ML20135G735

Final ASP Analysis - Catawba 2 (LER 414-96-001)

ML20135G735
Person / Time
Site:	Catawba
Issue date:	05/14/2020
From:	Christopher Hunter NRC/RES/DRA/PRB
To:
Littlejohn J (301) 415-0428
References
LER 1996-001-00
Download: ML20135G735 (18)

Text

Ant)endix B LER No. 414/96-001 AB.ndx9 LER No. 414/96-001 Event

Description:

Loss of offsite power with emergency diesel generator B unavailable Date of Event: February 6, 1996 Plant: Catawba 2 B.9.1 Event Summary At 1231 on February 6, 1996, Unit 2 was at 100% power when ground faults on the 2A main transformer X-phase and 2B main transformer Z-phase potential transformers resulted in a loss of offsite power (LOOP).

The reactor scrammed and emergency diesel generator (EDG) 2A (train A) started and loaded. EDG 2B was out of service because of a faulty capacitor in the battery charger for that diesel generator. EDG 2B (train B) was returned to service and its emergency bus (2ETB) energized at 1522 (2 h 51 min into the event). Using parts from the 2A main transformer, the 2B main transformer was repaired, and offsite power was restored to Unit 2 on February 8, 1996, at 0120 (36 h 49 min into the event)."8 The conditional core damage probability (CCDP) estimated for this event is 2. 1 x10~.

B.9.2 Event Description At 1231 on February 6, 1996, Unit 2 was at 100% power when ground faults on the resistor bushings for the 2A main transformer X-phase and 2B main transformer Z-phase potential transformers resulted in a LOOP.

The reactor scrammed, and EDG 2A started and loaded. EDG 2B was out of service because of a faulty ac capacitor in the diesel battery charger. EDG 2B was returned to service and its emergency bus (2ETB) energized at 1522 (2 h 51 min into the event). Not all emergency loads on bus 2ETB were energized due to activities in progress to implement a cross-tie to Unit 1. At 1800 (5 h 29 min into the event), the cross-tie activities for train B were completed, and the source for bus 2ETB was transferred to transformer SATB (a Unit 1 B-train offsite power source supplied power to Unit 2 transformer SATB). Initial efforts to complete the cross-tie to bus 2 ETB were unsuccessful because of a procedural inadequacy. At 2000 (7 h 29 min into the event), cross-tie activities were completed for EDG Train A and power was transferred to transformer SATA (a Unit I A-train offsite power source supplied power to Unit 2 transformer SATA). Personnel repaired the 2B main transformer using parts from the 2A transformer and restored offsite power to Unit 2 on February 8, 1996, at 0120 (36 h 49 min into the event). Repairs on the 2A main transformer were not completed until 0327 on February 11, 1996 (62 h 56 min from the start of the event).

At 1236, or 5 min after the LOOP, operators manually closed the main steam isolation valves (MSIVs). At 1238, a safety injection (SI) actuation occurred because of low steam line pressure in the 2A steam generator (SG). At 1247, the pressurizer power-operated relief valve (PORV) 2NC34A began to cycle; at 1310 (39 min into the event), the pressurizer level went off-scale high as the reactor coolant system (RCS) became water solid. At 1320, the pressurizer relief tank (PRT) pressure increased and the PRT rupture disc ruptured as PORV 2NC34A continued to cycle. A steam bubble was reestablished in the pressurizer at 1926 (6 h 55 min into the event or 6 h 16 min after becoming water solid). PORV 2NC34A fully stroked approximately 43 B.9-1 NUREG/CR-4674, Vol. 25

LER No. 414/96-001 AnDendix B times on steam and an additional 31 times on water. A Nuclear Regulatory Commission (NRC) Inspection Team estimated that this PORV came off its closed seat about 110 times. (Evaluations by the licensee of stroke-time tests and visual external inspection concluded that no damage to PORV 2NC34A occurred. The PORV was supplied by Control Components Incorporated. The PRT rupture disc was replaced on February 9, 1996, at 1428.)

