ML20210S841
| ML20210S841 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 08/29/1997 |
| From: | De Agazio A NRC (Affiliation Not Assigned) |
| To: | Feigenbaum T NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO) |
| References | |
| NUDOCS 9709120157 | |
| Download: ML20210S841 (22) | |
Text
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NUCLEAR REQULATORY COMMISSION WAsMINeToN, D.C. BeteHoot
% August 29, 1997 Mr. Ted C, Folgenbaum Executive Vice President and Chief Nuclear Offloor North Atlantic Energy Service Corporation clo Mr. Terry L Herpater P.O. Box 300 Seabrook, NH 03874
SUBJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECUR8OR ANALYSIS OF OPERATIONAL CONDITION AT SEABROOK
Dear Mr. Feigenbaum:
Enclosed for your review and comment is a copy of the prolkninary Accident Sequence Prwursor (ASP) analysis of an operational condition which was discovered at Seabrook on May 21,1996 (Enclosure 1), and was reported in Licenseo Event Report (LER) No. 443/96-003. This analysis was prepared by our contractor at the Oak Ridge National Laboratory.
The results of this preliminary analysis indicate that this event may be a procursor for 1996. In assessing operational events, an effort was made to make the ASP models as realistic as possible regarding the specific features and response of a given plant to various accident sequence initiators. We realize that licensees may have additional systems and emergency procedures or other features at their plants that might affect the analysis. Therefore, we are providing you an opportunity to review and comment on the technical adequacy of the preliminary ASP analysis, including the depiction of plant equipment and equipment capabilities. Upon recolpt and evaluation of your comments, we will revise the conditional core damage probability calculations where necessary to consie the specific information you have provided. The object of the review process is to provide as realistic an analysis of the significance of the event as posalbie, in order for us to incorporate your comments, perform any required reanalysis, and prepare the final report of our analysis of this event in a timely manner, you are requested to complete your review and to provide any comments by September 30,1997. We have streamlined the ASP Program with the objective of significantly improving the time after an event in which the final procursor analysis of the event is made publicly available. As soon as our final analysis of the everd has been completed, we will provide for your information the final procursor analysis of the event and the resolution of your comments, in previous years, licensees have had to wait until publication of the Annual Procursor Report (in some cases, up to 23 months after an event) for the final procursor analysis of an event and the resolution of their comments.
We have enclosed severalitems to facilitate your review. Enclosure 2 contains specific guidance for performing the requested review, identifies the criteria which we will apply to determine whether any credit should be given in the analysis for the use of licensee identified additional equipment or specific actions in recovering from the event, and describes the specific information that you should provide to suppert such a claim. Enclosure 3 is a copy of LER No. 443/96-003, which documented the event.
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Mr. Ted C. Feigenbaum 2- August 29, 1997 This request is covered by the existing OMB clearance number (3150 0104) for NRC staff follow-up review of events documented in LERs. Your response to this request is voluntary and does not constitute a licensing requirement.
Sincerely, Original signed by:
Albert W. De Agazio, Sr. Project Manager Project Directorate 13 Division of Reactor Projects t/Il i
Office of Nuclear Reector Regulation Docket No. 50-443 Serial No. SEA 97 021
Enclosures:
- 1. Preliminary Accident Sequence Procursor Analysis
- 2. Guidance for Licensee Review of Preliminary ASP Analysis
- 3. LER No. 443/96-003 (
cc w/encls: See next page Distribution Docket File SMays I PUBLIC PO'Reilly
. pol-3 RF BBoger REaton SLittle ADeAgazio OGC ACRS JRogge,RI DOCUMENT NAME: G:\DEAGAZIO\ ASP 0596.WPD v.e iw . ,y.eed d u ni, ins i.in m.b..: *c' - copy without encio.ur.. *r - copy with .noio.ur . 'N' u No copy /
OFFICE PM:PDI 3 /'t)[V LA:PDI 1,n (A)D:PDI-A /
NAME ADeAgazidP Slittl # REaton M DATE 08/ 9 9 /97 08/ AB /97 08/ 1,9 /97 0FFICIAL RECORD COPY 1 ..
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., *1 North Atlantic Energy Service Corporation Seabrook Station, Unit No. 1 Cc!
Lillian M. Cuoco. Esq. Mr. Dan McElhinney Senior Nuclear Counsel Federal Emergency bnagement Agency Northeast Utilities Service Company Region I P.O. Box 270 J.W. McCormack P.O. &
Hartford, CT 06141-0270 Courthouse Building, Room 401 Boston, MA 02109 Mr. Peter Brann Assistant Attorney General Mr. Peter LaPorte, Director State House, Station #6 ATTN: James Muckerheide Augusta, ME 04333 Massachusetts Emergency Management Agency Resident Inspector 400 Worcester Road U.S. Nuclear Regulatory Connission P.O. Box 1496 Seabrook Nuclear Power Station Framingham, MA 01701-0317 P.O. Box 1149 Seabrook, NH 03874 Jeffrey Howkrd, Attorney General G. Dana Bisbee, Deputy Attorney Jane Spector General Federal Energy Regulatory Commission 33 Capitol Street 825 North Capital Street, N.E. Concord, NH 03301 Room 8105 Washington, DC 20426 Mr. D. M. Goebel Vice President-Nuclear Oversight Town of Exeter Northeast Utilities Service Company 10 Front Street P. O. Box 270 Exeter, NH 03823 Hartford, CT 06141-0270 Mr. George L. Iverson, Director Mr. J. K. Thayer New Hampshire Office of Emergency Recovery Officer, Nuclear Engineering Management and Support -
State Office Park South Northeast Utilities Service Company 107 Pleasant Street P.O. Box 128 Concord, NH 03301 Waterford, CT 06385 Regional Administrator, Region ! Mr. F. C. Rothen U.S. Nucient Regulatory Commission 475 Allendale Road Vice President - Nuclear Work Services Northeast Utilities Service Company King of Prussia, PA 19406 P.O. Box 128 Waterford, CT 06385 office of the Attorney General One Ashburton Place 20th Floor Mr. A. M. Callendrello Boston, MA 02108 Licensing Manager - Seabrook Station North Atlantic Energy Service Corp, Board of Selectmen P.O. Box 300 Town of Amesbury Seabrook, NH 03874 Town Hall Amesbury, MA 01913 l
1
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North Atlantic Energy Service Corporation Seabrook Station, Unit No. I cc!
Mr. W. A. DiProfio Nuclear Unit Director Seabrook Station North Atlantic Energy Service Corporation P.O. Box 300 Seabrook, NH 03874 Mr. Frank W. Getman, Jr.
