ML18142A700
| ML18142A700 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/05/1978 |
| From: | White L Rochester Gas & Electric Corp |
| To: | Stello V Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18142A700 (6) | |
Text
REGULATORY INFQRl"IATION DISTRIBUTION SYSTEM (RIDS >
DISTRIBUTION FOR INCOMING MATERIAL 50-244 REC:
STEL'LO V NRC ORG:
WHITE L D ROCHESTER GAS
- 8. ELEC DOCDATE: 05/05/78 DATE RCVD: 05/09/78 DOCTYPE:
LETTER NOTARIZED:
NO COPIES RECEIVED
SUBJECT:
LTR 1
ENCL 0
RESPONSE
TO NRC LTR DTD 01/25/78... FURNISHING INFO RE APPLICANT"S SCHEDULE FOR PROVIDING AN EVALUATION OF THE EFFECTS OF ASYMMETRIC LOSS OF COOLANT ACCIDENT LOADS'HAT NILLi WITHIN THE TWO-YEAR TIME FRAME ESTABLISHED'SSESS THE SAFETY OF SUBJECT FAC1LI~
PLANT NAME: RE GINNA UNIT 1
REVIEWER INITIAL:
XJM DISTRIBUTOR INITIAL:~
DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS <+%++~<<<<++<++<++
GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.
(DISTRIBUTION CODE A001>
I'OR ACTION:
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'l/l/l///lllllllIlllll ROCHESTER GAS AND ELECTRIC CORPORATION e
S9 EAST AVENUE, ROCHESTER, N.Y. I4649 1
RIK LEON D. WHITE, JR.
VICK RRKSIDKNT May 5, 1978 Director of Nuclear Reactor Regulation ATTN:
Mr. Victor Stello, Jr., Director Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555 TKLK~K TT'RKA COCK TIS%46.2700C i nCD M~Cll en+
C Xl gCAM CD CP R. E.
GINNA NUCLEAR POWER PLANT PRIMARY COOIANT LOOP SUPPORT ANALYSIS
~l~~M~M~ MR<t r~Z mm Rochester Gas and Electric was requested by your letter of January 25, 1978, to inform you within 90 days of the detailed schedule for providing an evaluation of the effects of asymmetric loss of coolant accident (LOCA) loads as described in Enclosure 2
of your letter.
This submittal sets forth a realistic program of analytical evaluations that, will, within the two-year time frame you have established, assess the safety of the plant.
Prior to issuance of your letter of January 25,
- 1978, a task group of utilities was formed and, with the assistance of Westinghouse Electric Corporation, outlined the steps necessary to complete an evaluation of the type you have requested.
The evaluation program prepared by the task group has been broken into three parts, or phases, to gain the cost and time advantages of generic analyses applicable to several plants of similar design.
Phase A, which is essentially
- complete, consists of a detailed examination of primary loop characteristics for each plant involved in the task group in order to categorize and group the plants for the analyses.
Plant. groupings are based on plant. characteristics such as reactor vessel
- diameter, steam generator
- type, steam generator and reactor coolant pump support structures and primary loop temperatures.
Phase B will evaluate the postulated ruptures which have the greatest, effect upon the steam generator and reactor coolant. pump supports.
This evaluation will consist. of generic scoping analyses for the plant groupings determined in Phase A using standard loop analysis procedures.
The breaks to be postulated for the analyses are in the hot leg at the reactor outlet nozzle, in the crossover leg between the steam generator and the reactor coolant. pump, and 7Si300079
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SHEET NO.
in the cold leg at either the pump discharge or the.reactor vessel inlet nozzle.
During phase B bounding reactor coolant loop hydraulic forcing functions will be developed and generic subcompartment pressurization studies will be performed.
The results of these scoping analyses will determine if specific plant analyses will be required to assess the adequacy of the support systems.
Phase C will address the reactor vessel response for postu-lated pipe ruptures at the reactor vessel inlet and outlet nozzle.
Analyses will be 'performed on a group basis where asymmetric cavity pressure
- loads, vessel internal hydraulic loads, and vessel support characteristics are similar.
The need for specific plant analyses will be based on the results of the generic analyses.
Prior to beginning Phase C, studies will be performed to determine the effects of increased break opening times, more detailed downcomer annulus nodalization and structural damping on the MULTIFLEX calculations.
In addition, certain reactor cavity pressurization studies and the development of a mechanistic crack opening analysis will be performed.
The anticipated schedule for.the program is presented below:
Pro ram Effort, Com letion Date Phase A
Phase B Generic Group Analyses Phase B RCL Hydraulic Forcing Functions Phase B Subcompartment Pressurization Phase B Specific Plant Analyses Phase C Generic Group Analyses MULTIFLEXModifications Reactor Cavity Pressurization Studies Mechanistic Pipe Break Analysis Phase C Specific Plant Analyses Essentially Complete June 15, 1978 June 30, 1978 November 1, 1978 Early 1979 First Quarter 1979 January 1,
1979 January 1,
1979 January 1,
1979 January 1,
1980 Should the generic analyses indicate that some corrective action is necessary, potential plant modifications may be included in specific plant analyses.
These modifications may be later incorporated in the plant, or augmented in-service inspection or probability analyses may be proposed as an alternate solution.
In either case, it is our intent to demonstrate the safety of long-term continued operation.
Please inform us if you have additional questions concerning the program outlined above.
Sincerely yours, L. D. Whi e, Jr.
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