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| {{#Wiki_filter:~CATEGORY REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9705060022 DOC.DATE: 97/04/28 NOTARIZED: | | {{#Wiki_filter:~ CATEGORY REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) |
| NO DOCKET¹FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH.NAME AUTHOR AFFIL ATION ARBUCKLE,J.D. | | ACCESSION NBR:9705060022 DOC.DATE: 97/04/28 NOTARIZED: NO DOCKET ¹ FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFIL ATION ARBUCKLE,J.D. Washington Public Power Supply System BEMIS,M.R. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION |
| Washington Public Power Supply System BEMIS,M.R. | |
| Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION | |
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| ==SUBJECT:== | | ==SUBJECT:== |
| LER 97-005-00:on 970327,valid reactor scram signal received due to low water level condition during prepa ations for SRV testing.Caused by risks 5 consequences of decisions not completely identified.Res ored water level.W/970428 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR (ENCL I SIZE': TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES: RECIPIENT ID CODE/NAME PD4-2 PD INTERNAL: ACRS AEOD/SPD/RRAB NRR/DE/ECGB NRR/DE/EMEB NRR/DRCH/HICB NRR/DRCH/HQMB NRR/DSSA/SPLB RES/DET/EIB EXTERNAL: L ST LOBBY WARD NOAC POORE,W.NRC PDR COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME COLBURNgT AB ILE CENTE NRR~EE NRR/DRCH/HHFB NRR/DRCH;HOLB NRR/DRPM/PECB NRR/DSSA/SRXB RGN4 FILE 01 LITCO BPYCE,J H NOAC QUEENER,DS NUDOCS FULL TXT COPIES LTTR~a~CL 1 1 2 2 1 1 1 1 1 1 1 1 1 1 1 1 1.1 1 1 1 1 1 1 D 0 U N NOTE TO ALL"RIDS" RECIPZENTS: | | LER 97-005-00:on 970327,valid reactor scram signal received due to low water level condition during prepa ations for SRV testing. Caused by risks 5 consequences of decisions not completely identified.Res ored water level.W/970428 ltr. |
| PLEASE HELP US TO REDUCE WASTE.TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPZES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 25 WASHINGTON PUBLIC POWER SUPPLY SYSTEM PO.Box 968~Richlaail, iVashington 99352-0968 April 28, 1997 G02-97-082 Docket No.50-397 U.S.Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.20555 Gentlemen: | | DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ( ENCL I SIZE': |
| | TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc. |
| | NOTES: |
| | RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ~a~CL PD4-2 PD 1 1 COLBURNgT 1 1 INTERNAL: ACRS 1 1 AB 2 2 AEOD/SPD/RRAB 1 1 ILE CENTE 1 1 NRR/DE/ECGB 1 1 NRR ~EE 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH; HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 RES/DET/EIB 1 1 RGN4 FILE 01 1 . 1 D |
| | EXTERNAL: L ST LOBBY WARD 1 1 LITCO BPYCE,J H 1 1 NOAC POORE,W. 1 1 NOAC QUEENER,DS 1 1 0 NRC PDR 1 1 NUDOCS FULL TXT 1 1 U |
| | N NOTE TO ALL "RIDS" RECIPZENTS: |
| | PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR OR REDUCE THE NUMBER OF COPZES RECEIVED BY ORGANIZATION REMOVED FROM DISTRIBUTION LISTS YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 25 |
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| | WASHINGTON PUBLIC POWER SUPPLY SYSTEM PO. Box 968 ~ Richlaail, iVashington 99352-0968 April 28, 1997 G02-97-082 Docket No. 50-397 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen: |
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| ==Subject:== | | ==Subject:== |
| NUCLEAR PLANT WNP-2, OPERATING LICENSE NPF-21 LICENSEE EVENT REPORT NO.97-005-00 Transmitted herewith is Licensee Event Report No.97-005-00 for WNP-2.This report is submitted pursuant to 10 CFR 50.73 and discusses the items of reportability, corrective action taken, and action to preclude recurrence. | | NUCLEAR PLANT WNP-2, OPERATING LICENSE NPF-21 LICENSEE EVENT REPORT NO. 97-005-00 Transmitted herewith is Licensee Event Report No. 97-005-00 for WNP-2. This report is submitted pursuant to 10 CFR 50.73 and discusses the items of reportability, corrective action taken, and action to preclude recurrence. |
| Should you have any questions or desire additional information regarding this matter, please call'e or D.A.Swank at (509)377-4563.tfully P.V'ce President, Nuclear Operations Mail Drop PE23 Enclosure cc: EW Merschoff-NRC RIV KE Perkins, Jr.-NRC RIV, Walnut Creek Field Office TG Colburn-NRR NRC Senior Resident Inspector-927N DL Williams-BPA/399 PD Robinson-Winston 8: Strawn 9705060022 970428 PDR ADOCK 05000397 PDR (J g IIIIIIIIIIIIIIIIIIIIIIIIIII IIIIIIIIIIII l | | 'e Should you have any questions or desire additional information regarding this matter, please call or D.A. Swank at (509) 377-4563. |
| LICENSEE EVENT REPORT tLER)FACE(IY RAISE (ll DOCKET IRSES(R<a I'ACE (Sl Washington Nuclear Plant-Unit 2 0 5 0 0 0 3 9 7 1 CF 4&9<41 VALID REACTOR SCRAM SIGNAL WHILE SHUTDOWN WITH CONTROL RODS FULLY INSERTED DUE TO REACHING LOW WATER LEVEL SETPOINT DURING PREPARATIONS FOR SAFETY RELIEF VALVE TESTING 03 EVERT DAlS (EI 27 97 97 LER IRNRNR (el SEIXRNllAL fss(SER 0 0 5 0 0 04 28 97<ACR(IY RAIEES N/A OT)KR FACE(TXS SIVOLVED SD DOCKET)ARISE)Eel 0 5 0 0 0 5 0 0 0 0 Ties R(FORT IS NJSMRlU)NISOA)<T TO THs REOWIO(WIS Of 10 (f(s 011 FO(VER LEVEL (1(S 3 I 20.402(b)I 20.405(e)(l)(i) | | tfully P. |
| I 20.405(s)(1)QN P P P I 20.405(s)(I)(B) 20.405(s)(1)(iv)'c'N I'c I 20.405c I 60.36(c)(1) | | V'ce President, Nuclear Operations Mail Drop PE23 Enclosure cc: EW Merschoff - NRC RIV KE Perkins, Jr. - NRC RIV, Walnut Creek Field Office TG Colburn - NRR NRC Senior Resident Inspector - 927N DL Williams - BPA/399 PD Robinson - Winston 8: Strawn 9705060022 970428 PDR ADOCK 05000397 PDR (J g IIIIIIIIIIII IIIIIIIIIIIIIIIIIIIIIIIIIII |
| I 50.36(c)(2)
| | |
| I 60.73(e)(2)(i) | | l LICENSEE EVENT REPORT tLER) |
| I I 50.73(s)(2)R I 50.73(e)(2)(w)) | | DOCKET IRSES(R <a I'ACE (Sl FACE(IY RAISE (ll Washington Nuclear Plant Unit 2 0 5 0 0 0 3 9 7 1 CF 4 |
| I X I 60.73(s)(2)(iv) | | &9<41 VALID REACTOR SCRAM SIGNAL WHILE SHUTDOWN WITH CONTROL RODS FULLY INSERTED DUE TO REACHING LOW WATER LEVEL SETPOINT DURING PREPARATIONS FOR SAFETY RELIEF VALVETESTING EVERT DAlS (EI LER IRNRNR (el OT)KR FACE(TXS SIVOLVED SD SEIXRNllAL DOCKET )ARISE)Eel |
| I 60.73(s)(2)(v) | | <ACR(IY RAIEES fss(SER N/A 0 5 0 0 0 03 27 97 97 0 0 5 0 0 04 28 97 0 5 0 0 0 Ties R(FORT IS NJSMRlU) NISOA)<T TO THs REOWIO(WIS Of 10 (f(s 011 20.402(b) 20.405c X I 60.73(s)(2)(iv) 73.71(b) 3 I I I I I I 20.405(e)(l)(i) I 60.36(c)(1) I 60.73(s)(2)(v) I 73.71(c) |
| I 50.73(e)(2)(vi) | | FO(VER LEVEL I 20.405(s)(1)QN I 50.36(c)(2) I 50.73(e)(2)(vi) I I DTNE(t (SPeofy b( Abet(est be<cvf e(xf (1(S i) Text, NAC Fo(m 366AI P P P I 20.405(s)(I)(B) I 60.73(e)(2)(i) I 60.73(e)(2)(v)wlA 20.405(s) (1) (iv) I I 50.73(s)(2)R I I 60.73(e)(2)(UN)B I 50.73(e)(2)(w)) I 50.73(e)(2)(x) |
| I 60.73(e)(2)(v)wlA I I 60.73(e)(2)(UN)B I 50.73(e)(2)(x) | | 'c 'N I'c L)CERE(Z CONTACT fQI T)ss L<R (la Z.D. Arbucklef Licensing Technical Specialist 509 377-4601 (CNEFLETE CNC LSR f fOR EACH CONF(NRRT AAINIE DNCRSND S( T)es R(FORT (lsl R(FOR<A<LE R(PORTA(KE To Ievs)s TO HVOS |
| I I 73.71(b)I 73.71(c)I I DTNE(t (SPeofy b(Abet(est be<cvf e(xf i)Text, NAC Fo(m 366AI L)CERE(Z CONTACT fQI T)ss L<R (la Z.D.Arbucklef Licensing Technical Specialist (CNEFLETE CNC LSR fOR EACH CONF(NRRT f AAINIE DNCRSND S(T)es R(FORT (lsl R(FOR<A<LE To Ievs)s 509 377-4601 R(PORTA(KE TO HVOS SINf LEMERTAL RESORT EXFECT(D (14(YES (if yse, comp(ete EXPECTED SUBM(SSIDN DATE);:i)4)'4~vaFd 4'>.IXI No EXPECTED SU Bhl SS(ON DATE (15)ABSTRACI')6)
| | ;:i)4)'4~vaFd 4 '>. |
| On March 27, 1997 at 1247 hours with the plant shutdown in Mode 3 (Hot Shutdown), a valid reactor scram signal was received due to a low water level condition during preparations for safety relief valve testing.The low level condition occurred following closure of a safety relief valve that had been open for pressure control.Closure of the valve resultel in a large drop in reactor pressure vessel water level.During the time the valve was open, vessel pressure decreased from 1030 psig to about 880 psig.When the valve was closed, vessel level droppeF(rapidly from an initial value of+61 inches to+20 inches tn approximately 20 seconds.Vessel level continued to decrease to+13 inches, the Reactor Vessel Water Level-Low (Level 3)setpoint, over the next 30 seconds.Plant operators responded immediately to the low level condition and promptly restored vessel water level to within normal limits.The scram signal was reset at 1249 hours.The root cause of this event was risks and consequences of decisions not completely identified and assessed.The attempt to set up for safety relief valve testing without a contingency plan contributed significantly to the event.Further corrective action includes coaching the Operations crew involved, developing expectations for plant shutdown schedules, and providing simulator training on safety relief valve actuation at normal level and pressure with the plant shutdown.This event posed nlo threat to the health and safety of the public or plant personnel.
