ML20217G665

Forwards RAI Re Seismic & Fire Analyses of IPEEE Submittal. Info Requested within 60 Days of Receipt of Ltr

ML20217G665
Person / Time
Site:	Wolf Creek
Issue date:	10/08/1997
From:	Thomas K NRC (Affiliation Not Assigned)
To:	Maynard O WOLF CREEK NUCLEAR OPERATING CORP.
References
TAC-M83696, NUDOCS 9710140123
Download: ML20217G665 (6)

Text

a* i October 8, 1997  :

i Mr. Otto L. Maynard a President and Chief Executive Off cer  !

Wolf Creek Nuclear Operating Corporation l Post Office Box 411 Burlington, Kansas 66839 i

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE WOLF-CREEK INDIVIDUAL ~ PLANT EXAMINAT10N 0F EXTERNAL EVEN15 SUBMITTAL (TAC NO. i M83696)  ;

Dear Mr. Maynard:

= Based on the NRC's ongoing review of the Wolf Creek Individual Plant  !

Examination of External Events (IPEEE), the staff has determined that .

additional information is needed related to the seismic and fire analyses to i complete the review of the IPEEE submittal. -The information needed to-  ;

complete the' review is contained in the enclosure,  !

i To assist the staff in meeting its review schedule, please respond to this  !

request in writing within 60 days of recei t of this letter. If you have any l

_ questions, please contact me at (301) 415 362, '

Sincerely. .

i Original Signed By ,

Kristine M. Thomas, Project Manager Project Directorate TV 2-Division of Reactor Projects Ill/IV ,

Office of Nuclear Reactor Regulation Docket No. 50 4B2 )lSTRIBUT ON '

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%,,....+ October 8, 1997 Mr. Otto L. Maynard President and Chief Executive Officer j Wolf Creek Nuclear Operating Corporation '

Post Office Box 411 Burlington, Kansas 66B39

SUBJECT:

RE0 VEST FOR ADDITIONAL INFORMATION REGARDING THE WOLF CREEK INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS SUBMITTAL (TAC ND, MB3696) r

Dear Mr. Maynard:

I i Based on the NRC's ongoing review of the Wolf Creek Individual Plant l Examination of External Events (IPEEE), the staff has determined that additional information is needed related to the seismic and fire analyses to <

complete the review of the IPEEE submittal. The information needed to ,

complete the review is contained in the enclosure, 4

To assist the staff in meeting its review schedule, please respond to this request in writing within 60 days of receipt of this letter, If you have any questions, please contact me at (301) 415 1362, Sincerely, NITM%

Kristine M. Thomas, Project Manager Project Directorate IV 2 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation Docket No. 50 4B2

Enclosure:

Request for Additional Information cc w/ enc 1: See next page

. i Mr. Otto L. Maynard -

2- October 8,1997 i cc w/att: l Jay Silberg. Esq. Chief Operating Officer i Shaw Pittman, Potts & Trowbridge Wolf Creek Nuclear Operating Corporation 2300 N Street. NW P. O. Box 411 ,

Washington D.C. 20037 Burlington Kansas 66839 i i

Regional Administrator. Region IV Supervisor Licensing  !

U.S. Nuclear Regulatory Comission Wolf Creek Nuclear Operating Corporation  :

611 Ryan Plaza Drive. Suite 1000 P.O. Box 411 <

Arlington Texas 76011 Burlington, Kansas 66839

+

Senior Resident inspector U.S. Nuclear Regulatory Comission l U.S. Nuclear Regulatory Comission Resident inspectors Office i P. 0. Box 311 8201 NRC Road Burlington, Kansas 66839 Steedman. Missouri 65077 1032 Chief Engineer '

Utilities Division 6 Kansas Corporation Comission 1500 SW Arrowhead Road Topeka, Kansas 66604 4027 Office of the Governor State of Kansas

• Topeka, Kansas 66612 ,

Attorney General Judicial Center .

301 S.W. 10th -

2nd Floor Topeka, Kansas 66612

- County Clerk Coffey County Courthouse Burlington Kansas 66839 ,

Vick L. Cooper, Chief

- Radiation Control Program Kansas Department of Health and Environment

- Bureau of Air and Radiation Forbes Field Building 283 Topeka. Kansas 66620 i

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l* REQUEST FOR ADDITIONAL INFORMATION

]NDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GEkERATING STATION DOCKET NO. 50 482 A. Seismic

1. NUREG 1407 (Section 3.2.5.8. page 14) suggests that non seismic failures and human actions are to be clearly identified and the assessment assure that they are low enough in probability to not compromise the seismic-margins. Typically, some of the more risk significant human actions for PWRs are: connect service water as an auxiliary feedwater source, feed and bleed, control steam generator relief valves for steam generator cooling. RCS cooldown and depressurization to use the RHR system. reduce containment spray pum) flow, room cooling recovery, and establish cold leg recirculation. T1e submittal does not provide sufficient assurance that human actions can be accomplished after an SME. Per NUREG 1407 show the analysis and logic used to verify that human actions do not control the probability of losing a success path after an earthquake.

