ML20199J499

Forwards Request for Addl Info Re Resolution of Unresolved Safety Issue A-46 Program at Plant.Requests Response within 60 Days of Date of Ltr

ML20199J499
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	01/29/1998
From:	Jabbour K NRC (Affiliation Not Assigned)
To:	Reid D VERMONT YANKEE NUCLEAR POWER CORP.
References
REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-87-02, GL-87-2, TAC-M69490, NUDOCS 9802050330
Download: ML20199J499 (8)

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i January- 29, 1998 Mr. Donald A. Reid

-Senior Vice President, Operations Vermont Yankee Nuclear Power Corporation 185 Old Ferry Road Brattleboro, VT 05301

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE RESOLUTION OF UNRESOLVED SAFETY ISSUE A VERMONT YANKEE NUCLEAR POWER STATION (TAC N0, H59490)

Dear Mr. Reid:

By letter of July 1.1996 Vermont Yankee Nuclear Power Corporation provided the plant specific summary report in accordance with its commitment related to Generic Letter 87-02 regarding the resolution of the Unresolved Safety Issue (USI) A 46 program at Vermont Yankee Nuclear Power Station.

The NRC staff has reviewed the summary report, and, based on its review, finds that additional information, as outlined in the enclosure, is needed before we can complete our review of the report.

Please provide your responses within 60 days from the date of this letter. If you have any questions regarding this matter, please contact me at (301) 415-1496.

Sincerely, Original signed by KahtanN.Jabbour,SeniorProjectManager Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-271

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D. Reid: Vermont Yankee Nuclear Power Station CC:

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. Regional Administrator, Region.I Chii:f. Safety Unit i U. S. Nuclear Regulatory Commission Office of the Attorney General I 475 Allendale Road One Ashburton Place, 19th Floor-King of Prussia, PA 19406 Boston, MA 02108 1

Mr. David R. Lewis Mr. Peter LaPorte, Director Shaw, Pittman, Potts & Trowbridge ATTN: James Muckerheide 2300 N Street, N.W. Massachusetts Emergency Management Washington, DC 20037-1128 Agency 400 Worcester Rd.

-Mr. Richard P. Sedano. Commissioner P.O. Box 1496 Vermont Department of Public Service Framingham, MA 01701-0317 i 120 State Street, 3rd Floor Montpelier, VT 05602 Mr. Raymond N. McCanaless Vermont Division of Occupational i Public Service Board and Radiological Health i State of Vermont Administration Building 120 State Street Montpelier, VT 05602 Montpelier, VT- 05602 )

i Hs. Deborah B Katz Chairman, Board of Select.11en Box 83 Town of Vernon Shellburne Falls, MA 01370 P.O. Box 116 Vernon, VT 05354-0116 Mr. Gautam Sen Licensing Manager Mr. Richard E. McCullough Vermont Yankee Nuclear Power Operating Experience Coordinator Corporation Vermont Yankee Nuclear Power Station 185 Old Ferry Road

. P.O. Box 157 Brattleboro, VT 05301 Governor Hunt Road Vernon, VT 05354- Resident Inspector Vermont Yankee Nuclear Power Station G. Dana Bisbee. Esq. U. S. Nuclear _ RegulLtory Commission Deputy Attorney General P.O. Box 176 33 Capitol Street Vernon, VT 05354 Concord, NH 03301-6937 Mr. Jonathan H. Block, Esq.

Main Street

-P.O. Box 566 Putney, VT 05346-0566

REOUEST FOR ADDITIONAL INFORMATION VERMONT YANKEE NUCLEAR POWER STATION UNRESOLVED SAFETY ISSUE A-46 .

Docket No. 50-271

Reference:

Vermont Yankee letter, J. K. Thayer, to Document Control Desk (NRC).

dated July-1, 1996-(BVY 96-86), " Vermont Ytokee Summary Report for Resolution of USI A-46"

1. The conservatism of the seismi: margin methodology as described in EPRI NP-6041 report is not certain at this time. Its rpplication is, therefore, not endorsed by the NRC'for the analysis of safety-rflated systems and components, including the resolution of USI A-46 mechanical, electrical, and structural component outliers. You are requested to reevalJate the portion of your USI A-46 program, where the above methodology has been used, to ensure that all the outliers identified in Vermont Yankee A-46 program, are resolved by using methodologies that are consistent with the plant licensing basis or other approaches acceptable to the staff.

2. Section 9 provides a general discussion on the resolution of equipment and relay outliers identified in Table 5-4 and Table 8-1 of the Seismic Evaluation Report (attached to the reference stated above). -However, the discussion in Section 9 is very vague regarding the methodologies to be used and the status of this activity in the program. You are requested to provide the current status, and the percentage of completion, of the outlier resolution in your program. You

- are also requested to provide specific information regarding the approaches and methodologies used in the resolution of a representative sample of outliers listed in these tables.