Cocheco Falls Millworks 100 Main Street, Suite 201 Dover, NH 03820 Mr. B. D. Kenyon President - Nuclear Group Northeast Utilities Service Group P.O. Box 128 Waterford, CT 06385 Mr. B. L. Drawbridge Executive Director Services &
Senior Site Officer North Atlantic Energy Service Corp.
Seabrook, NH 03874 o
. , i l
l Enclosure 1 Preliminary Accident Sequence Precursor Analysis
w
.,, ,i LER No. 443/96-003 LER No. 443/96 003 Event
Description:
Turbine dris en emergency feedwater pump unavailable i
because of a mechanicrJ seal failure I
Date of Event: May 21,1996 Plant: Seabrook Event Summary Seabrook was at 100% power when personnel were performing a scheduled operating test on the turbine-driven emergency feedwater (TDEFW) pu np. The pump was manually tripped after spaiks were obsened coming out ofits outboard mechanical seal. The sparks were ultimately attnbuted to the improper installation of the mechanical seal assembly during the previous refueling outage in Nosember December 1995 (Ref.1, 2). This long term unavailability of the TDEFW pump (3,875 h) would have affected the units' response to a loss of ofTsite power (LOOP) or a transient esent. The estimated increase in the core damage probability (CDP)over the 5 month period for this event (i c., the importance)is 1.2 = 10d. The base probability of core damage (the CDP) for the same period is 3.1 = 104.
Event Description Seabrook was at 100% power on May 21,1996, when personnel started the TDErW pump for its scheduled quarterly sun cillance test. The operator tripped the pump locally during the test after sparks were obsened emanating from the outboard mechanical seat area of the pump. The mechanical seal was disassembled and inspected The sparks were the result of mechanical interference within the seal assembly. The outboard seal gland had 0.007 in. clearance from the top of the shaft sleeve and the throttle bushing inside diameter. The sparks were caused because the shaft sleeve rubbed against the inside diameter of the throttle bushing, causing a 0.005 in. gouge in the shan sleeve and chipping of the throttle bushing. Licensee personnel concluded that because of the improper installation of the seal, the TDEFW pump would not have been able to perform its safety function for the required mission time (24 h) since the November December 1995 refueling outage.
After repairing the TDEFW pump, personnel inspected the mechanical seal of the motor-driven emergency feedwater (MDEFW) pump and discovered it to have a similar alignment along with the conesponding indications of mechanical rubbing. The MDEFW pump outboard mechanical seal gland had a 0.0035 in.
clearance between the shaft sleeve and the top of the throttle bushing inside diameter. The MDEFW pump throttle bushing was not chipped like the throttle bushing was on the TDEFW pump. Presumably because the MDEFW pump had not failed a quarterly surveillance test, the system engineer concluded that the MDEFW pump was capable of performing its design function.
The design clearances and tolerances of the TDEFW pump's mechanical seals were insufficient to prevent damage during operation unless the installation technique used non-customary methods (ie, use of dial indicators and feeler gauges). The design permitted the allowable to!crances to be greater than the available clearance. Hence, the design did not preclude the interference between the seal and the shan sleeve. This 1
t o LER No. 443/96 003 design deficiency also applies to the MDEFW pump mechanical seals. Contributing to this event was the failure to adequately incorporate previous knowledge regarding seal installation into maintenance procedures or training. As a result, maintenance persormel were unaware of a prior seal failure (in 1987), or the need to take precision measurements to verify the proper installation of the seal assembly.
Additional Event Related Information The emergency feedwater (EFW) system consists of two 100% capacity trains that feed a common discharge header (Ref. 3). One train uses the TDEFW pump and the other train uses the MDEFW pump. All four steam generators can be fed by either EFW pump. The TDEFW pump is supplied stearn from thc A and B steam generators. The MDEFW pump is powered from 4160V emergency bus E6 supported by the B emergency diesel generator (EDG).
Seabrook also maintains a start up feedwater pump with a capacity approximately equivalent to the combined capacity of both EFW pumps (Ref. 3). The start up feedwater pump can be started from the control room, except during a LOOP. Two normally closed motor operated valves (MOVs) must be opened to establish feedwater flow. Following a LOOP, the nonnat power source to the start up feedwater pump is not supplied power from an emergency bus. Therefore, the normal breaker alignment for the start up feedwater pump must be altered from 4160V bus 4 to 4160V emergency bus E5 (emergency bus E5 is powered by the A EDO). The normal and alternate start up feedwater pump breakers are key interlocked, requiring one breaker to be racked out before the interlock key can be removed. The interlock key is required to rack in the allemate source breaker (from bus ES) to the start up feedwater pump.
Modeling Assumptions Even though previous surveillance tests were successfully completed, the licensee concluded that the TDEFW pump would not have been able to perform its safety function for the required mission time (24 h) since the November December 1995 refueling outage (Ref.1,2). Hence, the TDEFW pump was considered inoperable and its failure probability was adjusted to 1.0 (TRUE) for a 3,875 h condition assessment. The 3,875 h condition assessment is based on the TDEFW pump being required from the end of the outage on December 9,1995, until the discovery of the mechanical seal failure on May 21,1996. Two days (48 h)were subtracted from the total number of hours that the TDEFW pump was unavailable to account for a reador trip in January.
The licensee indicated that the MDEFW pump would have performed its safety function for the required mission time. However, because the outboard mechanical seal on the MDEFW pump had wear similar to that 7 of the TDEFW pump, the potential for a common cause failure increased. The EFW common cause factor was developed based on data distnbutions fo. . nixed pump types contained in INEL.94-0064, Common- >
Cause Fallurc Dara Collection and Analysis Sysicm (Ref. 4. Table 9 19; Alpha Factor Distribution Summary
- All AFP Types Fail to Start, CCCG = 2, a2 = 0.0884). Because a2 is equivalent to the p factor of the multiple Greek letter method used in the Integrated Reliability and Risk Analysis System (IRRAS) models, the common cause failure probability of the EFW system pumps (AFW PMP CF ALL) was adjusted from 3.8 x 10" to 8.84 = 104 based on the common cause failure potential.
