ML11276A135

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Transmittal of Accident Sequence Precursor Analysis for Event on July 16, 2010
ML11276A135
Person / Time
Site: Susquehanna 
(NPF-014)
Issue date: 10/12/2011
From: Bhalchandra Vaidya
Plant Licensing Branch 1
To: Rausch T
Susquehanna
Vaidya B, NRR/DORL/LPL1-1, 415-3308
References
Download: ML11276A135 (10)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 12, 2011 Mr. Timothy S. Rausch Senior Vice President and Chief Nuclear Officer PPL Susquehanna, LLC 769 Salem Boulevard Berwick, PA 18603-0467

SUBJECT:

TRANSMITTAL OF ACCIDENT SEQUENCE PRECURSOR ANALYSIS FOR SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1

Dear Mr. Rausch:

The enclosure provides the final result of the Accident Sequence Precursor (ASP) analysis of an event which occurred at Susquehanna Steam Electric Station, Unit No.1 as documented in licensee event report 387/10-003-01 and inspection reports 50-387/10-04, 50-387/10-08. On July 16, 2010, at approximately 1520, Unit 1 received a condenser bay flood alarm. Plant operators verified that flooding was occurring into the 656' elevation of the condenser bay.

Reactor power was reduced to 40 percent via control rod insertions and a recirculation runback.

Operator attempts to isolate condenser waterboxes remotely were unsuccessful. Unit 1 was subsequently manually scrammed, main steam isolation valves were shut, and the main condenser was isolated so that the circulating water system could be shutdown. Concurrently, plant operators manually closed waterbox isolation valves and isolated the leak. The ASP analysis calculated a conditional core damage probability (CCDP) of 3.7 x 10.5.

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 most likely to lead to inadequate core cooling and severe core damage (precursors).

As described in the NRC Regulatory Issue Summary (RIS) 2006-24, "Revised Review and Transmittal Process for Accident Sequence Precursor Analyses," the Office of Nuclear Regulatory Research implemented several process changes to the ASP Program. In accordance with the RIS, this event has a CCDP less than or equal to 1 x10*4; therefore, a formal licensee review was not requested.

For more information about the ASP Program, see the annual ASP Program status report at http://www. nrc.gov/reading-rm/doc-collections/commission/secys/20 1 0/secy20 10-0125/2010 0125scy.pdf.

T. S. Rausch

- 2 The enclosure containing the final analysis report is provided for your information.

Please contact me at 301-415-3308, if you have any questions.

Sincerely, Bhalchandra K. Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-387

Enclosure:

Final Precursor Analysis cc w/encl: Distribution via Listserv

Final Precursor Anal~sis 4ait.mnk4.i!ti,igiji44!i~i*'iia".i",;**".t&.jl~l!A@ii;(ra(."';i4i¥igg; Susquehanna, Unit 1 Manual Reactor Scram due to Leakage from the Circulating Water System and Subsequent Flooding of the Condenser Bay Event Date: 07/16/2010 LER: 387/10-003-01 IRs: 50-387/10-04,50-387/10-08 dCDP= 4x 1 0.6 EVENT

SUMMARY

Brief Event Description. On July 16, 2010, at approximately 1520, Unit 1 received a condenser bay flood alarm. Plant operators verified that flooding was occurring into the 656' elevation of the condenser bay. Reactor power was reduced to 40 percent via control rod insertions and a recirculation run back. Operator attempts to isolate condenser waterboxes remotely were unsuccessful. Unit 1 was subsequently manually scrammed, main steam isolation valves (MSIVs) were shut, and the main condenser was isolated so that the circulating water (CW) system could be shutdown. Concurrently, plant operators manually closed waterbox isolation valves and isolated the leak.

Plant response following the manual reactor scram was not as expected. The integrated control system (ICS) feedwater level control (FWLC) is designed to switch to single element control on low main steam flow. Due to steam condensation and flashing on the flow instrument, measured main steam flow remained above the transition point and ICS FWLC remained in three element control. The effect of this was that while feedwater (FW) Pumps Band C automatically switched to the idle mode and the level setpoint set-down occurred as expected, FW Pump A underwent demand oscillations prior to its transition to discharge pressure mode.

Inventory continued to be added to the reactor vessel until level reached the high level turbine trip setpoint and peaked at 55 inches. Exceeding the setpoint resulted in a trip of all FW pump turbines, the high-pressure coolant injection (HPCI) turbine, the reactor core isolation cooling (RCIC) turbine, and the main turbine. It took approximately 14 minutes for reactor vessel water level to steam down less than the trip setpoint. Once level was restored below the setpoint, HPCI and RCIC were manually reinitiated for pressure and level control respectively.

