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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
[Table view] |
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. nnan UNITED STATES
[ NUCLEAR REGULATORY COMMISSION y yE WASHINGTON, D. C. 20555
,f SAFETY EVALUATION REPORT BY OFFICE OF NUCLEAR REACTOR REGULATION FIRST LEVEL UNDERVOLTAGE PROTECTIOh TESTING YANKEE NUCLEAR POWER STATION YANKEE ATOMIC ELECTRIC COMPANY DOCKET NO.50-029 INTRODUCTION By letter dated June 30, 1986, (Reference 1), as supplemented October 22, 1986 (Reference 2) and December 4, 1986, (R.eference 3), Yankee Atomic Electric Company (YAEC), licensee for the Yankee Nuclear Power Station, submitted a basis for not performing monthly tests of the first level (loss-of-voltage) undervoltage protection and for not including the present refueling interval testing in the Technical Specifications (TS). Presented below is the staff's evaluation of this proposal.
BACKGROUND On June 3, 1977 (Reference 4), the NRC sent a letter to all licensees which established multi-plant Action Item B-23, Potential Equipment Failures Associated with Degraded Grid Voltage. As a result of this issue, YAEC installed on each emergency bus a degraded voltage relay which alarms in the control room. This is in addition to the loss-of-voltage relay which automatically starts the diesel generator for the affected bus on loss of bus voltage. In an SE dated January 4, 1985, (Reference 5), the NRC accepted this design, but requested Technical Specifications (TS) for monthly testing of both the first level (loss-of-voltage) and second level (degraded voltage). On November 30, 1985, (Reference 6), TS on the second level protection monthly testing were implemented. By letter dated June 30, 1986 as supplemented October 22, 1986 and December 4, 1986, YAEC provided its basis why monthly testing of the first level protection is not required and why inclusion of the existing refueling interval testing into the TS is not required.
i EVALUATION Attachment 1 is a one line diagram showing the emergency bus arrangement at l Yankee. This diagram was included in'a Licensee Event Report dated
! July 18, 1986 (Reference 7). The plant's Auxiliary Electrical System features three electrical divisions. The two outer' divisions are permanently connected to two 115 kV incoming lines, and the center division to the main generator.
The center bus is without power immediately following a generator trip. Power is restored by manually closing a tie breaker to either of the two outside 2400 volt buses, unless the trip is caused by a fault in the generator in which case the center bus is automatically tied to 2400 volt Bus 3.
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The first level undervoltage protection on the emergency buses consists of the following: )
One loss-of-voltage inverse time relay on each 480-volt Class 1E bus.
These relays (induction disc type) are set to actuate in 1.8 seconds on a complete loss of power, or 3.0 seconds at 277 volts (58%), or 7.0 seconds at 370 volts (77%) with a tap setting of 105 volts and time dial of 2. This trip setting corresponds to 399 volts or 83.25% of nominal.
Actuation of this relay under accident condition will energize an auxiliary relay which initiates the following actions:
- a. Actuates a lock-out relay which isolates the 480-volt Class 1E bus from the offsite source and starts the diesel generator.
- b. Trips the high pressure safety injection (HPSI) pump, while maintaining the remaining loss-of-coolant accident (LOCA) loads
[ low pressure safety injection (LPSI) pump and the motor control center (MCC)] on the bus.
- c. Once the diesel generator attains satisfactory voltage and frequency, its output breaker will close. This will deenergize the auxiliary relay which will remove the HPSI trip.
- d. Ten seconds after the bus is reenergized via diesel generator, the HPSI pump is loaded onto the bus. The load shedding is retained since no other loads are sequenced on following the HPSI pump start.
The NRC staff's safety evaluation of January 4, 1985, found the first level undervoltage protection design to be acceptable. However, the evaluation noted that Technical Specifications which require monthly functional testing of the relays should be submitted.
, In their letter dated June 30, 1986, the licensee provided their justification l for the current 18 month testing interval for the first level (loss-of-voltage) undervoltage protection relays. They state that their justification js based on the demonstrated high reliability of the first level relay system, the unique and highly reliable Plant Auxiliary System arrangement, and the sim-l plicity of the existing design. Further, they state that the modifications l required to permit on-line testing of the first level undervoltage protection are complex, would complicate the existing circuitry and could reduce circuit reliability.
