Text

Final Accident Sequence Precursor Analysis of October 2006, Operational Event
	ML083010185
Person / Time
Site:	Surry
Issue date:	01/08/2009
From:	Stang J; Plant Licensing Branch II
To:	Christian D; Virginia Electric & Power Co (VEPCO)
	Wright D, NRR/DORL, 301-415 -1864
Shared Package
ML083010199	List: ML083010185;
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 8, 2009 Mr. David A. Christian President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 SUB~IECT:

SURRY POWER STATION, UNIT NO.2 RE: FINAL ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF OCTOBER 2006, OPERATIONAL EVENT

Dear Mr. Christian:

The enclosure provides the final results of the Accident Sequence Precursor (ASP) analyses of an event which occurred at Surry Power Station, Unit NO.2 (Surry 2) as documented in Licensee Event Report (LER) 281/06-002. Additionally, Region II conducted a special inspection and issued Inspection Report (lR) 05000281/2006011 on December 1,2006. The LER and IR documents are publicly available through the Agencywide Document Access Management System (ADAMS), Accession Nos. ML063460178 and ML063350402, respectively. The subject event occurred on October 7,2006, during which Surry 2 was manually tripped based on indications associated with main steam flow, feedwater flow, main steam pressure, and steam generator level perturbations. Normal offsite power was lost to both Surry 1 emergency buses and one of the Surry 2 emergency buses due to flying debris that impacted the A and C Reserve Service Station Transformers' (RSST) electrical conductors.

Emergency Diesel Generator (EDG) 3, shared between Surry 1 and 2, started and automatically loaded to the 2J Emergency Bus as designed. As expected, EDG 2 did not start since the B RSST was not affected, and continued to supply power to the Surry 2, 2H Emergency Bus. The ASP analysis calculated a point estimate conditional core damage probability (CCDP) of 2x10-6.

An uncertainty analysis for this operating condition was also performed resulting in a mean CCDP of 2x10-6 with 5% and 95% uncertainty bounds of 1x10-7 and 8x1Q-6, respectively. The final ASP analysis is publically available through ADAMS No. ML082610715.

The Nuclear Regulatory Commission (NRC) established the ASP Program in 1979 in response to the Risk Assessment Review Group report (see NUREG/CR-0400, dated September 1978).

The ASP Program systematically evaluates U.S. nuclear power plant operating experience to identify, document, and rank the operating events that were most likely to have led to inadequate core cooling and severe core damage (precursors), accounting for the likelihood of additional failures.

The NRC currently uses the ASP Program to:

Monitor performance against the safety goal established in the agency's Strategic plan (see NUREG-1100, Volume 24, "Performance Budget: Fiscal Year 2009," issued February 2008).

Provide feedback to improve Standardized Plant Analysis Risk (SPAR) models.

Evaluate the generic implications of precursors and trend industry performance.

Support generic safety issue (GSI) resolution.

D. Christian

- 2 For more information about the ASP program, see the annual ASP program report at http://www.nrc.gov/reading-rm/doc-collections/commission/secys/2007/secy2007-0176/2007 0176scy.pdf.

The enclosure is provided for your information and no response is requested. If you have any questions please contact Donna Wright at 301-415-1864.

S* cere~

J Stang, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-281

Enclosure:

Final ASP Analysis cc w/encl:

Distribution via Listserv

Spurious Actuation Results in Unit 2 Trip and Partial Loss of Surry Unit 2 Offsite Power LER: 281/2006-002 CCDP= 2x10-6 Event Date: 10/7/2006 IR: 50-281/2006-11 I

EVENT

SUMMARY

Event Description. On October 7, 2006, at approximately 1711 hours0.0198 days 0.475 hours 0.00283 weeks 6.510355e-4 months , Unit 2 was manually tripped based on indications associated with main steam flow, main steam pressure, and steam generator feedwater flow and level perturbations. Normal offsite power was lost to both Unit 1 and one of the Unit 2 emergency buses due to flying debris that impacted the A and C Reserve Service Station Transformers' (RSST) electrical conductors. Exhaust steam discharging from opened Unit 2 cross-under piping relief valves (CURV) impacted the adjacent turbine building siding creating flying debris. The dedicated Unit 1, Emergency Diesel Generator (EDG) 1, started and loaded safety system Emergency Bus 1H. EDG 3, shared between Units 1 and 2, started and automatically loaded to the 2J Emergency Bus as designed. The alternate AC diesel generator (MC DG) automatically started but was not manually loaded by the operators to an emergency bus, because a breaker lockout signal had occurred on the 1J Emergency Bus normal supply breaker. As expected EDG 2 did not start since the B RSST was not affected, and continued to supply power to the Unit 2, 2H Emergency Bus. This left the Unit 1, 1J Emergency Bus deenergized.

Due to loss of Emergency Bus 1J power, a semi-vital bus also lost power on Unit 1. This semi vital bus powers non-safety related loads associated with secondary side systems. To stabilize the unit from the steam/feedwater transient induced by the loss of normal power to secondary side equipment, Unit 1 operators lowered power to approximately 71%. Unit 2 was stabilized in hot shutdown. At 1911 EDG 3, the shared EDG, was placed on the 1J Emergency Bus, leavinq the 2J Emergency Bus deenergized. At 2137, EDG 3 was transferred back from the 1J Emergency Bus to the 2J Emergency Bus. Following troubleshooting, the affected breaker lockout contacts associated with protective relaying were reset and the 1J Emergency Bus was energized at 2154 hours0.0249 days 0.598 hours 0.00356 weeks 8.19597e-4 months by closing its normal supply breaker.

The A RSST was not damaged, as determined by licensee inspections, and was returned to service supplying power to the Unit 1, 1J emergency bus at 0209 hours0.00242 days 0.0581 hours 3.455688e-4 weeks 7.95245e-5 months on October 8. The C RSST Bus Bar experienced minor damage and was repaired. The C RSST was returned to service supplying the F Transfer Bus at 1446 hours0.0167 days 0.402 hours 0.00239 weeks 5.50203e-4 months and normal offsite power was restored to all safety system emergency buses at 1656 hours0.0192 days 0.46 hours 0.00274 weeks 6.30108e-4 months on October 8. Further details can be found in the References 1 and 2. In addition, Appendix A provides a table containing a sequence of key events.

Cause. The cause for the perturbations in main steam flow, main steam pressure, and steam generator (SG) feedwater flow and level was due to problems with the turbine electro-hydraulic control system.

LER 281/06*002 Additional Event Details.

Unit 2 feedwater isolation occurred due to reactor trip signal in conjunction with a low Tave signal.

A Unit 2 power-operated relief valve (PORV) opened and closed 48 minutes into the event.

Unit 2 operators closed main steam trip valves due to the Main Turbine 4 Stop Valve not indicating full closed. Later investigation revealed that the valve was shut and that the invalid signal was due to an indicator problem.

Full circulating water system capability (i.e., to maintain proper canal level) was maintained throughout the event.

Recovery Opportunities. Offsite power was restored to the 2J Emergency Bus at approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the initiation of the event occurred. However, the C RSST was restored approximately 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months earlier, and therefore operators could have restored offsite power the 2J Emergency Bus approximately at that time.

Analysis Rules. The ASP program uses Significance Determination Process (SDP) results for degraded conditions when available. However, the ASP program performs independent initiating event analysis when an initiator occurs and a condition analysis when there are no performance deficiencies identified for a particular event. In addition, the ASP program analyzes separate degraded conditions that were present during the same period and similar degraded conditions on an individual system or component that had different performance deficiencies.