At 1641 (4 h I11 min into the event), control room operators received a report of a leak in the penetration room. [It was subsequently determined that three pit sump check valves from the turbine-driven auxiliary feedwater pump (TDAFWP) were leaking into the penetration room.] The TDAFWP was secured at 1759 (5 h 29 min into the event). Water in the pit sump for the TDAFWP was pumped to the turbine building sump. Back leak-age through check valves 2WvL894, 2WL836, and 2WL834 allowed the discharge from the sump for the TDAFWP to fill floor drain sump C, which overflowed onto the auxiliary feedwater (AFW) pump room floor to a level of several inches. [This area is separated from the AFW pump pits by a concrete curb approximately 46 cm (18 in.) high. The floor drain sump "C" is not powered by emergency power and, therefore, would be unavailable until offsite power is restored.] Operators manually closed valves 2WL835 and 2WL836, thereby stopping the water leak-age.

Because of a leak in the instrument air system, the containment was purged by using the containment air release and addition system on February 7, 1996, at 1033. Air leak-age was a recurring problem, as shown by venting data. This data show that Unit 2 was being vented every 12 h prior to this event. During this event, containment temperature increased in response to the loss of containment chilled water to the ventilation units (containment chilled water is not a diesel-backed load). When the PRT rupture disc ruptured, containment pressure increased further [pressure peaked at .0062 MPa (0.9 psig]). This pressure increase was sufficient to partially open some, but not all, of the ice condenser lower inlet doors. Energy absorption was limited to contact with ice in the lowest portion of the ice condenser. Because there was no flow through the ice condenser, the intermediate and upper deck doors did not open.

B.9.3 Additional Event-Related Information Each AFW pump is mounted in a separate pit for net positive suction head (NPSH) requirements. To prevent flooding of these pits, each motor-driven pump pit is supplied with a 190 L/min (50-gpm) sump pump that discharges to the liquid radwaste system. For the TDAFWP, the turbine oil is cooled through the lube oil cooler by a small portion of the discharge flow. The TDAFWP turbine oil cooler flow and a portion of the turbine seal water empty directly into the pit for the TDAFWP. If the sump is not drained, failure of the TDAFWP could occur in as little as three hours. To provide extra assurance that the TDAFWP` will not fail as a result of flooding, the pit for the TDAFWP is outfitted with two 190 L/min (50-gpm) sump pumps. One of these sump pumps can be powered during a LOOP from either EDG 2A or from the standby shutdown facility; the other sump pump is powered from EDG 2B.

A standby shutdown facility (SSF) is located in a separate building on the Catawba site. This facility, which is not normally manned, is capable of providing limited high-pressure injection for RCS makeup and reactor coolant pump (RCP) seal cooling [provided an RCP seal loss-of-coolant accident (LOCA) does not occur].

The SSF includes a separate diesel generator that can power SSFB loads in the event of a station blackout.

The diesel generator for the SSFC can also power one of the sump pumps for the TDAFWP. The SSF NUREG/CR-4674, Vol. 25 B.9-2

Appendix B Appenix No.

BLER4 14/96-001 systems are single trains and, therefore, are susceptible to a single failure. In conjunction with the TDAFWP and the availability of SGs, the SSFs can maintain hot standby conditions for both units. An operator was sent to man the SSF facility during this event; however, the SSF was never started.

The licensee evaluated the flooding of the AFW pump room in its individual plant examination (IPE).

(Recall, the AFW pump room was in danger of being flooded by operating the TDAFWP.) The IPE flood analysis for the AFW pump room evaluates a break in a pipe outside the sump pits. Water will reach the base of the auxiliary shutdown panel at the same time it reaches the top of the curb around the AFW pump pits.

[The curb walls around the pit are 46 cm (18 in. high.)] The lowest point of switches, fuses, or terminal strips within the auxiliary shutdown panel is 20 cm (8 in.) from the base. When water reaches this level, the IPE assumes that equipment controlled from the auxiliary shutdown panel is unavailable. Because the floor area outside the AEW pump room is about 207 m' (2,23 1.6 ft2 ) and the curb is 46 cm (18 in.) high, the estimated time to flood the turbine-driven pump pit area is about 33 h. The leak-age into the turbine-driven pump pit is within the capability of the operating sump pump. If this pump failed, an additional 3 h would be available before the leakage or the water accumulation in the turbine driven pump pit could fail the pump. After the pump pits are flooded, there is an additional area of 103 m' (1,110 ft') in the room for water to cover. The IPE further estimates that 41.6 min are available to isolate a flood of 9,194 L/min (2,429 gpm). Therefore, flooding of the TDAFWP is not considered credible because considerable time was available to mitigate the flooding.

