ML20135G722

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Final ASP Analysis - ANO 1 (LER 313-96-005)
ML20135G722
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 05/14/2020
From: Christopher Hunter
NRC/RES/DRA/PRB
To:
Littlejohn J (301) 415-0428
References
LER 1996-005-00
Download: ML20135G722 (15)
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LER No. 313/96-005 Annendix B LRN.339-0 B.6 LER No. 313/96-005 Event

Description:

Reactor trip and subsequent steam generator dryout Date of Event: May 19, 1996 Plant: Arkansas Nuclear One, Unit I B.6.1 Summary Arkansas Nuclear One, Unit 1 (ANO 1) was operating at 100% power when the plant experienced an automatic trip on high reactor coolant system (RCS) pressure that resulted from reduced main feedwater (MFW) flow. Following the scram, six of eight main steam safety valves (MSSVs) lifted on the B once-through steam generator (OTSG). One of these safety valves stuck open when pressure was reduced, and about 18 min after the trip, the operators, in accordance with the plant emergency operating procedures, isolated the faulted B OTSG from its MFW source and steam outlet. With the pressure and temperature decreasing on the secondary side of the OTSG, the OTSG was allowed to "dry out" because the RCS temperature was maintained relatively constant. About 5 h and 41 min after the trip, the stuck-open safety valve was gagged closed. After that, the B OTSG was refilled, and the plant was returned to normal hot shutdown conditions. The estimated conditional core damage probability (CCDP) for this event is 5.6 x~ 10'.

B.6.2 Event Description According to the licensee event report (LER),' the plant was operating at 100% power when a degradation of the power supply to the turbine hydraulic control valve for MFW pump (MFWP) A caused a rapid decrease in pump speed. A second decrease in pump speed resulted in the pump going to minimumn speed. The integrated control system (ICS) responded to the change in feedwater flow by increasing the speed on MFWP B. The lower heat removal rate resulting from the reduced feedwater flow caused by MFWP A going to minimum speed in turn caused the pressure in the reactor to increase. The increasing pressure then caused the reactor to automatically trip on high pressure, just before the operators attempted to manually trip the reactor. At this time, MFWP A was at minimum speed; the feedwater cross-over valve was closed (normal position) because no MFWP trip signal was present; and MFWP B was at maximum speed, in its "Diagnostic-Manual" mode, and not responding to additional ICS signals. Following the trip, A OTSG had a low water level inventory, and B OTSG had a high water level inventory. However, because of a back pressure wave induced by closing the main turbine stop valves, the sensed water level in the B OTSG indicated low, thereby actuating the emergency feedwater (EFW) system. MFWP B tripped on high discharge pressure -14 s after the reactor scram, and MFWP A responded to ICS demand signals when its control circuit fault cleared; however, because the demand signal was very high, MFWP A tripped on mechanical overspeed about 37 s after the reactor trip.

Secondary-side steam pressure in the A OTSG remained below the MSSV setpoints because of the reduced inventory in the steam generator that resulted from the lower feedwater flow rate caused by the MFWPT A speed decrease; conversely, the high inventory in the B OTSG (caused by MFWP B going to maximum B.6-1 NUREG/CR-4674, Vol. 25

LER No. 313/96-005 Appendix Apni B speed) resulted in a high secondary-side steam pressure. Consequently, six of the eight MSSVs on the B OTSG opened to reduce pressure. These valves opened prior to the A and B MFW pump trips. The operators noted -64 s after the reactor trip that one of the MSSVs bad failed to reclose following the pressure reduction, thus causing an accelerated RCS cooldown rate. Operators manually initiated high pressure injection (HPI) about 6 min after the reactor trip in accordance with the plant's Emergency Operating Procedures (EOPs) when the water level in the pressurizer dropped below 30 in. Following that, also in accordance with EOPs, the faulted B OTSG was isolated from its feedwater source and steam outlet about 18 min after the trip. The secondary side of the OTSG continued to "blow down" through the open MSSV; however, the operators controlled the RCS cooldown rate by maintaining the RCS temperature above 271.1 0 C (520'0 F). This blowdown is also referred to as "drying out" the OTSG, and the lack of steam in the OTSG results in the OTSG shell cooling down below the RCS temperature.' This tube-to-shell temperature differential is governed by plant Technical Specifications and is limited to 15.6'C (60'F) for ANO 1. During this transient, however, the shell-to-tube temperature differential increased to 23Y3C (747F). The OTSG vendor, Framatome Technology, Inc., and the licensee both analyzed the 23.3C (74 0 F) temperature difference and concluded that no excessive stresses were induced on the OTSG or the reactor pressure vessel.

