|
---|
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
- - - _ - - - - - . _ - _ . - - - . - - - - - - - - - ~ - - -
a Ct00 1 9
g* k UNITED STATES
NUCLEAR REGULATORY COMMISSION WASHINGTON. O.C. soseHooi k . . . . . /j ENCLOSURE 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ,
OF THE !
SECOND 10 YEAR INTERVAL INSERVICE IFSPECTION PR(XiRAM PLAN INTERIM REQUEST FOR 1ELIEF SOUTH CAROLINA CTRIC & GAS CO.
VIRGIL C. SUMME R NUCLEAR STATION. UNIT 1 DOCKE1 NUMBER: 50 395
1.0 INTRODUCTION
The Technical Specifications (TS) for V. C. Summer Nuclear Station. Unit 1.
state that the inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1. 2. and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel (B&PV) Code and applicable addenda as required by 10 CFR 50.55a(g) except where specific written relief has been 50.55a(g)(6)(1), 10 CFR granted50.55a(a) by(the Commission
- 3) states pursuanttotothe that alternatives 10 CFR requirements of paragraph (g) may be used, when authorized by the NRC. if (1) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with tie specified requirements would result in hardship or unusual difficultly without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4). ASME Code Class 1.-2. and 3 components (including supports) shall meet the requiremerts, except the design and access provisions and the Code.Section XI. " Rules pre service for Inservice examination requirements, Inspection setPower of Nuclear forth inPlant the ASME Components." to the extent practical within the limitations of design.
and materials of construction of the components. The regulations geometry, require t hat inservice examination of components and system pre.w re tests conducted during the first 10-year interval and subsecuent it tervcis co iy with the requirements in the latest edition and aodenca of SecHr. XI o@f tar ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120 month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for V.
C. Summer Nuclear Station. Unit 1. second 10-year inservice inspection (ISI) intervai is the 1989 Edition.
Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility. Information shall be submitted to the Comission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation ' the determination, pursuant to 10 CFR 50.55a(g)(6)(ih the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law, will not
- u 18sM Pe h G PDR
O 4
g 2
endanger life. property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
In a letter dated February 25. 1997 South Carolina Electric & Gas Co.
(licensee) submitted its request to use Code Case N 566 in lieu of the Code recuirements. However. Code Case N 566 is currently under review by the staff anc has not been approved for use by reference in Regulatory Guide 1.147, inservice inspection Code Case Acceptabi1ity. ASME Sectton XI. 01viston 1 nor authorized as an alternative pursuant to 10 CFR 50.55a(a) P :(1). Because of a-1997 fall refuelin information (RAI) gregarding outage. in response to an NRC request for additional the use of Code Case N 566, the licensee submitted an interim alternative to the requirements of lWA-5250(a)(2) in a letter dated July 30, 1997, for V. C. Summer Nuclear Station. Unit 1.
2.0 EVALUATION
{ The staff, with technical assistance from its contractor, the Idaba National Engineering and Environmental Laboratory (INEEL). has evaluated the
' information provided by the licensee in support of its interim alternative to the requirements of IWA 5250(a)(2) in a letter dated July 30, 1997, for V. C.
a Summer Nuclear Station. Unit 1. Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the j enclosed Technical Letter Report (TLR),
! Interim Re 5250(a)(2),quest for Relief The Code.Section XI. !WA-i requires that ifIWA b250(a)(2):
leakage occurs at a bolted connection, the bolting shall be removed. VT-3 visually examn.ed for corrosion. and evaluated in accordance with IWA-3100. Pursuant to 10 CFR 50.55a(a)(3)(1). the licensce
] proposed the following alternative to the requirements of IWA 5250(a)(2):
- 'The source of all leakage at bolted connections detected by VT 2 examination during a system pressure test shall be evaluated to
! determine the susceptibility of the bolting to corrosion and potential failure. This evaluation will consider the following variables at a j minimum:
- 1. Location of leakage
- 2. History of leakage
- 3. Fastener materials
- 4. Evidence of corrosion, with the connection assembled
- 5. Corrosiveness of the process fluid
- 6. History and studies of similar fastener material in a similar environment
- 7. Other components in the vicinity that may be degraded due to the leakage.