For each human action provide the location and the timing of performance. In addition, explain what affect an earthquake would have on perfornance of these actions.

2. The submittal discussion regarding seismic degradation of fire protection eouipment is limited to the interaction of piping in safety related areas. The evaluation should also include an examination of potential loss of fire protection capability itself, due=to seismic events. Examples of items found in past studies include (but are not limited to):

Fire protection pumps and tanks CO2 tanks and bottles sprinkler standoffs (either penetratirq suspended ceilings or simply hanging from the ceiling) use of cast iron fire mains to provide fire water to fire pumps Provide the location of each of these items (if at the WCGS). how they are anchored and/or supported, and whether or not they are seismically qualified.

3.- EPRI NP 6041 (Step 3 of Section 2) recommends development of generic anchorage capacities as part of the preparatory work, it states that "it is impossible to make judgments on the adequacy of seismic

. ruggedness without an understanding of the seismic demand corresponding

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to the HCLPF level and some measure of equipment anchorage capacit .

The submittal anchorages in only amentions theSection few places (e.g., performance 3.5.6. , of afor" bounding)evalut

a. Provide an example of a bounding anchorage evaluation, and

b. For each equipment category of Section 3.5 of the submittal, provide the guidance used by the SRT and the WCGS IPEEE to screen out anchorages. procedure it used during the

4. The evaluation of USl A 45 states that all "all components and structures relatin design basis SSE".g to decay heat removal are adequate 'or the plant Provide an evaluation of USI A 45 tor the 0.39 RLE as requested in NUREG 1407, including, but not limited to, the basis for i screening out the RWST, pond and service water system, l

5. The submittal notes that horizontal SME in structure res)onse spectra '

demand exceeds the design b3 sis spectra demand by as-muc1 as_50%. '

Determine and provide seismic capacities for those equipment items that have demands on the order of 40% to 50% higher than the design basis.-in order to verify the judgments made during the walkdown screening.

6. The submittal notes that three items could not be assigned capacities more than 0.3g PGA by the SRT. The submittal also states that if it were not for these " indeterminate" items, the plant HCLPF capacity would be greater than 0.3g PGA, Among these items are:

fcur 60 cell batteries and racks because of spacing between the batteries and rails twelve LSELS/ESFAS cabinets because they are not bolted together The submittal also asserts that these items could be shown to have a larger seismic capacity than the SSE if detailed seismic margin assessments would be performed. Consistent with the guidance of NUREG 1407 for focused scope plants and EPRI NP 6041 for performance of an EPRI seismic margin study, determine and provide the seismic capacity of these components.

B. ELCit

1. Section 7.1.2 states that hot shorts are significant for a number of fire areas. Provide the fire areas for which hot shorts are significant. Describe the method used to account for hot shorts in the PRA of unscreened compartments. This description should include how cabinets and components were reviewed to determine the following: (1) if a hot short could occur (2) the probability of hot shorts, and (3) the way in which the core damage frequency associated with hot shorts was developed.

e l 2. The treatment of propagation of fire from cabinets was inconsistent.

Propagation of fire from cabinets was assumed to not occur in the-control room, whereas it could occur in all other cabinets of the plant.

It is noted that the study distinguished among open cabinets, sealed cabinets in high traffic areas and sealed cabinets in low traffic areas.

The justification for the propagation of fire from sealed cabinets of 0,69 is clearly stated. However, the derivation of 0.15, as the propagation probability from sealed cabinets in high-traffic areas, is not clear.

Describe the testing, inspection and/or surveillance program in place at WCGS that would ensure that cabinet seals are (a) always in place in the cabinets for which credit was taken in the IPEEE, and (b) effective in preventing fire propagation. Provide a derivation with explanation of the probability of 0.15 for sealed cabinets in high traffic areas.

Provide a derivation with explanation for the assumptions that fires will not propagate from cabinets in the control room.

3. The general turbine area and the two radiation access areas were 4 screened out by including the automt. tic suppression system unavailability as a multiplication factor on the fire ignition frequency and conditional-core damage probabiHty. FIVE allows this method of screening if it can be demonstrated that the suppression system is code compliant, and the time of extinguishment is less than the time of damage.

For these areas - provide justification that the ~ suppression systems are installed in a manner that complies with all applicable NFPA standards, and that the suppression systems would be effective to prevent damage to cables, include in the res damage cables or equipment,ponse theto and the time damage criteria extinguish the used, fire. the time to

4. A review of the submittal with respect to the PRA of unscreened compartments reveals that it was generally assumed that fires that propagate out of cabinets would damage cables and equipment within 10 to 20 feet of the cabinet. It was also assumed that halon suppression would be effective unless a system failure occurred in preventing damage

-to equipment outside this radius. This implicitly assumes that the systems are designed, i.nstalled and maintained in a code compliant manner.

Provide justification that the suppression systems, taken credit for in the PRA of unscreened compartments are designed, installed and maintained in accordance with appropriate industry standards, such as those published by the NFPA.

C. [LFQ No additional information required.

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