ML20035H059
| ML20035H059 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 04/27/1993 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9305030047 | |
| Download: ML20035H059 (82) | |
Text
,
Docket No.52-003 April 27, 1993 l
i APPLICANT: Westinghouse Electric Corporation PROJECT:
Westinghouse AP600 Design l
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS THE DESIGN OF THE AP600 i
i On April 20, 1993, representatives of the Nuclear Regulatory Commission, the t
Department of Energy, and the Westinghouse Electric Corporation met to discuss the systems designs and unusual features of the AP600. is a list
[
of attendees. is a copy of the non-proprietary slides presented by Westinghouse. The proprietary version of the presentation was submitted by Westinghouse in its letter dated April 19, 1993.
Westinghouse opened the meeting with a brief discussion of the status of the review of the AP600 application. Then the applicant discussed the design of the passive systems, addressing defense-in-depth concerns and the capabilities of these systems during short and long term shutdowns. Westinghouse also addressed the isolation functions of the nonsafety systems.
t During the applicant's final presentation, it described the results of the probabilistic risk assessment (PRA) performed on the AP600 design, discussing i
passive system reliability, initiating events, sensitivity studies regarding nonsafety systems, and insights drawn from the PRA on system importance.
(Original signed by) j Thomas J. Kenyon, Project Manager Standardization Project Directorate t
Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
I 1.
List of Attendees 2.
Non-proprietary Slides cc w/ enclosures:
See next page DISTRIBUTION w/ enclosures:
Docket File PDST R/F TKenyon PShea PDR BPerch, 8H7 i
t DISTRIBUTION w/o enclosures:
TMurley/FMiraglia WRussell, 12G18 DCrutchfield RBorchardt GBagchi, 7H15 TEssig EJordan, MNBB3701 ACRS (11)
RHasselberg JMoore, 15B18 GGrant, 17G21 EFox, 9H15 i
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DATE: 042Q d4' '/93 04/ /93 04/t1/93 0
OFFICIAL RECORD COPY:
DOCUMENT NAME: APR_20. SUM 300113 9305030047 930427
-***4 4
i
e Docket No.52-003 Westinghouse Electric Corporation 4
cc:
Mr. Nicholas J. Liparulo Nuclear Safety and Regulatory Analysis Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. B. A. McIntyre Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Box 355 Pittsburgh, Pennsylvania 15230 Mr. John C. Butler Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Box 355 Pittsburgh, Pennsylvania 15230 l
Mr. M. D. Beaumont Nuclear and Advanced Technology Divis' ion Westinghouse Electric Corporation One Montrose Metro 11921 Rockville Pike Suite 350 Rockville, Maryland 20852 Mr. Sterling Franks l
U. S. Department of Energy NE-42 Washington, D.C.
20585 Mr. S. M. Modro EG&G Idaho Inc.
Post Office Box 1625 i
Idaho Falls, Idaho 83415 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Room 8002 Washington, D.C.
20503 4
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MEETING ATTENDEES l
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AP600 SYSTEMS DESIGN OVERVIEW April 20, 1993 1
Name Oraanization l
Andrea Sterdis Westinghouse Terry Schulz Westinghouse Cynthia Haag Westinghouse Jack Wheeler DOE Goutam Bagchi NRC/NRR/ECGB Edwin Fox NRC/NRR/PEPB William Russell NRC/NRR/ADT Thomas Kenyon NRC/NRR/PDST Bill Borchardt NRC/NRR/PDST Mark Beaumont Westinghouse /Rockville i'
Tom Essig NRC/NRR/PDST Sterling Franks DOE j
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o WESTINGHOUSE ELECTRIC CORPORATION PRESENTATION TO UNITED STATES 1
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WESTINGHOUSE ROCKVILLE NUCLEAR LICENSING CENTER k
i APRIL 20,1993
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PASSIVE SYSTEM DESIGNS PASSIVE SYSTEMS DEFENSE-IN-DEPTH PASSIVE SYSTEMS CAPABILITIES DURING SHUTDOWN PASSIVE SYSTEMS LONG TERM SHUTDOWN CAPABILITIES SAFETY RELATED ISOLATION OF NONSAFETY SYSTEMS 0930A i
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1 Greatly Simplify Systems to improve Safety, Cost, Construction, 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 Provide Non-Safety Systems Redundant active equipment powered by nonsafety diesels Minimize unnecessary use of passive safety systems Reduced risk to utility & public
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AP600 SAFETY SYSTEMS lRH 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 Safety Equipment Design Features Reliable / experience based equipment 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
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Passive Decay Heat Removal Natural circulation HX connected to RCS Passive Safety injection 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 I&C rooms) o ii.
