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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 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 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 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 ML20216F4921998-03-11011 March 1998 Correction to Safety Evaluation Re Revised SG Tube Rapture Analysis 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 ML20197G0041997-12-11011 December 1997 Safety Evaluation Accepting First 10-yr Interval Insp Program Plan,Rev 4 & Associated Requests for Relief for Plant 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 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 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 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 ML20247D1471989-07-18018 July 1989 SER Supporting Util Proposed Implementation of ATWS Design, Per 10CFR50.62 Requirements 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 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 ML20207Q2211987-01-16016 January 1987 SER Accepting Util 861117 Submittal on Utilization of Charcoal Absorber Matl in Safety & nonsafety-grade Air Filtration Units ML20211J7141986-11-0505 November 1986 Reevaluation & Affirmation of No Significant Change Finding Pursuant to Braidwood Station Unit 1 OL Antitrust Review ML20215D7341986-10-0101 October 1986 Safety Evaluation Re Util 860623 Request That One Startup Test Be Modified & Five Startup Tests Be Eliminated.Mod to Rod Drop Measurement Test & Elimination of Certain Other Startup Tests Acceptable ML20214N7201986-09-0909 September 1986 Safety Evaluation Conditionally Supporting Rod Swap Technique & Util Nuclear Analysis Methods for Control Rod Worth Measurements ML20206R0521986-06-25025 June 1986 Safety Evaluation Supporting Util 840229 & 860421 Responses to Generic Ltr 83-28,Items 3.2.1 & 3.2.2 Re post-maint Testing (All Other safety-related Components) ML20199K4021986-06-25025 June 1986 Safety Evaluation of Applicant 831105 & 840601 Responses to 830708 Generic Ltr 83-28,Item 2.1 (Part 1),requiring Identification of Reactor Trip Sys Components as safety- Related.Licensee Program Approved ML20197D5661986-05-0505 May 1986 SER Accepting Util Responses to Generic Ltr 83-28,Item 1.2, Post-Trip Review - Data & Info Capabilities IR 05000456/19840441986-02-25025 February 1986 Supplemental SER Re Electrical Separation Deficiencies Revealed During Const Appraisal Team Insps 50-456/84-44 & 50-457/84-40 ML20154C3641986-02-25025 February 1986 Suppl to Safety Evaluation Supporting Results of Tests Conducted by Wyle Labs Contained in Test Rept 17769-01 to Justify Less Separation Between Class 1E & non-Class 1E Cables than Required by Reg Guide 1.75 ML20209J1091985-11-0505 November 1985 SER Supporting Licensee Response to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Test Requirements That May Degrade Rather than Enhance Safety ML20138A8711985-10-0707 October 1985 Sser Supporting Util 850725 Proposed FSAR Change, Incorporating Nuclear Const Issues Group Rev 2 to Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants Into FSAR Table 3.8-2 & Section 3.10.3.2.2 ML20209G6381985-09-17017 September 1985 SER Supporting Util 831105 & 850215 Responses to Generic Ltr 83-28,Items 3.1.1 & 3.1.2, Post-Maint Testing Verification... & 4.1 & 4.5.1, Reactor Trip Sys Reliability.... Proposed Programs Meet Requirements ML20129H9071985-07-11011 July 1985 SER Accepting 850605 Submittal Re Generic Ltr 83-28,Item 1.1 on post-trip Review Program & Procedures ML20128M9391985-05-17017 May 1985 SER Based on Util 831105 Response to Generic Ltr 83-28, Item 1.1 Re post-trip Review Program Description & Procedure 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 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 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 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 ML20216D3841999-07-12012 July 1999 Revised NFM9900022, Braidwood Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function 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 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 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 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 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 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 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 ML20196A0721999-03-16016 March 1999 Cycle 8 COLR in ITS Format & W(Z) Function 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 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 ML20205M7061998-12-31031 December 1998 Unicom Corp 1998 Summary Annual Rept. with BW990001, Monthly Operating Repts for Dec 1998 for Braidwood Generating Station,Units 1 & 