A few unresolved issues (i.e., potential performance deficiencies causing degraded conditions) along with two GREEN findings have been identified for this event and are described in Reference 2. Therefore, this analysis focuses solely on the risk of the reactor trip and partial loss of offsite power (LOOP) that occurred at Unit 2.

ANALYSIS RESULTS Conditional Core Damage Probability The conditional core damage probability (CCDP), for this event is 2.2x10-6. The results of an uncertainty assessment on the CCDP are summarized below.

Surry 2 I

5%

Mean 95%

1.1x10-7 2.2x10*6 8.0x10-6 The Accident Sequence Precursor Program acceptance threshold is a CCDP of 1x 10-6 or the CCDP equivalent of an uncomplicated reactor trip with a non-recoverable loss of secondary plant systems (e.g., feedwater and condensate), whichever is greater. This CCDP equivalent for Surry 2 is 4x10-7.

Dominant Sequence The dominant accident sequences, Loss of Condenser Heat Sink (LOCHS) Sequence 06 (9.7x 10-7) and LOCHS Sequence 04 (9.0x10-7), contribute to 85% of the total internal events CCDP.

2

LER 281/06-002 Sequences involving a loss of coolant due to a postulated stuck-open PORV was the largest contributor to the overall risk from this event. The opening of a PORV 48 minutes following the actual reactor trip increased the probability of a stuck-open PORV. The loss of offsite power to one safety bus and the reliance of the associated EDG to provide power during the extended partial LOOP was an important contributor to mitigate the consequences of the postulated loss of coolant accident.

The dominant sequences are shown graphically in Figure B-1 of Appendix B. The events and important component failures in LOCHS Sequence 06 are:

LOCHS occurs due to the partial loss of offsite ac electrical power (this event is given because it occurred during the actual event),

Reactor shutdown succeeds, Auxiliary feedwater succeeds, PORV opens on demand (this event is given because it occurred during the actual event),

PORV(s) fail to close with the inability to close the PORV block valve (e.g.,

unavailability of ac power or valve fail to close), and High pressure injection (HPI) fails (e.g., unavailability of ac power to one HPI train and service water cooling to the other train).

The events and important component failures in LOCHS Sequence 04 are:

LOCHS occurs due to the partial loss of offsite ac electrical power, Reactor shutdown succeeds, Auxiliary feedwater succeeds, PORV opens on demand (this event is given because it occurred during the actual event),

PORV(s) fail to close with the inability to close the PORV block valve (e.g.,

unavailability of ac power or valve fail to close),

High pressure injection succeeds, Containment spray recirculation succeeds, and High pressure recirculation (HPR) fails (e.g., unavailability of ac power to one HPR train and valve failure in other train).

Results Tables The conditional probabilities for the dominant sequences are shown in Table 1.

The event tree sequence logics for the dominant sequences are presented in Table 2a.

Table 2b defines the nomenclature used in Table 2a.

The most important cutsets for the dominant sequences are listed in Table 3a and 3b.

Definitions and probabilities for modified or dominant basic events are provided in Table 4.

MODELING ASSUMPTIONS Analysis Type The Revision 3-Plus (Change 3.41) of the Surry 1 and 2 Standardized Plant Analysis Risk (SPAR) model (Reference 3) created in March 2008 was used for this assessment. This 3

LER 281/06*002 event was modeled as a Unit 2, loss of condenser heat sink initiating event with the unavailability of offsite power to Emergency Buses 1J, 1H, and 2J.

Unique Design Features EDG 1 provides power to Emergency Bus 'IH only (see Figure 1).

EDG 2 provides power to Emergency Bus 2H only (see Figure 1).

EDG 3 (swing diesel) provides power to Emergency Buses 1J and 2J (see Figure 1).

The swing EDG is preferentially aligned to Unit 2.

The AAC DG can be aligned to either Emergency Bus 1J or 2H given that the diesels that normally align to these buses have failed (see Figure 1).

The diesel generators do not require cooling water.

The swing diesel generator (EDG 3) will be available for the unit of concern unless the dedicated diesel generator for the other unit is not available. In that case, the swing diesel will be aligned to the other unit.

Bus 2H Bus 2J Bus 1J I

I I

I Surry 1 and 2 simplified diagram of EDG power distribution.

Bus 1H Figure 1.

Unit 1 Unit 2 Modeling Assumptions Summary Key Modeling Assumptions. Offsite power was unavailable to supply power to Emergency Buses 1J, 'IH, and 2J and assumed to be unrecoverable for the dominant accident sequences in this analysis. This assumption is based on the timing of the dominant accident sequences and the long recovery time (l.e., at least 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months were needed to recover the C RSST). Emergency Bus 2H never lost offsite power; therefore, EDG 2 was not required to operate and Emergency Bus 1J was modeled as de-energized due to the failure of the AAC DG to load.

4

LER 281/06-002 Other Modeling Assumptions. A Unit 2 PORV was modeled as cycling open and closed during the event.

Fault Tree Modifications/SPAR Model Corrections Several fault tree modifications were required for this event analysis. These modifications were required for two reasons: (1) the SPAR model is specifically designed for Unit 1 analyses (i.e., model changes were required to perform a Unit 2 analysis) and (2) errors existed in the modeling which safety buses the AAC DG supplied could supply power (see Unique Design Features).

Modeling for Unit 2 Analyses, The current SPAR model for Surry is designed for Unit 1 analyses. Because this analysis is for a Unit 2 initiating event analysis, fault trees covering the emergency power system needed to be changed. This was performed by using the Unit 1 emergency bus (Bus 1Hand 1J) fault trees to represent the Unit emergency buses (Bus 2H and 2J). To do this, the appropriate EDG(s) were moved to their representative fault tree (e.g., EDG 1 was moved to the Bus 2H fault tree). In addition to modifying EDG logic in the emergency power system fault trees, the existing LOOP-related house events (LOOP-1J, LOOP-1 H, LOOP-2J, LOOP-2H, LOOP) were moved. The logic diagrams for the modified fault trees (ACP-1 H, ACP-1J, ACP-2H, ACP 2J, ACP-F, DIV-H-AC, and DIV-J-AC) are provided in Appendix C.

Modeling of the AAC DG. The current SPAR model incorrectly models the AAC DG as powering Emergency Bus 2J instead of Emergency Bus 2H. However, for this analysis (because we are using a Unit 1 model for a Unit 2 analysis), the AAC DG was modeled as supplying power to Emergency Bus 1H (representing Emergency Bus 2H) and Emergency Bus 2J (representing Emergency Bus 1J in this analysis). See Appendix C for the logic diagrams of modified fault trees.

Modeling of Service water availability. Full circulating water capability was available throughout the event to maintain the proper canal level for the gravity-fed service water system. The circulating water is supplied by a separate offsite power source than the safety busses. The main circulating water (MCW) fault tree was modified to delete the safety bus LOOP flag events (LOOP-'I Hand LOOP-2J). In addition, the undeveloped basic event MCW-SYS-FC-UNAVL was modified to FALSE and added to the LOOP Flag sets. See Appendix C for the logic diagrams of this modified fault tree.

