ML20035F278
| ML20035F278 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 04/15/1993 |
| From: | Wambuch T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9304210108 | |
| Download: ML20035F278 (37) | |
Text
..._ _
April i5, 1993 Docket No.52-002 APPLICANT: ABB-Combustion Engineering, Inc. (ABB-CE)
PROJECT:
CE System 80+
SUBJECT:
SUMMARY
OF SENIOR MANAGEMENT MEETING HELD ON MARCH 1, 1993, REGARDING STATUS OF SYSTEM 80+ REVIEW j
l A meeting was held at the Nuclear Regulatory Commission (HRC) offices in Rockville, Maryland, on March 1,1993. The list of attendees is provided as.
ABB-CE made a presentation of the material provided in Enclo-sure 2.
ABB-CE stated that they would be requesting a meeting with the Commissioners in late April or early May and would like the next senior i
management meeting just prior to that. ABB-CE expects that the final safety evaluation report will be available by the fall of 1993. The NRC stated that 4 to 5 months would be needed by the staff after closure of the open items.
1 Specific items were discussed, and the significant staff comments are dis-cussed below. The staff stated that a policy decision may be required if the l
mechanistic approach for intersystem loss-of-coolant accident (LOCA) resolu-l tion is pursued, rather than the " practical criteria" for piping design l
specified in SECY-90-016,. " Evolutionary Light Water Reactor (LWR) Certifica-l tion Issues and Their Relationship to Current Regulatory Requirements." The staff indicated that it was satisfied with the hardware resolution of the common-mode failure issue for instrumentation & control diversity provided that the reanalysis of the CESSAR-DC accidents under these conditions proved satisfactory.
For the common-moda failure event, the staff will consider the reliance on leak detection to protect against the large break (>12 inch diameter) LOCA's and main steamline breaks inside containment.
However, the staff advised that the leakage limit in the action statement of the technical j
specifications must be the same as the licensing basis for the most limiting l
lines (0.5 gpm). The use of an acceptance criterion of core coolability for l
large breaks requires additional discussion with the reactor systems branch.
A meeting is planned. The criteria of 10 CFR 50.46 will be used for other breaks (<12 inches). Additional meetings were planned for specific open items as indicated in Enclosure 2.
gg.
l Thomas V. Wambach, Project Manager Standardization Project Directorate i
Associate Directorate for Advanced Reactors I
and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
DISTRIBUTION w/o enclosures:
As stated ACRS (11)
GGrant, 17G21 TMurley/FMiraglia JMoore, 0GC MFranovich WBeckner, 10E4 cc w/ enclosures:
TBoyce MRubin, 8E23 RJones, 8E23 See next page JLyons, 8DI JWermiel, 8H3 MWaterman, 8H3 BBoger, 10H4 WRussell, 12G18 EJordan, MNBB 3701 l
BDLiaw, 7026 GBagchi, 7H15 J0'Brien, RES l
DISTRIBUTION w/ enclosures:
SMagruder
- Docket-File-PDST R/F DCrutchfield RBorchardt
~PDR TEssi TWambach PShea 0FC:
LA:PDST:ADAP, PM:6ST:ADAR SC:PDST:)QAR 04/6/93pk TEssig f 3 Fo3 TWambach:tz NAME: PSh e
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0FFICIAL RECORD _ COPY: MSM301.TVW 9304210108 930415 l
PDR ADOCK 05200002 A
ABB-Combustion Engineering, Inc.
Docket No.52-002 cc:
Mr. Regis A. Matzie, Vice President Huclear Systems Development Combustion Engineering, Inc.
1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 Mr. C. B. Brinkman, Acting Director Nuclear Systems Licensing Combustion Engineering, Inc.
1000 Prospect Hill Road Windsor, Connecticut 06095-0500 Mr. C. B. Brinkman, Manager Washington Nuclear Operations Combustion Engineering, Inc.
12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Stan Ritterbusch Nuclear Systems Licensing Combustion Engineering, Inc.
1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C.
