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| | issue date = 07/30/1987 | | | issue date = 07/30/1987 |
| | title = LER 87-014-00:on 870630,discovered That Auxiliary Feedwater (AFW) Could Be Lost to Operating Unit in Event of High Energy Line Break in Main Steam Valve House.Caused by Poor Review of AFW Supply.Review continuing.W/870730 Ltr | | | title = LER 87-014-00:on 870630,discovered That Auxiliary Feedwater (AFW) Could Be Lost to Operating Unit in Event of High Energy Line Break in Main Steam Valve House.Caused by Poor Review of AFW Supply.Review continuing.W/870730 Ltr |
| | author name = BENSON D L | | | author name = Benson D |
| | author affiliation = VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) | | | author affiliation = VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:POW 28-06-_01 NRC Form 366 * (9-83,) U.S. NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 3160,0104 EXPIRES: 8/31/88 LICENSEE EVENT REPORT (LER) FACILITY NAME (1) !DOCKET NUMBER (2) PAGE 131 | | {{#Wiki_filter:POW 28-06-_01 NRC Form 366 |
| * Surry P9wer *station. | | * U.S. NUCLEAR REGULATORY COMMISSION (9-83,) |
| Unit 1 o I 5 I o Io I o 12 I 8 I O 1 loF O I 3 TITLE (41 Inadequate R*eview of AFW Suoply FollowinQ a HELB in Safecrnards EVENT DATE (5) LER NUMBER (6) REPORT DATE (71 OTHER FACILITIES INVOLVED (Bl MONTH DAY YEAR YEAR FACILITY NAMES DOCKET NUMBER(SI o I 6 31 o a 1 al 1 .:.... ol 1 14 -o I o ol 1 | | APPROVED 0MB NO. 3160,0104 EXPIRES: 8/31/88 LICENSEE EVENT REPORT (LER) |
| * 3 I o a I 1 01s1010101 I I OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more of the folloV(inlJ) | | FACILITY NAME (1) DOCKET NUMBER (2) PAGE 131 |
| (11) MODE (9) N 20.402(b) 20.406(cl 50,73(11(2)(ivl 73.71(bl -POWER 1 | | * TITLE (41 Surry P9wer *station. Unit 1 !o I 5 I o Io I o 12 I 8 I O 1 loF O I 3 Inadequate R*eview of AFW Suoply FollowinQ a HELB in Safecrnards EVENT DATE (5) LER NUMBER (6) REPORT DATE (71 OTHER FACILITIES INVOLVED (Bl MONTH DAY YEAR YEAR FACILITY NAMES DOCKET NUMBER(SI oI 6 31 oa 1 al 1 .:.... ol 1 14 - oI o ol 1 |
| * 20.406(1)(1
| | * 3 I o aI 1 01s1010101 I I OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more of the folloV(inlJ) (11) |
| )(I) LEVEL -<101 1 I O I O 20.405t1H1Hm
| | POWER MODE (9) 1* |
| ,...._ _,__ &0.36(c){1) 50.73(11(21(v) 73.71(cl ,__. -50.38(cl(2) 50.73(1)(2l{viil . OTHER (Spacify in Abstr*ct bl/ow and in T*Kt, NRC Form ,__. --=~== :::::::::;: | | N 20.402(b) 20.406(1)(1 )(I) -,__. |
| 60.73(11(21(11
| | 20.406(cl |
| &0.73(1)(2)(vllll(A) 366A) -,...._ 50,73(11{2){ill 50.73(11(21(vllll(BI | | &0.36(c){1) |
| --50.73(11(2)(iiil 50.73(1)(21(,i)
| | ~ |
| LICENSEE CONTACT FOR THIS LER (121 NAME TELEPHONE NUMBER AREA CODE D. L. Benson, Stat~n Manager . 81 01 4 31 51 7 I -I 31 11 8 Ill CAUSE SYSTEM COMPONENT I I I I I I I I COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 TURER I I I I I I SUP~LEMENTAL REPORT EXPECTED (141 SYSTEM I I COMPONENT.
| | 50,73(11(2)(ivl 50.73(11(21(v) 73.71(bl 73.71(cl |
| I I I I I I . MANUFAC-TURER I I I I I I EXPECTED SUBMISSION MONTH DAY YEAR :1 YES /If v**. complete EXPECTED SUBMISSION DATE) DATE (151 nl1 ~11 RIR ABSTRACT (Limit to 1400 spaces, i.B., approximately fifteen single-space typewritten lines} (161 NRC Form 386 19-83) On June 30, 1987 with Unit One at 100% power and Unit Two at 69% power, *it was determined as a result of a company initiated | | =~== : : : : :;: - |
| * safety system functional inspection, that a unique scenario could exist whereby auxiliary feedwater (AFW) {EIIS-BA} | | LEVEL - |
| could be lost to*an operating unit. In the event of a high energy line break (HELB) in the main steam valve house (MSVH) of an operating unit, the AFW pumps {EIIS-P} on that unit could be rendered inoperable due to the harsh environment. | | 1 IO I O 50.38(cl(2) 50.73(1)(2l{viil . OTHER (Spacify in Abstr*ct |
| With the unaffected unit at cold shutdown and only one of its motor driven AFW pumps available for cross connect to the affected unit, an assumed loss of offsite power, with a single failure could result in a loss of AFW to the affected unit. Administrative controls have been established to ensure that . when a unit is above 350 degrees/450 psig, two motor driven AFW pumps will be available from the other unit. Engineeririg is reviewing the original s~fety analysis for the HELB (loss of AFW) to assess the adequacy of the evaluation and determine if additional corrective actions are required. | | <101 20.405t1H1Hm ,__. |
| 97oso~o 00 1il i~~&eo PDR " PDR s . ...._\ t P*' p., , | | 60.73(11(21(11 &0.73(1)(2)(vllll(A) |
| NRC Form 366A* (9-83) FACILITY NAME (11 POW 28-06~01 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION* | | - bl/ow and in T*Kt, NRC Form 366A) |
| U,S, NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 3150--0104 EXPIRES: 8/31/88 DOCKET NUMBER (21 LER NUMBER (61 PAGE (31 ?Urry Power Station, Unit 1 o Is Io Io I o 12 1 a I o a 11 -o I 114 -o I o -o I 2 oF o 13 TEXT /ff tnOf9 _,,. i& n,quintd, usa additional NRC Farm 366A 's/ (171 . NRC FORM 366A (9-831 1. 0 Description of the Event* On June 30, 1987 with Unit One at 100% power and Unit Two at 69% power, it was determined as a result of a company initiated safety system functional inspection, that a unique scenario could exist whereby auxiliary feedwater | | - 50,73(11{2){ill 50.73(11(2)(iiil - |
| * (AFW) {EIIS.:..BA} | | LICENSEE CONTACT FOR THIS LER (121 50.73(11(21(vllll(BI 50.73(1)(21(,i) |
| could be lost to an operating unit. In the event of a high energy line break (HELB) in the main steam valve house (MSVH) of an operating unit, the AFW pumps {EIIS-P} on that unit could be rendered inoperable due to the harsh environment. | | NAME TELEPHONE NUMBER AREA CODE D. L. Benson, Stat~n Manager |
| With the unaffected unit at cold .shutdown and only one of its motor driven AFW pumps available.* | | . 81 01 4 31 51 7 I - I 31 11 8 Ill COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 CAUSE SYSTEM COMPONENT MANUFAC- SYSTEM COMPONENT. MANUFAC-TURER TURER I I I I I I I I I I I I I I I I I I I I I I I I I . I I I SUP~LEMENTAL REPORT EXPECTED (141 MONTH DAY YEAR EXPECTED SUBMISSION |
