ML20141B999

Requests That Proprietary AP600 Response to Requests for Addl Info & Supplemental Info to RAI Responses,Be Withheld, Per 10CFR2.790.W/encl 1 to Westinghouse Ltr DCP/NRC0921

ML20141B999
Person / Time
Site:	05200003
Issue date:	06/19/1997
From:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19317C413	List: ML20141B999 NSD-NRC-97-5195, Forwards Proprietary Response to NRC Requests for Addl Info on AP600.Encl 2 Provides Hardcopy of TEXAS-V Input Deck,Ref in Response to RAI 720.408.Response Withheld,Per 10CFR2.790
References
AW-97-1126, NUDOCS 9706240263
Download: ML20141B999 (45)

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Westinghouse Energy Systems 3

$ 3 4 gnn,g n g 3 ,33 Electric Corporation AW-97-1126 June 19,1997 Document Control Desk

'U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: MR. T. R. QUAY APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION ; ROM PUBLIC DISCLOSURE

SUBJECT:

AP600 RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION AND SUPPLEMENTAL INFORMATION TO THE RAI RESPONSES

Dear Mr. Quay:

The application for withholding is submitted by Westinghouse Electric Corporation (" Westinghouse")

pursuant to the provisions of paragraph (b)(1) of Section 2.790 of the Commission's regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.

The proprietary material for which withholding is being requested is identified in the proprietary version of the subject report. In conformance with 10CFR Section 2.790, Affidavit AW-97-1126 accompanies this application for withholding setting forth the basis on which the identified proprietary information may be withheld from public disclosure.

Accordingly, it is respectfully requested that the subject information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations.

Correspondence with respect to this application for withholding or the accompanying affidavit should reference AW-97-Il26 and should be addressed to the undersigned.

Very truly yours, (m O i

Brian A. McIntyre, Manager Advanced Plant Safety and Licensing jml cc: Kevin Bohrer NRC OWFN - MS 12E20 ___

T 9706240263 970619 PDR ADOCK 05200003.

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l AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

SS COUNTY OF ALLEGiiENY:

Before me, the un.dersigned authority, personally appeared tienry A. Sepp, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation (" Westinghouse") and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge,'information, and belief:

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1 lienry A. S'epp, Ma(nage/ r R Regulatory and Licensing Engineering Engineering Sworn to and subscribed before m this / f d day of gd_ ,1997 p

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Amb "&A-' 4gbeM:ypdn  ! o Notary Public AssmanoWNmaY l

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AW-97-1126 '

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-(1) I am Manager, in the Nuclear Services Division, of the Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power

j. plant . licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit.

(2) i a.n making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse application for ,

withholding accompanying this Affidavit. l (3) I have personal knowledge of the criteria and procedures Wized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial or Snancial information.

l (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's I' regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

l (i) The information sought to be withheld from public disclosure is owned and has been I held in confidence by Westinghouse.

(ii) The information is of a type customarily held in conGdence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in con 0dence if it falls in one or more of several types, the release of which might result in the loss of an existing c potential competitive advantage, as follows:

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(a) The information reveals the distinguishing aspects of a process (or component, l structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a .

competitive economic advantage over other companies.

l (b) It consists of supporting data, including test data, relative to a process (or <

component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

! (c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from

. closure to protect the V stinghouse competitive position.

• is info-m - marketable in many ways. The extent to which such to competitors diminishes the Westinghouse ability to rvices involving the use of the information.

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AW-97-l l26 (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) Enclosed is Letter NSD-NRC-97-5195, June 19,1997 being transmitted by Westinghouse Electric Corporation (_W) letter and Application for Withholding Proprietary Information from Public Disclosure, Brian A. McIntyre (_W), to Mr. T. R. Quay, Office of NRR. The proprietary information as submitted for use by Westinghouse Electric Corporation is in response to questions concerning the AP600 plant and the associated design certification application and is expected to be applicable in other licensee submittals in response to certain NRC requirements for ma .pr

AW-97-1126 justification oflicensing advanced nuclear power plant designs.

This information is part of that which will enable Westinghouse to:

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(a) Demonstrate the design and safety of the AP600 Passive Safety Systems.

(b) Establish applicable verification testing methods.

(c) Design Advanced Nuclear Power Plants that meet NRC requirements.

(d) Establish technical and licensing approaches for the AP600 that will ultimately result in a certified design.

(e) Assist customers in obtaining NRC approval for future plants.

Funhar this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of similar information to its customers for purposes of meeting NRC requirements for advanced plant licenses.

(b) Westinghouse can sell support and defense of the technology to its customers in the licensing process.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar advanced nuclear power designs and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

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o - Ti:e development of the technology described in part by the information is the result of  ;

applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs.would have to be performed and a significant manpower effort, j 1

having the requisite talent and experience, would have to be expended for developing analytical methods and receiving NRC approval for those methods.

Further the deponent sayeth not.

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i Enclosure 1 to Westinghouse Letter DCP/NRC0921 1

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l NRC REQUEST FOR ADDITIONAL INFORMATION

g. m Ouestion: 230.133 (OITS #5286)

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 B, 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 B, value of 0.50 and the 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%

lower than the calculated HCLPF value of 0.77 g. Westinghouse needs to provide the rationale for the nonconservative evaluation of variabilities. 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.

Response

a. Westinghouse did not use a nonconservative evaluation of variabilities. Comparisons to typical values from past generic seismic probabilistic risk assessment studies can be misleading since they reucct a large population of different types of components having different types of failure modes, and therefore, potentially have large variability. The values used in the AP600 seismic margin analyses (SMA) for total variabilities reDecting both ,

randomness and uncertainty components are appropriate for the critical fragility modes of the primary components. The basis of this position is given below.

The governing failure modes associated with the reactor pressure vessel and steam generator are related to bolts within the primary component supports:

Steam generator upper support ring girder dange joint bolts tension failure Reactor pressure vessel support box hold down bolts - shear failure Inelastic energy absorption or ductility was not considered since the governing failure modes are local without large energy absorption capability. This reserve margin factor associated with ductility generally has a large variability which significantly contributes to the variability ,. A review of ABWR fragility data associated with the reactor pressure vessel primary component supports reported in their Standard Safety Analysis Report (Seismic Capacity Analysis - Amendment 31; 23A6100) indicated , values similar to those in the AP600 SMA.

Conservative Door response spectra was not used to compensate for " low ," values. No credit is taken for the margin at a specific site when compared to the envelop spectra for the different soil conditions. Frequency l variation in modeling was considered with the logarithmic standard deviation defined recognizing that this affects l floor response levels and introduces error. The method used to reDect this potential error in the seismic margin analysis is discussed in the PRA report section 55.2. Further, as previously stated to the NRC (eg., response

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4 NRC REQUEST FOR ADDITIONAL. INFORMATION in' to RAI 230.106) the seismic response for the steam generator and reactor pressure vessel supports is obtained from time history analyses and not response spectra analyses. A deterministic reduction factor was used to adjust the seismic margin values to account for differences between the unbroadened time-history analysis input and the latest floor response spectra (AP600 SSAR. Revision 7, Figure 3.7.2-17, Sheets I through 9). The value of 0.95 was used for the steam generator, and 0.91 for the reactor pressure vessel. Instead of making the adjustment deterministically, using the logarithmic standard deviation to account for the response spectra correction, c would become 0.31 for both the steam generator and reactor pressure vessel governing failure modes that are associated with these supports.

In conclusion, the variabilities used for the AP600 SMA are realistic and representative of values that would be obtained for a specific plant design. Further, the variability associated with these components in question were properly accounted for.

b. The buckling equation used for the fragility analysis of the containment vessel is the same as that used for the design of the containment vessel. The compressive stress evaluations are discussed in the AP600 SSAR, Section 3.8.2, Revision 11.

PRA Revision: None.

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l l NRC REQUEST FOR ADDITIONAL INFORMATION

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i Question: 230.134 (OITS #5287)

Conservative Deterministic Failure Margin (CDFM) Method l l

Westinghouse states that the inelastic energy absorbing factor, F,, is estimated for the column structural elements in the shield building 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% for concrete structures) is used. To account for both the F, factor and a I higher damping value is considered to be a double counting of the nonlinear response effects and should be avoided. i l

Response

l Westinghouse agrees that double counting of nonlinear response effects should be avoided. Consideration of both damping and ductility margin factors for the shield building roof did not result in double counting. The effective ,

damping value obtained in the inelastic energy absorption analysis was 6.9%. It was recognized in the seismic l margin analysis that there would be an added margin with cracking of the concrete that would increase the damping I value from 7% to 10%. If the effective damping value was above 10% this added margin would not exist, and if I used, then double counting would exist. As stated, this is not the case. i l

PRA Revision: None.

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NRC REQUEST FOR ADDITIONAL INFORMATION h

Question: 230.135 (OITS #5288)

Test Results ,

1 Regarding the use of test data, the test response spectra should be used at about the 99-percent exceedance probability Icvel for the capacity according to Appendix Q to Reference 1. This results in a lower HCLPF value. Westinghouse needs to provide the rationale for not using the 99-percent exceedance probability level for test response spectra.

Reference 1: "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Revision 1 EPRI NP-6041-SL, August 1991, i

Response

Provided below is the rational for not using 99% exceedance probability:

1. Westinghouse followed a deterministic approach where the lower bounds of the equipment test response spectra were used. This results in lower HCLPF values.

2. Should individual equipment TRS be used with 99% probability of exceedance, then higher HCLPF values will  ;

be obtained.

3. Westinghouse provided the lowest HCLPF values based on several test programs. Westinghouse intentionally i looked for the lowest qualified test levels. Should the test programs with the higher seismic levels be used, then l higher HCLPF values will be obtained. j j

PRA Revision: None. l i

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i d NRC REQUEST FOR ADDITIONAL INFORMATION ,

Question: 230.136 (OITS #5289) ,

I 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 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 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.

Reference 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.

Response

As stated in the Seismic Margin Analysis , PRA subsection 55.2.2.3, generic fragility data were used when insufficient information was available to define the HCLPF value. Sufficient information to define the HCLPF value is not available for the components identified in the RAI. It is also noted that generic fragility data was used for snubbers, but these components do not control the HCLPF value for the primary component supports.