20585 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Washington, D.C.
20503 Mr. Raymond Ng 1776 Eye Street, N.W.
Suite 300 Washington, D.C.
20006 Joseph R. Egan, Esquire Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W.
Washington, D.C.
20037-1128
MEETING ATTENDEES March 1, 1993 ORGANIZATION AFFILIATION Michael Franovich NRR/PDST William Beckner NRR/SPSB Tom Boyce NRR/PDST Mark Rubin NRR/SRXB J. E. Lyons NRR/DSSA/SPLB Robert C. Jones NRR/SRXB Jared Wermiel NRR/DRCH/HICB Mike Waterman NRR/DRCH/HICB Bill Borchardt NRR/PDST Thomas Wambach NRR/PDST Bruce Boger NRR/DRCH W. T. Russell NRR/ADT-Tom Murley NRR Dennis Crutchfield NRR/ADAR B. D. Liaw NRR/DE G. Bagchi NRR/DE/ECGB John O'Brien NRC/RES Michael T. Cross ABB-CE Lyle D. Gerdes ABB-CE l
Thomas D. Crom ABB-CE l
Fred L. Carpentino ABB-CE i
Charles B. Brinkman ABB-CE l
Regis A. Matzie ABB-CE J. Longo, Jr.
ABB-CE Peter M. Lang DOE l
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PROPOSED AGENDA NRC SENIOR MANAGEMENT MEETING
- OPENING REMARKS (NRC)- 1:00 PM l
- INTRODUCTION (ABB-CE) i
- PROGRAM STATUS (NRC/ABB-CE)
{
- ISSUES:
l
- ITAAC PROGRESS
- INTERSYSTEM LOCA
- 1&C DIVERSITY
- LEAK-BEFORE-BREAK LEAKAGE DETECTION
- STRUCTURAL DESIGN DETAIL
- INSERVICE TESTING A It It Nuclear Systems pggpgp
SEVERE ACCIDENT DETERMINISTIC ANALYSIS I
-lSSUE:
-RESPOND TO THE DECEMBER,1992, RAls WITHOUT RELYING ON FUTURE EXPERIMENTS (i.E, ADDRESS ISSUES BY DEMONSTRATING ROBUST DESIGN FEATURES)
-STATUS: STAFF CONCURRENCE ON:
-CONTAINMENT OVERPRSSURE ANALYSIS
-REACTOR CAVITY WALL ANALYSIS
-CORE-CONCRETE INTERACTION ANALYSIS 1
l 4
Ak IR ER l
Nuclear Systems 7%E9E9
DESIGN BASIS SAFETY ANALYSIS
-lSSUE:
-RE-ANALYZE CHAPTER 15 EVENTS TO REFLECT 1) METHODS AGREEMENTS WITH STAFF AND 2) INCREASED POWER LEVEL.
-STATUS:
-INCREASED POWER LEVEL HAS NO SIGNIFICANT IMPACT ON ANALYSIS RESULTS
-REMOVAL OF THREE-SECOND TIME DELAY BETWEEN TURBINE TRIP AND LOSS OF OFFSITE POWER HAS MADE ANALYSIS MUCH i
MORE DIFFICULT, BUT NO SHOW STOPPERS l
l il A IB Et Nuclear Systems pggppy
o NEW SOURCE TERM
-lSSUE:
-APPLICATION OF NEW SOURCE TERM TECHNOLOGY TO DOSE CALCULATIONS FOR DESIGN BASIS ACCIDENTS AND PROTECTIVE ACTION GUIDELINES t
-STATUS:
-ANALYSIS IS ON SCHEDULE
-TENTATIVE AGREEMENT ON ANALYSIS METHODS FOR DESIGN BASIS ACCIDENTS REACHED AS A RESULT OF THE JANUARY 29, 1993 MEETING WITH NRC STAFF j
-POTENTIAL DESIGN IMPACT: RECLASSIFICATION OF THE CURRENT SAFETY-GRADE ANNULUS VENTILATION SYSTEM AND REMOVAL OF CHARCOAL FILTERS
-lMPACT OF NEW SOURCE TERM ON SAFETY EQUIPMENT i
QUALIFICATON UNDER INVESTIGATION.