| for cross connect to the affected unit, an assumed loss of offsite power, with a single failure could result in a loss of AFW to' the affected unit. The existing Technical Specification requires that only one motor driven AFW pump be available when the unit is at cold shutdown. | | :1 YES /If v**. complete EXPECTED SUBMISSION DATE) |
| This specification was developed in 1980 to provide assurance that AFW could be supplied from one unit to another in the event that the AFW pumps in one unit were destroyed by fire (Appendix R). The HELB event was not considered at that time, and Appendix R does not require cqnsideration of the single failure criterion. | | DATE (151 nl1 ~11 RIR ABSTRACT (Limit to 1400 spaces, i.B., approximately fifteen single-space typewritten lines} (161 On June 30, 1987 with Unit One at 100% power and Unit Two at 69% |
| This event is reportable in accordance with lOCFR 50. 73(a) (2) (v). 2.0 Safety Consequences and Implications This scenario is not analyzed in the UFSAR. ever, a main steam line break is a condition IV event and is of very low probability. | | power, *it was determined as a result of a company initiated |
| Engineering is reviewing the original safety analysis to determine whether an unreviewed safety question exists. Additionally, Functional Restoration Procedure H.l provides guidance in-the event of total loss of FW and AFW to the Steam Generators. | | * safety system functional inspection, that a unique scenario could exist whereby auxiliary feedwater (AFW) {EIIS-BA} could be lost to*an operating unit. In the event of a high energy line break (HELB) in the main steam valve house (MSVH) of an operating unit, the AFW pumps {EIIS-P} on that unit could be rendered inoperable due to the harsh environment. With the unaffected unit at cold shutdown and only one of its motor driven AFW pumps available for cross connect to the affected unit, an assumed loss of offsite power, with a single failure could result in a loss of AFW to the affected unit. |
| Further protection is provided through an extensive non-destructive testing program of the main steam postulated break points in the main steam valve house . | | Administrative controls have been established to ensure that . |
| NRC Form 366A 19-831 FACILITY NAME (1) e .pow 28-06-01 e .. LICENSEE EVENT REPORT (LERI TEXT CONTINUATION DOCKET NUMBER 121 U.S. NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 3150-0104 EXPIRES: 8/31/88 PAGE 13) SURRY POWER STATION, UNIT 1 o 15 Io Io Io 12 I 810 81 7 -o 11 I 4 lo In n I ':l OF n h TEXT (If mo,e-,,.c. | | when a unit is above 350 degrees/450 psig, two motor driven AFW pumps will be available from the other unit. |
| ia ,wqund, u* tldditioMI NRC Form 366A'1/ l17l NRc FOAM 366.A 19-831 3.0 Cause Section l4B.5.1.7 of the UFSAR considers a HELB in the MSVH and assumes that should all three of the affected units AFW pumps become inoperable; redundancy is not lost because there are three AFW pumps* available from the unaffected unit. The UFSAR does not consider .the possibility that only one AFW pump from the unaffected unit might be available when that unit is at cold shutdown. | | Engineeririg is reviewing the original s~fety analysis for the HELB (loss of AFW) to assess the adequacy of the evaluation and determine if additional corrective actions are required. |
| 4.0 Immediate Corrective Action Administrative controls have been established to ensure that when a unit is above 350 degrees/450 psig, two motor driven AFW pumps will be available from the other unit. ,. 5.0 Additional Corrective Actions Engineering is reviewing the original safety analysis for the HELB (loss of AFW) to assess the adequacy of the evaluation and to qetermine if additional corrective actions are required. | | 97oso~o001il i~~&eo PDR " PDR . . . _\ t s P*' p., , |
| 6.0 Actions Taken to Prevent Recurrence None. 7.0 Similar Events None 8.0 Manufacturer/Model Number N/A | | NRC Form 386 19-83) |
| * U.S.GPO: 1986*0*824-538/455 | | |
| .. e Vepco . July 30, 1987 U.S. Nuclear Regulatory Commission Document Control Desk 016 Phillips Building Washington, D.C. 20555 Gentlemen:
| | POW 28-06~01 NRC Form 366A* |
| Pursuant to Surry Power Station Technical and Power Company hereby submits the for Surry Unit 1 and 2. VIRGINIA ELECTRIC AND POWER COMPANY P, D. Box 315 Surry, Virginia 23883 Serial No: 87-016 Docket No: 50-280 50-281 License No: DPR-32 DPR-37 Specifications~ | | (9-83) U,S, NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION* APPROVED 0MB NO. 3150--0104 EXPIRES: 8/31/88 FACILITY NAME (11 DOCKET NUMBER (21 LER NUMBER (61 PAGE (31 |
| Virginia Electric following Licensee Event Report REPORT NUMBER 87-014-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by Safety Evaluation and Control. Very truly yours, David L. Benson Station Manager Enclosure pc: Dr. J. Nelson Grace Regional Administrator Suite 2900 101 Marietta Street, NW Atlanta, Georgia 30323}}
| | ?Urry Power Station, Unit 1 TEXT /ff tnOf9 _ , , . i& n,quintd, usa additional NRC Farm 366A 's/ (171 o Is Io Io I o 12 1a I o a11 - o I 114 - oI o -o I 2 oF o 13 |
| | : 1. 0 Description of the Event* |
| | On June 30, 1987 with Unit One at 100% power and Unit Two at 69% power, it was determined as a result of a company initiated safety system functional inspection, that a unique scenario could exist whereby auxiliary feedwater * (AFW) {EIIS.:..BA} could be lost to an operating unit. In the event of a high energy line break (HELB) in the main steam valve house (MSVH) of an operating unit, the AFW pumps {EIIS-P} on that unit could be rendered inoperable due to the harsh environment. With the unaffected unit at cold .shutdown and only one of its motor driven AFW pumps available.* |
