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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20216D9541999-07-28028 July 1999 Safety Evaluation Authorizing Proposed Alternatives for Second 10-year Interval Pursuant to 10CFR50.55a(a)(3)(ii) ML20202J0891999-02-0303 February 1999 Safety Evaluation Accepting Rev 2 of Third Interval Inservice Testing Program for Pumps & Valves for James a FitzPatrick Nuclear Power Plant ML20196J3501998-12-0404 December 1998 SER Accepting License Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20196F9251998-11-25025 November 1998 Safety Evaluation Re Thrid 10-year Interval Inservice Insp Program Relief Requests for Plant ML20195K4211998-11-17017 November 1998 Safety Evaluation Authorizing Proposed Alternative in Relief Request VRR-05 Per 10CFR50.55a(a)(3)(i) & PRR-01,PRR-02R1, PRR-03,PRR-04,VRR-02,VRR-03 & VRR-04 Per 10CFR50.55a(a)(3)(ii) ML20195E1051998-11-13013 November 1998 Safety Evaluation Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves, Issued 950817 ML20155H5801998-11-0303 November 1998 Safety Evaluation Authorizing Postponement of Beginning of Augmented Exam Requirements of 10CFR50.55a(g)(6)(ii)(A)(2) Re ASME Code,Section XI ML20155H5321998-11-0303 November 1998 Safety Evaluation Authorizing Alternative to ASME Code Requirements for CRD Bolting ML20154L6591998-10-14014 October 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Snubber Visual Inservice Exam Intervals & Sampling Rates Requirements Contained in ASME Code,Section Xi,Subsection Iwf,Article IWF-5000 ML20153D2591998-09-21021 September 1998 SER Accepting Proposed Alternative Testing of Containment Following ECCS Suction Strainer Replacement ML20236V8181998-07-29029 July 1998 Safety Evaluation Accepting Request for Relief from Implementation of Requirements of 10CFR50.55a Related to Containment Repair & Replacement Activities for James a FitzPatrick Nuclear Power Plant ML20236X5881998-07-29029 July 1998 Safety Evaluation Supporting Amend 245 to License DPR-59 ML20154L9201998-07-10010 July 1998 SER Accepting Rev to Reactor Vessel Surveillance Capsule Withdrawal Schedule for James a Fitzpatrick Nuclear Power Plant ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively ML20198S3031998-01-14014 January 1998 Supplemental SE Accepting RG 1.97,rev 2 Recommendations for Containment Isolation Valve Position Indication Instrumentation at NPP ML20128G2441993-02-0909 February 1993 Correction to NRC SE Associated W/Ts Amend 184,dtd 921217. SE Restates Portion of Section 2.0 ML20126F4771992-12-23023 December 1992 Safety Evaluation Granting Licensee Relief from ASME Code Requirements for Repair of RWCU Equalizing Line Until Next Refueling Outage ML20059K4331990-09-13013 September 1990 Safety Evaluation Accepting Util 881110,890328 & 900129 Submittals Re IGSCC Insp & Repair for Facility Reload 8/ Cycle 9 Refuel Outage ML20056B4011990-08-20020 August 1990 Safety Evaluation Approving Licensee Relief Request R14 & Denying Requests R15 & R5A Re Hydrostatic Test Requirements ML20058P4331990-08-13013 August 1990 Safety Evaluation Accepting ATWS Recirculation Pump Trip Sys Design Mod ML20206F5701988-11-18018 November 1988 Safety Evaluation Re Compliance w/10CFR50.62 ATWS Rule Re Alternate Rod Injection & Recirculating Pump Trip Sys ML20206D5231988-11-10010 November 1988 Safety Evaluation Supporting 880309 Request for Relief from Hydrostatic Test Requirement for HPCI & Rcic,Provided That Alternative Testing Performed ML20148B1001988-03-14014 March 1988 Safety Evaluation Accepting Util Justification for Deviations from Reg Guide 1.97 for post-accident Monitoring Variables ML20236G5801987-10-27027 October 1987 Safety Evaluation Supporting Util 850930,860827 & 1208 Submittals of Second 10-yr Inservice Insp Program Plan & Associated Relief Requests from ASME