ML20082G734
| ML20082G734 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/10/1995 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-95-021, CON-NRC-95-21 VPNPD-95-035, VPNPD-95-35, NUDOCS 9504130376 | |
| Download: ML20082G734 (4) | |
Text
r i
JM Wisconsin Electnc POWER COMPANY 231 w Mchgm PO Box 2046, MWee.WI 53201-2046 (414)221-2345 VPNPD-95-035 NRC-95-021 April 10, 1995 Document Control Desk U.
S.
NUCLEAR REGULATORY COMMISSION Mail Station F1-137 Washington, DC 20555 Gentlemen:
DOCKETS59-266 AND 50-301 REVIEW OF PRELIMINARY ACCIDENT SEOUENCE PRECURSOR ANALYSIS OF AN OPERATIONAL CONDITION AT POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 In a letter dated March 7, 1995, you provided a copy of the preliminary Accident Sequence Precursor (ASP) analysis of an operational condition which was discovered at Point Beach Nuclear Plant (PBNP) Units 1 and 2, on February 8, 1994.
The condition involved the inoperability of both Emergency Diesel Generators and is documented in Licensee Event Report 266/94-002-00.
You requested we review the analysis and provide comments to you within 30 days of our receipt of the letter.
Our comments on the preliminary ASP analysis are enclosed.
If you require additional information, please contact Mr. Stan Guokas of our staff at (414) 221-3973.
Sincerely, l'
<&f c.
Bob Link Vice President Nuclear Power Enclosures cc:
NRC Resident Inspector NRC Regional Administrator KVA/ cms 2bbbbb 00 9504130376 950410
~
l l PDR ADOCK 05000266 P.
PDR i
A subsM,,7dInmis.h ExyOw;wauw
e Comments on Section 1.1.5,
" Analysis Results" Tha second paragraph discusses a sensitivity study that assumes a 2-h battery life instead of the 1-h used for the base case. A 2-h battery life changes the conditional core damage probability associated with this event from 2.0E-5 to 1.0E-5.
We believe a 2-h battery life is appropriate for this event based on the following:
1.
Non-safety related batteries were installed at Point Beach Nuclear Plant (PBNP) in 1993.
This allowed the power supply for the main turbine bearing emergency oil pumps (a non-essential load) to be moved from the safety related to the non-safety related batteries.
PBNP calculations N-89-025 and N-89-026, performed prior to installation of the non-safety related batteries, determined that safety related batteries, DOS and D06, would be able to fulfill their design function for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, including supplying the main turbine bearing emergency oil pumps.
of the total load of 695 amps, 348 amps were drawn by the main turbine bearing emergency oil pumps.
Therefore, after switching these pumps to the non-safety related batteries, which occurred prior to the event in question, the total load on D05 and D06 is 347 amps.
2.
Calculations N-89-025 and N-89-026 did not take credit for load shedding that would be performed in accordance with Emergency Contingency Action (ECA) 0.0,
" Loss of All AC. Power."
ECA 0.0, Revision 11, dated September 24, 1993, was the revision in effect at the time of the event.
Step 31.a of ECA 0.0 directs the operators to check that battery bus voltage is greater than 120Vdc.
The Response Not Obtained column for this step directs the operators, in part, to minimize DC loads by securing selected non-safety related DC loads.
These are the electrical generator air side seal oil back-up pump (75 amps) and steam generator feed pump DC lube oil pumps (11 amps each).
3.
Safety related batteries D-05 and D-06 have a nominal rating of 385 amps for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (1540 amp-hours), 325 amps for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (1625 amp-hours) and 225 amps for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (1800 amp-hours).
Therefore, with a load of 347 amps, the batteries can be expected to last for approximately 4 to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
After the load shedding of 97 amps discussed above, the load is reduced to 250 amps.
The batteries can then be expected to last for approximately 6 to 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.
The batteries were in a nominal condition at the time of this event.
Comments on Table A.1.1,
" Definitions and Probabilities for Selected Basic Events for LER 266/94-002" ITEM PPR-SRV-CO-L:
The preliminary ASP analysis conservatively assumes that for every occurrence of a loss of offsite power (LOOP) the pressurizer power operated relief valves (PORVs) will open.
