ML20199G772
| ML20199G772 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/14/1999 |
| From: | Lyon C NRC (Affiliation Not Assigned) |
| To: | Richard Anderson NORTHERN STATES POWER CO. |
| References | |
| GL-88-20, TAC-M83644, NUDOCS 9901250014 | |
| Download: ML20199G772 (4) | |
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Anuary 14, 1999 g
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Mr. Rog:r O. And:rson, Director Nucl:ar En:rgy Engine: ring Northern States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT - REQUEST FOR ADDITIONAL INFORMATION (RAI) RELATED TO IPEEE REPORT (TAC NO.
M83644)
Dear Mr. Anderson:
Supplement 4 to Generic Letter (GL) 88-20, " Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," was issued in June 1991. The IPEEE is to specifically address seismic and internal fire events, and other external events such as high winds and floods. Supplement 5 to GL 88-20 was issued in September 1995 providing the staff position regarding modified procedures for satisfying the intent of the seismic IPEEE.
Northern States Power Company (NSP) submitted its IPEEE report on March 1,1995, addressing internal fires, high winds, floods, and other credible events. On November 20, 1995, NSP submitted Revision 1 of its IPEEE report. On November 3,1997, the NRC requested additionalinformation in order to complete the staff's review. NSP responded to the NRC request by letter dated May 29,1998.
Based on the staff's ongoing review of the Monticello IPEEE submittal and NSP's responses to the previous RAI, the staff is unable to conclude at this time that NSP has met the intent of Supplement 4 to GL 88-20. The staff has developed the enclosed questions related to the seismic and fire analyses of the IPEEE as a follow-up to the previous RAI. These questions were reviewed by a Senior Review Board consisting of NRC staff and consultants with probabilistic risk assessment expertise in external events.
The enclosed request was discussed with Mr. M. Voth of your staff on December 15,1998. A mutually agreeable target date of March 19,1999, for your response was established. If circumstances result in the need to revise the target date, please contact me at (301) 415-2296 at the earliest opportunity.
Sincerely, ORIGINAL SIGNED BY Carl F. Lyon, Project Manager 9901250014 990114 Rojed Nectorak W-1 PDR ADOCK 05000263 Division of Reactor Projects - Ill/IV P
PDR Office of Nuclear Reactor Regulation Docket No. 50-263
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d Mr. Roger O. Anderson, Director Monticello Nuclear Generating Plant Northern States Power Company cc:
J. E. Silberg, Esquire Kris Sanda, Commissioner Shaw, Pittman, Potts and Trowbridge Department of Public Service 2300 N Street, N. W.
121 Seventh Place East Washington DC 20037 Suite 200 St. Paul, Minnesota 55101-2145 U.S. Nuclear Regulatory Commission Resident Inspector's Office Adonis A. Nebiett 2807 W. County Road 75 Assistant Attomey General l
Monticello, Minnesota 55362 Office of the Attomey General 445 Minnesota Street Plant Manager Suite 900 l
Monticello Nuclear Generating Plant St. Paul, Minnesota 55101-2127 ATTN: Site Licensing Northern States Power Company 2807 West County Road 75 Monticello, Minnesota 55362-9637 Robert Nelson, President Minnesota Environmental Control Citizens Association (MECCA) 1051 South McKnight Road St. Paul, Minnesota 55119 Commissioner Minnesota Pollution Control Agency 520 Lafayette Road St. Paul, Minnesota 55119 Regional Administrator, Region ill U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Commissionerof Health Minnesota Department of Health 717 Delaware Street, S. E.
Minneapolis, Minnesota 55440 Darla Groshens, Auditor / Treasurer l
Wright County Government Center i
10 NW Second Street i
Buffalo, Minnesota 55313 l
January 1995 e
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l MONTICELLO IPEEE REPORT Supplemental Request for Additional Information Based on your individual plant examination of external events (IPEEE) submittal for Monticello and response to the request for additional information (RAl), the staff is unable to conclude at this time that you have met the intent of Supplement 4 to Generic Letter (GL) 88-20. Your l
responses to the following seismic and fire RAls (supplement to previous RAls) are necessary l
in order to complete our review.
l Seismic:
1 The staff requested additional information related to the IPEEE report in a letter dated November 3,1997. The NSP responses by letter dated May 29,1998, were not sufficiently complete to allow us to evaluate your submittal. In particular, the response to RAls 1 and 2 l
focused on the two success paths that were shown for the design-basis earthquake.
