ML20070H969
| ML20070H969 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 07/31/1994 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20070H967 | List: |
| References | |
| 23A6100, 23A6100-R07-A35, 23A6100-R7-A35, NUDOCS 9407220188 | |
| Download: ML20070H969 (54) | |
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d Amendment 35-Revision 7 - Page Change Instruction The following pages have been changed, please make the specified changes in your SSAR. Pages are listed as page pairs (front and back).
REMOVE PAGE REMOVE PAGE No.
ADD PAGE NO.
No.
- ADD PAGE NO.
Replace COVER page in front of each ABWR TAB 9.2 binder 9.2-51, 52 9.2 -51, 52 CIIAI M 1 TAB 9.3 TAB 1.6 1.6 - 1 thru 6 1.6 - 1 thru 6 CHAPTER 19 TAB 1.8 1.8-37 thru 48 1.8 -37 thru 48 19A - 7, 8 19A - 7, 8 TM1A TAB Aon.19B 1A-7 thru 10 1A-7 thru 10 CllAPTER 3 19B - 5, 6 19B - 5, 6 O
19B - 123,124 19B -123,124 TAB 3.7 3.7-37,38 3.7-37,38 3.7-47,48 3.7-47,48 19E.2 - 159,160 19E.2 - 159,160 19E.2 - 175,176 19E.2 - 175,176 g
3.8-51,52 3.8-51, 52 TAE Apo. 3A
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21-311 21-311 SA - 25, 26 SA - 25, 26.21-336 21-336 CIIAPTFR 7 TAB 7.6 7/, -47,48 7.6-47,48 TAB 7.7 7.7-69,70 7.7-69,70
- Figure 7.7-7 Sheet 2
' 7.7-93,94' 7.7-93,94 CIMrER 9 TAB 9.1 9.1-15,16 9.1 - 15,16 (1)
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'k 23A6100 Rev.1 ABWR standardsatoryAnalysis nepois
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1.6 Materialincorporated by Reference Table 1.6-1 is a list of all GE topical reports and any other report or document which is incorporated in whole or in part by reference in the ABWR SSAR.
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I Materialincorporated by Reference - Amendment 31 1.61
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23/6100 Rev. 7 ABWR StandardSafetyAnalysis Report O
Table 1.6-1 Referenced Reports ABWR SSAR Report No.
Title Section No.
I 22A7007 GESSAR II, 238 Nuclear Island, BWR/6 Standard Plant, General 3.7 l
Electric Company, March 1980, & Amendments 1-21.
19.2 19.3 19D.3 19D 7 19E.2 19E.3 APED-5750 Design and Performance of General Electric Boiling Water 5.4 Reactor Main Steam Line Isolation Valves, General Electric Company, Atomic Power Equipment Department, March 1969.
NEDO-10029 An Analytic Study on Brittle Fracture of GE-BWR Vessel Subject 5.3 l
to the Design Basis Accident, June 1969.
NEDO-10299A H.T. Kim, Core Flow Distribution in a Modern BWR as Measured 4.4 in Monticello, October 1976.
NEDO-10527 C.J. Paone and J.A. Woo \\ ley, Rod Drop Accident Analysis for 15.4 Large Boiling Water Reactors, Licansing Topical Report, March 1972.
NEDO-10585 F.G. Brutchscy, et al., Behavior ofIodine in Reactor Water 15.2 During Plant Shutdown and Startup, August 1972.
NEDO-10722 H.T. Kim, Core Flow Distribution in a Large Boiling Water 4.4 l
Reactor as Measured in Quad Cities Unit 1, December 1972.
l NEDO-10722A H.T. Kim, Core Flow Distribution in a Large Boiling Water 4.4 Reactor as Measuredin Quad Cities Unit 1, August 1976.
l NEDO-10802 - A R.B. Linford, AnalyticalMethods of Plant Transients Evaluations 4.4 l
for the GE BWR, December 1986.
l NEDO-10802-01A R.B. Linford, AnalyticalMethods of Plant Transients Evaluations 4.4 l
for the GE BWR, Amendment 1, December 1986.
l NEDO-10802-02A R.B. Linford, Analytical Methods of Plant Transients Evaluations 4.4 l
for the GE BWR, Amendment 2, December 1986.
NEDO-10871 J.M. Skarpelos and R.S. Gilbert, Technical Derivation of BWR 11.1 1971 Design Basis Radioactive Material Source Terms, March 1973.
NEDO-10958-A H.T. Kim, GeneralElectric Thermal Analysis Basis (GETAB):
4.4 Data, Correlation and Design Applications (LTR), January 1977. 4B NEDO-11209-04-A GE Nuclear Energy Quality Assurance Program Description, the 17.1 latest NRC-accepted version.
NEDE13426P T. R. McIntyre, et. al., Mark III Conformatory Test Program -1/3 3B Scale Impact Tests - Test Series 5805, August 1975.
1.62 MaterialIncorporated by Reference - Amendment 35
23A6100 Rev. 7 ABWR standardsafetyAnalysis neport O
Table 1.6-1 Referenced Reports (Continued)
Title Section No.
NEDO-20206 D.R. Rogers, BWR Turbine Equipment N-16 Radiation Shielding 12.2 Studies, December 1973.
NEDO-20340 J. Camw, Process Computer Performance Evaluation Accuracy, 4.3 June 1974.
NEDO-20533 W.J.Bilanin, The GE Mark Ill Pressure Suppression 6.2 Containment Analytical Model,. June 1974.
NEDO-20533-1 W.J.Bilanin, The GE Mark III Pressure Suppression 6.2 Containment Analytical Model, Supplement 1, September 1975 l
NEDE-20566-A General Electric Company Analytical Model for Loss-of-Coolant 6.3 Analysis in Accordance with 10CFR50, Appendix K, September 1986.
NEDM-20609-01 P.P Stancavage and D.G. Abbott, liquid Discharge Doses L/DSR 12.2 Code, August 1976.
NEDO-20953A J.A. Woolley, Three-Dimensional BWR Core Simulator, January 4A.4 1977.
NEDO-21052 F.J. Moody, Maximum Discharge Rate of Liquid-Vapor Mixtures 6.2 from Vessels, General Electric Company, September 1975.
NEDO-21143-1 H. Careway, V. Nguyen, and P. Stancavege, Radiological 15.2 Accident-The CONACO3 CODE, December 1981.
15.6 l
NEDO-21159 Airborne Releases from BWRs for EnvironmentalImpact 11.1 l
Evaluations, March 1976.
NEDO-21159 Airbome Releases from BWRs for EnvironmentalImpact 12.2 Evaluations-Amendment 2 -lodine NEDE-21175-P BWR/6 Fuel Assembly Evaluation of Combined Safe Shutdown 3.9 Earthquake (SSE) and Loss-of-Coolant Accident (LOCA)
Loadings, November 1976.
l NEDC-21215 Brunswick Steam Electric Plant Unit 1 Safety Analysis Report 4.4 for Plant Modifications to Eliminate Significant In-Core Vibrations, March 1976.
l NEDC-21251 J. Charnley, KKM Safety Analysis Report, April 1976.
4.4 NEDE-21354-P BWR Fuel Channel Mechanical Design and Deflection, 3.9 September 1976.
NEDE-21471 1 L. Lasher,et. af, Analytical Model for Estimating Drag Forces on 3B Rigid Submerged Structures Caused Supplement for X-Quencher Air Discharge, October 1979.
.~
F. Moody, Analytical Modelfor Estimating Drag Forces on Rigid 3B Submerged Structures Caused by a LOCA, September 1977.
MaterialIncorporated by Reference - Amendment 35 1.6-3
23A6100 Rev. 7 ABWR StandardSafetyAnalysis Report O
Table 1.6-1 Referenced Reports (Continued)
Title Section No.
NEDO-21506 Stability and Dynamic Forces of the GE Boiling Water Reactor 4.1 l
(LTR), October 1976.
l NEDE-21514 - 1&2 BWR Scram System Reliability Analysis, December 1976, 19D.6 General Electric Company.
NEDE-21526 J. Dougherty, SCAM - Subcompartment Analysis Method, 6.2 January 1977.
NEDE-21544-P R.J Ernst, et. al., Mark II Pressure Suppression Containment 3B Systems: An AnalyticalModelof the PoolSwellPhenomenon, December 1977.
NEDO-21778-A Transient Pressure Rises Affecting Fracture Toughness 5.3 l
Requirements for Boiling Water Reactors, December 1978.
NEDO-21985 Functional Capability Criteria for Essential Mark ll Piping, 3.9 September 1978, prepared by Battelle Columbus Laboratories for General Electric Company.
NEDE-22056 Failure Rate Data Manual for GE BWR Components, Rev. 2 19.3 January 17,1986, Class Ill, General Electric Company.
19D.3 19E.2 NEDO-22155 GE Report, Generation and Mitigation of Combustible Gas 6.2 Mixtures in Inerted BWR Mark I Containments, June 1982.
