ML20212B749
| ML20212B749 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 10/20/1997 |
| From: | Huffman W NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9710280191 | |
| Download: ML20212B749 (61) | |
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i o October 20, 1997 ; APPLICANT: Westinghouse Electric Corpcration PROJECT: AP600. ,
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS THE AP600 GENERAL DESIGN FEATURES AND SAFETY PHILOSOPHY . The subject meeting was held on October 7, 1997, at the Rockville, Maryland, offices of the Nuclear Regulatory Commission (NRC) between representatives of Westin house and the NRC staff. The purpose of the meeting was to provide a genera overview of the AP600 plant design and safety philosophy to the NRC acting director of the Division of Reactor Program Management, Office of Nuclear Reactor Regulation. ; Attachment 1 is the list of meeting attendees. Attachment 2 are the presenta-tion slides provided by Westinghouse in support of the meeting. ' original signed by: William C. Huffman, 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 Q1STRIBUTION w/ attachments: Docket File PDST R/F TKenyon
/fl PUBLIC BHuffman DTJackson JSebrosky DScaletti JNWilson DISTRIBUTION w/o attachments:
SCollins/fHiraglia, 0-12 G18 BSheron, 0-7 D25 RZinnerman, 0-12 G18 JRoe . DMatthews TQuay WDean, 0-5 E23 .ACRS (11) JMoore, 0-15 B18 DOCUMENT NAME: A: ROE-MTG. SUM n m.i. . ..p, .e ini. 4.. . . in.n.. 6. h. 6..i e . cop, withoui .it. chm.nt/.nck ," s Copg with .tt. chm.nti.nclosur. 'N' s No copv OfflCE PM:PDST:DRPM D:PDST:DRPM- p l NAME- WCHuffman ue h TRQuay149
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DATE 10/,)o/97 10/ M 97 OfflCIAL RECORD COPY s Erd; E m c s i m e m y
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4 , . 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 Nuclear 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. Cirdy L. Haag Doc-Search Associates Advanced Plant Safety & Licensing Post Office Box 34 Westinghouse Electric Corporation Cabin John, MD 20818 Energy Systens Business Unit Box 355 Dr. Craig D. Sawyer, Manager Pittsburgh, PA 15230 Advanced Reactor Programs 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
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WESTINGHOUSE - NRC MEETING ON AP600 DESIGN OVERVIEW OCTOBER 7, 1997 i MEETING ATTENDEES Natit ORGANIZATION t BRIAN MCINTYht WESTINGHOUSE ' TERRY SCHULZ WESTINGHOUSE SUE FAN 10 WESTINGHOUSE ' JACK ROE NRC TCO QUAY NRC DINO STALETTI NRC J0E SEBROSKY NRC ' BitL HUffMAN NRC CHUCK THOMPSON DOE j i Attachment 1
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I l l I WESTINGHOUSE PRESENTATION SLIDES 1 ON i AP600 DESIGN OVERVIEW Attachment 2 l I
AP600 1 DESIGN OVERVIEW T. L SCHULZ SYSTEMS ENGINEERING OCTOBER 7,1997 TL510A/97 I
AP600 PLANT SYSTEMS OVERVIEW $ 4
- 1. Design Approach
- 2. Plant Overview
- 3. Safety-Related System Design .
