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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20203F4511999-02-12012 February 1999 SER Finding Licensee Adequately Addressed GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves, for Virgil C Summer Nuclear Station ML20198F4241998-12-18018 December 1998 Safety Evaluation Granting Relief Request for Approval to Repair ASME Code Class 3 Service Water Piping Flaws in Accordance with GL 90-05 for VC Summer Nuclear Station ML20155G4551998-11-0404 November 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Use Code Case N-416-1 with Licensee Proposed Addl Exams ML20154Q9571998-10-21021 October 1998 SER Accepting Request Seeking Approval to Use Alternative Rules of ASME Code Case N-498-1 for Class 1,2 or 3 Sys ML20237A7181998-08-13013 August 1998 SER Accepting Util Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20248J0191998-06-0404 June 1998 Safety Evaluation Accepting Licensee Inservice Testing Program Interim Pump Relief Request Per 10CFR50.55a(a)(3) (II) ML20212C0381997-10-19019 October 1997 Safety Evaluation Accepting License Request for Deviation from Commitment to Meet Section III.G.2.c of App R to 10CFR50 Re Fire Protection of Safe Shutdown Capability for Plant ML20217E3491997-09-22022 September 1997 Safety Evaluation Authorizing Licensee Proposed Alternative for Current Interval Insp Program Plan ML20133J5551997-01-15015 January 1997 Safety Evaluation Granting Licensee Request Proposing Not to Perform Increased Frequency Testing on a Charging Pump at Virgil C Summer Nuclear Station ML20128G2931996-10-0202 October 1996 Safety Evaluation Supporting Amend 135 to License NPF-12 ML20128F4221993-02-0909 February 1993 Safety Evaluation Re Nuclear Physics Methodology for Reload Design.Request to Perform Reload Analyses Approved ML20056A7931990-08-0606 August 1990 Safety Evaluation Re Capability of Plant to Meet Requirements of Branch Technical Position Rsb 5-1, Design Requirements of RHR Sys. Design Satisfies License Condition 4 ML20245F5061989-06-22022 June 1989 Safety Evaluation Re Request for Relief from Section XI Re Hydrostatic Test Requirement ML20244D7361989-06-12012 June 1989 Safety Evaluation Accepting Util Response to Generic Ltr 83-28,Item 4.5.3 Re Reactor Trip Sys Reliability for All Domestic Operating Reactors ML20195B4421988-10-28028 October 1988 Safety Evaluation Supporting Amend 74 to License NPF-12 ML20151K0901988-07-28028 July 1988 Safety Evaluation Supporting Util Proposed Implementation of ATWS Rule Pending Resolution of Tech Spec Issue ML20151K7771988-07-27027 July 1988 Safety Evaluation Supporting Util Request to Deviate from Recommendations of Reg Guide 1.97 Re Instrumentation to Monitor Containment Temp ML20151R8561988-04-19019 April 1988 Safety Evaluation Supporting Related Inservice Testing Program & Request for Relief of Utils ML20236R4111987-11-13013 November 1987 Safety Evaluation Supporting Conformance to Reg Guide 1.97, Rev 3 ML20236K7701987-11-0505 November 1987 SER Accepting Util 831104 & 870401 Responses to Item 2.2.1 of Genreic Ltr 83-28 Re Equipment Classification Programs ML20237H3661987-07-22022 July 1987 Corrected Page to Safety Evaluation Issued W/Amend 67, Changing Second Paragraph & Deleting Third Paragraph on Page Three ML20214S8881987-06-0303 June 1987 Safety Evaluation Rept Granting Relief from Hydrostatic Testing After Repair to ASME Code Section Xi,Class 1,reactor Coolant Pump Seal Injection Line ML20209H3331987-01-30030 January 1987 SER Supporting Util 831104 Response to Generic Ltr 83-28, Item 4.5.2 