"When the evaluation of thG above variables is concluded and if the evaluation determines that the leaking condition has not degraded the
i r
! 3-I fasteners. then no further action is necessar However, reasonable t attempts to stop the leakage shall be taken. y.
i "If the evaluation of the variables above indicates the need for further ,
evaluation, or no evaluation is performed, then a bolt closest to the source of leakage shall be removed. The bolt will receive a VT 1 examination and be evaluated for corrosion in accordance with IWA- -
3100(a) and dis >ositioned in accordance with IWB 3140. If the leakage '
is identified wien the bolted connection is in service, and the information in the evaluation is supportive. the removal of the bolt for VT 1 examination may be deferred to the next refueling outage. When the removed bolting shows ovidence of rejectable degradation all remaining bolts shall be removed and receive a-VT-1 examination and evaluation in accordance with IWB 3140."
In accordance with lWA 5250(a)(2) if leakage occurs at a bolted connection. '
the bolting must be removed. VT 3 visually examined for corrosion, and ,
evaluated in accordance with IWA 3100. In lieu of this requirement, the (
lic3nsee has to corrosion. proposed Based ontothe evaluate the bolting items included to evaluation in the determine process, its susceptibility the staff concludes that the evaluation proposed by the licensee provides a sound :
engineering approach. In addition if the initial evaluation indicates the !
need for a more detailed analysis, the bolt closest to the source of the .
leaka 3100(ge a). will be removed. VT-1 examined, and evaluated in accordance with IWA-i Based on the bolting evaluation criteria contained in the interim relief request, the staff concludes that the licensee's proposed alternative to the '
requirements of IWA 5250(a)(2) is a conservative and technically sound engineering approach. As a result, significant patterns of degradation will -
be detected, providing an acceptable level of quality and safety.
3.0 CQN.G M 108 i The staff evaluated the licensee's Interim Request for Relief IWA-5250(a)(2) and concludes that the proposed alternative contained in the request provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(1) for the current interval or until Code Case N 566 is evaluated and authorized in an NRC Safety Evaluation. ,
d )
8 W
. . - - ~ _ _ _ _ . - _ - - _ - -_ . _ . _ . _ _ - _ - _ - _ - . - . . _ . , _ _ _ - . . - _ - _ . - - .
, ENCLOSURE 2 TECHNICAL LETTER REPORT ON THE SECOND 10 YEAR INTERVAL INSERVICE INSPECTION 1
INTEftlM REQUEST FOR REL!EF 1
EQR SOUTH CAROLINA ELECTRIQ,& GAS CO.
VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 .
4 QQCKET NUMBER! 50 395
1.0 INTRODUCTION
By letter dated February 25,1997, the licensee, South Corotiria Electric & Gas Co.
(SCE&G), proposed an alter.iative to the requireneents of the ASME Coda, Cection XI, for the V. C. Summer Nuclear Station, Unit 1, s6cond 10 year inservice Inspection (ISI) Interval. The alterna'ive proposed is to use Codo Case N 566 in !!ou 4
4 of the requirements of IWA 5250(a)(2), Code Case N 566 is currently under review by the Nuclear Regulatory Commies!on (NRC) staff and h48 not been approved for i
use by rafercnce in Regulatory Guldr 1.147, /nservice /nspection Code Case Acceptabilitye ASME Section XI, DMsion !. In respo'sso to an NRC request for saditional Information (RAI) regarding the use of thle Code Case, ths 2cesisee submitted an ir.terim request for relief from the requirewnts of IWA 5kb0fa)(2) in s letter dated July 30,1997. The Idaho National Engineeririg and Environmental Laboratosy (IN8 EEL) staff has evaluated the subject requeat for relief in the following i
section.
2.0 EyALUATION The Code of record for tha V, C. Summer, Unit 1, second 10 year ISIinterval, wh!ch began January 1994,!s the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code. The inforrnat'on provided iv,' tha licensee in support of the
l
- d 1 2 1 l request for relief from Code requirements has been evaluated and the basis for disposition is documented b6fow. ,
i :
Interim Raouest for Reflef IWA 5250(a)(2). Corrective Actions for Bolted Connections l
Code Raoulrement: Section XI, lWA 5250(a)(2), requires that if leakage occurs at a bolted connection, the bolting shall be removed, VT 3 visually examined for
- corrosion, and evaluated in accordance with lWA 3100.
2 l
Licensee's Pronosed Alternative: Pursuant to 10 CFR 50.55a(a)(3)(l), the licensee proposed the following alternative to the requirements of IWA 5250(a)(2):
l "The source nf allleakage at bolted connections detected by VT 2 examination during system pressury 6tt shall be evaluated to determine the susceptibility of the
' bolting to corrosion artti paiential failure. This evaluation will consider the following variables at a minimurm
, 1. Location of leakage
- 2. History of leakage
- 3. Fastener rraterials I 4. Evidence of corrosion, with the connection assembled
- 5. Corrosiveness of the process fluid
- 6. History and studies of similar fastener materialln a similar environment
- 7. Other components in the vicinity that may be degraded due to the leakage.
