ML20059D196
| ML20059D196 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 08/23/1993 |
| From: | AFFILIATION NOT ASSIGNED |
| To: | |
| References | |
| OLA-2-I-MFP-191, NUDOCS 9401070059 | |
| Download: ML20059D196 (16) | |
Text
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ESF Actuation
$. Description of Nonconf ormance '.25
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[ On 12/8/90 while Derforming a secondary plant startup ' ]
' 1 occurred on Steam Generator 1-3 while at 2% power and transferring fr.om I
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N OPS 12/10/90 APM 12/10/90 A 12. NCREval[a[ionAttachedo9 the NCR Test Corm nuation Sheets N 1. Plant Conditions Ill. Cause of Probles V. Corrective Actions A II. Description of Preblem IV. Analysis of Probla:s VI. Additional Information L T Y 14. Estinated 15. Responsible E S 13. Trend other Cocpletion Orge.nization C 2 Codes PlGl-]C l4l l l-l l l l l-l l l Date H N 5 N/A Complete Operations a 1 R 16. 10CTR$0.73 17. Potentially 10CTR21 18. 10CTR50.9 19. Reference Other , C E Reportable kepertable Reportable Repcrtable, if app.
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P 0 tes % ) No [] Yes l ) No1X) Yes [] No [x] N/A R 20. Rasis Refer to attached T A Initial 21. Time Limit 22. Method 23. Notified By 24 Time 25. Date
! * * "' Port N/A N/A T. I N/A N/A N/A l
L Followup 26. Required 27. Time Limit 28. LIR No. 29. Date
- Y'S lXI N II I
'P ft 30 days 1-90-015-01 01/25/91 U T 30. Other Agencies Notified None .
! 31. Re= Arks g m
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! U 32. Chai / 33. D 60. Other (p/s' , 41. Date 1 A 1 P SRFridley < l h' ate / /q f P 14 QA(p$) // 35. Date 62. Other (p/s) &3. Date j R v HJHansen ._ // /f n - '9/2//9/
- 36. QC p/s 37. e/ 64. Other (p/s) 45. Date A AJorgens(en 3 t. 4 L I
- 38. Re . MPPDCRs 39./ Days 46. Other (p/s)
' 47. Date I o lan _ _ __ _-_3_ _/;2 f / 8) / P5kC 1 68[MeettngLate: Reviev Q,-] / 69. GOhPkAC hct)f1Catton Date: M q G/ Corrective 50. Coorple d TR i (p/s) 51. Date 52. QA Verification by (p/s) 53. Date Action
/ 9401070059 930823 6 Distritut wn PDR ADDCK 05000275 O PDR I k'PC Katerials PSEC Secretary
' l Mata ge r . Q A Sta tion / Hydro Cer.strv:t ten Instaator
, Flant Marater. DCPP TIS ,
Approprnate QC GCVT P.AC Se c r e ta r y , Authorsted 1rspecter, Other i Eng ine e r in6 if applicatle Othe r 850 - 2 75 .3 2 OL 6 f-M FP- /9/ 8 23,th3 i racto10C5F}
m er.r-NUCLEAR REGULATORY COMM!ssiON < Docht No. O e775-Ol4~ Official Dh N n tra razuer cf 1_&$]F70 @AS tcyd 22FC7Nt(} S t 9 r _.__ ___ _ _._ em: , a E_ , _ . __ .DC1-90-OP-N083-f 4w . _ _ . _ _ "m( February 8,1991 ka n a / . . . _ _ C. ll f :'?_ f ) f- O _._ ,
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ciner _.__.____ amu __.____., _ _ _ Raporter . # A __ P-14 ESF ACTUATION DUE TO VALVE LEAKAGE MANAGEMENT
SUMMARY
On December 8, 1990, at 0031 hours PST, with Unit-1 in Mode 2 (Startup) at approximately 2% power, steam generator'(SG)'l-3 level exceeded the 67% narrow range level setpoint initiating a P-14 signal, an engineered safety feature (EST) T actuation. The root cause of the event was leakage through feedwater regulating and bypass valves FW-1-FCV-530 and FW-1-FCV-1530.- Leakage past feedwater check valve FW-1-531 and feedwater-recirculation control valve.FW-1-FCV-420. contributed to'this event. Other contributing causes include plant management's-
' failure to recognize the consequences of.the combined leakage of the feedwater check and regulating valves, and-to provide direction for either a timely repair of the'feedwater valves or an adequate strategy for operators to enable them to. start the Unit with leaking feedwater valves.