- 4. Nonsafety-Related System Design
- 5. System Shutdown Capability
- 6. Levels Of Defense S
T1.510/6/97 2
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.p AP600 DEVELOPMENT HISTORY [R1 1984 Conceptual design of modular, passive 10MWe P'#R 1985 EPRI and U.S. Department of Energy programs for conceptual design, development, and evaluation of AP600 i 1990 DOE and EPRI contracts for $120M Design Certification Program 1992 Standard Safety Analysis Report (SSAR) and Probabilistic Risk Assessment (PRA) report submitted to NRC 1993 ' Advanced Reactor Corporation First-of-a-Kind Engineering Award for $158M 1994 Comprehensive test program successfully concluded 1994 Draft Safety Evaluation Report received from NRC 1995 All Test Reports submitted to NRC 1996 Supplemental Draft Safety Evaluation Report received 1996 All Computer Code Validation Reports submitted to NRC 1998 Final Design Approval TL510W77 3
emm AP600 DESIGN APPROACH ~ - 4 o Plant Design Objectives Utility (URD), industry, NRC
- System Design a Design Certification Testing o Safety Analysis / Evaluations Safety - single failure, conservative assumptions PRA - multiple failures, realistic assumptions o Constructability, Operability, Maintainability Studies
- a Several Design iterations Completed TU 103497 6
AP600 PRA EVALUATIONS N
- PRA Used As Design Tool 6 major PRA quantifications / design iterations Significant interaction between design and risk analysis Initial PRA study done #11987 PRA submitted to NRC June 1992 Successive PRA studies became more detailed - PRA Related Changes -
Operation changes Start RNS after ADS actuation Containment closure provided during mid-loop Passive features provided during shutdowns Design changes ADS stage 4 valves made diverse from ADS 1/2/3 Diverse I&C functions expanded CMT check valves made normally open, diverse from accum Automatic SGTR protection provided without ADS TL510N97 7
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- AP600 PLANT DESIGN OBJECTIVES $
i - Greatly Simplified Plant Construction, Maintenance, Operation, Safety Increased Operation and Safety Margins l Cost of Power Less Than Fossil or Large Nuclear '
- Five Year Lead Time, Three Year Construction ,
- - Licensing Certainty No Plant Prototype; Proven Components & Systems
- Standard Design for Wide Range of Sites m,-.
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- Simplified Systems Throughout Plant 'l;
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Lower reactor power density Larger Pressurizer l Simplified RCS Loops with Canned Pumps ,
- Passive Safety Systems - Simplified Nonsafety Systems . Digital Based Instrumentation and Control Systems - Compact Control Room with Electronic Operator Interface . Optimized Plant Arrangesnent and Construction Integration of cost / construction / operation / maintenance i -
Extensive use of modular construction TLS 10A/9710
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AP600 INSTRUMENTATION AND CONTROL lE l-a l&C Design .; 3 systems; safety, control, diverse . Microprocessor, software based Multiplexed communications in safety & control systems < a Human Machine Interface Compact control room Wall panel, workstation video & limited dedicated displays Soft controls and limited fixed position controls Design for 1 reactor operator and 1 supervisor I 11.5 10/6/97 12
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AP600 I&C Safety Architecture OPERATIONS AND DATA DISPLAY AND CONTROL CENTERS SYSTEM PROCESSING SYSTEM HUMAN / OPERATOR INTERFACE M RATOR PLANT DIS 1. 48-DESIGN DISRAY ALARM UTED SYSTEM SYSTEM PLANT i COMPUTER ! SYSTEM SAFETY-RELATED PORTIONS i i j L AL l 1 r 1 r 1r REDUNDANT PLANT WIDE DATA HIGHWAY NONSAFETY-RELATED i s i t si a t PROTECTION PLANT DIVERSE SPECIAL NCORE CONTROL ACTUATION MONITORNG NSTRUMENTATION AND SAFETY MONITORING SYSTEM SYSTEM SYSTtiM SYSTEM SYSTEM