B.9.4 Modeling Assumptions This event was modeled as a LOOP initiator with failure of train B of the emergency power system. Because offs ite power was not restored for about 1'/2 days and both offsite power transformers required major repairs before power could be restored through these transformers, it was assumed that operators could not have restored offsite power during the event. Therefore, the following basic events were set to TRUE (i.e., failed):

1. operator fails to recover offsite power within 2 h (OEP-XHE-NOREC-2H),

2. operator fails to recover offsite power within 6 h (OEP-XHE-NOREC-6H),

3. operator fails to recover offsite power before battery depletion (OEP-XHE-NOREC-BD) , and

4. operator fails to recover offsite power given a seal LOCA (OEP-XHE-NOREC-SL).

In addition, the probability that the PORVs open during a transient (PORV) was set to TRUE because one PORV (2NC34A) lifted more than 74 times.

The ac power to the emergency buses was assumed to be potentially recoverable to the emergency buses by implementing a cross-tie to Unit 1 and by recovering EDG 2B. These actions were assumed to be independent for this analysis given that the event occurred during the day shift. (This assumption would have to be confirmed for an event occurring outside the day shift because it was unknown if sufficient personnel would be available during the period between 5:00 p.m. and 8:00 a.m. to perform all the actions in parallel that were performed during this event.)

The LOOP event tree for Catawba is shown in Fig. B.9. 1. Credit for the SSF at Catawba was accounted for by adding the fault tree shown in Fig. B.9.2 at the SSF branch point in the event tree shown in Fig. B.9. 1.

NUREG/CR-4674, Vol.25 B.9-3 NUREG/CR-4674, Vol. 25

LER No. 414/96-001 Avvendix B The failure probabilities for the basic events SSFB and SSFC were obtained from the Catawba IPE. Basic event SSFB is the failure to provide seal cooling to the reactor coolant pumps; basic event SSFC is the failure to provide power to the sump pump for the TDAFWP.

The recovery of power by implementing a cross-tie to Unit 1 was modeled by adding the basic event failure to cross-tie EDG B emergency bus within 3 h [OEP-XHE-NOCRS-B] to the Catawba fault trees for failure to recover power prior to core uncovery given an RCP seal LOCA (OP-SL, see Fig. B.9.3) and prior to battery depletion given no seal LOCA (OP-BD, see Fig. B.9.4). Failure to cross-tie to Unit 1 was modeled as a time-reliability correlation (TRC) as described in Human Reliability Analysis.' Because sequences of concern in the analysis involve a station blackout, the "recovery with hesitancy" TRC, as described in Chapter I1I of the reference, was used in the analysis. The probability distribution for this TRC is lognormal, with an error factor of 6.4. To reflect the observed time to implement the cross-tie, a median response of 60 min was assumed, following a 30-mmn delay. The probability of crew failure at 3 h, estimated using this TRC and response time, is 0.27..

A single sump pump must be available (requiring emergency power or power from the SSF) within 3 h to prevent failure of the TDAFWP as a result of flooding.

The probability of a seal LOCA was obtained from Evaluation of Station Black-out at NuclearPower Plants,'"

and RCP seal LOCA models were developed as part of the NUREG- 1150 probabilistic risk assessment efforts, as described in Revised LOOP Recovery and PWR Seal LOCA Models." This model assumes that it would take 2 h to uncover the core given a seal LOCA and that the seal LOCA would occur Wlithin 1 h of the station blackout. Therefore, the assumption was made that the core would be uncovered 3 h after the initiation of the station blackout.

The basic event for failing to recover EDG 2B was developed with an exponential repair model with a median repair time of 140 min and a delay of 30 muin (EPS-XHE-EDGB-NOR). (EDG 2B was recovered within 3 h.)

Based on this repair model, the probability of failing to recover EDG 2B is 0.48.

To account for the longer run time of EDG 2A during this event, the failure probability was modified from 0.042 to 0.045. Hence, the mission time for this event was increased from 6 h to 7.5 h while maintaining the same failure to start probability (0.03) and the same failure to run failure rate (0.002/h), as reported in the "ASP Models, PWR B, Catawba Units I and 2" (Ref. 12).

Although the AFW pump room was flooding and the source of the flooding was bearing and oil cooling water from the TDAFWP via the sump to the TDAFWP, the TDAFWP was considered operable with no change in the failure probability because (1) operators isolated the leak and (2) operators would have had at least 33 h to isolate the leak if the TDAFWP had been required to run. The 33 h estimate was obtained by multiplying the time provided in the IPE for isolating a flood (41.6 ruin) by the assumed flooding rate in the IPE 9,194 L/min (2,429 gpm) and dividing by the maximumn sump pump flow for the TDAFWP. This result is then converted to hours by dividing by 60 ruin/h [i.e., (41.6 ruin) x (9,194 L/ruin )(190 L/min) -(60 ruin/h)

ý33 h].

