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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217B6261999-10-0404 October 1999 Safety Evaluation Supporting Amend 202 to License DPR-51 ML20212L1141999-10-0101 October 1999 Safety Evaluation Granting Request for Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c ML20212F5261999-09-22022 September 1999 SER Approving Request Reliefs 1-98-001 & 1-98-200,parts 1,2 & 3 for Second 10-year ISI Interval at Arkansas Nuclear One, Unit 1 ML20212D5611999-09-14014 September 1999 Safety Evaluation Supporting Amend 200 to License DPR-51 ML20212D4091999-09-14014 September 1999 Safety Evaluation Supporting Amend 201 to License DPR-51 ML20211P9551999-09-0909 September 1999 Safety Evaluation Supporting Amend 199 to License DPR-51 ML20211K3311999-08-26026 August 1999 Safety Evaluation Supporting Amends 198 & 209 to Licenses DPR-51 & NPF-06,respectively ML20211F4281999-08-25025 August 1999 Safety Evaluation Concluding That Licensee Provided Acceptable Alternative to Requirements of ASME Code Section XI & That Authorization of Proposed Alternative Would Provide Acceptable Level of Quality & Safety ML20196J2771999-06-29029 June 1999 Safety Evaluation Supporting Amend 208 to License NPF-6 ML20195G9531999-06-10010 June 1999 Safety Evaluation Supporting Amend 197 to License DPR-51 ML20207G2901999-06-0707 June 1999 Safety Evaluation Supporting Amend 207 to License NPF-6 ML20207E7231999-06-0202 June 1999 Safety Evaluation Authorizing Proposed Alternative Exam Methods Proposed in Alternative Exam 99-0-002 to Perform General Visual Exam of Accessible Areas & Detailed Visual Exam of Areas Determined to Be Suspect ML20207B7071999-05-19019 May 1999 Safety Evaluation Supporting Amends 196 & 206 to Licenses DPR-51 & NPF-6,respectively ML20207A4651999-05-19019 May 1999 Safety Evaluation Supporting Amend 205 to License NPF-6 ML20206M7711999-05-11011 May 1999 SER Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 ML20206F0691999-04-29029 April 1999 Safety Evaluation Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205Q7721999-04-16016 April 1999 Safety Evaluation Supporting Amends 195 & 203 to Licenses DPR-51 & NPF-6,respectively ML20205M6941999-04-12012 April 1999 Safety Evaluation Granting Relief for Second 10-yr Inservice Inspection Interval for Plant,Unit 1 ML20205D6061999-03-31031 March 1999 Safety Evaluation Supporting Licensee Proposed Approach Acceptable to Perform Future Structural Integrity & Operability Assessments of Carbon Steel ML20205D4711999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Employ Alternative Welding Matls of Code Cases 2142-1 & 2143-1 for Reactor Coolant System to Facilitate Replacement of Steam Generators at Arkansas Nuclear One,Unit 2 ML20204H7451999-03-23023 March 1999 Safety Evaluation Supporting Amend 202 to License NPF-6 ML20204B1861999-03-15015 March 1999 Safety Evaluation Authorizing Licensee Request for Alternative to Augmented Exam of Certain Reactor Vessel Shell Welds,Per Provisions of 10CFR50.55a(g)(6)(ii)(A)(5) ML20202B4581999-01-26026 January 1999 Safety Evaluation Supporting Amend 200 to License NPF-6 ML20199F7931999-01-19019 January 1999 Safety Evaluation Supporting Amend 199 to License NPF-6 ML20206R8071999-01-13013 January 1999 Safety Evaluation Supporting Amend 198 to License NPF-6 ML20198S5231998-12-31031 December 1998 Safety Evaluation Supporting Amend 196 to License NPF-6 ML20198S0861998-12-31031 December 1998 Safety Evaluation Supporting Amend 197 to License NPF-6 ML20198S6061998-12-31031 December 1998 Safety Evaluation Supporting Amend 194 to License DPR-51 ML20198M7491998-12-29029 December 1998 Safety Evaluation Supporting Amend 195 to License NPF-6 ML20198M7841998-12-29029 December 1998 SER Accepting Util Proposal to Use ASME Code Case N-578 as Alternative to ASME Code Section Xi,Table IWX-2500 for Arkansas Nuclear One,Unit 2 ML20198M3691998-12-23023 December 1998 Safety Evaluation Supporting Amend 194 to License NPF-6 ML20196F4911998-12-0101 December 1998 SER Accepting Request for Relief ISI2-09 for Waterford Steam