ML20129B532
| ML20129B532 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/18/1996 |
| From: | Chandu Patel NRC (Affiliation Not Assigned) |
| To: | Sellman M ENTERGY OPERATIONS, INC. |
| Shared Package | |
| ML20129B537 | List: |
| References | |
| NUDOCS 9610230040 | |
| Download: ML20129B532 (25) | |
Text
..
October 18, 1996 y.
i Mr. Michael B. Sellaan Vice President Operations Entergy Operations, Inc.
P. O. Box B Killona, LA 70066 i
SUBJECT:
REVIEW 0F PRELININARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT WATERFORD STEAM ELECTRIC GENERATING STATION, UNIT 3
Dear Mr. Sellman:
i Enclosed for your information is a copy of the final Accident Sequence Precursor (ASP) analysis of the operational' event at Waterford Steam Electric Generating Station, Unit 3 reported in Licensee Event Report (LER)
No. 382/95 002.
This final analysis (Enclosure 1) was prepared by our i
contractor at the Oak Ridge National Laboratory (ORNL), based on review and l
evaluation of your comments on the preliminary analysis and comments received
]
from the NRC staff and from our independent contractor, Sandia National Laboratories (SNL). contains our responses to your specific comments. Our review of your comments employed the criteria contained in the material which accompanied the preliminary analysis. The results of the final analysis indicate that this event is a precursor for 1995.
Please contact me at (301) 415-3025 if you have any questions regarding the enclosures. We recognize and appreciate the effort expended by you and your staff in reviewing and providing comments on the preliminary analysis.
Sincerely, ORIGINAL SIGNED BY:
1 Chandu P. Patel, Project Manager Project Directorate IV-1 1
Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation l
l Docket No. 50-382
Enclosures:
As stated 1
l cc w/encis: See next page DISTRIBUTION:
I
% Docket. File.
PUBLIC PD4-1 r/f i
C. Patel W. Beckner C. Hawes J. Roe J. Dyer, RIV E. Adensam (EGA1)
P. O'Reilly S. Mays i
l Document Name: WATASPFN.LTR 0FC PM:PD4-1 n
(A)LA/PD4-1 Qls NAME CPatel:sp CHawes Omu j
DATE
/0 /18 /96 lD))h/96 COPY h/NO hES)NO a.
OFFICIAL RECORD' COPY
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9610230040 961018 Ogg2 E CENM M
ny\\
umTuo STATES g
g NUCLEAR REGULATORY COMMISSION 1
<t U,
WASHINGTON. D.C. speeHooi October 18, 1996 j
Mr. Michael B. Sellman i
Vice President Operations Entergy Operations, Inc.
1 P. O. Box B j
Killona, LA 70066
SUBJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT WATERFORD STEAM ELECTRIC GENERATING STATION, UNIT 3
Dear Mr. Sellman:
Enclosed for your information is a copy of the final Accident' Sequence
. Precursor;(ASP). analysis,3 re'popted~in Licensee'Esent Report' (LER) of. the operational. event'at Waterford -Steam Electric
~ Generating Station, Unit ~
No. 382/95 002. This final analysis (Enclosure 1) was prepared by our contractor at the Oak Ridge National Laboratory (ORNL), based on review and evaluation of your comments on the preliminary analysis and' comments received from the NRC staff and from our independent contractor, Sandia National Laboratories-(SNL). contains our responses to your specific comments. Our review of your comments employed the criteria contained in the material which accompanied the preliminary analysis. The results of the final analysis indicate that this event is a precursor for 1995.
Please contact me at (301) 415-3025 if you have any questions regarding the enclosures. We recognize and appreciate the effort expended by you and your staff in reviewing'and providing comments on the preliminary analysis.
Sincerely.
Cla d> s' PM Chandu P. Patel, Project Manager Project Directorate IV-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-382
Enclosures:
As stated cc w/encls: See next page
Mro Michael B. Sellman l
Entergy Operations, Inc.
Waterford 3 cc:
Administrator Regional Administrator, Region IV Louisiana Radiation Protection Division U.S. Nuclear Regulatory Commission Post Office Box 82135 611 Ryan Plaza Drive, Suite 1000 Baton Rouge, LA 70884-2135 Arlington, TX 76011 Vice President, Operations Resident Inspector /Waterford NPS Support Post Office Box 822 Entergy Operations, Inc.
K111ona, LA 70066 P. O. Box 31995 Jackson, MS 39286 Parish President Council St. Charles Parish Director P. O. Box 302 Nuclear Safety Hahnv111e, LA 70057 Entergy Operations, Inc.
P. O. Box B Executive Vice-President Killona, LA 70066 and Chief Operating Officer Entergy Operations, Inc.
1 Wise, Carter, Child & Caraway P. O. Box 31995 P. O. Box 651 Jackson, MS 39286-1995 Jackson, MS 39205 Chairman General Manager Plant Operations Louisiana Public Service Commission Entergy Operations, Inc.
One American Place, Suite 1630 P. O. Box B Baton Rouge, LA 70825-1697 Killona, LA 70066 Licensing Manager Entergy Operations, Inc.
P. O. Box B Killona, LA 70066 Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 i
i
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LER Nr. 382/95-002 LER No. 382/95-002 Event
Description:
Reactor trip, breaker failure and fire, degraded offsite power, and degraded shutdown cooling Date of Event: June 10,1995 Plant: Waterford 3 Event Summary A switchyard lightning arrestor failure caused a trip from 100% power at Waterford 3. Delayed opening of the 4.16-kV unit auxiliary transformer (UAT) feeder breaker paralleled the grid with the main generator which was speeding up. The resulting out-of-phase condition caused an overvoltage and fault-level currents that started a fire that damaged cables and switchgear for nonvital Train A.
Power wa initially lost to Train A safety loads, but was recovered when emergency diesel generator (EDG) A started and loaded. Condenser vacuum was subsequently lost as a result ofloss of power to balance of plant Train A equipment and the unexpected bypass of circulating water flow around the condenser. Plant cooldown was delayed when low hydraulic fluid levels prevented proper operation of shutdown cooling (SDC) system isolation valves. The conditional core damage probabihty (CCDP) estimated for this combined event is 2.5 x 10~' The increase in CCDP over a one-year period because of the unavailability of the SDC isolation valves is 1.7 x 10-5 l
Event Description Waterford 3 was operating at 100% power on June 10,1995. At 0858 h, a lightning arrestor failed at the Waterford Substation. The resulting grid disturbance caused the sudden pressure relay on Main Transformer A to actuate the main generator lockout relays. This actuation resulted in the trip of the main generator output breaker and exciter field breaker initiation of a fast dead bus transfer, and trip of the main turbine.
The B 6.9-kV and 4.16-kV buses successfully transferred to Startup Transformer (SUT) B. Ilowever, during the transfer of 4.16-kV bus A2 to SUT A, the A2 SUT feeder breaker closed before the A2 UAT breaker opened. The UAT and SUT breakers tripped and power was lost to bus A2.
