ML14314A987
| ML14314A987 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 11/28/2014 |
| From: | Jessie Quichocho Watts Bar Special Projects Branch |
| To: | James Shea Tennessee Valley Authority |
| Minarik A | |
| References | |
| TAC MF2956 | |
| Download: ML14314A987 (12) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 28, 2014 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 11 01 Market Street, LP 3D-C Chattanooga, TN 37402-2801
SUBJECT:
WATTS BAR NUCLEAR PLANT, UNIT 1- REQUEST FOR ALTERNATIVE 13-ISI-01 TO EXTEND THE SECOND REACTOR VESSEL WELD INSERVICE INSPECTION INTERVAL (TAC NO. MF2956)
Dear Mr. Shea:
By letter dated October 18, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13296A016), as supplemented by letter dated September 26, 2014 (ADAMS Accession No. ML14272A153), Tennessee Valley Authority (TVA or the licensee) proposed an extension of the inservice inspection (lSI) interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, lSI Program for Watts Bar Nuclear Plant, Unit 1.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR),
Section 50.55a(a)(3)(i), the licensee proposed to extend the lSI interval for examinations of the reactor pressure vessel welds (Category B-A) as well as the nozzle-to-vessel welds and inner radius sections (Category B-D) from 10 to 20 years. Furthermore, this proposal only applies to the second 10-year lSI Program interval. TVA stated in the request that the proposed alternatives would provide an acceptable level of quality and safety.
The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the subject request and concludes, as set forth in the enclosed safety evaluation that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i). Pursuant to 10 CFR 50.55a(a)(3)(i), alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if the licensee demonstrates that the proposed alternatives will provide an acceptable level of quality and safety. The staff concludes that increasing the lSI interval for the Examination Category B-A and B-D RPV welds from 10 to 20 years will result in no appreciable increase in risk. This conclusion is based on the fact that the plant-specific information provided by the licensee is bounded by the data in the Westinghouse Commercial Atomic Power (WCAP) Topical Report No. 16168-NP-A, Rev. 2 (WCAP-A), and the request meets all the conditions and limitations described in the WCAP-A. Therefore, the staff concludes that Request 13-ISI-01 provides an acceptable level of quality and safety, and the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the second 10-year lSI interval at Watts Bar, Unit 1. Accordingly, the examinations of the ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A and B-D RPV welds may be deferred until the end of 2024 plus or minus one refueling cycle and no later than May 27, 2027.
J.Shea All other requirements of the ASME Code,Section XI, not specifically included in the request for the proposed alternative, remain in effect.
If you have any questions, please contact the Project Manager, Jeanne Dion at 301-415-1349.
Sincerely, e:.Yer.~~
/'Watts Bar Special Projects Branch f/ Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390
Enclosure:
Safety Evaluation cc w/enclosure: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE 13-ISI-01 REGARDING SECOND 10-YEAR INSERVICE INSPECTION INTERVAL FOR REACTOR VESSEL WELDS TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNIT 1 DOCKET NUMBER 50-390
1.0 INTRODUCTION
By letter dated October 18, 2013 (U.S. Nuclear Regulatory Commission (NRC) Agencywide Documents Access and Management System (ADAMS) Accession No. ML13296A016) as supplemented by a letter dated September 26, 2014 (ADAMS Accession No. ML14272A153),
Tennessee Valley Authority (the licensee) submitted Request for Alternative 13-ISI-01 for the second 10-year inservice inspection (lSI) interval for Watts Bar Nuclear Plant, Unit 1 (Watts Bar, Unit 1). The licensee's request proposed an alternative to the lSI interval requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, Paragraph IWB-2412, "Inspection Program B," for inservice examination of the reactor pressure vessel (RPV) welds required by Table IWB-2500-1, Examination Categories B-A, "Pressure Retaining Welds in Reactor Vessel," and B-D, "Full penetration Welded Nozzles in Vessels." The licensee's proposed alternative would extend the second 10-year lSI interval from 10 to 20 years in order to allow for the deferral of the subject RPV weld examinations.