3. The equipment outlier list provided in Table 5-4 does not include equipment

-located at elevations 303.00 ft. and 318.75 ft. of the reactor building and at elevation 300.00 ft. of the turbine building, where-spectral exceedances of in-

= structure response spectra (IRS) over 1.5 times the Bounding Spectrum were-indicated in Table-5-2. You are requested to clarify what equipment items are involved in these building elevations and how their seismic adequacy have been verified.

4. You stated in the Seismic Evaluation Report that the current schedule for completion of the outliers is prior to startup from_ the 1999 refueling outage.

You ara requested to provide the basis of detennining the operability of the affected systems while-a n'aber of safety-related components in the safe Enclosure

2 shutdown path have been identified as outliers: thus rendering their seismic adequacy questionable and their conformance to the licensing basis uncertain.

5. In Section 8.4 of the Seismic Evaluation Report, you stated that the list of essential relays does not include any low ruggedness (bad actor) relays. You are requested to explain where the " bad actor" relays are being listed. For all

" bad actor" relays identified, also explain how their contact chatter was determined to be. acceptable, as stated in the report.

6. Regarding the adequacy of seismic demand determination (ground spectra and in-structure response spectra), you are requested to address the following on a plant specific basis.

Referrina to the in-structure resoonse soectra orovided in your 120-day-resoonse to the NRC's reauest in Sunalement No 1 to Generic Letter (GL ) 97-02. dated May 22.

1992 the followino information is reauested.

a. Identify structure (s) which have in structure response spectra (5% critical damping) for elevations within 40 feet above the effective grade, which are higher in amplitude than 1.5 times the Seismic Qualification Utility Group (500G) Bounding Spectrum,

b. With respect to the comparison of equipment seismic capacity and seismic demand, indicate which mdthod in Table 4-1 of Generic Implementation Procedure (GIP-2) was used to evaluate the seismic adequacy for equipment installed on the corresponding floors in the structure (s) identified in Item (a) above. If you have elected to use method A in Table 4-1 of the

. GIP 2, provide a technical justification for not using the in-structure response spectra provic'1 in your 120 day response. It appears that some A-46 licensees are making an incorrect comparison between their plant's safe shutdown earthquake (SSE) ground motion response spectrum and the SOUG Bounding Spectrum. The SSE ground motion response spectrum for most nuclear power plants is defined as the plant foundation level. The 50VG Bounding Spectrum is defined at the free field ground surface. For plants located at deep soil er rock sites, there may not be a significant difference between the ground motion amplitudes at the foundation level and those at the ground surface. However, for sites where a structure is 4

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founded on shallow soil, the amplification of the ground motion from the foundation level to the ground surface may be significant.

c. For the structure (s) identified in Item (a) above, provide the in-structure response spectra designated according to the height above the effective grade. If the in structure response spectra identified in the 120-day response to Supplement No.1 to GL 87-02 were not used, provide the response spectra that were actually used to verify the seismic adequacy of equipment within the structures identified in Item (a) above. Also, provide a comparison of these spectra to 1.5 times the Bounding Spectrum.

7. lne NRC staff has concerns about the way the A-46 cable trays and conduit raceways issue was being disposed of by licensees. The staff issued requests for additional information (RAI) to several licensees on this issue.

SOUG responded instead of the licensees because 50VG considered the RAI to be generic in nature. The staff issued a subsequent RAI to SOUG as a follow up to their response. However, the staff found that the correspondence with SOUG did not achieve the intended results in that they did not address the technical concerns of the staff. Therefore, we request that you address the following areas of staff concerns.

1he GIP procedure recommended performing what is called a limited analytic evaluation for selected raceways and cable trays. The procedure further recommended that when a certain cable tray system can be judged to be ductile >

and if the vertical load capacity of the anchorage can be established by a load check using three times the dead weight, no further evaluation is needed to demonstrate lateral resistance to vibration from earthquakes. .The staff has concerns with the manner in which these simplified GIP criteria were implemented at ycur plant.

The GIP procedure eliminates horizontal force evaluations by invoking ductility. However, all the so called non-ductile cable trays would ,

eventually become ductile by inelastic deformation, buckling or failure of the non-ductile cable tray supports and members. This proccdure is a basic departure from conventional methods of engineering evaluation and the GIP does not provide adequate bases for dealing with those cable trays that initially are judged to be non-ductile but are eventually called ductile by postulating failure of the lateral supports. If this procedure was followed for l

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eliminating cable trays for further assessment at your plant, then all'-the cable trays could conceivably be screened out from A-46 evaluation. We request

.that you r.espond to the following items to elicit information that would support our safety evaluation of cable trays at the Vennont. Yankee plant.