| | for cross connect to the affected unit, an assumed loss of offsite power, with a single failure could result in a loss of AFW to' the affected unit. |
| | The existing Technical Specification requires that only one motor driven AFW pump be available when the unit is at cold shutdown. This specification was developed in 1980 to provide assurance that AFW could be supplied from one unit to another in the event that the AFW pumps in one unit were destroyed by fire (Appendix R). The HELB event was not considered at that time, and Appendix R does not require cqnsideration of the single failure criterion. |
| | This event is reportable in accordance with 10CFR |
| | : 50. 73(a) (2) (v). |
| | 2.0 Safety Consequences and Implications This scenario is not analyzed in the UFSAR. How-ever, a main steam line break is a condition IV event and is of very low probability. Engineering is reviewing the original safety analysis to determine whether an unreviewed safety question exists. |
| | Additionally, Functional Restoration Procedure H.l provides guidance in- the event of total loss of FW and AFW to the Steam Generators. Further protection is provided through an extensive non-destructive testing program of the main steam postulated break points in the main steam valve house . |
| | . NRC FORM 366A (9-831 |
| | |
| | e e.pow .. 28-06-01 NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION 19-831 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION APPROVED 0MB NO. 3150-0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER 121 PAGE 13) |
| | SURRY POWER STATION, UNIT 1 o 15 Io Io Io 12 I 810 81 7 - o 11 I 4 lo In nI ':l OF n h TEXT (If mo,e-,,.c. ia ,wqund, u* tldditioMI NRC Form 366A'1/ l17l 3.0 Cause Section l4B.5.1.7 of the UFSAR considers a HELB in the MSVH and assumes that should all three of the affected units AFW pumps become inoperable; redundancy is not lost because there are three AFW pumps* available from the unaffected unit. The UFSAR does not consider |
| | .the possibility that only one AFW pump from the unaffected unit might be available when that unit is at cold shutdown. |
| | 4.0 Immediate Corrective Action Administrative controls have been established to ensure that when a unit is above 350 degrees/450 psig, two motor driven AFW pumps will be ,. available from the other unit. |
| | 5.0 Additional Corrective Actions Engineering is reviewing the original safety analysis for the HELB (loss of AFW) to assess the adequacy of the evaluation and to qetermine if additional corrective actions are required. |
| | 6.0 Actions Taken to Prevent Recurrence None. |
| | 7.0 Similar Events None 8.0 Manufacturer/Model Number N/A NRc FOAM 366.A 19-831 |
| | * U.S.GPO: 1986*0*824-538/455 |
| | |
| | e Vepco VIRGINIA ELECTRIC AND POWER COMPANY P, D. Box 315 Surry, Virginia 23883 |
| | . July 30, 1987 U.S. Nuclear Regulatory Commission Serial No: 87-016 Document Control Desk Docket No: 50-280 016 Phillips Building 50-281 Washington, D.C. 20555 License No: DPR-32 DPR-37 Gentlemen: |
| | Pursuant to Surry Power Station Technical Specifications~ Virginia Electric and Power Company hereby submits the following Licensee Event Report for Surry Unit 1 and 2. |
| | REPORT NUMBER 87-014-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by Safety Evaluation and Control. |
| | Very truly yours, David L. Benson Station Manager Enclosure pc: Dr. J. Nelson Grace Regional Administrator Suite 2900 101 Marietta Street, NW Atlanta, Georgia 30323}} |
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LER 97-002-00:on 970116,one Train of Auxiliary Ventilation Sys Declared Inoperable.Caused by Personnel Error.Properly Adjusted Damper 1-VS-MOD-58B & Exited Seven Day LCO on 970116.W/970214 Ltr
ML18153A0951997-01-0202 January 1997
LER 97-002-00:on 961213,automatic Reactor Trip Occurred During Planned Shutdown.Caused by Steam Flow/Feedwater Flow Mismatch.Rps Functioned as Designed & Plant Placed in Hot Shutdown
ML18153A0931996-12-12012 December 1996
LER 96-008-00:on 961112,water Gas Decay Tank Oxygen Analyzer Pressure Sensors Inoperable Due to Vendor Supplied Equipment Not Meeting Procurement specifications.Post-implementation Procedures Revised & Transducers replaced.W/961212 Ltr
ML18153A0691996-09-19019 September 1996
LER 96-007-00:on 960821,failed to Complete Fire Detection Zone Inspections within Required Time Period.Caused by Personnel Error.Counseled Personnel Re Fire Detection Zone Inspections & Revised Fire Watch training.W/960920 Ltr
ML18153A0481996-08-26026 August 1996
LER 96-005-00:on 960803,manual Reactor Trip.Caused by Loss of Electro Hydraulic Control Pressure.Repaired Two Compression Fitting Union Connections on Leaking Fitting & Performed Evaluations on Other tubing.W/960826 Ltr
ML18153A0521996-08-20020 August 1996
LER 96-004-01:on 960510,discovered Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies.Implemented Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures.W/960820 Ltr
ML18153A0321996-07-30030 July 1996
LER 96-006-01:on 960618,anti-corrosion Coating Had Not Been Reapplied to Station Battery 2B.Caused by Procedural Error in That Verbatim TS Compliance Not Reflected in Procedures. Coating Was Applied to batteries.W/960730 Ltr
ML18153A0281996-07-17017 July 1996
LER 96-006-00:on 960618,failed to Apply anti-corrosion Coating to Station Battery 2B.Caused by Procedural Error. Applied anti-corrosion Coating to Batteries & Revised TS 4.6.C.1.f Re Battery Coating requirements.W/960717 Ltr
ML18153A0141996-07-0202 July 1996
LER 96-004-00:on 960606,turbine/reactor Trip Occurred.Caused by High Level in Steam Generator B.Placed Plant in Hot Shutdown Condition,Calculated Shutdown Margin & Monitored Critical Safety Function Status trees.W/960702 Ltr
1999-08-27
[Table view]
Category:RO)
MONTHYEARML18152B4411999-08-27027 August 1999
LER 99-005-00:on 990731,effluent Radiation Monitors Were Declared Inoperable.Caused by Degraded Heat Trace Circuits for Monitors Sample Suction Line.Degraded Heat Trace Circuit Was Replaced & Addl Heat Trace Is Being Installed