The reactor tripped on low Departure from Nucleate Boiling Ratio (DNBR) signals, caused by low reactor coolant pump speed. Bus Al (6.9-kV) deenergized, which tripped two reactor coolant pumps, circulating water pumps, condensate pumps, and condenser air evacuation pumps. Main feedwater (MFW) pump A also tripped, apparently from loss of power to the pump speed pickups.
Vital 4.16-kV bus A3 deenergized when power was lost to bus A2. EDG A started and reenergized the required safety-related loads via the load sequencer. Emergency feedwater (EFW) actuated, and within 13 min both MFW isolation valves had been closed a: e result ofhigh steam generator (SG) level.
Approximately one min aller the trip, all turbine generator building (TGB) switchgear room fire alarm annunciators actuated. The TGB operator reported heavy smoke coming from the switchgear room 7 rhin later. Two auxiliary operators were directed to set up blowers to help dissipate the smoke, don protective clothing, and enter the switchgear room to investigate the cause of the smoke.
At 0935 (+37 min), the TGB auxiliary operator reported a fire in the 2A switchgear and in the cables above the switchgear. The fire was caused by the delayed opening of the A2 UAT breaker, which resulted in a voltage across the I
ENCLOSURE 1
j LER No. 382/95-002 breaker during opening beyond the breaker's design and a subsequent high-energy fault. The breaker failed internally and caused the fire (the breaker failure and fire are described in more detail in Additional Event-Related Information).
Upon notification of an actual fire in the switchgear room, the shift supervisor sounded the plant fire alann (post-event review indicated that the fire alann should have been sounded when smoke was first detected), dispatched the fire brigade, and directed the motor-operated disconnect for SUT A to be opened to ensure electricalisolation of the A2 bus.
The control room supen'isor left the control room to serve as fire brigade leader.
The fire brigade attempted to extinguish the fire using halon, carbon dioxide and dry-chemical fire extinguishers. When the fire brigade leader arrived at the fire scene, he immediately notified the control room to request ofTsite fire depanment assistance. The Ilahnville Fire Department was contacted at 094 I h (+43 min) via 911 for support.
The Ilahnville Fire Department arnved on-site 17 min later and recommended that water be used to extinguish the fire.
Carbon dioxide and dry chemical extinguishers were being unsuccessfully used by the fire brigade to fight the fire (although experience gained from the 1976 Browns Feny fire and other fires indicated that the use of water was necessary on large cable fires). The use of water was delayed for an additional 20 min. (Ref. 2 noted that interviews conducted with plant operators aner the event indicated a general reluctance on the part of the operators to apply water to an electrical fire, based on previous training that had emphasized the use of water was a last reson on electrical fires.)
The fire was extinguished within 4 min, once water wr.s used.
At 1112 h (+2.2 h), condenser vacuum was broken aller it had fallen to 20 in. IIg. A condenser low vacuum alann had actuated at 0940 h, shonly aller the fire was reponed The loss of vacuum was initially attributed to the unavailability of the two circulating water and condenser air evacuation pumps, resulting from the deenergization of bus Al at the beginning of the event, combined with several steam loads that were still discharging to the condenser, and the operators j
made a decision not to divert resources from fighting the fire to attempt to recover condenser vacuum. In actuality, when the two circulating water pumps deenergized at 0858 h, their associated motor-operated discharge valves also j
deenergized and remained open, resulting in a bypass of circulating water flow.
At i 147 h (+2.8 h), the main steam isolation valves were closed and the atmospheric dump valves used for decay heat removal. At 2348 h (14.8 h aner the event began), EFW was secured, and Condensate Pump B (the operable condensate pump) was used to supply water to Steam Generator B.
By 1257 h on June 11,1996, the plant had been cooled down and depressurized to shutdown cooling entry conditions.
At 1311 h, shutdown cooling suction header isolation valve SI-405B was commanded open while placing the shutdown cooling system in senice. This valve closed aner only partially opening and was declared inoperable. The equivalent valve in Train A, SI-405A was then opened. Several hours later, this valve's hydraulic pump was observed to be continually nmning instead of cycling as designed. Valve SI-405A was also closed and declared inoperable.
4 A containment entry was made to inspect the two valves, and low hydraulic fluid levels were found in both valve actuator reservoirs. Approximately 200 in? of hydraulic fluid were added to the resenoir for SI-405B, and the valve operated satisfactorily. Shutdown cooling loop B was placed in senice between 1800 h and 2400 h on June 12,1996.
When valve SI-405A was tested aner fluid had been added to its reservoir, the valve opened slowly. Additional troubleshooting indicated that the valve's hydraulic pump had been damaged by the continuous operation caused by the low hydraulic fluid level. The pump was replaced and the valve was retumed to senice shortly aner midnight on June 13,1995. Cooldown to Mode 5 began, with Train A components still powered by EDG A.
2
LER No. 382/95-002 Additional Event-Related Information The Waterford 3 fast dead bus transfer scheme consists of automatic or manual transfer ofin-house loads from the UATs to the SUTs. During a fast dead bus transfer, the UAT feeder breakers to the A1 and B16.9-kV and the A2 and B2 4.16-kV buses are designed to open in five cycles, and the SUT feeder breakers are designed to close in seven cycles, resulting in a two-cycle nominal deadband on the respective buses.
This scheme is a " simultaneous" (simultaneous trip and close signals with no interlock) bus transfer scheme (zero to two-cycle deadband) instead of the " sequential" (the tripping breaker interlocked with the closing breaker) bus transfer (greater than six-cycle deadband) commonly used in the United States. The simultaneous bus transfer scheme is used in all Swedish nuclear power plants. To prevent exceeding the fault duty of associated equipment and buses when two sources are in parallel, the Swedish design includes an interlock that limits the time period during which both breakers are permitted to remain closed to less than 0.1 sec. The Waterford 3 design does not include the interlock, and both breakers appeared to have remained closed for about 0.3 sec during the event.
During the time that the two breakers were simultaneously closed, the A2 bus connected SUT A to the main generator, which then provided power to the grid via the UAT and bus A2. During this time the main generator was rotating faster than the system frequency due to the load rejection. When the UAT breaker opened, the main generator was approaching 180 degrees out of phase with the system (~8 kV across the breaker). The interrupted current was ~28,800 A. This overvoltage due to the out-of-phase condition and the overcurrent resulted in an intemal breaker failure and the creation of ionizing gases, which caused the fire in the A2 switchgear. A preliminary investigation indicated that the most probable cause for the slow opening time of the UAT breaker was restricted movement of the trip latch roller bearing.
The amount of damage to the breaker and sunounding equipment indicates that (1) the fault cunent through the breaker was extremely high and (2) significant arcing occurred for some period of time. The are chutes and main contacts on all phases were destroyed, and the contact structures, breaker frame, and cubicle were also significantly damaged. The main bus and bus enclosure also appeared to have experienced severe arcing damage.