Additional details are provided in References 1, 2, and 3.

Key Event Details. The following event details are significant to the modeling of this event analysis:

  • A reactor scram and subsequent loss of condenser heat sink occurred due to circulating water leak. Operators manually scrammed the plant and isolated the condenser by shutting the MSIVs and closed the waterbox isolation valves locally. No recovery of the condenser is credited in this analysis.

1 Enclosure

LER 387/10-003-01 Additional Event Information. The following event details are provided as additional information about the event. This additional information was not factored in the modeling of this analysis due to the negligible risk impact.

In an attempt to dewater the turbine building, operations personnel transferred water from the condenser area to the condensate storage tank (CST) berm using temporary pumping equipment. However, the procedure provided no guidance as to a maximum level that should be transferred to the berm to limit interactions with other safety-related equipment. On July 17, 2010, the inspectors informed PPL that water was entering the buildings housing the 'B' and 'D' Emergency Diesel Generators through conduit and a junction box (which contained instrumentation cables associated with suction transfer of HPCI and RCIC from the CST to the suppression pool). It was determined that the switches and associated unscheduled junction boxes in the berm had been submerged when water was transferred from the condenser area and that these switches and associated junction boxes were not qualified for submergence.

On July 20, 2010, RCIC swapped its suction from the CST to the suppression pool, with CST level at 24% (transfer should occur at a level of 7.5%). It was subsequently determined that the short-term failure mechanism of the level switches controlling the Technical Specifications (TS) required function was a simulated low-level condition (Le.,

a fail-safe condition that ensures that the TS safety function is preserved).

ANALY515 RE5UL T5 Change in Core Damage Probability. The conditional core damage probability (CCDP) for this event is 3.7x10*6.

The Accident Sequence Precursor (ASP) Program acceptance threshold is a CCDP of 1 x 10.6 or the CCDP equivalent of an uncomplicated reactor trip with a non-recoverable loss of secondary plant systems (e.g., feed water and condensate), whichever is greater. This CCDP equivalent for Susquehanna, Unit 1 is 1.1 x 10.6.

Dominant Sequence. The dominant accident sequence, Loss of Condenser Heat Sink (LOCHS) 44 (CCDP =3.3x10*6) contributes 87% of the total internal events CCDP. Additional sequences that contribute greater than 1 % of the total internal events CCDP are provided in Appendix A.

The dominant sequence is shown graphically in Figure B-1 in Appendix B. The events and important component failures in LOCHS Sequence 44 are:

Loss of condenser heat sink transient occurs, Reactor trip succeeds, Safety relief valves reclose (if challenged),

High-pressure injection (HPCI and RCIC) fails, and RCS depressurization fails.

SAPHIRE 8 Report. The SAPHIRE 8 Worksheets (Appendix A) provide the following:

Summary of conditional event changes, including base and change case probabilities/freq uencies.

2

LER 387/10-003-01

  • Event tree dominant results Dominant sequences (including CCDPs).

Sequence logic for all dominant sequences.

  • Referenced fault trees (including definitions).

Cutset report for each dominant sequence.

Referenced events (including definitions and probabilities for key basic events)

MODELING ASSUMPTIONS Analysis Type. The Revision 8.17 of the Susquehanna, Unit 1 Standardized Plant Analysis Risk (SPAR) Model created in September 2010 was used for this event analysis. This event was modeled as a loss of condenser heat sink initiating event with complications.

Analysis Rules. The ASP program uses Significance Determination Process results for degraded conditions when available. However, the ASP Program performs independent initiating event analysis when an initiator occurs.

Key Modeling Assumptions. The following modeling assumptions and associated basic event modifications were required for this event analysis:

  • The probability of IE-LOCHS (Initiating Event-Loss of Condenser Heat Sink) was set to 1.0; all other initiating event frequencies were set to zero.
  • The basic events HCI-MUL TIPLE-INJECT (Probability of Multiple HPCllnjections) and RCI-RESTART (Restart of RCIC is Required) were set to TRUE because both the HPCI and RCIC pumps tripped automatically due high reactor vessel water level and were restarted manually by operators.

REFERENCES

1. Susquehanna Steam Electric Station, Unit 1, ilLER 387/10-003-001 - Unit 1 Manual Reactor Scram due to Leakage from the Unit 1 Circulating Water System and Subsequent Flooding of the Unit 1 Condenser Bay," dated September 14, 2010.
2. U.S. Nuclear Regulatory Commission, "Susquehanna Steam Electric Station - NRC Integrated Inspection Report 05000387/2010004 and 05000388/2010004; Preliminary White Finding," dated November 12, 2010
3. U.S. Nuclear Regulatory Commission, "Susquehanna Steam Electric Station - NRC Inspection Report 05000387/2010008 and 05000388/2010008; Final Significance Determination of White Finding with Assessment Follow-up," dated December 16, 2010.

