05000263/LER-2011-001, For Monticello Nuclear Generating Plant, Reactor Vessel Overfill in Appendix R Scenario
| ML110180086 | |
| Person / Time | |
|---|---|
| Site: | Monticello  (DPR-022) |
| Issue date: | 01/14/2011 |
| From: | O'Connor T Xcel Energy, Northern States Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-11-001 LER 11-001-00 | |
| Download: ML110180086 (4) | |
| 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)(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 |
| 2632011001R00 - NRC Website | |
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January 14, 2011 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed License No. DPR-22 LER 201 1-001, Reactor Vessel Overfill in Appendix R Scenario The Licensee Event Report (LER) for this occurrence is attached.
Summary of Commitments Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 L-MT-11-001 10 CFR 50.73 r i r f y J. OJConnor Si ice President, Monticello Nuclear Generating Plant Northern States Power - Minnesota Enclosure cc:
Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC
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 Records and FOlAlPrivacy Service Branch (T-5 F53), U.S. Nuclear Regulatory Commlssion, Washington, DC 20555-0001, or by internet e-mail to LICENSEE EVENT REPORT (LER) lnfocoiiects.resource@nrc.gov. and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-OW), Office of Management and (See reverse for required number of Budget. Washington, DC 20503. If a means used to impose an information C]
20.2203(a)(3)(i)
C]
50.73(a)(2)(i)(C) 50.73(a)(2)(vii)
C]
20.2203(a)(3)(ii)
C]
50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A)
C]
20.2203(a)(4) 50.73(a)(2)(ii)(B)
C]
50.73(a)(2)(viii)(B)
C]
50.73(a)(2)(iii) 50.73(a)(2)(ix)(A)
C]
20.2203(a)(2)(ii) 50.36(c)(l)(ii)(A)
C]
50.73(a)(2)(iv)(A) 50.73(a)(2)(x)
C]
20.2203(a)(2)(iii)
C]
50.36(~)(2)
C]
50.73(a)(2)(v)(A) 73.71 (a)(4)
C]
20.2203(a)(2)(iv)
C]
50,46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71 (a)(5)
During the performance of a fire protection assessment, a determination was made that the fire protection safe shut down analysis does not address a postulated reactor vessel overfill event.
In a postulated fire event that required the evacuation of the Control Room with a loss of offsite power, High Pressure Coolant lnjection and Reactor Core Isolation Cooling pumps would start if the low reactor water level setpoint is reached. For this fire, damage could result in the failure of the high reactor water level trip circuit for the High Pressure Coolant lnjection and Reactor Core Isolation Cooling systems. This could result in a reactor vessel overfill.
Compensatory measures for this issue have been implemented.
NRC FORM 366 (10-2010)
Monticello Nuclear Generating Plant
EVENT DESCRIPTION
On 12 November, 2010, the site determined that a postulated reactor vessel overfill scenario exists which had not been analyzed under the Appendix R Fire Protection Program. In the scenario, High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems start on low-low reactor water level and fail to trip on high reactor water level due to fire damage. This could result in a reactor vessel overfill as discussed below.
For a fire requiring evacuation of the Control Room, the post fire safe shutdown is accomplished from the Alternate Shut Down System (ASDS). For the postulated scenario, a fire in the Control Room or Cable Spreading Room would cause Operations personnel to evacuate the Control Room and proceed to the ASDS panel located in the Emergency Filtration Train (EFT) Building. The scenario requires the assumption of an unlikely loss of offsite power, with consequential decrease of water inventory in the reactor. When reactor water level reaches the low-low reactor water level setpoint, the main steam isolation valves close and HPCl and RCIC pumps start.
Safety related reactor water level switches LS-2-3-672E(LIS) and LS-2-3-672F(LIS) provide a high reactor water level trip signal. When a high reactor water level condition occurs both switches actuate (2-out-of-2 logic) to trip HPCl and RCIC. Fire damage to this circuit could prevent the high reactor water level trip. If HPCIIRCIC fail to trip on high reactor water level, then the reactor vessel would continue to fill until sufficient water fills the HPCl and RCIC steam lines to stall the HPCl and RCIC pumps.
After arriving at the ASDS panel, Operations personnel could procedurally initiate a reactor vessel blow down by manual operation of the safety relief valves (SRV) to allow for low pressure reactor water inventory makeup and decay heat removal. The HPCl and RCIC steam supply lines and SRVs connect to the main steam lines at the same elevation. When the SRVs are manually opened from the ASDS panel, the valves may be subjected to high pressure steamlwater flow.
I The SRVs and their associated tailpipes have been analyzed preliminarily and the loads found acceptable.
EVENT ANALYSIS
The event is reportable to the NRC under 10 CFR 50.73(a)(2)(ii)(B) - Degraded or Unanalyzed Condition. The site reported the event on November 12, 2010.
This event is not a Safety System Functional Failure.
I I U.S. NUCLEAR REGULATORY COMMISSION (10-201 0)
The Monticello risk assessment group reviewed the event for risk impact. Risk of core damage and large early release are not significantly impacted by effects on the SRV function during inadvertent overfill of the reactor vessel resulting from failure of the HPCI/RCIC high reactor water level trip due to a fire that defeats the trip logic. The SRVs are not significantly impacted from a vessel overfill event as they are capable of performing their intended function in both the safety mode as well as the depressurization mode. Additionally, industry operating experience and preliminary MNGP plant specific analysis support a conclusion that the SRV tailpipes will remain intact following SRV lifts while subject to liquid and/or two phase flow.
I Based on the above, the health and safety of the public have not been affected.
I This LER will be updated if non-conservative changes to the safety significance discussion above are required. Compensatory measures will remain in place until corrective actions are completed.
CAUSE
The cause of this event was that previous Appendix R analyses failed to consider an insufficient consideration of HPCl and RCIC automatic initiation and failure to trip.
CORRECTIVE ACTION
Corrective actions are being tracked in the Corrective Action Program.
Compensatory actions, in accordance with the Fire protection Program, have been taken in those areas where a fire could cause the postulated scenario.
A memorandum detailing the postulated scenario has been issued to Operations personnel as a briefing for this condition.
Other actions:
i. A site specific evaluation of this condition will be performed as part of Appendix R compliance as well as Regulatory Guide 1.189 compliance.
ii. Based on the completion of formal SRV tailpipe analysis, HPCI and RCIC will be evaluated as necessary for modifications.
I n M I L A R + i. V E N T S There have been no similar events in the last 3 years.
OTHER A discussion was held with NRC Region Ill, extending the due date for this LER to 14 January 201 1 (three days).