ML18152B4421999-08-27027 August 1999
LER 99-006-00:on 990802,determined That Plant Was Outside of App R Design Basis Due to Fire Barrier Deficiencies. Caused by Original Plant Design Deficiencies.Fire Watches Were Established & Mods Have Been Completed.With 990827 Ltr
ML18152B3771999-08-13013 August 1999
LER 99-004-00:on 990714,TS Violation Due to non-safety Related Fans Effect on CR Boundary Was Noted.Cause of Event Has Not Yet Been Determined.Cable Spreading Room Doors Were Operned to Reduce Pressure in Rooms
ML18152B4181999-05-18018 May 1999
LER 99-002-00:on 990425,MSSVs Tested Out of Tolerance for as Found Setpoint.Caused by Minor Setpoint Drift.No Immediate Action Required.Deviation Rept Submitted for Each Valve.With 990518 Ltr
ML18152B4111999-04-28028 April 1999
LER 99-003-00:on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With 990428 Ltr
ML18153A2741999-03-29029 March 1999
LER 99-002-00:on 990301,prematurely Released Fire Watches Resulted in Violation of TS 3.21.B.7.Caused by Inadequate Procedure.Procedure for Opening & Sealing Fire Stops Was Revised on 990212
ML18153A2681999-03-19019 March 1999
LER 98-013-01:on 981122,turbine/reactor Tripped on High Due to Short Circuit in Summator for MSL C Loop Channel III Flow Transmitter.Replaced 1-MS-FT1494 Summator & Module Repair Procedure Revised.With 9903190 Ltr
ML18152B7261999-01-21021 January 1999
LER 99-001-00:on 981222,auxiliary Feedwater Pipe Support Missed Surveillance.Caused by Personnel Error.Station Deviation Rept Was Submitted.Two Supports in Question Received Required Code Insp & Were Found Acceptable
ML18152B5811998-12-16016 December 1998
LER 98-013-00:on 981122,turbine/reactor Trip on High SG Level Occurred.Caused by Instrument Failure.Control Room Operators Placed Unit in Safe,Shutdown Condition
ML18152B5781998-12-16016 December 1998
LER 98-014-00:on 981126,manual Reactor Trip in Response to Main Feedwater Regulating Valve Failure Occurred.Caused by Dislocation of Retaining Clip in Positioner.Control Room Operators Placed Unit in Safe,Shutdown Condition
ML18152B7121998-12-0404 December 1998
LER 98-S01-00:on 981105,noted Failure to Deactivate Station Access Badge.Caused by Human Error.Licensee Will Now Deactivate Station Badges Before Clearance Is Revoked & Process for Badge Deactivations Have Been Strengthened
ML18152B7041998-12-0101 December 1998
LER 98-012-00:on 981102,noted That EDGs Were Concurrently Inoperable.Caused by Required Testing Per TS 3.16.B.1.a.2. Redundant EDG Was Returned to Svc within Two Hour Period, Following Satisfactory Testing.With 981201 Ltr
ML18152B6161998-11-0606 November 1998
LER 98-011-00:on 981008,diesel Driven Fire Pump Failed to Start During Performance of Monthly Operability Test.Caused by Faulty Overspeed Trip Device Failure.Diesel Driven Fire Pump Declared Inoperable
ML18152B6081998-10-23023 October 1998
LER 98-010-01:on 980715,intake Canal Level Probes Were Inoperable Due to Marine Growth.Caused by Design of Canal Level Instrumentation.Canal Level Probes Will Continue to Be Monitored More Closely
ML18152B7811998-07-31031 July 1998
LER 98-010-00:on 980715,low Intake Canal Level Instrument Channel I Was Declared Inoperable to Allow Testing of Intake Canal Level Probe 1-CW-LE-102.Subject Probe Was Cleaned by Diver,Tested & Channel I Was Returned to Operable Status
ML18153A2581998-06-0303 June 1998
LER 98-009-00:on 980509,nonisolable Leak of Reactor Coolant Pump Seal Injection Line Weld,Was Discovered.Caused by Lack of Fusion or Thermal Fatigue Coupled W/Vibration Stress Due to Loose Rod Hanger.Rcp Seal Injection Line Removed
ML18152B8241998-05-22022 May 1998
LER 98-008-00:on 980228,auxiliary Ventilation Fans Were Noted in Condition Outside of Design Basis.Caused by Failure to Recognize Potential Impact of Certain Design Basis Accident Scenarios.No Corrective Actions Needed
ML18152B7951998-04-29029 April 1998
LER 98-007-00:on 980330,radiation Monitors Were Declared Inoperable.Caused by Change in Operating Temperature Range. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-6
ML18153A2521998-04-22022 April 1998
LER 98-005-01:on 980212,fire Watch Insp Exceeded One Hour. Caused by Lack of Attention to Detail by Individual Involved.Individual Involved Was Coached on Requirement to Perform Fire Watch Patrols within Required Time Frame
ML18153A2511998-04-22022 April 1998
LER 98-006-00:on 980324,unisolable Through Wall Leak of RCP Thermowell Was Noted.Cause of Leak Is Unknown.Rtd Will Be Replaced
ML18153A2391998-03-13013 March 1998
LER 98-005-00:on 980212,fire Watch Insp Frequency Exceeded One H Occurred.Category 2 Root Cause Evaluation Being Conducted to Determine Cause of Event.Station Deviation Issued
ML18153A2341998-03-0909 March 1998
LER 98-003-00:on 980226,no Procedural Guidance for Maintaining EDG Minimum Fuel Supply During Loop,Was Identified.Caused by Absence of Procedural Instructions. Deviation Rept Submitted to Document Deviating Condition
ML18153A2301998-03-0606 March 1998
LER 98-004-00:on 980206,fire Watch Was Released Prematurely Resulting in Violation of Ts.Caused by Inadequate Planning of Repair Activity.Work Orders Will Include Ref to Applicable Procedures Developed to Assist in Repairs
ML18153A2251998-03-0404 March 1998
LER 98-002-00:on 980202,automatic Turbine Trip Resulted in Automatic Reactor Trip.Caused Degraded Generator Voltage Regulator sub-component Failure.Placed Plant in Safe Hot SD & Replaced Intermittent Relay & Relay Socket
ML18153A2201998-02-0606 February 1998
LER 98-001-00:on 980108,deficient Test Due to Faulty Test Equipment Resulted in TS Violation.Caused by Faulty Vibration Analyzer Cable or Loose Connection.Station Deviation Rept Was submitted.W/980206 Ltr
ML18153A2071998-01-13013 January 1998
LER 97-012-01:on 971028,loss of Power to Latching Mechanism on Several Doors Occurred.Caused by Tripping of Two Breakers in Security Distribution Panel.Reset Affected Breakers Which Restored Power to Security Systems & Affected Doors
ML18153A2101998-01-13013 January 1998
