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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20210P9181999-08-10010 August 1999 Safety Evaluation Authorizing Request for Reliefs CIP-01,02, 06,07,08,09,10 & 11 (with Certain Exceptions) & 12-18,for Second 10-year ISI Interval.Request CIP-04 & 05 Would Result in hardship,CIP-03 Not Required & CIP-11 Denied in Part ML20210P9441999-08-10010 August 1999 Safety Evaluation Accepting Licensee Assessment of Impact on Operation of Plant,Unit 1,with Crack Indications of 2.11, 6.36 & 1.74 Inches in Three Separate Jet Pump Risers ML20210N2341999-08-0505 August 1999 SER Accepting Response to NRC GL 87-02, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Unresolved Safety Issues (USI) A-46 ML20206G1871999-05-0404 May 1999 Safety Evaluation Approving Third 10-year ISI Program Requests for Relief (RR) RR-08,RR-15 & RR-17 ML20205F9031999-03-30030 March 1999 Safety Evaluation Supporting Proposed Rev to BSEP RERP to Licenses DPR-62 & DPR-71,respectively ML20203D7061999-02-0909 February 1999 SER Accepting Proposed Alternatives Contained in Relief Requests PRR-04,VRR-04,VRR-13,PRR-01,PRR-03,VRR-01.VRR-07, VRR-08 & VRR-09 Denied ML20154P8151998-10-16016 October 1998 SER Accepting Revised Safety Analysis of Operational Transient of 920117,for Plant,Unit 1 ML20154P8591998-10-16016 October 1998 SER Accepting Equivalent Margins Analysis for N-16A/B Instrument Nozzles for Plant,Units 1 & 2 ML20217K8461998-04-24024 April 1998 Safety Evaluation Approving Proposed Use of Code Case N-535 at Brunswick Unit 1 During Second 10-yr Interval,Pursuant to 10CFR50.55a(a)(3)(i).Authorizes Use of Code Case N-535 Until Code Case Included in Future Rev of RG 1.147 ML20217K3941998-04-24024 April 1998 SER Approving Relief Request for Pump Vibration Monitoring, Brunswick Steam Electric Plant,Units 1 & 2 ML20217E6841998-04-23023 April 1998 Safety Evaluation Accepting Code Case N-547, Alternative Exam Requirements for Pressure Retaining Bolting of CRD Housings ML20217E7471998-04-21021 April 1998 Safety Evaluation Accepting Alternative to Insp of Reactor Pressure Vessel Circumferential Welds ML20217B5241998-04-20020 April 1998 SE Accepting Licensee Request for Approval to Use Alternative Exam Requirement for Brunswick,Unit 1,reactor Vessel Stud & Bushing During Second 10-yr ISI Interval Per 10CFR50.55a(a)(3)(ii) ML20216B1041998-03-0404 March 1998 SER Approving Alternative to Insp of Reactor Pressure Vessel Circumferential Welds for Brunswick Steam Electric Plant, Unit 1 ML20198J0921997-09-18018 September 1997 Safety Evaluation Authorizing Licensee & Suppls & 16 Request for Approval of Alternative Reactor Vessel Weld Exam,Per 10CFR50.55a(g)(6)(ii)(A)(5) for Plant, Unit 2 for Next 2 Operating Cycles ML20198H2351997-09-0808 September 1997 Safety Evaluation Approving Licensee 970311 Request for Use of ASME Code Case N-509 & Relief from ASME Code Section IX Requirements for Exam of Hpcip Studs for Plant,Units 1 & 2 ML20137A4831997-03-18018 March 1997 SER Re CP&L Review of Power Uprate Process & Commitment Preventing Operation at Uprated Power Levels for Plant, Units 1 & 2 ML20129E0821996-09-26026 September 1996 Safety Evaluation Supporting Request to Use Certain Portions of Later Edition of ASME Code for Inservice Leakage Testing Valves for Brunswick Steam Electric Plant Units 1 & 2 ML20056D6761993-07-28028 July 1993 Safety Evaluation Concluding That Interior Masonry Walls May Be Downgraded to non-fire Related ML20128K7711993-02-11011 February 1993 Safety Evaluation Granting Relief from Certain