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MONTHYEARML20072G1601983-06-24024 June 1983 Rev 9 to ASME Code Section XI Inservice Insp & Testing Program Project stage: Other ML20072G1491983-06-24024 June 1983 Requests Relief from Requirement Re 10-yr Inservice Insp of Reactor Coolant Pump Casing Welds of ASME Boiler & Pressure Vessel Code,Section XI, Inservice Insp. Request Submitted as Rev 9 to Inservice Insp & Testing Program Project stage: Other ML20134E5931985-08-0909 August 1985 Forwards Rev to Request for Relief 29 & Request for Relief 66 from ASME Boiler & Pressure Vessel Code,Section XI Re Inservice Insp for First 10-yr Interval.Review Requested Prior to 10-yr Outage Scheduled for 850905.Fee Paid Project stage: Request ML20138E1661985-10-11011 October 1985 Safety Evaluation Re 850809 Inservice Insp of Components Relief Requests 29 & 66.Alternative Acceptable & Relief Should Be Granted Project stage: Approval ML20138E1511985-10-11011 October 1985 Forwards Safety Evaluation of 850809 Relief Request from Inservice Insp Requirements Re Hydrostatic Pressure Testing of Pipe Section in Charging Sys & Surface Exam of Valve Pressure Boundary.Relief Granted Project stage: Approval 1985-10-11
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20211J2851999-08-26026 August 1999 Safety Evaluation Supporting Amends 146 & 137 to Licenses DPR-42 & DPR-60,respectively ML20211D3981999-08-24024 August 1999 Safety Evaluation Supporting Requested Actions to Licenses DPR-42 & DPR-60,respectively ML20207B5931999-05-26026 May 1999 SER Accepting Licensee Proposed Alternative to ASME Code for Surface Exam (PT) of Seal Welds on Threaded Caps for Unit 1 Reactor Vessel Head Penetrations for part-length CRDMs ML20205B3351999-03-17017 March 1999 Safety Evaluation Supporting Amends 143 & 134 to Licenses DPR-42 & DPR-60,respectively ML20202J1731999-01-22022 January 1999 Safety Evaluation Concluding That NSP Proposed Alternative to Surface Exam Requirements of ASME BPV Code for CRD Mechanism Canopy Seal Welds Will Provide Acceptable Level of Quality & Safety ML20198L2211998-12-0707 December 1998 Safety Evaluation Supporting Amends 141 & 132 to Licenses DPR-42 & DPR-60,respectively ML20195D3821998-11-0404 November 1998 Safety Evaluation Supporting Amends 140 & 131 to Licenses DPR-42 & DPR-60,respectively ML20195D3761998-10-30030 October 1998 Safety Evaluation Supporting Amends 139 & 130 to Licenses DPR-42 & DPR-60 ML20154B9241998-09-22022 September 1998 Safety Evaluation Supporting Amends 138 & 129 to Licenses DPR-42 & DPR-60,respectively ML20237D6491998-08-13013 August 1998 Safety Evaluation Supporting Amends 137 & 128 to Licenses DPR-42 & DPR-60,respectively ML20237A8171998-08-0505 August 1998 SER Related to USI A-46 Program GL 87-02 Implementation for Prairie Island Nuclear Generating Plant,Units 1 & 2 ML20236V4071998-07-28028 July 1998 Safety Evaluation Supporting Amend 136 to License DPR-42 ML20247F9551998-05-0404 May 1998 Safety Evaluation Supporting Amends 135 & 127 to Licenses DPR-42 & DPR-60,respectively ML20217M6901998-04-29029 April 1998 Safety Evaluation Accepting Methodology for Relocation of Reactor Coolant Sys P/T Limit Curves & LTOP Sys Limits Proposed by NSP for Pingp,Units 1 & 2 ML20203H8331998-02-20020 February 1998 SE Accepting Proposed Alternative to ASME Code for Surface Exam of Nonstructural Seal Welds for Prairie Island Nuclear Generating Plant,Unit 2 ML20202B7211997-11-25025 November 1997 Safety Evaluation Supporting Amends 134 & 126 to Licenses DPR-42 & DPR-60,respectively ML20199H7251997-11-18018 November 1997 Safety Evaluation Supporting Amends 133 & 125 to Licenses DPR-42 & DPR-60,respectively ML20199C3671997-11-0404 November 1997 Safety Evaluation Supporting Amends 132 & 124 to Licenses DPR-42 & DPR-60,respectively ML20212G9371997-10-29029 October 1997 Revised SE Re