A It It Nuclear Systems 7%BDED o tea 2 ccmhvenon Engin
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SHUTDOWN RISK
-lSSUE:
-IDENTIFY INSIGHTS AND MODIFY DESIGN OR OPERATIONAL GUIDANCE ACCORDINGLY.
-STATUS:
-NRC RAls AND COMMENTS HAVE BEEN RESOLVED, INCLUDING FIRE AND FLOOD ANALYSIS.
l
-WITH ADDED INSTRUMENTATION AND OPERATIONAL RESTRICTIONS, SHUTDOWN RISK IS ONLY 30% OF THE TOTAL CORE DAMAGE FREQUENCY.
-SHUTDOWN RISK REPORT AND RAI RESPONSES WILL BE PRINTED AS PART OF THE PRA CHAPTER IN CESSAR-DC.
4 i
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- i
REACTOR COOLANT PUMP SEAL COOLING
-ISSUES:
4
-RELIABILITY OF SEAL COOLING DURING A STATION BLACKOUT 1
-SUSCEPTIBILITY TO INTERSYSTEM; LOCA FROM HIGH PRESSURE SEAL COOLER TUBE FAILURE THROUGH THE COMPONENT COOLING WATER SYSTEM
-STATUS:
-ADDED A HIGHLY RELIABLE, DIVERSE CHARGING PUMP WHICH l
CAN BE POWERED FROM EITHER EMERGENCY DIESELS OR THE i
COMBUSTION TURBINE GENERATOR
-ADDED OVERPRESSURE PROTECTION TO THE COMPONENT l
COOLING WATER SYSTEM l
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Nuclear Systems 7%EDED l
- --- 1" ~1o
e ELECTRICAL SYSTEMS REVIEW
-lSSUES:
-RESOLUTION OF DSER ISSUES
-RESOLUTIONS OF NEW NRC CONCERNS ON LEVEL-OF-DESIGN DETAIL AVAILABLE (E.G., VOLTAGE ANALYSIS, CHANNEL SEPARATION, BATTERY SIZING.
-STATUS:
-APPROACH TO RESOLUTION AGREED ON FEBRUARY 16.
-NRC WILL IDENTIFY CLOSE AND CONFIRMATORY DSER ITEMS BY THE FIRST WEEK OF MARCH.
-ABB-CE/DESI WILL REVISE DSER RESPONSES TO ADDRESS STATED NRC CONCERNS, INCLUDING ITAAC.
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= = = ~
I&C DIVERSITY
-ISSUES:
-METHODS FOR ANALYSIS OF ACCIDENTS WITH A COMMON MODE FAILURE
-DESIGN OF DIVERSE HARDWIRED BACKUP CONTROLS
-STATUS:
-SELECTION OF ANALYSIS METHODS IS ONGOING l
-STAFF CONCURS WITH THE USE OF A SINGLE PROGRAMMABLE LOGIC CONTROLLER IN EACH CONTROL PATH AT THE LAST LOOP CONTROLLER.