Re on-line Testing of Reactor Trip Sys Reliability ML20213A5611987-01-30030 January 1987 SER Accepting Util 831104 Response to Generic Ltr 83-28,Item 2.1 (Part 2) Re Vendor Interface Program for Reactor Trip Sys Components ML20212F2841986-12-22022 December 1986 Safety Evaluation Supporting Amend 57 to License NPF-12 ML20211M4161986-12-0909 December 1986 Safety Evalution Supporting Licensee 860123 Submittals Re Fast Neutron Fluence for Fracture Toughness Requirements for Protection Against PTS Events (10CFR50.61) ML20203N0151986-09-15015 September 1986 Safety Evaluation Re Generic Ltr 83-28,Items 3.1.3 & 3.2.3, Post-Maint Testing (RTS Components,All Other Safety-Related Components). Response Acceptable ML20199D4211986-06-0909 June 1986 SER on Util 831104 & 860423 Responses to Generic Ltr 83-28, Item 1.2 Re post-trip Review Data & Info Capabilities.Data & Info Capabilities Acceptable ML20211A2571986-05-22022 May 1986 Safety Evaluation Accepting Mods to App R,Clarified by Generic Ltrs 81-12 & 83-33,to Prevent Spurious Equipment Operation Caused by fire-induced Conductor or Cable Faults, Facilitate Operator Actions & Resolve Addl Circuit Concerns ML20154A0621986-02-24024 February 1986 Safety Evaluation Supporting 850930 & 1204 Responses to 850802 & 1104 Requests,Respectively,For Addl Info Re Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events ML20154D1921986-02-14014 February 1986 Sser 1 Re Licensee 851204 Response to Generic Ltr 83-28, Item 3.2.2 Concerning Procedures & Programs to Review Info on safety-related Equipment.Response Acceptable & Meets Intent of Generic Ltr 83-28 ML20136B2291985-11-0707 November 1985 Safety Evaluation Supporting Amend 46 to License NPF-12 ML20209H8411985-11-0404 November 1985 Safety Evaluation Re Response to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2,4.1 & 4.5.1.Response to Item 3.2.2 Incomplete & Addl Info Required ML20137S5781985-09-24024 September 1985 SER Approving Licensee 831104 & 0715 Responses to Generic Ltr 83-28,Item 1.1 Re post-trip Review.Program & Procedures for Restart from Unscheduled Reactor Trip Acceptable ML20133H7321985-08-0202 August 1985 SER Re Util 831104 Response to Generic Ltr 83-28,Items 4.2.1 & 4.2.2 Concerning Reactor Trip Sys Reliability. Licensee Should Add Undervoltage Trip Attachment to Trending Program ML20128A2181985-06-21021 June 1985 SER of Util Response to Generic Ltr 83-28,Item 1.1 Re post-trip Review.Listed Addl Info Required Before Review Can Be Completed 1999-02-12
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217D6401999-10-31031 October 1999 Rev 2 to WCAP-15102, VC Summer Unit 1 Heatup & Cooldown Limit Curves for Normal Operation RC-99-0202, Monthly Operating Rept for Sept 1999 for VC Summer Nuclear Station.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for VC Summer Nuclear Station.With ML20216J4191999-09-24024 September 1999 Part 21 Rept Re 990806 Abb K-Line Breaker Defect After Repair.Vendor Notified of Shunt Trip Wiring Problem & Agreed to Modify Procedure for Refurbishment of Breakers RC-99-0180, Special Rept on 990807,electric Driven Fire Pump XPP0134A Was Declared Inoperable.Caused by Pump Discharge Relief Valve Failing to Open as Normally Expected.Two Temporary Fire Pumps Were Installed to Provide Backup Suppression1999-09-0808 September 1999 Special Rept on 990807,electric Driven Fire Pump XPP0134A Was Declared Inoperable.Caused by Pump Discharge Relief Valve Failing to Open as Normally Expected.Two Temporary Fire Pumps Were Installed to Provide Backup Suppression RC-99-0183, Monthly Operating Rept for Aug 1999 for Virgil C Summer Nuclear Station,Unit 1.