)(i)

AND Analysis projected per IOCFR54.21(c)(l

)(ii) Analysis is projected I LRPD-04, Section 2.3.3 through the period of extended operation per I BWRVlP I , welds LRPD-03 Revision 0 Page 53 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations Attachment I - List of Potential TLAA and References BWRVIP-38, Shroud support c BWRVIP-41, Jet pump BWRVIP-48, Vessel ID attachment welds BWRVIP-49, Instrument penetrations BWRVIP-74, Reactor pressure vessel P-T Curves Fatigue Equivalent Margins for USE Reactor Vessel Welds BWRVIP-76, Core shroud Resolution Option -NoTLAA identified in BWRVIP-18.

TLAA for loss of preload for core plate bolts is projected through the period of extended operation by BWRVIP-25 Appendix B per IOCFR54.21(c)(l)(ii). Not a TLAA No TLAA for WNPS related to SLCIAP nozzle. Analysis remains valid through the period of extended operation per 1 OCFR54.21 (c)(l )(i). No TLAA for WNPS related to the jet pumps. Analysis remains valid for the period of extended operation in accordance with IOCFR54.21(c)(l)(i) The analysis remains valid for the period of extended operation in accordance with 1 OCFR54.2 1 (c)( 1 )( i). The analysis remains valid for the period of extended operation in accordance with lOCFR54.21(c)(l)(i).

P-T curves are addressed in Section

3.1.2. Fatigue

is addressed in EMA for USE are LRPD-04. addressed in Section 3.1.3. RPV welds are addressed in Sections 3.1.5 and 3-1 -6. The analysis remains valid for the period of extended operation in accordance with IOCFR54.21(c)(l)(i).

3.8.4 BWRVIP-26

3.8.5 BWRVl

P-27 I 3.8.6 BWRVl P-38 3.8.7 BWRVIP-41 3.8.8 BW RVI P-47 3.8.9 BWRVl P-48 BWRVl P-49 3.8.1 0 LRPD-03 Revision 0 Page 54 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations Attachment I - List of Potential TLAA and References c TLAA Description Other TL Crane Load cycles Reflood thermal shock of the core shroud. Reflood thermal shock of the reactor vessel Feedwater nozzle crack growth Upset, Emergency and Faulted Conditions Probability of a steam line break Fire protection e Exemptions from 10 CFR 50 Appendix R Exemption to use code case N- 640 for calculation of P-T limits Other 10 CFR e, Alternative Source Term [AST) Resolution Option 4A Not a TLAA. Attachment 4 shows allowable cycles will not be exceeded for the period of extended operation. Analysis remains valid for the period of extended operation in accordance with lOCFR54.2I(c)(l)(i). Not a TLAA Not a TLAA, not based on the life of the plant.

Not a TLAA, no analysis. Not a TLAA, no safety related decisions based on this probability. Not a TLAA remptions emptions Not a TLAA, an application of updated technology Not a TLAA. Section 3.9.1 3.9.2 3.9.3 3.9.4 3.9.5 3.9.6 4. ,l 4.2.1 4.2.2 Ref ere n ce 6.2.21 6.5.42 I UFSAR Section 3.3.5.4 UFSAR Section J. 6.5.51 Section C2.2.2 of the UFSAR Section 1.4 of the UFSAR 6.5.1 9 6.2.20 6.5.40 WNPS License Renewal Project TLAA and Exemption Evaluations Attachment 2 - List of Exemptions and References Yes - discussed in SSCA 6.7 LRPD-03 Revision 0 Page 55 of 98 Section 4.1.8 6.2.1 6 (approved)

SSCA 6.9 Yes - discussed in 6.5.34 (submitted)

Section 4.1.9 6.2.16 (approved)

SSCA 6.10 Yes - discussed in 6.5.35 (submitted)

Section 4.1.10 6.2.17 (approved)

SSCA 6.1 1 Yes - discussed in 6.5.36 (submitted)

Section 4.1.1 1 6.2.18 (approved)

~ Otl Exemption from 10 CFR 20.1 003, definition of total effective dose equivalent Code Cases N-588 and N-640 for P/T curves Alternative source term (AST) methodology x exemptions Yes (Not in scope for license renewal - only Part 50 exemptions are in scope. Exemption was reviewed and it is not based on anytime- limited assumptions.)

Yes - discussed in section 4.2.1 6.5.37 (submitted) 6.2.19 (approved) 6.5.38 (submitted) 6.5.39 (withdrew N-5881 6.2.20 (approved)

' 6.5.40 (submitted)

No - awaiting approval discussed in section 4.2.2 (01 #Error! Reference source not 6.5.41 (supp #I 1)

LRPD-03 Revision 0 Page 56 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations FSAR Section FSAR Text Attachment 3 - UFSAR TLAA Search Results Recommended Change

1.2 Definitions

36 Shutdown 1.5.2 Power Generation Design Criteria (Planned Operation) Nuclear Systems 1.6.5.9 Makeup Water Treatment System 2.4.4.3 Public Use 2.5.2.1 Introduction 2.5.2.3 Regional Geology Shutdown - The reactor is shutdown when the effective multiplication factor (keff) is sufficiently less than 1 .O that the full withdrawal of any one control rod could not produce criticality under the most restrictive potential conditions of temperature, pressure, core age, and fission product concentration.

3. The fuel cladding shall be designed to accommodate without loss of integrity the pressures generated by the fission gases released from the fuel material throughout the design life of the fuel. The Makeup Water Treatment System processes raw river water from the Connecticut River to maintain a supply of high quality water which may be used as a makeup for the station and reactor cycles.

Scale samples were taken from selected species for age-growth studies. These borings show that the area is overlaid by glacial deposits from the Pleistocene Age, with an average 30 feet of glacial overburden above the local bedrock, which consists of a hard biotite gneiss. Foliated igneous rocks of middle-and late-Devonian age underlie a large portion of the region.

~~ No change required.

No change required.

No change required. No change required.

No change required.

No change required.

WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results LRPD-03 Revision 0 Page 57 of 98 UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section 2.5.2.3.2 Geological History 3.2.3 Description ~ ~~ 3.3.5.1 Evaluation Methods 3.3.5.4 Thermal Shock FSAR Text In all cases, the faults dip steeply, and appear to be Triassic or younger in age. . . . All minor faults in the region appear to be high-angle and Triassic or younger in age. Sufficient plenum volume is provided to prevent excessive internal pressure from these fission gases or other gases liberated over the design life of the fuel. ~~ ___~ ___ The ASME Boiler and Pressure Code, Section Ill for Class A vessels, is used as a single guide to determine limiting stress intensities and cyclic loadings for the reactor vessel internals. The peak strain resulting in the shroud support plate is about 6.5%. This strain is higher than the 5.0% strain permitted by the ASME Code, Section Ill, for ten cycles. However, if the ASME Code curve is extended below ten cycles, the peak strain of 6.5% corresponds to about six allowable cycles.

Figure 3.3-10 illustrates both the ASME Code curve and the basic material curves from which it was established (with the safety factor of two on strain or twenty on cycles, whichever is more conservative).

The extension of the ASME Code curve represents a similar criteria to that used in the ASME Code, Section Ill, but applied to fewer than ten cycles of loading.

For this Type 304 stainless steel material, a 10% peak strain corresponds to one allowable cycle of loading.

Even a 10% strain for a single cycle loading represents a very conservative suggested limit because this has a Recommended Change

~ No change required.

No change required. Fuel design life isn't changing.

None. ASME section Ill is still the single guide. None. The shroud strain doesn't change.

