ML20244E364
| ML20244E364 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 06/12/1989 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| PROJECT-676A NUDOCS 8906200295 | |
| Download: ML20244E364 (148) | |
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j UNITED STATES f.f J.*,
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NUCLEAR REGULATORY. COMMISSION
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- June-12, 1989 g.,;.
v-N Docket No.50-60I
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sTAPPLICANT:.
Westinghouse Electric Corporation FACILITIES:
RESAR.SP/90 AP-600~
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SUBJECT:
SUMMARY
OF MEETING TO DISCUSS'THE WESTINGHOUSE STANDARD PLANT PROGRAMS-
/' L0n May 8,:1989,' representatives'of the NRC~and Westinghouse. met at the; Westinghouse Energy complex in Monroeville, Pennsylvania' to discuss the vendorf s standard '
plant programs. ' Enclosure leis' a list of. attendees. is the agenda:
' followed during' the meeting.
Enclosure'3'is a copy of the handouts provided
'by Westinghouse..
~ Westinghouse opened'the' meeting with an overview of its strategic plans.1 The applicant indicated their_ primary objective in the area of standardization. is to obtain the-Preliminary Design Approval (PDA) for RESAR SP/90 as quickly:as possible. After the PDA.is: issued,-the final design of the plant would proceed!
pending:a final design project. Westinghouse indicated that they would not be in aiposition to submit-the Final Design Approval and Design Certification
,(FDA/DC). application until the end of 1990 (best case). The staff indicated
'that the remainder of the review involved resolution of severe accident issues-and USIs/GSIs, the licensing; issues raised in the staff's draft SERs, and
- completion of review: meetings with the ACRS. The staff expects to complete
~
the PDA review in DecemberL1989.
Westinghouse' then discussed its February 22, 1989 request for the staff to perform'-an:early safety review of the conceptual design of the AP-600
- The l applicant stated they were looking for an; indication from the staff.if the
-overa11' approach of the design ~is consistent with NRC. regulations and policies e
Land where major open issues are likely to appear. Westinghouse agreed to
-provide the staff with a list:of potential policy and technical issues.
Dr. Murley stated that the staff will attempt to complete a policy review of this' nature by September 1989.
During the meeting, the staff informed Westinghouse that the staff would be reviewing' severe accident issues on a design-specific basis, using the same approach:that the staff used for the GE's ABWR.
-CONTACT:
T. Kenyon, NRR/DRSP
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8906200295 890612
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- both the RESAR SP/90 and the AP-600 designs. A detailed summary of these discussions can be found in Enclosure _4 of this report..
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Thomas J. Kenyon, Acting Project Manager Standardization ~and Non-Power Reactor Project. Directorate Division of Reactor Projects - III, IV, V 'and Special Projects cc:. See next page
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' -regarding the major licensing and severe-accident issues as they applied to both the RESAR SP/90 and the AP-600Ldesigns. A detailed summary of these l
' discussions can be found in Enclosure 4 of this report.
1 Th s J. Ke
, Acting Project Manager Standardization'and Non-Power Reactor Project Directorate Division of Reactor Projects - III, IV, V and Special Projects cc: See next page
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. June 12, 1989
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regarding the major licensing and severe accident issues-as they applied to
.both the RESARgSP/90 and the AP-600 designs. A detailed summary of these
' discussions can be found in Enclosure 4 of-this report.
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Thomas J. Kenyon, Acting Project Manager 1
Standardization and Non-Power Reactor Project Directorate i
Division of Reactor Projects - III, IV, Y and Special' Projects
.cc: See next page
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..,.7 ENCLOSURE,1 MEETING ATTENDEES 3
WESTINGHOUSE STANDARD PLANT PROGRAMS-MAY 8, 1989
.i i
. ATTENDEE ORGANIZATION DEPARTMENT-M. H. Shannon
.W Nuclear Safety.
R. P.-Vijuk-R ENGRG R. A. Bruce R
ENGRG E.: H. Weiss R
Marketing T. J. Kenyon NRC PDSNP' Ashok Thadani NRC NRR
. Tom Murley NRC NRR Les Rubenstein NRC NRR Jim Richardson NRC NRR-Theo Van deVenne W
System Eng.
Mark Beaumont 9 (Rockville)
Nuclear Safety Bill Johnson R
Nuclear Safety Tom Johnson R
System Eng.
Brian McIntre Westinghouse Nuclear Safety Bob Lutz W
Nuclear. Safety Dan Risher-U Nuclear Safety L. E. Hochester R
Nuclear Safety
.M. Y. Young U
Nuclear Safety L. E. Conway.
R ENGR.
M. M. Corletti R
ENGR.
T. L. Schulz-
'W ENGR.
S. M. Stahl R
Nuclear Safety.
Jim Scobel 9
Nuclear Safety G. R. Andre' R
Nuclear Safety
.C. M. Vertes R
Nuclear Safety A. F. Gagnor R
Nuclear Safety O
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Page 1
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Ag:nda_fer W/NRC Meeting May 8,;1989
~
~ Westinghouse Standard Plant Programs
.' Monday. May 8. 1989: (8:00 AM - 10:30 AM)
Introduction Bill Johnson-10 min.
r.
Overview of M Strategic Plans Brian McIntyre 30 min.
.o SP/90 E., -
- o. AP600:
SP/90 Licensing Issues Mike Shannon 40' min.
o NRC Additional Review Subjects '
o-Severe' Accident Issues o Schedule for SP/90, PDA o-New Industry Standards SP/90_ Design Program
. T. van de Venne 70 min, o Current Licensing Issues
- Anticipated Transients Without Scram p
4
.- Station Blackout
- Intersystem LOCA
- Source Term
- Probablistic Risk Assessment
- Containment
- Conditional Containment Failure o Probabilistic Risk Assessment o EPRI ALWR Requirements Document o On-going Japanese Activities 4
9
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Page 2 l'
Agenda-fGr,W/NRC Meeting May 8, 1989 Westinghouse Standard Plant Programs Monday May 8.'1989: (10:30 AM - 4:45 PM)
Advanced Analysis Methodology 60 min. total
~
o Automated Technical Specifications Ken Vavrek o TREAT /MAAP Code o SPNOVA/ BEACON / LOFT 5 Mike Young o Best Estimate LOCA Lunch (11:30 - 12:00)
AP-600 Safety Review o AP600 Program Ove'rview in Relationship Ron Vijuk 20 min.
to EPRI Requirements Document o Passive Safety Systems Design Description Terry Schulz 70 min.
Analysis Risk Assessment / Severe Accident Testing
- Larry Conway. 20 min.
o NRC Regulatory Baseline Mike Shannon 20 min.
o AP600 Scale Model Tom Johnson 30 min.
Summary Discussion all 25 min.
Tour of Test Facilities at R&D (3:15 - 4:45) m
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EhCLO5URE 3 WESTINGHOUSE PRESENTATIONS MAY 8, 1989 e
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6
w WESTINGH0VSE STRATEG C PLANS FOR
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DESIGN CERTIFICATION PROGRAMS IN THE 1990's BRIAN A. MclNTYRE, WANAGER AP600~ SAFETY AND Ll CENSING cert 1-11
~
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OVERV EW OF W STRATEG C P_ ass e
W POSITION ON 10CFR PART 52 "RULEMAKING ON EARLY SITE PERM TS, DESIGN CERTIFl CATION AND COMBINED
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LICENSES" SP/90. DESIGN CERTIFICATION PROGRAM e
e AP600 DESIGN CERTIFICATION PROGRAM l
eart211
10 CFR PART 52 ISSUE OPPORTUNITY FOR PRE 0PERATl0NAL HEARING CONCERNS ANY PREOPERATIONAL HEARING CAN DELAY PLANT OPERATION AFTER CONSTRUCTION IS COMPLETE..
e ISSUES 0THER THAN CONFORMANCE ISSUES CAN BE SUBJECT TO PRE 0PERATIONAL HEARING;
~
CONSEQUENCES CAN BE DELAY, LACK OF STABluTY RELATNE TO PREVIOUSLY APPROVED
~
DESIGN AND SITE.
e cert 6e 11
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10 CFR PART 52 ISSUES OP30RTLN!TY FOR PRIO3 RATIONA_ HEAR NG REQUIRED ACTION e
ANY PREOPERATIONAL HEARING ON CONFORMANCE ISSUES SHOUi.D BE MADE NON-ADJUDICATORY.
e ANY PRE 0PERATIONAL HEARING SHOULD BE UMITED-TO CONFORMANCE ISSUES ONLY.
PETITION TO MODIFY TERMS AND CONDITIONS OF UCENSE SHOULD NOT BE BASIS FOR ANY PRE 0PERATIONAL HEARING.
~
cert 8b 11 1
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SP/90 W DESIGN CERTIFICATION PROGRAM COMPLETE PDA IN 1989 o 3 DRAFT SER's RECIEVED o 1 DRAFT SER REMAINING (PRA BACK-END) o RESOLUTION OF OPEN ITEMS o ACRS SCHEDULE PDA SUBWITTAL BASED ON W/WHI PROGRAW
- o. INCLUDES INTERMEDIATE DESIGN FDA PROGRAW o WILL PROCEED PENDING A FINAL DESIGN PROJECT o NEXT STEP WOULD BE UCENSING REVIEW BASIS DOCUMENT o WILL FOLLOW WITH FDA/ DESIGN CERTIFICATION APPUCATION cert 9-11
7 4
W DESIGN CERTIFICATION PROGRAW AP600 KSIGN STATUS TOTAL PLANT CONCEPTUAL DESIGN COMPLETE e
WAJ0lt FEATUES
.- TWO-LOOP 600 MWe
- ItEACTOR C00l#ii SYSTEW INCLUDE PROVEN COMP 0NENTS
- DIGITAL IPS/lCS
- STANDAltD W FUEL ASSEMBLY
- PASSIVE. SAFETY INJECTION SYSTEM
- PASSIVE CONTAINMENT C00UNG SYSTEW
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W/NRC EAltLY SAFETY ltEVIEW INTERACTION DESIGN REPORT SUBWITTED 2/22/89 FOCUS ON FUNDAMENTAL SAFETY CHARACTERISTICS PROVIDE NRC FEEDBACK INTO DEVELOPMENT PROGRAW e
WILL PROVIDE INPUT TO UCENSING REVIEW BASIS 1
NRC FEEDBACK NEEDED onaut l
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4 W DESIGN CERTIFICATION PROGRAW
-AP600 DESIGN CDmFICATION PROGRAW W PROPOSAL FOR DETAILED DESIGN AND DESIGN CERTIFICATION PROGRAW SUBWITTED TO DOE 5/1/89 e
WILL APPLY FOR DESIGN CERTIFICATION IN ACCORDANCE WITH 10 CFR PART 52 LICENSING. REVIEW BASIS
- DOCUMENT APPUCABLE REQUIREMENTS
- ESTABUSH REVIEW PROCESS
- PROVIDE ~ PROCEDURE FOR PROBLEW RESOLUTION ONGOING KEY ISSUE REVIEW FDA REVIEW
- STANDARD SAFETY ANALYSIS REPORT SUBWITTAL
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- DRAFT SER/ FINAL SER
- ACRS REVIEW DESIGN CERTIFICATION RULEMAKING ced1111
SUMMARY
OF W STRATEGY W WILL C0WPLY WITH 10CFR52 FOR DESIGN CERTIFICATION COWPLETION OF SP/90 PDA REQUIRED FOR NEAR TERW VIABluTY OF PWR PLANT OPTION
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SUCCESSFUL AP600 DESIGN CERTIFICATION PROGRAW'IS OPPORTUNITY TO:
- REVITAUZE US NUCLEAR POWER OPTION
- RE-ESTABUSH US TECHNOLOGY LEADERSHIP
- DEMONSTRATE IMPROVED NUCLEAR REGULATORY PROCESS O
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S3/90 _CE\\S \\G SS ES M. H. Shannon, Manager Plant & Systems Eva uotion Licensing
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REV EW SJ8, EC S
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SEVERE ACC DE\\~
SSUES
\\ EW s)JS RY S~A\\ JAR)S REMA \\ \\G SP/50 AC"0N S 5
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SP/90 UCENSING ISSUES NRC: ADDITIONAL REVIEW SUBJECTS 60 YEAR UFE 4
- SP/90 DESIGN CURRENTLY 40 YEARS (50.51}
- TO CHANGE TO 60 YEARS AT FDA
- ADDRESS FATIGUE, CORROSION, THERMAL AGING FIRE PROTECTION
- IMPROVED SEPARATION OUTSIDE CONTAINMENT
- CONTAINMENT IS SINGLE FIRE AREA WHICH CANN0T BE LOST ENTIRELY
- NEW REQUIREMENTS MUST BE CAREFULLY REVIEWED e
TECHNICAL SPECIFICATIONS
- TECH. SPEC'S. PART OF FDA SUBMITTAL
- USE PRA TECHNIQUES DURING DEVELOPMENT
- AUTOMATED TECH. SPEC'S. (SPEC APPRAISAL) 990b-13
y SP/90 UCENSING lSSUES NRC ADDITIONAL REVIEW SUBJECTS (Con't) e TESTING AND MAINTENANCE
- UTIU1Y PROGRAMS
- ENSURE REUABluTY OF SYSTEMS & EQUlPMENT WITH PROGRAMS AND COWWITWENTS
- FACTORED INTO DESIGN PROCESS TO ENSURE
. ACCESSIBluTY
- ENSURE ALARA
- CONFlRW CONSISTENCY WITH PRA e
INDUSTRY USE OF MAAP
- W COMMITTED TO MAAP USE
- BEST AVAILABLE CODE FOR APPUCATION
- W PART OF EPRI EFFORT TO DEVELOP SEVERE ACCIDENT CRITERIA AND ENDORSES DIRECTION TAKEN
- PRA EVALUATION CRITERIA REQUIRED e
STATION BLACKOUT AND ELECTRICAL SYSTEMS
- DESIGN FEA1URES ADDED TO ADDRESS STATION BLACKOUT ISSUES
- WEETS R.G.1.155 ON STATION BLACK 0UT
i
.,1 SP/90 UCENSING ISSUES.
