ML20040F870

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Affidavit of Wl Jensen Supplementing 811214 Affidavit. Explains Basis for NRC Conclusion Re Min Time Available for Operator Action After Break in Reactor Coolant Pump Suction Piping Vs Break in Discharge Piping
ML20040F870
Person / Time
Site: Rancho Seco
Issue date: 02/05/1982
From: Jensen W
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20040F853 List:
References
ISSUANCES-SP, NUDOCS 8202100404
Download: ML20040F870 (7)


Text

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of SACRAMENTO MUNICIPAL UTILITY DISTRCT

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Docket No. 50-312 (SP)

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(Rancho Seco Nuclear Generating Station))

AFFIDAVIT OF WALTON L. JENSEN, JR.

I, Walton L. Jensen, Jr., being duly sworn, depose and state that:

1.

I am an employee of the U.S. Nuclear Regulatory Commission (NRC).

My present position is Senior Nuclear Engineer, Reactor Systems Branch, Division of Systems Integration within the Office of Nuclear Reactor Regulation. A copy of my professional qualifications is attached.

2.

The purpose of this affidavit is to supplement the information provided in my affidavit filed in this proceeding on November 24, 1981.

The information in this supplementary a'fidavit is provided to explain in more detail the basis for the Staff's conclusion that the minimum time available for operator action after a break in the reactor coolant pump suction piping would not be signficantly different from that available for operator action after a break in the reactor coolant pump-discharge piping.

3.

The NRC Staff has reviewed the information contained in the letter from Babcock and Wilcox to Sacramento Municipal Utility District on " Reactor Coolant Pump Suction break LOCA" dated Ma~rch 25, 1981, which 8202100404 820208 DRADOCKOSOOO3g2

i ndicated that more reactor coolant system water might be lost for sinall breaks at the reactor coolant pump suction than for breaks at the reactor coolant pump discharge in the event that all feedwater were temporarily lost. We conclude that this informatio is not significant with regard to the safe operation of Rancho Seco and that additional small break LOCA analyses at the reactor coolant pump suction need not be performed. This conclusion was derived from the following considerations:

a.

Regardless of the postulated break location in the cold leg piping, the reactor vessel water level would initially decrease to the same approximate elevation.

b.

The additional loss of primary system inventory during a break.in the pump suction piping would be from water in the cold leg piping.

c.

In the absence of Emergency Feedwater the operator has a minimum of 20 minutes to actuate High Pressure Injection (HPI) regardless of break location in the cold leg piping, d.

Emergency procedures instruct the operator to actuate HPI immedi$tely regardless of break location if a loss of all feedwater has occurred.

The discussion below addresses these areas:

In the event of a small break LOCA at Rancho Seco, the liquid level in the broken pipe would decrease to the elevation of the break by liquid discharge from the break. The break flow would then be steam which would

' be generated in the core by decay heat. The High Pressure Injection system at Rancho Seco has sufficient capability to replenish the water boiled in the core by the decay heat.

For breaks in the cold leg piping

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at the reactor coolant pump suction water would be lost from the reactor vessel to the break until the liquid level dropped to the reactor vessel inlet nozzle elevation. At this point only steam would be lost from the reactor vessel.

Liquid discharge would continue from the break until the break was uncovered. More water would be lost for a break in the reactor coolant pump suction piping than for a break in the discharge piping since the cold leg suction piping is located at a lower elevation than the cold leg discharge piping. The additional coolant loss, however, would be limited to the cold leg piping inventory below the reactor vessel inlet nozzle. The water loss from the reactor vessel which provides core cooling is limited by the elevation of the reactor vessel inlet nozzle so that loss of vessel water would be approximately the same regardless of the break location in the cold leg.

Following the event at TMI-2, B&W performed small break LOCA analyses beyond those which had been presented to the Staff as a licensing basis to show conpliance with 10 C.F.R. 5 50.46. These

- additional analyses were performed for the purpose of providing guidance to the operator and are documented in the B&W report titled " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177-FA Plant" dated May 7, 1979.

These analyses demonstrated that Emergency feedwater would be 2

required for breaks of 0.01 ft and smaller in the reactor coolant pump discharge piping to depressurize the reactor system sufficiently to actuate High P'ressure Injection.

It was further demonstrated that operator action within 20 minutes to manually actuate HPI would prevent core uncovery.

Operators at Rancho Seco are instructed to initiate HPI w

immediately in the event that all feedwater is lost in loss of feedwater e

procedures and to monitor auxiliary feedwater flow in the small break LOCA procedure for the purpose of maintaining full HPI flow if auxiliary feedwater is lost.

Pump suction breaks coincident with a loss of all feedwater were not analyzed in the May 7th report.

For the reasons discussed above, the amount of reactor vessel water that would be available to cool the core would be approximately the same after 20 minutes for a break at the pump suction as for a break at the pump discharge.

It should be noted that more HPI water would be available to makeup the water boiled by decay heat in the core for the pump suction break than was assumed for the pump discharge break. The pump discharge break analyses in the May 7th report assumed that the break was between the HPI nozzle and the reactor vessel and that 30% of the total HPI flow was lost through the break.

For a pump suction break, all of the HPI water would be available to flow to the core. We therefore conclude that a minimum of 20 minutes would be available to the operator to actuate HPI and prevent core uncovery for breaks in the pump suction as well as at the pump discharge, even if all feedwater is temporarily lost, and that the operating procedures at Rancho Seco are adequate for either event.

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The above statements and opinions are true and correct to the best of my. personal knowledge and belief.

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WaltofL ensen,.Jrg Subscribed and sworn jo before me this $^^ day of 7Aw,_,1982.

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My Commission Expires Q,L_

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WALTON L. JENSEN, JR.

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PROFESSIONAL QUALIFICAT]DNS I am a Senior Nuclear Engineer in the Reactor Systems Branch of the Nucleac Regulatory Commission.

In this position I am responsible for the technical analysis and evaluation of the public health and safety aspects of reactor.

systems.

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From June 1979 to December 1979, I was assigned to the Bu11etins and Orders Task Force of the Nuclear Regulstory Commission.

I participated in the preparation of HUREG-0565, " Generic Evaluation of Small Break Loss-of-Coolant Accident Behavior in Babcock & Wilcox Designed 177-FA Operating Plants."

From 1972 to 1976, I was assigned to the Containment Systems Branch of the NRC/AEC, and from 1976 to 1979, I was assigned to the Analysis Branch of the In these positions I was responsible' for the development,and evaluation NRC.

of computer programs and techniques to calculate the reactor system and containment system response to postulated loss-of-coolant accidents.

From 1967 to 1972, I was employed by the Babcock and Wilcox Company at Lynchburg, There I was lead engineer for the development of loss-of-coolant Virginia.

computer programs and the qualification of these programs by comparison with experimental data.

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from 1963 to 1957,1 was employed by the Atomic Energy Commission in the Division of Reactor Licensing.

I assisted in the safety reviews of large I

power reactors, and 1 led the reviews of several small research reactors.

I received an M.S. degree in Nuclear Engineering at the Catholic University of A,7. erica in 1968 and a B.S. degree in Nuclear. Engineering at iiississippi State Uni.versity.in 1963.

I am a graduate of the Oak Ridge School for Reactor Technology, c.

1963-1964.

I am a member of the American Nuclear Society.

I am the author of three scientific papers dealing with the response of B&W reactors to loss-of-Coolant Accidents and have authored one scientific paper dealing with containment analysis.

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