It is recognized that the Utility Requirements Document for Advanced Light Water Reactor (Reference 55-10 of the Seismic Margin Analysis, PRA Chapter 55) provides a summary of generic fragility data for preliminary analysis only; however, they are representative of the anticipated capacity. Westinghouse has identified a COL item which requires verification of as-built conditions conforming to the seismic margin evaluation. It is stated in Chapter 59, Section 59.10.6: "The Cor bined License applicant referencing the AP600 certified design will confirm that the as-built plant conforms to the design used as the bases for the seismic margin evaluation."

PRA Revision: None.

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4 NRC REQUEST FOR ADDITIONAL INFORMATION

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Question: 230 137 (OITS #5290)

(RAI based on] response to OITS #3432

a. Westinghouse states that, " Response for structures (SG supports and RPV supports) is from time hi'.cory analyses and not response spectra. Therefore, n. ode combination fragility parameters are not appropriatt." If the time history simulation is used, Reference 3 (Page 3-20) recommends that the associated uncertainty (B,) of hin(SadSa ,)wbe used in the vicinity of the fundamental structure frequency. Westinghouse 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 cri;ical st pport structures because the seismic load is dependent primarily on a single earthquake component." However, Reference 3 (Page 3-26) recommends that a randomen (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 B,(0.18) for the cases where the response is primarily from a single direction and a typical value of 8,(0.15) tot building response due to the effects of earthquake component combination. Provide the rationale for not using this randomness for the probabilistic fragility analysis.

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

Response

a. In figure 3-5 cf the cited Reference 3, Sam , is below Sam. As seen in Figures 3.7.1-6 to 3.7.1-8, Sam, is above S%. Therefore,it is not appropriate to use this uncertainty. ,

b. The cited Reference 3 considers variability in seismic response due to variability in analysis methods and modeling techniques. Westinghouse did recognize that the seismic response has variability due to potential analysis and modeling errors, and therefore, variabilities were considered for seismic response in the seismic margin analysis. This is discussed in the AP600 PRA Chapter SS, Seismic Margin Analysis, Section 55.2.2.3.

PRA Revision: None.

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NRC REQUEST FOR ADDITIONAL INFORMATION 5

Question: 230.138 (OITS #5291)

Typographical Errors 4

a. Probabilistic Fragility Analysis (1). A,,, and X, 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?

(2). Sa,,, is stated as the spectral acceleration value associated with mean-centered damping.

However, should it be median-centered value to compute the median damping factor?

b. Conservative Deterministic Failure Margin Method Put p in the right hand side of F, equation.

c. Provide Sections 55.6 through 55.8 a

d. Table 55-1 l

(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 (0.97g and 0.80g) for the main control room switch station.

Westinghouse needs to clarify which one will be used.

Response

a. (1) As stated on page 55-4 of PRA Revision 9, the HCLPF is defined by a log-normal probability distribution that is a function of median seismic capacity and composite standard deviation, c. The median seismic capacity can be related to the mean seismic capacity by the expression provided on the bottom of page 55-4. The adjustment from median to mean was used to ensure the median

capcity was co Tectly used in the analysis, since some generic data sources report mean rather than median capacity.

(2) We agree that it should be median-centered damping.

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b. The equation on page 55 9 of PRA Revision 9 shows p in the right hand side of the F, equation.

c. There was a typographical error in the draft Chapter 55 provided to NRC. What is now in PRA Revision 9 as Sections 55.6 and 55.7 were incorrectly labeled as 55.9 and 55.10. There are no missing sections in

Chapter 55.

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NRC REQUEST FOR ADDITIONAL INFORMATION E'

d. (1) Yes, there is a typo on the table. The Basis column for Room Number 11400 should say [1] rather )

l than [4]. This typo will be fixed in Revision 10 of the PRA. '

l (2) Yes, there is a typo on the table. The main control room switch station HCLPF value is 0.97g. The .

other MCR switch station (at 0.80g) was incorrectly labeled); it should be labeled MCR isolation I

dampers. This will be changed in Revision 10 of the PRA.

l I I i PRA Revision:

L Page $$-5 of PRA will be changed asfollows:

. l Sa, = spectral acceleration value associated with median-centered man :::::::d damping.

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Thefollowing two changes will be made to PRA Table $51, Seismic Margin HCLPF Values

(Sheet 2 of 5): Room Number 11400, the Basis column will be changed from [4] to [1].

(Sheet 5 of 5): Third row from the end of the table is labeled "MRC Switch Station" It v.itt be relabeled "MCR Isolation Dampers" j l

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i NRC REQUEST FOR ADDITIONAL INFORMATION I

l I Question: 230.139 (OITS #5525) l HCLPF Margin for Rigid Components with Non-ductile Supports I

The SSE design load and the review level earthquake (RLE) for the AP600 are 0.30g and 0.50g, respectively.

Therefore, a HCLPF margin of 1.67 is implied for all the safety related equipment and components. To achieve this i

HCLPF margin, a median margin factor of at least 4.2 is needed. This is based on assumptions that a relatively low variability of 8, = 0.40 is used for a fragility estimate, and the seismic design is performed up to the limits of the code design allowables.

For relatively flexible / ductile components, such as piping, the design criteria in the SSAR are considered to give a sufficient margin to achieve the above median factor of 4.2. Ho,vever, for dynamically rigid components whose

support structures are considered to have a non-ductile failure mode, such as elastic buckling and shear failure in

fillet welds or anchor bolt joints, the design requirements in the SSAR may not be sufficient to provide this safety margin.

For the purpose of a generic evaluation of the seismic margin factor, a few examples of the median strength ad the j allowable stresses for non-ductile failure modes are summarized below based on available statistical data and the

! AP600 SSAR.

! AP600 Additional Median Allowable for Median Margin to failure Mode Stren gth(*" SSEs 2> Marnin achieve.4.20 I i Elastic Buckling 1.03 F, 0.73 F, 1.41 2.98 l Shear in Fillet Weld'"' O.84 F, 0.42 F, 2.00 2.10 4

Shear in Bolts'"' O.65 F, 0.35 F, 1.86 2.26 i 1: e.g., "Probabilistic Description of Resistance of Safety-Related Nuclear Structures," l B, Ellingwood, NUREG/CR 3341, May 1983.

- 2: American Institute of Steel Construction (AISC), Specification for the Design, Fabrication and Erection of Steel Safety Related Structures for Nuclear Facilities, AISC-N690-1984. The stress limit coefficient of 1.4 was used for all three failure modes.

l 3: F, is the nominal ultimate stress of weld metal 4: For A325 Bolts 4

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l As indicated above, an additional margin factor (median) of 2.1 to 3.0 is necessary to achieve the aforementioned HCLPF margin of 1.67. Provide additional information on how to ensure this seismic margin for relatively rigid components with non-ductile support structures.

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Response

The structures whose HCLPF values were determined by probabilistic fragility methods that did not consider inelastic l energy absorption (ductiiity) are the primary component supports and the containment vessel Margin factors in I excess of the range of 2.1 to 3.0 exist for these items. Further, the margin in the stress limits are within or exceed this range as seen in the summary table below for the controlling stress areas.

Component Resultant Median pc HCLPF Stress to i

. Margin (g) (g) Code )

Factor Allowable Steam Generator 6.88 1.98 0.30 1.00 3.80 Support - Buckling Steam Generator 4.35 1.25 0.29 0.65 2.60 Support Bolt Tension l

Reactor Pressure 4.99 1,44 0.27 0.77 3.00 Vessel Support - Bolt Shear Pressunzer Support 5.95 1.65 0.39 0.67 2.70 Weld Pressurizer Support 5.84 1.63 0.39 0.6 T 3.30 Strut Containment Vessel 13.75 3.29 0.67 0.70 Buckling Stress to Level C - - - -

2.11 Limit 1.67 Factor of Safety in Level C Limit PRA Revision: None.

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- 1 NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion: 720.386 1 l

RAI Related to DSER Open items 19.1.3.1-4 and 19.1.3.1-6

) Westinghouse responded to the staffs second follow-on RAls (720.329 and 720.330), regarding DSER Open items 19.l.3.1-4 and 19.1.3.1-6, by stating that post-24 hour risk is not significant without providing adequate supporting i documentation. The staff had asked Westinghouse to identify accident sequences that require long-term cooling, the  ;

actions needed to be performed by the operators and the systems that must be available to perform these actions (including operational requirements). )

l Westinghouse states that risk associated with long-term cooling is not any different for AP600 than it is for operating reactors and that such risk has been addressed and accepted by the staff. Westinghouse's argument, however, is not i

consistent with the staffs position as documented in NUREG-1242 (NRC Review of EPRI's ALWR Utility )

Requirements Document, Vol. 2, Pt.1, pages I A.2-4 and 1 A.2-5) or with the industry's position as documented in l Section 2.10 (Revision 4) of EPRI's Utility Requirement Document (URD). EPRI's URD states that " mission time 3~

is only for calculation of equipment unreliabilities: actions that must be taken beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (e.g., changes in system alignment or replenishment of water sources) shall be considered explicitly." In response to comments from )

the staff, EPRI supplemented Section 2.10 in Revision 4 to require that (1) actions that must be taken beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> be considered explicitly and (2) sensitivity of the results to selection of mission times be evaluated for systems 4

that provide long term core cooling and containment heat removal.

For an appropriate response, Westinghouse will need to address previous RAls related to long-term cooling by (1) identifying and categorizing accident sequences which require long-term cooling,(2) assessing the frequency of each accident sequence category requiring long-term cooling (i.e., before long-term cooling failure probabilities se .

considered),(3) identifying and characterizing (for each of the accident sequence categories) the operator actions that need to be performed and the systems that must be available to perform these functions, and (4) identifying in ights J and potential operational requiremems. Long-term cooling is an important part of PRA insights and sho..ld be

, addressed in both the baseline and the focused PRA.

Response

The bases for the success criteria for long term cooling as modeled in the PRA have been documented in the AP600 SSAR (Reference 720.386-1) and in WCAP-14800 (Reference 720.386-2).

For events not requiring depressurization and inventory addition (e.g., transients), the PRA models credit PRHR i

! operation as sufficient to provide long term decay heat removal when active systems such as SFW are not available.

