ML20153F538

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Nonproprietary Slides Used During 930224 Presentation to NRC Re AP600 Testing Program
ML20153F538
Person / Time
Site: 05200003
Issue date: 02/24/1993
From:
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20153F504 List:
References
NUDOCS 9809290159
Download: ML20153F538 (104)
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Text

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WESTINGHOUSE ENERGY CENTER FEBRUARY 24,1993

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H. J. BRUSCHI, GENERAL MANAGER ADVANCED TECHNOLOGY BUSINESS AREA ,

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AP600 PROGRAM OVERVIEW [

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SCHEDULE -

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i DESIGN CERTIFICATION ISSUES IMPACTING FOAKE '

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o ITAAC TESTING REQUIREMENTS o SYSTEM AND EQUIPMENT DESIGN REQUIREMENTS o PLANT LAYOUT t

o REGULATORY TREATMENT OF NONSAFETY SYSTEMS .

o EQUIPMENT SPECIFICATIONS o CONTROL ROOM DESIGN 4

o EMERGENCY RESPONSE GUIDELINES i

o SOURCE TERM o

EQUIPMENT QUALIFICATION o PLANT LAYOUT

- r AP600 TESTING MEETINGS:. .

i o AP600 Testing Program Review Meetings (December '92 - April '93):

December 9,1992 -

OSU Tests Detailed Review  !

t i December 10,1992 - SPES-2 Tests Detailed Review February 25,1993 -

CMT Tests Detailed Review & Facility Visit March 9-10,1993 -

Test Program Review and Facility Visits l

March 23-24,1993 -

Containment Tests Review and Facility Visits  !

April 20,1993 -

ADS Tests Review and Facility Visit April 22,1993 SPES-2 Tests Review and Facliity Visit

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1 AP600 REGULATORY TREATMENT OF NONSAFETY SYSTEMS '-

o THE KEY AP600 LICENSING ISSUE o PLANT DESIGN PHILOSOPHY o ITAAC o TECHNICAL SPECIFICATIONS o RELIABILITY ASSURANCE PROGRAM o JANUARY 22 AGREEMENT IN PRINCIPLE IN LINE WITH -

WESTINGHOUSE APPROACH o WORKING CLOSELY WITH ALWR USC o DEVIL IN THE DETAILS

._ e INTERNATIONAL PARTICIPANTS S.L.KEANEY, MANAGER  ;

PROGRAM CONTROL AND i CONTRACT ADMINISTRATION

AP600 INTERNATIONAL PARTICIPATlON E

AGENCY /

COUNTRY _ ORGANIZATION SCOPE OF WORK Argentina

  • CNEA - Comision National TBD de Energia Atomica Bulgaria
  • NEK - Natsionalona TBD i Eleckricheska Kompania Croatia* Ministry of Industry,. Energy & TBD Shipbuilding; Croatian Electricity .

Generating Board ,

Czechoslovokia Czech Power Board Nuclear Safety Analysis Egypt

  • NPPA - Nuclear Power Plants TBD Authority
i AP600 INTERNATIONAL PARTICIPATION AGENCY /

COUNTRY ORGANIZATION SCOPE OF WORK

'1'  ; I ,

Finland

  • IVO - Imatron Voima Oy TBD Indonesia BATAN - Badan Tenaga Atom plant layout; structural Nasional analysis; BOP design; BPPT - Badan Pengkajan dan electrical design; testing Penerapan Teknologi support i PLN - Perusahann Umum Listrik Negara Italy SOPREN/ANSALDO - (W) NSSS system and component design; systems licensee testing; core design; fluid systems design; PCCS design studies

l AP600 INTERNATIONAL PARTICIPATION AGENCY /

COUNTRY ORGANIZATION SCOPE OF WORK italy ENEA - Comitato Nazionale per ADS testing; full height, la Ricerca e per lo Sviluppo full pressure integral systems Dell 'Energia Nucleare e delle tests; safety system analysis Energie Alternatice ,

ENEA-CRE (research branch of ADS 'esting; ADS valve testing ENEA); part of four party Technical Cooperation Agreement ENEA-DISP (regulatory branch) fluid systems design; in-service subcontractor to inspection requirements; con-SOPREN/ANSALDO tainment analysis

ENEL - Ente Nazionale per probabilistic safety studies; LOCA; L'energia Electnca severe accident analysis i

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AP600 INTERNATIONAL PARTICIPATION l

AGENCY /

COUNTRY ORGANIZATION SCOPE OF WORK Italy FIAT-CIEL - Componenti e impianti fluid systems; NSSS per L'Energia e L'Industria; component design (RV  !

subcontractor to internals, integrated head SOPREN/ANSALDO package, fuel handling and direct participant system) .

Belleli - engineering company; design of passive residual subcontractor to heat removal heat exchanger SOPREN/ANSALDO l a

, SIET - Societe Informazioni full height full pressure Esperienze Termoidrauliche; integral systems tests research company (shareholders =

ENEL and ENEA) t

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l AP600 INTERNATIONAL PARTICIPATION AGENCY /

COUNTRY ORGANIZATION SCOPE OF WORK

, Japan JAPC - Japan Atomic Power l&C design, BOP Company Latvia

  • Latvia Academy of Sciences - I' :i' T B D .

Lithuania

  • Lithuanian Ministry of Energy s TBD Poland IEA - Institute of Atomic Energy safety analysis; PRA; equipment design; BOP design Spain ENDESA - Empresa Nacional de fluid systems; NSSS design Electricidad SA (through EPRI) .

i ENUSA nuclear safety analysis

AP600 INTERNATIONAL PARTICIPATION AGENCY /

COUNTRY ORGANIZATION SCOPE OF WORK Spain INITEC - Empressa Nacional de PSARV module analysis; Ingenieria y Technologia structural steel framing; floor slabs, N1 basemat UNESA - Unidad Electrica SA I&C; Reactor Vessel; Pressurizer, Ni module design UTE - Initec/Agrupacion JV piping system analysis; piping modules design; electrical equip-ment specifications; plant design and layout inside c~ontainment 4

Thailand

  • EGAT - Electricity Generating TBD Authority of Thailand
  • Currently under negotiation

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PROBABILISTIC RISK ASSESSMENT j l

C. L. HAAG, SENIOR ENGINEER RISK MANAGEMENT AND

, OPERATIONS IMPROVEMENT

AP600 PRA AGENDA Passive System Reliability Initiating Event Evaluation Sensitivity Studies of Nonsafety Systems PRA Insights and System importance .

