05000366/LER-2013-004
| Edwin I. Hatch Nuclear Plant | |
| Event date: | 8-13-2013 |
|---|---|
| Report date: | 04-01-2014 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| 3662013004R03 - NRC Website | |
(02-2
- 014) Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or byLICENSEE EVENT REPORT (LER) internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of InformationS Eand Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget,CONTINUATION SHEET Washington, DC 20503. If a means used to impose an information coliection does not display a current valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
2013 004 3
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EIIS Code XX).
DESCRIPTION OF EVENT
On 8/13/2013, at 1545 EDT, Unit 2 was operating at 100 percent RTP when it was determined that inter- cable analyses for the safe shutdown analysis had not been done using the guidance of Reg Guide 1.189, Revisions 2 and NEI 00-01, Revision 2. An initial analysis for some of the components in the high/low pressure interface systems was performed considering inter-cable shorts in the Safe Shutdown Analysis.
Vulnerabilities were subsequently identified, in that a postulated fire can cause two valves required for safe shutdown to open during power operation. At risk are Residual Heat Removal (RHR) valves (ISV) with controls in Fire Areas 2203 and 0024. Specifically, during the postulated fire scenario on the affected unit, an inter-cable hot short could occur on the control cables for the RHR shutdown cooling isolation valve 1 E11-F008 or 2E11-F008 causing the affected valve to open at rated power. In addition, a spurious opening of RHR shutdown cooling isolation valve 1 El 1 -F009 or 2E1 1 -F009 could occur due to an intra- cable hot short on the control cables in the same postulated event. An intersystem LOCA could then occur when the low pressure system piping is subjected to higher reactor pressures as a result of the postulated conditions. This would result in a nonconforming condition that brought the operability for these valves and the associated RHR-shutdown cooling suction line into question should there be a postulated fire that results in inter-cable and intra-cable hot shorts. Even though the occurrence of the postulated fire in conjunction with the spurious opening of the valve followed by the simultaneous occurrence of an inter- cable short of one cable and an intra-cable short of another cable is a low probability event, it was prudent to establish compensatory measures to prevent the occurrence of this potential vulnerability.
The same concern does not exist for the corresponding Unit 1 Fire Area 1203 for the El 1 -F008 (ISV) valve. A fire in Fire Area 1203 may cause the spurious opening of valve 1 El 1 -F009 (ISV), but will not cause a spurious opening of valve 1 El 1-F008 (ISV), since the power cables for the 1 E 1l-F008 (ISV) valve are routed in a dedicated conduit and no control cables for the 1El 1-F008 (ISV) valve are located in fire area 1203. However, during the 2013 NRC Triennial Fire Protection Inspection a similar vulnerability was identified that affected the same valves on Unit 1 as well as the same vulnerability on the Unit 2 valves in a different location. This vulnerability involved a postulated fire in the cable spreading room that would also result in the spurious opening of the 1 El 1 -F008 or 2E1 1 -F008 valve caused by an inter-cable short in conjunction with the spurious opening of the 1 El 1-F009 or 2E1 1-F009 valve caused by an intra-cable short. Upon identification of these vulnerabilities, actions were taken to de-activate the Unit 1 valves in the "closed" position to restore Appendix R compliance. Actions had previously been taken to de-activate the Unit 2 valves.
The Plant Hatch Fire Hazards Analysis (FHA) and the Safe Shutdown Analysis Report (SSAR), contain the following statement:
"For a postulated fire in each area, all fire-induced circuit faults (hot shorts between multiple conductors within a sinale cable, open circuits within a sinqle cable, open circuits, and shorts to around) and their effects on the safe shutdown equipment were determined. Hot shorts between conductors of different cables need not be postulated.
Initially the FHA and SSAR limited the circuit analyses to intra-cable hot shorts (conductors within the same cable), but did not consider inter-cable hot shorts (conductors within co-located cables in the same tray).
This occurred as a result of the original vendor who developed the Plant Hatch Safe Shutdown Analysis (SSA) using an unsubstantiated assumption that the inter-cable hot shorts are not credible. Additionally, at the time of the original SSA no regulatory guidance was available that addressed inter-cable hot shorts.
Evolving regulatory guidance involving the interpretation of Appendix R requirements occurred from 1999 to 2009 with respect to circuit failure analyses. In November 2009 regulatory guidance was available that provided an acceptable means for meeting the Appendix R requirements with respect to circuit failure analyses. The Southern Nuclear Fire Protection Group was aware of the methodology introduced in that guidance, but failed to act upon it.
