ML20058A137
| ML20058A137 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/25/1978 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TAC-08731, TAC-08732, TAC-8731, TAC-8732, NUDOCS 7810300237 | |
| Download: ML20058A137 (4) | |
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NORTHERN STATES POWER COMPANY l
MIN N E A PO LI S. M I N N E S OTA 55401 l
9 Octc be r 25, 1978 4
Director of Nuclear Reactor Regulation U S Nuclear Regulatory Commission i
Washington, DC 20555 PRAIRIE ISLM;D NUCLEAR GENERATING P1 ANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Evaluation of Overall Asymmetric LOCA loads In a letter dated January 25, 1978 from Victor Stello, Director of Operating Reactors, n'orthern States Power Coupany wa's I
reques ted to provide the Commission with a schedule for com-j pleting an evaluation of the effects of asymmetric loss of -
coolant ac cident (LOCA) loads.
In a letter dated May 3,71978 :
j we info rced the Commission that we were working with a utility ~
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task group to complete the required analyses and we submit ted..
7-the anticipated schedule for the program being fo llowed by, ;
s the task group.
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Following completion of Phase A of the utility task group 2
i program, consisting of data collection and plant grouping, we
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reevaluated our position with respect to continued participa-tion in the group's activities and decided to discontinue our me=b e rs hip.
Continuation with the activities planned by the task group for Phase B was deemed to be unnecessary since i
similar analyses were completed by Westinghouse in 19 72 - 19 73 i
for the Prairie Island plant.
Methods of analysis were very similar to those currently used in new plant designs.
Contin-uation with the activities planned by the task group for Phase j
C was considered, but was not found to'be the most cost ef fective or expeditious method of resolving the question of l
asyntetric LOCA loads for a break at the reactor vessel nozzles for a plant of comparatively new design like Prairie Island.
Based on the results of Phase C analyses already completed for i
recent design Ves tinghouse PWR facilities, we have chosen to move ahead with -the installation of pipe restraints to limit the break area at the vessel nozzles to less than one square foot and perform a plant specific analysis to demonstrate the adequacy of this modification.
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NORTHERN STATES POWER COMPANY Director of Nuclear Reactor Regulation Page 2 Octobe r 25, 19 78 Because the design of the reactor coolant system and supports at Prairie Island is ve ry s imila r to new plant designs, we believe a LOCA at the reactor vessel nozzles can be tole ra ted if the break area is lialted to one square fo ot by suitably designed pipe restraints in the reactor shield wall.
This is similar to the codification planned for several recently coupleted pressurized water reactor facilities.
Installation of the pipe restraints at Prairie Island and cocpletion of a Phase C analysis to demonstrate that a nozzle area break can be tolerated will be done in accordance with the fellowing tentative schedule:
Autuen 1979 - Install restraints in Unit No. 2 reactor shield wall.
i Spring 1980 - Complete analysis of LOCA at reac-tor vessel nozzles assuming break area of one square foot, and Install restraints in Unit No. I reactor shield wall.
Reactor shield wall pipe restraints are currently being desig-ned by the Prairie Island a rch i te c t-e nginee r.
Westinghouse Electric Corporation has been authorized to proceed with the analytical work necessary to substantiate the adequacy of the Prairie Island design with the reactor shield wall restraints in place.
The re s traint design and analytical work is being pe r fo rced jointly for the Prairie Island and Kewaunce plants.
A joint effort is possible because Prairie Island and Kewaunee have identical reactor coolant system and containment designs.
Both plants were built in parallel by the same a r ch it ect-e ngi-neer a nd both utilize the same nuclear steam supply system design.
The Phase C analysis of the break at the reactor vessel nozzles as currently planned, will co ns is t of seven major sub-tasks:
- Reactor Internal and Coolant inop Hydraulic Forcing Functions Hydraulic forcing functions for the limited pipe rupture at the vessel inlet and outlet nozzles will be determined. The pump dis-charge nozzle break will also be checked in the event it controls the vessel internals loads with the nozzle break area limited.
The Westinghouse MULTIFLEX code will be used.
I with a break opeing time of 10 milliseconds.
A break" area ^ of one square foot will be assumed at the nozzles.
NORTHERN STATES POWER COMPANY Director of Nuclear Reactor Regulation Page 3 October 25, 1978
- Reactor Cavity Pressurization Forces The Westinghouse TMD cooputer code will be used to calculate pressure transients in the cavity for the nozzle breaks.
- Structural analysis of vessel, Supports, Nozzles, and Piping The calculated vessel motion and hydraulic loading will be used to evaluate the coolant piping.
Forcing functions to be applied to the structural nodels include:
cavity asytcetric pressure loads vessel internal hydraulic forces load applied by the loop piping to the vessel break release forces at the broken nozzle hydraulic forces
- Evaluation of Fuel Assemblies The reactor vessel motion analyses will assute the use of Westinghouse fuel. Core support plate time history notions will be available to ve rify that non-Wes tinghouse f uel also recains within allowable stress and defor-cation limits.
- ECCS Pioing and CRD Mechanism Evaluation Critical piping attached to the reactor vessel and CRD techanists will be evaluated using the calculated vessel motion to deconstrate adequacy.
- Evaluation of Reactor Internals The reactor internals and core support struct-urcs will be evaluated for structural integ-riety for the three breaks.
- Structurcl Concrete Evaluation The structural integrity of the concrete in the reactor shield wall will be evaluated.
While we believe the analysis will show acceptable cor. sequences for a limited area pipe break at the vessel nozzles or a full break at the pucp dischage nozzle, some additional modifications or detailed analysis, such as including plasticity in the s tructural models or a refined model for break opening time and area, may be found necessary.
We will inform the NRC Staff of any modifications of this nature.
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NORTHERN STATES POWER COMPANY Director of Nuclear Reactor Regulation Page 4 October 25, 19 78 Please contact us if you have a ny questions related to the program we have described for resolving this issue.
We would i
also like to e=phasize in closing that our decision to discon-tinue participation in the utility task group's activities in no way implies disagreement with their proposed plans.
We are j
in full agreement with the task gr up's announced goals and act ion plan for t heir neebe r u ti!itie,. Our plan for resolving the asymmetric loading question for Prairie Island is practical only because of t i.e rela tive ly recent design of our reac tor coo la nt systen piping and supports.
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L 0 '!ayer, PE Manager of Nuclear Support Services LOM/D'O!/deh cc: Director IE-III G Charnoff i
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