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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217K9931999-10-14014 October 1999 Safety Evaluation Supporting Amend 234 to License DPR-56 ML20217B4331999-10-0505 October 1999 Safety Evaluation Supporting Amend 233 to License DPR-56 ML20216H7091999-09-24024 September 1999 Safety Evaluation Supporting Amends 229 & 232 to Licenses DPR-44 & DPR-56,respectively ML15112A7681999-09-20020 September 1999 SER Accepting Revision 25 of Pump & Valve Inservice Testing Program,Third 10-year Interval for Plant,Units 1,2 & 3 ML20212D1281999-09-17017 September 1999 Safety Evaluation Supporting Proposed Alternatives CRR-03, 05,08,09,10 & 11 ML20211D5501999-08-23023 August 1999 Safety Evaluation Supporting Amends 228 & 231 to Licenses DPR-44 & DPR-56,respectively ML20206A2921999-04-20020 April 1999 Safety Evaluation Concluding That Proposed Changes to EALs for PBAPS Are Consistent with Guidance in NUMARC/NESP-007 & Identified Deviations Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 ML20205K7411999-04-0707 April 1999 Safety Evaluation Supporting Amends 227 & 230 to Licenses DPR-44 & DPR-56,respectively ML20196G7021998-12-0202 December 1998 SER Authorizing Proposed Alternative to Delay Exam of Reactor Pressure Vessel Shell Circumferential Welds by Two Operating Cycles ML20155C6071998-10-26026 October 1998 Safety Evaluation Supporting Amend 226 to License DPR-44 ML20155C1681998-10-22022 October 1998 Safety Evaluation Accepting Proposed Alternative Plan for Exam of Reactor Pressure Vessel Shell Longitudinal Welds ML20154H4771998-10-0505 October 1998 Safety Evaluation Supporting Amends 225 & 229 to Licenses DPR-44 & DPR-56,respectively ML20154J2401998-10-0505 October 1998 Safety Evaluation Supporting Amends 224 & 228 to Licenses DPR-44 & DPR-56,respectively ML20154G6821998-10-0101 October 1998 SER Related to Request for Relief 01A-VRR-1 Re Inservice Testing of Automatic Depressurization Sys Safety Relief Valves at Peach Bottom Atomic Power Station,Units 2 & 3 ML20154G6631998-10-0101 October 1998 Safety Evaluation Supporting Amends 223 & 227 to Licenses DPR-44 & DPR-56,respectively ML20153B9651998-09-14014 September 1998 Safety Evaluation Supporting Amend 9 to License DPR-12 ML20238F2661998-08-24024 August 1998 Safety Evaluation Supporting Amend 222 to License DPR-44 ML20237A7761998-08-10010 August 1998 SER Accepting Licensee Response to NRC Bulleting 95-002, Unexpected Clogging of RHR Pump Strainer While Operating in Suppression Pool Cooling Mode ML20236R8281998-07-15015 July 1998 Safety Evaluation Approving Proposed Alternative (one-time Temporary non-Code Repair) Pursuant to 10CFR50.55a(a)(3) (II) ML20248F4781998-06-0101 June 1998 Corrected Page 1 to SE Supporting Amends 221 & 226 to Licenses DPR-44 & DPR-56,respectively.Original Page 1 of SE Had Three Typos ML20247N5351998-05-11011 May 1998 SER Accepting Third 10-year Interval Inservice Program for Pump & Valves for Plant,Units 2 & 3 ML20198L3331997-12-18018 December 1997 Safety Evaluation Supporting Approval of Proposed Merger of Atlantic Energy,Inc,& Delmarva Power & Light Co ML20212G8301997-10-24024 October 1997 Safety Evaluation Supporting Amends 221 & 226 to Licenses DPR-44 & DPR-56,respectively ML20198S2161997-10-24024 October 1997 Safety Evaluation Accepting Proposed Change to Provisions Identified in Rev 14 of PBAPS QAP Description Re Nuclear Review Board Meeting Frequency ML20217J5631997-10-0909 October 1997 Safety Evaluation Supporting Amend 225 to License DPR-56 ML20217J6161997-10-0707 October 1997 Safety Evaluation Re Alternative to Reactor Pressure Vessel Circumferential Weld Insps for Plant,Unit 3 