2.With1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Braidwood Generating Station,Units 1 & 2.With ML20206B4001998-12-31031 December 1998 Annual & 30-Day Rept of ECCS Evaluation Model Changes & Errors for Byron & Braidwood Stations ML20206U9081998-12-17017 December 1998 Cycle 8 COLR in ITS Format & W(Z) Function BW980076, Monthly Operating Repts for Nov 1998 for Braidwood Generating Station,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Braidwood Generating Station,Units 1 & 2.With ML20196A4191998-11-19019 November 1998 Safety Evaluation Accepting QA TR CE-1-A,Rev 66 Re Changes in Independent & Onsite Review Organization by Creating NSRB ML20195D3561998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Braidwood Generating Station,Units 1 & 2.With ML20155B6711998-10-26026 October 1998 Safety Evaluation Accepting Requests for Relief Associated with Second 10-yr Interval ISI Program Plan ML20207H7671998-10-0505 October 1998 Rv Weld Chemistry & Initial Rt Ndt ML20154D4401998-10-0202 October 1998 Safety Evaluation Authorizing Second 10-yr Interval ISI Program Request for Relief 12R-30 for Plant,Units 1 & 2 ML20155C2601998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Braidwood Generating Station,Units 1 & 2 ML20195F5911998-09-11011 September 1998 Special Rept:On 980812,addl Unseated Wires Were Discovered. Cause Is Unknown at Present Time.Util Evaluated Number of Unseated/Ineffective Wires & Determined Effect on Containment Structural Integrity.Commitments,Encl ML20196B3711998-09-0808 September 1998 Cycle 8 Operating Limits Rept (Olr) ML20151X6671998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Braidwood Generating Station,Units 1 & 2.With 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 ML20237E2331998-08-21021 August 1998 Revised Pages of Section 20 of Rev 66 to CE-1-A, QA Topical Rept ML20237A1091998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Braidwood Generating Station,Unit 1 & 2 ML20236N7001998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Braidwood Generating Station,Units 1 & 2 ML20198A0151998-06-18018 June 1998 10CFR50.59 Summary Rept 960619 Through 980618, Vols I & Ii,Consisting of Descriptions & SE Summaries for Changes to Procedural UFSAR Changes,Tests & Experiments & FP Rept.Without Fp,Rept ML20249A5451998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Braidwood Generating Station Units 1 & 2 ML20247F7711998-05-0808 May 1998 Special Rept:On 980403 & 980503 Seismic Monitoring Sys Was Declared Inoperable.Caused by 5-volt Power Supply & Regulator Card Failure.Imd & Sys Engineering Are Continuing to Identify & Resolve Problems So Sys Can Be Operable ML20247L7591998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Braidwood Generating Station,Units 1 & 2 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 ML20216C6621998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Braidwood Generating Station,Units 1 & 2 1999-09-30
[Table view] |
Text
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t UNITED STATES j
, g NUCLEAR REGULATORY COMMISSION J
- WASHINGTON, D.C. 30006 4001
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4 l
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l I
$ REQUEST FOR RELIEF FROM ASME CODE REPAIR REQUIREMENTS l c FOR ASME CODE CLASS 3 PIPING COM)NWEALTH EDISON COMPANY BRAIDWOOD STATION. UNIT 1 l DOCKET NO. STN 50-456
1.0 INTRODUCTION
- According to 10 CFR 50.55a(g), nuclear power facility piping and components 2
are required to meet the applicable requirements of Section XI of the American
- Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code).
l Section XI of the Code specifies Code-acceptable repair methods for flaws that exceed Code acceptance limits in piping that is in-service. A Code repair is i required to restore the structural integrity of flawed Code piping,
- independent of the operational mode of the plant when the flaw-is detected.
! Those repairs not in compliance with Section XI of the Code are non-Code i repairs. However, the implementation of required Code (weld) repairs to ASME
- Code Class 1, 2 or 3 systems is often impractical for nuclear licensees since the repairs normally require an isolation of the system requiring the repair,
- and often a shutdown of the nuclear power plant.
Alternatives to Code requirements may be used by nuclear licensees when authorized by the Director of the Office of Nuclear Reactor Regulation if l the proposed alternatives to the requirements are such that they are shown to provide an acce) table level of quality and safety in lieu of the Code requirements (10 C i or if compliance with the Code
!- requirements wouldRresult 50.55a(a)(3)(h)p],or in hards i unusual difficulty without a compensating increase in the level of quality and safety ~[10 CFR
]. 50.55a(a)(3)(ii)].
