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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217C1311999-10-0808 October 1999 Safety Evaluation Supporting Amend 153 to License DPR-3 ML20211J3361999-08-27027 August 1999 Safety Evaluation Supporting Amend 152 to License DPR-3 ML20207F9491999-03-0505 March 1999 Safety Evaluation Supporting Amend 151 to License DPR-3 ML20202H5871999-02-0303 February 1999 Safety Evaluation Supporting Amend 150 to License DPR-3 ML20249A7901998-06-17017 June 1998 Safety Evaluation Supporting Amend 149 to License DPR-3 ML20237F1671993-02-19019 February 1993 Safety Evaluation Supporting Amend 147 to License DPR-3 ML20058F2201990-11-0202 November 1990 Safety Evaluation Accepting Util Response to Generic Ltr 83-28 Re post-trip Review - Data & Info Capability ML20058C4061990-10-22022 October 1990 Safety Evaluation Supporting Amend 137 to License DPR-3 ML20059G2411990-09-0606 September 1990 Safety Evaluation Supporting Amend 135 to License DPR-3 ML20058L6651990-08-0202 August 1990 Safety Evaluation Supporting Amend 134 to License DPR-3 ML20058L0321990-08-0202 August 1990 Safety Evaluation Supporting Amend 133 to License DPR-3 ML20055C8601990-06-18018 June 1990 Safety Evaluation Supporting Amend 132 to License DPR-3 ML20248H7391989-10-0303 October 1989 Safety Evaluation Not Accepting Procedure Generating Program for Plant.Program Should Be Revised to Reflect Items Described in Section 2 of Rept.Revision Need Not Be Submitted to NRC ML20247F1431989-09-0707 September 1989 Safety Evaluation Supporting Amend 124 to License DPR-3 ML20247E6831989-08-31031 August 1989 Safety Evaluation Supporting Amend 123 to License DPR-3 ML20246F2771989-07-11011 July 1989 Safety Evaluation Supporting Mods to ECCS Evaluation Model, Including Changes to FLECHT-based Reflood Heat Transfer Correlation,Steam Cooling Model & post-critical Heat Flux Heat Transfer Model ML20195D6701988-11-0101 November 1988 Safety Evaluation Supporting Amend 120 to License DPR-3 ML20205G1961988-10-25025 October 1988 Safety Evaluation Supporting Amend 119 to License DPR-3 ML20204G4871988-10-17017 October 1988 Safety Evaluation Supporting Amend 118 to License DPR-3 ML20205C4061988-10-14014 October 1988 Safety Evaluation Supporting Amend 117 to License DPR-3 ML20207L7051988-10-12012 October 1988 Safety Evaluation Supporting Amend 116 to License DPR-3 ML20207E8151988-08-0505 August 1988 Safety Evaluation Supporting Amend 115 to License DPR-3 ML20151M4911988-07-29029 July 1988 Safety Evaluation Supporting Amend 114 to License DPR-3 ML20151K3801988-07-25025 July 1988 Safety Evaluation Supporting Amend 113 to License DPR-3 ML20151K8571988-07-19019 July 1988 Safety Evaluation Supporting Amend 112 to License DPR-3 ML20153A8661988-06-29029 June 1988 Safety Evaluation Accepting Util Proposed Reflood Steam Cooling Model ML20196K2741988-06-28028 June 1988 Safety Evaluation Supporting Amend 111 to License DPR-3 ML20195K1501988-06-17017 June 1988 Safety Evaluation Supporting Amend 110 to License DPR-3 ML20195C5851988-06-13013 June 1988 Safety Evaluation Supporting Amend 109 to License DPR-3 ML20155K5141988-06-0909 June 1988 Safety Evaluation Supporting Amend 108 to License DPR-3 ML20154J7661988-05-18018 May 1988 Safety Evaluation Supporting Amend 107 to License DPR-3 ML20216J4081987-06-26026 June 1987 Safety Evaluation Supporting Amend 106 to License DPR-3 ML20216C1111987-06-18018 June 1987 Safety Evaluation Granting Three of Seven Requests Submitted by Util for Relief from Inservice Insp & Testing Requirements.Four Requests Withdrawn,Per 870122,0410 & 0507 Ltrs ML20215C5881987-06-0404 June 1987 Safety Evaluation Supporting Util 860505,870402,& 0506 Submittals Re Seismic Reevaluation of Plant.Concludes That Foundation Soils Under Reactor & Under Vapor Container Have Adequate Strength to Support Seismic Load from Earthquake ML20213G9161987-05-13013 May 1987 Safety Evaluation Supporting Amend 105 to License DPR-3 NUREG-0825, Safety Evaluation Supporting Util 840709,1231 & 851024 Repts Re Evaluation of Plant for Wind & Tornado Events as Requested in Integrated Plant Safety Assessment Rept, Sections 4.5 & 4.8.Risk from Wind/Tornado Events Assessed1987-05-13013 May 1987 Safety Evaluation Supporting Util 840709,1231 & 851024 Repts Re Evaluation of