Basic Event Probability Changes Table 4 provides all the basic events that were modified to reflect the best estimate of the conditions during the event. The basis for these changes is provided below:

ACP-BAC-LP-2J set to TRUE. This basic event represents the Unit 1 Emergency Bus 1J that lost its offsite power supply and had no emergency power source aligned to supply power; therefore, this event was set to TRUE.

DUAL-UNIT-LOOP set to FALSE. This basic event was set to FALSE because a dual-unit LOOP did not occur.

5

LER 281/06-002 LOOP-1J set to TRUE. This flag event represents the Unit 2 Emergency Bus 2J. The offsite power supply was lost to Bus 2J; therefore, this event was set to TRUE. Offsite power to the other Unit 2 Emergency Bus 2H was never lost.

LOOP-2H set to TRUE. This flag event represents the Unit 1 Emergency Bus 1H.

The offsite power supply was lost to Bus 1H; therefore, this event was set to TRUE.

LOOP-2J set to TRUE. This flag event represents the Unit 1 Emergency Bus 1J. The offsite power supply was lost to Bus 1J; therefore, this event was set to TRUE.

IE-LOCHS set to 1.0. Set the initiating event frequency to 1.0 due to loss of condenser heat sink event at Unit 2. All other initiating event frequencies were set to zero.

PPR-SRV-CO-TRAN set to TRUE. This event was set to TRUE because a Unit 2 PORV opened and closed 48 minutes into the event.

ZT-DGN-FR-L Mission Time set to 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . The mission time for this EDG template event was changed to 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months because of the offsite power recovery was recoverable within 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months after the event initiated.

REFERENCES

1. LER 281/06-002 Rev. 0, "Spurious Actuation Results in Unit 2 Trip and Loss of Offsite Power," October 7, 2006.

2. U.S. Nuclear Regulatory Commission, "Surry Power Station-NRC Special Inspection Report 05000280/2006011," December 1, 2006.

3. Idaho National Laboratory, "Standardized Plant Analysis Risk Model for Surry 1 and 2,"

Revision 3 Plus (Change 3.41), March 2008.

6

LER 281/06-002 Table 1. conditional core dama ~e probabilities 0 f the demmauIng sequences.

EventTree Name

~,~~~.

Nu CCDp1 LOCHS 06 9.7E-7 9.0E-7 2.2E-6 LOCHS 04 Total(all sequences)2 Contrlbutlol1(01..)

44.1 40.9 100

1. Values are point estimates.

2. Total CCDP includes all sequences (including those not shown in this table).

Logic

("'" denotes success; see Table2b for top event names)

LOCHS 06 IRPS IAFW PORV HPI LOCHS 04 IRPS IAFW PORV HPI ICSR HPR AFW Auxiliary Feedwater CSR Containment Spray Recirculation HPI High Pressure Injection HPR High Pressure Recirculation PORV PORVs are Closed RPS Reactor Trip 7

LER 281/06-002 Table 3. Conditional cutsets for the dominant sequences,

~

... >>"n nn:nf*

"n"

><nH"

~.*.....

CC......

Event Tree: LOCHS, Sequence 06 EPS-DGN-FR-DG3 PPR-SRV-OO-2 CPC-MDP-TM-10A 3.3E-7 33.7 EPS-DGN-TM-DG3 PPR-SRV-OO-2 CPC-MDP-TM-10A 18.6 1.8E-7 7,7 EPS-DGN-FS-DG3 PPR-SRV-OO-2 CPC-MDP-TM-10A 7.5E-8 4,5 EPS-DGN-FR-DG3 PPR-SRV-OO-2 CPC-MDP-FS-10A 2,9E-8 4.4E-8 EPS-DGN-FR-DG3 PPR-SRV-OO-SR1 CPC-MDP-TM-10A PPR-MOV-FC-BLK1 PPR-MOV-FC-BLK2 2,9E-8 3.0 EPS-DGN-FR-DG3 PPR-SRV-OO-SR2 CPC-MDP-TM-10A PPR-MOV-FC-BLK1 PPR-MOV-FC-BLK2 2,9E-8 3.0 3,0 EPS-DGN-FR-DG3 PPR-SRV-OO-SR3 CPC-MDP-TM-10A PPR-MOV-FC-BLK1 PPR-MOV-FC-BLK2 2.4E-8 EPS-DGN-TM-DG3 PPR-SRV-OO-2 CPC-MDP-FS-10A 3.0 Total (all cutsets)"

9.7E-7 100 Percent CCDP Minimum C:utsets (of basic events)

Contribution Event Tree: LOCHS, Sequence 04 EPS-DGN-FR-DG3 PPR-SRV-OO-2 HPR-XHE-XM-1115D 2.2E-7 24.2 1.2E-7 EPS-DGN-TM-DG3 PPR-SRV-OO-2 HPR-XHE-XM-1115D 13.4 10,0 9.0E-8 EPS-DGN-FS-DG3 PPR-SRV-OO-SR1 HPR-MOV-OO-1373 PPR-MOV-FC-BLK1 9.0E-8 10.0 EPS-DGN-FR-DG3 PPR-SRV-OO-SR2 HPR-MOV-OO-1373 PPR-MOV-FC-BLK1 9.0E-8 10.0 EPS-DGN-FR-DG3 PPR-SRV-OO-SR3 HPR-MOV-OO-1373 PPR-MOV-FC-BLK1 5.0E-8 5,6 EPS-DGN-FS-DG3 PPR-SRV-OO-2 HPR-XHE-XM-1115D 2,2E-8 2.4 EPS-DGN-FR-DG3 PPR-SRV-OO-2 HPR-MOV-OO-1373 Total (all cutsets)"

9.0E-7 100

1. Total CCDP Includes all cutsets (Including those not shown In this table).

8

LER 281/06-002 T bl 4 a e b bilitl d

0 fi iti e In! Ions an pro a Illes t b dT d d d or mo lie an ormnan asic even s.

Event Name Description Probabllityl Freq\\.len~y Modified (per year)

ACP-BAC-LP-2J DIVISION 2J AC POWER 4160V BUS 2J FAILS TRUE Yes CPC-MDP-FS-10A CPC SWS TRAIN A FAILURES TO START 2.0E-3 No CPC-MDP-TM-1OA CPC SWS MDP 10A UNAVAILABLE DUE TO T&M 1.5E-2 No DUAL-UNIT-LOOP LOOP AFFECTING BOTH UNITS FALSE Yes EPS-DGN-FR-DG3 DIESEL GENERATOR 3 FAILS TO RUN 2.4E-2 No EPS-DGN-FS-DG3 DIESEL GENERATOR 3 FAILS TO START 5.0E-3 No EPS-DGN-TM-DG3 DIESEL GENERATOR 3 UNAVAILABLE DUE TO T&M 1.2E-2 No HPR-MOV-OO-1373 MINFLOW ISOLN VLV 1-CH-MOV-1373 FAILS TO CLOSE 1.0E-3 No HPR-XHE-XM-1115D UNDEVELOPED EVENT TO MANUALLY CLOSE 1115D 1.0E-2 No LOOP-1J LOSS OF DIVISION 1J OFFSITE POWER TRUE Yes LOOP-2H LOSS OF DIVISION 2H OFFSITE POWER TRUE Yes LOOP-2J LOSS OF DIVISION 2J OFFSITE POWER TRUE Yes IE-LOCHS INITIATING EVENT-LOSS OF CONDENSER HEAT SINK 1.0 Yes' PPR-MOV-FC-BLK1 BLOCK VALVE 1535 CLOSED DUE TO PORV LEAKING 3.0E-1 No PPR-MOV-FC-BLK2 BLOCK VALVE 1536 CLOSED DUE TO PORV LEAKING 3.0E-1 No PPR-SRV-CO-TRAN PORVs/SRVs OPEN DURING TRANSIENT TRUE Yes PPR-SRV-OO-2 PORV 2 FAILS TO RECLOSE AFTER OPENING 1.0E-3 No PPR-SRV-OO-SR1 FAILURE OF SRV 1 TO RECLOSE 1.0E-3 No PPR-SRV-OO-SR2 FAILURE OF SRV 2 TO RECLOSE 1.0E-3 No PPR-SRV-OO-SR3 FAILURE OF SRV 3 TO RECLOSE 1.0E-3 No ZT-DGN-FR-L EDG FAILS TO RUN AFTER 1 HOUR OF OPERATION 1.6E-2 Yes2

1. Set the IE frequency to 1.0. All other initialing event frequencies were set to zero.

2. Adjusted the mission time to 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . See the Basic Event Probability Section for further details.