With the steam header B isolated, the normal supply for sealing steam for the gland seals on the main turbine was not available, and because the backup steam supply (the auxiliary boiler) was also unavailable due to control system problems, sealing steamn was eventually lost. As a result, about 35 min after the trip, the vacuum in the main condenser was lost. However, heat removal was still possible after the main condenser became unavailable by discharging steam through an atmospheric dump valve, which the operators did until the auxiliary boiler was available. At this time, the vacuum in the main condenser was reestablished, and the main condenser was again used for heat removal.

Plant maintenance crews successfully gagged closed the MSSV approximately 5 h and 41 min after the reactor trip. Using EFW, operators then began refilling B OTSG and cleared the main steam line isolation signal. The main steam isolation valve for B OTSG was opened about 2 h later and the main feedwater isolation valve about an hour after that. At that time, normal feedwater was established to the OTSG, and the plant was restored to a normal hot shutdown condition.

B.6.3 Additional Event-Related Information A short circuit in a digital speed sensing probe for MFWP A reduced voltage in the feedwater control system 24-V power supply. This, in turn, decreased control oil pressure for the MFWP turbine steam admission valve, causing the valve to partially close. The closing of the steam admission valve decreased the speed of MFWP A, thereby decreasing feedwater flow. The reduced feedwater flow in turn caused the ICS to demand maximum MFW; however, the MFW control system incorrectly interpreted this as a failure (invalid signal) in the ICS and transferred MFWP B control to the "Diagnostic Manual" mode. This effectively kept MFWP B operating in response to the last valid sensed signal (high demand). When the reactor tripped, the ICS sent a reduction signal to the feedwater control system. However, MFWP B did not respond because it was in "Diagnostic Manual" mode. The feedwater block valves closed in response to the rapid flow reduction signal.

As a result, the system pressure increased rapidly, and MFWP B tripped on overpressure.

25 B.6-2 NIJREG/CR-4674, Vol.

NUREG/CR-4674, Vol. 25 B.6-2

Annendix B LER LER No.

No. 313/96-005 313/96-005 Annendix B The MFW system at ANO 1 consists of two variable-speed turbine-driven pumps that take their common suction downstream of feedwater heaters E2A and E2B3 and discharge to the OTSGs. Either MFWP can discharge to both OTSGs by routing through the normally closed feedwater cross-over valve located before the feedwater flow control valves."' 4 Typically, these pumps are used to supply feedwater to the OTSGs from about 3% power to full power. The system also has an auxiliary motor-driven pump, which is used to supply feedwater to the OTSGs during plant startup and shutdown below 3% power. The auxiliary pump takes a suction from the MFWP suction header and discharges to the MFWvP A discharge header upstream of the cross-over valve. The MFWPs are rated at 60% of the plant's full-load capacity each, and the auxiliary feedwater pump is rated at 5% full-load capacity.

The MSSV that did not reclose failed to reseat because the locking device cotter pin was not engaged with the release nut. This allowed the release nut to travel down the spindle of the valve and block the manual lift top lever from returning to its normal position. This phenomenon has been documented in NRC Information Notice 84-33, as well as in other industry studies. Investigations indicate that either the failure of the cotter pin or the insufficient slot engagement by the cotter pin allows the release nut to rotate down the spindle while the MSSV is lifted. The NRC Augmented Inspection Team (in Sects. 3.2 and 6.3 of Ref. 2) sent to investigate this event found that of the 16 MSSVs at ANO I:

..one stuck-open, . . . because of a stem-nut utilized to facilitate manual lifting of the valve, not being properly pinned in place so that during lift and/or blowdown of the valve the nut traveled down the stem and contacted the lifting device. This contact precluded the valve from reseating. . .. 6 of the 15 other MSSVs in Unit I had less than desirable cotter pin engagement . .. 2 of the remaining 9 had marginal (i.e., cotter pin) engagement. (i.e., and the remaining 7 valves had acceptable cotter pin engagement) . .. despite marginal engagement, none of the nuts could be rotated by hand.