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AMSAC/ DIVERSE MONITORING D HER ;1EACTOR CORE REACTOR TRIP AUTOMATIC AhD REACTOR TRIP MANUAL PLANT ESF ACTUATION ^^^ ' ESF ACTUMON MONITORNG CONTROL MONITORNG TION MONITORING NONSAFETY-RELATED NONSAFETY-RELATE!D NONSAFETY-RELATED NONSAFETY-RELATED SAFETY-RELATED DIVERSE O i 10 TLS 6/16/97
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AP600... Simplified Design Based on Proven Technology 50% Fewer 35% Fewer 80% Less 80% Less 45% Less 79% Less
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AP600 SYSTEMS DESIGN E t Passive Safety-Related Systems Mitigates design basis accidents without use of nonsafety-related systems. Meets NRC safety goals without use of nonsafety-related systems Meets ALWR safety goals using nonsafety-related systems Active Nonsafety-Related Systems Reliably support normal operation Minimize use of passive safety-related systems Adverse interactions with safety-related systems evaluated TLs noe7 it
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. T AP600 SYSTEMS DESIGN APPROACH E "
Provide Simple Passive Safety Systems Use " natural" driving forces only One-time alignment of active valves ' No support systems after actuation Actuation is fail safe or powered by batteries No safety AC power, pumps, fans, diesels No operator actions reraired to cool core / containment Full safety design and igulatory oversight Provide Simple Active Non-Safety Systems Use active equipment with lessons learned from operating plants Redundant active equipment powered by nonsafety diesels Minimize unnecessary use of passive safety systems Reduce risk to utility & public . Graded design and regulatory oversight e TLS 10AS718
F AP600 SAFETY SYSTEMS S l Provide Passive Safety Systems Greatly simplify construction, maintenance, operation, ISI / IST n ^ Mitigate design basis accidents without use of NNS systems c Including'at power and shutdown events NRC PRA goals w/o NNS system; URD PRA goals w NNS system Safety Systems Design Features
.Only passive processes; no " active" equipment
. Conservative design for DBA; margins, single failure criteria Best estimate design for PRA; common mode failures No operator actions to cool core / containment
- Safety Equipment Design Features Reliable / experience based equipment '
Full NRC regulatory oversight . Reg Guide 1.26 Quality Group A, B, or C; Seismic I design -l Pre-opt. Testing / Inspection (SSAR chap.14 and ITAAC) Improved inservice testing / inspection capabilities : l Availability controls (Tech Spec with shutdown requirements) l l - Reliability controls (Reliability Assurance Program) T1.510A9719
~i AP600 PASSIVE SAFETY FEATURES lE Passive Decay Heat Removal Natural circulation HX connected to RCS Passive Safety injection N2 pressurized accumulators Gravity drain core makeup tanks (RCS pressu e)
Gravity drain refueling water storage tank (containment pressure) Automatic RCS depressurization . Passive Containment Cooling Steel containment shell transfers heat to natural circulation of air and evaporation of water drained by gravity
- Passive HVAC Compressed air for habitability of main control room Concrete wails for heat sink (MCR and I&C rooms)
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AP600 RCS MAKEUP W
- CVS Makeup Pumps ,
Nonsafety makeup for normal plant operation Can accommodate 3/8" break without SI Two motor driven centrifugal pumps Automatic control and loading on NNS diesel
- Core Makeup Tanks
. Safety makeup to RCS when CVS unavailable or for larger leaks Two tanks provide makeup by gravity at any RCS pressure Automatic actuation opens redundant valves (fail open), each CMT Provides significant makeup before ADS; 3 gpm leak for 100 hr
- PXS Tanks and ADS l
Safety injection for LOCA (DBA); feed / bleed coo'ing (PRA) l Two CMT, two Accumulators and one IRWST provide injection Long term injection from containment recirc (gravity) Staged ADS provides controlled depressurization of RCS Accum backup CMT (PRA) , 71.5 IOA97 2I