The second 10-year lSI interval for Watts Bar, Unit 1 began on May 27,2007, and is scheduled to end on May 26, 2016. The licensee's proposed alternative would allow for the deferral of the subject RPV examinations until 2024, plus or minus one refueling cycle. The staff reviewed Request for Alternative 13-ISI-01 pursuant to Section 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (1 0 CFR) to determine whether the licensee's proposed alternative to the lSI interval requirements of the ASME Code,Section XI will provide an acceptable level of quality and safety.
2.0 REGULATORY EVALUATION
2.1 Regulations and Guidance In accordance with 10 CFR 50.55a(g)(4), the licensee is required to perform lSI of ASME Code Class 1, 2, and 3 components and system pressure tests during the first 10-year interval and subsequent 10-year intervals that comply with the requirements in the latest edition and Enclosure
addenda of Section XI of the ASME Code, incorporated by reference in 10 CFR 50.55a(b),
subject to the limitations and modifications listed therein.
For the second 10-year lSI interval at Watts Bar, Unit 1, the Code of record for the inspection of ASME Code Class 1, 2, and 3 components is the 2001 Edition through the 2003 Addenda of the ASME Code,Section XI. The regulation in 10 CFR 50.55a{a)(3) states, in part, that the Director of the Office of Nuclear Reactor Regulation may authorize an alternative to the requirements of 10 CFR 50.55a(g). There are two justifications for an alternative to be authorized. First, per 10 CFR 50.55a(a)(3)(i), the licensee must demonstrate that the proposed alternative would provide an acceptable level of quality and safety. For the second possible justification for an alternative to be authorized, described in 10 CFR 50.55a(a){3)(ii), the licensee must show that following the ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Regulatory Guide (RG) 1.99, Revision (Rev.) 2, "Radiation Embrittlement of Reactor Vessel Materials," describes general procedures acceptable to the staff for calculating the effects of neutron irradiation embrittlement of the low-alloy steels currently used for light-water-cooled RPVs.
RG 1.174, Rev. 1, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing basis changes by considering engineering issues and applying risk insights.
RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," describes methods and assumptions acceptable to the staff for determining the RPV neutron fluence.
2.2 Background The lSI of Examination Categories B-A and B-D components consists of volumetric and surface examinations intended to discover whether flaws have initiated, whether pre-existing flaws have extended, and whether pre-existing flaws may have been missed in prior examinations. These examinations are required to be performed at regular intervals, as defined in Section XI of the ASME Code.
2.3 Summary of Westinghouse Commercial Atomic Power (WCAP)
Topical Report No. 16168-NP-A, Rev. 2 In June 2008, the Pressurized Water Reactor Owners Group (PWROG) issued the NRC-approved topical report WCAP-16168-NP-A, Rev. 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (ADAMS Accession No. ML082820046), which is in support of a risk-informed assessment of extensions to the lSI intervals for Examination Categories B-A and B-D components. Specifically, WCAP-16168-NP-A, Rev. 2 took data associated with three different PWR plants (referred to as the pilot plants), one designed by each of the three main vendors (Westinghouse, Combustion Engineering, and Babcock and Wilcox (B&W)) for PWR nuclear power plants in the United States, and performed studies on these pilot plants to justify the proposed extension of the lSI interval for the Examination Categories B-A and B-D components from 10 to 20 years.
The analyses in WCAP-16168-NP-A, Rev. 2 used the probabilistic fracture mechanics (PFM) methodology and inputs from the work described in NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (1 0 CFR 50.61 )"
(ADAMS Accession No. ML061580318) and NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)" (ADAMS Accession No. ML070860156). The PWROG analyses incorporated the effects of fatigue crack growth and lSI examination histories. Design basis transient data was used as input to the fatigue crack growth evaluation. The effects of lSI examination histories were modeled consistent with a previously-approved PFM Code in WCAP-14572-NP-A, 'Westinghouse Owners Group Application of Risk-Informed Methods to Piping lnservice Inspection" (ADAMS Accession No. ML012630375). These effects were considered in the PFM evaluations, using the Fracture Analysis of Vessels - Oak Ridge (FAVOR) computer code (ADAMS Accession No. ML042960391). All other inputs were identical to those used in the PTS re-evaluation underlying 10 CFR 50.61 a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,"
heretofore identified as the alternate PTS rule.
