ML20153G396

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Nonproprietary Presentation Matl from Westinghouse/Nrc 930420 Meeting on AP600 Design
ML20153G396
Person / Time
Site: 05200003
Issue date: 04/20/1993
From:
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20153G383 List:
References
NUDOCS 9809300062
Download: ML20153G396 (76)
v • d • e


Text

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WESTINGHOUSE ELECTRIC CORPORATION PRESENTATION TO t

UNITED STATES NN '

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~' NUCLEAR REGULATORY COMMISSION c

E i 3 WESTINGHOUSE ROCKVILLE NUCLEAR LICENSING CENTER 1 APRIL 20,1993 O O3 i

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AP600 AP600 OVERVIEW APRIL 20,1993 AGENDA INTRODUCTION ANDREA STERDIS AP600 PASSIVE SYSTEMS TERRY SCHULZ AP600 PROBABILISTIC RISK ASSESSMENT CINDY HAAG 0930A

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AP600 ,

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f INTRODUCTION i

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i ANDREA STERDIS-  ;

ADVANCED PLANT SAFETY & LICENSING -

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

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i t ( , . I J_ . ) . - L '- IL 1- a AP600 1992 1993 1954 1995 1996 ISST 1998 1999 2988 2901 23e2 2003 2004 SSAR FDA DC e atton . --

Q V. Comipleted ler Firm Ree6'ement Procus Rice AV Ptemmed SeertI Compleuen FOAKE (( M CatWeelPeth Q NRC lueellidag Compe9tive /\ U Component l I Selection COL RFP Appi- CO U U U 2002 ESP PLANT Plant Excavetion pg g Ceaunft First Comaese CO tude msnmT 4

Reference:

First-Of-A-Kind Engineering AP600 Advanced Light Water Reactor Design i Proposal to Advanced Reactor Corporation

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DESIGN CERTIFICATION ISSUES I IMPACTING FOAKE  :

o ITAAC TESTING REQUIREMENTS

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l o SYSTEM AND EQUIPMENT DESIGN REQUIREMENTS t

i o PLANT LAYOUT o REGULATORY TREATMENT OF NONSAFETY SYSTEMS o EQUIPMENT SPECIFICATIONS o CONTROL ROOM DESIGN o EMERGENCY RESPONSE GUIDELINES i

t o SOURCE TERM o EQUIPMENT QUALIFICATION o PLANT LAYOUT I

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AP600 '

i AP600 PASSIVE SYSTEMS i

I T. L. SCHULZ, FELLOW ENGINEER SYSTEMS AND EQUIPMENT ENGINEERING l

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AP600 j i

AP600 PASSIVE SYSTEM AGENDA-t PASSIVE SYSTEM DESIGNS i

PASSIVE SYSTEMS DEFENSE-IN-DEPTH t

PASSIVE SYSTEMS CAPABILITIES DURING SHUTDOWN i PASSIVE SYSTEMS LONG TERM SHUTDOWN CAPABILITIES SAFETY RELATED ISOLATION OF NONSAFETY SYSTEMS

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, AP600 ~ SYSTEMS DESIGN g  :

i Greatly Simplify Systems to improve Safety, Cost, Construction, l Maintenance, & Operation Provide Simple Passive Safety Systems  ;

Use " natural" driving forces only ,

One-time alignment.of active valves No support systems after actuation Reduced operator dependency

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. Provide Non-Safety Systems j Redundant active equipment powered by nonsafety diesels j Minimize unnecessary use of passive safety systems Reduced risk to utility & public TLS - 4.19 H l i

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-1. 1; r t- Lu9 AP600 SAFETY SYSTEMS E ,

a Provide Passive Safety Systems.

Greatly. simplified construction, maintenance, operation, ISI / IST Mitigate design basis accidents without use of NNS systems NRC PRA goals w/o NNS system; EPRI 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; multiple failures  ;

Greatly reduced need for operator actions l.

a Safety Equipment Design Features Reliable / experience based equipment j 6

Improved inservice testing / inspection Reg Guide 1.26 Quality Group A, B, or C; Seismic I design  :

Availability controlled by Tech Spec with shutdown requirements Reliability Assurance Program Tier i description and ITAAC-TLS - 4c19J93 .

