ML20128M238
| ML20128M238 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 10/10/1996 |
| From: | Stone J NRC (Affiliation Not Assigned) |
| To: | Carns N WOLF CREEK NUCLEAR OPERATING CORP. |
| References | |
| NUDOCS 9610160032 | |
| Download: ML20128M238 (22) | |
Text
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e October 10, 1996 Mr. Neil S. Carns President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, Kansas 66839
SUBJECT:
FINAL ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT WOLF CREEK GENERATING STATION
Dear Mr. Carns:
Enclosed for your information is a copy of the final Accident Sequence Precursor (ASP) analysis of the operational event at Wolf Creek Generating Station, reported in Licensee Event Report (LER) Nos. 482/96-001 and -002.
This final analysis (Enclosure 1) was prepared by our contractor at the Oak Ridge National Laboratory (ORNL), based on our review and evaluation of your comments on the preliminary analysis and comments received from the NRC staff and from our independent contractor, Sandia National Laboratories (SNL). contains our responses to your specific comments. Our review of your comments employed the criteria contained in the material which accompanied the preliminary analysis. The results of the final analysis indicate that this event is a precursor for 1996.
Please contact me at (301) 415-3063 if you have any questions regarding the enclosures. We recognize and appreciate the effort expended by you and your staff in reviewing and providing comments on the preliminary analysis.
Sincerely' b-N!!e k Nm![
9k$
./W James C. Steve, Senior Project Manager 0
Project Direttorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-482 DISTRIBUTION:
Docket JDyer, Region IV
Enclosures:
1.
Final Report PUBLIC P0'Reilly, AE00 2.
Comment Resolution PDIV-2 Reading SMays, AE00 JRoe cc w/encls:
See next page EAdensam WBateman ACRS OGC JStone EPeyton
\\
Dogument Name:
FINALASP.WC M
0FC PDIV-2 PDIV-2 g
NAME JSNAe/dh EOylk
\\
DATE 10/A/96 10/Cf /96 0;FICIAL RECORD COPY 60080 NRCflg CENTER W 9610160032 961010 PDR ADOCK 05000482 S
- ~
.4 October 10, 1996 Mr. Neil S. Carns President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, Kansas 66839
SUBJECT:
FINAL ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF EVENT AT WOLF CREEK GENERATING STATION
Dear Mr. Carns:
Enclosed for your information is a copy of the final Accident Sequence Precursor (ASP) analysis of the operational event at Wolf Creek Generating Station, reported in Licensee Event Report (LER) Nos. 482/96-001 and -002.
This final analysis (Enclosure 1) was prepared by our contractor at the Oak Ridge National Laboratory (0RNL), based on our review and evaluatinn of your comments on the preliminary analysis and comments received from the NRC staff and from our independent contractor, Sandia National Laboratories (SNL).
i contains our responses to your specific comments.
Our review of your comments employed the criteria contained in the material which accompanied the preliminary analysis.
The results of the final analysis indicate that this event is a precersor for 1996.
I Please contact me at (301) 415-3063 if you have any questions regarding the enclosures. We recognize and appreciate the effort expended by you and your
]
staff in reviewing and providing comments on the preliminary analysis.
Sincerely' b!Nise k.990Nmb i
.fW James C. Stone, Senior Project Manager O
Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-482 DISTRIBUTION:
Docket JDyer, Region IV i
Enclosures:
1.
Final Report PUBLIC PO'Reilly, AE00 2.
Comment Resolution PDIV-2 Reading SMays, AEOD JRoe cc w/encls:
See next page EAdensam WBateman ACRS OGC JStone EPeyton Document Name:
FINALASP.WC OFC PDIV-2 PDIV-2 NAME JSNMh EN DATE 10/D/96 10/cf /96 0:FICIAL RECORD COPY
e 1
Mr. Neil S. Carns October 10, 1996 cc w/encls:
Jay Silberg, Esq.
Vice President Plant Operations Shaw, Pittman, Potts & Trowbridge Jolf Creek Nuclear Operating Corporation 2300 N Street, NW P. O. Box 411 Washington, D.C.
20037 Burlington, Kansas 6t -l,9 l
a
)
Regional Administrator, Region IV Supervisor Licensing l'
U.S. Nuclear Regu,latory Commission Wolf Creek Nuclear Operating. Corporation 611 Ryan Plaza Drive, Suite 1000 P.O. Box 411 Arlington, Texas 76011 Burlington, Kansas 66839 Senior Resident Inspector U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Resident Inspectors Office P. O. Box 311 8201 NRC Road Burlington, Kansas 66839 Steedman, Missouri 65077-1032 Chief Engineer Supervisor Regulatory Compliance Utilities Division Wolf Creek Nuclear Operating Corporation Kansas Corporation Commission P.O. Box 411 1500 SW Arrowhead Road Burlington, Kansas 66839 Topeka, Kansas 66604-4027 Office of the Governor State of Kansas Topeka, Kansas 66612 Attorney General Judicial Center 301 S.W. 10th 2nd Floor Topeka, Kansas 66612 County Clerk Coffey County Courthouse Burlington, Kansas 66839 Public Health Physicist Bureau of Air & Radiation Division of Environment Kansas Department of Health and Environment Forbes Field Building 283 Topeka, Kansas 66620
]
_ -. _ _ _ _ _ ~
5 i ENCLOSURE 1 o
LER Nos. 482/96-001,.002 LER Nos. 482/96-001,-002 Event
Description:
Reactor Trip with a loss of Train A of the Essential Service Water and the Turbine-Driven Auxiliary Feedwater Pump Date of Event: January 30,1996 Plant: WolfCreek Event Summary With the unit at 98% power, cold air caused i e to build on the circulating water system traveling screens. The decreased circulating water flow caused the pressure in the condenser to increase. The operators began a controlled shutdown; however, they were eventually forced to manually trip the unit from approximately 80% power in anticipation of a loss of vacuum in the condenser. Later, a frazil ice buildup on the trash racks forced operators to secure the A essential service water system (ESWS) pump and declare ESWS Train /. aut of service. Unrelated to the icing conditions, the turbine-driven auxiliary feedwater pump (TDAFWP) was declared out of service for 9 h following the discovery of a packing leak after the reactor trip. The unavailability of the ESWS pump and the TDAFWP afTected the units' response 4
to a transient event; these unavailabilities would have atTected the units' response to a transient induced loss of offsite power (LOOP) event. The conditional core damage probability estimated for this event is 21 x 10" i
Event Description On January 30,1996, the plant was operating at 98% power at the beginning of a coast-down to a refueling outage. Ice began to block the circulating water traveling screens, which caused the pressure in the condenser to increase.