The fire that resulted from the breaker failure damaged the bus and surrounding cables and components. Two cubicles (the failed breaker was an end cubicle) were heavily damaged, and approximately 10 R of the cable bus duct was destroyed. Cables in approximately a 5-ft diameter column above the breaker had visible fire damage over their entire 10-ft vertical run. At the top of the vertical run, the cables were routed through a horizontal cable tray. Approximately 8 8 of cable in the horizontal tray had visible fire damage. General smoke and slight heat damage to the exterior of the remaining cubicles in the A2 bus occurred. In addition, damage included external heat to the jackets of four of the 15 feeder cables from the SUT to the A2 bus, and burn marks on the conduit of the cables that supply 6.9-kV power to the reactor coolant pump 1 A and 2A motors.
The TGB switchgear room contains both the A and B trains of nonvital switchgear. The ceiling of the room is approximately 2's f1 above the floor; the tops of the switchgear cubicles are approximately 7 A high. A 10-fi-high concrete block radiant heat shield, k>cated 6 n from the front of each set of cubicles, separates the two trains. The fire did not afTect the Train B switchgear or cables.
The TGB switchgear room had an ionization-type fire detection system, with detectors mounted on the ceiling, but no fire suppression system. The fire detection computer recorded the first fire alarm 55 see aner the reactor trip. Within 7 sec, all 36 fire detectors in the room had alarmed. Twenty-six min aller the trip, the first detector went into " device 1
communication error"; it apparently failed at that time and melted. By 0942 h (+44 min), all detectors in the room had apparently failed.
3
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LER No. 382/95-002 4
4 Subsequent to the fire, the licensee found tape over the fire alann annunciator buzzer located on the fire detection computer in the control room. Because of the tape, the alarm volume was low and nonintmsive. Due to the alarm panel's placement in the control room, alarm lights were also not readily visible. These factors, combined with the fact that the fire was not declared until aller the auxiliary operators entered the switchgear room and obsened it (36 min after l
the fire alarm annunciators actuated), contributed to the delay in responding to the fire.
Unlike many PWRs, the Waterford primary pressure relief system includes only code safety valves; no power-operated relief valves (PORVs) are incorporated in the design. The lack of PORVs prevents the use of feed and bleed for core cooling in the event both main and emergency feedwater systems are unavailable. If both of these systems were to fail at Waterfbrd, safety-related secondary-side atmospheric dump valves could be used to depressurize the steam generators to below the shutoff head of the condensate pumps. These pumps could then be used for decay heat removal.
Modeling Assumptions The event was modeled both as (1) a reactor trip, loss of main feedwater (caused by the loss of condenser vacuum 2.2 i
h after the trip), loss of ofTsite power to Train A safety-related components, and unavailability of SDC isolation valves SI-405A and SI-405B during the cooldown (initiating event assessment) and (2) a long-term unavailability of the SDC i
isolation valves (condition assessment).
l Reactor trip, loss of feedwater, and unavailable SDC isolation valves (initiating event assessment)
The ASP mockel for Waterfbrd 3 was revised to address the potential failure of the main feedwater isolation valves j
(MFIVs) to open. These valves were closed because of high SG levels shortly into the event. Failure of these valves 1
i to open would prevent use of the auxiliary feedwater (AFW) system and the condensate system for SG makeup. Short-term ex-control room recoveiy of EFW (beyond the use of the AFW pump), high-pressure injection (HPI), and the condensate system, had these systems failed, was not considered feasible because significant crew resources were being used to fight the fire.
I Redundant shutdown cooling isolation valves SI-405A and SI 405B were both assumed to have failed. This assumption may be conservative for SI-405A because it initially operated. However, the licensee determined that the valve's
)
hydraulic motor was sufliciently damaged to require replacement before the plant cooldown continued.
l The ASP models for a transient do not currently address the potential unauilability of offsite power to an mdividual train, as was obsen ed in this event. During the event, power to safety-rele x1 Train A loads was provided by EDG A.
3 The potential failure of the EDG to power Train A was modeled by adding a basic event to the model, EPS-DGN-FC-1 3AFR, to represent the potential failure of the EDG to start and run following the breaker failure.
j The mission time for the initiating event assessment was assumed to be the time from the reactor trip until shutdown i
cooling was established,-60 h. EDG A continued to supply Train A loads beyond this time. Ilowever, the added risk l
is considered to be small compared with the risk before shutdown cooling was established. [The Accident Sequence Precursor (ASP) program addresses shutdown-related events that are considered unusual and significant. Events such as this one, where one train is powered from its EDG, are not typically selected for analysis.]
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4 4
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i
LER No. 382/95-002 The following changes were made to basic events to reflect conditions observed during the event:
Basic event Revised probability Description (reason for chance)
AFW-TRAIN-FC-ALL 9.8 x 10
Nonsafety AFW system fails to proside flow to SGs (revised to reflect extended mission time) j COND-PFS FC-SYS 7.8 x 10
Secondary heat removat using condensate system fails (revised to reflect extended mission time and low probability ofinitial condensate system failure) 4 EFW-MDP-FC-A, B 5.0 x 10 EFW motor-driven pump train failures (revised to reflect extended mission time)
EFW-PMP-CF-ALL 2.0 x 10d Common cause failure of EFW pumps (revised to reflect extended mission time)
EFW-TDP-FC-TDP 4.1 x 10
EFW turbine-driven pump train failures (revised to reflect extended mission time)
EFW-XIIE-NOREC TRUE Ex-control room resources required for recovery utilized to fight fine 4
EPS-DON-FC-3AFR 1.4 x 10 EDG A fails to start and mn (revised to reflect extended mission time)
IIPI-XIIE-NOREC TRUE Ex-control room resources required for recovery utilized to fight fire MFW-SYS-TRIP TRUE MFW system trips (MFW unavailable because of loss of condenser vacuum)
MFW-VLV-CF-MFIV 2.6 x 10d Common cause failure of the MFW isolation valves to open (basic event added to model because these valves all'ect both the AFW and the condensate systems)
MFW-X11E-NOREC TRUE Operator fails to recover MFW (MFW not recoverable because ofloss of vacuum)
RHR-MOV-CF-SUCT TRUE Common cause failure of residual heat removal (RIIR) suction valves (set to TRUE to reflect the failure of SI-405A and SI-405B)
The mission time for the HPI pumps was not revised to reflect the 60-h mission time. If a transient-induced loss-of-coolant accident (LOCA) had occuned, the modeled plant response would have been accomplished in less than 24 h.
With the SDC isolation valves unavailable following a transient-induced (small-break) LOCA, the operators would have transferred to high-pressure recirculation (1 IPR) once the refueling water storage pool was depleted. This transfer would have occurred ~6 h following the LOCA.