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AP600 PRA -

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PASSIVE SYSTEM RELIABILITY I

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AP600 PRA PASSIVE SYSTEM MODELING

  • Input to calculate system reliability j Detailed design information System success criteria for each initiating event Initial system configuration Required support systems
  • Develop and quantify system fault trees ,
  • Example illustrates calculation of Passive RHR reliability l

_._-_____-______-___-_-___-_____________________________-__-______________________-__-_--_?

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AP600 PRA EXAMPLE PRHR SYSTEM INPUT

  • Detailed Design Information System Specification Document System Functions

System Description

Maintenance and Testing Equipment Description Instrumentation and Controls ,

Electrical Power System Interfaces Piping and Instrumentation Diagrams Major equipment drawings Pipe routing drawings Plant arrangement drawings Technical Specifications I

AP600 PRA EXAMPLE PRHR SYSTEM INPUT 4

  • Initiator u. ; -

Transient event

  • Success Criteria .. m .i PRHR to remove decay heat from..RCS j 1/2 AOVs on HX outlet line must open a initial System Condition Both AOVs normally closed AOVs fail open on loss of air or power
  • Mission Time 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1

= Support Systems Actuated by: Protection and Monitoring System Diverse Actuation System

i AP600 PRA FAILURE CONSIDERATIONS IN A PRHR

. FAULT TREE

  • Equipment Failures AOVs fail to open IRWST ruptures Plugging of flow venturi Instrumentation and control qquipment j
  • Test / Maintenance Consideration AOVs tested every 3 months System available during test Component maintenance unavailability

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L AP600 PRA FAILURE CONSIDERATIONS IN A PRHR  ;

FAULT TREE .  ; ..I 61 ' ~ .i a Operator Actions When automatic actuation fails:

Operator fails to recognize need for decay heat removal Operator fails to actuate PRHR AOVs

  • Common Cause Failures .

Failure of AOVs Instrumentation and Control j 4

4

AP600 PRA '

OTHER FAILURE CONSIDERATIONS OF PRHR

  • Gas Binding in PRHR HX Alarmed in control room Venting performed after maintenance / inspection H2 in RCS is saturated at 30 psig so it can not come out of solution PRHR HX not required at RCS pressure where accumulator could empty (<100 psig) .
  • Heat Transfer Performance .

Performed AP600-specific heat transfer test (full pressure / temp)

Verify with ITAAC (full pressure / temp)

Test HX every refueling (intermediate pressure / temp)

  • Appropriately not modeled in fault' tree

- -- - ---- ------ -- - -- - --- J

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AP600 PRA 5  :

SYSTEM RELIABILITY DATA i 4

i a Primary Source ALWR Utility Requirements Document (Volume Ill)

NUREG/CR-2815 (NREP) 1 NUREG/CR-4550 =

WASH-1400 i IEEE Std 500 i Westinghouse I

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AP600 PRA PRHR SYSTEM RELIABILITY ,

  • Calculated PRHR system reliability Unavailability calculated to be 7.7E-5
  • Equipment in PRHR system similar in duty and design to  ;

i operating plants which justifies the use of historical equipment t

reliabilities.

Single AOV fail to open 1.1E-3 Both AOVs fail to open 1.2E-6 l -

Common cause failure of AOVs 6.2E-5 l

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AP600 PRA 1 a

f INITIATING EVENT EVALUATION i I

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AP600 PRA i INITIATING EVENT EVALUATION j

= Initiating event frequencies for AP600 are based on historical data and AP600-specific analysis i

= Transients Detancd review of operating experience at 51 PWRs from 1984 to mid-1989 (INPO data). Adjusted data as appropriate to account for reduced number of loops. -

a Loss of Offsite Power Frequency based on ALWR URD data

= Loss of Coolant Accidents LOCAs are AP600-specific pipe break analysis

  • Support System Initiators Based on AP600-specific fault tree analysis. Includes loss of CCW, SW, and Compressed Air i

i AP600 PRA ,,,

INITIATING EVENT FREQUENCY DEPENDENCY VS NSS/DID SYSTEMS ,

initiatine Event NSS System DID System Transients:

Turbine trip i I x l less of feedwater flow x Secondary to primary side power mismatch x  ;

Core power excursion x Spurious S-signal x Loss of CCW x Loss of SW x .

Less of compressed air x Main steamline break downstream of MSIV Main steamline break upstream of MSIV Main steam line safety valve stuck open LOOP x LOCAs: ,

Large LOCA '

Medium LOCA CMT line break SI line break Small LOCA Very small LOCA x .

PRHR tube rupture N

Vessel rupture  !

t A IWS x

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AP600 PRA-EXAMPLE LOCA INITdATING EVENT FREQUENCY CALCULATION

  • Very Small LOCA Ruptures in pipes less than 3/4 inch diameter '

Pressurizer level instrumentation lines Miscellaneous primary stem lines < 3/4 inch Frequency calculation equation '

Pipe rupture failure rate x number of pipe sections .

Initiating event frequency is 5.5E-04 /yr l

AP600 PRA .