This is not a 40 year TLAA. None. Peak strain isn't changing. Conditions which lead to calculated peak strain still aren't expected to occur during plant lifetime.

LRPD-03 Revision 0 Page 58 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section FSAR Text large safety margin below the point at which even minor cracking is expected to begin. Because the conditions which lead to the calculated peak strain of 6.5% are not expected to occur even once during the entire reactor lifetime, the peak strain is considered tolerable.

The ASME Code, Section 111, allows 220 cycles of this loading, thus no significant deformations result.

The most irradiated point on the inner surface of the shroud is subjected to a total integrated neutron flux of 2.7 x 1 02' nvt (greater than 1 MeV) by the end of plant life. The peak thermal shock stress is 155,700 psi, corresponding to a peak strain of 0.57%. The shroud material is Type 304 stainless steel, which is not significantly affected by the expected level of irradiation.

The material does experience some hardening and an apparent loss in uniform elongation, but it does not experience a loss in reduction of area. Because reduction of area is the property which determines tolerable local strain, irradiation effects can be neglected. Recommended Change No Change. This is generic GE material.

The VY shroud fluence for 54 EFPY at 1912 MWE, based on the shroud flux values in GE-NE-0000-0014-0292-01 is still below the 2.7~10~' generic value in the UFSAF WNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 59 of 98 ~~ FSAR Section 3.3.6 Inspection and Testing FSAR Text 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 recirculation flow 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-induced equipment 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 Ill. 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.

Recommended Change ~~ -~ No change required.

WNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 60 of 98 FSAR Section 3.4.5.2.2 Materials of Construction 3.4.5.3.1 CRD Hydraulic Supply and Discharge Subsystems 3.4.7.1 (CRD) Development Tests 3.7.2 Power Generation Design Bases

4.2.2 Power

Generation Design Bases FSAR Text 3. Inconel750 is used for the collet fingers, which must function as leaf springs when cammed open to the unlocked position. Colmonoy 6 hard facing is applied to the area contacting the index tube and unlocking cam surface of the guide cap to provide a long-wearing surface adequate for design life.

Although the drives can function without cooling water, the life of their seals is shortened by exposure to reactor temperatures.

The development drive (one prototype) testing prior to 1970 included over 5,000 scrams and approximately 100,000 latching cycles during 5,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of exposure to simulated operating conditions. That usable seal lifetimes greater than 1,000 scram cycles may be expected.

1. The ability to achieve rated core power output throughout the design lifetime of the fuel without sustaining fuel damage.

~ ~ 1. The reactor vessel design lifetime shall be 40 years. Recommended Change No change needed. This is an active component with no guarantee of a 40 year life. No change needed, this is still true. No change needed, this is still true. No change needed-not based on life of the F.-rnt. No change needed, fuel design lifetime isn't changing.

The original reactor vessel design lifetime was WO years. The vessel is acceptable for 60 years of operation.

LRPD-03 Revision 0 Page 61 of 98 WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results I UPDATED FINAL SAFETY ANALYSIS REPORT 1 FSAR Section I 4.2.3 (RV) Safety Design Basis FSAR Text 2.To minimize the possibility of brittle fracture failure of the nuclear system process barrier, the following shall be required:

(1 ) the initial ductile-brittle transition temperature of materials used in the reactor vessel shall be known by reference or established empirically; (2) expected shifts in transition temperature during design service life due to environmental conditions, such as neutron flux, shall be determined and employed in the reactor vessel design; (3) operation margins to be observed with regard to the transition temperature shall be designated for each mode of operation. Recommended Change No Change.

1 VYNPS License Renewal Project TLAA and Exemption Evaluation LRPD-03 Revision 0 Page 62 of 98 UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section 4.2.4.1 Reactor Vessel lst paragraph FSAR Text The reactor vessel is a welded vertical cylindrical pressure vessel with hemispherical heads. The reactor vessel is designed and fabricated for a useful life of 40 years based upon the specified design and operating conditions. The vessel is designed, fabricated, inspected, tested, and stamped in accordance with the ASME Boiler and Pressure Vessel Code, Section Ill, its interpretations, and applicable requirements for Class A vessels as defined therein.

The reactor vessel and its supports are designed in accordance with the loading criteria of Appendix C, "Structural Loading Criteria." The materials used in the design and fabrication of the reactor pressure vessel are shown in Table 4.2.1. Although little corrosion of plain carbon or low alloy steels occurs at temperatures of 500°F to 600"F, higher corrosion rates occur at temperatures around 140°F. The 0.1 25-inch minimum thickness stainless steel cladding provides the necessary corrosion resistance during reactor shutdown and also helps maintain water clarity during refueling operations. Exterior exposed ferritic surfaces of pressure- containing parts have a minimum corrosion allowance of 1/32-inch. All carbon and low alloy steel nozzles exposed to the reactor coolant have a corrosion allowance of I/l&inch. Recommended Change The reactor vessel is a welded vertical cylindrical pressure vessel with hemispherical heads. The reactor vessel was originally isdesigned and fabricated for a useful life of 40 years based upon the specified design and operating conditions.

The vessel is acceptable for 60 years of operation.

The vessel is designed, fabricated, inspected, tested, and stamped in accordance with the ASME Boiler and Pressure Vessel Code, Section 111, its interpretations, and applicable requirements for Class A vessels as defined therein.

The reactor vessel and its supports are designed in accordance with the loading criteria of Appendix C, "Structural Loading Criteria." The materials used in the design and fabrication of the reactor pressure vessel are shown in Table 4.2.1. No change required to this paragraph.

WNPS License Renewal Project TLAA and Exemption Evaluation LRPD-03 Revision 0 Page 63 of 98 4.2.4.1 Reactor Vessel 5m paragraph FSAR Section FSAR Text Another way of minimizing the NDT temperature is by reducing the integrated neutron exposure at the inner surface of the reactor vessel. The coolant annulus between the vessel and core shroud and the core location in the vessel limit the integrated neutron exposure of reactor vessel material to less than 1 x IO1' nvt from neutrons with energy levels greater than 1 MeV, within the 40-year design lifetime of the vessel. This is not the expected exposure, nor is it the absolute limit of safe exposure; it is an exposure value that can be demonstrated to be safe and is Practical to maintain. The estimated exposure for the 40-year life is less than 2.3 x IO" nvt for neutron energies greater than 1 MeV at the vessel inner surface. (Reference 17). Recommended Change Another way of minimizing the NDT temperature is by reducing the integrated neutron exposure at the inner surface of the reactor vessel. The coolant annulus between the vessel and core shroud and the core location in the vessel limit the integrated neutron exposure of reactor vessel material to less than 1 x IO" nvt from neutrons with energy levels greater than 1 MeV, within the original 40-year design lifetime of the vessel. This is not the expected exposure, nor is it the absolute limit of safe exposure; it is an exposure value that can be demonstrated to be safe and is practical to maintain.

The estimated ex osure for the 6048-year life is less than 5.423 x 10' nvt for neutron energies greater than 1 MeV at the vessel inner P surface.-

..