Code Insp Requirements ML20238E1461987-09-0808 September 1987 Safety Evaluation of Util 870415 Proposed Design for Standby Liquid Control Sys.Design Meets Requirements of 10CFR50.62 Re ATWS ML20237L5241987-09-0101 September 1987 Safety Evaluation Supporting Util 831109,840629 & 850702 Responses to Generic Ltr 83-28,Items 2.1 & 4.5.2 Re Equipment Classification & Vendor Interface & Reactor Trip Sys Reliability, Respectively ML20236J8151987-07-30030 July 1987 Safety Evaluation Re Insps for & Repairs of IGSCC During Reload 7/Cycle 8 Refueling Outage.Facility Can Be Safely Operated for One 18-month Fuel Cycle in Present Configuration ML20212F9421986-12-31031 December 1986 Safety Evaluation Supporting Amend to License DPR-59, Changing Tech Specs Re Second Level of Undervoltage Protection ML20214R5871986-11-24024 November 1986 Safety Evaluation Accepting Util Actions to Ensure Structural Integrity of Vacuum Breakers in Mark I Containments ML20210T3101986-10-0202 October 1986 Safety Evaluation Accepting Util 860228 Submittal of Rev 2 to Offsite Dose Calculation Manual on Interim Basis ML20205E3601986-08-0606 August 1986 Safety Evaluation on Util 830806,1109 & 840330 Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1. Licensee Complied W/All Items ML20210K5571986-04-18018 April 1986 Safety Evaluation Supporting Util Request for Relief from First 10-yr Inservice Insp Requirements for Class 1,2 & 3 Components ML20137S9901985-09-26026 September 1985 Safety Evaluation Accepting MSIV Leakage Control Sys,Per GDC 54, Piping Sys Penetrating Containment ML20134D2071985-08-0909 August 1985 Safety Evaluation of Util 831109 & 840629 Responses to Generic Ltr 83-28,Item 1.1, Post-Trip Review (Program Description & Procedure). Program & Procedures Acceptable ML20133F0011985-07-30030 July 1985 Safety Evaluation Accepting Util 831109 & 840629 Responses to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing ML20129E1721985-07-0101 July 1985 Safety Evaluation Re Radiological Consequences of Hypothetical LOCA While Purging Containment at Power. Radiological Consequences Acceptable ML20129E1411985-07-0101 July 1985 Safety Evaluation Supporting Demonstration of Containment Purge & Vent Valve Operability.Info Submitted Demonstrates Ability of 20-inch & 24-inch Purge & Vent Valves to Close Against Buildup of Containment Pressure During Dba/Loca ML20127E7221985-06-17017 June 1985 SER Supporting Util 840629 Response to Generic Ltr 84-09, Recombiner Capability Requirements of 10CFR50.44(c)(3)(ii) ML20127B3251985-06-10010 June 1985 Interim Safety Evaluation Approving Util 830630 Procedures Generation Package (PGP) for Emergency Operating Procedures Upon Resolution of Exceptions Noted in Section 2.PGP Submitted Per Generic Ltr 82-33 Re Suppl 1 to NUREG-0737 ML20127C7961985-06-0606 June 1985 Safety Evaluation Re Insp & Repair of RCS Piping.Plant Can Be Safely Returned to Operation in Present Configuration for Duration of Cycle 7 ML20140G5731975-07-15015 July 1975 Safety Evaluation Supporting Tech Spec Changes to License DPR-59 to Revise Suppression Pool Water Temp Limits 1999-07-28
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARJAFP-99-0277, Monthly Operating Rept for Sept 1999 Fpr Ja FitzPatrick Nuclear Power Plant.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 Fpr Ja FitzPatrick Nuclear Power Plant.With ML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data JAFP-99-0261, Monthly Operating Rept for Aug 1999 for Jafnpp.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Jafnpp.With JAFP-99-0236, Monthly Operating Rept for July 1999 for Ja FitzPatrick Npp. with1999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Ja FitzPatrick Npp. with ML20216D9541999-07-28028 July 1999 Safety Evaluation Authorizing Proposed Alternatives for Second 10-year Interval Pursuant to 10CFR50.55a(a)(3)(ii) ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept JAFP-99-0211, Monthly Operating Rept for June 1999 for Ja FitzPatrick Nuclear Power Plant.