Hence, the quotes above show by allowing the release nut to rotate does not prevent the valve from reclosing after it has opened. The NRC's Augmented Inspection Team determined that the licensee's procedures for installing the cotter pins were inadequate. Moreover, the licensee's own inspection (Ref. 1, Section D) found that Cotter pins for two other valves were found not engaged in the release nuts. These valves were determined to have been operable since the release nuts could not be rotated due to the cotter pin ends being engaged on the nuts. Six valves had the pins partially engaged at the top end of the release nut slot. Seven valves were found with the cotter pins fuldly engaged.

Therefore, this was determined to be a singular incident.

B.6.4 Modeling Assumptions This event was examined- as the combination of two individual events. The first is the reactor trip and subsequent loss of main feedwater (LOFW) transient. The second is the potential for a steam generator tube rupture (SGTR) as a result of the "drying out" of B OTSG. The LOFW is a relatively simple and straightforward transient with few complications other than operator burdens in the recovery process. The potential for a SGTR, however, is neither simple nor straightforward. Both events are discussed below.

NUREG/CR-4674, Vol.25 B.6-3 NUREG/CR-4674, Vol. 25

LER No. 313/96-005 Atmendix B LOFW The LOFW transient began with the reactor trip, continued through the subsequent OTSG dryout, and concluded with MFW recovery. The transient concluded when the MFW isolation valve was opened, which according to Ref. 2, Attachment 1, was approximately 9 h and 36 min after the trip. The event was modeled as. a high-pressure reactor trip initiating event with subsequent LOFW (TE-TRANS and MFW-SYS-TRJP set to TRUE for this portion of the analysis). MFW was assumed recoverable; however, it should be noted that both MFWPs were tripped (one on mechanical overspeed and the other on high discharge pressure, but the second also had an undetermined failure in its control system) and were not used in lieu of EFW prior to the OTSG isolation. After that, in the long-term after the OTSG isolation, EFW was also used to supply the on-line OTSG. When the isolation was cleared, the B OTSG was refilled using EFW. MFW was not used until the OTSGs were supplied via the startup valves about 9 h and 42 min after the trip (Ref. 2, Attachment 1),

and EFW was not secured until almost 10 h after the reactor trip (Ref. 2, Attachment 1). The ANO 1 model used in conjunction with the Integrated Reliability and Risk Analysis System (IRRAS)5 already includes the motor-driven auxiliary feedwater pump as a supplement to the MFW when the MFWPs have tripped off or have failed.

SGTR The typical accident analysis for core damage examines loss-of-coolant accidents (LOCAs), of which the small-break LOCA (SLOCA) is a subset. The SGTR, in many aspects, is similar to the SLOCA. The SGTR is examined for its resulting effect on core integrity. According to NUREG-0844 (Ref. 6):

The leakage of primary coolant into the secondary has two major safety implications. The first is the potential for direct release of radioactive fission products into the environment, and the second is the loss of cooling water which is needed to prevent core damage. An extended uncontrolled loss of coolant outside containment would result in the depletion of the initial RCS inventory and emergency core cooling system (ECCS) water without the capability to recirculate the water.

The licensee and Framatome Technologies, Inc., examined this event by focusing on the differential temperature (At) experienced by the OTSG during the dryout, and they correlated that temperature to a pounds compressive force (Ref, 2, Section 4.2). The maximum At occurred approximately 2 h and 44 min after the trip (Ref. 2, Attachment 1). The stresses induced by the corresponding high At during the OTSG dryout were probably greater than those produced by the transient-induced differential pressure (Ap);

however, if the scope of the analysis follows the increased stress due to the maximum At, the underlying assumption still concerns tube integrity, and the analysis will ultimately result in examination of tube rupture or leakage. The analysis will follow the SGTR after that. In the absence of a simple correlation available to reconcile the stresses induced by the temperature increase, the transient was analyzed using the Ap increase rather than the At increase.