NRC ADDITIONAL REVIEW SUBJECTSJCont R
LEAK BEFORE BREAK / MATERIAL EMBRITTLEMENT
- CONSIDERED IN SP/90- DESIGN
- CORROSION RESISTANT MATERIALS
- lWPROVED T/H DESIGN OF SECONDARY SIDE
- R.V. DESIGNED FOR LOW NVT OF 1.4 x 10
- WELDS EUMINATED FROM CORE REGION
- W WILL DEMONSTRATE COMPUANCE WITH SRP 3.6.3 AT FDA SOURCE TERMS
- MECHANISTIC SOURCE TERMS SHOULD BE USED FOR DBA's
- W WILL SUPPORT STAFF EFFORT TO ESTABUSH MORE REAUSTIC SOURCE TERMS e
PHYSICAL SECURITY
- STRICT SEPARAT10N BY PHYSICAL BARRIERS
- UMITED ACCESS TO VITAL AREAS
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O SP/90 UCENSING ISSUES 1
NRC ADDm0NAL REVIEW SUBJECTS (Con't)
~
I OBE/DYNAWIC ANALYSIS WETHODS e
- OBE SHOULD NOT CONTROL DESIGN OF SAFETY SYSTEMS
- SP/90 OBE
- WILL BE ADDRESSED IN LRB e
lYPE C CONTAINMENT LEAKAGE RATE
- W AGREES CLR IS FUNCTION OF CONTAINMENT PRESSURE
- CLR ASSUMPTIONS SHOULD BE EXTENDED TO COVER TYPE A INTEGRATED CLR HYDROGEN GENERATION
- SP/90 INCLUDES HYDROGEN IGNITERS TO MmGATE CONSEQUENCES OF 100% METAL / WATER REACTION 9
W sP90e-13 1
e.
SP/90 UCENSING ISSUES SEVERE ACCIDENT ISSUES POST-TWI SAFETY ITEMS (10CFR50.34(f))
e
- ADDRESSED IN WODULE 2, " REGULATORY CONFORWANCE" (1983)
- RECENTLY-UPDATED: UNDERGOING INTERNAL REVIEW e
UNRESOLVED SAFETY ISSUES
- WAY 1988 DOCUMENT ADDRESSED ALL OUTSTANDING USl's, GENERIC' SAFETY ISSUES AND NEW GENERIC ISSUES
- UNDERGOING FINAL UPDATE:
RESPONSE TO NRC QUESTIONS AND LATEST STATUS e
PROBABluSTIC RISK ANALYSIS
- WODULE 16, "PROBABluSTIC SAFETY STUDY"
- LEVEL 3 PRA - INTERNAL EVENTS ONLY
- FDA PRA WILL INCLUDE EXTERNAL EVENTS i
e DETERMINISTIC SAFE 1Y ANALYSIS
- COMPLETION OF REVIEW / DEFENSE PROCESS sp00M3
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~
SP/90 LICENSING ISSUES SEVERE ACCIDENT ISSUES [ Con't)
CORE MELT FREQUENCY ~ 1.5 x 10
[lNTERNAL EVENTS}
NEW DESIGN FEATURES SIGNIFICANTLY REDUCE CMF
~
- 4 EMERGENCY FEEDWATER PUMPS
- 4 HHSI PUMPS WITH EWST
- AC INDEPENDENT RCP SEAL SUPPORT PUMP CAVITY DESIGN ENSURES MAINTAINING C00LABLE GE0 METRY e
MOST CONTAINMENT FAILURES ARE LONG TERM
[>24 HOURS)
PROBABluTY OF LARGE RELEASE ~ 3.0E-07 e#rn l
__ D
j SP/90 UCENSING ISSUES i
NEW INDUSTRY STANDARDS
.o ANSl/ANS 51.1: NUCLEAR SAFETY CRITERIA FOR DESIGN OF STATIONARY PWR's
- DSER OPEN ISSUES 2,3 AND 5 i
- USED AS GUIDEUNE FOR SAFETY CLASSIFICATION OF SYSTEMS, STRUCTURES AND COMPONENTS ON SP/90 & AP600
- UNDERGOING RIMS 10N BY NUPPSCO e
ANSl/ANS 2.3: STANDARD FOR ESTIMATING TORNADO AND EXTREME WIND CHARACTERISTICS AT NUCLEAR POWER SITES
- DSER OPEN ISSUE 6
- COVERED BY ONE OF EPRI REGULATORY OPilWlZATION SUBJECTS
- NRC OENTIFYING NEW POSm0N RIMSING R.G.1.76 e
ASCE 4-86:
STANDARD FOR SEISMIC ANALYSIS OF SAFETY j
RELATED NUCLEAR STRUCTURES
- DSER OPEN ISSUES 25 THROUGH 29
)
- RESEARCH TO RIMEW AND PREPARE POSm0N ON STANDARDS
- W/NRC TELECON SCHEDULED
~
sp90b13
~
SP/90 'JCENSING ISSUES NEW INDUSTRY STANDARDS [ Con't)
ACl-349-1985:
CODE REQUIREMENTS FOR NUCLEAR SAFETY RELATED CONCRETE STRUCTURES
- DSER OPEN ISSUES 35 AND 36
- 1976 EDITION ENDORSED BY R.G.1.142
- APP. B TO STANDARD (EMBEDMENTS & ANCHORAGE}
lSSUED LATER AND NOT ENDORSED
- W/NRC TELECON SCHEDULED ANSI /AISC-N690-1984: SPECIFICATION FOR THE DESIGN FABRICATl6N AND ERECTION OF STEEL SAFETY RELATED STRUCTURES FOR BLDGS.
- DSER OPEN ISSUES 16 AND 18
- AISC AND ASME COMMITTEES DISCUSSING N690 APPUCATION TO SUPPORTS WITH NRC
~
13
SP/90 UCENSING ISSUES SCEDULE FOR PDA l
l e
FINAUZE UPDATE OF MODULE 2. " REGULATORY CONFORMANCE"
--MAY 15,1989 e
RECENED DRAFT SER's
- MARCH 1988 - FRONT END PRA l
l
- JUNE 1988 - l&C, ELECTRIC POWER, ETC...
- MARCH 1989 - SAFETY SYSTEMS, AUXILIARY, CONTROL ROOM e
W RESPONSETO DSER OPEN ISSUES.
- SYSTEMS - MAY 15, 1989
- STRUCTURES (CH.3) - MAY 31,1989
- ACCENT ANALYSIS - JUNE 15,1989 e
RECEIPT OF BACKEND PRA DRAFT SER - ?
e HOLD ACRS WEETINGS
- SU8 COMMITTEE - JUNE, JULY AND AUGUST 1989
- FULL COWWITTEE - AUGUST 1989 RECEPT OF FINAL SER/PDA - SEPTEMBER 1989 e
1
r RESAR SP/90 PRESENTATION TO NRC STAFF MAY 8, 1989 e
T. VAN DE VENNE ENGINEERING MANAGER
~
APWR DEVELOPMENT i
4 J
j
SP/90 POSITION ON CURRENT LICENSING ISSUES 0
ANTICIPATED TRANSIENTS WITHOUT SCRAM 0-STATION BLACK 0UT-0 INTERSYSTEM LOCA 0-SOURCE TERM 0
PROBABILISTIC RISK ASSESSMENT 10 CONTAINMENT
- 04
-CONDITIONAL CONTAINMENT FAILURE O
MID-LOOP OPERATION 0
FIRE PROTECTION e
0319J/tvjv/050489 4
ANTICIPATED TRANSIENTS WITHOUT SCRAM SP/90 FEATURES l
-0 AMSAC' TYPE SYSTEM i
TURBINE TRIP EMERGENCY FEEDWATER ACTUATION 0
MODERATOR TEMPERATURE COEFFICIENT SIGNIFICANTLY MORE NEGATIVE THAN FOR-CURRENT PLANTS.
O SAFETY VALVE CAPACITY SIZING SIMILAR TO CURRENT PLANTS.
O B0 RATION BY OPERATOR ACTION.
0319J/tv Jv/050489
' ANTICIPATED TRANSIENTS WITHOUT SCRAM j
O CURRENT RESAR SP/90 CORE MELT FREQUENCY CONTRIBUTION IS 5.5 E-08/YR.
- 0 APPLICATION OF MOST RECENT DATA USED IS WOG ATWS EVALUATIONS WOULD INCREASE ATWS CORE MELT FREQUENCY CONTRIBUTION TO 1.3 E-07/YR.
O NO CREDIT-HAS BEEN TAKEN FOR REDUCED NUMBER OF TRANSIENTS THAT ARE EXPECTED TO BE THE NORM FOR FUTURE PLANTS 0
ATWS IS NOT A SIGNIFICANT CONTRIBUTOR TO SP/90 CORE MELT FREQUENCY e
.0319J/2ejv/050489
~
STATION BLACK 0UT SP/90 DESIGN FEATURES O
TWO AC/DC INDEPENDENT TURBINE-DRIVEN EMERGENCY FEEDWATER PUMPS.
