ML17265A210
| ML17265A210 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/06/1998 |
| From: | Vissing G NRC (Affiliation Not Assigned) |
| To: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR GL-87-02, GL-87-2, TAC-M69449, NUDOCS 9804130016 | |
| Download: ML17265A210 (9) | |
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April 6, 1998 Dr. Robert C. Mecredy Vice President, Nuclear Operations Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
REQUEST FOR ADDITIONALINFORMATIONON THE RESOLUTION OF UNRESOLVED SAFETY ISSUE (USI) A-46, FOR THE R. E. GINNANUCLEAR POWER PLANT (TAC NO. M69449)
Dear Dr. Mecredy:
In your letter of January 31, 1997, you provided a plant-specific summary report in accordance with your commitment relating to Generic Letter 87-02 on the resolution of the Unresolved Safety Issue A-46, "Environmental Qualification of Electrical Equipment" for the R. E. Ginna Nuclear Power Plant.
In order to complete our review, we have determined the need for additional information. The enclosed provides our request for additional information.
Please provide your response within 45 days from receipt of this letter.
Sincerely, Original Signed by:
- Guy S. Vissing, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation I
Docket No. 50-244
Enclosure:
Request for Additional
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April 6, 1998 Dr. Robert C. Mecredy Vice President, Nuclear Operations Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
REQUEST FOR ADDITIONALINFORMATIONON THE RESOLUTION OF UNRESOLVED SAFETY ISSUE (USI) A-46, FOR THE R. E. GINNANUCLEAR POWER PLANT (TAC NO. M69449)
'ear Dr. Mecredy:
In your letter of January 31, 1997, you provided a plant-specific summary report in accordance with your commitment relating to Generic Letter 87-02 on the resolution of the Unresolved Safety Issue AA6, "Environmental Qualification of Electrical Equipment" for the R. E. Ginna Nuclear Power Plant.
In order to complete our review, we have determined the need for additional information. The enclosed provides our request for additional information.
Please provide your response within 45 days from receipt of this letter.
Sincerely, Original Signed by:
Guy S. Vissing, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosure:
Request for Additional Information cc w/encl: See next page U
Docket File S. Bajwa ACRS PUBLIC S. Little C. Hehl PDI-1 R/F G. Vissing J. Zwolinski (A)
OGC DOCUMENT NAME: G:)GinnahM69449.RAI To receive a copy ofthis document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy OFFICE PH:PDI-1 E
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 20555-0001 April 6, 1998 Dr. Robert C. Mecredy Vice President, Nuclear Operations Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
REQUEST FOR ADDITIONALINFORMATIONON THE RESOLUTION OF UNRESOLVED SAFETY ISSUE (USI) A<6, FOR THE R. E. GINNANUCLEAR POWER PLANT(TAC NO. M69449)
Dear Dr. Mecredy:
In your letter of January 31, 1997, you provided a plant-specific summary report in accordance with your commitment relating to Generic Letter 87-02 on the resolution of the Unresolved Safety Issue A-46, "Environmental Qualification of Electrical Equipment" for the R. E. Ginna Nuclear Power Plant.
In order to complete our review, we have determined the need for additional information. The enclosed provides our request for additional information. Please provide your response within 45 days from receipt of this letter.
Sincerely, Docket No. 50-244 Guy S. Vissing, Seni Project Manager Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
Request forAdditional Information ccw/encl: See next page
Dr. Robert C. Mecredy Rochester Gas and Electric Company R.E. Ginna Nuclear Power Plant Peter D. Drysdale, Sr. Resident Inspector R.E. Ginna Plant U.S. Nuclear Regulatory Commission 1503 Lake Road Ontario, NY 14519 Regional Administrator, Region I
U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. F. William Valentino, President New York State Energy, Research, and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223 Charles Donaldson, Esquire Assistant Attorney General New York Department of Lalw 120 Broadway New York, NY 10271 Nicholas S. Reynolds Winston 8 Strawn 1400 S Street N.W.