Operation of PRHR causes the IRWST water to heat up, resulting in inventory loss through evaporation. IRWST inventory that evaporates due to PRHR operation is condensed on the containment liner, and collected in the IRWST gutter, which directs the water either to the sump during normal operation and automatically re directs the water back to the IRWST during an accident. The valves that re direct the flow are fail safe on loss of instrument air, loss of Class IE de power or loss of the PMS signal. The valves also are actuated automatically by DAS. Reference 720.386-1, subsection 6.3.2.1.1, notes that the IRWST gutter and its isolation valves are safety related, so that IRWST evaporation due to PRHR operation is collected and returned to the IRWST (even under focused PRA modeling assumptions) to ensure maintenance of sufficient IRWST inventory to support PRHR HX operation.

i T Westinghouse

. . _ _ _ _ _ _ _m .-

9 j NRC REQUEST FOR ADDITIONAL INFORMATION In the unlikely event that all AC power is lost and not recovered for 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />, ADS will automatically actuate unless the operators block this actuation. The operators are instructed to block this ADS actuation if the pressurizer level is stable, the CMTs are full, and the IRWST level is high and stable. The operators are also instructed to de-energize all loads on the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> batteries such that sufficient electrical power is preserved for ADS ac;uation in the unlikely case that might be necessary later. Therefore, PRHR, once actuated, is able to provide decay heat removal essentially indefinitely for these events without the need for subsequent operator action, except for blocking of ADS in case all  !

ac power is lost for 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />.

For events involving a loss of RCS inventory (including scenarios involving failure of operator action to biock ADS in case all ac power is lost for 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />), or loss of decay heat removal including failure of PRHR, long term cooling capability is provided by safety injection and RCS venting through the ADS. In the long term, natural recirculation from the containment sump is provided after the IRWST has injected into the RCS. For accident sequences requiring this long term recirculation cooling mode for success, the analysis in WCAP-14800: (1) identifies and categorizes success sequences requiring long term recirculation cooling; (2) estimates the frequency of each sequence with successful long term recirculatica; and (3) provides an analytical basis for declaring success for all sequences that have the potential to be risk significant if success is not demonstrated. The analyses provided in WCAP-14800 credit no operator actions or additional equipment operation beyond initiation of recirculation cooling (i.e., opening of sump recirculation valves). Further, those analyses include sufficient conservatisms in assumed initial and boundary I conditions to bound uncertainty related to the thermal / hydraulic modeling performed.

1 The analyses demonstrate successful long term recirculation cooling under bounding conditions of decay heat levels I anticipated during long term recirculation, with hardware failures that cover all potentially risk-significant success paths modeled in the PRA. The potential failure of operator action to block ADS in case all ac power is lost for 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> is also covered by these analyses, since the analyses assume decay heat levels substantially higher than would be expected after 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />. He potential effect on core damage frequency and large release frequency, for foth 5he focused PRA and the baseline PRA, due to sequences for which long term recirculation capabihty was got demonstrated by the analyses,is very small, and no new insights beyond those already reported in the PRA were obtained through this analysis.

References 720.386-1 and -2 validate the PRA success criteria for long term cooling. No operator actions, beyond those that may be performed as backup actions to initiate system operation (as already modeled in the PRA) are required to ensure continuous long term cooling.

References:

720.386-1 AP600 Standard Safety Analysis Report, Revision l3.

120.386-2 WCAP l4800, AP693 PRA ThermaVHydraulic Uncertainty Evaluationfor Passive System Reliability, June 1997.

PRA Revision: None.

720.386-2 W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION j

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4 Question: 720.392 As for the internal events analysis, the SMA is performed assuming a 24-hour mission time. However, in the SMA 5

a discussion, in the form of a sensitivity study, is presented assuming 72-hour mission time. In such " sensitivity" study, however, actual plant behavior during a loss of offsite power (LOSP) event (at 0.09g review level) is not modeled. Although it is mentioned that the operator may decide to stop the ADS actuation timer at approximately 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />, this is not modeled in the event tree. Furthermore, due to the assumption of active system unavailability, l

eventually the plant will need to be depressurized to use IRWST injection and sump recirculation for long-term 1 2 cooling (even if the operator " blocks" ADS actuation at 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />). Therefore, an important mixed cutset in the l l analysis should be: Seismically induced LOSP (0.09g) initiating event followed by non-seismic failure of ADS or f IRWST injection or sump recirculation. Please address this issue in your SMA.

i l Response: i l

In a EQ-LOSP event,if the PRHR is successful, and no RCS leak exists, the operators evaluate the plant condition.

and the RCS and IRWST inventories to decide whether to stop the ADS timer actuation before 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> into the event, and to conserve several hours of battery capacity to allow possible actuation of the ADS / gravity injection / cavity recirculation later in the event, if the need arises (e.g., not enough water in IRWST or RCS leak).

. In the mean time, PRHR would continue to provide the success path. This operator action is labeled as OP-ADS- l STOP. ,

If the PRHR is successful, and the operators do not take action (OP-ADS-STOP fails), then the ADS actuates by a

timer in 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> into the event (the gravity injection valves are also opened at the same time), followed by opening  ;

of two recirculation paths in 40 minutes. Thus, if auto ADS actuation is successful, then the IRWST injection, and l sump recirculation functions must work. If ADS function totally fails, then IRWST injection and sump recirculation l l are not needed. Note that the capability for long term cooling was demonstrated by analysis as shown in WCAP-j 14800,"AP600 PRA Thermal / Hydraulic Uncertainty Evaluation for Passive System Reliability."

f It is also important to note that the ability of PRHR to remove decay heat and preclude the need for IRWST injection and reci rculation cooling is supported by a recent desigrtchange that makes the IRWST gutter and isolation valve system safety-related. Per SSAR Revision 13, subsection 6.3.2.1.1, the valves in these gutters will close on PMS or DAS FRHR actuation signal so that, as PRHR heats up the IRWST and water evaporates into containment, a sufficient amount is condensed on the containment shell and returned, by the gutter system, to the IRWST, His provide sufficient PRHR capability to allow continuous operation for long-term cooling without the need for depressurization and IRWST injection.

The mixed cutsets resulting from the failure of OP-ADS-STOP were neither modeled, nor discussed, since they do not affect the plar:t HCLPF, as stated in the SMA submittal section. Dere are no models that are custom made to examine such cutsus for the 72-hour period, since these failures are assessed to be of small risk significance.

However, to be responsive to the current RAI, an evaluation of the dominant mixed SMA cutsets, based on available models is presented below.

3 W85tiflgh00S8

-_ _ __ __ = -. . - - - - . - - - . - . . _ __ ---- - - . . - - .

t NRC REQUEST FOR ADDITIONAL INFORMATION N

l l

l The following mixed cutset sequences may exist, if the operator action OP-ADS-STOP fails:

i l 1. EQ-LOSP OCCURS (0.09g)

• PRHR success

• OP-ADS-STOP fails

• full ADS fails (SYS-ADAB)

2. EQ-LOSP OCCURS (0.09g)

• PRHR success

• OP-ADS-STOP fails

• IRWST injection fails (SYS IW2ABB)

3. EQ-LOSP OCCURS (0.09g)

• PRHR success

• OP-ADS-STOP fails

• sump recirculation fails (SYS-RECIRB).

Fault tree models to represent failure of ADS, IRWST injection, or sump recirculation for this exact case are not available. However, the models already used in the EQ-LOSP event tree for the same functions (namely models SYS- ADAB, SYS IW2ABB, and SYS-RECIRB) are a close approximation to make statements about the dominant mixed cutsets, and are used for this limited evaluation.

For the purposes of this discussion, a failure probability of 1.0E-04 is assigned for OP-ADS-STOP. This is consistent with the range of HEPs (0.00001 to 0.0001) given in Table 20-3 of NUREG-1278 for the time interval 60-1500 minutes.

The dominant cutsets of the three systems, SYS-ADAB, SYS-IW2ABB, and SYS-RECIRB are given in the attached Tables 720.392-1, -2, and -3, as taken from the AP600 focused PRA models.

Thus the mixed cutsets for the above three sequences would be in the form:

(EQ LOSP (0.09g) occurs) { PRHR is successful) {OP-ADS STOP (1.0E-04) fails) (additional component failures occur),

where component failures refer to the dominant cutsets from Tables 720.392-1,-2,-3. Note that these cutsets contain single CCF basic events; most of these dominant cutsets refer to actuation failures due to CCF in I&C or its DC power supply.

Observations on " component failures":

The applicable dominant component failures from Tables 720.392-1,-2, and -3 for the mixed cutsets of the form mentioned above are discussed below.

1. The dominant cutsets of ADS failure from Table 720.3921 are all related to total failure of ADS, in which case the question goes away about operator failure to block ADS actuation since the ADS can not be actuated at all.

72m2 2 T westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION g eu:

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l

2. For IRWST injection function, only the following three dominant cutsets are applicable, given that ADS actuated:

2 3.00E 05 18.68 COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3 2.60E-05 16.19 COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES I l

5 1.20E-05 7.47 COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK The remaining dominant cutsets are common cause actuation failures that already affect ADS as discussed in item I above.

3. For sump recirculation, only the following two dominant cutsets are applicable, given that ADS actuated:

2 2.60E-05 19.03 COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 4 1.20E-05 8.78 COMMON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PLUGGING The remaining dominant cutsets are common cause actuation failures that already affect ADS as discussed in item I above.

l The five dominant component failures identified in items 2 and 3 have failure probabilities at the order of IE-05.

Thus, the failure probabilities of the mixed cutsets would be at the order of IE-09, after these component failure probabilities are multiplied by OP-ADS-STOP failure probability.

Conclusion:

The dominant mixed cutsets for the cases where ADS is automatically actuated due to lack of operator intervention following a EQ-LOSP event with PRHR successful were discussed. The combined probability of the operator error and a common cause component failure for each cutset is estimated to be at the order of IE-09 or less, depending on the failure probability that is assigned to the operator action.

PRA Revision: None.