Amends 125 & 117 to Licenses DPR-42 & DPR-60 ML20211E7901997-09-15015 September 1997 Safety Evaluation Supporting Amends 130 & 122 to Licenses DPR-42 & DPR-60,respectively ML20141B0331997-06-12012 June 1997 Safety Evaluation Supporting Amends 129 & 121 to Licenses DPR-42 & DPR-60,respectively ML20148D5441997-05-16016 May 1997 Safety Evaluation of Prairie Island Nuclear Generating Plant Individual Plant Exam ML20138J9961997-05-0606 May 1997 Safety Evaluation Accepting Proposed Alternative to ASME Code for Surface Exam of CRD Mechanism Canopy Seal Welds ML20137S5561997-04-0101 April 1997 Safety Evaluation Approving License Request for Transfer of Licenses for Monticello & Prairie Island,Units 1 & 2 Nuclear Generating Plants & Prairie Island ISFSI ML20134N7411997-02-19019 February 1997 Safety Evaluation Supporting Amends 126 & 118 to Licenses DPR-42 & DPR-60,respectively ML20147D8981997-02-10010 February 1997 Safety Evaluation Supporting Amends 125 & 117 to Licenses DPR-42 & DPR-60,respectively ML20128L6181996-10-10010 October 1996 Safety Evaluation Supporting Amend 124 to License DPR-42 ML20117J0851996-05-21021 May 1996 Safety Evaluation Supporting Amends 123 & 116 to Licenses DPR-42 & DPR-60,respectively ML20093H5251995-10-0606 October 1995 Safety Evaluation Supporting Amends 120 & 113 to Licenses DPR-42 & DPR-60,respectively ML20086E2161995-07-0303 July 1995 Safety Evaluation Supporting Amends 119 & 112 to Licenses DPR-42 & DPR-62,respectively ML20083M7571995-05-15015 May 1995 Safety Evaluation Supporting Amends 118 & 111 to Licenses DPR-42 & DPR-60,respectively ML20082M5711995-04-18018 April 1995 Safety Evaluation Supporting Amends 117 & 110 to Licenses DPR-42 & DPR-60,respectively ML20081F3411995-03-10010 March 1995 Safety Evaluation Supporting Amends 116 & 109 to Licenses DPR-42 & DPR-60,respectively ML20081A9081995-03-0808 March 1995 Safety Evaluation Supporting Amends 115 & 108 to Licenses DPR-42 & DPR-60,respectively ML20077K2541995-01-0505 January 1995 Safety Evaluation Supporting Amends 113 & 106 to Licenses DPR-42 & DPR-60,respectively ML20072C0901994-08-10010 August 1994 Safety Evaluation Supporting Amends 111 & 104 to Licenses DPR-42 & DPR-60,respectively ML20069A1181994-05-17017 May 1994 Safety Evaluation Supporting Amends 110 & 103 to Licenses DPR-42 & DPR-60,respectively ML20058N8021993-12-0808 December 1993 Safety Evaluation Approving Third 10-yr IST Program Requests for Pumps & Valves,Per 10CFR50.55a(f)(6)(i) & 10CFR50.55a(a)(3)(i) ML20058H0151993-12-0303 December 1993 Safety Evaluation Supporting Amends 109 & 102 to Licenses DPR-42 & DPR-60,respectively ML20057A6141993-09-0303 September 1993 Safety Evaluation Supporting Amends 108 & 101 to Licenses DPR-42 & DPR-60,respectively ML20046B6351993-07-29029 July 1993 Safety Evaluation Supporting Amends 107 & 100 to Licenses DPR-42 & DPR-60,respectively ML20044D3151993-05-0404 May 1993 Safety Evaluation Supporting Amends 105 & 98 to Licenses DPR-42 & DPR-60,respectively ML20035H6041993-05-0303 May 1993 SE Accepting Util Responses Re Test Plan & Justification for Use of Dynamic Load Factor for Special Handling Device ML20035H1821993-04-27027 April 1993 SE Supporting Implementation of Reg Guide 1.97 Re Instrumentation to Follow Course of Accident,Per GL 82-33 ML20035A2281993-03-22022 March 1993 SE Supporting Conclusions in Licensee 901127 Rept That Analysis of as-built Configuration That Demonstrated Const Error Causing Insignificant Impact on Responses of Both D5/D6 Bldgs Acceptable,As Built ML20128P4861993-02-0505 February 1993 Safety Evaluation Supporting Amends 104 & 97 to Licenses DPR-42 & DPR-60,respectively ML20127C0291993-01-0404 January 1993 Safety Evaluation Accepting pressure-retaining Components of safety-related Auxiliary Fluid Sys Associated W/Edgs ML20127C0071993-01-0404 January 1993 Supplemental SE Accepting Changes & Additions Described in Rev 1 to Design Rept for Station Blackout/Electrical Safeguards Upgrade Project ML20127C0151993-01-0404 January 1993 Safety Evaluation Accepting Instrumentation & Control Sys Aspects of Unit 2 Load Sequencer Sys in Station Blackout/ Electrical Safeguards Upgrade Project ML20127C0241993-01-0404 January 1993 Safety Evaluation Re Audit of Load Sequencer Implementation. Four of Five Items Reviewed Acceptable & Closed.One Open Item Remained Re Electromagnetic Environ Qualification for Lower Frequency Range of 30 Hz to 10 Khz 1999-08-26