ML20211L6241997-10-0303 October 1997 Safety Evaluation Authorizing Licensee Proposed Use of Code Case N-516-1 to Weld Modified Suction Strainer in Suppression Chamber at Plant ML20217D8161997-09-30030 September 1997 Safety Evaluation Supporting Amend 224 to License DPR-56 ML20211D6201997-09-17017 September 1997 SER Accepting VT-2 Examiner Qualification Request for PECO Energy Company,Peach Bottom Atomic Power Station,Units 2 & 3 ML20216G5601997-09-0404 September 1997 Safety Evaluation Supporting Amends 220 & 223 to Licenses DPR-44 & DPR-56,respectively ML20217M8001997-08-19019 August 1997 Safety Evaluation Supporting Amends 219 & 222 to Licenses DPR-44 & DPR-56,respectively ML20149L2841997-07-23023 July 1997 Safety Evaluation Accepting Licensee Relief Request RR-22 for Plant,Units 2 & 3 ISI Program ML20140B0371997-05-30030 May 1997 Safety Evaluation Accepting QAP Description Change ML20135B4111997-02-19019 February 1997 Safety Evaluation Supporting Amends 218 & 221 to Licenses DPR-44 & DPR-56,respectively ML20149L8681996-11-15015 November 1996 SER Accepting Core Spray Piping Insp & Flaw Evaluation for Plant,Unit 2 ML20128G6941996-09-27027 September 1996 Safety Evaluation Supporting Amend 217 to License DPR-44 ML20117B4521996-08-16016 August 1996 Safety Evaluation Supporting Amend 216 to License DPR-44 ML20115H2361996-07-15015 July 1996 Safety Evaluation Supporting Amends 277 & 278 to Licenses DPR-44 & DPR-56,respectively ML20117N9691996-06-18018 June 1996 Safety Evaluation Supporting Amends 214 & 219 to Licenses DPR-44 & DPR-56,respectively ML20149L2441996-01-29029 January 1996 Safety Evaluation Accepting Insp & Evaluation Methodology for Operation of Unit 3 Core Shroud for Duration of Current Operating Cycle,Performed in Response to GL 94-03 ML20096E0461996-01-16016 January 1996 Safety Evaluation Supporting Amends 213 & 218 to Licenses DPR-44 & DPR-56,respectively ML20096D0041996-01-11011 January 1996 Safety Evaluation Supporting Amends 212 & 217 to Licenses DPR-44 & DPR-56,respectively ML20096C9541996-01-11011 January 1996 Safety Evaluation Supporting Amends 211 & 216 to Licenses DPR-44 & DPR-56,respectively ML20094B1041995-10-17017 October 1995 Safety Evaluation Supporting Amend 215 to License DPR-56 ML20092D7501995-08-30030 August 1995 Safety Evaluation Supporting Amends 210 & 214 to Licenses DPR-44 & DPR-56,respectively ML20087A7931995-07-18018 July 1995 Safety Evaluation Supporting Amend 211 to License DPR-56 ML20086Q0701995-07-10010 July 1995 Safety Evaluation Supporting Amend 210 to License DPR-56 ML20085M9541995-06-19019 June 1995 Safety Evaluation Supporting Amends 207 & 209 to Licenses DPR-44 & DPR-56,respectively ML20087G0101995-06-13013 June 1995 Safety Evaluation Supporting Amend 206 to License DPR-44 ML20086E0481995-06-13013 June 1995 Safety Evaluation Supporting Amends 205 & 208 to Licenses DPR-44 & DPR-56,respectively 1999-09-24
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEAR05000278/LER-1999-005-03, :on 990920,uplanned Esfas During Planned Mod Activitives in Main CR Were Noted.Caused by Inattention to Detail by Individuals Performing Work.All CR Mods Were Ceased to Allow Review of Mod Work Packages.With1999-10-20020 October 1999
- on 990920,uplanned Esfas During Planned Mod Activitives in Main CR Were Noted.Caused by Inattention to Detail by Individuals Performing Work.All CR Mods Were Ceased to Allow Review of Mod Work Packages.With
ML20217K9931999-10-14014 October 1999 Safety Evaluation Supporting Amend 234 to License DPR-56 ML20217B4331999-10-0505 October 1999 Safety Evaluation Supporting Amend 233 to License DPR-56 05000278/LER-1999-004-03, :on 990901,3A RPS Bus Was Inadvertently Deenergized,During Planned Mod Activities on Main CR Panel. Caused by Electrician Failing to Self Check Work.All CR Work Was Ceased Immediately & Shutdown Meeting Held1999-10-0101 October 1999
