Letter Sequence Approval |
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TAC:52901, (Approved, Closed) TAC:52902, (Approved, Closed) TAC:52903, (Approved, Closed) TAC:53737, (Approved, Closed) TAC:53738, (Approved, Closed) TAC:53739, (Approved, Closed) TAC:53964, (Approved, Closed) TAC:53965, (Approved, Closed) TAC:53966, (Approved, Closed) |
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MONTHYEARML20133D0401985-01-25025 January 1985 Responds to NRC Re Violations Noted in Insp Repts 50-259/84-23,50-260/84-23 & 50-296/84-23.Corrective Actions: Vendor Manual Control Program Being Developed & Implemented to Ensure Plant Procedures Updated W/Current Vendor Info Project stage: Other ML20127D6641985-06-17017 June 1985 Safety Evaluation Re Licensee Response to Item 1.1 of Generic Ltr 83-28 on post-trip Review Program & Procedures. Post-trip Review Program & Procedures Acceptable Project stage: Approval ML20133K5511985-07-23023 July 1985 Provides Status Update on Util Commitments Re Vendor Manual Control Program Per Insp Repts 50-259/84-23,50-260/84-23 & 50-296/84-23.Phase I Considered Complete.Schedule for Phase II Manuals Will Be Submitted by 850801 Project stage: Other ML20137G1081985-08-0909 August 1985 Submits Schedule for Phase II Vendor Manual Control Program, Per Discussing Insp Repts 50-259/84-23, 50-260/84-23 & 50-296/84-23.Schedule for Vendor Program Will Be Submitted by 851101 Project stage: Other ML20148K9051988-03-23023 March 1988 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1 & 3.2.2, Post-Maint Testing of Reactor Trip Sys & All Other Safety-Related Components Project stage: Approval 1985-07-23
[Table View] |
Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1 & 3.2.2, Post-Maint Testing of Reactor Trip Sys & All Other Safety-Related ComponentsML20148K905 |
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Browns Ferry |
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03/23/1988 |
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NRC OFFICE OF SPECIAL PROJECTS |
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ML20148K895 |
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References |
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GL-83-28, TAC-52901, TAC-52902, TAC-52903, TAC-53737, TAC-53738, TAC-53739, TAC-53964, TAC-53965, TAC-53966, NUDOCS 8804010057 |
Download: ML20148K905 (5) |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20212E6341999-09-23023 September 1999 Suppl to SE Resolving Error in Original 990802 Se,Clarifying Fact That Licensee Has Not Committed to Retain Those Specific Compensatory Measures That Were Applied to one-time Extension ML20212D3831999-09-20020 September 1999 Safety Evaluation Supporting Proposed Rev to Withdrawal Schedule for First & Third Surveillance Capsules for BFN-3 RPV ML20210N1221999-08-0202 August 1999 Safety Evaluation Accepting Licensee Request for Relief from ASME B&PV Code,Section XI Requirements.Request 3-ISI-7, Pertains to Second 10-year Interval ISI for Plant,Unit 3 ML20209J0771999-07-16016 July 1999 Safety Evaluation Concluding That Licensee Provided Adequate Information to Resolve ampacity-related Points of Concern Raised in GL 92-08 for BFN & That No Outstanding Issues Re GL 92-08 Ampacity Issues for Browns Ferry NPP Exist ML20196F8811999-06-23023 June 1999 Safety Evaluation Accepting GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves ML20205F9341999-04-0101 April 1999 Safety Evaluation Authorizing Licensee 990108 Relief Request PV-38,from Requirements of ASME BPV Code Section XI IST Testing,Valve Program for