ML20203D446

Requests That Proprietary AP600 Response to FSER Open Items & Markup of PRA App B.4,be Withheld,Per 10CFR2.790

ML20203D446
Person / Time
Site:	05200003
Issue date:	12/09/1997
From:	Sepp H WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19317C780	List: ML20203D446 NSD-NRC-97-5476, Forwards Proprietary Westinghouse Response to FSER Open Items 720.418F & 720.419F.Open Items Pertain to AP600 Core Concrete Interaction Topic.Encl Withheld,Per 10CFR2.790
References
AW-97-1191, NUDOCS 9712160196
Download: ML20203D446 (25)

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Westinghouse Energy Systems Ba 355 Pinsburgh Pennsylvada 15230 0355 Electric Corporation AW-971191

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December 9,1997

- Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 NITENTION:

MR. T. R. QUAY APPLICNflON FOR WITilllOLDING PROPRIETARY INFORMATION FROM PUllLIC DISCLOSURE SUIDECT:

AP600 RESPONSE TO FSER OPEN ITEMS AND MARKUP OF PRA APPENDIX B.4

Dear Mr,

Quay:

The application for withholding is submitted by Westinghouse Electric Company, a division of CBS Corporation (" Westinghouse"), pursuant to the provisions of paragraph (b)(1) of Section 2,790 of the Commission's regulations. It contains commercial strategic infonnation proprietary to Westinghouse and customari'y held in con 0dence.

The proprietary material for which withholding is being requested is identi0ed in the proprietary version of the subject report. in conformance with 10CFR Section 2.790, Affidavit AW-971191 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 withhcid from public disclosure in accordance with 10CFR Section 2.790 of tie Commission's regulations.

Very truly yours, enry A. Sepp, blanager Regulatory Licensing and Engineering jml cc:

Kevin Bohrer NRC OWFN - MS 12E20 9712160196 971209 PDR-ADOCK 05200003 E

PDR.

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AW-97-1191

- AFFIDAVIT

'COMMONWEALTil OF PENNSYLVANIA:

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j COUNTY OF ALLEGilENY:

4 liefere me, the undersigned authority, personally appeared llenry A; Sepp, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Af6 davit on behalf of Westinghouse Electric Company, a division of CBS Corporation (" Westinghouse"), and that the--

averments of fact set fonh in this A f6 davit are true and correct to the best of his knowledge,

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information, and belief:

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llenry A. Sepp, Manager Regulatory Licensing and Engineering Sworn to and subscribed

before me this kM day 1

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,1997

. N0!adal Seal

' J.3not A. Sct*W 5 No'ary Pt@ne MCToev;lio Boro, AH0Gheny Ccu My conn.mn Emires May 22. En

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Notary Public

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AW.971191.

4 (1).

I am Manager, Regulatory Licensing and Engineering, in the Nuclear Services Division, of the Westinghouse Electric Company, a division of CilS Corporation (" Westinghouse"), and as -

such, I have been specifically delegated the fimetion of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing-and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Ilusiness Unit.

(2)

I am 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 accompanylag this Affidavit.

(3)

I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy Systems Business Unit it. % nating information as a trade secret, privileged or as confidential commercial or financial information.

(4)

Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's

. 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.

(i)

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

(ii)

The information is of a type customarily held in confidence 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 wiien and whether to hold certain types cf 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 confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential I

campetitive advantage, as follows:

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AW.971191 5

(a)

--The infonnation reveals the distinguishing aspects of a process (or component, 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.

-(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 econcmic 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.

(c)

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 disclosure to protect the Westinghouse competitive position.

(b)

'It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Wes6 ghouse ability to sell products and services involving the use of the infonnation.

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s AW 97-1191 a

(c)

Use by our competitor would put Westinghouse at a competitive disadvantay; 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.

4 (c)

Unrestricted d6 closure 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 con 6dence and, under the provisions of 10CFR Section 2.790, it is to be received ir con 0dence by the Commission.

