ML17265A208
| ML17265A208 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/03/1998 |
| From: | Vissing G NRC (Affiliation Not Assigned) |
| To: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| TAC-M99702, NUDOCS 9804080077 | |
| Download: ML17265A208 (7) | |
Text
Dr. Robert C. Mecredy
'ice President, Nuclear Ovations
'ochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649 April 3, 1998
SUBJECT:
REQUEST FOR ADDITIONALINFORMATION-REVIEW OF THE REQUEST FOR AMENDMENTDATED SEPTEMBER 29, 1997 - CHANGE TO THE TECHNICAL SPECIFICATION RELATED TO THE MAINSTEAM LINE ISOLATIONSIGNAL SETPOINTS (TAC NO. M99702)
Dear Dr. Mecredy:
We are continuing to review your request for amendment dated September 29, 1997, concerning the Technical Specification change to the main steam line setpoints.
We discussed this with your staff on March 18, 1998, concerning the need for additional information in order to complete our review. The enclosed request for additional information was discussed with your staff. Your staff indicated that they could respond to this request by early April 1998. We look forward to that response.
Sincerely, Original Signed by:
Docket No. 50-244 Guy S. Vissing, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional Information cc w/encl: See next page Docket File PUBLIC PDI-1 R/F J. Zwolinski (A)
S. Bajwa S. Little G. Vissing T. Collins OGC ACRS C. Hehl, Region I
DOCUMENT NAME: G:iGINNAiM99702.RI2 To receive a copy ofthis document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N"= No cop OFFICE DATE PM:PDI-1 Gvissing: Icc 03/
P/98 E
LA:PDI-1 SLitt a 03/
/98 D:PDI-1 SBajwa
/
/98 Official Record Copy DSSA:SRXB TCDIIins 03/
/98 03/
/98
'7804080077 980403 PDR ADOCK 05000244 P
I
'V lt l
Dr. R,obert C. IVlecredy
. Vice President, Nuclear 0~rations Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649 April 3, 1998
SUBJECT:
REQUEST FOR ADDITIONALINFORMATION-REVIEW OF THE REQUEST FOR AMENDMENTDATED SEPTEMBER 29, 1997 - CHANGE TO THE TECHNICAL SPECIFICATION RELATEDTO THE MAINSTEAM LINE ISOLATIONSIGNAL SETPOINTS (TAC NO. M99702)
Dear Dr. Mecredy:
We are continuing to review your request for amendment dated September 29, 1997, concerning the Technical Specification change to the main steam line setpoints.
We discussed this with your staff on March 18, 1998, concerning the need for additional information in order to complete our review. The enclosed request for additional information was discussed with your staff. Your staff indicated that they could respond to this request by early April 1998. We look forward to that response.
Sincerely, Original Signed by:
Docket No. 50-244 Guy S. Vissing, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional Information cc w/encl: See next page I T U
LEILJIIK:
Docket File PUBLIC PDI-1 R/F J. Zwolinski (A)
S. Bajwa S. Little G. Vissing T. Collins OGC ACRS C. Hehl, Region I
DOCUMENT NAME: G:>GINNA>M99702.RI2 To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No cop OFF1CE PM:PDi-1 E
LA:PDI-1 D:PDi-1 DSSA:SRXB kAME GVissingtlcc SLitt e DATE 03/
F/98 03/
/98 SBajwa 4
/ F /98 Official Record Copy TCol lins 03/
/98 03/
/98
~8 RE0(
4 Cy A,
O~
go
++*++
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055$ 4001 Apri 1 3, 1998
'r.
Robert C. Mecredy Vice President, Nuclear Operations Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
REQUEST FOR ADDITIONALINFORMATION-REVIEWOF THE REQUEST FOR AMENDMENTDATED SEPTEMBER 29, 1997 - CHANGE TO THE TECHNICAL SPECIFICATION RELATED TO. THE MAIN.STEAM LINE,ISOLATION,SIGNAL.......
SETPOINTS (TAC NO. M99702)
Dear Dr. Mecredy:
We are continuing to review your request for amendment dated September 29, 1997, concerning the Technical Specification change to the main steam line setpoints.
We discussed this with your staff on March 18, 1998, concerning the need for additional information in order to complete our review. The enclosed request for additional information was discussed with your staff. Your staff indicated that they could respond to this request by early April 1998. We look forward to that response.
Sincerely, Docket No. 50-244
Enclosure:
Request forAdditional Information cc w/encl: See next page G
S. Vissing, Senior roject Manager Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
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 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223 Regional Administrator, Region I
. - -.-- U.S. Nuclear Regulatory Commission-
'75 Allendaie 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 & 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
4l
E D
OALI FO T
TED CH NGEI E
IN E
L EIS I
L TON E
S 0
M ND EN D
You stated in the February 6, 1998, letter that LOFTRAN code was used to reanalyze
~ - - - -
- steam-line break events using%he revised values.(i.e allowable. value and setpoint);
'lease provide the version of the code used in this amendment.
Has this code and version been approved by the NRC (give reference to NRC-approved Westinghouse topical report)?
2.
Please provide results of your reanalyses.
3.
UFSAR indicates that the reactor control system is designed to accommodate a 10%
step or 5% ramp load increases...
without a reactor trip in the range of 12.8% to 100%
reactor power.
However, you stated in the September 29, 1997, letter that a reactor trip would not occur in the range of 15% to 100% reactor power.
Please clarify.
Enclosure