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Ja FitzPatrick Nuclear Power Plant.With JAFP-99-0175, Annual Summary of Changes,Tests & Experiments for 1997/1998. with1999-06-0202 June 1999 Annual Summary of Changes,Tests & Experiments for 1997/1998. with JAFP-99-0181, Monthly Operating Rept for May 1999 for Jafnpp.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Jafnpp.With JAFP-99-0166, Monthly Operating Rept for Apr 1999 for Ja FitzPatrick Nuclear Plant.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Ja FitzPatrick Nuclear Plant.With JAFP-99-0142, Monthly Operating Rept for Mar 1999 for Ja FitzPatrick Nuclear Power Plant.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Ja FitzPatrick Nuclear Power Plant.With JAFP-99-0092, Monthly Operating Rept for Feb 1999 for Ja FitzPatrick Nuclear Power Plant.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Ja FitzPatrick Nuclear Power Plant.With ML20207E9311999-02-26026 February 1999 Part 21 Rept Re Sprague Model TE1302 Aluminum Electrolytic Capacitors with Date Code of 9322H.Caused by Aluminum Electrolytic Capacitors.Affected Capacitors Replaced ML20202J0891999-02-0303 February 1999 Safety Evaluation Accepting Rev 2 of Third Interval Inservice Testing Program for Pumps & Valves for James a FitzPatrick Nuclear Power Plant ML20199M0891999-01-22022 January 1999 Part 21 Rept Re Failure of Square Root Converters.Caused by Failed Aluminum Electrolytic Capacitory Spargue Electric Co (Model Number TE1302 with Mfg Date Code 9322H).Sent Square Root Converters Back to Mfg,Barker Microfarads,Inc JAFP-99-0011, Monthly Operating Rept for Dec 1998 for Ja FitzPatrick Nuclear Power Plant.With1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Ja FitzPatrick Nuclear Power Plant.With ML20198F9991998-12-0404 December 1998 Assessment of Licensing Basis for Use of Containment Overpressure Credit for Net Positive Suction Head Analyses Power Authority of State of New York,James a Fitzpatrick Nuclear Power Plant ML20196J3501998-12-0404 December 1998 SER Accepting License Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves JAFP-98-0396, Monthly Operating Rept for Nov 1998 for Ja FitzPatrick Npp. with1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Ja FitzPatrick Npp. with ML20196F9251998-11-25025 November 1998 Safety Evaluation Re Thrid 10-year Interval Inservice Insp Program Relief Requests for Plant ML20195J7521998-11-18018 November 1998 Rev 7 to Jaf Colr ML20195K4211998-11-17017 November 1998 Safety Evaluation Authorizing Proposed Alternative in Relief Request VRR-05 Per 10CFR50.55a(a)(3)(i) & PRR-01,PRR-02R1, PRR-03,PRR-04,VRR-02,VRR-03 & VRR-04 Per 10CFR50.55a(a)(3)(ii) ML20195E1051998-11-13013 November 1998 Safety Evaluation Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves, Issued 950817 ML20197G6221998-11-0606 November 1998 Non-proprietary Rev 7 to HI-971661, Licensing Rept for Reracking of Ja FitzPatrick Sfp ML20155H5321998-11-0303 November 1998 Safety Evaluation Authorizing Alternative to ASME Code Requirements for CRD Bolting ML20155H5801998-11-0303 November 1998 Safety Evaluation Authorizing Postponement of Beginning of Augmented Exam Requirements of 10CFR50.55a(g)(6)(ii)(A)(2) Re ASME Code,Section XI JAFP-98-0360, Monthly Operating Rept for Oct 1998 for Ja FitzPatrick Nuclear Power Plant.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Ja FitzPatrick Nuclear Power Plant.With ML20155C2821998-10-30030 October 1998 Non-proprietary Rev 0 to GENE-187-30-1598 Np, CRD Bolting Flaw Evaluation for Ja FitzPatrick Nuclear Power Plant ML20154L6591998-10-14014 October 1998 Safety Evaluation Accepting Licensee Proposed Alternative to Snubber Visual Inservice Exam Intervals & Sampling Rates Requirements Contained in ASME Code,Section Xi,Subsection Iwf,Article IWF-5000 JAFP-98-0322, Monthly Operating Rept for Sept 1998 for Ja Fitzpatrick Nuclear Power Plant.