Inservice Testing Program Requirements for Several Pumps & Valves ML20198E5081992-11-23023 November 1992 Safety Evaluation Accepting Licensee 120-day Response to Suppl 1 to GL 87-02 ML20246D6811989-08-18018 August 1989 Safety Evaluation Supporting Installation & Design of Nitrogen Pneumatic Sys,Per Generic Ltr 84-09,by Adding New Check Valves to Existing Drywell Noninterruptible Instrument Air Lines ML20246C4201989-06-27027 June 1989 SER Accepting Util Response to Generic Ltr 83-28,Item 4.5.3 Re Reactor Trip Sys Reliability for All Operating Reactors ML20247P6201989-06-0101 June 1989 Safety Evaluation Supporting Util SAFER/GESTR-LOCA Analysis ML20247M1911989-05-25025 May 1989 Safety Evaluation Re Denial of Amend Request to Licenses DPR-71 & DPR-62 ML20246P9401989-05-10010 May 1989 Safety Evaluation Accepting Plant Second 10-yr Interval Inservice Insp Program ML20246J5531989-05-0909 May 1989 Safety Evaluation Concluding That Plant Can Be Safely Operated for Another 18-month Fuel Cycle in Configuration Following Reload 5,per Improvements,Insps & Repairs to Plant IGSCC ML20245D3761989-04-25025 April 1989 Safety Evaluation Supporting Licensee IGSCC Program for Refuel 7 Outage ML20236D5481989-03-17017 March 1989 Safety Evaluation Accepting Util Response to Generic Ltr 83-28,Item 4.5.2 Re Reactor Trip Sys Reliability ML20236D5381989-03-17017 March 1989 Safety Evaluation Accepting Util 831107 Response to Generic Ltr 83-28,Item 2.2.1 Re Equipment Classification Programs for safety-related Components ML20236D4641989-03-15015 March 1989 Safety Evaluation Re Generic Ltr 83-28,Item 2.1 (Parts 1 & 2) Concerning Equipment Classification & Vendor Interface for Reactor Trip Sys Components ML20235Z2841989-03-0808 March 1989 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Items 3.2.1 & 3.2.2 ML20235Z3451989-03-0808 March 1989 Safety Evaluation Supporting Util Compliance W/Atws Rule, 10CFR50.62 Re Power Testability Features of Alternate Rod Insertion Sys & Recirculating Pump Trip Design ML20235M5771989-02-16016 February 1989 Safety Evaluation Supporting Control Room Habitability Sys of Plant & Acceptability of Existing Tech Spec Re Control Room Pressurization Requirement ML20147G0661988-03-0202 March 1988 Safety Evaluation Supporting Proposed Functional Testing Plan for Snubbers ML20236D1311987-10-22022 October 1987 Safety Evaluation Re Util Request for Relief from Schedular Requirements for Performance of Visual Insp & Hydrostatic Test of CRD Withdrawal & Insert Lines.Granting of Request Recommended ML20235A7331987-09-18018 September 1987 Safety Evaluation Re Installation of Alternate Rod Injection (ARI) Sys & Adequacy of Plant Reactor Coolant Recirculating Pump Trip (RPT) Sys,In Compliance W/Atws Rule 10CFR50.62. ARI & RPT Acceptable NUREG-0661, Safety Evaluation Re Util 840831 Submittal of Addendum to Plant Unique Analysis Rept on Mark I Containment Mod Program.Safety/Relief Valve Load Cases C3.2 & C3.3 Adequately Addressed & Resolved1987-05-0707 May 1987 Safety Evaluation Re Util 840831 Submittal of Addendum to Plant Unique Analysis Rept on Mark I Containment Mod Program.Safety/Relief Valve Load Cases C3.2 & C3.3 Adequately Addressed & Resolved ML20212H5671987-01-16016 January 1987 Safety Evaluation Supporting Util Response to Generic Ltr 83-08 Re Restoring Safety Margins of Vacuum Breakers by Replacing Critical Parts W/Adequate Matls ML20207A8531986-11-0505 November 1986 Safety Evaluation Supporting Operation for Full Fuel Cycle W/O mid-cycle Insp for Crack Growth ML20215N3771986-10-30030 October 1986 Safety Evaluation