Plant for Wind & Tornado Events as Requested in Integrated Plant Safety Assessment Rept, Sections 4.5 & 4.8.Risk from Wind/Tornado Events Assessed ML20213D9671987-05-0707 May 1987 Safety Evaluation Supporting Amend 104 to License DPR-3 ML20207S6231987-03-10010 March 1987 Safety Evaluation Supporting Util 860122,0812,1028 & 870204 Submittals Re Fracture Toughness Requirements for Protection Against PTS Events ML20211N5881987-02-19019 February 1987 Safety Evaluation Re First Level Undervoltage Protection Testing.Testing Unnecessary ML20211L3951987-02-17017 February 1987 Safety Evaluation Supporting Amend 103 to License DPR-3 Re Max Nominal Enrichment of Fuel ML20207N8811987-01-0707 January 1987 Safety Evaluation Supporting Amend 102 to License DPR-3 ML20207N4261987-01-0606 January 1987 Safety Evaluation Supporting Amend 101 to License DPR-3 ML20207J9451986-12-30030 December 1986 SER Accepting Util 831105 & 850709 Responses to Generic Ltr 83-28,Item 2.1 (Part 2), Vendor Interface Program - Reactor Trip Sys Components ML20215E1201986-12-0909 December 1986 Safety Evaluation Supporting Util 830419 & 0830,840119, 851022 & 860930 Responses Re Conformance to Reg Guide 1.97. Plant Design Acceptable W/Exception of Neutron Flux Variable ML20214X3391986-12-0101 December 1986 Safety Evaluation Supporting Amend 100 to License DPR-3 ML20214J8521986-11-18018 November 1986 Sser Accepting SPDS Contingent Upon Resolution of Concerns Re Maint & Improvement of Placement & Visual Access of Containment Isolation Panel & Minor Human Factors Engineering Concerns ML20215E6471986-10-0202 October 1986 Safety Evaluation Supporting Util Requests for Exemption from Specific Requirements in App R to 10CFR50.Existing Fire Protection Provides Level of Protection Equivalent to Technical Requirements of App R ML20210S1791986-09-23023 September 1986 Safety Evaluation Supporting Amend 99 to License DPR-3 ML20212Q1151986-08-27027 August 1986 Safety Evaluation Supporting Util 830412 Proposal to Provide Integrated Safe Shutdown Sys Which Could Be Used for Safe Shutdown in Event of Fire at Facility ML20212N0161986-08-20020 August 1986 Safety Evaluation Supporting Amend 98 to License DPR-3 1999-08-27
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217C1311999-10-0808 October 1999 Safety Evaluation Supporting Amend 153 to License DPR-3 ML20211J5111999-08-31031 August 1999 Rev 29 to Yankee Decommissioning QA Program ML20211J3361999-08-27027 August 1999 Safety Evaluation Supporting Amend 152 to License DPR-3 ML20209D5391999-06-22022 June 1999 Rev 29 to Yaec Decommissioning QA Program ML20207F9491999-03-0505 March 1999 Safety Evaluation Supporting Amend 151 to License DPR-3 ML20202H5871999-02-0303 February 1999 Safety Evaluation Supporting Amend 150 to License DPR-3 ML20154P9691998-10-16016 October 1998 Rev 28 to Yankee Atomic Electric Co Decommissioning QA Program ML20249A7901998-06-17017 June 1998 Safety Evaluation Supporting Amend 149 to License DPR-3 ML20216C4581998-02-27027 February 1998 Response to NRC Demand for Info (NRC OI Rept 1-95-050) ML20203L1931998-02-25025 February 1998 Duke Energy Corp,Duke Engineering & Svcs,Inc,Yankee Atomic Small Break LOCA Technical Review Rept ML20203L2451998-02-23023 February 1998 Assessment Rept of Engineering & Technical Work Process Utilized at De&S Bolton Ofc ML20203L1621998-02-18018 February 1998 Rept of Root Cause Assessment Review ML20203L2691998-02-16016 February 1998 Duke Engineering & Svcs Assessment Process Review Rept ML20199B4601998-01-20020 January 1998 Special Rept:On 980105,meteorological Monnitoring Instrumentation for Air Temp Delta T Inoperable for More than 7 Days.Caused by Breakdown in Wiring Between Junction Box at 199 Foot Level.Wiring Replaced ML20203J3001997-12-31031 December 1997 Ynps 1997 Annual Rept ML20217N0981997-08-21021 August 1997 LER 97-S02-00:on 970725,discovered Uncontrolled Safeguards Documents.Caused by Personnel Error.Matls Retrieved & Stored in Safeguards Repositories ML20210H0991997-08-0707 August 1997 LER 97-S01-00:on 970709,potential Compromise of Safeguards Info Occurred.Caused by Human error.Stand-alone Personal Computer & Printer Not Connected to Network,Have Been Located within Text Graphics Svc Dept ML20149K7781997-07-24024 July 1997 Special Rept:On 970520 & 0714,air Temp