W8Silflgh0US8

L NRC REQUEST FOR ADDITIONAL INFORMATION 7 4;q h

Table 720.392-1 Dominant Cutsets of SYS-ADAB

Title:

ADAs.wtK File: adab.wlk ( File created by linking Adab.wlk WLINK ** Ver. 3.11 **)

Reduced Sum of Cutsets: 1.3130E-04 NUMBER CUT 5ET PROS PERCENT 8ASIC EVENT NAME EVENT PROS. IDENTIFIER 3 4.70E-05 35.80 COMMON CAUSE FAILURE OF CLASS 1E 8ATTERIES 4.70E-05 (CX-8Y-PN 2 3.00E-05 22.85 Cone 40N CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES TO OPERATE 3.00E-05 ADX-EV-SA 3 2.40E-05 13.28 Cope 40N CAUSE FAILURE OF CLASS lE INVERTERS 2.40E-05 CCX-IV-xR 4 1.10E-05 8.38 Cone 40N CAUSE FAILURE OF PM5 ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCx-PMXMODl-Sw 5 8.62E-06 6.57 COMMON CAUSE FAILURE OF PMS ESF OUTPUT ORIVER CARDS 8.62E-06 CCX-EP-SAM 6 3.63E-06 2.76 COMMON CAUSE FAILURE OF 4/4 STAGE 2 & 3 LINE MOVs TO OPEN 1.10E-03 ADX-MV-GO OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MANO2 7 3.32E-06 2.53 COs#40N CAUSE FAILURE OF 4/4 STAGE 2 & 3 LINE MOVs TO OPEN 1.10E-Ol ADX-MV-GO OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 8 1.20E-06 .91 Cone 40N CAUSE FAILURE OF PMS AND PLS SOFThARE 1.20E-06 CCX-5FTW 9 1.10E-06 .84 Cope 40N CAUSE FAILURE OF 4/4 STAGE 2 & 3 LINE MOVs TO OPEN 1.10E-03 ADX-Mv-GO FAILURE OF PMS, PLS AND DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-INo-FAIL 10 3.40E-07 .26 COse00N CAUSE FAItuRE OF PMS ESF INPUT LOGIC GROUPS (HARDWARE) 1.03E-04 CCX-INPUT-LOGIC OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MAN 02 11 3.11E-07 .24 COMMON CAUSE FAILURE OF PMS ESF INPUT LOGIC GROUPS (HARDWARE) 1.03E-04 CCX-INPUT-LOGIC OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 12 1.27E-07 .10 COso40N CAUSE FAILURE OF CMT/5 UMP LEVEL HEATED RTD SENSORS 3.84E-05 CMx-v5-FA OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MANO2 >

13 1.16E-07 .09 Cone 40N CAUSE FAILURE OF CMT/5 UMP LEVEL HEATED RTD SENSORS 3.84E-05 CMX-VS-FA OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 14 1.03E-07 .08 Cone 40N CAUSE FAILURE OF PMS ESF INPUT LOGIC GROUPS (HAR0 WARE) 1.03Ec04 CCX-INPUT-LOGIC FAILURE OF PMS, PLS AND DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-IND-FAIL 15 3.84E-08 .03 COs*40N CAUSE FAILURE OF CMT/5 UMP LEVEL HEATED RTD SENSORS 3.84E-05 CMX-v5-FA l 1.00E-03 FAILURE OF PMS, PLS AND DAS INDICATION FOR OPERATOR ACTIONS ALL-IND-FAIL t

720.392-4 '

W Westingfl00Se

NRC REQUEST FOR ADDITIONAL INFORMATION

t:

g 16 3.63E-08 .03 COso4CN CAUSE FAILURE OF PMS ESF INPUT LOGIC 50FlhARE 1.10E-05 CCX-IN-tfin.1C-5W OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MANO2 l 17 3.63E-08 .03 COMMON CAUSE FAILURE OF PMS ESF ACTUATION LOGIC SOFTWARE 1.10E-05 (CX-FMAMoo2-5W OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MANO2 18 3.34E-08 .03 HARDWARE FAILURE OF ST. #2 LINE 1 5.64E-02 AD2 Moo 01 HARDWARE FAILURE OF ST. #2 LINE 2 5.64E-02 A02M0002 H/JtDWARE FAILURE OF ST. #3 LINE 1 5.64E-02 AD3 Moo 03 HARDWARE FAILURE OF ST. #3 LINE 2 5.64E-02 AD3M0004 OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3. 30E-03 LPM-MAN 02 19 3.32E-08 .03 COpO40N CAUSE FAILURE OF PMS ESF INPUT LOGIC SOFTWARE 1.10E-05 CCX-IN-LOGIC-5W OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 20 3.32E-08 .03 COMMON CAUSE FAILURE OF PMS E5F ACTUATION LOGIC SOFTWARE 1.10E-05 CCX-PMAMOO2-5W OPEP.ATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 21 3.06E-08 .02 HARDWARE FAILURE OF ST. #2 LINE 1 5.64E-02 AD2M0001 .

HARDWARF. FAILURE OF ST. #2 LINE 2 5.64E-02 AD2NOo02 HARDWARE FAILURE OF ST. #3 LINE 1 5.64E-02 AD3 Moo 03 HARDWARE FAILURE OF ST. #3 LINE 2 5.64E-02 AD3MoOO4 OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 22 1.84E-OS .01 COnetoN CAUSE FAILURE OF CMT/5 UMP LEVEL HEAT ED RTD SENSORS 3.84E-OS CMX-VS-FA COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-XMTR 23 1.21E-08 .01 Cope 40N CAUSE FAILURE OF 4/4 STAGE 2 & 3 LINE MOvs TO OPEN 1.10E-03 ADx-Mv-GO CometON CAUSE FAILURE OF PMS ESF MANUAL INPUT MULTIPLEXER SOFThA. 1.10E-05 CCX-FMxM004-SW 24 1.10E-08 .01 CODO40N CAUSE FAILURE OF PMS ESF ACTUATION LOGIC SOFTWARE 1.10E-05 CCX-PMxMOO2-SW FAILURE OF PMS, PLS AND DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-INO-FAIL 25 1.10E-08 .01 COMMON CAUSE FAILURE OF PMS ESF INPUT LOGIC SOFTWARE 1.10E-05 (CX-IN-LOGIC-5W FAILURE OF PMS, PLS AND OAS INDICATION FOR OPERATOR ACTION > 1.00E-03 ALL-INo-FAIL 26 1.01E-08 .01 HARDWAAE FAILURE OF ST. #2 LINE 1 5.64E-02 AD2 Moo 01 HARDWARE FAILURE OF ST. #2 LINE 2 5.64E-02 A02 Moo 02 HARDWARE FAILURE OF ST. 83 LINE 1 5.64E-02 AD 3M0003 HAADWARE FAILURE OF ST. 83 LINE 2 5.64E-02 Ao3MooO4 FAILURE OF PMS. PLS ANO DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-IND-FAIL 27 2.53E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMBM0011 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMooll HARDWARE FAILURE OF ST. #4 LINE 3 5.80E-04 AD4M0009 28 2.53E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAM0011 FAILURE Of OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCM0011 HARDWARE FAILURE OF ST. 84 LINE 4 5.80E-04 AD4M0010 720.392-5 W_

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NRC REQUEST FOR ADDITIONAL INFORMATION n" uit PJ iii 1

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_ e 29 2.53E-09 .00 HARDWARE FAILURE OF ST. 84 LINE 1 5.80E-04 AD4M0007 FAILURE OF OUTPUT LOGIC GROUP 1 1/0 2.09E-03 PMBM0011 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDM0011 30 2.53E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMOO11 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCM0011 HARDWARE FAILURE OF ST. 84 LINE 2 5.80E-04 AD4M0008 31 2.41E-09 .00 HARDWARE FAILURE OF ST. #2 LINE 1 5.64E-02 AD2 mod 01 HARDWARE FAILURE OF ST. #2 LINE 2 5.64E-02 AD2M0002 HARDWARE FAILURE OF ST. #3 LINE 1 5.64E-02 AD3M0003 FAILURE OF ACTUATION LOGIC GROUP 1 4.07E-03 PMsM0021 OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MANO2 32 2.41E-09 .00 HARDWARE FAILURE OF ST. #2 LINE 1 5.64E-02 AD2MOOO1 HARDWARE FAILURE OF ST. #2 LINE 2 5.64E-02 AD2M0002 FAILURE OF ACTUATION LOGIC GROUP 1 4.07E-03 PMAM0021 OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MAN 02 HARDWARE FAILURE OF ST. #3 LINE 2 5.64E-02 AD3McD04 33 2.41E-09 .00 HARDWARE FAILURE OF ST. 82 LINE 1 5.64E-02 AD2M0001 FAILURE OF ACTUATION LOGIC GROUP 1 4.07E-03 PMDM0021 OPERATOR FAILS TO REf0GNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MANO2 HARDWARE FAILURE OF ST. 83 LINE 1 5.64E-02 AD3M0003 HARDWARE FAILURE OF ST. #3 LINE 2 5.64E-02 A03 Moo 04 34 2.41E-09 .00 FAILURE OF ACTUATION LOGIC GROUP 1 4.07E-03 PMCM0021 OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MAN 02 HARDWARE FAILURE OF ST. #2 LINE 2 5.6dE-02 AD2 MOD 02 HARDWARE FAILURE OF ST. #3 LINE 1 5.64E-02 AD3M0003 HARDWARE FAILURE OF ST. 83 LINE 2 5.64E-02 AD3 MOD 04 35 2.21E-09 .00 HARDWARE FAILURE OF ST. #2 LINE 1 5.64E-02 AD2M0001 HARDWARE FAILURE OF ST. #2 LINE 2 5.64E-02 AD2 MOD 02 HARDWARE FAILURE OF ST. 83 LINE 1 5.64E-02 AD3MOOO3 FAILURE OF ACTUATION LOGIC GROUP 1 4.07E-03 PMBMCO21 OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 36 2.21E-09 .00 HARDWARE FAILURE OF ST. #2 LINE 1 5.64E-02 AD2 MOO 01 HARDWARE FAILURE OF ST. #2 LINE 2 5.64E-02 A02M0002 FAILURE OF ACTUATION LOGIC GROUP 1 4.07E-03 PMAMOO21 OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 HARDWARE FAILURE OF ST. 83 LINE 2 5.64E-02 AD3M0004 720.392-6 W-Westinghouse h

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NRC REQUEST FOR ADDITIONAL INFORMATION

..e - w.