Re Util 860320 Response to Generic Ltr 84-09, Recombiner Capability Requirements of 10CFR50.44(c)(3)(ii). Licensee Should Remove All Potential Oxygen Sources from Containments ML20203N0081986-09-17017 September 1986 Safety Evaluation Supporting Util 850919 Request for Relief from Installing Excess Flow Switch & Automatic Shutoff Valve in Diesel Fire Pump Fuel Line to Provide Protection in Event of Fuel Line Rupture ML20212N0201986-08-22022 August 1986 Safety Evaluation Denying Util 860325 Request for Relief from Inservice Insp Requirements of ASME Code Section XI, Table IWC-2500-1 for Volumetric Exam of HPCI Pump Studs ML20211G6081986-06-12012 June 1986 Safety Evaluation Supporting IGSCC Insp,Repair & Replacement Program During Dec 1985 Refueling Outage ML20205S2541986-06-0404 June 1986 Safety Evaluation Accepting Rev 2 to Nuclear Const Issues Group Spec 1, Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants, for non-ASME Code Welds ML20211A8281986-06-0303 June 1986 Safety Evaluation Re Rev 4 to Offsite Dose Calculation Manual.Rev Acceptable ML20198S7281986-05-29029 May 1986 Safety Evaluation Supporting 851203 Proposal to Modify Tech Spec 3/4.5.3 to Clarify Min Amount of Condensate Storage Tank Water Required to Ensure Operability of Core Spray Sys During Operating Conditions 4 or 5.Rev to Tech Specs Encl ML20133N4141985-10-23023 October 1985 Safety Evaluation Re Util 831107 & 850828 Responses to Generic Ltr 83-28,Items 3.1.2 & 3.2.1 & 850701 Request for Addl Info.Responses Re Vendor & Engineering Test Guidance & Testing Requirements After Maint Acceptable ML20134P5211985-08-28028 August 1985 Safety Evaluation Approving Use of ASME Code Case N-411 for Damping Curves ML20128M2911985-07-16016 July 1985 Safety Evaluation Supporting Licensee Response to Generic Ltr 83-28, Salem ATWS Event, Items 3.1.3 & 3.2.3 Re post-maint Testing 1999-08-05
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217N3271999-10-21021 October 1999 Part 21 Rept Re non-linear Oxygen Readings with Two (2) Model 225 CMA-X Containment Monitoring Sys at Bsep.Caused by High Gain Produced by 10K Resistor Across Second Stage Amplifier.Engineering Drawings Will Be Revised BSEP-99-0168, Monthly Operating Repts for Sept 1999 for Bsep,Units 1 & 2. with1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Bsep,Units 1 & 2. with ML20212D0431999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Brunswick Steam Electric Plant,Units 1 & 2 ML20210P9441999-08-10010 August 1999 Safety Evaluation Accepting Licensee Assessment of Impact on Operation of Plant,Unit 1,with Crack Indications of 2.11, 6.36 & 1.74 Inches in Three Separate Jet Pump Risers ML20210P9181999-08-10010 August 1999 Safety Evaluation Authorizing Request for Reliefs CIP-01,02, 06,07,08,09,10 & 11 (with Certain Exceptions) & 12-18,for Second 10-year ISI Interval.Request CIP-04 & 05 Would Result in hardship,CIP-03 Not Required & CIP-11 Denied in Part ML20210N2341999-08-0505 August 1999 SER Accepting Response to NRC GL 87-02, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Unresolved Safety Issues (USI) A-46 ML20210R1191999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Bsep,Units 1 & 2 ML20210R1311999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Bsep,Unit 2 BSEP-99-0118, Monthly Operating Repts for June 1999 for Bsep,Units 1 & 2. with1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Bsep,Units 1 & 2. with BSEP-99-0095, Monthly Operating Repts for May 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20210M8581999-05-14014 May 1999 B214R1 RPV Hydrotest Bolted