(iv)

The infonnation 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 DCP/NRCl171 'NSD-NRC-97-5476), Decembei 9,1997, being transmitted by Westinghouse Electric Company, a division of CBS Corporation

(" Westinghouse"), letter and Application for Withholding Proprietary Information from Public Di:: closure, llenty A. Sepp (F, to Mr. T. R. Quay, Oflice 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 certi0 cation application and is expected to be applicable in other licensee submittals in av..,e

- AW-971191 response to certain NRC requirements forjustification oflicensing advanced nuclear power plant designs.

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

(a)

Demonstrate the design and safety of the Ap600 Passive Sr.fety Systems.

(b)

E:tablish 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.

(c)

Assist customers in obtaining NRC approval for future plants.

Further 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 informatica 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 fbr 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 ii. formation.

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AW 97-1191

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.The development of the technology described in pan 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 suni 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, 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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.s Enclosure I to Westinghouse Letter DCP/NRCll71 December 9,1977 8 S$

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. NRC FSER OPEN ITEM

. mmmmw Ouestion: 720.418F (OITS # 6130)

Westinghouse has modified the design'of the reactor cavity floor and the concrete curb surrounding the reactor cavity sump, and has not submitted any information regarding this updated design. Westinghouse needs to submit -

the design details and drawings for the reactor cavity floor and sump, and confirm that: (1) the elevations of the -

reactor cavity subcompartment RCDT subcompartment and interconnecting ventilation doct, as well as the now areas between these regions are consistent with the Argonne National Laboratory analysis. (2) core debris will not pass into the sump via interconnected pipelines embedded in the concrete floor and/or sump curb, and (3) the sump curb (height and width) is adequately sized to prevent molten core debris from overnowing or ablating the curb. The later point should be demonstrated for: the core debris masses in Table B 5 of the PRA. both RPV failure modes, and both concrete types considered in Appendix B of the PRA, Westinghouse should also explain why r.eglecting the impact -

of RCDT supports / sump ion debris spreading / height is appropriate. This is open item 19.2.3.3.3-1.

. Response, Section D.4 of AP600 PRA Appendix B will be revised to address this open item. The appropriate drawings are.

being provided separately as Enclosure 3 (Westinghouse proprietary) of Westinghouse letter DCP/NRCll71 transmitting this RAI response.

PRA Revision:

See Enclosure 2 of Westinghouse letter DCP/NRCI171 for the n.arkup of section B.4 of PRA Appendix B. The changes noted by the markup will be incorporated in Revis on 11 of the PRA, 720.418F 1 g

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NRC FSER OPEN ITEM nmmmm t

4 Ouestion: 720A19F- (OITS # 6131)-

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- Westinghouse's results of liner and basemat penetration are' based on infortnation provided in response to RAI 720,411. and need to be incorporated into Appendix B of the PRA.- In addition. Westinghouse should confirm the locwion of liner and basernat penetration for each RPV failure scenario and concrete type, This is open item 19.2.3.3.3 2.

Response

Section B.4 of AP600 PRA Appendix B will be revised to address this open item.

PRA Revision:

- See Enclosure 2 of Westinghouse letter DCP/NRCI171 for the taarkup of section B.4 of PRA Appendix B. The changes noted by the markup will be incorporated in Revision 11 of the PRA.

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4 to Westinghouse Letter DCP/NRC1171 December 9,1997 l

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= MARK' UP OF PRA APPENDIX B. SECTION B.4 2 -

5.4 Core Coacrete lateeeetions J

If the reactor vessel fails when the RCS is at a low pressure, the~

ten core debris will pour from the ;

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. reactor vessel onto the reactor cavity floor. If a steam explosion does not occur, the pour will spread over.