Vol.25 B.9-4 NUREG/CR-4674, NUREG/CR-4674, Vol. 25 B.9-4

ADDendix B LER No. 414/96-001 B.9.5 Analysis Results The CCDP estimated for this event is 2.1 x 10-'. The dominant core damage sequence, highlighted as sequence number 39 on the event tree in Fig. B.9. 1, involves the following:

• given the loss of offsite power, the reactor successfully trips;

• both trains of emergency power fail;

• AFW provides sufficient flow;

• the PORVs open and then successfully reseat;

• the safe shutdown facility fails;

• the RCP seals fail; and

• offsite power is not recovered after the RCP seal failure.

The second highest core damage sequence (No. 4 1) involves the following:

• given the loss of offsite power, the reactor successfully trips;

• both trains of emergency power fail; and

• AFW fails to provide sufficient flow.

Definitions and probabilities for selected basic events are shown in Table B.9. 1. The conditional probabilities associated with the highest probability sequences are shown in Table B.9.2. Table B.9.3 lists the sequence logic associated with the sequences listed in Table B.9.2. Table B.9.4 describes the system names associated with the dominant sequences. Minimal cut sets associated with the dominant sequences are shown in Table B.9.5.

B.9.6 References I1. Memorandum from S. D. Ebneter, Regional Administrator, to E. L. Jordan, Director, Office for Analysis and Evaluation of Operational Data, transmitting "Supporting Documents for the Catawba Loss of Offsite Power Event (February 6-8, 1996)," February 15, 1996.

2. U.S. Nuclear Regulatory Commission, "NRC Inspection Report Nos. 50-413/96-03 and 50-414/96-03 and Notice of Violation," March 12, 1996.

3. U.S. Nuclear Regulatory Commission, "Preliminary Notification of Event or Unusual Occurrence PNO-II-6-006," February 6, 1996.

4. U.S. Nuclear Regulatory Commission, "Preliminary Notification of Event or Unusual Occurrence PNO-II-6-006A," February 7, 1996.

5. U.S. Nuclear Regulatory Commission, "Preliminary Notification of Event or Unusual Occurrence PNO-II-6-006B," February 7, 1996.

NUREG/CR-4674, Vol.25 B.9-5 NUREG/CR-4674, Vol. 25

LER No. 414/96-001 Appendix B

6. U.S. Nuclear Regulatory Commission, "Preliminary Notification of Event or Unusual Occurrence PNO-II-96-006C," February 8, 1996.

7. 50.72 Report Number 29945, February 6, 1996.

8. LER 414/96-00 1, "Loss-of-Offsite Power Due to Electrical Component Failures," March 7, 1996.

9. E. M. Dougherty and J. R. Fragola, Human Reliability Analysis, John Wiley and Sons, New York, 1988.

10. P. W. Baranowsky, Evaluationof Station Black-out at Nuclear Power Plants,NUREG- 1032, U.S. Nuclear Regulatory Commission, June, 1988.

11. Revised LOOP Recovery andPWR Seal LOCA Models, ORNL/NRC/LTR-9/1 1, August 1989.

12. "ASP Models, PWR B, Catawba Units 1 and 2," Revision 1, November 1994.

Vol. 25 B.9-6 NUREG/CR-4674, Vol. 25 B.9-6

Amendix B LER No. 414/96-001 ApDendix B LER No. 414/96-001 w

a'- u5ehev a v a Ovo a v lea a vvOokehe Oýeu

'C a V00 a 000000000000000000000000000000000000000000 a

w . . . . . . I . I T. . I . i 2 7 ULILiU Jill UL illi F ICO, C.)

0 fit a

2 I

Fig. B.9. 1. Dominant core damage sequences for LER No. 414/96-00 1.

NUREG/CR-4674, Vol.25 B.9-7 NUREG/CR-4674, Vol. 25

LER No. 414/96-001 ADDendix ADoendix B B

LER No. 414/96-001 STANDBY SHUTDOWN FACILITY FAILS S4F I

SSFC FAILS TO POWER SSFB FAILS WITH SUMP PUMP FOR NO POWER TO BUS ITA WITHIN 15 MIN TDAFWP WITHIN 3 H SSFB SSFC Fig. 13.9.2. Fault tree modeling the Standby Shutdown Facility (SSF).

