ML20135G970
| ML20135G970 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/14/2020 |
| From: | Christopher Hunter NRC/RES/DRA/PRB |
| To: | |
| Littlejohn J (301) 415-0428 | |
| References | |
| LER 1996-007-00, LER 1996-008-00 | |
| Download: ML20135G970 (12) | |
Text
Armendix B Anpenix BLER Nos. 373/96-007. -008 B.8 LER Nos. 373/96-007, -008 Event
Description:
Concrete sealant fouls cooling water systems Date of Event: June 28, 1996 Plant: LaSalle 1 and 2 B.8.1 Event Summary A foam sealant was inadvertently injected into the service water tunnel for Units 1 and 2, resulting in fouling of nonessential service water and potential fouling of essential (safety-related) cooling water systems (Fig. B. 8. 1) The conditional core damage probability (CCDP) estimated for this event is 7.0 x 10' at LaSalle
- 2. Because of the location of the sealant and the location of the essential service water (ESW) system pump intakes, the significance of this event would be slightly lower for LaSalle I than for LaSalle 2.
B.8.2 Event Description The licensee event report (LER) for this event`~ reports that between May 21 and June 21, 1996, contractors, as requested by the station's Consolidated Facility Maintenance Group, began sealing cracks in the walls and floors of the lake screen house. The repair process involved drilling holes into the side of each crack along its length and injecting an expandable foam sealant (Furmanite) into these boles to seal the crack. While performing the repairs, workers started fixing cracks on the top of the service water tunnel. This tunnel supplies cooling water to both the ESW and the nonessential service water system (NESW). Because the workers believed that they were working on a concrete floor laid over soil, they proceeded to drill five holes through the ceiling of the service water tunnel. Consequently, instead of injecting sealant into a void under the building floor, the material was injected into the service water tunnel. In all, personnel injected between 2.3 and 3.4 M3 (80 and 120 ft3 ) of sealant into the tunnel.
On June 19, 1996, a high differential pressure developed across the NESW strainers for both units. An automatic back-wash for the strainers for both units failed, and differential pressure across the strainers exceeded the normal backwash setpoint by as much as 0.55 MPa (8 psid). Reactor power was reduced to about 77% on both units to reduce loading of the service water system, and the strainers were isolated one at a time for repairs. After the repairs were completed, operators were able to manually backwash each strainer successfully. Initially, it was believed that the strainers had been fouled by "corn cob" material being used in sandblasting the exterior of the raw water intake building.
On June 24, 1996, normal surveillance tests were attempted on the station's diesel fire water pumps (DFPs).
While DFP OA performed satisfactorily, DFP OB experienced a high cooling water outlet temperature, indicative of flow blockage, after about 5 min of operation. DEP OB was shut down, and both pumps were declared inoperable. Later on the same day, high differential pressures were experienced again across the NESW strainers, which again failed to successfully automatically backwash. Operators reduced plant service water heat loads and manually backwashed the strainers.
NUREG/CR-4674, Vol.25 B.8-I B.8-1 NUREG/CR-4674, Vol. 25
LER Nos. 373/96-007, -008 Apni Appendix B On June 25 and June 26, test runs of the five emergency diesel generators and the four residual heat removal (RHR) systems were made, and no operability problems were noted. Unit I scrammed due to instrument calibration problems late on June 26. On June 27, a Unit 2 service water strainer was inspected and found to be operable. On June 28, divers performed an inspection of the service water tunnel and discovered that a significant amount of debris remained present (this was later determined to be the Furmanite sealant). ESW systems taking suction from the service water tunnel were declared inoperable. Unit I was scrammed firom 1%power, and Unit 2 was reduced from 100% to 5% power and then scramnmed.
B.8.3 Additional Event-Related Information Both units were maintained in hot shutdown so that decay heat removal systems relying upon ESW would not be demanded, and the service water tunnel was extensively cleaned. Between 80 and 120 ft' of sealant material was removed from the tunnel, including one large piece measuring about 8 ft wide by 15 ft long by up to 1 ft thick.
