Appendix A

209 Pages·2009·6.75 MB·English

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NUREG-1503
Vol.2
Final Safety Evaluation Report
Related to the Certification
of the Advanced Boiling Water
Reactor Design
Appendices
Manuscript Completed: July 1994
Date Published: July 1994
Associate Directorate for Advanced Reactors and License Renewal
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
MASTER .
Di3Ti-*i»U VKX* '>> TJ-iit DOC'JMEN7 m UKLiMlTZD
ABSTRACT
This safety evaluation report (SER) documents the ABWR design features that are substantially the same as
technical review of the U.S. Advanced Boiling Water those previously considered. The SERs for the other BWR
Reactor (ABWR) standard design by the U.S. Nuclear designs have been published and are available for public
Regulatory Commission (SRC) staff. The application for inspection at the NRC Public Document Room,
the ABWR design was initially submitted by the General 2120 L Street, N.W., Washington, D.C. 20037. Unique
Electric Company, now GE Nuclear Energy (GE), in features of the ABWR design include internal recirculation
accordance with the procedures of Appendix O of Part SO pumps, fine-motion control rod drives, microprocessor-
of Title 10 of the Code of Federal Regulations (10 CFR based digital logic and control systems, and digital safety
Part 50). Later GE requested that its application be systems.
considered as an application for design approval and
subsequent design certification pursuant to On the basis of its evaluation and independent analyses, the
10 CFR § 52.45. NRC staff concludes that, subject to satisfactory resolution
of the confirmatory items identified in Section 1.8 of this
SER, GE's application for design certification meets the
The U.S. ABWR design is similar to the international requirements of Subpart B of 10 CFR Part 52 that are
ABWR design, which was being built at the Kashiwazaki applicable and technically relevant to the U.S. ABWR
Kariwa Nuclear Power Generation Station, at the time of standard design. A copy of the report by the Advisory
the staffs review, by the Tokyo Electric Power Company, Committee on Reactor Safeguards required by
Inc. The ABWR is a single-cycle, forced-circulation, 10 CFR S 52.53 is provided in Chapter 21. A final design
boiling water reactor (BWR) with a rated power of approval, issued on the basis of this SER, does not
3926 megawatts thermal (MWt) and a design power of constitute a commitment to issue a permit or license, or in
4005 MWt. Many features of the ABWR design are any way affect the authority of the Commission, the
similar to those of BWR designs that the staff had Atomic Safety and Licensing Board, and other presiding
previously approved. To the extent feasible and officers, in any proceeding pursuant to Subpart G of
appropriate, the staff relied on earlier reviews for those 10 CFR Part 2.
iii NUREG-1503
CONTENTS
VOLUME 2: APPENDICES
Page
APPENDIX A A-l
LIST OF ABBREVIATIONS A-l
APPENDIX B B-l
REFERENCES B-l
APPENDIX C C-l
CHRONOLOGY OF CORRESPONDENCE C-l
APPENDIX D D-l
FSER CONTRIBUTORS D-l
APPENDIX E E-l
STAFF POSITION ON SHELL BUCKLING DUE TO INTERNAL PRESSURE E-l
APPENDIX F F-l
STAFF POSITION ON STEEL EMBEDMENTS F-l
APPENDIX G G-l
STAFF POSITIONS AND TECHNICAL BASES ON THE USE OF AMERICAN NATIONAL
STANDARDS INSTITUTE (ANSI)/AMERICAN INSTITUTE OF STEEL CONSTRUCTION
(AISC) N690, "NUCLEAR FACEJTIES - STEEL SAFETY-RELATED STRUCTURES" G-l
APPENDIX H H-l
DYNAMIC LATERAL SOIL PRESSURES ON EARTH RETAINING WALLS AND EMBEDDED
WALLS OF NUCLEAR POWER PLANT STRUCTURES H-l
APPENDIX I 1-1
EVALUATION OF ABWR PUMP AND VALVE INSERVICE TESTING PLAN
(SSAR TABLES 3.9-8 AND 3.9-9) 1-1
APPENDIX J M
HUMAN FACTORS ENGINEERING PROGRAM REVIEW MODEL AND ACCEPTANCE CRITERIA FOR
EVOLUTIONARY REACTORS J-l
APPENDIX K K-l
IMPORTANT SAFETY INSIGHTS K-l
NUREG-1503
FIGURES
Page
G-l Structural stability curves for the axially loaded compression numbers 0-3
J.l HFE program review model elements J-4
J.2 HFE program review stages J-5
NUREG-1503
vi
Appendix A
LIST OF ABBREVIATIONS
A
ABWR advanced boiling water reactor CA compressed air
AC alternating current CAMS containment atmosphere monitoring system
ACC accumulator CAV cumulative absolute velocity
ACI automatic closure and interlock CBERU contrbuilding emergency recirculation unit
ACIWA ac-independent water addition CBRU control-building recirculation unit
ACRS Advisory Committee on Reactor CBSREA control building safety-related equipment
Safeguards area
ACS atmospheric control system CCDF complementary cumulative distribution
ACU air conditioning units function
ADS automatic depressurization system CCI core concrete interaction
AHU air handling units CCS condensate cleanup system
AISC American Institute of Steel Construction CDF core damage frequency
ALARA as low as is reasonably achievable CDM Certified Design Material
ALWR advanced light water reactor CDRL core damage radiation level
