ML20148S645

Rebuttal Testimony of Aj Levine Concerning Questions 2-1 & 2-2 Re Subj Facil

ML20148S645
Person / Time
Site:	Black Fox
Issue date:	11/20/1978
From:	Levine A GENERAL ELECTRIC CO.
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ML20148S636	List: ML20148S645
References
NUDOCS 7812040018
Download: ML20148S645 (8)

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

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PUBLIC SERVICE COMPANY OF OKLAHOMA, )

ASSOCIATED ELECTRIC COOPERATIVE, INC. ) Docket Nos. STN 50-556 AND WESTERN FARMERS ELECTRIC ) STN 50-557 COOPERATIVE, INC. )

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(Black Fox Station, Units 1 and 2) )

Rebuttal Testimony of Aaron J. Levine Concerning Questions 2-1 and 2-2 November 20, 1978

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• REBUTTAL TESTIMONY OF AARON J. LEVINE CONCERNING QUESTIONS 2-1 and 2-2 l

My name in Aaron J. Levine and my business address is 175 Curtner Avenue, San Jose, California. I am the Manager of Projects Licensing, Unit 1 for the Safety and Licensing Operation of the General Electric Company, Nuclear Energy Business Group. I have previously testified in this proceeding with respect to Questions 2-1 and 2-2 on October 20, 1978 (See Tr. 6322-6488).

I.

Tests Using a 218 Inch Core (Question 2-1)

The confirmatory tests to be performed at the Lynn facility will use actual reactor equipment for a 218 inch diameter core to confirm the adequacy of the core spray distribution methodology to be used for a 238 inch and other diameter cores. During the course of cross-examination by Intervenors' counsel (Tr. 6408-6414), a question was raised as to whether or not confirmatory tests using a 218 inch diameter core would be representative of the Black Fox Station which will use a 238 inch diameter core. ThisSection I of my Rebuttal Testimony addresses this point.

GE intends to conduct a full-scale 360* tests using a 238 inch diameter core at its facility in Vallecitos. However, prior to conducting the tests at Vallecitos, GE will conduct tests at its facility in Lynn, Massachusetts to determine the

T separability and' independence of thermal and hydraulics effects using a 218 inch. core. These effects can be determined in a steam environment without regard to core size because this determination is not a function of whether the core size is 218, 238 or.251 inches. As I have previously testified, thermal effects are virtually completed within 5 nozzle diameters.of the core spray nozzle, and thereafter, the hydraulic effects dictate the behavior of the spray droplets. Once such separability has been confirmed, the thermal effects.can be accounted for in an air environment by modifying the nozzle design to account for the effect of that phenomenon. Then full-scale tests at vallecitos using a 238 inch diameter core will assure that proper core spray distribution is. achieved.

II.

Operating Capability of the Core Spray System for the Black Fox Station Under Certain Postulated Conditions (Question 2-1) l The Emergency Core Cooling System (ECCS) for the Black Fox Station is designed to satisfy the requirements of 10 CFR S 50.46. As such, the ECCS is designed to accommodate the worst postulated single failure in the system and still maintain the calculated peak cladding temperature (ICT) below 2200'F for the postulated design basis accident (DBA). The ECCS consists of a high pressure core spray pump, a low pressure core spray pump, three low pressure coolant injection

pumps and an automatic depressurization system.

At Tr. 6433-34, I was asked whether or not the high pressure core spray system, for a certain period of time in a post-LOCA situation, would provide the only cooling function for the core. My answer should have been "yes, for certain postulated conditions," instead of "no". During the very early sequence of events postulated during a LOCA, the high pressure spray system is the only source of coolant available until reactor coolant system pressure has decreased enough to permit flow from the lower pressure systems to enter the reactor vessel.