Approximately I h later, operators began a controlled shutdown. Operators manually tripped the reactor when a loss ofvacuum in the condenser became imminent. The circulating water pumps were then secured due to the low water level in the intake bay for the circulating water pumps.' During the reactor trip, five control rods failed to insert fully.
Indications showed that the five control rods stopped inserting between 3,75 in. and 11.25 in. from the bottom of the core. As required by the emergency operating procedures, operators began an emergenq boration of the core. All five control rods dnfted to the bottom of the core over the next 80 min.
Approximately 90 min after the reactor trip, the TDAFWP was reported to have an inboard shaft gland leak. The pump was secured and declared out of service and, as required by Technical Specifications, the operators proceeded to take the plant to mode 4. The TDAFWP was repaired in 9 h and returned to a functional status, though operational testing was still required. Complicating the situation, the auxiliary boiler tripped on at least two occasions. The auxiliary boiler provides heating to both the reactor water storage tank (P.WST) and the condensate storage tank (CST) to prevent the water in the tanks from freezing. The lowest temperature that the water in the CST reached was 47.l*F as recorded by the plant computer. The emergency backup water supply to the CST is the ESWS.
1 Four hours after the reactor trip, inadequate ESWS warming line flow caused by original design errors and further reduced by an improper system alignment - resulted in a frazil ice buildup on the trash rack in front of the traveling screen for the A ESWS pump intake bay, This condition eventually forced the A ESWS pump to be secured due to low water level in its intake bay. Operations personnel were confused about the actual cause of the fluctuating water level in the ESWS intake bays and, after the water level in the intake bay for train A recovered, the operators attempted to l
continue pump operation. However, the water level in the intake bay did not remain high enough to allow continuous pump operation. The level in the B ESWS pump intake bay was also quite low due to a frazil ice buildup. At one point, 4
i September 13,1996 i
__.__._..._.-.____.___._______.-__m
+ r i
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l LER Nos. 482/96-001,-002 1
i j
bay for the B ESWS pump was only 4 f1 above the minimum water level required for suHicient net positive suction head
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(NPSH) and likely was moments from a failure of ESWS Train B Fortunately, the control room operators increased i
the heat load on ESWS Train B at this point and the intake water level for Train B began to recover. Ultimately, the ESWS Train A was out of service for 37 h, while ESWS Tram B pump and one of three one-half-capacity senice water pumps remained in operation.
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The operators were unable to enter mode 4 within the time required by Technical Specifications because of the inellicient j
use of the cooldown procedure. The licensee eventually melted the ice blockage and restored ESWS Train A to normal standby alignment using sparging air to better mix the ESWS return water with the cold water in the intake bay. Cooling
)
flow to Train A components was eventually reestablished by the nonsafety-related service water system (SWS). After evaluating the situation, the utility opted to enter the scheduled refueling outage early.
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Additional Event-Related Information l
During normal plant operation, the nonsafety-related SWS provides cooling to the ESWS loads and the turbine building loads, including the main feedwater (MFW) pump lube oil coolers The SWS draws water from the same intake bays j
as the circulating water pumps. The nonsafety-related SWS consists of three half-capacity pumps and one low-flow 4
startup pump. One SWS half-capacity pump remamed in operation after the trip and throughout the event.
The ESWS is a two-train safety-related system started and isolated from the SWS following a safety injection (SI) signal i
or a LOOP event. The ESWS loads are normally split and include the component cooling water (CCW) heat exchangers j
and the coolers for the diesel generators. The ESWS loads also include the room coolers for the emergency core cooling 1
system (ECCS) pumps, the charging pumps, the CCW pumps, the Auxiliary Feedwater (AFW) pumps, the control room,
}
switchgear rooms, and containment. Additionally, the ESWS provides the backup water supply to the AFW system in i
case of a condensate storage tank failure. The CCW system normally operates in a split mode and cools the residual heat i
removal (RHR) system heat exchanger, the RHR seal cooler, the charging pump bearing oil cooler, the safety injection j
pump bearing oil cooler, and the reactor coolant pumps.
j The ESWS was intended to be able to provide a flow of warm water (via a " warming line") to the ESWS intake bay.
Design input assumption errors resulted in ulequate warming line flow and lower warming line temperature than j
mtended. After the initial indication of a-mldup in the circulating water bays, the operators manually started the l
A and B trains of the ESWS system. Operm f ailed to align the ESWS properly and to isolate it from the SWS when, j
for expediency, they were directed to align the ESWS from memory. Although the expedient lineup was appropriate l
for the circumstances, a subsequent independent verification of the ESWS lineup with the procedure was not performed as required. The improper alignment resulted in further reductions of warming line flow to the ESw s intake bays for l
the pumps. This allowed frazil ice to build up on the trash racks and added to the confusion of the operators. While the l
Train A ESWS pump was declared inoperable due to the ice buildup, the water level in the Train B ESWS intake bay i
oscillated 6 to 15 ft below normal as a result of the ice buildup on its trash rack. This situation was not fully communicated to the shift supenisor.