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T he cavity floor and begin to transfer heat to the concrete floor of the reactor cavity. Due to the predicted t

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- mode of reactor vessel failure and the shape of the AP600 reactor cavity, analyses of the possible spreading -

,j of the core debris over the cavity floor were conducted using the MELTSoREAD code (reference B 6).

t De results of the MELTSPREAD analyses were then used as input to the MAAP4 code for analysis of core concrete interactions.

Thei9600 reactor cavity is at elevation 11' 6"and consists of two interconnected volumes. The volume whkh includes the reactor nsselis octagonalin shape. The other volume is rectangular in shape and

houses the reactor cociant drain sank and also contains the reactor cavity sump. The two volumes are connected by a $ foot w.Je tunnel whosefloor h iso at elevation 71'6"and a 3fs L vide ventdation duct :

whose bottom is 2 feet abon the cavityfloor. The cavity sump k situated between the tunnel and the ventdation duct at the side of the ren..or coolant drain tank room closest to the reactor nsssL - There it a 3 foot :hkk well that separates the reactor cavity drain tank region from the reat.or nssel region of the envity; The floor of the cavky sump is at elevatie.a 69' 6" and is complesely encompassed by a curb whose top is at elevation 73' 0"(18 inch high curbk A schematic layout of she cavity region is.

- provided in Figure B 3, The embedded steel containment liner beneath the reactor car'ty region is ellipsoidalin shape. The minimum thiancefrom the reactor cavityjloor to the steelliner wcurs at the and of the reactor coolant drain tank room furthermostfrom the reactor nssel and is 2.78 feet (0.847 metersh ne minimum l

distancefrom the esvity sump to the steelliner is 2.69 feet, which isjust slighdy less than the minimum distancefrom the envity floor. The minimum distance is used in the following analyses, as opposed to '

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the minimum nerical distance whkh are slightly greater than the minimum values presented, to simptffy --

the analyses to a two dimensionalproblem. The minimum distancefrom the reactor cavityfloor to the bottom of the basemet, whkk is parallel to the cavityfloor, k Iifeet (3.35 meters).

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The reactor cavity sump is covered with a stainless steel plate that supports the reactor cavity drain

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pumps. While there are a number ofpenetrations in the steelplate (e.g, manway, piping, etc. they will l_

not represent a significant flow path for viscous core debris to enter the cavity.. De steel plate is expected I; remain intact unless thermally anacked by core debris that secumulates on top of the plate.

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There is no piping buried in the concrete beneath the reactor cavity. The sump drain lines are not

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enclosed in either the reactor cavityfloor or reactor cavity sump concrete.

An investigation of the spreading of core debris that pours into the reactor cavity was conducted for reactor

l, sessel failure that occurs at low RCS pressure-due4e The innstigation considered the vessel failure mode l

'and location, as well as the recognition that the oxide and metal components of the in vessel core debris j

. are predicted to be separatedc Since the oxide and metal components of the core debris have very different L

physical characteristies (e.g., viscosity, heat capacity, etc.), the separated in vessel layers-mey influence the -

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- spreading of the core debris in the rerctor cavity. De melt spreading analysis was conducted for both

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= reactor vessel failure modes described in Section 8.1 (hinged and localized failures) using the initial conditions given in Table B.5; For the hinged sessel failure mode, the entire in vessel core debris mass was deposited on the cavity floor j_

at a constant rate over a time scale of 10 seconds. The ten second time release period used in this analysis is assumed to be representative of a rapid release in which the metal phase is teleased distinct from and t

