ML17265A470
| ML17265A470 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/03/1998 |
| 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 9812070265 | |
| Download: ML17265A470 (8) | |
Text
Dr. Robert C. Mecredy Vice President, Nuclear rations Decem r 3, 1998 Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
SECOND REQUEST FOR ADDITIONALINFORMATION,RESOLUTION OF USI A<6, R. E. GINNA NUCLEAR POWER PLANT (TAC NO. M69449)
Dear Dr. Mecredy In your letter of January 31, 1997, you provided the plant-specific summary report in accordance with its commitment relating to Generic Letter 87-02 on the resolution of Unresolved Safety Issue (USI) A-46 program at the Ginna Nuclear Power Plant (Ginna). The NRC staff has reviewed the summary report and determined that additional information was necessary in order to complete the review of the Ginna USI A-46 response.
A request for additional information (RAI) was submitted on April6, 1998. You responded to the RAI in your submittal on May 27, 1998.
The staff has reviewed the above May 27, 1998, submittal, and found that you have not completely resolved the staff's questions related to the Ginna USI A-46 program implementation.
We have determined that supplemental information is needed in order to complete the review.
Enclosed is a list of items in the second RAI.
In accordance with our discussion on this issue, your staff has indicated that you can provide a response on December 18, 1998.
Ifyou cannot meet this date, please advise us.
Sincerely, Original signed by:
Docket No. 50-244
Enclosure:
Request forAdditional Information cc w/encl: See next page DISTRIBUTION:
Guy S. Vissing, Senior Project Manager Project Directorate I-'1'ivision of Reactor Projects
- I/II Office of Nuclear Reactor Regulation Docket File PUBLIC J. Zwolinski S. Bajwa S. Little G. Vissing 0
Y"I Q 0 4f Q OGC ACRS C. Hehl, Rl R. Wessman G. Bagchi
~< ot II&RM CE7E>> CSPV DOCUMENT NAME: G:hGINNAI/I69449.2RI To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N"= No co OFFICE NAME PM:PDI-1 E
LA:PDI-Gvissing/rsI SL I L'C D:PDI-1 SBaiwa DATE 11/
/98 11/ 9/98 1
/98 Official Record Copy 98i2070265 981203 PDR
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Dr. Robert C. Mecredy Rochester Gas and Electric Company R.E. Ginna Nuclear Power Plant CC:
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 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 FaIls Road, Room 11 Rochester, NY 14620 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223
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SECOND REQUEST FOR ADDITIONALINFORMATION RELATED TO RESOLUTION OF USI A-46 R. E. GINNA NUCLEAR POWER PLANT The staff raised concerns in its April6, 1998, request for additional information (RAI) about the exceedance of Ginna in-structural response spectra (ISRS) over 1.5 times the free-field response spectra, as well as over 1.5 times the Bounding Spectrum (BS). The staff questioned the validity of GIP-2 Method A.1, concerning seismic capacity vs.
demand, for Ginna plant-specific situations.
In your letter of May 27, 1998, you provided a generic response to the above staff RAI, specifically, Questions No. 1 and No. 2. This generic response does not address plant-specific situations.
The staff noted the following from Figures 3.2 to 3.5 of your Seismic Evaluation Report:
For the standby auxiliary feedwater pump building, the ISRS at 25 feet above its grade elevation is about three (3) times of 1.5 x BS, in the frequency range of interest.
For the control building, the ISRS at 18 feet above its grade elevation is more than three (3) times of 1.5 x BS, in the frequency range of interest.
For the diesel generator building, the ISRS at 32 feet above its grade elevation is about 2.7 times of 1.5 x BS, in the frequency range of interest.
For the intermediate building, the ISRS at 18 feet above its grade elevation is about 2.4 times of 1.5 x BS, in the frequency range of interest.
Based on the above, you are requested to provide the following additional information:
a.
Provide a detailed discussion of how the "grade elevation" for each of the power block structures, including buildings designated as DG, IB, SH, TB, AB, AF, CB, and RC, was established.
Also provide a detailed description of the foundation and the surrounding embedment for each of the above building structures, and a detailed discussion of how they were modeled in the seismic analysis. This information is necessary in that the definition of "40 feet above the grade elevation" relies on a correct estimation of the grade elevation.
b.
Explain, from seismic analysis perspective, why the ISRSs of the above stated building structures would exceed 1.5 x BS by such a large margin, at an elevation only from 18 to 32 feet above the grade elevation.
These amplification factors appear to suggest that the building structures do not behave as "typical nuclear plant structures," as referred to in Section 4.2.3 of GIP-2. You are requested to address and resolve these plant-specific situations.
In your response to the staffs RAI Questions 43 and 04, for a number of equipment items, the equipment frequencies were stated to have been judged by SRT to be greater than 8 Hz by inspection.
Provide the basis for the SRT judgment regarding equipment natural frequency, especially when the estimated magnitude for natural frequency is relied upon to determine the applicability for the use of GIP-2 Method A.1. You are requested Enclosure
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to provide further justification for the frequency estimation, or provide analytical calculations to justify such estimation, for the equipment items identified as FT-4084, FT-4085, PSF01A8B, SAFWPCIP, SAFWPDIP, BUS14, DCPDPAB01ALB, and DCP DPAB02A8 B.
In your response to the staff's RAI, Question 04, you indicated that the seismic capacity vs. demand evaluation for the Undervoltage Relay Cabinet Bus 14 (ARA1RC14) was based on shake-table testing.
You are requested to provide a detailed discussion of the testing and to justifythe adequacy of such testing.
For the 480 VAC Motor Control Center (MCC), you indicated that the MCC can withstand a single frequency test consisting of a 1.35g, 5 beat, 5 cycle/beat input, performed at the significant structural frequencies.
It is known that single-axis, single-frequency nine beat tests, mostly performed prior to the issuance of IEEE Standard 344-1975, are considered inadequate for equipment seismic qualification due to their inability to excite multi-axis, multi-frequency responses of equipment (the very reason that plants are includ'ed in the USI A-46 program).
You are requested to justify the seismic adequacy of this motor control center.
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