ML20217C778

Discusses Documentation of AP600 Informal Correspondence to Westinghouse Re Probabilistic Risk Assessment & External Vessel Cooling

ML20217C778
Person / Time
Site:	05200003
Issue date:	03/23/1998
From:	Joseph Sebrosky NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9803270165
Download: ML20217C778 (7)

Text

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March 23, 1998 APPLICANT: Westinghouse Electric Company PROJECT: AP600

SUBJECT:

DOCUMENTATION OF AP600 INFORMAL CORRESPONDENCE TO WESTINGHOUSE CONCERNING PROBABILISTIC RISK ASSESSMENT (PRA)

AND EXTERNAL VESSEL COOLING On March 11 and 12,1998, the staff provided Westinghouse with the comments in Attachment 1 and Attachment 2, respectively. Attachment 1 contains comments concerning the shutdown PRA insights. Attachment 2 contains comments conceming extemal reactor vessel cooling.

original signed by:

Joseph M. Sebrosky, Project Manager Standardization Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No.52-003 Attachments: As stated cc w/atts: See next page plSTRIBUTION w/ attachments:

Docket File PDST R/F TKenyon PUBLIC BHuffman JSebrosky DScaletti JNWilson SMagruder JHWilson MDunsaniwskyj RPalla,0 8.H7 l MPohida, O- 10 E4 DISTRIBUTION w/o attachments: \ }

SCollins/FMiraglia,0-5 E7 BSheron,0-5 E7 BBoger,0-5 E7 ,

JRoe DMatthews TQuay ACRS (11) JMoore,0-15 B1B 3 n 9 C F D !? R R U M e n p y  ;

DOCUMENT NAME:A:\PRA. FAX To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy OFFICE PM:PDST:DRPM l SC,SB:DSSA l l [  ;

NAME JMSebrosky:sg J84 TRQuay 94' DATE 03/W/98 l/ 03/g/98 OFFICIAL RECORD COPY 9803270165 980323 PDR ADOCK 05200003 A PDR j

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9 Westinghouse Electric Corporation Docket No.52-003 cc: Mr. Nicholas J. Liparulo, Manager Mr. Frank A. Ross Nuclear Safety and Regulatory Analysis U.S. Department of Energy, NE-42 Nuclear and Advanced Technology Division Office of LWR Safety and Technology Westinghouse Electric Corporation 19901 Germantown Road P.O. Box 355 Germantown, MD 20874 Pittsburgh, PA 15230 Mr. Russ Bell Mr. B. A. McIntyre Senior Project Manager, Programs Advanced Plant Safety & Licensing Uuclear Energy Institute Westinghouse Electric Corporation 1776 i Street, NW Energy Systems Business Unit Suite 300 Box 355 Washington, DC 20006-3706 Pittsburgh, PA 15230 Ms. Lynn Connor Ms. Cindy L. Haag Doc-Search Associates Advanced Plant Safety & Licensing Post Office Box 34 Westinghouse Electric Corporation Cabin John, MD 20818 Energy Systems Business Unit Box 355 Dr. Craig D. Sawyer, Manager Pittsburgh, PA 15230 Advanced Reactor Programs j GE Nuclear Energy '

Mr. M. D. Beaumont 175 Curtner Avenue, MC-754 Nuclear and Advanced Technology Division San Jose, CA 95125 Westinghouse Electric Corporation One Montrose Metro Mr. Robert H. Buchholz 11921 Rockville Pike GE Nuclear Energy Suite 350 175 Curtner Avenue, MC-781 Rockville, MD 20852 San Jose, CA 95125 Mr. Sterling Franks Barton Z. Cowan, Esq.

U.S. Department of Energy Eckert Seamans Cherin & Mellott NE-50 600 Grant Street 42nd Floor 19901 Germantown Road Pittsburgh, PA 15219 Germantown, MD 20874 Mr. Ed Rodwell, Manager Mr. Charles Thompson, Nuclear Engineer PWR Design Certification AP600 Certification Electric Power Research Institute NE-50 3412 Hillview Avenue 19901 Germantown Road Palo Alto, CA 94303 Germantown, MD 20874 Mr. Robert Maiers, P.E.

Pennsylvania Department of Environmental Protection Bureau of Radiation Protection Rachel Carson State Office Building P.O. Box 8469 Harrisburg, PA 17105-8469

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Additional insight from the shutdown PRA March 11,1998 Following an extended loss of RNS during safe / cold shutdown with the RCS intact and PRHR unavailable, it is essential to establish and maintain a venting capability wiht ADS stage 4 for gravity injection and containment recirculation.

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Attachmert 1

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' ,ct:r Ccist System end Ccnset:d Systems F gg r

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[ between the conical ponion of the insulation and the sphencal ponion of the reactor sessel

, is not less than 9 inches.

5 The structural frame supporting the insulation is designed to withstand the bounding sesere accident loads without exceeding deucction enteria. The fasteners holding the insulation panels to the frame are also designed for these loads.

Each water inlet assembly is

/p 3/ 0Atnormallythe bottom of the closed to prevent insulation an air circulation pathare water through inlet the vessel assemblies.

insulation. The inlet

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, I dgfilled assemblies with water. This are self permits actuating ingress of water during passive devices.

a severe accident, The inlet assem while preventing O / g b excessive heat loss dunng nonnal operation.