From the results of the studies, the PWROG concluded that the ASME Code,Section XI 10-year inspection interval for Categories B-A and B-D components in PWR RPVs can be extended to 20 years. Their conclusion from the results for the pilot plants was considered to apply to any plant designed by the three vendors as long as the critical, plant-specific parameters (defined in Appendix A of WCAP-16168-NP-A, Rev. 2) are bounded by the parameters of the pilot plants.
2.4 Summary of the July 26, 2011, NRC Safety Evaluation (SE) for WCAP-16168-NP-A, Rev. 2 The original SE in WCAP-16168-NP-A, Rev. 2 that was published in 2008, was superseded by the July 26, 2011, SE (ADAMS Accession No. ML111600303) to address the PWROG's request for clarification of the information needed in applications utilizing WCAP-16168-NP-A, Rev. 2.
The staff's conclusion in the July 26, 2011 SE indicates that the methodology presented in WCAP-16168-NP-A, Rev. 2 is consistent with RG 1.174, Rev. 1 and is acceptable for referencing in requests to implement alternatives to ASME Code inspection requirements for PWR plants in accordance with the limitations and conditions in the July 26, 2011 SE. In addition to showing that the subject plant parameters and inspection history are bounded by the critical parameters identified in Appendix A in WCAP-16168-NP-A, Rev. 2, the licensee's application must provide the following plant-specific information:
(1) Licensees must demonstrate that the embrittlement of their RPV is within the envelope used in the supporting analyses. Licensees must provide the 95th percentile total through-wall cracking frequency (TWCFTOrAd and its supporting material properties at the end of the period in which the relief is requested to extend the lSI from 10 to 20 years. The 95th percentile TWCFrorAL must be calculated using the methodology in NUREG-1874. The Charpy transition temperature for the RPV, RTMAx-x. and the shift in the Charpy transition temperature produced by irradiation defined at the 30 ft-lb energy level, ~T30 , must be calculated using the methodology documented in the latest revision of RG 1.99 or other NRC-approved methodology.
(2) Licensees must report whether the frequency of the limiting design basis transients during prior plant operation are less than the frequency of the design basis transients
identified in the PWROG fatigue analysis that are considered to significantly contribute to fatigue crack growth.
(3) Licensees must report the results of prior lSI of the RPV welds and the proposed schedule for the deferred RPV weld exams for the 20-year lSI interval. The 20-year inspection interval is the maximum interval allowed per WCAP-16168-NP, Rev. 2. In its request for an alternative, each licensee shall identify the years in which the deferred inspections will be performed. The dates provided in licensees' requests must be within plus or minus one refueling cycle of the dates identified in the revised implementation plan provided to the NRC in PWROG Letter OG-10-238, dated July 12, 2010 (ADAMS Accession No. ML11153A033).
(4) Licensees with B&W plants must (a) verify that the fatigue crack growth for 12 heat-up/cool-down transients per year that was used in the PWROG fatigue analysis bound the fatigue crack growth for all of its design basis transients and (b) identify the design bases transients that contribute to significant fatigue crack growth.
(5) Licensees with RPVs having forgings that are susceptible to underclad cracking and with RT MAx-Fo values exceeding 240 degrees Fahrenheit (°F) must submit a plant-specific evaluation to extend the inspection interval for the ASME Code,Section XI, Examination Category B-A and B-D RPV welds from 10 to a maximum of 20 years because the analyses performed in the WCAP-A are not applicable.
(6) Licensees seeking second or additional interval extensions shall provide the information and analyses requested in Section (e) of 10 CFR 50.61a.
WCAP-16168-NP-A, Rev. 3, which contains the July 26, 2011 SE for WCAP-16168-NP-A, Rev.
2, was issued in October 2011 (ADAMS Accession No. ML11306A084, referred to as the WCAP-A in the rest of this SE).
3.0 PROPOSED ALTERNATE FOR WATTS BAR. UNIT 1 3.1 Description of Proposed Alternative In Request for Alternative 13-ISI-01, the licensee proposed to defer the RPV weld examinations required by the ASME Code,Section XI, Table IWB-2500-1, Examination Categories B-A and B-D for Watts Bar, Unit 1 until 2024, plus or minus one refueling cycle.