The Augmented Inspection Team report' and other information provided by personnel at Commonwealth Edison'-' indicate that on July 4 and July 5, 1996, ESW pumps were run for several hours for testing and all performed adequately, except for the U 1 A fuel pool emergency makeup (FP EM) pump and the U2 B FPEM pump, which were out of service. On July 5, the 2A residual heat removal system service water (RHRSW) strainer was opened for inspection. A large number of sealant pieces of significant size were found and the strainer was described as being filled 50-60% with debris. Examination of several tubes with a boroscope revealed that all tubes examined were blocked with sealant (not all tubes could be examined using the boroscope). Vendor information indicated that debris particles greater than 2 in. in size would not be removed during strainer backwash because the internal diameter (ID) of the backwash arm was 2 in. Some pieces of sealant debris were observed to have at least one dimension greater than the ID of the backwash
,arm. Only small amounts of debris were found in the balance of the ESW system. On July 7, both units were placed in cold shutdown.
Subsequent analysis determined that, given the layout of the service water tunnel and the sealant injection points, the vulnerability of Unit I ESW systems to fouling was low.' Potentially vulnerable systems taking suction from the Unit 2 end of the service water tunnel include the Unit I and Unit 2 NESW systems, 2A and 2B RHRSW systems, the 2A diesel generator cooling water system, the HPCS diesel generator cooling water pump, 2A and 2B FPEM pumps, and OB DFP.
B.8.4 Modeling Assumptions A probabilistic risk assessment (PRA) of the event prepared by the licensee4 indicates that, based on the judgment of personnel responding to the events, the combined probability that either the June 19 or the June 24 strainer failures could have resulted in a total loss of normal service water and subsequent scram was about 0.5 (IE-TRANS-NESW). This assumrption appears reasonable, and the event was analyzed as an at-power scram with nonsafety-related service water unavailable. Unavailability of the NESW renders dependent systems, either directly or indirectly, unavailable. These systems include the condensate system (CDS), main feedwater (MFW), power conversion systems (PCS), control rod drive hydraulic supply system (CRD), and systems required for containment venting (CVS).
NUREG/CR-4674, Vol. 25 B.8-2
Appendix B LER Nos. 373/96-007, -008 No test data could be identified to show that the potentially affected systems were fully operable after the June 19, 1997, event. Some testing was performed around June 25, 1996, that indicated that ESW was operable after the second fouling event. An exhaustive system test was performed on July 4, 1996, after a cleanup of the sealant material. Results from this testing also indicated that ESW was operable.
The 2A RHRSW system strainer was reported to be 50-60% fouled with sealant debris subsequent to the testing on July 4, 1997. This fouling occurred despite the fact that the system had been tested after the two reported strainer fouling events and much of the original quantity of debris possibly removed from the tunnel (the run times and results could not be clearly determined from the available information','). In addition, the 013 DFP was found by surveillance testing on June 24 to be inoperable, due to sealant fouling of its cooling water supply.
it is difficult to assess the likelihood that additional safety-related, systems dependent on the service water tunnel could have been rendered inoperable by sealant fouling, but it is clear that the sealant material did present a potential concern. Based on actual conditions and the behavior of the sealant material as reported in laboratory testing, Commonwealth Edison calculated a probability of failure of 0,01 for division 2 core standby coolant supply systems dependent on ESW due to sealant fouling. Failure of the ESW trains was incorporated into the ASP model fault trees with events DIV IFOUL, DIV2FOUL, and DIV3FOIJL. The primary impacts on plant safety systems from -fouling would be to fail the service water supply to an RHR heat exchanger or to fail the cooling water supply to the high-pressure core spray system. An independent failure probability of 0.01 was used for each basic event.
In addition, the B FPEM pump was assumed to be failed (SSW-MDP-FC-FPEMU) based on additional information obtained about the event.6 This basic event does not contribute materially to the total CCDP, but has been added for completeness.