Plant,Units 1,2 & 3 ML20153H3141998-09-28028 September 1998 SE Accepting Util Proposed Alternatives as Contained in Relief Requests 3-SPT-3 & 3-SPT-4 for Second Interval & Code Case N-546 ML20247M1901998-05-20020 May 1998 Safety Evaluation Related to Browns Ferry,Unit 3 Nuclear Power Plant Individual Plant Exam ML20217E5431997-09-22022 September 1997 Safety Evaluation Re Flaw Evaluation of Core Spray Internal Piping for Plant,Unit 3 ML20217C2191997-09-11011 September 1997 Safety Evaluation Supporting Amend 249 to License DPR-52 ML20149L2301997-07-28028 July 1997 Safety Evaluation Accepting License Relief Request from Certain ASME Code Requirements Delineated in Change to 10CFR50.55a for Plant,Units 1,2 & 3 ML20148U3291997-07-0808 July 1997 Safety Evaluation Denying Relief Request 3-ISI-1.Based on Review,Nrc Concludes That Licensee Does Not Have Sufficient Basis for Relief from Code Required Successive Exam ML20136D7831997-03-10010 March 1997 Safety Evaluation Authorizing Use of ASME Code Case N-416-1 ML20132B2701996-12-11011 December 1996 Safety Evaluation Accepting Licensee Emi/Rfi site-survey Consistent W/Industry Stds & Practice ML20058K2411993-12-0707 December 1993 Supplemental Safety Evaluation Supporting Structural Steel Thermal Growth Design Critera at Plant ML20058J1091993-12-0303 December 1993 Safety Evaluation Accepting Licensee 921228 Submittal of Suppl Response to GL 88-01, NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping ML20059C2451993-10-22022 October 1993 Safety Evaluation Accepting IST Program Requests for Relief ML20246M3891989-05-0909 May 1989 Safety Evaluation Supporting Util 890317 Proposed long-term Solution to Correct Rust Problems in Containment Spray Headers ML20245H3371989-05-0101 May 1989 Safety Evaluation Re Generic Ltr 83-28,Item 2.1 (Part 2) About Vendor Interface Program ML20245D5711989-04-19019 April 1989 Safety Evaluation Re Rust in Lower Containment Spray Header Due to Leaking Isolation Valves ML20247K5561988-09-23023 September 1988 Safety Evaluation Supporting Amends 155,151 & 126 to Licenses DPR-33,DPR-52 & DPR-68,respectively ML20154H2721988-04-19019 April 1988 Safety Evaluation Describing Relationship of Util Nuclear Performance Plan,Vol 3,Part Ii,Section 2.6, QA to TVA-TR75-1A, QA Program Description for Design,Const... W/Proper Implementation,Qa Program Description Acceptable ML20148K9051988-03-23023 March 1988 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1 & 3.2.2, Post-Maint Testing of Reactor Trip Sys & All Other Safety-Related Components ML20246M2341988-02-29029 February 1988 Safety Evaluation Supporting Amends 145,141 & 116 to Licenses DPR-33,DPR-52 & DPR-68,respectively ML20214H9091987-05-13013 May 1987 Safety Evaluation Supporting Amends 133,129 & 104 to Licenses DPR-33,DPR-52 & DPR-68,respectively ML20206L9821986-06-20020 June 1986 Safety Evaluation Supporting Requests for Establishment of Common Date to Implement Inservice Insp Program Requirements.Start Date of 880301 Accepted for Second 10-yr Insp Interval ML20127D6641985-06-17017 June 1985 Safety Evaluation Re Licensee Response to Item 1.1 of Generic Ltr 83-28 on post-trip Review Program & Procedures. Post-trip Review Program & Procedures Acceptable ML18029A5381985-05-0606 May 1985 Safety Evaluation Re Mark I Containment long-term Program Pool Dynamic Loads Review for Plant.Pool Dynamic Loads Meet Acceptance Criteria of NUREG-0661 or Alternative Criteria Acceptable ML18029A5401985-05-0606 May 1985 Safety Evaluation Re Mark I Containment long-term Program Structural Review for Plant.Mods Follow NUREG-0661 Guidelines & Acceptable.Analyses Verified & Approved ML20148T7811980-12-0101 December 1980 Generic Safety Evaluation Re BWR Scram Discharge Sys ML20062A4221978-09-29029 September 1978 Safety Evaluation Rept Supporting Amend 43 to Lic DPR-33 Covering Pump Seizure Accident,Inadvertent Cold Loop Startup Transient & Monitoring of Safety Margins for 1 Loop Oper ML20247C1721976-02-23023 February 1976 Safety Evaluation Supporting Operation of Plant After Restoration & Mods of Facilities Following 750322 Fire 1999-09-23