Operating procedure OP L-3 was revised to include steps to check' for main feedwater regulating and bypass valve leakage during startup, and to provide guidance for taking action to deal with any leakage identified. OP L-3 was also revised to require isolating FW-1-FCV-420 when it is not in service.- FW-1-FCV-420 was outage. isolated and will be repaired during the next refueling Corrective maintenance-was performed on FW-1-FCV-530, FW-1-FCV-1530 and FW-1-531 during a Unit'l forced outage. Further maintenance will be performed.on FW-1-FCV-530.during the. next refueling outage. A memorandum was issued to' plant staff. regarding this event. Operators were informed to.obtain management; concurrence on recovery plans. 90NCRWP\900PN083.JCN Page 1 of 15
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DC1-90-OP-N083 I 'I i February 8,1991 P-14 ESF ACTUATION DUE TO VALVE LEAKAGE I. Plant Conditiong Unit 1 was in Mode 2 (Startup) at approximately 2 percent reactor power. The plant was in the process of starting up from a reactor trip which occurred on December 5, 1990. II. Descrintion of Event A. Event: On December 8, 1990, at 0031 hours'PST, with Unit 1 in Mode 2 (Startup) at 2 percent reactor power, ! steam generator (SG) 1-3 exceeded the 67 percent I narrow range level setpoint, which initiated a P-14 signal. This engineered safety feature (ESP) l actuation occurred during the transfer of the feedwater flow controls from auxiliary feedwater to automatic main feedwater level control and resulted in main feedwater isolation, main feedwater_ pump (MFP) 1-2 trip, and main turbine trip. Figure 1 provides a schematic of the main feedwater system. . Three days earlier, on December 5, 1990, Unit 1 had experienced a reactor trip. During recovery from that trip, backleakage was experienced through SG 1-3 inlet feedwater check valve FW-1-531. Operations m&nagement believed this leakage to be within the capability of the operators control since similar leakage had been observed during. previous startups. Therefore, valve corrective actions were deferred until the next Unit i refueling outage. On December 7, 1990, at approximately 2300 hours, the control room operators had just completed the procedural steps to place MFP 1-2 in service with , pump speed adjusted for a feedwater/ main steam i differential pressure of approximately zero psid. Main feedwater isolation valve FW-1-FCV-440 was open. Feedwater regulating and bypass valves, FW l FCV-530 and FW-1-FCV-1530, respectively, indicated
- zero demand with the feedwater controller in l automatic.
l 90NCRWP\900PN083.JCN Page 2 of 15 l 1
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'L DC1-90-OP-N083
( February 8,1991 On December 8, 1990, shortly after 0000 hours, the control room operators observed that SG 1-3 level l was decreasing, with indicated auxiliary feedwater (AFW) flow to SG 1-3 greater than flow to the other three SGs. This problem was initially diagnosed by the operators as backleakage through feedwater check valve FW-1-531 because, as discussed above, backleakage was experienced during recovery from the December 5, 1990 reactor trip. Backleakage was also experienced during previous startups when low differential pressures existed across the check valve. To increase SG 1-3 level, operators increased speed on MFP 1-2 to prevent the backleakage and feed SG 1-3 from the main feedwater system. The level in SG 1-3 rapidly increased beyond its normal no-load level setpoinL. The control room operatars attempted to close feedwater regulating and bypass , valves FW-1-FCV-530 and FW-1-FCV-1530 by manually I decreasing valve demand, even though the demand was at zero. This action appeared to terminate the l initial steam generator level increase, so the decision was made to proceed with the Unit 1 startup. The main feedwater controls were placed in automatic and the level in SG 1-3, as indicated on the narrow and wide range level instrumentation, immediately began to increase. Wide range level appeared to stpbilize shortly into the transient. The control room operators concluded that the narrow range level increase was caused by SG swell and since wide' range level did not show an incrcasing trend, the reactor power was slowly increased. Reactor coolant system (RCS) temperature was increasing and the steam dumps opened. The narrow range level increased to 67 percent, causing the P-14 actuation. The Operations crew on shift determined that the P-14 actuation was : caused by steam generator overfeeding due to leakage ' past feedwater regulating valve FW-1-FCV-530 and/or feedwater regulating bypass valve FW-1-FCV-1530, and ; subsequent SG swell during steam dump actuation. At that point, it could not be definitively determined which valve or whether both valves were leaking because the valves are in parallel. l Operators restored plant conditions to those which I existed prior to the P-14 actuation and continued l 90NCRWP\900PN083.JCN Page 3 of 15 1
~ _. _ - _ _ ._ . .. .. _ _. .. _ _ _ __ _._- _ . _ _ . . ._. ._._. . _ _ ** DC1-90-OP-N083 February 8; 1991 with plant startup without consulting with plant.