9

LER 281/06*002 APPENDIX A SEQUENCE OF KEY EVENTS Table A-1. Seguence of Key Events for Surry 2 Loss of Condenser Heat Sink Transient.

Time Event 17:11:18; Unit 2 was at 100% power when the main turbine governor and intercept valves spuriously 10/17/06 close due to a main turbine electro-hydraulic control (EHC) system problem. Subsequently, the governor valves reopened followed by the intercept valves opening within a few seconds.

This resulted in Unit 2 generator output decreasing from 852 MWe to 0 MWE and back to over 200 MWe. Additionally, this transient impacted the main steam and feedwater systems resulting in multiple alarms including steam flow - feedwater flow mismatch for all three steam generators (SGs), generator motoring - turbine low differential pressure, high steam flow and level error for all three SGs.

17:11:27 Unit 2 high pressure turbine exhaust pressure increased above normal operating pressure because the high pressure turbine control valves reopened prior to the low pressure turbine intercept valves. Consequently, all 12 moisture separator reheater (MSR) CURVs (6 on each side of the turbine building) opened to relieve the excessive pressure (maximum was 300 psig) created by the valve opening sequence. During this time, sections of the turbine building outside wall were removed and ejected into the air by the steam exhaust flow from the CURVs opening.

17:11:32 Some of the turbine building wall sections ejected by the steam flow contacted conductors associated with the A and C RSSTs which subsequently resulted in a fault on these transformers and a loss of power to their loads. The A RSST is the normal power feed to the 1J 4160V Emergency Bus and the C RSST is the normal power feed to the 1Hand 2J 4160V Emergency Buses. The B RSST which was unaffected continued to power the 2H 4160V Emergency Bus during the event.

The loss of the 1J Emergency Bus also resulted in the loss of the Unit 1 semi-vital bus. Unit 1 operators entered the applicable abnormal operating procedures, started a third condensate pump and initiated a power reduction to approximately 73% due to affected secondary side plant systems.

17:11:34 Alarm for Unit 2 pressurizer PORV open was received (spurious due to loss of the semi-vital bus).

17:11:38 Due to the severe feedwater and steam system transients, related SG level error alarms, Tave - Tref deviation alarm, and the sound of steam flow from the turbine building, the Unit 2 control room operators initiated a manual reactor trip, which caused a main turbine trip, and entered their emergency procedure E-O, "Reactor Trip or Safety Injection."

After a turbine trip the station service loads are normally transferred automatically to the RSSTs. However, since A and C RSSTs were faulted, power was loss to the A and C Reactor Coolant Pumps. The forced flow from these reactor coolant pumps is the source of pressurizer spray flow.

17:11:43 Unit 2 auxiliary feedwater pumps started based on steam generator low-low levels.

17:11:56 Unit 2 feedwater isolation occurred due to reactor trip signal in conjunction with a low Tave signal.

17:14:18 Unit 2 operators closed main steam trip valves due to the Main Turbine 4 Stop Valve not indicating full closed.

17:16 Unit 2 operators transitioned to emergency procedure (EP) 0.1, "Reactor Trip Response."

17:18 Unit 1 semi-vital bus power supply was swapped from 1J1 Bus to 1H1 Bus (powered from the 1H Emergency Bus).

17:51 Unit 1 was stable at 71% power.

A-1

LER 281/06*002 Time Event 17:59:20 Unit 2 pressurizer PORV opened and closed.

18:15 Unit 1 operators attempted to supply power to the 1J Emergency Bus from the MC DG; however, the 1J Emergency Bus normal supply breaker, 15J8. fails to close.

19:11 Operators transferred EDG 3 from the 2J to 1J Emergency Bus.

21:37 Operators transferred EDG 3 back to the 2J Emergency Bus.

21:54 Operators were successful in energizing the 1J Emergency Bus via the MC DG.

02:09; A RSST was re-energized from offsite power and aligned to D Transfer Bus and the 1J 10108106 Emergency Bus. The MC DG was subseguently removed from service.

08:59 Unit 1 commenced power increase to 100% rated thermal power.

10:42 13:50 Unit 1 was at 100% rated thermal power.

C RSST was re-energized following repairs to the transformer's conductors.

14:46 F Transfer Bus was re-energized; offsite power now restored to onsite emergency buses.

15:12 EDG 1 was removed from the 1H Emergency Bus.

16:56 EDG 3 was removed from the 2J Emergency Bus.

A-2

LER 281/06-002 APPENDIX B EVENT TREE WITH DOMINANT SEQUENCE HIGHLIGHTED LOSS OF REACTOR AUXILIARY 00""'. ~""" I'~.,'"

HEAT SINK SYSTEM

-l IE LOCHS RPS AFW MAIN PORVs Rep SEAL HIGH CONTAINMENT HIGH PRESSURE SPRAY PRESSURE CLOSED MAINTAINED FEEDWATER ARE COOLING INJECTION RECIRC RECIRC HPI FAB CSR HPR MFW PQRV Lose I

I I

I I,

I PORVl I

I I

I LOW PRESSURE RECIRC n-LPR

/I END-STATB OK RCPSL OK CD CO CD OK RCPSl OK 10 CD 11 CD CD 13 OK 14 CD 15 CO 15 CD 17 ATWS Figure B-1. Loss of condenser heat sink event tree with dominant sequences highlighted.

B-1

LER 281/06-002 APPENDIX C MODIFIED FAULT TREE LOGIC DIAGRAMS lOGIC

'" acp-1h OR SURRY 1 & 2 4KV BUS 1H POWER SYSTEM ACP-BAC-LP-1 H. (9.600E-006). DIVISION 1HAC POWER 4160V BUS 1H FAILS acp-1h-1 AND LOSS OF POWER TO 1H 4160VAC BUS

'" acp-1h-3 OR OFFSITE POWER IS UNAVAILABLE LOOP-1H, (O.OOOE+OOO). LOSS OF OFFSITE POWER IE HAS OCCURRED A. acp-f TRAN 4.16KV BUS F IS UNAVAILABLE A. eps-dg2 TRAN DIESEL GENERATOR 2 IS UNAVAILABLE A. eps-sbo TRAN STATION BLACKOUT DIESEL IS UNAVAILABLE acp-1h-2 AND ROOM COOLING IS UNAVAILABLE ACP-XHE-XM-RCOOL. (1.000E+OOO), OPERATOR FAILS TO ESTABLISH ROOM COOLING WIO ESGR A. esgr TRAN LOSS OF EMERGENCY SWITCHGEAR ROOM COOLING Figure C-1. Modified fault tree for Emergency Bus 1H (representing Emergency Bus 2H).