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G4461999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pingp.With ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20216E7151999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Pingp,Units 1 & 2. with ML20211J2851999-08-26026 August 1999 Safety Evaluation Supporting Amends 146 & 137 to Licenses DPR-42 & DPR-60,respectively ML20211D3981999-08-24024 August 1999 Safety Evaluation Supporting Requested Actions to Licenses DPR-42 & DPR-60,respectively ML20211C2531999-08-0404 August 1999 Unit 1 ISI Summary Rept Interval 3,Period 2 Refueling Outage Dates 990425-990526 Cycle 19 971212-990526 ML20210Q4891999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Pingp,Units 1 & 2. with ML20211B5971999-07-31031 July 1999 Cycle 20 Voltage-Based Repair Criteria 90-Day Rept 05000282/LER-1999-007-01, :on 990625,loss of CR Special Ventilation Function Was Noted.Caused by Broken Door Latch Pins on CR Chiller Door.Ts Amend Request to Establish Allowed OOS Time Was Submitted1999-07-23023 July 1999
- on 990625,loss of CR Special Ventilation Function Was Noted.Caused by Broken Door Latch Pins on CR Chiller Door.Ts Amend Request to Establish Allowed OOS Time Was Submitted
ML20209J1131999-07-15015 July 1999 Safety Evaluation of Topical Rept NSPNAD-8102,rev 7 Reload Safety Evaluation Methods for Application to PI Units. Rept Acceptable for Referencing in Prairie Island Licensing Actions ML20209F9811999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Prairie Island Nuclear Generating Plant,Units 1 & 2.With ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20196F4081999-06-23023 June 1999 Revised Pages 71,72 & 298 to Rev 7 of NSPNAD-8102, Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units ML20195G5181999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Prairie Island Nuclear Generating Plant,Units 1 & 2.With . Page 3 in Final Rept of Incoming Submittal Was Not Included ML20207B5931999-05-26026 May 1999 SER Accepting Licensee Proposed Alternative to ASME Code for Surface Exam (PT) of Seal Welds on Threaded Caps for Unit 1 Reactor Vessel Head Penetrations for part-length CRDMs ML20196L2501999-05-13013 May 1999 Rev 0 to PINGP Unit 1 COLR Cycle 20 05000282/LER-1999-005-01, :on 990508,containment Inservice Purge Sys Was Not Isolated During Heavy Load Movement Over Fuel.Caused by Missing Procedure Step in D58.1.6.PINGP 1224 Was Initiated to Communicate Event & Forestall Repeating Event1999-05-0808 May 1999
- on 990508,containment Inservice Purge Sys Was Not Isolated During Heavy Load Movement Over Fuel.Caused by Missing Procedure Step in D58.1.6.PINGP 1224 Was Initiated to Communicate Event & Forestall Repeating Event
ML20206L6191999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Pingp,Units 1 & 2. with ML20205N1081999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Pingp,Units 1 & 2. with ML20205B3351999-03-17017 March 1999 Safety Evaluation Supporting Amends 143 & 134 to Licenses DPR-42 & DPR-60,respectively ML20205Q5101999-03-15015 March 1999 Inservice Insp Summary Rept Interval 3,Period 1 & 2 Refueling Outage Dates 981109-1229 Cycle 19,970327-981229 05000306/LER-1999-001-01, :on 990206,TS Required Reactor Protection Logic Test Was Missed.Caused by Personnel Error.Sd Banks Were Inserted at 0544,RT Breakers Were Opened & Test Was Performed.With1999-03-0808 March 1999
