ML092080229
| ML092080229 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/31/2009 |
| From: | Nancy Salgado Plant Licensing Branch 1 |
| To: | John Carlin Ginna |
| pickett , NRR/DORL, 415-1364 | |
| References | |
| TAC MD9962 | |
| Download: ML092080229 (7) | |
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Mr. John T. Carlin Vice President RE. Ginna Nuclear Power Plant RE. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519
SUBJECT:
RE. GINNA NUCLEAR POWER PLANT: SAFETY EVALUATION FOR RELIEF REQUEST NO. 18, REACTOR VESSEL WELD EXAMINATION EXTENSION (TAC NO. MD9962)
Dear Mr. Carlin:
By letter dated October 3, 2008, RE. Ginna Nuclear Power Plant, LLC, the licensee for the RE.
Ginna Nuclear Power Plant, requested Nuclear Regulatory Commission (NRC) approval to use an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, paragraph IWB-2412, Inspection Program B.
Specifically, the alternative was requested pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(a)(3)(i). The licensee requested approval for the use of an alternative to extend the inservice inspection (lSI) interval for examinations of the reactor pressure vessel (RPV) circumferential shell and shell-to-f1ange welds (Category B-A) as well as the nozzle-to-vessel welds and nozzle inner radius sections (Category B-D) from 10 years to 20 years.
The NRC staff has completed its review of the information provided by the licensee for Relief Request No. 18. The staff concludes that the information provided by the licensee supports the granting of an alternative pursuant to 10 CFR 50.55a(a)(3)(i) because the alternative provides an acceptable level of quality and safety. This approval is only to extend the licensee's current lSI interval to 2019 with the licensee performing the subject examinations in 2011.
Sincerely, I~ ;t'~
Nancy L. Salgado, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST NO. 18 RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R.E. GINNA NUCLEAR POWER PLANT, LLC R.E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
1.0 INTRODUCTION
By letter dated October 3, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML0828800328), R.E. Ginna Nuclear Power Plant, LLC, the licensee for the R.E. Ginna Nuclear Power Plant (Ginna), resubmitted a request for Nuclear Regulatory Commission (NRC) approval to use an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, paragraph IWB-2412, Inspection Program B. Specifically, the alternative was requested pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(a)(3)(i). The licensee requested approval for the use of an alternative to extend the inservice inspection (lSI) interval for examinations of the reactor pressure vessel (RPV) circumferential shell and shell-to-flange welds (Category B-A) as well as the nozzle-to-vessel welds and nozzle inner radius sections (Category B-D) from 10 years to 20 years.
2.0 REGULATORY REQUIREMENTS 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 1O-year intervals that comply with the requirements in the latest edition and 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 current lSI interval at Ginna, which ends in 2009, the code of record for the inspection of ASME Code Class 1, 2, and 3 components will be Section XI of the ASME Code 1995 Edition with the 1996 Addenda. 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). For an alternative to be authorized, as 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.
Enclosure
-2
2.1 Background
The lSI of Category B-A and B-D components consists of visual and ultrasonic 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.
2.2 Summary of WCAP-16168-t\\lP In 2006, the Pressurized Water Reactor (PWR) Owners Group submitted Topical Report WCAP-16168-NP, Revision 1 (ADAMS Accession No. ML060330504, referred to as the WCAP in the rest of this document), "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," to the NRC in support of making a risk-informed assessment of extensions to the lSI intervals for Category B-A and B-D components. In the report, the PWR Owners Group took data associated with three different PWR plants (referred to as the pilot plants), one designed by each of the main vendors for nuclear power plants in the United States, and performed the necessary studies on each of the pilot plants required to justify the proposed extension for the lSI interval for Category B-A and B-D components from 10 to 20 years.
The analyses in the WCAP used probabilistic fracture mechanics tools and inputs from the work described in the NRC's pressurized thermal shock (PTS) risk re-evaluation, specifically NUREG 1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report," (ADAMS Accession No. ML061580318) and NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS),"
(ADAMS Accession No. ML070860156). The PWR Owners Group analyses incorporated the effects of fatigue crack growth and inservice inspection. Design basis transient data was used as input to the fatigue crack growth evaluation. The effects of lSI were modeled consistently with the previously-approved probabilistic fracture mechanics codes contained in WCAP-14572 NP-A, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," (ADAMS Accession Nos. ML012630327, ML012630349, and ML012630313). These effects were put into evaluations performed with the Fracture Analysis of Vessels-Oak Ridge (FAVOR) code (ORNL/NRC/LTR0418 (ADAMS Accession No. ML042960391)). All other inputs were identical to those used in the PTS risk re-evaluation.
