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| ==2.0 BACKGROUND== | | ==2.0 BACKGROUND== |
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| During routine system inspections, r embers of the Wisconsin Electric Power Company (the licensee) staff discovered a 4-gallon per minute leak in a branch connection off the west header of the Service Water System (SWS) at the Point Beach Nuclear Power Plant. The leak is located approximately 18 inches off the west header on an elbow which is a part of 6-inch carbon steel piping that supplies service water to the decontamination area. By letter dated July 19, 1996, the licensee requested relief from the ASME Code Section XI repair or replacement requirements under impracticality provisions of 10 CFR 50.55(a). | | During routine system inspections, r embers of the Wisconsin Electric Power Company (the licensee) staff discovered a 4-gallon per minute leak in a branch connection off the west header of the Service Water System (SWS) at the Point Beach Nuclear Power Plant. The leak is located approximately 18 inches off the west header on an elbow which is a part of 6-inch carbon steel piping that supplies service water to the decontamination area. By {{letter dated|date=July 19, 1996|text=letter dated July 19, 1996}}, the licensee requested relief from the ASME Code Section XI repair or replacement requirements under impracticality provisions of 10 CFR 50.55(a). |
| The licensee based its request for relief on the results of a flaw evaluation that was performed by the licensee in accordance with the guidelines and acceptance criteria contained in GL 90-05. | | The licensee based its request for relief on the results of a flaw evaluation that was performed by the licensee in accordance with the guidelines and acceptance criteria contained in GL 90-05. |
| 3.0 LICENSEE'S RELIEF RE00EST 3.1 Components for Which Relief is Reouested ASME Code Class 3 service water system piping (six inch elbow). | | 3.0 LICENSEE'S RELIEF RE00EST 3.1 Components for Which Relief is Reouested ASME Code Class 3 service water system piping (six inch elbow). |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20212D5961999-09-15015 September 1999 Safety Evaluation Supporting Licensee IPEEE Process.Plant Has Met Intent of Suppl 4 to GL 88-20 ML20196J4251999-06-30030 June 1999 Safety Evaluation Authorizing Proposed Alternatives Described in Relief Requests VRR-01,ROJ-16,PRR-01 & VRR-02 ML20207D6751999-02-22022 February 1999 Assessment of Design Info on Piping Restraints for Point Beach Nuclear Plant,Units 1 & 2.Staff Concludes That Licensee Unable to Retrieve Original Analyses That May Have Been Performed to Justify Removal of Shim Collars ML20206R9001999-01-13013 January 1999 SER Accepting Nuclear Quality Assurance Program Changes for Point Beach Nuclear Plant,Units 1 & 2 ML20198C7671998-12-10010 December 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME BPV Code,1986 Edition,Section XI Requirement IWA-2232, to Use Performance Demonstration Initiative Program During RPV Third 10-yr ISI for Plant,Unit 2 ML20236Q3161998-07-10010 July 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME Code Requirements PTP-3-01 & PTP-3-02 ML20236L6771998-07-0707 July 1998 Safety Evaluation Approving Wepco Implementation Program to Resolve USI A-46 at Point Beach NPP Units 1 & 2 ML20217F3131998-04-17017 April 1998 Safety Evaluation Accepting Proposed Alternative to ASME Code for Surface Exam of Nonstructural Seal Welds,For Plant, Unit 1 ML20217A8501998-03-19019 March 1998 SER Accepting Proposed Changes Submitted on 980226 by Wiep to Pbnp Final SAR Section 1.8 Which Will Impact Commitments Made in Pbnp QA Program Description.Changes Concern Approval Authority for Procedures & Interviewing Authority ML20216J0101998-03-17017 March 1998 Safety Evaluation Accepting Third 10-yr Inservice Insp