The licensee addressed this specific switchgear room fire in the Waterford Individual Plant Examination for External Events (IPEEE) (Ref. 3). In that document the licensee concluded that the fire - while extensive and not suppressed until 5
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4 LER No. 382/95-002 I
the cables from the UAT to the switchgear were fully involved - did not cause significant damage outside the 1
plume / ceiling jet. Fire modeling also confinned that a large TGB switchgear fire would not generate a hot gas layer that could fail cables outside the plume. Because of this, the IPEEE assumed that TGB switchgear fires would only cause i
damage to one train ofofTsite power.. This assumption was used in this analysis as well. A sensitivity analysis, described i
in the Analysis Results, addresses the potential impact if the fire, or common cause breaker problems, had also resulted i
in a nonrecoverable loss of ofTsite power to Train B.
I;mg-term unavailabilhy of the SDC isolation valves (condithm assessment)
The SDC isolation valves were assumed to have been unavailable since the last refueling outage, in Spring 1994. The longest time period used to assess a condition (unavailability) in the ASP Program is one year, during which the plant j
is typically assumed to have been at power 70% of the time. In this event, however, Waterford was at power for the full 1-year period, resulting in an unavailability of 8760 h. (Because a duration of 8760 h is longer than that used in the j
j analysis of a typical long-term condition, the analysis results cannot be directly compared with those of other long-term condition assessments.) This assumption presumes that the loss of hydraulic fluid from the valve actuators occurs during i
valve operation (not when the valves are inoperative) and that the fluid level during the previous use of the valves was j
barely acceptable. If the hydraulic fluid was lost when the valves were in standby, then the analysis duration is overestimated (the valves would then become unavailable at one-half of the duration since last use; this would result in i
a 50% reduction in the increase in core damage probability caused by the failed valves).
Consistent with the previous assessment, shutdown cooling isolation valves SI-405A and s!-405B were both assumed j
to be failed. This assumption was reflected by setting basic event RHR-MOV-CF-SUCT to TRUE. Plant response to all initiators addressed in the ASP model was considered impacted by the unavailability of the SDC isolation valves.
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Analysis Resuhs
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The CCDP estimated for trip, fire and resulting loss of ofTsite power to Train A, loss of feedwater, and unavailability 4
of the SDC isolation valves is 2.5 x 10 The dominant sequence, highlighted on the event tree in Fig.1 (transient sequence 19), contributes about 83% to the conditional probability estimate for the initiating event and involves
+ the successful reactor trip, 4
+ failure of EFW (including the AFW pump) to provide secondary-side cooling,
- MFW unavailability, and i
+ failure of the condensate system as an attemate source ofcooling water.
3
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The dominant cut sets involve failure to provide an attemate source of water to the EFW pumps following depletion of
?
the condensate storage pool within the 60 h mission time and failure of the condensate system to provide flow to the steam generators (failure to initiate and equipment failure both contribute).
1 Table i provides the definitions and probabilities for selected basic events for the initiating event assessment. The j
conditional probabilities associated with the highest probability sequences are shown in Table 2, while Table 3 lists the sequence logic associated with the sequences listed in Table 2. Table 4 describes the system names associated with the i
dominant sequences. The minimal eut sets associated with each sequence are shown in Table 5.
The calculation for the reactor trip and fire is sensitive to the assumption that the fire or potential common cause breaker failures would not afTect the availability of offsite power to Train B. If the fire could have affected Train B, or if slow breaker opening also resulted in the loss of Train B switchgear (which is believed to be unlikely), then the event could
]
have been more significant. For example, an assumption of a 0.03 probability of nonrecoverable loss of offsite power to Train B (similar to Train A) results in an estimated CCDP of 1.4 x 10d (such an event would be considered significant from an ASP standpoint).
l i
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4
LER No. 382/95-002 l
The unavailable SDC isolation valves (the condition assessment) result in an overall increase in core damage probability 4
l for the assumed 1-year pericxi of 1.7 x 10 over the nominal core damage probability (CDP) estimated for the same 4
period of 8 8 x 10 This is the sum of the changes to the sequence prababilities (importance) shown in Table 7, which are calculated by subtracting the total CDP sequence value from the total CCDP sequence value for each sequence. The i
dominant core damage sequence involves l
l a small-break LOCA initiating event, l
successful EFW and iIPI operation, successful depressurization, failure to initiate SDC (which would avoid the use of high-pressure sump recirculation), and a
failure of high-pressure recirculation.
l For most ASP analyses of conditions (equipment faihrs over a period of time during which postulated initiating events could have occurred), sequences and cut sets associated with the observed failures dominate the CCDP (the probability of core damage over the unavailability period, given the observed failures). The increase in core damage probability (CDP) because of the failures is therefore essentially the same as the CCDP, and the CCDP can be considered a reasonable measure of the significance of the observed failures.
For this event, however, sequences unrelated to the SDC isolation valves dominate the CCDP estimate. The increase l
in CDP given the failed SDC isolation valves,1.7 x 104,is, therefore, a better measure of the significance of the SDC l
valve problems.
Definitions and probabilities for selected basic events for the condition assessment are shown in Table 6.
The conditional probabilities associated with the highest probability sequences are shown in Table 7. Table 8 lists the sequence logic associated with the sequences listed in Table 7. Table 9 describes the system names associated with the dominent sequences. Cut sets associated with each sequence are shown in Table 10.
Acroriyms AFW auxiliaiy feedwater ASP accident sequence precursor ATWS anticipated transient without scram CCDP conditional core damage probability CDP core damage probability j
cd core damage DNBR departure from nucleate boiling ratio l
EDG emergency diesel generator l
IIPI high pressure injection l
IIPR high pressure recirculation IPEEE individual plant examination for external events l
kV kilovolts LOCA loss-of-coolant accident LOOP loss of oilsite power MFIV main feedwater isolation valves MFW main feedwater PORV power-operated relief valve RCS reactor coolant system RIIR residual heat removal RWSP refueling water storage pool 7
I LER No. 382/95-002 SDC shutdown cooling SG steam generator SRV safety relief valve SUT startup transformer TGB turbine generator building UAT unit auxiliary transformer References 1.
LER 382/95-002, Rev. O, " Reactor Tnp and Non-Safety Related Switchgear Fire," July 7, I 995.
2.
NRC Augmented Inspection Team Report 50-382/95-15, July 5,1995 3.
Il'aterfon) 3 Irulividual Plant Eraminationfor External Events, July 1995.