management. . On December 21, 1990, as a resultfof further ! investigation.regarding feedwater system leakage, Operations determined that feedwater recirculation' y control valve FW-1-FCV-420 wasfleakingqto the condenser. 'This valve was' manually isolated. . B. Inoperable Structures,1 Components, or Systems 1that. " Contributed to the Eve...: None. ': C. Dates and Approximate Times for Major Occurrences:; -
- 1. Dec. 8, 1990, 0025 hrs:' 'MFP 1-2 isLplaced:
- in service.
- 2. Dec. 8, 1990, 0031 hrs:- . Event \ Disc ~overy. !
.date. P-14: ESF signal occurred'.
i Four hour'non- .
- 3. Dec. 8, 1990, 0146 hrs:
emergency report. was made-to:the ' NRC'in~accordance withe 10~CFR 50.72. -; l D. Other Systems or Secondary Functions Affected:-j ! None. l l E. Method of Discovery: , The event was apparent to control room? personnel'due l to control room alarms and indications. , F. Operator Actions: The operators stabilized.the plant iniMode-2,. verified.that the SG levels wereLbeing~ maintained 4byc the AFW system, and. reestablished ~ plant 1 conditions; ' to those existing prior to the.P-14. signal. G. Safety System Responses:: Ji MFP'l-2 tripped.: 1.
} ,
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l DC1-90-OP-N083-I February 8,1991
- 2. The main turbine tripped.
l
- 3. The main feedwater isolation valves.
automatically closed. 4.- The main feedwater regulating valves and. regulating: bypass valves-automatically' closed. III. Cause of the Event , A. Immediate Cause: l l The immediate cause for the SG'high-high level I signal was that'SG 1-3 level exceeded-the P-14 - actuation setpoint due to overfill and subsequent swell during'startup.- B. Root Cause: I The root cause of the event was leakage through. l feedwater regulating valve FW-1-FCV-530fand feedwater regulating bypass valve FW-1-FCV-1530. FW-1-FCV-530 was stroke tested during a UnitJ1 forced outage which began on December 23, 1990,_'and it was determined that leakage through this. valve l was due to valve position controller. drift. 'FW FCV-1530 was disassembled and inspected during;the same forced outage. ' Leakage through this' valve was l also due to valve pouition controller drift. C. Cdhtributory Cause: i 1. Backleakage through SG 1-3 inlet feedwater-l check' valve FW-1-531'due to a slight-l misalignment of.the check valve disc contributed to this event'since.the leakage through FW-1-531 in combination.with leaknge through FW-1-FCV-530 and FW-1-FCV-1530 initiated the level transient in-SG'l-3.
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- 2. On December-21,-1990, as a-result' of'further investigation regarding feedwater system ~
-leakage, Operations' determined-that-feedwater' recirculation control valve FW-1-FCV-420 was leaking to the condenser. . Leakage past the valve contributed to this event in.that it-provided a. leak path to.the condenser.
