ML110330085
| ML110330085 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 01/27/2011 |
| From: | Pope A Progress Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RA-11-002 | |
| Download: ML110330085 (8) | |
Text
Progress Energy 10 CFR PO Box 1551 411 Fayetteville Street Mall Raleigh NC 27602 Serial: RA- 11-002 January 27, 2011 United States Nuclear Regulatory Commission ATTENTION:
Document Control Desk Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 / LICENSE NO. DPR-23
SUBJECT:
REQUEST FOR RELIEF FROM ASME BOILER AND PRESSURE VESSEL CODE, SECTION XI, FOR THE FOURTH TEN-YEAR INSERVICE INSPECTION PROGRAM INTERVAL (RELIEF REQUEST NO. RR-23)Ladies and Gentlemen:
Pursuant to 10 CFR 50.55a(a)(3)(i), Carolina Power and Light Company (CP&L), now doing business as Progress Energy Carolinas (PEC), Inc., hereby requests NRC approval of a proposed alternative to the interval requirements for the reactor vessel Inservice Inspection (ISI). Specifically, the alternative would extend the ISI intervals for reactor vessel welds (Examination Category B-A) and nozzle-to-vessel welds (Examination Category B-D) from 10 years to 20 years. The proposed alternative is applicable to the 4 th ISI interval.The details of the 10 CFR 50.55a request are attached.
Approval is requested by July 31, 2011. The requested approval date would allow PEC the opportunity to implement a contingency plan to perform the required reactor vessel examination during the next H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, refueling outage scheduled for early 2012, if necessary.
This letter contains no regulatory commitments.
If you have any questions or require additional information, please contact Dana Covill at (919) 546-2631.Sincerely, ette Pope Acting Manager -Nuclear Regulatory Affairs DBM Attachment c: USNRC Region II USNRC Resident Inspector
-HBRSEP, Unit No. 2 B. Mozafari, NRR Project Manager -HBRSEP, Unit No. 2 A -'-
United States Nuclear Regulatory Commission RA- 11-002 Page 2 bc: Bob Duncan Chris Kamilaris Jeff Colbom Kevin Riley Dana Covill Kelvin Henderson Annette Pope Curt Castell (For RNP Licensing/Nuclear Records Files)Ted Huminski Michael Blew File: (Corporate)
United States Nuclear Regulatory Commission Attachment to RA- 11-002 Page 1 of 6 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 REQUEST FOR RELIEF FROM ASME BOILER AND PRESSURE VESSEL CODE, SECTION XI (RELIEF REQUEST NO. RR-23)1. ASME Code Component(s)
Affected The affected component is the reactor vessel, specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI (Reference
- 1) examination categories and item numbers covering examinations of the reactor vessel (RV). These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV, Code,Section XI.Examination Category Item No. Description B-A B 1.11 Circumferential Shell Welds B-A B 1.12 Longitudinal Shell Welds B-A B 1.21 Circumferential Head Welds B-A B 1.22 Meridional Shell Welds B-A BI.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inner Radius Areas 2. Applicable Code Edition and Addenda The Fourth Ten-Year Interval Inservice Inspection (ISI) Program Plan is prepared to American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 1995 Edition through the 1996 Addenda. (Reference 1). Throughout the remainder of this request, the ASME BPV Code,Section XI, is referred to as "the Code." 3. Applicable Code Requirement IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of reactor vessel pressure retaining welds identified in Table IWB-2500-1 once each ten year interval.
The H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, fourth 10-year inservice inspection interval is scheduled to end in 2012.4. Reason for Request An alternative is requested from the requirement of IWB-2412, Inspection Program B, that volumetric examination of essentially 100% of reactor vessel pressure retaining, Examination Category B-A and B-D welds, be performed once each ten-year interval.