B.9-8 NUREG/CR-4674, Vol.25 NUREG/CR-4674, Vol. 25 B.9-8

ADDendix B LER No. 414/96-001 ADDendax B LER No. 414/96-001 I

FAILURE TO RECOVER OFESITE POWER BEFORE SEAL LOCA 0 PISL I I FAILURE TO RECOVER FAILURE TO CROSSTIE FAILURE TO RECOVER OFFSITE POWER EDG 2B EMERGENCY EDG 2B WITHIN 3H BEFORE SEAL LOCA BUS WITHIN 3 H OEP-XHE-NOREC-SL OE P-XHE-NOCRS-B EPS-XHE-EDGB-NOR Fig. B.9.3. Fault tree modeling the recovery of offsite power before the core becomes uncovered given a seal LOCA (OP-SL).

B.9-9 NUREG/CR-4674, Vol. 25

LER No. 414/96-001 ADnendix Annendix B B

LER No. 4 14/96-001 FAILURE TO RECOVER OFFSITE POWER BEFORE BATTERY DEPLETED

'0[BD FAILURE TO RECOVER FAILURE TO CROSSTIE FAILURE TO RECOVER OFFSITE POWER BEFORE EDG 2B3 EMERGENCY EDG 2B WITHIN 3 H BATTERY DEPLETED BUS WITHIN 3 H OEP-XHE-NOREC-BD OEP-XHE-NOCRS-B EPS-XHE-EDGB-NOR Fig. B3.9.4. Fault tree modeling the recovery of offsite power before the batteries are depleted (OP-BID).

NUREG/CR-4.674, Vol. 25 B.9-10

ADoendix B Appenix BLER No. 414/96-001 Table B.9.1. Definitions and Probabilities for Selected Basic Events for LER 414/96-001 Modified Event Base Current for this name Description probability probability Type event IE-LOOIP Initiating Event-Loss of Offsite 6.9 E-006 1.0 E+000 TRUE Yes Power IE-SGTR Initiating Event-Steam Generator 1.6 E-006 0.0 E+000 IGNORE No Tube Rupture IE-SLOCA Initiating Event-Small 1.0 E-006 0.0 E+000 IGNORE No Loss-of-Coolant Accident IE-TRANs Initiating Event-Transient 5.3 E-004 0.0 E+000 IGNORE No AFW-TDP-FC-1A AFW Turbine-Driven Pump Fails 3.2 E-002 3.2 E-002 No AFW-XHE-NOREC-EP Operator Fails to Recover AFW 3.4 E-00 1 3.4 E-00 I No During Station Blackout AFW-XHE-XA-NWS Operator Fails to Align Nuclear 1.0 E-003 1.0 E-003 No Service Water EPS-DGN-CF-ALL Common-Cause Failure of ED~s 1.1 E-003 1.1 E-003 No EPS-DGN-FC-IA EDO A Fails 4.2 E-002 4.5 E-002 Yes EPS-DGN-FC-IB EDO B Fails 4.2 E-002 1.0 E+000 TRUE Yes EPS-XHE-EDGB-NOR Operator Fails to Recover EDG B 1.0 E+000 3.1 E-001 NEW Yes Within 3 h EPS-XHE-NOREC Operator Fails to Recover 8.0 E-00 1 1.0 E+000 Yes Emergency Power HPI-MDP-CF-ALL Common-Cause Failure of the 7.8 E-004 7.8 E-004 No High Pressure Injection (HPI)

________________Pumps ______

HPI-MDP-FC-IA HPI Motor-Driven Pump Train A 4.0 E-003 4.0 E-003 No Fails HPI-MOV-CC-DISCH HPI Cold Leg Injection Valve 3.0 E-003 3.0 E-003 No Fails HPR-MOV-CC-RHRB Residual Heat Removal (RHR) 3.1 E-003 3.1 E-003 No Discharge Motor-Operated Valve

__________________(MOV) into HPI Train B Fails HPR-MOV-CC-SMPA Sump Isolation MOV 185A Fails 3.0 E-003 3.0 E-003 No to Open________

NUREG/CR-4674, Vol.25 B.9-11 NUREG/CR-4674, Vol. 25

LER No. 414/96-001 Appendix B Table B.9.1. Definitions and Probabilities for Selected Basic Events for LER 414/96-001 (Continued)