0 ONE BACKUP SEAL INJECTION PUMP WITH A DEDICATED POWER SOURCE THAT IS INDEPENDENT OF 0FF-SITE AND ON-SITE AC POWER.
O FOUR HOUR CLASS 1.E BATTERIES.
O INSTALLED, MANUALLY OPERATED CONNECTIONS BETWEEN THE BACKUP SEAL INJECTION PUMP POWER SOURCE AND THE CLASS 1E BATTERIES.
i 0
THE STANDBY ON-SITE NON-SAFETY POWER SOURCE AS PROVIDED FOR IN CHAPTER 11 0F THE EPRI ALWR REQUIREMENTS DOCUMENT WILL BE EVALUATED DURING THE l
FINAL DESIGN PHASE.
O THE PRESENT SP/90 CONFIGURATION MEETS THE REQUIREMENTS OF REG. 1.155.
l 0319J/tv-Jv/050489
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STATION BLACK 0UT SP/90 PRA RESULTS 1
3 O
INITIATING EVENT FREQUENCY 3.5 E-4/YR 0
RESAR SP/90 CORE MELT FREQUENCY 8.1 E-7/YR (ABOUT 50% OF TOTAL CORE MELT FREQUENCY) 0 DOMINANT SEQUENCES LEADING TO CORE MELT SEAL LOCA OCCURS 5.1 E-7/YR NO SECONDARY COOLING 2.9 E-7/YR 0
MORE REALISTIC SEAL LOCA PROBABILITY BASED ON FRAMATOME TESTING WILL REDUCE STATION BLACK 0UT CONTRIBUTION TO CORE MELT FREQUENCY TO 3.5 E-7/YR 0319J/tvJv/050489
J
^
INTERSYSTEM LOCA SP/90 FEATURES l
1 0-RHR SUCTION LINE IS ONLY CREDIBLE PATH FOR INTERFACING LOCA.
O RHR ISOLATION VALVES ARE INCLUDED IN ISS TEST HEADER l
AND WILL BE LEAK TESTED DURING STARTUP.
O RHR SUCTION PIPING DESIGN PRESSURE HAS BEEN INCREASED SUCH THAT GROSS FAILURE WOULD NOT OCCUR EVEN WHEN SUBJECTED TO RCS OPERATING PRESSURE.
0 RHR SUCTION PIPING IS IN OPEN CONNECTION WITH THE i
IN~ CONTAINMENT-EWST SUCH THAT PRESSURE IS RELIEVED FOLLOWING FAILURE OF RHR ISOLATION VALVES.
O RHR PUMP AND PIPING ARE ARRANGED TO ASSURE SUFFICIENT EWST INVENTORY TO ALLOW CONTINUED CORE COOLING WITH NON-AFFECTED ISS SUBSYSTEMS.
0319J/te-jv/050489
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1 0
SP/90' MEETS 10CFR100 DOSE LIMITS FOR DESIGN BASIS l
ACCIDENTS USING CONSERVATIVE SOURCE TERMS.
l O
MORE REALISTIC SOURCE TERMS FOR DESIGN BASIS ACCIDENTS WOULD REDUCE THE ADVANTAGES ASSOCIATED WITH A DOUBLE CONTAINMENT AND WOULD SIMPLIFY IMPLEMENTATION OF PASSIVE CONTAINMENT COOLING FEATURES.
l 0399J/tv-jv/050489
o PROBABILISTIC RISK ASSESSMENT 01
~SP/90 LEVEL 3-PRA FOR BOTH INTERNAL AND EXTERNAL -
i EVENTS WILL BE PERFORMED AT FDA STAGE.
O SAR WILL INCLUDE ASSUMPTIONS WITH REGARD TO EQUIPMENT RELIABILITY, MAINTENANCE AND TESTING.
0 EQUIPMENT RELIABILITY IS ASSURED BY ENGINEERING
~~
SPECIFICATIONS, NOT BY RELIABILITY SPECIFICATIONS l
e 0319J/tv Jv/050489
n...
CONTAINMENT SP/90 FEATURES 0
SP/90 GENERALLY COMPLIES WITH EPRI ALWR REQUIREMENTS RELATED TO CC"TAINNENT INTEGRITY (CHAPTER 5)
LARGE DRY CONTAINMENT- (3.1 MILLION CU FT)
CONTAINMENT. LOADS EVALUATED USING MAAP CODE 13% HYDROGEN FOR 75% ZR-WATER REACTION IGNITERS FOR LOCAL HYDROGEN CONTROL
. CAVITY SIZED FOR LONG-TERM DEBRIS COOLING C.tVITY DESIGNED TO RETAIN CORE DEBRIS AC-INDEPENDENT RCS DEPRESSURIZATION O
PRESENT SP/90 DESIGN ALSO INCLUDES IGNITERS FOR-GLOBAL HYDROGEN CONTROL; FINAL CONFIGURATION WILL DEPEND 0N RESOLUTION OF AMOUNT OF ZR-WATER REACTION.
O DIFFERENCES WITH EPRI REQUIREMENTS WILL BE RECONCILED DURING SP/90 FINAL DESIGN PHASE SAFETY GRADE FAN COOLERS VS. SPRAY ONLY IMPLEMENTATION OF ALTERNATE WATER SUPPLY e
e
CONTAINMENT
~
L i
VENTING CAPABILITY
~
1
}h 0L NOT REQUIRED TO MEET STATED SEVERE ACCIDENT RELEASE AND DOSE.G0ALS FOR SP/90 PLANT 9
0 CONDITIONAL CONTAINMENT FAILURE PROBABILITY GOAL
~
INCLUDED IN GE LRB LEADS TO THE NEED FOR VENTING; THE TECHNICAL BASIS FOR THIS GOAL IS NOT CLEAR.
O V$NTING IS NOT A CURE-ALL; THERE ARE DISADVANTAGES, BOTH FROM A TECHNICAL AND A PUBLIC ACCEPTANCE POINT OF VIEW.
l 0
RESOLUTION SHOULD BE AT THE INDUSTRY LEVEL AND NOT AS PART OF A.N INDIVIDUAL VENDOR LICENSE APPLICATION.
0- '
WESTINGHOUSE DOES NOT INTEND TO INCLUDE CONTAINMENT VENTING IN SP/90 PDA APPLICATION.
0319j/te-Jv/050489 i
CONDITIONAL CONTAINMENT FAILURE O
WESTINGHOUSE-ACKNOWLEDGES THE NEED FOR A RUGGED CONTAINMENT AS CLEARLY SPECIFIED IN THE EPRI ALWR l
REQUIREMENTS DOCUMENT.
i 0
SPECIFICATION OF CONDITIONAL. CONTAINMENT FAILURE PROBABILITY MAY GIVE PRIORITY TO MITIGATION OVER PREVENTION.
O CERTAIN EVENTS FEATURE CONDITIONAL CONTAINMENT
. FAILURE PROBABILITY EQUAL TO OR APPROACHING 1.0.
(INTERFACING LOCA, SG TUBE RUPTURE)
I 0.
FOR THE SP/90 PLANT, CONDITIONAL CONTAINMENT FAILURE Is L
PROBABILITY FOR CORE MELT SEQUENCES INVOLVING LOSS OF ALL AC AND LOSS OF COOLING IS ABOUT ONE IR TWO; I
HOWEVER ALL OF THESE ARE LATE FAILURES.
R319J/2vJv/050489
= -
CONDITIONAL CONTAINMENT FAILURE EXAMPLE 0-ASSUMPTION - ADDITIONAL COMBUSTION TURBINE (CT) IS INSTALLED IN THE PLANT.
O DESIGN 'A'
- CT DEDICATED TO CONTAINMENT HEAT REMOVAL FUNCTION UNCHANGED CORE MELT FREQUENCY REDUCED CONDITIONAL CONTAINMENT FAILURE PROBABILITY LOWER SEVERE RELEASE FREQUENCY 0
. DESIGN 'B'
- CT CAN P0,WER EITHER A STARTUP FEEDWATER PUMP OR THE CONTAINMENT HEAT REMOVAL FUNCTION.
~
REDUCED CORE MELT FREQUENCY ESSENTIALLY H0 CHANGE IN CONDITIONAL CONTAINMENT FAILURE PROBABILITY LOWER SEVERE RELEASE FREQUENCY 0
DESIGN 'B' IS CLEARLY PREFERABLE BUT THE PROPOSED CONDITIONAL CONTAINMENT FAILURE CRITERION WOULD RESULT IN DESIGN 'A' BEING SELECTED.
0319J/tv-jv/050489 ausa
~
MID-LOOP OPERATION 4
SP/90 FEATURES O
WATER LEVEL DURING MID-LOOP OPERATION IS AT LEAST 9 INCHES ABOVE ACTUAL MID-PLANT EVALUATION.
O WITH VORTEX BREAKER, AIR ENTRAINMENT STARTS TO OCCUR AT APPROXIMATELY 3 INCHES BELOW MID-PLANE ELEVATION, BUT IS LIMITED TO LESS THAN 10%.
O RHR SUCTION LINES ARE SLOPED CONTINUOUSLY DOWNWARDS TOWARDS RHR PUMPS AND ARE, THEREFORE, SELF-VENTING.
O RHR PUMP SUCTION LINES ARE DESIGNED TO PROVIDE ADEQUATE PUMP NPSH AT FULL FLOW ASSUMING SATURATION IN THE HOT LEG.
O DEDICATED, REDUNDANT NARROW RANGE LEVEL INSTRUMENTS WITH MAIN CONTROL ROOM INDICATION AHD ALARM ARE PROVIDED.
O EACH OF THE FOUR REDUNDANT ISS SUBSYSTEMS INCLUDES RHR FLOW MEASUREMENT AND MAIN CONTROL ROOM INDICATION.
0319J/tv-jv/050489
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FIRE PROTECTION
^
SP/90 STATUS
.BTR 9.5-1 AND EPRI ALWR REQUIREMENTS SPECIFY ACCEPTABLE SEPARATION FOR SAFE SHUTDOWN EQUIPMENT.
l
-THREE HOUR FIRE BARRIER 20 FT. HORIZONTAL SEPARATION WITH DETECTION AND q
AUTOMATIC SUPPRESSION.
ONE HOUR ENCLOSURE WITH DETECTION AND AUTOMATIC SUPPRESSION.
0 OUTSIDE CONTAINMENT, SP/90 RELIES ALMOST EXCLUSIVELY ONTHREEHOURFIREBARRIERS.
O INSIDE CONTAINMENT, SP/90 UTILIZES EXISTING WALLS AS FIRE BARRIERS WHENEVER POSSIBLE; HOWEVER, SEPARATION AND ENCLOSURES ARE PRIMARY MEANS OF FIRE PROTECTION.
O DSER STATES THAT SP/90 FIRE PROTECTION IS NOT ACCEPTABLE FOR AN ADVANCED PLANT.
0319J/tv-jv/050489
L ;-
FIRE PROTECTION 1
PRA EXPERIENCE
~0 RESULTS.FOR A TYPICAL APPENDIX R PLANT INDICATE THAT-MAJORITY OF RISK IS DUE TO FIRES OUTSIDE CONTAINMENT.
CONTRIBUTION FROM FIRES INSIDE CONTAINMENT IS NEGLIGIBLE.
0 EVEN IF A. FIRE INSIDE CONTAINMENT WERE TO LEAD TO A CORE MELT, SEVERE RELEASES ARE UNLIKELY TO OCCUR SINCE THE CONTAINMENT SPRAY FUNCTION IS UNAFFECTED '.