.(1) Provide.the total n'aber of raceways that were classified as ductile in your A-46 evaluation and for which you did not perform a horizontal load- "

. evacuation. Indicate the approximate percentage of such raceways as compared with the entire population of raceways. Discuss how the ductility concept is used-in your walkdown procedures.

(2). Provide descriptions of all the typical configurations of your ductile raceways (dimensions, member sizes, supports, etc.)

(3) Provide justification for stating that ductile raceways need not be evaluated for horizontal load. When a reference is provided, state the page number and paragraph. The reference should be.self contained, and not refer to another reference.

(4) In the evaluation of the cable trays and raceways, if the attachments are assumed to be ductile in one horizontal direction, does it necessarily follow that the same system is ductile in the perpendicular direction? If yes, provide the basis of this conclusion. if no, discuss how the seismic adequacy'of the. attachments was evaluated.

(5) Provide a definition of ductility:in engineering terms as used at Vermont Yankee for the USI A 46 evaluation. Clarify how this definition is applied to actual system configurations at Vermont-Yankee consistently for the purpose of analytical-evaluation. Provide an assessment of the maximum

' ductility utilized for the weakest cable tray support.

(6) JDiscuss any raceways and cable trays including supports in your plant that-are outside of the experience data. Explain what criteria are used for

establishing their safety adequacy and.specify your plan for resolution ~ of

' outliers that did not meet;the acceptance criteria. Provide examples of '

the configurations of such raceways and cable trays including supports.

Also, indicate the percentage of cable trays and raceways outside the experience data in relation to the population of raceways 'and cable trays

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examined during the walkdowns of the safe shut down path. How are they .i o going to be. evaluated and disposed?

8. If Thermal-Lag panels are attached to a cable tray system, discuss how the changes'in weight have been incorporated in the GIP evaluation of these systems and their supports.

9. You stated that the reviews of the cable tray and conduit systems at Vermont Yankee were performed per the guidelines of Section II.8 of the GIP. However, the discussion provided in page 7-3 of the summary report regarding seismic i-supports appears.to differ from the GIP's description. Previde the basis for

your deviation fros' the GIP.

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L10c You stated that toward the end of the plant construction phase after all the

cable tray runs and related deadweight supports were installed, seismic

.' supports were added to the cable tray- runs in the Reactor Building (page 7-2).

Provide your evaluation of lateral seismic suppets and sample calculations as

. well as sketches or drawings.

You indicated that seismic supports of four cable trays were evaluated and they .

were found to have adequate margin for lateral seismic loads (Reactor Building.

elevation 252'-5") (page 7-3. summary report). Provide the evaluation of these supports'with adequate drawings or sketches.

11. You stated that rigid mount supports are used in isolated locations where trays n .are adjacent:to walls (page 7-1 of the summary report). You also indicated

' that cable trays with typically 24 to 26 feet long horizontal runs are

' laterally supported by seismic supports which are rigid mount frames e constructed of structural steel sections with welded connactions (Page 7-2),

Discuss how the supports are evaluated-and accepted. Clarify whether they are part of the outliers listed in Table 7-1 of the summary report. Provide sketches or drawings of the so called rigid supports and the calculations used for their evaluation.

l L 12.' Provide a sketch or drawing of the end anchor related to the cable tray system

?you discussed Lin Section 7.3 of-the summary report (page 7-3) and provide a calculation showing how this system and end conditions are evaluated and-accepted as adequate.

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13. -The summary report did not discuss anchor bolts except in relation to cable trays. Explain the absence of documented evaluations of anchorages for other equipment.

14. In the summary report you stateti that you were committed to implement the GIP-2, including the clarifications, interpretations, and exceptions in SSER 2, and to communicate to the NRC staff any significant or programatic deviations from the GIP guidance. You further stated that there are no significant or programmatic deviations from the GIP guidance.

Provide the worst-case items (from a safety point of view) which deviate from the GIP-2 guideline but were categorized as not being significant. In addition, we request that you clarify the definition of " safety significant" that the walkdown crew used: and provide the basis for why the definition is adequate.

15. You stated that resolution of the outliers continues. When the resolution is 4 complete, we request that you submit a completion report including the resolutions you discussed in the summary report.

16. Provide a clarification on the number of cable trays-evaluated: since the text indicates that 12 cable trays are evaluated for limited analytical review while Table 7-1 shows 13. Also, you discussed four cable trays in the Reactor Building, elevation 252, while Table 7-1 lists five cable trays. Explain this discrepancy.

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