LER 97-009-01:on 971014,declared Intake Canal Level Probes Inoperable Due to Marine Growth.Caused by Inadequate Maint of Intake Canal Level Probes.Subject Probes Were Cleaned, Tested Satisfactorily & Returned to Operable Status
ML18153A1911997-11-26026 November 1997
LER 97-011-00:on 971030,determined That Periodic Test Procedures for Testing Reactor Trip Bypass Breakers Did Not Test Manual Undervoltage Trip.Caused by mis-interpretation of Term in-service. Procedures Revised
ML18153A1971997-11-26026 November 1997
LER 97-012-00:on 971028,loss of Power to Latching Mechanism on Several Doors Occurred.Caused by Tripping of Breaker in Security Distribution Panel in Central Alarm Station (CAS) Panel.Breakers in Affected CAS Panel Reset
ML18153A1921997-11-25025 November 1997
LER 97-010-00:on 971028,discovered Missed Fire Protection Surveillance Pt.Caused by Personnel Error.Satisfactorily Completed PT Procedure 0-OPT-FP-009 & Diesel Driven Fire Pump 1-FP-P-2 Declared operable.W/971125 Ltr
ML18153A1831997-11-12012 November 1997
LER 97-009-00:on 971014,declared Intake Canal Level Probes Inoperable Due to Marine Growth.Cause Indeterminate.Divers Inspected,Cleaned & Returned Probes to Operable Status & Initiated Interdepartmental Team to Investigate Cause
ML18153A1791997-11-0707 November 1997
LER 97-008-00:on 971011,invalid Actuation of ESF Occurred. Caused by Personnel Errors.Main CR Bottled Air Sys Isolated & Containment Hydrogen Analyzer Heat Tracing Actuation Signal Reset
ML18153A1721997-10-30030 October 1997
LER 97-007-00:on 970930,determined That Plant Was Outside App R Design Basis Due to Vital Bus Isolation Issue.Caused by Personnel Error.Installed Circuit Protective Device During Oct 1997 Refueling Outage
ML18153A1421997-06-10010 June 1997
LER 97-001-01:on 970123,shutdown Occurred Due to Drain Line Weld Leak.Inspected & Tested Turbine Trip Actuation circuitry.W/970610 Ltr
ML18153A1391997-05-28028 May 1997
LER 97-005-00:on 970502,Unit 1 Power Range Nuclear Instrumentation Was Inoperable Due to Personnel Error.Sro & STA That Were Involved in Event Were Counseled
ML18153A1291997-04-18018 April 1997
LER 97-006-00:on 970320,loss of Refueling Integrity Due to Inadequate Containment Closure Process & Verification.Fuel Movement Stopped IAW Action Statement Requirements of TS 3.10.B.W/970418 Ltr
ML18153A1281997-04-15015 April 1997
LER 97-004-00:on 970317,main Steam Safety Valve Was Outside as Found Setpoint Tolerance.Specific Cause Unknown,However, Minor Setpoint Drift Can Be Expected.No Immediate Corrective Actions performed.W/970415 Ltr
ML18153A1241997-04-0808 April 1997
LER 97-002-01:on 970116,one Train of Auxiliary Ventilation Sys Was Inoperable Outside of Ts.Caused by Personnel Error. Submitted Deviation Rept Re Reverse Rotation of Fan & Work Request to Adjust linkage.W/970408 Ltr
ML18153A1191997-03-19019 March 1997
LER 97-001-00:on 970218,manual Reactor Trip & ESF Actuation Occurred Due to Loss of EHC Control Power.Caused by Momentary Short.Relay Card Was replaced.W/970319 Ltr
ML18153A1201997-03-19019 March 1997
LER 97-003-00:on 970219,loss of Pressurizer Heaters Resulted in Manual U1 Trip & U2 ESF Actuation.Caused by Loss of Group C Pressurizer Proportional Heaters.Reactor Trip Breakers Were Verified open.W/970319 Ltr
ML18153A1131997-02-20020 February 1997
LER 97-001-00:on 970123,shutdown Occurred Due to Steam Drain Line Weld Leak.Management Was Notified & Shift Supervisor Invoked Requirements of TS 4.15.C.1.W/undtd Ltr
ML18153A1101997-02-13013 February 1997
LER 97-002-00:on 970116,one Train of Auxiliary Ventilation Sys Declared Inoperable.Caused by Personnel Error.Properly Adjusted Damper 1-VS-MOD-58B & Exited Seven Day LCO on 970116.W/970214 Ltr
ML18153A0951997-01-0202 January 1997
LER 97-002-00:on 961213,automatic Reactor Trip Occurred During Planned Shutdown.Caused by Steam Flow/Feedwater Flow Mismatch.Rps Functioned as Designed & Plant Placed in Hot Shutdown
ML18153A0931996-12-12012 December 1996
LER 96-008-00:on 961112,water Gas Decay Tank Oxygen Analyzer Pressure Sensors Inoperable Due to Vendor Supplied Equipment Not Meeting Procurement specifications.Post-implementation Procedures Revised & Transducers replaced.W/961212 Ltr
ML18153A0691996-09-19019 September 1996
LER 96-007-00:on 960821,failed to Complete Fire Detection Zone Inspections within Required Time Period.Caused by Personnel Error.Counseled Personnel Re Fire Detection Zone Inspections & Revised Fire Watch training.W/960920 Ltr
ML18153A0481996-08-26026 August 1996
LER 96-005-00:on 960803,manual Reactor Trip.Caused by Loss of Electro Hydraulic Control Pressure.Repaired Two Compression Fitting Union Connections on Leaking Fitting & Performed Evaluations on Other tubing.W/960826 Ltr
ML18153A0521996-08-20020 August 1996
LER 96-004-01:on 960510,discovered Hydrogen Analyzers Inoperable.Caused by Procedural Deficiencies.Implemented Permanent Changes to Hydrogen Analyzer Instrument Calibr Procedures.W/960820 Ltr
ML18153A0321996-07-30030 July 1996
LER 96-006-01:on 960618,anti-corrosion Coating Had Not Been Reapplied to Station Battery 2B.Caused by Procedural Error in That Verbatim TS Compliance Not Reflected in Procedures. Coating Was Applied to batteries.W/960730 Ltr
ML18153A0281996-07-17017 July 1996
LER 96-006-00:on 960618,failed to Apply anti-corrosion Coating to Station Battery 2B.Caused by Procedural Error. Applied anti-corrosion Coating to Batteries & Revised TS 4.6.C.1.f Re Battery Coating requirements.W/960717 Ltr
ML18153A0141996-07-0202 July 1996
LER 96-004-00:on 960606,turbine/reactor Trip Occurred.Caused by High Level in Steam Generator B.Placed Plant in Hot Shutdown Condition,Calculated Shutdown Margin & Monitored Critical Safety Function Status trees.W/960702 Ltr
1999-08-27
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML18152A2811999-10-12012 October 1999
Technical Basis for Elimination of Nozzle Inner Radius Insps (for Nozzles Other than Reactor Vessel),Technical Basis for ASME Section XI Code Case N-619.