AND WILL BE ABLE TO PREVENT CONTAINMENT FAILURE.
(
0319J/tv-jv/050489
,J.
FIRE PROTECTION SP/90 POSITION 0
OUTSIDE CONTAINMENT, SP/90 FIRE PROTECTION IS SUBSTANTIALLY IMPROVED. RELATIVE TO CURRENT PLANTS.
IN FACT, IT IS DIFFICULT TO ENVISION A SIGNIFICANTLY HIGHER DEGREE OF SEPARATION.
O I'NSIDE CONTAINMENT, SP/90 FIRE PROTECTION IS AT LEAST EQUAL TO THE BEST CURRENT PLANTS.
O THEREFORE, PLANT RISK DUE TO FIRE IS EXPECTED TO BE LOW FOR THE SP/90 PLANT, AND NO NEW OR ADDITIONAL FIRE PROTECTION CRITERIA BEYOND BTP 9.5-1 ARE REQUIRED.
t i
I i
0319J/tv-Jv/050489 i
o
.i SP/90 PROBABILISTIC RISK ASSESSMENT I
0 PRA-RESULTS O
BNL REVIEW 0
OPERATOR ACTION O
CONCLUSION 0319J/tvJv/050489
SP/90 PRA RESULTS 1
EVENT IEE IME 1
' TRANSIENTS 10 1.6E-7 LOSS OF 0FFSITE POWER 0.12 8.1E-7
-SG TUBE RUPTURE 3.1E-2 2.1E-7 SECONDARY LINE BREAK 8.0E-4 3.0E-9 SMALL LOCA.
5.6E-3 2.0E-8
' LARGE LOCA 4.0E-4 2.2E-8 I
ATWS 3.0E-4 5.5E-8 INTERFACING LOCA 1.0E-6 5.9E-9 VESSEL FAILURE 1.0E-7 1.0E-7 LOSS OF COOLING
_2,0E-5 3.4E-Z TOTAL
~10 1,5E-6 0319]/tv-jv/050489
c' SP/90 PRA
~
~..
BNL REVIEW C0144ENTS 0
INTEGRATED PROTECTION SYSTEM HAS NOT BEEN MODELLED EXPLICITLY AND IS, THEREFORE, NOT PROPERLY ACCOUNTED FOR IN THE FAULT TREES, IN PARTICULAR " TRANSIENTS."
O PROBABILITY OF MECHANICAL FAILURE OF THE CONTROL IS MUCH HIGHER THAN ASSUMED BY WESTINGHOUSE.
THIS AFFECTS PRIMARILY THE ATWS EVENT TREE.
O NO SIGNIFICANT DISCREPANCIES IN OTHER EVENT TREES WERE IDENTIFIED.
~
i 0319j/2ejv/050489
F"r :
o
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CORE MELT FREQUENCY COMPARISON-I BNL Review SP/90 PSS Transient 3.04E-6(51%)*
1.61E-7(10.5%)
LOOP 1.00E-6(17%)
8.08E-7 (52.5%)
SGTR 2.71E-8(0.45%)
2.14E-8 (1.4%)
SSBK 2.09E-8(0.35%)
2.98E-9 (0.2%)
SLOCA 2.47E-8(0.41%)
1.97E-8 (1.3%)
LLOCA 2.24E-8(0.38%)
2.24E-8(1.5%)
ATWS' 1.33E-6(22%)
5.53E-8(3.6%)
ISL
- 2.81E-8(0.47%)
5.85E-9 (0.4%)-
VEF 1.00E-7 (1.7%)
1.00E-7 (6.5%)
LC 3.74E-7 (6 3%)
3.43E-7 (22.3%)
. Total-5.97E-6 1.54E-6
- Number in parenthesis inoicates contribution to total core melt frequency.
9 e
9 e
SP/90 PRA e INTEGRATED PROTECTION SYSTEM O
IPS HAS NOT BEEN MODELED EXPLICITLY, PRIMARILY BECAUSE OF A LACK OF DETAILED DESIGN INFORMATION.
O THE ONLY SEQUENCE THAT COULD BE SIGNIFICANTLY IMPACTED BY IPS RELIABILITY IS ' TRANSIENTS.'
0 HOWEVER, THERE IS CONSIDERABLE DIVERSITY IN THE ACTUATION OF THE PROTECTIVE FUNCTIONS FOR
' TRANSIENTS.
c a STARTUP FEEDWATER - CONTROL SYSTEM.& OPERATOR EMERGENCY FEEDWATER - PROTECTION SYSTEM, AMSAC, & OPERATOR-BLEED AND FEED - OPERATOR O
IN ALL CASES, OPERATOR ACTION IS NOT REQUIRED FOR ABOUT 30 MINUTES.
O FOR THESE REASONS, IPS RELIABILITY IS NOT EXPECTED TO HAVE A SIGNIFICANT IMPACT ON SP/90 PRA.
0319j/te-Jv/050489
SP/90 PRA CONTROL ROD INSERTION 0
IN THEIR REVIEW, BNL ASSUMED THAT FAILURE OF TWO RCC'S TO ENTER THE CORE CONSTITUTES FAILURE TO SCRAM.
\\
0 IN REALITY, SCRAM IS SUCCESSFUL EVEN IF MANY MORE RCC'S FAIL TO ENTER THE CORE; THE BNL RESULTS ARE, THEREFORE, INCORRECT.
O IN THE SP/90 PRA, COMMON MODE MECHANICAL FAILURE OF THE RCC'S WAS NOT ACCOUNTED FOR.
APPLYING METHODOLOGY DEVELOPED FOR THE WOG ON ATWS ISSUES WOULD DOUBLE THE SP/90 ATWS CONTRIBUTION, BUT IT WOULD REMAIN A SMALL PERCENTAGE OF THE TOTAL.
~
0399]/tv jv/050489
1 l
CORE MELT FREQUENCY COMPARISDN I
BNL Review SP/90 PSS BNL Review SP/90 PSS (Adjusted)
(Adjusted)
Transient 3.04E-6(51%)*
1.61E-7 (10.5%)
1.9E-7 1.6E-7 LOOP 1.00E-6(17%)
8.08E-7(52.5%)
1.0E-6 8.1E-7 SGTR 2.71E-8(0.45%)
2.14E-8(1.4%)
2.7E-8 2.1E-8 SSBK 2.09E-8(0.35%)
2.98E-9(0.2%)
2.1E-8 3.0E-9 SLOCA 2.47E-8(0.41%)
1.97E-8 (1.3%)
2.5E-8 2.0E-8 LLOCA 2.24E-8(0.38%)
2.24E-8(1.5%)
2.2E-8 2.2E-8 ATWS 1.33E-6(22%)
5.53E-8 (3.6%)
1.6E-7 1.3E-7 ISL 2.81E-8 (0.47%)
5.~85E-9(0.4%)
2.8E-8 5.9E-9 VEF 1.00E-7(1.7%)
1.00E-7(6.5%)
1.0E-7 1.0E-7 LC 3.74E-7 (6.3%)
3.43E-7 (22.3%)
3.7E-7 3.4E-7 Total 5.97E-6 1.54E-6 2.0E-6 1.6E-6
- Number in parenthesis indicates contribution to total core melt frecuency.
4
,s,
.SP/90 PRA 1
l OPERATOR ACTION 0:
ALTHOUGH THE EVENT TREES STILL CONTAIN OPERATOR ACTION N0 DES, THE RESULTS OF THE SP/90 PRA ARE RELATIVELY l
I
' INSENSITIVE TO OPERATOR ERROR.
i l
O
HOST SP/90 OPERATOR ACTIONS ARE SIMPLE.
MANUAL SCRAM BLEED & FEED CVCS B0 RATION
-O' THE ONLY SP/90 OPERATOR ACTION THAT IS COMPARABLE IN COMPLEXITY TO TODAY'S FLANTS IS FOR STEAM GENERATOR TUBE RUPTURE.
HOWEVER, AN AUTOMATIC SG OVERFILL PROTECTION SYSTEM PROVIDES BACKUP IN CASE OF OPERATOR ERROR.
t 0319J/tv-jv/050489
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SP/90 PRA EFFECT OF OPERATOR PERFORMANCE ONLPRA RESULTS RESAR SP/90 INCREASED BASE OPERATOR ERROR L
TRANSIENT 1.6E-7 1.9E-7 LOOP 8.1E-7 8.1E-7 SGTR 2.1E-8 1.1E-7 SSBK 3.0E-9 3.2E-9 SLOCA 2.0E-8 2.3E-8 LLOCA 2.2E-8 2.2E-8
~
.5.5E-8 3.4E-7 ISL-5.9E-9 5.9E-9 VEF-1.0E-7 1.0E-7 LC 3.4E-7 3.4E-7 T0iAL 1.5E-6 1.9E-6 19j/tv Jv/050489
SP/90 PRA CONCLUSIONS f
0 BY APPLYING PRA METHODS SYSTEMATICALLY THROUGHOUT THE
' DESIGN PROCESS, THE PROBABILITY OF ALL TRADITIONAL PWR CORE MELT SEQUENCES HAS BEEN REDUCED DRAMATICALLY.
e 0
THIS IMPROVEMENT HAS BEEN ACHIEVED IN SEVERAL WAYS:
SEPARATION OF FUNCTIONS HIGHER REDUNDANCY ADDITIONAL DIVERSITY LESS OPERATOR ACTION PHYSICAL SEPARATION 0
ALTHOUGH THE RESAR SP/90 IS NOT AS DETAILED AS THOSE PERFORMED FOR OTHER PLANTS, THERE IS SUFFICIENT CONSERVATISM IN MOST ASSUMPTIONS SUCH THAT WESTINGHOUSE HAS HIGH' CONFIDENCE THAT THE GOAL OF ONE COREMELT PER HILLION YEARS CAN BE APPROACHED OR MET.
0319j/tv-Jv/050489
o
^
SP/90 PRA-L 2-CONCLUSIONS _
O PROBABILITY 0F'EARLY CONTAINMENT FAILURE DUE TO
-OVERPRESSURIZATION IS EXTREMELY LOW (~1.0E-9/YR).
1 i
O PROBABILITY OF CONTAINMENT BYPASS IS VERY LOW
- (~3.0E-8/YR).
0 CONDITIONAL CONTAINMENT FAILURE PROBABILITY IS HIGH FOR SOME SEQUENCES; HOWEVER, ALL OF THESE ARE LATE FAILURES
(>24 HOURS) AND LESS CONSERVATIVE RECOVERY ASSUMPTIONS WOULD REDUCE THIS CONDITIONAL PROBABILITY.
0 CALCULATED RELEASE FOR WORST CASE SITES ARE SUCH THAT THERE IS HIGH CONFIDENCE THAT THE EPRI ALWR REQUIREMENTS SEVERE RELEASE GOAL CAN BE MET.
l 0319]/tvjv/050489
7..
1 e
.4 s
SP/90
) -
EPRI ALWR REQUIREMENTS RELATED TO PLANT SAFETY 1
i 1
~
0 RCP SEAL DESIGN 0
CONTAINMENT HEAT REMOVAL 0
CHARC0AL FILTRATION 0,, COMBUSTION TURBINE 0319j/tv-Jv/050489 r.,
.a r
,,,,..,..,,.1
REACTOR COOLANT PUMP SEAL
~
t 0
STANDARD WESTINGHOUSE RCP SEALS REQUIRE EITHER SEAL INJECTION OR THERMAL BARRIER COOLING TO MAINTAIN FULL INTEGRITY.