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,a ahead of the oxide phase, nis is roughly equivalent to the assumption tha; the angle by which the lower head hinges downward increases linearly with time until the head contacts the cavity Hoor. Because of the assumed rapid hinging failure that conveys the melt largely inside the lower head, no effects of metal water interactions are modelled distinct from normal he.t transfer from the rnelt to the water in the MELTSPREAD code. For the localized failure, a modes was developed to calculate the boildown of cavity water level, time dependent melt release rate and mel* superheat, nis model treated the reactor vessel failure elevation as a function of the cavity water d pth (i.e., the failure c evation was maintainedjust above the cavity water level) by calculating the cavity water boiloff rate as a function of the amount of core debris released to the reactor cavity and the amount of superheat in the core debris. De THIRMAL code was then used to investigate the effects of metal water interactions span arrival of materials at the bottom of the pool as the initial portion of the metallic melt relocates downward through the pool. Specifically, the combination of the slow initial releese rate and the large water depth would be expected to result in breakup and freezing of the melt as it falls through the water pool, thereby collecting on the cavity floor as a debris bed of solidified particulate. De melt arrival conditions for the MELTSRPEAD analysis were thus based on both calculated release conditions and the THIRMAL results.

The results of the THIRMAL analyses show that most of the core debris (-94 percent) is expected to reach the noor of the cavity in a partially frozen state while the remainde is expected to be fully frozen. None of the initial release of core material from the vessel is expected to be in a fully molten state. However, as the core debris accumuhtes on the cavity Hoor, the decreased water depth will result in an increasing fraction of the core debris arriving in a molten state. The debris arriv%g in a molten state can fill the interstices and may erode some of the previously solidified debns. Rus, whi!c the initial formation of a porous debris bed cannot be ruled out, the continued addition of molten core material willlikely result in a partially frozen debris layer that can further spread over the cavity Moor, ne rtsults of the THIRMAL analysis were used as the initial conditions for the MELTSPREAD analysis of the localized reactor vessel failure case.

The MELTSPREAD analyses were performed using a reactor cavity model shown in Figure B 3. The model used in the MELTSPREAD cnalysis, which is based on an AP600-liL' cavity, accurately represents the AP609 reactor cavity configuration in terms of spreading area and geometry.

For the hinged vessel failure case, the analysis results show that the core debris is spread relatively l

uniformly over the reactor cavity noor, as shown in Figure B-4. However, the distribution of the metal and oxide components of the core debris are not uniformly distributed over the reactor cavity Goor. In the region directly under the reactor vessel, the core debris consists primarily of the oxide component (e.g.,85 to 90 percent oxide). At the opposite end of the reactor cavity, the core debris consists mainly of the metal component of the core debns released from the reactor vessel (e.g.,75,o 85 percent metal). The core debris i

l is still almost totally molten at the end of the spreading analysis. The steel liner over the cavity floor is completely eroded away and the core debris has begun to penetrate into the concrete basemat. De l

penetration depth at the end of the MELTSPREAD analysis was approximately 1.2 inches (3 cm) under the reactor vessel and about 2.75 inches (7 cm) at the opposite end of the reactor cavity.

A different behavior is predicted for the lowlized reactor vessel failure case. The MELTSPREAD analysis predicts that the core debris will accumulate at the reactor vessel end of the reactor cavity as shown in Figure B 5. He distribution of the metal and oxide competents of the core debris we not uniformly distributed over the reactor cavity floor. In the region directly under the reactor vessel, the core debris consists primarily of the oxide component (e.g.,70 to 80 percent oxide). At the opposite end of the reactor l

cavity, the core debris consists mainly of the metal component of the core debns released from the reactor vessel (e g.,80 to 90 percent metal). The core debris is almost totally frozen at the end of the spreading l

analysis. The steel liner over the cavity door is not crodci(except for one node of the cavity model under l

the reactor sessel) and the core debris has only begun to penetrate the concrete basemat in one node.

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The results of the MELTSPREAD analyses were used to assess the effectiveness of the curbing around sne reactor cavity sump to prevent the accumulation of sign (ficant amounts of core debris in the sump.

