05000266/LER-2001-005, Page 3 of LER 01-005-00 and 4 Pages of Information on Point Beach Procedures
| ML030870834 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/31/2002 |
| From: | Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2003-0094 | |
| Download: ML030870834 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| 2662001005R00 - NRC Website | |
text
ýC FOR M 3 6 6 A
U S. NUCLEAR REGULAIORY COMMISSION LICENSEE EVENT REPORT (LER)
"TEXT CONTINUATION DOCKETf NUJ~MBER (2)I LER NUMBER (6)
PAGE (3)
~~~~FACILITY NAME (I)
Y NM (1)
SEQUENTIAL
- - REVISION 3oint Beach Nuclear Plant. Unit 1 05000266 I!
NUMBER NUMBER 3 OF 6 EXT (If more space,s reoured use odhfionoI copies Zf NRC Form 36M) (17) discharge flow. The significance of the timing of these actions was realized by the NMC in its self-initiated, voluntary review and update of the PRA. This condition had not been identified in the baseline PRA.
Operator training included lesson plans which identified the need and basis for maintaining minimum flows through the AFWS pumps and discussed the opening and closing logic for the recirculation valves. Operating crew simulator training included loss of instrument air scenarios. However, the specifics of the simulator program are such that failing closed the recirculation valves and shutting the AFWS discharge valves does not automatically fail the AFW pump.
Therefore, the crew simulator training may not have sensitized the operators to this vulnerability.
The PRA's capacity to integrate system performance with potential human actionfs to obtain a spectrum of plant responses allowed for identification of this vulnerability. The NMC has concluded that this vulnerability would not likely have been identified through normal surveillance or quality assurance activities. The root cause investigation of this condition identified that previous reviews in this area were generally focused on the necessity of providing adequale flow to the steam generators to remove decay heat. Because of the small margin in the capacity of the motor driven AFWS pumps in particular, it is essential in many scenarios that the recirculation valves are shut in order to assure adequate flow to the steam generators.
Corrective Actions
a A Root Cause Evaluation (RCE 01-069) Team was chartered to evaluate the vulnerability and why the risk significance of this condition was not recognized previously. The report of this team is scheduled to be provided for senior management review in late January 2002. The preliminary findings of this team with regard to root cause and contributing factors are included in the "Cause" section of this report.
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Beginning at 1520 on November 30, 2001, the operating crews were briefed on the concerns identified with a lost of IA and AFWS pump requirements to maintain adequate minimum pump flow. Temporary information tags were placed adjacent to the Control Room controls for all four AFW pumps to provide a reminder of the minimum flow requirements for each AFW pump.
Te mporary procedure changes were completed on November 30 to EOP-0, "Reactor Trip or Safety Injection" and EOP 0.1 "Reactor Trip Response o reflect the guidance provided earlier to operators via the temporary 4IM information tags. On December 14. 2001, these changes were made permanent. The step was added as a iolfout page item so that operators would stop the pumps any time the minimum flow requirements were not met.
Each operating crew received just in time training, briefings and simulator training concerning this event scenario to reinforce proper AFWS flow control.
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T On December 20, 2001, EOP 0 ano EOP 0.1 were further revised to link problems with IA as indicated by the IA header pressure low alarm with the continuing need to closely monitor and maintain adequate AFWS pump fl9 v This revision was also included in ECA 0.0. "Loss of All AC Power".
W/J Plant modifications to enhance sys:em reliability, including providing a backup air or nitrogen supply to the minimum recirculation valves. are -eing evaluated.
I Simulator modifications to enhance modeling the potential failure of the AFWS pumps following loss of instru ment air scenarios are being pursued pjvtlrý.l a711ep a
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To: Duane Schoon Procedure that are being changed via Temp Change Process to address the AFW recirc orifice issue.
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Procedure #
EOP 0 EOP 0.0 EOP 0.1 EOP 0.2 EOP 0.3 EOP 0.4 EOP 1.0 EOP 1.1 EOP 1.2 EOP 1.3 EOP 1.4 S
EOP 2 C-- EOP 3 EOP 3.1 EOP 3.2 EOP 3.3 a ECA 0.0 ECA 0.1 ECA 0.2 ECA 1.1 ECA 1.2 ECA 2.1 ECA 3.1 ECA 3.2 ECA 3.3 CSP S.1 CSP S.2 CSP C.1 CSP C.2 CSP C.3 CSP H.1 CSP H.2 CSP H.3 CSP H.3 CSP H.4 CSP H.5 CSP P. 1 Applicable unit
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1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 I &2 1 &2' 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2T 1 &2.
1 &2 1 &2
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1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2
CSP P.2 CSP Z.1 CSP Z.2 CSP Z.3 CSP 1.1 CSP 1.2 CSP 1.3 SSEP 2,,a' SEP 2Y" SEP 3.0 AOP 3.0 AOP 10
/AOP 10A AOP 23 t 4 0162A
/O1 62B 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &2 1 &
1 &2 1 &2Z 1 &2 Common to both 0if Common to both units 1 &2 Common to both units Common to both units Total procedures being changed = 102 procedures Ken Sokol 10/30/2002 j
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NUCLEAR POWER BUSINESS UNIT CRITICAL SAFETY PROCEDURES CSP INDEX Revision 58 November 20, 2001 INDEX UNIT 1 PROCEDURE NUMBER PROCEDURE TITLE REVISION EFFECTIVE NUMBER DATE CSP-ST.0 Critical Safety Function Status Trees.......................
1 C CSP-S.1 Response to Nuclear Power Generation/ATWS........ 22 C CSP-S.2 Response to Loss of Core Shutdown........................... 6 C CSP-C.1 Response to Inadequate Core Cooling....................... 22 C CSP-C.2 Response t6 Degraded Core Cooling......................... 19 C CSP-C.3 Response to Saturated Core Cooling...................... 8 C CSP-H.1 Response to Loss of Secondary Heat Sink............ 21 CSP-H.2 Response to Steam Generator Overpressure........... 9 CSP-H.3 Response to Steam Generator High Level................. 11 CSP-H.4 Response to Loss of Normal Steam Release Capabilities..........................................................
7 CSP-H.5 Response to Steam Generator Low Level.................... 8 CSP-P.1 CSP-P.2 CSP-Z.1 CSP-Z.2 CSP-Z.3 CSP-I.1 CSP-L2 CSP-L3 Response to Imminent Pressurized Thermal Shock Condition...........................................................
23 Response to Anticipated Pressurized Thermal Shock Condition.................................................
14 Response to High Containment Pressure............... 16 Response to Containment Flooding....................... 5 Response to High Containment Radiation Level....... 11 Response to High Pressurizer Level....................... 8 Response to Low PressurizeriLevel....................... 7 Response to Voids in Reactor Vessel................... 12 (A - Administrative Hold)
(T - Temporary Change)
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