05000259/LER-2012-002
| Browns Ferry Nuclear Plant (Bfn) Unit 1 | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| 2592012002R01 - NRC Website | |
I. PLANT CONDITION(S)
At the time the condition was identified, Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, were in Mode 1. BFN, Unit 1, was at approximately 60 percent rated thermal power, BFN, Unit 2, was at approximately 49 percent rated thermal power, and BFN, Unit 3, was at approximately 50 percent rated thermal power.
II. DESCRIPTION OF EVENT
A. Event
On August 18, 2010, at 1557 Central Daylight Time (CDT), during the National Fire Protection Association (NFPA) 805 transition review, it was discovered that the current BFN Appendix R analysis does not adequately evaluate fire induced circuit damage. Fire damage to the control circuits of a motor [MO] operated valve [V] (MOV) has the potential to bypass the open/close limit switch and/or torque switch causing the valve actuator motor to stall and subject the valve train to forces which exceed design. This has the potential to damage the actuator or the moving parts of the MOV (i.e., Main Steam Drain Line [SB] valves, Residual Heat Removal (RHR) [BO] Heat Exchanger [HX] outlet valves, and Emergency Equipment Cooling Water (EECW) pump [P] cross-tie valves) such as the stem, stern threads, actuator mount, or anti-rotation device which could prevent the ability to move the valve with the associated hand wheel. The failure to manually close these valves could result in the loss of decay heat removal function and loss of credited Emergency Diesel Generators (EDG) [EK] to power required Appendix R safe shutdown equipment.
These issues have significant safety impact since the capability to manually close these valves is necessary to ensure adequate core cooling during performance of BFN Safe Shutdown Instructions (SSTs). Compensatory actions in the form of fire watches have been established in accordance with the BEN Fire Protection Report to mitigate this condition.
This condition was originally determined to be not reportable. During subsequent review, the Tennessee Valley Authority (TVA) determined this condition did meet reporting requirements. This condition was reported to the NRC on February 5, 2012, at 1706 Central Standard Time (CST).
B. Inoperable Structures, Components, or Systems that Contributed to the Event There were no inoperable structures, components, or systems that contributed to this condition.
C. Dates and Approximate Times of Major Occurrences August 18, 2010, at 1557 CDT � During the NFPA 805 transition review, it was discovered that the current BFN Appendix R analysis does not adequately evaluate fire induced circuit damage for MOVs.
February 5, 2012, at 1706 CST� BFN reported condition to the NRC.
D. Other Systems or Secondary Functions Affected
There were no other systems or secondary functions affected.
E. Method of Discovery
This condition was discovered during a NFPA 805 transition review.
F. Operator Actions
There were no operator actions.
G. Safety System Responses
There were no safety system responses.
III. CAUSE OF THE EVENT
A. Immediate Cause
The current BFN Appendix R analysis does not adequately evaluate fire induced circuit damage for MOVs.
B. Root Cause
The cause was the failure to review and resolve issues related to industry and NRC guidance on fire induced circuit damage for MOVs as a result of process deficiencies.
C. Contributing Factors
There were no contributing factors.
IV. ANALYSIS OF THE EVENT
TVA is reporting this condition in accordance with 10 CFR 50.73(a)(2)(ii)(B), as any event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.
The NRC issued Information Notice (IN) 92-18, Potential for Loss of Remote Shutdown Capability During a Control Room Fire, on February 28, 1992. IN 92-18 addresses the potential loss of capability to maintain the reactor in a safe shutdown condition in the unlikely event that a control room [NA] fire forced reactor operators to evacuate the control room. If a fire in the control room could cause hot shorts (i.e., short circuits between control wiring and power sources), there would be certain MOVs required to shut the reactor down and maintain it in a safe shutdown condition. BFN evaluated the effects of hot shorts from the control room and concluded the issues identified in IN 92-18 were not applicable to BEN due to the MOVs wiring configuration.
BFN did not evaluate the potential impact of hot shorts outside the control room on MOVs. NFPA 805 transition requires evaluation of prior dispositions to the concern of a hot short in the control circuit of a MOV that can bypass the limit switch and/or torque switch. In this condition, the potential exists to damage the MOV motor and/or valve.
The current BFN Appendix R fire safe shutdown analysis credits local manual operation of MOVs that have control circuits routed in the Fire Area (FA) of concern. If a fire induced hot short occurs that bypasses the MOV limit switch and/or torque switch, the damage to the MOV could potentially result in an inability to operate the MOV manually using the associated hand wheel. The current BFN Appendix R electrical separation analysis did not consider or disposition this concern.
V. ASSESSMENT OF SAFETY CONSEQUENCES
Failure to Close Main Steam Line Drain Valves (Only applies to Unit 2) Valves 2-FCV-001-0057 and 2-FCV-001-0058 are credited in the event that the Primary Containment Isolation System [JE] valves for the Main Steam Line Drain (2-FCV-001-0055 and 2-FCV-001-0056) are damaged by a fire. Circuits for these valves (2-FCV-001-0057 and 2-FCV-001-0058) are exposed to damage in the same FAs; therefore, their use is only credited using the hand wheel. Potential for this damage is limited to FA 2-2 and FA 3-3 in the Unit 2 and Unit 3 reactor buildings.
For Unit 1, the Main Steam Line Drain valves do not have manual operation for the Appendix R function. Therefore, this issue does not apply to Unit 1.
For Unit 2, the limit switches for these valves (2-FCV-001-0057 and 2-FCV-001-0058) can be damaged in FA 2-2 where operation with the hand wheel is required. If this line cannot be isolated after the reactor pressure vessel (RPV) is flooded in the Alternate Shutdown Cooling (ASDC) mode in 20 minutes per the SSIs, suppression pool inventory will be lost to the Main Condenser via the RHR pump, the RPV, and the Main Steam Line Drain. The action to isolate the line in the safe shutdown analysis is required in 540 minutes; therefore, it is not an immediate concern. However, isolation of the Main Steam Line Drain is eventually required if operation in the ASDC mode is continued for long term.