The icing on the ESWS trash racks was the result of a phenomenon knon as frazil ice. According to the Augmented i
Inspection Report, the process starts when a body of water having a large surface area, such as the intake bay area, is subcooled by a loss of heat (as can happen on a very clear, windy, cold night). This condition, which existed at Wolf Creek, a: lowed tiny crystals of ice to form on the surface of the water. The heavy wind that existed on January 30 propelled the ice crystals below the intake surface. The water flow induced by the running ESWS pumps allowed the tiny ice crystals to readily accumulate on the metal surface of the trash racks.
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2 1
i I.
LER Nos. 482/96-001,-002 4
i Modeling Assumptions
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This event is modeled as a transient event with the TDAFWP and one train of the ESWS unavailable. The five control rods that failed to insert fully eventually drifled to the bottom of the core. The model was not specifically altered to i
reflect the control rod problem. The five control rods were considered fully inserted for modeling purposes based on J
their proximity to the bottom of the core and because the operators commenced an emergency boration as directed by i
the emerjfency giracediard
- ' ' * ~ ~
i 1
Room cooling for all the ECCS equipment is provided by the ESWS. Procedures are in place to provide altemate room cooling in the event of the loss of the normal room coolers. Considering the inclement cold and windy weather i
contributing to the event, the loss of any room cooler was not considered a factor in considering a component failed for analysis purposes. The ESWS also removes heat from the CCW system via the CCW heat exchangers. The loss of the ability to remove heat via the CCW heat exchangers causes a lou of bearmg oil or seal cooling to the SI pump c 2 Sc s
j RHR pumps. However, in the injection mode, ECCS pump coolmg was presumed to be adequate based on the flow of I
cold RWST water into the core. The ECCS injection system operator nonrecovery basic events were maintained at their nominal va'ues to reflect increased operator attention to these s,wtems during continued operation with degraded cooling.
A " place-holder" basic event (IIPI EWS-FAIL) was used to reflect a failure of the iIPI system as a result of the failure of the ESWS w here appropriate for clarity. The recirculation mode of RHR would be impacted by the potential loss of heat removal through the RHR heat exchangert This potential loss is accounted for in the models by increasing the common cause failure probability of the RHR heat exchangers (RilR-HTX-CF-ALL) to 0.1 based on the failure of ESWS Train A and the similar failure symptoms affecting Train B.
The SWS system provides cooling to the feedwater pump lube oil coolers. Because only one of three one-half-capacity SWS pumps remained in operation and the water supply from the intake bay was seriously threatened, the ability of the j
l operators to recover main feedwater was not considered possible. The operator nonrecovery value (MFW-XHE-
)
NOREC) therefore, is raised to 1.0 from the nominal value of 0.34 to reflect the inabihty of the operators to recover the l
main feedwater system, ifit were to fail. [The probability of the main feedwater system tripping (MFW-SYS-TRIP) is I
not changed because the tube oil coolers for the MFW pumps are cooled by the SWS, which still had one half-capacity j
j pump running ]
j Two basic events are added to the Wolf Creek mo lel to account for both trains of the ESWS Both events (EWS-MDP-4 FC-1 A and EWS-MDP FC-1B) are assigned a nominal failure probability of 1.78 x 10 based on data from the Wolf
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Creek Individual Plant Examination (IPE). Basic event ESW-MDP-FC-l A is set to "TRUE" (i e., failed) based on the j
unavailability of the ESWS Train A pump due to the inability to maintain the water level in the ESWS Train A pump intake bcy. Because ESh provides cooling to the emergency diesel generators (EDG), the ESWS Trc'.. A failure l
causes the model to recognize the A EDG as failed. The failure probability for ESW-MDP-FC-1B is not adjusted from the nominal failure probability given in the IPE because there is no means to indicate that an otherwise operable component is in jeopardy of imminent failure from an external cause. However, two sensitivity studies explore the
-impact of the degraded operating condition of the B ESWS pump: (1) the failure probability of ESWS Train B is increased by a factor of ten to 138 x 10-', and (2) the failure probability is changed to 01.
j A common cause failure event is also added fer the ESWS (EWS-MDP-CF.ALL). Based on the failure of the ESWS Train A purg and the operating condition of the ESWS Train B pump, the basic event probability is increased to 0.15 q
(based on the beta factor for the RHR pump). Finally, an event is added to account for the operator's failure to recover d
the ESWS ifit should fail (EWS XHE-NOREC). Because of the extreme operating conditions, however, this probability is set to TRUE (i c., no recovery).
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9 LER Non. 482/96-001,-002 The TDAFWP failure (AFW-TDP-FC-1C)is aho set to TRUE. The pump may have operated for the entire 24-h mission time with the inboard packing failure, however, predicting how long the pump could have continued to provide feedwater flow to the steam generators is ditTicult. Therefore, considering the pump uas physically disabled for repairs after the operators declared the pump to be out of service,it is appropnate to consider the pump to be failed 13ased on the initial operability of the 1DAFW pump (97 mm) and the subsequent unavailabihty of the pump (537 min), a combined operator nonrecovery value for a station blackout (Sf30) ( AF{W-XIIE-NOREC-EP) is projected as follows P(Operator fails to recover AFW) = [97(0 34) + 537(1.0)l/634 = 0 899, where the nominal value of AFW-XIIE-NOREC-EP is 0 34, while the pump is undergoing repairs, the value can be assumed to be 1.0 for an S130. If the entire 24-h mission time is considered, AFW-XIIE-NOREC-EP can be projected to be 0.586 by a similar calculation. A sensitivity study is explored for this value For the case with no LOOP, the operator nonrecovery value (AFW-XIF' NOPFn W '-11 at its nominal value of 0 26. Tius method is appropnate considering both motor-driven AFW pumps were avadaole and the TDAFWP was as ailable for at least the initial 97 mm of the event.