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  • f AP600 PASSIVE SAFETY FEATURES _ _ _

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f Passive Decay Heat Removal i Natural circulation HX connected to RCS l

- Passive Safety injection l N2 pressurized accumulators  !

Gravity drain core makeup tanks (RCS pressure) ]

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 walls for heat sink (MCR and.l&C rooms) 4 TLS - 4.19,93

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PASSIVE SAFETY SYSTEMS PCCVsf v A n A IRVST A

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4 M n I d n CMT CMT ACC ** ACC L J VESTINGHOUSC - 1/92

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. AP600 DECAY. HEAT REMOVAL E Startup Feedwater System Non-safety feedwater for normal ~ shutdowns and transients Two motor driven pumps feed all SGs '

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Water supplied from deaerating heater or CST Automatic start and flow control, auto load on NNS diesels

- Passive RHR Heat Exchanger Safety cooling when SFW is unavailable and non-LOCA accidents Two heat exchangers connected directly to RCS Forced flow with RCP; natural circ without RCP Automatic actuation; two fail-open valves PRHR HX located in IRWST, provides heat sink, boils in 2-3 hr-Passive containment cooling provides ultimate heat sink

- RCS Feed and Bleed Provides backup to SFW and PRHR HX for PRA events Fee-J from CMT/Accum/lRWST, bleed from ADS Automatic actuation of CMT on high RCS temp with low SG level rt.s - 4. a n

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- AP600 -

PASSIVE RHR HX PRESSURIZER n I I STEAM ,

IRWST GM.

PRHR PRHR A ty "x ' "x 2 s j l'

4TH DS I M s e ;J P HL RCP REACTOR CORE VESSEL WESTINGHOUSE - 3/92

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~l AP600 RCS MAKEUP- .

--- E CVS Makeup Pumps Non-safety makeup for normal plant operation Can accommodate 3/8" break without Si Two motor driven centrifuga! pumps Automatic start and connection to diesel  !

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- Core Makeup Tanks j Safety makeup to RCS when CVS unavailable or with larger leaks Two tanks provide makeup by gravity at any RCS pressure Automatic actuation by opening redundant air operated valves, fail l open, for each CMT '

Provides significant makeup before ADS act; 3 gpm leak / 40 hr

- PXS Tanks and ADS r Safety injection for LOCA Also PRA backup to CMT & CVS i Two CMT, two Accumulators and one IRWST provide makeup j Four stages ADS provide controlled depressurization of RCS TLS - 4.19.93 h

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AP600 PASSIVE SAFETY INJECTION A

ADS ".-

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-1 CORE MAKEUP l REFUEL TANK (1 OF 2) .

CAVITY g, PRESSURIZER ($# )

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LOOP IRWST COMPART.

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(1 OF 2) l [ hl CORE REACTOR VESSEL WESDNCHOUSE - 3/92 ,

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AP600 CONTAINMENT COOLING E

- Containment Fan Coolers Nonsafety heat removal during normal operation and transients-2 coolers, each with redundant fans Chilled water provides heat sink Automatic control and loading on NNS diesels

. Passive Containmemt Cooling System Safety heat removal when fan coolers are unavailable or during large energy releases Steel co..lainment shell cooled by air flow / water evaporation Water drains by gravity from elevated tank, air circulates by natural circulation Automatic actuation opens redundant air operated valves. fail open Other Containment Cooling Features -

Boiling of water sprayed on outside of containment vessel from fire protection pumps Natural circulation of air, without any water

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PASSIVE CONTAINMENT COOLING SYSTEM

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NOTES. (1) VEIRS DISTRIBUTE VATER FILM VESTINGOJSE - 9/90 i

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Passive Safety System Defense-in-Depth Capabilities

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AP600

l PASSIVE SYSTEM DID CAPABILITIES i

l Passive Safety Systems Provide Defense-in-Depth Capabilities

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i Some provided in original design; others provided in design changes incorporated to improve PRA I

More probable events have greater protection Supported by best estimate analysis N- - - - - - - - _ - - - - - - - - - - _ - - _ - - - _ - - - - _ _ _ - - - - - _ - - - - - - - _ - _ _ . - - _ _ _ - - - - - - - - - - - - - - _ - - _ _ _ _ - - _ - - _ - - - - _ - - - - - _ - - _ _ - - _ - _ - - - - - - _ - - _ - - - - - - - - _ - _