Connection Corrective Action Evaluation, Rev 0 ML20211L3711999-05-10010 May 1999 Rev 0 to ESR 98-00333, Unit 2 Invessel Feedwater Sparger Evaluation ML20206G1871999-05-0404 May 1999 Safety Evaluation Approving Third 10-year ISI Program Requests for Relief (RR) RR-08,RR-15 & RR-17 BSEP-99-0075, Monthly Operating Repts for Apr 1999 for Brunswick Steam Electric Plant,Unit 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Brunswick Steam Electric Plant,Unit 1 & 2.With ML20206N1791999-04-23023 April 1999 Rev 0 to 2B21-0554, Brunswick Unit 2,Cycle 14 Colr BSEP-99-0059, Monthly Operating Repts for Mar 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20205F9031999-03-30030 March 1999 Safety Evaluation Supporting Proposed Rev to BSEP RERP to Licenses DPR-62 & DPR-71,respectively ML20206N1831999-02-28028 February 1999 Rev 0 to Suppl Reload Licensing Rept for Bsep,Unit 2 Reload 13 Cycle 14 BSEP-99-0043, Monthly Operating Repts for Feb 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20203D7061999-02-0909 February 1999 SER Accepting Proposed Alternatives Contained in Relief Requests PRR-04,VRR-04,VRR-13,PRR-01,PRR-03,VRR-01.VRR-07, VRR-08 & VRR-09 Denied BSEP-99-0005, Monthly Operating Repts for Dec 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With BSEP-98-0231, Monthly Operating Repts for Nov 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With BSEP-98-0218, Monthly Operating Repts for Oct 1998 for Bsep,Units 1 & 2. with1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Bsep,Units 1 & 2. with BSEP-98-0210, Special Rept:On 980824,temp Element 2-CAC-TE-1258-22 Failed. Cause of Failed Temp Element Cannot Be Conclusively Determined.Temp Element Will Be Replaced & Cable Connections Repaired1998-10-30030 October 1998 Special Rept:On 980824,temp Element 2-CAC-TE-1258-22 Failed. Cause of Failed Temp Element Cannot Be Conclusively Determined.Temp Element Will Be Replaced & Cable Connections Repaired ML20154P8151998-10-16016 October 1998 SER Accepting Revised Safety Analysis of Operational Transient of 920117,for Plant,Unit 1 ML20154P8591998-10-16016 October 1998 SER Accepting Equivalent Margins Analysis for N-16A/B Instrument Nozzles for Plant,Units 1 & 2 BSEP-98-0202, Monthly Operating Repts for Sept 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20151Y6211998-09-14014 September 1998 BSEP Rept Describing Changes,Tests & Experiments, for Bsep,Units 1 & 2 ML20151Y6371998-09-14014 September 1998 Changes to QA Program, for Bsep,Units 1 & 2 BSEP-98-0185, Monthly Operating Repts for Aug 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With1998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Brunswick Steam Electric Plant,Units 1 & 2.With ML20151T5021998-08-0505 August 1998 Project Implementation Plan, Ngg Yr 2000 Readiness Program, Rev 2 BSEP-98-0164, Monthly Operating Repts for July 1998 for BSEP Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for BSEP Units 1 & 2 ML20236T1961998-07-0101 July 1998 Rev 1 to 2B21-0088, Brunswick Unit 2,Cycle 13 Colr ML20236T1921998-07-0101 July 1998 Rev 1 to 1B21-0537, Brunswick Unit 1,Cycle 12 Colr BSEP-98-0142, Monthly Operating Repts for June 1998 for BSEP Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for BSEP Units 1 & 2 ML20236T1971998-06-30030 June 1998 Rev 2 to 24A5412, Supplemental Reload Licensing Rept for Brunswick Steam Electric Plant Unit 2 Reload 12 Cycle 13 ML20249B9691998-06-11011 June 1998 Rev 1 to VC44.F02, Brunswick Steam Electric Plant,Units 1 & 2,ECCS Suction Strainers Replacement Project,Nrc Bulletin 96-003 Final Rept