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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217G9961999-10-14014 October 1999 SER Accepting First 10-year Interval Inservice Insp Requests for Relief for Plant,Units 1 & ML20212A7441999-09-10010 September 1999 Safety Evaluation Concluding That Alternatives Contained in Relief Request 12R-07 Provide Acceptable Level of Quality & Safety ML20210U8111999-08-0404 August 1999 SER Granting Licensee Relief Requests VR-1,VR-3 & Portion of VR-2 Pursuant to 10CFR50.55a(a)(3)(ii).Relief Request VR-4 Does Not Require Explicit NRC Approval Because Relief Applies to Valves Not Required by 10CFR50.55a ML20210K9861999-07-30030 July 1999 Safety Evaluation Accepting Licensee 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20207B6481999-05-25025 May 1999 SER Accepting Revised SGTR Analysis for Byron & Braidwood Stations.Revised Analysis Was Submitted to Support SG Replacement at Unit 1 of Each Station ML20205B5091999-03-26026 March 1999 SER Accepting Relief Requests 12R-24,Rev 0 & 12R-34,Rev 0, Related to Second 10-year Interval Inservice Insp for Byron Station,Units 1 & 2 ML20205C5101999-03-21021 March 1999 Revised Safety Evaluation Supporting Improved TS Amends Issued by NRC on 981222 to FOLs NPF-37,NPF-66,NPF-72 & NPF-77.Revised Pages Include Editorial Corrections ML20204G3831999-03-19019 March 1999 Safety Evaluation Accepting Second 10-yr Interval ISI Request for Relief 12R-11 ML20155B6711998-10-26026 October 1998 Safety Evaluation Accepting Requests for Relief Associated with Second 10-yr Interval ISI Program Plan ML20154D4401998-10-0202 October 1998 Safety Evaluation Authorizing Second 10-yr Interval ISI Program Request for Relief 12R-30 for Plant,Units 1 & 2 ML20238F3281998-08-31031 August 1998 SER Approving Second 10-year Interval Inservice Insp Program Request for Relief 12R-14 for Braidwood Station,Units 1 & 2 ML20238F6551998-08-28028 August 1998 SE Authorizing Licensee Request for Relief NR-20,Rev 1 & NR-25,Rev 0 Re Relief from Examination Requirement of Applicable ASME BPV Code,Section XI for First ISI Interval Exams ML20217K6331998-04-20020 April 1998 Safety Evaluation Accepting Methodology & Criteria Used in Generating Flaw Evaluation Charts for RPV of Braidwood IAW Section XI of ASME Code ML20217K7171998-04-20020 April 1998 Safety Evaluation Accepting Requests for Relief NR-22,NR-23 & NR-24 for First 10-yr Insp Interval ML20216F4921998-03-11011 March 1998 Correction to Safety Evaluation Re Revised SG Tube Rapture Analysis ML20212H1851998-03-0606 March 1998 SE Approving Temporary Use of Current Procedure for Containment Repair & Replacement Activities Instead of Requirements in Amended 10CFR50.55a Rule ML20197B7531998-03-0404 March 1998 SER Accepting License Request for Relief from Immediate Implementation of Amended Requirements of 10CFR50.55a for Plant,Units 1 & 2 ML20199G2591998-01-28028 January 1998 Safety Evaluation Rept Accepting Revised SG Tube Rupture Analysis ML20199H0031998-01-21021 January 1998 SER Accepting Pressure Temp Limits Rept & Methodology for Relocation of Reactor Coolant Sys pressure-temp Limit Curves & Low Temp Overpressure Protection Sys Limits for Byron Station,Units 1 & 2 & Braidwood Station,Units 1 & 2 ML20199C1401998-01-16016 January 1998 SER Accepting Request to Integrate Reactor Vessel Weld Metal Surveillance Program for Byron,Units 1 & 2 & Braidwood,Units 1 & 2 Per 10CFR50 ML20199C1231998-01-13013 January 1998 Safety Evaluation Granting Second 10-yr Inservice Insp Program Plan Relief Request ML20197G0041997-12-11011 December 1997 Safety Evaluation Accepting First 10-yr Interval Insp Program Plan,Rev 4 & Associated Requests for Relief for Plant ML20198H3211997-12-0303 December 1997 Safety Evaluation Re Licensee Submittal of IPE for Plant, Units 1 & 2,in Response to GL 88-02, IPE for Severe Accident Vulnerabilities ML20198R3061997-10-27027 October 1997 Staff Evaluation Rept Concluding That Licensee IPE Complete Wrt Info Requested by GL 88-20 & Process Meets Intent of Subj GL