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j AP600 LEVELS OF DEFENSE I FUNCTION CURRENT PWR AP600 4

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 PORV - LARGER PZR i PROTECTIO'N '

- Hi PRES TRIP - Hi PRES TRIP j - PZR SAFETY VALVES - PZR SAFETY VALVES j RCS HEAT REMOVAL - MAIN FEEDWATER SYS - MAIN FEEDWATER SYS

- AUX FEEDWATER SYS - STARTUP FEEDWATER SYS

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

- MANUAL FEED / BLEED (ACCUM / NRHRS, ACC)

) HIGH PRESSURE - CVCS PUMPS - CVCS PUMPS i INJECTION - HHSl PUMPS -CMT i - ACCUM / IRWST (ADS)

- ACCUM / NRHRS (ADS) i LOW PRESSURE -ACCUM -ACCUM

, INJECTION - LHSI PUMPS - IRWST (ADS)

- NRHRS PUMPS

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LONG TERM RECIRC - LHSI PUMPS FEEDING - CONTAINMENT SUMP (ADS) 4 4

HHSl PUMPS - NRHRS PUMPS i CONTAINMENT HEAT - FAN COOLERS - FAN COOLERS

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REMOVAL CONT SPRAY PUMPS / HX - EXTERNAL AIR + WATER DRAIN

- EXTERNAL WATER FIRE SYSTEM j - EXTERNAL AIR ONLY COOLING i

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STARTUP FEEDWATER SiSTEM I

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NORMAL RESIDUAL HEAT REMOVAL SYSTEM T T IRWST Y -Q-- CONTAINMENT IRC SUMP cves M l i

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AUTO SFWS, CVCS SUCCESS

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._. AUTO PRHR HX, CMT, SAFETY CASE PCCS > SUCCESS NO ADS

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AU TO CM T, PARTIAL ADS >- SUCCESS MAN NRHRS INJECT ADS, NO FLOOD l

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.. AUTO CMT, FULL ADS, IRWST,PCCS > SUCCESS ADS, FLOOD i

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.. AUTO ACCUM, IRWST, > SUCCESS l 4 PCCS ADS, FLOOD '

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., CORE DAMAGE 7  : SUCCESS

__ WESTINGHOUSE PROPRIETARY CLASS 2 FAILURE n .- , ,-- . , . , -vr, . . , - , - . , . . , _ . - - -- _ , - . , _ . . , , ~ . - - _ , , _

Plant: AP600 Event: LOSS OFFSITE POWER at FULL POWER

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Actuation / Electrical Systems Non-Safety Safety Diverse System Order of Use PLS DC AC PMS(1) DC DAS HW 1 I

- o Reactor Shutdown , ,

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1. Control Rods - - -

A - - -  !

2. Cor.trc! Rods - - - - -

A -

3. Control Rods - - - - - -

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4. Ride Out (2) M Yes Yes - -

A M

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i o RCS Inventory Control l 4- 1. CVS - A Yes Yes - - - - i 2.CMT - - -

A - - -

  • 3.CMT - - - - -

A -

!_ 4.CMT - - - - - -

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5. CMT, RNS, part ADS M Yes Yes A Yes - -

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. 6. CMT, IRWST, full ADS - - -

A Yes - -

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

Yes -

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

Yes -

M

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4 o RCS Heat Removal

! 1.SFW A Yes Yes - - - -

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

Yes -

M j o Containment Cooling

1. Fan Coolers A Yes Yes - - - -

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2. CV extemal air, water drain . - - -

A - - -

3

3. CV external air, water drain - - - - -

A -

4. CV external air, water drain - - - - - -

M

5. CV external water fire sys only M Yes Yes - - - -
6. CV external air only

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Notes:

,,. 1) Safety components have safety related MCB manual controls via both individual soft control switches and ..r

, dedicated system level switches.

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2) Reactor is shut down by negative moderator temperature coefficient as the coolant heats up. Requires automatic j RCS pressure relief, turbine trip, and PRHR HX actuation. Also requires manual CMT or CVS boration.

