ML20236D244

Forwards Request for Addl Info Re Probabilistic Safety Study for RESAR SP/90 Applications.Response Requested within 60 Days

ML20236D244
Person / Time
Site:	05000601
Issue date:	07/24/1987
From:	Kenyon T Office of Nuclear Reactor Regulation
To:	Johnson W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 8707300371
Download: ML20236D244 (5)

Text

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.. .. July 24o 1987

~ Docket No. 50-601 L

I Mr. W. J. Johnson, Manager Nuclear Safety Department Westinghouse Electric Corporation Water Reactor Division Box 355 Pittsburgh, Pennsylvania 15230

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THC PROBABILISTIC j SAFETY STUDY FOR THE RESAR SP/90 APPLICATIONS As'a result of our ongoing review of the Probabilistic Safety Study (PSS) for the RESAR SP/90 application, we require additional information in order:to complete our review of the "back-end" study. Enclosed are review questions Q 720.39 - 720.48.

Please respond to this request within 60 days of the date of this letter. If you have any questions regarding this matter, call me at (301)492-8206.

Sincerely original signed by Thomas J. Kenyon, Project Manager Standardization and Non-Power Project Directorate Division of Reactor Projects III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page DISTRIBUTION,

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gg ji E WASH'NGTON, D. C. 20555 sY July 24, 1987 Docket No. 50-601 l

Mr. W. J. Johnson, Manager '

Nuclear Safety Department Westinghouse Electric Corporation Water Reactor Division j Box 355 )

Pittsburgh, Pennsylvania 15230 l I

Dear Mr. Johnson:

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SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROBABILISTIC SAFETY STUDY FOR THE RESAR SP/90 APPLICATION As a result of our ongoing review of the Probabilistic Safety Study (PSS) for the RESAR SP/90 application, we require additional information in order to complete our review of the "back-end" study. Enclosed are review questions Q 720.39 - 720.48. ,

Please respond to this request within 60 days of the date of this letter. If j you have any questions regarding this matter, call me at (301)492-8206.

Sincerely

/

Thomas J. Ken n,Proj

) Manager Standardization and Non-Power Project Directorate -

Division of Reactor Projects I3I, IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page n

F Docket No. STN 50-601

.' RESAR-SP/90 cc:

Trevor Pratt Brookhaven National Laboratory Building 130 Upton, New York 11973 Mr. William Schivley Westinghouse Electric Corporation ECE-410 Mail Stop 4-08 Box 355 Pittsburgh, Pennsylvania 15230 ,

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l Enclosure REQUEST FOR ADDITIONAL INFORMATION REGARDING THE PROBABILISTIC SAFETY STUDY RESAR SP/90 720.39 Provide an inventory of elements for 40 species in kg [per ' Table 1  !

(p. 4, STCP User's Guide (MOD 1)) NUREG/CR-4587, July 1986].

720.40 Provide engineering drawings of:

i) Reactor cavity and junction flow areas; ii) Reactor building cross-sections (concrete only) showing wall penetrations; iii) Elevation floor plans; l iv) Overall ventilation system for the reactor building; and i I

v) The 4-V MAAP nodalization illustrating junction flow area, structure heat sink composition, areas and thicknesses, and junction elevations.

720.41 Provide the containment failure pressure distribution or mean and standard deviation as used in the PSS.

720.42 What is the PORV capacity per valve at design pressure (Table 5.4-13)?

720.43 What is the probability of restoration of AC power as a function of time from accident initiation? l 720.44 If the containment fails b mode, failure location (s),yand overpressure, area of the breach? what is the likely failure 720.45 Do sprays survive a hydrogen burn? What is the probability of their successful operation?

720.46 Provide the following detailed output from the MAAP calculations: -

1) The transient output in table fono and plots for TE, SE, and SEFC; ii) The molar gas concentrations in the four compartments; iii)

The temperature of the gas and of the surrounding structure in the four compartments; iv) The containment leak rate; v) The gas release rate from corium/ concrete interactions and the total release; vi) The aerosol generation rate in corium/ concrete interactions; vii) The fraction of I, Cs, and Te released in-vessel prior to l

! vessel failure, retained in-vessel after vessel failure, and the ex-vessel release to containment after vessel failure; also revaporization from vessel after vessel failure as a function of time; and viii) The temperature of corium as a function of time 720.47 What is the penetration size and failure mode due to over-temperature?

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. 720.48 For severe accidents where the reactor vessel fails while the primary coolant system is still.at high pressure, direct containment heating has been identified as a potential means for early '

containment failure. There are large. uncertainties in the methods used to make current predictions of the likelihood and consequences of direct heating. However, if the vessel can be depressurized before failure occurs, direct heating is precluded. Some recent studies by.the NRC and its contractors and by the industry _have 1 indicated that there a~re a number of potential schemes for ensuring {

such depressurization. Have these " accident management" schemes been 4

considered for the SP/90 design, and if so, please describe them along with any conclusions that have been reached regarding their viability?

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