ML20140G133

Forwards RAIs 230.133 - 230.188 Re AP600 Advanced Reactor Design in Civil Engineering & Geosciences Branch Area

ML20140G133
Person / Time
Site:	05200003
Issue date:	05/08/1997
From:	Diane Jackson NRC (Affiliation Not Assigned)
To:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 9706130372
Download: ML20140G133 (5)

Text

__ _ _ .

. . 1 May 8, 1997 l l

Mr. Nicholas J. Liparulo, Manager Nuclear Safety and Regulatory Activities Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, PA 15230

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION (RAI) ON THE AP600 ADVANCED REACTOR DESIGN IN THE CIVIL ENGINEERING AND GEOSCIENCES BRANCH (ECGB) REVIEW AREA

Dear Mr. Liparulo:

The Nuclear Regulatory Commission (NRC) ECGB staff has determined that it needs edditional information in order to complete its review of the Westing-house AP600 advanced reactor design. Enclosed are RAls #230.133 - 230.138 regarding the ECGB review of the Westinghouse submittal, dated March 18, 1997, of a draft seismic margin analysis (SMA) chapter and the responses to RAls.

The staff notes that RAI #230.211 (0ITS #3437 - draft safety evaluation report Section 19) was not answered in the March 18, 1997, letter and requests that the RAI be answered. To maintain the schedule, the staff believes Westing-house needs to fully respond to the RAls before the end of June to support a design review meeting in early July.

If you have any questions regarding this matter, you may contact me at (301) 415-8548.

Sincerely, original signed by:

Diane T. Jackson, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003

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DATE 05/ 9'/97 L 05/l/97 9706130372 970500 0FFICIAL RECORD COPY PDR ADOCK 05200003 A PDR

Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Mr. Ronald Simard, Director Advanced Plant Safety & Licensing Advanced Reactor Programs Westinghouse Electric Corporation Nuclear Energy Institute Energy Systems Business Unit 1776 Eye Street, N.W.

P.O. Box 355 Suite 300 Pittsburgh, PA 15230 Washington, DC 20006-3706 Ms. Cindy L. Haag Ms. Lynn Connor Advanced Plant Safety & Licensing Doc-Search Associates Westinghouse Electric Corporation Post Office Box 34 Energy Systems Business Unit Cabin John, MD 20818 Box 355 Pittsburgh, PA 15230 Mr. James E. Quinn, Projects Manager LMR and SBWR Programs Mr. M. D. Beaumont GE Nuclear Energy Nuclear and Advanced Technology Division 175 Curtner Avenue, M/C 165 Westinghouse Electric Corporation San Jose, CA 95125  :

One Montrose Metro l 11921 Rockville Pike Mr. Robert H. Buchholz  !

Suite 350 GE Nuclear Energy l Rockville, MD 20852 175 Curtner Avenue, MC-781 Mr. Sterling Franks l U.S. Department of Energy Barton Z. Cowan, Esq.

NE-50 Eckert Seamans Cherin & Mellott 19901 Germantown Road 600 Grant Street 42nd Floor Germantown, MD 20874 Pittsburgh, PA 15219 Mr. S. M. Modro Mr. Ed Rodwell, Manager Nuclear Systems Analysis Technologies PWR Design Certification Lockheed Idaho Technologies Company Electric Power Research Institute ,

Post Office Box 1625 3412 Hillview Avenue l Idaho Falls, ID 83415 Palo Alto, CA 94303 Mr. Frank A. Ross Mr. Charles Thompson, Nuclear Engineer U.S. Department of Energy, NE-42 AP600 Certification Office of LWR Safety and Technology NE-50 19901 Germantown Road 19901 Germantown Road Germantown, MD 20874 Germantown, MD 20874

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REQUEST FOR ADDITIONAL INFORMATION  !

CIVIL ENGINEERING AND GEOSCIENCES BRANCH I

PRA SSAR CHAPTER 55: SEISMIC MARGIN ANALYSIS WESTINGHOUSE AP600 DOCKET NO.52-003 230-133 Probabilistic Fragility Analysis

a. In the application of the probabilistic fragility analysis to the reactor pressure vessel (RPV) and steam generators, it appears that the variability of the floor responses are not properly accounted for. The calculated 8, values of 0.27 and 0.29 in Table 55-1 are considered to be too low in comparison with typical values of 0.5 to 0.6 from past, generic seismic probabilistic risk assessment (SPRA) studies.

