05000461/LER-2011-001
| Clinton Power Station, Unit 1 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| Initial Reporting | |
| ENS 46603 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition |
| 4612011001R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
General Electric — Boiling Water Reactor, 3473 Megawatts Thermal Rated Core Power Energy Industry Identification System (EllS) codes are identified in the text as [XX].
EVENT IDENTIFICATION
Postulated Spurious High Pressure Core Spray Initiation Result Unanalyzed
A. CONDITION PRIOR TO EVENT
Unit: 1 Event Date: 2/8/11 Event Time: 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> CST Reactor Mode: 1 Mode Name: Power Operation Power Level: 96.9 percent
B. DESCRIPTION OF EVENT
On 11/18/10, a self assessment performed in preparation for a NRC inspection identified that a manual action in off-normal procedure CPS 4003.01, "Remote Shutdown," was not consistent with the station's 10 CFR 50, Appendix R Safe Shutdown Manual Action Feasibility Study. The calculation for the study states that the High Pressure Core Spray (HPCS) [BG] Pump [P] is tripped in response to a fire in the Main Control Room (fire zone CB-6) [NA], by depressing the internal breaker [BKR] (4 kilovolts) trip plunger at the local panel [PL] to avoid a Reactor Pressure Vessel [RPV] overfill event. This action was implemented in CPS 4003.01 until 1/22/08 at which time the procedure was revised to locally close HPCS injection valve [INV] 1E22-F004 in lieu of tripping the breaker. The procedure revision was initiated due to personnel safety concerns of arc flashing while operating the plunger. However, this change in the manual action was not evaluated and/or reconciled within the Fire Protection Program in accordance with the station's Appendix R Safe Shutdown Manual Action Feasibility Study.
On 2/8/11, at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />, during Engineering's review of the current methodology to close the HPCS injection valve rather than open the breaker, it was determined that the entire HPCS system could conservatively be postulated to spuriously initiate due to concurrent fire induced hot short cable [CBL] damage to the two automatic initiation logic instrument cables routed in the same raceway in Fire Zone CB-6. In this event, the RPV would fill and flood the Main Steam [SB] Lines (MSLs) because the shorting would prevent HPCS pump shutdown and shutting the injection valve to terminate HPCS flow into the RPV. In response to the continuing HPCS injection, Main Steam Safety Relief Valves (SRVs) [RV] would lift after RPV pressure reaches the SRV setpoint and RPV water would be discharged through the SRVs and downcomers to the suppression pool.
This would result in high pressure, high temperature water discharge through SRVs, which would flash to two phase flow. SRVs and their associated downcomers have not been analyzed in the current design basis for the stresses expected due to this two phase flow. This analysis gap has potential impact on the quenchers and the pressure suppression function of the pool should the two phase flow loads cause quencher failure.
This condition is an analytical nonconformance due to a missing analysis rather than a physically degraded component. This scenario could occur due to fire with the unit being controlled from either the Main Control Room or the Remote Shutdown Panel (RSP).
This scenario is very similar to the event postulated in NRC Unresolved Item (URI) 2005006-01 which discussed fire induced electrical faults in the control cables and control logic of the HPCS pump and injection valve from a fire in the Division 3 switchgear [SWGR] room (Fire Zone CB-5a) which could result in spurious actuation of the HPCS pump and reactor core injection. In this scenario, the faults would also impair the capability to shut off the HPCS pump and stop it from injecting water into the core.
Engineering completed an evaluation to address this event and determine additional actions to be taken. The evaluation reviewed the adequacy of the proposed alternative compensatory measures, as compared to the existing required alternate compensatory measures developed for Multiple Spurious Operations (MSO) (required shiftly Operator rounds, the shiftly transient combustible surveillance, Fire Marshal tours, administrative controls on combustible material and Operations fire brigade training). The evaluation demonstrated that the alternative compensatory measures would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire and that compliance with 10 CFR 50, Appendix A, General Design Criterion III, and 10 CFR 50.48(a), "Fire Protection," is met.
