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) | |
Text
)LQDO Precursor Analysis Accident Sequence Precursor Program --- Office of Nuclear Regulatory Research Non-seismic 16-inch fire system piping header transited Oconee Nuclear through the auxiliary building and posed a potential flooding Station, Units 1, 2, & 3 problem should the piping rupture during a seismic event
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$SULO30, 2001 50-269/00-08, 50-270/00-08, Unit 1 = 4.3x10-6 and 50-287/00-08 Units 2 & 3 = 1.2x10-6 January 24, 2003 Condition Summary Description. 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 event. The postulated flood could disable both the high pressure injection (HPI) pumps and the component cooling system (CCS) pumps. 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 flooding. However, the licensee later acknowledged (1976) that the system was always wet.
Duration. Since the piping design was always non-seismic and was water-filled, the duration is considered as the maximum for PRA analysis, one year.
Cause. 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 opportunity. 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 building. Therefore, no recovery for the HPI or CCS pumps is credited. 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 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 frequency. 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 3. 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- 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 sequence. The events and important component failures in LOHPSW Sequence 20 include:
- Successful Reactor Trip,
- Successful emergency feedwater system,
- Failure to provide high pressure injection cooling,
- Failure of reactor coolant pump seal cooling, and
- Operator successfully recovers high pressure service water, leading to core damag SENSITIVE - NOT FOR PUBLIC DISCLOSURE
IR 50-255/01-08
Results tables
- Table 1 provides the importance values for some dominant sequence Table 2a provides the event tree sequence logic for the dominant sequence Table 2b defines the nomenclature used in Table 2 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 assessment. 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 seals. Therefore, these other frequencies are set to zero. 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 credited. 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 failures. 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 sequenc 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 SENSITIVE - NOT FOR PUBLIC DISCLOSURE
IR 50-255/01-08
- 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/yea Basic event probability changes Table 4 provides the basic events that were modified to reflect the event condition being analyzed. The bases for these changes are as follows:
- Failure of Component Cooling MDP Train A (CCS-MDP-FC-A). The value was set to TRU Failure of Component Cooling MDP Train B (CCS-MDP-FC-B). The value was set to TRU HPI Train A Fails (HPI-MDP-FC-A). The value was set to TRU HPI Train B Fails (HPI-MDP-FC-B). The value was set to TRU HPI Train C Fails (HPI-MDP-FC-C). The value was set to TRUE
- MFW Unavailable (MFW-SYS-TRIP). The value was set to TRU Operator Fails to Restore MFW (MFW-XHE-ERROR). The value was set to TRU 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:
- Changes in the probability of failing to recover systems mentioned above for 1 year to account for estimated core uncovery times for conditional assessment sequence Bases for these updates are described in the footnotes to Table 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 cutsets. 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).
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IR 50-255/01-08 References Oconee Nuclear Station - NRC Integrated Inspection Report 50-269/00-08, 50-270/00-08, and 50-287/00-08, dated April 30, 2001. 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. 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 1998. 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 SENSITIVE - NOT FOR PUBLIC DISCLOSURE
IR 50-255/01-08 Table 1. Conditional Probabilities Associated with Highest Probability Sequences (Importance)
Conditional Event tree Sequence core damage name Number probability (CCDP)
UNIT 1 20 3.9E-006 LOHPSW Total (all sequences) 4.3E-006 UNITS 2 &3 20 1.1E-006 LOHPSW Total (all sequences) 1.2E-006 Table 2a. Event Tree Sequence Logic for Dominant Sequence Event tree Sequence Logic name n (/ denotes success; see Table 4b for top event names)
LOHPSW 20 /RT, /EFW, HPI, RCP-SEALS, /REC-HPSW Table 2b. 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
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IR 50-255/01-08 Table Conditional cut sets for Dominant TRANS Sequence Event Tree: LOHPSW Sequence 20 UNIT ! CDP Percent Minimal cut sets1 contribution 1.8E-006 4 RCS-MDP-LK-SEALS-S SSF-MDP-TM-SWP 8.6E-007 2 RCS-MDP-LK-SEALS-S SSF-DGN-FR-SSF 6.8E-007 1 RCS-MDP-LK-SEALS-S SSF-DGN-TM-SSF 2.8E-007 RCP-MDP-LK-SEALS-S SSF-DGN-FS-SSF 2.2E-007 RCS-MDP-LK-SEALS-S SSF-XHE-XA-SSF 3.9E-006 Total2 UNITS 2 & 3 Percent Minimal cut sets1 CDP contribution 4.7E-007 4 RCS-MDP-LK-SEALS-S SSF-MDP-TM-SWP 2.2E-007 2 RCS-MDP-LK-SEALS-S SSF-DGN-FR-SSF 1.8E-007 1 RCS-MDP-LK-SEALS-S SSF-DGN-TM-SSF 7.2E-008 RCP-MDP-LK-SEALS-S SSF-DGN-FS-SSF 5.7E-008 RCS-MDP-LK-SEALS-S SSF-XHE-XA-SSF 1.1E-006 Total2 Notes: See Table 4 for definitions and probabilities for the basic events. Total CCDP includes all cut sets (including those not shown in this table).