Washington, DC 20005-3502 Ms. Thelma Wideman, Director Wayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31 Lyons, NY 14489 Ms. Mary Louise Meisenzahl Administrator, Monroe County Office of Emergency Preparedness 111 West Falls Road, Room 11 Rochester, NY 14620
REQUEST FOR ADDITIONALINFORMATION R.E. Ginna Nuclear Power Plant USI A<6
Reference:
Letter from Robert C. Mecredy (RGRE) to Document Control Desk (NRC), dated January 31, 1997, "Resolution of Generic Letter 87-02, Supplement 1 and Generic Letter 88-20, Supplements 4 and 5 (Seismic Events Only)."
In Figures 3.1 through 3.7 of the Seismic Evaluation Report (attached to the reference report), the in-structure response spectra for various elevations of safety-related structures are found to exceed 1.5 times the free-field response spectra.
In light of the guidelines provided by Section 4.2.3 of the generic implementation procedure (GIP-2), which
... restricts and limits.the. use.of the. ground. response spectrum for comparison.to the Seismic.
Qualification Utilities Group (SQUG) Bounding Spectrum (GIP-2 Method A.1) under the above exceedance condition, you are requested to provide justification for using Method A.1 in verifying the seismic adequacy of affected equipment items.
Note that in establishing the SQUG Bounding Spectrum (from the SQUG Reference Spectrum), there was an underlying assumption that within about 40 ft. above the grade elevation, for typical nuclear plant structures, the amplification of ground motion wilt be about 1.5. For cases where this assumption is invalid, the use of Method A.1 willnot be appropriate and may lead to unconservative results.
2.
Referring again to Figures 3.1 through 3.7 of your submittal, the in-structure response spectra for the control building, diesel generator building, intermediate building, screen house, and standby auxiliary feedwater pump building, are shown to significantly exceed 1.5 x Bounding Spectrum in the frequency ranges of interest.
In consideration of: (1) the limitations stated in Section 4.2.3 of GIP-2 (ref. Item 1) which states in part that the use of the ground response spectrum for comparison to the Bounding Spectrum is based on the conditions that the amplification factor between the free-field response spectra and the in-structure response spectra willnot be more than about 1.5, (2) the fundamental natural frequency of the equipment, including its supports, should be greater than 8 Hz, and (3) the equipment should be mounted at an elevation within about 40 ft. above the grade elevation, you are requested to either provide a justification for your use of Method A.1 or to reevaluate the seismic adequacy for the equipment items which are housed in these buildings.
The floor spectra of standby auxiliary feedwater pump building exceeds 1.5 x Bounding Spectrum at frequencies greater than about 6 Hz. On page seven of the Seismic Evaluation Report, it is stated that equipment in the building meets the capacity vs. demand requirements as long as.its fundamental frequency is greater than about 6 Hz. You are requested to list the affected equipment and provide for staff review representative sample calculations demonstrating how the fundamental frequencies are estimated.
You stated on page seven that the floor response spectra in the AuxiliaryBuilding exceeded 1.5 x the Bounding Spectrum at about 20 Hz and that the exceedances were not very large, therefore the equipment met the capacity vs. demand requirement.
However, it is shown, in Figure 3.1 "AuxiliaryBuilding FRS v. 1.5 x Bounding Spectrum," that the floor Enclosure response spectra start exceeding 1.5 x the Bounding Spectrum at about 10 Hz and the magnitude of exceedance is about 33% at 20 Hz. Therefore, you are requested to demonstrate that the equipment has greater capacity than the demand.
5.
The attached audit report, "Independent Audit Review of Seismic Walkdown and Evaluation of Safe Shutdown Electrical Equipment at Ginna Nuclear Power Station Unit 1," dated August 29, 1991, that was prepared by the auditor, Robert P. Kennedy, documents the review findings of the report, "Seismic Evaluation of Safe-Shutdown Electrical Equipment at Ginna Nuclear Power Station Unit 1," prepared for RGS E, by Stevenson
&Associates, Wobum, Mass., DRAFT, dated August 12, 1991.
It is stated in the audit report that the seismic margin assessment (SMA) methodologyas descnbed in the EPRI report NP-6041,...,
has been used to evaluate the seismic adequacy of the Reactor Water Make-up Tank (RWMT), the Battery Racks, and the Battery Room Block Wall.