ML20211L2151997-10-0303 October 1997 Safety Evaluation Supporting Licensee Relief Request,Per 10CFR50.55a(a)(3)(i) ML20217C1681997-09-22022 September 1997 Safety Evaluation Accepting Request for Relief from ASME Code,Section Iii,Div 2 for Repair of Damaged Concrete Reinforcement Steel NUREG-1335, Staff Evaluation Rept Concluding That Licensee IPE Complete Wrt Info Requested by GL 88-20 & Associated Guidance, NUREG-13351997-08-28028 August 1997 Staff Evaluation Rept Concluding That Licensee IPE Complete Wrt Info Requested by GL 88-20 & Associated Guidance, NUREG-1335 ML20141L9321997-05-29029 May 1997 Safety Evaluation Accepting Use of ASME Boiler & Pressure Vessel Code,Section Ix,Code Cases 2142-1 & 2143-1 for Reactor Coolant Sys for Plants ML20141B5551997-05-13013 May 1997 SE Accepting First 10-yr Interval Inservice Insp Program Plan Request for Relief NR-29 for Braidwood Station,Units 1 & 2 ML20140H8871997-05-0808 May 1997 Safety Evaluation Supporting Request for Relief from ASME Code Repair Requirements for ASME Code Class 3 Piping Ceco ML20134L7811996-11-18018 November 1996 Safety Evaluation Granting Listed Relief Requests,Per 10CFR50.55a(f)(6)(i) Based on Impracticalities in Design of Valves That Limit IST in Traditional Manner Using Position Indicating Lights ML20129F9101996-10-25025 October 1996 Safety Evaluation Accepting Request to Apply LBB Analyses to Eliminate Large Primary Loop Pipe Rupture from Structural Design Basis for Plant,Units 1 & 2 ML20059E2871993-12-30030 December 1993 Safety Evaluation Supporting Amends 57,57,45,45,93,77,152 & 140 to Licenses NPF-37,NPF-66,NPF-72,NPF-77,NPF-11,NPF-18, DPR-39 & DPR-48 Respectively ML20056D4921993-07-27027 July 1993 Safety Evaluation Re Fuel Reconstitution ML20127N1851993-01-25025 January 1993 Safety Evaluation Accepting Inservice Testing Program for Valves,Relief Request VR-4 ML20059L3371990-09-14014 September 1990 SER Granting Interim Relief for 1 Yr or Until Next Refueling Outage to Continue Current Testing Methods While Licensee Investigates Feasibility of Acceptable Alternatives ML20059L4581990-09-14014 September 1990 Sser Supporting Util Changes to Inservice Testing Program ML20058L9961990-08-0606 August 1990 Safety Evaluation Denying Licensee Response to Station Blackout Rule.Staff Recommends That Licensee Reevaluate Areas of Concern Identified in SER ML20058M0001990-08-0606 August 1990 Safety Evaluation Denying Licensee Response to Station Blackout Rule.Staff Recommends That Licensee Reevaluate Areas of Concern Identified in SER ML20248D5911989-08-0707 August 1989 SER Accepting Util 881130,890411,27 & 0523 Submittals Re Seismic Qualification of Byron Deep Wells ML20247D1471989-07-18018 July 1989 SER Supporting Util Proposed Implementation of ATWS Design, Per 10CFR50.62 Requirements ML20244D8191989-06-13013 June 1989 SER Supporting Util ATWS Mitigating Sys Actuation Circuitry Designs ML20247B3281989-04-24024 April 1989 Safety Evaluation Re Mechanical Draft Cooling Tower Tests ML20244A7221989-04-11011 April 1989 Safety Evaluation Concluding That Rev 1 to First 10-yr Interval Inservice Insp Program Plan Constitutes Basis for Compliance w/10CFR50.55a & Tech Spec 4.0.5.Response to Items 2.2.2 & 2.2.3 of Inel Technical Evaluation Rept Requested ML20155F1591988-10-0606 October 1988 Safety Evaluation Re Mixed Greases W/Greater than 5% Unqualified Contaminant in Limitorque Valve Operators. Insufficient Info Presented to Draw Conclusions ML20236L2001987-10-30030 October 1987 Safety Evaluation Supporting Amends 11 to Licenses NPF-37 & NPF-66,respectively & Amend 1 to License NPF-72 ML20237G8561987-08-10010 August 1987 SER on Util 870303 & 0522 Ltr Re Optpipe Computer Code Used in Snubber Reduction Program.Code Acceptable for Piping Dynamic Analysis Using Both Uniform & Independent Support Motion Response Spectrum ML20210R2061987-02-0606 February 1987 Safety Evaluation Supporting Util 850517,0802,0823,1211 & 860429 Submittals Re Environ Effects of High Energy Line Breaks in Auxiliary Steam or Steam Generator Blowdown Sys. Design of Blowdown Sys Acceptable ML20210T2571987-02-0606 February 1987 SER Re Util 850802 Submittal Describing Design Details of Steam Generator Blowdown & Auxiliary Steam Sys to Detect & Isolate High Energy Line Breaks.Sys Design Acceptable, However,Two Deviations from IEEE-STD-297 Criteria Apparent ML20209C3571987-01-23023 January 1987 SER Supporting Facility Design,Per Generic Ltr 83-28,Item 4.5.2, Reactor Trip Sys Reliability,On-Line Testing 1999-09-10