B.8.5 Analysis Results The CCDP estimated for this event is 7.0 x 10-6 .The dominant sequence highlighted on the event tree in Fig. B.8.2 involves
- failure of ADS.
Definitions and probabilities for basic events are shown in Table B.8. 1. The conditional probabilities associated with the highest probability sequences are shown in Table B.8.2. Table 1B.8.3 lists the sequence logic associated with the sequences listed in Table B.8.2. Table 13.8.4 describes the system names associated with the dominant sequences. Minimal cut sets associated with the dominant sequences are shown in Table B3.8.5.
Failure of the ESW trains was incorporated into the ASP model fault trees with events DIV IFOUL, DIV2FOUL, and DIV3FOUL. The primary impacts on plant safety systems from fouling would be to fail the service water supply to an RHR heat exchanger or to fail the cooling water supply to the high-pressure B.8-3 B.8-3NUR.EG/CR-4674, Vol. 25
LER Nos. 373/96-007,-008 Annendix Annendix B B LER Nos. 373/96-007. -008 core spray system. An independent failure probability of 0.01 for each basic event provides a CCDP of 7.0 x 10'. If divisions 1 and 3 are assumed to be 10 times less likely to fail due to fouling similar to the Commonwealth Edison PRA (i.e., DIV IFOUL = DIV3FOUL = 0.001, DIV2FOUL = 0.01), then the CCDP is 4.2 x 10.6. If divisions I and 3 are assumed to be 10 times more likely to fail due to fouling (DIV IFOUL
= DIV3FOUL = 0. 1, DIV2FOUL =0.0 1), then the CCDP is 3.3 x 10'~. (Commonwealth Edison calculated a probability of failure of 0.01 for division 2 core standby coolant supply systems dependent on ESW due to sealant fouling; the probability of failure for division 1 and 3 core standby coolant supply systems dependent on ESW due to sealant fouling was 0.001.) IE-TRANS-NESW sequence 31 is the largest contributor to the CCDP for both sensitivity studies.
Because the times and dates of sealant material injections are not known, it appears difficult to determine whether. or not enough material was injected or accumulated at any given time to potentially cause failure of both ESW trains. The LER for this event indicated that the potential for "common mode failure of the Essential Service Water System" existed. However, because licensee's comrment number 4 (sec pg. G.8-3) and the referenced CoinEd report suggest otherwise, the potential for comnmon-cause failure was removed from the ASP modeling of the event. A sensitivity study was performed to determine the potential impact of failing the ESW trains due to a commnon cause. With a common-cause failure value of 0. 1, the CCDP for this event is bounded at 9.5 x 10'. As in all of the previous cases, the dominant sequence is sequence 3 1.
B.8.6 References
- 1. LER No. 373/96-008, "Foreign Material Injected Into Service Water Tunnel Causes Dual Unit Shutdown Due to Inadequate Work Control," November 25, 1996.
- 2. LER No. 373/96-007, "Unit I Reactor Scram on Main Steam Flow High Trip Isolation During Surveillance," July 24, 1996.
- 3. "NRC Region III Augmented Inspection Team Review of the Potential Loss of the Ultimate Heat Sink Due to Foreign Material in the Safety-Related Service Water Intake Tunnel Inspection Report," U.S.
Nuclear Regulatory Commission, August 2, 1996.
- 4. "Probabalistic Risk Assessment Report of the Impact of Foam Sealant Injection in the LaSalle County Nuclear Station Service Water Tunnel," Commonwealth Edison, September 23, 1996.
- 5. Marshall and Rasmuson, Common-Cause Failure Data Collection and Analysis System Volume 6 -
Common-Cause Failure ParameterEstimates, INEL-94/0064, December 1995.
- 6. Teleconference involving personnel from the U.S. Nuclear Regulatory Commission, Oak Ridge National.
Laboratory, and Commonwealth Edison, April 11, 1997.