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Ja Fitzpatrick Nuclear Power Plant.With ML20153D2591998-09-21021 September 1998 SER Accepting Proposed Alternative Testing of Containment Following ECCS Suction Strainer Replacement ML20153B5611998-09-0101 September 1998 Rev 1 to JAF-SE-98-013, RHR & Core Spray Suppression Pool Suction Strainer Replacement ML20151X6891998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Ja FitzPatrick Nuclear Power Plant ML20237E8361998-08-25025 August 1998 Rev 6 to Colr ML20237E9471998-08-0808 August 1998 Rev 6 to Colr JAFP-98-0264, Monthly Operating Rept for July 1998 for Ja FitzPatrick Nuclear Plant1998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Ja FitzPatrick Nuclear Plant ML20236X5881998-07-29029 July 1998 Safety Evaluation Supporting Amend 245 to License DPR-59 ML20236V8181998-07-29029 July 1998 Safety Evaluation Accepting Request for Relief from Implementation of Requirements of 10CFR50.55a Related to Containment Repair & Replacement Activities for James a FitzPatrick Nuclear Power Plant ML20153B5781998-07-28028 July 1998 Rev 0 to JAF-SE-98-025, HPCI & RCIC Suppression Pool Suction Strainer Replacement ML20236X3831998-07-14014 July 1998 Rev 2 to JAF-RPT-MULTI-02671, Summary of Detailed Evaluation for NRC Generic Ltr 96-06 ML20154L9201998-07-10010 July 1998 SER Accepting Rev to Reactor Vessel Surveillance Capsule Withdrawal Schedule for James a Fitzpatrick Nuclear Power Plant JAFP-98-0222, Monthly Operating Rept for June 1998 for Ja FitzPatrick Nuclear Power Plant1998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Ja FitzPatrick Nuclear Power Plant JAFP-98-0193, Monthly Operating Rept for May 1998 for Ja FitzPatrick Nuclear Plant1998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Ja FitzPatrick Nuclear Plant ML20248F3531998-05-21021 May 1998 Part 21 Rept Re Electronic Equipment Repaired or Reworked by Integrated Resources,Inc from Approx 930101-980501.Caused by 1 Capacitor in Each Unit Being Installed W/Reverse Polarity. Policy of Second Checking All Capacitors Is Being Adopted JAFP-98-0168, Monthly Operating Rept for Apr 1998 for Ja FitzPatrick Nuclear Power Plant1998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Ja FitzPatrick Nuclear Power Plant JAFP-98-0128, Monthly Operating Rept for Mar 1998 for Ja FitzPatrick Nuclear Power Plant1998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Ja FitzPatrick Nuclear Power Plant ML20217A4631998-03-23023 March 1998 Safety Evaluation Accepting Use of Three Heats/Lots of Hot Rolled XM-19 Matl in Core Shroud Repair Assemblies Re Licenses DPR-16 & DPR-59,respectively JAFP-98-0091, Monthly Operating Rept for Feb 1998 for JAFNPP1998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for JAFNPP ML20202G9081998-02-0606 February 1998 Safety Evaluation Re Amend to License DPR-59 to Revise TS Tables 3.2-2 & 4.2-2 JAFP-98-0058, Monthly Operating Rept for Jan 1998 for Ja FitzPatrick Nuclear Power Plant1998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Ja FitzPatrick Nuclear Power Plant 1999-09-30
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p na v gt u'jo,, UNITED STATES e o NUCLEAR REGULATORY COMMISSION h 7: WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO REQUEST FOR REEVALUATION OF REQUEST FOR RELIEF FROM FIRST 10-YEAR INSERVICE INSPECTION REQUIREMENTS POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333
1.0 INTRODUCTION
Technical Specification 4.6.F for the James A. FitzPatrick Nuclear Power Plant states that inservice examination of ASME Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Connission.
