05000263/LER-2009-001, Containment Overpressure Not Ensured in the Appendix R Analysis
| ML092720864 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/18/2009 |
| From: | O'Connor T Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-09-082 LER 09-001-00 | |
| Download: ML092720864 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(ii) |
| 2632009001R00 - NRC Website | |
text
XcelEnergya September 18, 2009 L-MT-09-082 10 CFR Part 50.73 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed License No. DPR-22 LER 2009-001 Revision 1, "Containment Overpressure not ensured in the Appendix R Analysis" A revised Licensee Event Report (LER) for this occurrence is attached. The revision of this LER is to correct the Safety Significance for the event.
This letter contains no new commitments and no revisions to existing commitments.
Tim J. O'Connor Sit ce President, Monticello Nuclear Generating Plant Northern States Power - Minnesota Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC 2807 West County Road 75 e Monticello, Minnesota 55362-9637 Telephone: 763-295-5151
- Fax: 763-295-1454
NRC FORM 366 U.S. NUCLEAR REGULATORY APPROVED BY OMB NO. 3150-0104 EXPIRES 8-31-2010 (9-2007)
COMMISSION
, the NRC may not conduct or sponsor, and a digits/characters for each block) person is not required to respond to, the information collection.
FACILITY NAME (1)
DOCKET NUMBER (2)
PAGE (3)
Monticello Nuclear Generating Plant 05000263 1 of 4 TITLE (4) Containment Overpressure not Ensured in the Appendix R Analysis EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
MO DAY YEAR YEAR SEQUENTIAL REV MO DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NO 05000 FACILITY NAME DOCKET NUMBER 04 02 2009 2009 001 01 05 29 2009 05000 OPERATING 5
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) (11)
MODE (9) 20.2201 (b) 20.2203(a)(3)(ii)
X 50.73(a)(2)(ii)(B) 50.73(a)(2)(ix)(A)
POWER 0
20.2201 (d.
20.2203(a)(4) 50.73(a)(2)(iii) 50.73(a)(2)(x)'
LEVEL (10) 20.2203(a)(1).
50.36(c)(1)(i)(A) 50.73(a)(2)(iv)(A) 73.71 (a)(4) 20.2203(a)(2)(i) 50.36(c)(1 )(ii)(A)
__50.73(a)(2)(v)(A) 73.71 (a)(5) 20.2203(a)(2)(ii) 50.36(c)(2) 50.73(a)(2)(v)(B)
OTHER
- - Specify in Abstract below or 20.2203(a)(2)(iii) 50.46(a)(3)(ii) 50.73(a)(2)(v)(C) in NRC Form 366A 20.2203(a)(2)(iv) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(D) 20.2203(a)(2)(v) 50.73(a)(2)(i)(B) 50.73(a)(2)(vii) 20.2203(a)(2)(vi) 50.73(a)(2)(i)(C) 50.73(a)(2)(viii)(A)
T20.2203(a)(3)(i) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(B)
LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (Include Area Code)
Ron Baumer [763-295-1357 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE S
MANU-REPORTABLE FACTURER TO EPIX FACTURER TO EPIX SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR
___SUBMISSION YES (If yes, complete EXPECTED SUBMISSION DATE).
X NO DATE (15)
ABSTRACT During the review of calculations to respond to a request for additional information from the NRC in support of the station's Extended Power Uprate license amendment, station personnel discovered that although the plant credits Containment Overpressure the Appendix R analysis does not ensure Containment Overpressure is maintained. The cause of the event was the calculation that credited the Containment Overpressure was either not reviewed by fire protection program personnel or program personnel failed to internalize the need for a revision to the Appendix R analysis. Until the issue is resolved in the station's transition to NFPA-805, compensatory measures have been put in place to address the vulnerabilities.U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
SEQUENTIAL REVISION Monticello Nuclear Generating Plant 05000263 YEAR NUMBER NUMBER 2 of 4 2009 001 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
Event Description
On April 2, 2009, the plant was shutdown and in Mode 5 for a refueling outage. During the review of calculations to respond to a request for additional information from the NRC in support of the station's Extended Power Uprate license amendment, station personnel discovered that although the plant credits Containment Overpressure (COP), the Appendix R analysis does not ensure Containment Overpressure is maintained. For certain scenarios, COP is required in order to ensure adequate net positive suction head is available for vessel injection and suppression pool cooling pumps. The station procedure for performing a shutdown from outside the control room did not contain any steps to require operators to ensure adequate net positive suction head was maintained. Although this issue was discovered in April 2009, the issue has existed since design basis accident reanalysis in 2002.