3.2 Components for Which the Proposed Alternative is Requested The licensee requested deferral of the Watts Bar, Unit 1 second 10-year lSI interval examinations for the following RPV examination categories and item numbers from Table IWB-2500-1 of the ASME Code, Section XI:
Examination Category Item Number Description B-A B1.11 Circumferential Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Head Welds B-A B1.30 Shell-to-Flange Weld
B-A B1.40 Head-to-Flange Weld B-0 B3.90 Nozzle-to-Vessel Welds B-0 B3.100 Nozzle Inner Radius Section 3.3 Licensee's Basis for the Proposed Alternative The licensee stated that the methodology used to demonstrate the acceptability of extending the inspection interval for the Examination Category B-A and B-0 RPV welds is contained in the WCAP-A. This methodology uses the calculated TWCFrorAL as a measure of the risk of RPV failure. The licensee indicated that the inspection interval for the affected components can be extended from 10 to 20 years, meeting the change in risk guidelines in RG 1.174. The licensee addressed the plant-specific information discussed in Section 2.4 of this SE as follows:
(1) The licensee provided detailed TWCFrorAL calculations along with critical input parameters (neutron fluence and material properties) for demonstrating that the embrittlement of the Watts Bar, Unit 1 RPV is bounded by the Westinghouse pilot plant analysis. The TWCFrorAL value was calculated as 1.58 x 1o* 10 per operating reactor year (/ry); this value is less than the value of 1.76 x 1o-os /ry for the Westinghouse pilot plant study in the WCAP-A.
(2) The frequency of the Watts Bar, Unit 1 RPV limiting design basis transients are bounded by the frequencies identified in the PWROG fatigue analysis.
(3) The results of the previous RPV inspections for Watts Bar, Unit 1 are provided, which confirm that satisfactory examinations have been performed. The RPV examinations, which are currently scheduled to occur prior to the end of the second 10-year lSI interval in 2015, will be deferred until 2024, plus or minus one refueling cycle. The proposed examination date of 2024 is a deviation from the latest revised implementation plan schedule for Watts Bar, Unit 1 in PWROG Letter OG-1 0-238.
Plant-Specific Information Items (4), (5), and (6) were not addressed in the licensee's application. Based on its evaluation of plant-specific information items (1 ), (2), and (3) above, the licensee concluded that the Watts Bar, Unit 1 RPV is bounded by the Westinghouse pilot plant analysis in the WCAP-A, and therefore, the use of this proposed alternative will provide an acceptable level of quality and safety. Accordingly, the licensee requested that the NRC authorize Request for Alternative 13-ISI-01, pursuant to 10 CFR 50.55a(a)(3)(i).
4.0 STAFF TECHNICAL EVALUATION The NRC staff reviewed Request for Alternative 13-ISI-01 to determine whether the Watts Bar, Unit 1 RPV is bounded by the Westinghouse pilot plant study performed in the WCAP-A. The staff conducted its review by performing an independent evaluation of Plant-Specific Information Items (1 ), (2), and (3), as provided in the licensee's submittal. The staff also reviewed the licensee's submittal to verify that Plant-Specific Information Items (4), (5), and (6) are not applicable to Watts Bar, Unit 1 and therefore do not require a plant-specific evaluation.
Regarding Plant-Specific Information Item (1 ), the staff reviewed the licensee's statement in Table 1 of the submittal that the plant-specific value for TWCFrorAL (1.58 x 1o* 10 /ry) is bounded by the TWCFroTAL of 1.76 X 10"08 /ry from the Westinghouse pilot plant study in the WCAP-A.