We previously seviewed the first 10-year interval inservice inspection (ISI) program plan for the t itzPatrick facility and requests for relief from certain requirements of triu applicable ASME Code and addenda. By letter dated January 31, 1984, we granted relief from examination requirements which we had determined to be impractical to perform at the FitzPatrick facility. We also denied relief in those cases where the necessary findings could not be made.
By letter dated March 15, 1985, the Power Authority of the State of New York (the licensee) requested clarification and reevaluation of the items denied in our January 31, 1984 letter. In reviewing its relief requests, the licensee found that some were no longer necessary due to capabilities or conditions that developed while the program was under review. Other requests were found to be too broad in scope and, in those cases, the licensee intends to submit relief requests for specific welds.
Specifically, the licensee intends to submit relief requests for the following welds:
-Item B.I.1: Reactor Vessel Shell Beltline Region Welds: Code Category B-A.
Item B.I.2: Reactor Vessel Shell Welds other than in Beltline:
Code Category B-B.
Item B.1.4: Reactor Vessel Nozzle Inner Radii: Code Category B-D.
Item C2.1, C2.2, C2.5, Inaccessible Piping Welds: Code Categories C-F and C-G.
8604290359 86041833 DR ADOCK O W-
Items B.1.11; B4.12; B5.9; B4.12; Pressure Retaining Bolting Less Than 2 Inches in Diameter: Code Category B-G-2 In addition, by letter dated March 4, 1985, the licensee proposed a modified inservice inspection program plan to combine the inspection interval of the Class 2 and 3 components with that of the Class 1 components. PASNY has implemented an ISI program based on the 1974 Edition, Summer 1975 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code. Inspection coninenced during the Spring / Summer 1980 refueling outage. Prior to this time, inservice inspections at the Plant were performed in accordance with the 1970 Edition of Section XI. The 1970 edition has required inspection of only those components which are now considered Class 1, while the newer Code expanded the scope of inspection to include Class 2 and Class 3 components. Therefore, while the FitzPatrick initial 10-year interval for Class 1 components was completed on July 28, 1985, the actual interval for the Class 2 and 3 components is at its midpoint.
2.0 EVALUATION 2.1 Reouest for Combining ISI Programs for Class 1 and Classes 2 and 3 We have reviewed the licensee's letter dated March 4, 1985 which proposes to combine the Class 1, 2 and 3 components into one inspection interval to eliminate the inefficiencies inherent in maintaining what are effectively two separate inspection programs where each could require updating to a different code edition. The licensee proposed the following in lieu of completion of the Class 2 and Class 3 interval:
- a. Completion of 50% of the required inspections of Class 2 and 3 components, as determined by the current FitzPatrick ISI program, during the Reload 6/ Cycle 7 refueling outage. This would result in 50% of the inspections performed in roughly half of a 10-year interval, which meets or exceeds the requirements of IWC-2412 and IWD-2410. (Note that the Reload 6/ Cycle 7 refuelin May 1985 and these inspections have been completed)g outage ended in .
- b. Pressure and hydrostatic tests of the Class 2 and 3 components required by IWC-5000 and IWD-5000 for the current interval would be performed during the Reload 7/ Cycle 8 refueling outage (currently scheduled for early 1987). The additional time is required to prepare test procedures that conform to the requirements of FitzPatrick's original construction code, B31.1, 1967 edition.