- on 990901,3A RPS Bus Was Inadvertently Deenergized,During Planned Mod Activities on Main CR Panel. Caused by Electrician Failing to Self Check Work.All CR Work Was Ceased Immediately & Shutdown Meeting Held
ML20217G3541999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pbaps,Units 2 & 3. with ML20216H7091999-09-24024 September 1999 Safety Evaluation Supporting Amends 229 & 232 to Licenses DPR-44 & DPR-56,respectively ML15112A7681999-09-20020 September 1999 SER Accepting Revision 25 of Pump & Valve Inservice Testing Program,Third 10-year Interval for Plant,Units 1,2 & 3 ML20212D1281999-09-17017 September 1999 Safety Evaluation Supporting Proposed Alternatives CRR-03, 05,08,09,10 & 11 05000278/LER-1999-003-03, :on 990814,HPCIS Was Declared Inoperable Due to Erratic Behavior Resulting in Loss of Single High Train Safety Sys.Caused by Weakness in Procedural Guidance. Readjusted Hydraulic Governor Needle Valve.With1999-09-13013 September 1999
- on 990814,HPCIS Was Declared Inoperable Due to Erratic Behavior Resulting in Loss of Single High Train Safety Sys.Caused by Weakness in Procedural Guidance. Readjusted Hydraulic Governor Needle Valve.With
ML20212A5871999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Peach Bottom,Units 2 & 3.With ML20211D5501999-08-23023 August 1999 Safety Evaluation Supporting Amends 228 & 231 to Licenses DPR-44 & DPR-56,respectively ML20212H6311999-08-19019 August 1999 Rev 2 to PECO-COLR-P2C13, COLR for Pbaps,Unit 2,Reload 12 Cycle 13 ML20210N7641999-07-31031 July 1999 Monthly Operating Repts for Jul 1999 for PBAPS Units 2 & 3. with 05000277/LER-1999-005-01, :on 990616,failure to Maintain Provisions of FP Program Occurred.Caused by Less than Adequate Engineering Rigor in Both Development & Review Analysis.Fire Watch Immediately Established.With1999-07-16016 July 1999
- on 990616,failure to Maintain Provisions of FP Program Occurred.Caused by Less than Adequate Engineering Rigor in Both Development & Review Analysis.Fire Watch Immediately Established.With
ML20209H1121999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Pbaps,Units 2 & 3. with ML20195H8841999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Pbaps,Units 2 & 3. with 05000278/LER-1999-002-02, :on 990406,safeguard Sys to Unrelated Door Was Inadvertently Disabled by Security Alarm Station Operator. Caused by Noncompliance with Procedures & Less than Adequate Shift Turnover.Briefed Personnel on Event.With1999-05-0606 May 1999
- on 990406,safeguard Sys to Unrelated Door Was Inadvertently Disabled by Security Alarm Station Operator. Caused by Noncompliance with Procedures & Less than Adequate Shift Turnover.Briefed Personnel on Event.With
ML20206N1661999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Pbaps,Units 2 & 3. with ML20206A2921999-04-20020 April 1999 Safety Evaluation Concluding That Proposed Changes to EALs for PBAPS Are Consistent with Guidance in NUMARC/NESP-007 & Identified Deviations Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 05000278/LER-1999-001-03, :on 990312,ESF Actuation of Rcics Occurred Due to High Steam Flow Signal During Sys Restoration.Temporary Change to Restoration Procedure Was Initiated to Open RCIC Outboard Steam Isolation Valve in Smaller Increments1999-04-0808 April 1999
- on 990312,ESF Actuation of Rcics Occurred Due to High Steam Flow Signal During Sys Restoration.Temporary Change to Restoration Procedure Was Initiated to Open RCIC Outboard Steam Isolation Valve in Smaller Increments