Extension of the inspection interval for Examination Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in occupational radiation exposure and examination costs.5. Proposed Alternative and Basis for Use Progress Energy Carolinas (PEC), Inc. proposes to defer the Code required volumetric examination of the HBRSEP, Unit No. 2, reactor vessel full penetration pressure retaining Category B-A and B-D welds for the fourth inservice inspection, currently scheduled for 2012, until 2021 plus or minus one refueling outage.These dates are a slight change from the information provided to the Staff in PWR Owners Group letter OG-06-356 (Reference 2). Due to an extended outage in 2010, the HBRSEP, Unit No. 2, refueling outage United States Nuclear Regulatory Commission Attachment to RA- 11-002 Page 2 of 6 originally scheduled for 2011 has been moved to 2012. Therefore, the refueling outage originally planned for 2020 will move to 2021. PWR Owners Group letter OG-06-356 (Reference
- 2) indicates that six exams were planned in the industry for 2020 and one exam was planned for 202 1. With the HBRSEP, Unit No. 2, refueling outage moving to 2021, there would now be five exams planned in the industry for 2020 and two exams planned for 2021 thus providing a more uniform yearly sampling of exams.In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current inspection interval can be extended based on a negligible change in risk as specified in Regulatory Guide 1.174 (Reference 3).The methodology used to demonstrate the acceptability of extending the inspection intervals for Category B-A and B-D welds based on a negligible change in risk is contained in WCAP-16168-NP-A, Revision 2 (Reference 4). This methodology was used to develop a pilot plant analysis for reactor vessels in Westinghouse (W), Combustion Engineering (CE), and Babcock and Wilcox (B&W) plant designs and is an extension of the work that was performed as part of the NRC Pressurized Thermal Shock Risk Study (Reference 5). The critical parameters for demonstrating that this pilot plant analysis is applicable on a plant specific basis, as identified in Appendix A of WCAP-16168-NP-A, Revision 2, are identified in Table 1. By demonstrating that each plant specific parameter is bounded by the corresponding pilot plant parameter, the application of the WCAP risk results to the HBRSEP, Unit No. 2 reactor vessel is deemed to be acceptable as shown in Table I below.Table 1 Critical Parameters for Application of Bounding Analysis for HBRSEP, Unit No. 2 Additional Evaluation Parameter Pilot Plant Basis Plant Specific Basis Required?Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization No (PTS) Transients in the NRC PTS Risk (Reference
- 5) Study (Reference 6)Study are applicable Through Wall Cracking Frequency 1.76E-08 events per year 1.34E-10 events per No (TWCF) (Reference
- 4) year (Calculated per Reference 5)Frequency and Severity of Design Basis 7 heatup/cooldowns per year Bounded by 7 No Transients (Reference
- 4) heatup/cooldowns per year Cladding Layers (Single/Multiple)
Single Layer (Reference
- 4) Single Layer
- No* Single layer is assumed since it is more conservative.
This is explained in WCAP-16168 (more layers means more attenuation, which means less of an adverse effect)
United States Nuclear Regulatory Commission Attachment to RA- 11-002 Page 3 of 6 Additional information relative to the HBRSEP, Unit No. 2 reactor vessel inspection is provided in Table 2. This information confirms that satisfactory examinations have been performed on the HIBRSEP, Unit No. 2 reactor vessel.Table 2 Additional Information Pertaining to Reactor Vessel Inspection for HBRSEP, Unit No. 2 Inspection methodology:
The most recent inservice inspection in 2001 of the Category B-A and B-D welds was performed to ASME Code,Section XI, Appendix VIII requirements.
Future inservice inspections will continue be performed to this standard.Number of past inspections:
Three 10-Year inservice inspections have been performed.
Number of indications found: Of the two recordable indications found in the beltline welds during the most recent inservice inspection, only one indication was in the beltline region. This indication was within the inner 3/8h of the vessel thickness and was acceptable per Table IWB-3510-1 of Section XI of the ASME Code. There are no indications within the inner 1 inch or 1/10th thickness of the reactor vessel beltline region. Therefore, evaluation of the flaw limits in the Alternate PTS Rule (Reference
- 7) is not required.Proposed inspection schedule for The fourth inservice inspection is scheduled for 2012. This inspection will be balance of plant life: performed in 2021 within +/- one refueling outage. (The H. B. Robinson Unit 2 operating license has been extended to 2030 per Reference 8.) This schedule represents a slight change, as discussed above, to the current ISI schedule documented in OG-06-356 (Reference
- 2) and also the revised ISI implementation plan documented in OG-09-454 (Reference 13).