BSEP-98-0129, Monthly Operating Repts for May 1998 for Bsep,Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Bsep,Units 1 & 2 ML20151S9041998-05-31031 May 1998 Revised Pages to Monthly Operating Rept for May 1998 for Brunswick Steam Electric Plant,Unit 1 BSEP-98-0104, Monthly Operating Repts for Apr 1998 for Brunswick Steam Electric Plant,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Brunswick Steam Electric Plant,Units 1 & 2 ML20151S8991998-04-30030 April 1998 Revised Pages to Monthly Operating Rept for Apr 1998 for Brunswick Steam Electric Plant,Unit 1 ML20247N7501998-04-30030 April 1998 Rev 0 to BSEP Unit 1,Cycle 12 Colr ML20247N7721998-04-30030 April 1998 Rev 0 to J1103244SRLR, Supplemental Reload Licensing Rept for BSEP Unit 1,Reload 11,Cycle 12 ML20217K8461998-04-24024 April 1998 Safety Evaluation Approving Proposed Use of Code Case N-535 at Brunswick Unit 1 During Second 10-yr Interval,Pursuant to 10CFR50.55a(a)(3)(i).Authorizes Use of Code Case N-535 Until Code Case Included in Future Rev of RG 1.147 ML20217K3941998-04-24024 April 1998 SER Approving Relief Request for Pump Vibration Monitoring, Brunswick Steam Electric Plant,Units 1 & 2 ML20217E6841998-04-23023 April 1998 Safety Evaluation Accepting Code Case N-547, Alternative Exam Requirements for Pressure Retaining Bolting of CRD Housings ML20217E7471998-04-21021 April 1998 Safety Evaluation Accepting Alternative to Insp of Reactor Pressure Vessel Circumferential Welds ML20217B5241998-04-20020 April 1998 SE Accepting Licensee Request for Approval to Use Alternative Exam Requirement for Brunswick,Unit 1,reactor Vessel Stud & Bushing During Second 10-yr ISI Interval Per 10CFR50.55a(a)(3)(ii) BSEP-98-0080, Monthly Operating Repts for Mar 1998 for Bsep,Units 1 & 21998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Bsep,Units 1 & 2 ML20216B1041998-03-0404 March 1998 SER Approving Alternative to Insp of Reactor Pressure Vessel Circumferential Welds for Brunswick Steam Electric Plant, Unit 1 1999-09-30
[Table view] |
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[ psp"c% 1 UNITED STATES g .) ,j NUCLEAR REGULATORY COMMISSION o o t WASHINGTON, D.C. 20565-0001
\p...vl/ ..
l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i
i ALTERNATIVE TO INSPECTION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL WELDS j BRUNSWICK STEAM ELECTRIC PLANT UNIT N0. 1 CAROLINA POWER & LIGHT COMPANY l
DOCKET N0: 50-325
1.0 INTRODUCTION
- l By letter dated November 17, 1997. Carolina Power & Light Company (CP&L or the licensee) requested an alternative to performing the reactor pressure vessel
) (RPV) circumferential shell weld examinations requirements of both the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME l
Code).Section XI, 1980 Edition, through the Winter 1981 Addenda (inservice inspection), and the augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) for the Brunswick Steam Electric Plant (BSEP)
Unit 1. The alternative was proposed pursuant to the provisions of 50.55a(a)(3)(i) and 10 CFR 50.55a(g)(6)(ii)(A)(5), and is consistent with information contained in Information Notice (IN) 97-63 " Status of NRC Staff Review of BWRVIP-05." l The alternative proposed by CP&L is the performance of inspections of essentially 100 percent of the BSEP Unit 1 RPV shell longitudinal seam welds and essentially 0 percent of the RPV shell circumferential seam welds during Refueling Outage 11, which will result in partial examination (2 - 3 percent) of the circumferential welds at or near the intersections of the longitudinal and circumferential welds.