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML18039A9041999-10-15015 October 1999 LER 99-010-00:on 990917,automatic Reactor Scram on Turbine Stop Valve Closure Occurred.Caused by High Water Level in Main Steam Moisture Separator 2C2.Unit 2C2 Reservoir Level Transmitter & Relays Were Replaced & Tested Satisfactorily ML18039A8981999-10-14014 October 1999 LER 99-009-00:on 990915,manual Reactor Scram Was Noted Due to EHC Leak.Caused by Failure of Stainles Steel Tubing Connection.Removed Damaged Tubing & Connection Plug ML18039A8951999-10-0808 October 1999 LER 99-008-00:on 990905,HPCI Was Inoperable Due to Failed Flow Controller.Caused by Premature Failure of Capacitor 2C3.Replaced Controller & HPCI Sys Was Run IAW Sys Operating Instructions ML18039A8751999-09-30030 September 1999 LER 99-005-00:on 990901,SR for Standby Liquid Control Sampling Was Not Met.Caused by Deficient Procedure for Chemical Addition to Standby Liquid Control.Revised Procedure.With 990930 Ltr ML20217F9671999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML20212E6341999-09-23023 September 1999 Suppl to SE Resolving Error in Original 990802 Se,Clarifying Fact That Licensee Has Not Committed to Retain Those Specific Compensatory Measures That Were Applied to one-time Extension ML20212D3831999-09-20020 September 1999 Safety Evaluation Supporting Proposed Rev to Withdrawal Schedule for First & Third Surveillance Capsules for BFN-3 RPV ML20212B8561999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Browns Ferry Nuclear Plant.With ML18039A8821999-08-31031 August 1999 Increased MSIV Leakage Tech Spec Change Submittal - Seismic Evaluation Rept. ML18039A8391999-08-0606 August 1999 BFN Unit 2 Cycle 10 ASME Section XI NIS-1 & NIS-2 Data Repts. ML20210N1221999-08-0202 August 1999 Safety Evaluation Accepting Licensee Request for Relief from ASME B&PV Code,Section XI Requirements.Request 3-ISI-7, Pertains to Second 10-year Interval ISI for Plant,Unit 3 ML20210R0931999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML18039A8201999-07-26026 July 1999 LER 99-004-00:on 990625,facility Core Spray Divisions I & II Inoperable at Same Time Due to Personnel Error.Electrical Supply Breaker to Core Spray Division II Pump 3B Returned to Normal Racked in Position ML18039A8171999-07-20020 July 1999 LER 99-007-00:on 990623,discovered That SR for Monitoring of Primary Containment Oxygen Concentration Had Not Been Met. Caused by Failure of Operators to Adequately Communicate. Required Surveillances Were Performed.With 990720 Ltr ML18039A8161999-07-19019 July 1999 LER 99-006-00:on 990618,noted That Main Steam SRV Exceeded TS Setpoint Tolerance.Caused by Pilot Vlve disc-seat Bonding.Util Replaced All 13 SRV Pilot Cartridges with Cartridges Certified to Be Witin +/-1%.With 990719 Ltr ML20209J0771999-07-16016 July 1999 Safety Evaluation Concluding That Licensee Provided Adequate Information to Resolve ampacity-related Points of Concern Raised in GL 92-08 for BFN & That No Outstanding Issues Re GL 92-08 Ampacity Issues for Browns Ferry NPP Exist ML18039A8121999-07-12012 