Depending on initial conditions, a fire in the cable spreading room or control room, with or without a fire induced loss of off site power, could cause two in-series valves to spuriously open leading to venting of containment. Venting of containment during this scenario could result in the loss of required net positive suction head (NPSH) for the 12 Core Spray (CS) [BM]
and the 12 Residual Heat Removal (RHR) [BO] pumps [P]. With no operator action, inadequate NPSH could prevent vessel injection and or suppression pool cooling.
Event Analysis
The event was reportable under 10 CFR 50.73(a) (2) (ii) Degraded or Unanalyzed Condition.
Since the issue was for past operability, there was no 10 CFR 50.72 report for this event.
The event is not considered a safety system functional failure since all required safety related systems would have been available to perform their safety functions.
Safety Significance
There was no impact to the health and safety of the public; and the risk significance is less than the red threshold requirements for 10 CFR 50.48.
Monticello maintains administrative controls for the introduction of combustible materials and ignition sources to limit the probability and severity of a fire. Fire detection equipment or continuous manning is provided in the affected areas for rapid detection. Fire extinguishers and hose stations are in place and fire brigade response is expected to be rapid and successful for any fire in the affected areas. These prevention, detection and suppression measures limit the frequency of fire induced loss of required COP to a very small value.U.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6 PAGE (3)
SEQUENTIAL REVISION Monticello Nuclear Generating Plant 05000263 YEAR NUMBER NUMBER 3 of 4 2009 001 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
A Significance Determination Process (SDP) evaluation was performed per the guidance of NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. In summary, this analysis determined that the change in Core Damage Frequency attributable to this event is less than the red risk threshold. Accordingly enforcement discretion via the provisions of Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) is warranted.
Cause
Fire protection engineering personnel either did not review or were unaware/did not internalize that COP was required for an Appendix R fire and that the Appendix R analysis might require revision to ensure COP was maintained. Although the calculation procedure in 2002 required a fire protection review for the potential to affect the fire protection analysis, the requirements were not clear enough to ensure the personnel performing the calculation revision obtained a fire protection review.
Corrective Action
The calculation process has been revised since 2002. Specifically, completion of a Fire Protection Program checklist is now required by the fleet calculation procedure. Therefore, the process has already been revised to prevent this type of error, and no further action is required to prevent recurrence.
Final resolution of this condition will be completed with station's transition to the risk informed, performance-based fire protection program in NFPA 805. The NFPA 805 transition will either evaluate the non-compliant condition as acceptable due to detailed fire modeling and/or risk assessment or correct the condition via physical modification to the plant.
Procedure C.4-C has been revised to provide guidance to maintain the required NPSH for the CS and RHR pumps.
Until this issue is resolved, compensatory measures will be used to ensure the ability to safely shut down is maintained. Existing transient combustible material control procedures for the CR and CSR are adequate and use will be continued. Additional/new compensatory measures will include:
No hot work in the CR or CSR without shift manager approval. If hot work is required, SM and work supervisor must walk down the job and ensure adequate compensatory measures are taken.
The CSR detection and suppression systems and CR smoke detectors will remain functional. If the CSR detection or suppression systems are non-functional, thenU.S. NUCLEAR REGULATORY COMMISSION (9-2007)
LICENSEE EVENT REPORT (LER)
FACILITY NAME (1)
DOCKET (2)
LER NUMBER (6)
PAGE (3)
SEQUENTIAL REVISION Monticello Nuclear Generating Plant 05000263 YEAR NUMBER NUMBER 4 of 4 2009 001 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) station a continuous fire watch. If the CR smoke detectors are non-functional, station one fire watch to cover all of the affected back panel areas. Even though the CR is continuously manned, one fire watch will be stationed. This fire watch should carefully inspect all the back panel areas at least once every half hour.
Failed Component Identification None
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