LOGIC

'" acp-1j OR SURRY 1 & 2 4KV BUS 1J POWER SYSTEM ACP-BAC-LP-1J. (9.600E-006). DIVISION 1J AC POWER 4160V BUS 1J FAILS acp-1j-1 AND LOSS OF POWER TO 1J 4160V AC BUS

'" acp-1j-2a OR OFFSITE POWER IS UNAVAILABLE LOOP-1J. (1.000E+OOO). LOSS OF OFFSITE POWER IE HAS OCCURRED A. acp-d TRAN 4.16KV BUS D IS UNAVAILABLE

'" acp-1j-5 OR DIESEL GENERATOR 3 IS UNAVAILABLE EPS-XHE-XM-DG31J, (O.OOOE+OOO), OPERATOR FAILS TO ALIGN DG 3 TO BUS 1J A. eps-dg3 TRAN DIESEL GENERATOR 3 IS UNAVAILABLE acp-1j-2 AND ROOM COOLING IS UNAVAILABLE ACP-XHE-XM-RCOOL, (1.000E+OOO), OPERATOR FAILS TO ESTABLISH ROOM COOLING WIO ESGR A. esgr TRAN LOSS OF EMERGENCY SWITCHGEAR ROOM COOLING Figure C-2. Modified fault tree for Emergency Bus 1J (representing Emergency Bus 2J).

LOGIC

'" acp-2h OR SURRY 1 & 2 4KV BUS 2H POWER SYSTEM ACP-BAC-LP-2H, (9.600E-006), DIVISION 2H AC POWER 4160V BUS 2H FAILS acp-2h-1 AND LOSS OF POWER TO 2H 4160V AC BUS LOOP-2H, (1.000E+OOO), LOSS OF OFFSITE POWER IE HAS OCCURRED A. eps-dg1 TRAN DIESEL GENERATOR 1 IS UNAVAILABLE acp-2h-2 AND ROOM COOLING IS UNAVAILABLE ACP*XHE-XM-RCOOL, (1.000E+OOO), OPERATOR FAILS TO ESTABLISH ROOM COOLING WIO ESGR A. esgr TRAN LOSS OF EMERGENCY SWITCHGEAR ROOM COOLING Figure C-3. Modified fault tree for Emergency Bus 2H (representing Emergency Bus 1H).

C-1

LER 281/06*002 LOGIC A acp-2j OR SURRY 1 & 2 4KV BUS 2J POWER SYSTEM ACP-BAC-LP-2J, (9.600E-006), DIVISION 2J AC POWER 4160V BUS 2J FAILS acp-2j-1 AND LOSS OF POWER TO 2J 4160V AC BUS LOOP-2J, (1.000E+OOO), LOSS OF DIVISION 2J POWER HAS OCCURRED acp-2j-3 AND FAILURE OF POWER TO BUS 2J A acp-2j-4 OR OFFSITE POWER IS UNAVAILABLE EPS-XHE-XM-SB02J, (1.000E+OOO), OPERATOR FAILS TO ALIGN SBO DIESEL TO BUS 2J

... eps-sbo TRAN STATION BLACKOUT DIESEL IS UNAVAILABLE A acp-2j-5 OR DIESEL GENERATOR 3 IS UNAVAILABLE EPS-XHE-XM-DG32J, (1.000E+OOO), OPERATOR FAILS TO ALIGN DG 3 TO BUS 2J

... eps-dg3 TRAN DIESEL GENERATOR 3 IS UNAVAILABLE acp-2j-2 AND ROOM COOLING IS UNAVAILABLE ACP-XHE-XM-RCOOL, (1.000E+OOO), OPERATOR FAILS TO ESTABLISH ROOM COOLING WIO ESGR

... esgr TRAN LOSS OF EMERGENCY SWITCHGEAR ROOM COOLING Figure C-4. Modified fault tree for Emergency Bus 2J (representing Emergency Bus 1J).

LOGIC A acp-f OR 4.16KV BUS F IS UNAVAILABLE ACP-BAC-LP-F, (9.600E-006), 4.16KV BUS F IS UNAVAILABLE Figure CoS. Modified fault tree for Non-vital Bus F.

LOGIC A div-h-ac OR DIVISION 1H POWER FAILS ACP-BAC-LP-1H, (9.600E-006), DIVISION 1H AC POWER 4160V BUS 1H FAILS II div-h-ac-1 AND LOSS OF POWER TO 1H 4160V AC BUS A div-h-ac-3 OR OFFSITE POWER IS UNAVAILABLE LOOP-1H, (O.OOOE+OOO), LOSS OF OFFSITE POWER IE HAS OCCURRED

... acp-f TRAN 4.16KV BUS F IS UNAVAILABLE

... eps-dg2 TRAN DIESEL GENERATOR 2 IS UNAVAILABLE

... eps-sbo TRAN STATION BLACKOUT DIESEL IS UNAVAILABLE div-h-ac-z AND ROOM COOLING IS UNAVAILABLE ACP-XHE-XM-RCOOL, (1.000E+OOO), OPERATOR FAILS TO ESTABLISH ROOM COOLING WIO ESGR

... esgr TRAN LOSS OF EMERGENCY SWITCHGEAR ROOM COOLING Figure C-6. Modified fault tree for Division 1H (representing Division 2H).

.LOGIC A div-j-ac OR DIVISION 1J POWER FAILS ACP-BAC-LP-1J, (9.600E-006), DIVISION 1J AC POWER 4160V BUS 1J FAILS III div-j-ac-t AND LOSS OF POWER TO 1J 4160V AC BUS A div-j-ac-2a OR OFFSITE POWER IS UNAVAILABLE LOOP-1J, (1.000E+OOO), LOSS OF OFFSITE POWER IE HAS OCCURRED

... acp-d TRAN 4.16KV BUS D IS UNAVAILABLE A div-j-ac-5 OR DIESEL GENERATOR 3 IS UNAVAILABLE EPS-XHE-XM-DG31J, (O.OOOE+OOO), OPERATOR FAILS TO ALIGN DG 3 TO BUS 1J

... eps-dg3 TRAN DIESEL GENERATOR 3 IS UNAVAILABLE div-j-ac-2 AND ROOM COOLING IS UNAVAILABLE ACP-XHE-XM-RCOOL, (1.000E+OOO), OPERATOR FAILS TO ESTABLISH ROOM COOLING WIO ESGR

... esgr TRAN LOSS OF EMERGENCY SWITCHGEAR ROOM COOLING Figure Co?~ Modified fault tree for Division 1J (representing Division 2J).

C-2

LER 281/06-002 LOGIC

'" mew OR SURRY 1 & 2 MAIN CIRCULATING WATER IS UNAVAILABLE MCW-SYS-FC-UNAVL, (O.OOOE+OOO), MAIN CIRCULATING WATER IS UNAVAILABLE Figure C-8. Modified fault tree for main circulating water.

