ML24289A119
| ML24289A119 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 10/24/2024 |
| From: | Christopher Hunter NRC/RES/DRA/PRB |
| To: | |
| References | |
| Download: ML24289A119 (7) | |
Text
1 Final ASP Analysis - Reject Accident Sequence Precursor Program - Office of Nuclear Regulatory Research South Texas Project, Unit 2 Automatic Reactor Trip and Actuation of Two of Three EDGs Event Date: 5/12/2024 LER:
499-2024-001 CCDP =
8x10-7 IR:
TBD Plant Type:
Westinghouse Four-Loop Pressurized-Water Reactor (PWR) with Dry Ambient Pressure Containment Plant Operating Mode (Reactor Power Level):
Mode 1 (15% Reactor Power)
Analyst:
Reviewer:
Completion Date:
Christopher Hunter Matthew Humberstone 10/24/2024 1
EVENT DETAILS 1.1 Event Description On May 12, while returning to power after a refueling outage, an automatic reactor trip occurred due to a unit auxiliary transformer (UAT) lockout. During the trip, all control rods fully inserted, and all three engineered safety feature (ESF) busses were energized by emergency diesel generators (EDGs) 21 and 23 (trains A and C) and standby transformer 2 (train B). All equipment responded as expected except for steam generator (SG) power-operated relief valve (PORV) 2C, which fully opened when the manual control was depressed slightly. In addition, the load center E2A1 supply breaker failed to close automatically following the partial loss of offsite power (LOOP) and automatic ESF sequencing.
When load center E2A1 breaker failed to close, the capability to provide auxiliary feedwater (AFW) to SG 2A was lost and SG PORV 2A remained in the fully closed position. In addition, the open breaker resulted in a loss of the supply fan 21A, which provides ventilation cooling flow for EDG 21. So, although EDG 21 operated successfully for the 26 minutes, it may have not been able to fulfil its safety function for the complete PRA mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (if needed).
PORV 2C was declared inoperable due to erratic control when placed in Manual. The PORV would not modulate to control pressure but would only operate in the fully open or closed position. Additional information is provided in licensee event report (LER) 499-2024-001, Automatic Reactor Trip and Actuation of Two of Three EDGs, (ML24184C083).
1.2 Cause The causes of this event have not been identified yet; the licensee causal analysis is ongoing.
The results of the causal analysis will be provided in a future supplement to the LER.
1.3 Sequence of Key Events Table 1 provides the sequence of key events:
LER 499-2024-001 2
Table 1. Sequence of Key Events May 12, 2024 4:41 pm Switchyard breakers Y590 and Y600 tripped. Unit 2 reactor tripped and EDGs 21 and 23 automatically actuated and sequenced on the LOOP to their associated ESF buses. The following busses were deenergized: 13.8 kilo-volt (kV) auxiliary busses 2F, 2G, 2H, and 2J, 13.8 kV standby busses 2F and 2H, and load center 2W. All four reactor coolant pumps (RCP) lost power. Load center E2A1 supply breaker failed to close automatically following LOOP and automatic ESF sequencing resulting in the deenergization of class 1E 480 V load center E2A1.
4:45 pm Isolated main steam using safety grade switches from the main control room (MCR).
5:05 pm SG PORV 2C placed in manual for venting; however, the PORV opened to full travel without required input from operator. SG PORV 2C was declared inoperable.
5:07 pm Load center E2A1 supply breaker automatically closed after placing it in "pull to lock" and returning the hand switch to automatic. Class 1E 480 V load center E2A1 is energized.
5:29 pm Energized standby bus 2F from standby transformer 2 (i.e., restoration of offsite power to ESF bus A).
5:33 pm Energized standby bus 2H from standby transformer 1 (i.e., restoration of offsite power to ESF bus C).
5:45 pm Energized auxiliary bus 2F from standby transformer 2.
5:47 pm Energized auxiliary bus 2H from standby transformer 1.
5:50 pm Auxiliary bus 2J reenergized from offsite power.
5:53 pm Auxiliary bus 2G reenergized from offsite power.
5:58 pm Started RCP 2D.
7:17 pm Started RCP 2A.
7:18 pm Load center 2D1 energized from offsite power.
7:21 pm Load center 2D2 energized from offsite power.