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Virgil C Summer Nuclear Station,Unit 1.With ML20211K6161999-08-31031 August 1999 Rev 2 to VC Summer Nuclear Station,Colr for Cycle 12, Dtd Aug 1999 RC-99-0168, Special Rept:On 990804,electric Driven Fire Pump XPP0134A & Diesel Driven Fire Pump XPP0134B,were Removed from Svc to Allow for Plant Mod.Fire Pumps Were Returned to Operable Condition on 990818,after Mod Was Completed1999-08-19019 August 1999 Special Rept:On 990804,electric Driven Fire Pump XPP0134A & Diesel Driven Fire Pump XPP0134B,were Removed from Svc to Allow for Plant Mod.Fire Pumps Were Returned to Operable Condition on 990818,after Mod Was Completed ML20210M7071999-07-31031 July 1999 Rev 1 to VC Summer Nuclear Station COLR for Cycle 12 ML20211C2201999-07-31031 July 1999 Rev 1 to WCAP-15102, VC Summer Unit 1 Heatup & Cooldown Limit Curves for Normal Operation RC-99-0163, Monthly Operating Rept for July 1999 for VC Summer Nuclear Station,Unit 1.With1999-07-31031 July 1999 Monthly Operating Rept for July 1999 for VC Summer Nuclear Station,Unit 1.With RC-99-0137, Monthly Operating Rept for June 1999 for VC Summer Nuclear Station,Unit 1.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for VC Summer Nuclear Station,Unit 1.With RC-99-0122, Monthly Operating Rept for May 1999 for VC Summer Nuclear Station.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for VC Summer Nuclear Station.With ML20206H2971999-05-0505 May 1999 Part 21 Rept Re Common Mode Failure for magne-blast Breakers.Vc Summer Nuclear Station Utilizes These Breakers in Many Applications,Including 7.2-kV EDG Electrical Buses RC-99-0103, Monthly Operating Rept for Apr 1999 for VC Summer Nuclear Station.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for VC Summer Nuclear Station.With ML20206K2421999-04-30030 April 1999 Rev 0 to COLR for Cycle 12 for Summer Nuclear Station RC-99-0087, Part 21 Interim Rept (SSH 99-0001) Re 990218 Failure of Circuit Breaker Located in Cubicle 14 of XSW1DB to Close During Surveillance Testing.Caused by Positive Interlock Angle Was Incorrect.Breaker Was Returned to GE Factory1999-04-15015 April 1999 Part 21 Interim Rept (SSH 99-0001) Re 990218 Failure of Circuit Breaker Located in Cubicle 14 of XSW1DB to Close During Surveillance Testing.Caused by Positive Interlock Angle Was Incorrect.Breaker Was Returned to GE Factory RC-99-0083, Monthly Operating Rept for Mar 1999 for VC Summer Nuclear Station,Unit 1.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for VC Summer Nuclear Station,Unit 1.With RC-99-0063, Special Rept:On 990302 & 16,meteorological Site Number One 10 Meter Temp Element Was Declared Inoperable.Caused by Erratic Operation.Cabling & 10 Meter Electrical Connectors Were Replaced1999-03-26026 March 1999 Special Rept:On 990302 & 16,meteorological Site Number One 10 Meter Temp Element Was Declared Inoperable.Caused by Erratic Operation.Cabling & 10 Meter Electrical Connectors Were Replaced ML20196K5421999-03-22022 March 1999 Rev 2 to VC Summer Nuclear Station,Training Simulator Quadrennial Certification Rept,1996-99, Books 1 & 2. Page 2 of 2 Section 2.4.4 (Rev 2) of Incoming Submittal Were Not Included RC-99-0055, Special Rept:On 990302,Meteorological Site Number One 10 Meter Temp Element (RTD) Was Declared Inoperable Due to Erratic Operation.Cause of Original RTD Failure Is Unknown. Equipment Was Declared Operable on 9903121999-03-16016 March 1999 Special Rept:On 990302,Meteorological Site Number One 10 Meter Temp Element (RTD) Was Declared Inoperable Due to Erratic Operation.Cause of Original RTD Failure Is Unknown. Equipment Was Declared Operable on 990312 RC-99-0050, Monthly Operating Rept for Feb 1999 for VC Summer Nuclear Station,Units 1.