3

LER 387110-003-01 Appendix A: SAPHIRE 8 Worksheets Summary of Conditional Event Changes Event Description HCI-MULTIPLE-IN,IECT PROBABILITY OF MULTIPLE HPCIINJECTIONS Condo Value True Nominal Value 1.500E-1 IE-LOCHS LOSS OF CONDENSER HEAT SINKs 1.000E+O 2.000E-1 RCI-RESTART RESTART OF RCIC IS REQUIRED True 1.500E-1

a.

All other initiating events frequencies were set to zero.

Dominant Sequence Results Only items contributing at least 1.0% to the total CCDP are displayed.

EVENT TREE SEQUENCE CCDP

% CONTRIBUTION DESCRIPTION LOCHS 44 3.270E-6 87.2%

IRPS, ISRV, HPI, DEP LOCHS 14 2.682E-7 7.2%

IRPS, ISRV, IHPI, SPC, IDEP, CDS, ILPI, CSS, PCSR, CVS, Ll08 LOCHS 47-06-16 1.415E-7 3.8%

RPS, IPPR, IRRS, PCS1, ISLC, INX, ITAF, DE2 Total 3.749E-6 100.0%

Referenced Fault Trees Fault Tree Description CDS CONDENSATE CSS CONTAINMENT SPRAY CVS CONTAINMENT VENTING DE2 MANUAL REACTOR DEPRESS DEP MANUAL REACTOR DEPRESS HPI HIGH PRESSURE INJECTION (HPCI or RCIC)

Ll08 LATE INJECTION PCS1 POWER CONVERSION SYSTEM PCSR POWER CONVERSION SYSTEM RECOVERY RPS REACTOR SHUTDOWN SPC SUPPRESSION POOL COOLING Cutset Report - LOCHS 44 Only items contributing at least 1% to the total are displayed.

CCDP TOTAL%

CUTSET 3.270E-6 100 Displaying 171 of 171 Cutsets.

1.200E-6 36.69 IE-LOCHS,ADS-XHE-XM-MDEPR,HCI-MOV-CC-IVFRO,HCI-XHE-XL INJECT, RCI-TDP-FS-RSTRT, RCI-XHE-XL-RSTRT 2

6.000E-7 18.35 lE-LOCHS,ADS-XHE-XM-MDEPR,HCI-MOV-CC-1VFRO,HC1-XHE-XL INJECT,RCI-TDP-TM-TRAIN 3

4.200E-7 12.84 IE-LOCHS,ADS-XHE-XM-MDEPR,HC I-MOV-CC-IVFRO, HC I-XHE-XL INJECT,RCI-TDP-FS-TRAIN 4

2.461E-7 7.53 IE-LOCHS,ADS-XHE-XM-MDEPR,HCI-MOV-CC-IVFRO,HCI-XHE-XL INJECT,RCI-TDP-FR-TRAIN 5

1.200E-7 3.67 I E-LOCHS,ADS-XH E-XM-MDEPR,HCI-TDP-TM-TRAIN,R CI-TDP-FS RSTRT, RCI-XHE-XL -RSTRT 6

7.185E-8 2.2 IE-LOCHS,ADS-XHE-XM-MDEPR,HCI-XHE-XO-ERRO R 1,RCI-XHE-XO-ERROR 7

7.000E-8 2.14 IE-LOCHS,ADS-XHE-XM-MDEPR,HCI-TDP-FS-TRAIN,RCI-TDP-FS RSTRT,RCI-XHE-XL-RSTRT 8

6.000E-8 1.83 IE-LOCHS,ADS-XHE-XM-MDEPR,HCI-MOV-CC-IVFRO,HCI-XHE-XL A-1

LER 387110*003*01 CCDP TOTAl%

CUTSET INJECT,RCI-MOV-CC-INJEC 9

6.000E-8 1.83 IE-LOCHS,ADS-XHE-XM-MDEPR,HCI-MOV-CC-IVFRO,HCI-XHE-XL INJECT,RCI-XHE-XO-ERROR 10 4.570E-8 1.4 IE-LOCHS,DCP-BDC-CF-ALL 11 4.200E-8 1.28 IE-LOCHS,ADS-XHE-XM-MDEPR,HCI-TDP-TM-TRAIN,RCI-TDP-FS-TRAIN

11.