The emergency power system is directly alketed by the ESWS support system As previously noted, the event j
circumstances made it appear appropriate to adjust the ESWS operator nonrecoven value (EWS-XI LE-NOREC) to 1.0.
This adjustment seems to dictate that the operator nonrecoserv value for the emergency pow er sy stem j
(EPS-X11E-NOREC) also be adjusted to 10 Iloweser, not all possible EDG failures imolse a loss of ESWS support.
Therefore, EPS-XIIE-NOREC is letl at the nominal value of 0 80 A sensitiuty case was reviewed adjusting the emergency power operator nonrecovery value to 10 The Reactor Safety Study reports the probabihty of a LOOP being induced by a LOCA (transient) as 10 x 10(Reactor Safety Studv, WASIl-1400, NUREG-75/014. Table 115-3) Additionally, a search of the Sequence Codmg and Search System' for transient mduced LOOPS over a 10-year period between 1984 and 1993 revealed five transient-induced LOOPS out of 3985 trips. This calculation yields a rate of 1.25 x 10 per transient, which tends to substantiate the WASil-1400 value. The gnd based LOOP probability of shon-tenn and long-term otTsite power recovery and the probability of a reactor coolant pump (RCP) seal LOCA following a postulated station blackout were developed based on data distributions contained in NUREG-1032. Evaluation ofStation Blackout Accidents at Nuclear Power Plants The RCP seal LOCA models were developed as pan of the NUREG-1150 PRA etTorts.130th are desenbed in Revised LOOP Recovery and PliR Seal LOCA Models, ORNI/NRC/LTR-89/11, August 1989. The initiating cause of a LOOP is assumed to be a grid-related disturbance caused by the plant tnp 13ecause of the severe cold and wind, it was funher assurned that if a LOOP were to occur because of the transient, otTsite power would not be restored within 30 min. The possibility of a LOOP is added to the trai sient imtiating event tree following a reactor tnp (OFFSITE Fault Lee) This possibility leads to a transfer to an event tr e similar to the LOOP initiatmg event tree (TRANLOOP).
Analysis Results The estimated conditional core damage probability (CCDP) associated with this event is 2.1 x 10" The dominar* core damage sequence, highlighted as sequence number 21-39 on the event tree in Figs 1 and 2, contributes appronmately 66% to the CCDP estimate. This event involves a successful reactor irip, a
subsequent loss of olisite power, both trains of emergency power fail, and AFW fails to provide sutlicient flow.
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l I
l LER Nos. 482N6-001,-002 This sequence is driven by the loss of the TDAFWP and the common cause failure of the ESWS pumps In an actual station blackout, the operators would hkely have continued to operate the TDAFWP with the gland leak until it failed, while working in parallel to restore emergency power. The combined transient-induced LOOP sequences contribute 89%
of the total estimated CCDP.
The most significant transient sequence that does not involve a LOOP contnbutes approximately 6% to the CCDP estimate. This transient sequence (sequence number 20 on the' event treEm l'ig i f involves
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a successful reactor tnp, failure of AFW, failure of MFW, and failure of feed and bleed.
This transient sequence is driven by the common cause failme of the ESWS pumps and by a failure of the operator to establish make-up water to the CST (AFW-XHE-XA-MW)
A sensitivity study assumes the failure rate of ESWS Tram B is mereased by a factor of ten to 1.78 x 10 This was an 2
attempt to reflect the increased (time dependant) possibility of the remaining ESWS pump failing when intake water level decreased to 411 above the minimum required to mamtain NPSil, and poor communications did not allow this information to be properly relayed to shift n.aagement. The ESWS pump common cause failure factor due to the frazil ice buildup was leil at the previously increased value (0.15). The estimated CCDP associated with this case increases to 2.3 x 10 4 The dominant transient scquence remains the same, and the relative contnbution that transient-induced LOOP sequences add to the estimated CCDP remain about the same.
Further sensitivity studies indicate the increasing likelihood of core damage as the reliability of the second pump is decreased. If the probability of ESWS Tram B pump failure is assumed to be 0.1, the estimated CCDP associated with this case increases to 3.4 x 10" Further,if the probability of ESWS Tram B pump failure is assumed to be 0.5, the estimated CCDP associated with this case increases to 6.9 x 10~' Both of these cases are intended to capture the time sensitivity of the situation regarding the possible imminent failure of the B ESWS pump. Again, the ESWS pump common cause failure factor due to the frazil ice buildup was left at the previously increased value (0.15). This sensitivity case is calculated to reflect the potential for failure of the ESWS Train B pump. The dominant transient sequence remains the same as the base case (i e, sequence 21-39) for both cases.
When the operator nonrecoveiy value for emergency power (EWS-XIIE-NOREC) is assumed to be 1.0, the LLik increases to 2.6 x 10 compared with the base case value of 2.1 x 10 " If the operator nonrecovery value for.^5W during an SBO (AFW-XIIE-NOREC-EP) is reduced to 0.586 based on the TDAFWP availability dunng the 24-h mission period, then the CCDP is reduced to 1.6 x 10-'
Definitions and probabilities for selected basic events are shown in Table 1. The conditional probabilities associated with the highest probability sequences are shown in Table 2. Table 3 lists the sequence logic associated with the sequences listed in Table 2. Table 4 describes the system names associated with the dominant sequences. Minimal cut sets associated with the dominant sequences are shown in Table 5.
