ML20198F805
| ML20198F805 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 06/26/1974 |
| From: | Tedesco R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| CON-WNP-0981, CON-WNP-981 NUDOCS 8605290098 | |
| Download: ML20198F805 (5) | |
Text
-
l LADA 4 i
i y 3 61574 Docket No. 50-460 V. A. Moore, Assistant Director for Light Water Reactors, Group 2, L REQUEST POR ADDITIONAL INFORMATION FOR WFPSS NO. 1 Plant Name: WPPSS No. 1 Docket No.: 50-460 Licensing Stage: CP NSSS Supplier: Babcock & Wilcox Architect Engineer: United Engineers and Constructors Contalmaant Type: Dry Responsible Branch & Project Manager: LVR 2-3;
- 7. Cox Requested Completion Date: June 26, 1974 Applicant's Response Date: August 21, 1974 Review Status: Incomplete The containment Systems Branch has reviewed the applicable portions of the PSAR for the Washington Public Power Supply System No.1 (WPPSS-1).
2nclosed is a request for additional information which specifically identifies the infornation we will need to complete our review.
The most significant review items noted are with respect to the methods used by the applicant to deternine the energy release to the containment following blowdewn and some aspects of the modeling used in the sub-compartment pressure analysis.
A generic necting with the applicant was held March 14, 1974* to discuss concerns about the method used by B&W to predict the mass and energy release to the containment for the containment pressure analysis. We concluded at that tine the complete steam-water mixing in the CRAFI code had not been justified. Because this applicant has continued to base.the centainment design pressure on an analysis that assumed complete steam-water mixing during the post-blowdown period of the accident, we have concluded that there is not adequate margin between calculated Don K. Davis and Walt Jensen, "Cuanary of Generic Meeting with Babcock
& Wilcox (B&U) ou Containment Peak Pressure Analycis" meno dated
/,pril 2, 1974.
IhiV h
8605290098 740626 PDR ADOCK 05000460 A
p 2 6 54 V. A. Moore and design containment pressure. This may require an increase in the capability of the containment heat removal systems.
Original signed bl Bobert I. Tedeseo Pobert u. Tedesco Assistant Director for Containment Safety Directorate of Licensing
Enclosure:
As stated cc: w/o encl.
A. Giambusso
- 11. Mcdonald w/ enc 1.
J. IIendrie S. Ilanauer J. Glynn A. Schwencer T. Cox R. U ecker D. Eisenhut S. Varga J. Carter l
G. Lainas T. Greene L - Reading CS - Reading CSB - Reading Docket File No. 50-460 l
i om'cr >
L:CSB SB
,_,L: CS[K 1
TGreene:lt G
nas RLTedesco
. 6/.d;.71...
6c:;].74....
m.
. 6/.23/.7.4...
m.c.,,m......,e -..
REQUEST FOR ADDITIONAL INFORMATION WPPSS-1 DOCKET No.*50-460 6.1 Response to Question 6.31.d is inadequate. Although the pressurizer subcompartment pressure analyses was reanalyzed assuming a break of 2
0.9 ft, the design of the surge line piping has changed from a 14-inch line to a 16-inch line. Therefore, provide the pressure analysis assuming a break of the 16-inch line; i.e., an area of 2
1.4 ft,
p 6.2 Response to Question 6.31.c is inadequate.
For the steam generator subcompartment pressure analyses, past experience with similar type plants have indicated that when the nodes (volumes) of the steam 3
generator compartment exceeded 100,000 ft,as in your analysis, t
the compartment was analyzed using two or more nodes to better i
represent the subcompartment. Therefore, provide an analyses for the steam generator subcompartment pressure response using nodes that are not in excess of 100,000 ft Include the volume and vent areas used in the analysis.
6.3 The response to Question 6.31.f is incomplete, In order to determine the adequacy of the design pressure for the pipe penetration in the reactor vessel shielded wall, provide a pressure analyses in which the break is assumed to occur in the pipe penetration.
State the design pressures.
9
6.4 Discuss how the insulation for components within the reactor cavity and penetration and the pressurizer subcompartments was considered in determining the volumes and vent areas for the subcompartment pressure response analyses. We believe that the analysis be done conservatively by assuming that the insulation remain intact.
- 6.5 Response to Question 6.27 is inadequate. We have concluded that the assumption of complete steam-water mixing in the CRAFT code has not been fully justified and therefore the mass and energy release. rates for the containment pressure analysis is unacceptable for the post-blowdoun period.
It is our recommended position that the analysis of the energy release be determined without the assumption of quenching of the steam in the primary reactor coolant system during the reflood' and post-reflood period.
6.6 The interface resistance (contact or gap resistance) between steel and concrete for the steel-lined hemispherical dome heat sink used in the containment pressure calculation is assumed to be 1/16". Provide justification for the use of this value.
6.7 Provide an analysis that demonstrates adequate mixing of hydrogen within the subcompartment and that the hydrogen concentration vill remain below the guideline limits stated in Regulatory Guide 1.7.
Provide the assumptions and describe the mathematical model used in the analysis.
9 w
~
w
3 6.8 In regard to the subcompartment pressure analyses, identify the vent areas assumed to exist during the transient that were not available prior to the accident; i.e.,
identify any additicnal vent areas that may be provided by rupture disks or doors.
For these vent areas, discuss how the vent area will yary with time and describe any tests that will be conducted to confirm the vent opening time.
Provide an analysis that will consider the potential for and the effects of missiles generated from the proposed relieving devices.
- Recommended positions s
i 6
I e
g