| | SINfLEMERTAL RESORT EXFECT(D (14( EXPECTED SU Bhl SS(ON YES (if yse, comp(ete EXPECTED SUBM(SSIDN DATE) IXI No DATE (15) |
| The Reactor Protection System functioned as designed to cause a reactor scram signal and plant operator actions to recover level were prompt and correct.The lowest vessel level reached during the event was+13 inches, before being recovered to the normal operating level.All control rods were fully inserted throughout this event. | | ABSTRACI')6) |
| LICENSEE EVENT REPORT tLER)TEXT CONTINUATION FAClLITY NAHE (1)DOCKET RlNBER (2)~Washington Nuclear Plant-Unit 2 TEXT (17)05 0 0 0 3 9 7 97 005 0 0 OP 4 nD ri in On March 27, 1997 at 1247 hours with the plant in Mode 3 (Hot Shutdown)and approximately four hours after a'manual scram (LER 97-004-00), a valid reactor scram signal was received due to a low water level condition during preparations for safety relief valve testing.The low level condition occurred following closure of a safety relief valve that had been open for pressure control.Closure of the valve resulted in a larger-than-anticipated"shrink" in reactor pressure vessel water level.Prior to the event and in preparation for safety relief valve testing, auxiliary steam loads were secured and the Auxiliary Boiler[BLR]was placed in service to provide seal steam.Pressure had decreased to approximately 650 psig by the time all steam loads, with the exception of a sin"!e Reactor Feedwater (RKV)System[JB]pump, were secu'ed.Vessel level was raised to between+40 and+45 inches and the decision was made to close the Main Steam Isolation Valves (MSIVs)PSV]to increase reactor pressure.Prior to valve closure, a test run of the Reactor Core Isolation Cooling (RCIC)System[BN]was successfully performed to ensure that the system would be available if needed for level control.When the MSIVs were closed, vessel level was being controlled with a combination of Reactor Water Cleanup (RWCU)System[CE]blowdown flow and Control Rod Drive System[AA]injection flow.At the time, pressure was increasing at a rate of approximately five psig/minute and level was increasing approximately one inch/minute. | | On March 27, 1997 at 1247 hours with the plant shutdown in Mode 3 (Hot Shutdown), a valid reactor scram signal was received due to a low water level condition during preparations for safety relief valve testing. The low level condition occurred following closure of a safety relief valve that had been open for pressure control. |
| Approximately seven minutes after closing the MSIVs and with vessel pressure at approximately 700 psig, a Reactor Vessel Water Level-High (Level 8)trip signal was received.Control rod drive flow was reduced and RWCU system blowdown was maximized in an effort to slow the level increase.Reactor level was subsequently stabilized and slightly decreasing. | | Closure of the valve resultel in a large drop in reactor pressure vessel water level. |
| At the time the safety relief valve was opened, vessel level was approximately | | During the time the valve was open, vessel pressure decreased from 1030 psig to about 880 psig. When the valve was closed, vessel level droppeF( rapidly from an initial value of +61 inches to +20 inches tn approximately 20 seconds. Vessel level continued to decrease to +13 inches, the Reactor Vessel Water Level |
| +58 inches and reactor pressure was approximately 1030 psig.During the time the valve was open, vessel pressure decreased to about 880 psig.When the valve was closed, vessel level dropped rapidly from an initial value of+61 inches to+20 inches in approximately 20 seconds.Vessel level continued to decrease to+13 inches, the Reactor Vessel Water Level-Low (Level 3)setpoint, over the next 30 seconds.The RCIC system was manually started to raise water level, but not in time to avoid the scram signal.The system was started about 15 seconds prior to the scram signal.The lowest vessel level reached during the event was+13 inches, before being recovered to the normal operating level. | | - Low (Level 3) setpoint, over the next 30 seconds. |
| LICENSEE EVENT REPORT (LER)TEXT CONTINUATION FAClllTY NANE (1)DOCKET NQIBER (2)YfAII THXY (17)'Washington Nuclear Plant-Unit 2 05 0 0 0 3 9 7 97 005 0 0 3 o" 4 mm i rr ive Acti n Plant operators responded to the low level condition and promptly restored vessel water level to within normal limits.The scram signal was reset approximately two minutes after it was received.F herE I n 2.Pursuant to 10 CFR 50.73(a)(2)(iv) this event is reportable as,"Any event or condition that resulted in a manual or automatic actuation of any Engineered Safety Feature, including the Reactor Protection System."!During the evolution, some shrinkage in vessel water level was anticipated by the Operations crew.However, the crew did not anticipate the amount of vessel level decrease which occurred.3.Following a scram, pressure typically decreases to the 600 to 700 psig range.This decrease occurs regardless of rapid operator actions to secure auxiliary steam loads.Normal post-scram operations involve securing steam loads and then starting a cooldown.The RFW system pumps are the preferred source of water to the vessel, but draw too much steam to maintain a high pressure condition. | | Plant operators responded immediately to the low level condition and promptly restored vessel water level to within normal limits. The scram signal was reset at 1249 hours. |
| To maintain rated pressure it is usually necessary to close the MSIVs or Main Steam (MS)System Valve MS-V-146[V], which effectively removes the preferred feed and pressure control (turbine bypa s valves)sources.The RCIC system can then be used for level and pressure control.4.The decision to conduct the safety relief valve testing following a scram did not evaluate the crew members previous experience with the plant in these conditions. | | The root cause of this event was risks and consequences of decisions not completely identified and assessed. |
| No detailed contigency planning was conducted prior to establishing the plant conditions to perform safety relief valve testing.There is procedural guidance for conditions where high pressure is maintained following a scram.However, the crew did not adequately use this guidance for control of pressure and level with the MSIVs closed.This led to the high level condition and the inability to use the RCIC system for pressure and level control. | | The attempt to set up for safety relief valve testing without a contingency plan contributed significantly to the event. Further corrective action includes coaching the Operations crew involved, developing expectations for plant shutdown schedules, and providing simulator training on safety relief valve actuation at normal level and pressure with the plant shutdown. |
| LICENSEE EVENT REPORT tLER)TEXT CONTINUATION FAClLTTY NANE (1)THXT (G)Washington Nuclear Plant-Unit 2 DOCKET NNBER (Z)0 5 0 0 0 3 9 7 97',,'Cy MIMI-005 l%YSNM HVMMI 0 0 4 oF 4 The root cause of this event was risks and consequences of decisions not completely identified and assessed.A contributing cause was documents not followed.Safety relief valve testing is usually performed with the reactor on line at low power which allows for normal pressure and level control.In this event, the attempt to set up for this test without a contingency plan to perform this test following a scram contributed significantly to the transient. | | This event posed nlo threat to the health and safety of the public or plant personnel. The Reactor Protection System functioned as designed to cause a reactor scram signal and plant operator actions to recover level were prompt and correct. The lowest vessel level reached during the event was +13 inches, before being recovered to the normal operating level. All control rods were fully inserted throughout this event. |
| The effects of the transient may have been minimized had the Operations crew used applicable steps of the normal shutdown-to-hot shutdown procedure and maintained the RCIC system in operation prior to closing the MSIVs.II Expectations will be developed for plant shutdown schedules to provide guidance pertaining to actions that will be performed if a scram or other transient occurs prior to performance of a scheduled activity.Specific simulator training, unique to this event, will be performed. | | |
| The training will cover pressure and level control, during hot shutdown conditions, and the effects of safety relief valve actuation at normal level and pressure.The lessons learned from this event have been discussed with the Operations crews.The Operations crew involved was coached on maintaining focus of all critical plant parameters during transient conditions. | | LICENSEE EVENT REPORT tLER) TEXT CONTINUATION FAClLITY NAHE (1) DOCKET RlNBER (2) |
| mn f f n n The safety consequences of this event are minimal.At the time of the event, the reactor was in a shutdown condition. | | ~ Washington Nuclear Plant - Unit 2 05 0 0 0 3 9 7 97 005 0 0 OP 4 TEXT (17) nD ri in On March 27, 1997 at 1247 hours with the plant in Mode 3 (Hot Shutdown) and approximately four hours after a'manual scram (LER 97-004-00), a valid reactor scram signal was received due to a low water level condition during preparations for safety relief valve testing. The low level condition occurred following closure of a safety relief valve that had been open for pressure control. Closure of the valve resulted in a larger-than-anticipated "shrink" in reactor pressure vessel water level. |
| All control rods had already been fully inserted prior to the scram signal.The Reactor Protection System functioned as designed to cause a reactor scram signal and plant operator actions to recover level were prompt and correct.Therefore, the event posed no threat to the health and safety of the public or plant personnel.