Acronyms AFW auxiliary feedwater CCDP conditional core damage probability CCW component cooling w ater i
5
LER Nos. 482/96-001,.002 CST conoensate storage tank ECCS emergency core cooling system EDO emergency diesel generator ESWS emergency service water system i
IIPI hign piessure injection IIPR high pressure recirculation IPE
-individual plant examination
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LOCA loss of coolant accident LOOP loss of offsite power NPSII net positive suction head MFW main feedwater PORV power operated relief valve PRA probabilistic risk assessment PWR pressurized water reactor RCP reactor coolant pump RCS reactor coolant system RIIR residual heat removal RWST reactor w ater storage tank SBO station blackout SI safety injection SWS service water system TDAFWP turbine-driven auxiliary feedwater pump References 1.
LER 482/96-001, Rev. O, " Loss of Circulating Water Due to Icing on Traveling Screens Causes Reactor Tnp,"
February 28,1996.
2.
LER 482/96-002, Rev. O, " Loss of A Train Essential Service Water Due to Icing on the Trash Racks," February 29,1996 3.
NRC Inspection Report 50-482/96-05, March 7,1996.
4.
WASil-1400, NUREG-75/014, Table 115-3, Reactor Safety Study, October 1975.
5.
Sequence Coding and Search Systemfor Licensee Event Reports, Vols 1-4, NUREG/CR-3905 LD, Augast 1984.
6.
WolfCreek Generating Station Individual Plant Examination Summary Report, September 1992.
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l ER Nos. 482/96-001,4H)2 Table 1. Definitions and Probabilities for Selected Basic Eients for LER Nos. 482/964101,-002 Modified Event Base Current for this name Description probability probability Type event IE-LOOP Loss of Offsite Power Initiaimg 6.9 E-006 0 0 E+000 IGNORE Yes Event IE-SGTR Steam Generator Tube Rupture 1.6 E-006 0.0 E+000 IGNORE Yes initiating Esent IE-SLOCA Small Loss of Coolant Accident 1.0 E-006 0.0 E+000 IGNORE Yes
)
Initiatmg Event 1E-TRANS Transient Initiatmg Event 5.3 E-004 1.0 E4000 TRUE Yes AFW-MDP-CF-AB Common Cause Failure of All 2.1 E-004 2.1 E-004 No Motor-Dnsen Pumps AFW-PMP-CF-ALL Common Cause Feilure of AFW 2 M E-004 2 8 E-004 No l
Pumps AFW-TDP-FC-IC AFW Turbine-Drisen Pump Fails 3 2 E-002 1.0 E+000 TRUE Yes AIV-lNK-FC-CSTI Failure of Condensate Storage 4.1 E-005 4.1 E-005 No Tank AFW-XilE-NOREC Operator fails to Recoser AFW 2.6 E 001 2 6 E-001 No System AFW-XilE-NOREC-EP Operator Fails to Recover AFW 3 4 E-001 9 0 E-001 Yes During Station Blackout AFW-XIIE-XA MW Operator Fails to Initiate Makeup 1.0 E-003 1.0 E-003 No Water CVC-MDP-FC-l A Charging Train A Fails 3 9 E-003 3 9 E-003 No CVC-MDP-FC-1B Charging Train B Fails 6 8 E 003 6 8 E-003 No EPS-DGN-FC-1B Diesel Generator B Fails 4.2 E-002 4 2 E-002 No EPS-XilE-NOREC Operator Fails to Recover 8 0 E-001 8 0 E-001 No Emergency Power EWS-MDP-CF-ALL EWS MDP Common Cause 2.6 E-004 1.5 E-001 Yes Failure EWS-MDP-FC-I A Failure of ESWS Train A 1.7 E-003 1.0 E+000 TRUE Yes EWS-MDP-FC-1B Failure of ESWS Train B 1.7 E 003 1.7 E-003 No EWS-XilE-NOREC Operator Fails to Recoser ESWS 8 4 E-001 10E+000 TRUE Yes 9
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LER Nos. 482/96-001,.002 Table 1. Definitions and Probabilities for Selected Basic Esents for LER Nos. 482/96-001,-002 Modified j
Event Base Current for this j
name Description probability probability Type-ew nt l
llPI EWS-Fall llPI Fasts m injection Mode Given 10 Em 1.0 E+000 TRUE Yes ESWS is Failed
]
HPI XIIE-NOREC Operator Fails to Recoser the llPI 8.4 E-001 8.4 E-001 No System
)
l IIPI-XilE-XM-FB Operator Fails to Irutiate feed an.1 1.0 E-002 1.0 E-002 No Biced Coohng IIPR XIIE-NOREC Operator Ft.'. to Recoser the llPR 10 E+000 10 E+000 No System LOOP Transient-induced LOOP 10 E-003 1.0 E-003 No MFW-SYS-TRIP Main Feedwater System Tr 2 0 E-001 2 0 E-001 No MFW-XIIE-NOREC Operator Fails e Recov.r Main 3 4 E-001 10E+000 TRUE Yes Feedwater OEP-XllE-NOREC-211 Operator Fails to Recoser Offsite 2.2 E-001 1.1 E-001 Yes i
Power Within 2 Ilours OEP-XIIE-NOREC-611 Operator Fails to Recoser OtTsite 6.7 E-002 3.6 E-004 Yes Power Within 6 Ilours OEP-XIIE-NOREC-BD Operator Fails to Recover Offsite 2.5 E-002 3 6 E-003 Yes Power Before Batteries Deplete OEP-XilE NOREC-SL Operator Fails to Recover Offsite 5 8 E401 4 4 E-001 Yes Power (Scal LOCA)
PPR-MOV-OO-BLK!
