ML20216C753

Informs That Staff Unable to Conclude at Present Time That Wisconsin Electric Has Met Intent of GL 88-20,Supplement 4. RAI Re Seismic & Fire Analyses of IPEEE Encl

ML20216C753
Person / Time
Site:	Point Beach
Issue date:	03/10/1998
From:	Gundrum L NRC (Affiliation Not Assigned)
To:	Grigg R WISCONSIN ELECTRIC POWER CO.
References
GL-88-20, TAC-M83661, TAC-M83662, NUDOCS 9803160143
Download: ML20216C753 (6)

Text

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I March 10, 1998 Mr. Richard R. Grigg i Chief Nuclear Officer )

Wisconsin Electric Power Company 231 West Michigan Street, Poom P379 Milwaukee, WI 53201 i

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SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REQUEST FOR I ADDITIONAL INFORMATION RE: INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (TAC NOS. M83661 AND M83662) {

Dear Mr. Gngg:

i Based on the ongoing review of the Point Beach Nuclear Plant's Individual Plant

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Examination of External Events (IPEEE) June 30,1995, submittal and Wisconsin Electric's I

Augest 15,1996, response to previous requests for additional information (RAls), the staff is j unable to conclude at this time that Wisconsin Electric has met the intent of Generic l

Letter 88-20, Supplement 4. Therefore, we have developed the enclosed RAI related to the seismic and fire analyses of the lPEEE. Fiease provide your response to the enclosed RAI ,

I within 60 days of receipt of this letter.

Sincerely, i

ORIGINAL SIGNED BY l Linda L. Gundrum, Project Manager Project D rectorate ill-1 Division of Reactor Projects - til/IV l Office of Nuclear Reactor Regulation g Docket Nos. 50-266 and 50-301

Enclosure:

RAI i cc w/ encl: See next page l

DISTRIBUTION: i Docket File (50-266, 50-301) PUBLIC j PDill-1 Reading E. Adensam (EGA1) l A. Rubin, T-10 E50 J. McCormick-Barger, Rlli  !

DOCUMENT NAME: G:\WPDOCS\PTE NCH\PTB83661.RAI To receive a copy of this document, ind6cate in the box "C' "E" e Copy with .;W,.. .;/ n,;usure Y = No copy

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L"~ Mr. Richard R. Grigg Point Beach Nuclear Pit.nt

. Wisconsin Electric Power Company Unit Nos.1 and 2 cc:

Emest L. Blake, Jr.

Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W.

Washington, DC 20037.

Mr. Scott A. Patulski Vice President Point Beach Nuclear Plant Wisconsin Electric Power Company 6610 Nuclear Road '

Two Rivers, Wisconsin 54241 Mr. Ken Duveneck Town Chairman Town of Two Creeks 13017 State Highway 42 Mishicot, Wisconsin 54228 Chairman Public Service Commission of Wisconsin P.O. Box 7854 Madison, Wisconsin 53707-7854 Regional Administrator, Region lil U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Resident inspector's Office U.S. Nuclear Regulatory Commission ,

6612 Nuclear Road 1 Two Rivers, Wisconsin 54241

? N Sarah Jenkins Elec$c Division Public Service Commission of Wisconsin P.O. Box 7854 Madison, Wisconsin 53707-7854 j

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REQUEST FOR ADDITIONAL INFORMATION Seismic RAls

1. Based on your response to seismic RAI No. 7 for the Point Beach plant, your reported high confidence of low probability of failure (HCLPF) capacity for the plant does not exceed even the safe shutdown earthquake (SSE) spectrum for vibration frequencies

. less than about 4 Hz. The reported HCLPF spectrum is also lower, over all frequencies, than the review level earthquake (RLE) spectrem defined by a NUREG/CR-0098 median spectral shape anchored to a PGA [ peak ground acceleration) value of 0.3g. Although safety enhancements were planned that resulted from the USl A-46 and IPE programs, no plant improvements / resolutions have been proposed for several remaining components that have low seismic capacities, and which cause the plant HCLPF to be significantly lower than the RLE. (Examples include anchorage concerns, interaction concems, potential failures of block walls, and potential failures of flat-bottomed tanks.)

Please provide a discussion on why the plant is considered to have adequate resistance to potential seismically induced severe accidents.

2. The submittal reports that release category G was found to be associated with nearly the entire seismic core damage frequency (CDF). Hence, the conditional probability of large, early release is assessed as being nearly 100%, given a seismically induced core damage. This finding re;ults from the assumption that most sequences result in failure of the automatic containment isolation function. In the containment vulnerability assessment, it effectively invalidates this finding by assuming that containment will be

!solated manually by operators at least 90% of the time in the event of a core damage where automatic isolation is failed; the basis for this assumption is not provided, yet it is I used to reduce the large release risk by an order of magnitude. Despite the arguments in the IPEEE, it seems that failure of automatic containment isolation may be a potential seismic vulnerability.

Please provide the fragility assessment for seismic failure of the automat:c containment isolation system. On what basis is the 0.10 human error probability (HEP) for manual containment isolation justified? Is the basis consistent with that used in developing operator error fragilities for the Level-1 analysis? At what locations do the required operations for manual isolation take place? How many operators are required to manually isolate the containment? Please provide the timing when the acCons are required. Will all isolation valves be accessible / operable following an earthquake, or will potential seismic failures possibly lead to some components being inaccessible and/or inoperable?

3. The relay chatter evaluation has not been fully expanded beyond the scope of USl A-46 in order to address all IPEEE-only systems. The submittal states that relay chatter is acceptable for the unanalyzed systems on the grounds that operators could reset the relays. In this regard, please report what fraction of the seismic CDF is contributed by relay chatter. Please provide the seismic CDF for the case where no operator recovery from relay chatter is assumed.