Interval Relief Request RR-1-18 for Plant ML20198L1151998-01-0808 January 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Point Beach Nuclear Plant,Units 1 & 2 ML20198L4671998-01-0202 January 1998 SER Approving Request for Relief VRR-4B to Inservice Testing Program Wisconsin Electric Power Co,Point Beach Nuclear Plant,Units 1 & 2 ML20197J9341997-12-12012 December 1997 Safety Evaluation Accepting Licensee Request for Relief from Performing Inservice Volmetric Exam of Inaccessible Portions of RPV Lower Shell to Lower Head Ring Weld During 10-yr ISI Interval of Plant,Unit 2 ML20137U4991997-04-10010 April 1997 Safety Evaluation Accepting Proposed Alternatives Contained in Requests for Relief RR-1-17 & RR-2-21 ML20129G6901996-10-0303 October 1996 SER Accepting Request for Relief from ASME Code Repair Requirements for ASME Code Class Three Piping at Plant ML20062J4991993-10-28028 October 1993 Safety Evaluation Granting IST Relief Requests Per 10CFR50.55a(a)(3)(ii) & 10CFR50.55a(f)(4)(iv) ML20062F1361990-09-25025 September 1990 SE Accepting Util Responses to Generic Ltr 83-28,Item 1.2, Post-Trip Review - Data & Info Capability ML20248A0101989-09-18018 September 1989 Safety Evaluation Re Containment Liner Leak Chase Channel Venting.Concurs W/Licensee That Plant Does Not Need to Vent Containment Liner Weld Leak Chase Channels During Test ML20246H0121989-07-0707 July 1989 Safety Evaluation Accepting Util 880325 & 1117 Responses to NRC Bulletin 88-002, Rapidly Propagating Fatigue Cracks in Steam Generator Tubes ML20245B0311989-06-14014 June 1989 Safety Evaluation Supporting Util Response to Generic Ltr 83-28,Item 4.5.3 Re on-line Functional Testing of Reactor Trip Sys.Existing Intervals for on-line Functional Testing Consistent W/High Reactor Trip Sys Availability ML20207E4191988-08-0404 August 1988 Safety Evaluation Supporting Compliance W/Atws Rule 10CFR50.62, Requirements for Reduction of Risk from ATWS Events for Light Water Cooled Nuclear Power Plants ML20151R6771988-08-0202 August 1988 Safety Evaluation Granting Request for Relief from ASME Code,Section XI Evaluation Requirements ML20151N2191988-07-27027 July 1988 Safety Evaluation Supporting Util Proposal Re Design of Switchgear Room,Per Sections Iii.G & Iii.L of App R to 10CFR50 ML20150C1311988-06-21021 June 1988 Safety Evaluation Accepting Responses to Generic Ltr 83-28, Item 2.1,confirming That Program Exists for Identifying, Classifying & Treating Components Required for Performance of Reactor Trip Function as safety-related ML20154H5791988-05-12012 May 1988 Safety Evaluation Supporting Conclusions That Rev 1 to Offsite Dose Calculation Manual (ODCM) Uses Methods Consistent W/Staff Requirements,However Some Discrepancies Identified.Odcm & Environ Manual Should Be Revised ML20148H4551988-03-24024 March 1988 Safety Evaluation Accepting Util 840405 Response to Generic Ltr 83-28,Item 2.1,(Part 2) Re Vendor Interface Programs & Reactor Trip Sys Components ML20235K9241987-07-0909 July 1987 Safety Evaluation Re Reactor Pressure Vessel Flaw.Flaw Conditionally Acceptable Per Subarticle IWB-3123 of Section XI of ASME Code & Therefore Requires Augmented Inservice Insps Based on 10CFR50.55(g)(4) ML20213G5801987-05-0707 May 1987 Safety Evaluation Re Util 861027 Request for Relief from Exam Requirements of Section XI of ASME Boiler & Pressure Vessel Code for Shell & Nozzle Welds in Regenerative Hxs. Request Granted ML20206K6011987-04-10010 April 1987 SER Supporting Util 860513 Proposed Replacement of Hydraulic Snubbers W/Energy Absorbers on Main Steam Bypass Line ML20210P2781987-02-0505 February 1987 Safety Evaluation Supporting Util 831107 & 860411 Responses to Generic Ltr 83-28,Item 4.5.2 Re Reactor Trip Sys Reliability on-line Testing.Plant Designed to Permit on-line Functional Testing of Diverse Trip Features of Breakers ML20214U6081986-11-26026 November 1986 