| | Prior to the event and in preparation for safety relief valve testing, auxiliary steam loads were secured and the Auxiliary Boiler [BLR] was placed in service to provide seal steam. Pressure had decreased to approximately 650 psig by the time all steam loads, with the exception of a sin"!e Reactor Feedwater (RKV) System [JB] pump, were secu'ed. Vessel level was raised to between |
| imil Ev n There have been no recent similar events where not completely identifying and assessing the risks and consequences of decisions resulted in a low-level scram signal while in a shutdown condition.}} | | +40 and +45 inches and the decision was made to close the Main Steam Isolation Valves (MSIVs) PSV] to increase reactor pressure. Prior to valve closure, a test run of the Reactor Core Isolation Cooling (RCIC) System [BN] was successfully performed to ensure that the system would be available if needed for level control. |
| | When the MSIVs were closed, vessel level was being controlled with a combination of Reactor Water Cleanup (RWCU) System [CE] blowdown flow and Control Rod Drive System [AA] |
| | injection flow. At the time, pressure was increasing at a rate of approximately five psig/minute and level was increasing approximately one inch/minute. |
| | Approximately seven minutes after closing the MSIVs and with vessel pressure at approximately 700 psig, a Reactor Vessel Water Level - High (Level 8) trip signal was received. Control rod drive flow was reduced and RWCU system blowdown was maximized in an effort to slow the level increase. |
| | Reactor level was subsequently stabilized and slightly decreasing. At the time the safety relief valve was opened, vessel level was approximately +58 inches and reactor pressure was approximately 1030 psig. During the time the valve was open, vessel pressure decreased to about 880 psig. When the valve was closed, vessel level dropped rapidly from an initial value of +61 inches to +20 inches in approximately 20 seconds. Vessel level continued to decrease to +13 inches, the Reactor Vessel Water Level - Low (Level 3) setpoint, over the next 30 seconds. |
| | The RCIC system was manually started to raise water level, but not in time to avoid the scram signal. The system was started about 15 seconds prior to the scram signal. The lowest vessel level reached during the event was +13 inches, before being recovered to the normal operating level. |
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| | LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FAClllTY NANE (1) DOCKET NQIBER (2) |
| | YfAII |
| | 'Washington Nuclear Plant - Unit 2 05 0 0 0 3 9 7 97 005 0 0 3 o" 4 THXY (17) mm i rr ive Acti n Plant operators responded to the low level condition and promptly restored vessel water level to within normal limits. The scram signal was reset approximately two minutes after it was received. |
| | F herE I n Pursuant to 10 CFR 50.73(a)(2)(iv) this event is reportable as, "Any event or condition that resulted in a manual or automatic actuation of any Engineered Safety Feature, including the Reactor Protection System." |
| | ! |
| | : 2. During the evolution, some shrinkage in vessel water level was anticipated by the Operations crew. However, the crew did not anticipate the amount of vessel level decrease which occurred. |
| | : 3. Following a scram, pressure typically decreases to the 600 to 700 psig range. This decrease occurs regardless of rapid operator actions to secure auxiliary steam loads. |
| | Normal post-scram operations involve securing steam loads and then starting a cooldown. |
| | The RFW system pumps are the preferred source of water to the vessel, but draw too much steam to maintain a high pressure condition. |
| | To maintain rated pressure it is usually necessary to close the MSIVs or Main Steam (MS) |
| | System Valve MS-V-146 [V], which effectively removes the preferred feed and pressure control (turbine bypa s valves) sources. The RCIC system can then be used for level and pressure control. |
| | : 4. The decision to conduct the safety relief valve testing following a scram did not evaluate the crew members previous experience with the plant in these conditions. No detailed contigency planning was conducted prior to establishing the plant conditions to perform safety relief valve testing. |
| | There is procedural guidance for conditions where high pressure is maintained following a scram. However, the crew did not adequately use this guidance for control of pressure and level with the MSIVs closed. This led to the high level condition and the inability to use the RCIC system for pressure and level control. |
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| | LICENSEE EVENT REPORT tLER) TEXT CONTINUATION FAClLTTY NANE (1) DOCKET NNBER (Z) l%YSNM |
| | ',,'Cy MIMI HVMMI THXT (G) |
| | Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 97 -005 0 0 4 oF 4 The root cause of this event was risks and consequences of decisions not completely identified and assessed. A contributing cause was documents not followed. |
| | Safety relief valve testing is usually performed with the reactor on line at low power which allows for normal pressure and level control. In this event, the attempt to set up for this test without a contingency plan to perform this test following a scram contributed significantly to the transient. |
| | The effects of the transient may have been minimized had the Operations crew used applicable steps of the normal shutdown-to-hot shutdown procedure and maintained the RCIC system in operation prior to closing the MSIVs. |
| | II Expectations will be developed for plant shutdown schedules to provide guidance pertaining to actions that will be performed if a scram or other transient occurs prior to performance of a scheduled activity. |
| | Specific simulator training, unique to this event, will be performed. The training will cover pressure and level control, during hot shutdown conditions, and the effects of safety relief valve actuation at normal level and pressure. |
| | The lessons learned from this event have been discussed with the Operations crews. |
| | The Operations crew involved was coached on maintaining focus of all critical plant parameters during transient conditions. |
| | mn f f n n The safety consequences of this event are minimal. At the time of the event, the reactor was in a shutdown condition. All control rods had already been fully inserted prior to the scram signal. |
| | The Reactor Protection System functioned as designed to cause a reactor scram signal and plant operator actions to recover level were prompt and correct. Therefore, the event posed no threat to the health and safety of the public or plant personnel. |
| | imil Ev n There have been no recent similar events where not completely identifying and assessing the risks and consequences of decisions resulted in a low-level scram signal while in a shutdown condition.}} |
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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17292B7421999-07-20020 July 1999 LER 99-001-00:on 990628,ESF Signal Closed All Eight MSIVs While Plant Was Shutdown.Caused by Failure of Relay RPS-RLY-K10D.Subject Relay Was Replaced & Tested on 990630. with 990720 Ltr ML17292B4451998-10-27027 October 1998 LER 98-012-01:on 980715,failure to Comply with Requirements of TS SR 3.8.4.7 Was Noted.Caused by Inadequate Work Practices.Training Session Was Held with Personnel.With 981027 Ltr ML17284A7561998-09-0303 September 1998 LER 98-013-00:on 980805,ESF Actuations Were Noted Due to Deenergization of Vital Electrical Bus SM-8.Caused by Inadequate Direction in Troubleshooting Plan.Will Conduct Training for Engineering Personnel.With 980903 Ltr ML17284A7571998-09-0202 September 1998 LER 98-014-00:on 980807,completion of TS 3.8.1.F Required Shutdown Due to Inoperability of EDG-2 Was Noted.Caused by Degraded Voltage Regulator for DG-2.Replaced Voltage Regulator & Associated Scrs.With 980902 Ltr ML17284A7551998-09-0202 September 1998 LER 98-015-00:on 980808,discovered Reactor Coolant Pressure Boundary Leak During Shutdown Conditions.Caused by Leakage from Socket Weld (Fwb 63) on Elbow Connection.Failed Piping Connection Was Replaced.With 980902 Ltr ML17284A7311998-08-17017 August 1998 LER 98-012-00:on 980716,determined That 24-month SR 3.8.4.7 