NUREG/CR-4674, Vol. 25 .4 B.6-4

Appendix B Appenix BLER No. 3 13/96-005 About 18 min after the'plant had shut down, the OTSG was isolated and allowed to blow down through the stuck-open MSSV. While the secondary-side pressure was decreasing, the primary-side (RCS) pressure was kept nearly constant."' This resulted in the OTSG tubes being exposed to an increasing Ap, which stopped increasing only when the safety valve was gagged closed. The OTSG secondary-side pressure decreased to

0. 138 MPa (20 psig),7 while the primary-side pressure was stabilized near the normal operating pressure`'

of 14.72 MPa (2,155 PSI g).4 This means that the tubes of the B OTSG were subjected to a maximum Ap of 14.58 MPa (2,135 psid). The licensee has supplied data' that indicates that the probability of an SGTR is 1.6 x 10`~ for this Ap.

Following a nominal SGTR, the RCS is depressurized by the operators to below the MSSV setpoint. This allows the MSSVs to close and equalizes pressure between the primary and secondary sides of the SG, which terminates flow through the break. During this event, the high SG Ap (which could have induced the tube rupture) was the result of a stuck-open MSSV. The stuck-open MSSV would have prevented SG and RCS pressures from equalizing, and therefore prevented isolation of the ruptured SQ. In this case, the operators would have had to continue cooling down and depressurizing the RCS until the unit could have been placed on the DHR system. Basic events MSS-VCF-HW-ISOL and MSS-XHE-NOREC were set to TRUE (i.e.,

probability of occurring = 1.0) to reflect the inability to isolate the ruptured SQ following the postulated SGTR.

EVENT MODEL The analyses for LOFW and SGTR were combined to analyze the entire event as follows:

[P(SGTR) x estimated CCDP for SGTRJ + ([I - P(SGTR)I x estimated CCDP for LOFW]}.

B.6.5 Analysis Results The estimated CCDP for this event is 5.6 x 10' This estimation was derived from the equation given in the previous section and is calculated as follows:

[P(SGTR) x estimated CCDP for SGTRI + (I- P(SGTR)] x estimated CCDP for LOFWI},

where probability of tube rupture = 0.00 16 1 - probability of tube rupture = 0.9984 CCDP due to SGTR = 3.1 - 10`

CCDP due to LOFW = 6.37 x 1-The dominant core damage sequence, highlighted as sequence number 3 on Fig. B.6. 1, contributes -54% to the estimated CCDP. The dominant sequence involves the following steps:

B.6-5 NUREG/CR-4674, Vol. 25

LER No. 313/96-005 Appendix B LER No. 313/96-005 Appendix B

  • SGTR initiating event occurs,
  • the reactor is successfully tripped,
  • EFW is successful,
  • HPI is successful,
  • failure of SG isolation,

" successful depressurization to the DHR initiation pressure, and

" failure of DHR.

The LOFW contributes -.11% to the estimated total CCDP.

Definitions and probabilities for selected basic events are shown in Table B.6. 1. The conditional probabilities and sequence logic associated with the highest probability sequences are shown in Table B.6.2. Table B.6.3 describes system names associated with the dominant sequences. Minimal cut sets associated with the dominant sequences are shown in Table B.6.4. Because the LOFW transient contributed about 11% to the total CCDP, information regarding the analysis was included in the above tables for clarification, better understanding of the event, and calculation of the total CCDP.

B.6.6 References

1. LER No. 313/96-005, Rev. 0, "Automatic Reactor Trip and Engineered Safety Features Actuations Caused by Failure of a Speed Sensing Probe in the Control Circuitry of a Main Feedwater Pump Turbine and Failure of a Main Steam Safety Valve to Re-Seat," June 18, 1996.
2. NRC Augmented Inspection Team Report No. 50-3 13, -368/96-19, June 12, 1996.
3. ANO 1 ProbabilisticRisk Assessment - IndividualPlant Examination Submittal, April 1993.
4. ANO 1 Safety Analysis Report, Amendment 13, September 25, 1995.
5. U.S. Nuclear Regulatory Commission, Systems Analysis Programsfor Hands-On IntegratedReliability Evaluations (SAPHIRE), Version 5. 0, NUREG/CR-61 16 (EGG-2716), Volumes 1-10, July 1994.
6. U.S. Nuclear Regulatory Commission, NRC IntegratedProgram of Unresolved Safety Issues A-3, A-4, and A-S Regarding Steam Generator Tube Integrity, NUREG-0844, September 1988.
7. Personal communication, P. D. O'Reilly, U.S. NRC, with T. Reis, U.S. NRC.
8. D. C. Mims, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, July 23, 1997.