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for VC Summer Nuclear Station,Units 1.With ML18106B0931999-02-25025 February 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Caused by Crack Due to Improper Location of Heated Bar.Only One Part Out of 7396 Pieces in Forging Lot Was Found to Be Cracked.Affected Util,Notified ML20203F4511999-02-12012 February 1999 SER Finding Licensee Adequately Addressed GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves, for Virgil C Summer Nuclear Station ML18106B0551999-02-0101 February 1999 Part 21 Rept Re Possible Matl Defect in Swagelok Pipe Fitting Tee,Part Number SS-6-T.Defect Is Crack in Center of Forging.Analysis of Part Is Continuing & Further Details Will Be Provided IAW Ncr Timetables.Drawing of Part,Encl ML18106B0441999-01-29029 January 1999 Part 21 Rept Re Possible Defect in Swagelok Pipe Fitting Tee Part Number SS-6-T.Caused by Crack in Center of Forging. Continuing Analysis of Part & Will Provide Details in Acoordance with NRC Timetables ML20206R5241998-12-31031 December 1998 Santee Cooper 1998 Annual Rept RC-99-0052, Vsns 1998 Annual Operating Rept. with1998-12-31031 December 1998 Vsns 1998 Annual Operating Rept. with RC-99-0004, Monthly Operating Rept for Dec 1998 for VC Summer Nuclear Station,Unit 1.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for VC Summer Nuclear Station,Unit 1.With ML20206R5191998-12-31031 December 1998 Scana Corp 1998 Annual Rept ML20198F4241998-12-18018 December 1998 Safety Evaluation Granting Relief Request for Approval to Repair ASME Code Class 3 Service Water Piping Flaws in Accordance with GL 90-05 for VC Summer Nuclear Station RC-98-0223, Special Rept 98-001:on 981130,steam Line High Range Gamma Monitor (RMG-19C) Was Declared Inoperable Due to Indeterminate Alarm.Caused by Failures in Detector & Meter Reset Circuitry.Established Preplanned Alternate Method1998-12-16016 December 1998 Special Rept 98-001:on 981130,steam Line High Range Gamma Monitor (RMG-19C) Was Declared Inoperable Due to Indeterminate Alarm.Caused by Failures in Detector & Meter Reset Circuitry.Established Preplanned Alternate Method RC-98-0222, Monthly Operating Rept for Nov 1998 for VC Summer Nuclear Station,Unit 1.With1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for VC Summer Nuclear Station,Unit 1.With ML20155G4551998-11-0404 November 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Use Code Case N-416-1 with Licensee Proposed Addl Exams RC-98-0208, Monthly Operating Rept for Oct 1998 for VC Summer Nuclear Station,Unit 1.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for VC Summer Nuclear Station,Unit 1.With ML20207J5701998-10-31031 October 1998 Non-proprietary Rev 1 to WCAP-14955, Probabilistic & Economic Evaluation of Rv Closure Head Penetration Integrity for VC Summer Nuclear Plant ML20154Q9571998-10-21021 October 1998 SER Accepting Request Seeking Approval to Use Alternative Rules of ASME Code Case N-498-1 for Class 1,2 or 3 Sys ML20154K7901998-09-30030 September 1998 Non-proprietary Rev 0 to WCAP-15101, Analysis of Capsule W from Sceg VC Summer Unit 1 Rv Radiation Surveillance Program RC-98-0184, Monthly Operating Rept for Sept 1998 for VC Summer Nuclear Station.