I i
i A H Et N uclear Sys': ems 7%EDED
MANAGEMENT MONITORING ISSUES
-FIRE PROTECTION
-SABOTAGE ANALYSIS
-IN-SERVICE TESTING
-TECHNICAL SPECIFICATIONS
-ITAACs
-BALANCE OF PLANT SYSTEMS
-STRUCTURAL ANALYSIS A ED Et Nuclear Systems 7%EDED
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I RECENT SUBMil I AL COMPLETED i
i
-JANUARY 15 -
INITIAL REPORT ON NEW j
SOURCE TERM
-JANUARY 28 -
DRAFT ITAACs
-FEBRUARY 1 -
DRAFT ITAACs l
-FEBRUARY 2 -
DS.ER RESPONSES COMPLETE I
l
-FEBRUARY 9 -
SAFETY ANALYSIS RESULTS -
~
PART 1 f
-FEBRUARY 16 -
RESPONSES TO SEISMIC
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MARGINS METHODS "RAls" 1
-FEBRUARY 24 - REVISED LEVEL 1 PRA INCLUDING SHUTDOWN RISK A BB R l
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SUBMITTAL SCHEDULE (CONT'D) i 4
-MARCH 3-SHUTDOWN RISK REPORT AND i
RAIRESPONSES i
i
-MARCH 4 -
SEVERE ACCIDENT ANALYSIS REPORT AND RAI RESPONSES
-MARCH 5 -
COMMON MODE FAILURE l
ANALYSIS METHODS REPORT
-MARCH 5-CESSAR-DC, AMENDMENT L l
-MARCH 8 -
OPERATING EXPERIENCE l
REVIEW PROCESS i
-MARCH 8 -
PROTOTYPE ITAACs AND I
ROADMAPS A It Et Nuclear Systems pgEpEp T-"E
SUBMITTAL SCHEDULE (CONT'D) l
-MARCH 9 -
SAFETY ANALYSIS RESULTS -
PART 2 i
i
-MARCH 10 -
CESSAR-DC, AMENDMENT M i
j
-MARCH 31 -
PROTECTIVE ACTION GUIDELINE DOSE CALCULATION I
-APRIL 1 -
HUMAN FACTORS ENGINEERING INSTRUMENTATION SURVEY
- APRIL 19 -
CESSAR-DC, AMENDMENT N COMPLETE ITAACs i
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1 ANALYSES AVAILABLE FOR AUDIT 1
- MAY 31 -
SOURCE TERM CODE USER DOCUMENTATION AND VERIFICATION
-JUNE 30 - STRUCTURAL DESIGN DETAll I
-JUNE 30 - SEISMIC MARGIN ASSSESSMENT i
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A ED Et Nuclear Systems 7% EDEN
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INTERSYSTEM LOCA
-lSSUE:
-THE PROPOSED ABB-CE APPROACH TO ISLOCA SUBMITTED IN RESPONSE TO THE SER QUESTIONS WAS NOT ACCEPTABLE TO THE NRC STAFF
-THIS APPROACH WAS BASED ON MINIMlZlNG OFF-SITE DOSES AND A MECHANISTIC RESOLUTION.
-THE PRIMARY STAFF FOCUS WAS BASED ON MINIMlZING UNDETECTED LOSS'OF RCS COOLANT INVENTORY AND WAS BASED ON THEIR RESOLUTION WITH GE.
-THE STAFF APPROACH WOULD BE MORE DETRIMENTAL TO THE PWR DESIGN (HIGHER RCS PRESSURE) AND COULD NOT BE DEFENDED (IN OUR JUDGEMENT) DUE TO PRACTICALITY TEST.
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- INTERSYSTEM LOCA (CONT'D) 1
-RESOLUTION HIGHLIGHTS:
-ABB-CE AND THE NRC STAFF HAVE HAD CONTINUING DISCUSSIONS ON THIS SUBJECT.
-AT A 2/22 MEETING, A PROCESS WAS DETAILED THAT APPEARS TO HAVE HIGH POTENTIAL FOR SUCCESS.
-THE BASIC STEPS IN THIS PROCESS ARE:
-ACCEPTING PREMISE OF POTENTIAL FOR LEAKING AND/OR i
INADVERTENT OPERATION OF ISOLATION VALVES.
l
-ACCEPTING PREMISE OF PROBABILITY ARGUMENT FOR i
LIMITING SCOPE OF CHANGES.
I
-FOR THOSE SYSTEMS (ESPECIALLY THOSE DIRECTLY l
CONNECTED TO RCS) PROVIDING AN UPGRADE TO STRUCTURAL INTEGRITY.
i A It It i
Nuclear Systems 7%EDED
- .=- -
INTERSYSTEM LOCA (CONT'D) l
-RESOLUTION SCHEDULE:
-ABB-CE WILL SUBMIT A TEST CASE TO NRC STAFF BY 3/15/93.