Delta T Channel Indicated Temp Difference Between Top & Bottom of Meteorological Tower.Caused by Reversed Input Wiring to Channel.Restored Air Temp Delta T Channel Operability ML20141E4671997-05-30030 May 1997 Rev 28 to Operational QA Program ML20135C8461996-12-31031 December 1996 Yankee Nuclear Power Station 1996 Annual Rept ML20132G6771996-12-20020 December 1996 Rev 27 to YOQAP-I-A, Operational QA Program ML20058N4771993-12-20020 December 1993 Rev 0.0 to Yankee Nuclear Power Station Decommissioning Plan ML20059K8491993-12-15015 December 1993 Clarifications to Pages 2,41,43 & 44 of 44 in Section I, Organization of YOQAP-I-A,Rev 24, Operational QA Program ML20059C5011993-10-29029 October 1993 Special Rept:On 931019,meteorological Instrumentation Channel for Delta T Declared Inoperable.Caused by Ceased Aspirator Motor Located at Top of Tower.Motor Replaced ML20056H1741993-06-10010 June 1993 Preliminary Assessment of Potential Human Exposures to Routine Tritium Emissions from Yankee Atomic Electric Co Nuclear Power Facility Located Near Rowe,Ma ML20237F1671993-02-19019 February 1993 Safety Evaluation Supporting Amend 147 to License DPR-3 ML20125C7161992-12-0707 December 1992 Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl ML20198D2541992-05-13013 May 1992 Yankee Nuclear Power Station Certified Fuel Handler Recertification Program ML20198D2481992-05-13013 May 1992 Yankee Nuclear Power Station Certified Fuel Handler Initial Certification Program ML20062H1981990-11-30030 November 1990 Plant Specific Fast Neutron Exposure Evaluations for First 20 Operating Fuel Cycles of Yankee Rowe Reactor ML20058H2841990-11-0303 November 1990 Special Rept:On 901101,control Rod 24 Found Disconnected from Drive Shaft.Drive Shaft Latching Will Be Initiated ML20058F2201990-11-0202 November 1990 Safety Evaluation Accepting Util Response to Generic Ltr 83-28 Re post-trip Review - Data & Info Capability ML20062E8331990-10-31031 October 1990 Monthly Operating Rept for Oct 1990 for Yankee Atomic Power Station ML20058G1471990-10-31031 October 1990 Vol 2 to Star Methodology Application for PWRs Control Rod Ejection Main Steam Line Break ML20058C4061990-10-22022 October 1990 Safety Evaluation Supporting Amend 137 to License DPR-3 ML20062B6751990-09-30030 September 1990 Monthly Operating Rept for Yankee Atomic Power Station for Sept 1990 ML20059G2411990-09-0606 September 1990 Safety Evaluation Supporting Amend 135 to License DPR-3 ML20059E3071990-08-31031 August 1990 Safety Assessment of Yaec 1735, Reactor Pressure Vessel Evaluation Rept for Yankee Nuclear Power Station. Detailed Plan of Action W/Listed Elements Requested within 60 Days After Restart to Demonstrate Ability to Operate Longer ML20059E8001990-08-31031 August 1990 Monthly Operating Rept for Aug 1990 for Yankee Atomic Power Station ML20058P7841990-08-14014 August 1990 Part 21 Rept Re Misapplication of Fluorolube FS-5 Oil in Main Steam Line Pressure Gauges.All Four Indicators Replaced W/Spare Gauges Which Utilize High Temp Silicone Oil ML20058N6581990-08-13013 August 1990 Special Rept Re Diesel Fire Pump & Tank Inoperable for Greater than Seven Days for Draining,Cleaning & Insp.During Period Redundant Pumping Capacity Available Via Two Remaining Electric Driven Fire Pumps ML20058L0321990-08-0202 August 1990 Safety Evaluation Supporting Amend 133 to License DPR-3 ML20058L6651990-08-0202 August 1990 Safety Evaluation Supporting Amend 134 to License DPR-3 ML20056A1961990-08-0101 August 1990 Special Rept:Two Fire Pumps Inoperable at Same Time.Caused by Necessity to Accomplish Surveillance to Verify Capability to Start Pump on Emergency Diesel Generator 3 & Planned 18-month Insp of Diesel Per Tech Specs ML20055G6801990-07-31031 July 1990 Yankee Plant Small Break LOCA Analysis ML20055G7011990-07-31031 July 1990 Yankee Nuclear Power Station Core 21 Performance Analysis ML20055E1591990-07-31031 July 1990 Reactor Pressure Vessel Evaluation Rept ML20055J3221990-07-25025 July 1990 Decommissioning Funding Assurance Rept & Certification ML20055G7051990-07-19019 July 1990 Rev 0 to Yankee Cycle 21 Core Operating Limits Rept ML20055F6751990-06-30030 June 1990 Monthly Operating Rept for June 1990 for Yankee Atomic Power Station 1999-08-31
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4 .