Table 3 of the submittal provides the details of the TWCFTOTAL calculation for the Watts Bar, Unit 1 RPV. The staff performed an independent calculation to verify that the licensee's TWCFToTAL value was correctly determined in accordance with NUREG-1874, as required by the WCAP-A. The staff also verified that the RTMAx-x and t.. T 30 values were calculated using the methodology in RG 1.99, Rev. 2, consistent with the WCAP-A. Additionally, the staff reviewed the unirradiated material properties used for the TWCFTOTAL calculation against those used as the basis for the 40-year (32 effective full-power years (EFPY)) embrittlement analysis of the Watts Bar, Unit 1 RPV in Section 5.2, "Integrity of the Reactor Coolant Pressure Boundary," of the Final Safety Analysis Report (FSAR). The staff determined that the unirradiated material property inputs for the RPV beltline materials are consistent with those documented in the FSAR and previous docketed pressure-temperature limits report submittals. The staff also found that the 32 EFPY neutron fluence used for the TWCFTOTAL calculation is the same as the NRC-approved value calculated based on the analysis of dosimeters from the latest surveillance capsule removed from the Watts Bar, Unit 1 RPV. This surveillance capsule test report, WCAP-16760-NP, "Analysis of Capsule Z from the Tennessee Valley Authority, Watts Bar, Unit 1 Reactor Vessel Radiation Surveillance Program," November 2007 (ADAMS Accession No. ML073200244) was approved by the staff in a SE dated May 28, 2008 (ADAMS Accession No. ML081440258). Based on this assessment, the staff finds that the TWCFTOTAL value for Watts Bar, Unit 1 is bounded by the WCAP-A results, and therefore, the licensee has satisfactorily addressed Plant-Specific Information Item (1) for Watts Bar, Unit 1.
Regarding Plant-Specific Information Item (2), the staff reviewed the licensee's statement in Table 1 of the submittal that the frequency and severity of the design basis transients for Watts Bar, Unit 1 are bounded by the Westinghouse pilot plant study in the WCAP-A. The Westinghouse pilot plant study in the WCAP-A analyzed fatigue crack growth based on seven heatup and cooldown cycles per year; this was determined to be the bounding design basis transient frequency and severity for fatigue crack growth for the Westinghouse design. The staff reviewed the licensee's statement in Table 1 of the submittal against the information in Section 5.2 of the FSAR pertaining to the number of heatup and cooldown cycles analyzed for the design of the reactor coolant system for the 40 year licensed operating term. The staff verified that the cumulative number of heatup and cooldown cycles for 40 years, as listed in the FSAR, corresponds to less than seven heatup and cooldown cycles per year from the WCAP-A.
Based on its review of this information, the staff determined that there is adequate assurance that the number of heatup and cooldown cycles for Watts Bar, Unit 1 will remain less than that assumed for the Westinghouse pilot plant study in the WCAP-A. Therefore, the staff finds that the licensee has satisfactorily addressed Plant-Specific Information Item (2) for Watts Bar, Unit 1.
Regarding Plant-Specific Information Item (3), the staff reviewed the information provided in Table 2 of the licensee's submittal pertaining to the results of previous inspections of the RPV Examination Category B-A and B-D welds. Table 2 states that one 10-year lSI has been performed to date, and the RPV volumetric examinations performed during this first 10-year lSI interval met the performance demonstration requirements of Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the ASME Code,Section XI. The staff finds this information acceptable because paragraph (e), "Examination and Flaw Assessment Requirements," of the alternate PTS rule, 10 CFR 50.61 a, requires that volumetric examinations performed in support of demonstrating the applicability of the rule (which is the basis for the risk-informed analysis of the WCAP-A) must use procedures, equipment, and personnel that have been qualified in accordance with Appendix VIII of the ASME Code,Section XI, as
specified in 10 CFR 50.55a(b)(2)(xv). The staff's evaluation of the licensee's first 10-year lSI interval RPV beltline volumetric examination results, relative to the acceptance criteria of 10 CFR 50.61a, are discussed below.
The licensee stated in Table 2 that there were zero indications identified in the RPV beltline region during the most recent lSI; therefore, the lSI results for Watts Bar, Unit 1, inherently satisfy the requirements of 10 CFR 50.61 a. The staff determined that greater specificity was needed regarding the presence of RPV beltline indications in order to determine whether the previous lSI results are bounded by 10 CFR 50.61 a. Specifically, even if there are no indications that fall outside the acceptance standards of the ASME Code,Section XI, Table IWB-351 0-1, such ASME Code-acceptable indications would still require an assessment in accordance with 10 CFR 50.61a(e)(1) to ensure that they are bounded by the limits of Tables 2 and 3 of 10 CFR 50.61a, if the flaws are located within a depth from the clad-to-base metal interface of 1 inch or 10 percent of the RPV wall thickness, whichever is greater. This assessment is necessary in order to demonstrate that the risk-informed analyses of the WCAP-A are applicable. Therefore, in Request for Additional Information (RAI)-1, the staff requested that the licensee clarify whether any relevant indications were found in the RPV beltline region within a depth from the clad-to-base metal interface of 1 inch or 10 percent of the RPV wall thickness, whichever is greater, that would require screening in accordance with the criteria of 10 CFR 50.61a(e)(1).
