ML16038A001
| ML16038A001 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 02/16/2016 |
| From: | Travis Tate Plant Licensing Branch 1 |
| To: | Heacock D Dominion Nuclear Connecticut |
| Guzman R, NRR/DORL/LPL | |
| References | |
| CAC MF5868 | |
| Download: ML16038A001 (10) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 16, 2016 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Nuclear Connecticut, Inc.
Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 3-ALTERNATIVE REQUEST IR-3-27 FOR IMPLEMENTATION OF EXTENDED REACTOR VESSEL INSERVICE INSPECTION INTERVAL (CAC NO. MF5868)
Dear Mr. Heacock:
By letter dated March 9, 2015, Dominion Nuclear Connecticut, Inc. (the licensee), submitted Alternative Request IR-3-27, which proposed an alternative to the inservice inspection (ISi) 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 the Millstone Power Station, Unit No. 3 (MPS3).
Pursuant to Title 1O of the Code of Federal Regulations (10 CFR), Part 50, Section 55a(z)(1 ),
the licensee requested the use of a proposed alternative to extend the MPS3 reactor vessel third ISi from 10 years to 20 years, on the basis that the alternative provides an acceptable level of quality and safety. The current third 10-year ISi interval ends on April 22, 2019, for MPS3.
The U.S. Nuclear Regulatory Commission staff has completed its review of the licensee's proposed Alternative Request IR-3-27 for MPS3. Pursuant to 10 CFR 50.55a(z)(1 ), the staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety. The staff also concludes that extending the third ISi interval for Examination Categories B-A and B-D components from 1O years to 20 years will not result in any considerable increase in risk; and the licensee's alternative ISi schedule for the specified welds is acceptable for extension to April 22, 2029. The examination of the Categories B-A and B-D components for MPS3 shall be conducted prior to the end of the extended third interval. All other requirements of the ASME Code,Section XI, not specifically included in the request for the proposed alternative, remain in effect.
D. Heacock If you have any questions, please contact the project manager, Richard Guzman, at (301) 415-1030 or Richard.Guzman@nrc.gov.
Sincerely, Travis L. Tate, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423
Enclosure:
Safety Evaluation cc w/enclosure: Distribution via Listserv
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UNITE.b .STATES NUCLE~R*.REGULATORV..COMMISSION
- !WASHINGTON, o.c. 20555~0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE REQUEST FOR IMPLEMENTATION OF EXTENDED REACTOR VESSEL THIRD INSERVICE INSPECTION INTERVAL REQUEST FOR RELIEF DOMINION NUCLEAR CONNECTICUT, INC .. ET AL.
MILLSTONE POWER STATION. UNIT NO. 3 DOCKET NO. 50-423
1.0 INTRODUCTION
By le'tter dated March 9, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15075A025), Dominion Nuclear Connecticut, Inc. (the licensee),
- submitted a relief request that proposed an alternative to the inservice inspection (ISi) interval
- '.requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (,C"SME Code),Section XI, Paragraph IWB-2412, "Inspection Program B," for the Millsto~*e Power Station, Unit No. 3 (MPS3). Inspection Program B requires volumetric examin.ation of essentially 100 percent of reactor pressure-retaining welds identified in Table i'WB-2500-1 once each 10-year interval. *Pursuant to Title 10 of the Code of Federal Regulations (10 CFR}, Part 50, Section 55a(z)(1 ), the licensee requested the use of a proposed alternative to extend the MPS3 reactor vessel third ISi interval from 10 years to 20 years, on the basis that the alternative provides an acceptable level of quality and safety.
2.0 REGULA TORY REQUIREMENTS ISi of the ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as a way to detect anomaly and degradation indications so that structural integrity of these components can be maintained. This is required by 10 CFR 50.55a(g}, except where specific relief has been granted by the U.S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6}(i}. The regulation in 10 CFR 50.55a(z) states that alternatives to the requirements of paragraphs (b) through (h) of 10 CFR .50.55a, or portions thereof, may be used when authorized by the Director of the Office *
.of Nuclear Reactor Regulation. A proposed altern'ative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that (1) the proposed alternative would provide an acceptable level of quality and s~fety, or (2) compliance with the specified requirements of this section would result in hardship or unusual difficulty~ without a compensating increase in the level of quality and safety.