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LER No. 382/95-002 Table 1. Definitions and Probabilities for Selected Basic Esents for the Initiating Event Assessment for LER 382/95-002 Modified Esent Base Current for this name Description ptohability probability Type event IE-LOOP Loss of Offsite Power initiating 8.6 E-006 0.0 E+000 IGNORE No Event IE-SOTR Steam Generator Tube Rupture 1.6 E 006 0.0 E+000 IGNOIE No initiating Event IE-SLOCA Small Loss-of-Coolant Accident 1.0 E-006 0.0 E+000 IGNORE No initiating Event IE-TRANS Transient Initiating Event 6.8 E-004 1.0 E+000 Yes AIV. TRAIN-FC-ALL AFW Pump Train Fails to 8.7 E-003 9.8 E-003 Yes Provide flow COND-PFS-FC-SYS Secondary llent Removal Using 1.5 E402 7.8 E-003 Yes Condensate System Fails COND-XIIE-XM Operator Fails to Initiate 1.0 E-002 1.0 E-002 No Secondary Cooling EFW MDP-FC-A EFW Motor-Driven Pump A 3.9 E-003 3.9 E-003 No Failures EFW-MDP-FC-H EFW Motor-Driven Pump B 3.9 E-003 5.0 E-003 Yes Failures EFW-FMP-CF-ALL Common Cause Failure of Elv 1.4 E-004 1.4 E-004 No Pumps
)
)
EIM-TDP-FC-TDP EFW Turbine-Driven Pump 3.6 E-002 4.0 E-002 Yes 1
Train Failures El%XIIE-NOREC Operator Fails to Recover EFW 2.6 E-001 1.0 E+000 TRUE Yes j
System EFW-XilE-XA-CCW Operator Faik to hiitiate 1.0 E-003 1.0 E-003 No Backup Water Source EPS-IXIN-FC-3AFR EDG 3A Fails to Start and Run 0.0 E+000 1.4 E-001 NEW Yes llPI-IIDV-OC-SUCB Refueling Water Storage Pool 1.4 E-004 1.4 E-004 No (RWSP) Suction Train B i
Failures IIPI-MDP-CF-ALL Common Cause Failure ofliigh 1.0 E-004 1.0 E-004 No Pressure injection Motor-Driven Pumps 10
I,ER No. 382/95-002 Table 1. Definitions and Probabilities for Selected Basic Events for the Initiating Event Assessment for LER 382/95-002 l
1 Modified Event Base Current for this name Description probability probability Type event IIPI-MDP-FC-B IIPI Motor-Driven Pump B 3.9 E-003 3.9 E-003 No Train failures
!!PI-MOV-CF-ALL Common Cause Failure of 5.5 E-005 5.5 E-005 No injection Motor 4)perated Valves
!!PI-XIIE-NOREC Operator Fails to Recover the 8 4 E-001 1.0 E+000 TRUE Yes
!!PI System MFW-SYS-TRIP Main Feedwater System Trips 2.9 E 001 1.0 E+000 TRUE Yes I
i MI'W-VLV-CF-MFIV Common Cause Failure of 0.0 E+000 2.6 E-004 NEW Yes MFIVs to Open MFW XilE-NOREC Operator Fails to Recover Main 3.4 E-001 1.0 E4000 TRL:8 Yes Feedwater PCS-VCF IIW Turbine Dypass Valves /
1.0 E-003 1.0 E-003 No Condensate / Circulation Failures PCS-X11E-XM-CIX)WN Operator Fails to initiate 1,0 E-003 1.0 E-003 No Cooldown PPR-SRV-CO-TRAN Safety Relief Valves (SRVs) 2.0 E-002 2.0 E-002 No Open During Transient PPR-SRV-OO-1 SRV 1 Fails to Rescat 1.6 E-002 1.6 E-002 No PPR-SRV OO-2 SRV 2 Fails to Rescat 1.6 E-002 1.6 E402 No RIIR-MOV-CF-SUCT Common Cause Failure of RilR I.2 E-003 1.0 E~000 TRUE Yes 4
Suction Valves i
i1
LER No. 382/95-002 Table 2. Sequence Condithmal Probabilities for the initiating Event Assessment for LER 382/95-002 Conditional core Event tree damage Percent name Sequence name probability contribution (CCDP)
TRANS 19 2.0 E-005 82.9 TRANS 18 2.2 E-006 9.1 TRANS 24 6.6 E-007 2.6 TRANS 08 4.7 E-007 1.9 Total (all sequences) 2.5 E-005 Table 3. Sequence legic for Dominant Sequences for the Initiating Es ent Assessment for LER 382/95-002 i
Event tree name Sequence name logic TRANS 19
/RT, EFW, MFW, /SRV-RES, COND TRANS 18
/COND, COOLDOWN I
TRANS 24
/IIPI, COND TRANS 08
/RT, /EFW, SRV, SRV-RES,IIPI 12
LER No. 382/95-002 Table 4. System Names for the Initiating Event Assessment for LER 382/95-002 System name logic COND Secondary 1leat Removal Using Condensate System Fails COOLDOWN RCS Cooldown to RIIR Pressure Using Turbine-Bypass Valves, etc.
EFW No or Insuflicient EFW Flow IIPI No or Insuflicient iIPI System Flow MFW Failure of the Main Feedwater System RT Reactor Fails to Trip During Transient SRV SRVs Open During Transient SRV-RES SRVs Fail to Rescat l
l 13
LER No,382/95-002 Table 5. Condithmal Cut Sets for fligher Probability Sequences for the Initiating Event Assessment for LER 382/95-002 Cut set Percent Conditional number contribution probability
- Cut sets' TRANS Sequence 19 2.0 E-005 1
48.2 1.0 E-005 EFW XilE-XA CCW, EFW-XIIE-NOREC, MFW-SYS-TRIP, MFW-XI!E-NOREC, COND-XI!LXM 2
37.6 7.8 E-006 EFW XIILXA-CCW, EFW-XIIE-NOREC, MFW-SYS-TRIP, MFW XIIE-NOREC, COND-PFS-FC-SYS 3
6.8 1.4 E-006 EFW-PMP-CF ALI, EFW XilE-NOREC, MFW SYS TRIP, MFW XilE-NOREC, COND-XI!E XM 4
5.3 1.1 E-006 EFW-PMP-CF All EFW XilE NOREC, MFW SYS-TRIP, MFW XIIE-NOREC, COND PFS-FC.SYS 5
1.2 2.6 E-007 EFW-XllE-XA.CCW, EFW XilE-NOREC, MFW-SYS-TRIP, MFW XIIE-NOREC, MFW VLV-CF MFIV TRANS Sequence 18 2.2 E-006 1
43.6 1.0 E-006 EFW-XllE XA CCW, EFW-XI!E-NOREC, MFW-SYS-TRIP, MFW-XilE-NOREC, PCS XI!LXM CDOWN 2
43.6 1.0 E-006 EFW XilE XA-CCW, EFW Nile-NOREC, MFW SYS-TRIP, MFW XilE-NOREC, PCS-VCF-llW 3
6.1 1.4 E-007 EFW-PMP-CF-ALL, EFW-X11E-NOREC, MFW-SYS-TRIP,
)
MFW X!!E-NOREC, PCS-XilLXM CDOWN 4
6.1 1.4 E-007 EFW-PMP-CF-ALL, EFW XilE NOREC MFW-SYS-TRIP, MFW XilE-NOREC, PCS-VCF-ilW TRANS Sequence 24 6.6 E-007 1