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G9961999-10-14014 October 1999 SER Accepting First 10-year Interval Inservice Insp Requests for Relief for Plant,Units 1 & BW990066, Monthly Operating Repts for Sept 1999 for Braidwood Station, Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Braidwood Station, Units 1 & 2.With ML20217H5221999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Byron Station, Units 1 & 2.With ML20217P6351999-09-29029 September 1999 Non-proprietary Rev 6 to HI-982083, Licensing Rept for Spent Fuel Rack Installation at Byron & Braidwood Nuclear Stations 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 ML20212A7441999-09-10010 September 1999 Safety Evaluation Concluding That Alternatives Contained in Relief Request 12R-07 Provide Acceptable Level of Quality & Safety ML20212B9261999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Byron Station,Units 1 & 2.With BW990056, Monthly Operating Repts for Aug 1999 for Braidwood Station, Units 1 & 2.With1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Braidwood Station, Units 1 & 2.With ML20210R6421999-08-13013 August 1999 ISI Outage Rept for A2R07 ML20210U8111999-08-0404 August 1999 SER Granting Licensee Relief Requests VR-1,VR-3 & Portion of VR-2 Pursuant to 10CFR50.55a(a)(3)(ii).Relief Request VR-4 Does Not Require Explicit NRC Approval Because Relief Applies to Valves Not Required by 10CFR50.55a ML20210R3431999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Byron Station, Units 1 & 2.With BW990048, Monthly Operating Repts for Jul 1999 for Braidwood Station, Units 1 & 2.With1999-07-31031 July 1999 Monthly Operating Repts for Jul 1999 for Braidwood Station, Units 1 & 2.With ML20210K9861999-07-30030 July 1999 Safety Evaluation Accepting Licensee 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20210E2251999-07-21021 July 1999 B1R09 ISI Summary Rept Spring 1999 Outage, 980309-990424 M990002, Revised NFM9900022, Braidwood Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function1999-07-12012 July 1999 Revised NFM9900022, Braidwood Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function ML20216D3841999-07-12012 July 1999 Revised NFM9900022, Braidwood Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function ML20209G1751999-07-0808 July 1999 SG Eddy Current Insp Rept,Cycle 9 Refueling Outage (B1R09) ML20209H3711999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Byron Station, Units 1 & 2.With ML20207H7771999-06-30030 June 1999 Rev 0 to WCAP-15177, Evaluation of Pressurized Thermal Shock for Byron,Unit 2 ML20207H7851999-06-30030 June 1999 Rev 0 to WCAP-15183, Commonwealth Edison Co Byron,Unit 1 Surveillance Program Credibility Evaluation BW990038, Monthly Operating Repts for June 1999 for Braidwood Station, Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Braidwood Station, Units 1 & 2.With ML20207H7941999-06-30030 June 1999 Rev 0 to WCAP-15180, Commonwealth Edison Co Byron,Unit 2 Surveillance Program Credibility Evaluation ML20207H8071999-06-30030 June 1999 Rev 0 to WCAP-15178, Byron Unit 2 Heatup & Cooldowm Limit Curves for Normal Operations ML20207H7561999-06-28028 June 1999 Pressure Temp Limits Rept (Ptlr) ML20207H7621999-06-28028 June 1999 Pressure Temp Limits Rept (Ptlr) BW990029, Monthly Operating Repts for May 1999 for Braidwood Stations, Units 1 & 2.With1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Braidwood Stations, Units 1 & 2.With ML20195J8001999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Byron Station,Units 1 & 2.With ML20209H7481999-05-31031 May 1999 Revised Monthly Operating Repts