Westinghouse Proprietary Class 2 10/27/92 1

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. -- NORMAL SHUTDOWN > SUCCESS l NO ADS l

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, _. AUTO CMT, PRHR HX, PARBAL ADS > SUCCESS MAN NRHRS INJECT ADS, NO FLOOD AUTO CMT, PRHR HX, SAFETY CASE FULL ADS, IRWST, > SUCCESS PCCS ADS, FLOOD

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_, MAN PART ADS AUTO ACCUM > SUCCESS  ;

MAN NRHRS INJECT ADS, NO FLOOD 1

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MAN FULL ADS

. AUTO ACCUM, IRWST, & SUCCESS PCCS ADS, FLOOD l

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- SUCCESS a i

- f FAILURE
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MANUAL SG ISOL, SUCCESS

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- AUTO CMT, PRHR HX, SAFETY CASE CVCS / SFWS ISOL  :- SUCCESS NO ADS I

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' AU TO CM T, PARTIAL ADS  :- SUCCESS MAN NRHRS INJECT ADS, NO FLOOD

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, IRWST,PCCS = SUCCESS ADS, FLOOD

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MAN FULL ADS

-. AUTO ACCUM, IRWST, *- SUCCE55 PCCS ADS, FLOOD l

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- CORE DAMAGE

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.- Westinghouse Proprietary FAILURE GW GSY 001, REV. 0

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AP600

' ADS DEFENSE-IN-DEPTH

= DBA Performance Conservative decay heat, line resistances, pressure drop calc, containment pressure Successful IRWST gravity injection achieved with single failure Limiting failure is one 4th stage valve or one battery train (causes failure of one 1st & 3rd stage valves)

. i e t  ; r; . i 1 . : 1 < t u l

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 toietate common mode failure 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

= ADS sizing basis provides substantial margin and failure tolerance

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AP600 1

t Passive Safety System Capabilities During Shutdowns

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<.  :. 1 1 1 . 1 m. .._ w AP600 PASSIVE CAPABILITY DURING SHUTDOWNS  !

Passive Safety Functions Provided During All Shutdown Modes  ;

i

. Hot Shutdown / Hot Standby / Cold Shutdown Same As At

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

. Cold Shutdown Mid-Loop PRHR HX ineffective (RCS open)

CMT / accum unnecessary Tech. Spec require:

Containment integrity ADS stages 1,2,3 open IRWST MOV available j

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

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LOSS OF OFFSITE POWER l -.

(HOT / COLD SHUTDOWN)("

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4 AUTO NRHRS COOL, l CVCS SUCCESS l 3

NO AOS l j  ?

MANUAL PRHR HX AUTO CMT, PCCS > SUCCESS 4

NO ADS i

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1 9

i AUTO CMT, PART ADS l SAFETY CASE IRWST,PCCS SUCCESS

ADS, FLOOD i

2.

t, MAN PART ADS AUTO IRWST, PCCS  : SUCCESS ADS, FLOOO I

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" CORE DAMAGE i

NOTE (1) RCS PRESSURE BOUNDARY 15 INTACT IN THIS CASE 4

1

- SUCCESS m nne 5

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L~ AP600 -

LOSS OF OFFS TE POWER l-i (M D-LOOP) I i

l-3 LOOP 1

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{ AUTO NRHRS COOL i M AN CVCS p- SUCCESS l~ NO ADS i  :

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! MAN NRHRS INJECT i j (1)

  • SUCCESS I
ADS, NO FLOOD l MAN IRWST (1) i SAFETY CASE AUTO PCCS > SUCCESS j ADS, FLOOD u l i~ l j_ f

,! AUTO IRWST (1)

AUTO PCCS SUCCESS j_ ADS, FLOOD

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!._ CORE DAMAGE i

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) NOTE (1) ADS STAGES 1,2,3 W1LL BE OPEN DURING MID-LOOP

_ success i

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WESTINGHOUSE PROPRIETARY CLASS 2 FAILURE

a .

-AP600 -

LOSS OF OFFSITE POWER (EFUEL NG)

LOOP i

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9 AUTO NRHRS COOL

> SUCCESS NO ADS s

MAN SFPCS COOL

  • SUCCESS NO ADS i

REFUELING CAVITY SAFETY CASE (1)  : SUCCESS NO ADS l

__ t CORE DAMAGE d

OTE (1) EITHER CLOSE CONTAINMENT R PROVIDE ADDITIONAL MAKEUP AFTER 72 88t

SUCCESS l

6 FAILURE

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AP600  !.