24.1 1.(> E-007 Ef W XI!E-XA-CCW, t.FW X11E-NOREC, MFW SYS-TRIP, M M-XIIE-NOREC, PPR SRV-OO-1 COND-X11E-XM 2
24.1 1.6 E-007 EFW-XilE-XA.CCW, EFW-XilE-NOREC, MFW-SYS-TRIP, MFW X11E-NOREC, PPR-SRV-OO-2, COND-XilE-XM 3
18.8 1.2 E-007 EFW-XIIDXA-CCW, EFW-XIIDNOREC, MFW-SYS-TRIP, MFW XI!E-NOREC, PPR-SRV-OO-1, COND-PFS-FC-SYS 4
18.8 1.2 E-007 EFW-XilE-XA-CCW, EFW X11LNOREC, MFW.SYS-TRIP, MFW-XIIE-NOREC, PPR-SRV-OO-2, COND-PFS-FC-SYS 5
3.4 2.2 E-008 EFW-PMP-CF ALL, EFW-X11LNOREC, MFW-SYS-TRIP, MFW-X1IE-NOREC, PPR-SRV OO-1, COND-XIIE-XM 6
3.4 2.2 E-008 EFW-PMP-CF AIL EfW-XilE-NOREC, MFW-SYS TRIP, MFW XilE-NOREC, PPR SRV-OO-2, COND-XilE-XM 14
LER No. 382/95-002 Table 5. Conditional Cut Sets forliigher Probability Sequences for the Initiating Es ent Assessment for 4
LER 382/95-002 Cut set Percent Conditional number contribution probability' Cut sets' 7
2.6 1.7 E-008 EFW PMP-CF-A1.L EFW XilE-NOREC, hiFW-SYS-TRIP, hlFW XilE NOREC, PPR-SRV4)O-1, COND-PFS-FC-SYS a
8 2.6 1.7 E-008 EFW-PMP-CF AIA EFW-X11E-NOREC, AIFW-SYS-TRIP, MFW-XilE-NOREC, PPR SRV-OO-2, COND-PFS-FC-SYS TRANS Sequence 08 4.7 E-007 I
36.6 1.7 E-007 PPR-SRV-CO-TR.AN, PPR SRV4X)-1, EPS DGN-FC 3AFR, 4
IIPI-MDP-FC-II, llPI-XilE-NOREC 2
36.6 1.7 E-007 PPR-SRV-CO-TRAN, PPR SRV-OO 2, EPS-DGN-FC-3AFR, llPI MDP-FC-B,llPI X11E-NOREC 3
6.7 3.2 E-008 PPR-S R V-CO-TRAN, PPR-S RV4X)- 1, I I PI-M DP-C F-Al L, llPI-XilE-NOREC 4
6.7 3.2 E-008 PPR SRV-CO-TRAN, PPR-SRV-OO-2,llPI MDP-CF AI.L.
IIPI-X1tE-NOREC
~
5 3.7 1.7 E-008 PPR-SRV-CO TRAN, PPR-SRV-OO-1,llPI MDP-CF All, ilPI-XIIE-NOREC 6
3.7 1.7 E-008 PPR-SRV-CO TRAN, PPR-SRV-OO-2, llP141DP-CF-ALI, llPI-XilE-NOREC 7
1.3 6.2 E-009 PPR-SRV-CO-TRAN, PPR-SRV-OO-l. EPS-DGN FC-3 AFR, it PI-MOV-OC-SUCB, llP!-XIIE-NOREC 1
8 1.3 6.2 E-009 PPR-SRV CO TRAN, PPR SRV4)O-2, EPS DGN-FC 3AFR, llPI4 TOV-OC-SUCB, llPl XIIE-NOREC Total (all sequences)
M N M5
~...
- The conditional probability for each cut set is determined by multiplying the probability of the initiating event by the probabilities of the basic events in that minimal cut set. The probabihty of the initiating events are given in Table 1 and begin with the designator 'IE" %e probabilities for the basic events are also given in Table L i
- Basic events EFW XilE-NOREC, MFW-SYS-TRIP, MFW-XilE-NOREC, and RllR-MOV CF SUCT are all type TRUE events which are not normally included in the output of fault tree reduction programs. These events have been added to aid in understanding the sequences to potential core damage associated with the event.
1 15 i
i i
~ _..
4
)
LER "o 382/95-002 i
)
Table 6. Definitions and Probabilities for Selected Basic Esents for the Condition Assessment for LER 382/95-002 Modified Event Base Current for this name Description probability probability Type event l
IIPI-MDP-FC-B llPI Motor-Driven Pump B Train Failures 3.9 E-003 3.9 E-003 No llPR-AOV-CC-SMPA Containment Sump A Failures 1.1 E-003 1.1 E 003 No IIPR-AOV-CC-SMPB Containment Semp B Failures 1.1 6003 L1 E-003 No IIPR AOV-CF SMP Common Cause Failure (CCF)of Sump 1.0 E-004 1.0 E-004 No Air Operated Valves llPR-IIDV-CF RWSP CCF of the Isolation flydraulic Discharge 2.0 L004 2.0 E-004 No Valves to the RWSP llPR-ilDV-OO RWSPA RWSP Train A Isolation Hydraulic 2.0 E 003 2.0 E-003 No Diuharge Valve (IIDV) Failures IIPR IIDV-OO-RWSPB RWSP Train B Isolation llDV Failures 2.0 E-003 2.0 E-003 No llPR SMP-FC-SUMP Containment Recirculation Sump Failures 5.0 E-005 5.0 E-005 No
!!PR XIIDNOREC Operator fails to Recover the IIPR System 1.0 E+000 1.0 E+000 TRUE No liPR-XIIDNOREC-L Operator Fails to Recover the IIPR System 1.0 E+000 1.0 E+000 TRUE No During a LDOP 1
MES-VCF-ilW-ISOL Ruptured Steam Generator Isolation 1.0 E-002 1.0 E-002 No l
Failures MSS-XIIE-NOREC Operator Recovery Action for Steam 1.0 E-001 1.0 E-001 No Generator Isolation PPR SRV-CO-L SRVs Open During a LOOP 1.6 E 001 1.6 LOOL No i
l PPR-SRV-CO TRAN SRVs Open During Transient 2.0 E-002 2.0 E-002 No l
PPR-SRV OO-l SRV 1 Fails to Rescat 1.6 E-002 1.6 LOO 2 No PPR SRV OO-2 SRV 2 Fads to Rescat 1.6 E-002 1.6 E-002 No RilR-MOV-CF-SUCT Common Cause Failure of RIIR Suction 1.2 E 003 1.0 E+000 TRUE Yes Valves RIIR XIIE NOREC Operator fails to Recover the RIIR 3.4 E 001 3.4 E-001 No i
System l
RilR X}{LNOREC-L Operator Fails to Recover the RIIR 3.4 E 001 3.4 L001 No System During a 1.DOP RWSP-REFILL Operator Fails to Refill RWSP 8.5 E-003 8.5 E-003 No 16
LER No. 382/95-002 Table 7. Sequence Conditional Probabilities for the Condition Assessment for LER 382/95-002 Conditional core Event tree damage Core damage Importance Percent name Sequence name probability probability (CCDP-CDP) contribution' (CCDP)
(CDP)
SLOCA 03 1.1 E-006 8.5 E-009 1.1 E-006 65.4 TRANS 05 4.8 E-007 3.4 E-009 4.8 E-007 28.7 LOOP 05 8.2 E-008 3.6 E-008 4.5 E-008 2.6 SGTR 04 4.1 E-008 2.9 E-010 4.1 E-008 2.4 Total (all sequences) 9.I E-005 8.9 E-005 1.7 E4X)5
- Percent contribution to the total importance.