t SAFETY EVALUATION BY THE OFFICE OF THE NUCLEAR REACTOR REGULATION RELATED TO
, GENERIC LETTER 83-28, ITEMS 3.1.1, 3.1.2, 3.2.1, 3.2.2, 4.1 and 4.5.1 YANKEE ATOMIC ELECTRIC COMPANY :
I YANKEE NUCLEAR POWER PLANT i DOCKET NO.50-029 i
, 1.0 Introduction On February 25, 1983, both of the scram circuit breakers at Unit 1 of the
. Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal
] . from the reactor protection system. This incident occurred during the plant
- .t startup, and the reactor was tripped manually by the operator about 30 seconds g after the initiation of the automatic trip signal. The failure of the circuit
-? breakers has been determined to be related to the sticking of the undervoltage j trip attachment. Prior to this incident, on February 22, 1983, at Unit 1 of
, the Salem Nuclear Power Plant, an automatic trip signal was generated based 4 on steam generator low-low level during plant startup. In this case the i reactor was tripped manually by the operator almost coincidentally with the automatic trip.
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- Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO), directed the staff to investigate and report on the
- generic implications of these occurrences at Unit 1 of the Salem Nuclear Plant.
l The results of the staff's inquiry into the gendric implications of the Salem unit incidents are reported in NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant." As.a result of this investigation, the Director, Division of Licensing, Office of Nuclear Reactor Regulation re-4 quested (by Generic Letter 83-28 dated July 8,1983) all licensees of operat-
- ing reactors, applicants for an operating license, and holders of construction permits to respond to certain generic concerns. These.are categorized into
! four areas; (1) Post-Trip Review, (2) Equipment Classification and Vendor
.. Interface, (3) Post-Maintenance Testing, and (4) Reactor Trip Systera (RTS) i Reliability Improvements. Within each of these areas various specific actions j , were delineated.
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j This safety evaluation (SE) addressed the following actions of Generic Letter j 83-28:
3.1.1 and 3.1.2, Post-Maintenance Testing (Reactor Trip System Components) i --
3.2.1 and 3.2.2, Post-Maintenance Testing (All Other Safety-Related Com- '
ponents)
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4.1, Reactor Trip System Reliability (Vendor-Related Modifications) '
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. Safety Evaluation 2 i'
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=f 4.5.1, Reactor Trip System Reliability '(System Functional Testing)
- By a letter dated November 5, 1983, Yankee Atomic Electric Company (the lic-i ensee) described their planned or completed actions regarding
- .the above items I
for Yankee Nuclear Power Plant.
2.0 Evaluation i
! 2.1 General-
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! Generic Letter 83-28 included various NRC staff positions regarding the
- specific actions to be taken by operating reactor licensees.and operating
- license applicants. The Generic Letter 83-28 positions and discussions 2 . of licensee compliance regarding Actions 3.1.1, 3.1.2, 3.2.1, 3.2.2, 4.1
.g and 4.5.1-for Yankee are presented in the sections that follow. .j
- 6 f ? 2.2 Actions 3.1.1 and 3.1.2, Post-Maintenance Testing (Reactor Trip.
i System Components) :
j Position
} Licensees and appitcants shall submit'the results of their review of test l and maintenan e procedures and Technical Specifications to assure that !
1 post-maintenance operability testing of safety-related components in the i j reactor trip system (RTS) is required to be conducted and that the test-j ing demonstrates that the equipment is capable of performing _its safety j functions before being returned to service.