- c. Commencing with the Reload 7/C'ycle 8 refueling outage, all classes of components would be included in an ISI Program based on a single approved edition (and addenda) of Section XI of ASME Code,
- d. 100% of the inspections of Class 1 components, as required by Section XI of the ASME Code and the FitzPatrick ISI program, would
be completed by the end of the first 10-year interval. (Note that these inspections were completed during the Reload 6/ Cycle 7 retueling outage). The inspections for the second 10-year interval vi'l commence during the Reload 7/ Cycle 8 refueling outage.
Based on our review of the above commitments by the licensee, we tind that the number of inspections of Class 2 and 3 components already performed during the last refueling outage as well as those to be performed during the upcoming Reload 7/ Cycle 8 refueling outage provide reasonable assurance of the structural integrity of components and supports. In addition, Regulation 10 CFR 50.55a(g)(4)(iv) allows Inservice Inspection programs to use portions of Code editions and addenda provided that all related requirerrents of the respective editions or addenda are met. We find that the proposed revised ISI program conforms to this regulation and, therefore, approve this request for relief.
2.2 Request for Relief Concerning Items B4.9, B5.4, B6.4; Integrally Welded Supports f'or Piping, Pump, and Valves; Code Category B-K-1.
Section XI of the ASME Code requires that the volumetric examinations performed during each inspection interval shall cover 25% of the integrally welded supports. The areas shall include the integrally welded external support attachments. This includes the welds to the pressure-retaining boundary and the base metal beneath the weld zone and along the support attachment member for a distance of two support thicknesses.
The licensee has requested relief from the volumetric examination of all Class 1 integrally-welded external support attachments for piping, pumps and valves on the basis that the physical design of integrally welded supports (tillet or partial penetration welds) does not permit meaningful volumetric examination. This fact has been recognized by Section XI of the ASME Code and the requirement for volumetric examination of integrally welded supports has been dropped from later Addenda of the Code (e.g.,1977 Edition, Sumer 1978 Addenda).
Pursuant to 10 CFR 50.55a(g)(4)(iv), the licensee has elected to utilize the 1977 Edition, through Sumer 1978 Addenda for the examination method for code category B-K-1. This code year and addenda requires that a surface examination be performed on support attachments for which the support base material design thickness is 5/8" and greater, and which conform to the configuration of integral attachments referenced in Figures IWB-2500-13 and IWB-2500-15. Accordingly, the licensee proposes to inspect those supports for which relief is requested by surface examination.
The above regulation accepting the use of the alternate ASME Code, 1977 Edition thru Addenda of Sumer 1978, also states that all relevant requirements of the more recent edition must be met. Accordingly, the
licensee must increase the frequency of examination of the subject welds from once per plant lifetime to once per inspection interval as per the 1977 Edition thru Summer 1978 Addenda.
Based on the above evaluation, we find that the licensee's proposed examinations of the subject welds conform to the regulations and provide reasonable assurance of the structural integrity of these welds. We, therefore, approve the licensee's request for relief with the provision that examination of the subject welds be conducted once per inspection interval.
2.3 Request for Relief Concerning Item 84-6: Branch Pipe Connection Exceeding Six Inch Diameter, and Including Residual Heat Removal (RHR) Weid #20-10-141, Code Category B-J.
Section XI of the ASME Code requires that volumetric weld examinations shall be performed during each inspection interval and shall cover all the area of 25% of the circumferential joints including the adjoining branch connection joints. In the case of pipe branch connections, the areas shall include the weld metal, the base metal for one pipe wall thickness beyond the edge of the weld on the main pipe run, and at least 2 inches of the base metal along the branch run, as per IWB-2500-9 with the acceptance standard of IWB-3514.
The licensee, in accordance with the requirements of Inspection and Enforcement Bulletin (IEB) 83-02 and Generic Letter 84-11, has performed ultrasonic examination on the welds listed below:
- 1. 12-03-2-5 4. 12-02-2-21 7. 12-02-2-73
- 2. 12-02-2-11 5. 12-02-2-62 8. 12-02-2-79
- 3. 12-02-2-16 6, 12-02-2-68 9, 20-10-141 The 12 inch welds are all located in the Reactor Water Recirculation System (#1-8); the 20 inch weld is located in the RHR System (#9).