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for VC Summer Nuclear Station.With ML20154K8041998-09-30030 September 1998 Non-proprietary Rev 0 to WCAP-15103, Evaluation of Pressurized Thermal Shock for VC Summer Unit 1 RC-98-0166, Monthly Operating Rept for Aug 1998 for VC Summer Nuclear Station,Unit 1.With1998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for VC Summer Nuclear Station,Unit 1.With ML20237A7181998-08-13013 August 1998 SER Accepting Util Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves RC-98-0153, Monthly Operating Rept for July 1998 for VC Summer Nuclear Station,Unit 11998-07-31031 July 1998 Monthly Operating Rept for July 1998 for VC Summer Nuclear Station,Unit 1 RC-98-0131, Monthly Operating Rept for June 1998 for VC Summer Nuclear Station1998-06-30030 June 1998 Monthly Operating Rept for June 1998 for VC Summer Nuclear Station ML20248J0191998-06-0404 June 1998 Safety Evaluation Accepting Licensee Inservice Testing Program Interim Pump Relief Request Per 10CFR50.55a(a)(3) (II) RC-98-0113, Monthly Operating Rept for May 1998 for VC Summer Nuclear Station,Unit 11998-05-31031 May 1998 Monthly Operating Rept for May 1998 for VC Summer Nuclear Station,Unit 1 RC-98-0100, Monthly Operating Rept for Apr 1998 for VC Summer Nuclear Station,Unit 11998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for VC Summer Nuclear Station,Unit 1 ML20217G7411998-04-22022 April 1998 Rev 1 to VC Summer Nuclear Station COLR for Cycle 11 RC-98-0076, Final Part 21 Rept Re a DG EG-B for Vsns,As Followup to .Power Control Svcs of Engine Sys,Inc Provided Response on 980318.Evaluation Concludes That Failure of EG-B Is one-time non-repeatable Failure1998-04-17017 April 1998 Final Part 21 Rept Re a DG EG-B for Vsns,As Followup to .Power Control Svcs of Engine Sys,Inc Provided Response on 980318.Evaluation Concludes That Failure of EG-B Is one-time non-repeatable Failure RC-98-0084, Monthly Operating Rept for Mar 1998 for Virgil C Summer Nuclear Station,Unit 11998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Virgil C Summer Nuclear Station,Unit 1 ML20212H1421998-03-0202 March 1998 Interim Part 21 Rept SSH 98-002 Re EG-B Unit That Was Sent to Power Control Svcs for Determination of Instability & Refurbishment of a Dg.Cause of Speed Oscillations Unknown. Completed Hot Bore Checks on Power Case 1999-09-08
[Table view] |
Text
. . - . - -- . . - _ _ ~_
arc p4 4 UNITED STATES j
i g NUCLEAR REGULATORY COMMISSION l o t WASHINGTON, D.C. 2066M001
. . . . . ,o l I
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION j RELIEF REQUEST FOR APPROVAL TO REPAIR ASME CODE CLASS 3 SERVICE WATER PIPING FLAWS IN ACCORDANCE WITH GENERIC LETTER 90-05 FOR ,
I SOUTH CAROLINA ELECTRIC AND GAS COMPANY ;
I VIRGIL C. SUMMER NUCLEAR STATION DOCKET NUMBER 50-395
1.0 INTRODUCTION
The Code of Federal Regulations,10 CFR 50.55a(g), requires nuclear power facility piping and i 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 cf the Code are non-Code repairs. However, implementing required Code (weld) repairs to ASME Code Class 1,2 or 3 systems is often impractical, since the repairs normally require isolating the system to be repaired, and often shutting down the nuclear power plant.
Licensees may u.ae alternatives to Code requirements when the U. S. Nuclear Regulatory Commission (NRC) authorizes this. The NRC may authorize this if the proposed alternatives to the requirements provide an acceptable level of quality and safety in lieu of the Code requirements (10 CFR 50.55a(a)(3)(i)], or if compliance with the Code requirements would l
result in hartjship or unusual difficulty without a compensating increase in the level of quality and safety [10 CFR 50.55a(a)(3)(ii)].