Safety Evaluation Supporting Util 850516 Capsule T Summary Rept Re Use of Reactor Vessel Pressure Temp Limits Specified in Figures 15.3.1-1 & 15.3.1-2 of Tech Specs.Temp Limits Valid & May Continue to Be Used ML20206S7091986-09-16016 September 1986 Safety Evaluation on Util 850426 Response to Open Items Re Generic Ltr 81-14, Seismic Qualification of Auxiliary Feedwater Sys (Afws). Reasonable Assurance Exists That Afws Will Perform Required Safety Function Following SSE ML20214L9311986-09-0404 September 1986 Corrected Safety Evaluation Re Projected Values of Matl Properties for Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.Licensee Projections Acceptable ML20207D6781986-07-11011 July 1986 Safety Evaluation Accepting Util Responses to Generic Ltr 82-33 Re post-accident Monitoring Instrumentation Compliance W/Guidelines of Reg Guide 1.97,Rev 2,subj to Listed Condition.Portions of Rev 1 to EGG-EA-6771 Encl ML20138N7801985-10-31031 October 1985 Safety Evaluation Granting Util 840706 Relief Requests for Second 10-yr Inservice Insp Interval.Review of Requests for Relief from ASME Code Section XI Requirements Summarized in Encl Tables ML20134A4821985-10-24024 October 1985 Safety Evaluation Supporting Util Response to Generic Ltr 83-28,Items 3.1.1,3.1.2,4.1 & 4.5.1 Re post-maint Testing (Reactor Trip Sys Components) & Reactor Trip Sys Reliability.Programs Outlined in Acceptable ML20134A6051985-10-22022 October 1985 Safety Evaluation Re Util 831107 & 850910 Responses to Generic Ltr 83-28,Item 1.1, Post-Trip Review Program Description & Procedures. Program & Procedures Acceptable ML20138H1721985-10-18018 October 1985 Safety Evaluation Accepting Util 831107 Response to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing ML20133G4171985-07-29029 July 1985 Safety Evaluation Accepting Util 831108 Response to Generic Ltr 83-28,Item 1.1 Re post-trip Review.Response to Listed Deficiencies,Including Development of Systematic Safety Assessment Program for Unscheduled Reactor Trips Required ML20129H7871985-05-16016 May 1985 Safety Evaluation Supporting Licensee Response to Generic Ltr 83-28,Items 4.2.1 & 4.2.2 Re Reactor Trip Sys Reliability,Provided Corrective Action Taken If Higher than Normal Valves Observed in Trip Force & Response Time Values ML20205H2171984-09-10010 September 1984 Supplemental Safety Evaluation Re Util 820820 & 860113 Requests for Relief from Inservice Insp Requirements. Volumetric Exam Acceptable Method for Detecting O.D. Initiated Flaws.Relief from Surface Exams Should Be Granted ML20204F5381983-04-25025 April 1983 Safety Evaluation of Util Preferred Ac Power Sys Conformance GDC 17.Proximity of Low Voltage Transformers Does Not Fully Meet GDC 17 Requirements for Physical Separation,But Deluge Sprinkler Sys Adequate 1999-09-15
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARNPL-99-0569, Monthly Operating Repts for Sept 1999 for Pbnp,Units 1 & 2. with1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pbnp,Units 1 & 2. with ML20212D5961999-09-15015 September 1999 Safety Evaluation Supporting Licensee IPEEE Process.Plant Has Met Intent of Suppl 4 to GL 88-20 NPL-99-0051, Monthly Operating Repts for Aug 1999 for Pbnp,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Pbnp,Units 1 & 2. with NPL-99-0449, Monthly Operating Repts for July 1999 for Pbnp,Units 1 & 2. with1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Pbnp,Units 1 & 2. with ML20196J4251999-06-30030 June 1999 Safety Evaluation Authorizing Proposed Alternatives Described in Relief Requests VRR-01,ROJ-16,PRR-01 & VRR-02 ML20209D2691999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Pbnps,Units 1 & 2 ML20196F3341999-06-22022 June 1999 Safety Evaluation for Implementation of 422V+ Fuel Assemblies at Pbnp Units 1 & 2 ML20195F9781999-06-10010 June 1999 Unit 2 Refueling 23 Inservice Insp Summary Rept for Form NIS-1 