2
. J l 1.ER No. 443/96 003 The Seabrook Indwidual Mont Examinarfon (IPE) indicates that the start up feedwater pump is a backup source of feedwater for the EFW system. To credit the use of the start up feedwater pump, a basic esent was added to the IRRAS model for the Seabrook plant based on the IPE value for a failure of the start up l
fecdwater pump to start and run (Ref. 5, Table 7.9 1) or a f ailure of the associated valves to open (basic event l EFW hiDP FC SFP). Decause an operator is required to open two normally closed hiOVs to establish flow from the start up feedwater system, another basic event was added to account for the failure of the operator to manipulate the required hiOVs (ERWX11E Xht SFPi Finally, during a LOOP, an operator must realign the supply breaker for the stan up feedwater pump to the A EDO. A basic event was therefore added to represent the failure of an operator to complete this realignment (ERWXilE Xht BRKR). This last basic event aves based on the assumption that it would take an operator approximately 15 min, following a LOOP, to perfean the activity and that approximately 45 min were available before a steam generator dry out would occur, leading to core damage. A lognormal distribution was used to calculate the failure probability for ERWX11E Xht BRKR.
The operator non recovery probability for the EFW system during a LOOP (EFW XIIE NOREC L) was adjusted from 0.26 to 0.80 because this action is not independent from other operator actions. The operator must nrst realign the supply breaker for the start up feedwater pump to the A EDG (EFW X11E Xht BRKR).
If the operator fails to realign this breaker, the start up feedwater pump would not be available in a LOOP scenario (LOOP sequence 15). Further, if the operator does indee.d fail to realign this breaker, it is more likely that the operator will fail to recover the EFW system durir.g a LOOP. Finally, during a SBO, the only source of EFW is the TDEFW pump, therefore, with the TDEFW pump unavailable, there is no orportunity to recover EFW. Based on this, the operator non recosery factor during a SBO (EFW XilE NOREC EP) was set to"TRUE"(recovery not possible).
Analysis Results The increase in the CDP over a 3,875 hour0.0101 days <br />0.243 hours <br />0.00145 weeks <br />3.329375e-4 months <br /> period for this event is 1.2 = 10". The nominal CDP for the same period is 3.1 = 10-5. The dominant core damage sequence for this event (sequence 39 on Fig.1) involves
+ a postulated LOOP, e a successful reactor trip,
- a failure of emergency power, and failure of emergency feedwater.
This station blackout sequence (sequence 39 on Fig.1) accounts for 77% of the total contribution to the increase in the CDP. The next most dominant sequence (sequence 15 on Fig.1) contributes 15% to the total increase in the CDP. This sequence involves a LOOP with the success of emergency power, a failure of EFW, recovery of offsite power, and a failure of feed and bleed decay heat removal.
An attemate study investigating the conditional core damage probability (CCDP) associated with the reactor trip that occurred in January with the unavailable TDERV pump was conducted. The TDERV pump failure probability (EFW TDP FC l A) was set to "TRUE"(failed). Using the same material assumptions as those made for the previous condition assessment, the CCDP for this initiating event is 4.0 = 104. The dominant core damage sequence invohes a failure to trip the reactor and a failure of the EFW system.
3
- n l
l Lf.R No. 443/96 003 Definitions and probabilities for selected basic events are shown in Table 1. The conditional probabilities j associated with the highest probability sequences are shown in Table 2. Table 3 lists the sequence logic I associated with the sequences listed in Table 2. Table 4 describes the system names associated with the dominant sequences. hiinimal cut sets associated with the dominant sequences are shown in Table 5.
Acronyms ATWS anticipated transient without scram CCDP conditional core damage probability CDP core damage probability EDG emergency diesel generator EFW emergency feedwater system IPE integrated plant examination IRRAS Integrated Reliability and Risk Analysis System LOOP loss of offsite power hiDEFW motor driven EFW (pump) hiFW main feedwater hiOV motor operated valve PORV power operated relief valve SDO station blackout TDEFW turbine-driven EFW (pump)
References
- 1. LER 443/96 003, Rev. O " Emergency Feedwater Pump hiechanical Seal Failure," June 21,1996.
- 2. LER 443/96 003, Rev.1," Emergency Feedwater Pump hicchanical Seal Failure," September 12,1996.
- 3. FinalSafety Analysts Report, Serbrook Nuclear Station.
- 4. INEL 94/0064," Common Cause Failure Data Collection and Analysis System", December,1995.
- 5. Seabrook Nuclear Station,IndnidualPlant Examination.
4
w LER No. 443/96 003 l{ 5555858888855885588885585888885585888888
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Fig. I Dominant core damage sequence for LER No. 443/96 003.
- 5 l
LER No. 443/96-003 4
Table 1. Definitions and Probabilities for Selected Basic Events for LER No. 443/96 003 1 ;
i 4
Modified Event Bne Current for this name Descriptl0n probability probability Type event IE LOOP trutisting Esent-less ofofisite t 6 E406 8 6 E 006 No Power (LOOP) lESOTR trutiatmg Event-Steam 0,nerator 1.6 E 006 1.6 E 006 No Tute Rupture IESLOCA initiating Event-Small tess+f. 10 E 006 1.0 E 006 No Coolant Accident (SLOCA)
IE TRANS trdtietmg Event-Transient $.3 E 004 S.3 E 004 No j (TRANS)
El%MDP l'C lB Elv Motor-Driven INmp Fails 3.9 E 003 3.9 E-003 No Erw MDP l'C 5iP Stan up Feedwster 1%mp Fails 2.1 E 002 2.1E002 NEW No i
l EfW PMP CF ALL Common Cause Failure of EFW 3 8E 001 8.8 E402 Yes
. INmps(Excludes Start up ieedwster Pump)
EfW TDP l'C lA EfW Turbine Dnven Pump Fails 3 9 E 002 1.0 Em TRUE Yes EFW X11E NOREC Operator Fails to Recover Etw 2.6 E 001 2 6 E 001 No l EFW XHF-NOREC EP Operator fails to Recover EFW 14 E 001 1.0 E6 TRUE Yes During a Station Blutout EfW XIIE NOREC L Operator rails to Recover EI'W 2 6 E 001 8 0 E 001 . Yes Dunns a LOOP ETW X1tE-NREC ATW Operator Fails to Recover Efw I.0 E6 1.0 E6 No -
During an A1WS I
EFW X1tE XM BRXR Operator rails to Realign Start- 1.6 E 001 1 '> E 001 NEW No up Feedwster Pump Supply Breaker Ef%XHE XM SFP Operator rails to Open Sten up 1.0 E402 1.0 E 002 NEW No Fealwster Pump MOVs EPS DON CF ALL Common Couse Failure of EDos 1,6 E 003 1.6 E 003 No EPS-DON FC lA A EDO Fails 4.2 E-002 4.2 E 002 No EPS DON FC 1B B EDO Fails 4.2 E 002 4.2 E 002 No EPS XitE-NOREC Operator Fails to Recover 8.0 E 001 8 0 E-001 No t.mergency Power 6
- LER No. 443/96-003 Table 1. Definitions and Probabilities for Selected Basic Events for LER No.443/96-003 Modified Event Base Current for this
- name Description probability probability Type event ilPI XIIE XM TB Operator f ails to trutiste feed- 1.00002 1.0E002 No i and Bleed llPI XilE XM Fill Operator rails to trutiste reed- 1.0E002 10E42 No and Bleed Dunns LOOP
, MIV SYS TRIP Mein feedwater(MIV) $ystem 2 0 E 001 2.0 E 001 No Trips MIV XilE NOREC Operator fails to Recoser Mfw 3 4 E@l 3.4 E@l No i
- OEP XilE NOREC 6tl Operator Fails to Recover offsite 6.7 E 002 6.7 E 002 No q Power Within 6 h 5 PPR $RV CC 1 Power Operated Relief Valve 6.3 E@) 6.3 E 003 No (PORV)I fails to Open on Demand l
PPR SRV CC 2 PORV 2 Fails to Open on 6.3 E 003 6.3 E 003 No Demand 1
! RPS-NONREC Non Recoverable Reactor 2 0 E 00$ 2.0 E 00$ No
- Protection System Failures 1
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LER No. 443/96 003 i j l Table 2. Sequence Conditional Probabilities for LER No. 443/96 003 !