Had Not Been Fulfilled within Specified Frequency.Caused by Inadequate Work Practices.License Requested & Received Enforcement Discretion Re Battery Svc test.W/980817 Ltr ML17284A7121998-07-23023 July 1998 LER 98-006-01:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of 10CFR50,App R Calculations for High Impedance Faults.Caused by Inadequate Work Practices.Implemented Procedural Changes ML17284A6951998-07-17017 July 1998 LER 98-011-00:on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted. W/980717 Ltr ML17284A6961998-07-15015 July 1998 LER 98-010-00:on 980615,TS Required Shutdown Due to Inoperability of TIP Sys Isolation Valve Was Noted.Caused by Improper Installation of TIP Tubing.Reattached Affected Tubing & Inspected Other TIP tubing.W/980715 Ltr ML17284A6731998-07-0101 July 1998 LER 98-009-00:on 980606,nuclear Steam Supply Shutoff Sys Group 3 & 4 Isolations During Testing Was Noted.Caused by Procedural Deficiency.Counseled Individuals Involved in preparation.W/980701 Ltr ML17284A6651998-06-24024 June 1998 LER 98-007-00:on 980530,inadvertent Full Scram & Division 1 ECCS Injection Was Noted.Caused by Failure to Meet Mgt Work Practice Expectation When Encountering Deficient Procedure. Incident Review Board Convened to Review event.W/980624 Ltr ML17284A6641998-06-24024 June 1998 LER 98-008-00:on 980531,inadvertent Full Scram During RPV Leak Testing in Mode 4 Was Noted.Caused by Change in Mgt Techniques.Revised Procedures to Take Into Account Addl Water Head in Pressure Sensing lines.W/980624 Ltr ML17284A6631998-06-19019 June 1998 LER 98-006-00:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of App R Calculations for High Impedance Fault Analysis.Caused Indeterminate. Implemented Procedural Changes Involving Operator Action ML17284A6551998-06-0404 June 1998 LER 98-005-00:on 980506,potential for Failure of RHR Sys Valve to Close on Isolation Signal Was Noted.Caused by Design Deficiency.Caution Tag Was Placed on RHR-V-40 Control Switch to Inform Plant Operators of limitation.W/980604 Ltr ML17284A6421998-06-0101 June 1998 LER 98-004-00:on 980502,determined That Primary Containment Penetration Overcurrent Protection Does Not Meet Reg Guide 1.63 Requirements.Caused by Inadequate Design Changes. Installed Addl Fuse in RHR-MO-9 circuit.W/980601 Ltr ML17292B3281998-04-0909 April 1998 LER 98-002-00:on 980311,reactor Scram & Plant Transient Occurred,Due to Failed Closed Main Steam Isolation Valve. Caused by Loss of Pneumatic Actuating Supply Pressure. Problem Evaluation Request Written for Failure of MS-V-22D ML17292B3291998-04-0909 April 1998 LER 98-003-00:on 980311,WNP-2 Experienced SCRAM Signal as Result of Low Rpv.Caused by Less than post-SCRAM Operational Strategy for Resetting SCRAM Signal in Conditions.Changes in post-SCRAM Operational Strategy implemented.W/980409 Ltr ML17292B2661998-03-0404 March 1998 LER 98-001-00:on 980203,automatic Start of HPCS EDG Was Noted.Caused by Operator Error.Operations Crew Stabilized Plant at Approximately 75% Reactor Power & Investigation of Event Was initiated.W/980304 Ltr ML17292B1111997-11-10010 November 1997 LER 97-011-00:on 971010,HPCS Battery Charger Failed.Caused by Failure of a Phase Firing Control Circuit Board Due to Aging During 7 Yrs of Use.Hpcs Sys Was Immediately Declared inoperable.W/971110 Ltr ML17292B1151997-11-0707 November 1997 LER 97-010-00:on 970906,discovered That TS SR 3.4.5.1 for Identified Portion of RCS Total Leakage Would Not Be Able to Perform within Time Limits of SR 3.0.2.Caused by Inadequate Methods.Method of Meeting SR 3.4.5.1 Established ML17292B0641997-09-24024 September 1997 LER 97-004-01:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Indication of Entry Into Region a of power-to-flow Map.Caused by Inadequate Attention to Detail.Established Event Evaluation teams.W/970924 Ltr ML17292B0241997-08-18018 August 1997 LER 97-009-00:on 970717,discovered Error in Recently Performed Inservice Testing procedure,OSP-TIP/IST-R701. Caused by Procedure Inadequacy.Plant Procedure OSP/TIP/IST-R701 Will Be changed.W/970818 Ltr ML17292B0291997-08-15015 August 1997 LER 97-008-00:on 970716,wire Seal Used to Lock Containment Instrument Air Pressure Control valve,CIA-PCV-2B,found Not Intact.Cause of Misadjustment of CIA-PCV-2B Unknown.Event Will Be Communicated to Plant employees.W/970815 Ltr ML17292B0201997-08-15015 August 1997 LER 97-S01-00:on 970718,failure to Take Compensatory Measure for Inoperative Microwave Security Zone Occurred. Caused by Personnel Error.Training Will Be Conducted W/ Appropriate Members of Security force.W/970815 Ltr ML17292A9481997-07-23023 July 1997 LER 97-007-00:on 970611,voluntary Rept of Automatic Start of DG-1 & DG-2 Was Experienced.Caused by Undervoltage Condition on Electrical Busses SM-7 & SM-8.Circulating Water Pump CW-P-1C Control Switch Placed in pull-to-lock.W/970723 Ltr ML17292A9201997-06-26026 June 1997 LER 97-006-00:on 970527,non-performance of Surveillance Requirement 3.6.1.3.2 for Blind Fanges,Was Noted.Caused Because Misunderstanding of Intent of Specs.Added Five Structural Assemblies for SP.W/970626 Ltr ML17292A8331997-04-28028 April 1997 LER 97-004-00:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Entry Into Region a of power-to-flow Map Following Planned Trip of Single Mfp. Event Evaluation teams,established.W/970428 Ltr ML17292A8311997-04-28028 April 1997 LER 97-005-00:on 970327,valid Reactor Scram Signal Received Due to Low Water Level Condition During Preparations for SRV Testing.Caused by Risks & Consequences of Decisions Not Completely Identified.Restored Water level.W/970428 Ltr ML17292A8251997-04-21021 April 1997 LER 97-003-00:on 970320,notification of Noncompliance W/Ts as TS SRs for Response Time Testing Were Not Being Met for Specified Instrumentation in Rps,Pcis & Eccs.Requested Enforcement Discretion for One Time exemption.W/970421 Ltr ML17292A7431997-03-20020 March 1997 LER 97-002-00:on 970218,determined That Rod Block Monitor (RBM) Calibr Values Were Not Set IAW Tech Specs.Caused by Calibr Procedures Inadequacies.Revised & re-performed RBM Channel Calibr procedures.W/970330 Ltr ML17292A7401997-03-13013 March 1997 LER 97-001-00:on 970211,reactor Water Cleanup Sys Blowdown Flow Isolation Setpoint Was Slightly Above TS Allowable Valve Occurred Due to Calculation Error.Lds Fss LD-FS-15 LD-FS-16 Were Declared inoperable.W/970313 Ltr ML17292A6641997-01-22022 January 1997 LER 96-009-00:on 961220,miscalculation of Instantaneous Overcurrent Relay Settings Resulted in Inoperability of safety-related Equipment.Caused by Utilization of Inappropriate Design.Testing Was completed.W/970122 Ltr ML17292A6461997-01-0606 January 1997 LER 96-008-00:on 961205,failure to Comply with TS Action Requirement for Emergency Core Cooling Sys Actuation Instrumentation Occurred Due to Unidentified Inoperability Condition.Pmr initiated.W/970106 Ltr ML17292A6371996-12-19019 December 1996 LER 96-007-00:on 961122,electrical Breakers Were Not Seismically Qualified in Test/Disconnect Position.Circuit Breaker Mfg Did Not Consider Raced Out Breaker Position During Testing.Relocated Circuit breakers.W/961217 Ltr ML17292A4121996-08-0808 August 1996 LER 96-006-00:on 960709,average Power Range Monitor Rod Block Downscale Surveillance Not Performed Prior to Entry Into Mode 1.Caused by long-standing Misinterpretation of Requirements of Tss.Procedures revised.W/960808 Ltr ML17292A3801996-07-24024 July 1996 LER 96-004-00:on 960624,plant Was Manually Scrammed by Control Room Personnel Due to Reactor Water Level Transient Experienced During Testing of Digital Feedwater Sys.Caused by Programming Error.Sys Was corrected.W/960724 Ltr ML17292A3771996-07-24024 July 1996 LER 96-005-00:on 960624,determined Missed Surveillance Test Re Channel Check of Average Power Range Monitor.Caused by Inadequate Procedures.Revised Surveillance Procedure Re When APRM Checks Must Be performed.W/960724 Ltr ML17292A3641996-07-12012 July 1996 LER 96-003-00:on 960615,required Surveillance Test Not Performed When Required by TS 3.4.1.3.Caused by Inadequate Procedures.Implementing Surveillance Procedure & Reactor Plant Startup Procedures revised.W/960712 Ltr ML17292A3361996-06-20020 June 1996 LER 96-002-00:on 960504,critical Bus SM-8 Lost Power When Supply Breaker 3-8 Tripped.Caused by Personnel Error. Operators Counselled & Procedures revised.W/960620 