Licensees may also subinit requests for relief from certain Code requirements when they determine that conforming with certain Code requirements is impractical for its facility (10 CFR
, 50.55a(g)(5)(iii)]. Pursuant to 10 CFR 50.55a(g)(6)(i), the Commission will evaluate
! determinations of impracticality, and may grant relief and may impose alternative requirements
- as it determines is authorized by law.
Generic Letter (GL) 90-05," Guidance for Performing Temporay Non-Code Repair of ASME i Code Class 1,2 and 3 Piping," dated June 15,1990, provides guidance for the staff in evaluating licensee relief requests for temporary non-Code repairs of Code Class 3 piping. For
. the purpose of the GL, impracticality is defined to exist if the flaw detected during plant j operation is in a section of Class 3 piping that cannot be isolated for completing a Code repair within the time period permitted by the limiting condition for operation (LCO) of the affected
- 9812290122 981218 ADOCK 05000395 PDR j P PM i
l system as specified in the plant Technical Specifications (TS), and performance of Code repair necessitates a plant shutdown.
l
2.0 BACKGROUND
Licensee personnel found a through-wall defect in a 4-inch carbon steel, moderate energy, service water (SW) system pipe. South Carolina Electric & Gas Company (SCE&G) estimated the leakage to be about five drops per minute. The affected pipe is a branch connection which functions as a nuclear safety-related emergency make-up source from the SW system "A" Train to the component cooling water system surge tank. The leak exists in the heat-affected zone of ;
the butt weld which connects the 4-inch pipe to the weld-o-let on the adjacent 20-inch SW system piping. SCE&G installed a lightweight, temporary patch constructed of hose clamps and rubber gasket material over the defect to impede any leakage which might occur. SCE&G performed a structural analysis, and ultrasonically examined additional piping locations, as GL 90-05 requires. These measurements showed slight wall thinning, but none of the locations had a measured wall thickness that violated the calculated, minimum-allowable wall thickness.
SCE&G's May 13,1998, letter submitted a relief request in accordance with 10CFR50.55a(a)(3)(i) and GL 90-05, seeking NRC approval to delay repairing the ASME Code Class 3 piping through-wall defect. SCE&G determined that the time needed to complete the Class 3 Code repair would exceed the 72-hour TS LCO 3.7.4 action statement allowance. They estimated that the repair activity would take 88 hours0.00102 days <br />0.0244 hours <br />1.455026e-4 weeks <br />3.3484e-5 months <br /> to complete, and could not be performed without shutting down the "A" train SW system. Exceeding the SW system 72-hour LCO action l statement requires a plant shutdown.
3.0 LICENSEE'S RELIEF REQUEST SCE&G requested relief from repairing the SW piping through-wall defect in accordance with the requirements of ASME Code,Section XI, Article IWA-4000. SCE&G determined that performing the Code repair within the 72-hour SW systern LCO 3.7.4 Action Statement was impractical. SCE&G will make an ASME Code-approved repair by replacement during the next scheduled outage exceeding 30 days, or Refueling Outage (RFO) 11, which is currently scheduled for the Spring of 1999.
3.1 Licensee's Component identification The defective piping is 4-inch carbon steel, moderate energy (design conditions are 65 psig and 95*F), ASME Code Class 3 SW system pipe.
3.2 ASME Code,Section XI RequirementsSection XI,1989 edition, Subarticle IWA-4310 requires the following:
Defects shall be removed or reduced in size in accordance witn this Article. The component shall be acceptable for continued service if the resultant section thickness created by the cavity is equal to or greater than the minimum design thickness. If the resulting section thickness is reduced below the minimum des.gn thickness, the component shall be repaired in accordance with this Article. Alternatively, the component may be evaluated and accepted in accordance with the design rules of i
l 1
1 i
either the Cont'ruction Code, or Section Ill, when the Construction Code was not Section Ill. The repair program and the associated evaluation analyses shall be subject to review in accordance with IWA-4130(c).