From the results of the studies, the PWR Owners Group concluded that the ASME Code,Section XI 1O-yearinspection interval for Category B-A and B-D components in PWR reactor vessels can be safely 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 (Westinghouse, Combustion Engineering, and Babcock and Wilcox) as long as the critical, plant-specific parameters (defined in Appendix A of the WCAP) are bounded by the pilot plants.
2.3 Summary of NRC Safety Evaluation Report of WCAP-16168-NP The NRC staff's conclusion in its safety evaluation (SE) (ADAMS Accession No. ML0929200462) of the WCAP indicates that the methodology presented in the WCAP, in concert with the guidance provided by Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," is acceptable for referencing in requests to implement alternatives to ASME Code inspection requirements for PWR plants in accordance with the limitations and conditions
- 3 in the SE. In addition to showing that the subject plant is bounded by the pilot plants' information from Appendix A in the WCAP, the key points of the SE are summarized below.
- The dates identified in the request for an alternative should be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC. Any deviations from the implementation plan contained in OG-06-356 (ADAMS Accession No. ML082210245) should be discussed in detail in the request for an alternative lSI interval. The maximum interval for a proposed lSI is 20 years.
- The request for an alternative lSI interval can use any NRC-approved method to calculate aT30 and RTMAX-X. However, if the request uses NUREG-1874 methodology to calculate a T30, then the request should include the analysis described in paragraph (6) of subsection (f) to the voluntary PTS rule. The analysis should be done for all of the materials in the beltline area with at least three surveillance data points.
- If the subject plant has RPV forgings that are susceptible to underclad cracking or if the RPV includes forgings with RTMAX-FO values exceeding 240 OF, then the WCAP analyses are not applicable. The licensee must submit a plant-specific evaluation for any extension to the 10-year inspection interval for ASME Code,Section XI, Category B-A and B-D RPV welds.
3.0 ALTERNATIVES PROPOSED FOR GINNA 3.1 Description of Proposed Alternatives In Relief Request No. 18, the licensee proposes to allow performance of the ASME Code required Category B-A and B-D weld lSI of Ginna during the 2011 outage, with the next exam to be performed in 2031 (a 20-year interval from the last inspection). This schedule is consistent with the information in PWR Owners Group letter, OG-06-356.
3.2 Component for Which Relief is Requested The affected component is the Ginna RPV. The following examination categories and item numbers from IWB-2500 and Table IWB-2500-1 of the ASME Code,Section XI, are addressed in this request:
For Relief Request No. 18 Examination Category Item Number Description B-A B1.11 Circumferential Shell Weld B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas
- 4 3.3 Basis for Proposed Alternatives The basis for the first alternative is found in the NRC-approved version of the WCAP, WCAP 16168-NP-A, Rev. 2 (ML0828200462, referred to as WCAP-A in the rest of this document).
Plant-specific parameters for the subject plant are summarized in the attachment to the licensee's letter of October 3, 2008. The format of the information is patterned after that found in Appendix A of the WCAP.
All of the critical parameters listed in Tables 1, 2, and 3 of the attachment to the licensee's submittal are bounded by the WCAP-A pilot plant evaluations.
3.4 Duration of Proposed Alternatives The duration of the proposed alternative is for examinations for the Fourth Interval Inservice Inspection Program are proposed to be performed in the 2011 outage, with following examinations proposed to be performed in 2031.
4.0 STAFF TECHNICAL EVALUATION The NRC staff has reviewed the attachment to the licensee's request for an alternative submittal, dated October 3, 2008, to make this evaluation. The "Frequency and Severity of Design Transients" of Ginna were found to be bounded by WCAP-A. Also, the Ginna RPV is single-layer clad and, therefore, was bounded by WCAP-A.
Table 2 of the submittal includes additional information pertaining to previous RPV inspections and the schedule for future ones. One recordable indication is in the inner 1/8th of the vessel inside diameter in the beltline region. The indication has a depth of 0.22", length of 1.54", and is 0.38" subsurface. This indication was found to be acceptable in accordance with IWB-3500 of Section XI of the ASME Code.
The calculation of through-wall cracking frequency (TWCF) TWCF95-ToTAL was performed using Table 3 as a basis. The licensee's submittal used the NUREG-1874 methodology to calculate l::. T30. The NRC staff verified these values and calculated l::. T30 values using the methodology of RG 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials." The TWCF was found to be acceptably low as calculated through the methodology prescribed in the WCAP and in Table 3 of the submittal, using the RG 1.99, Rev. 2 methodology.
At the time of issuance of the SE for WCAP-16168-NP, it was the NRC's intent to establish a process by which licensees could receive approval to implement 20-year lSI intervals for the subject component examinations through the end of their facility's current operating license.
This objective led to the provision established in the WCAP-16168-NP SE that licensees submit a license condition which would require the licensee to evaluate future volumetric lSI data in accordance with the criteria in the draft and/or final alternative PTS Rule, 10 CFR 50.61a.
