ML17249A780
| ML17249A780 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/26/1980 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | White L ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| TASK-15-01, TASK-15-02, TASK-15-03, TASK-15-07, TASK-15-08, TASK-15-15, TASK-15-2, TASK-15-7, TASK-15-8, TASK-RR NUDOCS 8003240456 | |
| Download: ML17249A780 (7) | |
Text
KMM7i35[~'r- [I[> ~OIIy Docket No. 50-244 DISTRIBUTION Docket NRC PSE Local PDR ORB REading
'<.NRR Reading DEisenhut RVollmer OELD OIQE (3)
OLZiemann TWAmbach Hr. Leon D. White, Jr.
HSmith Vice President NSIC Electric and Steam Production TERA Rochester Gas and Electric CorporationACRS (16S) 89 East Avenue DCrutchfield Rochester, New York 14649 FPB 2 6 1980
Dear fir. White:
I
SUBJECT:
SEP - Design Basis Events for Ginna We are continufng our review of the design basis events analyses for Ginna, and )ave found that the additional information described in the enclosure to >his letter is needed.
We request your response within 90 days after your receipt of this letter.
j Sincerely, QggIllal IIjI.I>C~
Dennis L. Ziemann, Chief Operating Reactors Branch ¹2 Division of Operating Reactors
Enclosure:
Request for Additional Information cc w/enclosure:
See next page OFFICE)'URNAME DATEP TIIAiFac+cc 2/g</00 P
- O..N H
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/00 DOR ORB ¹2 DLZiemann Zggi'GO NRC FORM 318 (9.78) NRCM 0240
%VS. GOVERNMENT PRINTING OFFICE: 1979'289'369
~1 0"
II lip
Mr. Leon D. White, Jr.
February 26, 1980 CC Harry H-Voigt, Esquire
- LeBoeuf, Lamb, Leiby & MacRae 1757 H Street, N-W-
Mashington, D.
C.
20036 Mr. Michael Slade 12 Trai liood Circle Rochester, New York 14618 Rochester Cotnnittee for Scientific Information Robert E. Lee, Ph.D.
P. 0. Box 5236 River Cantus Station Rochester, New York
.14627 Jef rey Cohen New York State Energy Office Sxan Street Building Core 1, Second Floor Eni re State P 1 az a
- Albany, New York 12223 Director, Technical Developm n.
S:a e of New York Energy Office Ag ncy Building 2 Emire State Plaza
- Albany, New York 12223 Rochester Publi c Library 1'.5 South Avenue P,cchester, New York 14604 Supervisor o
the Town of Ontario 1G7 Ridg Road West Or<<ario, New York 14519 Program Director, Technical Assessmen.
Division Office of Radiation Programs (AV-459)
U. S.
Environmental Protection Agency Crystal Mall 82 Arlington, Yirginia '20460 U. S.
Environmental Protection Ace ncy Region II Office ATTN:
EI S COORD INATOR 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,
Chair."zn At~ c Safety and Licensing Board U-S.
Nuclear Regulatory Conmssicn Mashing'.on, D. C.
20555 Dr. Richard F. Cole A chic Safety and Licensing Board U. S.
Nuclear Regulatory Cormr.ssicn Washing. on, D.
C.
20555 Dr. Emath A. Luebke A cc:i c Safety and Licens ing Board U-S.
Nuclear R gulatory CottTr:ssicn Washing'.on, D.
C.
20555 Mr. Tha".zs B.
Cochran Nat.ral Resources Defense Council, Inc.
1725
- Street, N.
W.
Sui=e 600 Washing. on, D.
C.
20006
Request for Additional Information - GINNA l.
The transient analyses p; ~vided in the topical report XNOF-77-40 generally resulted in higher calculated MDNBR values than those calculated values cor-responding to the reference cycle analysis.
One apparent reason for this trend is the increased initial ONBR assumed for the two sets of analyses.
Verify that the Cycle 8 analyses, based on an initial ONBR of 2.00, reflects the ratio at the assumed 102ll power level and assumed pressure and tempera-ture condi tions.
2.
Provide the following additional information regarding the analysis of a turbine trip event:
a.
Provide a single failure analysis to determine the limiting failure con-current with a turbine trip.
Evaluate the effects of the worst single failure on the turbine trip analYsis, b.
Justify why a range of reactivity feedback effects was not considered since this is normally analyzed for this type of plant.
c.
Identify whether the scram characteristics include assuming the most re-active rod stuck out of the core.
3.
Provide the following additional information regarding the analyses of a main steam line break:
a.
Discuss the main feedwater arid auxiliary feedwater flow assumptions used in the analysis.
These assumptions should conservatively maximize flow I
to the broken loop steam generator.
b.
Discuss the potential for single failures in the auxiliary feedwater con-trol system which may result in runout flow being continuously directed to the broken loop steam generator.
c.
Specify the initial core flow assumed in the analysis and demonstrate the assumptions are conservative.
d.
Specify how stored energy in the primary system (e.g., thick metal) was treated.
4.
Provide the following information regarding the analysis of a complete loss of forced coolant flow:
a Provide an evaluation of vat'ious sinqle failures. and consider theil impact on the consequences of this event.
b.
Identify whether the most reactive rod was assumed stuck out of the core following reactor scram.
5.
Provide the following information regarding the analysis of a primary pump rotor seizure event:
a.
Identify whether the analysis considered a loss of offsite power, turbine trip, and-coastdown of the remaining reactor coolant pump.
If these as-sumptions were not made for the reference analysis provide an evaluation of the locked rotor event addressing those items.
b.
Provide a single failure analysis in order to determine the most limiting failure for this event.
Eya]uate tne effects of the limiting s nqle
- failure, c.
Identify that non-safety grade equipment relied upon to mitigate the consequences of the accident.
Specifically address the assumed oper-ability of the pressurizer spray, relief and steam dump system.
d.
Discuss 'the long term coolability of the core.
e.
Specify whether the most reactive rod was assumed stuck out of the core following reactor trip.
6.
Provide'he following additional information regarding the analysis o
the inadvertent opening of a steam generator relief/safety valve:
a.
Discuss the basis of the assumed steam vent flow; i,e., what valve fails open and why this particular case was chosen.
b.
Discuss the main feedwater and'auxiliary feedwater flow assumptions used in the analysis.
These assumptions should conservatively maximize flow to the broken loop steam generator.
c.
Discuss the potential for single failures in the auxiliary feedwater control system which may result in runout flow being continuously directed to the broken loop steam generator.
- 7. Provide the following additional information regarding the analysis of a con-tinuous rod withdrawal at power:
a.
Discuss the basis for the selection of the high reactivity insertion
b.
Identify whether the scram characteristics include assuming the most reactive rod stuck in its fully wi.thdrawn position, c.
The results of -the Cycle 8 analysis for a slow rod withdrawal show a
marked increase over the calculated MONBR for the previous analysis.
Identify those input parameter dlffeI'ences between the analyses which contributed to the variation in the calculated results.
If the change in input parameters is not in the conservative direction then justify the use of less conservative values for the Cycle 8 analysis.
The refer-ence (FSAR) analysis of a slow rod withdrawal assumed the most negative Doppler coefficient.
Justify the conservatism of the assumed coefficient multiplier or provide an analysis assuming a Doppler coefficient of -1.5 x 10 a g/'F.
d.
Current plants utilizing a Westinghouse NSSS routinely consider a low in-itial power case corresponding to 10~ power.
In light of the fact that the MDNBR decreases for the partload cases, py'oylde an eyaluation of assuming 10" power initially.