IR 05000269/2000008
| ML20113E944 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/12/2020 |
| From: | Christopher Hunter NRC/RES/DRA/PRB |
| To: | |
| Hunter C (301) 415-1394 | |
| References | |
| IR 2000008 | |
| Download: ML20113E944 (12) | |
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)LQDO Precursor Analysis Accident Sequence Precursor Program --- Office of Nuclear Regulatory Research Oconee Nuclear Station, Units 1, 2, & 3 Non-seismic 16-inch fire system piping header transited through the auxiliary building and posed a potential flooding problem should the piping rupture during a seismic event
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Unit 1 = 4.3x10-6 Units 2 & 3 = 1.2x10-6 January 24, 2003 Condition Summary Descriptio The inspection report 50-269/00-08, 50-270/00-08, and 50-287/00-08 (Ref. 1)
indicated that the non-seismic 16-inch fire system piping header transited through the auxiliary building and posed a potential flooding problem should the piping rupture during a seismic even The postulated flood could disable both the high pressure injection (HPI) pumps and the component cooling system (CCS) pump The non-seismic fire system piping was initially identified as dry by the licensee in 1972 response to NRC questions, and therefore not adversely effect safety-related equipment due to floodin However, the licensee later acknowledged (1976) that the system was always wet.
Duratio Since the piping design was always non-seismic and was water-filled, the duration is considered as the maximum for PRA analysis, one year.
Caus The cause of this event is a deficiency in the design of piping systems that can cause flooding and loss of safety-related equipment under a postulated seismic event.
Recovery opportunit According to the inspection report (Ref. 1), a study showed that it would take 45 minutes to identify and isolate a pipe break in the auxiliary buildin Therefore, no recovery for the HPI or CCS pumps is credite However, the Standby Shutdown Facility (SSF) provides a source of cooling water to the reactor coolant pump seals as a backup to the CCS pumps.
Analysis Results
Assumptions For this analysis, the main feedwater system is tripped and all HPI and CCS pumps are assumed to fai,R 50-255/01-08
SENSITIVE - NOT FOR PUBLIC DISCLOSURE The reactor coolant pump seals are of the Bingham high temperature type for Unit 2 and Unit 3, but were the low temperature Westinghouse type for Unit 1 for more than one year in accordance with the inspection report (Ref. 1). Although recovery of seal cooling is available for all three units within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> using the Standby Shutdown Facility (SSF) and, therefore, both the lower temperature seals (Unit 1) and the high temperature seals (Units 2 and 3) should not fail, the worst case seal failure probabilities (Unit 1: 0.73 and Units 2 & 3: 0.19) are used in this analysis, resulting in differing importance measures.
Importance The risk significance of portions of equipment and cable trains is determined by performing an initiating event assessment using the Standardized Plant Analysis Risk Model for Oconee Units 1, 2, and 3 (ASP PWR D), Revision 3i (Ref. 2) where the transient initiating event frequency is replaced by the seismic frequenc The current probability of the basic events that are assumed failed (TRUE) are used in the analysi This method is outlined in NUREG/CR-6544, Development of a Methodology for Analyzing Precursors to Earthquake-Initiated or Fire-Initiated Accident Sequences, Section 3.7 (Ref. 3). For this analysis, loss of all HPI and CCS pumps, the resulting importance [increase in mean core damage probability (Mean CDP)] is 4.3 x 10-6 for Unit 1 and 1.2 x 10 -6 for Units 2 and The screening threshold is <10-6.
Uncertainty The uncertainty about each mean is (see Figure 4):
Unit 1, 5% bound, 5.1 x 10-7 and 95% bound, 1.3 x 10-5.