3.3 Licensee's Proposed Alternatives Based on GL 90-05 SCE&G installed a lightweight, temporary patch constructed of hose clamps and rubber gasket material over the defect to impede any leakage which might occur. SCE&G performed an augmented inspection of five additanal piping locations, in accordance with GL 90-05. The licensee selected these locations because they are subjected to flow conditions similar to those at the defect. SCE&G used ASME Code methods to calculate the minimum allowable wall thickness for each location. SCE&G then took ultrasonic testing (UT) measurements at each l location in a band around the pipe circumference. These measurements showed slight wall l thinning, but none of the locations had a measured wall thickness that violated the calculated minimum allowable wall thickness. SCE&G halted the augmented inspection at the initial five locations since they did not identify any more flaws.
SCE&G Quality Services will qualitatively assess the patch at least once per week for any degradation in the patch or base piping. They will visually note any changes in the amount of leakage from the patch. SCE&G will perform an engineering evaluation when they note any changes. At least once every 3 months, SCE&G will remove the patch and evaluate the affected pipe for structuralintegrity. This will be accomplished by making UT measurements of the pipe wallin the affected area. SCE&G will perform an engineering evaluation of the results l to estimate the degradation, and prescribe remedial actions,if necessary. These activities will continue until the plant reaches Mode 5 as part of a scheduled 30-day outage. SCE&G will document these evaluations, and will place them into the plant records. SCE&G will promptly initiate a Code repair if structuralintegrity of the pipe cannot be assured.
3.4 Licensee's Basis for Relief The licensee's basis for relief is as follows:
- 1) An evaluation for the performance of a code repair resulted in a determination that the required time to complete the repair would be in excess of the time allowance of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> under the action statement for the Limiting Condition for Operation (LCO-3.7.4). The repair activity is estimated to take 88 hours0.00102 days <br />0.0244 hours <br />1.455026e-4 weeks <br />3.3484e-5 months <br /> to complete and cannot be performed without shutting down the "A" train SW System. This would resun :n "A"
, train ECCS [ emergency core cooling system) equipment inoperability in excess of l the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement of LCO 3.5.2.
i l 2) A "Through wall Flaw" Evaluation of flaw stability as outlined in GL 90-05 showed the
- flaw to be acceptable for continued operation with a stability parameter of 29 ksi*in"5 l versus the acceptance criteria of 35 ksi+in"5 i
a 3) An augmented inspection of five additionallocations was performed as required by
] GL 90-05. All five inspections measured acceptable wall thickness with respect to
- calculated minimum allowable wall thickness.
l j_ 4) The following is an evaluation based on failure of the pipe:
4
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4 a) The current leakage, estimated at 5 drops per minute, is minor and does not impact system flow requirements. The worst case leakage would occur if a postulated crack-break developed at the defect location. The size of this crack would be equivalent to a 5/8 inch hole and would conservatively result in a 95 gpm [ gallon per minute] leak. The SW system flow margin is approximately 2,000 gpm during norrnal operation and approximately 2,800 gpm during post accident operation. The postulated 95 gpm leak is well within these margins and its acceptability is further supported by an evaluation of the pump curve which shows an insignificant effect on total developed head at an additional 95 gpm of flow, b) There are no essential to safe shutdown components or equipment within the immediate area which would be affected by spraying water.
c) The pipe is not an anti-falldown concern because it is approximately 6 feet above the floor and would fall to the floor without impacting any essential to safe shutdown equipment.
d) Flooding is not a concern because the volume of the Tendon Jacking and Tcndon Access Gallery areas is the limiting factor in the design flood level of elevation 412'-8". When this area becomes full, the water levelin the Intermediate Building general area will begin to rise. A conservative leakage estimate from this 4 inch pipe assuming a catastrophic failure would be 3,000 gpm. Assuming Operator action to stop the leak within 30 minutes, the total volume of water would be 90,000 3allons which is less than the Tendon Jacking and Tendon Access Gallery volume of 168,000 gallons.