1 Conditional Event tree Sequence core damage Core damage Importance Percent name number probability probability contribution' (CCDP CDP)
LOOP 39 9.0 E 005 1.1 E 006 8.8 E 005 77.1
, LOOP 15 1.6 E 005 7.9 E 008 1.6 E 005 14.5 TRANS 21 8 5.1 E 006 7.7 E 008 5.0 E 006 4.4 TRANS 20 2.3 E 006 1.4 E 008 2.3 E 006 2.0 LOOP 19 1.2 E 006 6.4 E 009 1.2 E 006 - 1.1 Total (all sequences) 15 E 004 3.1 E 005 1.2 E 004 i
' Percent contribution to the total importance.
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1 8
y 1 LER No. 443/96-003 Table 3. Sequence Logic for Dominant Sequences for LER No. 443/96 003 Event tree name Sequence Logic number LOOP 39 /RT L. EP, EFW.L.EP LOOP 15 /RT L, /EP, EFW.L, /0P 611, F&B TRANS 21 8 RT, /RCSPRESS, EFW.ATWS TRANS 20 /RT, EFW, MFW, F&B LOOP 19 /RT L,/EP, EFW L, OP-6H. F&B L Table 4. System Names for LER No. 443/96 003 System name Logic EFW No or Insumcient EFW Flow EFW.ATWS No or Insumcient EFW Flow During an ATWS EFW.L No or Insumcient EFW Flow During a LOOP EFW.L.EP No or Insumcient EFW Flow During a Station Blackout EP Failure of Both Trains of Emergency Power F&B Failure to Provide Feed.md Bleed Cooling F&B L Failure to P4 ovide Feed and Bleed Cooling During LOOP MFW Failure of the MFW System OP 6H Operator Fails to Recos er Offsite Power Within 6 Hours RCSPRESS Failure to Limit Reactor Coolant System Pressure to
<3200 PSI RT Reactor Fails to Trip During Transient RT L Reactor Fails to Trip During LOOP 9
.l s LER No.443/96 003 Table 5. Conditional Cut Sets for Higher Probability Sequences for LER No. 443/96 003 Cut set Percent number contribution CCDP' Cut sets' LOOP Sequence 39 9.0 E 005
] 1 $2.4 4.7 E 005 EPS DON FC 1A, EPS DONJC 1B, EPS XilE NOREC, Ef%TDP FC 1A, El%XllE NOREC EP 2 47.6 4.3 E 005 tPS dan <F ALL, EPS XilE NOREC, EFW TDP-FC 1A, g EF%X11E NOREC EP LOOP Sequence 15 1.6 E 005
)
1 21.7 3.5 E 006 tlW PMP CF ALL, EFW X1tE XM BRKR, El%XilE NOREC L.
! /OEP XilE-NOREC4tl,llPI XIIE XM IB 2 13.7 2.2 E 006 Elv PMP CF ALL, Elv X)lE XM BRKR, Elv XilE.NOREC L, OEP XilE NOREC411.PPR SRV CC 2
> 3 13.7 2.2 E 006 EFW PMP CF ALL, El%XilE XM BRKR,EF%XilE NOREC L.
, OEP XilE NOREC411, PPR SRV CC 1 4 9.9 1.6 E 006 EPS don FC lA, EPS dan rC 1B, EfMTDP FC IA, EIWX1tE XM BRKR,EfWXilE NOREC l,
/OEP XilE.NOREC411. IIPI X)lE XM l'B j- $ 6.2 j.0E-006 /EPS DON FC l A, EPS DON IC lB, Ef%TDP FC 1A, j EfW XilE XM BRKR,El%XilE NOREC L,
/OEP XilE NOREC411.PPR SRV CC 2 6 6.2 1.0 E 006 /EPS-DON IC l A, EPS-DON FC 1B Ef%TDP-FC IA,
. EFW X11E XM BRKR, EFW XilE-NOREC L, l OEP XilE.NOREC411 PPR SRV CC 1 7 5.3 8.8 E 007 EPS DON FC 1A,/EPS DON-FC 1B, EFW PMP-CF ALL, EFEX11E NOREC L./OEP X11E NOREC411,itPI XIIE XM FB
- 8 3.3 5.5 E 007 EPS-DGN FC l A /EPS DnN FC lB, EF%PMP CF ALL, i
EFW X1tE NOREC L,IGT-XilE NOREC4}{, PPR SRV CC l 9 3.3 5.5 E 007 EPS DON FC lA,/EPS DON FC IB,Ei%PMP CF Al.L.
EFW XilE NOREC L, OEP X11E NOREC411, PPR SkV CC 2 TRANS Sequence 218. .5.1 E 006 I RPS-NONREC, Ef%PMP CF ALL,EFW XIiE NREC ATW i I 70.2 3.6 E 006 2 16.8 8.6 E-007 RP$t.!ONREC, EFW TDP FC l A, EFW MDP FC SFP, Ef%XilE-NREC Alw 3 7.9 4.1 E 007 RPS NONREC, El%TDP FC.lA, EFW XilE XM SFP, Ef%XilE NREC ATW j 10
7 F h. !
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, LER No. 443/96 003 L
Table 5 Coralitioni Cut Sets for Higher Probability Sequences Vor LER No. 443/96-003
~
h, I Cut set Percent
} number contribution CCDI" Cut sets' 4 3.1 1.6 E 007 RPS NONREC, EFW TDP-FC 1 A, EFW MDP FC-!B.