for May 1999 for Braidwood Station,Units 1 & 2 ML20207B6481999-05-25025 May 1999 SER Accepting Revised SGTR Analysis for Byron & Braidwood Stations.Revised Analysis Was Submitted to Support SG Replacement at Unit 1 of Each Station ML20195B2591999-05-19019 May 1999 Rev 66a to CE-1-A,consisting of Proposed Changes to QAP for Dnps,Qcs,Znps,Lcs,Byron & Braidwood Stations ML20206R6991999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Byron Station Units 1 & 2.With BW990021, Monthly Operating Repts for Apr 1999 for Braidwood Station, Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Braidwood Station, Units 1 & 2.With M980023, Rev 2 to NFM9800233, Byron Station Unit 2 COLR for Cycle 8A (BY2C8A)1999-04-28028 April 1999 Rev 2 to NFM9800233, Byron Station Unit 2 COLR for Cycle 8A (BY2C8A) ML20195C7961999-04-28028 April 1999 Rev 2 to NFM9800233, Byron Station Unit 2 COLR for Cycle 8A (BY2C8A) BW990016, Monthly Operating Repts for Mar 1999 for Braidwood Generating Station,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Braidwood Generating Station,Units 1 & 2.With ML20205P7001999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Byron Station,Units 1 & 2.With ML20205B5091999-03-26026 March 1999 SER Accepting Relief Requests 12R-24,Rev 0 & 12R-34,Rev 0, Related to Second 10-year Interval Inservice Insp for Byron Station,Units 1 & 2 ML20205C5101999-03-21021 March 1999 Revised Safety Evaluation Supporting Improved TS Amends Issued by NRC on 981222 to FOLs NPF-37,NPF-66,NPF-72 & NPF-77.Revised Pages Include Editorial Corrections ML20204G3831999-03-19019 March 1999 Safety Evaluation Accepting Second 10-yr Interval ISI Request for Relief 12R-11 ML20206A8831999-03-17017 March 1999 Rev 0 to NFM9900043, Byron Unit 1,Cycle 10 COLR in ITS Format W(Z) Function M990004, Rev 0 to NFM9900043, Byron Unit 1,Cycle 10 COLR in ITS Format W(Z) Function1999-03-17017 March 1999 Rev 0 to NFM9900043, Byron Unit 1,Cycle 10 COLR in ITS Format W(Z) Function ML20196A0721999-03-16016 March 1999 Cycle 8 COLR in ITS Format & W(Z) Function ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20207J4371999-03-0808 March 1999 ISI Outage Rept for A1R07 ML20204H9941999-03-0303 March 1999 Non-proprietary Rev 4 to HI-982083, Licensing Rept for Spent Fuel Rack Installation at Byron & Braidwood Nuclear Stations BW990010, Monthly Operating Repts for Feb 1999 for Braidwood Generating Station,Units 1 & 2.With1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Braidwood Generating Station,Units 1 & 2.With ML20204C7671999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Byron Station,Units 1 & 2.With ML20206U9011999-02-15015 February 1999 COLR for Braidwood Unit 2 Cycle 7. Page 1 0f 13 of Incoming Submittal Was Not Included BW990004, Monthly Operating Repts for Jan 1999 for Braidwood Generating Station,Units 1 & 2.With1999-01-31031 January 1999 Monthly Operating Repts for Jan 1999 for Braidwood Generating Station,Units 1 & 2.With ML20199G8271998-12-31031 December 1998 Rev 1 Comm Ed Byron Nuclear Power Station,Unit 1 Cycle 9 Startup Rept 1999-09-30
[Table view] |
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3* 4 . UNITED STATES y NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO INTEGRATION OF REACTOP PRESSURE VESSEL SURVEILLANCE PROQBAM COMMONWEALTl! EDISON COMPANY BYRON AND BRAIDWOOD STATIONS. UNITS i AND 2 DOCKET NOS. STN 50-454 STN 50-455. STN 50-456. AND STN 50-457
1.0 INTRODUCTION
By letter dated May 6,1997, as supplemented by letter of May 6,19g7, Commonwealth Edison Company (Comed), *he licensee for Byron and Braidwood Stations, submitted a request to integrate the reactor pressure vessel (RPV) weld metal surveillance programs for several plants pursuant to Title 10 of the Code of Federal Reaulations, Par 150 (10 CFR Part 50),
Appendix H,Section Ill.C. The request, if approved, would result in the integration of the RPV weld metal surveillance program for Braidwood, Units 1 and 2, and it would also result in the integration of the Byron, Units 1 and 2, RPV weld metal surveillance programs.