Enclosure
Pursuant to 10 CFR 50.55a(g)(4), components (including supports) that are class-ified as ASME Code Class 1, 2, and 3 must meet the requirements, except design and access provisions and preservice examination requirements, as set forth in Section XI of editions and addenda of the ASME Code, which become -effective subsequent to editions specified in paragraphs (g)(2) and (3) of this section, to the extent practical, within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the successive 120-month inspection intervals (following the initial 120-month inspection interval) must comply with the requirements in the latest edition and addenda of the ASME Code, which was incorporated by reference in 10 CFR 50.55a(a), 12 months before the start of the 120-month interval (or the optional ASME Code Cases listed in NRC Regulatory Guide (RG) 1.147, Revision 17, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1,) subject to the conditions listed in 10 CFR 50.55a(b).
RG 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials, describes general procedures acceptable to the staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor pressure vessels (RPVs).
RG 1.174, Revision 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.
The licensee has requested relief .from ASME Code requirements pursuant to 10 CFR 50.55a(g)(6)(i). The MPS3 third 10-year ISi interval is based on the ASME Code,Section XI, 2004 Edition.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request, and the NRC to grant, the relief requested by the licensee. The end date for the current MPS3 third 10-year interval ISi program is April 22, 2019.
3.0 TECHNICAL EVALUATION
3.1 Background The ISi of Examination Categories B-A and B-D components consists of visual and ultrasonic examinations intended to discover whether new flaws have initiated, whether preexisting -flaws have extended, and whether preexisting 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. '
3.2 Summary of WCAP-16168-NP, Revision 2 By letter dated May 8, 2008 (ADAMS Accession No. ML081060051), the NRC staff issued a final safety evaluation (SE) (ADAMS Accession No. ML081060045), which found that Topical Report WCAP-16168-NP, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (the WCAP), was acceptable for referencing in licensing applications for pressurized-water reactors (PWRs) designed by Westinghouse Electric Company (Westinghouse); Combustion Engineering, Inc.; and Babcock & Wilcox, Inc. (B&W).
The WCAP was developed to support a risk-informed assessment of extensions to the ISi intervals for ASME Code,Section XI, Categories B-A and B-D components from 10 years to 20 years, using data from three different PWR plants (referred to as the pilot plants) representing each of the vendors.
The analyses in the WCAP used probabilistic fracture mechanics tools and input from the work described in 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 (PWROG) analyses incorporated the effects of fatigue crack growth and ISi data. Design-basis transient data was used as an input for the fatigue crack growth evaluation. The effects of ISi data were modeled consistently with the previously approved probabilistic fracture mechanics codes WCAP-14572, Revision 1-NP-A, 'Westinghouse Owners Group Application of Risk-Informed Methods to Piping lnservice Inspection Topical Report" (ADAMS Accession Nos. ML012630349 and ML012630327), and Supplement 1 to WCAP-14572, Revision 1-NP-A, "Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk Informed lnservice Inspection" (ADAMS Accession No. ML012630313). These effects were inputs into the evaluations performed with the "Fracture Analysis of Vessels - Oak Ridge" computer code. All other inputs were identical to those used in the PTS risk reevaluation underlying 10 CFR 50.61 a, "Alternative fracture toughness requirements for protection against pressurized thermal shock events."
The PWROG concluded, as a result of these studies, that the ASME Code,Section XI, 10-year ISi interval for Categories B-A and B-D components in PWR RPVs can be safely extended from 10 years to 20 years. This conclusion, based on the results from the pilot plant analyses, was considered to apply to any plant designed by the three PWR vendors represen_ted in the pilot plant study, as long as certain critical plant-specific criteria (defined in Appendix A of the WCAP) are bounded by the analysis for the applicable pilot plant.