IE .

ji 37 2.21E-09 .00 HARDWARE FAILURE OF ST. #2 LINE 1 5.64E-02 AD2 MOD 01 FAILURE OF ACTUATIN LOGIC GROUP 1 4.07E-03 PMDM0021 OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-0 3 ADN-MAN 01 HARDWA1E FAILURE OF ST. #3 LINE 1 5.64E-02 AD3M0003 HAR0 WARE FAILURE OF ST. #3 LINE 2 5.64E-02 AD3 MOD 04 38 2.21E-09 .00 FAILURE OF ACTUATION LOGIC GROUP 1 4.07E-03 PMCMOo21 OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 HARDWARE FAILURE OF ST. #2 LINE 2 5.64E-02 AD2M0002 HARDWARE FAILURE OF ST. #3 LINE 1 5.64E-02 AD3M0003 HARDWARE FAILURE OF ST. 83 LINE 2 5.64E-02 AD3MOC04 39 1.41E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMBMOOll FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMDO301ASA HARDWARE FAILURE OF ST. #4 LINE 3 5.80E-04 AD4M0009 40 1.41E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMBMOOll FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PM0010185A HARDWARE FAILURE OF ST. 84 LINE 3 5.80E-04 AD4M0009 41 1. 41E -09 .00 FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PM80301ASA FAILURE Ca; OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDM0011 HARDWARE FAILURE OF ST. #4 LINE 3 5.80E-04 AD4M0009 42 1.41E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PM8030185A FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMOOll HARDWARE FAILURE OF ST #4 LINE 3 5.80E-04 AD4M0009 43 1.41E -09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMODll FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMCO301ASA HARDWARE FAILURE OF ST. #4 LINE 4 5.80E-04 AD4M0010 44 1.41E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMODll FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMC030185A HARDWARE FAILURE OF ST. #4 LINE 4 5.80E-04 AD4M00lO 45 1.41E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMA0301ASA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCM00ll HARDWARE FAILURE OF ST. 84 LINE 4 5.60E-04 AD4M0010 46 1.41E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-Ol PMA030185A FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMOD11 HARDWARE FAILURE OF ST. 84 LINE 4 5.80E-04 AD4M0010 47 1.41E-09 .00 HARDWASE FAILURE OF ST. #4 LINE 1 5.80E-04 AG4 MOD 07 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PM8MODil FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PM00101ASA W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

lii W!

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48 1.41E-09 .00 HARDhmRE FAILURE OF ST. #4 LINE 1 S.80E-04 AD4M0007 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PM8MODll FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMD030185A

[

49 1.41E-09 .00 HARDWARE FAILURE OF ST. #4 LINE 1 S.80E-04 AD4 MOD 07 FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PM80301ASA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMODll 50 1.41E-09 .00 HARDWARE FAILURE OF ST. #4 LINE 1 S.80E-04 AD4 MOD 07 FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PM803018SA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDM00ll l

f i

o 720.392-8 W_

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NRC REQUEST FOR ADDITIONAL. INFORMATION 4 "::.

n m 21 Table 720.392-2 Dominant Cursets Of SYS-IW2ABB

Title:

IW2Asa.wLK File: iw2abb.wlk ( File created by linking iw2abb.wlk WLINK ** Ver. 3.11 **)

Reduced sum of Cutsets: 1.6060E-04 NUMBER CUTSET PROB PE RCENT BASIC EVENT NAME EVENT PROB. IDENTIFIER 1 4.70E-05 29.27 CO M ON CAUSE FAILURE OF CLASS lE BATTERIES 4.70E-05 CCX-BY-PN 2 3.00E-05 18.68 CO M ON CAUSE FAILURE OF 4 IRWST IN3ECTION CHECK VALVES 3.00E-05 IWX-CV-AO 3 2.60E-05 16.19 COMMON CAUSE FAILURE OF 4 IRWST IN3ECTION SQUIS VALVES 2.60E-05 Iwx-EV-SA 4 2.40E-05 14.94 COM ON CAUSE FAILURE OF CLASS 1E INVERTERS 2.40E-05 CCX-IV-XR 5 1.20E-05 7.47 COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK 1.20E-05 IWX-FL-GP 6 1.10E-05 6.85 COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCX-FMXMOO1-5W 7 8.62E-06 5.37 COmON CAUSE FAILURE OF PMS ESF OUTPUT ORIVER CARDS 8.62E-06 (CX-EP-5AM 8 1.20E-06 .75 COMMON CAUSE FAILURE OF PMS AND PLS SOFTWARE 1.20E-06 CCX-5FTW 9 3.11E-07 .19 COmON CAUSE FAILURE OF PMS ESF INPUT LOGIC GROUPS (HARDWARE) 1.03E-04 CCX-INPUT-LOGIC OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 10 1.16E-07 .07 COMMON CAUSE FAILURE OF CMT/ SUMP LEVEL HEATED RTD SENSORS 3.84E-05 CMX-VS-FA OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 11 1.03E-07 .06 CO MON CAUSE FAILURE OF PMS ESF INPUT LOGIC GROUPS (HARDWARE) 1.03E-04 CCX-INPUT-LOGIC FAILURE OF PMS. PLS AND DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-IND-FAIL 12 5.76E-08 .04 IWRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA-PLUG IWRST OISCHARGE LINE "B" STRAINER PLUGGED 2.40E-04 IWS-PLUG 13 3.84E-08 .02 COMMON CAUSE FAILURE OF CMT/ SUMP LEVEL HEATED RTO SENSORS 3.84E-05 CMX-VS-FA FAILURE OF PMS, PLS AND DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-IND-FAIL 14 3.32E-08 .02 CO MON CAUSE FAILURE OF PMS ESF INPUT LOGIC SOFTWARE 1.10E-05 CCX - I N - L OGIC- 5W OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E*03 ADN-MAN 01 15 3.32E-08 .02 COMON CAUSE FAILURE OF PMS ESF ACTUATION LOGIC SOFTWARE 1.10E-05 CCX-PMXMOO2-5W OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 16 1.10E-08 .01 COMMON CAUSE FAILURE OF PMS ESF ACTUATION LOGIC SOFTWARE 1.10E-05 (CX-FMXMOO2-5W FAILURE OF PMS. PLS AND DAS INOICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-IND-FAIL W_

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NRC REQUEST FOR ADDITIONAL INFORMATION 17 1.10E-08 .01 COMMON CAUSE FAILURE OF PMS ESF INPUT LOGIC SOFTWARE 1.10E-05 CCx-IN-LOGIC-Sw FAILURE OF PMS. PLS AND DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-IND-FAIL 18 1.13E-09 .00 COMMON CAUSE FAILURE OF PMS ESF INPUI LOGIC GROUPS (HARDWARE) 1.03E-04 CCX-INPUT-LOGIC CO8640N CAUSE FAILURE OF PMS ESF MANUAL INPUT MULTIPLEXER 50FTWA. 1.10E-05 CCX-PMxMOD4-Sw 19 1.05E-09 .00 IwRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IwA-PLUG FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMOO11 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMOD11 20 1.05E-09 .00 IWRST DISCHARGE LINE *B" STRAINER PLUGGED 2.40E-04 Iwe-PLUG FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PM8 MOD 11 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMOD11 21 8.78E-10 .00 Iwa5T DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA-PLUG IRwST CHECK VALVE 1228 FAILS TO OPEN 1.75E-03 IwSCV122AO FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMOD11 22 8.78E-10 .00 IwRST DISCHARGE LINE "A" STRAINER PLUGGEO 2.40E-04 IwA-PLUG IRwST CHECK VALVE 1248 FAIL 5 TO OPEN 1.75E-03 IwSCV124AO FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMOD11 23 8.78E-10 .00 IRwST CHECK VALVE 122A FAILS TO OPEN 1.75E-03 IwACV122AO IwRST DISCHARGE LINE "S" STRAINER PLUGGED 2.40E-04 IWB-PLUG FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMOMOD11 24 8.78E-10 .00 IRwST CHECK VALVE 124A FAILS TO OPEN 1.75E-03 IwACV124AO IwRST DISCHARGE LINE "B" STRAINER PLUGGED 2.40E-04 IW8-PLOG FAILudE OF OUTPUT LOGIC GROUP 1 I/O 2.09E- 03 PM8 MOD 11 25 7.35E-10 .00 IwRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IwA-PL UG IRWST CHECK VALVE 1228 FAILS TO OPEN 1.75E-03 IwSCV122AO IRwST CHECK VALVE 1248 FAILS TO OPEN 1.75E-03 IwSCV124AO 26 7.35E-10 .00 IRwST CHCCK VALVE 122A FAILS TO OPEN 1. 75E -0 3 IwACV122AO IRwST CHECK VALVE 124A FAILS TO OPEN I./5E-03 IWACV124AO IwRST DISCHARGE LINE "B" STRAINER PLUGGED 2.40E-04 IwS-PLUG 27 7.32E-10 .00 IwRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IwA-PLUG HARDWARE FAILURE OF IRwST VALVE 1238 1.46E-03 IRwM0007 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCM0011 28 7.32E-10 .00 IwRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IwA-PLUG HARDWARE FAILURE OF IRwST VALVE 1258 1.46E-03 IRwMOD08 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMOD11 29 7.32E-10 .00 HARDWARE FAILURE OF IRwST VALVE 123A 1.46E-03 IRwM0005 IwRST DISCHARGE LINE "B" STRAINER PLUGGED 2.40E-04 Iw8-PL UG FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMOM0011 720.392-10 3 Westingil00Se i