Ltr ML17292A2861996-05-24024 May 1996 LER 96-001-00:on 960425,inadvertent ESF Actuations Occurred Due to Tripping of Temporary Power Supply to IN-3.Caused by Personnel Error.Operations Restored to IN-3 Loads & Reset ESF actuations.W/960524 Ltr ML17291B0891995-10-19019 October 1995 LER 95-011-00:on 950919,failed to Comply W/Ts SR for RCIC Sys Due to Analysis Deficiency That Resulted in Inadequate Surveillance Test Procedure.Surveillance Procedure Revised to Correct deficiency.W/951019 Ltr ML17291A9021995-07-0707 July 1995 LER 95-010-00:on 950609,HPCS DG Declared Inoperable Due to Discovery That TS Test Method Incomplete.Caused by Inadequate Testing Procedure.Test Procedure for HPCS DG Reviewed & Special Test Procedures written.W/950707 Ltr ML17291A9031995-07-0707 July 1995 LER 95-009-00:on 950607,inadvertent MSIV Closure Occurred During Surveillance Test Due to Poor Communication Between Test Team.Determined That MSIV Closure Not Valid Because Closure Not Triggered by Plant conditions.W/950707 Ltr ML17291A8501995-06-0808 June 1995 LER 95-006-01:on 950405,reactor Scram Occurred During Surveillance Testing Due to Protective Sys Relay Failure. Replaced Failed Relay Before Plant Startup ML17291A8101995-05-12012 May 1995 LER 95-008-00:on 940125,TS Wording Lead to Potential TS Violation.Caused by Lack of Clarity in Ts.Submitted Improved TS for Plant to Provide Addl clarity.W/950512 Ltr ML17291A7841995-05-0505 May 1995 LER 95-007-00:on 950222,emergency Diesel Start Occurred Due to Voltage Transient on BPA Grid.Confirmation Was Received at 17:51 H That Disturbance Had Originated in BPA Grid ML17291A7801995-05-0404 May 1995 LER 95-006-00:on 950405,main Turbine Trip Occurred During Performance of Surveillance Test Due to Protective Sys Relay Failed.Replaced Failed Relay Before Plant startup.W/950504 Ltr ML17291A7851995-05-0303 May 1995 LER 95-005-00:on 950222,inoperable IRM Had Been Relied Upon to Meet TS Requirements During Reactor Startup.Caused by Lack of Neutron Source to Test Instrumentation. Sys Knowledge Gained Will Be incorporated.W/950503 Ltr ML17291A7071995-03-25025 March 1995 LER 95-004-00:on 950226,malfunction in Main Turbine DEH Control Sys Caused All Four High Pressure Turbine Governor Valves to Rapidly Close.Caused by Blown Fuse.Suspected Faulty Circuit Card replaced.W/950325 Ltr ML17291A7011995-03-20020 March 1995 LER 95-002-00:on 950218,automatic Reactor Scram Occurred. Caused by Erroneous Positioning of Control During Performance of Scheduled Periodic Functional Test.Control repositioned.W/950320 Ltr 1999-07-20
[Table view] Category:RO)
MONTHYEARML17292B7421999-07-20020 July 1999 LER 99-001-00:on 990628,ESF Signal Closed All Eight MSIVs While Plant Was Shutdown.Caused by Failure of Relay RPS-RLY-K10D.Subject Relay Was Replaced & Tested on 990630. with 990720 Ltr ML17292B4451998-10-27027 October 1998 LER 98-012-01:on 980715,failure to Comply with Requirements of TS SR 3.8.4.7 Was Noted.Caused by Inadequate Work Practices.Training Session Was Held with Personnel.With 981027 Ltr ML17284A7561998-09-0303 September 1998 LER 98-013-00:on 980805,ESF Actuations Were Noted Due to Deenergization of Vital Electrical Bus SM-8.Caused by Inadequate Direction in Troubleshooting Plan.Will Conduct Training for Engineering Personnel.With 980903 Ltr ML17284A7571998-09-0202 September 1998 LER 98-014-00:on 980807,completion of TS 3.8.1.F Required Shutdown Due to Inoperability of EDG-2 Was Noted.Caused by Degraded Voltage Regulator for DG-2.Replaced Voltage Regulator & Associated Scrs.With 980902 Ltr ML17284A7551998-09-0202 September 1998 LER 98-015-00:on 980808,discovered Reactor Coolant Pressure Boundary Leak During Shutdown Conditions.Caused by Leakage from Socket Weld (Fwb 63) on Elbow Connection.Failed Piping Connection Was Replaced.With 980902 Ltr ML17284A7311998-08-17017 August 1998 LER 98-012-00:on 980716,determined That 24-month SR 3.8.4.7 Had Not Been Fulfilled within Specified Frequency.Caused by Inadequate Work Practices.License Requested & Received Enforcement Discretion Re Battery Svc test.W/980817 Ltr ML17284A7121998-07-23023 July 1998 LER 98-006-01:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of 10CFR50,App R Calculations for High Impedance Faults.Caused by Inadequate Work Practices.Implemented Procedural Changes ML17284A6951998-07-17017 July 1998 LER 98-011-00:on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted. W/980717 Ltr ML17284A6961998-07-15015 July 1998 LER 98-010-00:on 980615,TS Required Shutdown Due to Inoperability of TIP Sys Isolation Valve Was Noted.Caused by Improper Installation of TIP Tubing.Reattached Affected Tubing & Inspected Other TIP tubing.W/980715 Ltr ML17284A6731998-07-0101 July 1998 LER 98-009-00:on 980606,nuclear Steam Supply Shutoff Sys Group 3 & 4 Isolations During Testing Was Noted.Caused by Procedural Deficiency.Counseled Individuals Involved in preparation.W/980701 Ltr ML17284A6651998-06-24024 June 1998 LER 98-007-00:on 980530,inadvertent Full Scram & Division 1 ECCS Injection Was Noted.Caused by Failure to Meet Mgt Work Practice Expectation When Encountering Deficient Procedure. Incident Review Board Convened to Review event.W/980624 Ltr ML17284A6641998-06-24024 June 1998 LER 98-008-00:on 980531,inadvertent Full Scram During RPV Leak Testing in Mode 4 Was Noted.Caused by Change in Mgt Techniques.Revised Procedures to Take Into Account Addl Water Head in Pressure Sensing lines.W/980624 Ltr ML17284A6631998-06-19019 June 1998 LER 98-006-00:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of App R Calculations for High Impedance Fault Analysis.Caused Indeterminate. Implemented Procedural Changes Involving Operator Action ML17284A6551998-06-0404 June 1998 LER 98-005-00:on 980506,potential for Failure of RHR Sys Valve to Close on Isolation Signal Was Noted.Caused by Design Deficiency.Caution Tag Was Placed on RHR-V-40 Control Switch to Inform Plant Operators of limitation.W/980604 Ltr ML17284A6421998-06-0101 June 1998 LER 98-004-00:on 980502,determined That Primary Containment Penetration Overcurrent Protection Does Not Meet Reg Guide 1.63 Requirements.Caused by Inadequate Design Changes. Installed Addl Fuse in RHR-MO-9 circuit.W/980601 Ltr ML17292B3281998-04-0909 April 1998 LER 98-002-00:on 980311,reactor Scram & Plant Transient Occurred,Due to Failed Closed Main Steam Isolation Valve. Caused by Loss of Pneumatic Actuating Supply Pressure. Problem Evaluation Request Written for Failure of MS-V-22D ML17292B3291998-04-0909 April 1998 LER 98-003-00:on 980311,WNP-2 Experienced SCRAM Signal as Result of Low Rpv.Caused by Less than post-SCRAM Operational Strategy for Resetting SCRAM Signal in Conditions.Changes in post-SCRAM Operational Strategy implemented.W/980409 Ltr ML17292B2661998-03-0404 March 1998 LER 98-001-00:on 980203,automatic Start of HPCS EDG Was Noted.Caused by Operator Error.Operations Crew Stabilized Plant at Approximately 75% Reactor Power & Investigation of Event Was initiated.W/980304 Ltr ML17292B1111997-11-10010 November 1997 LER 97-011-00:on 971010,HPCS Battery Charger Failed.Caused by Failure of a Phase Firing Control Circuit Board Due to Aging During 7 Yrs of Use.Hpcs Sys Was Immediately Declared inoperable.W/971110 Ltr ML17292B1151997-11-0707 November 1997 LER 97-010-00:on 970906,discovered That TS SR 3.4.5.1 for Identified Portion of RCS Total Leakage Would Not Be Able to Perform within Time Limits of SR 3.0.2.Caused by Inadequate Methods.Method of Meeting SR 3.4.5.1 Established ML17292B0641997-09-24024 September 1997 LER 97-004-01:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Indication of Entry Into Region a of power-to-flow Map.Caused by Inadequate Attention to Detail.Established Event Evaluation teams.W/970924 Ltr ML17292B0241997-08-18018 August 1997 LER 97-009-00:on 970717,discovered Error in Recently Performed Inservice Testing procedure,OSP-TIP/IST-R701. Caused by Procedure Inadequacy.Plant Procedure OSP/TIP/IST-R701 Will Be changed.W/970818 Ltr ML17292B0291997-08-15015 August 1997 LER 97-008-00:on 970716,wire Seal Used to Lock Containment Instrument Air Pressure Control valve,CIA-PCV-2B,found Not Intact.Cause of Misadjustment of CIA-PCV-2B Unknown.Event Will Be Communicated to Plant employees.W/970815 Ltr ML17292B0201997-08-15015 August 1997 LER 97-S01-00:on 970718,failure to Take Compensatory Measure for Inoperative