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i Passive Safety System Long Term Shutdown Capabilities I

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% AJ AP600 POST 72 HOUR ACTIONS

  • Long Term Passive Safety System Operation 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 -

hours Limited offsite support after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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

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AP600 i

l POST 72 HOUR ACTIONS '

  • Safety System Extended Support Actions i

Provide makeup water into containment j

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Only needed after one month assuming DBA containment leakage t

I Provide makeup water to the passive containment cooling water storage tank Air cooling alone maintains containment pressure below design pr ssure Provide electrical power to supply the post-accident and spent fuel pit monitoring instrumentation i Provide electrical power to the hydrogen recombiners Only needed for events where containment hydrogen buildup is a concern

I 1 I

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AP600 POST 72 HOUR ACTIONS

=

Safety System Extended Support Actions (continued)  :

Provide breathable, compressed air for the control room air supply and pressurization system Only required in case of serious core damage and containment leakage Provide control room cooling and air recirculation Only required in hot weather conditions Provide ventilation cooling to post-accident monitoring equipment rooms 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

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i AP600 '

i Safety Related Isolation Of Nonsafety Systems

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AP600 J NONSAFETY SYSTEM ISOLATION FUNCTIONS

  • Nonsafety Systems' Provide Some Safety Related Isolation Functions ,

RCS pressure boundary isolation  !

Containment isolation Other isolation functions provided to mitigate DBA's j

  • These isolation Capabilities Are Fully Safety Related Single failure capability i Reg Guide 1.26 quality group A, B, or C Seismic I Tech Spec controls  !

Described in SSAR and ITAAC ,

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AP600 l NONSAFETY SYSTEM ISOLATION FUNCTIONS ,

= Example: CVS Functions

  • CVS Functions

-Safety Functions RCS pressure boundary isolation Containment penetration isolation Boron dilution accident tennination

  • Excessive makeup isolation DID functions RCS makeup for leaks RCS pressure reduction

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AFETY RELATED _

130 L ATisN AP600 - CHEMICAL AND VOLUME CONTROL SYSTEM. -

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CCWS RCS ~U ~

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. 4 i 1 i . : t AP600 NON-SAFETY SYSTEMS Provide Non-Safety: 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 EPRI safety goals i Non-Safety Systems Design Features  ;

Redundancy for more probable failures, automatic actuation  :

Power from offsite / onsite (nonsafety diesels) sourses Separated from safety systems Non-Safety Equipment Design Features Reliable / experienced based equipment  ;

Reg Guide 1.26 Quality Group D; limited hazzard protection Short term availability by plant procedures w/o shutdown requirements Long term availability by Reliability Assurance Program Less detailed Tier i description and ITAAC TLS - 4.19.93

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1 PROBABILISTIC RISK ASSESSMENT C. L. HAAG, SENIOR ENG!NEER i

RISK MANAGEMENT AND i

OPERATIONS !MPROVEMENT 0810A

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AP600 PRA -

AGENDA Passive System Reliability 9

Initiating Event Evaluation

  • Sensitivity Studies of Nonsafety Systems PRA Insights and System importance t I

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Input to calculate system reliability 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

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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

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EXAMPLE PRHR SYSTEM INPUT -

= lnitiator Transient event

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

i

  • Support Syste as t Actuated ':w Protection and Monitoring System Diverse Actuation System

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FAILURE CONSIDERATIONS IN A PRHR '

FAULT TREE '

  • 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

.  ;  ; i

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OTHER FAILURE CONSIDERATIONS OF PRHR .

i 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 i 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) 1

  • Appropriately not modeled in fault tree

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AP600 PRA SYSTEM RELIABILITY DATA

  • Primary Source

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ALWR Utility Requirements Document (Volume 111)

= Secondary Sources NUREG/CR-2728 (IREP)

NUREG/CR-2815 (NREP)

NUREG/CR-4550 t WASH-1400 t IEEE Std 500 Westinghouse i

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AP600 PRA PRHR SYSTEM RELIABILITY-i 1

. Equipment in PRHR system similar in duty and design to  !

operating-plants which justifies the use of historical equipment-reliabilities.