ML20209D2751999-05-31031 May 1999 Revised MORs for May 1999 for Pbnps,Units 1 & 2 NPL-99-0328, Monthly Operating Repts for May 1999 for Pbnp,Units 1 & 2. with1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Pbnp,Units 1 & 2. with NPL-99-0273, Monthly Operating Repts for Apr 1999 for Point Beach Nuclear Plant,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Point Beach Nuclear Plant,Units 1 & 2.With ML20196F3521999-04-30030 April 1999 Non-proprietary WCAP-14788, W Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wepc Point Beach Units 1 & 2 (Fuel Upgrade & Uprate to 1656 Mwt - NSSS Power) NPL-99-0193, Monthly Operating Repts for Mar 1999 for Pbnp,Units 1 & 2. with1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Pbnp,Units 1 & 2. with NPL-99-0134, Monthly Operating Repts for Feb 1999 for Pbnp,Units 1 & 2. with1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Pbnp,Units 1 & 2. with ML20207D6751999-02-22022 February 1999 Assessment of Design Info on Piping Restraints for Point Beach Nuclear Plant,Units 1 & 2.Staff Concludes That Licensee Unable to Retrieve Original Analyses That May Have Been Performed to Justify Removal of Shim Collars ML20206R9001999-01-13013 January 1999 SER Accepting Nuclear Quality Assurance Program Changes for Point Beach Nuclear Plant,Units 1 & 2 NPL-99-0008, Monthly Operating Repts for Dec 1998 for Pbnp,Units 1 & 2. with1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Pbnp,Units 1 & 2. with NPL-99-0091, 1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with1998-12-31031 December 1998 1998 Annual Results & Data Rept for Pbnps,Units 1 & 2. with ML20198C7671998-12-10010 December 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME BPV Code,1986 Edition,Section XI Requirement IWA-2232, to Use Performance Demonstration Initiative Program During RPV Third 10-yr ISI for Plant,Unit 2 NPL-98-1006, Monthly Operating Repts for Nov 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Point Beach Nuclear Plant,Units 1 & 2.With ML20195J5101998-11-16016 November 1998 Proposed Revs to Section 1.3 of FSAR for Pbnp QA Program ML20198J5941998-11-0303 November 1998 1998 Graded Exercise,Conducted on 981103 NPL-98-0948, Monthly Operating Repts for Oct 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Point Beach Nuclear Plant,Units 1 & 2.With NPL-98-0880, Special Rept:On 980913,fire Alarm Control Panels Inoperable for More That Fourteen Days.Troubleshooting of D-401 Panel Following Installation of Replacement Batteries Revealed No Apparent Cause for Spurious Alarms.Panel D-401 Restored1998-10-21021 October 1998 Special Rept:On 980913,fire Alarm Control Panels Inoperable for More That Fourteen Days.Troubleshooting of D-401 Panel Following Installation of Replacement Batteries Revealed No Apparent Cause for Spurious Alarms.Panel D-401 Restored ML20154M9121998-10-14014 October 1998 Unit 1 Refueling 24 Repair/Replacement Summary Rept for Form NIS-2 ML20154L6751998-10-14014 October 1998 Unit 1 Refueling 24 ISI Summary Rept for Form NIS-1 NPL-98-0826, Monthly Operating Repts for Sept 1998 for Point Beach Nuclear Plant,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Point Beach Nuclear Plant,Units 1 & 2.With ML20151W3851998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Pbnp Units 1 & 2 NPL-98-0653, Monthly Operating Repts for July 1998 for Point Beach Nuclear Plant,Units 1 & 21998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W4471998-07-31031 July 1998 Corrected Page to MOR for July 1998 for Pbnp Unit 2 ML20151W4541998-07-31031 July 1998 Corrected Page to MOR for July 1998 for Pbnp Unit 1 ML20236Q3161998-07-10010 July 1998 Safety Evaluation Accepting Licensee Proposed Alternative to ASME Code Requirements PTP-3-01 & PTP-3-02 ML20236L6771998-07-0707 July 1998 Safety Evaluation Approving Wepco Implementation Program to Resolve USI A-46 at Point Beach NPP