EFW-XilE NREC47W
- 3 * ' i
~
TRANS Sequence 20 2.3 E 006 D '-
} } 28.7 6.7 E 007 EFW PMP CF ALL, EFW MDP FC SFT. EfW-XilE-'NOREC, MFW SYS-TRIP, MEW XHE NOREC,14Pi XIIE XM-FB e
u' 2 18.1 4.2 E 007 EFW PMP CF-ALL, EFW MDP FC SFP, EFW X1tE NOREC, MP'/ SYS-TRIP, MF"4 X11E NOREC, PPR SRV CC 2
=-
3 18.I 1.2 E-007 EFW PMP-CF-ALL, EFW-MDP-FC-SFP, EFW XilE-NOREC,
_ MFW SYS TidP, MFW XI!E NOREC, FPR SRV CC 1
. 4 13.6 3.2 E-007 EFW PMPCF-ALL, EFW X}{E XM SFP, EFW X11E NOREC, MFW-SYS TRIP, MFW X11E-NOREC, HPI X1iE XM FB 5 ,
8.6 2.0 E 007 EFW-FMP-CF ALL, EFW XHE XM SFP, EFW XHE NOREC, MFW SYS-TRIP, MFW XHE NOREC, PPR4RV CC-2 6 8.6 2.0 E 007 EFW-PMP-CF All., EFW X11E-XM SFP,'IN XHE NOREO, MFW SYS TRIP, MFW XHE NOREC, PPR SRV CC-1 7 1,'4 3.0 E 008 Elv TDP FC lA, EFW.MDP FC-1B, ErW MDP FC SFP, EFW XHE-NOREC, MFW SYS TRIP, MFW XHE-NOREC, llPl XHE-XM FB
<_ ' ~
sh LOOP Sequence 19 1.2 E 006 s s 20.3 2.5 E 007 EFW FMP-CF Al L, EFW XHE XM ORKR, EFW-XHE-NOREC-L, g-1 OEP-XHE-NOREC4H, llPI X11E XM-FBL p 2 12.8 1.6 E 007 EFW PMP CF-ALL, EFW-XHE XM BRKR, EFW XHE-NOREC-L.
OEP XHE-NOREC4H, PPR-SRV CC 2 3 12.8 1.6 E 007 EFW PMP CF ALL, EFW XHE XM BRKR, EFW XHE-NOREC L, OEP XHE NOREC411, PPR-SRV CC-1 4 0.2 1.1E-007 /EPS-DGN FC lA, EPS-DON FC-18, EAV TDP FC-I A, Elv XHE-XM-BRKR, EFW XHE-NOREC-L, OEP XIIE NOREC4H, HPI XHE-XM FBL 5 5.8 7.2 E-008 /EPS DGN-FC l A EPS-DON FC 1B, EFW TDP-FC-1 A.
EFW XHE-XM BRKR, EFW-XHE NOREC L, OEP-XHE NOREC4H, PPR SRV CC-2 4- 6 5.0 7.2 E 008 /EPS DGN FC-I A. EPS-L ON-FC 18, EFV. U.wFC 1 A, D
EFW XHE-XM-BRKR, EFW XHE-NOREC L.,
= . 11 W
_ A_ - - - _ _ - _ _ . . . - . . _ - _ - ._
LER No. 443/96-003 Table 5. Conditional Cut Sets for Higher Probability Sequences for LER No. 443/96-003 Cut set Percent number contribution CCDP' Cut sets' 7 4.9 6.3 E-008 EPS DON FC 1 A, /EPS-DGN-FC-18. EFW PMP CF ALL.
EFW XI!E NOREC-L, OEP XHE-NOREC4H,llPI XilE-XM-FBL 8 3.1 3.9 E 008 EPS-DGN-FC.l A, /EPS-DGN FC IB, EFW PMP CF ALL, ETW X1tE-NOREC-L,OEP XIIE-NOREC411, PPR SRV CC 1 9 3.1 3.9 E 008 EPS-DGN FC 1 A, /EPS-DGN-FC la, EFW-PMP-CF.ALL, EfW XilE NOREC-L, OEP X11E-NOREC411, PPR SRV CC 2 Total (all sequences) 1.5 E-004 < ^
'The CCDP is determined by multiplying the probability that the porton of the sequencs that makes the precursor sisible (e s., the system with a failure is demanded) will occur durmg the dura 6on of the event by the probabilities of the remaining basic events in the minimal cut set. This can be approximated by I . e*, where p is determined by multiplying the expected number ofinitiators that occe,- during the duration of the event by the probabilities of the basic events in that minimal cut set. The expected number of initiators is gisen by At, where A is the frequency of the initiating event (given on a per hour basis), and t is the duration time of the event (3,87$ h). This approxir.ation is consersative d
for prectirsors made sisible by the initiating event. The frequencies of interest for this evet are:
Aw, a $.3 x 10 /h, A m - 8.6 x 10 % The importance is determined by subtracting the CDP for the same period but with plant equipment assumed to be operating non:inally.
6 Besic event EFW TDP-FC IA is a type TRUE event. This type of event is not normally included in the output of the fault tree reduction process. This event has been added to aid in understanoms the sequences to potential core damage associated with the esent.
12
.; 7: .
l-Enclosure 2 Guidance for Licensee Review of Preliminary ASP Analysis
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GUIDANCE FOR LICENSEE REVIEW 0F PRELIMINARY ASP ANALYSIS
Background
The preliminary procursor analysis of an operational event that occurred at your plant has been provided for your review. This analysis was performed as a part of the NRC's Accident Sequence Precursor (ASP) Program. The ASP Program uses probabilistic risk assessment techniques to provide estimates of operating event significance in terms of the potential for core damage. The types of events evaluated include actual initiating events, such as a loss of off-site power (LOOP) or loss-of-coolant accident (LOCA), degradation of plant conditions, and safety equipment failures or unavailabilities that could increase the probability of core Jamage from postulated accident sequences.
This preliminary analysis was conducted using the information contained in the plant-specific final safety analysis report (FSAR), individual plant examination (IPc), and the licensee event report (LER) for this event.