In an integrated surveillance program, the representative materials chosen for surveillance for a reactor are irradiated in one or more other reactors that have similar design and operating foetures. Integrated surveillance programs must be approved by the Director, Office of Nuclear Reactor Regulation on a case by-case basis. Several of the criteria which must be met to obtain approval for an integratsd program are detailed in 10 CFR Part 50, Appendix H. These criteria are discussed below.
2.0 SACKGROUND The purpose of a RPV material surveillance program is to monitor changes in the fracture toughness properties of ferritic materials in the RPV beltline region of light water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment. Under the RPV surveillance program, fracture toughness test data are obtained from material specimens exposed in surveillance capsulet, which are periodically withdrawn from the RPV, The data obtained from the survel.1ance program are used as described in 10 CFR Part 50, Appendix G (fracture toughness requirements) and 10 CFR 50.61 (fracture toughness requirements for protection against pressurized thermal shock).
As discussed in 10 CFR Part 50, Appendix H, several of the criteria for approving the integration of RPV surveillance programs include:
1, The reactor in which the materials will be irradiated and the reactor for which the materials are being irradiated must have sufficiently similar design and operating features to permit accutate comparisons of the predicted amount of radiation damage.
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- 2. Each reador must have an rigt dosimetry program. ;
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- 3. There must be i;M ammgement for data shadng between plants. l'
- 4. There must be a contingency plan to assure that the survoiNance program for each reactor l
~
wel not be jeopardized by operation at a reduced power level or by an extended outage of another reactor from which data are expected.
t
- 6. There must be substantisi advantages to be gained such as reduced power outages or reduced personnel exposure to radiation, as a diroot result of not requiring surveillerwe
- capsules in all reactors in the set.
- 6. No reduction in the requirements for number of matedals to be irradiated, specimen types, or number of spoolmens po' reactor is permitted.
4 i
The surveillance programs at Byron and Braidwood, Units 1 and 2, are currently independent.- l
, The licensee is requesting to integrate the wold metal surveillance programs of Byron, Units 1 and 2 (both surveillance wolds were fabricated using weld wire heat number 442002) and to ,
integrate the weld metal surveillance programs of Braidwood, Units 1 and 2 (both surveillance welds were fabricated using weld wire heat number 442011). The surveillance material for the i limhing base metal materials for these unho are unique to each reactor vessel (Brandwood, Unit 1 !
host number 4gD867/4gC8131 1; Braidwood, Unit 2 heat number 500102/50Cg71 1; Byron, Unit i heat number 5P.5g33; Byron, Unit 2 heat number 4gD330/4gC2g61 1). As a result, each remotor vessel will still retain a completely independent surveillance program for these materials.
3.0 EVALUATION l To support their request to integrate the wold metal surveillance programs for Braidwood, Units 1 and 2, and for Byron, Units 1 and 2 Comed addressed the criteria within the regulations for an integratoo surveillance program, as discussed below. The staffs conclusion are provided in i
Section 4.0 of this evaluationi To dem enstrate that the reactor in which the materials will be irradiated and the reactor for which the materials are being irradiated are sufficisntly similar in design and operating features to 3 t
permh accurate comparisons of the predicted amount of radiation damage, Comed assessed the simMarity in RPV design and operating features between Byron, Units 1 and 2, and between Braidwood, Units 1 and 2. This assessment evaluated, !n part, the weld wire heat, vessei
- fabricator, fabrication data, type of fuel, fuel loading pattom, fluences at end-of-license, capsule :
locations and lead factors, vessel inlet temperatures, vessel dimension and geometry, damage rate, spectral balance, gamma heating, and irradiation temperature. Based on this asseparnent, Comed concluded that the RPV design and operating features are sufficiently similar to warrant ;
integrating the weld metal surveillance programs orBraidwood, Units 1 and 2, and Byron, Units 1 t and 2. l To demonstrate that the reactor has an adequate dosimetry program, Comed stated that their ,
dosimetry program is consistent with the requirements of Draft Regulatory Guide DG 1053, !