For the case of RPV, by assuming a 8, value of 0.50 and the i median value of 1.44 g (Table 55-1), the corresponding HCLPF (high confidence, low probability of failure) value would be 0.46 g, which is about 40 percent lower than the calculated HCLPF value of 0.77 g. Westinghouse needs to provide the rationale for the nonconservative evaluation of variabili- l ties. If it is intended to use conservative floor response spectra to compensate for this nonconservative assumption (i.e., low B.), explain quantitatively that the net results for the HCLPF calculations are still conservative,

b. Westinghouse needs to provide the buckling equation used for the fragility analysis of containment vessel.

230-134 Conservative Deterministic Failure Margin (CDFM) Method Westinghouse states that the inelastic energy absorbing factor, F , is estimated for the column structural elements in the shield b"uilding roof, for which the EPRI CDFM approach is used. It also states that an additional margin factor is considered to account for a higher damping value due to inelastic responses. However, the formulation for the F, factor in the standard safety analysis report (SSAR) should be used to modify the linear responses for which a linear (lower) damping value (e.g., 7 percent for concrete structures) is used. To account for both the F factor and a higher damping value is considered to be a doub,le counting of the nonlinear response effects and should be avoided.

230-135 Test Results Regarding the use of test data, the test response spectra should be used at about the 99-percent exceedance probability level for the capacity according to Appendix Q to Reference 1. This results Enclosure

l

! i i

in a lower HCLPF value. Westinghouse needs to provide the ratio-4 nale for not using the 99-percent exceedance probability level for l

test response spectra.

l 230-136 Generic Fragility Data The use of generic data is indicated for several components based on Reference 2. However, the suggested generic fragility values are intended for a preliminary analysis only. These generic values should not be used for critical components which are important to plant risks. In addition, for components with new design features, it should be confirmed that the new design features do not potentially contribute to lowering fragility values. An example may include the fuel rods, for which some differences in design (e.g., different outside diameter and additional gas space below the fuel pellets) are observed compared I with the typical four-loop design.

The use of generic data is considered inappropriate for the following components:

Reactor Internals: Critical component (this component represents the plant HCLPF) and new design features.

CRDM: Past PRA/SMA studies indicate significant )

variations in estimated fragility values of l CRDM.

Valves: Describe the classification (e.g., motor-operated or manual) and elevation of location.

Main Control Room Operation: New design features should be addressed.

230-137 Response to 0ITS #3432

a. Westinghouse states that, " Response for structures (SG supports and RPV supports) is from time history analyses and not response spectra. Therefore, mode combination fragility parameters are not appropriate." If the time history simu-lation is used, Reference 3 (Page 3-20) recommends that the associated uncertainty (8,) of %1n(Saa/Sa%) be used in the vicinity of the fundamental structure frequency. West-inghouse needs to provide the rationale for not using this uncertainty for the probabilistic fragility analysis.

b. Westinghouse states that, "The combination of earthquake components is not considered for the critical support struc-tures because the seismic load is dependent primarily on a single earthquake component." Howeaer, Reference 3

O (Page 3-26) recommends that a randomness (B,) for response be included in the fragility analysis since the actual response will be higher or lower and provides an upper bound value of 8, (0.18) for the cases where the response is primarily from a single direction and a typical value of 8, (0.15) for building response due to the effects of earth-quake component combination. Provide the rationale for not using this randomness for the probabilistic fragility analy-s i s .~

230-138 Typographical Errors

a. Probabilistic Fragility Analysis (1). A and X i are stated as the mean peak ground capacity and the i-th design mean margin factor, respectively.

However, should they represent median (not mean) values to use log-normal distributions?

i (2). Sam is stated as the spectral acceleration value associ-  ;

ated with mean-centered damping. However, should it be '

median-centered value to compute the median damping factor?

b. Conservative Deterministic Failure Margin Method j i

Put in the right hand side of F, equation. l l

c. Provide Sections 55.6 through 55.8  ;

d. Table 55-1 (1). Clarify where the valve HCLPF value at Room Number 11400 is obtained. Is it from Reference 2 or deterministic approach?

(2). This table provides two HCLPFs (0.979 and 0.80 9 ) for the main control room switch station. Westinghouse needs to clarify which one will be used.

REFERENCES

1. "A Methodology for Assessment of Nuclear Power Plant Seismic Margin,"

Revision 1, EPRI NP-6041-SL, August 1991.

2. Advanced Light Water Reactor Utility Requirements Document, Volume III, ALWR Passive Plant, Chapter 1, Appendix A, PRA Key Assumptions and Groundrules. Revisions 5 & 6, Issued 12/93.

3. " Methodology for Developing Seismic Fragilities," EPRI TR-103959, June, 1994.

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