05000397/LER-2006-002
| Docket Numbersequential Revmonth Day Year Year Month Day Yearnumber No 0500 | |
| Event date: | 11-03-2006 |
|---|---|
| Report date: | 01-02-2007 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat |
| 3972006002R00 - NRC Website | |
Plant Conditions
At the time of the event, Columbia Generating Station (Columbia) was in Mode 4 (cold shutdown) with reactor vessel level [AD-RPV] at 70 inches and reactor coolant temperature at 114 degrees Fahrenheit. There were no structures, systems, or components (SSCs) inoperable at the start of the event that contributed to the event.
Event Description
On November 3, 2006, at approximately 0309 PST, an isolation of the Residual Heat Removal (RHR) Shutdown Cooling (SDC) [BO] common suction header occurred when the inboard primary containment isolation valve (RHR-V-9) [ISV] closed. The RHR SDC isolation occurred while performing PPM 2.7.6, "Reactor Protection System," Section 5.9, Shifting Reactor Protection System (RPS) [JC] B to ALT B or Normal Power Supply. During the RPS B transfer, electrical disconnect RHR-DISC-V/9 [DISC] was opened, per Step 5.9.6 of Plant Procedures Manual (PPM) 2.7.6, to disable RHR-V-9 and maintain SDC. Step 5.9.17 of PPM 2.7.6 required that disconnect RHR-DISC-V/9 be closed after the RPS B transfer was complete. Contrary to the intent of the procedure, opening disconnect RHR-DISC-V/9 did not remove control power from the RHR-V 9 isolation logic and created a sealed-in close signal during the transfer that isolated SDC when power was restored to RHR-V-9.
At the time of the event, RHR SDC subsystem B was operating in the SDC mode and RHR SDC subsystem A was available for SDC service but not in operation. Reactor Recirculation Pump 1A (RRC-P-1A) [AD-P] was running to support reactor core circulation and was unaffected by the RHR SDC isolation. Alternate means of decay heat removal were available at the time of the event.
The SDC isolation was caused by closure of RHR-V-9, the inboard primary containment isolation valve in the common suction line for both RHR SDC subsystems. Closure of RHR-V-9 subsequently tripped RHR pump 2B (RHR-P-2B). Operators received the "RHR B Pump Trip" alarm and entered abnormal condition procedure ABN-RHR-SDC-LOSS, "Loss of Shutdown Cooling," and ABN-LEVEL, "Unplanned Water Level Change.
Operators reopened RHR-V-9 and restored SDC with RHR SDC subsystem B at 0355 PST. During the time that RHR SDC was out of service, reactor coolant temperature increased to 148 degrees Fahrenheit and vessel level reached 95 inches. Reactor temperature and level were restored to their previous operating bands by 0451 PST.
Immediate Corrective Action Prior to the event, operators had recently been briefed on a loss of SDC and were prepared to respond.
Control Room Operators entered ABN-RHR-SDC-LOSS and ABN-LEVEL to manage the event. PPM 2.7.6 was reviewed to determine what step was being performed at the time the SDC isolation occurred. Valve travel to the fully closed position removed the sealed-in close isolation signal and RHR SDC subsystem B was placed back into service.
The applicable drawings were reviewed and confirmed that the actuations/isolations that occurred should have been expected.
26158 R3
Cause of Event
The cause of this event was an inadequate procedure step in PPM 2.7.6 that resulted in an unintended and undetected sealed-in isolation signal to RHR-V-9. RHR-V-9 has two disconnects, RHR-42-8BA2A and RHR DISC-V/9. Power for RHR-V-9 control logic is provided through disconnect RHR-42-8BA2A. Using disconnect RHR-DISC-V/9 to maintain SDC during the RPS transfer created a sealed-in isolation signal to RHR-V-9 that isolated SDC upon restoration of power to the valve.
The unintended action is the result of inaccurate technical information incorporated into PPM 2.7.6 from another approved procedure. Using RHR-DISC-V/9 to maintain SDC during RPS B transfers was originally established in SOP-RHR-SDC-BYPASS, "Bypassing RHR Shutdown Cooling Isolation Logic in Mode 4 and 5," Revision 0 dated August 10, 2004 and later incorporated in PPM 2.7.6, Revision 20 dated May 30, 2005.