NEDE-22277-P-l G. A. Watford, Compliance of the GE BWR Fuel Design to 20.3.7 Stability Licensing Criteria, October 1984.
NEDE-23819 P.D. Knecht, BWR/6 Drywell and Containment Maintenance and 12.4 Testing Access Time Estimates, May 1978.
NEDE-23996-1 P.D. Knecht, Maintenance Access Time Estimates, BWR/6 12.4 Auxiliary and Fuel Buildings, May 1979.
NEDE-23996-2 A. Chappori, Maintenances Access Time Estimates, BWR/6 12.4 Radwaste Building, May 1979.
NEDO-24057 Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 3.9 Plants, November 1977.
NEDO-24057-P Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 3.9 l
Plants, November 1977.
NEDE-24131 Basis for 8x8 Retrofit Fuel Thermal Analysis Application, 4D.2 September 1978.
NEDO-24154 Qualification of the One-Dimensional Core Transient Model for 4.4 BWRs, Vol.1 & 2, October 1978.
NEDO-24154-P Qualification of the One-Dimensional Core Transient Model for 4.4 BWRs, Vol. 3 October 1978. (Proprietary) 1.6-4 MaterialIncorporated by Reference -- Amendment 35
23A6100 Rav. 7 ABWR StandardSafetyAnalysis Report Table 1.6-1 Referenced Reports (Continued)
Title Section No.
l NEDE-24222 J. Weiss, Assessment of BWR Mitigation of ATWS, December 15E j
1979.
19.3
-l NEDE-24302-P Mark ll Containment Program, Generic Chugging Load 3B Definition Report, April 1981.
NEDE-24326-1-P General Electric Environmental Qualification Program, 3.9 Proprietary Document, January 1983.
3.11 NEDE-24351 D. Hale, Fatigue Crack Growth in Piping and RPV Steels in 3E l
Simulated BWR WaterEnvironment Update, July 1981.
NEDE-24679 Study of Advanced BWR Features, Plant Definition / Feasibility 12.4 Results, Vol.lli, Part G, October 1979.
NEDO-24708 P. W. Marriot, AdditionalInformation Required for NRC Staff 7.3 l
Generic Report on Boiling Water Reactors, August 1979.
NEDE-25100-P Mark ll Containment Supporting Program, Caorso SRV 3B Discharge Tests Phase I Test Report, May 1979.
I NEDE-25118 Mark ll Containment Supporting Program, Caorso SRV 3B Discharge Tests Phase II ATR, August 1979.
NEDO-25132A E. W. Bradley, Gamma & Beta Dose to Man from Noble Gas 12.2 Release to the Atmosphere GEMAN Code, April 1980.
NEDO-25153 L. E. Lasher, Analytical Model for Estimating Drag Forces on 3B Rigid Structures Caused by Steam Condensation and Chugging, July 1979.
NEDE-25250 A. Javid, Generic Criteria for High Frequency Cutoff of BWR 3.9 Equipment, January 1980. (Proprietary)
NEDO-25257 E. W. Bradley and V. D. Nguyen, Radiation Exposure from 12.2 Airborne Effluents-the REFAE Code, July 1980.
NEDE-25.?73 F. T. Dodge, Scaling Study of the General Electric Pressure 3B Suppression Test Facility-Mark Ill Long Range Program, Task 2.2. 7, SwRI, March 1980. (Proprietary) l NEDE.'.0090 Alto Lazio Station Reliability Analysis, December 1984 19D.6 NEDC-30130-A Bill Zarbis, Steady-State Nuclear Methods, May 1985.
4.3 (Proprietary) 4.4 NCDC-30259 H.A. Careway, D.B. Townsend, B.W. Shaffer, A Technique for 15.6 Evaluation of BWR MSIVLeakage Contribution to Radiological l
Dose Rate Calculations, September 1983.
NEDE-30637 B.M. Gordon, Corrosion ar'd Corrosion Controlin BWRs, 5.2 December 1984.
NEDE-30640 Evaluation of Proposed Modification to the GESSAR 11 Design, 19P Class Ill, June 1984.
MaterialIncorporated by Reference - Amendment 35 1.64
23A6100 RGv. 7 ABWR StandardSafetyAnalysis Report O
Table 1.6-1 Referenced Reports (Continued)
Title Section No.
NEDO-30832 J.E. Torbeck,, Elimination of Limit on BWR Suppression Pool 3B Temperature forSRVDischarge With Quenchers, December 1984.
NEDC-30851P-A W. P. Sullivan, Technical Specification improvement Analyses 19D.6 l
for BWR Reactor Protection System, March 1988.
NEDE-31096-A GE Licensing Topical Report ATWS Response to NRC ATWS 19B.2 Rule 10CFR 50.62, February 1987.
NEDE-31152-P GE Bundle Designs, December 1988.
4.2 NEDO-31331 Gerry Burnette, BWR Owner's Group Emergency Procedure 18A l
Guidelines, March 1987.
NEDC-31336 Julie Loong, GeneralElectric Instrument Setpoint Methodology, 7.3 October 1986.
l NEDC-31393 ABWR Containment Horizontal Vent Confirmatory Test, Part I, 3B March 1987.
NEDO-31439 C. VonDamm, The Nuclear Measurement Analysis & Control 20.3 Wide Range Neutron Monitoring System (NUMAC-WRNMS),
l May 1987 NEDC-31858P Louis Lee, BWROG Report for Increasing MSIV Leakage Rate 15.6 Limits and Elimination of Leakage Control System,1991 l
NEDE-31906-P A. Chung, Laguna Verde Unit I ReactorInternals Vibration 7.4 l
Measurement, January 1991.
NED0-31960 Glen Watford, BWR Owners' Group Long-Term Stability 4.4 Solutions Licensing Methodology, June 1991.
NEDC-32267P ABWR Project Application Engineering Organization and 17.1 Procedures Manual, December 1993.
O 1.6-6 MaterialIncorporated by Reference - Amendment 35
23A6100 Rev. 7 ABWR st:nd:rdsxteryAn:tysis nipois O
V Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title l
B3.5 1960 American Standard Tolerance for Ball and Roller Bearings B30.2 (See ASME B30.2)
B30.9 (See ASME B30.9)
B30.10 (See ASME B30.10)
B30.11 (See ASME B30.11)
B30.16 (See ASME B30.16)
B31.1 (See ASME B31.1)
B96.1 (See ASME B96.1)
C1/ASOC 1985 Specifications of General Requirements for a Quality Program C37.01 (See IEEE C37.01)
C37.04 (See IEEE C37.04)
C37.06 1987 Preferred Ratings of Power Circuit Breakers C37.09 (See IEEE C37.09)
C37.11 1979 Power Circuit Breaker Control Requirements C37.13 (See IEEE C37.13)
C37.16 1988 Preferred Ratings and Related Requirements for LowVoltage AC Power Circuit Breakers Trip Devices for AC and General-Purpose DC Low-Voltage Power C37.17 1979 Circuit Breakers C37.20 (See IEEE C37.20)
C37.50 1989 Test Procedures for Low Voltage AC Power Circuit Breakers Used in Enclosures C37.100 (See IEEE C37.100)
C57.12 (See IEEE C57.12)
C57.12.11 (See IEEE C57.12.11)
C57.12.80 (See IEEE C57.12.80)
C57.12.90 (See IEEE C57.12.90)
C62.41 (See IEEE C62.41)
C62.45 (See IEEE C62.45)
C63.12 (See IEEE C63.12)
D975 / ASTM '
1981 Diesel Fuel Oils, Specifications for Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 35 1.8-37
23A6100 Rsv. 7 ABWR St:nd:rdS:letyAn: lysis Ripirt O
Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title HEI 1970 Standards for Steam Surface Condenser,6th Ed., Heat Exchangers Institute HFS-100 1988 Human Factors Engineering of Visual Display Terminal Workstations MC11.1 1976 Quality Standard for Instrument Air N5.12 1972 Protective Coatings (Paint) for Nuclear Industry N13.1 1969 Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities l
l N14.6 1986 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials N18.7 1976 Administrative Controls and Quality Assurance for the Operation Phase of Nuclear Power Plants N45.2.1/
1973 Cleaning of Fluid Systems and Associated Components During (RG 1.37)
Construction Phase of Nuclear Power Plants l
N45.2.2/
1972 Packaging, Shipping, Receiving, Storage, and Handling of items (RG 1.38) for Nuclear Power Plants During the Construction Stage N45.2.3 1973 Housekeeping During the Construction Phase of Nuclear Power Plants N45.2.4 1972 Quality Assurance Program Requirements for Nuclear Power Plants N45.2.5 1974 Supplementary Quality Assurance Requirements for Installation, inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants f
l N45.2.8 (RG 1976 Quality Assurance Requirements for Installation, Inspection, and l
1.116)
Testing of Mechanical Equipment and Systems N45.4 (See ASME N45.4)
N101.2 1972 Protective Coatings (Paints) for Light Water Nuclear Containment Facilities N101.4 1972 QA for Protective Coatings Applied to Nuclear Facilities N195 (See ANS 59.51)
N237 (See ANS 18.1)
N270 (See ANS 52.2)
N509 (See ASME N509)
N510 (See ASME N510)
N690 (See AISC N690) l 1.8-38 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 35 l
l