ML20205K7411999-04-0707 April 1999 Safety Evaluation Supporting Amends 227 & 230 to Licenses DPR-44 & DPR-56,respectively ML20205P5851999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Peach Bottom Units 2 & 3.With ML20207G9971999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Peach Bottom Units 2 & 3.With 05000278/LER-1998-009-01, :on 981227,unplanned Esfa Were Noted.Caused by Transformer Insulator Failure.Replaced Failed Insulator. with1999-01-20020 January 1999
- on 981227,unplanned Esfa Were Noted.Caused by Transformer Insulator Failure.Replaced Failed Insulator. with
ML20199E3471998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Peach Bottom,Units 1 & 2.With ML20206P1651998-12-31031 December 1998 Fire Protection for Operating Nuclear Power Plants, Section Iii.F, Automatic Fire Detection ML20205K0381998-12-31031 December 1998 PECO Energy 1998 Annual Rept. with ML20206D3651998-12-31031 December 1998 1998 PBAPS Annual 10CFR50.59 & Commitment Rev Rept. with ML20206D3591998-12-31031 December 1998 1998 PBAPS Annual 10CFR72.48 Rept. with 05000277/LER-1998-008-01, :on 981130,circuit Breaker SU-25 Tripped.Caused by Less than Adequate Procedural Guidance.Operators Verified Sys Integrity & Successfully Returned Sys to Svc.With1998-12-30030 December 1998
- on 981130,circuit Breaker SU-25 Tripped.Caused by Less than Adequate Procedural Guidance.Operators Verified Sys Integrity & Successfully Returned Sys to Svc.With
05000277/LER-1998-007-02, :on 981107,failure to Meet TS & Associated LCO Requirments of Absolute Difference in APRM & Calculated Power of Less than 2% Was Noted.Caused by Substitute Valves Being Used.Removed Substitute Valves.With1998-12-0404 December 1998
- on 981107,failure to Meet TS & Associated LCO Requirments of Absolute Difference in APRM & Calculated Power of Less than 2% Was Noted.Caused by Substitute Valves Being Used.Removed Substitute Valves.With
ML20196G7021998-12-0202 December 1998 SER Authorizing Proposed Alternative to Delay Exam of Reactor Pressure Vessel Shell Circumferential Welds by Two Operating Cycles ML20196E8261998-11-30030 November 1998 Response to NRC RAI Re Reactor Pressure Vessel Structural Integrity at Peach Bottom Units 2 & 3 ML20198B8591998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Pbaps,Units 2 & 3. with 05000278/LER-1998-005-03, :on 981025,inadvertent Unit 3 Electrical Bus E33 Trip (Esfa) During Performance of Unit 2 Electrical Bus E32 Surveillance Test Was Noted.Caused by Personnel Error. Sp S12M-54-E32-XXF4 Was Completed.With1998-11-20020 November 1998
- on 981025,inadvertent Unit 3 Electrical Bus E33 Trip (Esfa) During Performance of Unit 2 Electrical Bus E32 Surveillance Test Was Noted.Caused by Personnel Error. Sp S12M-54-E32-XXF4 Was Completed.With
ML20206R2571998-11-17017 November 1998 PBAPS Graded Exercise Scenario Manual (Sections 1.0 - 5.0) Emergency Preparedness 981117 Scenario P84 ML20198C6751998-11-0505 November 1998 Rev 3 to COLR for PBAPS Unit 3,Reload 11,Cycle 12 ML20195E5341998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Pbaps,Units 2 & 3. with ML20155C6071998-10-26026 October 1998 Safety Evaluation Supporting Amend 226 to License DPR-44 ML20155C1681998-10-22022 October 1998 Safety Evaluation Accepting Proposed Alternative Plan for Exam of Reactor Pressure Vessel Shell Longitudinal Welds ML20155H7721998-10-12012 October 1998 Rev 1 to COLR for Peach Bottom Atomic Power Station Unit 2, Reload 12,Cycle 13 05000277/LER-1998-006-02, :on 980915,automatic RWCU Isolation Occurred While Placing RWCU Sys in Svc.Caused by Unexpected Surge of Water.Procedure Change Was Initiated to Open MO-2-12-74 & RWCU Sys Was Successfully Returned to Svc.With1998-10-0909 October 1998