WNPS License Renewal Project TLAA and Exemption Evaluation LRPD-03 Revision 0 Page 64 of 98 ~ ~~ Recommended Change FSAR Section 4.2.4.9 Reactor Vessel Insula tion ~ ~ The reactor vessel insulation has an average maximum heat transfer rate of approximately 80 Btulhr-ft2 at the operating conditions of 550°F for the vessel and 135°F for the outside air. The maximum insulation thicknesses are 4 inches for the upper head, 3-1/2 inches for the cylindrical shell and nozzles, and 3 inches for the bottom head. The upper head insulation is designed to permit complete submersion in water during shutdown without loss of insulating material, contamination of the water, or adverse effect on the insulation efficiency of the insulation assembly after draining and drying. The lower head and cylindrical shell insulation is permanently installed.

k&be The insulation panels for the cylindrical shell of the vessel are held in place by vessel insulation supports located at two elevations on the vessel. The support brackets for each support are full-penetration welded to the vessel at 12 evenly spaced locations around the circumference

.. FSAR Text The reactor vessel insulation has an average maximum heat transfer rate of approximately 80 Btu/hr-fQ at the operating conditions of 550°F for the vessel and 135°F for the outside air. The maximum insulation thicknesses are 4 inches for the upper head, 3-1/2 inches for the cylindrical shell and nozzles, and 3 inches for the bottom head. The upper head insulation is designed to permit complete submersion in water during shutdown without loss of insulating material, contamination of the water, or adverse effect on the insulation efficiency of the insulation assembly after draining and drying. The lower head and cylindrical shell insulation is permanently installed for the 40-year design life of the vessel. The insulation panels for the cylindrical shell of the vessel are held in place by vessel insulation supports located at two elevations on the vessel. The support brackets for each support are full-penetration welded to the vessel at 12 evenly spaced locations around the circumference.

WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results LRPD-03 Revision 0 Page 65 of 98 UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section 4.2.5 Safety Evaluation 3 paragraph

4.2.5 Safety

Evaluation ~~~ ~~ ~~ FSAR Text The reactor vessel is designed for a 40-ear life and will not be exposed to more than 1 x 10 nvt of neutrons with energies exceeding 1 MeV. The reactor vessel is also designed for the transients which could occur during the 40-year life. IX The design transients used in the original ASME Ill design of the Vermont Yankee reactor vessel are specified in Section 5 and Attachment C of Reference

3. Reference 4 provides an up-to-date list of design transients, vessel cyclic limits, references to current design specifications, and stress reports for reactor components. Following fabrication of the reactor vessel the NRC and nuclear power industry established updated methods to determine initial transition temperature and neutron shift. Fracture toughness requirements (Reference 16), material drop weight, and Charpy impact test results were used to determine a reference nil-ductility temperature (RTndt) for all pressure boundary components.

The guidance of Regulatory Guide 1.99, Revision 2 (Reference

5) was employed to conservatively establish adjusted RTndt (ARTndt) for the plates and welds adjacent to the core. Recommended Change The reactor vessel was originally isdesigned for a 40- year life and wihvould not be exposed to more than 1 x 10 nvt of neutrons with energies exceeding 1 MeV. The reactor vessel was isalso designed for the transients which could occur during the 40-year life.

Vessel operation up to 60 years was reviewed and the maximum fluence to the vcssel inner wall is 5.39 x 1 017n/~rn2, still well belaw the original design value. No change. No change.

VYNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 66 of 98 FSAR Section 4.2.6 Inspection and Testing 4.3.4 (RR) Desciption

4.3.4 Description

page 4.3-7 FSAR Text Vermont Yankee's approach is to monitor startup/shutdown and feedwater onloff flow cycles and perform UT exams on a frequency that will assure potential crack growth is smaller in relation to ASME XI limits. ~~ Since the removal of Reactor Recirculation System valve internals requires unloading of the nuclear fuel, the valves are provided with high quality back seats and trim to facilitate stem packing renewal without draining the vessel and to provide adequate leak- tightness during normal operation. The design objective of the back seats and trim is to provide a minimum 20-year service life. ~ ~ ~ The pump casing is designed in accordance with the ASME Boiler and Pressure Vessel Code, Section Ill, Class C, as far as this code can be applied. This class is used because the pump casing does not experience temperature transients as severe as those that portions of the reactor vessel and certain piping connections experience; therefore, it is not necessary to make the cyclic analysis required for Class A equipment.

The design objective for the recirculation pump casing is a useful life of 40 years, accounting for corrosion, erosion, and material fatigue.

The pump drive motor, impeller, wear rings, and seals are designed for as long a life as is practical. Recommended Change No change. No change. Note this is a "short lived" item. e No change. The original design objective for the recirculation pump casing was isa useful life of 40 years, accounting for corrosion, erosion, and material fatigue. The pump drive motor, impeller, wear rings, and seals are designed for as long a life as is practical.

The pump casing was reviewed for license renewal loss of material due to corrosion, erosion, and cracking due to material fatigue are managed for the period of extended operation.

WNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-U3 Revision 0 Page 67 of 98 FSAR Section 4.6.3 Description (MSIV) 24'h paragraph, page 4.6-7 4.6.3 Description (MSIVs) 4.6.4 Test Program 4.1 0.3.2 Unidentified Leakage Rage FSAR Text The isolation valve is designed to pass saturated steam at 1250 psig and 575°F with a moisture content of approximately 0.23%, oxygen content of 30 ppm, and a hydrogen content of 4 ppm. The design objective for the valve is a minimum of 40 years service at the specified operating conditions.

The estimated operating cycles per year is 100 cycles during the first year and 50 cycles per year thereafter. In addition to minimum wall thickness required by applicable codes, a corrosion allowance of 0.120 inches minimum is added to provide for 40 years services.

~~ Design specification ambient operating conditions are 135°F normal, 150°F maximum, at 100% relative humidity, in a radiation field of 15 Whr gamma and 25 Rad/hr neutron plus gamma, continuous for design life.

The inboard valves are not exposed to these maximum conditions continuously, and the outboard valves are in much less severe ambient conditions.

~ A full-size, 20 inch valve produced for actual use in a BWR was tested in a range of steamlwater blowdown conditions simulating postulated accident conditions.

The test valve was opened and closed more than 400 times (200 cycles) during the test program, . . . An analysis was undertaken to estimate, on the basis of information available when the plant was built from the Pipe Study and other sources, the probability of a line break occurring in a reactor piping system as a result of progressive crack growth. Recommended Change The isolation valve is designed to pass saturated steam at 1250 psig and 575°F with a moisture content of approximately 0.23%, oxygen content of 30 ppm, and a hydrogen content of 4 ppm. The design objective for the valve was isa minimum of 40 years service. The estimated operating cycles per year is 100 cycles during the first year and 50 cycles per year thereafter.

In addition to minimum wall thickness required by applicable codes, a corrosion allowance of 0.120 inches minimum is added to provide for 40 years services.

AllowaSle operating cycles and corrosian allowance were reviewed and remain valid for the period of extended operation (60-years).

No change. No change. No change.

LRPD-03 Revision 0 Pane 68 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluation FSAR Text Attachment 3 - VYNPS TLAA Search Results Recommended Change UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section 4.10.3.2 Unidentified Leakage Rate 4.10.3.3 Total Leakage Rate 5.2.3.7 Primary Containment Normal Heating, Ventilation, and Air Conditioning Systems Leaks X No. of components Component- Year X Design life X Probability of break for given leak rate Results of this study (see Appendix "I", Figure 1-3) demonstrate the importance of adequate leak detection capability in maintaining a high piping system reliability.

A relatively higher risk of a steam line break, as compared to a waterline break, is also indicated, as a result of the lower leak rate and, hence, more difficult detection of a steam line crack of a given length.

Suggested Change: No change is required to this description.

The actual calculation in Appendix I is changed below. A flow recorder continually plots time versus discharge flow rate from each sump; an increase in leakage rate is detectable by an increase in sump discharge flow time and an increased frequency in discharge flow cycles. \ ~ ~~ Maintaining the drywell ambient temperature in the range of 135°F to 165°F except for the upper drywell regions during normal plant operation assures that the insulation on motors, isolation valves, operators and sensors, instrument cable, electrical cable and gasket materials or sealants used at the penetrations have a sustained life without deterioration.