LEAKAGE EVENTS NRC/Brookhaven reported 39 leakage events (1 gpm - 100 gpm)

  • Westinghouse reviewed events and determined 5 at power events apply to AP600 NRC # LER # Description Leak (gpm) 17 323-89006 Pzr SV seal 10.0 20 339-91011 RHR valve packing 10.0 18 323-91004 1.9 11 302-90001 PORV block valve packing 1.3 28 369-90025 PORV packing 1.0
  • For leaks < 1 gpm, below Tech Spec limit, continue plant operation
  • For leaks 1 - 100 gpm, proceed with orderly shutdown

3 LEAKAGE EVENTS g' .

i Leak  : CVCS Makeup *

  • Orderly Shutdown 1 - 100 gpm operational c.m yA , ,e Repair / recover CVCS "" = Orderly Shutdown before Si actuation.

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Orderly Shutdown after CMT actuation no 1 r

" ' eT ADS Actuation

AP600 PRA SENSITIVITY STUDIES OF NONSAFETY SYSTEMS ,

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AP600 PRA AP600 NON-SAFETY SYSTEM SENSITIVITY CASE
- n l Estimated Core Damage Frequency At Power Shutdown Total

. i Base Case 3.3E-7 /yr 8.9E-8 /yr 4.2E-7 /yr

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Sensitivity Case 2.6E-6 /yr 5.4E-7 /yr 3.1 E-6 /yr NRC Goal 1.0E-4 /yr Note: Sensitivity case removes credit for CVS, SFW, RNS, offsite power and DGs following an initiating event.

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AP600 PRA AP600 NON-SAFETY SYSTEM SENSITIVITY CASE Estimated Release Frequency-At Power Shutdown Total l Base Case 2E-8 /yr 1E-9 /yr 2E-8 /yr Sensitivity Case 2E-7 /yr 7E-8 /yr 3E-7 /yr NRC Goal 1 E-6 /yr Note: Sensitivity case removes credit for CVS, SFW, RNS, offsite power and DGs following an initiating event I

AP600 PRA t

PRA INSIGHTS AND SYSTEM IMPORTANCE t

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AP600 PRA PRA INSIGHTS VERSUS IMPORTANT ANALYSIS

Identified insights in AP600 PRA report (Chapter 17)

Insights are changes made to the design, operation, or PRA success criteria .  : l' -

Insights are not intende.d;to.be a listing of the risk important features of the plant

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= importance Analysis .

Used in response to some RAls

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AP600 PRA PRA SYSTEM IMPORTANCE RAI 720.13 - Requested system level importance Results of RAI 720.13:

  • Gravity injection Largest increase in core damage and release frequencies System required for Si line break and large LOCAs .
  • Passive RHR Second largest increase in core damage and release '

frequencies Accumulators, CMTs, ADS Stages 1-3, ADS Stage 4 Small increase in frequencies due to system redundancy

APG00 PRA PRA SYSTEM IMPORTANCE

- Startup Feedwater, Normal RHR, and DGs Negligible impact on core damage and release frequencies

- CVCS Relatively minor importance on core damage Small increase in release frequency due to LOCA e. vents with a large, pre-existing opening in containment

l AP600 PRA AP600 PRA INSIGHTS -

= Success criteria changes Accumulator or CMT for small or medium LOCAs One accumulator for large LOCA Multiple ADS valve failures  ;

= Operation changes Start NRHR after any ADS Require passive core cooling features during shutdowns lST test intervals (ADS valves)

= Design changes NRHR valves made remote 4th stage ADS valves diverse Expanded diverse l&C capabilities Added redundant IRWST injection check valves Added redundant / diverse IRWST recirc valves Made CMT check valves normally open

- _ _ - - - - _ - _ _ _ - - . - - - - - - - - - - _ - - - _ - - - - _ _ _ _ _ - - - - - - _ _ _ _ _ - - _ _ - - _ ______I

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r REGULATORY TREATMENT OF

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NONSAFETY SYSTEMS .

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T. L. SCHULZ, FELLOW ENGINEER

! SYSTEMS AND EQUIPMENT ENGINEERING .

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AP600 REGULATORY TREATMENT OF NONSAFETY SYSTEMS

  • Passive Systems Defense-in-Depth Capabilities
  • Passive Systems Capabilities During Shutdowns l
  • Passive Systems Long Term Capabilities (Hurricane Andrew / post 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />)
  • Nonsafety DID System Safety isolation Functions
  • Reliability of important Systems / Components

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PASSIVE SYSTEM DID CAPABILITIES

  • Passive Safety Systems Provide Defense-in-Depth Capabilities Some provided in original design; others provided in design

! changes incorporated to improve PRA

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Supported by best estimate analysis t

'AP600 LEVELS OF DEFENSE FUNCTION CURRENT PWR . AP600 REACTOR SHUTDOWN CONTROL RODS (BREAKERS) - CONTROL RODS (BREAKERS)

- RIDEOUT (NEG MTC,

- CONTROL RODS (MG SETS)

AMSAC, AFWS, CVCS) - RIDEOUT (MORE NEG MTC, DAS, PRHRS / SFWS, CMT / CVCS)

RCS OVERPRESSURE - PZR PORY - LARGER PZR PROTECTION - Hi PRES TRIP - HI PRES TRIP

- PZR SAPETY VALVES - PZR SAFETY VALVES RCS HEAT REMOVAL - MAIN FEEDWATER SYS - MAIN FEEDWATER SYS

- AUX FEEDWATER SYS - STARTUP FEEDWATER SYS

- MANUAL PEED / BLEED - PRHR HX (PZR PORV, HHSI) - AUTO FEED 4LEED (CMT/IRWST, ADS)

- MANUAL FEE 04LEED (ACCUM / NRHRS, ADS)

HIGH PRESSURE - CVCS PUMPS - CVCS PUMPS INJECTION - HHSI PUMPS -CMT

- ACCUM /1RWST (ADS)

- ACCUM / NRHRS (ADS)

LOW PRESSURE -ACCUM -ACCUM INJECTION - LHSI PUMPS - IRWST (ADS)

- NRHRS PUMPS LONG TERM RECIRC - LHSI PUMPS FEEDING - CONTAINMENT SUMP (ADS)