However, since that time, further guidance from the NRC's Office of General Counsel has resulted in a modification of this NRC position.
Based on the current guidance, the NRC staff will grant lSI interval extensions for the subject components on an interval-by-interval basis, i.e., only a facility's current lSI interval will be extended for up to 20 years. Licensees will have to submit subsequent requested alternatives,
- 5 for NRC review and approval, to extend each following lSI interval from 10 years to 20 years, as needed. Based on this new NRC position, the requirement in the staff's SE on WCAP-16168 NP for a license condition to address the evaluation of future lSI data is no longer necessary, and the license condition requested by Ginna LLC for Ginna by letter dated October 7,2008 (ADAMS Accession No. ML082880033), will not be issued in conjunction with this requested alternative. However, subsequent requested alternatives which seek to extend additional lSI intervals from 10 to 20 years for the subject component examinations should include the evaluation of a facility's most recent lSI data in accordance with the criteria in the final alternative PTS Rule, 10 CFR 50.61a, in order to obtain NRC staff approval. In addition, for purposes of technical and regulatory consistency, the WCAP-16168-NP SE will be revised to reflect these changes in NRC position regarding the implementation of lSI interval extensions based on WCAP-16168-NP.
Hence, the NRC staff also reviewed the licensee's proposed schedule as it applies to the Fourth Ginna lSI Interval. The staff finds the proposal to perform the subject examinations in 2011 consistent with the information provided in OG-06-356, therefore, acceptable.
In summary, the licensee has demonstrated through the submittal that the RPV components identified under Relief Request No. 18 for Ginna is bounded by WCAP-A. The submittal demonstrates that there is no significant additional risk associated with extending the current lSI interval for Category B-A and B-D components from 10 years to 20 years.
5.0 CONCLUSION
The NRC staff has completed its review of the submittals for Relief Request No. 18 for Ginna.
The staff concludes that extending the current lSI interval for the identified Category B-A and B-D components from 10 years to 20 years shows no appreciable increase in risk. The staff comes to this conclusion based on the fact that the plant-specific information provided by the licensee is bounded by the data in WCAP-A, and the request meets all the conditions and limitations described in WCAP-A and the SE to the WCAP. Therefore, the staff concludes that Relief Request No. 18 provides an acceptable level of quality and safety and the alternative can be granted pursuant to 10 CFR 50.55a(a)(3)(i) until the end of the current extended lSI interval in 2019 with the licensee performing the subject examinations in 2011.
All other requirements of the ASME Code,Section XI, not specifically included in the request for the proposed alternatives, remain in effect.
Principal Contributor: Carolyn Fairbanks Date: Jiily 31, 2fff)
July 31, 2009 Mr. John T. Carlin Vice President RE. Ginna Nuclear Power Plant RE. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519
SUBJECT:
RE. GINNA NUCLEAR POWER PLANT: SAFETY EVALUATION FOR RELIEF REQUEST NO. 18, REACTOR VESSEL WELD EXAMINATION EXTENSION (TAC NO. MD9962)
Dear Mr. Carlin:
By letter dated October 3, 2008, RE. Ginna Nuclear Power Plant, LLC, the licensee for the RE.
Ginna Nuclear Power Plant, requested Nuclear Regulatory Commission (NRC) approval to use an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, paragraph IWB-2412, Inspection Program B.
Specifically, the alternative was requested pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(a)(3)(i). The licensee requested approval for the use of an alternative to extend the inservice inspection (lSI) interval for examinations of the reactor pressure vessel (RPV) circumferential shell and shell-to-flange welds (Category B-A) as well as the nozzle-to-vessel welds and nozzle inner radius sections (Category B-D) from 10 years to 20 years.
The NRC staff has completed its review of the information provided by the licensee for Relief Request No. 18. The staff concludes that the information provided by the licensee supports the granting of an alternative pursuant to 10 CFR 50.55a(a)(3)(i) because the alternative provides an acceptable level of quality and safety. This approval is only to extend the licensee's current lSI interval to 2019 with the licensee performing the subject examinations in 2011.
Sincerely, IRAI Nancy L. Salgado, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosure:
Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION; PUBLIC RidsNrrPMDPickett RidsAcrsAcnwMailCenter LPL1-1r/f RidsNrrDciCvib CFairbanks, CVIB RidsNrrDorlLpl1-1 RidsOgcMailCenter GDentel, R1 SLittie ADAMS Accession No. ML092080229 OFFICE LPL 1-1\\PM LPL 1-1\\LA CVIB/BC LPL 1-1\\BC NAME DPickett SLittie MMitchell as siqned on NSalgado DATE 7/29/09 7/29/09 07/16/09 7/31/09 Officlal Record Copy