90NCRWP\900PN083.JCN Page 5 of 15 i i {. L
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.DC1-90-OP-N083 February Bi 1991
- 3. During recovery from the Unit 1 reactor trip on December 5, 1990, backleakage was experienced-through SG 1-3 inlet feedwater check valve FW--
1-531. Operations management believed that this leakage was within the capabilityLof the operators control since similar leakage had been observed during previous startupsifrom post-trip' outages in February and June 1990. Plant management's failure,to recognize the consequences of the combined leakage of the
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feedwater check and regulating valves. contributed to this event.
- 4. ' Procedures for placing the main feedwater system in service'(feeding forward.to,the steam generators) did not provide guidance for contingency actions'to investigate feedwater regulating valve leakage.' Plant. management's-failure.to provide direction forleither a timely repair of the feedwater valves or an adequate strategy,for operators to enable them :
to startup the unit with leaking feedwater valves was a contributing cause. IV. Analysis of the Event l A. Leaking Feedwater. Regulating Valves: The high-high SG water level P-14 protective- , interlock is designed to protect the main; turbine 1 l from water intrusion and subsequent damage caused.by l SD overfill. This feature is designed primarily1for lj equipment protection, and functioned as designed. J Leakage past the feedwater: regulating-valves.is i bounded by the analyzed event of a failed _ full-open j feedwater regulating valve._ j i B. Leaking Feedwater Check Valve: Feedwater check. valve FW-1-531.is a safety: category I valve installed tofprotectLthe SG and RCS'from-excessive cooldown due to blowdown following:a postulated line rupture in the safety; category II: portion of a'feedwater line. The' complete failure of FW-1-531 would constitute a main feedwater line failure, which is a previously analyzed'FSARjUpdate: accident. The condition of-the' check valve has been evaluated-90NCRWP\900PN083.JCN Page 6 of 15 m
I DC1-90-OP-N083 ' February 8,1991 , for potential effect on feedwater line integrity and for potential effect on water supply to the SGs. Valve integrity was investigated by physical l inspection of the valve. The valve inspection showed no physAcal damage to the disc or seat that l would impair disc movement or lead to catastrophic ! disc failure. The leakage was due to misalignment of the disc and the seat which resulted in lack of l contact on the full circumference of the seat. At higher pressure differentials, the disc would elastically deflect and provide a seal, as was evident by successful surveillance test results. The inspection verified there was no impairment in integrity or in the ability to close in response to l a differential pressure of greater than 500 paig. The effect of FW check valve leakage on fe'edwater delivery to the SGs was investigated by evaluating the check valve leakage on the performance of the AFW system. Chapter 6 of the FSAR Update summarizes the flow requirements for the various design basis accidents. As shown in FSAR Update Table 6.5-2, a minimum flow of 440 gpm is required to be delivered to two SGs 10 minutes after a main feedwater line break. The main feedwater line break accident is the most limiting Chapter 15 accident with respect to AFW flow requirements. The feedwater line break accident was reviewed to evaluate the effect of a leak in the feedwater line check valve. The worst case scenario is censiderad to be a feedwater line break occurring in SG 2, one of the two steam 4 generators which supplies steam for the turbine ' driven auxiliary feedwater (TDAFW) pump. The l leaking feedwater line check valve is in che i feedwater line to SG 3 which is the second source of steam to the TDAFW pump turbine. In this case, the feedwater line break would result in a rapid loss of steam pressure in the affected SG and the check { valve leak would reduce auxiliary feedwater flow to I the other steam generator (SG 3). The availability of the steam sources for the TDAFW pump turbine were evaluated. SG 2 would be faulted due to the feedwater line break and would not be considered a viable steam source. Further, the delivery of auxiliary feedwater to SG 3 would be reduced due to backleakage through the feedwater check valve. SG 3 inventory and steam pressure may s 90NCRWP\900PN083.JCN Page 7 of 15