- on 980915,automatic RWCU Isolation Occurred While Placing RWCU Sys in Svc.Caused by Unexpected Surge of Water.Procedure Change Was Initiated to Open MO-2-12-74 & RWCU Sys Was Successfully Returned to Svc.With
ML20154J2401998-10-0505 October 1998 Safety Evaluation Supporting Amends 224 & 228 to Licenses DPR-44 & DPR-56,respectively ML20154H4771998-10-0505 October 1998 Safety Evaluation Supporting Amends 225 & 229 to Licenses DPR-44 & DPR-56,respectively ML20154G6821998-10-0101 October 1998 SER Related to Request for Relief 01A-VRR-1 Re Inservice Testing of Automatic Depressurization Sys Safety Relief Valves at Peach Bottom Atomic Power Station,Units 2 & 3 ML20154G6631998-10-0101 October 1998 Safety Evaluation Supporting Amends 223 & 227 to Licenses DPR-44 & DPR-56,respectively ML20154H5541998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Pbaps,Units 2 & 3. with 05000278/LER-1998-004-03, :on 980820,automatic RWCU Isolation Occurred While Placing B RWCU Sys Demineralizer in Svc.Caused by less-than-adequate Control of Equipment.Isolated B Demineralizer & Returned RWCU Sys to Svc1998-09-18018 September 1998
- on 980820,automatic RWCU Isolation Occurred While Placing B RWCU Sys Demineralizer in Svc.Caused by less-than-adequate Control of Equipment.Isolated B Demineralizer & Returned RWCU Sys to Svc
05000277/LER-1998-005-02, :on 980824,noted Failure to Meet TS Actions for Suppression chamber-to-drywell Vacuum Breaker Not Being Fully Seated.Caused by Personnel Failing to Take All TS Required Actions.Temporary Procedure Changes Were Made1998-09-18018 September 1998
- on 980824,noted Failure to Meet TS Actions for Suppression chamber-to-drywell Vacuum Breaker Not Being Fully Seated.Caused by Personnel Failing to Take All TS Required Actions.Temporary Procedure Changes Were Made
ML20153B9651998-09-14014 September 1998 Safety Evaluation Supporting Amend 9 to License DPR-12 1999-09-30
[Table view] |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION i
WASHINGTON, D.C. 2006H001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE TO THE AUGMENTED EXAMINATION OF THE REACTOR PRESSURE VESSEL LONGITUDINAL WELDS PEACH BOTTOM ATOMIC POWER STATION. UNIT 2 PECO ENERGY COMPANY DOCKET NO. 50-277
1.0 INTRODUCTION
Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(2), the licensees are required to perform augmented examinations of reactor pressure vessels. The licensees are required to implement an augmented examination of " essentially 100%" of the reactor pressure vessel (RPV) shell welds.
The shell welds are specified in the 1989 Edition of the American Society of Mechanical Engineers (ASME) Code,Section XI, Table IWB-25001, Examination Category B A, " Pressure l
Retaining Welds in Reactor Vessel," item B1.10. This ASME category includes item 81.12, i
longitudinal shell welds. The 10 CFR 50.55a(g)(6)(ii)(A)(2) defines " essentially 100%"
examination as "more than 90% of the examination volume of each weld." Under 10 CFR 50.55a(g)(6)(ii)(A)(5), the licensees unable to completely satisfy the requirements of the augmented reactor vessel examination may propose an attemative that provides an acceptable level of quality and safety. A licensee may use its proposed altemative when authorized by the l
Director of the Office of Nuclear Reactor Regulation, in a letter dated April 2,1998, the PECO Energy Company (the licensee) submitted a request to use an alternative to perform certain augmented examinations of the RPVlongitudinal shell welds at Peach Bottom Atomic Power Station (PBAPS), Unit 2. In a letter dated August 12, 1998, the licensee provided additional and clarifying information to support the original request.
