ML20294A552
| ML20294A552 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 12/09/2020 |
| From: | Christopher Hunter NRC/RES/DRA/PRB |
| To: | |
| References | |
| LER 325-2020-003 | |
| Download: ML20294A552 (6) | |
Text
1 Final ASP Analysis - Precursor Accident Sequence Precursor Program - Office of Nuclear Regulatory Research Brunswick Steam Electric Plant, Unit 1 Loss of Offsite Power during Hurricane Isaias Event Date: 8/3/2020 LER:
325-2020-003 CCDP =
2x10-5 IR:
TBD Plant Type:
General Electric Type 4 Boiling-Water Reactor (BWR) with a Mark I Containment Plant Operating Mode (Reactor Power Level):
Mode 1 (19% Reactor Power)
Analyst:
Reviewer:
Completion Date:
Christopher Hunter Mehdi Reisi Fard 11/12/2020 1 EXECUTIVE
SUMMARY
On August 3, 2020, a loss of offsite power (LOOP) occurred on Brunswick Steam Electric Plant Unit 1 that resulted in a reactor scram. Emergency diesel generators (EDGs) 1 and 2 automatically started and loaded to their respective emergency buses. Operators manually started reactor core isolation cooling (RCIC) and high-pressure coolant injection (HPCI) to maintain reactor water level and provide pressure control, respectively. An Unusual Event was declared at 11:12 p.m. At the time of the event, Unit 1 was in the process of shutting down for maintenance associated with a ground on the main generator and was not synced to the grid.
As a result of the reactor trip, reactor water level reached low level 1, which results in an isolation signal of the Group 2 (i.e., floor and equipment drains), Group 6 (i.e., monitoring and sampling) and Group 8 (i.e., shutdown cooling) valves. The Group 8 valves were closed at the time of the event. Per design, the LOOP also caused a Group 1 (i.e., main steam isolation valve) isolation. The Unit 1 LOOP did not affect Unit 2, which remained at 100-percent power.
Operators restored offsite power to the Unit 1 nonsafety-related buses from the UAT in approximately 37 minutes. Offsite power was restored to the two safety buses via the station auxiliary transformer (SAT) and the Unusual Event was exited at 2:54 p.m. on August 4th.
This accident sequence precursor (ASP) analysis reveals that the most likely core damage sequence is a loss of offsite power (LOOP) initiating event and the successful operation of the EDGs providing safety-related alternating current (AC) power with subsequent (postulated) failure of both HPCI and RCIC and operators failing to depressurize the reactor. This accident sequence accounts for approximately 72 percent of the total conditional core damage probability (CCDP) for this event. FLEX mitigation strategies were credited in this analysis; however, the risk impact was minimal because postulated station blackout (SBO) scenarios are not a dominant risk contributor. The ASP analysis results reinforce that a LOOP resulting from natural phenomenon outside the licensees control is a substantial risk contributor; however, plant and operator response were timely and appropriate to mitigate the risk of the event.
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2 EVENT DETAILS 2.1 Event Description On August 3, 2020, a LOOP occurred on Brunswick Steam Electric Plant Unit 1 that resulted in a reactor scram. EDGs 1 and 2 automatically started and loaded to their respective emergency buses. Operators manually started RCIC and HPCI to maintain reactor water level and provide pressure control, respectively. An Unusual Event was declared at 11:12 p.m. At the time of the event, Unit 1 was in the process of shutting down for maintenance associated with a ground on the main generator and was not synced to the grid.
As a result of the reactor trip, reactor water level reached low level 1, which results in an isolation signal of the Group 2 (i.e., floor and equipment drains), Group 6 (i.e., monitoring and sampling) and Group 8 (i.e., shutdown cooling) valves. The Group 8 valves were closed at the time of the event. Per design, the LOOP also caused a Group 1 (i.e., main steam isolation valve) isolation. The Unit 1 LOOP did not affect Unit 2, which remained at 100-percent power.1 Operators restored offsite power to the Unit 1 nonsafety-related buses from the UAT in approximately 37 minutes. Offsite power was restored to the two safety buses via the SAT and the Unusual Event was exited at 2:54 pm. on August 4th. Additional information is provided in licensee event report (LER) 325-2020-002, Automatic Specified System Actuations due to Loss of Offsite Power, (ADAMS Accession No. ML20265A162).
2.2 Cause The licensee determined that the electrical fault that resulted in the transformer bus powering the SAT to trip and the subsequent LOOP was caused storm-generated debris from Hurricane Isaias.
3 MODELING 3.1 SDP Results/Basis for ASP Analysis The ASP Program performs independent analyses for initiating events. ASP analyses of initiating events account for all failures/degraded conditions and unavailabilities (e.g., equipment out for maintenance) that occurred during the event, regardless of licensee performance.2 Additional LERs were reviewed to determine if concurrent unavailabilities existed during the August 3rd event. No windowed events were identified. Discussions with Region 2 staff indicate that no licensee performance deficiency associated with this event has been identified; however, the LER remains open.
3.2 Analysis Type An initiating event analysis was performed using Revision 9.33 of the standardized plant analysis risk (SPAR) model for Brunswick Steam Electric Plant (Unit 1) created on July 31, 2020. This event was modeled as a switchyard-centered LOOP initiating event.
1 The Unit 2 EDGs started but did not load onto their respective safety buses because they remained energized via offsite power throughout the event.