HHSI PUMPS - NRHRS PUMPS CONTAINMENT HEAT - FAN COOLERS - FAN COOLERS REMOVAL - CONT SPRAY PUMPS / HX - EXTERNAL AIR + WATER DRAdM

- EXTERNAL WATER FIRE SYSTEM

- EXTERNAL AIR ONLY COOLedG TLS - 2/10/93

'AP600 DECAY HEAT REMOVAL Startup Feedwater System Non-safety feedwater for normal shutdowns and transients Two motor driven pumps feed all SGs Water supplied from deaerating heater or CST Automatic start and flow control Automatic load on NNS diesels Passive RHR Heat Exchanger Safety dooling for events whers SFW is unavailable and during non-LOCA accidents Two heat exchangers connected directly to RCS Forced flow with RCP; natural cire without RCP Automatic actuation; two fail-open valves PRHR HX located in IRWST, provides heat sink IRWST remains subcooled for 2 - 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Passive containment cooling provides ultimate heat sink RCS Feed and Bleed Provides backup to SFW and to PRHR HX for PRA multiple failure events Feed from CMT/Accum/lRWST, bleed from ADS Automatic actuation of CMT on high RCS temp with low SG level TLS - 2/10/93

AP600 -

MAIN & STARTUP FEEDWATER SYSTEMS D .

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STORAGE TANK ,

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FEEDVATEg TEEDVATER v . l 300 STER NS PUMPS GE3ID V ,

HIGH PRES $URE HEATERS i

CONDENSATE 11DRAGE TANK  ;

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CORE MAKEUP nErutt TANK (1 OF 2) cauty

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LOSS OF OFFSITE POWER LOOP l

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AUTO SFWS, CVCS

& SUCCESS NO ADS I

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AUTO PRHR HX, CMT, SAFETY CASE PCCS SUCCESS NO ADS T

AU TO CM T, PARTI AL ADS = SUCCESS MAN NRHRS INJECT ADS, NO FLOOD i

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AUTO CMT, FULL ADS, IRWST PCCS SUCCESS ADS, FLOOD l

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t MAN PART ADS AUTO ACCUM, IRWST, O SUCCESS PCCS AOS, FLOOD l

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CORE DAM ACE

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j - iMant: AP600 Event: LOSS MAIN FEEDWATER at FULL POWER a

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j Non Safety Safety l l Diverse F unction System Order of Use PLS DC AC j f lPMS(1) DC lDAS HW i

s l o Reactor Shutdown l 1. Control Rods - - -

A - - -

i

2. Control Rods - - - - -

A -

1 3. Control Rods - - - - - - M l j 4. Ride Out (2) M Yes Yes - -

A M o RCS Inventory Control j

1. CVS A Yes Yes - - - -

2.CMT - - -

A - - -

3.CMT - - - - -

A -

4.CMT - - - - - -

M

5. CMT, RNS, pan ADS M Yes Yes A Yes - -
6. CMT, IRWST, full ADS - - -

A Yes - - '

7. Accum, RNS, part ADS M Yes Yes -

Yes - M

8. Accum, IRWST, full ADS - - - - -

Yes -

M o RCS Heat Removal 1.SFW A Yes Yes - - - -

2. PRHR HX - - -

A - - -

3. PRHR HX - - - - - A -

4 PRHR HX - - - - - - M

5. CMT. RNS. part ADS M Yes Yes A Yes - -
6. CMT, IRWST, full ADS - - -

A Yes - -

7. Accum, RNS. part ADS M Yes Yes -

Yes - M

8. Accum, IRWST, full ADS - - - -

Yes - M o Containment Cooling

1. Fan Cookm A Yes Yes - - - -
2. CV extemal air, water drain - - -

A - - -

3. CV extemal air, water drain - - - - - A -
4. CV extemal air, water drain - - - - - - M
5. CV extemal water fire sys only M Yes Yes - - -
6. CV extemal air only - - - - - -

Notes:

1) Safety components have safety related MCB martal cortrois vu both indNdual soft control swechos ans dedicated system level sweenes.
2) Raactor is shut down by negative moderator terrporatre coeffloert as the coolant heats up Re@wes mammans RCS pressure reiset, turtune tnp. and PRHR HX actuanon. Also toquaes manual CMT or CVS borsson

. . ~ . . . . ..- - - ..- - - . . . . - - - - - - . - ... _- - .. -- - - - -... .- - - - - _ . - - . - . - - -

AP600 -

RCS _ EA < (0-3/8")

S LEAK i

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AUTO CVCS, SFWS NORMAL SHUTDOWN > SUCCESS NO ADS I

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AUTO CMT, PRHR HX, PARTIAL ADS > SUCCESS MAN NRHRS INJECT ADS, NO FLOOD i

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t AUTO CMT, PRHR HX, SAFETY CASE FULL ADS IRWST, O SUCCESS PCCS ADS, FLOOD i

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t MAN PART ADS AUTO ACCUM > SUCCESS MAN NRHRS INJECT ADS, NO FLOOD l

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9 MAN FULL ADS AUTO ACCUM, IRWST, 5 SUCCESS PCCS ADS, FLOOD l

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AP600 ADS DEFENSE-IN-DEPTH i

= DBA Performance i

Conservative decay heat, line resistances, pressure drop calc,

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containment pressure i .

Successful IRWST gravity injection achieved with single failure i

Limiting failure is one 4th stage valve or one battery train i

(causes failure of one 1st & 3rd stage valves) i

4 AP600 ADS DEFENSE-IN-DEPTH (Continued) .

  • PRA Performance Best estimate decay heat, line resistances, pressure drop calc, containment pressure Successful IRWST gravity injection achieved with multiple failures Can tolerate common mode failurb of all stage 1/2/3 valves or all stage 4 valves Successful RNS pump injection achieved with opening of any one 2/3/4 stage line i

e ADS sizing basis provides substantial margin and failure tolerance

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i PASSIVE CAPABILITY DURING SHUTDOWNS

. Passive Safety Functions Provided During All Shutdown Modes

  • Hot Shutdown / Hot Standby / Qold Shutdown Same As At Power ,.

l -

Tech Spec require PRHR HX, CMT, IRWST, and ADS to be available .g ..

i Cold Shutdown Mid-Loop

=

PRHR HX ineffective (RCS open) ., .

l CMT / accum unnecessary ,,q.