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- 2) Reactor is shut down by negative moderator temperature coefficient as the coolant beats up. Requires automat'c-RCS pressure relief, turbine trip, and PRHR HX actuation. Also requires manual CMT or CVS boration.
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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 failure of all stage 1/2/3 valves or all stage 4 valves Successful RNS pump injection achieved with opening of any i
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PASSIVE CAPABILITY DURING SHUTDOWNS Passive Safety Functions Provided During All Shutdown Modes 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 Refueling Shutdown Refueling cavity provides >72 hours with equipment hatch open Equipment hatch can be closed without AC power d
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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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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.
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AP600 POST 72 HOUR ACTIONS 4
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 storage tank Air cooling alone maintains containment pressure below design pressure Provide electrical power to supply the post-accident and spent fuel pit l monitoring instrumentation Provide electrical power to the hydrogen recombiners Only needed for events where containment hydrogen buildup is a concern
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AP600 POST 72 HOUR ACTIONS Safety System Extended Support Actions (continued)
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' 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 re. quired in hot weather conditions l
Provide makeup water to the spent fuel pit 7 days at BOL,21 days at EOL 72 hr for worst case. emergency core unload l
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AP600 Safety Related Isolation Of Nonsafety Systems L
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AP600 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 These isolation Capabilities Are Fully Safety Related Single failure capability Reg Guide 1.26 quality group A, B, or C Seismic I Tech Spec controls Described in SSAR and ITAAC
AP600 O
NONSAFETY SYSTEM ISOLATION FUNCTIONS Example: CVS Functions CVS Functions Safety Functions RCS pressure boundary isolation Containment penetration isolation Boron dilution accident termination Excessive makeup isolation DID functions RCS makeup for leaks RCS pressure reduction i
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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 Non-Safety Systems Design Features Redundancy for more probable failures, automatic actuation Power from offsite / onsite (nonsafety diesels) sourses Separated from safety systems
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C. L. HAAG, SENIOR ENGINEER RISK MANAGEMENT AND l
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AP600 PRA AGENDA Passive System Reliability
. initiating Event Evaluation Sensitivity Studies of Nonsafety Systems PRA Insights and System importance
AP600 PRA PASSIVE SYSTEM REllABILITY
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AP600LPRA PASSIVE SYSTEM MODELING 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 t
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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 4
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EXAMPLE PRHR SYSTEM INPUT Initiator
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Transient event Success Criteria
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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 Mission Time 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l
Support Systems a
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AP600 PRA FAILURE CONSIDERATIONS IN A PRHR FAULT TREE Equipment Failures AOVs fail to open IRWST ruptures Plugging of flow venturi Instrumentation and control equipment Test / Maintenance Consideration a
AOVs tested every 3 months System available during test Component maintenance unavailability l
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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
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<e Alarmed in control room Venting-performed after maintenance / inspection H2 in RCS is. saturated.at 30 psig so it can not come out of i
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)
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AP600 PRA SYSTEM RELIABILITY DATA Primary Source ALWR Utility Requirements Document (Volume lil)
Secondary Sources NUREG/CR-2728 (IREP)
NUREG/CR-2815 (NREP)
