ML20234C996

Affidavit of P Huang on Mass Transfer & Flashing Fraction for Steam Generator Tube Rupture at 5% Power.* Related Info Encl

ML20234C996
Person / Time
Site:	Seabrook
Issue date:	11/20/1987
From:	Huang P WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20234C672	List: ML20234C670 ML20234C720 ML20234C827 ML20234C848 ML20234C869 ML20234C872 ML20234C894 ML20234C912 ML20234C996 ML20234D014
References
OL, OL-1, NUDOCS 8801060408
Download: ML20234C996 (4)

Text

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Dated November 20, 1987 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION before the ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )

. )

PUBLIC SERVICE COMPANY OF ) Ibcket Nos. 50 443-OL NEW HAMPSHIRE, et al. ) Docket Nos. 50-444-OL

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Seabrook Station, Units ) On-site Bnergency Planning 1 and 2) .) and Safety Issues 1

. AFFIDAVIT OF PING HUANG ON MASS TRANSFER AND FLASHING FRACTION FOR A STEAM GENERATOR TUBE RUPTURE AT 5 PERCENT POWER I, Ping Huang, being duly sworn, depose and state:

1. I am employed by Westinghouse Electric Corporation as a Senior 4

Engineer in 0perational Safeguards Engineering in the Nuclear

~

Technolgy Systems Division in the Power Systems Business Unit.

2. My professional qualifications are attached hereto and marked "A".

3 The purpose of my affidavit is to present the integrated break flow and flashing fraction from an analysis of a steam generator tube rupture during steady state reactor operation at 5% of full power when the reactor has not operated above that power level.

4. The' values for these parameters were determined for the reactor coolant system (RCS) and steam generator conditions at 5% of full power using the break flow model in the LOFTRAN program for a thirty c )

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, minute transient as was used in the steam generator tube rupture analysis reported in the FSAR.

5. Conservative values of the depressurization rate prior to reactor trip and RCS pressure and break flow rate after reactor trip were used.

6. Prior to reactor trip, the full power RCS depressurization rate which is lower than the depressurization rate at 5% power prior to reactor trip was used. This lower depressurization rate, which takes no credit for the effects of higher break flow rates resulting from lower temperatures in the hot and cold legs of the RCS at 5%

power, is conservative because it results in in higher mass transfer-from the RCS to the faulted steam generator.

7. In order to simplify the analysis, the equilibrium RCS' pressure and break flow rate corresponding to the condition for which the flow out of the break equals the Safety Injection flow into the RCS was used throughout the transient after reactor trip until the event is

. terminated at thirty minutes rather than to model the pressure and flow rate transients as was done in the FSAR analysis. This is conservative because it does not take credit for the transient that would occur after reactor trip. . 'Ihe depressurization of the RCS due to the loss of RCS inventory will result in a safety injection (SI) actuation signal shortly after reactor trip. SI flow causes the an increase in RCS pressure until the equilibrium conditions are reached. Thus the mass transfer from the RCS to the faulted steam generator based on asstning the equilibrium conditions throughout

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~ the transient after reactor trip'is higher than the value that would

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have been obtained if the pressure and flow rate transients had been explicitly modeled.

8. The integrated mass transfer from the Reactor Coolant Systen (RCS) to the faulted steam generator resulting from the analysis of a steam generator tube r*Jpture for the Seabrook Nuclear Power Plant for steady state operation at 5% of full power does not exceed 140,000 lbs. The break flow flashing fraction does not exceed 7.5%.

Further affiant sayeth not fjay; lch@y -

Ping Huang SUBSCRI1ED AND to b 9fore me this g day o #N414N(1987.

. Lilli-&A_lDLL J' t/ NOTARY PUBLIC-

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._____________-___________-___-_------__--_A

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Ping H. Ikaa

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i My name is Ping H. Huang. My business ad:!ress is P.O. Sex 355, Pittsburgh, Pennsylvania 15230. ! am employed by ' westinghouse Electric Corporation as a Senior Engineer in Operational Safeguards Engineering of the Nuclear Technology System Division. I sm a registered Pr0fessi0nal Engineer in the State of Pennsylvania.

I graduat,ed f:v.i. Nati nal Tsinghuo University of Tai'an in 1976 '4th a B.S. Degree in Nuclear Engineering. In 1990, I received a M.S. Degree in Nuclear Engineering from Pennsylvania State University. I have been a part time graduate student at Carnegie Mellen University from 1981 to present to pursue a %ctorate regree in Nuclear Science and Engineering.

From March 19% to January 1983, I was employed by the Nuclear Fuel Division of Westinghouse Electric Corporation as a Nuo'. car E gineer.

My responsibilities included core design and fuel management for the fuel relcads, and development of new products to reduce fuel cycle ecs s and improve the margin to safety limits.

n January 1963, I accepted a position in the Nuclear Safety Department of the Nuclear Technology System Civision of Westinghouse Electric Corporation as a Safety A.alysis Engineer. My experience includes 1.arge and kall Break less of Coclant Accident analysis, Steam Generator Tube Rupture (SGTR) analysis and deveicpment of Emergency Operating Procedures. Since December 1983 to present, my major assignments at Westinghouse have been related to the analysis of the Steam Generater Tube Rupture Accident. I have regularly performed  ;

SGTR analysis and evaluation for Westinghouse pressurizer water ,

reactors. I have also been involved in the development of a new SGTR )

analysis methodology to resolve all the licensing issues related to l the SGTR accident. The new SGTR analysis methodology was approved by the Nuclear Regulatory Coccission in March 1987.

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