Tech Spec require:

Containment integrity 3,  !

ADS stages 1,2,3 open l

IRWST MOV available  !

i

= Refueling Shutdown Refueling cavity provides >72 hours with equipment hatch open Equipment hatch can be closed without AC power I

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SAFETY CASE IRWST,PCCS SUCCESS '

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, OTE (1) EITHER CLOSE CONTAINMENT OR PROVIDE ADDITIONAL MAKEUP AFTER 72 ~4 i - N@N I

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AP600 -

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Passive Safety-System Long Term Shutdown Capabilities t,

After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Hurricane Andrew i I

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' POST 72 HOUR ACTIONS

= Long Term Passive Safety System Operation .

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Core cooling and ultimate heat sink remain available indefinitely

(>> 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) without operator action or offsite support Other safety functions require limited offsite support after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Limited offsite support after 72 hhrs. ,

Uses readily accessible and transportable equipment and supplies from offsite

. Safety-related connections provided to engage offsite support equipment Installed nonsafety systems NOT required to sustain safety system functions Recovery to cold conditions accomplished when nonsafety systems are made available

AP600

~

I POST 72 HOUR ACTIONS  !

i a Safety System Extended Support Actions Provide makeup water into containment Only needed after one month assuming DBA containment leakage Provide makeup water to the passive containment cooling water l storage tank Air cooling alone maintains containment pressure below design pressure Provide electrical power to supply the post-accident and spent fuel pit monitoring instrumentation Provide electrical power to the hydrogen recombiners .

Only needed for events where containment hydrogen buildup is a concem i I

4 l

AP600 POST 72 HOUR ACTIONS .

i t

a Safety System Extended Support Actions (continued) -

j Provide breathable, compressed air for the control room air supply and pressurization system .

Only required in case of serious core damage and containment leakage .

l i .ip- .  ;

Provide control room cooling and air recirculation  !

Only required in hot weather conditions Provide ventilation cooling to post-accident monitoring equipment l rooms l

Only required in hot weather conditions Provide makeup water to the spent fuel pit 7 days at BOL,21 days at EOL 72 hr for worst case emergency core unload

+

i HURRICANE ANDREW /AP600 CAPABILITY ~l CORE COOLING / DECAY HEAT REMOVAL t

Turkey Point Actions Shutdown in advance (T < 350 F)

  • D/Gs tested in advance
  • RHR system used to cool core

. Power from D/Gs after loss of grid

= Maintained potential for cooling via SGs with auxiliary feed. pump

  • Stayed on D/G for nearly 7 days until reliable offsite source restored .

AP600 Caoabilities -

  • Redundant D/Gs and normal RHR provide equivalent capability as used at Turkey Point
  • Passive systems backup normal systems and D/Gs i

i l

HURRICANE ANDREW /AP600 CAPABILITY AP600 PASSIVE SYSTEM BACKUP CAPABILITY a Passive Systems not needed unless both D/Gs are lost => Station Blackout -

  • Upon station blackout PRHR would be actuated - natural circulation cooling to IRWST - boiloff to containment after 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> -- heat removal through containment shell -- condensate retum to IRWST
  • PRHR maintains RCS at T < 450 F, p < 425 psia, no primary inventory loss
  • Continue on PRHR indefinitely as long as 1E battery capacity can be maintained - 20 kw AC generator within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is sufficient
  • If 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1E battery nears depletion, ADS is actuated (RCS @ ~T = 425 F, p = 310 psia) i
  • Continue to cool core indefinitely via recirculation witilin containment

l HURRICANE ANDREW /AP600 CAPABILITY 6

STRUCTURAL DESIGN t i

Turkey Point Safety Related Structures

  • No damage
  • Tomado design loads more limiting than wind design loads l

AP600 Safety Related Structures

  • Also tomado limited, no damage  ;

4 Turkey Point Non-Safety Related Structures

  • Built to South Florida building code
  • Majority of structures survived Andrew well
  • Central receiving facility had damage to structure and some contents .

! AP600 Non-Safety Related Structures

  • Standard design - limiting wind is South FR)rida - URD 110 mph /50 year

~

AP600 i

t Nonsafety Defense-in-Depth System Safety isolation Functions ,

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AP600 NONSAFETY SYSTEM ISOLATION FUNCTIONS i Nonsafety Systems Provide Some Safety Related Isolation

^

Functions RCS pressure boundary isolation  !

Containment isolation Other isolation functions provided to mitigate DBA's

  • These Isolation Capabilities Are Fully Safety Related Single failure capability  !

-- Reg Guide 1.26 quality group A, B, or C Seismic l Tech Spec controls Described in SSAR and ITAAC  !

i l

. i AP600 ~l j NONSAFETY SYSTEM ISOLATION FUNCTIONS t

  • Example: CVS Functions -

t

  • CVS Functions Safety Functions RCS pressure boundary isolation ,

- Containment penetration isolation Boron dilution accident' termination

- Excessive makeup isolation l

DID functions

- RCS makeup for leaks RCS pressure reduction t

4 (shay suno ukoi.n .

AP600 -

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AP600

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Reliability of important Systems / Components i t

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AP600 RELIABILITY OF SYSTEMS / COMPONENTS

  • Importance of Systems / Components identified in AP600 PRA-PRA functions identified in SSAR descriptions
  • Reliability Controlled By Design / Construction / Operation Activities Design; PRA insights / importance, historical experience, equipment vendor inputs, feasibility testing (where required)

- Construction; QA/QC, ITAAC Operation; Tech Spec, IST/ISI, 0-RAP

AP600 -

3

)

RELIABILITY OF SYSTEMS / COMPONENTS Design Phase Reliability Activities i

i

  • PRA Insights / Importance Passive safety systems important Importance ranking -

lRWST injection .