Units 2 and 3: 5% bound, 1.4 x 10-7 and 95% bound, 3.6 x 10-6
Dominant sequence The seismic Initiating event tree, Loss of High Pressure Service Water Sequence 20 is the dominant sequenc The events and important component failures in LOHPSW Sequence 20 include:
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Successful Reactor Trip,
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Successful emergency feedwater system,
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Failure to provide high pressure injection cooling,
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Failure of reactor coolant pump seal cooling, and
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Operator successfully recovers high pressure service water, leading to core damag IR 50-255/01-08
SENSITIVE - NOT FOR PUBLIC DISCLOSURE
Results tables
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Table 1 provides the importance values for some dominant sequences.
Table 2a provides the event tree sequence logic for the dominant sequences.
Table 2b defines the nomenclature used in Table 2a.
Table 3 provides the conditional cut sets for the dominant sequence
Table 4 provides the definitions and probabilities for selected events.
Assessment summary SPAR model used in the analysis The Revision 3i SPAR model (Ref. 2) was used for this assessmen For this seismic-induced initiating event analysis none of these initiating events, except LOHPSW and LOOP, are applicable to this analysis with loss of reactor coolant pump seal Therefore, these other frequencies are set to zer The loss of high pressure service water initiating event (IE-LOHPSW) frequency is replaced by the seismic frequency for the plant (see below for details of seismic-Induced analysis considerations). Recovery of HPSW is credite Since a seismic event is also assumed to cause a LOOP, an additional analysis was made with the IE-LOOP frequency replaced by the seismic frequency and with the same assumed failure The results of the LOOP analyses for all three units was negligible and, therefore, no further analysis is included in this repor Modifications to event tree and fault tree models The top initiating event in the Loss of High Pressure Service Water (LOHPSW) was revised to identify it as Seismic-Induced Loss of High Pressure Service Water and to identify the dominant sequence 20 (see Figure 1). HPSW is recovered for this sequence.
The fault tree for the standby shutdown facility (SSF) was added to include the alternate cooling water source to the reactor coolant pump seals (see Figures 2 and 3).
Seismic-induced analysis methodology The seismic-induced analysis is based on NUREG/CR-6544 (Ref. 3), For this analysis all HPI and CCS pumps are assumed failed due to flooding from the seismically ruptured fire water piping in the auxiliary building with no recover IR 50-255/01-08
SENSITIVE - NOT FOR PUBLIC DISCLOSURE
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Initiating Seismic Frequency - The initiating seismic frequency (Fi) was developed from NRC NUREG-1488 (Ref.4) and based on a mean ground acceleration of 0.35g (approximately 300 cm/sec2).
Fi = 3.0 x 10-5/year.
Basic event probability changes Table 4 provides the basic events that were modified to reflect the event condition being analyze The bases for these changes are as follows:
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Failure of Component Cooling MDP Train A (CCS-MDP-FC-A). The value was set to TRUE.
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Failure of Component Cooling MDP Train B (CCS-MDP-FC-B). The value was set to TRUE.
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HPI Train A Fails (HPI-MDP-FC-A). The value was set to TRUE.
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HPI Train B Fails (HPI-MDP-FC-B). The value was set to TRUE.
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HPI Train C Fails (HPI-MDP-FC-C). The value was set to TRUE
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MFW Unavailable (MFW-SYS-TRIP). The value was set to TRUE.
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Operator Fails to Restore MFW (MFW-XHE-ERROR). The value was set to TRUE.
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Loss of High Pressure Service Water (IE-LOHPSW). This value was set to 3.0 x 10-5 (see seismic-induced analysis methodology above).
Model update The Rev 3i SPAR model for Oconee was updated to account for:
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Changes in the probability of failing to recover systems mentioned above for 1 year to account for estimated core uncovery times for conditional assessment sequences.
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Bases for these updates are described in the footnotes to Table 4.
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The Rhodes model was not updated, as the duration for recovery of reactor coolant pump seal cooling was within a two hour period (same for all three Units) and no loss of offsite power occurred in the sequences or cutset However, the probability of RCP seal failure was revised to reflect the higher probability for the Unit i seals and the high temperature seals used in Units 2 & 3 (from the SPAR Model, Section 6.2.3, Ref. 2).