In particular, the height of the core debris affacent to the reactor cavity sump curbing (cells 12 through 15 in Fig *sre B 3) was examined to determine the potent.'alfor core debris to enter the cavity sump. For the hinged reactor vesselfailure case, the maximum helght of core debris adjacent ts the sump curbing during the initialflow of core debris from the reactor vessel region to the reactor coolant drain tank region is about 32 inches. This occurs during a very brief time interval (at about 10 seconds after reactor vesselfailure) when thepow is parallelto the curbing. Based on the high viscosity of the core debris, little of the core debris is expected to spillover onto the cavity sump cover pla'r. Following the initial wave, the analyses predict that the core debris is repected off of the back wall of the reactor-coolant drain tank portion of the cavity. The maximum height of the repected wave in the area adjacent to the react 3r cavity sump k 24.7 inches. This occurs at about 25 seconds efter reactor vesselfailure, At this time, core debris is expected topow onto the reactor cavity urap cover. After the passage of the core debris repected wave, the equilibrium height of core debris in the region of the cavity sump is about 22 inches. Since this is higher than the cavity sump curb, the continualpresence of core debris on top of the sump cover will result in thermalfailure of the cover and the subsequentpow of core debris into the cavity sump. For the locali:ed reactor vesselfailure case, the maximum depth of cor. debris in the cavity sump at any time in the transient is about 10 inches, Due to the characteristics of the core debris pow into the reactor cavity for the localized failure mode, there is no transient behavior with large amplitude waves of core debris transiting the cavity. Thus,for the locali:rd reactor vesselfailure mode, the reacur cavity curbing prevents the accumulation of core delsris in the cavity sump.

The results of the MELTSRPEAD analyses were used to establish initi;l conditions for assessment of core concrete interactions using the MAAP4 code models. Since MAAP4 can only treat the core abris that is uniformly spread over a cavity floor, two parallel MAAP 4 analyses were done for each vessel failure mode.

The first analysis for each vessel failure mode treats the core debris under the reactor vessel while the second analy* treats the core debris that is in the RCDT end of the cavity. In all cases, tne results of the MELTSPREAD analysis werg used to define the initial conditions for the core concreie interactions using the MAAP4 models. Since one portion of the reactor cavity initially contains oxide-rich core debris and ths other end contains metti rich core debris, the rate of concrete decomposition is controlled by different factors in each analysis. He oxide-rich debris contains most of the fission product decay heat and this controls the concrete decomposition. He metal rich debris does not contain decay heat but is subject to exothermic heat of reaction of steam with the unoxidized metal and this controls the concrete decomposition.

The core concrete interactions for the AP600 design were analy:edfor two concrete types: basaltic concrete and common limestone. sand concrete. The common limestone. sand concrete f.as a sign (ficantly higher noncondensible gas generation rate, compared to basaltic concrete and should therefore present a more severe containment pressuri:ation transient. On the other hand, the basaltic concrete has a sign (ficantly higher core penetration rate, due to the physicd properties of the concrete, and should therefore present a more severe basemat penetration failure mode, compared to common limestone. sand concrete. The comparison of the results of the containmentfailure modesfor the two types of concrete was used to determine the needfor a specification of a concrete typefor the containment basemat.

For the hinged reactor vessel failure case, the core concrete interaction analysis shows (see Figures B 6a and4-4 through B 64) that the concrete basemat in the region of the cavity under the reactor vessel is croded much more rapidly than the region of the RCDT As documented earlier, core debri; may enter the cavity sump for the hinged reactor vesselfailure case. Close examination of the MELTSPREAD analyses indicates that the core debris in the cavity sump would consist prsmarily of the metal component of the core debrh, similar to the reactor coolant drain tank side of the reactor cavity. Since the MAAP core concrete interaction results show that the core concrete penetration on the reactor coolant drain tank side of the cavity is minimal, compared to As oxide melt penetration on the reactor vessel side of 3

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e the cavity, the core debris penetration from the materialin the cavky sump would not be controuing.