Units 1 & 2 NPL-98-0558, Monthly Operating Repts for June 1998 for Pbnp,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Pbnp,Units 1 & 2 ML20151W4261998-06-30030 June 1998 Corrected Page to MOR for June 1998 for Pbnp Unit 2 ML20151W4221998-05-31031 May 1998 Corrected Page to MOR for May 1998 for Pbnp Unit 2 NPL-98-0481, Monthly Operating Repts for May 1998 for Point Beach Nuclear Plant,Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W4011998-04-30030 April 1998 Corrected Page to MOR for April 1998 for Pbnp Unit 2 NPL-98-0356, Monthly Operating Repts for April 1998 for Point Beach Nuclear Plant,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for April 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20217F3131998-04-17017 April 1998 Safety Evaluation Accepting Proposed Alternative to ASME Code for Surface Exam of Nonstructural Seal Welds,For Plant, Unit 1 ML20216D7071998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W3981998-03-31031 March 1998 Corrected Page to MOR for March for Pbnp Unit 2 NPL-98-0209, Special Rept Re Fire Barrier Inoperable for Greater than Seven Days.Compensatory Measures Implemented in Accordance W/Fire Protection Program Requirements During Time That Barriers Were Inoperable1998-03-30030 March 1998 Special Rept Re Fire Barrier Inoperable for Greater than Seven Days.Compensatory Measures Implemented in Accordance W/Fire Protection Program Requirements During Time That Barriers Were Inoperable ML20217A8501998-03-19019 March 1998 SER Accepting Proposed Changes Submitted on 980226 by Wiep to Pbnp Final SAR Section 1.8 Which Will Impact Commitments Made in Pbnp QA Program Description.Changes Concern Approval Authority for Procedures & Interviewing Authority ML20216J0101998-03-17017 March 1998 Safety Evaluation Accepting Third 10-yr Inservice Insp Interval Relief Request RR-1-18 for Plant NPL-98-0159, Monthly Operating Repts for Feb 1998 for Point Beach Nuclear Plant,Units 1 & 21998-02-28028 February 1998 Monthly Operating Repts for Feb 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20151W3891998-02-28028 February 1998 Corrected Page to MOR for Feb 1998 for Pbnp Unit 2 ML20216D7121998-02-28028 February 1998 Revised Corrected MOR for Feb 1998 for Point Beach Nuclear Plant,Unit 2 NPL-98-0084, Monthly Operating Repts for Jan 1998 for Point Beach Nuclear Plant,Units 1 & 21998-01-31031 January 1998 Monthly Operating Repts for Jan 1998 for Point Beach Nuclear Plant,Units 1 & 2 ML20198L1151998-01-0808 January 1998 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Point Beach Nuclear Plant,Units 1 & 2 1999-09-30
[Table view] |
Text
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psun g- -4 UNITED STATES g
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2006t4001
\...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF FROM ASME CODE REPAIR RE0VIREMENTS FOR ASME CODE CLASS 3 PIPING WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301
1.0 INTRODUCTION
Title 10 of the Code of Federal Reaulations, Section 50.55a(g) requires nuclear power facility piping and components to meet the applicable requirements of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (hereafter referred to as the Code). ,
Section XI of the Code specifies Code-acceptable repair methods for flaws that .
exceed Code acceptance limits in piping that is in-service. A Code repair is required to restore the structural integrity of flawed Code piping, independent of the operational mode of the plant when the flaw is detected. '
Those repairs not in compliance with Section XI of the Code are non-Code repairs. However, the implementation of required Code (weld) repairs to ASME Code Class 1, 2 or 3 systems is often impractical for nuclear licensees since :
the repairs normally require an isolation of the system requiring the repair, !
and often a shutdown of the nuclear power plant.