July 1999 LER 99-005-00:on 990617,ESF Actuation & HPCI Declared Inoperable.Caused by Personnel Error.Reset HPCI & Returned Sys to Operable Status with 25 Minutes.With 990712 Ltr ML20209H4381999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML18039A8101999-06-28028 June 1999 LER 99-004-00:on 990530,safety Features Sys Actuations Occurred Due to RPS Trip.Caused by Failure of MG Set AC Drive Motor Starter Contractor Coil.Licensee Placed 2B RPS Bus on Alternate Feed & Half Scram Was Reset ML20196F8811999-06-23023 June 1999 Safety Evaluation Accepting GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power Operated Gate Valves ML18039A8071999-06-14014 June 1999 LER 99-003-00:on 990515,automatic Reactor Scram Due to Turbine Trip Was Noted.Caused by Failure of Mechanical Trip Cylinder to Latch When Hydraulically Reset.Operations Crew Stabilized Reactor Following Scram ML18039A8021999-06-14014 June 1999 LER 99-002-00:on 990501,SRs for Single CR Withdrawal During Cold SD Were Not Adequately Implemented.Caused by Procedural Inadequacy.Revised Applicable Plant Surveillances.With 990614 Ltr ML18039A8011999-06-14014 June 1999 LER 99-001-00:on 990515,automatic Reactor Scram Occurred Due to Tt.Caused by Failure of Mechanical Trip Cylinder to Latch When Hydraulically Reset.Operations Crew Stabilized Reactor Following Scram.With 990614 Ltr ML20196B8051999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML18039A7791999-05-0606 May 1999 LER 99-003-00:on 990408,declared Plant HPCI Sys Inoperable Due to Loose Wire.Caused by Failure to Properly Tighten Screw at Some Time in Past.Loose Wire Was Tightened ML18039A7761999-04-30030 April 1999 Revised Surveillance Specimen Program Evaluation for TVA Browns Ferry Unit 3. ML20206R0731999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Bfnp.With ML18039A7561999-04-23023 April 1999 Bfnp Risk-Informed Inservice Insp (RI-ISI) Program Submittal. ML18039A7671999-04-0808 April 1999 Rev 0 to TVA-COLR-BF2C11, Browns Ferry Nuclear Plant Unit 2 Cycle 11 Colr. ML18039A7461999-04-0707 April 1999 LER 99-001-00:on 990308,determined That Two Trains of Standby Gas Treatment (SGT) Were Inoperable.Caused by Trip C SGT Blower Motor Breaker.Initiated Shutdown of Plant,Reset C SGT Blower Motor Breaker & Declared Train Operable ML20205N8341999-04-0101 April 1999 Part 21 Rept Re Automatic Switch Co Nuclear Grade Series X206380 & X206832 Solenoid Valves Ordered Without Lubricants That Were Shipped with Std Lubrication to PECO & Tva.Affected Plants Were Notified ML20205F9341999-04-0101 April 1999 Safety Evaluation Authorizing Licensee 990108 Relief Request PV-38,from Requirements of ASME BPV Code Section XI IST Testing,Valve Program for Plant,Units 1,2 & 3 ML20205T5441999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Bfnp.With ML20205S0601999-03-31031 March 1999 Rept on Status of Public Petitions Under 10CFR2.206 with Status Change from Previous Update,990331 ML20205S0661999-03-31031 March 1999 Rept on Status of Public Petitions Under 10CFR2.206 with No Status Change from Previous Update,990331, Atlas Corp ML18039A7361999-03-11011 March 1999 Rev 4 to TVA-COLR-BF2C10, Bfnp,Unit 2,Cycle 10 Colr. ML20204C7891999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Browns Ferry Nuclear Plant,Units 1,2 & 3.With ML18039A6951999-02-19019 February 1999 LER 99-002-00:on 990122,LCO Was Not Entered During Calibration Testing of 3D 480 Volt Rmov Board.Caused by Personnel Error.Tva Has Briefed Operations Personnel to Preclude Recurrence of Event.With 990219 Ltr ML18039A6871999-02-12012 February 1999 LER 99-001-00:on 990114,Unit 