ML18152B3531999-09-30030 September 1999
Monthly Operating Repts for Sept 1999 for Surry Power Station,Units 1 & 2.With 991012 Ltr
ML18152B6651999-08-31031 August 1999
Monthly Operating Repts for Aug 1999 for Surry Power Station Units 1 & 2.With 990915 Ltr
ML18152B4421999-08-27027 August 1999
LER 99-006-00:on 990802,determined That Plant Was Outside of App R Design Basis Due to Fire Barrier Deficiencies. Caused by Original Plant Design Deficiencies.Fire Watches Were Established & Mods Have Been Completed.With 990827 Ltr
ML18152B4411999-08-27027 August 1999
LER 99-005-00:on 990731,effluent Radiation Monitors Were Declared Inoperable.Caused by Degraded Heat Trace Circuits for Monitors Sample Suction Line.Degraded Heat Trace Circuit Was Replaced & Addl Heat Trace Is Being Installed
ML18151A3981999-08-13013 August 1999
SPS Unit 2 ISI Summary Rept for 1999 Refueling Outage.
ML18152B3771999-08-13013 August 1999
LER 99-004-00:on 990714,TS Violation Due to non-safety Related Fans Effect on CR Boundary Was Noted.Cause of Event Has Not Yet Been Determined.Cable Spreading Room Doors Were Operned to Reduce Pressure in Rooms
ML18152B3791999-07-31031 July 1999
Monthly Operating Repts for July 1999 for Surry Power Station,Units 1 & 2.With 990811 Ltr
ML18152B3911999-06-30030 June 1999
Monthly Operating Repts for June 1999 for Surry Power Station,Units 1 & 2.With 990713 Ltr
ML18152B4341999-05-31031 May 1999
Monthly Operating Repts for May 1999 for Surry Power Station,Units 1 & 2.With 990614 Ltr
ML20195E2401999-05-31031 May 1999
Rev 2 to COLR for SPS Unit 2 Cycle 16 Pattern Ag
ML18152B4181999-05-18018 May 1999
LER 99-002-00:on 990425,MSSVs Tested Out of Tolerance for as Found Setpoint.Caused by Minor Setpoint Drift.No Immediate Action Required.Deviation Rept Submitted for Each Valve.With 990518 Ltr
ML18152B4161999-04-30030 April 1999
Monthly Operating Repts for Apr 1999 for Surry Power Station Units 1 & 2.With 990512 Ltr
ML18152B4111999-04-28028 April 1999
LER 99-003-00:on 990331,potential Loss of Charging Pumps Was Noted.Caused by Main CR Fire.Station Deviation Was Issued on 990331.With 990428 Ltr
ML18152B6511999-03-31031 March 1999
Monthly Operating Repts for Mar 1999 for Surry Power Station Units 1 & 2
ML18153A2741999-03-29029 March 1999
LER 99-002-00:on 990301,prematurely Released Fire Watches Resulted in Violation of TS 3.21.B.7.Caused by Inadequate Procedure.Procedure for Opening & Sealing Fire Stops Was Revised on 990212
ML18153A2681999-03-19019 March 1999
LER 98-013-01:on 981122,turbine/reactor Tripped on High Due to Short Circuit in Summator for MSL C Loop Channel III Flow Transmitter.Replaced 1-MS-FT1494 Summator & Module Repair Procedure Revised.With 9903190 Ltr
ML18152B7331999-02-28028 February 1999
Monthly Operating Repts for Feb 1999 for Surry Power Station,Units 1 & 2.With 990310 Ltr
ML18152B5421999-01-31031 January 1999
Monthly Operating Repts for Jan 1999 for Surry Power Station,Units 1 & 2.With 990210 Ltr
ML18151A3031999-01-29029 January 1999
ISI Summary Rept for 1998 Refueling Outage,Including Form NIS-1, Owners Rept for ISIs & Form NIS-2, Owners Rept for Repairs & Replacements.
ML18152B7261999-01-21021 January 1999
LER 99-001-00:on 981222,auxiliary Feedwater Pipe Support Missed Surveillance.Caused by Personnel Error.Station Deviation Rept Was Submitted.Two Supports in Question Received Required Code Insp & Were Found Acceptable
ML18152B6011998-12-31031 December 1998
Monthly Operating Repts for Dec 1998 for Surry Power Station,Units 1 & 2.With 990115 Ltr
ML18152B5781998-12-16016 December 1998
LER 98-014-00:on 981126,manual Reactor Trip in Response to Main Feedwater Regulating Valve Failure Occurred.Caused by Dislocation of Retaining Clip in Positioner.Control Room Operators Placed Unit in Safe,Shutdown Condition
ML18152B5811998-12-16016 December 1998
LER 98-013-00:on 981122,turbine/reactor Trip on High SG Level Occurred.Caused by Instrument Failure.Control Room Operators Placed Unit in Safe,Shutdown Condition
ML18152B7121998-12-0404 December 1998
LER 98-S01-00:on 981105,noted Failure to Deactivate Station Access Badge.Caused by Human Error.Licensee Will Now Deactivate Station Badges Before Clearance Is Revoked & Process for Badge Deactivations Have Been Strengthened
ML18152B7041998-12-0101 December 1998
LER 98-012-00:on 981102,noted That EDGs Were Concurrently Inoperable.Caused by Required Testing Per TS 3.16.B.1.a.2. Redundant EDG Was Returned to Svc within Two Hour Period, Following Satisfactory Testing.With 981201 Ltr
ML18152B7081998-11-30030 November 1998
Rev 0 to COLR for Surry 1 Cycle 16,Pattern Un.