- 5) A least once per week the patch and its surrounding area will be visually inspected for degradation and leakage. At least once every three months the patch will be removed and UT measurements of the pipe wall in the affected area will be obtained and evaluated for degradation.
4.0 STAFF EVALUATION AND CONCLUSIONS 4.1 Root Cause Analysis. Structural Inteority Evaluation. and Operability Determination The licensee determined that one location on a 4-inch ASME Code Class 3 SW system line had a through-wall defect. The leak exists in the heat-affected zone of the butt weld connecting the 4-inch pipe to the weld-o-let on the adjacent 20-inch SW system piping. SCE&G believed that microbiologically induced corrosion caused the defect.
SCE&G evaluated the structural integrity of the flaweo piping in accordance with GL 90-05, and found that the flaw satisfies the GL 90-05 through-v.all criteria for non-Code repair. A visual inspection of the pipe indicated that the leak resulted from localized corrosion, and was not a crack. UT measurements around the defect indicated that the affected area was less than 3/16 inch in diameter. SCE&G's through-wall flaw evaluation, prepared using GL 90-05 guidance, showed the flaw to be acceptable for continued operation with a stability parameter of 29 ksi*in" versus the acceptance criteria of 35 ksiain" Wall thickness measurements on the affected 90-degree elbow revealed only the expected general wall thinning and the presence of i
corrosion pits. All measured wall thickness was above the 0.093 inch calculated minimum, except for the defect itself. SCE&G calculated the minimum allowable wall thickness using ASME Code methodology. This methodology included a stress intensification factor that accounted for the existing through-wall defect to conservatively ensure pipe integrity.
SCE&G determined the system operability was not impaired because the leakage wa.s minor, and did not impact system flow requirements. SCE&G estimated that worst-case leakage at the defect location would be about 95 gpm. The postulated 95 grm leak is well within system flow requirements. SCE&G also concluded that there is no essentw. to-safe-shutdown I components or equipment within the immediate area which spraying water would affect, and flooding was not a concern.
4.2 Auamented Inspection SCE&G ultrasonically examined five additional piping locations for evidence of leakage. These measurements showed slight wall thinning, but none of the locations had a measJred Wall thickness which violated the calculated, minimum-allowable wall thickness.
4.3 Proposed Temocrary Non-Code Repair end Monitorina Provisions SCE&G applied a soft patch, held in place by hose clamps, over the flawed area to temporarily repair the flawed line. SCE&G's structuralintegrity evaluation of the flawed piping indicated l that the piping is acceptable for temporary repair, as GL 90-05 permits. SCE&G will visually inspect the flawed area at least once per week for degradation and leakage, and will take and I evaluate UT measurements of the pipe wallin the affected area at least once every 3 months. )
4.4 Staff Conclusions The staff has determined that the licensee's flaw evaluation is consistent with GL 90-05 guidelines and acceptance criteria. The staff, therefore, finds the licensee's structural integrity and operability assessments to be acceptable. Furthermore, the staff finds that making an immediate Code repair is impractical for the V. C. Summer Nuclear Station, since the Class 3 l Code repair would require isolating the SW system for longer than the 72-hour TS LCO 3.7.4 action statement allowance. Exceeding the SW system 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO action statement requires a plant shutdown. SCE&G indicated that they will perform a Code replacement of the defective piping during the next scheduled outage exceeding 30 days, or during RFO 11, currently scheduled for the Spring of 1999. Therefore, in accordance with 10 CFR 50.55a(g)(6)(i), the staff finds that the Code requirements are impractical, grants the requested relief, and imposes the alternatives, until the next refueling outage (RFO 11) is complete. The relief granted is authorized by law and will not endanger life or property or the common defense and security and is,otherwise in public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Principal Contributor: M. Padovan Date: December 18, 1998 l-l
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