O TESTING HAS DEMONSTRATED THAT LOSS OF SEAL SUPPORT SYSTEM RESULTS IN LIMITED SEAL LEAKAGE (ABOUT 20 GPM MAXIMUM PER PUMP).
~
O PAST EVALUATIONS HAVE SHOWN THAT A SAFE SHUTDOWN SEAL IS TECHNICALLY FEASIBLE, BUT WILL REQUIRE SIGNIFICANT DEVELOPMENT AND MAY COMPROMISE BASIC RCP RELIABILITY.
O WESTINGHOUSE WILL RETAIN THE PRESENT RCP DESIGN AND WILL CONTINUE TO INCLUDE DIVERSE SEAL PROTECTION IN THE SP/90 DESIGN.
0319J/tv jv/050489
SP/90 i
L OTHER EPRI ALWR REQUIREMENTS ISSUES 0
CONTAINMENT HEAT REMOVAL SAFETY GRADE FAN COOLERS VS. SPRAY ONLY SP/90 DESIGN CAN EASILY BE MODIFIED
~
IMPLEMENTATION TO BE BASED ON STAFF'S POSITION O
CHARC0AL FILTRATION SP/90 GENIRALLY PROVIDES FOR CHARC0AL FILTRATION DELETION OF CHARC0AL FILTRATION IS SIMPLE IMPLEMENTATION TO BE BASED ON STAFF'S POSITION j
0 COMBUSTION TURBINE ADDITIONAL COMBUSTION TURBINE AC POWER SOURCE COST / BENEFIT EVALUATION DURING FINAL DESIGN 0319J/tvJv/050489
APWR STATUS ENGINEERING PROGRAM
.0 DETAILED ENGINEERING IN NON-SITE SPECIFIC AREAS IS CONTINUING.
l 1
L.0 -
AS PART OF THE APWR UPGRADING PROGRAM PHASE 1, THE DESIGN.0F THE REACTOR COOLANT AND INTEGRATED SAFEGUARDS SYSTEMS WAS COMPLETED.
IN ADDITION, DESIGN SIMPLIFICATION IN SEVERAL AREAS WAS INVESTIGATED.
0 THE CURRENT APWR UPGRADING PROGRAM PHASE.2 INCLUDES THE DESIGN OF THE EMERGENCY FEEDWATER, CHEMICAL AND VOLUME CONTROL, AN6 BORON RECYCLE SYSTEMS.
AN IMPROVED FEEDWATER CONTROL SYSTEM IS ALSO UNDER EVALUATION.
0399J/ > jv/050489
I PROJECT IMPLEMENTATION 0.
KANSAI AND HOKURIKI ELECTRIC POWER COMPANY ARE JOINTLY INVESTIGATING THE TAKAYAMACHI SITE (200 MILES NORTHWEST OF TOKYO).
l 0
TWO YEAR SITE EVALUATION STARTED IN DECEMBER OF 1988.
PRELIMINARY P'ANS CALL FOR TWO PHASES OF TWO APWR UNITS O
L EACH.
0319J/tvJv/050489
~
I AP600 SYSTEMS DESIGN 1.
PASSIVE SAFETY SYSTEMS DESIGN - T.SCHULZ 2.
SAFETY & RISK ANALYSIS T.SCHULZ 3.
PASSIVE SAFETY SYSTEM TESTS ~
- L.CONWAY e
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AP600 SYSTEMS DESIGN APPROACH o SYSTEMATIC APPROACH - CRITERIA / GOALS 8
o AGGRESSIVELY SIMPLIFY SYSTEMS
- DESIGN, PROCUREMENT, CONSTRUCTION,
{
OPERATION, INSPECTION, MAINTENANCE o EMPLOY PASSIVE SAFETY SYSTEMS
- SIMPLIFIED SAFETY SYSTEMS
- NON-SAFETY SUPPORT SYSTEMS o OPTIMIZE / INTERGRATE SYSTEMS WITH.....
- SAFETY AND RISK ANALYSIS
- PLANT ARRANGEMENT
- MODULARIZED CONSTRUCTION 71505/05/1989
- - ~ ~
~---
4 AP600 PASSIVE SAFETY FEATURES 1
0 PASSIVE DECAY HEAT REMOVAL
~
- NATURAL CIRCULATION HX CONNECTED TO RCS 0
PASSIVE SAFETY INJECTION
- N2 PRESSURIZED ACCUMULATORS
- GRAVITY DRAIN CORE MAKEUP TANKS, HIGH. PRES
- GRAVITY DRAIN IN-CONTAINMENT REFUELING WATER STORAGE TANK, LOW PRES
- GRAVITY DRAIN RECIRCULATION FROM CONTAINMENT
~
o PASSIVE CONTAINMENT COOLING
- STEEL CONTAINMENT SHELL TRANSFERS HEAT TO NATURAL CIRCULATION AIR COOLING AND GRAVITY DRAIN WATER COOLING.
o PASSIVE CONTAINMENT SPRAY
- N2 PRESSURIZED ACCUMULATORS O
EMERGENCY HVAC
- COMPRESSED AIR FOR HABITABILITY OF ECR
- CONCRETE WALLS FOR HEAT SINK TL505/05/1989 l-
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-o NORMAL SHUTDOWNS
-- AB0VE 350 F / 400 PSIG; STARTUP FEEDWATER TO SG FROM DEAERATING HEATER (2 PUMPS, BLACKOUT DIESEL)
BELOW 350 F / 400 PSIG; SPENT FUEL COOLING SYSTEM (2 PUMPS & HX, CCW/SW COOLING, BLACKOUT DIESEL)
O TRANSIENTS (SAFETY)
ANY RCS CONDITION; PASSIVE RHR HX LOCATED IN IRWST (1 HX, NATURAL CIRCULATION)
BACKUP COOLING BY FEED & BLEED WITH CMT/ACCUM/
IRWST INJECTION & RCS DEPRESSURIZATION o LOCA (SAFETY) 4 INJECTION FROM CMT/ACCUM/IRWST ALONG WITH RCS DEPRESSURIZATION LONG TERtf HEAT REMOVAL BY STEAMING TO CONTAINMENT' WITH PASSIVE CONTAINMENT COOLING PROVIDING HEAT SINK.
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- IMPROVED RHR CONNECTION TO THE RCS HOT LEG DEVELOPED AND TESTED (1/4 SCALE)
- SHORT SECTION 0F LARGER PIPE (20") CONNECTED TO BOTTOM 0F HOT LEG
- ALLOWS PUMP OPERATION WITH LOWER WATER LEVELS BEFORE.ANY AIR IS ENTRAINED AND WITH HOT LEG ESSENTIALLY EMPTY UITHOUT AIR BINDING
- RAISED SG ALLOWS HIGHER NORMAL MID LOOP LEVEL (80%)
- NARROW RANGE HOT LEG LEVEL INSTRUMENT WITH MCR READOUT AND ALARM
- ALL NORMAL RCS DRAIN OPERATIONS CONTROLLED AND MONITORED FROM MCR
- SUCTION LINE DESIGNED TO PROVIDE PUMP WITH ADEQUATE NPSH AT FULL FLOW WITH SATURATED WATER
- SUCTION LINE ROUTED S0 THAT IT IS SELF VENTING
- RUGGED PUMP DESIGN CAN TOLERATE SOME AIR INGESTION AND CONTINUE OPERATING; CAN BECOME AIR B0UND WITHOUT DAMAGE ammnm
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.AP600 - RCS INVENT 0RY CONTROL o NORMAL OPERATION i
- cVc5 MAKEUP PUMPS (2 PUMPS, NNS, BLACKOUT DIESEL) i e TRANSIENTS AND SMALL LEAKS (SAFETY) 4
- CORE MAKEUP TANKS (2 TANKS, RCS PRES, GRAVITY)
O LOCA (SAFETY) 1
- CORE MAKEUP TANKS (2 TANKS, RCS PRES, GRAVITY)
- ACCUMULATORS (2 TANKS, 700 PSIG, N2)
- AUTOMATIC RCS DEPRESSURIZATION ALLOWS LONG TERM INJECTION FROM:
- INCONTAINMENT REFUELING WATER STORAGE TANK (1 TANK, CONTAINMENT ATM)
- RECIRCULATION FROM CONTAINMENT BY GRAVITY t
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AP600 - CONTAINMENT HEAT REMOVAL o NORMAL OPERATION 1
- FAN COOLERS COOLED BY CHILLED WATER (2 COILS, 4 FANS, NNS, BLACKOUT DIESEL)
- o. TRANSIENTS (SAFETY)
- IRWST ABSORBS DECAY HEAT FROM PRHR FOR ~2 HR BEFORE BOILING TO CONTAINMENT
- STEAM CONDENSES ON INSIDE OF CONTAINMENT SHELL, MOST CONDENSATE DRAINS BACK TO IRWST CONTAINMENT SHELL COOLED BY NATURAL CIRCULATION-AIR
& GRAVITY DRAIN OF WATER FROM ELEVATED TANK O' LOCA (SAFETY)
- LARGE CONTAINMENT VOLUME ABSORBS INITIAL MASS /
ENERGY
- STEAM FROM LOCA & RCS DEPRES ENTERS CONTAINMENT, CONDENSES ON INSIDE OF CONTAINMENT SHELL, AND DRAINS BACK TO IRWST/ SUMP
- CONTAINMENT SHELL COOLED BY NATURAL CIRCULATION AIR
& GRAVITY DRAIN OF WATER FROM ELEVATED TANK TLs05/06/1989
AP600 - DESIGN SIMPLIFICATION I
l l-PLANT FEATURES STD 2 LOOP AP600 PUMPS - SAFETY 25 0
- NNS 188 139
)
HVAC FANS 52 27 NVAC FILTER UNITS 16 7
VALVES - NSSS (>2")
512 215 BOP (>2")
2041 1530 PIPE - NSSS (>2")
44,300 FT 11,042 FT
~
- BOP.(>2")
97,000 FT 67,000 FT l CONT PIPE PEN-TOTAL 93 48 0 PEN 38 13 EVAPORATORS 2
0 DIESEL GENERATORS 2 (SC) 1 (NNS)
BLDG. VOL.- CONTAINMENT 2.7 MIL FT3 3.0 MIL FT3
- SEISMIC 6.7 MIL FT3 1.6 MIL FT3
- NON SEISMIC 6.2 mil. FT3 5.6 MIL FT3 l
1 TL501/18/1989
1 AP600 TRANSIENT & ACCIDENT ANALYSIS EVENT NRC APPROVED RESULTS CODES / METHODS o PRESSURIZER SIZING NA No SAFETY VALVE LIFT (LOSS LOAD, BLACKOUT)
WITHoUT PORY l
o COMPLETE LOSS OF RC FLOW YES No DNB o LOSS MAIN FEEDWATER YES No SAFETY VALVE LIFT o FEEDLINE BREAK, DE MFL YES No SAFETY VALVE LIFT o STEAMLINE BREAK, DE MSL YES No DNB o SMALL LOCA, 0"-8" CL YES No CORE UNCoVERY o LARGE-LOCA, DE CL YES PCT < 1500 F o CONTAINMENT, LG :LOCA YES+ MOD.
PRES < 40 PSIG PEAK,
< 11 PSIG 9 24 HR o.SG TUBE RUPTURE, DE SGT YES No SG OVERFILL o 0FFSITE DOSE, LOCA YES+ MOD.