Thus, based on the core debris penetration in the reactor vesselportwn of the cavity and conservatively 7

using a 2.78 foot disaance to the containment liner, key resultsfor the hinged reactor vesselfailure case

. are:

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l Core Concrete Interaction Resuks

- Hinged Vessel failure Case Parameur Basakic Concrete Common Limestone.5,and Concrete Time of Mek.nrough of 7.5 Hours 15 Hours Embedded IJner Time of Mek.Through of 3.4 Days 28.9 Days Basemat Two indicators ofpotential chauenge to c.ontainment integrity are shown in the above table of results:

penetration of the steelliner and penetration of the entire basemat. Detailed structural analyses were not performd to determine the containment fragility for various depths of basemat penetration as a function ofcontainment pressure. Thus, the two values are termed "indkators of a potentialchauenge",

it is highly unlikely that the containment fission product boundary would lost when the core first penetrates the basemat, as discussed below. On the other hand, it is also highly unlikely that the containment integrity wiu stHI be intactJust prior to the time that the core debris peneter.:es the entire basemat depth. Containment basematfailure and the subsequent release offission products from the '

containment is likely to occur at some point in time between the two " indicators".

For the basahic concrete case, c :4 hours (8.64E4 seconds) into the accident, the downward crosion is i

l almost 3.9 feet (1.2 meters) on the reactor vessel side of the cavity while the erosion is only about 1.15 feet l

(0.35 meters) on the RCDT side, ne core debris depth or. reactor vessel side (including the decomposed concrete products) measures about 3.6 feet (1.1 meters). Dus, while the containment shell is assumed to be only 2.378 feet (0.847 meters) from the original cavity floor and would be penetrated by the leading edge of the core debris, the containment shell would still be covered by a substantial layer of core debris and would not communicate with the containrr. nt atmosphere for passable fission product releases at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />"

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:: d:_. 32^^^ :::cni = d _: a 5:=: ;5: 1: =:iin' '::::!:

For the case of the localized reactor vessel failure, the core concrete interaction analysis also shows (see Mgures B.7ar: d B ?b through B.74) that the concrete basemat in the region of the cavity under the reactcc vessel is eroded much more rapidly than the region of the RCDT. In this case, no core debris is predictd to enter the reactor cavity sump. Key resuks for the localized reactor vesselfauure case are:

Core Concrete Interaction Results lacalized Vessel Failure Case Forameter Basakie Concrete Common Lamestone 5and Concrete Time of Mek.Through of 8.9 Hours 16.4 Hours Embedded Liner Time of Mek.Drough of 3.4 Days

!I Days Basemat For the basakic concrete case, at24 hours (3.64E4 seconds) into the accident, the downward erosion is almost 3.9 feet, or 1.2 met rs,(almost identicsl to the hinged '.essel failure case) on the reactor vessel side 4

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of the cavity while the erosion is only about 1.0 foot (0.31 meters) on the RCDT sioe. De core debris depth on reactor vessel side (including the decomposed concrete products) measures about 3.3 feet (1.0 l

meters), nus, while the containment shell is only 2.378 feet (0.847 meters) from original cavity floor and -

4' would be penetrated by the leading edge of the core debris, the containment shell would still be covered by a substantial layer of core debris and would not communicate with the containment atmosphere for

- possible fission product releases at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Tb !::f!:; :d;; cf i: :=: f:S; :: exf!: :I ::i: :; ;

6 x::i7; : d:" S !: 5,ci;d h S tx: : :: de.:: 37^Ja n: ': = ::+!h!: : = !G S: : :.5 S =; f:n ':!d-

- Based on these analyses, it can be concluded that: a) the goal of protecting the containment fission s

' product boundary during thefirst 24 houn of a core melt accident is met, b) it h toot necessary to specVy a concrete typefor the containment basemat since credible containment basematfauures that could lead tofission product releases to the aan.osphere are likely to occur at times well beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and c) the reactor cavity sump is adequately protected such that it is not a weakness la containment basemat integrity during postulated accidents that lead to core concrete interactions.

H.4.1 Containment Pressurlastion Due to Core Concrete lateractions The containment pressurization due to steam and noncondensible generation during the episodes of core concrete interactions described above was assessed to determine the effect of core concrete interactions on the containment integrity.