THIS CASE WILL ATTEMPT TO INCLUDE ALL INGREDIENTS OF PROPOSED SOLUTION.
-NRC STAFF WILL REVIEW TEST CASE FOR APPROVAL AND l
PROVIDE ABB-CE WITH COMMENTS BY 3/22/93.
-lN PARALLEL WITH NRC REVIEW, ABB-CE WILL CONTINUE ITS DESIGN UPGRADE AND DOCUMENTATION AND TARGET l
(SUBJECT TO NRC COMMENTS ON TEST CASE) A 3I31/93 i
SUBMITTAL.
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INTERSYSTEM ILLUSTRATION RCS
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IN CONJUNCTION WITH A LIMITING FAULT IS A BEYOND DESIGN BASIS EVENT i
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MARCH 1, 1993 9
4 6
ABB APPROACH FOR DEMONSTRATING CAPABILITY 0F THE DIVERSE EQUIPMENT TO COPE WITH A C0H40N MODE SOFulARE FAILURE:
(1)
ASSUME THAT THE POSTULATED CMF FAILS ALL AUTOMATIC AND MANUAL CONTROL FUNCTIONS OF THE PLANT PROTECTION SYSTEMS.
l (2)
ASSESS THE CAPABILITY TO COPE WITH TRANSIENTS AND ACCIDENTS USING THE REMAINING AVAILABLE MEANS.
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Nuplex 80+TM Defense-in-Depth (Current)
Success Path Critical Function Non-Safety Safety Manual Reactivity Control Rod Control, CVCS (Boration)
Safety injection System, Reactor Trip Reactor Trip Breakers Breakers Vital Auxiliaries AC Main Transformer, Alternate A/C Emergency Diesels, Aux Transformers DC Station Battery Station Battery RCS Inventory Control CVCS (Charging / Letdown)
Safety injection System Safety injection System RCS Pressure Control Hee:ers/ Spray, Safety injection System, Safety injection CVCS (Charging)
Safety Depressurization System System Core Heat Removal Forced Circulation Natural Circulation RCS Heat Removal Main Feed Emergency Feed, Emergency Feed Shutdown Cooling &
Safety injection Safety injection System Containment isolation Control Valves isolation Valves isolation Valves Containment Environment Fan Coolers, H, Ignitors Containment Spray, Cont. Spray Recombiners Radiation Emission Monitor and Copntrol Radiation isolation of Release Isolation Valves Releaso Paths Paths N EEEE F
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i RESULTS OF THE DIVERSITY AND DEFENSE IN DEPTH EVALUATION REPORT 28 TRANSIENTS Tr ACCIDENTS WERE EVALUATED USING CONSERVATIVE EVALUATION METHODS AND ACCEPTANCE CRITERIA:
l (1)
FOR 19 EVENTS, THE DIVERSITY IN NUPLEX 80+
WAS SHOWN TO BE ADEQUATE.
THE LAWRENCE LIVERMORE REVIEW CONCURRED WITH THIS CONCLUSION.
(2)
FOR 9 EVENTE, FURTHER EVALUATION WAS REQUESTED.
L 9
THE 9 EVENTS ARE:
i 1
EVENT CLASSIFICATION
{
EVENT WITHOUT CMF 1.
LOSS OF FLOW MODERATE FREQUENCY I
2.
SHEARED SHAFT LIMITING FAULT 3.
SHAFT SEIZURE LIMITING FAULT l
4.
CEA EJECTION LIMITING FAULT 5.
LETDOWN LINE BREAK LIMITING FAULT j
6.
STEAM GENERATOR TUBE LIMITING FAULT RUPTURE 7.
MAIN STEAM LINE BREAK LIMITING FAULT 8.
FEEDWATER LINE BREAK LIMITING FAULT 9.