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

ADDendix B LER No. 313/96-005 Appendix B LER No. 313/96-005 Fig. B.6. 1 Dominant core damage sequence for LER No. 313/96-005.

B.6-7 NUREG/CR-4674, Vol.25 NUREG/CR-4674, Vol. 25

LER No. 313/96-005 Appendix B Table B.6.1. Definitions and Probabilities for Selected Basic Events for LER No. 313/96-005 Event name Description Base Current Type Modified probability probability for this event TE-LOOP Initiating Event-Loss of 5.8 E-006 5.8 E-006 No Offsite Power IE-SGTR Initiating Event-SGTR 1.6 E-006 1.0 .E+000 TRUE Yes TE-SLOCA Initiating Event-SLOCA 1.0 E-006 1.0 E-006 No IE-TRANS Initiating Event-Transient 1.3 E-004 1.0 E+000 TRUE Yes (TRANS)

DHR-MDP-CF-ALL Common-Cause Failures of 4.5 E-004 4.5 E-004 No' DHR Pumps DHR-MOV-CC-SUC DHR Suction Path Failures 6.0 E-003 6.0 E-003 No DHR-MOV.-CF-DISCH Common-Cause Failures of the 2.6 E-004 2.6 E-004 No Discharge MOVs DHR-XHE-NOREC Operator Fails to Recover the 1.0 E-00 1 1.0 E-00 1 No DHR System DHR-XHE-XM-DHR Operator Fails to Initiate the 1.0 E-003 1.0 E-003 No DHR System EFW-MDP-FC-1A Failure of the Emergency 3.8 E-003 3.8 E-003 No Feedwater (EFW) Motor-Driven Pump EFW-MOV-CF-DISAL EFW Discharge Valves Fail 5.5 E-005 5.5 E-005 No From Common Causes EFW-TDP-FC-1B Failure of the EFW Turbine- 3.2 E-002 3.2 E-002 No Driven Pump EFW-TNK-FC-CST Failure of the Condensate 1.0 E-004 1.0 E-004 No Storage Tank EFW-XHE-NOREC Operator Fails to Recover 2.6 E-00 1 2.6 E-001I No EFW System EFW-XHE-XA-CST Operator Fails to Align a 1.0 E-003 1.0 E-003 No

_________________Backup Water Supply I_______ I___I___I_

B.6-8 NUREG/CR-4674, Vol.25 Vol. 25 B.6-8

Annendix B LER No. 313/96-005 Table B.6.I Definitions and Probabilities for Selected Basic Events for LER No. 313/96-005 (Continued)

Event name Description Base Current Type Modified probability probability for this event HPI-CKV-00-MST Makeup Storage Tank Stop 3.0 E-003 3.0 E-003 No Check Valve Fails to Seat HPI-MDP-CF-ABC High Pressure Injection (HPI) 1.2 E-005 1.2 E-005 No Motor-Driven Pumps Fail to Run due to Common Cause HP1-MDP-FC-IC HP! Train C Fails 3.9 E-003 3.9 E-003 No HPI-MOV-CC-SUCA Train A Suction Isolation 3.1 E-003 3.1 E-003 No Motor-Operated Valve Fails HPI-MOV-CC-SUCC Train C Suction Isolation 3.1 E-003 3.1 E-003 No Motor-Operated Valve Fails HPI-MOV-CF-SUCT HPI Suction Isolation Motor- 2.6 E-004 2.6 E-004 No Operated Valves Fail due to Common Cause HPI-XHE-NOREC Operator Fails to Recover HPI 8.4 E-00 1 8.4 E-00 I No HPI-XHE-XM-BOR Operator Fails to Initiate 1.0 E-003 1.0 E-003 No Emergency Boration HPI-XHE-XM-HPIC Operator Fails to Initiate HPI 1.0 E-002 1.0 E-002 No

_____________________Cooling _______ ___________

MFW-SYS-TRIP Main Feedwater (MFW) 2.0 E-00 I 1.0 E+000 TRUE Yes System Trips ____