Insufficient information was provided for the single success path identified for beyond safe shutdown earthquake (SSE) events in Table A.2.5-2 of your IPEEE submittal, as amended by Table 1 in your May 29,1998 letter. Our concem centers on the question of the adequacy of I
this success path. We understand that high pressure coolant iniection (HPCI), reactor core isolation cooling (RCIC), Division 11 AC power, Division ll DC power (long term), Division il emergency diesel generator emergency service water (EDG ESW), Division ll residual heat removal system service water (RHR SW), and Division ll ESW would not be available at the 0.3g review level earthquake (RLE). In essence, this leaves only a single division of equipment for success above the SSE. In accordance with NUREG-1407," Procedural and Submittal Guidance for IPEEE for Severe Accident Vulnerabilities," and Electric Power Research Institute (EPRI) NP 6041, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin,"
success paths must be successful assuming a loss of offsite power (LOSP) and small loss-of-coolant accident (LOCA) for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the earthquake. In addition, success paths must be robust against non-seismic failures and human actions such that they may have a reasonable chance of success should these occur. In particular, NUREG-1407 states that non-seismic failures and human actions should "have low enough probabilities to not affect the seismic margin evaluation." It is not clear from your submittal and responses that non-seismic failures and human actions were adequately considered in constructing the beyond SSE success path.
Accordingly, please provide a detailed discussion of how the beyond SSE success path will assure adequate vessel level, core cooling, containment pressure control, and essential room cooling assuming LOSP and a small LOCA for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Your response should include two kinds of LOCAs, a stuck open safety relief valve (SRV) and a 1-inch instrument line break in containment. The response should also include a discussion of human actions -- first, l
assuming that no non-seismic failures occur, and then a discussion of human actions assuming i
a single random failure of one of the remaining two RHR pumps (or trains). For these human actions, please identify those actions that are not proceduralized (i.e., currently not included in the plant emergency operating procedures).
Enclosure l
s t
2 Fire:
In response to Fire RAI 5, it is stated " Analyses have demonstrated that ADS (automatic depressurization system) actuation will not lead to fuel cladding temperatures in excess of Appendix K limits, provided that the ADS panel and the core spray pump are actuated within 10 minutes." This statement indicates that there is a potential fcr high fuel cladding temperature in the case of a cable spreading room fire (likelihood of this event notwithstanding),
if the operators fail to take the proper actions within 10 minutes of ADS actuation. In the case of a cable spreading room fire, a large number of control and instrumentation circuits may be affected by the fire, leading to a large number of alarms, actuation or de-actuation of various equipment, and anomalies on the control board that have not been experienced before. Under such conditions, the operators may get confused, and may take a long time to properly diagnose the conditions of various systems, and take corrective actions.
The discussion regarding ADS actuation from a cable spreading room fire states that multiple hot shorts have to take place for the ADS to actuate, and therefore the conditional probabilities of these failures can be multiplied (assuming independence among the hot shorts), which leads to a small overall probability of failure. The assumption that the hot shorts are independent may not be valid, since circuits and cables are often similar, and if one fails in a certain mode, the probability of others failing in the same mode may be very high, especially when considering, for example, hot shorts of multiple circuits in the same cabinet, it must also be added that the assumption regarding fire spread to the outside of an electrical cabinet is not supported for all cases by the Sandia tests [1]. In the fire IPEEE, the electrical cabinets have been treated as " virtual rooms." This assumption is not considered as valid.
Please provide additional information on mitigative actions that the operators have to take to avert high cladding temperature for the above scenario. Please provide the basis for the assumption that the operators will clearly understand the exact nature of the situation (including recognizing that the fire hhs occurred and that the ADS has been actuated), given the possibility that a large number of control and instrumentation circuits may have been affected, resulting in a large number of alarms and equipment actuations (both actual and spurious). Please identify which valves and systems would be and which would not be affected by the fire. If there is a procedure that specifically addresses cable spreading room fires, please provide a summary of the procedure. Please provide a summary of the analysis that demonstrates how the operators are capable of controlling the plant when the ADS is actuated from a cable spreading room fire within the 10-minute time window. Please provide a description of the analysis without using the assumptions regarding cabinets as " virtual rooms" and independence of hot shorts.
Reference:
1.
J. M. Chavez, "An Experimental Investigation of Internally ignited Fires in Nuclear Power Plant Control Cabinets: Part 1: Cabinet Effect Tests," Sandia National Laboratories, NUREG/CR-4527, November 1986.