PRHR HX ADS stages 1,2,3 ADS stage 4 Accum CMT

AP600 .

L RELIABILITY OF SYSTEMS / COMPONENTS i

Design Phase Reliability Activities (Continued)

  • Equipment Requirements Developed To Avoid Historical Problems Utility inputs Equipment vendor inputs .

Incorporate into E-Spec .

Do not specify component reliability -

Vendors don't know quanitative reliability Vendors unwilling to quote; at best will increase cost to cover their risk

  • Feasibility Testing As Required Only Required for IRWST check valves and ADS valves Demonstrates operability not reliability Use historical reliability for all valves

i

.i i

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i INSPECTIONS, TESTS, ANALYSES ..

i AND ACCEPTANCE CRITERIA t ll B. A. MclNTYRE, MANAGER -

ADVANCED PLANT SAFETY AND LICENSING i esim i

l-k

i 4 AP600 ITAAC PROGRAM '

i o AP600 ITAAC PROGRAM WILL BE DIFFERENT l

o SCREENING CRITERIA  !

o DECEMBER 15,1992 SUBMITTAL- ,

i o 36 SYSTEM ITAAC l

, o 12 SAFETY SYSTEMS' '

o 24 NONSAFETY SYSTEMS o DEFENSE IN DEPTH AND' SUPPORT o NONSYSTEMITAAC o HUMAN FACTORS o NUCLEAR ISLAND BUILDING o SAFETY RELATED PIPING o INTERFACE l

. TIER 1 Design Description /ITAAC Screening Criteria Checklist

Purpose:

The purpose of this checklist is to determine those systems and structures for which Tier 1 design descriptions and associated ITAACs must be prepared.

1 I. System or Structure:

II.' Evaluation:

A. Are there any stnictures, systems, or components classified as Class A, B, or C7 YES NO Justification:

(

. B. Are there any structures, systems, or components classified as Class D because they provide defense in-depth functions YES NO as defined in GW G1010, AP600 Nuclear Safety Classification and Seismic Requirement Methodology?

Justification:

1 4

III. Conclusion Tier 1 Design Description and associated ITAAC will be developed for this system. YES NO Justification:

System Engineer /Date

~

Technical Review Team wmeeimmam.

AP600 ITAAC I

o REACTOR  ;

o Fuel Handling and Refueling System i o Reactor Coolt,nt System  !

o Reactor System

, t o NUCLEAR SAFETY SYSTEMd" !"... I i!! il .

o Automatic Depressurization System o Containment System '

o Passive Containment Cooling System '

o Passive Core Cooling System o Steam Generator System o Main Control Room Habitability System s o INSTRUMENTATION AND CONTROL o Diverse Actuation System o Data Display and Processing System o incore Instrumentation System o Plant Control System o Protection and Safety Monitoring System o Radiation Monitoring System

i

i AP600 ITAAC ,

i o AUXILIARY SYSTEMS o Component Cooling Water System o Chemical and Volume Control System i o Standby Diesel and Auxiliary Boller Fuel Oil System

o Fire Protection System o Mechanical Handling System i o Primary Sampling System o Normal Residual Heat Removal System  ;

o Spent Fuel Pit Cooling System -

o Service Water System o Containment Hydrogen Control System .

o STEAM AND POWER CONVERSION SYSTEMS o Main and Startup Feedwater System o Main Steam System .- n !' . -

I i

AP600 ITAAC c.Lii,..;,. ,

o ELECTRICAL POWER o Main AC Power System o Non Class 1E DC and UPS System o Plant Lighting System a i e' o Class 1E DC and UPS System o Onsite Standby Power System . .

o HEATING, VENTILATING, AND AIR CONDITIONING SYSTEMS o Nuclear Island Nonradioactive Ventilation System -

o Central Chilled Water System i o Annex / Auxiliary Building Nonradioactive Ventilation System o Diesel Generator Building Ventilation System t

e AP600 ITAAC PROGRAM i

o PROGRAY PLANS o KEEP ITAAC OFF CRITICAL PATH .

o REVIEWED WITH STAFF, FEBRUARY 16 o MAINTAIN INVOLVEMENT IN INDUSTRY ITAAC ACTIVITITES o MARCH 9, GE BUILDING o MARCH 9, GE ELECTRICAL DISTRIBUTION i

o ABB/CE INDUSTRY REVIEWS

! o NUMARC ACTIVITIES .

3

AP600 ITAAC PROGRAM i

o UPDATE AP600 TO CURRENT INDUSTRY STANDARDS o FLUID AND MECHANICAL SYSTEMS o APRIL 15, .1993 o BUILDING, ELECTRICAL AND I&C o JUNE 1,1993 .

o

i i

TESTING E. J. PIPLICA, MANAGER TEST ENGINEERING l

J AP600 TEST PROGRAM a DESIGN CERTIFICATION TESTS

- PASSIVE CONTAINMENT COOLING SYSTEM TESTS

- PASSIVE CORE COOLING SYSTEM TESTS

'.~
; t a DESIGN VERIFICATION (ENGINEER,1NG) TESTS

- COMPONENT DESIGN VERIFICATION TESTS J

RELATIONSHIP TO THE DESIGN PROCESS i - PERFORM A RANGE OF TESTS TO OBTAIN HIGH FIDEllTY DATA I

- TEST DATA USED TO DEVELOP OR VERIFY MODELS USED IN EXISTING j COMPUTER CODES

- COMPUTER ANALYSIS IS PERFORMED MODELING THE TEST FACILITY GEOMETRY

- SAME COMPUTER CODES ARE USED TO ANALYZE THE PERFORMANCE OF THE AP600

- IF NECESSARY, THE AP600 DESIGN IS OPTIMlZED BASED ON THE RESULTS OF COMPUTER ANALYSIS I

--- - - - - _ - -- - - D

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t AP600 TEST OBJECTIVES / CONCLUSIONS  :

i L

i Pesehe sedeep syeeman Teene Test Objecth C-  ; _ _ . Test Resulte Required for Date