23A6100 R2v. 7 ABWR st:ndxrdsihtyAn:Iysis Rip:rt C
Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title N OA-1 (See ASME NOA-1) l NOA-1a (See ASME NOA-la) l NQA-2a 1990 Quality Assurance Requirements of Computer Software for Nuclear Facility Application OM3 1990 Requirements for preoperational and initial Startup Vibration Test Program for Water-Cooled Power Plants OM7 1986 Requirements for Thermal Expansion Testing of Nuclear Plant Piping Systems [ September 1986 (Draft-Revision 7)]
American Petroleum institute (API) t 620 1986 Rules for Design and Construction of Large, Welded, Low-Pressure Storage Tanks 650i 1980 Welded Steel Tanks for Oil Storage American Society of Heating, Refrigerating and Air-Conditioning Engineers,Inc. (ASHRAE) 30 1978 Methods of Testing Liquid Chilling Packages 33 1978 Methods of Testing Forced Circulation Air Cooling and Air Heating Coils American Society of Mechanical Engineers (ASME)
AG-1t 1991 Code on Nuclear Air and Gas Treatment 830.2' 1983 Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Grider, Top Running Trolley Hoist)
B30.9t 1984 Slings t
B30.10 1982 Hooks t
B30.11 1980 Monorails and Underhung Cranes B30.16i 1981 Overhead Hoists t
B31.1 1986 Power Piping t
B96,1 1986 Specification for Welded Aluminum-Alloy Storage Tanks N45.4 1972 Leakage-Rate Testing of Containment Structures for Nuclear Reactors N 509' 1989 Nuclear Power Plant Air-Cleaning Units and Components N
N510' 1989 Testing of Nuclear Air-Cleaning Systems N OA-1t 1983 Quality Assurance Program Requirements for Nuclear Facilities Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 35 1.8-39
23A6100 Rev. 7 ABWR StandmlS:t;ty An: lysis R; port Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title t
l N QA-la 1983 Addenda to ANSI /ASME NOA-1-1983 OMa 1988 Operation and Maintenance of Nuclear Power Plants (Addenda to OM-1987)
Secll 1989 BPVC Section 11, Material Specifications Sec ill 1989 BPVC Section lil, Rules for Construction of Nuclear Power Plant Components Sec Vill 1989 BPVC Section Vill, Rules for Construction of Pressure Vessel SecIX 1989 BPVC Section IX, Qualification Standard for Welding and Brazing Procedures Welder, Brazers and Welding and Brazing Operators SecXI 1989 BPVC Section XI, Rules for Inservice inspection of Nuclear Power Plant Components American Society for Testing and Materials (ASTM) l C776 1979 Sintered Uranium Dioxide Pellets l
C934 1980 Design and Quality Assurance Practices for Nuclear Fuel Rods E84 REV. A 1991 Methods of Test of Surface Burning Characteristics of Building Materials E119 1988 Standard Test Methods for Fire Tests of Building Construction and Materials E152 1981 Standard Methods of Fire Tests of Door Assemblies (See ASME BPVC Section til for ASTM Material Specifications)
American Welding Society (AWS) t A4.2 1986 Procedures for Calibrating Magnetic Instruments to Measure the Delta Ferrite content of Anstenitic Stainless Steel Weld Metal D1.1t 1986 Steel Structural Welding Code D14.1t 1985 Welding of industrial and Mill Cranes and other Material Handling Equipment American Water Works Association (AWWA)
D100t 1984 Welded Steel Tanks for Water Storage CMAA70 1983 Specification for Electric Overhead Traveling Cranes 9
1.8-40 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 3S
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23A6100 Rev. 7 ABWR standardsafetyAn: lysis R:p:rt
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Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title Insulated Cable Engineer Association (ICEA)
P-46-426/IEEE 1982 Ampacities including Effect of Shield Losses for Single S-135 Conductor Solid-Dielectric Power Cable 15 kV through 69 kV P-54-440/ NEMA 1987 Ampacities of Cables in Open-Top Cable Trays WC-51 S-61-402/ NEMA 1973 Thermoplastic Insulated Wire & Cable for the Transmission and W C-5 Distribution of Electrical Energy S-66-524/ NEMA 1982 Cross Linked Thermosetting Polyethylene insulated Wire and W C-7 Cable for Transmission and Distributor of Electrical Energy institute of Electrical and Electronics Engineers (IEEE) t l
C37.01 1979 Application Guide for Power Circuit Breakers C37.04t 1979 AC Power Circuit Breaker Rating Structure f
C37.09i 1979 Test Procedure For Power Circuit Breakers t
C37.13 1989 Low Voltage Power Circuit Breakers C37.20t 1987 Switchgear Assemblies and Metal-Enclosed Bus n
C37.90.2 1987 Trial-Use Standard, Withstand Capability of Relay Systems to Radiated Electromagnetic Interference form Transceivers t
C37.100 1992 Definitions for Power Switchgear Transformers i
C57.12 1987 General Requirements for Distribution, Power, and Regulating Transformers t
C57.12.11 1980 Guide for installation of Oil-immersed Transformers (10MVA &
Larger,69-287 kV Rating) t C57.12.80 1978 Terminology for Power and Distribution Transformers i
C57.12.90 1987 Test Code for Distribution, Power, and Regulating Transformers t
C62.41 1991 Guide for Serge Voltage in Low-Voltage AC Power Circuits C62.45t 1987 Guide on Surge Testing for Equipment Connected to Low-Voltage AC Power Curcuits t
C63.12 1987 American National Standard for Electromagnetic Compatibility Limits-Recommended Practice 1982 Application Criteria for Digital Computers in Safety Systems for Nuclear Facilities (to be replaced by the issued version of P 7-4.3.2, " Standard Criteria for Digital Computers Used in Safety Systems of Nuclear Power Generation Stations")
i 80 1986 Guide for Safety in AC Substation Grounding Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 35 1.8-41
23A6100 Rev. 7 ABWR Stand;rd SafetyAclysis R:pirt O
Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title 81
'983 Guide for Measuring Earth Resistivity, Ground Impedance, and 1
Earth Surface Potentials of a Ground System S-135 (See ICEA P-46-426) 1 1986 Recommended Practice for Electric Power Distribution for 141 Industrial Plants (IEEE Red Bock) 242 1986 Recommended Practice for Protection and Coordination of t
Industrial and Commercial Power Systems 279 1971 Criteria for Protection Systems for NPGS 308 1980 Criteria for Class 1E Power Systems for NPGS t
317 1983 Electrical Penetration Assemblies in Containment Structures for t
NPGS 323' 1974 Qualifying Class 1E Equipment for NPGS 334 1974 Motors for NPGS, Type Tests of Continuous Duty Class 1E t
338' 1977 Criteria for the Periodic Testing of NPGS Safety Systems 344 1987 Recommended Practices for Seismic Qualifications of Class 1E 1
Equipment for NPGS 352 1987 General Principles for Reliability Analysis of Nuclear Power t
Generating Station Protection Systems 379 1977 Standard Application of the Single-Failure Criterion to NPGS 1
Safety Systems 382 1985 Qualification of Actuators for Power-Operated Valve Assemblies i
with Safety-Related Functions for NPP 3831 1974 Type Test of Class 1E Cables; Field Splices and Connections for NPGS 384 1981 Criteria for Independence of Class 1E Equipment and Circuits T
387 1984 Criteria for Diesel-Generator Units Applied as Standby Power t
Supplies for NPGS 399 1990 Recommended Practice for Industrial and Commercial Power i
Systems Analysis (IEEE Brown Book) 450' 1987 Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations 484 1987 Recommended Practice for the Design and Installation of Large t
Lead Storage Batteries for NPGS 485 1983 Recommended Practice for Sizing Large Lead Storage Batteries i
for NPGS 1.8-42 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 35
23A6100 Rev. 7 ABWR StandirdSaf;tyAn:Iysis R:pirt V
Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title 500 1984 Guide to the Collection and Presentation of Electronic, Sensing Component, and Mechanical Equipment Reliability Data for Nuclear Power Generating Stations.