MFW-XHE-NOREC Operator Fails to Recover 1.6 E-002 1.6 E..002 No MFW MSS-VCF-HW-ISOL Ruptured SG Isolation Fails 1.0 E-002 1.0 E+000 TRUE Yes MSS-XHE-NOREC Operator Fails to Isolate 1.0 E-00 1 1.0 E+000 TRUE Yes Ruptured SG PCS-ICC-FA-TT Failure of the Main Turbine to 1.0 E-003 1.0 E-003 No Trip____________

PCS-PSF-HW Hardware Failures Causing 1.0 E-005 1.0 E-005 No Failure to Depressurize PCS-XHE-XM-SO Operator Fails to Initiat RCS 4. EI0 . E0N IDepressurization I________________ ____ ______

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

LER No. 313/96-005 Appendix B Table B.6.1. Definitions and Probabilities for Selected Basic Events for LER No. 3 13/96-005 (Continued)

Event name Description Base Current Type Modified probability probability for this event PPR-MOV-00-BLK Power-Operated Relief Valve 4.0 E-003 4.0 E-003 No (PORV) Block Valve Fails to Close PPR-SRV-CC-PORV PORV Fails to Open on 6.3 E-003 6.3 E-003 No Demand PPR-SRV-CC-RCS PORV/SRVs Fail to Limit 4.4 E-004 4.4 E-004 No Reactor Coolant System (RCS)

Pressure PPR-SRV-CO-TRAN PORV Opens During Transient 8.0 E-002 8.0 E-002 No PPR-SRV-00-PORV PORV Fails to Reclose After 3.0 E-002 3.0 E-002 No Opening PPR-XHE-NOREC Operator Fails to Close the 1.1 E-002 1.1 E-002 No Block Valve RCS-PHN-MODPOOR Moderator Temp Coefficient 1.4 E-002 1.4 E-002 No not Negative Enough RCS-XHE-XM-DEPRH Operator Fails to Depressurize 1.0 E-003 1.0 E-003 No the RCS to RHR Entry Conditions RPS-NONREC Nonrecoverable RPS Failures 2.0 E-005 2.0 E-005 No RPS-REC Recoverable RPS Failures 4.0 E-005 4.0 E-005 No RPS-XHE-XM-SCRAM Operator Fails to Manually 1.0 E-002 1.0 E-002 No Trip the Reactor I____ I______

NUREG/CR-4674, Vol. 25 B61 B.6-1 0

LER No. 313/96-005 Annendix BLENo3136-0 Table B.6.2. Sequence Conditional Probabilities for LER No. 313/96-005 Event Sequence Conditional Percent Logic tree number core damage contribution' name probability

______I______ (CCDP) ________ __________

SGTR 3 1.6 E-003 53.5 /RT, /EFW, /HPI, JRCS-SG, SGISOL,

_____ ________ ________ DEP-S, DHR 8 1.0 E-003 31.8 /RT, /EFW, /HPI, /RCS-SG, SGISOL,

_ ____ ___ __ ___ ___ ___ DEP-S, DEP-P 9 4.1 E-004 13.0 /RT, /EFW, /1-PI,RCS-SG Subtotal (SGTR) 3.1 E-003 TRANS 2 1-16 3.1 E-007 49.4 RT, MFW-A, IEFW-ATWS, RCSPRESS 2 1-17 2.0 E-007 31.4 RT, MFW-A, EFW-ATWS 20 8.9 E-008 13.9 /RT, EFW, MFW, HPI-COOL 21-15 2.0 E-008 3.2 RT, MFW-A, /EFW-ATWS,

________ ________/RCSPRESS, BORATION 8 9.9 E-009 1.5 /RT, /EFW, PORV, PORV-RES, HPI Subtotal (TRANS) 6.3 E-007

'Percent contribution to the subtotal CCDP.