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1 AP600 DECAY HEAT REMOVAL M Startup Feedwater System Non-safety feedwater for normal shutdowns and transients ; Two motor driven pumps feed both SGs ; Water supplied from condensate storage tank ; Automatic start and flow control, auto load on NNS diesels Passive RHR Heat Exchanger
. Safety cooling when SFW unavailable and during non-LOCA accidents One heat exchanger connected directly to RCS Forced flow with RCP; natural circulation without RCP Automatic actuation opens redundant valves (fail open)
PRHR HX located in IRWST, provides heat sink, boils in 3 hours Passive containment cooling provides long term heat sink l .. l - RCS Feed and Bleed Provides backup to SFW and to PRHR HX for PRA events Feed from CMT/Accum/lRWST, bleed through ADS Automatic CMT actuation on high RQS temp with low SG level l TLs04/97 21
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E AP600 CONTAINMENT COOLING ljp1 Containment Fan Coolers Nonsafety-related heat removal during normal operation and transients ; 2 coolers, each with redundant fans - Heat sink provided by chilled water, CCW, SW, cooling towers Automatic control and loading on diesels Passive Containment Cooling System
.Provides safety-related heat removal Fan coolers unavailable or during large energy releases Containment shell cooled by water evaporation Water drains by gravity from ele"ated tank Air circulates by natural circulation ,
Automatic actuation opens redundant air operated valves, fail open Other Containment Cooling Features Alternate / long term water supplies PCS ancillary water storage tank and recirc pumps Fire protection pumps or temporary supplies ! Natural circulation of air, without any water, prevents containment failure TLS 10/6M7 25
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~i AP600 PASSIVE SYSTEM RELIABILITY W r
Conservative System Conservative Design / Analysis Development Equip. Design AP600 Testing (Experience, (System, Integral,; EO Testing) V
- In-Plant Activities Conservative (Startup/ITAAC, _
PASSIVE SYSTEM , Safety IST/ISI. Tech RELIABILITY T/H Anolysis Spec RAP) n Emergency PRA Proceedure Success Criter,o i T/H Analysis T/H Analysis PRA CMF/SRF, (Level 1/2/3) TL310N97 29
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-l AP600 NON-SAFETY DID SYSTEMS E .
- Provide Non-Safety Defense-in-Depth Systems Reliably support normal operation Minimize challenges to passive safety systems Not required to mitigate design basis accidents Not required for NRC PRA goals; used for URD PRA goals
- - Non-Safety DID Systems Design Features Redundancy for more probable failures, automatic control
- Power from offsite / onsite sources (nonsafety diesels)
Separation from safety systems
- Non-Safety DID Equipment Design Features Reliable / experience based equipment Reduced NRC regulatory oversight .
Reg Guide 1.26 Quality Group D; limited hazard protection Availability by procedures w/o shutdown requirements (SSAR) Reliability controls (less detailed Reliability Assurance Program) Pre-Opt. tests and Less detailed ITAAC - T1.5104B7 31
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SAFETY VS NONSAFETY SYSTEMS E ~ Safety RTNSS Defense important in Depth - Type System passive active active-Industry Codes ASME 111 ANSI B31.1 ANS B31.1 Seism:c Design i U Bldg Code U Bldg Code Onsite Power Supply 1E non 1E non 1E - - QA Program App B graded graded 4 - SSAR Description /P&lD yes yes yes SSAR Pre-Opt Testing yes yes yes ITAAC Detail full less less PRA, Baseline yes yes yes Focused yes no no - Reporting,10CFR21 yes , no no (NUMARC, IMPO, Lic) yes yes yes - Inspections (NRC, SALP) yes yes yes Availability Controls Tech Spec Admin Proc na Reliability Assurance Prog yes , yes yes TLSSDM/1135
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AP600 SHUTDOWN CAPABILITY M
- Safety Functions Provided During All Shutdown Modes First line of defense is nonsafety-related DID systems .