The requirement for inservice inspections, which include RPV circumferential weld inspection, derives from the Technical Specifications (TS) 4.0.5 for BSEP Unit 1 which state that "the inservice inspection (ISI) and testing of the
, ASME Code Class 1, 2 and 3 components be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, and aoplicable addenda, as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the NRC. .~. Pursuant to the requirements of 10 CFR 50.55a(g)(4).
ASME Code Class 1.2. and 3 components shall meet the requirements, except the 9804270396 900421 PDR ADOCK 05000325 P PDR
4
- design and access provisions and the preservice examination requirements, set forth in the ASME Code.Section XI. " Rules for Inservice Inspection of Nuclear
! Power Plant Components" to the extent practical within the limitations of design. geometry, and materials of construction of the components. The l regulations require that inservice examination of components and system l pressure tests conducted during the first 10-year interval and subsequent l intervals comply with the requirements in the latest edition and addenda of l the ASME Code.Section XI. incorporated by reference in 10 CFR 50.55a(b) on l the date 12 months prior to the start of the 120-month interval, subject to l the limitations and modifications listed therein. The applicable ASME Code'.
Section XI. for Brunswick. Unit 1. during the current 10-year ISI interval is the 1980 Edition through the Winter 1981 Addenda.
Section 50.55a(g)(6)(ii)(A) to Title 10 of the Code of Federal hegulations l
(10 CFR 50.55a(g)(6)(ii)(A)) requires that licensees perform an expanded RPV l shell weld examination as specified in the 1989 Edition of Section XI of the l ASME Code. on an " expedited" basis. " Expedited." in this context. effectively l meant during the inspection interval when the Rule was approved or the first
! period of the next inspection interval. The final ~ Rule was published in the Federal Register on August 6, 1992 (57 FR 34666). By incorporating into the regulations the 1989 Edition of the ASME Code the NRC staff required that licensees perform volumetric examination of " essentially 100 percent" of the RPV pressure-retaining shell welds during all inspection intervals. Section 50.55a(a)(3)(i) (10 CFR 50.55a(a)(3)(i)) indicates that alternatives to the requirements in 10 CFR 50.55a(g) are justified when the proposed alternative provides an acceptable level of quality and safety.
By letter dated September 28, 1995, as supplemented by letters dated June 24 and October 29. 1996, and May 16. June 4. and June 13. 1997, the Boiling Water Reactor Vessel and Internals Project (BWRVIP), a technical committee of the BWR Owners Group (BWROG). submitted the proprietary report. "BWR Vessel and Internals Project. BWR Reactor Vessel Shell Weld Inspection Recommendations '
(BWRVIP-05)." which proposed to reduce the scope of inspection of the BWR RPV welds from essentially 100 percent of all RPV shell welds to 50 percent of the axial welds and 0 percent of the circumferential welds. By letter dated October 29. 1996, the BWRVIP modified their proposal to increase the 1 examination of the axial welds to 100 percent from 50 percent while still I proposing to inspect essentially 0 percent of the circumferential RPV shell welds, except that the intersection of the axial and circumferential welds would have included approximately 2-3 percent of the circumferential welds.
On May 12. 1997, the NRC staff and members of the BWRVIP met with the Commission to discuss the NRC staff's review of the BWRVIP-05 report. In
accordance with guidance provided by the Commission in Staff Requirements Memorandum (SRM) M970512B, dated May 30, 1997, the staff has initiated a broader. risk-informed review of the BWRVIP-05 proposal.
In IN 97-63, the staff indicated that it would consider technically justified alternatives to the augmented examination in accordance with 10 CFR 50.55a(a)(3)(1). 10 CFR 50.55a(a)(3)(ii). and 50.55a(g)(6)(ii)(A)(5). from BWR licensees who are scheduled to perform inspections of the BWR RPV circumferential welds during the Fall 1997 or Spring 1998 outage seasons.
Acceptably justified alternatives would be considered for inspection delays of up to 40 months or two operating cycles (whichever is longer) for BWR RPV circumferential shell welds only.