IR 50-255/01-08
SENSITIVE - NOT FOR PUBLIC DISCLOSURE References 1.
Oconee Nuclear Station - NRC Integrated Inspection Report 50-269/00-08, 50-270/00-08, and 50-287/00-08, dated April 30, 2001.
2.
James K. Knudsen and Scott T. Beck, Standardized Plant Analysis Risk Model for Oconee Units 1, 2, and 3 (ASP PWR D), Revision 3i, Idaho National Engineering and Environmental Laboratory, June 2000.
3.
J. Budnitz, et al., Development of a Methodology for Analyzing Precursors to Earthquake-Induced and Fire-Induced Accident Sequences, NUREG/CR-6544, U.S. Nuclear Regulatory Commission, Washington, DC, April 199.
P. Sobel, Revised Livermore Seismic Hazard Estimates for 69 Nuclear Power Plant Sites East of the Rocky Mountains, NUREG-1488, U. S. Nuclear Regulatory Commission, Washington DC, October 199 IR 50-255/01-08
SENSITIVE - NOT FOR PUBLIC DISCLOSURE Table Conditional Probabilities Associated with Highest Probability Sequences (Importance)
Event tree name Sequence Number Conditional core damage probability (CCDP)
UNIT 1 LOHPSW
3.9E-006 Total (all sequences)
4.3E-006 UNITS 2 &3 LOHPSW
1.1E-006 Total (all sequences)
1.2E-006
Table 2 Event Tree Sequence Logic for Dominant Sequence Event tree name Sequence no.
Logic (/ denotes success; see Table 4b for top event names)
LOHPSW
/RT, /EFW, HPI, RCP-SEALS, /REC-HPSW Table 2 Definitions of Fault Trees Listed in Table 2a
/RT Reactor trips successfully
/EPW Successful emergency feedwater HPI High Pressure injection fails RCP-SEALS Reactor coolant pump seal cooling fails
/REC-HPSW Successful recovery of high pressure service water
IR 50-255/01-08
SENSITIVE - NOT FOR PUBLIC DISCLOSURE Table Conditional cut sets for Dominant TRANS Sequence Event Tree: LOHPSW Sequence 20 UNIT ! CDP Percent contribution Minimal cut sets1 1.8E-006 4 RCS-MDP-LK-SEALS-S SSF-MDP-TM-SWP 8.6E-007 22.0 RCS-MDP-LK-SEALS-S SSF-DGN-FR-SSF 6.8E-007 17.1 RCS-MDP-LK-SEALS-S SSF-DGN-TM-SSF 2.8E-007 7.1 RCP-MDP-LK-SEALS-S SSF-DGN-FS-SSF 2.2E-007 5.6 RCS-MDP-LK-SEALS-S SSF-XHE-XA-SSF 3.9E-006 Total2 UNITS 2 & 3 CDP Percent contribution Minimal cut sets1 4.7E-007 4 RCS-MDP-LK-SEALS-S SSF-MDP-TM-SWP 2.2E-007 21.2 RCS-MDP-LK-SEALS-S SSF-DGN-FR-SSF 1.8E-007 16.8 RCS-MDP-LK-SEALS-S SSF-DGN-TM-SSF 7.2E-008 6.9 RCP-MDP-LK-SEALS-S SSF-DGN-FS-SSF 5.7E-008 5.4 RCS-MDP-LK-SEALS-S SSF-XHE-XA-SSF 1.1E-006 Total2 Notes:
1.
See Table 4 for definitions and probabilities for the basic events.
2.
Total CCDP includes all cut sets (including those not shown in this table).