Electric Station,Unit 3 & Arkansas Nuclear One,Unit 2 ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program ML20154J2471998-10-0909 October 1998 SER Accepting Inservice Testing Program,Third ten-year Interval for License DPR-51,Arkansas Nuclear One,Unit 1 ML20153F5891998-09-23023 September 1998 Safety Evaluation Supporting Amends 193 & 193 to Licenses DPR-51 & NPF-6,respectively ML20237A2451998-08-0707 August 1998 Safety Evaluation Supporting Amend 192 to License NPF-6 ML20236R3871998-07-13013 July 1998 Safety Evaluation Supporting Amends 192 & 191 to Licenses DPR-51 & NPF-6,respectively ML20248D7491998-05-28028 May 1998 Safety Evaluation Accepting Licensee Request for Relief from ASME Code Repair Requirements for ASME Code Class 3 Piping ML20217A7211998-04-17017 April 1998 Safety Evaluation Supporting Proposed Alternative for ANO-1 to Implement Code Case N-533 (w/4 H Hold Time at Test Conditions Prior to VT-2 Visual Exam) ML20216E5251998-04-10010 April 1998 Safety Evaluation Supporting Amend 191 to License DPR-51 ML20216D7371998-04-10010 April 1998 Safety Evaluation Supporting Amend 190 to License DPR-51 ML20217P8281998-04-0707 April 1998 Safety Evaluation Accepting Relief Authorization for Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250 Bolting Exam for Plants,Per 10CFR50.55a(a)(3)(i) ML20216F0151998-03-12012 March 1998 Safety Evaluation Supporting Amend 189 to License NPF-6 ML20216D5131998-03-12012 March 1998 Safety Evaluation Supporting Amend 190 to License NPF-6 ML20216D6111998-03-12012 March 1998 Safety Evaluation Supporting Amend 188 to License NPF-6 ML20198G0441997-12-23023 December 1997 Safety Evaluation Supporting Amend 187 to License NPF-6 ML20199H3711997-11-19019 November 1997 SER Accepting Approving Request Relief from Requirements of Section XI, Rule for Inservice Insp of NPP Components, of ASME for Current or New 10-year Inservice Insp Interval IAW 50.55(a)(3)(i) of 10CFR50 ML20216E9921997-09-0404 September 1997 Safety Evaluation Accepting 970623 Request for Relief Re Authorization for Use of ASME Code Case N-416-1 & N-532,ISI Program for Listed Plants ML20141H8411997-07-30030 July 1997 Safety Evaluation Accepting Use of Code Case N-508-1 for All Four Plants for Rotation of Serviced Snubbers & Pressure Relief Valves for Purpose of Testing in Lieu of ASME Code ML20148K9761997-06-14014 June 1997 Safety Evaluation Granting Licensee Exemption from Requirements of 10CFR50,App R,Section III.0 1999-09-09
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217L8931999-10-31031 October 1999 Rev 1 to BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs 05000313/LER-1999-313, :on 990911,AI of EFWS Occurred During Plant Shutdown as Result of Securing Running RCPs Due to Reverse Rotation of Idle Pump.Caused by Failure of Motor anti-rotation Device.Device Was Replaced.With1999-10-11011 October 1999
- on 990911,AI of EFWS Occurred During Plant Shutdown as Result of Securing Running RCPs Due to Reverse Rotation of Idle Pump.Caused by Failure of Motor anti-rotation Device.Device Was Replaced.With
ML20217B6261999-10-0404 October 1999 Safety Evaluation Supporting Amend 202 to License DPR-51 ML20212L1141999-10-0101 October 1999 Safety Evaluation Granting Request for Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c 0CAN109902, Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With 05000368/LER-1999-005-02, :on 990831,two Trip Functions of One Core Protection Calculator Channel Potentially Inoperable Longer than TS Allow Was Discovered.Caused by Resistance Fluctuations.Util Plans to Replace Switch.With1999-09-30030 September 1999
- on 990831,two Trip Functions of One Core Protection Calculator Channel Potentially Inoperable Longer than TS Allow Was Discovered.Caused by Resistance Fluctuations.Util Plans to Replace Switch.With