2 ASP analyses also account for any degraded condition(s) that were identified after the initiating event occurred if the failure/degradation exposure time(s) overlapped the initiating event date.
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3.3 SPAR Model Modifications No SPAR model modifications were needed for this analysis.
3.4 Analysis Assumptions The following modeling assumptions were determined to be significant to the modeling of this initiating event assessment:
The probability of IE-LOOPSC (loss of offsite power (switchyard-centered)) was set to 1.0 due to the loss of offsite power. All other initiating event probabilities were set to zero.
Basic event OEP-VCF-LP-SITESC (site LOOP (switchyard-centered)) was set to FALSE because the event was single unit LOOP.
The probability of basic event FLX-XHE-XM-ELAP (operators fail to declare ELAP when beneficial) was set to it nominal value of 1x10-2 to activate the credit for FLEX mitigation strategies for SBO scenarios for which an extended loss of AC power (ELAP) is declared. Sensitivity analyses show that the amount of FLEX credit has a minimal impact on the analysis results.
Offsite power was restored to the two safety buses via the SAT in approximately 15 and 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />, respectively. It is believed that operators could have restored offsite power more quickly if postulated SBO had occurred. However, an exact determination on when the offsite power could have been restored is not available. Given this uncertainty, basic events OEP-XHE-XL-NR30MSC (operators fail to recover offsite power in 30 minutes) and OEP-XHE-XL-NR02HSC (operators fail to recover offsite power in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) were set to TRUE. Although this assumption is potentially conservative, sensitivity analyses show credit for the offsite power recovery has a minimal impact on the analysis results.
4 ANALYSIS RESULTS 4.1 Results The mean CCDP for this analysis is calculated to be 2.0x10-5. The ASP Program threshold for initiating events is a CCDP of 10-6 or the plant-specific CCDP of an uncomplicated reactor trip with a non-recoverable loss of feed water or the condenser heat sink, whichever is greater. This CCDP equivalent for Brunswick Steam Electric Plant (Unit 1) is 1.5x10-5. Therefore, this event is a precursor.
The parameter uncertainty results for this analysis provided below:
Table 1. Parameter Uncertainty Results 5%
Median Point Estimate Mean 95%
Sample Size Method 1.65E-6 9.68E-6 1.69E-5 1.97E-5 7.17E-5 5000 Monte Carlo
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Figure 1. CCDP Uncertainty Distribution 4.2 Dominant Sequences3 The dominant accident sequence is a switchyard LOOP sequence 25 (CDP = 1.2x10-5), which contributes approximately 72 percent of the total CCDP. The sequences that contribute at least 5.0 percent to the total CCDP are provided in the following table. The event tree with the dominant sequence is shown graphically in Figure A-1 of Appendix A.
Table 2. Dominant Sequences Sequence CDP Description LOOPSC 25 1.22x10-5 72.3% Switchyard LOOP initiating event occurs; successful reactor trip; EDGs successfully provide power to at least one 4-kV safety-related bus; all high-pressure injection sources fail; and reactor depressurization fails resulting in core damage LOOPSC 6 1.46x10-6 8.7%
Switchyard LOOP initiating event occurs; successful reactor trip; EDGs successfully provide power to at least one 4-kV safety-related bus; at least one high-pressure injection source is successful; suppression pool cooling fails; reactor depressurization succeeds; low-pressure injection is successful; containment spray fails; and containment venting fails resulting in core damage LOOPSC 28 9.88x10-7 5.9%
Switchyard LOOP initiating event occurs; successful reactor trip; the SRVs fail to open resulting in core damage 3
The CCDPs in this section are point estimates.
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4.3 Key Uncertainties A review of the analysis assumptions and results did not reveal key modeling uncertainties.
Because of the minimal risk impact of postulated SBO scenarios, the typical modeling uncertainties (e.g., hardware and human reliability of FLEX mitigation strategies, stuck-open SRV probabilities and modeling) were not significant to this analysis.
LER 325-2020-003 A-1 Appendix A: Key Event Tree Figure A-1. Brunswick (Unit 1) Switchyard LOOP Event Tree IE-LOOPSC LOSS OF OFFSITE POWER INITIATOR (SWITCHYARD-CENTERED)
RPS REACTOR SHUTDOWN EPS FS = FTF-SBO EMERGENCY POWER SRV-O PRESSURE RELIEF SRV TWO OR MORE STUCK OPEN SRVS HPI HIGH PRESSURE INJECTION SPC SUPPRESSION POOL COOLING DEP MANUAL REACTOR DEPRESS LPI LOW PRESS COOLANT INJECTION (LCS OR LPCI)
VA ALTERNATE LOW PRESS INJECTION SPC SUPPRESSION POOL COOLING CSS CONTAINMENT SPRAY CVS CONTAINMENT VENTING LI LONG-TERM LOW PRESS INJECTION End State (Phase - CD) 1 OK 2
OK 3
OK 4
CD 5
OK LI04 6
CD 7
OK CS1 8
OK 9
OK LI04 10 CD 11 CD 12 CD 13 OK 14 OK 15 OK 16 CD 17 OK LI04 18 CD 19 OK SP1 20 OK CS1 21 OK 22 OK LI04 23 CD 24 CD 25 CD P1 26 LOOP-P1 P2 27 LOOP-P2 28 CD 29 SBO 30 ATWS 31 CD