B.6-11 NUREG/CR-4674, Vol. 25

LER No. 313/96-005 Annendix B Anoendix B LER No. 313/96-005 Table B.6.3. System Names for LER No. 313/96-005 System name Description BORATION Emergency Boration Fails DEP-P Failure to Depressurize the RCS to DHR Using PORVs DEP-S Failure to Depressurize the RCS to DHR Using Spray DHR No or Insufficient Flow From the DHR System EFW No or Insufficient Emergency Feedwater (EFW) System Flow EFW-ATWS No or Insufficient EFW System Flow HPI No or Insufficient Flow From the High Pressure Injection (HPI)

System HPI-COOL Failure to Provide HPI Cooling MFW Failure of the Main Feedwater System MFW-A Failure of the Main Feedwater System During ATWS PORV PORV Opens During Transient PORV-RES PORV Fails to Reseat RCS-SG Failure to Initiate RCS Depressurization RCSPRESS Failure to Limit RCS Pressure RT Reactor Fails to Trip During Transient SGISOL Failure to Isolate SG NIJREG/CR-4674, Vol. 25 B61 B.6-12

Appendix B LER No. 313/96-005 Appendix B LER No. 313/96-005 Table B.6.4. Conditional Cut Sets for Higher Probability Sequences for LER No. 313/96-005 Cut set 1 Percent Conditional Cut setSb number Jcontribution probability" SGTR Sequence 03 1.6 E-003 ..... ...

1 59.4 1.0 E-003 MSS-VCF-HW-ISOL, MSS-XHE-NOREC, DI-R-XHE-XM-DHR 2 35.6 6.0 E-004 MSS-VCF-H-W-ISOL, MSS.-XHE-NOREC, DHR-MOV-CC-SUC.

DHR-XHE-NOREC 3 2.6 4.5 E-005 MSS-VCF-HW-ISOL. MSS-XHE-NOREC. DHR-MDP-CF-ALL.

DHR-XHE-NOREC 4 1.5 2.6 E-005 MSS-VCF-HW-ISOL, MSS-XHE-NOREC, DHR-MOV-CF-DISCH, DHR-XHE-NOREC SGTR Sequence 8 1.0 E-003........

1 100.0 1.0 E-003 MSS-VCF-HW-JSOL. MSS-XHE-NOREC. RCS-XHE-XM-DEPRH SGTR Sequence 9 4.1 E-004......

1 97.5 4.0 E-004 PCS-XHE-XM-SG 2 2.4 1.0 E-005_ PCS-PSF-HW TRANS Sequence 21-16 3.1 E-007 ......

1 88.9 2.8 E-007 RPS-NONREC, MFW-SYS-TRIP, RCS-PHN-MODPOOR 2 6.3 2.0 E-008 RPS-NONREC, MFW-SYS-TRIP. PCS-ICC-FA-TT 3 2.7 8.8 E-009 RPS-NONREC, MFW-SYS-TRIP, PPR-SRV-CC-RCS 4 1.7 5.6 E-009 RPS-REC, MFW-SYS-TRIP, RPS-XHE-XM-SCRAM, RPS-PHN-MODPOOR TRANS Sequence 2 1-17 2.0 E-007...............

1 82.9 1.7 E-007 RPS-NONREC, MFW-SYS-TRIP. EFW-TDP-FC-IB,

____I______ EFW-XHE-NOREC NUREG/CR-4674, Vol.25 B.6-13 B.6-13 NUREG/CR-4674, Vol. 25

LER No. 313/96-005 ADDendix LER No. 3 13/96-005 Annendix B B Table B.6.4. Conditional Cut Sets for Higher Probability Sequences for LER No. 313/96-005 (Continued)

Cut set Percent Conditional Cut sets' number contribution probability4 2 10.0 2.0 E-008 RPS-NONREC, MFW-SYS-TRIP, EFW-MDP-FC-IA, EFW-XHE-NOREC 3 2.5 5.2 E-009 RPS-NONREC, MFW-SYS-TRI P. EFW-XHE-XA-CST, EFW-XHE-NOREC 4 1.6 3.3 E-009 RPS-REC, RPS-XHE-XM-SCRAM, MFW-SYS-TRIP, EFW-TDP-FC- I B, EFW-XHE-NOREC TRANS..... Sequence..20.8.9....0.