For Unit 3, the cables which can bypass the limit switches of valves are not located in FA 3-3, where hand wheel operation is credited. Therefore, this issue does not apply to Unit 3.
An evaluation was performed for the Unit 2 fire scenario. The total fire frequency, including the severity factor and automatic suppression activity for the failure to isolate the Main Steam Line Drain pathway, is 3.3E-5/yr. After applying a 0.3 manual non-suppression probability to the total fire frequency, the fire frequency for Unit 2 is reduced to 1.0E-5/yr.
Failure to Close RHR Heat Exchanger Outlet Valves (Applies to Units 1, 2, and 3) The residual heat removal service water (RHRSW) [BI] pumps and supply headers are shared between the units. There are four supply headers each having two RHRSW pumps and each serving one RHR Heat Exchanger on each unit. The RHR Heat Exchangers must be isolated in order to prevent flow diversion from the RHR Heat Exchangers in use. The RHR Heat Exchanger Outlet MOVs are credited for this purpose. The RHR Heat Exchangers are required to be placed in service within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to remove decay heat from the suppression pool.
If the RHR Heat Exchanger Outlet valves cannot be closed as required in the SSIs, cooling water flow to the credited heat exchangers could be significantly reduced and suppression pool temperature could increase beyond the net positive suction head limits of the RHR pumps. This could result in a loss of the decay heat removal function.
In the event that a valve which is credited to be closed with the hand wheel becomes mechanically damaged, there would be time available to diagnose the problem and isolate the flow path with a manual heat exchanger inlet isolation butterfly valve which is accessible near the MOVs.
An evaluation was performed on the failure to close the RHR Heat Exchanger Outlet valves. Including the severity factor, automatic suppression, and manual suppression, the total fire frequency for the potential to bypass the limit switches on the RHR Heat Exchanger Outlet valves is 2.0E-4/yr for each unit. Ninety percent of this frequency was due to a fire on the 480V Reactor Motor Operated Valve Board [EC] 1B, which assumed a very conservative zone of influence and no severity factor credit. Application of a forty percent reduction in the fire frequency of this one fire would reduce the total ignition frequency to below 1.0E-4/yr. There are additional considerations, such as valve failure position and hot short probabilities, which would result in further reductions in the evaluated risk associated with this condition.
Failure to Close EECW Pump Crosstie Valves (Applies to Units 1, 2, and 3) The function of the EECW pump crosstie valves is to allow the RHRSW pumps Al, B1, Cl, and D1 to be shifted from RHRSW service (normal alignment) to EECW service.
RHRSW pumps Cl and D1 have MOVs (0-FCV-067-0048 and 0-FCV-067-0049), and RHRSW pumps Al and B1 have manual valves. The normal position of these crosstie valves is closed. For the Appendix R safe shutdown analysis, RHRSW pumps Al, B1, Cl, and D1 are credited only for their normal RHRSW service, with the crosstie valve closed. Fire damage to valve control cables for 0-FCV-067-0048 and 0-FCV-067-0049 can cause the valves to spuriously open, which connects the C or D RHRSW header to the North or South EECW header. This does not cause a problem unless a RHR Heat Exchanger is placed in service on the same RHRSW header or a RHRSW outlet valve spuriously opens. This will result in EECW flow being diverted to the RHRSW system, and possibly starve cooling water flow to the EDGs, or RHRSW flow being diverted to the EECW system, resulting in reduced cooling water flow for decay heat removal.
An evaluation was performed on the spurious opening of the EECW pump crosstie valves. The total fire frequency from fires that could cause 0-FCV-067-0048 and 0-FCV-067-0049 to fail open is 8.25E-5/yr. This evaluation takes credit for automatic suppression, manual suppression, severity factors to the first target of identified targets, and application of a hot short probability. This evaluation does not take credit of any compensatory action which may be taken in the control room or in the field.
For the above conditions, roving fire watches have been established in accordance with the Fire Protection Report in order to decrease the probability of a serious fire.
VI. CORRECTIVE ACTIONS - The corrective actions are being managed by TVA's corrective action program.
A. Immediate Corrective Actions
There were no immediate corrective actions.
B. Corrective Actions
1. The conditions identified in this LER will be resolved as part of the transition to NFPA 805.
2. Verify that BFN has established a formal review of incoming regulatory correspondence in accordance with the procedure that manages TVA's interface with the NRC. This action has been completed. (This action
- addresses the failure to review and address NRC guidance on fire induced circuit damage for MOVs.) 3. Establish corporate fire protection governance and oversight of BFN fire protection activities. This action has been completed. (This action addresses the failure to review and resolve issues related to industry and NRC guidance on fire induced circuit damage for MOVs.)
VII. ADDITIONAL INFORMATION
A. Failed Components
There were no failed components.
B. Previous Similar Events
A search of BFN LERs for Units 1, 2, and 3 for approximately the past five years did not identify any similar events. However, LERs 50-259/2012-001-00 and 50-259/2012-003-00 were submitted as a result of conditions that were discovered during NFPA 805 transition reviews.
A search was performed on the BFN corrective action program. There was a previous Problem Evaluation Report (PER) associated with this condition, PER 229734.
C. Additional Information
The corrective action document for this report is PER 245385.
D. Safety System Functional Failure Consideration
This condition is not considered to be a safety system functional failure in accordance with NEI 99-02.
E. Scram With Complications Consideration
This condition did not include a scram.
VIII. COMMITMENTS
There are no commitments.