AM accident management CE ABB-Combustion Engineering
ANL Argonne National Laboratory CED common engineering documents
ANS American Nuclear Society CET containment event trees
ANSI American National Standards Institute CF&CAE condensate feedwater, and condensate air
AOO(s) anticipated operational occurrences extraction
APR automatic power regulator CFR Cofle of refera] Regulations
APRM average power range monitor CFS condensate and feedwater system
ARI alternate rod insertion CH chugging
ARS amplified response spectra CIV containment isolation valve
ASB Auxiliary Systems Branch cm centimeters
ASCE American Society of Civil Engineers CMAA Crane Manufacturers Association of
ASD allowable stress design America
ASD adjustable speed drive CMP configuration management plan
ASF automatic suppression function CMU control room multiplexing unit
ASHRAE American Society of Heating, CO condensation oscillation
Refrigeration, and Air Conditioning COL combined license
Engineers COPS containment overpressure protection
ASME American Society of Mechanical Engineers system
ASTM American Society for Testing and CPG containment performance goal
Materials CP construction permit
ATIP automatic transversing in-core probe CPU central processing unit
ATLM automated thermal limit monitor CR control room
ATWS anticipated transient without scram CRHA control room habitability area
CRD control rod drive
— B — CRDS control rod drive system
CRHA control room habitability area
CRT cathode-ray tube
BNL Brookhaven National Laboratory CS control system
BPU bypass unit CS core support
BPWS blanked position withdrawal sequence CS crown and segment technique
BTP Branch Technical Position CSNI Committee on the Safety of Nuclear
BWR boiling water reactor Installations
BWROG BWR Owners Group CST condensate storage tank
A-l NUREG-1503
Appendix A
EDO emergency diesel generator
EDO Executive Director for Operations
CT completion times EF error factor
CTG combustion turbine generator EFU emergency filtration unit
cuw
reactor water cleanup EHR extra hard rock
CVCF constant voltage constant frequency EMC electronic magnetic compatibility
CWS circulating water system EMI electromagnetic interference
EMS essential multiplexing system
D EOF Emergency Operations Facility
EOP emergency operating procedures
EPA electrical protection assemblies
DAC design acceptance criteria EPG emergency procedure guidelines
DAL design action list EPRI Electric Power Research Institute
DBA design-basis accident ERM Engineering Review Memorandum
DBLOCA design-basis loss-of-coolant accident EQ environmental qualification
DBT design basis tornado EQD environmental qualification document
DC design certification ESD electrostatic discharge
DCH direct containment heating ESF engineered safety feature
DD design description ESW essential service water
DCC damage control center
DCD design control document
DCM damage control measure
DCM Tier 1 Design Certification Material for
tneOEABWR FATT fracture appearance transition temperature
DCV drywell connecting vent FCI fuel-coolant interactions
DEPSS drywell equipment and piping support FCS flammability control system
structure FCU fan cool unit
DET decomposition event tree FDA final design approval
DF decontamination factor FDDI fiber distribution data interface
DFSER draft final safety evaluation report FDWC feedwater control
DG diesel generator FDDI fiber distributed data interface
DGCW diesel generator cooling water FIST full integral simulation test
DGL diesel generator lubrication FIVE fire-induced vulnerability evaluation
DGSA diesel generator starting air FMCRD fine-motion control rod drive
DMC digital measurement and control FMEA failure modes and effects analysis
DOD Department of Defense FOST fuel oil storage and transfer
DOE Department of Energy FPC fuel pool cooling and cleanup
dp deltap FRS floor response spectra
DRAP design reliability assurance program FS full-scale
DSER draft safety evaluation report FSER final safety evaluation report
DSIL drywell spray initiation limit ft feet
DTM digital trip module FWLB feedwater line break
DTS drain transfer system
G
GDC general design criteria/criterion
GE GE Nuclear Energy
EAB exclusion area boundary GI generic issue
EB electrical building GL generic letter
EBVS electrical building ventilation system GSI generic safety issue(s)
ECCS emergency core cooling system(s) GWd gaseous waste management system
NUREG-1503 A-2
Appendix A
H
HCLPF high confidence low probability of failure LBB leak-before-break
HCTL heat capacity temperature limit LCS leakage control system
HCW high-conductivity waste LCS local control switches
HCU hydraulic control unit LCW low-conductivity waste
HECW HVAC emergency cooling water LD lower drywell
HELB high-energy line breaks LDF lower drywell flooder
HELSA high-energy line separation analysis LDS leak detection and isolation system
HEPA high-efficiency particulate air LER licensee event reports