~
A licensee may also submit requests for relief from certain Code requirements
! when a licensee has determined that conformance with certain Code requirements I iii)]. Pursuant to l is impractical for6its1 facility [10 CFR 50.55a(g)(5)(determinations the Consission will evaluate of 4
10 CFR 50.55a(g)(d)m(ay),
impracticality an grant relief and may impose alternative requirements j as it determines is authorized by law.
l Generic Letter (GL) 90-05, entitled " Guidance for Performing Temporary Non- )
- Code Repair of ASME Code Class 1, 2 and 3' Piping," dated June 15, 1990, !
. provides guidance for the staff in evaluating relief requests submitted by licensees for temporary non-Code repairs of Code Class 3 piping. The staff i uses the guidance in GL 90-05 as its criteria for making its safety evaluation
- of relief requests for temporary non-Code repairs of Code Class 3 piping. !
[
! 9705130305 970508 l
!- PDR ADOCK 05000456 i P PDR .
i
i t
2- t
2.0 BACKGROUND
f In December of 1996, Commonwealth Edison Company (Comed, the licensee) !
identified a leak on a drain line of Train A of the essential service water !
system (SX) at Braidwood Station, Unit 1. The leak was located on a 6-inch by 2 8-inch reducer located on line ISX93AA-8" downstream from the 1A SX System '
4 Strainer. The licensee's operability assessment and its bases were provided j to the staff by letter dated December 31, 1996. By letter dated January 9, 1 '1997, the licensee requested relief from the ASME Code,Section XI, repair or
- replacement requirements under impracticality provisions of 10 CFR -
t (g .
! 50.55(a)lua) flaw eva tion that was performed by the licensee in accordance with theThe licensee j guidelines and acceptance criteria contained in GL 90-05.
3.0 LICENSEE'S RELIEF REOUEST 3.1 . Components for Which Relief is Reauested
- j. ASME Code Class 3 essential service water system , 6-inch by 8-inch reducer located on line ISX93AA-8" downstream from the 1A SX System Strainer.
3.2 Section XI Edition for Braidwood Station. Unit 1 1983 Edition of the ASME Code,Section XI, inclusive of the 1983 summer Addenda.
3.3 ASME.Section XI. Code Reauirement l The ASME Code,Section XI, requires that repairs t replacements of ASME Code Class components be performed in accordance with tales found in Articles i IWA-4000 or IWA-7000, respectively. The intent of these rules serve to
! provide an acceptable means of restoring the structural integrity of a
~ degraded Code Class system back to the original design requirements.
j .3.4 Content of the Relief Reauest
. Relief is sought from performing a repair or replacement of 6-inch by 8-inch reducer per the requirements of Article IWA-4000 or IWA-7000, respectively.
Relief was sought until the next Braidwood Station, Unit 1, scheduled outage which_ began on March 29, 1997.
3.5 Basis for Relief i
Request for relief has been submitted under the impracticality provisions of
- 10 CFR 50.55a(g). The licensee has stated that the reducer was structurally
[ capable of meeting its intended function based upon flaw evaluation using the
! "Through-Wall Flaw" approach of GL 90-05. This evaluation demonstrated that
- the flaw stability and structural integrity would be ensured until permanent
- Code. repair was performed. The' licensee attempted to perform a Code repair or i replacement during the week of January 6,1997, without success. Successful
- completion of the Code repair or replacement depended on the isolation of this
} piping from the cooling lake. Complications in isolating this portion of the 3
i i
s e ,- a
__m _ . . . . _ _ _ _ _ _ _ . . _ _ _ _ _ . _ _ - - . _ . _.___ _
l j'
.SX pipe from the cooling lake using the upstream butterfly valve were
- encountered. The butterfly valve failed to provide isolation to perform Code
repair or replacement without shutting the unit down and, thus, the Code repair or replacement was impractical. Replacement was completed during the
- refueling outage which began on March 29, 1997.
3.6 Licensee's Alternative Proaram
- 1. Performing of temporary non-code re) air of the leaking area using gasket
- material mechanically attached to tie pipe. The temporary non-code repair was to remain in place until a Code repair or replacement is completed during the .t refueling outage.
- 2. Monitoring of the degraded area by plant operators.