ML18152B5721998-11-30030 November 1998
Monthly Operating Repts for Nov 1998 for Surry Power Station,Units 1 & 2.With 981214 Ltr
ML18152B6161998-11-0606 November 1998
LER 98-011-00:on 981008,diesel Driven Fire Pump Failed to Start During Performance of Monthly Operability Test.Caused by Faulty Overspeed Trip Device Failure.Diesel Driven Fire Pump Declared Inoperable
ML18152B6241998-10-31031 October 1998
Monthly Operating Repts for Oct 1998 for Surry Power Station Units 1 & 2.With 981111 Ltr
ML18152B6081998-10-23023 October 1998
LER 98-010-01:on 980715,intake Canal Level Probes Were Inoperable Due to Marine Growth.Caused by Design of Canal Level Instrumentation.Canal Level Probes Will Continue to Be Monitored More Closely
ML18152B6881998-09-30030 September 1998
Monthly Operating Repts for Sept 1998 for Surry Power Station Units 1 & 2.With 981012 Ltr
ML18153A3271998-08-31031 August 1998
Monthly Operating Repts for Aug 1998 for Surry Power Station,Units 1 & 2
ML18152B7811998-07-31031 July 1998
LER 98-010-00:on 980715,low Intake Canal Level Instrument Channel I Was Declared Inoperable to Allow Testing of Intake Canal Level Probe 1-CW-LE-102.Subject Probe Was Cleaned by Diver,Tested & Channel I Was Returned to Operable Status
ML18153A3161998-07-31031 July 1998
Monthly Operating Repts for July 1998 for Surry Power Station Units 1 & 2.W/980807 Ltr
ML18152B7621998-06-30030 June 1998
Monthly Operating Repts for June 1998 for Surry Power Station,Units 1 & 2.W/980707 Ltr
ML18153A2581998-06-0303 June 1998
LER 98-009-00:on 980509,nonisolable Leak of Reactor Coolant Pump Seal Injection Line Weld,Was Discovered.Caused by Lack of Fusion or Thermal Fatigue Coupled W/Vibration Stress Due to Loose Rod Hanger.Rcp Seal Injection Line Removed
ML20248F7441998-05-31031 May 1998
Reactor Vessel Working Group,Response to RAI Regarding Reactor Pressure Vessel Integrity
ML18153A3141998-05-31031 May 1998
Monthly Operating Repts for May 1998 for Surry Power Station,Units 1 & 2.W/980610
ML18152B8241998-05-22022 May 1998
LER 98-008-00:on 980228,auxiliary Ventilation Fans Were Noted in Condition Outside of Design Basis.Caused by Failure to Recognize Potential Impact of Certain Design Basis Accident Scenarios.No Corrective Actions Needed
ML18152B8161998-04-30030 April 1998
Monthly Operating Repts for Apr 1998 for Surry Power Station Units 1 & 2.W/980508 Ltr
ML18152B7951998-04-29029 April 1998
LER 98-007-00:on 980330,radiation Monitors Were Declared Inoperable.Caused by Change in Operating Temperature Range. Preplanned Alternate Method of Monitoring Was Initiated IAW TS Table 3.7-6
ML18153A2511998-04-22022 April 1998
LER 98-006-00:on 980324,unisolable Through Wall Leak of RCP Thermowell Was Noted.Cause of Leak Is Unknown.Rtd Will Be Replaced
ML18153A2521998-04-22022 April 1998
LER 98-005-01:on 980212,fire Watch Insp Exceeded One Hour. Caused by Lack of Attention to Detail by Individual Involved.Individual Involved Was Coached on Requirement to Perform Fire Watch Patrols within Required Time Frame
ML20217P9941998-04-0707 April 1998
Safety Evaluation Granting Licensee Third 10-yr Inservice Insp Program Relief Requests SR-018 - Sr-024
ML18153A2951998-03-31031 March 1998
Monthly Operating Repts for Mar 1998 for Sps,Units 1 & 2.W/ 980408 Ltr
ML18153A2391998-03-13013 March 1998
LER 98-005-00:on 980212,fire Watch Insp Frequency Exceeded One H Occurred.Category 2 Root Cause Evaluation Being Conducted to Determine Cause of Event.Station Deviation Issued
ML18153A2341998-03-0909 March 1998
LER 98-003-00:on 980226,no Procedural Guidance for Maintaining EDG Minimum Fuel Supply During Loop,Was Identified.Caused by Absence of Procedural Instructions. Deviation Rept Submitted to Document Deviating Condition
ML18153A2301998-03-0606 March 1998
LER 98-004-00:on 980206,fire Watch Was Released Prematurely Resulting in Violation of Ts.Caused by Inadequate Planning of Repair Activity.Work Orders Will Include Ref to Applicable Procedures Developed to Assist in Repairs
ML18153A2251998-03-0404 March 1998
LER 98-002-00:on 980202,automatic Turbine Trip Resulted in Automatic Reactor Trip.Caused Degraded Generator Voltage Regulator sub-component Failure.Placed Plant in Safe Hot SD & Replaced Intermittent Relay & Relay Socket
1999-09-30
[Table view] |
Text
POW 28-06-_01 NRC Form 366
- U.S. NUCLEAR REGULATORY COMMISSION (9-83,)
APPROVED 0MB NO. 3160,0104 EXPIRES: 8/31/88 LICENSEE EVENT REPORT (LER)
FACILITY NAME (1) DOCKET NUMBER (2) PAGE 131
- TITLE (41 Surry P9wer *station. Unit 1 !o I 5 I o Io I o 12 I 8 I O 1 loF O I 3 Inadequate R*eview of AFW Suoply FollowinQ a HELB in Safecrnards EVENT DATE (5) LER NUMBER (6) REPORT DATE (71 OTHER FACILITIES INVOLVED (Bl MONTH DAY YEAR YEAR FACILITY NAMES DOCKET NUMBER(SI oI 6 31 oa 1 al 1 .:.... ol 1 14 - oI o ol 1
- 3 I o aI 1 01s1010101 I I OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more of the folloV(inlJ) (11)
POWER MODE (9) 1*
N 20.402(b) 20.406(1)(1 )(I) -,__.
20.406(cl
&0.36(c){1)
~
50,73(11(2)(ivl 50.73(11(21(v) 73.71(bl 73.71(cl
=~== : : : : :;: -
LEVEL -
1 IO I O 50.38(cl(2) 50.73(1)(2l{viil . OTHER (Spacify in Abstr*ct
<101 20.405t1H1Hm ,__.