United States Nuclear Regulatory Commission Attachment to RA-1 1-002 Page 4 of 6 Table 3 provides additional information relative to the calculation of the TWCF for HBRSEP, Unit No. 2.WCAP-15827, Revision 0, WCAP-15805, Revision 0 and the HBR2-UFSAR (References 9, 10 and 11, respectively) were used as the information sources for the inputs listed in Table 3 for HBRSEP, Unit No. 2.Table 3 Details of TWCF Calculation for 50 EFPY of Operation for HBRSEP, Unit No. 2 Inputs Reactor Coolant System Temperature, TRcs[°F1:
I N/A Twa 1 [inches]:
10.10 Material or Cu Ni R.G. Fluence [1019 Region & Component Description Flux Type [wt%] [wt%] 1.99 CF RTNDT(u) [°F] Neutron/cm 2 , E>1 Pos. ['F] MeV]1 Upper Shell Plate (W10201-1)
A302A 0.13 0.11 1.1 62.9 69 2.50 2 Upper Shell Plate (W10201-2)
A302A 0.15 0.25 1.1 84.8 30 2.50 3 Upper Shell Plate (W10201-3)
A302A 0.11 0.08 1.1 51.8 36 2.50 4 Inter. Shell Plate (W10201-4)
A302A 0.12 0.09 2.1 67.1 20 6.00 5 Inter. Shell Plate (W10201-5)
A302A 0.10 0.12 2.1 38.8 20 6.00 6 Inter. Shell Plate (W10201-6)
A302A 0.09 0.09 2.1 45.9 45 6.00 7 Lower Shell Plate (W9807-3)
A302A 0.12 0.10 1.1 58.0 50 2.05 8 Lower Shell Plate (W9807-5)
A302A 0.15 0.10 1.1 70.5 33 2.05 9 Lower Shell Plate (W9807-9)
A302A 0.14 0.15 1.1 70.5 9 2.05 10 Upper Ax. Weld (1-273A) ARCOS1B5 0.22 0.05 1.1 100.8 -56 1.06 11 Upper Ax. Weld (1-273B) ARCOS1B5 0.22 0.05 1.1 100.8 -56 1.85 12 Upper Ax. Weld (1-273C) ARCOS1B5 0.22 0.05 1.1 100.8 -56 0.61 13 Inter. Ax. Weld (2-273A) ARCOS B5 0.22 0.05 1.1 100.8 -56 4.45 14 Inter. Ax. Weld (2-273B) ARCOS B5 0.22 0.05 1.1 100.8 -56 1.47 15 Inter. Ax. Weld (2-273C) ARCOS B5 0.22 0.05 1.1 100.8 -56 2.58 16 Lower Ax. Weld (3-273A) ARCOS B5 0.22 0.05 1.1 100.8 -56 2.05 17 Lower Ax. Weld (3-273B) ARCOS B5 0.22 0.05 1.1 100.8 -56 1.66 18 Lower Ax. Weld (3-273C) ARCOSIB5 0.22 0.05 1.1 100.8 -56 1.66 19 US -IS Circ. Weld (10-273)*
LINDE 1092 0.21 1.01 1.1 230.2 -56 2.50 20 IS -LS Circ. Weld (11-273)*
LINDE 1092 0.19 0.98 1.1 217.1 -77 2.05*Note: US, IS, and LS are Upper Shell, Intermediate Shell and Lower Shell, respectively.
U.United States Nuclear Regulatory Commission Attachment to RA- 11-002 Page 5 of 6 Table 3 (cont.) Details of TWCF Calculation for 50 EFPY of Operation for H. B. Robinson, Unit 2 Outputs Methodology Used to Calculate AT 3 0: Regulatory Guide 1.99, Revision 2 (Reference 12)Controlling RTmA.XXX Fluence [ 1019 Maei2Rein#
RT -xFluence A 3 0] TC 9.0 Neutron/cm2, AT30 [OF] TWCF95FXX (From Above) E>1 MeV F Axial Weld -AW 1 602.20 1.85 1.169 73.15 1.43E-17 Circumferential Weld -CW 19 690.57 2.50 1.246 286.88 4.111E-l 1 Plate -PL 1 607.08 2.50 1.246 78.39 1.92E- 11 TWCF95-TOTAL (C(AwTWCF95.AW
+ OQPLTWCF 9 5.PL + ctcwTWCF 9 5-cw): 1.34E-10 6. Duration of Proposed Alternative This request is applicable to the HBRSEP, Unit No. 2 inservice inspection program for the remainder of the 60-year extended operating license (Reference 8).7. Precedents Several requests for relief on this subject have been granted, including:
Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) -Relief Request for Extension of the Reactor Vessel Inservice Inspection Date to the Year 2020 (Plus or Minus One Outage) (TAC No. ME30 10), July 12, 2010 McGuire Nuclear Station, Unit 1 (McGuire 1) -Relief Request 09-MN-003 for Extension of the Reactor Vessel Inservice Inspection (ISI) Date to the Year 2020 (Plus or Minus One Outage) (TAC No.ME1822) June 28, 2010 R. E. Ginna Nuclear Power Plant: Safety Evaluation for Relief Request No. 18, Reactor Vessel Weld Examination Extension (TAC No. MD9962) July 31, 2009 8. References
- 1. ASME Boiler and Pressure Vessel Code,Section XI, 1995 Edition through the 1996 Addenda, American Society of Mechanical Engineers, New York.2. OG-06-356, "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." MUFIP 5097-99, Task 2059," October 31, 2006.3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002.4. WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," June 2008.5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," March, 2007.6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004.
United States Nuclear Regulatory Commission Attachment to RA-1 1-002 Page 6 of 6 7. Code of Federal Regulations, 10 CFR Part 50.6 1 a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events." 8. NUREG-1785, "Safety Evaluation Report Related to the License Renewal of H. B. Robinson Steam Electric Plant, Unit 2 (Docket No. 50-26 1) Carolina Power & Light Company", March 2004.9. WCAP-15827, Revision 0, "H. B. Robinson Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," March 2003.10. WCAP-15805, Revision 0, "Analysis of Capsule X from the Carolina Power and Light Company H. B.Robinson Unit 2 Reactor Vessel Radiation Surveillance Program," March 2002.11. HBR2-UFSAR, Revision 23, "H. B. Robinson Unit 2, Updated Final Safety Analysis Report," Carolina Power & Light Company.12. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.13. OG-09-454, "Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor vessel In-Service Inspection Interval." PA-MSC-0120," December 1, 2009.