C-3

LER 281/06-002 APPENDIX C BEST ESTIMATE GEM RUN I NIT I

A TIN G EVE N T ASS E SSM E N T Code Ver : 7:27 Fam SURY 3P Model Ver : 2008/01/31 User INL Init Event: IE-LOCHS Ev ID: BEST-ESTIMATE Total CCDP:

2.2E-006 Desc : LOCHS with Partial LOOP Mean CCDP:

2.2E-006 BASIC EVENT CHANGES Event Name Description Base Prob Curr Prob Type DUAL-UN IT-LOOP LOOP AFFECTING BOTH UNITS 1.0E+000 +O.OE+OOO FALSE IE-ISL-HPI ISLOCA 2-CKV HPI INTERFACE 3.5E-006 +O.OE+OOO IE-ISL-LPI ISLOCA 2-CKV LPI INTERFACE 3.5E-006 +O.OE+OOO IE-ISL-RHR RHR pipe ruptures 4.0E-006 +O.OE+OOO IE-LLOCA LARGE LOCA 2.5E-006 +O.OE+OOO IE-LOACB-1H LOSS OF 4160 VAC BUS 1H 4.5E-003 +O.OE+OOO IE-LOACB-1J LOSS OF 4160 VAC BUS 1J 4.5E-003 +O.OE+OOO IE-LOCCW LOSS OF COMPONENT COOLING WA 4.0E-004 +O.OE+OOO IE-LOCHS LOSS OF CONDENSER HEAT SINK 8.0E-002 1.0E+000 IE-LOCW LOSS OF CIRCULATING WATER 4.0E-004 +O.OE+OOO IE-LODCB-1A LOSS OF VITAL 125 VDC BUS 1 1.2E-003 +O.OE+OOO IE-LOIAS LOSS OF INSTRUMENT AIR 1.0E-002 +O.OE+OOO IE-LOMFW LOSS OF MAIN FEEDWATER 1.0E-001 +O.OE+OOO IE-LOOP LOSS OF OFFSITE POWER

+O.OE+OOO +O.OE+OOO IE-LOSC LOSS OF ALL RCP SEAL COOLING +O.OE+OOO +O.OE+OOO IE-MLOCA MEDIUM LOCA 2.0E-004 +O.OE+OOO IE-SGTR STEAM GENERATOR TUBE RUPTURE 4.0E-003 +O.OE+OOO IE-SLOCA SMALL LOCA 6.0E-004 +O.OE+OOO IE-TRANS GENERAL PLANT TRANSIENT 8.0E-001 +O.OE+OOO IE-XLOCA VESSEL RUPTURE 1.0E-007 +O.OE+OOO LOOP-1J LOSS OF OFFSITE POWER IE HAS +O.OE+OOO 1.0E+000 TRUE LOOP-2H LOSS OF DIVISION 2H OFFSITE

+O.OE+OOO 1.0E+000 TRUE LOOP-2J LOSS OF DIVISION 2J OFFSITE

+O.OE+OOO 1.0E+000 TRUE PPR-SRV-CO-TRAN PORVs/SRVs OPEN DURING TRANS 4.0E-002 1.0E+000 TRUE ZT-DGN-FR-L Emergency Diesel Generator (

1.8E-002 1.6E-002 SEQUENCE PROBABILITIES Truncation Cumulative 100.0%

Individual

1. 0%

Event Tree Name Sequence Name CCDP

%Cont LOCHS LOCHS LOCHS LOCHS LOCHS 06 04 17-28 17-13 05 9.7E-007 9.0E-007 1.6E-007 1.lE-007 4.3E-008 D-1

LER 281/06*002 SEQUENCE LOGIC Event Tree Sequence Name Logic LOCHS 06

/RPS

/AFW PORV HPI LOCHS 04

/RPS

/AFW PORV

/HPI

/CSR HPR LOCHS 17-28 RPS RCSPRESS LOCHS 17-13 RPS

/RCSPRESS

/MFW2 BORATION LOCHS 05

/RPS

/AFW PORV

/HPI CSR Fault Tree Name Description AFW AUXILIARY fEEDWATER BORATION EMERGENCY BORATION CSR CONTAINMENT SPRAY RECIRC HPI HIGH PRESSURE INJECTION HPR HIGH PRESSURE RECIRC MFW2 MAIN FEEDWATER PORV PORVs ARE CLOSED RCSPRESS RCS PRESSURE LIMITED RPS REACTOR TRIP SEQUENCE CUT SETS Truncation:

Event Tree: LOCHS Sequence:

06 CCDP

% Cut Set 3.3E-007 33.78 1.8E-007 18.62 7.5E-008 7.76 4.4E-008 4.50 2.9E-008 3.04 2.9E-008 3.04 Cumulative: 100.0%

EPS-DGN-FR-DG3 CPC-MDP-TM-10A EPS-DGN-TM-DG3 CPC-MDP-TM-I0A EPS-DGN-fS-DG3 CPC-MDP-TM-10A EPS-DGN-fR-DG3 CPC-MDP-FS-10A PPR-MOV-FC-BLK1 EPS-DGN-fR-DG3 CPC-MDP-TM-10A PPR-MOV-fC-BLK1 EPS-DGN-fR-DG3 CPC-MDP-TM-10A 0-2 Individual:

1.0%

CCDP:

9. 7E-007 Cut Set Events PPR-SRV-OO-2 PPR-SRV-OO-2 PPR-SRV-OO-2 PPR-SRV-OO-2 PPR-MOV-fC-BLK2 PPR-SRV-OO-SR2 PPR-MOV-fC-BLK2 PPR-SRV-OO-SR3

LER 281106*002 2.9E-008 3.04 2.4E-008 2.48 1.6E-008

1. 70 1.6E-008

1. 68 1.6E-008

1. 68 1.0E-008

1. 03 Event Tree: LOCHS Sequence:

04 CCDP

% Cut Set 2.2E-007 24.22 1.2E-007 13.35 9.0E-008 10.02 9.0E-008 10.02 9.0E-008 10.02 5.0E-008 5.56 2.2E-008 2.42 2.0E-008 2.18 2.0E-008 2.18 2.0E-008 2.18 1.7E-008

1. 94 1.2E-008

1. 34 1.lE-008

1. 20 1.lE-008

1. 20 PPR-MOV-FC-BLK1 EPS-DGN-FR-DG3 CPC-MDP-TM-10A EPS-DGN-TM-DG3

'CPC-MDP-FS-10A EPS-DGN-FR-DG3 HPI-SYS-FC-CSTBY HPI-XHE-XM-STBY PPR-MOV-FC-BLK1 EPS-DGN-TM-DG3 CPC-MDP-TM-10A PPR-MOV-FC-BLK1 EPS-DGN-TM-DG3 CPC-MDP-TM-10A PPR-MOV-FC-BLK1 EPS-DGN-TM-DG3 CPC-MDP-TM-10A EPS-DGN-FS-DG3 CPC-MDP-FS-10A EPS-DGN-FR-DG3 PPR-SRV-OO-2 EPS-DGN-TM-DG3 PPR-SRV-OO-2 PPR-MOV-FC-BLK1 PPR-SRV-OO-SR3 PPR-MOV-FC-BLK1 PPR-SRV-OO-SR2 PPR-MOV-FC-BLK1 PPR-SRV-OO-SR1 EPS-DGN-FS-DG3 PPR-SRV-OO-2 EPS-DGN-FR-DG3 HPR-MOV-OO-1373 PPR-MOV-FC-BLK1 EPS-DGN-FR-DG3 PPR-SRV-OO-SR3 PPR-MOV-FC-BLK1 EPS-DGN-FR-DG3 PPR-SRV-OO-SR1 PPR-MOV-FC-BLK1 EPS-DGN-FR-DG3 PPR-SRV-OO-SR2 EPS-DGN-FR-DG3 PPR-SRV-OO-2 EPS-DGN-TM-DG3 HPR-MOV-OO-1373 PPR-MOV-FC-BLK1 EPS-DGN-TM-DG3 PPR-SRV-OO-SR2 PPR-MOV-FC-BLK1 EPS-DGN-TM-DG3 PPR-MOV-FC-BLK2 PPR-SRV-OO-SR1 PPR-SRV-OO-2 HPI-SYS-FC-ASTBY PPR-SRV-OO-2 HPI-XHE-XM-U2XTIE PPR-MOV-FC-BLK2 PPR-SRV-OO-SR3 PPR-MOV-FC-BLK2 PPR-SRV-OO-SR2 PPR-MOV-FC-BLK2 PPR-SRV-OO-SR1 PPR-SRV-OO-2 CCDP:

