ML20236A996
| ML20236A996 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 10/15/1987 |
| From: | Frank Akstulewicz Office of Nuclear Reactor Regulation |
| To: | Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| NUDOCS 8710230286 | |
| Download: ML20236A996 (4) | |
Text
,
'0ciobir 15',11'987'-
Docket :No,':
50-213-Distribution.
UccketJ1ay ELJordan~'
.NRC T roii]aliPDR EDHaga'n I
> Mr. Edward J. Mroczka, Senior Vice~ President '
ISAPD-RF JPartlow
' Nuclear Engineering and Operations DCrutchfieldt Econner-
. Connecticut-Yankee Atomic Power Company FAkstulewicz LJHarmon
'. Post Office Box,270 --
MShuttleworth
. WJones
~ Hartford, Connecticut 06141-0270 OGC-Bethesdai GPA/PA5
'EButcher LLois
Dear Mr. Mroczka:
ACRSL(10)/
PRandall 1
ARM /LFMB j
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION CONCERNING. REANALYSIS'0F-1 o
NON-LOCA DESIGN BASIS TRANSIENTS l
q Re:
Haddam Neck Plant l
By letters dated June 30, 1986 and.May 8,.1987, Connecticut Yankee Atomic:-
y Power Company (CYAPCO) submitted a reevaluation of the design bases non-LOCAL transients for the Haddam Neck Plant. During its review of.these submittals,
~
the staff identified additional issues which need to be. resolved: prior to-j
' completing our review.
Enclosure'11to this letter contains our request for q
additional information.'
1 t
a-3 In order to meet the current review schedule, the staff. would need responses
~1 to these items within-30 days of receipt of thi.s. letter. 'If this-is not-l possible, please-have-a member of your licensing staff' contact me at
?
(301) 492-4790.
h i
The reporting'and/or recordkeeping requirements contained'in this letter affec't l
fewer than ten respondents; therefore,' OMB clearance is'not required under P. L.96-511.
Sincerely,.
I e
original signed by R
1 i
[
J Francis M. Akstulewicz, Project Manager t
l Integrated-Safety Assessment I
Project Directorate
'l l
Division of. Reactor Projects, - III, IV, V and Special Projec_ts
Enclosure:
As stated cc: See next page PM:ISAPD nD:ISAPDD
-FAkstulewit'z:dr~
- fCThomas p/g/87 p/jf/87 i-l 8710230286 871015 DR ADOCK 05000213
-i PDR l
1 1
e-1 Mr. Edward'J. Mroczka Haddam' Neck Plant Connecticut-Yankee Atomic Power Company cc: Gerald Garfield, Esquire Kevin McCarthy, Director Day, Berry & Howard Radiation Control ~ Unit Counselors at Law Department of Environmental L,
City Place-Protection f
Hartford, Connecticut 06103-3499 State.0ffice Building Hartford, Connecticut 06106 I
Superintendent Richard M. Kacich, Manager L
Haddam Neck Plant Generation Facilities Licensing RFD #1 Northeast-Utilities Service Company.
Post Office Box 127E Post Office Box 270 East Hampton, Connecticut 06424 Hartford. Connecticut 06141-0270-Wayne D. Romberg Donald O. Nordquist, Director Vice President, Nuclear Operations Quality Services Department i
Northeast Utilities Service Company-Northeast Utilities Service Department.
I Post Office Box 270 Post Office Box 270 Hartford, Connecticut 06141-0270 Ha rtford,. Connecticut 06141-0270 Board of Selectmen j
Town Hall l
Haddam, Connecticut 06103 1
I Bradford S. Chase, Under Secretary J
Energy Division Office of Policy and Management.
80 Washington Street i
Hartford, Connecticut 06106
~l 1
j Resident Inspector Haddam Neck Nuclear Power Station 1
c/o U.S. NRC P. O. Box 116 East Haddam Post Office East Haddam, Connecticut 06423 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 1
631 Park Avenue King of Prussia, Pennsylvania 19406 i
--__-_-a
Request for Additional Information Concerning Reanalysis of Non-LOCA Design Basis Transients I
'1.0 General A.
Explain why NUSCo did not analyze transients-in.the category of Increase in Reactor Coolant Inventory, which constitute the fifth category of SRP Chapter 15 transients.
B.
Provide.and' justify the nodalization diagram and' the conservatism used in code model selection and in the input assumptions for.each transient.
~
C.
Explain and justify how the' submitted analyses in NUSCo 151~ bound those which were not submitted.
D.
Identify for each transient analyzed the core burnup selected to yield the most limiting combination of physics parameters, j
E.
Explain how the scenario selected for each transient with the identified initiator is a moderate-frequency transient and most limiting.
2.0 Startup of an Isolated or an Idled Loop.
Explain and justify the statement " Effects of imperfect thermal mixing are included in the ar,alysis by using bounding assumptions to conserv-atively model this effect." What single active failures were-examines to enable NUSCo to conclude that no such single active failure could impact the pre-trip event consequences and how was this conclusion reached?
3.0 Excessive Feedwater A.
Justify the different times assumed for operator actions (37 seconds for 4-loop operation and 45 seconds for 3-loop operation) to trip the main feedwater pumps for the cases with potential for SG over-fill. Does use of a greater operator reaction time lead to steam generator overfill?
B.
Justify not using the split-core model.
4.0 Steam Line Break A.
Explain the nodalization used in the thermal-hydraulic analysis and justify the conservatism when compared with a split-core r
model in the thermal-hydraulic calculations.
B.
Justify the Boron transport model used in the analysis.
n
.' 5.0 Steam Generator Tube Rupture A.
Explain in detail why tripping the reactor at event initiation maximized the radiological consequences.
Explain further why the assumption of concurrent loss of offsite power is conservative.
B.
Provide Figure 15.6-3 as referenced on page 4.92.
C.
Explain what is meant by " blowdown" using steam generator safety valves.
D.
Demonstrate, by carrying out the calculations further in time, that the assumption of 30 minutes for isolation of affected SG is reasonable and conservative. Do the same for the rest of the transient.
It appears that the discussion pertaining to the cooldown phase is based upon assumptions.
E.
Explain the procedure which the licensee plans to use the intact steam generators to bring the rest of the plant to the RHRS entry conditions.
How was the steam release data (730,000 lbm, Table 4.10-2) from the intact SG obtained.
6.0 Loss of Normal Feedwater Flow A.
Explain what is meant by a " conservative core residual heat generation rate.
B.
Demonstrate that the case analyzed is the worst case and explain why loss of offsite power would represent a more conservative assumption than would offsite power available.
7.0 Reactor Coolant Pump Rotor Seizure and Shaft Break Reactor Coolant Pump Rotor Seizure A.
Why was this analysis performed at 100% power and not higher initial power than 100% for conservatism?
B.
Which peaking factors and axial power profiles are used for this analysis? How many axial nodes were modeled in the core and at what location are the minimum inner surface heat transfer co-efficient and the peak gap conductance which would maximize the predicted cladding inner surface temperature?
C.
Explain how the new scram reactivity as defined (most reactive rod assembly stuck out) in the Technical Specifications would impact this transient.
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