2.0 BACKGROUND
- STAFF ASSESSMENT OF BWRVIP-05 REPORT:
The staff's independent assessment of the BWRVIP-05 proposal is documented in a letter dated August 14, 1997, to Mr. Carl Terry, BWRVIP Chairman. The staff concluded that the industry's assessment does not sufficiently address risk.
and additional work is necessary to provide a complete risk-informed evaluation.
The staff's assessment was performed for BWR RPVs fabricated by Chicago Bridge and Iron (CB&I). Combustion Engineering (CE), and Babcock & Wilcox (B&W). The staff assessment identified cold overpressure events as the limiting transients that could lead to failure of BWR RPVs. Using the pressure and temperature resulting from a cold overpressure event in a foreign reactor and the parameters identified in Table 7-1 of the staff's independent assessment, the staff determined the conditional probability of failure for axial and 1 circumferential welds fabricated by CB&I. CE, and B&W. Table 7-9 of the I staff's assessment identifies the conditional probability of failure for the reference cases and the 95 percent confidence uncertainty bound cases for axial and circumferential welds fabricated by CB&I. CE and B&W. B&W fabricated vessels were determined to have the highest conditional probability of failure. The input material parameters used in the analysis of the reference case for B&W fabricated vessels resulted in a reference temperature (RTer) at the vessel inner surface of 114.5*F. In the uncertainty analysis, the neutron fluence evaluation had the greatest RTer value (145 F) at the inner surface. Vessels with rte 1 values less than those resulting from the staff's assessment will have less embrittlement than the vessels simulated in the staff's assessment and should have a conditional probability of vessel failure less than or equal to the values in the staff's assessment, l
i I
The failure probability for a weld is the product of the critical event frequency and the conditional probability of the weld failure for that event.
Using the event frequency for a cold overpressure event and the conditional probability of vessel failure for CB&I fabricated circumferential welds, the best-estimate failure frequency from the staff's assessment is 6.0 X 10-inn per reactor year and the uncertainty bound failure frequency is < 2.8 X 10'*D per reactor year.
3.0 LICENSEE TECHNICAL JUSTIFICATION:
! The licensee indicated in the November 17, 1997, letter that the basis for requesting the alternative inspections is the BWRVIP-05 report, which stated that the prcbability of failure of BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. This conclusion was also demonstrated in the staff's independent assessment of the BWRVIP-05 report. The BWRVIP-05 report indicates that, for a typical BWR RPV the
, failure probability for axial welds is 2.7 X 10 and the failure probability 7
for circumferential welds is 2.2 X 1042 for 40 years of plant operation.
The licensee calculated the RTm value for the BSEP Unit 1 circumferential weld at the end of the requested relief period using the methodology in Regulatory Guide (RG) 1.99. Revision 2. The RT, values calculated in accordance with RG 1.99 Revision 2. depend upon the neutron fluence, the amounts of copper and nickel in the circumferential weid, and its unirradiated RTuor. The licensee determined the highest neutron fluence at the end of the next two operating cveles at the inner surface of the circumferential beltline weld to be 0.063 X 10' n/cm2 . The amounts of copper and nickel in the limting circumferential beltline weld is 0.06 percent and 0.87 percent. respectively. .
The plant-specific unirradiated RT, for the circumferential beltline weld is !
10*F. Using these parameters and the methodology in Regulatory Guide 1.99.
Revision 2. the licensee determined that the RT, value for the circumferential weld at the end of the relief period is 64.26 F. The licensee noted that the RTm resulted from the plant-specific unirradiated RTm value that was provided in CP&L's letter dated November 16. 1995, and that use of the generic initial RT, value of -56 *F would yield an RT, value of 14.63 F. The larger RT, value that results from plant-specific unirradiated RT, is still less than the most limiting reference case (B&W fabricated vessels) in the staff's assessment. Since the RT, of the BSEP. Unit 1. beltline circumferential weld is.less than the limiting RTm value in the staff's Insufficient or no failures to accurately determine reference case failure probability.
l l
l 5-independent assessment, the licensee concluded that the BSEP, Unit 1, vessel circumferential welds are bounded by the staff's independent' assessment, thus providing additional assurance that the vessel welds are bounded by the BWRVIP-05 report.