12

4. 1 02E-8 1.25 IE-LOCHS,ADS-XHE-XM-MDEPR,HCI-TDP-FR-TRAIN,RCI-TDP-FS RSTRT,RCI-XHE-XL-RSTRT 13 3.500E-8 1.07 IE-LOCHS,ADS-XHE-XM-MDEPR,HCI-TDP-FS-TRAIN, RCI-TDP-TM-TRAIN Cutset Report - LOCHS 14 Only items contributing at least 1% to the total are displayed.
11.

~

TOTAL%

CUTSET 2.682E-7 100 Displaying 31 of 31 Cutsets.

1 2.500E-7 93.23 IE-LOCHS,CFAILED,CVS-XHE-XM-VENT,RHR-XHE-XM-ERROR 2

4.644E-9 1.73 IE-LOCHS,CFAILED,CVS-XHE-XM-VENT,RHR-MDP-CF-START 3

3.785E-9 1.41 IE-LOCHS,CFA1LED,CVS-XHE-XM-VENT,RHR-MOV-CF-HXBPS Cutset Report - LOCHS 47-06-16 Only items contributing at least 1% to the total are displayed.

11.

CCDP TOTAL %

CUTSET 1.415E-7 100 Displaying 29 of 29 Cutsets.

1 1.700E-8 12.01 IE-LOCHS,ADS-XHE-XM-MDEPR2,RPS-SYS-FC-PSOVS 2

1.360E-8 9.61 I E-LOCHS,ADS-SRV-CC-VAL V1, RPS-SYS-FC-PSOVS 3

1.360E-8 9.61 IE-LOC HS,ADS-SRV-CC-VAL V2,RPS-SYS-FC-PSOVS 4

1.360E-8 9.61 IE-LOCHS,ADS-SRV-CC-VAL V3,RPS-SYS-F CopSOVS 5

1.360E-8 9.61 IE-LOCHS,ADS-SRV-CC-VALV4, RPS-SYS-F C-PSOVS 6

1.360E-8 9.61 IE-LOCHS,ADS-SRV-CC-VAL V5, RPS-SYS-FC-PSOVS 7

1.360E-8 9.61 I E-LOCHS,ADS-SRV-CC-VAL V6, RPS-SYS-FC-PSOVS 8

3.800E-9 2.68 IE-LOCHS,ADS-XHE-XM-MDEPR2,RPS-SYS-FC-RELA Y 9

3.040E-9 2.15 IE-LOCHS,ADS-SRV-CC-VALV1,RPS-SYS-FC-RELAY 10 3.040E-9 2.15 IE-LOCHS,ADS-SRV-CC-VAL V2,RPS-SYS-FC-RELA Y 11 3.040E-9 2.15 IE-LOCHS,ADS-SRV-CC-VAL V3,RPS-SYS-FC-RELA Y 12 3.040E-9 2.15 IE-LOCHS,ADS-SRV-CC-VAL V4,RPS-SYS-FC-RELA Y 13 3.040E-9 2.15 IE-LOCHS,ADS-SRV-CC-VALV5,RPS-SYS-FC-RELA Y 14 3.040E-9 2.15 IE-LOCHS,ADS-SRV-CC-VAL V6,RPS-SYS-FC-RELA Y 15 2.500E-9 1.77 IE-LOCHS,ADS-XHE-XM-MDEPR2,RPS-SYS-FC-CRD 16 2.000E-9 1.41 IE-LOCHS,ADS-SRV-CC-VALV1,RPS-SYS-FC-CRD 17 2.000E-9 1.41 IE-LOCHS,ADS-SRV-CC-VAL V2,RPS-SYS-FC-CRD 18 2.000E-9 1.41 IE-LOCHS.ADS-SRV-CC-VAL V3,RPS-SYS-F C-C RD 19 2.000E-9 1.41 IE-LOCHS,ADS-SRV-CC-VALV4,RPS-SYS-FC-CRD 20 2.000E-9 1.41 I E-LOCHS,ADS-SRV-CC-VALV5, RPS-SYS-FC-CRD 21 2.000E-9 1.41 IE-LOCHS,ADS-SRV-CC-VAL V6,RPS-SYS-FC-CRD Referenced Events Event Description Probability ADS-SRV-CC-VALV1 ADS VALVE FAILS TO OPEN 8.000E-3 ADS-SRV-CC-VALV2 ADS VALVE FAILS TO OPEN 8.000E-3 ADS-SRV-CC-VALV3 ADS VALVE FAILS TO OPEN 8.000E-3 A-2