LOSS OF COOLANT ACCIDENT LIMITING FAULT i
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(1)
EVALUATED USING BEST-ESTIMATE ASSUMPTIONS OPERATING PARAMETERS i
EQUIPMENT OPERABILITY BREAK SIZES / LOCATIONS OPERATOR RESPONSE (2)
USING THE FOLLOWING ACCEPTANCE CRITERIA l
o 100% OF 10CFR100 FOR LOSS OF FLOW
[
o CORE C00 LABILITY PER NUREG-0562 FOR NON-LOCAs PER 10CFR 50.46 FOR LOCAs l-o CONTAINMENT PRESSURE ASME SERVICE LEVEL C i
-. b
LEAK-BEFORE-BREAK LEAKAGE DETECTION 4
-lSSUE:
-NEED FOR TECHNICAL SPEC FOR 0.5GPM RCS LEAKAGE IN SUPPORT OF LBB I
-STATUS:
l
-ABB-CE RECOMMENDING LEAK DETECTION METHODOLOGY BASED ON PLANT OPERATING PROCEDURES, CONSISTENT l
WITH PRIOR NRC APPROVALS.
I
-NRC IN PROCESS OF REVIEWING APPLICABILITY OF PRIOR APPROVALS TO SYSTEM 80+
i
-NRC TO RESPOND TO THE PROPOSED APPROACH PRESENTED AT 2/8/93 MEETING WITH ABB-CE A D Et l
Nuclear Systems pggDED l
STRUCTURAL DESIGN DETAIL
-lSSUE:
-PRESENT SUFFICIENT DESIGN DETAll TO PROVIDE STAFF CONFIDENCE IN ADEQUACY OF STRUCTURAL DESIGN IN THEIR SAFETY EVALUATIONS i
-STATUS:
-DETAIL DESIGN PROGRESSING IN ACCORDANCE WITH PLAN AND SCHEDULE
-NEXT MEETING WITH STAFF PLANNED FOR MARCH l
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Nuclear Systems 7%EDED 4
SCHEDULE l
-1/30/93 COMPLETED THE 3-D MODEL OF THE 4
NUCLEAR ISLAND STRUCTURES i
-2/30/93 IDENTIFIED GLOBAL LOADS AND l
COMBINATIONS
-2/15/93 COMPLETED EVALUATION OF REACTOR CAVITY STRUCTURAL CAPACITY FOR l
SEVERE ACCIDENTS l
l l
-2/26/93 COMPLETE LOADING ANALYSES IN SUPPORT OF WALL / FLOOR DESIGN l
l
-3/93 MEETING WITH NRC STAFF l
AHR l
Nuclear Systems pggggy l
.-_~-
SCHEDULE (CONT'D) i
-3/15/93 PRELIMINARY DETAILS FOR WALLS, i
FLOORS, BASEMAT AND CONNECTORS i
FOR CRITICAL AREAS
-3/31/93 COMPLETE LOADING ANALYSIS IN l
SUPPORT OF BASEMAT DESIGN t
-3/31/93 MEETING WITH NRC STAFF; CONFIRM CRITICAL DESIGN AREAS; AGREE ON i
CESSAR-DC CONTENT.
l A It Et Nuclear Systems 7%EDED j
1_ _ _
INSERVICE TESTING (IST)
PROGRAM
-ISSUE:
-DEVELOP IST FOR DESIGN CERTIFICATION
-STATUS:
-PRELIMINARY EXAMPLE OF IST PROGRAM SUBMITTED TO NRC i
STAFF 1/15/93 FOR COMPONENT COOLING WATER AND l
i
-NRC STAFF HAS REQUESTED, VIA CONFERENCE CALL 2/18/93, l
FURTHER REFINEMENTS TO THE IST PROGRAM AND l
CESSAR-DC SECTION 3.9.6.
j
-LATTER ITEM REQUESTED, INCLUDING SUPPORTING TIER 2 l
INFORMATION, REQUIRED FOR TIER 1 IN SECTION 3.9.6.