- on 990206,TS Required Reactor Protection Logic Test Was Missed.Caused by Personnel Error.Sd Banks Were Inserted at 0544,RT Breakers Were Opened & Test Was Performed.With
ML20207J6951999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Prairie Island Nuclear Generating Plant ML20202J7711999-02-0404 February 1999 Simulator Certification Rept for Prairie Island Plant Simulation Facility,1998 Annual Rept ML20207L2811999-01-31031 January 1999 Revised Monthly Operating Repts for Jan 1999 for Pingp,Units 1 & 2 ML20202G3761999-01-31031 January 1999 Non-proprietary Rev 7 to NSPNAD-8102-NP, Prairie Island Nuclear Power Plant Reload SE Methods for Application to PI Units ML20202J1731999-01-22022 January 1999 Safety Evaluation Concluding That NSP Proposed Alternative to Surface Exam Requirements of ASME BPV Code for CRD Mechanism Canopy Seal Welds Will Provide Acceptable Level of Quality & Safety 05000306/LER-1998-006-01, :on 981219,unplanned Actuation of ESF Equipment During Performance of Sp.Caused by Personnel Error.Control Room Took Prompt Action & Returned Plant to Proper Status & Second pre-job Briefing for SP-2126 Was Conducted1999-01-18018 January 1999
- on 981219,unplanned Actuation of ESF Equipment During Performance of Sp.Caused by Personnel Error.Control Room Took Prompt Action & Returned Plant to Proper Status & Second pre-job Briefing for SP-2126 Was Conducted
ML20205H0561998-12-31031 December 1998 Northern States Power Co 1998 Annual Rept. with ML20206P7861998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Prairie Island Nuclear Generating Plant.With ML20198J6441998-12-17017 December 1998 Rev 0 to PINGP COLR Unit 2-Cycle 19 05000306/LER-1998-005-02, :on 981109,RT from 22% Power During Planned SD Operation Was Noted.Caused by Tt.Fw Heater Drain Level Control Was Thoroughly Inspected & Calibrated.With1998-12-0909 December 1998
- on 981109,RT from 22% Power During Planned SD Operation Was Noted.Caused by Tt.Fw Heater Drain Level Control Was Thoroughly Inspected & Calibrated.With
ML20198L2211998-12-0707 December 1998 Safety Evaluation Supporting Amends 141 & 132 to Licenses DPR-42 & DPR-60,respectively ML20206N2731998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Prairie Island Nuclear Generating Plant,Units 1 & 2.With 05000282/LER-1998-016, :on 981029,negative Flux Rate RT Occurred Upon CR Insertion After Failure of CRD Cable.Caused by Internal Short Circuit Developing in CRDM Patch Cables at Reactor Head Connector.Replaced CRDM Patch Cables.With1998-11-24024 November 1998
- on 981029,negative Flux Rate RT Occurred Upon CR Insertion After Failure of CRD Cable.Caused by Internal Short Circuit Developing in CRDM Patch Cables at Reactor Head Connector.Replaced CRDM Patch Cables.With
ML20196D7341998-11-20020 November 1998 Third Quarter 1998 & Oct 1998 Data Rept for Prairie Island Isfsi ML20195D3821998-11-0404 November 1998 Safety Evaluation Supporting Amends 140 & 131 to Licenses DPR-42 & DPR-60,respectively ML20155K6301998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Prairie Island Nuclear Generating Plant,Units 1 & 2.With ML20195D3761998-10-30030 October 1998 Safety Evaluation Supporting Amends 139 & 130 to Licenses DPR-42 & DPR-60 05000306/LER-1998-004-01, :on 980910,shield Building Integrity Was Breached.Caused by Inadequate TS Change.Revised Affected Procedures.With1998-10-0505 October 1998
- on 980910,shield Building Integrity Was Breached.Caused by Inadequate TS Change.Revised Affected Procedures.With