LER 387/10-003-01 Event ADS-SRV-CC-VALV4 ADS-SRV-CC-VAL V5 ADS-SRV-CC-VALV6 ADS-XHE-XM-MDEPR ADS-XHE-XM-MDEPR2 CFAILED CVS-XHE-XM-VENT DCP-BDC-CF-ALL HCI-MOV-CC-IVFRO HC1-TDP-FR-TRAI N HCI-TDP-FS-TRAIN HCI-TDP-TM-TRAIN HCI-XHE-XL-INJECT HCI-XHE-XO-ERROR1 IE-LOCHS RCI-MOV-CC-INJEC RCI-TDP-FR-TRAIN RCI-TDP-FS-RSTRT RCI-TDP-FS-TRAIN RCI-TDP-TM-TRAIN RCI-XHE-XL-RSTRT RCI-XHE-XO-ERROR RHR-MDP-CF-START RHR-MOV-CF-HXBPS RHR-XHE-XM-ERROR RPS-SYS-FC-CRD RPS-SYS-FC-PSOVS RPS-SYS-FC-RELA Y Description Probability ADS VALVE FAILS TO OPEN B.OOOE-3 ADS VALVE FAILS TO OPEN B.OOOE-3 ADS VALVE FAILS TO OPEN B.OOOE-3 OPERATOR FAILS TO MANUALLY DEPRESSURIZE THE REACTOR 5.000E-4 OPERATOR FAILS TO DEPRESSURIZE THE REACTOR (ATWS) 1.000E-2 CONTAINMENT FAILURE CAUSES LOSS OF ALL INJECTION 5.000E-1 OPERATOR FAILS TO VENT CONTAINMENT 1.000E-3 4-0F-4 125 VDC BUSES FAIL FROM COMMON CAUSE 4.570E-B HPCIINJECTION VALVE FAILS TO REOPEN 1.500E-1 HPCI PUMP TRAIN FAILS TO RUN GIVEN IT STARTED 4.102E-3 HPCI PUMP FAILS TO START 7.000E-3 HPCI TRAIN IS UNAVAILABLE BECAUSE OF MAINTENANCE 1.200E-2 OPERATOR FAILS TO RECOVER HPCIINJ. VALVE REOPENING B.OOOE-1 OPERATOR FAILS TO START/CONTROL HPCIINJECTION 1.437E-1 LOSS OF CONDENSER HEAT SINK 1.000E+O RCIC INJECTION VALVE CAUSES FAILURE TO START 1.000E-3 RCIC PUMP FAILS TO RUN GIVEN THAT IT STARTED 4.102E-3 RCIC FAILS TO RESTART GIVEN START AND SHORT-TERM RUN B.OOOE-2 RCIC PUMP FAILS TO START 7.000E-3 RCIC TRAIN IS UNAVAILABLE BECAUSE OF MAINTENANCE 1.000E-2 OPERATOR FAILS TO RECOVER RCIC FAILURE TO RESTART 2.500E-1 OPERATOR FAILS TO START/CONTROL RCIC INJECTION 1.000E-3 RHR PUMPS FAIL FROM COMMON CAUSE TO START 9.288E-6 RHR HTX BYPASS VALVES FAIL FROM COMMON CAUSE 7.570E-6 OPERATOR FAILS TO START/CONTROL RHR 5.000E-4 CONTROL ROD DRIVE MECHANICAL FAILURE 2.500E-7 HCU SCRAM PILOT SOVS FAIL 1.700E-6 TRIP SYSTEM RELAYS FAIL 3.800E-7 A-3

LER 387/10-003-01 Appendix B: Key Event Tree I

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- 2 The enclosure containing the final analysis report is provired for your information.

Please contact me at 301-415-3308, if you have any questions.

Sincerely, IRAI Bhalchandra K. Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-387

Enclosure:

Final Precursor Analysis cc w/encl: Distribution via Listserv DISTRIBUTION:

PUBLIC LPL1-1 R/F RidsNRRDorILPI1-1 RidsNRRPMSusquehanna RidsNrrLASLittie (hard copy)

RidsOGCMailCenter RidsAcrsAcnw&mMailCenter GDemoss, RES/DRA RidsRg1 MailCenter PKrohn, R1 ADAMS Package: Accession No. ML11276A138 Transmittal Letter: Accession No. ML11276A135

Enclosure:

Final Precursor Analysis: Accession No. ML112411361 OFFICE LPL 1-1/PM LPL 1-1/LA RESIDRAID LPL1-1/BC LPL 1-1/PM I NAME DATE BVaidya 10/12/11 SLittle (ABaxter for) 10/12/11 RCorreia by memo dated 09/23/11 NSalgado (RGuzman for) 10/12/11 BVaidya 10/12/11 i

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