The staff has noted that this methodology may yield analytical results which are not as conservative as those obtained by following the GIP-2 guidelines.
Because of the uncertainty of its conservatism, the methodology described in the above referenced EPRI report, has not been endorsed by the staff for the analysis of safety-related systems and components, including the resolution of mechanical, electrical, and structural component outliers in the A-46 program.
You are requested to reevaluate the portion of your A-46 program, where the above methodology has been employed, to ensure that all the analyses and outlier resolutions performed for the Ginna A-46 program utilize methodologies that are consistent with the plant licensing basis or other approaches acceptable to the staff.
6.
In the above audit report, a number of review findings were identified by the auditor. You are requested to discuss the implementation status of each of the findings and how these findings were resolved to the satisfaction of the auditor. Specifically, address those findings discussed on pages two and three, and in the first paragraph of page four of the audit report.
7.
In Appendix A, to the Seismic Evaluation Report, "Resumes and Seismic Capability Engineers," there is no evidence or certificate provided to demonstrate that Mr. LeonardA. Sucheski has completed the necessary SQUG training courses on the seismic adequacy verification of nuclear power plant equipment.
Provide evidence to support his qualification for participating in the USI A-46 implementation program.
8.
You stated on page 13 that a number of items of electrical equipment were re-anchored in the early eighties as a result of the Systematic Evaluation Program (SEP).
It appears that you did not perform the anchor bolt tightness check for those anchor bolts during the A-46 walkdown. Explain what was meant by "re-anchored."
Does re-anchored mean that the old anchor bolts were replaced by larger diameter new bolts as a result of the SEP activities, or that the same old bolts were re-torqued.
Identify the procedures used for the re-anchoring process during the A<6 walkdown. You also stated that you did not perform the bolt tightness check for those anchor bolts which were installed using a QNQC controlled procedure because you believe that the procedure was more stringent than that specified in GIP-2. Submit the QNQC controlled procedure and provide a comparison of its requirements to the GIP-2 requirements to demonstrate why the bolt tightness check need not be performed for these bolts.
9.
On page 26, you stated that the anchor bolts for the Volume Control Tank had a safety factor of 1.06. Submit the calculations for these anchor bolts. The calculations should include the overturning bending moment demand and the resisting moment capacity provided by the bolts.
10.
You stated on page 26 that the Refueling Water Storage Tank is a flat-bottom, vertical tank, 26.5'n diameter, and 81'all, and was anchored at the bottom with 30 2.25-inch-diameter "cast'in-place bolts', arid'the taiik liad a niomerit demand of 79,000 k-ft and a moment capacity of only 6,000 k-ft. You also stated on page 28 that a modification had been made to attach the tank to the floor slab at 271'levation.
Submit the details of the attachment of the tank to the slab, and the calculations used to verify that the tank would not have buckling or yielding stress problems after the modification.
11.
On page three of Appendix D, you stated that, for ultimate strength evaluation of cable trays, the plastic moment is taken as the allowable moment multiplied by the ratio of the minimum yield stress divided by the allowable stress, and then multiplied by 1.5 (shape factor). Since the shape factor of 1.5 is only applicable for solid rectangular bars, and cable trays usually are made from channels which do not resemble a solid rectangular shape, explain why it is appropriate to use 1.5 for the shape factor.
12.
On page four of Appendix D under the title of lateral load evaluation, you stated that "Acceleration values for the Cooper Nuclear Station corresponding to each of the options are summarized below." Explain why you used acceleration values for the Cooper Nuclear Station for Ginna Nuclear Station.
13.
On page three of Appendix H, "Outlier Resolution Plan and Schedule," you stated that the resolution for the outlier block walls was to perform additional analysis to determine the capacity of the walls, and the procedure would be finalized in 1997. Submit the procedures and criteria used for the wall analysis.
14.
In Appendix C, the check list for Seismic Interaction Review under the title of acceptance, you used the designation of "UnK."at several places.
We assume that "Unk" stands for unknown. State how you intend to dispose of these unknowns.
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