PORV I Block Vahe Fails to 3 0 E403 3 0 E-003 No Close PPR-MOV OO-BLK2 PORV 2 Block Vahe Fails to 3.0 E-003 3 0 E-003 No Close PPR-SRV-CC-1 PORV i Fails to Open on Demand 6.3 E-003 6.3 E-003 No PPR-SRV-CC-2 PORV 2 Fails to Open on Demand 6.3 E-003 6.3 E-003 No PPR SRV40-SBO PORVs Open During SBO 10 E6 1.0 E+000 No PPR-SRV-CO-TRAN PORVs Open During Transient 4.0 E-002 4.0 E 002 No PPR-SRVoO l PORV i Fails to Reclose After 3 0 E402 3 0 E-002 No Opening 10
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E LER Nos. 482/96-001,-002 j
Table 1. Definitions and Probabilities for Selected Basic Es ents for LER Nos. 482/96-001,-002 i
i Modified Event Base Current for this name-Description probability probability Type event PPR-SRV-OO-2 PORV 2 Fails to Reclose Aller 3.0 E402 3.0 E-002 No i
o emos r
PPR-XilE-NOREC Operator Fails to Close Block 1.1 E-002 1.1E-002 No Valve 4
i RCS-MDP-LK-SEALS RCP Seals Fail Without Coolmg 2.7 E-002 2.1 E-001 L.,
and injection RilR-IITX-CF-ALL Common Cause Failure of RilR I 4 E405 1.0 E-001 Yes Ileat Exchangers j
RilR-MDP-FC-1 A RiiR Train A Fails 3 9 E-003 3.9 E-003 No 2
RilR X11E-NOREC Operator Fails to Recoser the RilR IOE+000 1.0 E400 No System 7
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LER Nos. 482/96-001, (H)2 1
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l Table 2. Sequence Conditional Probabilities for LER Nos. 482/96-001,-002 l
I i
I Conditional core i
3 Esent tree damage Percent i
i name Sequence name probability Contribution
~"
~
(CCDP)
TRANS 21-39 1.4 E-004 65.9
)
TRANS 21-36 2.1 E-005 10.0 TRANS 21-37 1.4 E-005 6.7 TRANS 29 1.2 E-005 5.6 TRANS 21-38 9 3 E-006 4.4 i
TRANS 08 4.2 E-006 20 TRANS 21-33 4.1 E-006 1.9 l
TRANS 05 3.5 E-006 1.6 Total (all sequences) 2.1 E-004 i
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LER Nos. 482/96-001,-002 f
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r Table 3. Sequence legic for Dominant Sequences for LER Nos. 482/96-001.-002 1
i Event tree name Sequence name logic TRANS 21-39
/RT-L, OFFSITE, EP, AFW-L-EP l
TRANS 21-36
/RT-L, OFFSITE, EP, /AFW-L-EP, PORV-SBO,
/PORV-EP, SEALLOCA, /OP SL,IIPI l
TRANS 21-37
/RT-L, OFFSITE, EP, /AFW-L-EP, PORV-SBO,
/PORV-EP, SEALLOCA, OP-SL r
l TRANS 20
/RT, /OFFSITE, Alj'. MFW, F&B TRANS 21 38
/RT-L, OFFSITE. EP, /Al w'-L-D. PORV-SBO, PORV-EP l
TRANS 08
/RT. /OFFSITE, /AFW, POD '/. PORV-RES, IIPl TRANS 21-33
/RT-L, OFFSITE, EP, /AFW-1.-EP, PORV-SBO,
/PORV-EP, SEALLOCA, /OP-SL, /1IPI,
/COOLDOWN, RI1R,1 IPR i
TRANS 05
/RT, /OFFSITE, /AFW, PORV, PORV-RES,
/IIPI, /COOLDOWN, R1IR,IIPR i
I l
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LER Nos. 482/96-401,-002 Table 4. System Names for LER Nos.482/96-001,-002 l
System name legic AFW No or InsufEcient AFW AFW-L-EP No or Insu$cient AFW Flow During Station Blackout COOLDOWN Reactor Coolant System Cooldown to RilR Pressure Using Turbine Bypass Valves, etc.
EP Failure of Both Trains of Emergency Power F&B Failcre of Feed and Bleed Cooling 1IPI No or Insumeiert Flow from the Iligh Pressure Injection System i
iIPR No or Insumcient Iligh Pressure Recirculation Flow MFW Failure of the Main Feedwater System OFFSITE Transient-Induced Loss of Offsite Power OP-SL Operator Fails to Recover OiTsite Power (Seal i
LOCA)
].
PORV Power Operated Relief Valves (PORVs) Open During Transient PORV-EP PORVs Fail to Reclose (no Electric Power) 3 PORV-RES PORVs Fail to Rescat PORV-SBO PORVs Open Dunng Station Blackout RlIR No or Insumcient Flow from the RilR System i
RT Reactor Fails to Trip During Transient SEALLOCA RCP Seals Fail During LOOP 4
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LER Nos. 482/96-001,-002 1
Table 5, Conditional Cut Sets for Higher Probability Sequences for LER Nos,482/96-001,-002 1