No change. No change.

LRPD-03 Revision 0 Page 69 of 98 WNPS License Renewal Project TLAA and Exemption Evaluation b Attachment 3 - WNPS TLAA Search Results FSAR Section FSAR Text Recommended Change I I 6.4.1 High Pressure Coolant Injection System 14'h paragraph, page 6.4-5 The system is designed for a service life of 40 years, accounting for corrosion, erosion, and material fatigue. The system was isdesigned for an original service life of 40 years, accounting for corrosion, erosion, and material fatigue. The HPCl system was reviewed for license renewal, and Corrosion, erosion, and material fatigue were evaluated for 60 years. 6.5.2.1. LOCA Analysis Methods and Results 6.6 Inspection and Testing 7.2.3.1 0 Wiring LOCA analysis methods developed by General Electric Company (Reference

2) which conform to 10CFR50.46 requirements were used to demonstrate 1 OCFR50.46 conformance for the Vermont Yankee Nuclear Power Station (WNPS) for the first 16 cycles of plant operation.

The portions of the Core Standby Cooling Systems requiring pressure integrity are designed to specifications for inservice inspection to detect defects which might affect the cooling performance.

The reactor vessel nozzles and the core spray and feedwater spargers receive particular attention. Records are kept of the number of design basis thermal cycles these components receive. Wiring and cables for Reactor Protection System instrumentation are selected to avoid excessive deterioration due to temperature and humidity during the design life of the plant. No change. No change. No change.

WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results LRPD-03 Paae 70 nf 98 Revision 0 UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section 7.3.4.9 Environmental Capabilities 7.5.6.2.3 Physical Arrangement FSAR Text Verification that the isolation equipment has been designed, built, and installed in conformance to the specified criteria is accomplished through quality control and performance tests in the vendor's shop or after installation at the plant before startup, during startup, and thereafter during the service life of the equipment.

Each LPRM detector assembly contains four miniature fission chambers with an associated solid sheath cable. Each fission chamber produces a current which when coupled with the LPRM signal-conditioning equipment, provides the desired scale deflection throughout the design lifetime of the chamber. Each individual chamber of the assembly is a moisture-proof, pressure-sealed unit. Each assembly also contains a calibration tube for a Traversing lncore Probe (TIP).

The enclosing tube around the entire assembly contains holes along its length. These holes allow circulation of the reactor coolant water to cool the fission chambers.

Numerous tests have been performed on the chamber assemblies including tests of linearity, lifetime, gamma sensitivity, and cable effects.(l)

These tests and experience in operating reactors, including Vermont Yankee, provide confidence in the ability of the LPRM Subsystem to monitor neutron flux to the design accuracy throughout the design lifetime. Recommended Change ~ ~ No change. No change. NOTE: AMRM-32, Section 2.0, states "The LPRM have limited lifetimes and are replaced as determined by OP-4407 based on calibration current measurements.

The TIP guide tubes inside the LPRM are an integral part of the detectors and are also replaced when a detector is replaced. As short lived components, neither the detectors nor the TIP guide tubes are subject to aging management review."

WNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 71 of 98 FSAR Section 7.5.9.3 Power Generation Evaluation 9.2.4.2 Low Purity Wastes 10.1 3.1 Power Generation Objectives 14.5.4.1 Pressure Regulator Failure 14.5.6.1 Recirculation Pump Seizure FSAR Text An adequate number of TIP machines is supplied to assure that each instrument location assembly can be probed by a TIP and the central one can be probed by every TIP to allow intercalibration. The system has been field tested in an operating reactor to assure reproducibility for repetitive measurements, and the mechanical equipment has undergone life testing under simulated operating conditions to assure that all specifications can be met. For the purpose of analyzing future radiological impacts during the plant's life, it is assumed that 1 % of the combined processed stream treated each year would be discharged from the station. ~~ ~~ ~~ The power generation objective of the Station Makeup Water System is to maintain a supply of treated water that may be used as makeup for the station and reactor cycles. An analysis of the impact of thermal stress from this event on the RPV fatigue life has been made (NEDO- 22243-1, "Safety Evaluation of MSlV Low Turbine Inlet Pressure Isolation Setpoint Change for Vermont Yankee Nuclear Power Station," May 1983).

The analysis concluded that the additional usage factor associated with the transient is insignificant. For Cycle 22 and future cycles, Vermont Yankee is expected to have rated power Operating Limit MCPRs at least 0.20 higher than the Safety Limit MCPR. Recommended Change No change. No change. No change. ~~ No change. No change.

Page 72 of 98 TLAA and Exemption Evaluation I .- . . WNPS Licer - WNPS License Renewal Project TLAA and Exemption Evaluation LRPD-03 Revision 0 Page 72 of 98 Attachment 3 - WNPS TLAA Search Results .I UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section A.9.2.1 General B.2.2.4 Quality Assurance and Inspection of the Reactor Primary System B.2.2.4 Quality Assurance and Inspection of the Reactor Primary System B.2.3.2 Addition of Cooling Tower Complex B 4.10 Resolution

~ FSAR Text Pipe support design methodology has been consistent throughout the life of the plant and is described in Section 12.2 and Section C.3.9 of Appendix C of this FSAR. "The Committee continues to emphasize the importance of quality assurance in fabrication of the primary system and of inspection during service life. Provisions are being made to the maximum extent considered feasible for inspection of primary system components during service life, consistent with the . requirements of "Draft ASME Code for lnservice Inspection of Nuclear Reactor Coolant Systems." "At the time of the previous review by the Committee, the applicant planned to use the Connecticut River as a heat sink by drawing cooling water for the main condenser from the river, heating it in the condenser, and returning the heated water to the river. Since that time, limitation by state agencies on the allowable temperature rise and maximum temperature of water returned to the river has led the applicant to propose the use of cooling towers to reject a portion of the waste heat from the plant to the atmosphere." Where deflection is not the limiting factor, the ASME Boiler and Pressure Vessel Code, Section Ill, was used as a guide to determine limiting stress intensities and cyclic loadings for the core internal structure.

____________~~~

~ Recommended Change No change. No change. No change. No change. Does not involve a TLAA. No change.

LRPD-03 Revision 0 Page 73 of 98 WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results UPDATED FINAL SAFETY ANALYSIS REPORT ~~~ ~~ FSAR Section B.5.3 Fuel Orientation C.2.2 Loading Conditions and Allowable Limits FSAR Text ~~ Experience has shown that the distinguishing features will be visible during the design lifetime of the fuel. In all cases, fueling procedures require that the fuel assembly number be verified. Certain of the limits described in these criteria, Le., deformation limit and fatigue limit, are included for completeness, but do not necessarily require application to all components.

Where it is clear that fatigue or excess deformation are not of concern for a particular structure or component, a formal analysis with respect to that limit is not required. Recommended Change No change. Fuel lifetime isn't changing.

No change. NOTE: Appendix C is identified as "historical" and not maintained current in Section C.l .I. All changes for Appendix C may be ignored and the "historical" record maintained.

WNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 74 of 98 FSAR Section FSAR Text C.2.2.2 Allowable Limits (definition of allowable limits) Recommended Change Generic Definition (Current FSAR) Emergency (low probability) Faulted (extremely low probability) (sixth paragraph)

SFrnin is related to the event probability by the following equation:

&0 = 40 year event encounter Drobability Upset (likely) 1.0 > ~40> 10" i 0-l > ~40 > 1 o-~ 10" > P40 > IO-^ 9 SF min = ....................... (Equation A) 3 - log10 P40 where: IO-' > P40 > IO5 (Equation A applies) IO-' > P40 > 1 .O > P40 > IO-' (SFmin = 2.25) (SFrnin = 1.125) (Page C.2-6 of 65, first paragraph) These expressions show the probabilistic significance of the classical safety factor concept as applied to reactor safety. The SFmin values corresponding to the current governing accident event probabilities are summarized as follows: Item Governing Loading Conditions P40 SFrnin (Last paragraph of section C.2.2) The minimum safety factor decreases as the event probability diminishes and if the event is too improbable (incredible:

P40 < then no safety factor is appropriate or required.

WNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 75 of 98 FSAR Section FSAR Text Recommended Change (Page C.2-6 of 65, first paragraph)

These expressions show the probabilistic significance of the classical safety factor concept as applied to reactor safety. The SFmin values corresponding to the current governing accident event probabilities are summarized as follows: Item Governing Loading Conditions PtjQ4 SFmin (Last paragraph of section C.2.2) The minimum safety factor decreases as the event probability diminishes and if the event is too improbable (incredible:

P6(m <: then no safety factor is appropriate or required.

VYNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 76 of 98 FSAR Section C.2.4.1 Criteria (2.2.4.2 Vessel Fatigue Analysis C.2.5.3 Fatigue Analysis FSAR Text Stress analysis requirements and load combinations for the reactor vessel have been evaluated for the cyclic conditions expected throughout the 40 year life, with the conclusion that ASME code limits are satisfied.

The vessel design report contains the results of the detailed design stress analyses performed for the reactor vessel to meet the code requirements. Selected components, considered to possibly have higher than code design primary stresses as a result of rare events or a combination of rare events, have been analyzed in accordance with the requirements of the loading criteria in this appendix. Results of the most critical of those analyses are included in a following section. The conclusion is that the limits in the criteria have been met. An analysis of the reactor vessel shows that all components are adequate for cyclic operation by the rules of Section Ill of the ASME Code. Exhibit 4 of the Reactor Pressure Vessel Design Report gives a summary of the analysis.

The analysis indicates that for the more critical components on the vessel that the primary plus secondary stress intensity range is less than 3 SM and that a plastic analysis is not required.

Also, the usage factors for the conservatively specified operating cycles is substantially less than the code allowed 1 .O. A fatigue analysis was performed using as a guide the Recommended Change Stress analysis requirements and load combinations for the reactor vessel were originally lxwebe~ evaluated for the cyclic conditions expected throughout the a 40 year life, with the conclusion that ASME code limits are satisfied.

The vessel is acceptable for 60 years of operation The vessel design report contains the results of the detailed design stress analyses performed for the reactor vessel to meet the code requirements. Selected components, considered to possibly have higher than code design primary stresses as a result of rare events or a combination of rare events, have been analyzed in accordance with the requirements of the loading criteria in this appendix. Results of the most critical of those analyses are included in a following section. The conclusion is that the limits in the criteria have been met.

No change. No change. Appendix C is identified as "historical" in WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results LRPD-03 Revision 0 Page 77 of 98 UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section FSAR Text 4SME Boiler and Pressure Vessel Code, Section 111. The method of analysis used to determine the cumulative fatigue usage as described in APED 5460, "Design and Performance of GE-BWR Jet Pumps," September 1968. The most significant fatigue loading occurs in the jet pump - shroud - shroud support area of the internals.

The analysis was performed for the Dresden plant, where the configuration (stilt type shroud support) was similar to the Vermont Yankee plant. Therefore, the calculated fatigue usage is expected to be a good approximation for this plant. Loading Combinations and Transients Considered

1. Normal Startup and Shutdown Vessel fluid temperature goes from 70°F to 545°F at 1 OO"F/hr rate. 120 cycles. 2. Operating Basis and Design Basis Earthquake Considered using a conservative axi-symmetric load. Resulting stress negligible.

3. Ten-Minute Blowdown Vessel fluid initially at 545°F. Stuck-open relief valve causes fluid temperature to drop to 370°F in ten minutes. 1 cycle. 4. HPCl Operation Produced by loss of feedwater pumps. Fluid Recommended Change Section C.l .I; it should be left as historical and not updated.

WNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 78 of 98 I FSAR Section Recirculation Loop Piping, pages C.2-32 and C.2-33 FSAR Text temperature between vessel and shroud drops to 300°F in about five minutes. Vessel lower plenum may reach 100°F. 30 cycles. 5. LPCl Operation (DBA) Vessel flooded with 140°F cooling water subsequent to complete vessel blowdown in 30 seconds. 1 cycle. 6. Improper Start of Recirculation Loop 130°F water flows in reverse through one recirculation outlet nozzle for 55 seconds. Bulk water temperature between vessel and shroud steps from 545°F to 480°F and returns to 545°F. 1 cycle. Cumulative Fatigue Usage Umax = 0.33 (Uallowable

= 1 .O) Remarks The location of maximum fatigue usage is on the bottom side of the baffle plate at the point where the baffle plate attaches to the shroud in the vicinity of the minimum ligament.

Statement of Criteria B. For load combinations that have a very low probability of occurrence, maintain primary stresses below the following limit: Recommended Change I Statement of Criteria B. For load combinations that have a very low probability of occurrence, maintain primary stresses below the following limit:

WNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT L RP D -0 3 Revision 0 Page 79 of 98 FSAR Section FSAR Text 225 times 831 .I .O allowable stresses, where SF SF = 9 3 - log 10 P 40 and Pdo = Probability of load combination occurrence in 40- year plant life. Method of Analvsis B. Effects from the following loading combinations determined in accordance with rules of B31 .I .O: 1. The sum of the longitudinal stresses due to pressure, dead weight, and inertia effects of maximum hypothetical earthquake must be less than 1.8 times the hot allowable stress. The probability of this lofd occurrence during the 40- year plant life is IO- and SF = 1.5. 2. The sum of the longitudinal stresses due to maximum pressure, dead weight and inertia effects of design basis earthquake must be less than

1.5 times

the hot allowable stress. The probability of this load occurring during the 40-year plant life is IO2 and SF = 1.8. 3. The sum of the longitudinal stresses due to maximum pressure, dead weight and inertia effects of maximum hypothetical earthquake must be less Recommended Change 225 times B31 .I .O allowable stresses, where SF SF = 9 and 3 - log 10 P@$Q PeON = Probability of load combination occurrence in 6048-year plant life.

Method of Analvsis B. Effects from the following loading combinations determined in accordance with rules of B31 .I .O: 1. The sum of the longitudinal stresses due to pressure, dead weight, and inertia effects of maximum hypothetical earthquake must be less than 1.8 times the hot allowable stress.

The probability of this load occurrence during the 6044)-year plant life is and SF = 1.5. 2. The sum of the longitudinal stresses due to maximum pressure, dead weight and inertia effects of design basis earthquake must be less than 1.5 times the hot allowable stress. The probability of this load occurring during the 6048- year plant life is 1 O-' and SF = 1.8. 3. The sum of the longitudinal stresses due to maximum pressure, dead weight and inertia effects of maximum hypothetical earthquake must LRPD-03 Revision 0 Page 80 of 98 WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results UPDATED FINAL SAFETY ANALYSIS REPORT

~~ -~ ~ FSAR Section Main Steam Piping pages C.2-35 and C.2-36 than 2.0 times the hot allowable stress. The probability of this load combination occurring during the 40 year plant life is 25 x and SF = 1.36. Statement of Criteria

6. For load combinations that have a very low probability of occurrence, maintain primary stresses below the following limit: - 225 times 831 .I .O allowable stresses, where SF SF = 9 and 3 - log 10 P 40 P4,, = Probability of load combination occurrence in 40- year plant life.