i l l I I 'DC1-90-OP-N083 ( February 6; -1991 decay as a result of reduced auxiliary.feedwater flow. The inventory and pressure decay would be slow, however, and.SG 3 would provideEsteam for some period. Since the steam pressure response of the SG-has not been quantified, a conservative. approach was taken by assuming no steam would km 'available. from-SG 3. Thus, the TDAFW pump'is conservatively. assumed to be not operable. This feedwater line break' scenario was evaluated with the application:or a single failure. The
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limiting single failure is1 bus H_which takes, motor driven auxiliary-feedwater (MDAFW)fpumpL2 out of. l service. The bus:H failurejalso_ removes power!from the feedwater isolation valve onLthe_.SG 3 feedwater line. . Without power, the-feedwaterLisolation: valve - FW-1-FCV-440_would not close on the safety 1 injection- ' signal and would- not11solate_ the check valve leak. i With the leak not~ isolated,-full AFW delivery to_SG l 3 could'be impaired and,_ consequently, FSAR' Update j flow requirements may.not be satisfied. This single I failure results in MDAFW pump'3? feeding the: . ! feedwater line (SG 3)1with.the leaking check valve H and feeding the unaffected SG'4. . Due toLthef . l feedwater check valve leakage, the delivery of AFW' l to SG 3 could be reduced. This.would result inoSG.3 levels decreasing faster than'SG 4' levels. . This-disparity would be evident:to the control' room - operators. The operators <would close.the auxiliary l feedwater regulating valves.from.the' control room-
']
within 10 minutes ta) isolate'auxiliaryjfeedwater:to th% leaking feedwater line and to1 the damaged: steam generator as required by procedure. : Subsequently, flow from MDAFW pump-3 would'be delivered to SG 4 only. The flow rate from the pump.wouldinot reach-full design flow due to1the increased hydraulic' resistance in the auxiliary feedwater1 flow path to SG 4, since the pump is feeding only one,line-instead of two. The condition of one steamLgeneratoribeing supplied-from one MDAFW pump following.a.feedwater line break. has been evaluated for DCPPfby Westinghouse Electric. Corporation (Reference letter PGE-89-526)..The results of that evaluation.show that the conclusions of the FSAR Update remain valid for.this-postulated- j condition. The following' assumptions were 'made irr ' the evaluation: i 90NCRWP\900PN083.JCN Page 8 of 15 ! i q
_ _ - . ~ . _ _ . . __ . _ . _ _. . _ j .. I I 1 I- - DCl-90-OP-N083 . $ February 8,1991 ; 1 l j The TDAFW pump turbine is disabled'due to j j interruption of both steam sources l j following a feedwater line break. ] i i j A single failure of the MDAFW pump which' l 4 is not associated with the faulted steam j generator occurs. 4 j Ten minutes after reactor trip, operator-i action isolates the feedwater line break j and water is supplied to one intact SG -- . 1 Estimated flow to the intact SG is-325 j 9pm. - Thirty minutes after reactor trip, ! operator action increases the auxiliary: feedwater delivery to 440 gpm. The. 4 additional auxiliary feedwater is fed to l at-least'two SGs. I i This evaluation was performed for two cases: Lwith - off-site power and without off-site-power.;LThe i results were shown to be-within FSAR Update limits- l i by showing that no boiling'would occur in the hot-
- leg of the RCS and that the pressurizer would not j fill.
5 The operators are trainedEto verifyfadequate , auxiliary feedwater flow and'SG. levels following an accident. Emergency Operating Procedure L (EP) : F-0,
- " Critical Safety Function Status Trees," would i direct operators to EP FR-H.1, " Response to-Loss of !
Secondary Heat Sink." This procedure. instructs the j operators to restore at least-460 gpm (indicated ! flow) of auxiliary'feedwater flow tofthe'SGs"by. - t performing local manual. valve: alignments as-necessary to achieve.the minimum-flow requirements .
- to two SGs. The operators could close the SG-3.