Alternatives to Code requirements may be used by nuclear licensees when (
authorized by the Director of the Office of Nuclear Reactor Regulation if the proposed alternatives to the requirements are such that they are shown to provide an acceptable level of quality and safety in lieu of the Code ;
requirements [10 CFR 50.55a(a)(3)(1)], or if compliance with the Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety [10 CFR 50.55a(a)(3)(ii)].
A licensee may also submit requests for relief from certain Code requirements when a licensee has determined that conformance with certain Code requirements ;
is impractical for its facility [10 CFR 50.55a(g)(5)(iii)]. Pursuant to ,
10 CFR 50.55a(g)(6)(1), the Cornission will evaluate determinations of !
1mpracticality and may grant relief and may impose alternative requirements as l it determines is authorized by law.
Generic Letter (GL) 90-05, entitled " Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2 and 3 Piping," and dated June 15, 1990, provides guidance for the staff in evaluating relief requests' submitted by 9610070265 961003 PDR ADOCK 05000266 P PDR
~
1 l
(
l
! licensees for temporary non-Code repairs of Code Class 3 piping. The staff uses the guidance in GL 90-05 as its criteria for making its safety evaluation of relief requests for temporary non-Code repairs of Code Class 3 piping.
2.0 BACKGROUND
During routine system inspections, r embers of the Wisconsin Electric Power Company (the licensee) staff discovered a 4-gallon per minute leak in a branch connection off the west header of the Service Water System (SWS) at the Point Beach Nuclear Power Plant. The leak is located approximately 18 inches off the west header on an elbow which is a part of 6-inch carbon steel piping that supplies service water to the decontamination area. By letter dated July 19, 1996, the licensee requested relief from the ASME Code Section XI repair or replacement requirements under impracticality provisions of 10 CFR 50.55(a).
The licensee based its request for relief on the results of a flaw evaluation that was performed by the licensee in accordance with the guidelines and acceptance criteria contained in GL 90-05.
3.0 LICENSEE'S RELIEF RE00EST 3.1 Components for Which Relief is Reouested ASME Code Class 3 service water system piping (six inch elbow).
3.2 Section XI Edition for Point Beach. Unit 1 and 2 1986 Edition of the ASME Code,Section XI.
3.3 ASME Section XI Code Reauirement lhe ASME Code Section XI requires that repairs or replacements of ASME Code Class components be performed in accordance with rules found in Articles IWA-4000 or IWA-7000, respectively. The intent of these rules is to provide an i acceptable means of restoring the structural integrity of a degraded Code Class system back to the original design requirements.
3.4 Content of the Relief Reauest Relief is sought from performing a repair or replacement of the leaking six ;
inch elbow per the requirements of Article IWA-4000 or IWA-7000, respectively.
Relief is being sought until the next Point Beach Unit 2 scheduled outage which is scheduled to begin in October, 1996. At that time the licensee will perform a Code repair to return the system to compliance with the Code.
3.5 Basis for Relief 1
Request for relief has been submitted under the provisions of 10 CFR l 50.55a(a)(3)(ii) because compliance with the Code requirements by the licensee l in this case would result in hardship. The provisions of 10 CFR t
50.55a(a)(3)(ii) allow licensees to propose alternatives to Code requirements if " compliance with the . . . requirements . . . would result in hardship or
! unusual difficulty without a compensating increase in the level of quality and
safety." The licensee has completed a temporary repair using elastomer patch (1/8 inch block rubber) with a plastic backing plate (1/8 inch Teflon) held cnto the piping by 3/4 inch wide stainless steel straps. This temporary patch has effectively stopped the leak and there is no effect on system operability.
Also because the leak has been stopped there is no adverse effect on any safety-related equipment in the surrounding area.
3.6 Licensee's Alternative Proaram
- 1. Visual inspection of the patch to detect possible leakage will be performed weekly. Operators will be able to identify any significant leakage during nc'.1nal rounds.