Method of Analysis B. Effects from the following loading combinations determined in accordance with rules of I331 .I .O: 1. The sum of the longitudinal stresses due to pressure, dead weight, and inertia effects of maximum hypothetical earthquake must be less than 1.8 times the hot allowable stress.

The probability of this load occurrence during the 40- year plant life is and SF = 1.5. Recommended Change be less than 2.0 times the hot allowable stress. The probability of this load combination occurring during the 6048 year plant life is -25 x IO" and SF = 1.36. Statement of Criteria B. For load combinations that have a very low probability of occurrence, maintain primary stresses below the following limit:

225 times B31 .I .O allowable stresses, where SF SF = 9 and 3 - log 10 PG()4Q P604g = Probability of load combination occurrence in 6048-year plant life.

Method of Analysis 9. Effects from the following loading combinations determined in accordance with rules of 831 .I .O: 1. The sum of the longitudinal stresses due to pressure, dead weight, and inertia effects of maximum hypothetical earthquake must be less than 1.8 times the hot allowable stress. The probability of this load occurrence during the 6048-year plant life is IO" and SF = I .5.

WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results LRPD-03 Revision 0 Page 81 of 98 UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section RClC and HPCl pump casings page C.2-49 RClC and HPCl turbines page C.2-52 Table C.2.2 Primary Stress Limit FSAR Text 2. The sum of the longitudinal stresses due to maximum pressure, dead weight and inertia effects of design basis earthquake must be less than

1.5 times

the hot allowable stress. The probability of this load occurring during the 40-year plant life is lom2 and SF = 1.8. 3. The sum of the longitudinal stresses due to maximum pressure, dead weight and inertia effects.

of maximum hypothetical earthquake must be less than 2.0 times the hot allowable stress. The probability of this load combination occurring during the 40 year plant life is .25 x IO3 and SF = 1.36. 2. The minimum wall thickness of the pump shall be based on that to limit stress to the allowable working stress when subjected to design pressure plus corrosion allowance.

2. The minimum wall thickness of the turbine casing shall be based on that to limit stress to the allowable working stress when subjected to design pressure plus corrosion allowance. The methods of linear elastic stress analysis may be used in the fracture analysis where its use is clearly conservative or supported by experimental evidence.

Examples where "fracture mechanics" may be applied are for fillet welds or end of fatigue life crack propagation. Recommended Change ~ ~ 2. The sum of the longitudinal stresses due to maximum pressure, dead weight and inertia effects of design basis earthquake must be less than 1.5 times the hot allowable stress. The probability of this load occurring during the 6040- year plant life is IO-* and SF = 1.8. 3. The sum of the longitudinal stresses due to maximum pressure, dead weight and inertia effects of maximum hypothetical earthquake must be less than 2.0 times the hot allowable stress.

The probability of this load combination oFcurring during the 6048 year plant life is .25 x 10- and SF = 1.36. No change. No change. No change.

WNPS License Renewal Project TLAA and Exemption Evaluation UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 82 of 98 FSAR Section FSAR Text TABLE C.2.4 Fatigue Limit Recommended Change F.l Summary Description Vermont Yankee has made changes to the facility over the life of the plant that may have invoked the final General Design Criteria as design criteria. F.2.1 Group I Overall Plant Requirements No change. Complete records of the as-built design of the station, changes during operation and quality assurance records will be maintained throughout the life of the station. average temperature of interest.

No change. (2) It is acceptable to use the ASME Section Ill Design Fatigue Curves in conjunction with a cumulative usage factor of 1 .O (using Miner's hypothesis) in lieu of using the mean fatigue data curves with a limit on fatigue usage of 0.05, since the two methods are approximately equivalent.

LRPD-03 Revision 0 Page 83 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - VYNPS TLAA Search Results UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section H. 1 Summary Description H3.3 Recirculation Flow

1.1 Probability

of Leaking Failures FSAR Text It is the purpose of Section H.2 to review initial reactor core design criteria and, by presentation of analytical data, show the existence of adequate thermal margins. The thermal operating limits for the Vermont Yankee Nuclear Power Station are evaluated for each cycle of operation.

These are presented in the current cycles Core Performance Analysis Report and in the Technical Specifications.

However, this accuracy was only predictable at the beginning of life, in the recently calibrated condition, and with some subjective engineering estimates introduced into some of the component uncertainty contributors.

One of the objectives of the Pipe Rupture Study is to predict, based on our knowledge of stress levels and crack propagation rates in a BWR piping system, the rate of occurrence of through-wall cracks during the life of the plant for each piping system and component. ~~ ~~ Recommended Change No change. No change. No change.

LRPD-03 Revision 0 Page 84 of 98 WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results UPDATED FINAL SAFETY ANALYSIS REPORT FSAR Section I FSARText Recommended Change No change. 1.2 Critical Crack Size Fracture data for a wider range of materials and temperatures may be obtained from the terminal fracture behavior of wide-plate crack growth test specimens, as recorded by Brothers.

In another task under the pipe study, E. Kiss9 has conducted reversed-bending fatigue tests of 6-inch pipe at room temperature, with internal pressure in some cases.

Stress intensity factors as high as 11 1 ksi inch have been recorded for circumferential through- wall cracks, with no crack instability occurring.

The data for elevated temperature falls in approximately the same range of KC as the room temperature data. These points represent terminal fracture following several thousand cycles of plastic strain.

VYNPS License Renewal Project TLAA and Exemption Evaluation I UPDATED FINAL SAFETY ANALYSIS REPORT LRPD-03 Revision 0 Page 85 of 98 I FSAR Section 1.4 PROBABILITY OF LINE BREAK FSAR Text I Recommended Change A BWR typically has about 250 piping components of size 4 inches or larger which are located between the vessel and the first shutoff valve. Of this total, 100 components are associated with steam

$nd 150 with water. If a detectable leak rate is 5 gpm, then (from Figure 1-3) a crack in a steamline has a 3.7 x IO probability leading to line break, and a crack in a waterline has 6 x probability of line break. The probability of a steam line break in a 40-year plant design life is: 5.3 x IO4 Leaks X Com ponent-Year X 100 components X 40 years = 7.8 x IO4 This is equivalent to a system reliability of 0.9992. x 3.7 x IO4 For waterlines, the probability of a break is:

or a reliability of 0.99998. 5.3 x 1 O4 x 1 50 x 40 x 6 x 1 0" = 1.9 x 1 O",

LRPD-03 Revision 0 Page 86 of 98 WNPS License Renewal Project TLAA and Exemption Evaluation Attachment 3 - WNPS TLAA Search Results FSAR Text UPDATED FINAL SAFETY ANALYSIS REPORT Recommended Change FSAR Section K.1.2.2 Structural and Design Evaluations K.3.1 Design Objectives For waterlines, the probability of a break is: 5.3 x IO4 x 150 x 6048 x 6 x or a reliability of 0.999978.

= 21.9 x lo5, Although the repair is not considered an ASME Boiler & Pressure Vessel (B&PV) Code repair, the repair satisfies the Design By Analysis stress and fatigue criteria of the ASME Boiler & Pressure Vessel Code, Section Ill, Subsection NG (Reference 5). The function of the core shroud repair is to structurally replace all potentially sensitized 304 stainless steel circumferential core shroud welds, i.e., HI through H7 (See Figure K.1-I).

