05000482/LER-2011-011, Regarding Inadequate Analysis Assumptions Resulting in Deficient Control Room Evaluation Procedure
| ML12018A249 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 01/03/2012 |
| From: | Hedges S Wolf Creek |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| WO 12-0002 LER 11-011-00 | |
| Download: ML12018A249 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 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)(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)(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 |
| 4822011011R00 - NRC Website | |
text
W0LF CREEK
'NUCLEAR OPERATING CORPORATION Stephen E. Hedges Site Vice President January 3, 2012 WO 12-0002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Subject: Docket No. 50-482:
Licensee Event Report 2011-011-00, "Inadequate Analysis Assumptions Resulting in Deficient Control Room Evacuation Procedure" Gentlemen:
The enclosed Licensee Event Report (LER) is being submitted pursuant to 10 CFR 50.73(a)(2)(ii)(B) regarding an unanalyzed condition that could potentially affect post fire safe shutdown equipment at the Wolf Creek Generating Station.
Commitments contained in this LER have been stated on the attachment.
If you have any questions concerning this matter, please contact me at (620) 364-4156, or Mr. Gautam Sen at (620) 364-4175.
Sincerely, E. Hedges SEH/rlt Attachment Enclosure cc:
E. E. Collins (NRC), w/a, w/e J. R. Hall (NRC), w/a, w/e N. F. O'Keefe (NRC), w/a, w/e Senior Resident Inspector (NRC), w/a, w/e P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET
Attachment to WO 12-0002 Page 1 of 1 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by WCNOC in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr.
Gautam Sen at (620) 364-4175.
REGULATORY COMMITMENTS Requlatory commitment Due A complete review of the assumptions that are used in thermal June 15, 2012 hydraulic analysis SA-08-006, Rev. 2, will be performed to ensure that the assumptions are complete and accurate.
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2013 (10-2010)
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 Service Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail LICENSEE EVENT REPORT (LER) to infocollects.resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503.
If a means used to (See reverse for required number of impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required digits/characters for each block) to respond to, the information collection.
- 13. PAGE WOLF CREEK GENERATING STATION 05000 482 1 OF 4
- 4. TITLE Inadequate Analysis Assumptions Resulting in Deficient Control Room Evacuation Procedure
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR 05000 NUMBER NO.
05000 FACILITY NAME DOCKET NUMBER 11 03 2011 2011 011 00 01 03 2012 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
Ei 20.2201(b)
EL 20.2203(a)(3)(i)
El 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii)
[1 20.2201(d)
[] 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
El 20.2203(a)(4)
E3 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
[__ 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 E] 20.2203(a)(2)(ii)
E] 50.36(c)(1)(ii)(A)
El 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x)
El 20.2203(a)(2)(iii)
El 50.36(c)(2)
El 50.73(a)(2)(v)(A)
El 73.71 (a)(4)
El 20.2203(a)(2)(iv)
[1 50.46(a)(3)(ii)
[E 50.73(a)(2)(v)(B) 0l 73.71 (a)(5) 100
[] 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 NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILITY NAME TELEPHONE NUMBER (Include Area Code)
Gautam Sen, Manager Regulatory Affairs (620) 364-4175MANU-REPORTABLE MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT FACTURER TOEPIX
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On November 3, 2011, during the 2011 Triennial Fire Protection Inspection, it was determined that procedure OFN RP-01 7, "Control Room Evacuation," had two deficiencies. For a postulated fire in the control room, the procedure does not adequately protect the steam generators from overfill and does not adequately protect the pressurizer from filling to above 100% indicated water level, possibly causing the primary system to go solid.
The direct cause of the event is an inadequate analysis assumption translated into procedure OFN RP-017. When the event scenarios supporting OFN RP-017 were developed, the engineers who worked on the Post Fire Safe Shutdown analyses did not fully consider the potential adverse effect of automatic functions and improperly credited closure of the main steam isolation valves from the control room.
Procedure OFN RP-017 was revised to provide compensatory measures that would prevent overfill of either the steam generators or the pressurizer.
NRC FORM 366 (10-2010)LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE SEQUENTIAL REV WOLF CREEK GENERATING STATION 05000 482 YEAR NUMBER NO.
2 OF 4
2011 011 00 PLANT CONDITIONS AT THE TIME OF THE EVENT Mode 1 100 percent power No inoperable structures, components or systems contributed to this event.
DESCRIPTION OF THE EVENT On November 3, 2011, during the 2011 Triennial Fire Protection Inspection, it was determined that procedure OFN RP-017, "Control Room Evacuation," had two deficiencies. For a postulated fire in the control room, the procedure does not adequately protect the steam generators [EllS Code: SB] from overfill and does not adequately protect the pressurizer [EIIS Code: AB-PZR]
from filling to above 100% indicated water level.