NUREG/CR-4550 WASH-1400 IEEE Std 500 Westinghouse
AP6_0.0_PRA L
i PRHR SYSTEM RELIABILITY Equipment in PRHR system similar in duty and design to
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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-6 Common cause failure of AOVs 6.2E-5 Calculated PRHR system reliability
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Unavailability calculated to be 7.7E-5 l
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INITIATING-EVENT EVALUATION Initiating event frequencies for AP600 are based on historical data and AP600-specific analysis Transients 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 Frequency based on ALWR URD data l
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
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AP600 PRA INITIATING EVENT FREQUENCY DEPENDENCY VS NSS/DID SYSTEMS Initiatine Event NSS System DID System Transients:
Loss of feedwater flow x
Secondary to primary side power mismatch x
Core power excursion x
Spurious S-signal x
x Loss 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
Large LOCA Medium LOCA CMT line bitak Si line break Small LOCA Very small LOCA PRIIR tube rupture x
SGTR Vessel rupture NIWS x
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AP600 PRA EXAMPLE LOCA INITIATING EVENT FREQUENCY CALCULAT!ON 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)
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 4
For leaks < 1 gpm, below Tech Spec limit, continue plant operation For leaks 1 - 100 gpm, proceed with orderly shutdown a
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AP600 PRA RCS CVCS RECOVER LEAKAGE MAKEUP CVCS (0-Igpm)
(tYR)
Continue Operation Cor finue Operofion RCS teak Event f ree RCS Leak initiating Event
AP600 PRA RCS CVCS RECOVER LEAKAGE MAKEUP CVCS (1 - 100gpm)
(24HR)
Manual Shutdown Manual Shutdown RCS leak Event Tree RCS Leak initiating Evenf
AP600 PRA LEAK CMT RECOVER W/O CVCS CVCS Manual Shutdown Automolic ADS (small LOCA event)
Manual ADS (smoli ; OCA event)
RCS Leak ' Event Tree I'
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RCS LEAK EVALUATION Currently evaluating RCS leakage events with respect to RTNSS Anticipate core damage frequency increase will be negligible importance of CVCS should not change a
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INITIATING EVENT CONTRIBUTION TO CORE DAMAGE (AT POWER)
In6tisting Event Mme Case Semitivity Caw CIW
% of Total' CIW
% of Total Transienu/ LOOP 7.5E-8 22.4 6.2I!.7 24.0 t
Small LOCA 2.3E-8 6.9 1.3E-7 4.8 Very small LOCA 1.2E-8 3.h 1.3E-7 5.1 PRilR tube ruptwe 4.2E-8 12.6 1.4E-6 53.1 Medium LOCA 1.2E-8 3.6 1.3E-7 5.1 Safety injection line tweak 7.3E-8 21.9 7.7E-8 3.0 CMT line tweak 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 2.7t?M 0.1.
A7WS loss of feedwater 4.5E-8 13.6 4.9E-8 1.9 w/o scram Vessel rupsure 3.0E-8 9.0 3.0E.8 1.2 Total 3.3E-7 2.6E.6 u.........
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INmATING EVENT PERCENT CONTRIBUTION TO CORE DAMAGE (AT POWER)
Indtleting Event Rase Caw SensMivity Caw Transients / LOOP 22.4 24.0 Small LOCA 6.9 4.8 Very small LOCA 3.6 5.1 PRilR tube rupeure 12.6 53.1 Medium LOCA 3.6 5.1 Safety injection line tweak 21.9 3.0 CMT Ime tweak 0.8 1.2 Large LOCA 4.8 0.6 SG tube rupture 0.8 0.1 A'IWS loss of feedwater w/o 13.6 I.9 scram Vessel rupture 9.0 1.2 l
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AP600 PRA 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 m
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PRA INSIGHTS AND SYSTEM IMPORTA-NCE
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AP600 PRA PRA INSIGHTS VERSUS IMPORTANT ANALYSIS PRA Insights Identified 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 Importance Analysis Used in response to some RAls
..2.....
- AP600 PRA PRA SYSTEM IMPORTANCE RAI 720.13 - Requested system level importance Results of RAl 720.13:
Gravity injection
. Largest increase in core damage and release frequencies System required for Si line break and.large LOCAs t
Passive RHR Second largest increase in core damage and release frequencies P
Accumulators, CMTs, ADS Stages 1-3, ADS Stage 4
=
Small increase in frequencies due to system redundancy s
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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 LCCA events with a large, pre-existing opening in containment
,+
AP600 PRA AP600 PRA INSIGHTS 4
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
=
t NRHR valves.made remote 4th stage ADS. valves diverse 4
Expanded diverse l&C capabilities Added redundant IRWST injection check valves Added redundant / diverse IRWST recirc valves Made CMT check valves normally open
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