i Table 8. Sequence Logic for Dominant Sequences for the Condition Assessment l
for LER 382/95-002 Event tree name Sequence name logic SLOCA 03
/RT, /EFW, /IIPI, /COOLDOWN, RIIR,1 IPR T.UN5 05
/RT, /EFW, SRV, SRV-RES,
/1IPI, /COOLDOWN. RIIR,IIPR LOOP 05
/RT-L, /EP, /EFW-L, SRV-L, SRV-RES, /IIPI-L,
/COOLDOWN, R1IR-L, IIPR-L SGTR 04
/RT, /EFW-SGTR, /ilPI, /RCS-SG, SGISOL, /RCSCOOL, RilR, RWSPREFIL 17
l I
LER No. 382/95-002 Table 9. System Names for the Condition Assessment for LER 382/95-002 System name Logie COOLDOWN RCS Cooldown to RIIR Pressure Using Turbine-Bypass Valves, etc.
EFW No or Insufficient EFW Flow EFW L No or InsuHicient EFW Flow During a LOOP EFW-SGTR No or Insufficient EFW Flow During a Steam Generator 1
Tube Rupture event J
EP Failure of Both Trains of Emergency Power IIPI No or Insuflicient IIPI System Flow IIPI-L No or Insullicient IIPI System Flow During a LOOP 1 IPR No or Insufficient iIPR Flow IIPR-L No or Insufficient IIPR Flow During a LOOP RCS-SG Failure to Lower RCS Pressure to Less Than Steam i
Generator Relief-Valve Set Point q
(
RCSCOOL Failure to Cooldown RCS to Less Than RCS Pressure RIIR No or Insuflicient RHR System Flow RIIR-L No or Insuflicient RIIR System Flow During a LOOP RT Reactor Fails to Trip During a Transient i
RT-L Reactor Fails to Trip During a LOOP RWSPREFIL Operator Fails to Refill RWSP SGISOL Failure to Isolate Ruptured Steam Generator Before RWSP Depletion SRV SRVs Open During a Transient SRV-L SRVs Open During a LOOP SRV-RES SRVs Fail to Rescat 1
18
LER No. 382/95-002 i
Table 10. Conditional Cut Sets for liigher Probability Sequences for the Condition Assessment for LER 382/95-002 Change in Cut set Percent CCDP Cut sets' number contribution (Importance)'
SLOCA Sequence 03 1,1 E-006 1
53.4 6.0 E-007 RllR-MOV-CF-SUCT, RilR X11E-NOREC,IIPR-IIDV-CF-RWSP,
!!PR XIILNOREC l
2 26.7 3.0 E-007 RilR-MOV-CF-SUCT, RllR-XilE-NOREC, IIPR-AOV-CF-SMP,
!!PR X11LNOREC 3
13.3 1.5 E-007 RllR-MOV-CF-SUCT, RllR-XHLNOREC, llPR-SMP-FC-SUMP, llPR X11E-NOREC 1
4 2.0 2.3 E-00g RilR-MOV-CF-SUCT, RIIR-X11E-NOREC, ilPI-MDP-FC-D, 1 IPR-IIDV-OO-RWSPA, IIPR-XlIE-NOREC 1
5 1.1 1.2 E-008 RilR-MOV CF-SUCT, RIIR-XilE-NOREC, llPI-MDP-FC-B, IIPR AOV CC-SMPA,IIPR-X11E NOREC l
6 1.0 1.1 E-008 RllR-MOV-CF SUCT. Rl!R-XIIDNOREC,IIPR-IIDV-OO-RWSPA, IIPR-ilDV-OO-RWSPB, llPR-X11E-NOREC TRANS Sequence 05 4.8 E-007 1
26.7 1.2 E-007 PPR SRV-CO-TRAN, PPR-SRV-OO-1, RIIR-MOV-CF-SUCT, l
RilR-X11E NOREC,itPR-ilDV CF RWSP,!!PR-X11E NOREC l
2 26.7 1.2 E-007 PPR-SRV-CO-TRAN, PPR-SRV-OO-2, RilR-MOV CF-SUCT,
)
RilR XIIE-NOREC, llPR-IIDV-CF-RWSP, llPR-XIIE-NOREC 1
3 13.3 6.5 E-008 PPR-SRV CO TRAN, PPR-SRV-OO-1, RilR MOV-CF-SUCT, RIIR-X1IE-NOREC,1 IPR-AOV-CF-S MP, llPR-XIIE-NOREC
)
4 13.3 6.5 E-008 PPR-SRV-CO-TRAN, PPR SRV-OO-2, RilR-MOV-CF-SUCT, RilR XilLNOREC,IIPR-AOV-CF-SMP,llPR XilE-NOREC 5
6.6 3.2 E-008 PPR-SRV-CO-TRAN, PPR SRV.OO 1, RilR-MOV4F-SUCT, RilR-XIIE-NOREC, HPR-SMP FC-SUMP,11PR-XIIE-NOREC 6
6.6 3.2 E-008 PPR-SRV-CO-TRAN, PPR-SRV-OO-2, RilR-MOV-CF-SUCT, Ri!R XIIDNOREC,IIPR-SMP-FC-SUMP,!!PR-XIIDNOREC 7
1.0 5.1 E-009 PPR-SitV CO-TRAN, PPR SRV OO 1, Ri!R MOV-CF-SUCT, RllR XIIENOREC,llPI-MDP-FC-B,itPR ilDV OO-RWSPA, llPR XilLNOREC 8
1.0 5.I E-009 PPR-SRV-CO-TRAN, PPR-SRV.OO-2, RllR-MOV CF-SUCT, j
RilR-XIIE-NOREC, IIPl-MDP-FC-B, llPR-IIDV-OO-RWSPA, IIPR-XIIE-NORFf 19
4 LER No. 382/95-002 d
Table 10. Conditional Cut Sets for Higher Probability Sequer.ces for the Condition Assessment for LER 382/95-002 i
Change in j
Cut set Percent CCDP Cut sets' number contribution (Importancey LOOP Sequence 05 4.5 E-008 4
d l
1 26.7 1.3 E-008 PPR-SRV-CO-L, PPR-SRV.OO-1, RilR-MOV-CF-SUCT, Ri[R XI!E-NOREC-L llPR-ilDV CF.RWSP,IIPR-X1IE NOREC-L 4
2 26.7 1.3 E-008 PPR-SRV-CO-1, PPR-SRV OO-2, Rl!R MOV-CF-SUCT, RilR X1tE-NOREC-L,IIPR-IIDV-CF-RWSP,IIPR-XIIE-NOREC-L 3
13.3 6.6 E-009 IPR SRV CO-L PPR-SRV-OO 1, RllR-MOV-CF-SUCT, l
s
)
RilR-X1ILNOREC L llPR-AOV-CF-SMP,IIPR XilE-NOREC-L 4
13.3 6.6 E-009 PPR SR / CO-1, PPR-SRV-OO-2, RIIR-MOV-CF-SUCT, i
RIIR XIILNOREC-L,!!PR AOV-CF-SMP,llPR XIIE NOREC-L 1
4 5
6.6 3.2 E-009 PPR SRV-CO-L, PPR-SRV OO 1, RilR-MOV-CF-SUCT, l
Rl!R-X1IE-NOREC-1, IIPR-SMP-FC-SUM P. IIPR-X11E-NOREC-L 6
6.6 3.2 E-009 PPR-SRV-CO-1 PPR-SRV-OO-2, RIIR-MOV-CF-SUCT, l
RllR-XIIE-NOREC-L,llPR-SMP-FC-SUMP,IIPR XilE-NOREC-L i
7 1.0 5.1 E-010 PPR-SRV CO-1, PPR SRV-OO 1, RllR-MOV-CF-SUCT, j
RllR XIIE-NOREC-1,IIPI-MDP-FC-B, !!PR-IID-OO-RWSPA, IIPR-XIIE-NOREC-L g
8 1.0 5.1 E-010 PPR-SRV CO-1 PPR-SRV.OO-2, RliR-MOV CF-SUCT, d
j R11R XIIE-NOREC-L !!PI-MDP-FC-B, IIPR-IIDV-OO-RWSPA, j
IIPR XIIE-NOREC-L 3
SGTR Sequence 08 4.1 E-008 g
i 1