In its September 26, 2014, RAI response (ADAMS Accession No. ML14272A153), the licensee stated that during the most recent inservice examination of the RPV welds, there were no recordable beltline indications. The licensee further stated that there were no indications that required sizing and subsequent evaluation per IWB-3510, and there were no flaws that would require evaluation per the acceptance criteria 10 CFR 50.61 a(e)(1 ), (2), or (3). The staff found the licensee's response to RAI-1 acceptable because the licensee confirmed that there were no RPV beltline indications that would require evaluation in accordance with 10 CFR 50.61a(e)(1).
Therefore, the staff finds that the volumetric examination results for the Watts Bar, Unit 1 RPV are bounded by the requirements of 10 CFR 50.61 a, and the risk-informed analyses of the WCAP-A are applicable. Accordingly, RAI-1 is resolved.
Plant-Specific Information Item (3) states that the deferred RPV weld examination dates provided in licensees' requests to implement this alternative must be within plus or minus one refueling cycle of the dates identified in the revised implementation plan provided to the NRC in PWROG Letter OG-1 0-238. Furthermore, the 20-year inspection interval is the maximum interval allowed for the subject RPV welds per the WCAP-A methodology. The licensee proposed to defer the subject RPV weld examinations until 2024 plus or minus one refueling cycle. The staff noted that this proposed inspection schedule is a deviation from the implementation date of 2020 for Watts Bar, Unit 1 provided in PWROG Letter OG-1 0-238.
However, the new proposed examination date of 2024 plus or minus one refueling cycle still ensures that the deferred RPV weld exams will occur no later than 20 years from the start of the second 10-year lSI interval, which began on May 27, 2007. Therefore, the licensee's new deferred examination date of 2024 plus or minus one refueling cycle is still bounded by the WCAP-A. However, in order to complete its evaluation of Request for Alternative 13-ISI-01, the staff must evaluate the licensee's reason(s) for deviating from the implementation plan in PWROG Letter OG-10-238. Therefore, in RAI-2 the staff requested that the licensee discuss the reason for deviating from this implementation schedule.
In its September 26, 2014, RAI response, the licensee stated that the deviation from the 2020 implementation date provided in PWROG Letter OG-10-238 is requested in order to merge the RPV weld examination schedule with other ASME Code examinations, which require the core barrel to be removed. The licensee specifically noted that the volumetric examination of the RPV nozzle welds (Examination Category B-0) and visual examination of the interior attachments to the RPV and core support structure (Examination Categories B-N-2 and B-N-3) require removal of the core barrel and are scheduled for performance near the end of each lSI interval. To avoid the additional cost and radiation exposure associated with an additional removal of the core barrel mid-way through the third interval, the licensee requested that the second 10-year lSI interval RPV weld examinations be performed coincident with these other exams in 2024 plus or minus one refueling outage.
The staff found the licensee's response to RAI-2 acceptable because the licensee provided adequate justification for the deviation from the 2020 RPV weld exam implementation date provided in PWROG letter OG-10-238. Specifically, the staff finds that the removal of the core barrel in 2020 in order to examine the Examination Category B-D RPV nozzle welds in accordance with the 2020 implementation date in PWROG Letter OG-1 0-238 would impose a significant burden and hardship on the licensee since the removal of the core barrel would also be required in 2024 (plus or minus one refueling cycle) for the examinations of the RPV interior attachments and core support structure. Therefore, the licensee's proposal to align the examinations of the subject RPV welds with the examinations of the RPV interior attachments and core support structure, such that all of these exams occur in 2024 plus or minus one refueling cycle, is appropriate because it requires the licensee to perform one core barrel removal evolution for this lSI interval, and the deferred RPV weld examination date of 2024 plus or minus one refueling outage is still within the 20 year limit established in the WCAP-A.
Therefore, the staff finds that RAI-2 is resolved.