NRC
< 165 REM BEST EST
<1 REM o ATWT, LOSS MFW W/o TRIP YES RCS PRES ~2900 PSIG e CORE MELT FREQ, INTERNAL
~YES
~1x10-6 yn
/
' EVENTS o SIGNIFICANT RELEASE FREQ, YES
~3x10-9/YR INTERNAL EVENTS o CONT RESPONSE, CORE MELT YES INTACT CONT
~
TLs05/05/1989 I
AP600 - SMALL LOCA ANALYSIS
~
O CODE:
NOTRUMP, A WESTINGHOUSE DEVELOPED, NRC APPROVED SMALL LOCA EVALUATION CODE o MODEL:
AP600 FEATURES MODELED INCLUDING CANNED RCP, 2 CL PER HL, DVI, PRHR XX, CMT, ACCUM, AUTO DEPRES, IRWST o CASES:
FOUR CASES WERE ANALYSED
- NO BREAK AUTO DEPRES
- 2" BREAK IN'DVI LINE
- 3" BREAK IN CL
- 8" BREAK 0F DVI N0ZZLE o RESULTS:
NO CORE UNC0VERY OCCURED IN ANY OF THESE CASES, DEPRES WAS SUFFICIENT TO ACHIEVE IRWST INJECTION TL305/05/1989
p.-
. AP600 - NO BREAK DEPRESSURIZATION r
27 26 Top of Hot Leg 25
- -.~ ~.- - - -.~.-.-.-.-
24 Core 23 Mixture 22 Bottom of Hot be~g ~ ~ ~ ' ~ ' ~
~~
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Level 21 (Ft) 20 19 Top of Core
.-.-r-
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0 5
10 15 20 25 Time (Min) 10,000 1000 Pressurizer Pressure 100 (PSIA)
_._._._._.-._._.-._._3 10
~
14.7 PSIA I
I 0
5 10 15 20 25 Time (Min) 25522 82 CAP
'AP600 - CONTAINMENT PRESSURE. ANALYSIS.
O CODE:
COMPACT, A WESTINGHOUSE DEVELOPED, NRC APPROVED CONTAINMENT EVALUATION CODE j
o MODEL:
AP600 FEATURES MODELED INCLUDING AIR / WATER EVAPORATION COOLING, MULTI-NODAL NATURAL CIRCULATION INSIDE CONTAINMENT o CASES:
TWO CASES WERE ANALYSED
~
- DOUBLE ENDED STEAM LINE BREAK
- DOUBLE ENDED CL LOCA o RESULTS:
THE PEAK WAS 32 PSIG AT 750 SEC FOR THE LARGE STEAM LINE BREAK.
PEAK CONTAINMENT PRESSURE WAS 41 PSIG AT 16 SEC FOR THE LARGE LOCA.
IN ONE DAY THE PRES WAS 11 PSIG AND IN 3 DAYS 9 PSIG.
IF WATER WAS NOT RESUPPLIED AT 3 DAYS THE PRES RISE TO < DESIGN IN 13 DAYS.
O TL305/05/19pP l
AP600 CONTAINMENT PRESSURE LARGE LOCA 50 5
_ _ _ _ _ _ _ _ PES (GN PRES 5URE_ _ _ _ _ _ _,
W PEAK PRESSURE
=
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5 TIME (DAYS) 066 A 2715246 m_____________.
l
AP600 - POST ACCIDENT OFFSITE DOSES O
REFLECTS REVISED IODINE SPECIES & RELEASE TIMING
- IODINE SPECIES; 2.5% ELEMENTAL (91%)
2.0% ORGANIC (4%)
95.5% PARTICULATE (5%)
- TIMING; 10% AT T=0, 40% OVER FIRST 30 MIN o
ACCOUNTS FOR SEVERAL REMOVAL MECHANISMS
- SPRAY, ABSORPTION, SEDIMENTATION, SURFACE DEPOSITION o
AP600 PL. ANT CHARACTERISTICS
- LIMITED SPRAY DURATION (30 MIN)
- SPRAY ACTUATED ON HIGH RADIATION
- ELEVATED RELEASE (LATER) o DESIGN CONTAINMENT LEAK RATE
.2%
o PRELIMINARY CALCULATION RESULTS:
- THYROID DOSE AT SITE BOUNDRY (2HR) 165 REM
- THYROID DOSE IN LOW POPULATION 27 REM ZONE (30 DAY)
TLS05/05/1969
i
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AP600 CMF ANALYSIS J
o CORE MELT FREQUENCY; IPE METHODOLOGY, MODIFIED FOR AP600 SPECIFIC FEATURES 4
7 SUPPORT STATES RELATING TO LOSS OF AC AND DC POWER 10 EVENT TREES WITH PLANT DAMAGE STATES' BASED ON W EXPERIENCE AND AP600 SPECIFIC ANALYSIS
. INITIATING EVENT FREQUENCIES FROM WASH 1400 AND W OPERATING EXPERIENCE 40 FAULT TREES TO QUANTIFY THE PASSIVE SYSTEMS RELIABILITY
- - IREP DATA BASE USED WITH COMMON MODE FAILURE QUANTIFIED USING THE BETA METHOD
~
ACTIVE SYSTEM RELIABILITIES ESTIMATO FROM PREVIOUS W STUDIES (SP/90,...)
5p TLs05/01/1989
)
)
AP600 IfiITIATING EVENT FREQUENCY
)
r l
AP600 SP/90
(/YR)
(/YR) o TRANSIENTS W MFW 6.5 10 EXPERIENCE >1985 e TRANSIENTS W/0 MFW 0.57
^
o LOSS OFFSITE POWER 9.7E-2 1.2E-1 EXPERIENCE >1985 o SG TUBE RUPTURE
.3.6E-3 3.1E-2 EXPERIENCE >1985 AND 2 VS 4 SG o SECONDARY BREAKS LATER 8.0E-4 o LOCA - SMALL 3.0E-3 5.6E-3
<2" VS <6" o LOCA - MEDIUM 8.0E-4 2"-8"
^
o LOCA - LARGE 3.0E-4 4.0E-4
>8" VS >6" o LOSS COOLING WATER LATER 2.0E-5 o VESSEL FAILURE 1.0E-7 1.0E-7 o INTERFACING LOCA 1.9E-9 5.8E-9 3 VALVE VS IRWST o ATWT CALC 3.0E-4 TLS057D1/1999
CORE MELT FREQUENCY COMPARSION
)
EVENT CORE HELT FREQUENCY (/YR)
AP600 SP/90 CURRENT TRANSIENTS 1.5E-7 1.6E-7 1.3E-5 BLACKOUT 4.1E-8 8.0E-7 6.6E-6 SG TUBE RUPTURE 1.8E-8 2.1E-8 1.7E-6 LOCA - SMALL 4.8E-7 2.0E-8 8.0E-6, LOCA - MEDIUM 2.8E-7 5.0E-6
^^
LOCA - LARGE 6.1E-8 2.2E-8 8.0E-7 ATWT 8.8E-8 5.5E-8 2.2E-6 LOSS COOLING
~0 3.4E-7 1.1E-5 INTERFACING LOCA 1.6E-9 5.8E-9 1.0E-6 VESSEL RUPTURE 1.0E-7 1.0E-7 3.0E-7 TOTALS 1.2E-6 1.5E-6 5.0E-5 PER YR PER YR PER YR TLS05/01/1989 1
AE600 CONSEQUENCE ANALYSIS i
o ACTIVITY RELEASE FREQUENCY
- PROBABILITY OF EACH PLANT DAMAGE STATE QUANTIFIED FROM CORE MELT FREQUENCY CALCULATION
- FAILURE OF CONTAINMENT ISOLATION SEPARATELY CALCULATED AND COMBINED WITH DAMAGE STATES o CONTAINMENT REIPONSE TO CORE MELT
- MAAP GB) MODEL 0F AP600 PLANT WITH PASSIVE SAFET( FEATURES
- 8 DIFFERENT DAMAGE STATES ANALYSED INCLUDING SEVERAL CONTAINMENT BYPASS EVENTS o.0FFSITE DOSES
- 0FFSITE DOSES CALCULATED USING !4ACCS CODE (WHOLE BODY DOSE, MAX & MEAN)
CONSERVATIVE SITE PARAMETEP's, CATAWBA
- NO EVACUATION ASSUMED TLtO5/01/1999
AP600 SEVERE ACCIDENT o
CONTAINMENT CAN WITHSTAND CORE HELT:
CONTAINMENT PRESSURE STAYS BELOW 2 TIMES DESIGN e
PRESSURE WITH AIR COOLING ONLY.
- DIRECT CONTAINMENT HEATING IS PREVENTED BY RCS DEPRESSURIZATION WHICH IS REDUNDANT, DIVERSE, AND AC POWER INDEPENDENT.
- CONTAINMENT CONCRETE MELT THROUGH IS PREVENTED BY PASSIVE WATER SUPPLIES (IRWST, ACCUN, CMT, PCS) AND REACTOR CAVITY SHAPED FOR F'ORMATION OF C00LABLE DEBRIS BED.
- CONTAINMENT VOLUME IS SUFFICIENT SUCH THAT HYDROGEN CONCENTRATION IS LESS THAN 13% WITH ZIRC WATER REACTIONS UP TO 85% OF THE ACTIVE FUEL.
~
- DC POWERED IGNITERS HANDLE LOCAL HYDROGEN CONCENTRATIONS AND ZIRC WATER REACTIONS GREATER THAN 85%.
~
l l
TLs05/01/1999
.c AP600 SEVERE ACCIDENT L
o CONTAINMENT ISOLATION IMPROVED FEWER PENETRATIONS / FEWER OPEN PATHS OPEN PENETRATIONS USE FAIL CLOSED VALVES CONTAINMENT MINI PURGE USE REDUCED BECAUSE OF IN-CONTAINMENT FILTERS.
o "V" SEQUENCE CORE MELT FREQUENCY REDUCED EFFECTIVE LEAK TESTING, EVEN FOR RHR MOV's.
ADDED THIRD VALVE TO RHR (SFCS) SUCTION LINE, LOCKED CLOSED MANUAL VALVE OUTSIDE CONTAINMENT.
RHR (SFCS) DESIGNED FOR 900 PSIG, WILL NOT FAIL AT 2250 PSIG.
O SGTR CORE MELT FREQUENCY REDUCED AUTOMATIC DEPRESSURIZATION PREVENTS SG OVERFILL WITHOUT OPERATOR ACTIONS.
neuuna.
.~
AP600 SEVERE ACCIDENT WHOLE BODY MAX DOSE AT SITE BOUNDARY (.5 MI) WITHOUT EVACUATION OR. SHELTERING FOLLOWING CORE MELT:
CM EVENT AP600 CURRENT PLANT INTACT CONT
.4 REM
>5 REM 1.2E-6/YR 1E-5/YR WITHOUT WATER COOLING
.8 REM
>25 REM 1.0E-8/YR 1E-6/YR SGTR SEQUENCE
.8 REM.
>25 REM 1.0E-11/YR 1E-6/YR "V" SEQUENCE 48 REN
>25 REM 1.6E-9/YR 1E-6/YR WITH CONTAINMENT 200 REN
>25 REM ISOLATION FAILURE 7.3E-10/YR 1E-6/YR PROBABILITY REQUIRING 3E-9/YR 2E-5/YR EVACUATION (>l REM)
TLS05/01/1989
AP600 EMERGENCY PLANNING o
CONSIDER ELIMINATION OF EVACUATION AND SHELTERING PLANNING BASED ON THE FOLLOWING:
- USE OF PASSIVE SAFETY SYSTEMS
- SMALL ACTIVITY RELEASES FOR DBA (
1 REM)
ASSUMING MECHANISTIC SOURCE TERMS
- PROBABILITY OF 1 REM RELEASE FROM ALL EVENTS LESS THAN 10-7 ya
/
IT APPEARS THAT -THE AP600 CAN HEET.THESE GOALS:
- NO EARLY OR LATE CONTAINMENT FAILURE MODES HAVE BEEN IDENTIFIED, EVEN WITHOUT WATER COOLING.