In addition, the repair can accommodate a complete failure of the H8 shroud weld with the shroud support legs intact. The design life of the repair is 40 years. No change. No change required, the 40 year life of the repair is greater than the period of extended operation. Repair was made in 1995.

WNPS License Renewal Project TLAA and Exemption Evaluation I LRPD-03 Revision 0 Page 87 of 98 UPDATED FINAL SAFETY ANALYSIS REPORT K*4*3.1 Repair Hardware Structural Evaluation

~ FSAR Section The design by analysis stress and fatigue criteria of the ASME Boiler & Pressure Vessel Code, Section K.4.3.2 Flow Induced Vibration Ill, Subsection NG, are satisfied.

The maximum fatigue usage in the tie rod assembly due to OBE and thermal expansion (including open and shutdown) loads occurs in the threaded section of the spring rods. The fatigue usage at this location is less than 12%. The fatigue usage from shroud and flow induced vibration is negligible.

As discussed above, the evaluations show that stresses resulting from flow-induced vibration are small and pose no fatigue concern. Recommended Change No change. No change.

WNPS License Renewal Project TLAA and Exemption Evaluation LRPD-03 Revision 0 Page 88 of 98 Key word is in blue.

Temperature Limits Tech Spec Section Tech Spec Text B 3.6.A Pressure Temperature Limits Recommended Change All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.2 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

~ The guidance of Branch Technical Position - MTEB 5- No change till revised P/T curves are developed and 2, material drop weight, and Charpy impact test results were used to determine a reference nil-ductility temperature (RTNDT) for all pressure boundary components. For the plates and welds adjacent to the core, fast neutron (E > 1 MeV) irradiation will cause an increase in the RTNDT. For these plates and welds an adjusted RTNDT (ARTNDT) of 89°F and 73°F (% and 3/4 thickness locations) was conservatively used in development of these curves for core region components. Based upon plate and weld chemistry, initial RTNDT values, predicted peak fluence (2.3~10" nlcm2) for a gross power generation of 4.46~1 O8 MWH(t) (Battelle Columbus Laboratory Report BCL 585-84-3, dated May 15, 1984) these core region ARTN~T values conservatively bound the guidance of Regulatory Guide I .99, Revision 2. submitted.

WNPS License Renewal Project TLAA and Exemption Evaluation Technical Specifications LRPD-03 Revision 0 Page 89 of 98 Tech Spec Section 3.6. E. Structural Integrity and Operability Testing B 3.6. E. Structural Integrity and Operability Testing Tech Spec Text Due to convection cooling, stratification, and cool CRD flow, the bottom head area is subject to lower temperatures than the balance of the pressure vessel.

The RTND~ of the lower head is lower than the ARTNDT used for the beltline.

The lower head area is also not subject to the same high level of stress as the flange and feedwater nozzle regions. The dashed Bottom Head Curve is less restrictive than the enveloping curve used for the upper regions of the vessel and provides Operator's with a conservative, but less restrictive P/T limit for the cooler bottom head region. The actual shift in RTNDT of the critical plate and weld material in the core region will be established periodically during operation by removing and evaluating, in accordance with ASTM E185, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The structural integrity and the operability of the safety-related systems and components shall be maintained at the level required by the original acceptance standards throughout the life of the plant. Prior to operation, the reactor primary system was free of gross defects. In addition, the facility has been designed such that gross defects should not occur throughout plant life. Recommended Change No change required for the current P/T curves, re- evaluate when curves are developed for the period of extended operation.

OK as is. ~~ ~- Nochange. . No change.

WNPS License Renewal Project TLAA and Exemption Evaluation Technical Specifications Tech Spec Section B 4.7.A Primary Containment System LRPD-03 Revision 0 Page 90 of 98 B 4.7.D Primary Contianment Isolation Valves Correspondence Tech Spec Text Experience with this type of coating during plant operating cycles between 1972 and the present indicates that this inspection methodology and interval are adequate.

Content Since valve internals are designed for a 40-year lifetime, an inspection program which cycles through all valves in one-eighth of the design lifetime is extremely conservative.

The test closure time limit of five seconds for these main steam isolation valves provides sufficient margin to assure that cladding perforations are avoided and I OCFRI 00 limits are not exceeded. Recommended Change No change NOTE: Valve internals will still be designed for 40 years. Valve bodies will be in aging management programs to assure acceptability for 60 years. Therefore this doesn't change. No change. BVY 98-1 38 Implementation of ASME Code Case N560 Risk Informed Inspection of Class 1 piping. ERC 2000-029 Revision 3 to RPV Pressurenemperature Limits TLAA NONE No TLAA in this memo. Follow up to verify implementation.

If this includes small bore piping, a small bore piping program isn't needed. YES. RPV Pressurenemperature limits are addressed in this document.

Review this reference for possible input to RPV Pressurenemperature Limits section WNPS License Renewal Project TLAA and Exemption Evaluation ERC 2001-019 LRPD-03 Revision 0 Page 91 of 98 ERC 2002-035 ERC-2004-2005 DM 21 INF 92-007 INF 92-035 INF 93-020 INF 93-021 INF 98-045 SIL 0243 SIL 0318 Fatigue Pro study Fatigue Pro study, Phase 2 Update HELB for MELLA.

Design Memo #21: CRD Thermal Cycle design notes. FAC of Feedwater Piping Higher than predicted erosion/corrosion of RCPB in containment at BWRs. Thermal Fatigue Cracking of Feedwater Piping to Steam Generators Erosion/Corrosion Program Generic Comments Cavitation Erosion of letdown line orifices resulting in cracking of pipe welds Mitigation of SCC is Austenitic Stainless Steel small bore piping in BWRs. BWR Reactor Vessel Cyclic Duty Monitoring NONE. Letter is to a vendor about study being conducted.

NONE. Letter is to a vendor about study being conducted.

NONE. (Hit because it was signed by B. Slifer. YES. Cycle management is covered in the Metal Fatigue TLAA. Review this reference for input to material fatigue TLAA. NONE Information Notice to PWRs only. NONE Information only, no response required.

NONE Information only, PWRs only.

NONE Information only, no response required.

NONE Information only.

NONE Information only. YES Cyclic duty monitoring is covered in the Metal Fatigue TW. Review this reference for input to metal fatigue TLAA.

WNPS License Renewal Project TLAA and Exemption Evaluation LRPD-03 Revision 0 Page 92 of 98 WNPS QA Program (QAPM)

SILO409 Rev2 lncore Dry Tube Cracks have been found. SIL 0426 SIL 0638-01 Automatic Depressurization System Cycling Cracking of control blades.

I QAPM Section NONE. Cracks have not been found and no pressure boundary leakage has resulted.

No inspection of tubes is required till they reach their 20 year life (not 40 year life). Dry tubes are subject to aging management review in AMRM-32. NONE Applicable to BWR-5/6 only. NONE. Control blades are short lived. Table 1 K.5 ANSI N45.2.5 Section 4.9 FPCRM Section Table 1, N.ll ANSI N45.2.12 Section 4.5.1 FPCRM text TLAA I QAPM text The words "splicing crew" are interpreted to refer to all project members that are actively engaged in preparing and assembling cadweld mechanical splices at the final splice location. Separate test cycles will be established for each bar size and each splice position.

The QAPM Section A.6 corrective action program may be used instead of these requirements as long as the appropriate time limits are applied to significant conditions adverse to quality. Recommended change None - Not a TLAA None - Not a TLAA I WNPS Fire Protection Commitment Reference Manual I WNPS License Renewal Project TLAA and Exemption Evaluation Page29, (h) LRPD-03 Revision 0 Page 93 of 98 Page 69, H. PP 7011 Section No hits Page A-25, Appendix A, (h) PP 7011 text TLAA NA NA Page B-23, Appendix B, 4.4 Service or operating life should be a minimum of one half hour for the self-contained units.