1 Cut set Percent number Contribution Probability Cut sets' TRANS Sequence 21-39 1.4 E-004 l
1 76.8 1.1 E-004 LOOP, EWS-MDP-CF-AL1, EWS-XllE-NOREC, EPS-X1tE-NOREC, AFW-TDP-FC-lC, AFW XHE-NOREC-EP 2
21.5 3.0 E-005 LOOP, EPS-DGN-FC 1B, EWS-MDP FC A, EPS-XilE NOREC, AFW-TDP-FC-1C, AFW X11E NOREC EP TRANS Sequence 21-36 2.I E-005 1
98.8 2.1 E-005 ll)OP, EWS-NtDP-CF ALL EWS-X1tE NOREC, EPS XIIE-NOREC, PPR SkV-C04ho, RCS-MDP-LK SEALS. IIPI-EWS-FAIL llPl XI{E-NOREC 2
1.2 2.5 E-007 1AX)P EPS MDP FC I A. EWS-MDP IC lB. EWS-XilE-NOREC, EPS-XitE-NOREC, PPR-SRV-CO-SHO. RCS-MDP LK SEALS, HPI-EWS FAIL HPI-XilE-NOREC l
1
- TRANS Sequence 21-37 1.4 E-005 1
76 8 1.1E-005 IJ)OP, EWS4tDP CF ALL, EWS-XIIE-NOREC, EPS-XilE-NOREC, PPR-SRV-CO-SHO, RCS4fDP-LK-SEALS. OEP-XllE NOREC-SL 2
21.5 3.1 E-006 LOOP, EPS-DGN FC-1B. EWS4tDP-FC-1 A, EPS-X11E-NOREC, PPR-SRV-CO SBO, RCS4tDP-LK-SEAllOEP-XIIE-NOREC SL l
1 TRANS Sequence 20 1.2 E-005 4
1 54.2 6.6 E-006 AFW XIIE-XA41W, AFW-XllE-NOREC, MFW-SYS-TRIP, MFW XilE-NOREC, EWS4fDP-CF AIL EWS-XilE NOREC, llPI EWS-FAIL HPI-XIIE NOREC 2
15 2 1.8 E-006 AFW PMP-CF-AIL AFW-XilE-NOREC, MFW SYS-TRIP, MFW X1tE-NOREC, EWS4tDP-CF-ALL EWS-XIIE-NOREC, HPI EWS FAIL HPI-X1tE-NOREC 3
11.4 1.4 E-006 AFW-MDP CF-AB, AFW X1(E NOREC, MFW-SYS-TRIP, MFW-XI!E-NOREC, EWS41DP-CF-ALL, EWS-XIIE-NOR EC, llPI-EWS-FAIL, IIPI-XilE-NOREC 4
4.3 5.2 E-007 AFW-XIIE-XA-MW, AFW X11E-NOREC, MFW-SYS-TRIP, MFW X11E-NOREC,IIPl XIIE XM FB 5
2.7 3.3 E-007 AFW X1tE-XA MW, AFW XilE-NOREC, MFW SYF TRIP, MFW-X11E-NOREC, PPR-SRV CC 2 6
2.7 3.3 E-007 AFW X11E-XA-MW, AFW-XilE-NOREC, MFW-SYS-TRIP, MFW X11E-NOREC, PPR-SRV CC 1 7
2.2 2.7 E-007 AFW TNK-FC-CST!, AFW X11E NOREC,MFW SYS TRIP, MFW X11E NOREC, EWS4fDP CF.ALL, EWS-XilE.NOREC, HPI-EWS-FAIL llPI-XilE-NOREC l
15
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LER Nos 482i96 00I,-002 1
Table 5. Conditional Cut Sets for liigher Probability Sequences for LER Nos. 482/96-001 -002 1
Cut set Percent 4
. number.
Contribution Probability Cut sets" 8
1.2 1.5 E-007 AFW PMP CF-ALL AFW X11E-NOREC, MFW SYS TRIP, MFW X1tE-NOREC, HPI XIIE XM FB TRANS Sequence 21-38 9.3 E-006 1
38 4 3.6 E-006 LOOP, EWS-MDP CF-ALL, EPS-Nile NOREC, PPR SRV-CO-SBO, PPR-S RV.OO-2 2
38.4 3.6 E-006 L(X)P, EWS-MDP-CF ALL EPS X11E-NOREC, PPR-SRV-CO-SBO, PPR SRV-OO-l
)
3 10 7 1.0 E-006 LOOP. FW5i-MDP FC-1 A, EPS-DGN-FC 1B EPS-Nile-NOREC, PPk M CO-SBO, PPR-SRV OO-2 4
10.7 1.0 E-006 LOOP, EWS MDP-FC 1 A. LPS-DGN-FC-Ill, EPS-Nile NOREC, PPR SRV-CO-SBO, PPR SRV OO-l TRANS Sequence 08 4.2 E-006 1
38 8 1.7 E-006 PPR SRV CO-TRAN, PPR SRV OO-2, PPR X11E NOREC, EWS-MDP-CF-ALL, EWS-XllE NOREC,llPI-EWS-Fall.,
llPI XilE-NOREC 2
38.8 1.7 E-006 PPR-SRV CO-TRAN, PPR SRV4X)-1, PPR-XilE NOREC, EWS MDP-CF ALI, EWS-XilE-NOREC, HPI-EWS-Fall, llPI XilE NOREC 3
10.6 4.5 E-007 PPR-SRV CO-TRAN, PPR-SRV-(X) 2, PPR-MOV4X)-BLK2, EWS MDP CF-ALL, EWS-XilE-NOREC, llPI-EWS-FAII, llPI-XilE-NOREC 4
10.6 4.5 E-007 PPR-SRV-CO-TRAN, PPR SRV-OO-l, PPR-MOV-OO-BLKI, EWS-MDP-CF-A1.L EWS-XilE-NOREC, ilPI-EWS-Fall, llPI AllE NOREC TRANS Sequence 21-33 4.1 E-006 j
1 63.1 2.5 E-006 LOOP, EWS-MDP-CF-ALL, EPS XilE-NOREC, PPR SRV-CO-SBO, RCS-MDP-LK-SEALS, RilR-IITX CF-ALL, RllR-XilE-NOREC, llPR X1tE-NOREC 2
17.6 7.1 E-007 IDOP, EWS-MDP-FC 1 A. EPS-DGN-FC-113, EPS-X1tE-NOREC, PPR SRV-CO-SBO, RCS-MDP-LK SEALS, RilR-IITX-CF ALL, RilR XilE NOREC, HPR X11E-NOREC 3
7.4 3.0 E-007 LOOP, EWS MDP-FC 1 A EWS-MDP FC-Il3, EWS-XHE-NORFC, EPS XIIE NOREC, PPR SRV CO-SBO, RCS41DP LK-SEALS, RHR XilE NOREC,llPR X11E-NOREC 4
4.2 1.7 E-007 iDOP, EWS-MDP-CF ALL, EPS X1tE-NOREC, PPR-SRV-CO-SBO, RCS-MDP-LK SEALS, RilR XilE.NOREC,CVC MDP-FC IB, itPR-AllE NOREC 16
+.a.,
LER Nos. 482/96-001,-002 Table 5. Conditional Cut Sets for Higher Probability Sequences for LER Nos. 482/96-001. 002 Cut set P:rcent number Contribution.