Modified Event Base Current for this name Description probability probability Type event HPR-MOV-CF-SUCA High Pressure Recirc (HPR) 1.0 E+000 1.0 E+000 No Suction MOVs from RHR Train A Fail to Open Due to Common Cause HPR-XHE-NOREC-L Operator Fails to Recover the 1.0 E+000 1.0 E-i000 No HPR System During a LOOP HPR-XHE-XM-L Operator Fails to Initiate HPR 1.0 E-003 1.0 E-003 No During a LOOP OEP-XHE-NOCRS-B Failure to Cross-Tie EDO B 1.0 E+000 2.7 E-001 NEW Yes Emergency Bus Within 3 Hours OEP-XHE-NOREC.2H Operator Fails to Recover Offsite 1.4 E-001 1.0 E+000 TRUE Yes Power Within 2 h OEP-XHE-NOREC-6H Operator Fails to Recover Otfisite 9.9 E-004 1.0 E+000 TRUE Yes Power Within 6 h OEP-XHE-NOREC-BD Operator Fails to Recover Offsite 2.3 E-002 1.0 E+000 TRUE Yes Power Before Battery Depleted OEP-XHE-NOREC-SL Operator Fails to Recover Offsite 4.8 E-001 1.0 E+000 TRUE Yes Power (Seal LOCA)

PORV PORVs Open During Transient 7.0 E-00 1 1.0 E+000 TRUE Yes PPR-SRV-00-PRVI PORV I Fails to Reclose After 2.0 E-003 2.0 E-003 No

___________________Opening _____________

PPR-SRV-OO-PRV2 PORV 2 Fails to Reclose After 2.0 E-003 2.0 E-003 No Opening PPR-SRV-OO-PRV3 PORV 3 Fails to Reclose After 2.0 E-003 2.0 E-003 No Opening RCS-MDP-LK-SEALS RCP Seals Fail Without Cooling 2.4 E-00 1 7.0 E-00 I Yes and Injection RHR-MDP-CF-ALL RHR Pump Common-Cause 4.5 E-004 4.5 E-004 No Failures RHR-MDP-FC-IA RHR MDP IA Fails 4.1 E-003 4.1 E-003 No SEALLOCA RCP Seals Fail During LOOP 2.4 E-00 1 7. 0 E-001I Yes B.9-12 NUREG/CR-4674, Vol.25 NUREG/CR-4674, Vol. 25 B.9-12

LER No. 414/96-001 LRN.449-0 Annendix B Table B,.9.1. Definitions and Probabilities for Selected Basic Events for LER 414/96-001 (Continued)

Modified Event Base Current for this name Description. probability probability Type event SSFB SSF Fails with No Power to Bus 2.2 E-001 2.2 E-00OI NEW Yes JTAIIII SSFC SSF Fails to Power Sump Pump 9.5 E-002 9.5 E-002 NEW Yes of TDAFW Pump I____ I____ I___ I___

NUREG/CR-4674, Vol.25 B.9-13 NUREG/CR-4674, Vol. 25

LER No. 414/96-001 ADDendix B Amendix B LER No. 414/96-001 Table B.9.2. Sequence Conditional Probabilities for LER 414/96-001 Conditional core Event tree damage Percent name Sequence name probability contribution

_________ _________ (CCDP) _____

LOOP 39 8.5 E-004 40.1 LOOP 41 5.3 E-004 25.0 LOOP 32 3.6 E-004 17.2 LOOP 40 2.7 E-004 13.0 LOOP 10 7.8 E-005 3.7 Total (all sequenc es) 2.1 E-003 Table B.9.3. Sequence Logic for Dominant Sequences for LER 4 14/96-001 Event tree name Sequence name Logic LOOP 39 /RT-L, EP, /AFW-L, PORV-L, IPORV-RES, SSF, SEALLOCA, OP-SL LOOP 41 /RT-L, EP, AFW-L-EP LOOP 32 /RT-L, EP, /AFW-L, PORV-L, IPORV-RES, SSF, /SEALLOCA, OP-BD LOOP 40 /RT-L, EP, /AFW-L, PORV-L, PORV-EP LOOP 10 /RT-L, /EP, /AFW-L, PORV-L, PRV-L-EP, OP-2H, /HPI-L, HPR-L B.9-14 NUREG/CR-4674, Vol.25 Vol. 25 B.9-14

ADDendix B LER No.

