05000458/LER-2006-001, Unanalyzed Condition Regarding RCIC Availability in Post-Fire Safe Shutdown Scenario
| ML060720037 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 03/06/2006 |
| From: | Lorfing D Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| G9.25.1.3, G9.5, RBF 1-06-0045, RBG-46544 LER 06-001-00 | |
| Download: ML060720037 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 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) 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)(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 |
| 4582006001R00 - NRC Website | |
text
'Entergy Entergy Operations, Inc.
River Bend Station 5485 U.S. Highway 61 N St. Francisville, LA 70775 Tel 225 381 4157 Fax 225 635 5068 dlorfin@entergy.com David N. Lorfing Manager-Licensing March 6, 2006 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Licensee Event Report 50-458 / 06-001-00 River Bend Station - Unit 1 Docket No. 50-458 License No. NPF-47 File Nos.
G9.5, G9.25.1.3 RBG-46544 RBF1-06-0045 Ladies and Gentlemen:
In accordance with 10CFR50.73, enclosed is the subject Licensee Event Report.
This document contains no commitments.
Sincerely, David N. Lorfing Manager - Licensing DNL/dhw Enclosure
_j, 6 1 01)1
Licensee Event Report 50-458 / 06-001-00 March 6, 2006 RBG-46544 RBF1 0045 Page 2 of 2 cc:
U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Sr. Resident Inspector P. 0. Box 1050 St. Francisville, LA 70775 INPO Records Center E-Mail Mr. Jim Calloway Public Utility Commission of Texas 1701 N. Congress Ave.
Austin, TX 78711-3326 Mr. Jeff Meyers Louisiana Department of Environmental Quality Office of Environmental Compliance P.O. Box 4312 Baton Rouge, LA 70821-4312
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/3012007 (004)
, the NRC may digits/characters for each block)
Inotconduct or sponsor, and a person Is not required to respond to, the
___nforma__
___on__co~llcon
- 3. PAGE River Bend Station, Unit 1 05000 458 1 of 4
- 4. TITLE Unanalyzed Condition Regarding RCIC Availability in Post-Fire Safe Shutdown Scenario
- 6. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEUNILRVFACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER REOV MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 01 5
2006 2006 - 001 -
00 03 06 2006 05000
- 9. OPERATING MIODE I. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF10 CFR§: (Checka811'that apply) o 20.2201(b)
E 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
I 0
20.2201(d) 0 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A) o 20.2203(a)(1) 0 20.2203(a)(4)
El 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) 0 20.2203(a)(2)(i) a 50.36(c)(1)(I)(A) 0 50.73(a)(2)(iii) 0 50.73(a)(2)(lx)(A)
- 10. POWER LEVEL 0 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A) 0 50.73(a)(2)(iv)(A)
[I 50.73(a)(2)(x)
El 20.2203(a)(2)(iii) 0 50.36(c)(2)
El 50.73(a)(2)(v)(A)
E 73.71(a)(4) o 20.2203(a)(2)(iv) 0 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B)
El 73.71(a)(5) 100 E 20.2203(a)(2)(v) a 50.73(a)(2)(i)(A) 0 50.73(a)(2)(v)(C) 0 OTHER o 20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) 0 50.73(a)(2)(v)(D)
Specify In Abstract below
___or In NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPIHONE NUMBER (hdude Area Code)
David N. Lorl'ing, Manager - Licensing (acting) 1225-381-4157
- 13. COMPLETE ONE UNE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT
CAUSE
SYSTEM COMPONENT MANU_
l REPORTABLE
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE JFACTURER TO EPIX FACTURER TO EPI]X
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION o YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Umit to 1400 spaces, i.e., approximately 15 single-spaced typewnitten lines)
On January 5, 2006, an engineering assessment of the River Bend Station fire protection program identified a deficiency in the post-fire safe shutdown analysis. Specifically, the safe shutdown analysis assumes that the reactor core isolation cooling (RCIC) system is available for operation at the remote shutdown panel following a postulated main control room (MCR) fire. The assessment determined that motor-operated valve E51-MOVF063, the RCIC inboard steam supply isolation, would be rendered nonfunctional. This valve is normally open, but must be assumed to close spuriously due to postulated damage to its control circuitry during the MCR fire. In this specific scenario, the valve would not be controllable from the remote shutdown panel, and the RCIC system would not be available. This is being reported in accordance with 10CFR50.73(a)(2)(i)(B) as an unanalyzed condition that significantly degrades plant safety. This condition does not cause the RCIC system to be inoperable with respect to its functions required by plant Technical Specifications. At the time of this discovery, the plant was operating at 100 percent power. Interim compensatory actions have been implemented to preclude the postulated loss of the function of this valve.
NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPERU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FAILURE CONTINUATION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE I
I YEAR ISEOUENTIAL IREVSION Y
NUMBER NUMBER River Bend Station Unit 1 05000-458 2006 001 00 2
OF 4
REPORTED CONDITION On January 5, 2006, an engineering assessment of the River Bend Station fire protection program Identified a deficiency In the post-fire safe shutdown analysis. Specifically, the safe shutdown analysis assumes that the reactor core isolation cooling (RCIC) (BN) system is available for operation at the remote shutdown panel following a postulated main control room (MCR) fire. The assessment determined that motor-operated valve
(**ISV**) E51-MOVF063, the RCIC inboard steam supply isolation, would be rendered nonfunctional. This valve is normally open, but must be assumed to close spuriously due to postulated damage to its control circuitry during the MCR fire. In this specific scenario, the valve would not be controllable from the remote shutdown panel, and the RCIC system would not be available. This is being reported in accordance with 10CFR50.73(a)(2)(i)(B) as an unanalyzed condition that significantly degrades plant safety. This condition does not cause the RCIC system to be inoperable with respect to Its functions required by plant Technical Specifications.
INVESTIGATION AND IMMEDIATE ACTIONS 10CFR50 Appendix R, Section III.L.1 states that for alternate shutdown capability (i.e.
shutdown from outside the control room) reactor coolant system process variables shall be maintained within those predicted for a loss of normal AC power and the fission product boundary integrity shall not be affected. Generic Letter (GL) 86-10, "Implementation of Fire Protection Requirements", Response to Question 3.8.4, "Control Room Fire Considerations" states that the following conditions are assumed to occur during the evaluation of an MCR fire: fire induced spurious operation of safe shutdown components has occurred, offsite power is lost, and the automatic starting of the emergency diesel generators (DGs) does not occur.
Based on the assumptions imposed by GL 86-10, the following MCR fire scenario must be addressed: fire causes E51-MOVF063 to close, fire requires the MCR to be evacuated, during relocation to the Division 1 remote shutdown panel, offsite power is lost, and the DGs do not automatically start.
In order to achieve post-fire safe shutdown conditions during a main control Room fire, the River Bend post-fire safe shutdown analysis concluded that the RCIC system was available from the remote shutdown panel in order to maintain reactor water level and that the Division 1 and 3 DGs will be started locally. (At River Bend, the Division 2 DG is not analyzed to remain free from fire damage during an MCR fire). Since E51-MOVF063 is powered from Division 2 (both control power and motive power) and there is no Division 2 power available (credit cannot be taken for Div 2 power being available) to re-open the valve, steam will not be supplied to the RCIC turbine, as assumed in the analysis. Since E51-MOVF063 is located in the drywell, manually opening the valve isU.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FAILURE CONTINUATION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL I REVISION Y
NUMBER INUMBER River Bend Station Unit 1 05000-458 2006 001 00 3
OF 4
impractical. Therefore, RCIC would not be available to maintain the reactor water level.
Establishing reactor water level control is a time-critical function required to occur within ten minutes of a main control room evacuation in order to meet the Appendix R safe shutdown performance goal of maintaining reactor coolant inventory to prevent core damage.