This is provided in Table A.1.1 under primary name, PPR-SRV-CO-L, "PORVs open during LOOP" (1.0 Probability).
Based on the discussion below, the PORVs will not open on a LOOP. We recommend this probability be set conservatively at 0.05.
The pressurizer PORVs are designed to open at 2350 psia.
The PBNP reactor coolant system (RCS) was originally designed for a normal operating pressure of 2250 psia.
The normal operating pressure for both units was later changed to 2000 psia.
The RCS normal operating pressure in both units at the time of this event was 2000 psia.
This provides an additional 250 psi margin to reach the pressurizer PORV setpoint.
Plant data demonstrates that RCS pressure does not significantly increase after a loss of AC power event.
On August 16, 1987, PBNP Unit 2 experienced a reactor trip due to a loss of AC.
Pressurizer pressure reached 2067 psia.
Therefore, based on this information, we would not expect the pressurizer PORVs to open on a loss of offsite power.
The August 16, 1987, PBNP Unit 2 event was not a station blackout (SBO).
The emergency diesel generators automatically repowered the Unit 2 safeguards buses.
The auxiliary feedwater (AFW) flow rate during the August 16, 1987, event was approximately 400 gpm per steam generator during the first 4 minutes.
After the first 4 minutes, the flow was throttled back to 200 gpm per steam generator, the expected flow rate during a station blackout.
The flow rate of 400 gpm was provided by both electric AFW pumps and the unit-specific turbine-driven AFW pump.
During a SBO event, only the turbine-driven AFW pump would operate, providing a flow rate of 200 gpm per steam generator.
During a SBO event, the lower AFW flow rate during the first four minutes could cause RCS pressure to go slightly higher than it did in the August 16, 1987 event.
However, the analysis in the P3NP Final Safety Analysis Report (FSAR), Section 14.1.11, " Loss of All AC Power to the Auxiliaries," which conservatively assumes an AFW flow of 100 gpm per steam generator, demonstrates that the RCS pressure increase after a loss of all AC power is minimal.
Therefore, it has been concluded that a PORV would not be expected to open during a station blackout.
ITEM AFW-TDP-FC-1A:
The preliminary ASP analysis assumes the probability of failure of the AFW turbine-driven pump is 1.5E-1.
This is given under primary name AFW-TDP-FC-1A, "AFW turbine-driven pump fails."
As stated in the PBNP Individual Plant Examination (IPE) submittal to the NRC, dated June 30, 1993, the following probability data was used for the turbine-driven auxiliary feedwater pumps (TDAFWP):
- TDAFWP fails to start = 1.9E-3.
- TDAFWP fails to run for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> = 4.31E-2.
TDAFWP out for testing and maintenance = 9.1E-3.
TDAFWP common cause failure steam admission valves = 1.E-4.
This data was developed from a bayesian update of Idaho National Engineering Laboratory report EGG-SSRE-8875, " Generic Component Failure Data Base for Light Water and Liquid Sodium Reactor PRAs,"
dated February, 1990, with PBNP plant-specific data collected from January 1, 1985, to September 5, 1990.
Because any one of these failures would fail the TDAFWP, the failure rates are summed to give a TDAFWP failure probability of 5.42E-02.
Other much smaller contributors are; spurious closure of the normally open discharge valve, failure of the discharge check valve to open, and spurious closure of the AFW suction valve.
Based on the above, we recommend the probability of item AFW-TDP-FC-1A be set to 6.0E-02.
ITEM AFW-XHE-NOREC-EP:
The preliminary ASP analysis assumes the probability of failure of the operator to recover AFW during station blackout is 3.4E-1.
This is given under primary name AFW-XHE-NOREC-EP, " Operator fails to recover AFW during station blackout."
As stated in the PBNP IPE submittal, the probability that an operator will fail to control steam generator flow with minimum steam generator level indication is 2.4E-1.
We, therefore, recommend the probability of item AFW-XHE-NOREC-EP be set to 2.4E-1.