519t 1981 IEEE Standard Recommended Practices and Requirements for Harmonic Control in Electrical Power Systems t
603 1980 IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations t
622 1987 Recommended Practice for the Design and Installation of Electric Heat Tracing Systems in Nuclear Power Generating Stations 622At 1984 Recommended Practice for the Design and installation of Electric Pipe Heating Control and Alarm Systems in Nuclear Power Generating Stations l
730 1984 Standard for Software Quality Assurance Plans 741t 1986 Standard Criteria for the Protection of Class 1E Power Systems
-q and Equipment in Nuclear Power Generating Stations 765t 1983 Standard for Preferred Power Supply for Nuclear Power Generating Stations 802.2t 1985 Standards for Local Area Networks: Logic Link Control 802.5 1985 Token Ring Access Method and Physical Layer Specifications i
t 828 1983 Standard for Software Configuration Management Plans 829t 1983 Standard for Software Test Documentation 830t 1984 Standard for Software Requirements Specifications t
845 1988 Guide to Eva sation of Man-Machine Performance in Nuclear Power Genetting Station Control Rooms and Other Peripheries 944' 1986 Recommended Practice for the Application and Testing of Uninterruptable Power Supplies for Power Generating Station 946' 1985 Recommended Practice for the Design of Safety-Related DC Auxiliary Power Systems for Nuclear Power Generating Stations t
1012 1986 Standard for Software Verification and Validation 0
1023 1988 IEEE Guide to the Application of Human Factors Engineering to Systems, Equipment and Facilities of Nuclear Power Generating Stations 1033' 1985 Recommended Practice of Application of IEEE-828 to Nuclear Power Generation Stations
,l 1042 1987 Guide to Software Configuration Management l
1228 (Draft) 1992 Standard for Software Safety Plans Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 35 1.8-43
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23A6100 Rev. 7 ABWR St:ndardSttyAulysis R: port O1 Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title instrument Society of America (ISA)
S7.3t 1981 Quality Standard for Instrument Air
.367.02-80 1980 Nuclear-Safety-Related Instrument Sensing Line Piping and Tubing Standards for Use in Nuclear Power Plants National Electrical Manufacturers Association (NEMA)
AB1 1986 Molded Case Circuit Breakers FB1 1977 Fittings and Support for Conduit and Cable Assemblies t
ICS l 1983 General Standards for Industrial Control ICS 2' 1988 Standards for Industrial Centrol Devices, Controllers and Assemblies MG1 1987 Motors and Generators WC-5 (See ICEA S-61-402)
WC7 (See ICEA S-66-524)
WC 51 (See ICEA P-54-440)
National Fire Protection Association (NFPA) t 10 1981 Portable Fire Extinguishers -Installation 10A 1973 Portable Fire Extinguishers - Maintenance and Use 11t 1988 Low Expansion Foam and Combined Agent Systems-Foam Extinguishing System t
12 1985 Carbon Dioxide Extinguishing Systems 13t 1985 Installation of Sprinklers Systems t
14 1986 Installation of Standpipe and Hose Systems 15t 1985 Standard for Water Spray Fixed Systems i
16 1991 Deluge Foam-Water Sprinkler and Foam-Water Spray Systems 16At 1988 Recommended Practice for the installation of Closed Head Foam-Water Sprinkler Systems i
20 1990 Standard for the installation of Centrifugal Fire Pumps i
24 1984 Private Service Mains and their Appurtenances i
26 1988 Recommended Practice for the Supervision of Valves Controlling Water Supplies for Fire Protection 1.8M Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 35
-~
23A6100 Rev. 7 ABWR stand:rdS:1ityAn:Iysis Riport O
~ Table 1.8-21 Industrial'.: odes and Standards
- Applicable to ABWR (Cominued)
Code or Standard Number Year Title 37' 1984 Stationary Combustion Engines and Gas Turbines t
70 1987 National Electrical Coue-Handbook 1987 72t 1990 Protective Signaling Systems 72D 1986 Proprietary Protective Signaling Systems i
78 1986 Lightning Protection Code t
80 1986 Fire Doors and Windows t
80A 1993 Protection of buildings from Exterior Fire Exposures t
90A 1985 Installation of Air Conditioning and Ventilating Systems 91t 1983 Blower and Exhaust Systems 92At 1988 Smoke Control Systems t
101 1985 Life Safety Code 251' 1985 Fire Test, Building Construction and Materials 252t 1984 Fire Tests, Door Assemblies 255i 1984 Building Materials, Test of Surface Burning Characteristics Sitt 1987 Classification of Flammable Liquids 801t 1986 Facilities Handling Radioactive Materials 802' 1988 Nuclear Research Reactors 803t 1993 Fire Protection for Light Water Nuclear Power Plants 1961t 1979 Fire Hose 1963t 1985 Semw Threads and Gaskets for Fire Hose Connec lons Steel Structures Painting Council (SSPC)
PA 1 1972 Shop, Field and Maintenance Painting PA-2 1973 Measurements of Paint Film Thickness with Magnetic Gages SP-1 1982 Solvent Cleaning l
SP-5 1985 White Metal Blast Cleaning i
SP-6 1986 Commercial Blast Cleaning SP-10 1985 Near-White Blast Cleaning U.S. Department of Defense (DOD) l b000.2 1991 Defense Acquisition Management Policies and Procedures Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 35 1.845
23A6100 Rev. 7 ABWR Stard;rdSal;tyAn:1ysis Rep:rt O
Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title AD/A223168 1990 System Engineering Management Guide AR602-1 1983 Human Factors Engineering Program Dl-HFAC-80740 1989 Human Factors Engineering Program Plan ESD-TR-86-278 1986 Guidelines for Designing User Interface Software HDBK-761A 1990 Human Engineering Guidelines for Management Information Systems HDBK-763 1991 Human Engineering Procedures Guide, Ch. 5-7 & Appendix. A&B l
STD-2167A 1988 Defense System Software Development TOP 1-2-610 1990 Test Operating Procedure Part 1 U.S. Military (MIL)
F-51068 Latest Filter, Particulate High-Efficiency, Fire-Resistant Edition H-46855B 1979 Human Engineering Requirements for Military Systems, Equipment and Facilities HDBK-251 Latest Reliability / Design: Thermal Applications Edition HDBK-759A 1981 Human Factors Engineering Design for Army Material STD-282 1956 Filter Units, Protective Clothing Gas-Mask Components and Related Products: Performance-Test Methods STD-461C 1987 Electromagnetic Emission and Susceptibility Requirements for the Control of Electromagnetic interference STD-462 1967 Measurement of Electromagnetic Interference Characteristics STD-1472D 1989 Human Engineering Design Criteria for Military Systems, Equipment and Facilities STD-1478 1991 Task Performance Analysis Others ERDA 76-21 1976 Testing of Ventilation Systems, Section 9 of Industrial Ventilation Systems IEC 880 1986 Software for Computers in the Safety Systems of Nuclear Power Stations IEC 964 1989 Design for Control Rooms of Nuclear Power Plants, Bureau Central de la Commission Electrotechnique Internationale 1.8-46 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 3S
m.
23A6100 Rev. 7 ABWR sz1ntsai;tyAn:lysisR: port N
Table 1.8-21 Industrial Codes and Standards
- Applicable to ABWR (Continued)
Code or Standard Number Year Title ISO 7498 1984 Open Systems interconnection-Basic Refence Model, as the Data Link Layer and Physical Layer OSHA 1910.179 1990 Overhead and Gantry Cranes TEMA C 1978 Standards of Tubular Exchanger Manufacturers Association UL-44 1983 Rubber-Insulated Wires and Cables UL-489 1991 Molded-Case Circuit Breakers and Circuit Breaker Enclosures UL-845 1988 Standard for Safety Motor Control Centers - Low Voltage Circuit Breakers Crane Manufacturers Association of America, Specification No. 70 Aluminum Construction Manual by Aluminum Association NCIG-01 Rev.2 Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants UBC 1991 Uniform Building Code V'
- The listing of a code or standard does not necessarily mean that it is applicable in its entirety.
t Also an ANSI code (i.e. ANSl/ASME, ANSI /ANS, ANSI /IEEE etc.).
- ANSI, ANSI /ANS, ANSI /ASME, and ANSI /IEEE codes are included here. Other codes that approved by ANSI and another organization are listed under the latter.
f As modified by NRC accepted alternate positions to the related Regulatory Guide and identified in Table 2-1 of Reference 1 to Chapter 17.
l l
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Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 35 1,8-47
23A61CO Rev. 4 ABWR StandardSafetyAnalysis Report O
Table 1.8-22 Experience Information Applicable to ABWR lssue No.