"When the evaluation of the above variables is concluded and if the evaluation determines that the leaking condition has not degraded the fasteners, then no further action is necessary. However, reasonable attempts to stop the leakage shall be taken.
- lf the evaluation of the variables above indicates the need for further evaluation, or no evaluation is performed, then a bolt closest to the source of leakage shall be removed. The bolt will receive a VT 1 examination and be evaluated for corrosion in accordance with IWA 3100(a) and dispositioned in accordance with IWB 3140, if the leakage is identified when the bolted connection is in service, and the information in the evaluation is supportive, the removal of the bolt for VT 1 1
5 4
y ,. ,, -. -- . . . , - - - , . . -. ~ .# , ,, _,-_ --..mm . .. _ . - . .- - ._ ,,,, _.,.,, _ - - ,. .
.o' 3-1 examination may be deferred to the next refueling outage. When the removed bolting shows evidence of rejectionable degradation, all remaining bolts shall be removed and receive a VT.1 examination and evaluation in accordance with IWB 3140.*
Licensee's Basis for Reauestino Relief (as stated):
1 "Some of the problems associated with the current requirements of IWA 5250(aH2) ;
are summerized as follows: '
1.
lWA 3100 does not provide an acceptable standard for a VT 3 bolt inspection.
- 2. The requirement calls for bolt removal without rogard to the size of the leakage.
- 3. The requirement increases the radiological dose to workers for leaks that are often not a challenge to operational nor structurallimits.
- 4. Bolts sometimes cannot be removed without damaging the bolt or cannot be removed due to the component configuration.
- 5. It is not a requirement of the Code that the Owner must stop the leakage and inspection of the bolting is not necessarily going to stop the leak.
- 6. Removing one bolt at a time,if allowed by system conditions, may actually increase the leakage.
- 7. In many cases, implementation of the requirement would cause the plant an unnecessary transient or delay startup.
- In addition to the problems associated with the requirements of IWA 5250(a)(2), a Special Task Group of the ASME Committee has concluded that the Code does allow the acceptance of leakage by the analytical evaluation methods of IWB 3142.4, and that the actions required by IWA 5250 should not preclude this acceptance. Also, the Working Group Pressure Testing concluded that the system integrity of a bolted connection is not necessarily compromised by leakage and recommended the approval of Code Case N 566.
"This interim relief request is more prescriptive and more conservative than the Code Case, it also addresses many of the implementation and radiological hardships associated with IWA 5250(a)(2) and yet maintains the conclusion of the ASME Committee by assuring that a proper evaluation of the connection and/or the bolting is performed. The joint evaluation must consider specific factors which, if indicative of degradation, must be dispositioned in accordance with IWB 3140 of Section XI.
Due to the fact that this engineering evaluation is more comprehensive than the simple bolt inspection currently required by IWA 5250, coupled with the benefit that these alternative requirements ensure structuralintegrity is maintained, and reduces the operational, maintenance, and radiological hardships of the current requirements, this relief request should be considered as an acceptable alternative in accordance
,,,,o 4
With 10 CFR 50.55a(a)(?.)(l). This conclusion is further supported by the fact that the ASME has approved Code Case N 566 and this interim relief request is essentially a conservative subset of the Code Case."
Evaluatiom In accordance with IWA 5250(a)(2), if les'. age occurs at a bolted connection, the bolting must be removed, VT 3 visually examined for corrosion, and i evaluated in accordance with IWA 3100, in lieu of this requirement, the licensee has proposed to evaluate the botting to determine its susceptibility to corrosion. Based on the items included in the evaluation process, the INEEL staff believes that the evaluation proposed by the licensee provides a sound engineering approach. In add; tion, if the initial evaluation indicates the need for a more detailed analysis, the bolt closest to the source of the leakage will be removed, VT.1 examined, and evaluated in accordance with IWA 3100(a). ,.,
Based on the bolting evaluation criteria contained in the interim relief request, the INEEL staff concludes that the licensee's proposed alternative to the requirements of IWA 5250(a)(2) is a conservative and technically sound engineering approach. As a result, significant patterns of degradation will be detected and an acceptable level of quality and safety will be provided.
3.0 CONCLUSION
The INEEL staff evaluated the licensee's Interim Request for Relief IWA 5250(a)(2) and concludes that the proposed alternative contained in the request will provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
.