3 HPCI Was Noted Inoperable. Caused by Oil Leak on Stop Valve.Corrective Maint Was Performed to Repair Oil Leak.With 990212 Ltr ML18039A6931999-02-0303 February 1999 Rev 3 to TVA-COLR-BF2C10, Bfnp Unit 2 Cycle 10 Colr. ML18039A6941999-02-0303 February 1999 Rev 1 to TVA-COLR-BF3C9, Bfnp Unit 3 Cycle 9 Colr. ML18039A6671998-12-31031 December 1998 LER 98-004-00:on 981202,SR Intent Was Not Adequately Implemented.Caused by Procedural Inadequacy.Revised Procedures to Provide Proper SR Implementation.With 981231 Ltr ML18039A6661998-12-31031 December 1998 Ro:On 981215,HRPCRM 2-RM-90-273C Was Declared Inoperable. Caused by Downscale Indication.Containment RM Will Be Utilized as Planned Alternate Method of Monitoring Until Hrpcrm 2-RM-90-273C Can Be Returned to Operable Status ML20199K8951998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Browns Ferry Nuclear Plant.With ML20199F2721998-12-31031 December 1998 ISI Summary Rept (NIS-1), for BFN Unit 3,Cycle 8 Operation ML18039A6471998-12-15015 December 1998 LER 98-007-00:on 981116,unplanned ESF Following Loss of 4kV Unit Board 3B Occurred.Caused by Temporary Energization of Lockout Relay on 4kV Unit Board 3B When Resistor on Relay Monitoring Lamp Circuit Shorted.Replaced Resistor ML18039A6371998-12-0707 December 1998 LER 98-006-00:on 981116,MSSR Valves Exceeded TS Setpoint Tolerance.Caused by Pilot Valve Disc/Seat Bonding. Installed SRV Pressure Switches During Unit 3,cycle 8 Outage.With 981207 Ltr ML20199F2791998-12-0303 December 1998 Bfnp Unit 3 Cycle 8 ASME Section XI NIS-2 Data Rept ML20198D9621998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Bfn,Units 1,2 & 3. with ML18039A6071998-11-12012 November 1998 LER 98-005-00:on 981014,mode Changes Not Allowed by TS 3.0.4 Were Made During Reactor Startup.Caused by TS LCO 3.0.4 Not Being Properly Applied.Training Info Memo Re Proper Application for TS LCO 3.0.4 Was Prepared.With 981112 Ltr 1999-09-30
[Table view] |
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- l ENCLOSURE SAFETYEVALVATIONFORGENERICLETTER83-28, ITEMS 3.1.1, 3.1.2, 3.2.1, AND 3.2.2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 1,<2, AND 3 OOCKET NOS. 50-259, 50-260, AND 50-296
- 1. Introduction Generic Letter (GL) 83-28 was issued by the staff on July 8,1983. It described intermediate-term actions to be taken by the licensees and applicants to address the generic issues raised by the two Anticipated Transients Without Scram (ATWS) Events that occurred at Unit 1 of Salem Nuclear Power Plant. Items 3.1.1, 3.1.2, 3.2.1, and 3.2.2, of the GL required licensees and applicants to submit results of their review of ,
test and maintenance procedures and vendor and engineering reconnendations I to assure that appropriate test guidance is included in the test and I maintenance procedures or the Technical Specifications, where required.
By letters dated November 7,1983, and March 15, 1984, Tennessee Valley i Authority (TVA), the licensee for Browns Ferry Plant, provided infortnation I regarding their compliance to Items 3.1.1, 3.1.2, 3.2.1, 3.2.2, and 4.5.1 i of GL 83-28. We evaluated the licensee's responses to the l above items against the NRC positions described in the GL for completeness and adequacy. We found the licensee's response to be accept 6ble for Item 4.5.1, but the responses to Items 3.1.1, 3 tl.2, 3.2.1, and 3.2.2 were i incomplete, thus requiring additional inf6rmation to determine acceptability.
. A Safety Evaluation for the above items ad a Request for Additional Information were transmitted to the licensee by NRC letter dated April 14, 1986.
