ML18142A790
| ML18142A790 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/15/1978 |
| From: | White L Rochester Gas & Electric Corp |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18142A790 (10) | |
Text
yg I//Ng REGULATORY INFORMATION DISTRIBUTION SYSTEM
< RIDS)
DISTRIBUTION FOR INCOMING MATERIAL 50-244 REC:
ZIEMANN D L NRC ORG:
WHITE L D ROCHESTER GAS Sc ELEC DOCDATE: 03/15/78 DATE RCVD: 08/17/78 REVIEWER INITIAL:
XJM DISTRIBUTOR INITIAL:H.(
DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.
(DISTRIBUTION CODE A001)
DOCTYPE:
LETTER NOTARIZED:
NO COPIES RECEIVED
SUBJECT:
LTR 1
ENCL 0
RESPONSE
TO NRC" S REQUEST OF 12/16/77... FURNISHING INFO CONCERNING PERMANENT RESOULTION AND SCHEDULE FOR IMPLEMENTATION TO STAFF CONCERNS ABOUT UPPER PLENUM INJECTION OF EMERGENCY CORE COOLING SYSTEM WATER.
/
PLANT NAME: RE GINNA UNIT 1
FOR ACTION:
INTERNAL:
EXTERNAL:
~~LTR ONLY(7)
E+~~LTR ONLY<2)
HANAUER~~~FLTR ONLY( 1 )
EISENHUT+~~LTR ONLY< 1)
BAER'~+LTR ONLY(i)
GRIMES~~4~I TR ONLY< 1)
J.
MCGOUGH>+LTR ONLY(l>
LPDRiS ROCHESTER'Y4+LTR ONLY< 1)
TIC~HILTR ONLY(1 >
NSIC-r+LTR ONLY(1)
ACRS CAT B++LTR ONLY(16)
NRC PDR++LTR ONLY('1>
OELD%+LTR ONLY(i)
CHECK++LTR ONLY(1>
SHA04+LTR ONLY(1)
BUTLER++LTR ONLY(i)
J COLLINS4+LTR ONLY< 1)
NTROL NBR:
780790001 DISTRIBUTION:
LTR 40 ENCL 0 SIZE:
4P 44~4444%4+%%~~44~4~M%~$ ~%%%~4%%%4%~M THE END
~M4Q%%%+4%~4%%444%% %%%%%%%%%%~%%%~
li I
4'
//////////////r rziiriziz LA4$ oNI4Alo I
/////I// ///
///I//////I'OCHESTER GAS AND ELECTRIC CORPORATION o
89 EAST AVENUE, ROCHESTER, N.Y. 14649 LEON D. WHITE, JR.
VICE PRKSIDKNT Director of Nuclear Reactor Regulation ATTN:
Mr. Dennis L. Ziemann, Chief Operating Reactor Branch 42 U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Ziemann:
A letter from Mr. Edson G.
Case dated December 16, 1977 requested that we provide a permanent resolution, and a schedule for its implementation, to staff concerns about upper plenum injection (UPI) of emergency core cooling system (ECCS) water.
We believe that the models currently used to evaluate ECCS per-formance are conservative and comply, with Appendix K to 10 CFR Part 50.
We will, however, take the action outlined below in order to satisfy the staff requests in the letter of December 16, 1977.
A new ECCS evaluation myf~l will be developed to address the Safety Evaluation Report (SER) issued by the NRC on December 16, 1977.
This SER requested that explicit accounting of upper plenum injected water be simulated during a loss of coolant accident.
Previous ECCS analyses have assumed that UPI water adds directly to bottom reflooding with no accounting of steam-water interaction.
Such a model is conservative since no credit is taken for steam condensation and fuel rod cooling as the UPI water penetrates the core.
However, in an effort to be responsive to the
- SER, we have developed a work scope which will include steam-water interaction due to upper plenum injection during the reflood portion of the accident.
The work is segre-gated into three phases:
development of model, generic sensi-tivity study and individual plant analysis.
The development of the new model will use a]2) starting
[3]
point the Westinghouse evaluation codes WREFLOOD and LOCTA Added to these appropriate codes will be an average model approach for simulating the UPI water and its interaction with the system.