1.4 Additional Information The following event details are provided as additional information that was not explicitly accounted for in this analysis:
During the event, RCP 2B developed a high seal leak-off flow rate, indicating a degraded seal. However, the integrity of the pump seal was maintained as a result of the other intact seal stages. The licensee determined it was necessary to continue the cooldown to cold shutdown conditions to replace the seal. Sensitivity analysis conservatively assuming the failure of all second stage seals for all four RCPs results in a negligible increase in the mean conditional core damage probability (CCDP) for this event. The risk impact of potentially failed seal stage is limited due to the shutdown seals upgrade that have been installed on all RCPs.
2 MODELING 2.1 SDP Results/Basis for ASP Analysis The Accident Sequencer Precursor (ASP) Program performs independent analyses for initiating events. ASP analyses of initiating events account for all failures/degraded conditions and
LER 499-2024-001 3
unavailabilities (e.g., equipment out for maintenance) that occurred during the event, regardless of licensee performance.1 In response to this event, the NRC is schedule to perform a special inspection per Management Directive 8.3, NRC Incident Investigation Program (ML18073A200). The LER remains open.
No windowed events were identified.
2.2 Analysis Type An initiating event analysis was performed using a test and limited use revision of the version 8.80 SPAR model for South Texas Project, Unit 1 created on August 29, 2024. Note that this Unit 1 model is being used for a Unit 2 analysis. The following SPAR model changes were made to support this analysis:
Credit for other steam pathways (in addition to the SG PORVs) for decay heat removal (e.g., steam dumps and SG safety valves) were provided.
The pressurizer PORV success criteria for feed and bleed cooling was changed to 1 out of 2 valves based on a review of thermal-hydraulic calculations for South Texas Project.
Event tree and fault trees associated with crediting of the FLEX mitigation strategy were revised based on a review of the South Texas Project final integrated plan. In addition, FLEX credit was activated to support this analysis.2 The reactor coolant system (RCS) pressure relief success criteria given an anticipated transient without scram (ATWS) were modified based on a review of thermal-hydraulic calculations for South Texas Project. Specifically, either both pressurizer PORVs or safety valves are required to operate to ensure that RCS pressure does not exceed 3200 psi.
House events were added the applicable alternating current (AC) power fault tree logic to allow the modeling of the partial LOOP.
Corrections were made to the auxiliary feedwater (AFW) electrical room cooling fault tree.
2.3 SPAR Model Modifications To account for the failure of the load center E2A1 supply breaker to automatically close during ESF sequencing, a new AND gate labeled as ACP-LCCE1A-2 (feeder breaker from bus E1A fails to close) was inserted under the top gate of ACP-LCCE1A1 (480 VAC LCC bus E1A1) fault tree. Under this new AND gate, two new basic events were inserted. Basic events ACP-CRB-OO-LCCE1A (feeder breaker from bus E1A fails to close) and ACP-XHE-XM-E1A1RECOVERY (operators fail to close feeder breaker to load center E1A1) represent the breaker failure to automatically close and the potential for operators to manually close the breaker, respectively.
The modified ACP-LCCE1A1 fault tree is provided in Figure B-1 of Appendix B.
1 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.
2 Credit for the FLEX mitigation strategies is deactivated in the base SPAR models.
LER 499-2024-001 4
2.4 Analysis Assumptions The following modeling assumptions were determined to be significant for this analysis:
The probability of IE-LOCHS (loss of condenser heat sink) was set to 1.0 due to loss of condenser heat sink event initiating event caused by the partial LOOP. All other initiating event probabilities were set to zero.
Basic events HE-LOOP-1A (loss of offsite power to bus E1A) and HE-LOOP-1C (loss of offsite power to bus E1C) were set to TRUE because offsite power was lost to ESF buses A and C.
Basic event MSS-ARV-CC-7431 (failure of SG 1 ARV-7431) was set to TRUE due to the failure of SG PORV C being unable to operate properly in manual. Note that this is conservative modeling assumption since the PORV was still available to operate in automatic mode to support decay heat removal.
Basic event ACP-CRB-OO-LCCE1A was set to TRUE due to the failure of feed breaker to the load center E2A1 to close.
Basic event ACP-XHE-XM-E1A1RECOVERY was set to 0.1. NUREG-1792, Good Practices for Implementing Human Reliability Analysis, (ML051160213) states that 0.1 is an appropriate screening (i.e., typically conservative) value for most post-initiator human failure events. Sensitivity analyses show that this human error probability would need to higher than 0.5 for the mean CCDP to exceed the precursor threshold.