1 467

4. E-08 E W-X E-...S. NO EC M WS ST I......

4R N 4.6enc 40.1 E-00 EF -hK F - . ........-

OR....... T............-

MFW-XHE-NOREC. HPI-XHE-XM-HPIC 2 39.6 3.2 E-008 EFW-MDP-FAC-I.EWTD-CI, EFW-XHE-NOREC, F-YTRP MWSSTI.MFW-XHE-NOREC, PPR-SRV-CC-PORV 6 2.9 2.6 E-009 EWMPF-A EFW-TNK-FC.CST , EFW-XHE-NOREC,MFSYTRP MFW-SSTIMWXHE-NOREC. PPR-SRV-CC-POR 7 2.5 2.3 E-009 EFW-MOV-FCF-DSA, EFW-XHE-NOREC. MFW-SYS-TRIP.

MFW-XHE-NOREC, HPI-XHE-XM-HPIC 8 1.6 1.4 E-009 EFW-MOV-FCF-ISALF-D-C , EFW-XHE-NOREC.MWYSTP, MWSSTIMFW-XHE-NOREC, PPR-SRV-CC-PORV 9 1.0 9.1 E-0109 EFW-XHE-XA-CST, EFW-XHE-NOREC, MFW-SYS-TRIP, MFW-XHE-NOREC, HPR-MOV-CC-SUTHP-HENOE T7N eqec 21-5 2.0 E-009 J 98.0 2.0 E-008 EF)O-FDSL RPS H-NONREC. FY-RP HPI-XHE-XM-NOE, F-SSTRP BORC 91.9 4.0 E-010 RPS-RHECRP-XH-SEF-XM-SCRAM. MFW-SYS-TRIP, J______ ______ P-XHE-XM-OREC P-O-FSIT P-H-OE B.6-14 NUREG/CR-4674, Vol.25 NUREG/CR-4674, Vol. 25 B.6-14

Appendix B Appenix No.

BLER3 13/96-005 Table B.6.4. Conditional Cut Sets for Higher Probability Sequences for LER No. 313/96-005 (Continued)

Cut set 1 Percent Conditional Cut sets' number Jcontribution -probability" TRANS Sequence 8 9.9 E-009 ........ ....

1 58.8 5.8 E-009 PPR-SRV-CO-TRAN, PPR-SRV-OO-PORV. PPR-XHE-NOREC.

HPI-MOV-CF-SUCT, HPI-XHE-NOREC 2 21.4 2.1 E-009 PPR-SRV-CO-TRAN, PPR-SRV-OO-PORV, PPR-MOV-OO-BLK, HPI-MOV-CF-SUCT. HPI-XI-E-NOREC 3 2.7 2.7 E-010 PPR-SRV-CO-TRAN, PPR-SRV-OO-PORV. PPR-XHE-NOREC, HPI-MDP-CF-ABC, HPI-XHE-NOREC 4 2.7 2.7 E-0 10 PPR-SRV-CO-TRAN, PPR-SRV-OO-PORV, PPR-XHE-NOREC.

H-Pl-MOV-CC-SUCA, HPI-MD-FC-IC, HPI-XHE-NOREC 5 2.6 2.6 E-010 PPR-SRV-CO-TRAN. PPR-SRV-OO-PORV. PPR-XHE-NOREC.

HPI-CKV-OO-MST. HP1-MDP-FC- IC. HPI-XHE-NOREC 6 2.1 2.1 E-0 10 PPR-SRV-CO-TRAN, PPR-SRV-OO-PORV, PPR-XHE-NOREC, HPI-MOV-CC-SUCA, HPI.MOV-CC-SUCC, HPI-XI-E-NOREC 7 2.0 2.1 E-010 PPR-SRV-CO-TRAN, PPR-SRV-OO-PORV, PPR-XHE-NOREC, HPI-CKV-OO-MST. I-PI-MOV-CC-SUCC. H-PI-XHE-NOREC Subtotal SGTR 3.1 E-003 Subtotal TRANS 6.3 E-007 Total 3.1 E-003 aThe 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 cutset. The probabilities of the initiating and basic events are given in Table B.6. 1. Initiating events begin with the designator IE.

5 Basic events MFW-SYS-TRIP, MSS-VCF-HW-ISOL, and MSS-XI-E-NOREC are "TRUE" type events, which would not normally be included in the output of a fault tree analysis and reduction program; however, these events have been added to this table to help understand the sequences for potential core damage associated with LER No. 3 13/96-005.

B.6-15 B.6-15NUREG/CR-4674, Vol. 25

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