- SIGNIFICANT RELEASES DUE TO BYPASS ARE < 10.8 ya
/
o REDUCTION OF ENERGENCY PLANNING ZONE WILL BE DISCUSSED WITH THE NRC DURING PRELIMINARY AP600 REVIEW IN 1989 9
AP600 RISK ANALYSIS
SUMMARY
~/
- o. CORE MELT FREQUENCY ANALYSIS
- RELIABLE PASSIVE SAFETY SYSTEMS
. - REDUCED RELIANCE ON OPERATOR ACTIONS
- DIVERSITY WITH PASSIVE SAFETY SYSTEMS AND ACTIVE NORMAL SYSTEMS
- CMF 1.2E-6/YR (INTERNAL), ~3E-6/YR (EXTERNAL) o SEVERE ACCIDENT ANALYSIS
- NO CONTAINMENT OVERPRESSURIZATION CALCULATED
.- AIR COOLING ALONE REMOVES SUFFICIENT HEAT
- ADS PREVENTS DIRECT CONTAINMENT HEATING
- C00LABLE DEBRIS BED WITH PASSIVE WATER SUPPLIES
- HYDR 0 GEN. CONTROL (VOLUME, IGNITERS)
- IMPROVED CONTAINMENT ISOLATION
- REDUCED V-SEQUENCE AND SGTR SEQUENCE
, - SRF 3.0E-9/YR (INTERHAL), ~1E-8/YR (EXTERNAL)
- ELIMINATION OF EMERGENCY PLANNING ZONE
)
O TL905/01/1989
j i
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l d
LOCA AND TRANSIENT ANALYSIS M. Y. YOUNG, MANAGER.
SAFETY ANALYSIS TECHNOLOGY o
LOCA - B.EST ESTIMATE METHODOLOGY o
TRANSIENT - ADVANCED METHODS
+
0494H:LEN/050289
s..
4..
APPENDIX K RULE CHANGES THE NEW APPENDIX K RULE WAS EFFECTIVE ON 10/17/88.
THE KEY FEATURES OF THE NEW RULE ARE:
NO SPECIFIC MODELS/ CORRELATIONS ARE REQUIRED, MODELS MUST BE REALISTIC AND BASED ON RESEARCH DATA CODE MUST GIVE A REALISTIC REPRESENTATION OF THE LOCA PROCESS.
CODES WHICH CONTAIN UNREALISTIC OR OVERSIMPLIFIED MODELS, TO BE CONSERVATIVE, WILL NOT.BE ACCEPTED.
RULE REQUIRES CALCULATION OF NOMINAL,'50TH PERCENTILE PCT, AND AN ACCURATE ESTIMATE OF I
THE 95TH PERCENTI'LE PCT TAKING INTO ACCOUNT THE UNCERTAINTY IN THE PLANT,,THE ACCIDENT CONDITIONS, AND THE COMPUTER CODE.
CURRENT EM MODELS ARE STILL PERMITTED.
0494H:LEH/050289
DEVELOPMENT OF WCOBRA/ TRAC o
INITIALLY DEVELOPED FOR THE USNRC AT PACIFIC NORTHWEST LABORATORY UNDER THE NAME OF COBRA / TRAC o
COBRA / TRAC IS COMBINATION OF COBRA-TF TWO-FLUID, THREE FIELD VESSEL MODEL' TRAC-PD2, FIVE EQUATION DRIFT FLUX LOOP MODEL o
COBRA /TF WAS USED AS PART OF W/NRC/EPRI FLECHT-SEASET FLOW BLOCKAGE PROGRAM
-o W HAS CONTINUED TO MAKE SEVERAL CODE IMPROVEMENTS
~
TO COBRA / TRAC RESULTING IN WCOBRA/ TRAC 1
o WCOBRA/ TRAC HAS BEEN ASSESSED AGAINST
~
FLECHT-SEASET
- LOFT NRU
- ORNL CCTF
- G-2 UPTF
- CREARE TESTS o
NRC APPROVED WCOBRA/ TRAC FOR TWO-LOOP W PLANTS WITH UPI (8/29/88) UNDER SECY-83-472 I
o W PLANS TO LICENSE WCOBRA/ TRAC UNDER NEW APPENDIX K RULE 0696H:LEM/050289
SPN0VA.- BEACON - LOFTRAN5 INTEGRATED SERIES.
NEEDS:
ACCURATE ANALYSIS OF CORE TRANSIENT BEHAVIOR (SPNOVA - LOFTRAN5)
ACCURATE EVALUATION OF PLANT CONDITIONS AS RELATED TO SAFETY ANALYSIS BASIS (SPHOVA-BEACON)
ON-LINE CORE MONITORING OF KEY PARAMETERS (BEACON)
~
PREDICTIVE CAPABILITY TO ASSESS OPTIMUM
~
OPERATION /STARTUP STRATEGY (BEACON)
SOLUTION:
SUPER FAST STEADY STATE & TRANSIENT ANALYSIS SPNOVA USED IN ALL CODES osm m os,
1.
SPNOVA - BEACON - LOFTRAN5 INTEGRATED SERIES PHOENIX-P APPROVED 3D ANC APPROVED ALUCARD 3D SPNOVA 20 89 INCORE STATIC i
SPNOVA 3D SPNOVA BEACON LOFTRAN-5 1D/2D KINETIC MONITORING 3D NSSS i
TRANSIENTS 20 89 20 89 30 89 10 90 O
0501N LEN/050389
4
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AP600fAND CURRENT REACTOR NEEDS FOR BEST ESTIMATE LOCA AND-ADVANCED TRANSIENT METHODS AP600:
PROVIDES A RATIONAL BASIS FOR CERTIFICATION OF SAFETY SYSTEMS AND REACTOR PROTECTION DESIGN CURRENT REACTORS:
ADVANCED METHODS, IN ADDITION TO PROVIDING OPERATING MARGIN, ALSO PROVIDE MARGIN FOR INCREASED OPERATIONAL SAFETY IN SEVERAL L.
AREAS.
REDUCED-VbSSELFLUENCEANDEMBRITTLEMENT
~
L o
o INCREASED DIESEL RELIABILITY FOR LONG-TERM COOLING o
OPERATING STRATEGIES WHICH REDUCE THE POTENTIAL FOR SG TUBE DAMAGE o
REDUCED EXPOSURE o
REDUCED TRIP FREQUENCY I
0502N:MYY/050589
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AP600 DESIGN.WILL COMPLY WITH EPRI ALWR
.o REQUIREMENTS FOR PASSIVE PLANTS W IS CURRENTLY WORKING WITH EPRI AND OTHER o
EPRI CONTRACTORS TO DEVELOP THE'EPRI ALWR REQUIREMENTS DOCUMENT (VOLUME III) FOR' PASSIVE PLANTS o
W STAFF WORKING N AP600 HAVE PARTICIPATED IN THE EPRI WORK ON ALWR REQUIREMENTS FOR EVOLUTIONARY PLANTS 0;
EPRI WILL COMPLETE PASSIVE PLANT REQUIREMENTS BY EARLY 1990 - NRC REVIEW l
WILL HELP ESTABLISH BASES FOR AP600 DESIGN CERTIFICATION REVIEW
)
05778: cmp:890504 a
1 AP600. PLANT OBJECTIVES l
SIMPLE,. DEDICATED, 0
SAFETY INDEPENDENT, PASSIVE SAFETY SYSTEMS SUBSTANTIAL MARGIN'FOR DBA MEET DBA-W/0 OPERATOR ACTION.
CMF < 10-5/YR (ALL EVENTS)
SRF.< 10-6/YR (ALL EVENTS)
SIMPLIFIED DESIGN /
o.
RELIABILITY OPERATION / MAINTENANCE 90% AVAILABILTIY
~
60 YEAR DESIGN LIFE INCREASED DESIGN MARGINS PROVEN EQUIPMENT COMPETITIVE WITH COAL IN U.S.
o ECONOMICS 36 MONTH CONSTRUCTION SCHEDULE NO PROTOTYPE REQUIRED PRE-ENGINEERED / PRE-LICENSED STANDARD DESIGN FOR U.S.
SITES k
0579B:Hurley:890505: cmp
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AP600 EARLY SAFETY REVIEW INTERACTION PURPOSE o
ASSURE THE AP600 PROVIDES A SOUND APPROACH TO COMPLIANCE WITH LICENSING CRITERIA METHODOLOGY o
AP600 PLANT DESCRIPTION REPORT SUBMITTED 2/22/89 o
SMALL TEAM 0F NRC STAFF EXPERTS o
FEEDBACK THROUGH MEETINGS AND QUESTIONS / ANSWER TRANSMITTALS o
THREE MONTH DURATION STARTING NOW o
NRC
SUMMARY
REPORT 0579B:Murley:890505: cmp
O 4,
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1' AP600 TESTING PRESENTATION TO NRC STAFF May 8, 1989 L.E. Conway l
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AP600 TEST PROGRAM a
i
-CONCEPTUAL DESIGN TESTS DEMONSTRATE FEASIBILITY OF ALL UNIQUE DESIGN L:
FEATURES DEFINE AND CONFIRM PLANT ARRANGEMENT, STRUCTURES PROVIDE INPUT TO AND VERIFY TRANSIENT ANALYSIS ASSUMPTIONS l
TESTS FOR DESIGN CERTIFICATION IDENTIFIED
~
ADDRESS ALL SAFETY FUNCTIONS VERIFY ANALYTICAL METHODS / DEMONSTRATE SCALABILITY TO FULL SIZE PLANT DEMONSTRATE MANUFACTURING TECHNIQUES LARGER SCALE OR FULL SIZE COMPONENTS WHERE PRACTICAL 0320J:LEC/JV:050589
~
'g'.