- 2. The integrity of the piping will be assessed every three months until a permanent repair can be completed. Radiographs will be taken of the degraded area every three months to assess the piping.
- 3. An engineering evaluation will be performed of the strachral integrity of the system based on the results of the visual and radiographic examinations.
- 4. A Code repair will be completed during the next refueling outage which is scheduled to begin in October 1996.
4.0 STAFF EVALUATION AND CONCLUSIONS 4.1 Operability Determination. Root Cause Analysis and Structural Intearity Evaluation The licensee determined that the leakage was located in the SWS which is classified as an ASME Code Class 3 system. The leak is located approximately 18 inches off the west header on an elbow which is a part of the 6-inch carbon steel piping that supplies service water to the decontamination area. Upon discovery of the leak, the licensee performed an evaluation of the leak using the guidance provided in Generic Letter 90-05 and found that the leak met the criteria for a non-Code repair. The licensee determined that the operability of the system will not be impaired because the leak has been stopped using an elastomer patch with plastic backing plate held on the piping by 3/4 inch wide stainless steel straps. Because the leak has been stopped there will be no adverse affect on any other safety-related equipment in the surrounding area.
The licensee performed a root cause analysis of the flaw, and determined that the degradation resulted from microbiological 1y-induced corrosion (MIC). The leaking area was located on an elbow which is a part of six inch carbon steel piping. The flaw was inspected using ultrasonic and radiographic examinations to assess overall degradation of the affected elbow. The examination of the area of leakage indicated that the through-wall defect is localized. An evaluation of the ultrasonic and radiographic examination results and calculation of structural integrity confirmed that the flawed piping satisTies the "through-wall flaw" approach described in the NRC Generic Letter 90-05.
_ _ _ _ _ _ . _ _ _ _ _ ___ _ _ _ _ _ _ . - _ _ ~ .
i i
! 4 l 4.2 Auamented Inspection-1 .
^
The leak was located on an elbow located on the SWS. The area of leakage was !
ultrasonically and radiographically examined to assess the leak. No further ;
i areas of degradation were observed and the leak was determined to be localized !
- and caused by MIC. An augmented inspection was performed on five other l similar locations on the SWS in accordance with the guidance provided in Generic Letter 90-05. Two of the five augmented inspection locations and i inspections of additional locations have been completed. The examination 3
results identified only minimal thinning with no location below the minimum i
- required pipe wall thickness. '
4.3 Prooosed Temocrary Non-Code Renair and Monitorina Provisions At this time, the licensee is monitoring the leak every week. The licensee has commiitted to perform radiographic examination of the leakage area every three months to assess the elbow's wall degradation rate. Based on the ,
results of the examinations, an engineering evaluation will be performed to >
determine if further remedial measures or corrective actions are needed until the Code repair is completed.
4.4 Staff Conclusions The staff has determined that the licensee's flaw evaluation is consistent with the guidelines and acceptance criteria of GL 90-05. The staff therefore finds the licensees' structural integrity and operability assessments to be acceptable. The licensee has established a periodic inspection program to monitor flaw growth and ensure continued operability. The licensee is also monitoring the leak by visual inspection every week to ascertain that the leak has been stopped using a temporary elastomer patch. The licensee's actions constitute'an acceptable temporary alternative to the Code requirements.
Furthermore, the staff finds that performance of an immediate Code repair would constitute an undue burden (create undue hardship) upon the licensee since it would require an isolation of the affected SWS piping. Such an isolati n is not in the best interest of plant safety, given the magnitude of the leak and the licensee's alternative program. The staff therefore concludes that the licensee's proposed alternatives to the requirements would i provide an acceptable level of quality or safety, is authorized by law and i will not endanger life or property or the common defense and security,' and is i otherwise in the public interest, given due consideration to the burden upon the licensee and facility that could result if the Code requirements were imposed on the facility. Pursuant to 10 CFR 50.55a(a)(3)(11) the alternative is authorized until the next scheduled outage exceeding 30 days, but no later than the next refueling outage. At that time a Code repair will be ~ performed.
Principal Contributor: G. Georgiev, DE/ECGB Date: October 3, 1996
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