9.0E-007 Cut Set Events HPR-XHE-XM-1115D HPR-XHE-XM-1115D PPR-MOV-FC-BLK2 HPR-MOV-OO-1373 PPR-MOV-FC-BLK2 HPR-MOV-OO-1373 PPR-MOV-FC-BLK2 HPR-MOV-OO-1373 HPR-XHE-XM-1115D PPR-SRV-OO-2 PPR-MOV-FC-BLK2 HPR-XHE-XM-1115D PPR-MOV-FC-BLK2 HPR-XHE-XM-1115D PPR-MOV-FC-BLK2 HPR-XHE-XM-11l5D LPI-MDP-TM-1A HPI-XHE-XL-RWST2 PPR-SRV-OO-2 PPR-MOV-FC-BLK2 HPR-XHE-XM-1115D PPR-MOV-FC-BLK2 HPR-XHE-XM-11l5D D-3

LER 281/06-002 1.lE-008

1. 20 9.6E-009

1. 07 Event Tree: LOCHS Sequence:

17-28 CCOP 2.3E-008 1.9E-008

% Cut Set 14.57 12.15 1.7E-008 1.3E-008 10.95 8.24 3.9E-009 2.50 3.9E-009 2.50 3.9E-009 2.50 3.9E-009 2.50 3.9E-009 2.50 3.9E-009 2.50 3.2E-009 2.08 3.2E-009 2.08 3.2E-009 2.08 3.2E-009 2.08 3.2E-009 2.08 3.2E-009 2.08 2.9E-009

1. 88 2.9E-009

1. 88 PPR-SRV-OO-SR1 PPR-MOV-FC-BLK1 EPS-OGN-TM-OG3 PPR-SRV-OO-SR3 EPS-OGN-TM-OG3 PPR-SRV-OO-2 RPS-BME-CF-RTBAB RPS-TXX-CF-60F8 RPS-XHE-XE-NSGNL RPS-ROO-CF-RCCAS

/RPS-CCP-TM-CHA RPS-XHE-XE-NSGNL RPS-BME-CF-RTBAB PPR-SRV-CC-SRV1 RPS-BME-CF-RTBAB PPR-SRV-CC-SRV2 RPS-BME-CF-RTBAB PPR-SRV-CC-SRV3 RPS-BME-CF-RTBAB PPR-SRV-CC-SRV3 RPS-BME-CF-RTBAB PPR-SRV-CC-SRV2 RPS-BME-CF-RTBAB PPR-SRV-CC-SRV1 RPS-TXX-CF-60F8 RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV3 RPS-TXX-CF-60F8 RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV3 RPS-TXX-CF-60F8 RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV2 RPS-TXX-CF-60F8 RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV1 RPS-TXX-CF-60F8 RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV2 RPS-TXX-CF-60F8 RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV1 RPS-ROO-CF-RCCAS PPR-SRV-CC-SRVI RPS-ROO-CF-RCCAS PPR-SRV-CC-SRV2 RPS-ROO-CF-RCCAS PPR-SRV-CC-SRV3 RPS-ROO-CF-RCCAS PPR-SRV-CC-SRV3 RPS-ROO-CF-RCCAS D-4 PPR-MOV-FC-BLK2 HPR-XHE-XM-11150 LPI-MOP-TM-1A HPI-XHE-XL-RWST2 CCOP:

1.6E-007 Cut Set Events RCS-PHN-MOOPOOR

/RPS-CCP-TM-CHA RCS-PHN-MOOPOOR RCS-PHN-MOOPOOR RPS-CCX-CF-60F8 RCS-PHN-MOOPOOR PPR-MOV-FC-BLK1 PPR-MOV-FC-BLK2 PPR-MOV-FC-BLK2 PPR-MOV-FC-BLK1 PPR-MOV-FC-BLK1 PPR-MOV-FC-BLK2

/RPS-CCP-TM-CHA PPR-MOV-FC-BLK2

/RPS-CCP-TM-CHA PPR-MOV-FC-BLK1

/RPS-CCP-TM-CHA PPR-MOV-FC-BLK2

/RPS-CCP-TM-CHA PPR-MOV-FC-BLK2 PPR-MOV-FC-BLK1 PPR-MOV-FC-BLK2 PPR-MOV-FC-BLK2 PPR-MOV-FC-BLK1 PPR-MOV-FC-BLK2

LER 281/06-002 2.9E-009

1. 88 2.2E-009

1. 41 2.2E-009

1. 41 Event Tree: LOCHS Sequence:

17-13 CCOP

% Cut Set 3.2E-008 29.27 2.7E-008 24.42 2.4E-00B 22.00 1.BE-00B 16.55 2.1E-009

1. B9 Event Tree: LOCHS Sequence:

05 CCOP

% Cut Set 4.9E-009 11.38 4.9E-009 11.38 4.9E-009 11.38 4.5E-009 10.53 4.5E-009 10.53 4.5E-009 10.53 1.2E-009 2.75 1.lE-009 2.55 PPR-SRV-CC-SRV1 RPS-ROO-CF-RCCAS PPR-SRV-CC-SRV2

/RPS-CCP-TM-CHA RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV1

/RPS-CCP-TM-CHA RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV3

/RPS-CCP-TM-CHA RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV2

/RPS-CCP-TM-CHA RPS-XHE-XE-NSGNL PPR-SRV-CC-SRV3 Cut RPS-BME-CF-RTBAB RPS-TXX-CF-60F8 RPS-XHE-XE-NSGNL RPS-ROO-CF-RCCAS

/RPS-CCP-TM-CHA RPS-XHE-XE-NSGNL RPS-UVL-CF-UVOAB RPS-XHE-XE-SIGNL Cut PPR-MOV-FC-BLK1 CSR-XHE-XR-FLANGE PPR-MOV-FC-BLK1 CSR-XHE-XR-FLANGE PPR-MOV-FC-BLK1 CSR-XHE-XR-FLANGE PPR-MOV-FC-BLK1 LPR-SMP-PG-SUMP PPR-MOV-FC-BLK1 LPR-SMP-PG-SUMP PPR-MOV-FC-BLK1 LPR-SMP-PG-SUMP EPS-OGN-FR-OG3 PPR-SRV-OO-2 LPR-SMP-PG-SUMP PPR-SRV-OO-2 D-5 PPR-MOV-FC-BLK1 RPS-CCX-CF-60F8 PPR-MOV-FC-BLK1 RPS-CCX-CF-60FB PPR-MOV-FC-BLK2 RPS-CCX-CF-60F8 PPR-MOV-FC-BLK1 RPS-CCX-CF-60F8 PPR-MOV-FC-BLK1 RPS-CCX-CF-60F8 PPR-MOV-FC-BLK2 RPS-CCX-CF-60F8 PPR-MOV-FC-BLK2 CCOP:

