05000397/LER-2006-002, From Columbia Generating Station, Regarding Shutdown Cooling Isolation Due to Inadequate Procedure Step
| ML070090481 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/02/2007 |
| From: | Atkinson D Energy Northwest |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GO2-07-002, IR-05-004 LER 06-002-00 | |
| Download: ML070090481 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3972006002R00 - NRC Website | |
text
ENERGY 0
NORTHWEST Peopfe
- Vision
- Solut;ions P.O. Box 968
- Rich*and, WA
- 99352-0968 January 2, 2007 G02-07-002 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001
Subject:
COLUMBIA GENERATING STATION, DOCKET NO. 50-397 LICENSEE EVENT REPORT NO. 2006-002-00 References: 1)
Letter dated November 7, 2005, C.E. Johnson (NRC) to J.V. Parrish (EN), "Columbia Generating Station -
NRC Integrated Inspection Report 05000397/2005004."
Dear Sir or Madam:
Transmitted herewith is Licensee Event Report No. 2006-002-00 for Columbia Generating Station. This report is submitted pursuant to 10 CFR 50.73(a)(2)(v)(B) as specified by NRC in Integrated Inspection Report 05000397/2005004 (Reference 1). The enclosed report discusses items of reportability and corrective actions taken.
There are no new regulatory commitments being made. If you have any questions or require additional information, please contact Mr. GV Cullen at (509) 377-6105.
Respectfully, DK Atkinson Vice President, Nuclear Generation Mail Drop PE08
Enclosure:
Licensee Event Report 2006-002-00 cc: BS Mallett - NRC RIV RF Kuntz-NRC NRR INPO Records Center NRC Sr. Resident Inspector-988C (2)
RN Sherman - BPA/1 399 WA Horin - Winston & Strawn CE Johnson - NRC RIV/fax
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 6/30/2007 (6-2004)
, the NRC may digits/characters for each block) not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. FACILITY NAME
- 2.
OCKET NUMBER
- 3. PAGE Columbia Generating Station 05000397 1 OF4
- 4. TITLE Shutdown Cooling Isolation due to Inadequate Procedure Step
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER I
NUMBER NO 0500 11 03 2006 2006 - 002 - 00 01 02 - 2007 FACILITY NAME DOCKET NUMBER 0500
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 0l 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
[] 50.73(a)(2)(viii)(A)
El 20.2203 (a)(1)
[I 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
El 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iii)
[I 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x) 00 El 20.2203(a)(2)(iii)
El 50.36(c)(2)
[E 50.73(a)(2)(v)(A)
[E 73.71(a)(4)
El 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B)
El 73.71(a)(5)
El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER El 20.2203(a)(2)(vi)
El 50.73(a)(2)(i)(B)
El 50.73(a)(2)(v)(D)
Specify in Abstract below or in (If more space is required, use additional copies of NRC Form 366A)
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 R3U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REVISION Columbia Generating Station 05000397 NUMBER NUMBER 3 OF 4 1
2006-002-00
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 "U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REVISION Columbia Generating Station 05000397 NUMBER NUMBER 4 OF 4 1
2006-002-00 closed and alternate methods of decay heat removal available.
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.
LER 2003-003 reported an RHR SDC isolation due to the closure of the inboard RHR SDC primary 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.
LER 2003-005 reported an RHR SDC isolation due to the closure of the outboard RHR SDC primary containment isolation valve RHR-V-8. This event occurred during a planned surveillance test and was caused by an inadequate surveillance procedure.
EIIS Information (Denoted as [XX])
Text Reference System Component Reactor Recirculation Pump, RRC-P-1A AD P
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 P
Reactor Protection System JC 26158 R3