99.7 4.1 E-008 MSS-VCF IlW.lSO!, MSS-XI!E NOREC, RilR MOV-CF-bUCT, RilR XilE-NOREC, RWSP-REFILL Total (all sequences) 1.9 E-005 The change in conditional probabihty (importance) is determined by calculating the conditional probabihty for the period in which the condition existed and given the condition, and subtracting the conditional probability for the same period but with plant equipment assumed to be operating nominally. The conditional probability for each cut set within a sequence is determined by multiplying the probability that the portion of the sequence that makes the precursor visible (e.g., the system with a failure is demanded) will occur during the duration of the event by the probabilities of the remaining basic events in the minimal cut set. This can be approximated by 1 - e*, where p is determined by multiplying the expected number of initiators that occur during the duration of the event by the probabilities of the basic events in that minimal cut set. The expected number ofinitiators j
is given 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 (in this case,8760 h).
This approximation is conservative for precursors made visible by the initiating event. The frequencies ofinterest for this event are: 6 = 6.8 x 10%, b = 8.5 x 10%,6 = 1.0 x 10%, and 6 = 1.6 x 10%
6 Basic event Rl!R-MOV-CF-SUCT is a type TRUE event which is not normally included in the output of fault tree reduction programs. His events has been added to aid in understanding the sequences to potential core damage associated with the event.
20 s
d 4
m_
LER No. 382/95-002 LER No. 382/95-002 Event
Description:
Reactor trip, breaker failure and fire, degraded ofTsite power, and degraded shutdown cooling Date of Event: June 10,1995 Plant: Waterford 3 Licensee Comments
Reference:
Letter from J. J. Fisicaro (Entergy Operations, Inc.) to the U.S. Nuclear Regulatory Commission, " Review of Preliminary Accident Sequence Precursor Analysis," W3F1 0140, August 15,1996.
Comment 1:
The licensee provided many specific comments on the text of the Event Summary, Event Description, and Additional Event-Related Infonnation sections of the analysis documentation concerning the cause of the breaker failure, plant response to that failure, and the design of the bus transfer scheme at Waterford.
Response 1:
With the exception of Comment 4, the clarifications and corrections provided by the licensee were incorporated into the analysis documentation. Because most of the comments were editorial in nature, they have not been repeated below. Those comments of a technical nature are discussed below.
Comment 2:
'Ihe second paragraph,3rd sentence in the Event Description is not correct. There is nothing to support the UAT tripping on overcurrent. The overcurrent relays were set to trip @ > I second for a current of 30000 amps. The event was less than 29000 amps for approximately 0.3 seconds. The power would not have been lost to the A2 bus unless the SUT breaker had l
also tripped.
l Response 2:
The reference to the UAT feeder breaker tipping on overcurrent was changed to indicate that both the UAT and the SUT breakers tripped, and power was lost to bus A2.
I i
l Comment 3:
The 6th paragraph, I st sentence in the Event Description states that the A1 bus de-energized and all of its loads de-energized at the beginning of the event. This sentence should be moved to the beginning of the ever.t.
.1 ENCLOSURE 2 l
l I
LER No. 382/95 002 Response 3:
To preserve the sequence of events, this sentence was moved to the 3rd paragraph in the Event Description.
Comment 4:
The 9th paragraph in the Event Description discusses the use of water on the fire. The recommendation to use water was not made solely by the Volunteer Fire Department. The decision to use water was the result of a methodical analysis performed by the Waterford 3 Fine Brigade Leader and the Voluntary Fire Department Chief. Also, the Fire Brigade was not " reluctant" to use water. They had been trained to consider gas and dry chemical as the preferred options.
Response 4:
This comment pertains to the reluctance of the fire brigade to use water on the fire when carbon dioxide and dry chemical fire extinguishers were proving to be ineffective. The AIT report for the event (Reference 2 to the analysis documentation) noted that all operators indicated in later interviews that they were reluctant to use water on the electrical fire. The applicable paragraph in the Event Description was reworded instead to indicate that the source of this information was the AIT report.
1 Comment 5:
The 10th paragraph,2nd sentence in the Event Description states that a condenser low vacuum alarm had actuated at 0940 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.5767e-4 months <br />, "42 min after the 6.9 kV. " The 6.9 kV bus de-energized at 0858 hours0.00993 days <br />0.238 hours <br />0.00142 weeks <br />3.26469e-4 months <br /> when the transfer to the SUT failed.
Response 5:
This pan of the sentence was deleted.
i 1
Comment 6:
The 5th paragraph in the Additional Event-Related Information section discusses " fire stops."
The design of the Calvert Bus used at Waterford 3 does not employ the use of " fire stops."
Thus, the statement regarding the ineffectiveness of the venical fire stops is inaccurate.
Additionally, the statement that fire damage was limited by the fire stop in the horizontal section is also inaccurate.
Response 6:
All references to fire stops at Waterford 3 have been removed.
i 2