60.73(11(21(11 &0.73(1)(2)(vllll(A)
- bl/ow and in T*Kt, NRC Form 366A)
- 50,73(11{2){ill 50.73(11(2)(iiil -
LICENSEE CONTACT FOR THIS LER (121 50.73(11(21(vllll(BI 50.73(1)(21(,i)
NAME TELEPHONE NUMBER AREA CODE D. L. Benson, Stat~n Manager
. 81 01 4 31 51 7 I - I 31 11 8 Ill COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (131 CAUSE SYSTEM COMPONENT MANUFAC- SYSTEM COMPONENT. MANUFAC-TURER TURER I I I I I I I I I I I I I I I I I I I I I I I I I . I I I SUP~LEMENTAL REPORT EXPECTED (141 MONTH DAY YEAR EXPECTED SUBMISSION
- 1 YES /If v**. complete EXPECTED SUBMISSION DATE)
DATE (151 nl1 ~11 RIR ABSTRACT (Limit to 1400 spaces, i.B., approximately fifteen single-space typewritten lines} (161 On June 30, 1987 with Unit One at 100% power and Unit Two at 69%
power, *it was determined as a result of a company initiated
- safety system functional inspection, that a unique scenario could exist whereby auxiliary feedwater (AFW) {EIIS-BA} could be lost to*an operating unit. In the event of a high energy line break (HELB) in the main steam valve house (MSVH) of an operating unit, the AFW pumps {EIIS-P} on that unit could be rendered inoperable due to the harsh environment. With the unaffected unit at cold shutdown and only one of its motor driven AFW pumps available for cross connect to the affected unit, an assumed loss of offsite power, with a single failure could result in a loss of AFW to the affected unit.
Administrative controls have been established to ensure that .
when a unit is above 350 degrees/450 psig, two motor driven AFW pumps will be available from the other unit.
Engineeririg is reviewing the original s~fety analysis for the HELB (loss of AFW) to assess the adequacy of the evaluation and determine if additional corrective actions are required.
97oso~o001il i~~&eo PDR " PDR . . . _\ t s P*' p., ,
NRC Form 386 19-83)
POW 28-06~01 NRC Form 366A*
(9-83) U,S, NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION* APPROVED 0MB NO. 3150--0104 EXPIRES: 8/31/88 FACILITY NAME (11 DOCKET NUMBER (21 LER NUMBER (61 PAGE (31
?Urry Power Station, Unit 1 TEXT /ff tnOf9 _ , , . i& n,quintd, usa additional NRC Farm 366A 's/ (171 o Is Io Io I o 12 1a I o a11 - o I 114 - oI o -o I 2 oF o 13
- 1. 0 Description of the Event*
On June 30, 1987 with Unit One at 100% power and Unit Two at 69% power, it was determined as a result of a company initiated safety system functional inspection, that a unique scenario could exist whereby auxiliary feedwater * (AFW) {EIIS.:..BA} could be lost to an operating unit. In the event of a high energy line break (HELB) in the main steam valve house (MSVH) of an operating unit, the AFW pumps {EIIS-P} on that unit could be rendered inoperable due to the harsh environment. With the unaffected unit at cold .shutdown and only one of its motor driven AFW pumps available.*
for cross connect to the affected unit, an assumed loss of offsite power, with a single failure could result in a loss of AFW to' the affected unit.
The existing Technical Specification requires that only one motor driven AFW pump be available when the unit is at cold shutdown. This specification was developed in 1980 to provide assurance that AFW could be supplied from one unit to another in the event that the AFW pumps in one unit were destroyed by fire (Appendix R). The HELB event was not considered at that time, and Appendix R does not require cqnsideration of the single failure criterion.
This event is reportable in accordance with 10CFR
- 50. 73(a) (2) (v).
2.0 Safety Consequences and Implications This scenario is not analyzed in the UFSAR. How-ever, a main steam line break is a condition IV event and is of very low probability. Engineering is reviewing the original safety analysis to determine whether an unreviewed safety question exists.
Additionally, Functional Restoration Procedure H.l provides guidance in- the event of total loss of FW and AFW to the Steam Generators. Further protection is provided through an extensive non-destructive testing program of the main steam postulated break points in the main steam valve house .
. NRC FORM 366A (9-831
e e.pow .. 28-06-01 NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION 19-831 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION APPROVED 0MB NO. 3150-0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER 121 PAGE 13)
SURRY POWER STATION, UNIT 1 o 15 Io Io Io 12 I 810 81 7 - o 11 I 4 lo In nI ':l OF n h TEXT (If mo,e-,,.c. ia ,wqund, u* tldditioMI NRC Form 366A'1/ l17l 3.0 Cause Section l4B.5.1.7 of the UFSAR considers a HELB in the MSVH and assumes that should all three of the affected units AFW pumps become inoperable; redundancy is not lost because there are three AFW pumps* available from the unaffected unit. The UFSAR does not consider
.the possibility that only one AFW pump from the unaffected unit might be available when that unit is at cold shutdown.
4.0 Immediate Corrective Action Administrative controls have been established to ensure that when a unit is above 350 degrees/450 psig, two motor driven AFW pumps will be ,. available from the other unit.
5.0 Additional Corrective Actions Engineering is reviewing the original safety analysis for the HELB (loss of AFW) to assess the adequacy of the evaluation and to qetermine if additional corrective actions are required.
6.0 Actions Taken to Prevent Recurrence None.
7.0 Similar Events None 8.0 Manufacturer/Model Number N/A NRc FOAM 366.A 19-831
- U.S.GPO: 1986*0*824-538/455
e Vepco VIRGINIA ELECTRIC AND POWER COMPANY P, D. Box 315 Surry, Virginia 23883
. July 30, 1987 U.S. Nuclear Regulatory Commission Serial No: 87-016 Document Control Desk Docket No: 50-280 016 Phillips Building 50-281 Washington, D.C. 20555 License No: DPR-32 DPR-37 Gentlemen:
Pursuant to Surry Power Station Technical Specifications~ Virginia Electric and Power Company hereby submits the following Licensee Event Report for Surry Unit 1 and 2.
REPORT NUMBER 87-014-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by Safety Evaluation and Control.
Very truly yours, David L. Benson Station Manager Enclosure pc: Dr. J. Nelson Grace Regional Administrator Suite 2900 101 Marietta Street, NW Atlanta, Georgia 30323