Date Title Comment Type: Generic Letters 80-06 4/25/80 Clarification of NRC Requirement for Emergency Response Facilities at Each Site 80-30 12/15/80 Periodic Updating of Final Safety Analysis Reports (FSARs)
COL Applicant 80-31 12/22/80 Control of Heavy Loads 81-03 2/26/81 implementation of NUREG-0313m, Rev.1 81-04 2/25/81 Emergency Procedures and Training for Station Blackout Events COL Applicant 81-07 2/3/81 Control of Heavy Loads 81-10 2/18/81 Post-TMI Requirements for the Emergency Operations Facility 81-11 2/22/81 Error in NUREG-0619 l
81-20 4/1/81 Safety Concerns Associated with Pipe Breaks in the BWR Scram l
System 81-37 12/29/81 ODYN Code Reanalysis Requirements j
l 81-38 11/10/81 Storage of Low-Level Radioactive Wastes at Power Reactor COL Sites Applicant 82-09 4/20/82 Environmental Qualification of Safety-Related Electrical Equipment 82-21 10/6/82 Technical Specifications for Fire Protection Audits COL Applicant 82-22 10/30/82 inconsistency Between Requirements of 10CFR73.40(d) and Standard Technical Specifications for Performing Audits of Safeguard Contingency Plans 82-27 11/15/82 Transmittal of NUREG-0763, " Guidelines for Confirmatory in-Plant Tests of Safety-Relief Valve Discharges for BWR Plants,"
and NUREG-0783, " Suppression Pool Temperature Limits for BWR Contsinments."
82-33 12/17/82 Supplement 1 to NUREG-0737 82-39 12/22/82 Problems with the Submittals of 10CFR73.21 Safeguards COL Information Licensing Review Applicant I
83-05 2/83 Safety Evaluation of " Emergency Procedure Guidelines,"
COL Revision 2, NEDO-24934, June 1982 Applicant l
83-07 2/16/83 The Nuclear Waste Policy Act of 1982 COL Applicant 1.8-48 Conformance with Standard Review Plan and Applicacility of Codes and Standards - Amendment 34 l
23A6100 Rev. 7 ABW8 StaedardSafety Analysis Report
(~~S 1 A.2.7 Post-Accident Sampling [lI.B.3]
NRC Position A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a sample under accident conditions without incurring a l
radiation exposure to any indhidualin excess of 0.05 and 0.50 Sv to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.
A design and operational review of the radiological spectrum analysis facilities shall be performed to determine the capability to quantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) certain radionuclides that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which indicate high fuel temperatures), and nonvolatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.4 release.
The review should also consider the effects of direct radiation from piping and O
components in the auxiliary building and possible contamination and direct radiation
(/
from airborne effluents. If the reviewindicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.8 or 1.4 source term). Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift).
Response
Discharges From Plant and Containment-During the development of an accident, samples ofliquid and gaseous discharges from both the plant and containment will be obtained. Chemical and radiochemical analyses will be performed for protection of the health and safety of the public and the plant operators. These samples will be obtained from the Process Sampling System. The Post Accident Sampling Systems will not be required to obtain these samples.
Core Damage Assessment-During this initial period, instrumentation will pro ide sullicient information for the operators to perform their duties. For example, the
/
T containment high range radiation meters will give instant information concerning the U
radiation level in containment (To obtain data from the PASS several hours may be j
required for sampling and analpes.). Calculations can be performed to relate Response to TMI Related Matters - Amendment 35 1A-7 1
23A61:0 Rev. 4 ABWR Standard Safety Analysis Report O
containment radiation level with the probable extent of core damage. Core damage assessment instrumentation is described in Secdon 18.4.6. This section describes the safety parameter display system (SPDS). The principle purpose of the SPDS is to aid the control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to avoid a degraded core. The following cridcal safety functions are provided at the wide screen display panel in the main control room:
(1) Reacdvity control (2) Reactor core cooling and heat removal from the primary system (3) Reactor coolant system integrity (4) Radioactivity control (5) Contamination conditions This instrumentation and the PASS work together to obtain complementary information. After this initial period during the development of an accident, the ABWR PASS will be used to obtain samples of reactor water and containment atmosphere to assess the extent of core damage. The ABWR PASS has been designed to safely obtain j
samples with radioactivity levels up to 37,000 M Bq/g. Approximately one day after a serious core damage accident,it is expected that sample radioacdvity levels will be no more than this value. Early in such an accident, the plant instrumentation in the main control room would be indicating that abnormal conditions exist. If a reactor coolant sample were obtained which had excessive radioactivity, as measured by the area radiation monitor in the PASS area, the plant operators would use this high radiation information as confirmatory evidence that severe core damage has occurred and continue following the emergency operating procedures. It would not be necessary to perform any radiochemical analyses to reach this conclusion. During less severe accidents,in which only some cladding damage has occurred, samples may be obtained from either the Process Sampling System or PASS.
l NUREG-0737 Requirements-The ABWR PASS has been designed to meet the eleven requirements listed in NUREG-0737 except as noted below:
(1) The combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision is made to take a sample. Meets the requirements of NUREG-0737.
O 1A4 Response to TMIRelatedMatters - Amendment 34 l
l l
23A6100 Rev. 2 ABWR studardsdelyAnlysis Report v
(2) There shall be onsite capability to perform the following within the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time period:
o (a) Determine the presence and amount of certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of J
the degree of core damage. Meets the requirements of NUREG-0737.
(b) Hydrogen in containment atmosphere. Hydrogen in containment atmosphere is measured by the Containment Atmospheric Monitoring System. Meets the requirements of NUREG-0737.
(c) Dissolved gases, chloride and boron in liquids. Dissolved gases are discussed in item 4 below. Meets the requirements concerning chloride and boron of NUREG-0737.
(d) Inline monitoring capability is acceptable. No inline monitors are
?
provided in PASS.
(3) Sampling need not depend upon an isolated auxiliary system being put into operation. Meets the requirements of NUREG-0737.
(4) Reactor coolant samples and analyses for total dissolved gases and hydrogen are required. During a severe core damage accident for the ABWR, the reactor water will become mixed with the suppression pool water. The pressure in the reactor vessel will decrease to approximately the pressure within the wetwell and the drywell. As a result of this decreasein pressure, dissolved gases will '
partially pass out of the water phase into the gas phase. Data on gases dissolved in the reactor water under these conditions will have little meaning in responding to the accident. During accidents in which only a small amount of cladding damage has occurred or in accidents in which the reactor vessel has not been depressurized, pressurized reactor water samples may be obtained from the Process Sampling System. Therefore, the ability to obtain pressurized or depressurized reactor water samples for dissolved gas analyses has not been prosided.
(5) If both of the following are present:
L (a) There is only a single barrier between primary containment and the cooling water.
(b)
If the cooling water is sea water or brackish water, chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after sampling shall be provided. If both are not present, the time to complete the analyses is increased to 4 days. Analysis does not h
have to be done onsite. Meets the requirements of NUREG-0737. (Note d
that there are several barriers to prevent chloride intrusion from the power cycle cooling water into the reactor vessel. These barriers are: the i
Response to TMIRelated Matters - Amendment 32 1A-9 1
23A6100 Rea 7 ABWR StandardSafety Analysis Report O
main condenser tubing, the condensate polishing demineralizers and the pumps and valves in the condensate /feedwater systems. These pumps are stopped and these valves closed during upset conditions.
Thus, because both factors are not present, the time to complete the analysis is increased to 4 days.)
(6) It must be possible to obtain and analyze a sample without radiation exposures to any individual exceeding 0.05 Sv for whole body and 0.50 Sv for extremities.
Meets the requirements of 50.34(f)(2)(viii).
(7) Ability to sample and analyze for reactor coolant boron must be provided.
Meets the requirements of NUREG-0737.
(8)
Ifinline monitoring is used, backup sampling and analysis capability must be provided. Inline monitoring is not used. Meets the requirements of NUREG-0737.
(9)
(a) Capability to identify and quantify a specified number ofisotopes over a range of nuclide concentrations from approximately 37,000 Bq/g to 370,000 M Bq/g. Capability is provided to identify and quantify the desired isotopes in samples over a range from approximately 37,000 Bq/g to 37,000 M Bq/g. Samples obtained during the accident recovery phase would be within this range for most core damage accidents. If the gross radioacti ity levels are higher than 37,000 M Bq/g, this would confirm that severe core damage has occurred.
1 (b) Restrict background levels of radiation in the laboratory and proside proner ventilation. Meets the requirements of NURECr0737.
(10) Provide adequate accuracy, range and sensitivity to provide pertinent information. Meets the requirements of NURECr0737.
(11)
(a) Provide sample lines with proper features for sampling during accident conditions. Meets the requirements of NURECr0737.
(b) PASS ventilation exhaust should be filtered with charcoal adsorbers and HEPA filters. Meets the requirements of NURECr0737.
O 14-10 Response to TMI Related Matters - Amendment 4
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23A6100 Rev. 7 ABWR saadardsweryAutrsisneport N
Strain energy weighted modal damping can an used in the dynamic analysis. Strain energy weighting is used to obtain the modal damping coefficient due to the contributions of elements with different damping properties in the model. The element damping values are specified in Table 3.7-1. Strain energy weighted modal damping is
]
calculated as specified in Subsection 3.7.2.15.