The licensee has requested that the ultrasonic examinations performed on
. these branch pipe welds be accepted in lieu of the standard ASME Code volumetric examination.
The technique employed by the licensee was at least equivalent to the method referenced in Section XI in regard to examination angle (s), and exceeded the recommended Code requirement for instrument sensitivity.
Furthemore, the more stringent requirements for personnel qualifications imposed by IEB 83-02 and administered by the Electric Power Research Institute have enhanced the inspection quality provided these welds.
We have previously accepted the licensee's examinations of the subject welds for confomance to IEB 83-02 and Generic Letter 84-11. Since the ultrasonic examination technique employed was at least as stringent as CD
that required by the Code, we find that the examinations performed provide reasonable assurance of the structural integrity of the subject welds. We therefore approve the licensee's request for relief.
2.4 Request for Relief Concerning Item C2.1: Welds in Piping, and Fittings; Code Categories C-F and C-G.
Section XI of the ASME Code, 1974 Edition through Summer 1975 Addenda, requires the following for Code Categories C-F and C-G:
Category C-F: Pressure-Petaining Welds in Piping, Pumps, and Valves in Systems which Circulate Reactor Coolant Volumetric weld examinations shall cover 100% of the welds. This examination shall be scheduled over the lifetime of the plant (four intervals with three periods within each interval).
Category C-G: Pressure-Retaining Welds in Piping, Pumps, and Valves in Systems which Circulate other than Reactor Coolant Volumetric weld examination of 50% of the total number of welds shall be performed. The examination 1, hall cover 100% of the weld.
This examination shall be scheduled over the lifetime of the plant (four intervals with three periods within each interval).
For the equivalent categories in the 1977 Edition through Summer 1978 Addenda, the following is required:
Category C-F Surface examinations shall be performed on piping welds of nominal wall thickness 1/2 inch or less and on branch connections.
Examinations shall be performed each inspection interval. The welds selected for examination shall include 50% of the main steam system welds, and 25% of the welds in all other systems including Residual Heat Removal Systems (RHR), Emergency Core Cooling Systems (ECCS),
and Containment Heat Removal System (CHRS).
Category C-G:
Surface examinations shall be performed on pump casing welds and valve body welds. The examination shall be performed from either the inside or outside surface of the components. The welds selected for examination shall be 100% of the welds in all components in each piping run examined under examination Category C-F. The frequency shall be each inspection interval.
The licensee has requested relief from the volumetric examination of Class 2 piping that is 0.5 inch nominal wall thickness or less, for em
~
nominal pipe sizes over 4 inches. The licensee proposes to perform a surface examination on welds in Code Categories C-F and C-G in lieu of the volumetric examination, in accordance with Section XI,1977 Code Edition through Summer 1978 Addenda.
Regulation 10 CFR 50.55a(g)(4)(iv) allows Inservice Inspection programs to use portions of Code editions and addenda provided that all related requirements of the respective editions or addenda are met. Tables IWC 2411-1 and IWC 2412-1 require that all inspections of components requiring examination be completed during each inspection interval.
Based on the above evaluation, we find that the inspections to the 1977 Code Edition through Summer 1978 Addenda provide reasonable assurance of the structural integrity of the welds. We, therefore, approve the licensee's request for relief with the provision that the required examinations of welds in the RHR, ECCS and CHRS must be completed each inspection interval.
3.0
SUMMARY
CONCLUSION We conclude based on the considerations discussed above, that relief granted from the examination and testing requirements, and the alternate methods proposed and evaluated, give reasonable assurance that the integrity of the piping, pressure boundary components, and support structures is maintained; that granting relief where Code requirements are impractical is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest considering the burden that result if they were imposed upon the facility. t.'e further conclude that combining ten-year ISI programs of Class 1, and Class 2, and Class 3 components will give reasonable assurance that the inspection of components are in accordance with regulations in extent and frequency and are in the public interest.
Principal Contributor: B. Turovlin Dated: April 18,1986 cpr