Microwave Security Zone Occurred. Caused by Personnel Error.Training Will Be Conducted W/ Appropriate Members of Security force.W/970815 Ltr ML17292A9481997-07-23023 July 1997 LER 97-007-00:on 970611,voluntary Rept of Automatic Start of DG-1 & DG-2 Was Experienced.Caused by Undervoltage Condition on Electrical Busses SM-7 & SM-8.Circulating Water Pump CW-P-1C Control Switch Placed in pull-to-lock.W/970723 Ltr ML17292A9201997-06-26026 June 1997 LER 97-006-00:on 970527,non-performance of Surveillance Requirement 3.6.1.3.2 for Blind Fanges,Was Noted.Caused Because Misunderstanding of Intent of Specs.Added Five Structural Assemblies for SP.W/970626 Ltr ML17292A8331997-04-28028 April 1997 LER 97-004-00:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Entry Into Region a of power-to-flow Map Following Planned Trip of Single Mfp. Event Evaluation teams,established.W/970428 Ltr ML17292A8311997-04-28028 April 1997 LER 97-005-00:on 970327,valid Reactor Scram Signal Received Due to Low Water Level Condition During Preparations for SRV Testing.Caused by Risks & Consequences of Decisions Not Completely Identified.Restored Water level.W/970428 Ltr ML17292A8251997-04-21021 April 1997 LER 97-003-00:on 970320,notification of Noncompliance W/Ts as TS SRs for Response Time Testing Were Not Being Met for Specified Instrumentation in Rps,Pcis & Eccs.Requested Enforcement Discretion for One Time exemption.W/970421 Ltr ML17292A7431997-03-20020 March 1997 LER 97-002-00:on 970218,determined That Rod Block Monitor (RBM) Calibr Values Were Not Set IAW Tech Specs.Caused by Calibr Procedures Inadequacies.Revised & re-performed RBM Channel Calibr procedures.W/970330 Ltr ML17292A7401997-03-13013 March 1997 LER 97-001-00:on 970211,reactor Water Cleanup Sys Blowdown Flow Isolation Setpoint Was Slightly Above TS Allowable Valve Occurred Due to Calculation Error.Lds Fss LD-FS-15 LD-FS-16 Were Declared inoperable.W/970313 Ltr ML17292A6641997-01-22022 January 1997 LER 96-009-00:on 961220,miscalculation of Instantaneous Overcurrent Relay Settings Resulted in Inoperability of safety-related Equipment.Caused by Utilization of Inappropriate Design.Testing Was completed.W/970122 Ltr ML17292A6461997-01-0606 January 1997 LER 96-008-00:on 961205,failure to Comply with TS Action Requirement for Emergency Core Cooling Sys Actuation Instrumentation Occurred Due to Unidentified Inoperability Condition.Pmr initiated.W/970106 Ltr ML17292A6371996-12-19019 December 1996 LER 96-007-00:on 961122,electrical Breakers Were Not Seismically Qualified in Test/Disconnect Position.Circuit Breaker Mfg Did Not Consider Raced Out Breaker Position During Testing.Relocated Circuit breakers.W/961217 Ltr ML17292A4121996-08-0808 August 1996 LER 96-006-00:on 960709,average Power Range Monitor Rod Block Downscale Surveillance Not Performed Prior to Entry Into Mode 1.Caused by long-standing Misinterpretation of Requirements of Tss.Procedures revised.W/960808 Ltr ML17292A3801996-07-24024 July 1996 LER 96-004-00:on 960624,plant Was Manually Scrammed by Control Room Personnel Due to Reactor Water Level Transient Experienced During Testing of Digital Feedwater Sys.Caused by Programming Error.Sys Was corrected.W/960724 Ltr ML17292A3771996-07-24024 July 1996 LER 96-005-00:on 960624,determined Missed Surveillance Test Re Channel Check of Average Power Range Monitor.Caused by Inadequate Procedures.Revised Surveillance Procedure Re When APRM Checks Must Be performed.W/960724 Ltr ML17292A3641996-07-12012 July 1996 LER 96-003-00:on 960615,required Surveillance Test Not Performed When Required by TS 3.4.1.3.Caused by Inadequate Procedures.Implementing Surveillance Procedure & Reactor Plant Startup Procedures revised.W/960712 Ltr ML17292A3361996-06-20020 June 1996 LER 96-002-00:on 960504,critical Bus SM-8 Lost Power When Supply Breaker 3-8 Tripped.Caused by Personnel Error. Operators Counselled & Procedures revised.W/960620 Ltr ML17292A2861996-05-24024 May 1996 LER 96-001-00:on 960425,inadvertent ESF Actuations Occurred Due to Tripping of Temporary Power Supply to IN-3.Caused by Personnel Error.Operations Restored to IN-3 Loads & Reset ESF actuations.W/960524 Ltr ML17291B0891995-10-19019 October 1995 LER 95-011-00:on 950919,failed to Comply W/Ts SR for RCIC Sys Due to Analysis Deficiency That Resulted in Inadequate Surveillance Test Procedure.Surveillance Procedure Revised to Correct deficiency.W/951019 Ltr ML17291A9021995-07-0707 July 1995 LER 95-010-00:on 950609,HPCS DG Declared Inoperable Due to Discovery That TS Test Method Incomplete.Caused by Inadequate Testing Procedure.Test Procedure for HPCS DG Reviewed & Special Test Procedures written.W/950707 Ltr ML17291A9031995-07-0707 July 1995 LER 95-009-00:on 950607,inadvertent MSIV Closure Occurred During Surveillance Test Due to Poor Communication Between Test Team.Determined That MSIV Closure Not Valid Because Closure Not Triggered by Plant conditions.W/950707 Ltr ML17291A8501995-06-0808 June 1995 LER 95-006-01:on 950405,reactor Scram Occurred During Surveillance Testing Due to Protective Sys Relay Failure. Replaced Failed Relay Before Plant Startup ML17291A8101995-05-12012 May 1995 LER 95-008-00:on 940125,TS Wording Lead to Potential TS Violation.Caused by Lack of Clarity in Ts.Submitted Improved TS for Plant to Provide Addl clarity.W/950512 Ltr ML17291A7841995-05-0505 May 1995 LER 95-007-00:on 950222,emergency Diesel Start Occurred Due to Voltage Transient on BPA Grid.Confirmation Was Received at 17:51 H That Disturbance Had Originated in BPA Grid ML17291A7801995-05-0404 May 1995 LER 95-006-00:on 950405,main Turbine Trip Occurred During Performance of Surveillance Test Due to Protective Sys Relay Failed.Replaced Failed Relay Before Plant startup.W/950504 Ltr ML17291A7851995-05-0303 May 1995 LER 95-005-00:on 950222,inoperable IRM Had Been Relied Upon to Meet TS Requirements During Reactor Startup.Caused by Lack of Neutron Source to Test Instrumentation. Sys Knowledge Gained Will Be incorporated.W/950503 Ltr ML17291A7071995-03-25025 March 1995 LER 95-004-00:on 950226,malfunction in Main Turbine DEH Control Sys Caused All Four High Pressure Turbine Governor Valves to Rapidly Close.Caused by Blown Fuse.Suspected Faulty Circuit Card replaced.W/950325 Ltr ML17291A7011995-03-20020 March 1995 LER 95-002-00:on 950218,automatic Reactor Scram Occurred. Caused by Erroneous Positioning of Control During Performance of Scheduled Periodic Functional Test.Control repositioned.W/950320 Ltr 1999-07-20
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17284A9001999-10-31031 October 1999 Rev 0 to COLR 99-15, WNP-2 Cycle 15,COLR GO2-99-177, LER 99-S01-00:on 990903,failure to Take Compensatory Measure within Required Time Upon Failure of Isolation Zone Microwave Unit,Was Noted.Caused by Personnel Error.Provided Refresher Training on Compensatory Measures.With1999-10-0101 October 1999 LER 99-S01-00:on 990903,failure to Take Compensatory Measure within Required Time Upon Failure of Isolation Zone Microwave Unit,Was Noted.Caused by Personnel Error.Provided Refresher Training on Compensatory Measures.With ML17284A8941999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for WNP-2.With 991012 Ltr ML17284A8801999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for WNP-2.With 990910 Ltr ML17284A8691999-07-31031 July 1999 Monthly Operating Rept for July 1999 for WNP-2.With 990813 Ltr ML17292B7421999-07-20020 July 1999 LER 99-001-00:on 990628,ESF Signal Closed All Eight MSIVs While Plant Was Shutdown.Caused by Failure of Relay RPS-RLY-K10D.Subject Relay Was Replaced & Tested on 990630. with 990720 Ltr ML17292B7271999-06-30030 June 1999 Monthly Operating Rept for June 1999 for WNP-2.With 990707 Ltr ML17292B6961999-05-31031 May 1999 Monthly Operating Repts for May 1999 for WNP-2.With 990608 Ltr ML17292B6641999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for WNP-2.With 990507 Ltr ML17292B6391999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for WNP-2.With 990413 Ltr ML17292B5871999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for WNP-2.With 990311 Ltr ML17292B5571999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for WNP-2.With 990210 Ltr ML17292B5621999-01-31031 January 1999 Rev 1 to COLR 98-14, WNP-2 Cycle 14 Colr. ML17292B5341999-01-15015 January 1999 Part 21 Rept Re Incorrect Modeling of BWR Lower Plenum Vol in Bison.Defect