1.lE-007 Set Events CVC-XHE-XM-BOR

/RPS-CCP-TM-CHA CVC-XHE-XM-BOR CVC-XHE-XM-BOR RPS-CCX-CF-60FB CVC-XHE-XM-BOR CVC-XHE-XM-BOR CCOP:

4.3E-008 Set Events PPR-MOV-FC-BLK2 PPR-SRV-OO-SR1 PPR-MOV-FC-BLK2 PPR-SRV-OO-SR3 PPR-MOV-FC-BLK2 PPR-SRV-OO-SR2 PPR-MOV-FC-BLK2 PPR-SRV-OO-SR2 PPR-MOV-FC-BLK2 PPR-SRV-OO-SR1 PPR-MOV-FC-BLK2 PPR-SRV-OO-SR3 CSR-XHE-XR-FLANGE EPS-OGN-FR-OG3

LER 281106*002 8.9E-010 2.07 8.9E-010 2.07 8.9E-010 2.07 6.5E-010

1. 52 6.0E-010

1. 40 BASIC EVENTS Event Name CPC-MDP-FS-10A CPC-MDP-TM-10A CSR-MDP-CF-FSALL CSR-XHE-XR-FLANGE CVC-XHE-XM-BOR EPS-DGN-FR-DG3 EPS-DGN-FS-DG3 EPS-DGN-TM-DG3 HPI-SYS-FC-ASTBY HPI-SYS-FC-CSTBY HPI-XHE-XL-RWST2 HPI-XHE-XM-STBY HPI-XHE-XM-U2XTIE HPR-MOV-00-1373 HPR-XHE-XM-1l15D LPI-MDP-TM-1A LPR-SMP-PG-SUMP PPR-MOV-FC-BLKl PPR-MOV-FC-BLK2 PPR-SRV-CC-SRVl PPR-SRV-CC-SRV2 PPR-SRV-CC-SRV3 PPR-SRV-00-2 PPR-SRV-OO-SRl PPR-SRV-OO-SR2 PPR-SRV-00-SR3 RCS-PHN-MODPOOR RPS-BME-CF-RTBAB RPS-CCP-TM-CHA RPS-CCX-CF-60F8 RPS-ROD-CF-RCCAS RPS-TXX-CF-60F8 RPS-UVL-CF-UVDAB RPS-XHE-XE-NSGNL RPS-XHE-XE-SIGNL PPR-MOV-FC-BLKl PPR-MOV-FC-BLK2 PPR-SRV-00-SR3 CSR-MDP-CF-FSALL PPR-MOV-FC-BLKl PPR-MOV-FC-BLK2 PPR-SRV-OO-SRl CSR-MDP-CF-FSALL PPR-MOV-FC-BLKl PPR-MOV-FC-BLK2 PPR-SRV-00-SR2 CSR-MDP-CF-FSALL EPS-DGN-TM-DG3 CSR-XHE-XR-FLANGE PPR-SRV-00-2 LPR-SMP-PG-SUMP EPS-DGN-TM-DG3 PPR-SRV-00-2 (Cut Sets Only)

Description CPC SWS TRAIN A FAILURES TO START CPC SWS MOP lOA UNAVAILABLE DUE TO TEST & MAl COMMON CAUSE FAILURE OF ALL CSR MOPS TO START TEST FLANGES LEFT BLANKED AFTER l-PT-17.6 (VA OPERATOR FAILS TO INITIATE EMERGENCY BORATION DIESEL GENERATOR 3 FAILS TO RUN DIESEL GENERATOR 3 FAILS TO START DIESEL GENERATOR 3 UNAVAILABLE DUE TO T & M CHARGING PUMP P-1A IS IN STANDBY CHARGING PUMP P-1C IS IN STANDBY OPERATOR FAILS TO ALIGN HPI SUCTION TO THE U2 OPERATOR FAILS TO ALIGN AND START STANDBY CHA OPERATOR FAILS TO CROSSTIE UNIT 2 CHARGING TO MINFLOW ISOLN VLV l-CH-MOV-1373 FAILS TO CLOS UNDEVELOPED EVENT TO MANUALLY CLOSE 11150 OR LPI MOP TRAIN lA UNAVAILABLE DUE TO T & M CONTAINMENT SUMP PLUGS BLOCK VALVE 1535 CLOSED DUE TO PORV LEAKING BLOCK VALVE 1536 CLOSED DUE TO PORV LEAKING SRV-l (SV-155-1A) FAILS TO OPEN SRV-2 (SV-155-1B) FAILS TO OPEN SRV-3 (SV-155-1C) FAILS TO OPEN PORV 2 FAILS TO RECLOSE AFTER OPENING FAILURE OF SRV 1 TO RECLOSE FAILURE OF SRV 2 TO RECLOSE FAILURE OF SRV 3 TO RECLOSE MODERATOR TEMP COEFFICIENT NOT ENOUGH NEGATIV CCF OF RTB-A AND RTB-B (MECHANICAL)

CH-A IN T&M CCF 6 ANALOG PROCESS LOGIC MODULES IN 3 OF 4 CCF 10 OR MORE RCCAS FAIL TO DROP CCF 6 BISTABLES IN 3 OF 4 CHANNELS CCF UV DRIVERS TRAINS A AND B (2 OF 2)

OPERATOR FAILS TO RESPOND WITH NO RPS SIGNAL OPERATOR FAILS TO RESPOND WITH RPS SIGNAL PRE Curr Prob 2.0E-003 1.5E-002 9.8E-006 5.4E-005 2.0E-002 2.2E-002 5.0E-003 1.2E-002 3.3E-00l 3.3E-00l 1.OE-00l 2.0E-002 3.4E-00l 1.OE-003 1.OE-002 8.0E-003 5.0E-005 3.0E-00l 3.0E-00l 8.0E-003 8.0E-003 8.0E-003 1.OE-003 1.OE-003 1.0E-003

1. OE-003 1.4E-002 1.6E-006 5.0E-003 1.8E-006 1.2E-006 2.7E-006 1.OE-005 5.0E-00l 1.OE-002 D-6

January 8, 2009 D. Christian

- 2 For more information about the ASP program, see the annual ASP program report at http://www.nrc.gov/reading-rm/doc-collections/commission/secys/2007/secy2007-0176/2007 0176scy.pdf.

The enclosure is provided for your information and no response is requested. If you have any questions please contact Donna Wright at 301-415-1864.

Sincerely, lRAI John Stang, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-281

Enclosure:

Final ASP Analysis cc w/encl:

Distribution via Listserv DISTRBUTION:

Public LPL2-1 r/f RidsNrrDorlLpl2-1 (MWong) Resource RidsNrrLAMOBrien Resource (hard copy)

RidsNrrPMJStang Resource RidsNrrPMDWright Resource (hard copy)

RidsRgn2MailCenter Resource RidsOgcMailCenter Resource RidsAcrsAcnw_MailCTR Resource RidsResDraCHunter Transmittal Letter: ML083010185 orandum dated October 21, 2008 RES/DRASP NRRlLPL CHunter*

MWon I

I

ML083010185

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