05000397/LER-2007-004, Re Reactor Scram Due to Tripped Condensate Booster Pump
| ML072410391 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/21/2007 |
| From: | Atkinson D Energy Northwest |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| GO2-07-130 LER 07-004-00 | |
| Download: ML072410391 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 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 |
| 3972007004R00 - NRC Website | |
text
ENERGY Dale Atkinson Vice President, Nuclear Generation & CNO
.R P.O. Box 968, Mail Drop PE08 NORT W ESTRichland, WA 99352-0968 NW Ph. 509-377-4302 F. 509-377-4150 dkatkinson@energy-northwest.com August 21, 2007 G02-07-130 10 CFR 50.73 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. 2007-004-00
Dear Sir or Madam:
Transmitted herewith is Licensee Event Report No. 2007-004-00 for Columbia Generating Station., This report is submitted pursuant to 10 CFR 50.73(a)(2)(iv)(A). The enclosed report discusses items of reportability and corrective actions taken.
There are no new 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 and Chief Nuclear Officer
Enclosure:
Licensee Event Report 2007-004-00 cc:
BS Mallett - NRC RIV CF Lyon - 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 I collection.
- 3. PAGE Columbia Generating Station 05000397 1 OF 4
- 4. TITLE Reactor Scram due to Tripped Condensate Booster Pump
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO 0500 06 28 2007 2007-004-00 08 21 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) 1 El 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
E3 50.73(a)(2)(vii)
El 20.2201(d)
[I 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
[I 50.73(a)(2)(viii)(A) 0l 20.2203 (a)(1)
El 20.2203(a)(4)
El 50.73(a)(2)(ii)(B)
[1 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)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A)
ED 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x) 70 El 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71(a)(4)
E3 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii)
El 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 starting-up following the R-18 refueling outage. Columbia was operating in Mode 1 (Power Operation) at 70 percent power with reactor water level at
+36 inches. Condensate [SD] Booster Pump [P](CBP) 2A (COND-P-2A) was out of service for maintenance.
COND-P-2A being out of service with Columbia at 70% power fostered a condition where the loss of an additional CBP would require a plant shutdown.
Event Description
On June 28, 2007 at 1717 PDT Columbia experienced an unexpected trip of COND-P-2B. The trip occurred while operators were manually transferring the COND-P-2B duplex oil filter to the standby filter. This action was being performed in response to an elevated differential pressure reading across the operating filter.
Concern existed that with COND-P-2A out of service, a further increase of the differential pressure across the filter would automatically trip COND-P-2B causing a reactor scram.
The lube oil system for each CBP includes an auxiliary oil pump used while the CBP is not in service, a normal oil pump used while the CBP is in service, a duplex oil cooler, a duplex oil filter, and piping to supply oil to both the pump and motor bearings. The duplex oil filter transfer mechanism consists of 3-way plug valves on the inlet and outlet of the duplex filter. Both valves are operated simultaneously via a mechanical linkage with a transfer handle. Normally during a filter transfer, the transfer handle is rotated 180 degrees and ultimately points to the in-service filter. As the handle is rotated, the plug valves will begin to port oil to the standby filter allowing it to be filled and vented. When the handle has been rotated to its mid-transfer position (i.e., 90 degrees), even oil flow is supplied to both filters. Further rotation up to 180 degrees completes the transfer to the standby filter and isolates the now out-of-service filter. With proper installation of the plug valves an oil flow path is maintained continuously throughout a transfer (i.e., make-before-break). At the time of the event, the oil filter transfer mechanism was installed in a break-before-make configuration with the valve stop ring reversed from its desired orientation. As such, the action of transferring filters created a temporary loss of oil flow to COND-P-2B and caused a pump trip on low oil pressure.
With both COND-P-2A and COND-P-2B out of service, feedwater [SJ] pumps [P] RFW-P-1A and RFW-P-1 B tripped on low suction pressure. The reactor automatically scrammed on a low water level (+13 inches) immediately before a manual scram was initiated. Following the scram, the reactor water level reached -50 inches and the Main Steam [SB] Isolation Valves [V](MSIVs) closed. Reactor water level was restored using High Pressure Core Spray (HPCS)[BG] and Reactor Core Isolation Cooling (RCIC)[BN]. The reactor was maintained stable in Mode 3 with pressure control established using the Safety Relief Valves (SRVs).
The NRC was notified of the reactor scram at 1853 PDT as Event Notification 43457.