The approach for handling the UPI water interaction is similar to the approach used by the NRC in its SER and in our submittals of January 16, and February 15, 1978.
The simulation of the UPI water will be accounted for during the reflood portion of the transient and no blowdown effect will be considered.
07'9000 Rool a <
I/p
k l
lf u',
~
~
~g,
~
t
'o I
P
'N f
t
ROCHESTER GAS AND ELECTRI CORP.
~
DATE March 15, 1978 To Mr. Dennis L. ziemann, Chief SHEET NO.
The basis for eliminating modeling of the UPI water during the blowdown portion of th~4tg~nsient is based upon two generic 2-loop sensitivity studies These studies showed that for all postulated breaks from a 3 ft split break to the full double 2
ended cold leg guillotine (DECLG)
(C 1.0,
.8,
.6,
.4) safety injection does not occur during blow8own.
Therefore, no modifi-cation to the SATAN-VI code used during the blowdown portion of the transient is required.
The DECLG breaks have been shown to be the most limiting breaks.
For small breaks, the results show that the core pressure is above the shutoff head of the low head safety injection pumps (UPI) for 8 inch diameter breaks and smaller.
Therefore, no UPI will occur and it need not be modeled.
For breaks between 8 inches and 3 ft splits UPI will occur during the blowdown portion of the transient.
However, for such breaks the core recovers very quickly (>
1 ft/sec) due to accumulator in-jection.
Assuming the steam-water interaction model yields about the same flooding rate change as for the large breaks
(+0.2 in/sec),
the relative effect is much smaller for these breaks than the limiting breaks.
As indicated in our January and February, 1978 submittals there is little effect on peak clad temperature for the limiting breaks.
Therefore, since the small breaks have 500-700'F margin to the acceptance criterion, no modification will be made to either WFLASH or SATAN-VI to run breaks sizes between 8 inches and 3 ft In the new model the following effects will be considered; metal heat in the upper plenum, top quench front propagation, vaporization of UPI water, horizontal and vertical entrainment, decay heat in the fuel and integration with the existing WREFLOOD code to get the overall system and feedback effects.
Upon comple-tion of the development
- phase, generic sensitivity studies will be performed including a large break spectrum for double-ended cold leg guillotine ruptures (CD = 1.0,
.8,
.6,.4).
Single failure cri-teria will be examined assuming loss of offsite power and no pumped SI during blowdown.
Upon completion of the sensitivity studies, a worst break analysis will be performed for our plant using the limiting con-ditions determined by the generic sensitivity study.
0 P
V il (I I'l P
V
~
~
4 1
I
ROCHESTER GAS AND ELECTRIC CORP.
March 15'978 Mr. Dennis L. Ziemann, Chief SHEET NO.
For the work scope described above the estimated calendar time for completion is:
Development Sensitivity Plant Analysis Report 5-6 months 2 months 1 month 1 month 9-10 months
" This schedule is an estimate of the time required to accom-plish the basic scope of work and does not include the time re-quired for NRC staff review or the time required to respond to requests for additional information.
It will be desirable that the NRC staff review and approve the model outlined above prior to the time that sensitivity studies or specific plant analyses are performed.
We will therefore con-tact you to arrange a meeting to discuss our results as we near the end of the model development phase of our work. If you have questions which will affect the action outlined above please con-tact us prior to that time.
Sincerely yours, LDW:cern
Attachment:
References
J
'l 1
8 N'
REFERENCES
- Case, E.G., Letter, dated December 16,
- 1977, "Safety Evaluation Report on ECCS Evaluation Model for Westing-house Two-Loop Plants".
Collier, G., et al "Calculational Model for Core Re-flooding often a "Loss of Coolant Accident (WREFLOOD Code)"
WCAP-8170, June 1974.
Bordelon, et al, "LOCTA-IV Program:
Loss of Coolant Transient Analysis", WCAP-8301-June 1974.
Delsignore, et al, "Westinghouse ECCS Two-Loop Plant Sensitivity Studies (14 x 14)", WCAP-8854-A, May 1977.
"Westinghouse Emergency Core Cooling System Plant Sensitivity Studies",
WCAP-8340, July 1974.
~
Ay
~
'c