_h

NRC REQUEST FOR ADDITIONAL INFORMATION L: w, 30 7.32E-10 .00 HARDWARE FAILURE OF IRW5T VALVE 125A 1.46E-03 IRWM0006 IWRST DISCHARGE LINE "8" STRAINER PLUGGED 2.40E-04 IW8-PLUG FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PM8mO11 31 6.13E-10 .00 IWRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA-PLUG IRWST CHECK VALVE 1228 FAILS TO OPEN 1.75E-03 IW8Cv122AO HARDWARE FAILURE OF IRW5T VALVE 1258 1.46E-03 IRWMOD08 32 6.13E-10 .00 IWRST DISCHARGE LINE "A" 5 TRAINER PLUGGED 2.40E-04 IWA-PLUG HARDWARE FAILURE OF IRWST VALVE 1238 1.46E-03 IRWMOOO7 IRW5T CHECK VALVE 1248 FAILS TO OPEN 1.75E-03 IW8CV124AO 33 6.13E-10 .00 IRWST CHECK VALVE 122A FAILS TO OPEN 1.75E-03 IWACV122AO HARDWARE FAILURE OF IRWST VALVE 125A 1.46E-03 IRWMOOO6 IWRST DISCHARGE LINE "B" STRAINER PLUGGED 2.40E-04 IW8-PLUG 34 6.13E-10 .00 HARDWARE FAILURE OF IRWST VALVE 123A 1.46E-03 IRWM0005 IRWST CHECK VALVE 124A FAILS TO OPEN 1.75E-03 IWACv124AO IWRST DISCHARGE LINE "B" STRAINER PLUGGED 2.40E-04 Ika-PLUG 35 5.82E-10 .00 IWRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA-PLUG FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMOD11 FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMc0301ASA 36 5.82E-10 .00 IWRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA-PL UG FAILURE OF OUTPUT LOGIC GROUP 1 I/O . 2.09E-03 PMAMOO11 FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMCO30185A 37 5.82E-10 .00 IWRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA-PLUG FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMA0301ASA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCM00ll 38 5.82E-10 .00 IWRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA-PLUG FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMA0301BSA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-0 3 PMCM00ll 39 5.82E-10 .00 IWRST DISCHARGE LINE "8" STRAINER PLUGGED 2.40E-04 IWS-PLUG FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PM8MODil FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMD0301ASA 40 5.82E-10 .00 IWRST DISCHARGE LINE "B" STRAINER PLUGGED 2.40E-04 IW8-PLUG FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PM8M00ll FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMD030185A 41 5.82E-10 .00 IWR5T DISCHARGE LINE "8" STRAINER PLUGGED 2.40E-04 IW8-PLUG FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PM80301ASA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMoMooll 720.392-11 V_/ Westinghouse

+-

• t NRC REQUEST FOR ADDITIONAL INFORMATION di!!lunWS r-351 42 5.82E-10 .00 IWRST DISCHARGE LINE "B" STRAINER PLUGGED 2.40E-04 IWB-PLUG FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PM8030185A FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMODil 43 5.12E-10 .00 IWR5T DISCHARGE LINE "A" STRAINER PLUGGr_D 2.40E-04 IWA-PLUG HARDWARE FAILURE OF IRWST VALVE 123R 1.46E-03 IRWMOD07 HARDWARE FAILURE OF IRWST VALVE 1258 1.46E-03 IRWMoD08 44 5.12E-10 .00 HARDWARE FAILURE OF IRWST VALVE 123A 1.46E-03 IRWM0005 HARDWARE FAILURE OF IRWST VALVE 125A 1.46E-03 IRWMOD06 IWRST DISCHARGE LINE "B" STRAINtR PLUGGED 2.40E-04 IW8-PLUG 45 4.87E-10 .00 IWRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA-PLUG IRW5T CHECK VALVE 1228 FAILS TO OPEN 1.75E-03 IwSCV122AO FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMc0101ASA t 46 4.87E-10 .00 IwRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA- PL UG IRWST CHECK VALVE 1228 FAILS TO OPEN 1.75E-03 IW8CVI22AO FAILURE OF OUTPUT LOGIC GROUP 18 1.16L-0 3 PMc0301BSA 47 4.87E-10 .00 IWRST DISCHARGE LINE "A" STRAINER PLUGGED 2.40E-04 IWA-PLUG IRWST CHECK VALVE 1248 FAILS TO OPEN 1.75E-03 IW8CV124AO 1.16E-03 FAILURE OF OUTPUT LOGIC GROUP 1A PMA0301ASA 48 4.87E-10 .00 IWRST DISCHARGE LINE *A" STRAINER PLUGGED 2.40E-04 IWA-PLUG IRWST CHECK VALVE 1248 FAILS TO OPEN 1.75E-03 IW8CV124AO FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMA030185A 49 4.87E-10 .00 IRWST CHECK VALVE 122A FAILS TO OPEN 1.75E-03 IWACV122AO r 2.40E-04 IWRST DISCHA8tGE LINE "B" STRAINER PLUGGED IWB-PLUG FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 Pe0301ASA 50 4.87E-10 .00 IRWST CHECK VALVE 122A FAILS TO OPEN 1.75E-03 IWACV122AO IWRST DISCHARGE LINE "B" STRAINER PLUGGED 2.40E-04 IW8-PLUG FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMDO30185A a i

1 720.372-12 W_

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NRC REQUEST FOR ADDITIONAL INFORMATION

!!! uni Table 720.392-3 Dominant Cutsets of SYS-RECIRB

Title:

RECIRS.WLK File: recirb.wlk ( File created by linking recirb.wik WLINK ** Ver, 3.11 **)

Reduced Sun of Cutsets: 1.3660E-04 NUMBER CUT 5ET PROS PERCENT BASIC EVENT NAME EVENT PROS. IDENTIFIER 1 4.70E-05 34.41 Cope ON CAUSE FAILt>RE OF CLASS 1E BATTERIES 4.70E-05 CCx-BY-PN 2 2.60E-05 19.03 COpeeON CAUSE FAILURE OF 4 SQUIS VALVES IN RECIRC LINES 2.60E-05 IWx-EV4-5A 3 2.40E-05 17.57 COMMON CAUSE FAILURE OF CLASS 1E INVERTERS 2.40E-05 (Cx-IV-xR 4 1.20E-05 8.78 CO*OeON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PLUGGING 1.20E-05 rex-FL-GP S 1.10E-05 8.05 COMMON CAUSE FAILURE OF PM5 ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCx-PMxMOD1-5W 6 8.62E-06 6.31 COseeON CAUSE FAILURE OF PMS ESF OUTPUT DRIVER CARDS 8.62E-06 CCx-EP-SAM 7 4.78E-06 3.50 COMMON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRWST. BAT) 4.78E-04 IWX-xMTR OPERATOR FAILS TO ACTUATE CONT. SUMP RECIR. (LEVEL SIGNAL FAILS) 1.00E-02 REN-MAN 04 8 1.20E-06 .88 COMMON CAUSE FAILURE OF PMS AND PLS SOFTWARE 1.20E-06 CCx-5FTW 9 4.78E-07 .35 COneON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRwST. BAT) 4.78E-04 IWX-xMTR FAILURE OF PMS. PLS AND DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-IND-FAIL 10 1.03E-07 .08 COpeq0N CAUSE FAILURE OF PMS ESF INPUT LOGIC GROUPS (HARDWARE) 1.03E-04 (Cx-INPUT-LOGIC FAILURE OF PMS, PLS ANO DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-IND-FAIL 11 5.76E-08 .04 SUMP SCREEN A PLUGS AND PREVENTS FLOW 2.40E-04 REA-PLUG SUMP SCP.EEN 8 PLUGS AND PREVENTS FLOW 2.40E-04 RE8-PLUG 12 1.84E-08 .01 COooeON CAUSE FAILURE OF TANK LEVEL TRANSMIT 1FRS (IRWST. BAT) 4.78E-04 IWx-xMIR COnceON CAUSE FAILURE OF CONTAINMENT SUMP LEVEL TRANSMITTERS 3.84E-05 CCx-VS-FA 13 1.10E-08 .01 CoseeCM CAUSE FAILURE OF PMS ESF ACTUATION LOGIC SOFTWARE 1.10E-05 (Cx-PMxMOO2-5W FAILURE OF PMS, PL5 AND DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-INO-FAIL 14 1.10E-08 .01 COMMON CAUSE FAILURE OF PMS ESF INPUT LOGIC SOFTWARE 1.10E.-05 CCx-IN-LOGIC-5W FAILURE OF PMS, PLS AND DAS INDICATION FOR OPERATOR ACTIONS 1.00E-03 ALL-IND-FAIL 15 9.13E-09 .01 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMOMOO11 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMOO11 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMsMOD11 720.392-13 W Westinghouse e

NRC REQUEST FOR ADDITIONAL INFORMATION

+f it i

_ t.

16 9.13E-09 .01 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDM00ll FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMODil FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMOOll 17 7.64E-09 .01 9%KDWARE FAILURE CAUSE RECIRC. CV 119A FAILS TO OPEN 1.75E-03 REACV119GO FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-01 PMCMOOll FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMaMODil 18 7.64E-09 .01 HARDWARE FAILURE CAUSE RECIRC. CV 119s FAILS TO OPEN 1.75E-03 RESCV11960 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMOM0011 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMBMOD11 19 7.64E-09 .01 HARDWARE FAILURE CAUSE RECIRC. Cv 119A FAILS TO OPEN 1.75E-03 REACV119GO FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMODll FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMODil 20 7.64E-09 .01 HARDWARE FAILURE CAUSE RECIRC. Cv 1198 FAILS TO OPEN 1.75E-03 REBCV119GO FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMOO11 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMODll 21 6.40E-09 .00 HARDWARE FAILURE CAUSE RECIRC. CV 119A FAILS TO OPEN 1.75E-03 REACV119GO HARDWARE FAILURE CAUSE RECIRC. CV 1198 FAILS TO OPEN 1.75E-03 RE8CV119GO FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMBMOOll 22 6.40E-09 .00 HARDWARE FAILURE CAUSE RECIRC. CV 119A FAILS TO OPEN 1.75E-03 REACV119GO HARDWARE FAILURE CAUSE RECIRC. CV 119B FAILS TO OPEN 1.75E-03 REBCV119GO FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMOD11 23 6.38E-09 .00 HARDWARE FAILURE OF IRWST SQUIB VALVE 1208 1.46E-03 IRWM0012 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMOD11 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PM8MODll 24 6.38E-09 .00 HARDWARE FAILURE OF IRWST SQUIB VALVE 120A 1.46E-03 IRWMOO10 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMOOll FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMaMooll 25 6.38E-09 .00 HARDWARE FAILURE OF IRWST SQUIB VALVE 120s 1.46E-03 IRWM0012 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMOOll FAILURE OF OUTP"JT LOGIC GROUP 1 I/O 2.09E-03 PMAM0011 26 6.38E-09 .00 HARDWARE FAILURE OF IRWST SQUIB VALVE 120A 1.46E-03 IRWMOO10 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCM00ll FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 FMAMODll 27 6.02E-09 .00 SUMP SCREEN A PLUGS ANC PREVENTS FLOW 2.40E-04 REA-PLUG HARDWARE FAILURE CAUSES RECIRC MOV 1178 FAILS TO OPEN 1.20E-02 IRwMuo03 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMODll 720.392-14 3 Westingh00Se