AP600 FEASIBILITY / DEMONSTRATION TESTS TESTING COMPLETED OR'IN PROGRESS:
COMPLETED O
PCCS WIND TUNNEL TEST 0
PCCS AIR FLOW PATH RESISTANCE TEST - COMPLETED 0
PCCS HEATED PLATE TEST - COMPLETED 0
INTEGRATEDPCCS-TEST-INPROGRNSS 0
PASSIVE RHR HEAT EXCHANGER TEST L
0 RHR SUCTION N0ZZLE TEST - COMPLETED l
0 SG CHANNEL HEAD / PUMP SUCTION AIR FLOW TEST 0
HIGH INERTIA RCP JOURNAL BEARING TEST 0
HYDROBALL ICIS TEST 0320J:LEC/JV:050589
AP600 TESTING ANTICIPATED TESTING; 0
DNB TEST' 0
ADS HYDRODYNAMIC TEST 0
PCCS LARGE SCALE INTEGRAL TEST 0
PCCS WATER FILM DISTRIBUTION TEST
'0-PLANT WIND TUNNEL-TEST O
LONG-TERM C00 LING VISUAL EXPERIMENT 0
COMPONENT PERFORMANCE TESTS
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0 SG/RCP WATER FLOW TEST 0
1/7 SCALE RV/ INTERNALS FLOW TEST 0
HYDROBALL ICIS IRRADIATION TEST 0
PRELIMINARY DESIGN PROGRAM TEST EXTENSIONS 4
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0320J:LEC/JV:050589
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AP600 PASSIVE SAFETY SYSTEM 4-TEST RESULTS o WIND TUNNEL TEST OF SHIELD BUILDING / CONTAINMENT SYMMETRICAL AIR INLET AT T0P/0UTSIDE EDGE OF SHIELD BUILDING IS OPTIMUM LOCATION WATER STORAGE TANK SUPPORT BEAMS, ELEVATED EXHAUST ENHANCE AIR FLOW WIND FROM ANY DIRECTION 150, ALWAYS ENHANCES AIR FLOW o AIR FLOW PATH RESISTANCE TEST STREAMLINING TURN ~(DOWNCOMMER / UPCOMMER)1 AND BAFFLE SUPPORTS REDUCED OVERALL LOSS COEFFICIENT L
FROM 4.2 TO 2.4 o HEATED PLATE TEST EVAPORATIVE AND DRY PLATE HEAT TRANSFER RATES i
MATCH COMPUTER MODEL PLATE WETTING CHARACTERISTICS ARE EXCELLENT (HIGH OR LOW PLATE / WATER TEMPERATURES)
WATER FILM BEHAVIOR IS STABLE UNDER ALL CONDITIONS (N0 DRY PATCHING, NO FILM STRIPPING AT HIGH AIR VELOCITY)
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TEST-SECTION VIEW
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PASSIVE RH.R HEAT EXCHANGER TEST OBJECTIVES O
DEMONSTRATE / OPTIMIZE HEAT TRANSFER CAPABILITY OF THE PRHR HX'AT PROTOTYPIC TUBE SIDE FLOW RATES, INLET TEMPERATURES AND IRWST CONDITIONS O
ASSESS PRHR HX DESIGN FEATURES WHICH OPTIMIZE STEAM QUENCHING AND' TANK MIXING
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LICENSING BASELINE FOR.
ADVANCED PASSIVE PWR POWER PLANTS' 9
MAY 8, 1989 M. H. SHANNON, MANAGER I
PLANT & SYSTEMS EVALUATION LICENSING 03258/MHS:1/022789
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p PASSIVE ALWR LICENSING BASELINE L
I 1.
Provides Hierarchy of NRC Requirements and Guidelines Extent of review and public scrutiny Criteria established by NRC but NRC will consider alternative Criteria not yet established; expect designer to do so 2.
Living Document Vs. Fixed Baseline Compatibility with plant licensing process USIs, GSIs, Regulatory change...
3.
Incorporation of Industrial Codes and Standards Working level document for manufacturers / suppliers NRC endorsement based on code date NRC SRPs and Regulatory Guidas are now generally out of date - have not kept up with Standards Committees 0325B/MHS:2/022789
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PASSIVE ALWR LICENSING BASELINE-
- a
- Provides. for Supportable Alternatives to Current Regulatory
- Guidance e-ALWR Reculatory Issuti L
o.
RHR System o
ECC o
Containment _ Integrity o
Canned Motor. Pumps
'o -
Emergency Power Supply o
I&C Architecture o
Fire Protection o
. Physical Security o
Site Characteristics Interfaces ALWR Optimization Issues f
o Use EPRI/NRC Process 0325B/MHS:3/022789
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f i i PASSIVE.ALWR-LICENSING BASELINE R
Anolications of Baseline Who is going to use it and how Designer:
What are Criteria Input,into LRB Regulator:
Cross-Reference to SRP
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Impact of Changes I
Owner:
Provides Baseline for Plant L
0325B/MHS:4/022789 l
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ENCLOSURE'4 E
SUMMARY
OF ISSUE DISCUSSIONS FOR RESAR SP/90 AND AP-600 t
60 YEAR LIFE The application for the.RESAR SP/90 PDA is for a 40 year life and Westinghouse is-considering making the FDA/DC application 0 60 year life. The FDA/DC application for the AP-600 will be for a 60 ye v life. Westinghouse has agreed L
-to address appropriate technical subjects at that stage of review.
FIRE PROTECTION
(
- Westinghouse believes the improved separation outside of containment should resolve the fire protection issue for both plants in those areas.
However, the applicant believes that fire protection inside containment is at least equal to the best operating plants, and that no new or additional fire protection l-1 criteria beyond'BTP 9.5-1 are required. Westinghouse indicated that the plant had three areas where there were not three hour fire barriers:
inside containment, in the main centrol room,-and in the main steam tunnel. Although this was an open issue in the DSER, the applicant stated they felt that, with addition clarification, this issue could be satisfactorily resolved with the staff. The' staff indicated that further discussion of this issue was required.
TECHNICAL SPECIFICATIONS Westinghouse committed to use PRA techniques during development of technical specifications.
The applicant stated they would submit the technical specifications at the FDA/DC stage. The staff indicated that this was acceptable.
TESTING AND MAINTENANCE Westinghouse believes they cannot fully comply with the November 22, 1989 request regarding testing and maintenance since specific equipment will not l
have been selected and maintenance and testing procedures are usually within the scope of the utility operating the plant. Westinghouse indicated that they would factor the needs of testing and maintenance into the design process I
to ensure accessibility, but to go beyond that would be impractical because equipment will not have been selected and because utilities would find dictation of such procedures objectionable.
The staff stated that significant and unnecessary challenges have been made to plant safety systems due tt, operator errors during surveillance testing during cperation.
The staff believes that Westinghouse should consider what must be done to minimize such challenges.
The applicant indicated that they have taken stepstoreducesucherrorsbyprovidingself-testingcapabilitiesfortheir I & C equipment (i.e., Eagle 21, and through the passive safety systems of the AP-600.
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MAAP 1
Westinghouse indicated they were committed to using the MAAP code in their safety analysis. The staff stated they did not intend to review MAAP, but that some interaction was taking place between the staff and the developers of the MAAP code to obtain an understanding of the basic principles of the code.
STATION BLACK 0UT Westinghouse stated that the RESAR SP/90 design included design features to reduce the effects of a station blackout. These design features included two ac/dc independent turbine driven pumps, the backup seal injection pump with a dedicated power source, and four hour Class IE batteries.
The applicant indicated it would evaluate the need for a standby on-site non-safety power source-during the FDA stage. The staff indicated it would review the
- applicant's position upon submittal of the USI/GSI update.
The applicant indicated that the ac/dc independent passive safety systems of the AP-600 resolve the issue of station blackout for that design.
SOURCE TERM Westinghouse indicated, and the staff concurred, that it would comply with 10 CFR part 100 and TID 14844 in its evaluations of DBAs for its standard designs, but that it intended to use a more realistic source term in its evaluation of-severe accidents.
PHYSICAL SECURITY Westinghouse stated that both the RESAR SP/90 and the AP-600 included consideration of satejuards concerns in the development of their layouts.
OBE/SSE The applicant stated that the OBE should not control the design of safety systems and that they intended to follow the guidelines of the EPRI Requirements Document. The OBE = 1/3 SSE for both facilities.
Resolution of this issue is planned during the FDA/DC review.
POST-TMI ISSUES Westinghouse indicated that it was in the process of updating Module 2 of the RESAR SP/90, which addresses post-TMI issues and would be submitting the revision shortly.
UNRESOLVED SAFETY ISSUES / GENERIC SAFETY ISSUES (USIs/GSIs)
The applicant indicated that it addressed all outstanding USIs and medium and high priority GSIs in May 1988 and that it would submit an update to Module 2 of the RESAR SP/90 shortly, which would address all USIs and GSIs, and their applicability to the plant design.
\\
.-________-_-__Q
PRA The applicant indicated that it performed a level 3 PRA (using a site envelope) for the RESAR SP/90 for internal events only.
Westinghouse stated it intended to perform an external events PRA at the FDA/DC stage of review. The core melt frequencyfortheRESARSP/90isagroximately1.5X10~6 and the probability
)
of a large release is about 3 X 10 The appigant stated that the core melt frequency for the AP-600 is approximately 1.2 X 10 (internaleventsonly). Westinghouse indicated they calculated no containment overpressurization due to their passive containment cooling systems, so the probability of a large release (due to an Event V sgquence and/or the steam generator tube rupture sequence) was about 3 X 10'.
stated they had looked at the effects of some external events, but they have not completed their analysis.
CORE-CONCRETE INTERACTION - ABILITY TO COOL CORE DEBRIS 2
Westinghouse stated that they will comply with the EPRI criteria of 0.02 m /MWt and that the design will include some method that would ensure automatic flooding of the lower cavity in the event of a severe accident. The applicant is currently evaluating alternative designs to ensure compliance with that commitment. The staff indicated it would evaluate Westinghouse's position upon submittal.
ATWS j
The applicant stated that they felt a diverse scram system was unnecessary for l
the RESAR SP/90 design because its moderator temperature coefficient is more I
negative tharr that for current plants, so the plant is less sensitive to an ATWS event as evidence in the contribution of an ATWS event to the core melt probability (5.5X10~g).
In addition, the applicant takes credit for manual boration by the operator to mitigate such an event. Westinghouse indicated it was concerned about the potential for increased spurious trips. The staff indicated that diversity may be necessary, but it would consider the i
applicant's arguments in its evaluation of the ATWS issue.
INTERSYSTEM LOCA 1
l Should the isolation valves fail, Westinghouse indicated that the design pressure of the piping outside of the containment will be sufficient to withstand primary side pressure or will be vented to the EWST.
The applicant further stated that should the seal on one of the low pressure pumps fail, the reactor vessel inventory could still be maintained through the remaining l
inventory of the EWST as well as the core reflood tanks and accumulators. The staff indicated it was still concerned about this issue, but would consider the applicant's arguments in its evaluation of this issue.
VENTING The RESAR SP/90 design does not include a containment vent.
Westinghouse indicated that they believe that containment venting can add as much risk as eliminate it, and that it is not required to meet the stated severe accident
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release;and dose goals for the RESAR SP/90' plant. The applicant ' stated. they.
lhave 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before the containment would irreversibly fail due tof X,
.overpressurization, which is sufficient time to restore containment cooling:.and prevent failure. Westinghouse. indicated they believed the operating purge-
. system may be. sufficient to prevent overpressurization, but the applicant was i:
still evaluating use'of the small (6")' system for that purpose. The staff' E
. indicated that it still. felt that containment venting. capability was desirable,
-but,it.would consider the applicant's arguments in its evaluation of this 1
issue.
-CONDITIONAL CONTAINMENT FAILURE c
Westinghouse indicated their concern that specification of a conditional L
. containment: failure probability may give priority to mitigation over_ prevention L
- of severe. accidents. Westinghouse stated they felt their designs-included-
" rugged" containments in conformance with the EPRI ALWR Requirements Document.
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- DIRECT CONTAINMENT HEATING l
' Westinghouse stated that the configuration 'of the cavity. of their containments L
will prevent. core > debris from entering the upper containment. This should resolve ~ the concern over direct containment heating.
In addition, the ac independent depressurizat~ ion system will reduce the probability of a high
~
pressure mo.lten core ejection from the r'eactor vessel.
MID-LOO.0 OPERATION' Westinghouse committed to~1nstall a vortex breaker at the RHR hot leg connection to significantly reduce air entrainment during mid-loop operation.
This feature, in conjunction with other design features of the plant, should eliminate concerns over midsloop operation.
EMERGENCY PLANNING L
Westinghouse stated that the design of the AP-600 supports consideration by the
'NRC of elimir,ation of evacuation and sheltering planning. The staff indicated
. reduction of emergency plann_may support such consideration, it did not forsee that even though the design operating plants require such planning and exercising.
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_ _______J