Until this condition is resolved, interim compensatory actions have been taken.
Abnormal Operating Procedure (AOP)-0031, "Shutdown From Outside the Main Control Room" has been revised to add an action that, upon confirmation of an MCR fire, an operator will be dispatched to open the circuit breaker which powers E51-MOVF063. The intent of this action is to remove power from the valve before it can spuriously close.
The required safe shutdown position for the valve Is open. Establishing this action as a compensatory measure is consistent with the guidance provided In Regulatory Issue Summary (RIS) 2005-07, "Compensatory Measures to Satisfy the Fire Protection Program Requirements." Also, a prohibition on hot work within the MCR has been put in place pending the resolution of this issue.
CAUSAL ANALYSIS Development of the original post-fire safe shutdown analysis methodology assumed that, for each fire area, each and every spurious actuation would occur individually. The analysis evaluated each potential fire induced spurious actuation, and identified the single worst case situation. After having selected the worst case spurious actuation, the analysis developed a mitigating action to overcome the malfunction of the fire damaged component. In the case of the MCR fire, RBS determined that a stuck-open reactor safety-relief valve (SRV) was the single worst case spurious actuation. The basis for deciding on a stuck-open SRV as the worst case spurious was the potential for uncovering the reactor core and damaging the fuel cladding. Using the stuck open SRV assumption, It was calculated that the operators had approximately ten minutes to align the RCIC system at the Division 1 remote shutdown panel in order to keep reactor water level above the top of active fuel. Without the stuck-open SRV, the minimum time needed to start the RCIC system is significantly greater than the ten minutes calculated.
A loss of offsite power and the failure of the Division 2 DG were also assumed during the MCR fire. No concurrent spurious actions in addition to the open SRV were assumed.
When RIS 2004-03, "Risk-informed Approach for Post-fire Safe Shutdown Associated Circuit Inspections," was issued, it provided clarification of the position that multiple, simultaneous spurious actuations are to be included during Implementation of the post-fire safe shutdown analysis. It was during the development of the RBS response to RIS 2004-03 that the reported condition was discovered.
- - U.S. NUCLEAR REGULATORY COMMISSION (1-2001)
LICENSEE EVENT REPORT (LER)
FAILURE CONTINUATION
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE YEAR SEQUENTIAL REVISION NUMBER NUMBER River Bend Station Unit 1 05000-458 2006 -001
- - 00 4
OF 4
CORRECTIVE ACTION
Various options have been identified to correct this condition. An evaluation will determine which option is most appropriate. This action is being tracked in the station corrective action program.
SAFETY ANALYSIS
A post-fire safe shutdown circuit analysis using the guidance of RIS 2004-03, revision 1, was performed in late 2005. Safe shutdown related attributes Involving the RBS fire protection licensing basis, procedural guidance, component selection, cable selection, and manual action procedural Implementation were included in the analysis. All risk significant scenarios identified In the RIS were evaluated for applicability. The analysis considered multiple concurrent spurious component actuations due to fire-induced cable damage to the associated automatic logic circuits. The analysis considered component failure or spurious actuation to occur without regard to specific cable fault type or cable jacket material. This methodology resulted in a conservative analysis with respect to risk considerations. The inherent divisional separation between redundant safe shutdown components and cables due to physical plant construction, cable routing segregation has resulted in a robust safe shutdown analysis. The analysis concluded that the RBS post-fire safe shutdown analysis has little vulnerability to the issues associated with the RIS. It was the result of this investigation that the condition of valve E51-MOVF063 was discovered.
This condition does not cause the RCIC system to be incapable of performing its design function with regard to plant Technical Specifications. The RCIC system is not credited by the RBS Updated Safety Analysis Report for the mitigation of any design basis accident. This condition results in a noncompliance with the requirements of 10CFR50 Appendix R, for a specific scenario that has a low likelihood of occurrence.
(NOTE: Energy Industry Component Identification codes are annotated as (**XX**).)