Control Room personnel may be furnished breathing air by a manifold system piped from a storage reservoir, if practical.

Service or rated operating life shall be a minimum of one-half hour for the self-contained units. Service or operating life should be a minimum of one half-hour for the self-contained units. The licensee has proposed to provide a recharging capability or new apparatus which has a greater service life to insure a supply of emergency breathing air for a period of six hours.

NONE. Short lived component.

___~ ~ NONE. Short lived component.

~~ NONE. Short lived component.

NONE. Short lived component.

I WNPS Fire Protection and Appendix R Program I

LRPD-03 Revision 0 Page 94 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations Estimate of outages for VYNPS: Beginning of life = End of extended operation

= Current outage frequency

= Refuel outages Outages to date Best estimate 24 Conservative:

24 Attachment 4 - Estimate of Crane Cycles 1972 2032 Actual date is 3/21/2012 Actual date is 12/1/72 per PP7015 1.5 years Till EOL (2032) Total 19 29 43 53 Assumes 18 month cycles for rest of plant life.

Assumes one outage per year for the rest of plant life. Reactor building crane:

Main hook Auxhook RB bridge crane rated load (tons): 110 7.73 Main hook rating from NRC SER (NVY 84-139) still current per OP 2200, Appendix B Aux hook rating from MM2003-040 and OP 2200, Appendix B Loads and weights taken from FW 81-134, updated per OP 2200, Appendix B Results for WNPS RB crane: Best estimate 3007 0.40 Conservative:

7590 0.40 Cycles K LRPD-03 Revision 0 Page 95 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations Heavy loads Reactor vessel head Drywell head Steam dryer Steam separator (shroud head) Shield blocks New fuel storage vault plugs Spent fuel pool Attachment 4 - Estimate of Crane Cycles # Lifts per # Weight outage 1 54.00 2 1 44.00 2 1 22.00 2 1 33.00 2 2 64.00 4 2 67.00 4 2 71.50 4 3 3.00 6 1 0.60 2 # Lifts in 60 Load magnitude Load years 86 W probability 0.491 0.029 86 86 0.400 0.029 0.200 0.029 Mean effective load factor K 0.0034 0.0018 0.0002 0.0008 0.01 13 0.0129 0.0157 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 86 172 1 72 172 258 86 86 258 86 86 43 43 172 172 0.0143 0.0005 0.0000 0.0000 0.300 0.029 0.582 0.057 0.609 0.057 0.650 0.057 0.027 0.086 0.005 0.029 0.004 0.029 0.055 0.086 0.066 0.029 0.041 0.029 1 .ooo 0.014 0.31 8 0.014 0.073 0.057 0.023 0.057 Notes: Off and on each outage. gate Refueling slot plugs Vessel head insulation Spent fuel shipping cask Filter- demineralizer hatch Contaminated equipment storage area hatches Off and on each outage. Off and on each outage. Off and on each outage. 1 0.45 2 3 6.00 6 1 7.25 2 1 4.50 2 1 110.00 1 I 35.00 1 2 8.00 4 2 2.50 4 Off and on each outage. Off and on each outage. Off and on each outage. Off and on each outage. Off and on each outage. Off and on each outage.

Off and on each outage.

Empty in once each outage, full out once each outage. Empty in once each outage. Off and on each outage. Off and on each outage.

LRPD-03 Revision 0 Page 96 of 98 VYNPS License Renewal Project TLAA and Exemption Evaluations M N 0 P Q Attachment 4 - Estimate of Crane Cycles Heavy loads

1. Reactor head 1 strongback Stud tensioner 1 monorail Cattle chute 1 Dryerheparator 1 storage pit shield Plug Crane load block 1 Best estimate effective robabilit factor K Notes: # Lifts per # Lifts Load in 60 magnitude Weight 4.00 outage years W 2 86 0.036 I I I Best estimate total cycles: 3007 Best estimate mean effective load factor K: 0.4004 3.50 2 14.00 2 43.50 2 86 0.032 86 0.127 86 0.395 0.029 0.029 0.086 0.043 Off and on each outage.

0.0001 0.0018 Off and on each outage. Off and on each outage. 0.0014 0.0000 Off and on each outage. Assumed 3 empty cycles of 28.00 6.00 6 258 0.255 3 129 0.055 " '-1 No longer used, 40 cycles. 0.029 I 0.0000 main hook each outage. In and out each outaae. 2.50 5.00 2 86 0.023 0 40 0.045 I LRPD-03 .- . , WNPS Licen! - TI ~- LRPD-03 Revision 0 Page 97 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations I rayr SI VI uw Conservative estimate Mea,, # Lifts Load effective Heavy Loads

1. Weight outage years W probability factor K a Reactor vessel 1 54.00 4 212 0.491 0.028 0.0033 head b Drywell head 1 44.00 4 212 0.400 0.028 0.0018 c Steamdryer 1 22.00 4 212 0.200 0.028 0.0002 d Steam separator 1 33.00 4 212 0.300 0.028 0.0008 e Shield blocks 2 64.00 8 424 0.582 0.056 0.01 IO 2 67.00 8 424 0.609 0.056 0.0126 2 71.50 8 424 0.650 0.056 0.0153 f New fuel storage 3 3.00 12 636 0.027 0.084 0.0000 g Spent fuel pool 1 0.60 4 212 0.005 0.028 0.0000 1 h Refueling slot 3 6.00 12 636 0.055 0.084 0.0000 i Vessel head 1 4.50 4 212 0.041 0.028 0.0000 j Spentfuel 1 110.00 2 118 1 .ooo 0.016 0.0155 I 35.00 2 118 0.318 0.01 6 0.0005 1 k Filter- 2 8.00 8 424 0.073 0.056 0.0000 ~ , hatch # Lifts per in 60 magnitude Load load (shroud head) I vault plugs 1 gate 1 0.45 4 212 0.004 0.028 0.0000 Plugs 1 7.25 4 21 2 0.066 0.028 0.0000 insulation shipping cask demineralizer Notes: Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Full cask out, twice per outage, plus 12 to empty SFP. Empty cask in, twice per outage, plus 12 to empty SFP. Off and on, twice each outage LRPD-03 Revision 0 Page 98 of 98 WNPS License Renewal Project TLAA and Exemption Evaluations 1 Attachment 4 - Estimate of Crane Cycles I Heavy Loads

1. Weight Contaminated 2 2.50 equipment storage I area-hatches - 1 I m I Reactorhead I 1 I 4.00 # Lifts per outage 8 4 # Lifts Load in 60 magnitude Load years W probability 424 0.023 0.056 212 0.036 0.028 I n o p strong back Stud tensioner 1 3.50 monorail Cattle chute 1 14.00 Dryer/separator 1 43.50 storage pit shield 4 212 0.032 4 212 0.127 4 212 0.395 12 636 0.255 10 530 0.055 4 212 0.023 0 40 0.045 0.028 0.028 0.028 0.084 0.070 0.028 0.005 q Crane load block 1 6.00 I Conservative total cvcles: 1 7590 1 -1 r I Conservative mean effective load factor K: HP water blaster 1 2.50 Mean effective load factor K 0.0000 0.0000 0.0000 0.0001 0.001 7 0.0014 0.0000 0.0000 0.0000 s 0.4007 Vessel service 1 5.00 platform Notes: Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage Off and on, twice each outage 10 empty cycles Off and on, twice each outage No longer used, assumed 40 cycles.