HF human factors LLHS light load handling system
HI hydraulic institute LLNL Lawrence Livermore National Laboratory
HFE human factors engineering LLRT local leak rate tests
HFPP human factors program plan LOCA loss-of-coolant accident
HIC high-integrity containers LOOP loss-of-offsite power
HNCW HVAC normal cooling water LOPP loss-of-preferred power
HPCF high-pressure core flooder LPCI low-pressure coolant-injection
HPCS high-pressure core spray LPFL low-pressure flooder
HPIN high-pressure nitrogen gas supply LPMS loose parts monitoring system
HPME high-pressure core melt ejection system LPRM local power range monitor
HR hard rock LPZ low-population zone
HRA human reliability analysis LRB Licensing Review Bases
HSD hot shower drain LRFD load and resistance factor design
HSI human system interface LTS long term solutions
HVAC heating, ventilating, and air conditioning LVDT linear variable differential transformers
HWC hydrogen water chemistry LWMS liquid waste management system
HWH hot water heating LWR light water reactor
I M
I&C instrumentation and control m meters
IA instrument air M-O Mononobe and Okabe
IBD instrument block diagrams MACCS melcor accident consequence code system
ICC inadequate core cooling MAPLHGR maximum average planar linear heat
ICD interface control diagram generation rate
IE Inspection and Enforcement MC main condensers
IED improvised explosive device MCAE main control area envelope
IEEE Institute of Electrical and Electronics MCC motor control center
Engineers MCES main condenser evacuation system
IGSCC intergranular stress corrosion cracking MCPR minimum critical power ratio
ILRT integrated leakage rate tests MCR main control room
IN information notice MEB Mechanical Engineering Branch
in. inch MG motor-generator
IORV inadvertent open relief valve ML manufacturing license
ISI inservice inspection MOV(s) motor operated valves
ISM independent support motion MPL(s) master parts lists
ISO isometric drawing(s)/International Systems MPT main power transformer
Interconnection MRBM multi-channel rod block monitor
1ST inservice testing MS main steam
ITAAC inspections, tests, analyses, and acceptance MSIV main steamline isolation valves
criteria MSL main steamline
ITP initial test program' MSLB main steamline break
A-3 NUREG-1503
Appendix A
— M —
PCP process control program
PCPL primary containment pressure limit
MST main steam tunnel PCT peak cladding temperature
MTU metric ton of uranium PDA preliminary design approval
MUWC makeup water (condensate) PFD process flow diagrams
MUWP makeup water system (purified) PGA peak ground acceleration
MVA megavolt amps PGCS power generation control system
MWP makeup water system PIP plant investment protection
Mwt megawatt thermal PM preventive maintenance
POV powered operated valves
— N — PRA probabilistic risk assessment
PRC Piping Review Committee
PRM process radiation monitor
NB nuclear boiler PRM Program Review Model
NBS nuclear boiling system PRMS process radiation monitoring system
NDE non-destructive examination PRNM power range neutron monitor
NEMA National Electrical Manufacturers PS pressed and spun
Association PSA probabilistic safety assessment
NEMS non-essential multiplexor system PSB Power Systems Branch
NFPA National Fire Protection Association PSD power spectrum density
NMS neutron monitoring system PSDF power spectral density function
NNS non-nuclear safety PSI preservice inspection
NPB nuclear power block PSIS pounds per square inch gauge
NPP nuclear power plant PSS process sampling system
NPSH net positive suction head PSTF pressure suppression test facility
NQA Nuclear Quality Assurance
NRC Nuclear Regulatory Commission
- Q-
NRD nonradioactive drain
NRHX non-regenerative heat exchangers
NSSFC National Severe Storm Forecast Center Q question
NSSS nuclear steam supply system(s) QA quality assurance
QG quality group
— O —
— R —
OBE operating basis earthquake
OER operating experience review RAI request for additional information
OHLHS overhead heavy load handling system RAP reliability assurance program
OLU output logic unit RB reactor building
OL operating license RBM rod-block monitor
OM operations and maintenance RBV reactor building vibration
OPRM oscillation power range monitor RBVS reactor building ventilation system
O-RAP operational reliability assurance process RCCV reinforced concrete containment vessel
ORNL Oak Ridge National Laboratory RCIC reactor core isolation cooling
OSC Operational Support Center RCIS rod control and information system
OSI open systems interconnection RCPB reactor coolant pressure boundary
RCS reactor coolant system
— P — RCW reactor building cooling water
RCWS reactor building cooling water system
RFC recirculation flow control
RFCS recirculation flow control system
P&ID • piping and instrumentation diagram(s) RG Regulatory Guide
PASS post-accident sampling system RH relative humidity
PCHS power cycle heat sink RHR residual heat removal
NUREG-1503 A-4