Single AOV fail to open 1.1E-3 Both AOVs fail to open 1.2E 4 Common cause failure of AOVs 6.2E-5 l

=

Calculated PRHR system reliability Unavailability calculated to be 7.7E-5 -

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INITIATING: EVENT EVALUATION

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Initiating event frequencies for:: AP600 are based on historical-  !

data and AP600-specific analysis

Detailed 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.  !

=

Loss of.Offsite Power -

y Frequency based on ALWR URD data  !

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Loss of Coolant Accidents l:

LOCAs are AP600-specific pipe break analysis j i ,

= Support System initiators l 1

Based on AP600-specific fault tree analysis. Includes loss of

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1 : 6 1.' .a AP600 PRA INITIATING EVENT FREQUENCY DEPENDENCY VS NSS/DID SYSTEMS Initiatine Event NSS System DID System Transients:

Turbine trip x Loss of feedwater flow x <

Secondary to primary side power mismatch x

  • Core power excursion x Spurious S-signal x Loss of CCW x inss of SW x Loss of compressed air x Main steamline break downstmam of MSIV Main steamline break upstream of MSIV Main steam line safety valve stuck open LOOP x LOCAs:

Large LOCA Medium LOCA CMT line bmak i SI line break Small LOCA Very small LOCA x PRHR tube rupture SGTR Vessel rupture ATWS x

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  • Very Small LOCA Ruptures in pipes less than 3/4 inch diameter Pressurizer level instrumentation lines Miscellaneous primary system lines < 3/4 inch Frequency calculation equation Pipe rupture failure rate x number of pipe sections Initiating event frequency is 5.5E-04 /yr

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AP600 PRA LEAKAGE EVENTS NRC/Brookhaven reported 39 leakage events (1 gpm - 100 'gpm) j

=

' Westinghouse reviewed events and determined 5 at power events apply to AP600

)

NRC # LER # Descriotion -Leak (gpm) 17 323-89006 Pzr SV seal 10.0  ;

20 339-91011 RHR valve packing 10.0 18 323-91004 1.9 l 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  !

l For leaks 1 - 100-gpm, procecd with. orderly shutdown i I

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N RCS CVCS LEAKAGE MAKEUP RECOVER (0-Igpm) CVCS ,

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Continue Operation RCS Leak Event Tree RCS Leak initiating Event

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Currently evaluating RCS leakage events with respect to RTNSS Anticipate core damage frequency increase will be negligible

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I SENSITIVITY STUDIES OF NONSAFETY SYSTEMS I

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Estimated Core Damage Frequency At Power Shutdown Total ,

Base Case 3.3E-7 /yr 8.9E-8 /yr 4.2E-7 /yr Sensitivity Case 2.6E-6 /yr 5.4E-7 /yr 3.1 E-6 /yr NRC Goal 1.0E-4 /yr i

Note: Sensitivity case removes credit for CVS, SFW, RNS, offsite power and DGs following an initiating event

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INITIATING EVENT CONTRIBUTION TO CORE DAMAGE (AT POWER)

Initiating Event Base Case Senskivity Case CDF  % of Total CDF  % of Total Transients / LOOP 7.5E-8 22.4 6.2E-7 24.0 Small LOCA 2.3E-8 6.9 13E-7 4.8 Very small LOCA 1.2E-8 3.6 13E-7 5.1 PRHR tube rupture 4.2E-8 12.6 1.4E-6 53.1 Medium LOCA I.2E-8 3.6 13E-7 5.1 2

Safety injection line break 73E-8 21.9 7.7E-8 3.0 CMT line break 2.7E-9 0.8 3.0E-8 1.2 Large LOCA 1.6E-8 4.8 1.6E-8 0.6 SG tube rupture 2.6E-9 0.8 2.7E 9 0.1 A7WS loss of feedwater 4.5E-8 13.6 4.9E-8 1.9 w/o scram Vessel rupture 3.0E-8 9.0 3.0E-8 1.2 Total 33E-7 2.6E-6 3

+

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INITIATING EVENT PERCENT CONTRIBUTION TO CORE DAMAGE (AT POWER)

Initiating Event Base Case Sensitivity Case Transients / LOOP 22.4 24.0 Small LOCA 6.9 4.8 Very small LOCA 3.6 5.1 PRHR tube rupture 12.6 53.1 Medium LOCA 3.6 5.1 ,

Safety injecuon line treait 21.9 3.0 t

CMTline treak 0.8 1.2 j Large LOCA 4.8 I 0.6 SG tube rupture 0.8 0.1 ATWS loss of feedwater w/o 13.6 1.9 scram Vessel rupture 9.0 1.2 n_

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AP600 NON-SAFETY SYSTEM SENSITIVITY CASE -

Estimated Release Frequency .