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IR 50-255/01-08 Table 4. Definitions and probabilities for modified and dominant basic events Probabili Modifie Event name Description ty/Freque ncy 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 3.0E-05 YES2 EVENT SSF-MDP-TM-SWP SSF SERVICE WATER PUMP UNAVAILABLE DUE TO TEST 8.3E-02 NO2 AND MAINTENANCE 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 3.1E-02 NO AND MAINTENANCE SSF-DGN-FS-SSF SSF DIESEL GENERATOR FAILS TO START 1.3E-02 NO SSF-XHE-XA-SSF OPERATOR FAILS TO INITIATE STANDBY SHUTDOWN 1.0E-02 NO FACILITY RCS-MDP-LK-SEALS RCP SEALS FAIL W/O COOLING AND INJECTION (Unit 7.3E-01 YES3 1)
RCS-MDP-LK-SEALS RCP SEALS FAIL W/O COOLING AND INJECTION 1.9E-01 YES3 (Units 2 & 3)
NOTES: Basic events set to TRUE reflect the failed position for this analysis. Loss of high pressure service water initiating event revised to reflect the initiating seismic frequency. RCP seal failure probability from REV. 3 SPAR Model, section 6.2.3, Table 6- SENSITIVE - NOT FOR PUBLIC DISCLOSURE
SEISMIC INDUCED REACT OR FAILURE OF EMERGEN CY MAIN NO PORVs BLEED OPERATOR HIGH PIGGY-BACK LOSS OF HIGH TRIP RCP SEALS F EEDWATER F EEDWAT ER PORVs CLOSE PORTION FAILS TO PRESSURE RECIRCULATION PRESSURE OPEN OF HPI RECOVER INJECTION COOLING SERVICE WATER COOLIING HPSW IE-L OH PSW RT R CP-SEALS EF W M FW PORV POR V- RES HPI-COOL REC- HPSW HPI PBC # END -STAT E F REQUENC Y 1 OK 2 OK 3 OK 4 CD 5 CD 6 CD 7 OK 8 OK SENSITIVE - NOT FOR PUBLIC DISCLOSURE 9 OK 10 CD 11 CD
12 CD 13 OK 14 CD 15 CD 16 CD 17 CD 18 OK 19 CD 20 CD 21 CD 22 OK 23 CD 24 CD 25 CD 26 OK 27 CD 28 CD 29 CD 30 CD IR 50-255/01-08 31 CD Figure 1 Seismic-Induced Loss of High Pressure Service Water Event Tree Sequence 20
RCP SEAL COOLING FAILS SENSITIVE - NOT FOR PUBLIC DISCLOSURE 10 RCP-SEALS RCP SEALS FAIL W/O FAILURE OF RCP SSF REACTOR COOLING AND SEAL COOLING COOLANT MAKEUP INJECTION SEISMIC SYSTEM FAILS RCS-MDP-LK-SEALS-S RCP-SEAL-COOL SSF-SYSTEM Figure 2 RCP Seal Cooling Fails-SSF Seal Cooling Recovery IR 50-255/01-08
SSF REACTOR SENSITIVE - NOT FOR PUBLIC DISCLOSURE COOLANT MAKEUP SYSTEM FAILS
SSF-SYSTEM STANDBY SHUTDOWN OPERATOR FAILS TO SSF REACTOR COOLANT SSF RCM PUMP FACILITY POWER INITIATE STANDBY MAKEUP PUMP UNAVAILABLE DUE TO SHUTDOW N FACILITY FAILS T&M SSF-EPS SSF-XHE-XA-SSF SSF-PDP-FC-RCM SSF-PDP-TM-RCM IR 50-255/01-08 Figure 3 SSF Reactor Coolant Makeup Fault Tree
SENSITIVE - NOT FOR PUBLIC DISCLOSURE IR 50-255/01-08 CDP Figure 4 12