ML20245C321

Summary of 770525 Meeting W/Westinghouse & Util Group Members Re Efforts to Prevent Reactor Vessel Overpressurization.W/O Encls

ML20245C321
Person / Time
Site:	Beaver Valley, Salem, Kewaunee, Point Beach, Prairie Island, Surry, Turkey Point, Ginna, Robinson, San Onofre, Cook, Zion, 05000000, Trojan
Issue date:	07/11/1977
From:	Zech G Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
TAC-6714, NUDOCS 8710260123
Download: ML20245C321 (4)

Text

GW~

ij g*

p' 4 UNITED STATES NUCLEAR REGULATORY COMMISSION uc p

c7i WASHINGTON, D. C. 20565 -

) .g j July 11, 1977

)}k 1 -

j DOCKET NOS: E0-334, 50-315, 50-244, 50-305, 50-266, 50-301, 50-282, 50-306, 50-261, iS0-206d 50-280, 50-281, 50-344, 50-250, l 50-251, 50-295, 50-304 and 50-272 LICENSEE / FACILITY: Duquesne Light Company (Beaver Vally Unit 1),

Indiana & Michigan Power Co. (D.C. Cook Unit 1)

Rochester Gas & Electric Corp. (R. E. Ginna)

Wisconsin Public Service Corp. (Kewaunee)

Wisconsin Electric Power Co. (P. Beach 1 &2)

Northern States Power Co. (Prairie Island Units 1 &2)

Carolina Power & Light Co. (H.B. Robinson Unit 2)  !

Northeast Utilities ( Haddam Neck)

Southern California Edison Co. (San Onofre)

Virginia Electric & Power Co. (Surry Units 1&2)

Florida Power & Light Co. (Turkey Point Units 3&4)

Portland General Electric Co. (Trojan)

Commonwealth Edison Co. Zion Units 1&2)

Consolidated Edison Co. Indian Point Unit.2&3)

SUMMARY

OF MEETING HELD ON MAY 25, 1977, WITH WESTINGHOUSE AND THE LICENSEES OF WESTINGHOUSE - DESIGNED PWR FACILITIES CONCERNING THE STATUS OF THEIR EFFORTS TO PREVENT REACTOR VESSEL OVERPRESSURIZATION On May 25, 1977, we met with representatives of Westinghouse Electric i Corporation and members of the Westinghouse Utility Group on Reactor Vessel Overpressurization, representing the licensee of each Westinghouse designed operating PHR facility (except Yankee Rowe and Salem), to discuss the results of a generic analysis conducted to determine the severity of potential pressure transients at these facilities and to review proposed measures to prevent or mitigate such transients.

A list of attendees is attached.

Significant discussions and agreements are summarized below.

The Utility Group chairman summarized the earlier activities by its members in determining a solution to the problem of reactor vessel ,

I overpressurization and indicated that their consultant, Westinghouse, had completed the generic transient analysis for the Westinghouse PWR plants. l This analysis is intended to be a boundary-type of analysis, using input l

}

parameters such that all members of the Utility Group will be able to i reference the analysis in their forthcoming plant-specific submittals.

The conclusion reached by the licensee as a result of the analysis is that a single power-operated pressurizer relief valve will mitigate pressure transients. (However, some overshoot of Appendix G limits results in certain transient conditions) 8710260123 770711 J PDR ADOCK 05000206 i

\

P PDR L____________________________ __ l

~ ..

i 1

Meeting Sumnary Numerous Utilities )

i Westinghouse representatives then provided a presentation of the detailed I results of the transient analysis. Copies of the presentation and the l transient analysis results are attached.

The Westinghouse analysis- considered both the mass imput and the heat ,

input types of transients. As indicated above, the method of overpressure i protection assumed in the analysis is a pressurizer power operated relief valve. The effect of two power-operated relief valves on setpoint overshoot 1 was included, although credit is taken for only one valve due to single failure  !

cri teria. A discussion was provided of the various reference parameters used for each of the two basic types of transients analyzed. The results of the analysis show that the start of a Safety Injection (SI) pump while in a  !

water-solid condition is the worst-cause transient of the mass input types l reviewed. The staff was advised that for such a transient, some overshoot 1 of Appendix G limits is predicted to result. The peak pressure noted in the analysis was about 800 psig.

The results of the heat input type of transient show that the worst ]

case transient is caused by a reactor coolant pump start, while water {

solid, with a temperature differential between the reactor coolant -

system (RCS) and a steam generator. With a AT of up to 50'F the single PORV would mitigate the transient with a setpoint overshoot of about 100 psi (PORV setpoint of 600 psig), with an initial RCS temperature of 180 F.

The effects observed by varying the various input parameters were shown and a summary was than made of the two types of transients reviewed. With I regard to the conservatism in the analysis, we advised the licensees that if properly substantiated, credit could be taken for the conservatism )

identified. {

I The staff advised the licensees that, as indicated at our November 4,1976  !

l meeting and in separate correspondence with each of the licensees, the I Appendix G limits for each of the facilities should not be exceeded as a result of a reactor coolant system pressure transient. Since the Westing-house generic analysis is a bounding-type of analysis, each licensee should show in its plant-specific submittal that the parameters applicable to their plant will not result in Appendix G limits being exceeded. For those facilities ,

which cannot demonstrate that the Appendix G limits will be met, further  !

analyses and review may be necessary.

i l

l l

I

_ _ _ _ - - - - 1

,o<

)

l Meeting Summary -

3-Numerous Utilities >

1 The Westinghouse generic analysis is to be submitted about mid-June 1977 and following a review by each licensee to determine the applicability .

of the results of the analysis to their facility, a plant-specific analysis l will be provided to the staff in late June 1977.  ;

l In summary, the staff indicated that the following should be considered be each licensee and/ included in its submittal: a 1.. Previous submittals defered addressing acceptance criteria until ,

the generic analysis was completed. This information should now {

be provided. 1 I

2. The plant conditions, components and systems that are to be l controlled by administrative procedures should be identified and ]

appropriate Tech Spec changes proposed. This should include the  !

l components or systems assumed to be unavailable and therefore not considered in the transient analyses.

3. Include electrical schematics and logic diagrams for protective ]

system control circuitry. l I

i

4. A dual setpoint enabling alarm must be included; a computer-generated alarm is not acceptable. The alarm must be "hard-wired" to the PORV isolation valve to assure that it is opened when j required. In response to a question asked during the meeting, the a valve position indicator lights are not acceptable for this purpose.

5. Additional details should be provided on the modeling of PORV's.

The effects of flashing on assumed relief capacity should be con-sidered.

l

6. An explanation of how the criteria will be satisfied should be i I

provided regarding the loss of air / loss of offsite power.

1 i

l l

l j

- - - - - - ~ - - - - -

-f

.y

, i l

Meeting Summary Numerous Utilities I 1

7. The final design must be capable of satisfying Appendix G limits

i over the 40 year life of the plant. If the 40 year period can-not be met initially, a licensee may elect to propose meeting a' (

lesser Appendix G fluence level, but with the understanding that the design would be considered to be interim.in.na_ture and a final design must be proposedland implemented within the.next' l'i )

refueling outage.

f 4

Gary Zech, Project Manager Operating Reactors Branch #1 Division of Operating Reactors

Enclosure:

1. List of Attendees .

j

2. Utility Group on Overpressurization - Meeting Agenda

3. Transient analysis results ,

c i

_ . _ _ . _