At Power Shutdown Total Base Case 2E-8 /yr 1 E-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

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AP600 PRA PRA INSIGHTS VERSUS

-!MPORTANT ANALYSIS

'i dentified insights in AP600 PRA report (Chapter 17)

Insights are changes made to the design, operation, or PRA success criteria Insights are not intended to be a listing of the risk important- .;

features of the plant e importance Analysis Used in response to some RAls i

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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 ,

I

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c ta AP600 PRA PRA SYSTEM IMPORTANCE  :

Startup Feedwater, Normal RHR, and DGs

  • Negligible impact on core damage and release frequencies '

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

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t 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 IST 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 j

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" Electric Corporation Pittskten Pemsylvarta 15230-0355 1 DCP/NRCl409 NSD-NRC-98 5753 l

, Docket No.52-003

. , , August 13,1998

- Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC ;20555 4

ATTENTION: T.R. QUAY l

SUBJECT:

RESPONSE TO NRC LSTTERS CONCERNING REQUEST FOR WITHHOLDING TNFORMATION =

Reference:

1. Letter, Sebrosky to McIntyre, " Request for withholding proprietary information for Westingtouse letters dated December 14,1992, and December 17,1992," dated July 10,1998.
2. Letter, Huffman to McIntyre, " Request for withholdmg infonnation from public

..^ disclosure of Westinghouse AP600 design letters of Decembes 15,1992," dated July 14,1998.

3. Letter, Sebrosky to McIntyre, " Request for withholding inform 6 tion 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 I provided the NRC assessment of the Westinghouse claim that proprietary infonnation was i provid*% A letter dated December 14,1992, that provided the NRC with copies of presentation m9tial from a management meeting held December 14,1992, discussing the AP600 testing program.

. fhe NRC has no record of a nonproprietary version of the slides being provided. At the time this 4* L presentation was made, the information was proprietary since that description of the AP600. testing

- progr$un had commercial value to Westinghouse. At this time, almost six years later, this information

- E. does not have commercial value and is t.o 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 nei .,c M. O $gQ

~

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DCP/NRCl409 NSD-NRC-98 5753 -2 August 13, 1998 l testing program. The NRC has no record of a nonproprietary version of the slides being provided. At 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, almost six years later, this information does not have commercial 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 proprictary and that a nonproprietary versloet was not provided. At the J time this subject was being discussed with the NRC technical staff, the information was considered to be proprietary by Westinghouse since it contained information that had commercial value to 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 provided in a letter dated February 24,1993, that contained presentation materials from the Febmany 24,1993, Westinghouse /NRC AP600 senior management meeting. 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 staffindicated the material was used by the staff in the development of the AP600 draft safety evaluation report and-m 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 V for clarification only, not part 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 be proprietary by Westinghouse since it contained information that had commercial value to Westinghouse. If this presentation material was indeed used by the staffin development of the AP600 draft final safety 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.

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

provided in a letter dated April 19,1993, 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) 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 staff in the development of the AP600 draft safety evaluation report and therefore should remain on the docket.

Our 1995 request,' Reference 4, ir.dicated that the material provided in the Westinghouse letter of

- April 19,1993, was presentation material that was intended for clarification only, not part of the

-formal review material and requested that the material be retumed to Westinghouse. At the time this subject was being discussed with the NRC technical staff, the 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 development of the AP600 draft final safety 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.

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

This response addresses the proprietary issues delineated in the references.

AY Bnan A. McIntyre Manager

/~ **'

Advanced Plant Safety and Licensing

., jml ec; L W. Roc - NRC/NRR/DRPM J. M. Sebrosky - NRC/NRR/DRPM W. C. Huffman - NRC/NRR/DRPM H. A Sepp - Westinghouse i

I l

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