LER No. 414/96-001 4 14/96-001 AnDendix B Table B.9.4. System Names for LER 414/96-001 System name Logic AFW-L No or Insufficient AFW Flow During LOOP AFW-L-EP No or Insufficient AFW Flow During Station Blackout EP Failure of Both Trains of Emergency Power HPI-L No or Insufficient Flow from HPI System During a LOOP HPR-L No or Insufficient Flow from HPR System During a LOOP OP-2H- Operator Fails to Recover Offsite Power Within 2 Hours OP-BD Operator Fails to Recover Mffite Power Before

_______________Battery is Depleted OP-SL Operator Fails~to Recover Offsite Power (Seal LOCA)

PORV-EP PORVs Fail to Reclose (No Electric Power)

PORV-L PORVs Open During LOOP PORV-RES PORVs Fail to Reseat PRV-L-EP PORVs and Block Valves Fail to Reclose

_______________[Electric Power (EP) Succeeds]

RT-L Reactor Fails to Trip During LOOP SEALLOCA RCP Seals Fail During LOOP SSF Safe Shutdown Facility Fails B.9-15 B.9-15NUREG/CR-4674, Vol. 25

LER No. 414/96-001 AnDendix B Table B.9.5. Conditional Cut Sets for Higher Probability Sequences for LER 414/96-001 Cut set 1 Percent Conditional No. Jcontribution probability0 Cut setSb LOOP..Sequence...9.8.5......

1 68.1 5.8 E-004........

EP-DN-C-A

... -E.N.RS-B ESIN-CB.EP..... . V S..........

.O RCSSB

.. OE.X..NR..SL LOO 1.7enc 1.4 E-004 .PS.GN....LL EPXXE-RC

.S.B...P....NO.S.B LOO Seuene 4 5. E-04 .. 7..............

1 94.6 5.0 E-004 iEPS-DGN-FC-IA, EPS-DGN-FC-IB. EPS-XHE-NOREC, SB AF-TP-FH-NCR-B lA. AF-H-OEALC-EP V P-H-NRC-2 29.8 1.5 E-004 EPS-DGN-FC-1A, EPS-DGN-FC-IB, EPS-XHE-NOREC, SC AFW-XHE-NOCREC-EPV,AF-XE-A-NWS E-HENRE-L AFW-XHE-NOREC-EP 3 68.1 2.4 E-005 EPS-DGN-CFC-AL, EPS-DG-CI.ESXHE-NOREC, E-H-NCSB SSFB .

POEPXE- RSB /SEALLOCA, OEP-XHE-NOREC-SL P-H-DBD, O H - O E - D

......................... O ..............

LOOP Sequence 40 2.7 E-004 . ......................

I............

2 32.5 9.0 E-004 EPS-DGN-FC-1A. EPS-DGN-FC-IB. EPS-XHE-NOREC,POV PPR-SRV-OO-PRV2WXE-OECE 3 32.5 9.0 E-005 EPS-DGN-FC-1A, EPS-DON-FC- IB, EPS-XHE-NORECPR PPR-SRV-OOPRV3EAF-H-X-W B.9-16 NUREG/CR-4674, Vol.25 Vol. 25 B.9-16

Am)endix B LER No. 414/96-001 LER No. 414/96-001 Appendix B Table B.9.5. Conditional Cut Sets for Higher Probability Sequences for LER 414/96-001 (Continued)

Cut set Percent Conditional No. contribution probability' Cut setsb LOOP Sequence 10 7.8 E-.005 1 9.9 7.8 E-006 IEPS-DGN-FC-IA, PORV, PPR-SRV-OO-PRVI, OEP-XHE-NOREC-2H. EPS-DGN-FC- I B, RHR-MDP-FC- IA, HPR-XHE-NOREC-L 2 9.9 7.8 E-006 IEPS-DGN-FC-IA, PORY, PPR-SRV-OO-PRV3.