Based on the acceptable first 10-year lSI interval volumetric examination results, as discussed above, and the acceptable deferred RPV weld examination date of 2024 plus or minus one refueling cycle, the staff finds that the licensee has satisfactorily addressed Plant-Specific Information Item (3) for Watts Bar, Unit 1.
Regarding Plant-Specific Information Item (4), the staff noted that this item is only applicable to B&W plants since this is the only plant design for which plant-specific verification of the fatigue crack growth for all design basis transients is necessary. Watts Bar, Unit 1 is a Westinghouse plant, and it has already been established in the WCAP-A that the fatigue crack growth corresponding to seven heatup/cooldown cycles per year is bounding for all design basis transients, per the Westinghouse pilot plant study. Therefore, the staff finds that Plant-Specific Information Item (4) is not applicable to Watts Bar, Unit 1, and no plant-specific evaluation is required for this item.
Regarding Plant-Specific Information Item (5), the staff noted that the RT MAx-Fo value for Watts Bar, Unit 1 is less than 240 °F. This item is only applicable to plants with RTMAX-FO values exceeding 240 oF (regardless of the number of cladding layers). Therefore, the staff finds that Plant-Specific Information Item (5) is not applicable to Watts Bar, Unit 1, and no plant-specific evaluation is required for this item.
Regarding Plant-Specific Information Item (6), the staff finds that this item is not applicable to Watts Bar, Unit 1 because the licensee's application requested deferral of the subject RPV weld examinations for only the second 10-year lSI interval.
Based on its review of the licensee's submittal, and its findings regarding Plant Specific Information Items (1 ), (2), (3), (4), (5), and (6), as documented above, the staff has determined that the licensee adequately demonstrated that the Watts Bar, Unit 1 RPV is bounded by Westinghouse pilot plant study from the WCAP-A. Consequently, the licensee has demonstrated that the proposed alternative will provide an acceptable level of quality and safety, and it meets the guidance provided by RG 1.174, Rev. 1 for risk-informed decisions.
5.0 CONCLUSION
Pursuant to 10 CFR 50.55a(a)(3)(i), alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if the licensee demonstrates that the proposed alternatives will provide an acceptable level of quality and safety. The staff has completed its review of Request for Alternative 13-ISI-01 for Watts Bar, Unit 1. The staff concludes that increasing the lSI interval for the Examination Category B-A and B-D RPV welds from 10 to 20 years will result in no appreciable increase in risk. This conclusion is based on the fact that the plant-specific information provided by the licensee is bounded by the data in the WCAP-A, and the request meets all the conditions and limitations described in the WCAP-A. Therefore, the staff concludes that Request 13-ISI-01 provides an acceptable level of quality and safety, and the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the second 10-year lSI interval at Watts Bar, Unit 1. Accordingly, the examinations of the ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A and B-D RPV welds may be deferred until the end of 2024 plus or minus one refueling cycle and no later than May 27, 2027.
All other requirements of the ASME Code,Section XI, not specifically included in the request for the proposed alternative, remain in effect.
J. Shea All other requirements of the ASME Code,Section XI, not specifically included in the request for the proposed alternative, remain in effect.
If you have any questions, please contact the Project Manager, Jeanne Dion at 301-415-1349.
Sincerely, IRA/
Jessie F. Quichocho, Chief Watts Bar Special Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390
Enclosure:
Safety Evaluation cc w/enclosure: Distribution via ListServ DISTRIBUTION:
PUBLIC LPL2-2 R/F RidsNrrDorllp_Wb RidsAcrsAcnw_MaiiCTR RidsNrrPMWattsBar1 RidsNrrLABCiayton RidsRgn2MaiiCenter RidsNrrDeEvib JJenkins, NRR JJandovitz, EDO AMinarik, NRR ADAMS Access1on No.: ML14314A987 *b y_memo dae t d 0 co t ber 29 2014 OFFICE LPWB/PM LPL2-2/PM LPL2-2/LA NRR/DE/EVIB* LPL2-2/BC NAME AMinarik JDion BCiayton SRosenberg JQuichiocho DATE 11/17/14 11/19/14 11/18/14 10/29/14 11/28/14 OFFICIAL RECORD COPY