[0 I[' b (4 The , total now area of the water inlet assemblies have sufTicient margin to preclude significant t/'I / g t' [' pressure drop during ex vessel cooling during a severe accident. The minimum total Dow are: for the water inlets assemblies is 6 ft'. Due to the relatively low approach velocities in

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,p t , the Dow pisths leading to the reactor cavity, and due to the relatively large minimum flow area d p /p 4 through each water inlet assembly, with an area of at least 7 in', the water iniet assemblies

/ I, p p are not susceptible to clogging from debns inside containment.

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[O0 j ('f . Near theclosed top of the lower reactor insulation segment and a smallare These dampers are steam vent dampers.

I 3 od(, f normally to present vessel heat loss, buildup of steam pressure D[ , under the insulation will cause them to open to the sent position. The steam sent dampers

[g O) [f are passive self actuated devices and will operate when steam is generated under the t7 insulation with the cavity filled with water.

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' Extensise maintenance of the sessel insulation is not nonnally required. Penodic senfication that the sessel insulation moung parts can be performed dunng refueling outages.

Description loo p t ,f.3 4 3z _ ,, _ , ,of,E,,,xternal

, , Vess,el, Cooling c , ,,F,a,ded.,C,ompartm,ents

- cm a via ' < < ~ ",%

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. l n s',r e s c-n f t e we nf d eer st *fm .,.,s Emessel cooling during a sesere accident is provided by Gooding the reactor coolant system f g., loop cosupanment including a settical access tunnel. the reactor coolant drain tank room. and

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the reactor cavity. Water from these comp tments replenishes the . ster that mnes in j/W'hg p, i

p contact with the reactor s essel and is boiled and vented to containment. The opening t>ctween the vertical access tunnel and the reactor coolant drain tank room is approximately 100 ft' #.

Figure 53 8 depicts the Gooded compartments that provide the water for ex-sessel cooling.

,e 4g )g Th: :p=: ;; 5::nn 5: ==::: :: !=: d=fr :=h :::= =d h: re=::: =: - idedeem p // eb:: =:ie :=h th:: : den ::: p:=!:d: : "i=== acu ::= cf L r:5 :: p:: m t . :::: ::

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0=d h: ==::r =& =mpan :=: The doorway between the reactor cavity compartment and the reactor coolant drain t A room consists of a normally closed door and a damper

/ , W flp above the Jour The door and Jamper arrangement. shown an Figure 3.3 9. maintoons the proper airpos through the reactor cavity during normal operation. The damper prevents p I air from powing into the reactor coolant drain tank compartment. but opens to permot Ql96 L i flooJang of the reactor cavityfrom the reactor coolant drain tank compartment. The Jamper Draft Resision: 21

[ Westingh00$8 5 3-21 February 13.1998 v u , n ,: p ) - 4, Attachment 2

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3 5 // i, g I 5. Reactor Coolant System and Connected Systems

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1 lhh opening has an mansmum flow area of 3fri it Is constructed of hght-ue:ght material to l4 mentmt:e the force necessary to open the damper arid permit flooding and continued

, fll4 flow through the opening during ex-vessel coohng.

a U. k Determination of Forces on Insulation and Support System

/

p'/ M) The expected forces that may be expected in the reactor cavity region of the AP600 plant dunng a core damage accident in which the core has relocated to the lower head and the gM reactor cavity is reflooded have been conservatively established based on data from the ULPU test program (Reference 5). The particular configuration (Configuration III) reviewed closely f 4 models the full-scale AP600 geometry of water in the region near the reactor vessel. between 4/ the reactor vessel and the reactor vessel insulation. The ULPU tests provide data on the pressure generated in the region between the reactor vessel and reactor vessel insulation.

These data, along with observations and conclusions from heat transfer studies, are used to deselop the functional requirements with respect to in vessel retention for the reactor sessel insulation .and support system. Interpretation of data collected from ULPU Configuration til

. expenmerits in conjunction with the static head of water that would be present in the AP600 is used to estimate forces acting on the rigid sections ofinsulation. Further evaluation of the l forces on the reactor sessel insulatton and supports is provided in the AP600 Probabilistic Risk Assessment.

5.J.5.5 Design Evaluation A structural analysis of the AP600 reactor cavity insulation system demonstrates that it meets the functional requirements discussed abose. The analysis encompassed the insulation and support system and included a determination of the stresses in support members. bolts.

insulation panels and welds. as well as de'lection of support members and insulation panels The results of the analy ses show that the insi:lation is able to meet its functional requirements.

The reactor sessel insulation prosides an engineered pathway for water-cooling the sessel and for senting steam from tt reactor casirv. l l

The reactor sessel insulation is purchased equipment. The purchase specification for the l I reactor sessel insulation will require confirmatory static load analyses. I 5.J.6 Combined License Information 5.J.6.I Pressure-Temperature Limit Curses The pressure temp, cunes shown in Figures 5.3 2 and 5.3 3 are genenc cunes for AP600 reactor sessel design, and they are the limiting cunes based on copper and nickel matenal composition Howeser. for a specific AP600 these curves will be plotted based on material composition of copper and nickel. Use of plant-specific cunes will be addressed by the Combined License applicant dunng procurement of the reactor sessel.

Draft Resinion: 21 February 13,199M 3322 W WeSiingh0US8

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