Probability Cut sets'.
-c 5
2.4 9 8 E-008 LOOP, EWS-MDP-CF-AL1, EPS-XilE.NOREC, PPR-SRV CO-SBO, RCS MDP-LK-SEALS, RilR-MDP-FC-l A. RilR-XIIE-NOREC, IIPR XIIE-NOREO 6
24 9 8 E-008 LOOP, EWS-MDP CF-ALL, EPS-X1tE NOREC, PPR SRV-CO-SHO, RCS-MDP-LK SEALS, RilR XIIE-NOREC,CVC-MDP FC 1 A IIPR X11E-NOREC TRANS Sequence 05 3 5 E-006 1
37.I 1.3 E-006 PPR.SRV CO-1RAN, PPR SRV4X)-2 PPR XilE-NOREC, RilR X11E-NOREC, RilR-flTX-CF-ALL,itPR XIIE.NOREC 2
37 1 1.3 E-006 PPR-SRV-CO-TRAN PPR-SRV-OO-1, PPR-XilE-NOREC, LIIR-X1lE-NOREC, RilR IITX-CF.All,llPR-XIIE NORLC 3
10.I 3.6 E-007 PPR SRV-CO-TRAN, PPR-SRV-OO-2, RllR-XIIE-NOREC, PPR MOV4X)-BLK2, RilR-IITX-CF-AL1, IIPR-Nile-NOREC 4
10.1 3 6 E-007 PPR-SRV CO-TRAN, PPR SRV4X)-1, RilR-XIIE-NOREC, PPR MOV4X)-DLKI, RIIR-ilIN-CF-ALI,ilPR XilE-NOREC Total (all sequences) 2,1 E-004
- Ikuic ewres Alv TDP-FC-lC, EWS-MDP-FC 1 A, EWS-XilE-NOREC, and MFW-XIIE NOREC are all type TRUE esents which are not normally included in the output of fault tree reduction programs. These events have been added to and in understanding tlw wquences to potential core damage associated with the event.
17
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ENCLOSURE 2 a
LER Nos. 482/96-001,-002 LER Nos. 482/96-001,-002 Event
Description:
Reactor Trip with a loss of Tram A of the Essential Service Water and the Turbine-Driven Auxilian Feedwater Pump Date of Event: January 30,1996 Plant: WolfCreek Licensee Comments i
Reference:
Letter from N.S. Carns, Wolf Creek Nuclear Operating Corporation, to the U.S. Nucitar Regulatory Commission, WM 96-0075 Comment 1:
(Summag) The analysis states that, "The lowest temperature for the CST was not reported." The temperature in the CST would not have added to the I.ER Ilowever, the lowest temperature of the CST during the event was 47.1*F (2/2/96,1719 h).
Response 1:
The CST temperature was referenced because the ESWS is the backup water source for the AFW system. With the ESWS in jeopardy, the status of the CST is more prominent. The sentence was changed to reflect that the lowe t temperature recorded during the event was 47.l'F.
Comicant 2:
(Summag) The report states that the operators were directed to align the ESWS from memory which severely restricted warming line flow to the ESWS intake. This statement is not completely accurate.
The design waiming line flow requirement was 4000 ppm, but the as-built warming line flow is approximately 2500 gpm. The improper alignment initially established in the event resulted ia a warming line flow of 1700 gpm. Therefore, it is more properly the inadequate design and not the improper lineup that resulted in the rettricted warming line flow during the event. Additioaally, the WNOC operators were not directed to align the ESWS from memory. The system does not have an auto-start capability under the event circumstances. The operator did not have the proper procedure witn him and made what b believed was the correct lineup; however, the opere'or incorrectly established the warm weather ESWS alignment. Making a system realignment in this manner is appropriate for urgent situations provided the system is checked with the proper procedure after completing the scalignment. This follow-up check with the procedure was not performed.
Response 2:
The inacequate warming line flow that existed prior to the event was not appropriately addressed in the analysis. This area has now be:n addressed. It is fel: that if the operator was directed to urgently line up the ESWS for the cold weather alignment and check this against the procedure aftenvard, then the operator was essentially directed to perform the thup from memory. The lack of follow-up by the operator with the procedure or by shift managen..et in expecting to see a completed procedure confirms that the imeup was left to the operator's memon. The analysis was revisest to read," Design 18
eW, l.ER Nos. 482/96-00I,.002 input assumption errors resulted in inadequate w arming line flow and lower warming line temperature than intended. After the initial indication ofice buildup in the circulating water bays, the operators manually started the B train of the ESWS system Operators failed to properly align the ESWS and to isolate it from the SWS when, for expediency, they were directed to align the ESWS from memory.
The improper alignment resulted in further reductions of warming line flow to the ESWS intake bays for the pumps."
Comment 3:
(Summary) The discussion in the report concerning the icing conditions in the intake area is confusing.
There are three separate items discussed. (1) the process by which frazil ice is formed, (2) icing ca the trash racks (ESWS pump house), and (3) icing on the intake screens (circulating water pi house). The icing began at two different h> cations from two difTerent mechanisms. The icing problea in the ESWS was due to the frazil ice phenomenon The icing problem in the circulating water intake was the result of spraying water on the traveling screens with an ambient air temperature of 7'F.
Response 3:
The paragraph was not clear on the distinction between the icing in the circulating water intake bay and the icing in the ESWS intake bay. The paragraph was reworded to exclude any reference to the circulating water intake bays or pumps.
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