ML20202J7991998-09-30030 September 1998 Non-proprietary Version of Rev 3 to CEN-629-NP, Repair of W Series 44 & 51 SG Tubes Using Leaktight Sleeves,Final Rept ML20154H4061998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Prairie Island Nuclear Generating Plant.With ML20198R8061998-09-30030 September 1998 Rev 1 to NSPLMI-96001, Prairie Island Nuclear Generating Plant Ipeee ML20154B9241998-09-22022 September 1998 Safety Evaluation Supporting Amends 138 & 129 to Licenses DPR-42 & DPR-60,respectively ML20198P0571998-09-0303 September 1998 Rev 1 to 95T047, Back-up Compressed Air Supply for Cooling Water Strainer Backwash Valve Actuator ML20153B0761998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Prairie Island Nuclear Generating Plant.With 05000282/LER-1998-009-01, :on 980731,noted That Recent Testing of RCS Vent Paths Had Not Been Performed in Literal Compliance with Wording of TS 4.18.1.Caused by Misunderstanding of Wording in TS Section 4.18.Will Modify Surveillance Procedures1998-08-27027 August 1998
- on 980731,noted That Recent Testing of RCS Vent Paths Had Not Been Performed in Literal Compliance with Wording of TS 4.18.1.Caused by Misunderstanding of Wording in TS Section 4.18.Will Modify Surveillance Procedures
ML20237D6491998-08-13013 August 1998 Safety Evaluation Supporting Amends 137 & 128 to Licenses DPR-42 & DPR-60,respectively ML20237A3961998-08-11011 August 1998 Safety Evaluation on Westinghouse Owners Group Proposed Insp Program for part-length CRDM Housing Issue.Insp Program for Type 309 Welds Inadequate from Statistical Point of View 1999-09-30
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UNITED STATES 8
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT N0. 2 DOCKET NO. 50-306 INSERVICE INSPECTION OF COMPONENTS -
I5I RELIEF REQUEST NOS. 29 AND 66 Introduction By letter dated August 9, 1985, Northern States Power Company (the licensee) requested relief from two requirements of Section XI of the ASME Boiler and Pressure Vessel Code for the Prairie Island Nuclear Generating Plant Unit 2.
The licensee's relief requests pertain to the first 10-year interval which is due to expire at the start of cycle 11 (October 1985) for Unit 2.
During the first 10-year interval, the requirements cf the 1974 edition through Sumer of 1975 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code apply.
The licensee, pursuant to 10 CFR 50.55a(g)(5)(iii), submitted by letter dated August 9, 1985, information to support the determination that examinations imposed by the ASME Code are impractical for Prairie Island Unit 2 request.
Evaluation This evaluation addresses two relief requests related to the inservice in-spection of components identified in the licensee's letter dated August 9, 1985. These relief requests are identified as Relief Request No. 29 dealing with hydrostatic testing and Relief Request No. 66 dealing with examination of the internal pressure boundary surface of one valve.
1.
Relief Request No. 29 A.
Component Description The affected component is a Class 2 charging line pipe section down-stream of valves'2 VC-7-10 and 2 VC-7-11. This line includes 35 feet of 2" diameter pipe located outside containment and 4 feet of 3/4" diameter pipe located inside containment. This piping is considered part of the charging system.
B.
Code Requirement Pursuant to IWC 5220, Pressure piping systems shall be hydrostatically pressure tested to at least 1.25 times the system design pressure (2485 psig) at a test temperature not less than 100*F.
Licensee's Basis for Requesting Relief C.
The piping system described above is not isolable from the Class 1 piping unless a freeze plug is installed in the charging line to the cold leg downstream of the 3/4" bypass line and in the charging line to auxiliary 8510240601 8510113j6 DR ADOCK O
.. spray on the regenerative heat exchanger outlet. Although a freeze plug has been used in the past, the licensee is reluctant to utilize this technique because of the hazards associated with installing and maintaining the plug in the pipe system during the test period.
D.