" Calculation and Dosimetry Methods for Detstmining Pressure Vessel Neutron Fluence." Comed noted, however, that previously determined fluence values for Braidwood and Byron, Units 1 and
- 2, were not determined with the more recently approved neutron cross-section libraries
- l. >
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! (i.e., ENDF/B VI). However, Comed indicated that the use of the previously determined fluence values is justified based on the relatively small change anticipated from updating the neutron fluence evaluations to ENDF/B VI as a result, in part, of implementing low leakage fuel .
management. Comed stated that an update to the ENDF/li VI cross sections wiu be made by ,
Comed at the next scheduled capsule withdrawal for each set of unNs (capsule W for Byron, Unit '
1, et outage B1R08 in November 1997 and capsule W for Braidwood, UnN 1 at outage A1R07 in )
September 1998).
Since Comed owns both Byron, UnHs 1 and 2, Comed indicated that there is an adequate arrangement for data shar6ng between the plards. The identical sMustion exists for Braidwood, ,
1 Units 1 and 2.
j To demonstrate that the surveillance program for each reactor will not be jeopardized by operation at reduced power level or by an extended outage of another reactor from which data are expected, Comed indicated that all originally planned surveillance capsules will still be e tested; thereby retaining a completely independent surveillance program that will not depend on the other vessel for data, in addition, Comed indicated that no reduction in the number of materials to be irradiated, specimen types, or number of specimens per reactor was requested. l As a resuh, the surveillance program for each reactor will not be jeopardized by operation at ,
reduced power level or by an extended outage of another reactor from which the data will be '
integrated.
Comed did not request a reduction in the number of surveillance capsules in each reactor nor did ;
it request a reduction in the number of materials to be irradiated, specimen types, or number of specimens per reactor.
In addition to the Appendix H r:quirements for integrating the surveillance program, Comed also addressed the credibility criteria for the integrated surveillance data discussed in 10 CFR 50.61.
4.0 CONCLUSION
- s. As discussed above, Comed's request basically involves com*o ining the weld metal surveillance data from the two units at a site (i.e., combine Byron, Units 1 and 2 data and combine Braidwood, UnNs1 and 2 data) to assess the embrittlement trends for the RPV in both units (rather than just using the surveillance data from Unit 1 to assess the RPV in UnN 1 and the data -
from UnN 2 to assess the RPV in UnN 2). The staff concludes that this general approach is acceptable and notes that it is consistent with 10 CFR 50.61(c)(2) which indicates that licensees ,
should consider information from related surveillance programs in assessing the embrittlement of their vessel.
The analysis performed by the licensee is intended to demonstrate, in part, that the design and .
operating features for the units for which the surveillance programs are to be integrated are sufficiently similar. Based on the information provided by Comed, the staff concludes that the -
Braidwood, Units 1 and 2, weld metal surveillance data can be integrated and the Byron, Units 1 and 2, weld metal surveillance data can be integrated. The staff, however, notes the following: ;
. this conclusion is based on the operating characteristics (i.e., irradiation environment) 1
- between the units remaining sufficiently similar. This limitation is being considered as part of the NRC-approved methodology for the Pressure Temperature Lim!ts Report (PTLR), which
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- i is currenth under staff review Therefore, this limMation does not invalidate the conclusion l that the surveillance programs can be inlograted.
the oroditety (as defined in 10 CFR 50.61) of the surveillance data was not considered in determining the acceptatety of the proposal. The creditety of the surveillance data is being considered as part of the NRC approved methodology for the PTLR, which is currenty under i staff review. Therefore, this does not invalidate the conclusion that the surveillance .
programs can be integrated. i there is a significant difference between the unirradiated reference temperature for the !
surveillar.co wold metal (host 442002) at Byron Units 1 and UnN 2 (i.e., RTag = 30'F for
+10'F for Byron, Unit 2). As discussed in a meeting with the ,
Dyron, industry onUnN 1, and12,1 November RTag =997 (" Meeting Summary for November 12,1997, Meeting
-l Owners Group Representatives and NEl Regardir g Review of Responses to Generic Letter 92-01, Revision 1, Supplemord 1. Responses" dated November 19,1997), the staff identified ,
this as a potential issue which may resuN in rule changes; however, this finding does not j invalidate the conclusion that the surveillance programs can be integrated. e Principal Contributor: K. Karwoski l i
Date: January 16, 1998 j 9
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