Nodeling Techniques The models used for the analysis of 1995 and 1996 events were developed by the '
Idaho National Engineering Laboratory (INEL). The models were developed using the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) software. The models are based on linked fault trees. Four types of initiating events are considered: (1) transients, (2) loss-of-coolant accidents (LOCAs), (3) losses of offsite power (LOOPS), and (4) steam generator tube ruptures (PWR only). Fault trees were developed for each top event un the event trees to a supercomponent level of detail. The only support system currently modeled is the electric power system.
The models may be modified to include additional detail for the systems /
components of interest for a particular event. This may include additional equipment or mitigation strategies as outlined in the FSAR or IPE.
Probabilities are modified to reflect the particular circumstances of the event being analyzed.
Guidance for Peer Review Comments regarding the analysis should address:
e Does the " Event Descrietion" section accurately describe the event as it occurred?
e Does the " Additional Event-Related Information" section provide accurate additional information concerning the configuration of the plant and the operation of and procedures associated with relevant systems?
e Does the "Modeling Assumptions" section accurately describe the modeling done for the event? Is the modeling of the event appropriate for the events that occurred or that had the potential to occur under the event conditions? Thfs also includes assumptions regarding the likelihood of equipment reccvery.
, ;'n
- e i
Appendix H of Reference 1 provides examples of comments and responses for previous ASP analyses.
Criteria for Evaluating Consnents Modifications to the event analysis may be made based on the comments that you provide. Specific documentation will be required to consider modifications to the event analysis. References should be made to portions of the LER, AIT, or other event documentation concerning the sequence of events. System and component capabilities should be supported by references to the FSAR, IPE, plant procedures, or analyses. Comments related to o?erator response times 'y and capabilities should reference plant procedures, tie FSAR, the IPE, or applicable operator response models. Assumptions used in determining failure probabilities should be clearly stated.
Criteria for Evaluating Additional Recovery Measures .
Additional systems, equipment, or specific recovery actions may be considered for incorporation into the analysis. However, to assess the viability and effectiveness of the equipment and methods, the appropriate documentation must be included in your response. This includes:
- normal or emergency operating procedures.*
- piping and instrumentation diagrams (P& ids),'
- electrical one-line diagrams,'
- results of thermal-hydraulic analyses, and
- operator training (both procedures and simulator),' etc.
Systems, equipment, or specific recovery actions that were not in place at the time of the event will not be considered. Also, the documentation should address the impact (both positive and negative) of the use of the specific recovery measure on:
- the sequence of events,
- the timing of events,
- the probability of operator error in using the system or equipment, and
- other systems / processes already mod 61ed in the analysis (including operator actions).
For exa;nple, Plant A (a PWR) experiences a reactor trip, and during the subsequent recovery, it is discovered that one train of the auxiliary feedwater (AFW) system is unavailable. Absent any further information regrading this event, the ASP Program would analyze it as a reactor trip with one train of AFW uni.vailable. The AFW modeling would be patterned after information gathered either from the plant FSAR or the IPE.
- However, if information is received about the use of an additional system (such as a standby steam generator feedwater system) in recovering from this event, the transient would be modeled as a reactor trip with one train of AFW unavailable, but this unavailability would be
- Revision or practices at the time the event occurred.
~ 1
.g, mitigated by the use of the standby feedwater system. The mitigation effect for the standby feedwater system would be credited in the analysis provided that the following material was available:
standby feedwater system characteristics are documented in the FSAR or accounted for in the IPE, procedures for using the system during recovery existed at the time of the event,
- the plant operators had been trained in the use of the system prior to the event,
- a clear diagram of the system is available (either in the FSAR, IPE, or supplied by the licensee),
- previous analyses have indicated that there would be sufficient time available to implement the procedure successfully under the circumstances of the event under analysis,-
.the effects of using the standby feedwater system on the operation and recovery of systems or procedures that are already included in the event modeling. In this case, use of the standby feedwater system may reduce the likelihood of recovering failed AFW equipment or initiating feed-and-bleed due to time and personnel constraints.
Materials Provided for Review The following materials have been provided in the package to facilitate your i review of the preliminary analysis of the operational event.
e The-specific LER, augmented inspection team (AIT) report, or other pertinent reports, o A summary of the calculation results. An event tree with the dominant sequence (s) highlighted. Four tables in the' analysis indicate: (1) a summary of the relevant basic events, including modifications to the probabilities to reflect the circumstances of the event, (2) the dominant core damage sequences, (3) the system names for the systems cited in the dominant core damage sequences, and (4) cut sets for the dominant core damage sequences.
Schedule Please refer to the transmittal letter for schedules and procedures for submitting your comments.
References
- 1. L..N. Vanden Heuvel et al., Precursors to Potential Severe Core Damage Accidents: 1994, A Status Report, USNRC Report NUREG/CR-4674 (ORNL/NOAC-232) Volumer. 21 and 22, Martin Marietta Energy Systems, Inc., Oak Ridge National Laboratory and Science Applications International Corp.,
December.1995.
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l Enclosure 3 LER No. 443/96-003
r ,l* r 0 8.,
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North North Atlantie Energy Senice Corporation P.O. Box 300 I
- Atlantic Seah,oot,sii03874 (603)474 9521 The Northeast Utilities System SEP l 2156 Docket No. 50-443 NYN 46062 United States Nuclear Regulatory Commission Attn.: Document Control Desk Washington, D.C. 20555 Seabrook Station Licensee Event Report (LER) 96-03-01 Emergency Feedwater Pumn Mechanical Seal Failure Enclosed, please iind supplemental Licensee Event Report (LER) No. 96-003-01 for Seabrook Station.
This supplement is based upon North Atlantic's August 8,1996, response to a Notice of Violation (NOV) identified in NRC Inspection Report 96-04 regarding incorrect installation of mechanical seals in the turbine driven emergency feedwater (EFW) pump that rendered the pump inoperable. North Atlantic subsequently perfonwd a formal root cause evaluation which is described herein.
Should you require further information regarding this maner, please contact Mr. Anthony M. Callendrello, Licensing Manager, at (603) 474 9521, extension 2751.
Very truly yours, NORTH ATLANTIC ENERGY SERVICE CORP.