- 7. Teleconference involving personnel from the U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, and Commonwealth Edison, April 17, 1997.
Vol.25 B.8-4 NUREG/CR-4674, Vol. 25 B.84
Appendix B LER Nos. 373/96-007, -008 Figure removed during SUNSI review.
Fig. B.8.1. Plan of the concrete header for the service water tunnel at LaSalle. (Source: "PRA Report of the Impact of Foam Sealant Injection in the LaSalle County Nuclear Station Service Water Tunnel," Commonwealth Edison, September 23, 1996.)
B.8-5 NUREG/CR-4674, Vol. 25
LER Nos. 373/96-007,-008 Appendix Ap3endix B B
LER Nos. 373/96-007, -008 Fig. B.8.2. Dominant core damage sequence for LER Nos. 373/96-007, -008. (The loss of NESW fails the following systems: CDS, MFW, PCS, CRD, and CVS.)
NUREG/CR-4674, Vol. 25 B.8-6
Annendix B LER Nos. 373/96-007, 008 Table B.8.1. Definitions and Probabilities for Selected Basic Events for LER Nos. 373/96-007, -008 Modified Event Base Current for this name Description probability probability Type event IE-LOOP Initiating Event-Loss of Offsite 1.3 E-005 0.0 E+000 Yes Power JE-SLOCA Initiating Event-Small Loss-of- 4.8 E-007 0:0 E+000 Yes Coolant Accident IE-TRANS-NESW Initiating Event-Loss of NESW 5.0 E-001 5.0 E-001I NEW Yes Induced Transient ADS-SRV-CC-VALVS ADS Valves Fail to Open 3.7 E-003 3.7 E-003 No ADS-XHE-XE-ERROR Operator Fails to Depressurize 1.0 E-002 1.0 E-002 No Using the Automatic Depressurization System ADS-XHE-XE-NOREC Operator Fails to Recover ADS 7.1 E-00 1 7.1 E-001I No DIVIFOUL Failure of Division 1 Safety 1.0 E-002 1.0 E-002 NEW Yes Systems Due to the Loss of ESW DIV2FOUL Failure of Division 2 Safety 1.0 E-002 1.0 E-002 NEW Yes Systems Due to the Loss of ESW DIV3FOUL, Failure of HPCS due to Failing of 1.0 E-002 1.0 E-002 NEW Yes Div. 3 EDG Cooling Water System _______ ___________
HCS-MDP-FC-TRAIN HPCS Train Level Failures 1.3 E-002 1.3 E-002 No HCS-XHE-X(E-NOREC Operator Fails to Recover HPCS 7.0 E-001 7.0 E-001 No PCS-SYS-VF-MISC PCS Hardware Components Fail 1.7 E-001 1.7 E-001 No PPR-SRV-OO-2VLVS One or Two Safety Relief Valves 2.0 E-003 2.0 E-003 No (SRVs) Fail to Close RCI-TDP-FC-TRAIN Reactor Core Isolation Cooling 8.7 E-002 8.7 E-002 No (RCIC) Train Components Failures_____
RCI-XHE-XE-NOREC Operator Fails to Recover RCIC 7.0 E-00 1 7.0 E-00 I No RH-PC-LT Operator Fails to Recover the 2.6 E-003 2.6 E-003 No Residual Heat Removal (RHR)
System and the PCS over a 12-h Period RHR-MDP-CF-MDPS Common-Cause Failure of RHR 1.0 E-004 1.0 E-004 No