OEP-XHE-NOREC-2H, EPS-DGN-FC-1B, RHR-MDP-FC-1A, HPR-XHE-NOREC-L 3 9.7 7.6 E-006 IEPS-DGN-FC- IA, PORV, PPR-SRV-00-PRVI.,

OEP-XHE-NOREC-2H. EPS-DON-FO- I B, HPI-MDP-FC- lA, HPR-XH-E-NOREC-L 4 9.7 7.6 E-006 /EPS-DGN-FC-IA, PORV, PPR-SRV-OO-PRV3, OEP-XI-E-NOREC-2H. EPS-DCN-FC- IB, HPI-MDP-FC-IA, HPR-XHE-NOREC-L 5 7.5 5.9 E-006 /EPS-DGN-FC-IA, PORV, PPR-SRV-OO-PRVI, OEP-XHE-NOREC-2H, EPS-IX3N-FC- IB, HPR-MOV-CF-SUCA, HPR-MOV-CC-RHRB, HPR-XHE-NOREC-L 6 7.5 5.9 E-006 IEPS-DGN-FC-1A, PQRV, PPR-SRV-OO-PRV3, OEP-XHE-NOREC-2H, EPS-DGN-FC-IB, HPR-MOV-CF-SUCA, HPR-MOV-CC-RI-fRB, HPR-XHE-NOREC-L 7 7.2 5.7 E-006 /EPS-DGN-FC-1A, PORV. PPR-SRV-OO-PRVI, OEP-XHE-NOREC-2H, EPS-DGN-FC- 1B, HPR-MOV-CC-SMPA, HPR-XHE-NOREC-L 8 7.2 5.7 E-006 ,'EPS-DGN-FC- IA, PORV, PPR-SRV-OO-PRV3, OEP-XHE-NOREC-2H, EPS-DGN-FC- IB, HPR-MOV-CC-SMPA, HPR-XHE-NOREC-L 9 7.2 5.7 E-006 IEPS-DGN-FC-IA, PORV, PPR-SRV-OO-PRV1, OEP-XHE-NOREC-2H, EPS-DGN-FC- IB. HPI-MOV-CC-DISCH.

f-fPR-XH-E-NOREC-L 10 7.2 5.7 E-006 /EPS-DGN-FC-lA, PORV, PPR-SRV-OO-PRV3, OEP-XHE-NOREC-2H, EPS-DGN-FC- I B, HPI-MOV-CC-DISCH, HPR-XHE-NOREC-L 11 2.4 1.9 E-006 /EPS-DGN-FC-IA, PORV, PPR-SRV-OO-PRVI, OEP-XHE-NOREC-2H EPS-DGN-FC- IB. H-PR-XHE-XM-L 12 2.4 1.9 E-006 /EPS-DGN-FC-IA. PORV, PPR-SRV-OO-PRV3, IOEP-XHE-NOREC-2H. EPS-DGN-FC- IB, HPR-XI-E-XM-L NUREG/CR-4674, Vol.25 B.9-17 B.9-17 NUREG/CR-4674, Vol. 25

LER No. 414/96-001 Appendix B Table B.9.5. Conditional Cut Sets for Higher Probability Sequences for LER 414/96-001 (Continued)

Cut set Percent Conditional No. contribution probability" Cut setsb 13 [.8 1.4 E-006 JEPS-DGN-FC4IA, PORV, PPR-SRV-OOPRVI, OEP-HE-NOREC-2H-, EPS-DGN-FC- IB, HPI-MDP-CF-ALL.

HPR-XHE-NOREC-L 14 1.8 1.4 E-006 /EPS-DGN-FC-IA, PORV, PPR-SRV-OO-PRV3.

OEP-HE-NOREC-21-, EPS-DGN-FC-IB, I-PI-MDP-CF-ALL.

HPR-XHE-NOREC-L 15 1.0 8.6 E-007 /EPS-DGN-FC- IA, PORV, PPR-SRV-00-PRV 1, OEP-HE-NOREC-2H, EPS-DGN-FC- I B, RHR-MDP-CF-ALL, HPR-XHE-NOREC-L 16 1.0 8.6 E-007 /EPS-DGN-FC-IA. PORV, PPR-SRV-OO-PRV3.

OEP-XHE-NOREC-2H. EPS-DGN-FC- I B. RHR-MDP-CF-ALL, HPR-XHE-NOREC-L f Total (all sequences) 2.1 E-003

'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 probabilities for the initiating events and the basic events are given in Table B.9. 1, b"Basic events TE-LOOP, EPS-DGN-FC- IB, OPE-XHE-NOREC-2H, OEP-XHE-NOREC-6H, OEP-XHE-NOREC-BD, PORV, and OEP-XHE-NOREC-SL are all type TRUE events which are not normally included in the output of fault tree reduction programs but have been added to aid in understanding the sequences to potential core damage associated with the event.

B.9-18 NUREG/CR-4674, Vol.25

.NUREG/CR-4674, Vol. 25 B.9-18