The licensee submitted supplemental responses for Items 3.1.1, 3.1. 2, 3.2.1, and 3.2.2, in letters dated August 28, 1986 and December 4, 1986, respectively. Our evaluation of these responses follows.
- 2. Evalcation By letter dated November 7, 1983, Tennessee Valley Authority responded to the requested information concerning the status, plans, and schedules for 88040100$7 880323 PDR ADOCK 05000259 p PDR
conformance with the NRC positions contained in GL 83-28. Our evaluations revealed that Items 3.1.1, 3.1.2, 3.2.1, and 3.2.2 were incomplete, thus requiring more infomation. Subsequent licensee letters dated August 28, !
1986 and December 4, 1986, provided sufficient additional information to i allow an evaluation to be made against the NRC positions as stated in the I
GL. l l
Delineated below are the results of NRC's evaluation of items 3.1.1, 1 3.1.2, 3.2.1, and 3.2.2 of GL 83-28. j
- a. Item 3.1.1 Review of Test and Maintenance Procedures and Technical !
Specifications (Reactor Trip System Components) l Item 3.1.1 requires licensees and applicants to submit the results of their review of test procedures, maintenance procedures, and Technical Specifications to ensure that post maintenance operability 1 testing of safety-related components in the reactor _ trip system is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being ,
returned to service. )
The licensee stated in their initial response dated November 7, 1983, that existing procedures, standard practices, and QA programs requires that all critical systems, structures, or components (CSSC) be tested after maintenance before being returned to service. The licensee stated that the Technical Specifications for Browns Ferry lists the limiting conditions and surveillance requirements for the reactor protection system.
The licensee's supplemental response dated August 28, 1986, stated that post maintenance testing is accomplished for safety-related equipment by testing specified by a standing maintenance procedure.
or by performance of appropriate surveillance testing on the affected system, or by additional testing specified in a Maintenance Request (MR). One or more of the above tests rgay be used depending upon the circumstances. -
For the Reactor Protection System, the licensee indicated that maintenance procedures and surveillance instructions are currently being reviewed to ensure that post-maintenance testing adequately i demonstrates that. the equipment is capable of performing its intended i safety function. The licensee indicated that this effort will be completed before restart of any unit.
Although, the licensee has not completed the reiiew of all maintenance ,
and test procedures associated with the reactor protection system, l their supplemental response revealed that they intend to complete the review prior to restart. Based on the above, the licensee's response is acceptable.
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Enclosure 3
- b. Item 3.1.2, Check of Vendor and Engineering Recorrvnendations for Testing and Maintenance.(Reactor Trip System Components)
Item 3.1.2 requires licensees and applicants to submit the results of their check of vendor and engineering reconnendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required.
The licensee stated in their initial response dated November 7,1983, that they had programs in place for the review of technical infonnation received from General Electric Company (GE) and operating experience reports, which include vendor and engineering recortunendations. The licensee stated that all GE Service Instruction Letters (Sils) had been reviewed and that Browns Ferry had initiated the recommended actions.
The licensee's supplemental response of August 28, 1986, indicated that Browns Ferry is currently implementing a program establishing controlled vendor information for the Reactor Protection System and this vendor information will be reviewed to ensure that appropriate test guidance is reflected in plant procedures. The response further indicated that this effort will be completed before restart of any unit. Based on the licensee's continuing implementation of the above programs, the resconse is acceptable.
- c. Item 3.2.1, Review of Test and Maintenance Procedures ( All Other Safety-RelatedComponents)
Item 3.2.1 requires licensees and applicants to submit a report documenting the extending of test and maintenance procedures and Technical 1 Specifications review to ensure that post-taaintenance operability testing l of all safety-related equipment is required to be conducted and that l the testing demonstrates that the equipnient is capable of perfonning )
its safety functions before being returned to service. ]
The licensee stated in their initial retponse dated November 7, 1983, that existing Quality Assurance Manual procedures required maintenance instructions to contain requirements for post-maintenance operational testing of CSSC after maintenance before being returned to service.