i i
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Nuclear Systems 7%EDED i
INSERVICE TESTING PROGRAM (CONT'D)
-PROPOSED PLAN FOR RESOLUTION l
-SUBMIT IST PROGRAM FOR ALL SYSTEMS STAFF FOR REVIEW i
WITH THE FOLLOWING ADDITIONS BY MARCH 15:
[
-JUSTIFICATIONS FOR NON-QUARTERLY VALVE TESTING
-IDENTIFICATION OF CHECK VALVES WHICH HAVE SAFETY-RELATED REVERSE FLOW FUNCTION
-TYPICAL TEST CONNECTIONS FOR PUMPS, CHECK VALVES, CONTAINMENT ISOLATION VALVES, AND POWER OPERATED l
VALVES.
-SCHEDULE MEETING MARCH 22,1993 TO RESOLVE OUTSTANDING ISSUES TO 3/15/93 SUBMITTAL.
l
-lST PROGRAM TO BE ADDED TO CESSAR-DC AS TABLE 3.9-15.
l
-SUPPORTING TIER 2 INFORMATION REQUIRED FOR TIER 1 l
MATERIAL TO BE ADDED TO SECTION 3.9.6 BASED ON l
INDUSTRY / REGULATORY CONSENSUS.
i A R El l
Nuclear Sys': ems 7% EDEN l
SYSTEM 80+ ITAAC ACTIVITY JANUARY-FEBRUARY 1993 1
-JAN.11-22 PARTICIPATED IN NRC TEAM REVIEW OF TEN ABWR ITAAC 1
-JAN. 28 SUBMITTED 20 DRAFT SYSTEM ITAAC f
- FEB.1 SUBMITTED 27 DRAFT SYSTEM l
- FEB.1 -5 INDUSTRY REVIEW OF TEN SYSTEM i
ITAAC IN WINDSOR l
l
- FEB. 8-12 INDUSTRY REVIEW OF TEN SYSTEM l
ITAAC IN CHARLOTTE l
l A It It Nuclear Sys': ems 7%EDID l
- = = -
SYSTEM 80+ ITAAC ACTIVITY JANUARY-FEBRUARY 1993
-FEB.11 RECEIPT OF CRUTCHFIELD LETTER SUGGESTING CHANGE IN SYSTEM l
80+ ITAAC SUBMITTALS TO ACCOMMODATE THE SAN JOSE
/ WINDSOR / CHARLOTTE REVIEWS AND TO OBTAIN NRC TEAM REVIEW OF. PROTOTYPE ITAAC.
l
- FEB. 24-25 MANAGEMENT REVIEW OF ELEVEN l
PROTOTYPE ITAAC i
- FEB. 24 NRC SUGGESTED MODIFICATION l
OF SUBMITTAL DATE i
l A It II l
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1 WINDSOR & CHARLOTTE INDUSTRY INPUT t
-TEN TWELVE-HOUR DAYS OF REVIEW
-PARTICIPANTS:
-9 UTILITIES
-4 ARCHITECT / ENGINEERS
-lNPO l
-EPRI l
-NUMARC
-DOE
-SHAW PITTMAN 1
i
-NRC STAFF
-ElGHTEEN SYSTEMS REVIEWED USING CURRENT GUIDANCE i
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- " ~
MANAGEMENT REVIEW OF PROTOTYPE ITAAC i
-TWO FIFTEEN-HOUR DAYS
-PARTICIPANTS:
)
-ABB-CE MANAGEMENT
{
-DE&S i
j
-SWEC l
-APS
-NUMARC i
-SHAW PITTMAN l
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-ELEVEN SYSTEM ITAACs REVIEWED i
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A It It Nuclear Systems 7%EDED
PROTOTYPE ITAAC SUBMITTAL
-SHUTDOWN COOLING SYSTEM l
-REACTOR COOLANT SYSTEM
-REACTOR VESSEL INTERNALS i
-CONDENSER CIRCULATING WATER SYSTEM
-PLANT PROTECTION SYSTEM
-POTENTIAL OPEN ISSUE: DIVERSITY ANALYSIS l
l l
i A It 59 l
Nuclear Systems pggggy