WIiliam A. DiPr o Station Director cc: H. J. Miller, Regional Administrator A. W. De Agazio, NRC Project Manager, Seabrook Station J. B. Macdonald, Senior Resident inspector, Seabrook Station INPO Records Center 700 Galleria Parkway Atlanta, G A 30339 -
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, -Hrett9002t-980912 2, PDR ADOCK 05000443 S PDR I
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rau m 05000443 1 of 4 argency Feedwater Pump Mechanical Seal Failure DAlt (D)
DAY YLAM LtM NUMULM (5) i YLAH ntPQMI DAlt I()
bt OVE N ilAL MLVidiUN i
NUMBER MUN IH DAY NUMBER YLAH ' 41W Y hAME OfittM FAcILiist5 INvvLvtD (s)
~21 95 96 003 l ' M a* I NUWE R 01 09 12 96 ' ^"" ' awt WQ mau uuusica THI5 ettPUMT #5 5UMMit t LDt PUM5UdNI
}4Q.24Ul(D) ntuumtMtNi IO Ih i 5 UP 10 4tM 8:
20.24VJeaH4Hv) (Check one or more) (11)
~ 40.22VJiaH1) 100 - 20.2203(aH4)D) bQ. / Ji4H4)D) 1
~29.savJiaH2)ol
~~
bG./JiaHJHvm) go.22uJiaHJHu) 'K bO. /JiaH4Hul
~40.42GJiaHJ)to) bO. /JiakiHal 40.22VJiaH4) bv.iJiaHeHml
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, ,, _40.22VJiaH4Hml ,
DO. IJiaHJHsvl bV.Jbichin U1HLH JU 22WJiaH2)ovl bO.abicHJ) bO. /JiaH2Hvl bpecir v en Abst.act below bO. / JiaH4Hvn) or m NRC Form 366A LILLNuth GUNIAG E TOM IN35 LLH (12)
Anthony M. Callendrello, Ucensing Manager es.t a m. A. m.i Qtu (6031474 9521 x 2751 COMPLtIE suuruNt NI ONE nLINE M ANw AL ivest M ru Ai t FOR EACH COMPONENT FAILU LAvat REpa DESCHIBED ttu suuruNtNi ' IN THIS HEPOHT (131 M ANur AL lunt M nr K A t bOPPLLMENI AL MtPOM T LArkG itD (14)
,q LAPELitU MUNiH dete EXPECTED SUBMISSION DATEl 7
$- SUBMISSION uA1 it AC st to 1400 spaces, i.e., appronmately7b singhe spaced .
typewritten hnes) (16) 16 et approximately 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br /> the Steam Dri -
mg.
r (NSO). During stationed the performance locally at the pump of this testing sparks wereven Emergency re Feedwater Purap (FW.P.37Al w n the mechanical seal assembly. , manually tripped the pump turbme. observed emanating from the outboard mechanica
. A Nuclear lusling outage, (RFO4) which occurred in Novembe DThe pump mechanical seal was disassembled and determ i Specification survedlance tests that were conducted smcr- ecember 1995. This condition was not ap o that time.
FW.P 37A was restored to operable status ay 22 on Murmg post maintenance lormed a formal root cause analysis for this e vent.
for a previously identified event as the pr ivt memtenance techniques. imary causes for thes event.The evaluation identdied a mechanical seal design d f A contnbuting cause for this event wase found to be iciency and inade need the outboard mechanical seat on the turbme dnven em seal assembly.
The system engmeer evaluated the mechanical seals on thAn enspecten the of the mechanica mamtenance procedure which provices guidance e start for perfo up feedwater pump andection determined t
that an mspc em ter detait m the mechanical seat mstallation
)ph4C lailure Caused by Close Clearan s rmmg mamtenance on the emergency feedwater was ection. Additionally. North Atlantic wdi: evaluate apumps mecha wdl be iSuse trammg analysis for sessions, repeat events.and ces, mvestigate perform a reviewpossible monitormg of previous techniques problems i related nical todesign seal detect t change inc p 9 51 i major maintenance of Complex mechanical Quipment eient seel fadures, discuss this
f364A l U.S. NUCLEAR REGULATORY COMMIS$lON UCENSEE EVENT REPORT (LER) 7 NAME (Il TEXT CONTINUATION DOCKET NUMBER (2) LER NUMBER (61 PAGE13)
Seabrook Station 05000443 YEAR SLOVENTIAL REVISION NUMBER NUMBER 2 of 4 96 -
003 -
01 Gwii space is requrred, use addotionaltopres of NRC Form 366M (11) teseriotion of Event r 21,1996 at approximately 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br /> the Turbine Driven Emergency Feedwater P of qustterly surveillance testing. ump (FW P 37A) was started in d mschanical seal area. A Nuclear Systems Operator (NSOL ,
ped the pump stationed locally at The spsrks appeared to be due a mechanical inference within the mechanical seal assembly.
Following the event, a encs inspsetion revealed that the outboard mechanical seal gland on FW P 37A s fit.
was found in the botto seassmbly, the shaft sleeve was discovered to have contacted (rubbed) the The inside diameter of the ove hcd cpproximately 0.005* gouge in it and the throttle bushing was chipped. The inboard seal w .
elserence meancss rssultedb3 tween in contact the within the top of the shaft sleeve and the throttle bushing inside diameter. The cum seal assembly, pon the results fif the mechanical seal inspections on FW P 37A a similar inspection was performed rnstgancy Feedwater Pump, FW P 378. This pump was removed from serviceon and on the electric.
inspected June 4,1996. The in concludsd that similar alignments of the mechanical seats were observed Thison this pump.
suggests that the as suscsptible to the same mechanical rubbing as experienced on the turbine-driven pump i
. Both the inboted and sshaft msking siseves contact withhad concentric the throttle bushing. bumish marks on tnem, as was found ,
on the FW-P a 37A inboard slee roximately 0.0035' clearance between the shaft sleeve and thee inside top). A diameter of the mt reviewitsof if pstforming thefunction design as foundit calleddata bydothe upon to so. system engineer conclud(,J that the electric was emergency feedw lentic reported this event to the NRC on May 24,1996 at 0916 pursuant 10CFRLl?2(b)(1)(ii)(B) , as a condition for an extsnded period of time durin-) operation. This report was later updated o on and results identified in an inspection of the electric-driven emergency feedwater pump mechanical seals,199 ,
ase of Event llantic initially reported this event in June of 1906 with the root cause determined to be that o perating experience similar event in 1987 was not adequately incorporated into design changes, procedures, training ibuting cause for this event had been determined to be a vendor design deficiency of the mechanica in Rsport 96-04 identified that North Atlantic had performed a less rigo sutats with the safety significance of the event.
e
- al soci failure and concluded that two primary causes were determined for this event: North Atlantic perfo rsign deficiency pn ciscrtnces and tolerances of the turbine driven EFW pump's mechanical seals were inwufficient persticq unless the installation technique used non-customary methods (i e
. ., use of dial indicators and feeler gaugest.