3.3 Summary of NRC Staff Evaluation for WCAP-16168-NP, Revision 2 The NRC staff issued a revised SE dated July 26, 2011 (ADAMS Accession No. ML111600303), superseding the initial May 8, 2008, SE in the WCAP and addressing the PWROG's request for clarification of the information needed in applications utilizing the WCAP.
In this SE, the staff concluded that the methodology presented in the WCAP is consistent with the guidance provided in RG 1.174, Revision 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 specified in the SE. In addition to showing that the subject plant is bounded by the pilot plants/parameters identified in Appendix A in the WCAP, licensees
that submit a request for an alternative based on the WCAP should submit 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 (TWCFrnrAL) and its supporting material properties at the end of the period in which the relief is requested to extend the ISi from 1O years to 20 years. T~e 95th percentile total TWCF (TWCF9s-rorAL) must be calculated using the methodology in NU.REG-1874. The Radiographic Testing temperature-maximum (RTMAx-x) and the shift in the Charpy transition temperature produced by irradiation defined at the 30 toot-pound (ft-lb) energy level, LH30, must be calculated using the methodology documented in the latest revision of RG 1.99 or other NRG-approved methodology. RT MAx-x is the material property that characterizes the reactor vessel's resistance to fracture, initiating from flaws in plates (RT MAX-PL), forgings (RT MAx-rn), axial welds, and circumferential welds (RT MAx-Awtcw).
- 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 ISi of RPV welds and the proposed schedule for the next 20-year ISi interval. The 20-year inspection interval is a maximum interval. In its request for an alternative, each licensee shall identify the
. years in which future inspections will be performed. The dates provided must be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRG 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 of 12 heat-up/cool-down transients per year, which was used in the PWROG fatigue analysis, bounds the fatigue crack growth for all of its design-basis transients; and (b) identify the design-basis transients that contribute to significant fatigue crack*
growth.
- 5. Licensees with RPVs having forgings that are susceptible to underclad cracking and
- with RTMAx-Fo values exceeding 240 degrees Fahrenheit must submit a plant-specific evaluation to extend the inspection interval for ASME Code,Section XI, Categories B-A and B-D RPV welds from 1O years to a maximum of 20 years because the analyses performed in the WCAP 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, Revision 3, which contains a copy of the NRG staff's January 26, 2011, SE of the WCAP was issued in October 2011 (ADAMS Accession No. ML11306A084) and is hereafter referred to as the WCAP-A.
3.4 Description of Proposed Alternatives The licensee proposes to defer the ASME Code required Categories B-A and B-D weld ISi for MPS3 until 2027. The proposed ISi date is consistent with OG-10-238 which is the latest updated implementation plan for the PW ROG plants (ADAMS Accession No. ML11153A033).
3.5 Components for Which Relief is Requested The affected component is the MPS3 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:
Examination Category Item Number Description B-A B1 .11 Circumferential Shell Welds B-A B1 .12 Longitudinal 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-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section 3.6 Basis for Proposed Alternative The basis for the proposed alternative is WCAP-A. Plant-specific parameters for MPS3 are summarized in the attachment to the licensee's letter dated March 9, 2015. The format of the information is patterned after that found in Appendix A of the WCAP-A. Table 1 of the attachment lists the critical parameters identified in the WCAP-A and compares the results of the Westinghouse pilot plant to those at MPS3.
3.7 Duration of Proposed Alternative The licensee's proposed alternative is applicable to the MPS3 ISi program for the third and fourth 10-year inspection intervals.
3.8 NRC Staff Evaluation The NRC reviewed the licensee's proposal to extend the MPS3 ISi interval in order to determine whether the licensee met the risk-informed criteria set forth in the WCAP-A for a Westinghouse plant. By showing that MPS3 is bounded by the Westinghouse pilot plant analysis with respect to the six criteria discussed in Section 3.3 of this SE, the licensee would have a sufficient technical basis for extending the ISi in accordance with the provisions of the WCAP-A. The MPS3 RPV has a single layer cladding and is bounded by the Westinghouse pilot plant basis.