Licensee's Proposed Alternative This.section of pipe will be visually inspected in accordance with the rules of IWC 2000 and pressure tested to approximately 2600 psig following each refueling outage as an alternative.
E.
Evaluation Because of the design, the charging line downstream of valves 2 VC-7-10 and 2 VC-7-11 cannot be pressurized to the proper test pressure without the utilization of a freeze plug. We agree with the licensee regarding the potential hazards associated with the use of a freeze plug in that, if the freeze plug should break loose during the pressure test, it could damage the downstream piping.
It is therefore impractical to meet the code requirements for pressure testing the pipe section described above because of the risk associated with the use of a freeze plug. The 2600 psig test pressure is approx-imately 400 psig below the prescribed test pressure in IWC 5000 and approximately 350 psig above the normal operating pressure.
F.
Conclusion Based on the above evaluation, the staff concludes that, for the section of pipe discussed above, the code requirement regarding hydrostatic testing is impractical.
It is further concluded that the alternative discussed above will provide adequate assurance of structural integrity of the pipe section described above. Therefore, relief from IWC 5000 of Section XI of the ASME Boiler and Pressure Vessel Code should be granted for the section of pipe described above provided that, following each refueling cycle, the pipe section is hydrostatically tested to approximately 2600 psig and welds are visually inspected in accordance with IWC 2000 of Section XI of g.
the ASME code.
2.
Relief Request No. 66 A.
Component Description The component concerned with this relief request is a 10-inch motor operated gate valve. The manufacturer is Darling Company and the valve is located in the Unit 2 residual heat removal B return line (valveNo.MV32169).
b
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B.
Code Requirement IWB 2412 and Table IWB 2500, B-M-2 of Section XI of the ASME Boiler and Pressure Vessel Code, require visual inspection of the internal pressure boundary of valves of the same constructural design, e.g.,
globe or check valves, manufacturing method and manufacturer, that perform similar functions in the system.
C.
Licensee's Basis for Requesting Relief The licensee's bases for requesting relief from the Code requirement are as follows:
1.
The disassembly of the valve will require breaking a seal weld and replacing it after the examination of internal pressure boundary is completed. The remaking of the seal weld presents the potential of damaging the valve internals.
2.
Valve No. MV-32169 is the only gate valve existing in Unit 2 that s
was supplied by Darling Company. Therefore, this valve does not represent a population of valves existing in Unit 2 other than MV-32169 itself.
3.
A hazardous condition will exist during the examination of the valve since there will be a single check valve isolating the water in the refueling canal which is flooded during this period.
The check valve is designed to seal at a much higher pressure and temperature than will exist during the examination period of the valve.
- t 4
The later edition of the ASME Code does not require the extent of the examination to include the manufacturer.
5.
The examination of the same valve in Unit No. I performed in January 1985 showed no evidence of wear, erosion, corrosion or other anomalies.
In this case the seal weld between the body and bonnet was not remade because of difficulty and potential damage to the valve internals.
D.
Licensee's Proposed Alternative As an alternative, the licensee proposes to examine a motor operated valve from a different manufacturer during the Unit 2 refueling and if the residual heat removal system shows evidence of possible valve degradation during hydrostatic testing, to disassemble and examine the subject valve.
E.
Evaluation Access to perform an internal surface examination on valve MV-32169 is restricted due to the valve design and the potential hazardous conditions that exist during the examination. The result of the examination of the same valve in Unit 1 gives reasonable assurance
. that the internal surface of the valve in Unit 2 has not degraded to the point where its intended safety function would be affected. As a matter of fact, the examination showed no signs of degradation after 10 years of service.
In addition it is unlikely that it will be possible to remake the seal weld after the examination based on the experience gained from the examination of the Unit i valve.
F.
Conclusion Based on the above evaluation, the staff finds that it is impractical to meet the code requirement by the examination of this valve. As an alternative, the licensee has committed to examine a motor operated valve from a different manufacturer during the Unit 2 refueling outage.
In addition, if during hydrostatic testing the residual heat removal system shows evidence of possible valve degradation, then the subject valve would be disassembled and examined.
In conclusion, the staff finds this alternative acceptable and there-fore the relief should be granted.
Date: October 11, 1985 Principal Contributor:
D.C. Di Ianni l
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