ENCLOSURE

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4. The surrogate element has been reported as a dominant contributor to seismic CDF, which means that neither dominant sequences nor important contributors to seismic CDF are fully understood. Given this concem is associated with the important objective of performing an IPEEE, please discuss how insights regarding dominant sequences l and contributors would change if the surrogate element was not used or the screenmg -

level was much higher (e.g.,0.5g). Please discuss in detail the bases and any I additional calculation performed to support your conclusions.

5. The August 15,1996, response to seismic RAI No. 6 is incomplete with respect to seismic inadvertent actuation of fire protection system and seismically induced loss of fire suppression system capability. The evaluation for inadvertent actuation does not describe walkdown findings. The evaluation for seismic loss of fire suppression was limited to potential interactions with safety equipment, and did not consider loss of fire suppression capability itself; again, no walkdown findings were mentioned. Some examples of items found in past studies include (but are not limited to): I

• Unanchored CO, tanks or bottles

. Sprinkler standoffs penetrating suspended ceilings

. Fire pumps unanchored or on vibration isolation mounts

. Mercury or " bad actor" relays in fire protection system (FPS) actuation circuitry i

. Weak or unanchored 480V or 600V (nonsafety-related) electrical cabinets in close proximity to essential safety equipment (i.e., as potential fire sources)

• Use of cast iron fire mains to provide fire water to fire pumps.

i NUREG-1407 suggests a walkdown as a means of identifying any such items. Please provide the related results of your seismic-fire interadion study, Provide guidelines given to walkdown personnel for evaluating these issues (if they exist). Please identify equipment in fire suppression systems that may be damaged due to the review level earthquake and discuss your resolution of these items, if any.

Fire RAls

1. In your response to Fire RAI No. 2, on the possibility of a loss of offsite power (LOOP) due to a fire within the plant, you stated that "no fire compartments were identified that would disable all offsite power." It is not clear that your analysts have considered LOOP resulting from fires in the control room and cable spreading rooms. It is common for the controls of the breakers associated with offsite and Class 1E power to be located next to one another in the same panel. A small control panel fire may cause control circuit failures that would lead to these breakers failing open (thus, leading to LOOP). Similar failures may occur from a fire inside the cable spreading room, and other rooms where the cables or equipment associated with the control circuits of these breakers are present.

This omission could potentially be significant, therefore, potential vulnerabilities may have been overlooked as a result of this assumption.

. , 3 Please provide a detailed discussion on how the control circuits were analyzed for those breakers that can potentially, by their failures, cause a LOOP. Discuss the location of the cables associated with these circuits and provide the technical basis for not considering LOOP for fires in these areas.

2. The cable spreading room (compartment 318) and electrical equipment rooms (compartments 245 and 246) contain oil-filled transformers. There is a potential for a large fire from severe transformer, failure. Your submittal and the response to RAI NO. 21 did not address the potential for a large fire or explosion in a transformer. You have subdivided the cable spreading room into smaller areas for the detailed fire analysis. This applies only to small fires.- A transformer fire or explosion may have sufficient energy to jeopardize the integrity of fire boundaries. Potential vulnerabilities may have been overlooked from the omission of fire scenarios involving an energetic rupture of a transformer or an extremely large fire in this room. if such failures were :

considered, what would be the impact on the IPEEE results? If a boundary is breached because of a very severe fire in the cable spreading room, what would be the additional impact on safe shutdown equipment? Please provide a discussion on how the remote shutdown capability would be affected (either through direct impact, through blocking passages needed for operator access, or through affecting communication among operators).

3. For several fire scenarios, you have used optimistically small times for detection and suppression. Specifically, for compartment 156 (MCC [ motor control center) room),

Scenario 2, the time for automatic detection and suppression is 65 seconds. Such a short detection and suppression time is overly optimistic. You cite FIVE formulations in your response to RAI No. 34 as the basis for the timing. It appears that you have employed FIVE without confirming the reasonableness of the results. For this scenario, Pcci (probability of critical combustible loading damage) was assumed to be 2X10-2, which led to a small CDF. For this specific scenario, if the timing was modified, the CDF would increase by at least one order of magnitude. Similar conditions exist for Compartment 166 (MCC room), Scenarios 2 and 3, and Compartment 318 (cable '

spreading room), Scenario 3.

Please provide a reanalysis of the time periods using appropriate data, assumptions, and formulations, for fire scenarios associated with the compartments mentioned above.

1 Please provide a copy of the computations,1 sic data, and assumptions supporting these time periods. Also, please provide the analytical results of CDFs and associated '

fire scenarios and accident sequences using these more realistic detection and saopression times.

4. In response to RAI No. 3, you stated that hot shorts had been considered in the analysis since Appendix R ir. formation had been used, which included hot shorts. Theoretically speaking, if one conducted a comprehensive analysis of potential failure modes of equipment and cables from a fire and available safe shutdown paths given such failures, all relevant spurious actuations and other failures would be identified properly.

However, from your response to RAI No. 4, it appears that you did not consider, in your s Appendix R effort, the possibility of a loss-of-co >lant accide: t (LOCA) or inadvertent w

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steam dump from a fire event. This oversight indicates that the hot short (or spurious actuation) phenomenon has not been properly addressed in the Appendix R analysis of this plant.

Please provide either a detailed description of the analysis done as part of the Appendix R effort to address the hot short issues, and specifically the possibility of occurrence of a LOCA, or a comparison of the Appendix R safe shutdown and IPE models demonstrating that the initiating events identified in the IPE model have been properly addressed in the Appendix R analysis.

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