The licensee examined the RPV intemal surface to the extent practical, even though 90%
coverage was not attained.
2.0 REQUEST FOR ALTERNATIVE EXAMINATION The licensee requests an altemative examination to performing additional examinations to achieve the full 90% coverage.
3.0 REQUIREMENT Under 10 CFR 50.55a(g)(6)(ii)(A)(2) requirements, all licensees shall augment their reactor l
vessel examination by implementing once, as part of the inservice inspection interval in effect on l
Septernber 8,1992, the examination requirements for reactor vessel shell welas specified in item B1.10 of Examination Category B A, " Pressure Retaining Welds in Reactor Vessel,"in Table IWB-2500-1 of subsection lWB of the 1989 Edition of section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(A)(3) l and (4).
9811020100 981022 PDR ADOCK 05000277 P
PDR ENCLOSURE c
e.
4.0 BASIS FOR ALTERNATIVE EXAMINATION Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the licensees that are unable to completely satisfy the augmented RPV shell weld examination requirement may submit information to the Commission to support such determination and propose attematives to the examination requirements that would provide an acceptable level of quality and safety.
5.0 PROPOSED ALTERNATIVE EXAMINATION The licensee is unable to meet the 90% volume coverage requirement for each longitudinal weld of the PBAPS, Unit 2, reactor pressure vessel as required by 10 CFR 50.55a(g)(6)(ii)(A)(2).
Therefore, it is proposing an attemative to 90% volume coverage requirement for each longitudinal weld, in accordance with 10 CFR 50.55a(s)(3)(i), pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5).
I The licensee stated that it intends to inspect the PBAPS, Unit 2, RPV during the upcoming PBAPS, Unit 2 refueling outage (2R12), scheduled to begin in early October 1998. The proposed alternative is to rform an examination of the RPV longitudinal shell welds to the maximum extent practical frora the Inner Diameter (ID), within the constraints of vessel intemal rostrictions.
This examination would be performed for longitudinal shell welds specified in item B1.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel,"in Table IWB-2500-1 f
i of Subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code. Further examination from the ID is not practical without disassembly of the vessel intemal components. Taba 1 provides the longitudinal welds, the estimated volumetric examination coverage from the ID, and the ID restrictions that will potentially obstruct scanning.
The volumetric examination co<erage and restrictions from the ID are based on a detailed weld coverage scan plan completed in February 1998. As shown in Table 1, eight of 15 wsids achieve greater than 90% volumetric coverage, crediting ID inspections only. There are no vessel intemals which pose a restrictica to the 10 examination which are easily removable. The only removable components (i.e., not weided to the vessel) which limit ccan coverage are the feedwater spargers. However, it is impractical to remove the feedwater spargers due to the potential for damage to the sparger seals and nonles. Therefore, there are no components which can be reasonably removed to increase coverage from the ID.
For those longitudinal welds where greater than 90% mjumetric examination may not be achieved from the ID, the estimated supplemental coverge and physical constraints on the vessel Outer Diameter (OD) are identified in Table 1. As neted in Table 1, further review has determined that two (2) welds (RPV-V1 A and RPV VSA) wouid exceed 90% volume coverage with a supplemental OD examination. In some locations, the editional weld volume that can be accessed from the OD is a subset of the ID examinations.
The restrictions which prohibit unrestrained access to 100% of the longltudinal weld volume from the OD are the vesselinsulation and the vessel noules.
The percentage of longitudinal weld volume coverage estimated from the ID Examination reflects a significant portion of the total reactor pressure vessel weld length but represerits greater than f
l-90% of the longitudinal welds within the beltline region. The beltline is the region in the RPV that is adjacent to the core, receives significant amounts of neutron radiation, and is susceptible to l
E _...,
~
3-embrittlement. The beltline welds are RPV VIA, VIB, VIC, V2A, V2B 2nd V2C. Attempting to perform supplemental OD examinations would result in a minimal increase to longitudinal weld volume coverage and would only increase the amount of inspection of beltline weld RPV VIA by l
a small amount (10%). Examination from the OD would only permit two additional welds to exceed 90% volumetric coverage requirement. Additional disassembly and reassembly of portions of the reactor vessel biological shield and insulation would result in further increases in personnel radiation exposure as well as an increase in the general area dose rates in the drywell for the entire population of workers. Additional doses to the entire population to perform the supplemental (OD) examinations contained in Table 1 are estimated to be 24 man-Rem.