Reliable, address lessons learned Not required for safety case (SSAR) Passive systems provide safety-related defense Pass.ive Safety-Related Systems Function During Shutdowns Hot shutdown., hot standby, cold shutdown, mid-loop, refueling Same modes of operation as during accidents from power Less demanding conditions (lower decay / sensible heat) Passive Systems Availability Controlled By Tech Spec Defined in SSAR, chapter 16 - i TLS 10A97 37
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AP600 SHUTDOWN SYSTEMS $ MODE Automatic Core Passive RFR IRTST Centamment Containment Depressuration Wakeu;, Cocimg System Tank WODE 5 9 of 10 paths One CMT PPJC heat One irjechen !!av path and Gesure Two water flow RCS pressure boundary OPERABE OPERABE exchanger one itcirculation flow path paths OPERABE closed All paths closed OPERABE OPERABE WODE 5 Stages 1. 2 and 3 None None One injection flow path and Gesure Two water flow RCS pressure boundary open and 2 of 4 one rearculation flow path paths CPERABE open stage 4 OPERABE OPERABE WODE 5 Stages 1. 2. and 3 None None One injechen flow path and Qasure Two water Fw g RCS pressure boundary open and 2 of 4 onc recirculation flow paUn paths OPERAB!I ogn. reduced RCS stage 4 OPERABE OPERABE inventory NODE 6 Stages 1. 2. and 3 None Nene One injection flow path and Ocsure Two water flow Reactor internals in open and 2 of 4 one recirculation flow path paths CPERABE place refueling cavity stage 4 OPERABE OPFRABE not full NODE 6 None None None One injection f!cv path and Gesure None Reactor internals one recirculation flow path removed. refueling OPERABE cavity full TLS 10A97 38
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i l AP600 POST 72 HOUR ACTIONS .7 l Long Term Passive Safety-Related System Operation Licensing Basis No offsite equipment No offsite consumables < 7 days Long term suppod equipment can be nonsafety-related Analyzed for safe shutdown earthquake and extreme winds Alternate long term support provided by safety related connections - Offsite assistance after 3 days (equipment aiid consumables) Realistic / PRA Basis Core cooling and ultimate heat sink remain available indefinitely Without operator action or re-fill of tanks Normal nonsafety-related systems NOT. required to sustain safety functions Recovery of normal operations accomplished when installed normal systems are raturned to service TLS 10N97 39
AP600 POST 72 HOUR ACTIONS $ Safety-Related System Extended Support Actions . Makeup water to the containment , Not needed for one month assuming DBA containment leakage Makt up water to the passive containment cooling water storage tank - A:.cillary water storage tank,2 recirc pumps provide makeup water Air cooling alone maintains containment pressure below design Electrical power for post-accident monitoring instrumentation 2 ancillary diesels provided Cooling for control room 2 ancillary MCR cooling fans provided Only required in hot weather conditions Cooling for post-accident monitoring equipment rooms 2 ancillary cooling fans provided - Only required in hot weather conditions Makeup water to the spent fuel pit Makeup from PCS / ancillary water storage tank for full core offload Required after > 7 days TLS10A97de
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- AP600 Provides Multiple Levels of Defense First is usually nonsafety-related active system Reliable (redundant active components, onsite power)
Lessons learned from operating plants Not required for safety case in SSAR At least one is safety-related passive system Provides safety case in SSAR Other passive features provide additional defense-in-depth Example - passive feed-bleed backs up PRHR HX
- Multiple Levels of Defense Available During Shutdowns Available during hot standby through refueling shutdown One is nonsafety-related active system.-
May be in operation (RNS, CCS, SWS) At least one passive safety-related system also available Not used for normal operation mi.
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$AEIY WANUAL SC ISOL. I LEM 150 LATED '
kANUAL SC 130L. % LEAM ISOLAU CA3E RC5 COOL /DEPRES 8 RCS COOLfXM3ES i i i 4 AUTO .cs.
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' h AUTO CWT. PJLL ADS RW57. PCS I LEAM NOT ISCL RCS 66 CONT FLOOD i
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-l AP600 PRA RESULTS M i
f l Core Damage Frequency Large Release Frequency I At-Power Shutdoum Total At-Power h Total 1.7E47 5.5E-48 2.3E47 1.8E-48 1.4E-es 3.2E-88 i gw 7.7s 4.iE47 s.ie 5.se47 2 se47 a.ie47
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- AP600 TESTING PROGRAM OVERVIEW 4
i October 7,1997 { i Brian A. McIntyre , Westinghouse Electric Corporation f
AP600 Test Program Objectives The AP600 test program objectives were: . To provide information to verify component designs To simulate the AP600 thermal-hydraulic phenomena and behavior of the passive safety systems To provide high quality, qualified data to validate the l computer codes used in the Westinghouse safety l analyses l To support the U.S. Nuclear Regulatory Commission (NRC) design certification review of the AP600 l Each of these objectives was achieved. 1
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o , ! AP600 Test Program Categories , 4 i The AP600 tests can be grouped into three . general categories: l 1. Component Design Verification Tests I
- 2. Passive Containment Cooling System Tests
! 3. Passive Core Cooling System Tests 1 l l' l l - l. l. ( m aac e
. *?