__________ I__Pumps I_______ I_______ I____ I____ I NUREG/CR-4674, Vol.25 B.8-7 B.8-7 NUREG/CR-4674, Vol. 25
LER Nos. 373/96-007. -008 Annendixý Anoendix B LER Nos. 373/96-007. -008 Table B.8.1. Definitions and Probabilities for Selected Basic Events for LER Nos. 373/96-007, -008 (Continued)
Modified Event Base Current for this name Description probability probability Type event RHR-MDP-FC-TRNA RHR Train A Components Fail 3.8 E-003 3.8 E-003 No RHR-MDP-FC-TRNB RHR Train B Components Fail 3.8 E-003 3.8 E-003 No SSW-MDP-FC-FPEMU Failure of the Fuel Pool 3.3 E-003 1.0 E-t000 TRUE Yes
_______________Emergency Makeup Pump Train I B.8-8 NUREG/CR-4674, Vol.25 Vol. 25 B.8-8
Anoendix B Anoendix B LER Nos. 373/96-007.-008 LER Nos. 373/96-007. -008 Table B.8.2. Sequence Conditional Probabilities for LER Nos. 373/96-007, -008 Conditional core Event tree Sequence damage Percent name number probability (CCDP) contribution TRANS-NESW 31 6.2 E-006 88.7 TRANS-NESW 07 3.7 E-007 5.3 TRANS-NESW 62 2.3 E-007 3.3 Total (all sequences) 7.0 E-006 Table B.8.3. Sequence Logic for Dominant Sequences for LER Nos. 373/96-007, -008 Event tree name SequenceLoi numberLoi TRANS-NESW 31 /RPS, PCS, /SRV, MFW, HPCS,
____________RCIC, ADS, CRD TRANS-NESW 07 /RPS, PCS, /SRV, MFW,
____ _______ _ __ ___ ____ _ ICPS, RHR., CVS TRANS-NESW 62 /RPS, PCS, P2, HPCS, ADS B.8-9 B.8-9NUREG/CR-4674, Vol. 25
LER Nos. 373/96-007. -008 Avvendix Appendix B B
LER Nos. 373/96-007. -008 Table B.8.4. System Names for LER Nos. 373/06-007, -008 System name Logic ADS Automatic Depressurization Fails, CRD Insufficient CRD Flow to the RCS CVS Containment (Suppression Pool) Venting Fails HPCS HPCS Fails to Provide Sufficient Flow to the Reactor Vessel MFW Main Feedwater System Fails P2 One or Two SRVs Fail to Close PCs Power Conversion System Fails RCIC RCIC Fails to Provide Sufficient Flow to the RCS RHR Residual Heat Removal Fails RPS Reactor Shutdown Fails SRV None of the SRVs Fail to Close B.8-10 NUREG/CR-4674, Vol.25 NUREG/CR-4674, Vol. 25 B.8-1 0
Annendix R Nos. 373/96-007, -008 Annenix BLER Table B.8.5. Conditional Cut Sets for Higher Probability Sequences for LER Nos. 373/96-007, -008 Cut set number Jcontribution Percent TRANS-NESW Sequence 31 CCDPa 6.2 E-006 .:.:..
Cut setsb 1 44.4 2.7 E-006 IE-TRANS-NESW, /SRV, HCS-MDP-FC-TRAIN, HCS-XHE-XE-NOREC, RCf-TDP-FC-TRAIN, RCI-XHE-XE-NQREC, ADS-XHE-XE-ERROR 2 34.2 2.1 E-006 [E-TRANS-NESW. /SRV. DIV3FOUL, HCS-XHE-XE-NOREC, RCI-TDP-FC-TRAIN, RCI-XHE-XE-NOREC.
ADS-XHE-XE-ERROR 3 11.6 7.2 E-007 IE-TRANS-NESW, /SRV, HCS-MDP-FC-TRAIN, HCS-XHE-XE-NOREC. RCI-TDP-FC-TRAIN.
RCJ-XHE-XE-NOREC. ADS-SRV-CC-VALVS.