The station implemented an alternative compensatory action by adding Fire Zone CB-5a as a line item in eSOMS for Operator rounds.
No inoperable equipment or components directly affected this event.
Issue Report 1172335 was initiated to investigate this event and initiate corrective actions.
The NRC Operations Center was notified about this condition via Event Notification 46603 on 2/8/11 at 1406 hours0.0163 days <br />0.391 hours <br />0.00232 weeks <br />5.34983e-4 months <br /> EST.
C. CAUSE OF EVENT
The cause of this event is attributed to a failure to follow procedure AD-CL-101-1004, "CPS SPECIFIC `xxxx.xx' NUMBERED PROCEDURES WRITERS GUIDE." The writer's guide establishes performance of disciplinary reviews for specifically coded procedures to ensure the procedures are technically and functionally accurate for all functional areas. Procedure CPS 4003.01 is a procedure that specifically requires an engineering review of changes for impacts. Operations procedure writers did not ensure an engineering review of changes to procedure CPS 4003.01.
D. SAFETY CONSEQUENCES
This event is reportable under the provisions of 10 CFR 50.73(a)(2)(ii)(B) as a condition that resulted in the nuclear power plant being in an unanalyzed condition that degraded plant safety.
The risk significance of an overfill event is low. An overfill event represents a deviation from the intended shutdown strategy and reactor response, but does not result in any immediate core cooling challenges. The steam lines downstream of the SRV discharge flange are not part of the reactor coolant pressure boundary.
Damage to this piping inside the drywell would pressurize the drywell with flashing steam. This steam pressure would act through the drywell Loss of Coolant Accident (LOCA) vents rather than the SRV quenchers and remain bounded by LOCA drywell/containment performance. It would not impact safe shutdown per 10 CFR 50, Appendix R. The continued operation of HPCS will ensure adequate RPV water inventory to maintain adequate fuel temperatures to protect the fuel cladding, and plant capability to operate multiple SRVs will control pressure and dedicated safe shutdown equipment can be controlled from either the Control Room or RSP for a fire in any plant area. Review of calculations related to the quenchers identified that sufficient margin is available to provide assurance that the quenchers will not fail even with LOCA and earthquake loads being considered.
E. CORRECTIVE ACTIONS
Operations procedure writers and supervisor have been briefed on the requirements of procedure AD-CL- 101-1004, and the need to formally request review by other departments, as required.
Operations will validate that procedures related to arc flash concerns have had the cross-disciplinary reviews performed for procedure technical and functional accuracy.
Operations will review a sample of procedures with certain Class Codes related to Engineering, Environmental and In-Service Inspection procedure revisions for the past five years to ensure the cross- disciplinary reviews were performed to ensure the procedure was technically and functionally accurate, if necessary.
A transient and stress analysis of high pressure, high temperature, two phase SRV discharge flow will be completed.
F. PREVIOUS OCCURRENCES
The condition discussed in this report is similar to NRC URI 2005006-01 identified 6/30/2005. This document discusses postulated fire-induced electrical faults in the control cables and control logic of the HPCS pump and discharge valve from a fire in the Division 3 switchgear room which could result in spurious actuation of the HPCS pump and core injection. These faults could impair the capability to shut off the pump and stop it from continually injecting into the core.
In 2005, the ability to trip the HPCS breaker was still an action in procedure CPS 4003.01, and the ability to shutoff the HPCS pump before it could lead to an overfill event that would lead to SRVs opening as designed to relieve pressure, with water exiting a fully flooded reactor pressure vessel through open SRVs, through the downcomers to the suppression pool was considered achievable by Engineering and Operations.
In the July, 2008 CPS NRC Triennial Fire Protection Inspection (2008006), the inspectors followed-up on the URI, but again concluded that no action was required on the part of the licensee, and that the issue required resolution by NRC Office of Nuclear Reactor Regulation.
G. COMPONENT FAILURE DATA
None