  • SSAR FDA/DC Design Ver* =mari Poemare RHR HX (Wesanghouse STC) Deserenad heat transfer characteristcs of im PRHR heat exchanger and menng charactermeics in to IRWST r

e-3 Coe,,te.ed. x '

Der, ember 1989 s Phase 2 Completed. X  ;

Odober 1990 ,

Aumenmac Cm mason System Test (Caessoa, luly) f e Phase A Sparger Porturmance & Tar

  • Loade Contrmed the capacity of tw sparger and deteraned the dynamic Completed. X estece on the IRWST strucmare November 1992 o Phase a Veke Performance To semulate operaeon of to automaec depressuruseon system and to March 1994 X l mnhrm We eM and operabihty of im ADS valves Case adehaup Tank p__.;== AESD) To verty the yawty drain behavior of to core makeup tank over a full June 1993 X  !

tange of Row rates and pressures and to verty em operanon of the tank towelinstumenteeion ,

Lang Term Cootne tire Scais) To prowde data to evaluate em operaton of im PXS at low pressure December 1993 X tOregon Steen Ureveresyn oess (Ceneene uns,ersey to ==nend em eosing crocas heat nux corredrkm for W kml June 1993 X i aseemthee at lower now conemons tJ g 5d Penne ne meagres Syeesse Tees le provide data to evaluate em operaton of im PXS at hsgh pressure. December 1993 X  !

eget t time, en seesyt eusueng response to small break LOCA. tube ruptures and steamene treek eenesenes l

i I

i i

f

AP600 TEST OBJECTIVES / CONCLUSIONS Penelve Centehment Coeting Teets Teet C ._T: _ _ - c: ,- - r - . Test Reeules Ikt uired for Deee SSAR FDA/DC Design Venhcaton Inuyal PCCS Test (Wootnghouse STC) Demonswated operesion of the PCCS over tio lut range desgn base Completed. X oPeretng mrdsons July 1992 t,arge Scade PCCS (Weeanghouse STC) To demonstate operaton of the PCCS on a scaled stucane which Phase 1- X accurately models boti tie conteenment dome and sudewa8 heet May 1992 transler areas, and inside contesnment strucanes Phase 2 -

X June 1993 PCCS Weser Deereuton (Wenanghouse AESD)

Phase i Caneer of Dame. 20 Diameenr Demonstated tio eNectweness of water dannbuton on the center of Completed. X tie contamment dome June 199I Phase 2 Fid Scots lie Secton Demoravated the e#ecibonoss of wamr dennbuton on the Completed X consamment dome and upper sidowsN January 1992

  • Phase 3 Fue Scale lit Secton Verty design of water dumbuton system Apnl1993 X i PCCS Wind Tunnel (Unsversdy of Western Onneno) o Phase t Overes leiddng eftscas (lil00ei scafel Dessionstrated tie onnd irdsced pressure on the mntamment shaald Completed. X hudeng em to mar inlet /oudet conhguratons and srte sauceres Ju6y 199I o Phase 2 Deended tendes Fnal ver*m a si100ih scales Denomimed basene loodmg and demonstated the effect of wmd on Compassed. X osseamment annulus av Row February 1992 e Pmmee 4A at te2Sei acase a t m er.aeo le porturm tente at tugher Reynolds numbers and to demonstrate September 1993 X eAnt of oute geography

AP600 TESTS IN PROGRESS a INTEGRAL TESTS i

- SPES-2 FULL HEIGHT FULL PRESSURE TEST

- OSU 1/4 HEIGHT SCALED PRESSURE TEST ,

- PCCS LARGE SCALE HEAT TRANSFER TESTS

- SEPARATE EFFECTS TESTS

- CORE MAKEUP TANK TEST

- AUTOMATIC DEPRESSURIZATIONB TEST - PHASE 3

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'l CONCLUSIONS g i COMPREHENSIVE AP600 TEST AND ANALYSIS PROGRAM HAS BEEN

. DEVELOPED AND IS IN PROGRESS i

I DATA FROM TEST PROGRAMS ALREADY INCORPORATED IN SSAR ANALYSIS

- KEY NEW PLANT FEATURES ARE BEING TESTED TEST WILL CHARACTERIZE THE UNIQUE FEATURES AT LARGE SCALES SO THAT COMPUTER MODELS CAN BE DEVELOPED OR VERIFIED  ;

COMBINED TEST AND ANALYSIS PROGRAM WILL MEET THE LICENSING NEEDS TEST DATA WILL BE FORWARDED TO THE STAFF AS TEST SERIES ARE

- COMPLETED UPCOMING TESTS ARE BEING CAREFULLY FOLLOWED TO MEET TEST  !

OBJECTIVES AND SCHEDULES j

_---- - - _- ---_-- L

(D

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v) w- : ,% o  ;

SA a}Gw l

1 Westinghouse Energy Systems em sse l

Electric Corporation- Pasburgh Pennsylvarta 15230 0355 !

1 DCP/NRC1409 l NSD-NRC-98-5753 l Docket No.52-003 1 August 13,1998

' Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:' T. R. QUAY l 1

SUBJECT:

RESPONSE TO NRC LETTERS CONCERNING REQUEST FOR WITHHOLDING INFORMATION

Reference:

1. Letter, Sebrosky to McIntyre, " Request for withholding proprietary information for  ;

Westinghouse ietters dated December 14,1992, and December 17,1992," dated  !

July 10,1998.

^

. 2. Letter, Huffman to McIntyre, " Request for widiholding information from public

- disclosure of Westinghouse AP600 design letters of December 15,1992," dated July 14,1998.

v

3. Letter, Sebrosky to McIntyre, " Request for withholding information from public disclosure for Westinghouse AP600 design letter of February 24,1993, April 19, 1993, and July 14, 1993," dated June 18,1998.
4. Letter, McIntyre to Quay, " Status review of AP600 proprietary submittals," dated September 18,1995.