An evaluation of fire-induced circuit faults for valves classified as high/low (H/L) pressure interface valves was performed as the extent of condition review for this event. Specific failure modes had to be evaluated to determine if in the event of a postulated fire in a fire area, the failure could lead to spurious opening of any H/L operated valve. The evaluation results were made on an area by area assessment of postulated fault conditions to determine if both valves in a H/L valve set are impacted in the same fire area. Since this evaluation had not been previously performed, vulnerabilities identified as a result of this evaluation were considered an unanalyzed condition(s). For those impacted valve sets the resulting consequence must be considered in order to determine if this unanalyzed condition significantly degrades plant safety and if compensatory actions are necessary to address noncompliances identified. DOEJ-HR-SNC515132-EO01 summarized H/L pressure interfaces associated with the valves listed for each fire area. For each valve, the conclusion is reached that the spurious opening of the valves would not result in a loss of safety function that significantly degrades plant safety. Compensatory measures in the form of fire watches were established based on the presence of identified vulnerabilities.
During the 2013 NRC Triennial Fire Protection Inspection (TFPI), the inspectors identified the additional vulnerabilities in the cable spreading room for the respective units for the same RHR shutdown cooling isolation valves, and questioned why the "extent of condition" review performed did not identify these same vulnerabilities. The initial "extent of condition" review excluded the affected control cables in the cable spreading room since credit was taken for operator actions to use the remote shutdown panel switches to effectively isolate the affected control cables which would remove the potential vulnerabilities. During the TFPI, additional review revealed that a postulated fire in the cable spreading room would not necessarily result in actuation of the remote shutdown panel switches. For this reason the vulnerability of the control cables for the affected valves in this area had to be assumed. Upon identification of this vulnerability actions were taken to de-activate the Unit 1 valves in the "closed" position to restore Appendix R compliance as previously described.
CAUSE OF EVENT
The cause of the event is attributed to ineffective follow-up and monitoring of regulatory activities involving inter-cable hot shorts in identifying the vulnerability associated with these RHR isolation valves (ISV).
Additionally, certain gaps in the SSA were identified during a previous evaluation of the viability for Plant Hatch to transition to NFPA 805. The impact of the gaps between the Plant Hatch NFPA Draft Calculation SENH-09-003 and current compliance through the safe shutdown revalidation task was not recognized.
The Unit 1 vulnerability in the cable spreading room fire area was not identified in the initial "extent of condition" review as a result of an incorrect assumption by SNC personnel that the affected circuits would be isolated in the event of a postulated fire in the cable spreading room fire area.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This event is reportable in accordance with Title 10 of the Code of Federal Regulations (CFR), Part 50.73(a)(2)(ii)(B) because the event represents an unanalyzed condition that significantly degrades plant safety. The degraded barrier is due to the postulated intersystem LOCA that could result in a significant challenge for safe shutdown of the HNP-2 or HNP-1 reactor due to loss of reactor coolant. It is reasonable to expect the plant could safely shut down, but the conditions of the shutdown will be significantly more difficult should the postulated intersystem LOCA occur.
An assessment of the fire risk associated with the potential for an interfacing system loss of coolant accident (LOCA) in the RHR shutdown cooling suction path due to postulated cable shorts in Fire Area 2203 and 0024 was performed given the additional vulnerability in the cable spreading room. This assessment concluded that for the postulated intersystem LOCA scenario for this evaluation delta core damage frequency (CDF) is equal to the delta large early release frequency (LERF) which was calculated to be 6.20E-08 for Unit 1 and 8.26E-08 for Unit 2. Given this information this condition had very low safety significance.
CORRECTIVE ACTIONS
Immediate corrective actions were taken to de-energize inboard isolation valves 1 El1 -F008, 2E1 1 -F008, 1 El 1 -F009 and 2E1 1 -F009 in the "closed" position to remove the postulated vulnerabilities and restore compliance with Appendix R requirements. A more complete evaluation of fire-induced circuit faults for valves classified as high/low pressure interface valves was performed as the extent of condition review for this event. Specific failure modes had to be evaluated to determine if a postulated fire in a fire area could cause a failure leading to spurious opening of any H/L operated valve. The evaluation summarized H/L pressure interfaces associated with the valves listed for each fire area and concluded that the spurious opening of the valves would result in some vulnerabilities, but would not result in any additional loss of safety function that significantly degrades plant safety. The Fire Protection Program (FPP) procedure was also revised to include a requirement to monitor emerging regulatory issues through participation in industry initiatives and to document changes in regulatory positions using the corrective action program.
Plant Hatch issued a letter of intent to the NRC on October 4, 2013, to begin transition to NFPA 805.
Efforts are currently underway to reconstitute the Plant Hatch SSA utilitizing Reg Guide 1.189, Revision 2 and the NRC endorsed portions of NEI 00-01, Revision 2. As part of the reconstitution, noncompliances discovered will be entered into the CAP process to initiate compensatory measures and corrective actions as necessary.
ADDITIONAL INFORMATION
Other Systems Affected: None Failed Components Information:
Master Parts List Number: 1 El 1 -F008/-F009 EIIS System Code: BO 2E1 1 -F008/-F009 Manufacturer: Wm. Powell Co. Reportable to Epix: Yes Model Number: N/A Root Cause Code: E Type: 900# O.S.Y. Gate Valve EIIS Component Code: ISV Manufacturer Code: P305 Previous Similar Events: A review of LERs and Corrective Action Program documents did not reveal any events similar to those discussed in this report within the last three years.