AP600 Component Design '[ 1 Verification Tests . Reactor Coolant Pump (RCP} Air and Water ..
] -
I ~ . l Flow Tests . High Inertia RCP RotorTest I i Incore Instrumentation System Test Reactor Intemals Air Flow Test . i l l 1 i 4 I
- =
1 4 -
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. :l 5
AP600 Passive Containment
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i Cooling System Tests l Heated Plate Test . ! Integral PCS Test ; I Air Flow Path Resistance Test i j Large-Scale Heat Transfer PCS Test l Water Distribution Tests i'
- Wind Tunnel Tests l
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- l. - ._ . . - - _ _ _ _
AP600 Passive Core Cooling System Tests
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Passive Residual Heat Removal Heat Exchanger Test Automatic Depressurization System Test Check Valve Test Core Makeup Tank Test . Full-Pressure, Full-Height Integral Systems Test , Long-Term Cooling Integral Systems Test
. i l
AP900C48 es ,
j 3
%! b j . AP600 INTEGRAL TEST PROGRAM E" ~
4 i j . Two separate integral System Tests examined the behavior of the i l RCS and interconnected systems i j . Full-height, full-pressure test at SPES facility in Italy models APODO RCS (simulates full power operation), CMTs, accumulators, passive ! heat exchanger, IRWST, and automatic depressurization system at 1/395th. scale ) . A 1/4 height, lower pressure (~400 psig) test at Oregon State University ) l models the RCS, passive safety features, including containment sump j for long term y Aty injaMan cooling at ~1/200th. volume scale. i ! . Both integral tests were discussed with and agreed to by the NRC i i
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c-l 1 > l - l l SPES-2 FULL-HEIGHT, FULL PRESSURE TEST 3 l ! i
. Purpose of Tests l
- Obtain TM data at high pressure for computer code verification l
a 7 i s i i
. The SPES-2 test matrix addressed: -
! - SBLOCA simulations (twz 'sk+, location, interactions with i ! nor. safety systems) , l
- Steam line break i Steam generator tube rupture i
l i
. Design basis withAmithout active systems i
! . SGTR with inadvertent ADS actuation k l
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- - . - _ . . _ _ . _ . . _ . _ _ _ -***-e- -. _.__ _
i a .: AP600 TEST PROGRAM l 1 4 i SPES-2 TEST RESULTS i I l l
- Passive Safety System operation (with single failure simulated) .
1 l prevented cose uncovery for all small LOCA's l (up to 84nch DEG, DVI DEG) i j
- SGTR, RCS Qw:=arized and flow to SG terminated by passive .
systems with no operator action 4 ton-safety systems l
- SGTR and large SLB do not result in ADS actuation
- No adverse safety /non-safety system interactions i .
I i upesseeapf &12este 29
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OSU INTEGRAL S STEMS TEST . ! The OSU Test Matrix addressed: 4 4
= SBLOCA Staudation f
Break Size (1/2-inch to 4-inch, DEG DVI) : i ! - Break Location (CL, DVI, CMT, BL) j Non-Safety System interactions i I
- Long-Term Cooling Oc-TC^5 IRWST Draindown j - Transition to Containment Sump injection i - Containment Backpressure Simulation i
i i i i
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, .i y ..y }. !
' = - AP600 TEST PROGRAM , OSU TEST RESULTS . j 1/6 low pressure scaling shown to match FHFP responses l
. Passive Safety System cr~. h (with single failure simulated) prevented core uncovery for all small LOCA's l (up to 8-in. DEG, DVI DEG)
!
- Transition from IRWST to long term injection from containment sump occurred at ~8 hrs.
. Limiting PRA success criteria (multiple failures beyond design basis) identified . .
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o s1 g ng .
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AP600 TEST PROGRAM conclusions l
~ . A comprehensive AP600 test and analysis program has been completed . Key new features confirmed . Tests characterized the unique features at large scales so that computer models can be developed or verified . State-of-the-art safety analysis methods employed . The combined test and analysis program will meet AP600 licensing needs /
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