ML20216J6271999-09-27027 September 1999 Rev 0 to CALC-98-R-1020-04, ANO-1 Cycle 16 Colr ML20212F5261999-09-22022 September 1999 SER Approving Request Reliefs 1-98-001 & 1-98-200,parts 1,2 & 3 for Second 10-year ISI Interval at Arkansas Nuclear One, Unit 1 ML20212D5611999-09-14014 September 1999 Safety Evaluation Supporting Amend 200 to License DPR-51 ML20212D4091999-09-14014 September 1999 Safety Evaluation Supporting Amend 201 to License DPR-51 ML20211P9551999-09-0909 September 1999 Safety Evaluation Supporting Amend 199 to License DPR-51 0CAN099907, Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with ML20211K3311999-08-26026 August 1999 Safety Evaluation Supporting Amends 198 & 209 to Licenses DPR-51 & NPF-06,respectively ML20211F4281999-08-25025 August 1999 Safety Evaluation Concluding That Licensee Provided Acceptable Alternative to Requirements of ASME Code Section XI & That Authorization of Proposed Alternative Would Provide Acceptable Level of Quality & Safety 05000313/LER-1999-002-04, :on 990706,TS Allowable Outage Time for One EDG Was Exceeded Due to Idler Gear Stub Shaft Bracket Bolting Failure.Caused by Loss of Bolt pre-load.Installed New Modified Idler Stub Shaft Assembly.With1999-08-0505 August 1999
- on 990706,TS Allowable Outage Time for One EDG Was Exceeded Due to Idler Gear Stub Shaft Bracket Bolting Failure.Caused by Loss of Bolt pre-load.Installed New Modified Idler Stub Shaft Assembly.With
0CAN089904, Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with ML20210K8831999-07-29029 July 1999 Non-proprietary Addendum B to BAW-2346P,Rev 0 Re ANO-1 Specific MSLB Leak Rates 0CAN079903, Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with ML20196J2771999-06-29029 June 1999 Safety Evaluation Supporting Amend 208 to License NPF-6 05000313/LER-1999-001-04, :on 990521,automatic Actuation of CREVS Occurred Due to Higher than Normal Radiation at Detector When Radioactive Filter Was Moved in Adjacent Area.Training Has Been Provided to Personnel.With1999-06-21021 June 1999
- on 990521,automatic Actuation of CREVS Occurred Due to Higher than Normal Radiation at Detector When Radioactive Filter Was Moved in Adjacent Area.Training Has Been Provided to Personnel.With
ML20195G9531999-06-10010 June 1999 Safety Evaluation Supporting Amend 197 to License DPR-51 ML20207G2901999-06-0707 June 1999 Safety Evaluation Supporting Amend 207 to License NPF-6 ML20207E7231999-06-0202 June 1999 Safety Evaluation Authorizing Proposed Alternative Exam Methods Proposed in Alternative Exam 99-0-002 to Perform General Visual Exam of Accessible Areas & Detailed Visual Exam of Areas Determined to Be Suspect ML20196A0191999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Arkansas Nuclear One,Units 1 & 2.With ML20196A6251999-05-31031 May 1999 Non-proprietary Rev 0 to TR BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20195D1991999-05-28028 May 1999 Probabilistic Operational Assessment of ANO-2 SG Tubing for Cycle 14 ML20207A4651999-05-19019 May 1999 Safety Evaluation Supporting Amend 205 to License NPF-6 ML20207B7071999-05-19019 May 1999 Safety Evaluation Supporting Amends 196 & 206 to Licenses DPR-51 & NPF-6,respectively 05000368/LER-1999-004-01, :on 990414,noted That Two Trip Functions of One Core Protection Calculator Channel Were Potentially Inoperable.Caused by Excore Detector Drift.Affected Trip Functions Have Remained Bypassed or Tripped.With1999-05-12012 May 1999
- on 990414,noted That Two Trip Functions of One Core Protection Calculator Channel Were Potentially Inoperable.Caused by Excore Detector Drift.Affected Trip Functions Have Remained Bypassed or Tripped.With
ML20206M7711999-05-11011 May 1999 SER Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 0CAN059903, Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with ML20206F0691999-04-29029 April 1999 Safety Evaluation Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205Q7721999-04-16016 April 1999 Safety Evaluation Supporting Amends 195 & 203 to Licenses DPR-51 & NPF-6,respectively ML20205M6941999-04-12012 April 1999 Safety Evaluation Granting Relief for Second 10-yr Inservice Inspection Interval for Plant,Unit 1 ML20205D6061999-03-31031 March 1999 Safety Evaluation Supporting Licensee Proposed Approach Acceptable to Perform Future Structural Integrity & Operability Assessments of Carbon Steel