In direct integration analysis, damping is input in the form of a and damping constants, which give the percentage of critical damping,1 as a function of the circular frequency, co.
a pto A=g+7 (3.7-27) 3.7.3.8.1.8 Effect of Differential Building Movements In most cases, subsystems are anchored and restrained to floors and walls of buildings that may have differential movements during a seismic event. The movements may range from insignificant differential displacements between rigid walls of a common building at low elevations to relatively large displacements between separate buildings at a high seismicity site.
\\
Differential endpoint or restraint deflections cause forces and moments to be induced into the system. The stress thus produced is a secondary stress. It isjustifiable to place this stress, which results from restnint of free-end displacement of the system, in the secondary stress category because the stresses are self-limiting and, when the stresses exceed yield strength, minor distorties or deformations within the system satisfy the condition which caused the stress to occur.
The earthquake thus produces a stress-exhibiting property much like a thermal expansion stress and a static analysis can be used to obtain actual stresses. The differential displacements are obtained from the dynamic analysis of the building. The displacements are applied to the anchors and restraints corresponding to the maximum differential displacements which could occur. The static analysis is made three times:
once for one of the horizontal differential displacements, or < for the other horizontal differential displacement, and once for the vertical.
The inertia (primary) and displacement (secondary) loads are dynamic in nature and their peak values are not expected to occur at the same time. Hence, the combination
)
of the peak values ofine:rtia load and anchor displacement load is quite conservative. In addition, anchor movement effects are computed from static analyses in which the displacements are applied to produce the most conservative loads on the components.
Therefore, the primary and secondary loads are combined by the SRSS method.
Seismic Design - Amendment 35 3.7-37 1
4
23A6100 Rett. 3 ABWR StandardSafetyAnalysis Report O
3.7.3.8.1.9 Design of Small Branch and Small Bore Piping (1) Small branch lines are defined as those lines that can be decoupled from the analytical model used for the analysis of the main run piping to which the branch lines attach. As allowed by Subsection 3.7.3.3.1.3, branch lines can be decoupled when the ratio of run to branch pipe moment ofinertia is 25 to 1, or greater. In addition to the moment ofinertia criterion for acceptable decoupling, these small branch lines shall be designed with no concentrated masses, such as valves,in the first one-half span length from the main run pipe; and with sufficient flexibility to prevent restraint of movement of the main rtm pipe. The small branch line is considered to have adequate flexibility ifits first anchor or restraint to movement is at least one-half pipe span in a direction perpendicular to the direction of relative movement between the pipe run and the first anchor or restraint of the branch piping. A pipe span is defined as the length tabulated in Table NF-3611-1, Suggested Piping Support Spacing, ASME B&PV Code Section III, Subsection NF. For bmnches where the preceding criteria for sufficient flexibility cannot be met, the applicant will demonstrate acceptability by using an alternative criteria for sufIicient flexibility, or by accounting for the effects of the branch piping in the analysis of the main run piping.
(2) For small bore piping defined as piping 50A and less nominal pipe size, and small branch lines 50A and less nominal pipe size, as defined in (1) above,it is acceptable to use small bore piping handbooks in lieu of performing a system flexibility analysis, using static and dynamic mathematical models, to obtain loads on the piping elements and using these loads to calculate stresses per equations in NB, NC, and ND3600 in ASME Code Section III, whenever the following are met:
(a) The small bore piping handbook at the time of application is currently accepted by the regulatory agency for use on equivalent piping at other nuclear power plants.
(b) When the small bore piping handbook is serving the purpose of the Design Report it meets all of the ASME requirements for a piping design report. This includes the piping and its supports.
(c) Formal documentation exists showing piping designed and installed to the small bore piping handbook (1) is conservative in comparison to results from a detail stress analysis for all applied loads and load combinations using static and dynamic analysis methods defined in Subsection 3.7.3, (2) does not result in piping that is less reliable because ofloss of flexibility or because of excessive number of supports, (3) satisfies required clearances around sensitive components.
3.7-38 Seismic Design - Amendment 33
23A6100 Rev. 7 :
- ABWR standardsareryAnalysis neport.
O 3.7.6 References 3.7-1 Deleted.
3.7-2 Deleted.
3.7-3 Deleted.
3.7-4 L. K. Liu, Seismic Analysis of the Boiling WaterReactor, symposium on seismic analysis of pressure vessel and piping components, First National Congress on Pressure Vessel and Piping, San Francisco, California, May 1971.
3.7-5 EPRI NP-5930, A CnterionforDeterminingExceedance of the OperatingBasis Earthquake, July 1988.
3.7 EPRI TR-100082,' Standardization of Cumulative Absolute Velocity, December 1991.
3.7-7 EPRI NP-6695, GuidelinesforNuclearPlant Response to an Earthquake, December.
1989.
i O,
y' P. Koss, Seismic Testing ofElectrical Cable Support Systems, Structural Engineers of 3.7-8 California Conference, San Diego, September 1979.
e t
s Seismic Design-Amendment 35 3.747 I-
,n-.
33A6100 Rett, 4 ABWR StandardSafetyAnalysis Report O
Table 3.7-1 Damping for Different Materials Percent Critical Damping item SSE Reinforced concrete structures 7
Welded structural assemblies 4
Steel frame structures 4
Bolted or riveted structural assemblies 7
Equipment 3
piping systems' diameter greater than 300A nominal 3
diameter less than or equal to 300A 2
nominal Reactor pressure vessel, support skirt, 4
shroud head and separator Guide tubes and CRD housings 2
Fuel 6
Cable trays 20 (max)
(see Figure 3.7-27)
Conduits 7
HVAC ductwork
-companion angle 7
l
- pocket lock 7
- welded 4
- Damping values of ASME Code Case N 411-1, alternative damping values for Response Spectra Analysis of Class 1,2, i
and 3 Piping, Section ill, Division 1, may be used as permitted by Regulatory Gt;ide 1.84.These damping values are applicable in analyzing piping response for Seismic and other dynamic loads filtering through building structures in high frequencies range beyond 33 Hz.
l 9
3.748 Seismic Design - Amendment 34
23AS100 Rev. 7 ABWR standardsafety Analysis Report d
Table 3.8-2 Major Allowable Stresses in Concrete and Reinforcing Steel Concrete Reinforcing Steel Compression Tangential Shear Tension Service Load 16.54 MPa (1) Provided by concrete 206.8 MPa Combination v=0 c
(2) Provided by orthogonal 310.3 MPa (For test reinforcement pressure case) v,o = 1.2f = 1.96 MPa c
l Factored Load 23.44 MPa (1) Provided by concrete 372.4 Mpa Combination v=0 c
(2) Provided by orthogonal reinforcement v,o = 2.4{f'c = 3.92 MPa Od Table 3.8-3 Stress Intensity Limits Primary Stresses Primary &
Bending &
Secondary Gen. Mem Local Mem.
Local Mem.
Stresses P
P P+P P+P+Q m
t t
t Test Condition 0.75 Sy 1.15 Sy 1.15 Sy N/A Design Condition 1.0 Sm' 1.5Sm 1.5Sm N/A Post-LOCA The larger of The larger of The larger of 3Sm' Flooding 1.2 Smc or 1.0 Sy 1.8 Smc or 1.5 Sy 1.8 Smc or 1.5 Sy
- The allowable stress intensity Sm is the Sm listed in Table 1-10.0 and Sy is the yield strength listed in Table 1-2.0 of Appendix I of ASME Code Section Ill.
O Seismic Category I Structures - Amendment 35 3.8-51 i
23A6100 Rn 3 ABWR StandardSafetyAnalysis Report O
Table 3.8-4 Codes, Standards, Specifications, and Regulations Used in the Design and Construction of Seismic Category I internal Structures of the Containment Specification Specification l
Reference or Standard Number Designation Title 1
ACI 301 Specifications for Structural Concrete for Builders 2
ACI 307 Recommended Practice for Concrete Formwork 3
ACI 305 Recommended Practice for Hot Weather Concreting 4
ACI 211.1 Recommended Practice for Selecting Proportions for Normal Weight Concrete 5
ACI 315 Manual of Standard Practice for Detailing Reinforced Normal Weight Concrete 6
ACI 306 Recommended Practice for Cold Weather Concreting 7
ACI 309 Recommended Practice for Consolidation of Concrete 8
ACI 308 Recommended Practice for Curing Concrete 9
ACI212 Guide for use of Admixtures in Concrete 10 ACI 214 Recommended Practice for Evaluation of Compression Test results of Field Concrete 11 ACI 311 Recommended Practice for Concrete inspection 12 ACI 304 Recommended Practice for Measuring, Mixing, Transporting, and Placing Concrete 13 ACI 349 Code Requirements for Nuclear Safety-Related Concrete Structures (as modified by Table 3.8-10) 14 ACI 359 ASME Boiler and Pressure Vessel Code, Section ill, Division 2, Concrete Reactor Vessels and Containments 15 ANSI /AISCN690 Specification for the Design, Fabrication, and Erection of Steel Safety-Related Structures for Nuclear Facilities (as modified by Table 3.8-10) 16 AWS D1.1 Structural Welding Code 17 NClG-02 Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants 18 ANS!/ASME Quality Assurance Program Requirements for Nuclear NOA-1-1986 Facilities 19 (Deleted) 20 NRC Regulatory Quality Assurance Requirements for installation, Guide 1.94 Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants 3.8-S2 Seismic Category I Structures - Amendmer:t 33
23A6100 Rev. 2 ABWR Standard Safety Analysis Report
/m
[
b Q,/
3A.10.2 Enveloping Floor Response Spectra The site-envelope floor response spectra due to the 0.3g SSE are obtained according to the following steps.