Immediate Corrective Action
In response to this event, danger tags were hung on the filter transfer valves for all three CBPs and a night order issued instructing plant operators not to transfer any lube oil filters with either the associated CBP or auxiliary oil pump in service.
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
~2007-004-001 Modifications to plant operating procedures were made to provide guidance for transferring lube oil filters with both the CBPs and their associated auxiliary oil pump shutdown. The associated alarm response procedure was also revised to include guidance for investigating high lube oil differential pressure alarms as well as instructions on transferring lube oil filters.
Cause of Event
The cause of this event was a latent equipment condition involving the incorrect configuration of the COND-P-2B lube oil filter valves combined with the operational decision to transfer filters with COND-P-2B required to be in service.
From an equipment condition perspective, this event was caused by less-than-adequate configuration control for the COND-P-2B lube oil filter transfer valves. This condition was the result of an error prone valve design, a lack of proper detail in maintenance instructions, and a lack of visible and permanent alignment markings on valve components.
From an operational perspective, the crew's decision to transfer filters did not adequately assess the risk associated with performing such an operation. Prior to the transfer, no definitive determination of transient conditions was performed to establish an imminent need to perform the activity.
Assessment of Safety Consequences
This event did not pose a threat to the health and safety of the public.
With the incorrect configuration of the transfer valves, the act of rotating the transfer handle causes the oil filter outlet valve to isolate while the filter inlet remains open. This results in the positive displacement oil pump becoming deadheaded. With no relief valves between the lube oil pump and filter, the filter housing is pressurized in excess of 160 psi and can cause the pressure gauges to fail high. The maximum working pressure of the duplex filters is 150 psi; however, inspection of the housing revealed no structural deformation following this event.
The filter transfer resulted in a temporary loss of lube oil flow to COND-P-2B causing the pump to trip and feedwater to isolate. Reactor water level dropped until the low reactor water level scram setpoint (+13 inches) was reached. Coincident with the automatic scram of the reactor, the control room supervisor directed that the plant be manually scrammed. The systems necessary to support a safe shutdown were operable and performed their required functions. The reactor water level eventually reached -50 inches, causing an automatic isolation of the MSIVs. Reactor water level was restored using RCIC and HPCS with temperature and pressure control established through the Safety Relief Valves (SRVs)[SB][RV] discharging to the suppression pool. In response to increased suppression pool level, Residual Heat Removal (RHR)[BO] was aligned in the suppression pool cooling mode to compensate for SRV heat addition. The plant remained in a safe shutdown condition until corrective actions in support of restart could be performed.
Further Corrective Actions Energy Northwest will correct the physical orientation of the lube oil transfer valve stop ring for COND-P-2B and verify that the intended make-before-break configuration is achieved. Furthermore, Energy Northwest will add the desired flow plan as well as requirements for peer verification and post-maintenance testing to the model work order for CBP duplex filter transfer valves. Additional vendor supplied information regarding the 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 4 OF 4 1
2007-004-00 desired flow plan and associated configuring instructions will be added as controlled documents for the CBP filter assemblies.
Energy Northwest will revise plant operating procedures for transferring CBP oil filters to include pressure monitoring during on-line filter transfers. Energy Northwest will take further action to enhance operational decision-making processes for emergent equipment issues that fall outside the scope of formal programmatic guidance.
Energy Northwest has addressed the extent of condition. Empirical evidence indicates that the oil filter transfer valves for COND-P-2A and COND-P-2C are oriented correctly as well as those for other systems with similar configurations. However, a caution tag on COND-P-2C suggests that while the valves are correctly oriented the transfer handle is not. Energy Northwest will properly orient this transfer handle to ensure a consistent configuration among all CBP duplex oil filters.
Similar Events
There are no documented previous instances at Columbia involving the inadvertent removal of CBP lube oil filters from service. In June 2002 the lube oil transfer valves for COND-P-2A were found in the mid-transfer position. As these valves were correctly configured, both filters were in service and did not create an adverse condition. Nonetheless, these valves are common maintenance items for leakage requiring periodic disassembly. Disassembly of these valves provides the opportunity to introduce configuration errors during reassembly. The elimination of such configuration errors is an important focus of the corrective actions resulting from this event.
EIIS Information Text Reference System Component Condensate Booster Pump, COND-P-2AIB/C SD P
High Pressure Core Spray BG Main Steam Isolation Valves SB V
Reactor Core Isolation Cooling BN Reactor Feedwater Pump, RFW-P-1A/B SJ P
Residual Heat Removal BO Safety Relief Valves SB RV 26158 R3