The licensee stated that two complete 10-year ISls have been performed on MPS3. Ten indications were identified in the beltline region during the most recently completed ISi and were found in a nozzle shell longitudinal weld seam, two intermediate shell lqngitudinal weld seams, and a lower shell longitudinal weld seam. Each of the indications was found to be acceptable
per Table IWB-3510-1 of the ASME Code,Section XI, considering they were neither within the inner 1/10'h of the reactor vessel thickness nor deeper than 1 inch from the clad-base metal interface, making the indication acceptable per the requirements of 10 CFR 50.61 a.
The MPS3 third ISi of the reactor vessel full penetration pressure-retaining Categories B-A and B-0 welds is scheduled for 2016, preceding the end of the third 10-year ISi interval scheduled for April 22, 2019. The licensee requested to extend the end of the interval to 2029 and proposed to perform the third examination of the welds in 2027, plus or minus one refueling outage. The proposed date is consistent with PWROG letter OG-10-238, and the NRC staff finds the date acceptable, so long as the examination is completed prior to April 22, 2029.
Table 3 of the licensee's submittal provided the TWCF of the limiting axial weld, plate, and circumferential weld, as well as the critical parameters needed to perform the calculations. In order to calculate the shift in the Charpy transition temperature produced by irradiation defined at the 30 ft-lb energy level, Lffso, the licensee used the methodology provided in RG 1.99, Revision 2. The licensee reported that the TWCF9s-TOTAL for MPS3 was 1.44 x 10-3 , which is well below the Westinghouse pilot plant bounding value of 1. 76 x 1o-a year. The NRC staff performed independent calculations, which verified the results reported by the licensee; therefore, the staff finds the TWCF9s-TOTAL for MPS3 acceptable.
With regard to the frequency and severity of design-basis transients, the licensee was required to show that MPS3 has a number of heatup/cooldown transients bounded by that of the Westinghouse pilot plant basis (seven heatup/cooldown cycles per year). Table 4.3-2 of the MPS3 license renewal application shows that there had been 34 and 33 respective heatups and cooldowns between January 31, 1986, and April 30, 2003, correlating to two cycles per year.
The NRC staff agrees' that the frequency of the limiting design-basis transients during prior plant operation is less than the frequency of the Westinghouse design-basis transients identified in the PWROG fatigue analysis.
In summary, the licensee's submittal demonstrates that the RPV for MPS3 is bounded by the Westinghouse limitations set forth in the WCAP-A and the associated SE by the NRC staff. The licensee adequately confirmed that the MPS3 RPV meets the applicable criteria set forth in the WCAP-A. Since MPS3 is a Westinghouse plant, the fourth criterion shown in Section 3.3 of this SE related to B&W plants is not applicable. Furthermore, the licensee did not report any forgings that are susceptible to underclad cracking; thus, the fifth criterion is also not applicable to this plant. Lastly, the licensee is not curre'ntly seeking additional interval extensions; therefore, the sixth and final criterion is not applicable.
4.0 CONCLUSION
The NRC staff has completed its review of the licensee's submittal for an alternative ISi extension to allow use of alternate reactor inspection interval requirements for MPS3. The staff concludes that extending the third ISi interval for Categories 8-A and B-0 components from 1O years to 20 years will not result in any considerable increase in risk. This conclusion is based on the fact that the MPS3 RPV is bounded by the WCAP-A, and the request met all of the provisions set forth in the WCAP-A, and as described in the NRC staff's January 26, 2011, SE of the WCAP. Therefore, the licensee's proposed Aternative Request IR-3-27 provides an acceptable level of quality and safety. Pursuant to 10 CFR 50.55a(z)(1 ), the staff concludes that the licensee's alternative ISi sch.edule for the specified welds is acceptable for extension to
0 April 22, 2029. The examination of the Categories B-A and B-D components for MPS3 shall be conducted prior to the end of the extended third interval. All other. requirements of the ASME Code,Section XI, not specifically included in the request for the proposed alternatives, remain in effect.
Principal Contributor: Austin Young Date: February 16, 2016
D. Heacock If you have any questions, please contact the project manager, Richard Guzman, at (301) 415-1030 or Richard.Guzman@nrc.gov.
Sincerely,
/RA/
Travis L. Tate, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423
Enclosure:
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