ADS-XHE-XE-NOREC 4 8.9 5.5 E-007 IE-TRANS-NESW, /SRV, DIV3FOUL, HCS-XHE-XE-NOREC, RCI-TDP-FC-TRAIN, RCI-XHE-XE-NOREC, ADS-SRV-CC-VALVS. ADS-XHE-XE-NOREC TRANS-NESW Sequence 07 3.7 E-007 1 34.3 1.3 E-007 IE-TRANS-NESW, /SRV, RHR-MDP-CF-MIJPS. RH-PC-LT 2 34.3 1.3 E-007 IE-TRANS-NESW, /SRV. DIVIFOUL, DIV2FOUL, RH-PC-LT 3 13.0 4.9 E-008 IE-TRANS-NESW, /SRV, DIV IFOUL, RHR-MDP-FC-TRNB, RH-PG-LT 4 13.0 4.9 E-008 IE-TRANS-NESW, /SRV, RHR-MDP-FC-TRNA. DIV2FOUL.
RH-PC-LT 5 4.9 1.8 E-008 IE-TRANS-NESW, /SRV, RHR-MDP-FC-TRNA.
RHR-MDP-FC-TRNB. RI--PC-LT TRANS-NESW Sequence 62 2.3 E-007............
1 38.2 9.1 E-008 IE-TRANS-NESW, PPR-SRV-OO-2VLVS, HCS-MDP-FC-TRAIN. HCS-XHE-XE-NOREC, ADS-XHE-XE-ERROR 2 29.4 7.0 E-008 IE-TRANS-NESW, PPR-SRV-OO-2VLVS, DIV3FOUL, HCS-XHE-XE-NOREC, ADS-XHE-XE-ERROR 3 10.0 2.3 E-008 IE-TRANS-NESW, PPR-SRV-OO-2VLVS, H-CS-MDP-FC-TRAJN, HCS-XHE-XE-NOREC,
__________________ _________ADS-SRV-CC-VALVS, ADS-XHE-XE-NOREC B.8-11 R.8-11NUREG/CR-4674, Vol. 25
LER Nos. 373/96-007,-008 ADDendix B AnDendix B LER Nos. 373/96-007. -008 Table B.8.5. Conditional Cut Sets for Higher Probability Sequences for LER Nos. 373/96-007, -008 (Continued)
Cut set Percent number contribution CCDpa Cut setsb 4 7.7 1.8 E-008 IE-TRANS-NESW, PPR-SRV-OO-2VLVS, DIV3FOUL, HCS-XHE-XE-NOREC, ADS-SRV-CC-VALVS, ADS-XHE-XE-NOREC 5 6.5 1.5 E-008 PCS-SYS-VF-MISC, PPR-SRV-OO-2VLVS, HCS-MDP-FC-TRAIN, HCS-XHE-XE-NOREC, ADS-XHE-XE-ERROR 6 5.0 1.1 E-008 PCS-SYS-VF-MISC. PPR-SRV-OO-2VLVS. DIV3FOUL.
HCS-XHE-XE-NOREC, ADS-XHE-XE-ERROR 7 1.7 4.0 E-009 PCS-SYS-VF-MISC, PPR-SRV-OO-2VLVS, HCS-MDP-FC-TRAIN. HCS-XHE-XE-NOREC.
ADS-SRV-CC-VALVS, ADS-XHE-XE-NOREC 8 1.3 3.1 E-009 PCS-SYS-VF-MISC, PPR-SRV-OO-2VLVS, DIV3FOUL, HCS-XHE-XE-NOREC, ADS-SRV-CC-VALVS, ADS-XHE-XE-NOREC Total (all sequences) 1 7.0 E-006
'The conditional probability for each cut set is determined by multiplying the probability of the initiating event by the probabilities of the basic events in that minimal cut set. The probabilities for the initiating events and the basic events are given in Table B.8. 1.
blnitiating events, such as IE-TRANS-NESW, are not normally included in the output of the fault tree reduction process but has been added to aid in understanding the sequences to potential core damage associated with the event.
Unavailability of the NESW initiates the transient and renders numerous systems that are dependent on service water unavailable, including the CDS, WFW, PCS (delayed), CRD, and CVS.
25 B.8-12 NUREG/CR-4674, Vol. Vol. 25 B.8-12