1 feedwater isolation valve to. isolate the back: : { 1eakage through the check valve and_ resume auxiliary j feedwater flow to SG 3 and/or open the MDAFW' ! . discharge line cross-tie to provide _ flow to SG 1 L from MDAFW pump 3. These. actions:are consistent i with-the assumptions used in the Westinghouse evaluation. The evaluation of the consequences of a11eakLin.the , feedwater line' check valve, the results'of the i Westinghouse evaluation, and review of'the DCPP ~ ; emergency procedures show.there were no adverse s . I 90NCRWP\900PN083.JCN Page 9' of 15 5 I-i ; 1 C________..__- _ . _ . _ . ,_ --
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-i DC1-90-OP-N083 J
February 8,1991 safety consequences resulting from'the event or from l the leaking check valve. Further, the:P-14 : actuation and the leaking feedwater' valves resulted in no adverse safety consequences. : .Therefore,cthe. health and safety of the public were'not advereely l affected by this event. _; B. Reportability: ,
- 1. Reviewed under QAP-15.B .snd. determined to be 5 non-conforming in accordance with section 1 2.1.2. since. it was an unplanned l ESF -actuation ' ' I and is reportable as stated lbelow.- ;
i l 2. Reviewed under NUREG 1022 and determined to' bel j reportable 11n accordance'with 101CFR' 50.73 (a) (2) (iv) as f anj unplanned :ESF actuationi :
- 3. This event does not require a 110'CFR:21 report since there is'no generic hardware. problem.'
- 4. This-event:does not~ require'an~INPO-Nuclear- ,
Network entry since corrective actions were j equipment-specific to DCPP. 1
- 5. Reviewed under 10 CFR 50.9 and1 determined to ke: ,
! not reportable since this'avent does not have a l significant-implication'for-public= health and- 1 safety or common defense and security. , V. CORRECTIVE ACTIONS A. Immediate' Corrective' Actions: j None. j ? 'I B. Corrective Actions to prevent Recurrence:
- 1. Operating Procedure:L-3 will be revised to include steps to check for-main feedwater regulating and bypass. valve leakage, and-guidance on dealing with'any leakage found _
- j during this testing. q RESPONSIBILITY:.S. Fridley Complete ~
Operations (PGOE) AR A0211309,.AE # 01 Nc*c outage related. Not JCO related 4 90NCRWP\900PN083.JCN Page' 10- of 15 l l ( L
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, DC1-90-OP-N083 -6 i . February 8,1991 l NRC commitment: LER.1-90-015-01 Not a CMD commitment
- 2. Feedwater. heater recirculation control valve FW-1-FCV-420 was manually isolated and' . .
corrective maintenance will be performed during 1R4, scheduled to start February 3, 1991. Operations procedure OP L-3 was revised;to require ~ closing an additional manual isolation l valve in the flow path of FW-1-FCV-420 when FW-1-FCV-420.is'not in service' . RESPONSIBILITY:~J. Mellinger' -Complete- j Work Planning (PGNW) AR A0211309,,AE.#105 i Outage related: 1R4, Mcde 5: l Not JCO related NRC commitment:fLER 1-90-015-01 Not a CMD ccamitment
- 3. The following correctiveLmaintenance was performed during a Unit 1 forced outage which i began on December 23, 1990:
- a. Position controller"for the~feedwater-regulating valve FW-1-FCV-530 was adjusted i to correct for drift. However, the valve exhibited minor leakage.during_ restart '
and, in .accordance with revised OP L-3, 1 the valve was isolated until the Unit reached 12% power. During-the next refueling outage, further valve corrective maintenance will be performed including- < calibration of the position' controller. -
- b. Feedwater. regulating bypassLvalve FW .
FCV-1530 was disassembled,.the valve seat i was lapped and the valve plug was. ., replaced. The valve was reassembled and : given.a successful leak check. The valve 3
.positionEcontroller was. adjusted to correct'for. drift. ,
- c. SG 1-3 inlet check valve FW-1-531'was i disassembled and inspected,.and a misalignment of the check valve dis'c was corrected.