Contrary to the existing expectation in SWP-PRO-02, "Preparation, Review, Approval, and Distribution of Procedures," on responsibilities for technical adequacy, the adverse impact of originally designating disconnect RHR-DISC-V/9 rather than disconnect RHR-42-8BA2A in SOP-RHR-SDC-BYPASS was not recognized.
Therefore, a contributing cause of this event is inadequate scope within the Procedure Review Program as the guidelines in SWP-PRO-02 for procedure validation lack clear direction.
Assessment of Safety Consequences
This event did not pose a threat to the health and safety of the public or plant personnel.
The RHR SDC was out of service for approximately 46 minutes during which the reactor level increased from 70 inches to 95 inches and the reactor coolant temperature increased from 114 degrees Fahrenheit to 148 degrees Fahrenheit. The increase in reactor level is partially attributed to the isolation of the previously established letdown path through RHR subsystem B. The time-to-boil was calculated to be 2.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> as of 0309 on 11/3 at the onset of the isolation.
At the time of the event, RRC-P-1A was running to support reactor core circulation and was unaffected by the RHR SDC isolation. The drywell was purged and High Pressure Core Spray (HPCS) [BG] and Condensate System [SD] were unavailable. Technical Specification 3.4.10, "RHR Shutdown Cooling System — Cold Shutdown," allows both RHR SDC subsystems and recirculation pumps to be shutdown for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period on the basis that core heat generation is low enough and the associated heatup rate slow enough that RHR flow interruptions can be tolerated.
Alternate decay heat removal was available by venting to the suppression pool and ABN-RHR-SDC-LOSS, "Loss of Shutdown Cooling," provides instruction to manually restore RHR shutdown cooling and isolate containment. Throughout the event, the reactor pressure remained significantly below the shutoff head of all of the Low Pressure Coolant Injection (LPCI) [BO] and Low Pressure Core Spray (LPCS) [BM] pumps, thereby, ensuring that sufficient capability was available to maintain reactor water inventory.
The margins in time-to-boil, Technical Specification provisions, available compensatory measures, and crew preparation mitigated the potential safety consequences associated with the RHR SDC isolation.
This event is only applicable to shutdown conditions since SDC can only be used during shutdown at low reactor pressure. If this event occurred with a shorter time to boil, operators would have been able to respond more quickly to restore SDC. Nonetheless, the safety consequences would be limited with the reactor vessel 26158 R3 A Based on NRC Integrated Inspection Report 05000397/2005004, this event is reportable in accordance with 10 CFR 50.73 (a)(2)(v)(B).
Further Corrective Actions Energy Northwest will revise procedures SOP-RHR-SDC-BYPASS and PPM 2.7.6 to designate disconnect RHR-42-8BA2A rather than RHR-DISC-V/9 to disable RHR-V-9.
Energy Northwest will identify and perform a technical review of all SDC-related procedures for technical accuracy.
Energy Northwest will reinforce expectations for technical accuracy and completeness with Procedure Sponsors and Qualified Procedure Reviewers and ensure that clear and specific guidance is provided to those performing procedure validations.
The extent of condition could extend to other infrequently used procedures. Energy Northwest will identify appropriate infrequently used procedures and perform a further review to identify and correct any technical inadequacies of consequence.
Similar Events There have been two reported isolations of SDC at Columbia within the past 5 years.
containment isolation valve RHR-V-9. This event occurred during a planned maintenance activity and was caused by maintenance personnel performing work on the wrong relay when replacing a relay wire lug.
containment isolation valve RHR-V-8. This event occurred during a planned surveillance test and was caused by an inadequate surveillance procedure.
EllS Information (Denoted as [XX1) Text Reference � System� Component Reactor Recirculation Pump, RRC-P-1A �
AD
Reactor Pressure Vessel� AD� RPV High Pressure Core Spray BG Low Pressure Core Spray �
BM
Low Pressure Safety Injection BO RHR-DISC-V/9 � BO� DISC RHR-42-8BA2A BO DISC RHR SDC Isolation Valve, RHR-V-9 � BO� ISV RHR SDC Pump, RHR-P-2B� BO Reactor Protection System JC 26158 R3