NRC REQUEST FOR ADDITIONAL INFORMATION

$: M TU P 28 6.02E-09 .00 SUMP SCREEN 8 PLUGS AND PREVENTS FLOW 2.40E-04 REB-PLUG HARDWARE FAILURE CAUSES RECIRC MOV ll7A FAILS TO OPEN 1.20E-02 IRwMOD01 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMODll 29 5.34E-09 .00 HARIAARE FAILURE OF IRwST SQUIS VALVE 1208 1.46E-03 IRWMOD12 HARDWARE FAILURE CAUSE RECIRC. CV 119A FAILS TO OPEN 1.75E-03 REACV119GO >

FAILU2E OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMSMODil 30 5.34E-09 .00 HARDWARE FAILURE OF IRwST SQUIS VALVE 120A 1.46E-03 InwMOD10 HARDWARE FAILURE CAUSE RECIRC. CV 1198 FAILS .1 OPEN 1.75E-03 RE8CV119GO FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMSMoDll 31 5.34E-09 .00 HARDWARE FAILURE OF IRwST SQUIB VALVE 1208 1.46E-03 IRWMOD12 HARDWARE FAILURE CAUSE RECIRC. CV 119A FAILS TO OPLN 1.75E-03 REACV119GO FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAM0011 32 5.34E-09 .00 HARDWARE FAILURE OF IRwST SQUIS VALVE 120A 1.46E-03 IRWMOD10 HARDWARE FAILURE CAUSE RECIRC. Cv 1198 FAILS TO OPEN 1.75E-03 REBCV119GO FAILukE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAPODil 33 5.26E-09 .00 COMMON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRwST. SAT) 4.78E-04 IwX-AMTR ,

COnceON CAUSE FAILURE OF PMS ESF MANUAL INPUT MULTIPLEXER 50FTiwA. 1.10E-05 CCX-PMXM004-5W 34 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMoDil FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCM0011 FAILURE OF OUTPUT LOGIC GROU* 1A 1.16E-03 PM80301ASA 35 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDM0011 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMODll FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PM8030185A 36 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMODil taiiURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMCO301ASA FAILURE OF OCTPUT LOGIC GROUP 1 I/O 2.09E-03 PM8MOOll ,

37 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDM0011 FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMCO30185A FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMBMODll 38 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMD0301ASA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMODll '

FAILURE OF OUTPUT LOGIC GROUP 1 1/O 2.09E-03 PMEMODll 39 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 18 1.16e-03 PMD030185A FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMODil FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PM8M00ll ,

40 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMODll FAILURE OF OUTPUT LOGIC GROUP 1 1/0 2.09E-03 PMCMODll FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMA0301ASA ,

720.392-15 i

. _ . . _ , - .mm . . _ __ . _m . . . . . _ _ _

~

NRC REQUEST FOR ADDITIONAL INFORMATION 41 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMOOll FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMODll FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMA030185A 42 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMOOll FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMCO301ASA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PP%M00ll 43 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMDMOOll FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMCO301BSA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAM00ll 44 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PMD0301ASA FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMOOll FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAM0011 45 5.07E-09 .00 FAILURE OF OUTPUT LOGIC GROUP 18 1.16E-03 PMD030185A FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.0SE-03 PMCMODll FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMOO11 46 5.04E-09 .00 SUMP SCREEN A PLUG 5 AND PREVENTS FLOW 2.40E-04 REA- PLUG HARDWARE FAILURE CAUSE RECIRC. CV 1198 FAILS TO OPEN 1.75E-03 RESCv119GO HARDWARE FAILURE CAUSES RECIRC MOV 1178 FAILS TO OPEN 1.20E-02 IRWM0003 47 5.04E-09 .00 SUMP SCREEN 8 PLUG 5'ANO PREVENTS FLOW 2.40E-04 REB-PLUG HARDWARE FAILURE CAUSE RECIRC. CV 119A FAILS TO OPEN 1.75E-03 REACV119Ge HARDWARE FAILURE CAUSES RECIRC MOV ll7A FAILS TO OPEN 1.20E-02 IRWMOO01 48 4.45E-09 .00 HARDWARE FAILURE OF IRWST SQUIB VALVE 120A 1.46E-03 IRWM00lu HARDWARE FAIL _URE OF IRWST SQUIB VALVE 1208 1.46E-03 IRWMOO12 FAILURE OF OUTPUT LOGIC GRCUP 1 I/O 2.09E-03 PM8 MOO 11 49 4.45E-09 .00 HARDWARE FAILURE OF IRWST SQUIS VALVE 120A 1.46E-03 IRWM0010 HARDWARE FAILURE OF IRWST SQUIB VALVE 1208 1.46E-03 IRWM0012 FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMAMOOll

• 50 4.24E-09 .00 HARDWARE FAILURE CAUSE RECIRC. CV 119A FAILS TO OPEN 1.75E-03 REACV119Go .

FAILURE OF OUTPUT LOGIC GROUP 1 I/O 2.09E-03 PMCMOOll FAILURE OF OUTPUT LOGIC GROUP 1A 1.16E-03 PM80301ASA i

i 720.392-16 W-westinghouse

i 4

NRC REQUEST FOR ADDITIONAL INFORMATION Ef 9%

t Question: 720.400 (OITS #5497) 1 Provide the analysis (calculation note) that provides the basis for the 0.01 probability of consequential SGTR in l j ATWS events (OTH SGTR).

Response

i The probabilities of single and multiple consequential steam generators are calculated by starting with the correlation that is taken from NUREG-0844, US NRC, September 1988. This correlation provides the probability as a function of the primary to secondary side pressure differential resulting from the initiating event or a subsequent component failure.

The basis of the value used in the PRA are provided in subsections 31.3.8 and 31.3.9. l 1,

i The highlights of the multiple consequential SGTR calculation (OTH SGTR event) are summarized below:

3

1. The probability of 0.025 estimated by the reference NUREG is assumed to be an upper bound. A lognormal

~

distribution with a 95% value of 0.025 and an error factor of 3 is assumed to obtain the mean of 0.01.

1 2. The probability of multiple SGTRs are assumed to be no larger than that of a single SGTR. This assumption

, is severely conservative. Also assumed is the maximum peak differential pressure of 2600 psig across the tubes during the transient.

4 With this information, and the calculation as shown in PRA subsection 31.3.9, the value of 0.01 is calculated for the OTH SGTR evens.

1 1 PRA Revision: None.

T westinghouse

4 I NRC REQUEST FOR ADDITIONAL INFORMATION l 5 ??

I f ++

Ouestion: 720.408 i l

, The second vessel failure mode, i.e., hinged failure leading to a very high flow rate is considered in the earlier j submittal dated December 12,1996, but no calculations are provided in the April 21,1997 submittal. Please provide l l the calculations for the hinged failure mode, or justification for not providing the calculation i Respor;se:

i The TEXAS-V code was used to quantify an ex-vessel FCI in the AP600 reactor cavity. The peak pressure and impulse loads were quantified for use in a structural assessment of the reactor cavity during such a severe accident event. Two different scenarios were considered: 1) a localized RPV failure and 2) a hinged failure of the entire RPV lower head. The results for the localized failure were provided in the April 21,1997 submittal. The results for the hinged failure of the entire RPV lower head are provided in this response.

It is assumed that external cooling of the RPV is ineffective and the RPV fails at the point where the lower head i connects to the cylindrical portion of the RPV. The failure is assumed to be circumferential so that the entire lower l head becomes disconnected from the RPV. This results in initially a large flow of molten metal from the RPV j followed by molten oxide material. The molten metal released is at 1890K (80K superheat). The elevation of the water poolin the cavity is assumed to be at the elevation of the RPV failure to give the deepest possible water pool and, therefore, the worse case scenario. The molten debris will, therefore, discharge directly into the cavity water pool.

However, only the initial pour of molten metal is considered in the FCI since the initial interaction between the water and the debris is the most important. De fuel coolant interaction is assumed to trigger around the time that debris 4

comes in contact with the cavity floor. This occurs within the first few seconds of the scenario. Any debris which enters the pool after the time of the triggering event is not considered in the FCI.

t

The initial conditions and TEXAS.V input deck used for this failure mode are Westinghouse Proprietary and are included in Enclosure 2 of Westinghouse letter DCP/NRC0921, June 19,1997. A summary of TEXAS-V input is provided in Table 720.408 1. The 'EXAS V input deck (file called EXV_ FAST.IN)is provided in electronic format (Westinghouse proprietary) in Enclosure 3 of Westinghouse letter DCP/NRC0921.

The results are provided in tabular, graphical, and electronic form. The tabular output (see Enclosure 4 of Westinghouse letter DCP/NRC0921 for electronic copy; file called EXV_ FAST.OUT) was generated by the TEXAS.

V code and it includes the void fraction, melt volume fraction and pressure as a function of time and location (pool depth). See Table 720.408 2 for a legend for the tabular output, and Table 720.408-3 for the output. Figure 720.408-1 provides the family of pressure histories. Time "zero" in the figure is taken to be when the debris contacted the cavity floor and triggered the explosion. The pressure histories calculated by the TEXAS V code in

the water pool were also taken to be the cavity wall pressure. No pressure attenuation from the pool to the wall was used.

1 4

T Westinghouse 1

~

1 4.

NRC REQUEST FOR ADDITIONAL INFORMATION 5r jgn i 1 l The pressure impulse was derived for each pool elevation by integrating by hand (graphical) the area under each of  ;

the pressure histories in the figure. These results are only available in tabular format. The resulting pressure  !

impulses were as follows:

i Pressure Impulse (Hinged Failure Mode)

I Axial Distance from Pressure Impulse the Bottom of the Pool (m) (kPa's) {

l.25 490 ,

1 1.72 490 l 2.25 442.5 2.75 420  !

3.25 280 l J

i 1

PRA Revision: None.

1 720.e2 W westinghouse 1

l l

D.

NRC REQUEST FOR ADDITIONAL INFORMATION i

I TABLE 720.4081 1

SUMMARY

OF TEXAS INPUT (Hinged Failure Mode) 1 Input Quantity Input Value Melt Flow Rate 15100 kg/s j Melt Superheat 80'K Melt Temperature 1890* K Melt Density 7800 kg/m*

Coherent Jet Diameter 0.068 m Number of Coherent Jets 236 Jet Velocity 2.26 m/s

• Nominal Pool Area 20 m 2

• Based on hydrostatic head due to depth of debris accumulated above junction of RPV lower head and cylindrical section.