Therefore, based on the incremental cost and radiation dose, in conjunction with the limited additional volumetric coverage, the licensee has concluded that performing OD examinations to increase coverage beyond that achieved from the ID would result in undue hardship without a l
compensating increase in safety.
5.0 EVALUATION The staff has evaluated the alternatives proposed by the licensee for the volumetric examination l
of the above mentioned reactor vessel shell welds in regard to the following factors:
. Physical constraints at each weld that limit the required examination coverage l
. Maximum extent of volumetric coverage obtained with the existing constraints l
. Supplementing inner diameter examination with examination from the outside
. Results of previous vessel examinations
. Detection of the presence of degradation mechanisms, if any, from the examination l
Based on its evaluation, the NRC staff agrees that supplementing the ID examination with an OD examination would result in only two additional welds to achieve an increased volumetric coverage to meet the Code requirement. However, it constitutes a hardship in that a significant amount of disassembly and reassembly of the reactor vessel biological shield and insulation is i
I required without a compensating increase in the level of quality and safety, and the licensee is unable to completley comply with the requirements.
l Curing the fabrication process of the PBAPS, Unit 2 RPV, all of the shell welds were thoroughly examined using several examination methods as required by the original construction code.
Additionally, all of the shell welds received volumetric examinations prior to initial plant l
operations, as prescribed by ASME Section XI preservice inspection requirements. Selected shell welds have received volumetric examinations during the first inservice inspection interval in accordance with ASME Section XIinservice inspection requirements. No rejectable indications were identified during any of these examinations. The preservice inspection and inservice inspection identified planar and/or linea indications are listed in Table 1. Allindications in Table 1 are Code acceptable.
The General Electric (GE) GERIS 2000 System will be used to perform the remote controlled, automated UT examinations of the RPV. This tool has been used previously at PBAPS, Unit 3 and other Boiling Water Reactors for the purpose of RPV examinations. GE demonstrated this
1 1
4 system at the Performance Demonstration initiative (PDI), qualification Session No. 6102, in accordance with the 1992 Edition,1993 Addenda of ASME Boiler and Pressure Vessel Code,Section XI, Appendix Vill requirements. Appendix Vill was developed to ensure the effectiveness of UT examinations within the nuclear industry by means of a rigorous, item-specific, performance demonstration. The performance demonstration was conducted on an RPV mock-up containing flaws of various sizes and locations. The demonstration estabiished the capability of equipment, procedures, and personnel, which are similar or the same as those that will be used at PBAPS, Unit 2, to find flaws that could be detrimental to the integrity of the i
I RPV.
Greater than 90% of the portion of each longitudinal weld within the beltline region will be scanned. Hence the welds that are most susceptible to radiation embrittlement will receive greater than 90% examination.
The staff concludes from the results of previous volumetric examinations and the extent of the examination proposed by the licensee that there is reasonable assurance that any degradation mechanism that could exist in the welds would be detected during the proposed volumetric examination. Hence, the licensee's proposed altemative provides reasonable assurance of structuralintegrity, and provides an acceptable level of quality and safety.
6.0 CONCLUSION
Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the staff evaluated the licensee's proposed attemative of examining reacior vessel shell welds specified in item B1.12 of Examination Category B-A
" Pressure Retaining Welds in Reactor Vessel,"in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of the ASME Code,Section XI and the augmented examination requirements in 10 CFR 50.55a(g)(6)(ii)(A)(2). The staff has determined that the licensee is unable to completely comply with the requirements, and that the proposed altemative provides an acceptable level of quality and safety. The licensee's proposed attemative is authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5).
Attachment:
Table Principal Contributor. H. Conrad Date: October 22, 1998
,