gn pstmitted ice betwesn a stack the seal and the upshafoft tolerances sleeve. greater than the clearance. The design did not preclude, as it could ha al seals, This design deficiency also applies to the motor driven EFW oump tractive actions for previously identified event were not adequate to prevent recurren ce ras occurred to the secondary seal of the motor dnven EFW pump inboard seal 9ssembly e in January 1987 prior of the full-power operating license. These failures did not affect the primary seal and there was no loss of The initial failure was at the start of a quarterly surveillance run. There is no wntten , e record of this event bu 166A 14 s5)
36 A U.S. NUCLEAR REGULATORY COMMISSION UCENSEE EVENT REPORT (LER)
NAME (1)
TEXT CONTINUATION ,
DOCKET NUMBER (2) LER NUMBER (6) P?GE t3) 05000443 YEAR $EQUENilAL kE VISION Seabrook Station - NUMBER NUMBER 3 of 4 96 --
003 --
01 N spect is required, use additional copoes of NRC Form 366A) (11) tion of the system engineer was that the seal rub became evident as soon as the pump The .
wasseal started
'ly res waseggin replaced catastrophicand reinstalled, failure withseal.
of the secondary the pump vendor present. When the pump was re started for its maintenan ,
sngineer discovered ssal rub if not installed carefully. the close clearance between the throttle bushing and shaft sleeve an rpsrly installed to the top of its fit so that the throttle bushing, which is press 6d into the ,
sult to of the include 1987 seal the following failures, caution maintenance statement: procedure MS0523.21,
- Emergency Feedwater Pump Maintena rightening tart up.
- the mechanicalsealgland ring botting, the gland ring must be heldin the top ofits fit to prevent se r, this procedure revision did not indicate the need to use a dialindicator to verify that the gland ring was at t tst tightening the bolting, it is believed that this step was omitted since, at that time useorof it was bstisved to be within the skill of the worker, ,
a dialindicat for seal i.e., the worker will use the tools needed to properly accomplish North is specifisdAtlantic does not currently expect workers to utilize dial indicators, or other precision measuring equipm in a procedure. ,
ed its aswith the inadequate dsscribed in LER 96-003-00, corrective actions from the 1987 seal failures, and consistent with the cause of the rece collectively, there was a f ailure to incorporate operating experience from the 1987 em Specificcily, enginesr's andthere was no a mechanic's documentation of the 1987 seal failures. Knowledge of these prior events was limite recollection.
These individuals did not share the lessons leamed from the earlier vith othst personnel who worked on the EFW pump seals. As a result, the maintenance supervisor who performe
- in of the ssal assembly.tha ssels during ORO4 was unaware of the prior events or *s need to take precision sctive p are notaction germane process has changed to the current program. substantially since 1987 and problems with the process that was in existence on atelyin workforrequests ctptured future use. and Adverse Condition Reports (ACRs) will ensure that operating exp Notwithstanding, this event will be discussed at training sessions provided to its personnel to highlight the need to ensure adequate documentation of operating experience. North Atlantic does ve that additional actions are necessary to address the reason for why inadequate corrective actions we o itsd for the seal failures in 1987, nor for the failure to incorpor3te operating experience from these prior events.
ing Causes i maintenance techniques itcondary dsmonstrated seal assembly. that the predictive maintenance techniques for the EFW pumps were unable to identify degradati pphy) could have detected the gradual temperature rise on the shaf t sleeve cause .
ivsis of Event und condition of turbine driven emergency feedwater pump indicated that the pump would not have performed its iction aid hsve if called been ableupon to do its to satisfy so.safety The as found condition of the electric emergency feedwater pump indicated that the function.
snucry 1996 surveillance due to the inadvertent isolation of a pressure sensor.The startup feedwater pump was NAESCO was unable to determine 6eA M-95)
)ORM 344A
)
U.S. NUCLEAR REGULATORY COMMIS$10N UCENSEE EVENT REPORT (LER)
TEXT CONTINUATION 9TY NAME 11)
DOCKET NUMBER (2) LER NUMBER (El PAGE (31 05000443 YEAH SEQUENTIAL 4EV1510N Seabrook Station NUMBE R NUMBER 4 Of 4 96 -
003 --
01 lit more space is reqwred, use additionalcopres of N.9C form 366A) f111 p or when the isolation valve was closed. Furthermore, emergency procedures direct altemate methods of providing l Twetsr to the steam generators in emergency conditions.
Corrective Action m identification of this condition North Atlantic declared the turbine driven EFW pump inoperable and replaced the bo rd mechanical seal. The inboard seal was checked and properly adjusted by centenng the gland. The clearances of h mechenicsi seals on the motor driven EFW pump were also checked and adjusted. The system engineer evaluated the l
- hinicci sagls on the start up feedwater pump and determined that an inspection was not required. North Atlantic prminsd that no other safety related pumps at Seabrook Station utilize a similar seal design. Additionally, the following
) tstm corrsctive actions will preclude an event similar to this from occurring in the future:
Maint<sntnce procedure MS0523.21,
- Emergency Feedwater Pump Maintenance," will be reviewed and revised, as cppropriate, to incorporate greater detail in the mechanical seal maintenance / installation sections to ensure that the seal clacrences are adequate to preclude failure. The review and enhancement of procede e MS0523.21 has been completed.
Enginsering will evaluate an EFW pump mechanical seal design change to preclude catastrophic seal failure caused by closs clacrances.
North #tlantic willinvestigate possible monitoring techniques to detect incipient seal failures.
North Atlantic will discuss this event at training sessions provided to appropriate personnel to highlight the need to ensure adzquats documentation of operating experience.
The rot,t csuse analysis identified a number of issues associated with the disassembly and assembly of the EFW pumps during ths last two refueling outages. As a result North Atlantic will perform a review of previous problems related to major maintenance of complex mechanical equipment and consider performing a root cause analysis if there appears to be a rtpatitive problem.
Additional In formation hs tims of this event the plant was operating at 100% power with the Reactor Coolant System temperature of 587.5' enheit and pressure of 2235 psig.
Similar Events is tha second occurrence where an inadequate f ailure analysis resulted in a similar event .
94 001 01, Reactor Trip and Safety injection due to inadvertent MSIV Closure, describes an event where a reactor kafety in}setion were a result of a Main Steam isolation Valve closed with the plant operating at 100% power. The soot e of this event was determined to be due to an inadequate root cause evaluation of a similar main steam isolation valve it .
Manuf acturer Data e.
CAM 366A e4 95)
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