To estimate the effect of steam and noncondensible gas generation on the containment pressure and temperature, the AP600 containment was included in the separate effects MAAP4 analysis of core concrete interactions described above. Since the core concrete interaction assessment with MAAP4 was a " separate effects" analysis, the initial containment conditions were specified to be 21.S psia (0.15 MPa) at saturation sonditions represented by a steam-air mixture.-hWAPi : dyindwe e: j :ir!!r f= he x=:=

x :' f:!=: - ':_ d=xt;' d:=. S hi. :n S x :d:T;* een&gf+e6 6 : d of ! d:y, =

c:f!:;d by S E' ^ ^." x1, d:: : x :d:T; : px_=: c' 40 ph (0.275 "":) ::d : xxi : ^

x;n.=

c' 305'F (125'K).

The indicator of a challenge to containment lategrity for the containment pressurizatson due to the.

noncondensible gases producedfrom core concrete interactions is the Service Level"C" pressure, which 1

is 90psig f0.72MPa). This is wellbelow the 50% containmentfailure probsbility value of129psig (0.99 MPa). The MAAP4 analysis results are:

Containment Pressurstatson Due to Core Concrete lateractions Parameter Basaltic Concrete Common Limestone-Sand Concrete Hsaged Vessel Failure Case Time to Pressurite Containment to Not Ap/icable Not Applicable Service hvel"C" Containment Pressure at Meh.

29 psig 85 psig Through of the Basemat

~

l Localized Vessel Failure Case Time to Pressurite Containment to Not Applicable il Days Service hvel"C" Containment Pressure at Mek.

43 psig 96 psig Through of the Basemat 5

._ ~

- M l

From these results, it is concluded that the containment conditions at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after teactor vessel failure are not sensitive to the assumed mode of reactor vessel failure. In both cases, the containment pressurires

~

by about 14.5 psi (0.1 MPa)in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> due to the steam and noncondensible gases generated during core

- concrete interactions, f

The results also show that, in all cases except the localked reactor vesselfailure with a common -

limestone. sand concrete bosomat, the containment does not pressuri:e to Servke hvel C containment

^

chauence indkator value prior to the time that the core debris completely penetrates the containment basemet. Thus, for these cases there k no potential chauenge to containment integrity due to overpressurkation since; a) there k no longer a source of mass and energy input to the containment l

Aper the core debris penetrates the entire basemat, and b), basemat penetration assures that the containment wiR be depressurked through the basemasfailure. In the case of a localhed reactor vessel fausre and an assumed basemat concrete composition of common limestone. sed the analysk results show that the indicatorfor containment overpressurkation challenge (Serske Level"C")faus within the same timeframe as the complete basemat penetration indicatorfor basemat integrity chauence. Since the complete basemat penetration indkator of a basemat integrity chauenge is the optimhtic bounding.

Indicator and the containment Servke hvel C pressure is the possimistk bounding indkator for containment overpressure chauenge, the union of the two sets represents a very sm4 conditional probability. Based on the subjective uncertainties represented by these bounding indicators, it is likely that even in this case, containment basematfailure would occur and thereby preclude a containment overpressure failure.

A separate bounding calculation was also performed in which all of the gases from the core concrete interaction, as predicted by the MAAP4 arialyses described above, were assumed to be released to the containment with no condensatior, of the steam that is generated from the concrete decomposition. In this case, it is calculated that the containment would pressurize by 21.8 psi (0.15 MPa) in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

In both the MAAP4 and the bounding calculations, the containment pressure s still well below the point i

where the integrity of the containment might be challenged 2 before the containment basematfails and the containment is depressurked by basemat penetration.

Based on these analyses, it can be concluded that it is not necessary to specyy a concrete type for the containmentbasematsincecontainmentoverpressurefailureduetonon condensablegasgenerationfrom core concrete interactions k not likelyfor any credible severe accident scenarios.

M d

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