One might then ask - "what if the high pressure core spray system failed and it was not operational during the very early sequence of events postulated during the LOCA?" Such an occurrence is not a concern because its effect is evaluated under the single failure criterion set forth in NRC regulations, 10 CFR Part 50, Appendix A. The worst single failure is not a failed high pressure core spray system, but rather the fai3ure of the diesel which supplies power to two low pressure coolant injection pumps. The calculated peak cladding temperature is higher for that single failure than the one which assumes failure of the high pressure core spray system, and as indicated above, the ECCS design accommodates the worst single i

failure.

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- III. .

Simulator Nozzle Design-(Question 2-1)

I wish to answer more completely a question posed by. Mr. Shon (Tr. 6457) , namely, "How can one mechanically design a nozzle that will do the same thing as this vapor does to the droplets in the same distance, when this seems to occur.outside the nozzle?" This question pertains to the capability to demonstrate in air the effects of the steam environment on the spray distribution from the nozzles in a reactor suchias the Black Fox Station.

The fundamental difference in spray performance between air and steam is the presence of condensation in steam.

• The effect of condensation is to modify the spray distribution very_near the nozzle. Once the spray has traveled a short distance away from the nozzle, condensation effects are significantly less important and the spray behaves the same, whether it is in air or steam. This effect, that is, the change -

in distribution due to steam can be duplicated for testing in an air environment by changes in nozzle geometry, such as i throat diameter and nozzle 1e'ngth.

l IV.

Figure of Merit Definitions (Question 2-1)

The following exchanges between Mr. Shon and me occurred at Tr. 6457 and 6458:

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. "Q. In the attachment to the letter, which is an attachment in itself,.to.your testimony, in the questions listed therein in question ~8, in.

discussing minimum required' flow there is a -

parenthetical remark to the effect that three  !

different definitions should be used for this quantity.

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You just gave us one number. I think that was the .35 gallons per minute, wasn't it?

A. .That is the number that comes out of our tests,.to show that you can -- that the heat ,

transfer coefficient can be satisfied with that amount of flow.  !

Q. What were the three definitions, and how do they differ, do you know?

A. I just don't recollect at this time. I'd have to go back.and refresh my memory." ,

t I have nowfrefreshed my memory and I am able to provide the ,

three definitions.

The NRC. requested that GE provide a comparison of the [

minimum flow to the minimum required flow for. calculation of a Fig'ure of Merit (FOM). The*FOM is a ratio of the quantity of water provided by.the core spray to'the quantity which is calculated to be required to remove decay heat from a fuel, bundle. The FOM is sensitive to the assumed decay heat and the time at which spray flow is assumed to be available for.

cooling the bundles. The definition applies to the comparison +

of minimum flows and not to the minimum flow itself.

The first definition of the FOM uses the 1.2 times the American Nuclear Society decay heat curve as described in 10 CFR Part 50 Appendix K and the time at which rated spray flow occurs. The second definition uses the American .

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6-Nuclear Society revised decay heat curve issued in 1978 with a two sigma error band and the same time. The third definition uses the same decay heat as the second definition and 300 seconds rather than the time of rated spray. The 300 seconds is representative of the time one would expect, that using the core spray system alone, the calculated peak cladding temperature turns from increasing to decreasing.

In my answer to the first question quoted above, I stated that the minimum required flow rate indicated by tests was .35 gallons per minute. I also stated at Tr. 6346 that my memory was not clear whether the figure should be .35 or

.5 gpm. I.have checked this figure and testing was performed only to values as low as .5 gpm. Based on this data, we believe that credit for the heat transfer coefficient could be taken for flows as low as .35 gpm, although the design basis has been established at 3.25 gpm.

V.

Reverification in Program (Question 2-2)

In my previous testimony concerning Question 2-2, I stated that GE's code input reverification program was approximately 80 percent complete. The Reverification Program is now virtually complete including sample calculations to assess the integrated affect of those input changes identified in the Program. It is currently estimated from these sample

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7-calculations that as a result of the changes, the calculated ,

peak cladding temperature for the Black Fox plant would decrease 13*F from the last calculated value (2038'F). This results in a calculated peak cladding temperature of approximately 2025'F for the Black Fox Station.

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