1
! The licensee assessed the systems that could lead to a cold overpressurization l of the BSEP, Unit 1. RPV. These included the high pressure coolant injection i
(HPCI), reactor core isolation cooling (RCIC), standby liquid control (SLCS),
control rod drive (CRD) and reactor water cleanup systems (RWCU). Both the HPCI and RCIC pumps are steam driven and do not function during cold shutdown.
l The licensee stated that there were no automatic starts associated with SLCS.
Operator initiation of SLCS should not occur during shutdown; however, the j SLCS injection rate is approximately 41 gpm which woul.d allow the operators i sufficient time to control reactor pressure if manual initiation occurred.
The CRD and RWCU systems t.se a feed and bleed process to control RPV level and pressure during normal cold shutdown conditions. The CRD pumps injection rate is less than 60 gpm which allows sufficient time for operators to react to l unanticipated level changes.
In all cases, the operators are trained in methods of controlling water level within specified limits in addition to responding to abnormal water level conditions during shutdown. The licensee also stated that procedural controls for reactor temperature, level, and pressure are an integral part of operator
! training. Plant-specific procedures have been established to provide guidance to the operators regarding compliance with the Technical Specification pressure-temperature limits. On the basis of the pressure limits of the operating systems, operator training, and established plant-specific procedures, the licensee determined that a non-design basis cold overpressure transient is unlikely to occur during the requested delay. Therefore, the l
licensee concluded that the probability of a cold overpressure transient is '
considered to be less than or equal to that used in the staff's assessment of BWRIVP-05.
4,0 STAFF REVIEW 0F LICENSEE TECHNICAL JUSTIFICATION:
BSEP Unit 1 is a CB&I fabricated vessel, and the staff noted that the RTuor i value determined from the plant-specific unirradiated RTuo7 (64.26 F) is approximately 14 F higher than the limiting value determined in the staff's assessment for CB&I fabricated vessels (50 *F). However, RTuor is a measure of the amount of irradiation embrittlement, and since CB&I fabricated vessels have very low copper values, they have low amounts of irradiation embrittlement. For comparison, the staff confirmed that the RTuo7 value for
I the circumferential welds at the end of the relief period is less than the values in the limiting reference case and uncertainty analysis for the B&W fabricated vessels. Since the RT,m values are well below the values in the reference case and the uncertainty analysis for B&W fabricated vessels, the l
BSEP Unit 1 RPV will have less embrittlement than the B&W fabricated vessels and will have conditional probability of vessel failure less than or equal to that estimated in the staff's assessment.
The staff reviewed the information provided by the licensee regarding the BSEP. Unit 1. high pressure injection systems, operator training, and plant-specific procedures to prevent RPV cold overpressurization. The information provided sufficient basis to support approval of the alternative examination l request. Based on the high pressure injection systems analyses, operator
! training, and plant-specific procedures, the probability of a cold l
overpressurization transient occurring at BSEP. Unit 1. during the requested delay is low, which is consistent with the staff's assessment.
5.0 CONCLUSION
S:
i l Based upon its review. the staff reached the following conclusions: i
- 1) Based on the licensee's assessment of the materials in the circumferential weld in the beltline of the Brunswick. Unit 1. RPV. the l conditional probability of vessel failure should be less than or equal to that estimated from the staff's assessment of the limiting case of l B&W fabricated vessels.
- 2) Based on the licensee's high pressure injection systems analyses.
- operator training, and plant-specific procedures, the probability of I
cold overpressure transients should be sufficiently low during the requested delay period.
- 3) Based on the previous two conclusions, the staff concludes that Brunswick. Unit 1. RPV can be operated during the requested delay period with an acceptable level of quality and safety and the inspection of the circumferential welds can be delayed for two operating periods.
Therefore, the proposed alternative to performing the RPV examination requirements of the ASME Code.Section XI 1980 Edition, with Winter 1981 Addenda, and the augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) at Brunswick Unit 1 for circumferential shell welds for two operating cycles is authorized pursuant to 10 CFR 7
50.55a(a)(3)(i).