RESPONSIBILITY: T. Bennett Complete I L 90NCRWP\900PN083.JCN Page' 11 of 15
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DC1-90-OP-N083
' February 0,1991 Mechanical Maintenance (PGMT)
AR A0211309, AEs # 02, 03 & 04 l Outage related: 1R4, Mode 5 l Not JCO related . j NRC commitment: LER11-90-015-01 i Not a CMD commitment )
- 4. A memorandum from the Plant Manager will be transmitted to Department Heads and other Plant Supervision regarding this event to emphasize; that all equipmeny problems should.be adequately evaluated in a timely manner'and appropriate. compensatory measures.should:be identified. ,
i RESPONSIBILITY:-S. Fridley- Complete Operations (PGOM) AR A0211309, AE # 06 : Not-outage'related. Not JCO related NRC commitment: IJR 1-90-015-01 i Not a CMD commitment Following the P-14. actuation, Operatorsf 5. continued with Plant startup without consulting Plant Management. Although.this is not aLcause of the P-14 actuation, the JPM, Operations,.has reviewed with Operations Supervisors the importance of obtaining PlantLManagement concurrence for recovery plans'following ESF , actuations and other significant plant ' transients. 1 RESPONSIBILITY: D. Miklush Complete Operations' : Not outage related ' Not JCO related t Not an NRC commitment Not a CMD commitment : VI. ADDITIONAL INFORMATION A. Failed Components: if FW-1-FCV-530, FW-1-FCV-1530, FW-1-FCV-420 and FW ' ~! 531 and leaked contributing to this event. ; B. Previous NCRs on Similar Events:. , 90NCRWP\900PN083.JCN Page 12 of 15 ,
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f* '. l l DCl-90-OP-N083 February 8,1991
- 1. DC2-89-OP-N042, " Main Turbine Trip and Feedwater Isolation due to a Feedwater Regulating Valve not being Fully closed during Surveillance Testing". The corrective actions for this NCR cautioned operators to ensure l feedwater regulating and bypass' valves are l fully closed during surveillance testing and l was therefore too specific to prevent the December 8, 1990 P-14 trip as surveillance was not being performed e.t that time.
l
- 2. DC1-87-OP-N134, "High-high Level in SG 1-1 (P-l 14) Resulted in Turbine Trip". The corrective actions for this NCR was the issue of an
- operator awareness bulletin cautioning i operators of the possibility of level
) instability when switching'the feedwater system i between manual and automatic level control.
- 3. DC1-87-OP-N014, " Unit 1 Turbine / reactor Trip due to P-14". The corrective actions for this NCR was performance of startup testing of the feedwater control system. Since this event the feedwater level control system has been replaced with a digital system so these corrective actions could not have prevented the recent P-14 event.
C. Operating Experience Review:
- 1. NPRDS:
N/A.
- 2. NRC Bulletins, Generic Letters or Information Notices:
None.
- 3. INPO SERs and SOERs:
None. D. Trend Code: PG (Plant Staff, General) - C4 (Equipment Deficiency, Maintenance).
, E. Corrective Action Tracking:
90NCRWP\900PN083.JCN Page 13 of 15
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't DC1-90-OP-N083
# February 8,1991 ~
)
f
- 1. The tracking ~ action requestlis A0211309. l
- 2. The' investigative and corrective actions.are I outage'related and will occur duringL1R4.- ;
F. Footnotes and-Special Comments.- i
- The estimated. completion date'of this NCR is'May.15,
! 1991. . G.
References:
.j
- 1. The originating AR'is A0210733. j l i
- 2. LER 1-90-015-01, " Unit'1lESF Actuation,JP-14 '
i. (High-high Steam Generator Level) . Due To: Feedwater Valva Leakage". . , ! H. . Meeting Minutes: The TRG reconvened CH1' February 8,?1991)to reviewcthel incorporation'of changes.made toLLERL1-90-015-011 O based on comments from-the"NRC Resident Inspectors! r office. These comments were ofra nature'that DCPP- ; I was not sufficiently'self-criticaljand'didn't. , sufficiently involve management in the December 8,_ 1990 Plant restart.- The estimated closure dateof this NCR is May'15, 1991. I. Remarks: None. I 1 1 F r 1 l 90NCRWP\900PN083.JCN' Page- 14;;of 15+
,- DC1-90-OP-N083
- February 8,1991 Figure 1 MAIN FEEDWATER SYSTEM ONE UNE DIAGRAM Aus111.ry mC -
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mgg m* 1-2 Few_qqs rew-ISee F..dw.t.r Main Feedwat.r g% y ,CV- .....,. ,_,. 4 90NCRWi \900PN083.JCN Page 15 of 15
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