Dear Mr. Quay:

Reference 1 provided the NRC assessment of the Westinghouse claim that proprietary information was provided in a letter dated December 14,1992, that provided the NRC with copies of presentation material from a management meeting held December 14,1992, discussing the AP600 testing program.

The NRC has no record of a nonproprietary version of the slides being provided. At the time this l ._ presentation was made, ue information was proprietary since that description of the AP600 testing i program had commercial value to Westinghouse. At this time, almost six years later, this information does not have commcreial value and is no longer considered to be proprietary by Westinghouse.

Reference 1 also provided the NRC assessment of the Westinghouse claim that proprietary information was provided in a letter dated December 17,1992, that provided the NRC with copies of presentation

. material from a meeting with the technical staff held December 9-10,1992, discussing the AP600 afs?, og

+

G0/20*'d 2002Gl?t0CI8 01 2LLP PLC 21P 3Ti! A3OdNOW-Odd dd 22:Cl 86,CI 900

- , ~ . ~ ~ . - - - - - . -- ._ - - . _ - - . . - - - - - - _ - _ - .

a .

DCP/NRCl409 NSD-NRC 98-5753 August 13, 1998 l

testing program. The NRC has no record of a nonproprietary version of the slides being provided. At i the time this presentation was made, the information was proprietary since that description of the AP600 testing program had commercial value to Westinghouse. At this time, ahnost six years later, this information does not have comrnercial value and is no longer considered to be proprietary by Westinghouse.

, Reference 2 provided the NRC assessment of the Westinghouse claim that proprietary information was  !

provided in a letter dated December 15,1992, that contained a preliminary description of the AP600 '

refueling outage plan activities. The NRC assessment was that no material in the letter was specifically identified as being proprietary and that a nonproprietary version was not provided. At the time this subject was being discussed with the NRC technical staff, the information was considered to he' proprietary by Westinghouse since it contained information that had commercial value to l

Westinghouse. At this time, almost six years later, this information does not have commercial value  ;

and is no longer considered to be proprietary by Westinghouse. '

Reference 3 provided the NRC assessment of the Westinghouse claim that proprietary information was l provided in a letter dated February 24,1993, that contained presentation materials from the February

) 24,1993, Westinghouse /NRC AP600 senior management meeting. The NRC assessment was that the l material was similar to material that exists in the current (1998) nonproprietary version of the AP600 l probabilistic risk assessment and AP600 standard safety analysis report. In addition the staffindicated the material was used by the staff in the development of the AP600 draft safety evaluation report and l therefore should remain on the docket. Our 1995 request, Reference 4, indicated that the material

provided in the Westinghouse letter of February 24,1993, was presentation material that was intended l v

! for clarification only, not part of the formal review material and requested that the material be returned l i to Westinghouse. At the time this subject was being discussed with the NRC technical staff, the i information was considered to be proprietary by Westinghouse since it contained information that had  !

commercial value to Westinghouse. If this presentation material was indeed used by the staffin l development of the AP600 draft final safety evaluation report in November 30,1994, then at this time, l over five years later, this information is no longer considered to be proprietary by Westinghouse.

l Reference 3 provided the NRC assessment of the Westinghouse claim that proprietary information was

provided in a letter dated April lf,1993, that contained presentation materials from the April 20, 1993, AP600 overview. The NRC assessment was that the material was similar to material that exists in the current (1998) nonproprie'.ary version of the AP600 probabilistic risk assessment and AP600 standard safety analysis report. In addition the staff indicated the material was used by the staff in the l development of the AP600 draft safety evaluation report and therefore should remain on the docket.

l- Our 1995 request, Reference 4, indicated that the material provided in the Westinghouse letter of l April 19,1993. was presentatioa material that was intended for clarification only, not past of the formal review material and requested that the material be returned to Westinghouse. At the time this subject was being discussed with the NRC technical staff, the information was considered to bc l proprietary by Westinghouse since it contained information that had commercial value to l Westinghouse. If this presentation matedal was indeed used by the staffin development of the AP600 l draft final safety evaluation repon in Novereber 30,1994, then at this time, over five years later, this

. information is no longer considered to be propiietary by Westinghouse.

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DCP/NRCl409 NSD NRC-98 5753 August 13,1998 Reference 3 provided the NRC assessment of the Westinghouse claim that proprietary information was provided in a letter dated July 14,1993, that contained presentation materials from the July 14,1993, meeting where the AP600 main control room habitability was discussed. The NRC assessment was that the material was similar to material that exists in the current (1998) nonproprietary version of the AP600 probabilistic risk assessment and AP600 standard safety analysis report. In addition the staff indicated the material was used by the staffin the development of the AP600 draft safety evalt.ation report and therefore should remain on the docket. Our 1995 request, Reference 4, indicated that the material provided in the Westinghouse letter of July 14,1993, was presentation material that was intended for clarification only, not part of the formal review material and requested that the material l

be returned to Westinghouse. At the time this subject was being discussed with the NRC technical

! staff, the information was considered to be p prietary by Westinghouse since it contained information i that had commercial value to Westinghouse. .f this presentation material was indeed used by the staff in development of the AP600 draft final saft evaluation report in November 30,1994, then at this time, over five years later; this information is no longer considered to be proprietary by Westinghouse.

His response addresses the proprietary issues delineated in the references.

A Brian A. McIntyre, Manager V

l Advanced Plant Safety and Licensing l

jml cc: J. W. Roc - NRC/NRR/DRPM J. M. Sebrosky - NRC/NRR/DRPM W. C. Huffman - NRC/NRR/DRPM l H. A. Sepp - Westinghouse t

l l

l l

a 1797a wpf G0/PO*d 2002Gir10C18 01 2LLr PLC 21r 311to30dNOW-Dae ad C2:CI 86.C1 One

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