ML20205R6351999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Ano,Units 1 & 2. with ML20205D4711999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Employ Alternative Welding Matls of Code Cases 2142-1 & 2143-1 for Reactor Coolant System to Facilitate Replacement of Steam Generators at Arkansas Nuclear One,Unit 2 ML20204H7451999-03-23023 March 1999 Safety Evaluation Supporting Amend 202 to License NPF-6 05000368/LER-1999-003, :on 990224,noted That One Excore Nuclear Instrumentation Channel Was Inoperable Longer than Allowed by Ts.Caused by Detector Failure.Proposed Exigent TS Change Was Submitted to Allow Operation.With1999-03-23023 March 1999
- on 990224,noted That One Excore Nuclear Instrumentation Channel Was Inoperable Longer than Allowed by Ts.Caused by Detector Failure.Proposed Exigent TS Change Was Submitted to Allow Operation.With
ML20204B1861999-03-15015 March 1999 Safety Evaluation Authorizing Licensee Request for Alternative to Augmented Exam of Certain Reactor Vessel Shell Welds,Per Provisions of 10CFR50.55a(g)(6)(ii)(A)(5) 05000368/LER-1999-001, :on 990108,TS Requirements for as-lift Settings of MSSV Were Not Met.Caused by Errors in Determination of Effective Seating Area by Vendor During Test Device Qualification.Replaced All Mssvs.With1999-03-0909 March 1999
- on 990108,TS Requirements for as-lift Settings of MSSV Were Not Met.Caused by Errors in Determination of Effective Seating Area by Vendor During Test Device Qualification.Replaced All Mssvs.With
05000368/LER-1999-002-01, :on 990202,noted Failure to Verify Station Battery cell-to-cell & Terminal Tightness.Caused by Inadvertent Omission of Requirements During Procedure Revs. Svc Test Procedures Were Revised.With1999-03-0404 March 1999
- on 990202,noted Failure to Verify Station Battery cell-to-cell & Terminal Tightness.Caused by Inadvertent Omission of Requirements During Procedure Revs. Svc Test Procedures Were Revised.With
0CAN039904, Monthly Operating Repts for Feb 1999 for Ano,Units 1 & 2. with1999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Ano,Units 1 & 2. with ML20212G6381999-02-25025 February 1999 Ano,Unit 2 10CFR50.59 Rept for 980411-990225 ML20203E4891999-02-11011 February 1999 Rev 1 to 97-R-2018-03, ANO-2,COLR for Cycle 14 05000368/LER-1999-001-01, :on 990108,four Main Steam Safety Valves as- Found Lift Values Failed to Meet TS Requirements.Cause of Evaluation in progress.Short-term & long-term Corrective Actions Will Be Described in Suppl to Rept.With1999-01-28028 January 1999
- on 990108,four Main Steam Safety Valves as- Found Lift Values Failed to Meet TS Requirements.Cause of Evaluation in progress.Short-term & long-term Corrective Actions Will Be Described in Suppl to Rept.With
ML20202B4581999-01-26026 January 1999 Safety Evaluation Supporting Amend 200 to License NPF-6 05000313/LER-1998-005-03, :on 981225,two Manual RTs & Manual Actuations of EFWS Due to Reduced Cw Flow to Mc.Caused by Large Intrusions of Fish Exceeding Removal Capability of Traveling Screen.Increased Fish Removal Capability.With1999-01-21021 January 1999
- on 981225,two Manual RTs & Manual Actuations of EFWS Due to Reduced Cw Flow to Mc.Caused by Large Intrusions of Fish Exceeding Removal Capability of Traveling Screen.Increased Fish Removal Capability.With
ML20199F7931999-01-19019 January 1999 Safety Evaluation Supporting Amend 199 to License NPF-6 05000368/LER-1998-008-01, :on 981215,one CR Emergency Chiller,Part of CREVS Discovered to Be Inoperable,While Chiller in Other Train Was Out of Svc for Planned Maint.Caused by Equipment Failure.Operations Personnel Verified.With1999-01-14014 January 1999
- on 981215,one CR Emergency Chiller,Part of CREVS Discovered to Be Inoperable,While Chiller in Other Train Was Out of Svc for Planned Maint.Caused by Equipment Failure.Operations Personnel Verified.With
1999-09-09
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p ps vtoy,k UNITED STATES j
g NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. enana annt SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST TO USE CODE CASE N-533 FOR !
ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT 1
' DOCKET NO. 50-313 l
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1.0 INTRODUCTION
The Technical Specifications (TSs) for Arkansas Nuclear One, Unit 1 (ANO-1) state that the inservice inspection (ISI) of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 50.55a(a)(3) states that attematives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed altematives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulties ;
without a compensating increase in the level of quality and safety. ,
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120 month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for ANO-1 third 10-year ISI interval is the 1992 Edition, with pressure testing requirements from the 1993 Addenda. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.
Pursuant to 10 CFR 50.55a(g)(5),if the licensee determines that conformance with an examina-tion requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Comm!ssion in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose altemative requirements tc,at 9804220317 900417 DR ADOCK 0500 3 ENCLOSURE
l are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
2.0 BACKGROUND
in a letter dated June 25,1997, Entergy Operations, Inc. (the licensee), submitted to the NRC its third 10 year ISI plan and requests for altematives for ANO 1. As part of the plan, Request Number 97-005 requested relief from Article IWA-5242(a) of Section XI for Code Class 1 and 2 piping systems. The portion of Request Number 97-005 related to Code Class 1 piping systems pertains to the use of ASME Code Case N-533, "Altemative Requirements for VT-2 Visual Exemination of Class 1 Insulated Pressure-Retaining Bolted Connections,Section XI, Division 1,"
which has been widely requested by licensees and, with the limitations discussed below, has been previously approved by the staff. Given the potential benefits of applying Code Case N 533 during a refueling outage scheduled for early 1998, the staff has opted to separately evaluate the l licensee's request pertaining to Code Case N-533 pending the completion of its review of the remainder of the plan for the third 10 year ISI interval for ANO-1 (including that portion of i Request Number 97-005 pertaining to Code Class 2 piping systems).
1 3.0 EVALUATION i The staff, has evaluated the information provided by the licensee in support of its third 10-yerr interval ISI plan, request to implement Code Case N 533 as an altemative to the Code requirements for ANO-1.
Code Reauirement: l l
i IWA-5242(a) states that for systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure-retaining bolted connections for a direct VT-2 visual examination.
Code cases are periodically published by ASME to either clarify the intent of the Code rules or to i provide altemative rules to existing Code requirements. Use of these nonmandatory Code cases for inservice inspection is subject to general acceptance by the NRC staff and incorporation into Regulatory Guide (RG) 1.147. Pursuant to 10 CFR 50.55a, other Code cases may be used provided specific authorization is granted.
Licensee's Code Reimf Recuest Tne licensee has requested to use Code Case N-533, "A!temative Requirements for VT-2 Visual Examination of Class 1 Insulated Pressure Retaining Bolted Connections,Section XI, Division 1."
Licensee's Basis for Recuesting Relief (as stated)-
" Pursuant to 10 CFR 55a(s)(3)(ii), en altamative is requested on the basis that the original requirements would result in hardship without a compensating increase in the level of quality and safety.
3-Systems which are borated for the purpose of ctw*olling reactivity at ANO-1 include reactor coolant, decay heat removal, high pressure inlede. and make-up. These systems encompass a large portion of the overall181 program and phy" h cover a large expanse of the reactor building. Many areas in which this piping and the m..vciated bolted connections are located are difficult to access (e.g., scaffold and/or ladder installation is required) and many of these areas are located such that significant radiation exposures would be encountered, in order to identify leakage to be repaired during the outage, the preferred time frame to perform this inspection is the beginning of the outage subsequent to depressurization of the reactor coolant system. To perform these inspections at pressure would involve holding the reactor coolant system at operating pressure and temperature for an extended period of time to allow for scaffold construction, insulation removal and VT-2 inspection.
This is normally a relatively short time frame when the unit is transitioning to cold shutdown.
Holding the unit at normal operating temperature and pressure for an extended period of time would result in a significant delay in going to cold shutdown.
Performing this inspection at the end of the outage would be ineffective, since finding leakage at that time would constitute bringing the unit back to cold shutdown to perform the repair and then beginning the start up process over again. Also, in order to reinsulate the reactor coolant system at the end of the outage, the unit would have to remaM at hot standby while the insulation is reinstalled and the scaffolding is removed. Typically, the reinstallation of insulation and removal of scaffolding is performed prior to leaving cold shutdown.