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l The licensee states that the Westinghouse Type CV-7 undervoltage relays were '
installed in 1972 and have been tested every 18. months. The available main-tenance history since 1975 indicates that the tripping time in 13 out of 15 i relays tested have not deviated more that +0.06 seconds over 18 months of operation. In two relays, a deviation of +0.16 and +0.15 seconds, respectively, 4
'was found. Also, since 1972, there has not been a single case in which the i undervoltage relays have been challenged and failed to actuate. Further, the undervoltage relay on the 480 volt Emergency Bus 2 is automatically challenged on any reactor scram due to the generator trip.
The staff finds the above data does not by itself disclose whether the licensee's undervoltage protection relays are more or less reliable than average.
IEEE Standard 500-1984 (Reference 8) indicates protective relay failures per million i cycles of operation as follows:
Low Hich Fails to Open U T5 Tts Fails to Close 0.85 5.95 The licensee's relatively few cycles of operation (as compared to a million) does not-indicate that the relays are better or worse than average. However, the staff notes that the average probability of failure upon demand is very
. low. The following discussion centers on this observation.
The licensee states that two to six hours will be required to perform the monthly testing for all three CV-7 relays (Reference 3). This equates to
- approximately one to two hours for each relay. During the testing, the relay i would be unavailable to initiate an action, i.e., equivalent to a failure to
! actuate upon demand.
(conservative directionAssuming each relay)is for this analysis , unavailable for one hour
- failuretoactuateupondemandof1389X10,ghisequatestoaprobabilityof (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> divided by 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> per month). However, the ave i
upondemandis5.95X10-gagefailurerateofprotectiverelaystoactuate (highside). Assuming then that the licensee's relays are equivalent to the average relay, monthly on-line testing of the relays would increase their probability'of failure to actuate upon demand by i more than two orders of magnitude. Thus, the staff agrees that on-line
! monthly ~ testing of the relays is not necessary or desirable. However, the staff will require that the Technical Specifications be amended to require l, testing on a refueling cycle basis at least once per 18 months. Presently,
- the testing of the under-voltage (loss-of-voltage) relays is not included in l l the Technical Specifications, only in the operating procedures (OP-5762).
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, ~4-CONCLUSIONS Our review of the licensee's first level (loss-of-voltage) undervoltage pro-tection system, and other relevant features of the licensee's system, indicates that on-line monthly testing (as compared to testing on a refueling basis) of the first level undervoltage protection system is more likely to decrease over-all reliability than increase it. Therefore, we find that it is not necessary or even desirable to test the first level protection system on a monthly basis.
However, the staff will require the licensee to amend their Technical Speci-fications to provide for testing of the first level undervoltage protection system on a refueling basis at least once per 18 months consistent with the Westinghouse Standard Technical Specifications (Rev. 4). The licensee is required to submit such Technical Specifications for staff review and approval.
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REFERENCES:
- 1. Yankee Atomic Electric Company letter, G. Papanic, Jr. to E. McKenna, NRC, dated June 30, 1986.
- 2. Yankee Atomic Electric Company letter, G. Papanic, Jr. to E. McKenna, NRC dated October 22, 1986.
- 3. Yankee Atomic Electric Company letter, G. Papanic, Jr. to E. McKenna, NRC, dated December 4, 1986.
- 4. NRC letter, A. Schwencer to R. H. Groce, Yankee Atomic Electric Company, dated June 3, 1977.
- 5. NRC letter, J. Zwolinski to J. Kay, Yankee Atomic Electric Company letter, dated January 4,1985.
- 6. NRC letter, G. Lear to G. Papanic, Jr. Yankee Atomic Electric Company letter, dated November 30, 1985.
- 7. Licensee Event Report, LER 86-07, Yankee Nuclear Powre Station, July 18, 1986.
- 8. IEEE Standard 500-1984, Reliability Data for Nuclear Powr Generating Stations.
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1987 istribution Copies:
81 ' .'~;Ft M NRC PDR-Local PDR PAD #1 r/f.
-PAD #1'p/f
-TNovak, Actg. DD NThompson, DHFT OGC-Bethesda.
.EJordan BGrimes JPartlow Glear PShuttleworth EMcKenna AToalston FRosa ACRS (10)
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