Procedure OFN RP-017 did not specify the need to isolate normal feedwater [EIIS Code: SJ]
to prevent overfill of the steam generators and provided no guidance to mitigate normal feedwater from feeding the steam generators if the main steam isolation valves (MSIV) [EIIS Code: SB-V] fail to close. This could cause overfill of the steam generators and water admission to the turbine driven auxiliary feedwater (AFW) pump turbine [EIIS Code: BA-TRBJ.
Water admission to the turbine driven AFW pump turbine could cause loss of the pump, a component required for achieving safe hot shutdown. It was assumed, that for certain scenarios, the MSIVs would close using the "all-close" switches in the control room. The MSIVs are verified closed later in the procedure, at approximately 20 minutes. Preliminary thermal hydraulic analysis determined that at approximately 3 minutes after the reactor is tripped, if the main feed pumps [EIIS Code: SJ-P] are not tripped, and flow of feedwater into the steam generators is not thereby stopped, the steam generators could overfill, causing water to be admitted into the AFW pump turbine steam line resulting in potential damage to the pump.
Procedure OFN RP-017 controls water level in the pressurizer by throttling one of the four boron injection tank (BIT) outlet valves [EIIS Code: CB-V], EM HV8801B. A Safety Injection (SI) was not assumed to occur. While reviewing the thermal hydraulic analysis (SA-08-006, Rev. 2) scenarios supporting OFN RP-017, it was observed that the low pressurizer pressure SI set point was reached in some of the scenarios. Therefore, an SI signal was possible. If an SI occurs, all four of the BIT valves may open and throttling only one of the outlet valves may not prevent overfill of the pressurizer. The procedure does not provide guidance to mitigate train-A components from injecting water to the primary system due to a spurious or valid SI until late in the procedure. The train-A components could cause a pressurizer overfill condition prematurely and challenge the primary system pressure boundary.LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE SEQUENTIAL REV WOLF CREEK GENERATING STATION 05000 482 YEAR NUMBER NO.
3 OF 4
2011
-- 011 00 BASIS FOR REPORTABILITY Since a Post Fire Safe Shutdown (PFSSD) issue is identified in which no or insufficient guidance is available to Operations personnel to readily mitigate the postulated fire induced equipment maloperation, the issue is considered reportable under 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.73(a)(2)(ii)(B) as an unanalyzed condition that significantly degrades plant safety.
Since procedure OFN RP-017 would not have provided Operations personnel with the most conservative actions, Wolf Creek Nuclear Operating Corporation is reporting this condition pursuant to 10 CFR 50.73(a)(2)(ii)(B) for any event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.
ROOT CAUSE The direct cause of the event is an inadequate analysis assumption translated into procedure OFN RP-017. When the event scenarios were developed in support of procedure OFN RP-017, the engineers who worked on the PFSSD analyses did not fully consider the potential adverse effect of automatic functions and improperly credited closure of the main steam isolation valves from the control room. Therefore, the most conservative assumptions were not translated into the procedure.
CORRECTIVE ACTIONS
The apparent cause evaluation (CR 00045442) for this issue has been reviewed with the PFSSD engineers in effort to ensure an understanding of the improper assumptions that were applied in the development of thermal hydraulic analysis SA-08-006, Rev. 2.
Procedure OFN RP-017 was revised to provide compensatory measures that would prevent overfill of either the steam generators or the pressurizer.
A complete review of the assumptions that are used in thermal hydraulic analysis SA-08-006, Rev. 2, will be performed to ensure that the assumptions are complete and accurate. This action will be complete by June 15, 2012.LICENSEE EVENT REPORT (LER)
U.S. NUCLEAR REGULATORY COMMISSION (10-2010)
CONTINUATION SHEET
- 1. FACILITY NAME
- 2. DOCKET
- 6. LER NUMBER
- 3. PAGE SEQUENTLAL E REV WOLF CREEK GENERATING STATION 05000 482 YEAR NUMBER NO.
4 OF 4
2011
-- 011 00
SAFETY SIGNIFICANCE
This issue has low safety significance. There were no actual consequences since no fire has occurred in the control room that required evacuation. A fire in the control room of such magnitude and severity as to cause an evacuation and plant shutdown is extremely unlikely.
Based on the Fire Hazards Analysis (E-1 F9905), the combustible loading in the control room is low and interior finish materials meet or exceed the surface flammability requirements of applicable standards. Cables entering the control room are IEEE 383 rated. Large concentrations of cables in the control room trenches are protected with an automatic Halon extinguishing system, and automatic smoke detectors are located in the control cabinets and trenches.
OPERATING EXPERIENCE/PREVIOUS SIMILAR OCCURRENCES LER 2010-003-00 reported a condition where a postulated fire induced hot short could have prevented operation of the train 'B' diesel generator if a fire occurred in the control room. This condition was due to an inadequate review of control room circuitry for impact on the PFSSD analyses following a control room fire.
LER 2010-008-00 reported a condition where a postulated fire in the control room could cause a flow imbalance in the Essential Service Water system and cooling flow to other essential components could be reduced to below the minimum required flow. This was caused by a latent design deficiency.