Applies Only to Reload Fuel Assemblies Currently in Operation at WNP-2.BISON Code Model for WNP-2 Has Been Revised to Correct Error ML17292B5331999-01-15015 January 1999 Part 21 Rept Re XL-S96 CPR Correlation for SVEA-96 Fuel. Defect Applies Only to WNP-2,during Cycles 12,13 & 14 Operation.Evaluations of Defect Performed by ABB-CE ML17292B4791998-12-31031 December 1998 Washington Public Power Supply Sys 1998 Annual Rept. with 981215 Ltr ML17292B5351998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for WNP-2.With 990112 Ltr ML17292B5741998-12-31031 December 1998 WNP-2 1998 Annual Operating Rept. with 990225 Ltr ML17284A8231998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for WNP-2.With 981207 Ltr ML17284A8081998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for WNP-2.With 981110 Ltr ML17292B4451998-10-27027 October 1998 LER 98-012-01:on 980715,failure to Comply with Requirements of TS SR 3.8.4.7 Was Noted.Caused by Inadequate Work Practices.Training Session Was Held with Personnel.With 981027 Ltr ML17284A7831998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for WNP-2.With 981007 Ltr ML17284A7491998-09-10010 September 1998 WNP-2 Inservice Insp Summary Rept for Refueling Outage RF13 Spring,1998. ML17284A7561998-09-0303 September 1998 LER 98-013-00:on 980805,ESF Actuations Were Noted Due to Deenergization of Vital Electrical Bus SM-8.Caused by Inadequate Direction in Troubleshooting Plan.Will Conduct Training for Engineering Personnel.With 980903 Ltr ML17284A7571998-09-0202 September 1998 LER 98-014-00:on 980807,completion of TS 3.8.1.F Required Shutdown Due to Inoperability of EDG-2 Was Noted.Caused by Degraded Voltage Regulator for DG-2.Replaced Voltage Regulator & Associated Scrs.With 980902 Ltr ML17284A7551998-09-0202 September 1998 LER 98-015-00:on 980808,discovered Reactor Coolant Pressure Boundary Leak During Shutdown Conditions.Caused by Leakage from Socket Weld (Fwb 63) on Elbow Connection.Failed Piping Connection Was Replaced.With 980902 Ltr ML17284A7681998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for WNP-2.With 980915 Ltr ML17284A7311998-08-17017 August 1998 LER 98-012-00:on 980716,determined That 24-month SR 3.8.4.7 Had Not Been Fulfilled within Specified Frequency.Caused by Inadequate Work Practices.License Requested & Received Enforcement Discretion Re Battery Svc test.W/980817 Ltr ML17284A7261998-07-31031 July 1998 Monthly Operating Rept for July 1998 for WNP-2.W/980810 Ltr ML17284A7121998-07-23023 July 1998 LER 98-006-01:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of 10CFR50,App R Calculations for High Impedance Faults.Caused by Inadequate Work Practices.Implemented Procedural Changes ML17284A6951998-07-17017 July 1998 LER 98-011-00:on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted. W/980717 Ltr ML17284A6961998-07-15015 July 1998 LER 98-010-00:on 980615,TS Required Shutdown Due to Inoperability of TIP Sys Isolation Valve Was Noted.Caused by Improper Installation of TIP Tubing.Reattached Affected Tubing & Inspected Other TIP tubing.W/980715 Ltr ML17284A6731998-07-0101 July 1998 LER 98-009-00:on 980606,nuclear Steam Supply Shutoff Sys Group 3 & 4 Isolations During Testing Was Noted.Caused by Procedural Deficiency.Counseled Individuals Involved in preparation.W/980701 Ltr ML17284A6751998-06-30030 June 1998 Ro:On 980617,flooding of RB Northeast Stairwell with Consequential Flooding of Two ECCS Pump Rooms.Caused by Inadequate Fire Protection Sys Design.Pumped Out Water from Affected Areas to Point Below Berm Areas of Pump Rooms ML17284A6641998-06-24024 June 1998 LER 98-008-00:on 980531,inadvertent Full Scram During RPV Leak Testing in Mode 4 Was Noted.Caused by Change in Mgt Techniques.Revised Procedures to Take Into Account Addl Water Head in Pressure Sensing lines.W/980624 Ltr ML17284A6651998-06-24024 June 1998 LER 98-007-00:on 980530,inadvertent Full Scram & Division 1 ECCS Injection Was Noted.Caused by Failure to Meet Mgt Work Practice Expectation When Encountering Deficient Procedure. Incident Review Board Convened to Review event.W/980624 Ltr ML17284A6631998-06-19019 June 1998 LER 98-006-00:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of App R Calculations for High Impedance Fault Analysis.Caused Indeterminate. Implemented Procedural Changes Involving Operator Action ML17284A6551998-06-0404 June 1998 LER 98-005-00:on 980506,potential for Failure of RHR Sys Valve to Close on Isolation Signal Was Noted.Caused by Design Deficiency.Caution Tag Was Placed on RHR-V-40 Control Switch to Inform Plant Operators of limitation.W/980604 Ltr ML17284A6421998-06-0101 June 1998 LER 98-004-00:on 980502,determined That Primary Containment Penetration Overcurrent Protection Does Not Meet Reg Guide 1.63 Requirements.Caused by Inadequate Design Changes. Installed Addl Fuse in RHR-MO-9 circuit.W/980601 Ltr ML17284A6491998-05-31031 May 1998 Rev 0 to COLR 98-14, WNP-2,Cycle 14 Colr. ML17292B4031998-05-31031 May 1998 Monthly Operating Rept for May 1998 for WNP-2.W/980608 Ltr ML17292B3921998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for WNP-2.W/980513 Ltr ML17292B3291998-04-0909 April 1998 LER 98-003-00:on 980311,WNP-2 Experienced SCRAM Signal as Result of Low Rpv.Caused by Less than post-SCRAM Operational Strategy for Resetting SCRAM Signal in Conditions.Changes in post-SCRAM Operational Strategy implemented.W/980409 Ltr ML17292B3281998-04-0909 April 1998 LER 98-002-00:on 980311,reactor Scram & Plant Transient Occurred,Due to Failed Closed Main Steam Isolation Valve. Caused by Loss of Pneumatic Actuating Supply Pressure. Problem Evaluation Request Written for Failure of MS-V-22D ML17292B3371998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for WNP-2.W/980409 Ltr ML17292B2641998-03-0404 March 1998 Performance Self Assessment,WNP-2. ML17292B2661998-03-0404 March 1998 LER 98-001-00:on 980203,automatic Start of HPCS EDG Was Noted.Caused by Operator Error.Operations Crew Stabilized Plant at Approximately 75% Reactor Power & Investigation of Event Was initiated.W/980304 Ltr ML17292B2911998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for WNP-2.W/980313 Ltr ML17284A7971998-02-17017 February 1998 Rev 28 to Operational QA Program Description, WPPSS-QA-004.With Proposed Rev 29 ML17292B3591998-02-12012 February 1998 WNP-2 Cycle 14 Reload Design Rept. 1999-09-30
[Table view] |
Text
~ CATEGORY REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9705060022 DOC.DATE: 97/04/28 NOTARIZED: NO DOCKET ¹ FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFIL ATION ARBUCKLE,J.D. Washington Public Power Supply System BEMIS,M.R. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION
SUBJECT:
LER 97-005-00:on 970327,valid reactor scram signal received due to low water level condition during prepa ations for SRV testing. Caused by risks 5 consequences of decisions not completely identified.Res ored water level.W/970428 ltr.
DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ( ENCL I SIZE':
TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.
NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ~a~CL PD4-2 PD 1 1 COLBURNgT 1 1 INTERNAL: ACRS 1 1 AB 2 2 AEOD/SPD/RRAB 1 1 ILE CENTE 1 1 NRR/DE/ECGB 1 1 NRR ~EE 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH; HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 RES/DET/EIB 1 1 RGN4 FILE 01 1 . 1 D
EXTERNAL: L ST LOBBY WARD 1 1 LITCO BPYCE,J H 1 1 NOAC POORE,W. 1 1 NOAC QUEENER,DS 1 1 0 NRC PDR 1 1 NUDOCS FULL TXT 1 1 U
N NOTE TO ALL "RIDS" RECIPZENTS:
PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR OR REDUCE THE NUMBER OF COPZES RECEIVED BY ORGANIZATION REMOVED FROM DISTRIBUTION LISTS YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 25
WASHINGTON PUBLIC POWER SUPPLY SYSTEM PO. Box 968 ~ Richlaail, iVashington 99352-0968 April 28, 1997 G02-97-082 Docket No. 50-397 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:
Subject:
NUCLEAR PLANT WNP-2, OPERATING LICENSE NPF-21 LICENSEE EVENT REPORT NO. 97-005-00 Transmitted herewith is Licensee Event Report No. 97-005-00 for WNP-2. This report is submitted pursuant to 10 CFR 50.73 and discusses the items of reportability, corrective action taken, and action to preclude recurrence.
'e Should you have any questions or desire additional information regarding this matter, please call or D.A. Swank at (509) 377-4563.
tfully P.