IR 50-255/01-08
SENSITIVE - NOT FOR PUBLIC DISCLOSURE Table Definitions and probabilities for modified and dominant basic events Event name Description Probabili ty/Freque ncy Modifie d
CCS-MDP-FC-A FAILURE OF COMPONENT COOLING TRAIN A TRUE YES1 CCS-MDP-FC-B FAILURE OF COMPONENT COOLING TRAIN B TRUE YES1 HPI-MDP-FC-A HPI TRAIN A FAILS TRUE YES1 HPI-MDP-FC-B HPI TRAIN B FAILS TRUE YES1 HPI-MDP-FC-C HPI TRAIN C FAILS TRUE YES1 MFW-SYS-TRIP MAIN FEEDWATER UNAVAILABLE 8.0E-01 YES1 MFW-XHE-ERROR OPERATOR FAILS TO RECOVER MAIN FEEDWATER FLOW 4.0E-02 YES1 IE-LOHPSW LOSS OF HIGH PRESSURE SERVICE WATER INITIATING EVENT 3.0E-05 YES2 SSF-MDP-TM-SWP SSF SERVICE WATER PUMP UNAVAILABLE DUE TO TEST AND MAINTENANCE 8.3E-02 NO2 SSF-DGN-FR-SSF SSF DIESEL GENERATOR FAILS TO RUN 3.9E-02 NO SSF-DGN-TM-SSF SSF DIESEL GENERATOR UNAVAILABLE DUE TO TEST AND MAINTENANCE 3.1E-02 NO SSF-DGN-FS-SSF SSF DIESEL GENERATOR FAILS TO START 1.3E-02 NO SSF-XHE-XA-SSF OPERATOR FAILS TO INITIATE STANDBY SHUTDOWN FACILITY 1.0E-02 NO RCS-MDP-LK-SEALS RCP SEALS FAIL W/O COOLING AND INJECTION (Unit 1)
7.3E-01 YES3 RCS-MDP-LK-SEALS RCP SEALS FAIL W/O COOLING AND INJECTION (Units 2 & 3)
1.9E-01 YES3 NOTES:
1.
Basic events set to TRUE reflect the failed position for this analysis.
2.
Loss of high pressure service water initiating event revised to reflect the initiating seismic frequency.
3.
RCP seal failure probability from REV. 3 SPAR Model, section 6.2.3, Table 6-PBC PIGGY-BACK RECIRCULATION COOLING HPI HIGH PRESSURE INJECTION REC-HPSW OPERATOR FAILS TO RECOVER HPSW HPI-COOL BLEED PORTION OF HPI COOLIING PORV-RES PORVs CLOSE PORV NO PORVs OPEN M FW MAIN FEEDWATER EFW EMERGEN CY FEEDWATER RCP-SEALS FAILURE OF RCP SEALS RT REACTOR TRIP IE-L OHPSW SEISMIC INDUCED LOSS OF HIGH PRESSURE SERVICE WATER
END -STATE FREQUENC Y
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Figure 1 Seismic-Induced Loss of High Pressure Service Water Event Tree Sequence 20
SENSITIVE - NOT FOR PUBLIC DISCLOSURE IR 50-255/01-08
RCP-SEALS RCS-MDP-LK-SEALS-S RCP-SEAL-COOL SSF-SYSTEM RCP SEAL COOLING FAILS RCP SEALS FAIL W/O COOLING AND INJECTION SEISMIC FAILURE OF RCP SEAL COOLING SSF REACTOR COOLANT MAKEUP SYSTEM FAILS Figure 2 RCP Seal Cooling Fails-SSF Seal Cooling Recovery
SENSITIVE - NOT FOR PUBLIC DISCLOSURE IR 50-255/01-08
SSF-SYSTEM SSF-XHE-XA-SSF SSF-PDP-FC-RCM SSF-EPS SSF-PDP-TM-RCM SSF REACTOR COOLANT MAKEUP SYSTEM FAILS OPERATOR FAILS TO INITIATE STANDBY SHUTDOW N FACILITY SSF REACTOR COOLANT MAKEUP PUMP FAILS SSF RCM PUMP UNAVAILABLE DUE TO T&M STANDBY SHUTDOWN FACILITY POWER Figure 3 SSF Reactor Coolant Makeup Fault Tree
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CDP CDP Figure 4
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