(1) The calculated 2,3,5, and 10% damping response spectra of all SASSI 3-D cases are enveloped at all required locations in each of the three directions. In the vertical direction, where applicable, SRSS was used to include the coupling vertical response due to horizontal shaking.
(2) The envelope spectra in the two horizontal directions at each location obtained in step 1 are subsequently enveloped to form the bounding horizontal spectra.
(3) The envelope spectra were subsequently peak broadened byil5%.
The site-envelope peak broadened SSE floor response spectra at critical damping ratios 2,3,5, and 10% for the R/B are shown in Figures SA-128 through SA-165 for the horizontal direction and in Figures 3A-166 through SA-209 for the vertical direction including floor oscillator responses. The vertical responses for the reactor building walls also include the responses of the finite element model (Section SH.1) for the upper
( >'
\\
bound R1 case. The results for the C/B are shown in Figures SA-210 through SA-228.
For seismic design of equipment and piping, the alternative seismic input can be individual floor response spectra of each site condition considered in generating the site-envelope spectra.
Furthermore, vertical response spectra for floor oscillators should be used for the components supported by the floor slabs. However, for the components in the vicinity of the walls or on the walls, the building vertical response spectra at respective elevation can be used.
3A.10.3 Enveloping Maximum Absolute Accelerations The site-envelope absolute acceleration responses (enveloping maximum zero period accelerations) are shown in Tables SA-23a through SA-23d and 3A-24 for the R/B and C/B. The vertical responses include the coupling vertical response due to horizontal shaking where applicable.
3A.10.4 Enveloping Maximum Relative Displacements l
The site-envelope maximum relative displacements with respect to input motion at grade level in the free-field are shown in Tables SA-25a through SA-25d and SA-26 for the R/B and C/B. These results may be used for design of components supported on (o) the surrounding soil medium and connected to the respective building.
v' Seismic Soil Structure Interaction Analysis - Amendment 32 3A-25
23A6900 Rev. 7 ABWR Standard SafetyAnalysis Report O
The site-envelope maximum relative displacements of the nodal points with respect to the base of each respective stick model are shown in Tables 3A-27a through 3A-27d and 3A-28. These results may be used for design of components located within the respective building.
3A.10.5 Summary The site envelope maximum seismic responses presented in Subsection 3A.10 envelop the maximum seismic SSE responses of the ABWR plant structures and components for a wide range of subsurface properties and conditions as well as the effect of concrete cracking and side soil-wall separation. These responses are used to design the ABWR plant structures and components.
3A.11 References 3A-1 Appendix 3A, General Electric Company BWR/6-238 Standard Safety Analysis Report (GESSAR), Docket No. STN 50-447, July 30,1973.
SA-2 Appendix 3A, General Electric Company GESSAR II BWR/6 Nuclear Island Design (22A7007), March 1980.
3A-3 NUREG/CR-1161, Recommended Revisions to Nuclear Regulatory Commission Seismic Design Criteria, May 1980.
3A-4 Lysmer,J., Tabatabaie-Raissi, M., Tajirian, F., Vahdani, S., and Ostadan, F.,
SASSI-A System for Analysis of Soil-Structure Interaction, Report No.
UC/B/GT/81-02, Geotechnical Engineering, University of California, Berkeley, CA, April,1981; also Ostadan, F., Computer Program SASSI, CE (944), Theoretical, User's and Validation Manuals (1991), Bechtel Corporation, San Francisco, California SA-5 Schnabel, P.B., Lysmer,J., and Seed, H.B., SHAKE-A Computer Program for Earthquake Response Analysis of Horizontally Layered Sites, Report No. RC 72-12, Earthquake Engineering Research Center, University of California, Berkeley, CA,1972.
3A-6 Seed, H.B. and Idriss, I.M-Soil Moduli and Damping Factors for Dynamic Response Analysis, Report No. RC 70-10, Earthquake Engineering Research Center, University of California, Berkeley, CA,1970.
3A-7 Idriss, I.M-Response of Soft Soil Sites During Earthquakes, H. Bolton Seed Memorial Symposium Proceedings, Volume 2, Bi Tech Publishers, May 1990.
O 3A-26 Seismic Soil Structure Interaction Analysis - Amendment 3S
23A6100 Rev. 7 ABWR standardsaferyAnalysis Report O
V 0*
TEMPERATURE SENSORS 331 (REFERTOTABLE BELOW)
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TE-003B.F,K,P TE-003D H,M,S 151' TE-004A,EJN TE-004C,G,L,R 196* TE-005A,E,J.N TE-005C,G.L.R 241*
TE-0068,F,K,P TE-006D,H,M S 286 TE-0078.F,K,P TE-007D,H,M,S 331* TE-008A.E.J.N TE-008C.G.L.R NOTE: DIVISIONS 1,11, til AND IV TEMPERATURE SENSORS AT EACH LOCATION SHALL BE SEPARATED BY 15 - 30 CM.
r Figure 7.6-9. Instrumentation Location Definition for the Suppression Pool Temperature Monitoring System
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Figure 7.6-9 Instrumentation Location Definition for the Suppression Pool Temperature Monitoring System AII Other Instrumentation Systems Required for Safety-Amendment 35 7.647
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- JA6100 Rev. 7 ABWR StandardSafetyAnalysis Report s~-~.
I
\\
V The pressure control function provides ABWR automatic load following by forcing the turbine control valves to remain under pressure control supenision, while enabling fast bypass opening for transient events requiring fast reduction in turbine steam flow.
The steam bypass function controls reactor pressure by modulating three automatically operated, regulating bypass valves in response to the bypass flow demand signal. This control mode is assumed under the following conditions:
(a) During reactor vessel heat-up to rated pressure.
(b) While the turbine is brought up to speed and synchronized.
(c) During power operation when reactor steam generation exceeds the turbine steam flow requirements.
(d) During plant load rejections and turbine-generator trips.
(e) During cooldown of the nuclear boiler.
(7) I&C Interface
()
The external signal interfaces for the SB&PC System are as follows:
(a) Narrow range dome pressure signals from the SB&PC System to the Recirculation Flow Control System.
(b) Equivalent load or steam flow feedback signal from the Turbine Control l
System (which is also a triplicated fault-tolerant digital controller).
(c) Signals to and from the main control room.
(d) Bypass hydraulic power supply trouble signal from the Turbine Bypass System to the SB&PC Systern.
(c) Output signals from the SB&PC System to the performance monitoring and control function of the process computer.
(f)
Displayed variables and alanns from the SB&PC System to the main control room panel operator interface.
(g) Narrow and wide range pressure signals, MSIV position signals from the Nuclear Boiler System to the SB&PC System.
(h) Bypass valve position, servo current, position error and valve open and closed signals from the Turbine Bypass Systern.
)
(i)
Emergency bypass valve fast opening signals and bypass valve flow
,U demand signals from the SB&PC System to the Turbine Bypass System.
Control Systems Not Required for Safety-Amendment 3S 7.7-69
23A6900 Rev. 3 ABWR StandardSafetyAnalysis Report O
(j)
Electric power from the non-Class 1E power supply to the SB&PC System.
(k) Pressure setpoint change requests / commands from the turbine master controller, for automatic startup and shutdown sequences.
(1)
Governor-free demand signal to the reactor power compensator in the 3PR system.
(m) Reactor power compensation signal in accordance with speed error from the SB&PC System to the APR System.
(n) Main condenser vacuum low signal from the extraction system to the SB&PC System.
(8) Testability The ITDC input and output communication interfaces are continuously functioning during normal power operation. Abnormal operation of these components can be detected during operation. In addition, the ITDC is equipped with self-test and online diagnostic capabilities for identifying and isolating failure ofinput/ output devices, buses, power supplies, processors, and interprocessor communication paths. These online tests and diagnoses can be performed without disturbing the normal control functions of the SB&PC system.
(9) Environmental Considerations The SB&PC System is not required for safety purposes, nor is it required to operate during or after any design basis accident. The system is required to operate in the normal plant emironment for power generation purposes only.
The SB&PC System equipment is located in the main control room and subject to the normal control room emironment (Section 3.11).