V'ce President, Nuclear Operations Mail Drop PE23 Enclosure cc: EW Merschoff - NRC RIV KE Perkins, Jr. - NRC RIV, Walnut Creek Field Office TG Colburn - NRR NRC Senior Resident Inspector - 927N DL Williams - BPA/399 PD Robinson - Winston 8: Strawn 9705060022 970428 PDR ADOCK 05000397 PDR (J g IIIIIIIIIIII IIIIIIIIIIIIIIIIIIIIIIIIIII
l LICENSEE EVENT REPORT tLER)
DOCKET IRSES(R <a I'ACE (Sl FACE(IY RAISE (ll Washington Nuclear Plant Unit 2 0 5 0 0 0 3 9 7 1 CF 4
&9<41 VALID REACTOR SCRAM SIGNAL WHILE SHUTDOWN WITH CONTROL RODS FULLY INSERTED DUE TO REACHING LOW WATER LEVEL SETPOINT DURING PREPARATIONS FOR SAFETY RELIEF VALVETESTING EVERT DAlS (EI LER IRNRNR (el OT)KR FACE(TXS SIVOLVED SD SEIXRNllAL DOCKET )ARISE)Eel
<ACR(IY RAIEES fss(SER N/A 0 5 0 0 0 03 27 97 97 0 0 5 0 0 04 28 97 0 5 0 0 0 Ties R(FORT IS NJSMRlU) NISOA)<T TO THs REOWIO(WIS Of 10 (f(s 011 20.402(b) 20.405c X I 60.73(s)(2)(iv) 73.71(b) 3 I I I I I I 20.405(e)(l)(i) I 60.36(c)(1) I 60.73(s)(2)(v) I 73.71(c)
FO(VER LEVEL I 20.405(s)(1)QN I 50.36(c)(2) I 50.73(e)(2)(vi) I I DTNE(t (SPeofy b( Abet(est be<cvf e(xf (1(S i) Text, NAC Fo(m 366AI P P P I 20.405(s)(I)(B) I 60.73(e)(2)(i) I 60.73(e)(2)(v)wlA 20.405(s) (1) (iv) I I 50.73(s)(2)R I I 60.73(e)(2)(UN)B I 50.73(e)(2)(w)) I 50.73(e)(2)(x)
'c 'N I'c L)CERE(Z CONTACT fQI T)ss L<R (la Z.D. Arbucklef Licensing Technical Specialist 509 377-4601 (CNEFLETE CNC LSR f fOR EACH CONF(NRRT AAINIE DNCRSND S( T)es R(FORT (lsl R(FOR<A<LE R(PORTA(KE To Ievs)s TO HVOS
- i)4)'4~vaFd 4 '>.
SINfLEMERTAL RESORT EXFECT(D (14( EXPECTED SU Bhl SS(ON YES (if yse, comp(ete EXPECTED SUBM(SSIDN DATE) IXI No DATE (15)
ABSTRACI')6)
On March 27, 1997 at 1247 hours0.0144 days <br />0.346 hours <br />0.00206 weeks <br />4.744835e-4 months <br /> with the plant shutdown in Mode 3 (Hot Shutdown), a valid reactor scram signal was received due to a low water level condition during preparations for safety relief valve testing. The low level condition occurred following closure of a safety relief valve that had been open for pressure control.
Closure of the valve resultel in a large drop in reactor pressure vessel water level.
During the time the valve was open, vessel pressure decreased from 1030 psig to about 880 psig. When the valve was closed, vessel level droppeF( rapidly from an initial value of +61 inches to +20 inches tn approximately 20 seconds. Vessel level continued to decrease to +13 inches, the Reactor Vessel Water Level
- Low (Level 3) setpoint, over the next 30 seconds.
Plant operators responded immediately to the low level condition and promptly restored vessel water level to within normal limits. The scram signal was reset at 1249 hours0.0145 days <br />0.347 hours <br />0.00207 weeks <br />4.752445e-4 months <br />.
The root cause of this event was risks and consequences of decisions not completely identified and assessed.
The attempt to set up for safety relief valve testing without a contingency plan contributed significantly to the event. Further corrective action includes coaching the Operations crew involved, developing expectations for plant shutdown schedules, and providing simulator training on safety relief valve actuation at normal level and pressure with the plant shutdown.
This event posed nlo threat to the health and safety of the public or plant personnel. The Reactor Protection System functioned as designed to cause a reactor scram signal and plant operator actions to recover level were prompt and correct. The lowest vessel level reached during the event was +13 inches, before being recovered to the normal operating level. All control rods were fully inserted throughout this event.
LICENSEE EVENT REPORT tLER) TEXT CONTINUATION FAClLITY NAHE (1) DOCKET RlNBER (2)
~ Washington Nuclear Plant - Unit 2 05 0 0 0 3 9 7 97 005 0 0 OP 4 TEXT (17) nD ri in On March 27, 1997 at 1247 hours0.0144 days <br />0.346 hours <br />0.00206 weeks <br />4.744835e-4 months <br /> with the plant in Mode 3 (Hot Shutdown) and approximately four hours after a'manual scram (LER 97-004-00), a valid reactor scram signal was received due to a low water level condition during preparations for safety relief valve testing. The low level condition occurred following closure of a safety relief valve that had been open for pressure control. Closure of the valve resulted in a larger-than-anticipated "shrink" in reactor pressure vessel water level.
Prior to the event and in preparation for safety relief valve testing, auxiliary steam loads were secured and the Auxiliary Boiler [BLR] was placed in service to provide seal steam. Pressure had decreased to approximately 650 psig by the time all steam loads, with the exception of a sin"!e Reactor Feedwater (RKV) System [JB] pump, were secu'ed. Vessel level was raised to between
+40 and +45 inches and the decision was made to close the Main Steam Isolation Valves (MSIVs) PSV] to increase reactor pressure. Prior to valve closure, a test run of the Reactor Core Isolation Cooling (RCIC) System [BN] was successfully performed to ensure that the system would be available if needed for level control.
When the MSIVs were closed, vessel level was being controlled with a combination of Reactor Water Cleanup (RWCU) System [CE] blowdown flow and Control Rod Drive System [AA]
injection flow. At the time, pressure was increasing at a rate of approximately five psig/minute and level was increasing approximately one inch/minute.
Approximately seven minutes after closing the MSIVs and with vessel pressure at approximately 700 psig, a Reactor Vessel Water Level - High (Level 8) trip signal was received. Control rod drive flow was reduced and RWCU system blowdown was maximized in an effort to slow the level increase.
Reactor level was subsequently stabilized and slightly decreasing. At the time the safety relief valve was opened, vessel level was approximately +58 inches and reactor pressure was approximately 1030 psig. During the time the valve was open, vessel pressure decreased to about 880 psig. When the valve was closed, vessel level dropped rapidly from an initial value of +61 inches to +20 inches in approximately 20 seconds. Vessel level continued to decrease to +13 inches, the Reactor Vessel Water Level - Low (Level 3) setpoint, over the next 30 seconds.
The RCIC system was manually started to raise water level, but not in time to avoid the scram signal. The system was started about 15 seconds prior to the scram signal. The lowest vessel level reached during the event was +13 inches, before being recovered to the normal operating level.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FAClllTY NANE (1) DOCKET NQIBER (2)
YfAII
'Washington Nuclear Plant - Unit 2 05 0 0 0 3 9 7 97 005 0 0 3 o" 4 THXY (17) mm i rr ive Acti n Plant operators responded to the low level condition and promptly restored vessel water level to within normal limits. The scram signal was reset approximately two minutes after it was received.
F herE I n Pursuant to 10 CFR 50.73(a)(2)(iv) this event is reportable as, "Any event or condition that resulted in a manual or automatic actuation of any Engineered Safety Feature, including the Reactor Protection System."
!
- 2. During the evolution, some shrinkage in vessel water level was anticipated by the Operations crew. However, the crew did not anticipate the amount of vessel level decrease which occurred.
- 3. Following a scram, pressure typically decreases to the 600 to 700 psig range. This decrease occurs regardless of rapid operator actions to secure auxiliary steam loads.
Normal post-scram operations involve securing steam loads and then starting a cooldown.
The RFW system pumps are the preferred source of water to the vessel, but draw too much steam to maintain a high pressure condition.
To maintain rated pressure it is usually necessary to close the MSIVs or Main Steam (MS)
System Valve MS-V-146 [V], which effectively removes the preferred feed and pressure control (turbine bypa s valves) sources. The RCIC system can then be used for level and pressure control.
- 4. The decision to conduct the safety relief valve testing following a scram did not evaluate the crew members previous experience with the plant in these conditions. No detailed contigency planning was conducted prior to establishing the plant conditions to perform safety relief valve testing.
There is procedural guidance for conditions where high pressure is maintained following a scram. However, the crew did not adequately use this guidance for control of pressure and level with the MSIVs closed. This led to the high level condition and the inability to use the RCIC system for pressure and level control.
LICENSEE EVENT REPORT tLER) TEXT CONTINUATION FAClLTTY NANE (1) DOCKET NNBER (Z) l%YSNM
',,'Cy MIMI HVMMI THXT (G)
Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 97 -005 0 0 4 oF 4 The root cause of this event was risks and consequences of decisions not completely identified and assessed. A contributing cause was documents not followed.
Safety relief valve testing is usually performed with the reactor on line at low power which allows for normal pressure and level control. In this event, the attempt to set up for this test without a contingency plan to perform this test following a scram contributed significantly to the transient.
The effects of the transient may have been minimized had the Operations crew used applicable steps of the normal shutdown-to-hot shutdown procedure and maintained the RCIC system in operation prior to closing the MSIVs.
II Expectations will be developed for plant shutdown schedules to provide guidance pertaining to actions that will be performed if a scram or other transient occurs prior to performance of a scheduled activity.
Specific simulator training, unique to this event, will be performed. The training will cover pressure and level control, during hot shutdown conditions, and the effects of safety relief valve actuation at normal level and pressure.
The lessons learned from this event have been discussed with the Operations crews.
The Operations crew involved was coached on maintaining focus of all critical plant parameters during transient conditions.
mn f f n n The safety consequences of this event are minimal. At the time of the event, the reactor was in a shutdown condition. All control rods had already been fully inserted prior to the scram signal.
The Reactor Protection System functioned as designed to cause a reactor scram signal and plant operator actions to recover level were prompt and correct. Therefore, the event posed no threat to the health and safety of the public or plant personnel.
imil Ev n There have been no recent similar events where not completely identifying and assessing the risks and consequences of decisions resulted in a low-level scram signal while in a shutdown condition.