The licensee stated in supplemental response dated December 4,1986, i that post-maintenance testing is accomplished for safety-related equipment by testing specified by standard maintenance procedures, ,
by appropriate . surveillance procedures, or by additional testing l specified in maintenance requests. The licensee further indicated that surveillance and maintenance procedures are currently being reviewed and upgtaded under the procedure uparade program as des-cribed in the Browns Ferry Nuclear Performance Plan. The upgrading i effort will be in two phases (pre-startup and post-startup) and i includes the review of test guidance from vendors, engineering, General Electric Service Information Letters (SILs), INPO informa-tion, Licensee Event Reports (LERs), and other technical information 4
i l
l Enclosure 4
received from vendors to ensure that appropriate guidance is included in plant procedures. The licensee states that this upgrade program will ensure that post-maintenance testing adequately demonstrates that the equipment is capable of performing its safety function.
Based on the above, the licensee's response to this item is acceptable.
- d. Item 3.2.2 Check of Vendor and Engineering Recomendations for Testing and Maintenance procedures (All Other Safety-Related Components)
Items 3.2.2 requires licensees and applicants to submit the results of their check of vendor and engineering recomendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required.
The licensee's initial response of November 7, 1985 stated that standard practices had been established to control and review vendor manuals, vendor and engineering recommendations, and other industry information to ensure that vendor's guidance is appropriately incorporated into plant procedures, l l
The licensee's supplemental response dated December 4,1986, stated that maintenance and surveillance procedures were being reviewed and upgraded to ensure that vendor and engineering recommendations are reviewed and incorporated into test and maintenance procedures, where applicable. As discussed in items 3.1.2 and 3.2.1 above, the licensee's review is a two phase program where certain reviews will be performed prior to Browns Ferry startup and others af ter plant startup. Although, the licensee has not completed the review of vendor and engineering recocinendations for applicability to test and maintenance procedures, the response revealed that a program has been started to accomplish the review. Based on the above, the licensee's response to item 3.2.2 is acceptable. , .
- 3. Conclusions Based on the above statements, the staf f finds that the licensee's l responses to be acceptable and meeting the intent of GL 83-28. The l licensee stated that procedures have been established which require post-maintenance testing of the safety-related components, which includes the Reactor Protection System components. The licensee ilso indicated that they have a continuous program in operation to ensure that vendor and engineering information and recommendations are reviewed and incorporated into test and maintenance procedures. Although, the licensee's reviews are not all complete, the staff considers the licensee's responses to items 3.1.1, 3.1.2, 3.2.1, and 3.2.2 of GL 83-28 to be acceptable.
t Enclosure 5
- 4. Reference Documents
- a. NRC letter, D. G. Eisenhut to all Licensees of Operating Reactors, GL 83-28, Required Actions Based on Generic Implications of Salem i ATWS Events, dated July 8, 1983.
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- b. NRC memorandum, F. J. Miraglia to Directors, dated February 11, 1985, l 0
Licensee Action Review Schedule for GL 83-28, Enclosure 3. Review i Criteria for Regional Licensing Review of GL 83-28. ;
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- c. NRC Inspection Report Nos. 50-259/84-50, 50-260/84-50, and ;
50 296/84-50 dated January 16, 1985. l 1
- d. Tennessee Valley Authority letter dated November 7,1983, Response to l GL 83-28, Browns Ferry Nuclear Plant, j
- e. NRC letter dated April 14, 1986 to Tennessee Vality Authority . l transmittal of Safety Evaluation and Request for Additional l Information, Browns Ferry Nuclear Plant. l 1
- f. Tennessee Valley Authority letter dated August 28, 1986, Response to Items 3.1.1 and 3.1.2 of NRC Request for Additional information.
- g. Tennessee Valley Authority letter dated December 4, 1986, Response to Items 3.2.1 and 3.2.2 of NRC Request for Additional Information.
krincipal Contributor: T. E. Conlon Dated: J l
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