4 l- 3. Code repair or replacement of 6-inch by 8-inch reducer during the next
! refueling outage (scheduled to begin on March 29,1997).
i 4.0 STAFF EVALUATION i
- 4.1 Goerabi'ity Determination. Root Cause Analysis and Structural Intearity Evaluat- on
-The licensee determined that the 6-inch by 8-inch reducer is located on the
- drain line of Braidwood Station, Unit 1, Train A, essential service water i
! strainer. The reducer is made from ferritic carbon steel material (SA234-
- and is located on an ASME Code, Class 3 system. The reducer was i Grade visuallyWPB) ar j ultrasonically examined to identify the physical characteristics
- of the flaw and the extent of degradation. The flaw was characterized as a
, through-wall hole with diameter of approximately 1/8-inches. Ultrasonic examination was performed in order to obtain thickness measurements. The
[ results of the thickness measurements revealed that the wall thickness i surrounding the flaw ranged from 0.335 inches to 0.422 inches. The required l- piping minimum wall designed thickness is 0.036 inches. The licensee l performed an evaluation of the reducer using the guidance provided in GL 90-05
- and found that the reducer is structurally capable of performing its intended i function until the next scheduled refueling outage. The licensee also l j performed a root cause analysis of the degraded reducer. Based upon results of visual and ultrasonic examination, the licensee concluded that the degradation of the reducer was due to localized corrosion. In addition,
. augmented inspections of five additional locations were performed. The augmented inspections utilized ultrasonic and radiographic examination
- techniques in accordance with the guidance provided in GL 90-05. Three of i these five locations are sister components in both Units 1 and 2. All areas
, inspected had no localized wall thinning flaws below the applicable Code
- required minimum design wall thickness.
i
! 4.2~ Auamented Inspection
- The leak was located on a 6-inch by 8-inch carbon steel reducer located on the
! -essential. service water piping. The reducer was visually and ultrasonically
[ examined to assess the severity of wall degradation. The reducer was l:
J ,
j :
i determined to be structurally capable of performing its intended function until a Code repair or replacement was accomplished. An augmented ins >ection using ultrasonic and radiogra> hic examination was performed on five ot1er similar locations (three of tiose five locations were sister components in 3 both Units 1 and 2) in accordance with the guidance provided in GL 90-05. The examination results revealed that all locations met the minimum required wall
- thickness and, thus, the piping is acceptable for service.
l
- 4.3 Pronosed Temoorary Non-Code Renair and Monitorina Provisions At the time of its submittal, the licensee performed a temnorary non-Code i
repair of the leaking area using gasket material mechanically attached to the pipe. The licensee intended that the temporary non-code repair remain in
- place until a Code repair or replacement was completed during the next
- refueling outage. The degraded area was monitored by plant operators. The l degraded section of pipe was replaced during the refueling outage that began
- on March 29, 1997.
5.0 CONCLUSION
S
! The staff has determined that the licensee's flaw evaluation has been i consistent with the guidelines and acceptance criteria of GL 90-05. The staff, therefore, finds the licensee's structural integrity and operability assessments to be acceptable. The licensee has performed a non-code repair of i the leaking reducer using gasket material mechanically attached to the pipe.
! The temporary non-code repair remained in place until replacement was completed during the next refueling outage. While the non-Code repair was in l place, the degraded area was monitored by plant operators.
Furthermore, the staff finds that performance of an immediate Code repair
- would constitute an undue burden (create undue hardship) upon the licensee
! since it would require an isolation of the affected essential service water ;
2 system piping.
- unit dow(n. ESWS)
S hutting the unit down was not in the best interest of plantSuch an isolat i safety, given the magnitude of the flaw and the licensee's alternative program. The staff, therefore, concludes that authorization of the licensee's
! alternative program would provide an acceptable level of quality or safety, is 2
authorized by law and will not endanger life or property or the common defense i and security, and is othenvise in the public interest giving due consideration to the burden upon the licensee and facility that could result if the Code requirements were imposed on the facility. Pursuant to 10 CFR l 50.55a(g)(6)(1) the alternative is authorized until the next scheduled outage.
! As noted previously, replacement of the leaking section of piping was completed during the refueling outage which began on March 29, 1997.
i Principal Contributor: G. Georgiev k Date: May 8, 1997 i
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