ML20059K178

Requests for Addl Info Re AP600 Design Certification within 90 Days of Date of Ltr Receipt

ML20059K178
Person / Time
Site:	05200003
Issue date:	01/14/1994
From:	Kenyon T Office of Nuclear Reactor Regulation
To:	Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 9402010393
Download: ML20059K178 (7)

Text

,-

y pR MCh j k;7J g ' y.h}

UNITED STATES 1')

NUCLEAR REGULATORY COMMISSION

• f WASHINGTON. D.C. 20555 0001

/

January 13, 1994 Docket No.52-003 Mr. Nicholas J. Liparulo Nuclear Safety and Regulatory Activities Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON THE AP600 As a result of its review of the June 1992, application for design certifica-tion of the AP600, the staff has determined that it needs additional informa-tion in order to complete its review. The additional information is needed in the areas of software common-mode failure (Q420.123) and reactor systems (Q440.49-Q440.51). The requests for additional information regarding reactor systems were prompted by the recent changes in the design and operating procedures for the automatic depressurization system in the AP600 design.

Enclosed are the staff's questions.

Please respond to this request within 90 days of the date of receipt of this letter.

You have requested that portions of the information submitted in the June 1992, application for design certification be exempt from mandatory public disclosure. While the staff has not completed its review of your request in accordance with the requirements of 10 CFR 2.790, that portion of the submitted information is being withheld from public disclosure pending the staff's final determination.

The staff concludes that this request for additional information does not contain those portions of the information for 1

which exemption is sought. However, the staff will withhold this letter from public disclosure for 30 calendar days from the date of this letter to allow Westinghouse the opportunity to verify the staff's conclusions.

If, after that time, you do not request that all or portions of the information in the enclosures be withheld from public disclosure in accordance with 10 CFR 2.790, this letter will be placed in the NRC's Public Document Room.

NRC RE CENHH CSPY 310048 9402010393 940114 la PDR 'ADOCK 05200003 A

PDR d

i Mr. Nicholas J. Liparulo January 14, 1994 This request for additional information affects nine or fewer respondents, and therefore, is not subject to review by the Office of Management and Budget under P.L.96-511.

If you have any questions regarding this matter, you.can contact me at (301) 504-1120.

Sincerely, original signed by:

Frederick W. Hasselberg Thomas J. Kenyon, Project Manager Standardization Project Directorate Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page DISTRIBUTION:

Central-File

• PDST R/F RBorchardt DCrutchfield PDR*

MSiemien, 15B18 Alevin, BE23 GHsii, 8E23 TCollins, 8E23 HLi, 8H3 MChiramal, 8H3 MPohida, 10E4 WTravers ACRS (11)

RArchitzel PShea TKenyon FHasselberg KShembarger JMoore, 15B18 WDean, 17G21 GSuh (2), 12E4

• DO NOT DISPATCH FOR 30 DAYS OFC LA:PDST:ADAR PM:PDST:ADAR PM:PDST:ADAR SC:PDST:ADAR NAME PShea 9 S TKenyon:h[h FHasselkik RArchitzeL.M 01/N/9k 01/d/94 01/(7;/94 Ol/tb/94 Y DATE OFFICIAL RECORD COPY:

SRXB.RAI b

J

i Mr. Nicholas J. Liparulo January 14, 1994 This request for additional information affects nine or fewer respondents, and therefore, is not subjectfio review by the Office of Management and Budget under P.L.96-511.

If you have any questions regarding this matter, you can contact me at (301) 504-1120.

i Sincerely, ll, u

l-v% "M h

Thomas J. Kenyon, Project anager Standardization Project Directorate

?

Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation

Enclosure:

As stated I

cc w/ enclosure:

See next page

.?

i 9

t i

P

}

\\

Mr. Nicholas J. Liparulo Docket No.52-003 Westinghouse Electric Corporation AP600 cc: Mr. B. A. McIntyre Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit P.O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. John C. Butler Ad'.anced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Box 355 Pittsburgh, Pennsylvania 15230 Mr. M. D. Beaumont Nuclear and Advanced Technology Division Westinghouse Electric Corporation One Montrose Metro 11921 Rockville Pike Suite 350 Rockville, Maryland 20852 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.

20585 Mr. S. M. Modro EG&G Idaho Inc.

Post Office Box 1625 Idaho Falls, Idaho 83415 l

Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.

Room 8002 l

Washington, D.C.

20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 Mr. Victor G. Snell, Director Safety and Licensing AECL Technologies 9210 Corporate Boulevard Suite 410 Rockville, Maryland 20850

t REQUEST FOR ADDITIONAL INFORMATION

.ON THE WESTINGHOUSE AP600 DESIGN Instrumentation and Controls 420.123 The probabilistic risk assessment (PRA) for the AP600 design assumes that software common-mode failure among instrumentation and control (I&C) cards result in an unavailability for I&C cards within subsys-tems of 1.2x10 failures per demand (f/d).