4 Licensees and applicants shall submit the results of their check of ven-
] dor and engineering recommendations (regarding safety-related components
- in the RTS) to ensure that any appropriate' test guidance is-included in- '
! the test and maintenance procedures or.the. Technical Specifications, j where required.
! Discussion i
i The licensee's response states that the requirements of Technical Speci- 1 i fications and applicable engineering and vendor recommendations have been i reviewed to assure that post-maintenance testing.of safety-related com-j ponents in the reactor trip system are required to be conducted and that j the testing demonstrates that_the equipment is capable of. performing its
- safety function.
Additionally, the licensee's response states that the results of their
, review disclose that appropriate'information has.been incorporated into i applicable test and maintenance procedures to meet the requirements of j the above actions.
Based on the above, the licensee has complied'with the NRC Staff position for Actions 3.1 1 and 3.1.2 of Generic Letter 83-28.
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9 Safety Evaluation 3 2.3 Actions 3.2.1 and 3.2.2, Post-Maintenance Testing (All Other Safety-Related Components)
Position Licensees and applicants shall submit a report documenting the extending of test and maintenance procedures and Technical Specifications review to assure that post-maintenance operability testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety function before being returned to service.
Licensees and applicants shall submit the results of their check of ven-dor and engineering recommendations (regarding all other safety-related components) to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, i where required.
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Discussion The licensee's response states that the requirements of Technical Speci-fications and applicable engineering and vendor recommendations are in-corporated into test and maintenance procedures. These procedures are reviewed to ensure that post-maintenance operability testing of safety-related components is required to be conducted and that the testir.g demonstrates that the equipment is capable of performing its intended safety functions before being returned to service. Additionally, testing and maintenance procedures are developed utilizing appropriate vendor information and engineering recommendations.
The licensee's response further states that their review of vendor and engineering recommendations as required by Actions 3.2.1 and 3.2.2 sub-stantiates their belief that their system for incorporating appropriate vendor and engineering recommendations is effective and adequate.
' Based on the above, the licensee has complied with the NRC Staff position for Actions 3.2.1 and 3.2.2 of Generic Letter 83-28.
2.4 Action 4.1, Reactor Trip System Reliability (Vendor-Related Modifications)
Position All vendor-recommended reactor trip breaker modifications shall be re-viewed to verify ~that either: (1) each modification has, in fact, been implemented; or (2) a written evaluation of the technical reasons for not implementing a modification exists.
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4 For example, the modifications recommended by Westinghouse in NCD-Elec-18 for the DB-50 breakers and a March 31, 1983 letter for the DS-416 breakers shall be implemented or a justification for not implementing shall be made available. Modifications not previously made shall be incorporated or a written evaluation shall be provided.
DISCUSSION The reactor trip breakers at Yankee are Westinghouse ACBs, DB-25. No modifications of these have been recommended by the vendor.
. Based on the above, the licensee has complied with the NRC staff
- position for Action 4.1 of Generic Letter 83-28.
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( 2.5 Action'4.5.1, Reactor Trip System Reliability (System Functional Testing)
POSITION On-line functional testing of the reactor trip system, including independent testing of the diverse trip features, shall be performed.
The diverse trip features to be tested include the breaker undervoltage and shunt trip features on Westinghouse, B&W, and CE plants; the j circuitry used for power interruption with the silicon controlled '
rectifiers on B&W plants; and the scram pilot valves and backup scram valves (including all initiating circuitry) on GE plants.
DISCUSSION On-line testing cannot be conducted at Yankee. This action is therefore deferred to Item 4.5.2 regarding justification for no on-line testing, which will be reviewed separately. In addition, no diverse trip features (UV-trips) are installed at Yankee. The generic letter assumes that undervoltage devices are already installed in Westinghouse breakers. This is not true for Yankee breakers which have shunt trip attachments only and no undervoltage devices. Item 4.3 of Generic Letter 83-28 addresses the need for diverse trip features for breakers.
The staff's SER, provided in an August 9, 1984 letter from W. A. Paulson to J. A. Kay, found the Yankee plant design provided sufficient diversity, including provisions for reactor shutdown under loss-of-voltage conditions, to meet the requirements of Item 4.3.
Based on the above, Action 4.5.1 of Generic Letter 83-28 has been reviewed and is considered closed. However, the unique situation described above will be reviewed under Items 4.5.2 of Generic Letter 83-28.
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3.0 CONCLUSION
Based upon the foregoing discussions, the staff concludes that the licensee is in compliance with Actions 3.1.1, 3.1.2, 3.2.1, 3.2.2, 4.1, and 4.5.1 of Generic Letter 83-28.
Date: NAY 0 71986 Principal Contributor: John A. Schumacher DRP f
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