W85tingh0US8

l NRC REQUEST FOR ADDITIONAL INFORMATION 5?

TABLE 720.408-2 LEGEND FOR TABULAR OUTPUT Column Outout Ouantity 1 Position from floor (m) 2 Gas temperature (K) 3 Void fraction 4 Melt volume fraction 5 Average radius of the fuel (m) 6 Average radius of the fragmented fuel (m) 7 Gas velocity (m/s) 8 Liquid velocity (m/s) 9 Pressure (Pa) 4 l

8 720.m s W westinghouse

. f

• NRC REQUEST FOR ADDrilONAL INFORMATION r m i  !!

l Table 720.408-3

.000000E+00 13 2.500E-01 3.915E+a. 000E+00 .000E+00 .000E+00 .000E+00 1.983E M i.983E-04 1.828E405 7.500E-01 3.007E+02 .000E+0e .000E+00 .000E.00 .000E+00 3.565E 04 365E.04 i.282E.04 1.250E+00 3 8WEw2 1.421E-05 .000E+00 .000E+00 .000E+00 5.950E-02 4.060E-M I.737E+1 1.750E+00 3.WIE+02 1.99IE-03 .000E+00 .000E+00 .000E+00 4.073E-02 5.774E-03 1.694E+05 4 2.250E+00 4.lb9E+02 7.46tE-02 6.605E-02 2.659E-03 2.659E-03 5.198E-01 4.961E-01 1.625E+05 2.750E+00 435SE+02 3.7I8E-02 6.294E-02 6.421E-03 5.054E-03 3.031E-01 2.830E-01 1.548E+05 3.250E+00 3.852E+02 3346E-03 3.675E-02 3.400E-02 .000E+00 1.894E-01 1.74 t E-01 1.496E+05 3.750E+00 3.852E+02 1.586E-07 .000E+00 .000E+00 .000E+00 2.787E-01 2.787E-01 1.495E+05 4.750E+00 3.F6'4E+02 7.243E-01 .000E+00 .000E+00 .000E+00 4.759E-01 -6.500E-01 1.521 E+05  :

4. . .,0E#X) 3.8MD02 1.000E+00 .000E+00 .000E+00 .000E+00 2.957E-01 2.957E-01 1.515E+05 5.250E+00 3.8MEv02 1.000E+00 .000E+00 .000E+00 .000E+00 2.453E-01 2.453E-01 1.515E+05

+

5.750E+00 3.8ME+02 1.000E+00 .0003+00 .000E+00 .000E+00 !.935E-01 1.935E-01 1.515E+05 2.000063E-03 13 7.500E-01 3.915E+02 .000EM)0 .000E+00 .000E+00 .000E+00 .0LdE+00 .000E+00 1.827E+05 7 500E-01 6.510E+02 8.012E-05 .000E+00 .000E+00 .000E+00 -1334E+01 -2.866E+01 1.191E+08 1250E+00 6.514E+02 2.055E-03 .000E+00 .000E+00 .000E+00 -5.462E401 -4.450E+01 7.812E+07 1.750E+00 6.490E+02 1.624E-02 .000E+00 .000E+00 .000E+00 -8.242E+01 -8.617E+01 7.021E+07 2.250E+00 6.470E+02 4.222E-01 5.174E-02 1.424E-03 1.424E-03 -5.583E+01 -4.988E+01 5.814E+07 2350E+00 6.47tE+02 3312E-01 7.721E-02 2330E-03 2330E-03 6.614E+01 7.069E+01 8.122E+07 3.250E+00 6.517E+02 7.847E-02 3.710E-02 3312E-02 3312E-02 1.139E+02 1.M2E+02 9355E+07 ,

3.750E+00 6.501E+02 1.918E-02 2.801E-02 3337E-02 3337E-02 1.291E+02 1.291E+02 7.537E+07 [

4.250E+00 4.MIE+02 5.933E-01 .000E+00 .000E+00 .000E+00 2.242E+00 -3.6ME-01 2.676E+05 4.750E+00 3.866E+02 1.000E+00 .000E+00 J)00E+00 .000E+00 4.469E-01 4.469E-01 1.517E+05 5.250E+00 3.865E+02 1.000E+00 .000E+00 .000E+00 .000E+00 2.264E-01 2.264E-01 1.515E+05 t 5350E+00 3.865Ft02 1.000E+00 .000E+00 .000E+00 .000E+00 1.742501 1.742E-01 1.515E+05 4.000313E-03 13

[

720.408-5 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION 2.500E-01 3.915E+02 .000E+00 .000E+00 .000E+00 .000E+00 .000E+00 .000E+00 1.827E+05 7.500E-01 6.513E+02 1.096E-M .000E+00 .000E+00 .000E+00 -1.!M+01 -7.456E+00 1.361E+08 1.250E+00 6.524E+0'2 3.550E-03 .000E+00 .000E+00 .000E+00 1.: $+01 -2.503E-01 1.390E+08 1.750E+00 6.518E+02 1.649E-01 .000E+00 .000E+00 .000E+00 1.438E+02 1.391E+02 8.355E+07 2.250E+00 6.524E+02 2.688E-01 5.550E.02 1.240E-03 1.240E-03 6.846EM)I 5.214E+01 8.862E+07 2.750E+00 6.470E+02 5.120E-01 7.251E-02 1.267E-03 !.267E-03 8.991E+01 8.213E+01 2.966E+07 3.250E+00 6.218E+02 3.130E-01 3.561E-02 3.312E-02 3.312E-02 1.478E+02 1.427E+02 1.710E+07 3.750E+00 5.752E+02 2.057E-01 3.M4E-02 3.334E-02 3.334E-02 1.374E+02 1.563E+02 9.146E+06 4.250E+00 6.994E+02 3.596E-02 .000E+00 .000E+00 .000E+00 6.702E+01 6.702E+01 7.657E+07 4.750E+00 5.001E+02 9.501E-0I .000E+00 .000E+00 .000E+00 7.16IE+01 8.509E-01 2.752E+05 5.250E+00 3.958E+02 1.000E+00 .000E+00 .000E+00 .000E+00 6.857E+00 2.805E-01 1.627E+w

750E+00 3.871E+02 1.000E+00 .000E+00 .000E+00 .000E+00 1.619E-01 1.619E-01 1.524 E+05 o.014062E-03 13 2.500E-01 3.915E+02 .000E+00 .000E+00 .000E+00 .000E+00 .000E+00 .000E+00 1.827E+05 7.500E-01 4.508E+02 1.584E-02 .000E+00 1 E+00 .000E+00 -1.608E+01 -2.167E+00 9.161E+05 1.250E+00 6.470E+02 3.442E-02 .000E+00 .000E+00 .000E+00 2.921E+01 3.948E+01 2.681E+07 1.750E+00 6.N2E+02 1.918E-01 .000E+00 .0(X)E+00 .000E+00 7.990E+01 7.408E+01 5.329E+07 2.250E+00 6.470E+02 3.755E-01 2.INE-02 .000E+00 .000E+00 1.518E+02 1.497E+02 3.834E+07 2.750E+00 6.483E+02 3.958E-01 8.875E-02 9.317E-M 9.317E-M 1.717E+02 1.670E+02 3.648E+07 3.250E+00 6.728E+02 3.597E-01 4.858E-02 1.647E-03 1.M7E-03 1.Il8E+02 1.172E+02 2.473E+07 3.750E+00 6.989E+02 1.122E-01 3.449E-02 3.331E-02 3.331E-02 9.838E'91 8.844E+01 3.698E+07 4.250E+00 6.168E+02 3.259E-01 1.198E-03 .000E+00 .000E+00 1.825E -)2 1.817E+02 6.701E+o6 4.750E+00 7.412E+02 4.130E-01 .000E+00 .000E+00 .000E+00 7.398E+01 6.766E+01 6.500E+06 5.250E+00 6.602E+62 9.541E-01 .000E+00 .000E+00 .000E+00 1.986E+02 1.290E+01 4.800E+05 5.750E+00 5342E+02 9.996E-01 .0COE+00 .000E+00 .000E+00 1.868E+01 3.702E-01 3.886E+05 W Westinghouse

2 -.ma m --2 L-- -

o 1

4 a.

NRC REQUEST FOR ADDITIONAL INFORMATION run SF W i i i i . i i i i i i i i i i i l

' ~

x x=1.25 m ----

- x=1.75 m O -

x = 2.25 m x x = 2.75 m -

x = 3.25 m x _

J tn -- --

+, n -

\

' \

, \ -

(C y \

G- -

I

\ -

\

I LJ -

t 1 -

T I ) \

D

- - \ -

\

tn cn -

t n \ -

I LJ l 1

m I

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I

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1 -

1

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' lii,,Ik;i,l,, ii

, i 0 . 25 .5 .75 1

-2 TIME SECONDS x10 Figure 720.408-1 Local Explosion Pressure (Hinged Failure Mode)

W Westinghouse

9

-4

)

NRC REQUEST FOR ADDITIONAL INFORMATION Question: 720.410 For the dripping flow case, only the pressure trace (in the pool) is provided in the /.pril 21, 1997, submittal.

Westinghouse states in the letter that they do not have the code capability of plotting from the TEXAS code, and supplied the information in a tabular form. Please provide an electronic version of the table that is included in the April 21,1997, submittal. In addition, please provide picts vs time for cavity wall pressure, and impulse load. (If the latter data will be provided in a tabular form, also provide an electronic version.)

Response

An electronic version of the tabular TEXAS-V input and output for the localized RPV failure (see April 21,1997 i submittal) is provided as Enclosures 3 and 4, respectively, of Westinghouse letter DCP/NRC0921, dated June 19, 1997. The input file is called EXV-SLOW.IN and is Westinghouse proprietary. The output file is called EXV-SLOW.OUT, The April 21,1997 submittal included pressure histories plots. These pressure histories were generated for the water pool cells by the TEXAS-V code, These calculated pool pressures were also taken to be the cavity wall pressure histories. No pressure attenuation from the pool to the wall was used in the EVSE assessment. Thus, the previously submitted pressure history plots also represent the cavity wall pressure histories. The reported impulse loads are singled valved results which represent the integral of the area under each pressure-time plot. The impulse loads are only available in a tabular format.

PRA Revision: None.

4 k

N i

i 720.410 1 3 Westinghouse