In addition, the removal ani,einstallation of insulation with plant equipment in operation at
. system pressure and temperature increases the rid of personnel injury and presents a safety concem to plant personnel. The personnel risk anc radiation exposure is significant for the removal and reinstallation of insulation at these bol'.ed connections during pressure testing l activities. !
A VT-2 visual examination with the system depressurized would still provide adequate detection j of leakage because boric acid residue can be easily detected with insulation removed at the bolted connection. ,
I Based on the previously stated reasons, Entergy Operations requests relief from the inspection at operating pressure requirements detailed in IWA 5242(a). In lieu of these requirements, Entergy Operations proposes the altemative examination requirements that follow."
Licensee's Proposed Altemative Examination (as stated):
"A system pressure test with a minimum four hour hold time and VT-2 visual examination shall be performed each refueling outage without the removal of insulation on systems borated for the purpose of controlling reactivity.
Each refuelog outage, the insulation shall be removed from the bolted connections in systems borated for the purpose of controlling reactivity, and a VT-2 visual examination shall be performed on each of the connections. During this VT-2 examination, the connections are not required to be pressurized. Any evidence of leakage shall be evaluated in accordance with IWA-5250.
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These altemative examination requirements are the same as those specified in ASME Section XI Code Case N-533 as approved by the Board of Nuclear Codes and Standards, with the additional four hour hold time provision stated above.
These attematives provide reasonable assurance that safety and integrity will be maintained for bolted connections in systems borated for the purpose of controlling reactivity."
Evaluation:
Paragraph IWA-5242(a) requires the removal of all insulation from pressure-retaining bolted connections in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tests. The licensee has proposed to use Code Case N 533, which requires (1) performing the Code-required pressure test without removing the insulation, (2) examining Class 1 bolted connections, each refueling outage, at atmospheric or static pressure with insulation removed, and (3) evaluating any evidence of leakage in accordance with IWA-5250.
Paragraph IWA-5242(a) provides requirements to ensure that leakage or evidence of leakage at bolted connections is found. Performing a W-2 visual examination during system pressure tests as required by Code Case N-533, with the insulation in place will likely result in the detection of any significant leakage. The licensee has committed to a 4-hour hold period prior to performing the VT-2 visual examination during system pressure tests. This hold period is consistent with the staff's acceptance of Code Case N-533 for other licensees. Furthermore, performing a VT-2 visual examination after removal of the insulation at atmospheric or static pressure during outages, as specified by Code Case N-533, will allow for examination for evidence of borated water leakage. The Code Case states that any evidence of leakage must be evaluated in accordance with IWA-5250 of Section XI.
The NRC has accepted the use of Code Case N-533 for numerous licensees (provided the 4-hour hold time at test condidtions is observed prior to the VT-2 visual examination) in accordance with 10 CFR 50.55(a)(3)(i) after finding the proposed altemative provides an acceptable level of quality and safety. The licensee has requested the use of an altematve to the ASME Code (the use of Code Case N-533 with the added 4-hour hold period) pursuant to 10 CFR 50.55(a)(3)(ii) on the basis that the original requirement results in hardship without a compensating increase in the level of quality and safety. The hardship associated with the current requirement includes increasing the duration of outages, increasing the risk of personnel injury, and increasing the radiation exposure of plant personnel. Given the staff's finding that the proposed attemative provides comparable assurance of the detection of leakage from borated systems, the staff also agrees that the hardships described by the licensee are witheut compensating increases in the level of quality or safety.
4.0 CONCLUSION
The staff has reviewed the licensee's submittal and concluded that by using Code Case N-533 l the licensee will locate and evaluate leakage, or evidence of leakage, in a manner comparable to j the requirements of the Code, the licensee's proposed altemative provides an acceptable level of i quality and safety and that the existing requirement results in hardship without a compensating j l
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increase in the level of quality or safety. Therefore, the licensee's proposed attemative for ANO-1 to implement Code Case N-533 (with a 4-hour hald time at test conditions prior to the VT-2 visual examination), is authorized pursuant to 10 CFR 50.55a(s)(3)(ii). The use of this Code Case is authorized for the duration of the respective current applicable 10-year ISI
- interval, or until the Code Case is approved for general use by reference in RG 1.147. After that l time, the licensee may continue to use Code Case N-533 with the limitations, if any, listed in RG 1.147.
Principal Contributors: T. McLellan W. Reckley l Date: April 17, 1998
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