3 ANALYSIS RESULTS 3.1 Results The mean CCDP for this analysis is calculated to be 8.2x10-7. 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 and the condenser heat sink, whichever is greater.
This CCDP equivalent for South Texas Project is 7.8x10-7. Therefore, this event is not a precursor. The parameter uncertainty CCDP results are provided in the table below:
Table 2. Parameter Uncertainty Results 5%
Median Point Estimate Mean 95%
1.2x10-7 4.9x10-7 8.0x10-7 8.2x10-7 2.6x10-6 3.2 Dominant Sequences3 The dominant accident sequence is loss of condenser heat sink sequence 20 (CCDP =
5.2x10-7), which contributes approximately 65 percent of the total CCDP. The sequences that contribute at least 10 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.
3 The CCDPs presented in this section are point estimates.
LER 499-2024-001 5
Table 3. Dominant Sequences Sequence CCDP Description LOCHS 20 5.2x10-7 65.3%
Loss of condenser heat sink initiating event occurs; the reactor successfully trips; offsite power continues to supply ESF bus B; AFW fails to provide decay heat removal; and the failure of feed and bleed cooling results in core damage.
LOCHS 22-20 1.1x10-7 13.1%
Loss of condenser heat sink initiating event occurs; the reactor fails to trip resulting in an anticipated transient without scram; offsite power continues to supply ESF bus B; and RCS pressure relief is successful; AFW successfully supplies decay heat removal; and operators fail to initiate emergency boration, which is assumed to result in core damage.
3.3 Key Uncertainties A review of the analysis assumptions and results did not reveal key modeling uncertainties.
LER 499-2024-001 A-1 Appendix A: Key Event Tree(s)
Figure A-1. Loss of Condenser Heat Sink Event Tree IE-LOCHS LOSS OF CONDENSER HEAT SINK RPS REACTOR TRIP OEP OFFSITE ELECTRICAL POWER AFW AUXILIARY FEEDWATER SG-PORVS STEAM GENERATOR PORVS PORV PORVS ARE CLOSED LOSC RCP SEAL COOLING MAINTAINED HPI HIGH PRESSURE INJECTION FAB FEED AND BLEED SSCR SECONDARY COOLING RECOVERED SSC SECONDARY SIDE COOLDOWN RHR RESIDUAL HEAT REMOVAL CFC CONTAINMENT FAN COOLERS HPR HIGH PRESSURE RECIRC End State (Phase - CD) 1 OK 2
LOSC 3
OK 4
OK 5
CD 6
CD 7
OK 8
CD 9
CD 10 CD 11 OK 12 OK 13 CD 14 CD 15 CD 16 OK 17 OK 18 CD 19 CD 20 CD 21 LOOPPC 22 ATWS 23 CD
LER 499-2024-001 B-1 Appendix B: Modified Fault Tree(s)
ACP-LCCE1A1 STEX 1 & 2 480 LCC VAC BUS E1A1 FAULT TREE ACP-LCCE1A1-1 LOSS OF ROOM COOLING NHV External NORMAL HVAC ACP-XHE-XM-SWGRMA True OPERATOR FAILS TO ESTABLISH ROOM COOLING W/O HVAC ACP-E1A External FAILURE OF POWER ON 4160 VAC BUS E1A ACP-CRB-CO-1E1A1 2.50E-06 CRB PKCB1E1A-12 (4160V TO TRANSFORMER E1A1) TRANSFERS OPEN ACP-CRB-CO-E1A12E 4.15E-06 CRB PLCB1E1A1-2E (XFMR E1A1 TO LCC E1A1) TRANSFERS OPEN ACP-BAC-LP-LCE1A1 1.41E-05 480 LCC VAC BUS E1A1 FAILS ACP-TFM-FC-E1A1 4.63E-05 4.16KV/480V TRANSFORMER TR-E1A1 FAILS ACP-LCCE1A-2 FEEDER BREAKER FROM BUS E1A FAILS TO CLOSE ACP-CRB-OO-LCCE1A 2.12E-06 FEEDER BREAKER FROM BUS E1A FAILS TO CLOSE ACP-XHE-XM-E1A1RECOVERY 1.00E+00 OPERATORS FAIL TO CLOSE FEEDER BREAKER TO LOAD CENTER E1A1