Software common-mode failures within safety and control subsystems result in an unavail-ability of 1.1x10'5 f/d (Table E-5.3 of the PRA).

The probability of failures / demand for I&C cards is two decades lower than the value man test program (10y experts agree can be demonstrated by a practical f/d). The probability of failures / demand for subsystems (made up from a collection of cards and connections between cards) is at least one order of magnitude better than the practical test value, and two orders of magnitude better than the value being considered by the Nuclear Installatiop Inspectorate (NII) of the United Kingdom for the Sizewell B PPS (10' f/d).

The summary of results for the internal events Level 1 analysis at power (Section 8.2 of the PRA) show that the common cause hardware failure of I&C cards is a significant contributor to the core damage frequency for most of the at-power initiating events. The data in-Table E-5.3 gives values for hapdware board CHF unavailability in the range of 1.2x10'6 f/d to 4.

f/d.

A software common-mode board failure in the range of 10'jx10'f/d to 10'3 f/d (instead of the value of 10 f/d used in the PRA) could have a significant impact on the core damage frequency results for several of the initiating events given in Table 8-1.

Perform a sensitivity assessment of the effect on the core damage frequency for events studied in the PRA resulting from software common-mode failures on I&C card unavailability in the range stated above.

Provide a description of the test program that Westinghouse 10'proposingtodemonstrateacommon-modecardunavailabilityof is Reactor Systems 440.49 In late 1993, Westinghouse proposed changes to the design of the AP600 automatic depressurization system (ADS) and the manner in which it will be operated. As a result of these changes, the staff con-cludes that Westinghouse should reevaluate the design of the ADS test facility, both in terms of hardware and configuration.

Enclosure

The staff understands that Westinghouse's current plans are to test one valve in each of trains 1-3, with the second valve represented by an orifice.

For the first stage, the orifice.is upstream of the valve, while for stage 2 and 3, the orifice is downstream of the valve. The staff is concerned with this approach because:

a.

The " critical flow" behavior of an orifice is substantially i

different from that of a nozzle. Orifices do not " choke,"

although the flow rate does exhibit a limiting behavior as upstream pressure is increased. On the other hand, short nozzles do " choke." To a first approximation, a valve appears to resem-ble a converging-diverging nozzle more than it does an orifice.

Accordingly, in the absence of a second valve in the test train, i

the staff recommends that Westinghouse replace the orifice with a short nozzle that has approximately the same length and minimum flow area as the valve body, b.

Previously, both valves in an ADS stage were to be opened simul-taneously. Having a valve in the upstream position would maxi-mize the upstream pressure seen by an ADS valve, and allow testing over a greater range of pressures.

The new operating procedure, however, calls for the upstream " isolation" valve in an ADS stage to be opened first, and the downstream " control" valve to be opened thereafter. Thus, the configuration. suggests a fixed nozzle upstream of the valve to be opened. The staff i

recommends that, in concert with item (a) above, the' valve to be l

tested be positioned downstream of the nozzle simulating the open isolation valve m each stage.

q Westinghouse should also revlm the test matrix for the ADS valves in light of the proposed design change to ensure that the range of test I

parameters still adequately covers the range of thermal-hydraulic conditions that the valves are expected to experience in the AP600 pl ant.

i 1

440.50 The new configuration of the ADS announced in late 1993 appears to-i have implications regarding the test' matrices for the test programs for the OSU/ APEX and SPES-2 facilities. While the design change in stages 2 and 3 appears to be relatively easy to account for in the two integral facilities, the new design of the 4th stage of the ADS means that the single failure behavior of the AP600 (i.e., failurti 9f one 4th stage ADS valve) is substantially different, in that such a failure no longer completely eliminates the venting capacity of one:

train of the 4th stage, but simply reduces it.

Furthermore, this may completely change the limiting single failure for the AP600 design over a range of design basis accidents involving depressurization (see Q440.51).

j How does Westinghouse plan to account for the change in the ADS i

design and potential changes in limiting single failures for those tests in the SPES-2 and OSU/ APEX facilities in which the limiting single failure is to be simulated?

l I

'I

.z 440.51 The new configuration of the ADS announced in late 1993 appears to have implications regarding the single failure assumptions for AP600 accident analyses.

For most accidents, the most limiting single failure has been assumed to be one valve of the 4th stage of the ADS, resulting in the loss of venting capability in the affected train.

The new configuration of stage 4 of the ADS, however, results in a-single failure of an ADS valve that does not eliminate the venting capability of that entire train, which may mean that the failure of a stage 4 ADS valve is no longer the most limiting single failure for many, if not all, design basis accidents involving depressurization.

Reanalyze the Chapter 15 events involving actuation of the ADS to determine the most limiting single failure with the new ADS configu-ration.

9 r