(10) OperatorInformation During operation of the SB&PC System, the operator may obsene the performance of the plant via CRTs on the main control console or on large screen displays in the main control room. As described in (8) above, the self-test provision assures that all transducer / controller failures are indicated to the operator and maintenance personnel. The triplicated logic facilitates online repair of the controller circuit boards.
O 7.7-70 Control Systems Not Required for Safety - Amendment 33
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23A6100 Rev. 4 ABWR StandardSafety Analysis Report O
l The following figures are located in Chapter 21:
Figure 7.7-12 Steam Bypass and Pressure Control System IED (Sheets 1-2) 1 Figure 7.7-13 Steam Bypass and Pressure Control System IBD (Sheets 1-5)
Figure 7.7-14 Fuel Pool Cooling and Cleanup System IBD (Sheets 1-8)
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9 7.7 94 Control Systems Not Required for Safety-Amendment 34 l
l
r 23A6100 Rev. 7 ABWR StandardSafetyAnalysis Report
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(2) Prevent the uncontrolled loss of contaminated pool water to other relatively cleaner locations within the containment or fuel-handling area (3)
Provide liner leak detection and measurement These drainage paths are designed to permit free gravity drainage to the equipment drain tanks or sumps of sufficient capacity and/or pumped to the Radwaste Building.
A makeup water system and pool water levelinstmmentation are provided to replace evaporative and leakage losses. Makeup water during normal operation will be supplied from condensate. The Suppression Pool Cleanup (SPCU) System can also be used as a Seismic Category I source of makeup water in case of failure of the normal Makeup Water System.
Both FPC and SPCU Systems are Seismic Category I, Quality Group C design with the exception of the filter-demineralizer portion, which is shared by both systems.
Following an accident or seismic event, the filter-demineralizers are isolated from the FPC cooling portion and the SPCU System by two block valves in series at both the inlet and outlet of the common filter-demineralizer portion. Seismic Category I, Quality Group C bypass lines are provided on both FPC and SPCU Systems to allow continued flow of cooling and makeup water to the spent-fuel pool.
,m Connections from the RHR System to the FPC System provide a Seismic Category I, safety-related makeup capability to the spent-fuel pool. The FPC System from the RHR connections to the spent-fuel pool are Seismic Category I, safety-related. The manual valves which permit the RHR System to take suction from the spent-fuel storage pool and cool the pool are accessible following an accident in sufficient time to permit an operator to align the RHR System to prevent the spent-fuel stroage pool from boiling.
Furthennore, fire hoses can be used as an alternate makeup source. The fire protection standpipes in the Reactor Building and their water supply (yard main, one diesel engine driven pump and water source) are seismically designed. A second fire pump, driven by a motor powered from the combustion turbine generator,is also provided. Engineering analysis indicates that, under the maximum abnormal heat load with the pool gates closed and no pool cooling taking place, the pool temperature will reach about 100 C in about 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. This provides sufficient time for the operator to hook up fire hoses for pool makeup. The COL applicant will develop detailed procedures and operator training for providing firewater makeup to the spent-fuel pool. See Subsection 9.1.6.9 for COLlicense information.
The FPC components, housed in the Seismic Category I Reactor Building, are Seismic Category I, Quality Group C, including all components except the filter-demineralizer.
These components are protected from the effects of natural phenomena, such as:
earthquake, external flooding, wind, tornado and external missiles. The FPC System is (N
non-safety-related with the exception of the RHR System connections for safety-related
('j makeup and supplemental cooling. The RHR System connections will be protected from the effects of pipe whip, internal flooding, internally generated missiles, and the Fuel Storage and Handling - Amendment 35 9.1-15
23A6100 Rev. 5 ABWR StandardSafety Analysis Report O
effects of a moderate pipe rupture within the vicinity. See Subsection 9.1.6.10 for COL license information.
From the foregoing analysis, it is concluded that the FPC System meets its design bases.
9.1.3.4 Inspection and Testing Requirements No special tests are required because, normally, one pump, one heat exchanger and one filter-demineralizer are operating while fuel is stored in the pool. The spare unit is operated periodically to handle abnormal heat loads or to replace a unit for senicing.
Routine visualinspection of the system components,instnamentation and trouble alarms is adequate to verify system operability.
9.1.3.5 Radiological Considerations The water level in the spent-fuel storage pool is maintained at a height which is sufficient to provide shielding for normal building occupancy. Radioactive particulates removed from the fuel pool are collected in filter-demineralizer units which are located in shielded cells. For these reasons, the exposure of plant personnel to radiation from the FPC System is minimal. Further details of radiological considerations for this and other systems are described in Chapters 11,12, and 15.
9.1.4 Light Load Handling System (Related to Refueling) 9.1.4.1 Design Bases The fuel-handling system is designed to provide a safe and effective means for transporting and handling fuel from the time it reaches the plant until it leaves the plant after post-irradiation cooling. Safe handling of fuel includes design considerations for maintaining occupational radiation exposures as low as reasonably achievable (A1 ARA).
Design criteria for major fuel-handling system equipment are provided in Tables 9.1-2 through 9.1-4, which list the safety class, quality group and seismic category. Where l
applicable, the appropriate ASME, ANSI, Industrial and Electrical Codes are identified.
l Additional design criteria are shown below and expanded further in Subsection 9.1.4.2.
The transfer of new fuel assemblies between the uncrating area and the new-fuel l
l inspection stand and/or the new-fuel storage vault to the fuel storage pool is accomplished using a 49.82 kN auxiliary hoist on the R/B crane equipped with a general purpose grapple. From this point on, the fuel will either be handled by the telescoping grapple (or auxiliary hoist) on the refueling machine.
I l
9.1-16 Fuel Storage and Handling - Amendment 35 l
l
2h E
2 N:lti l
5 Table 9.2-3 Capacity Requirements for Condensate Storage Tank 3
l}
Dead space-top of pool 29,901L' l)
Normal operation variation and receiving volume for plant startup return water 999,240L 3
Minimum storage volume 247,500L 3
g Dead space-middle of pool 129,901L' E
Water source for station blackout 569.567L' Dead space-bottom of pool 129,901L' Total 2,108,321L 3
- These values are based on a bottom area of 130m,
t Water for operation of RCIC is taken from the condensate storage tank and the suppression pool as described in the EPGs of Appendix 18A.
g
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h Table 9.2-4a Reactor Building Cooling Water Division A EI3 C
Emergency Normal (LOCA)
Operating Shutdown at 4 Shutdown at Hot Standby Hot Standby (Suppression Operating Mode / Components Conditions Hours 20 Hours (No Loss of AC)
(Loss of AC)
Pool at 97#C Heat' Fle w
- Heat Flo w Heat Flow Heat Flow Heat Flo w Heat Flo w Essential Emergency Diesel Generator A -
13.40 229 13.40 229 RHR Heat Exchanger A 108.02 1,199 34.75 1,199 25.54 1,199 89.18 1,199 l
Others (essential)'
3.18 205 3.60 205 3.81 205 3.39 205 4.10 205 4.19 205 l
Non-Essential CUW Heat Exchanger
- 20.10 159 159 159 20.10 159 20.93 159 f
7.12 279 7.12 279 7.12 279 7.12 279 7.12 279 9.63 279 FPC Heat Exchanger A Inside Drywell" 5.86 320 5.86 320 5.86 320 5.86 320 3.39 320 Others (non-essential)"
2.64 160 2.64 160 2.64 160 2.64 160 0.84 59 0.75 59 Total Load 38.94 1.123 54.01 2,322 54.01 2,322 38.94 1,123 75.36 2,450 117.23 1,971 E
y 3
- Heat in GJ/h; flow in m /h, sums may not be equal due to rounding.
t HECW refrigerator, CAMS coolers, room coolers (RHR, RCIC, CAMS), RHR motor and seal coolers.
- The heat transferred from the CUW heat exchanger at the start of cooldown is appreciable, but during the criticallast part of a cooldown, the heat removed is very little because the temperature difference between the reactor water and the RCW System is small. Sometimes, the operators may remove the CUW heat exchangers from service during cooldown. Thus, the heat removed varies from about that during normal operation at the start of cooldown to very little at the end of cooldown.
g g
f includes FPC room cooler.
(
" Drywell (A & C) and RIP coclers.
{
t t Instruments and service air coolers; CUW pump cooler, CRD pump oil, and RIP MG sets. A hot water exchanger is in this division which removes Ei.
E heat from the RCW System.
i i
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23A6100 R2v. 2 ABWR standardsafetyAnalysisReport (8) Sampling equipment is designed for flushing and blowdown in order to remove sediment deposits, air and gas packets. Provisions are made to purge sample lines. All flushings are either returned to process or sent to the radwaste system, except where noted.
(9) Provisions are made to sample the bulk volume of tanks. The Standby Liquid Control System storage tank may be sampled from the top opening so that any low points and potential sediment traps can be avoided.
9.3.2.3 Sampling Panels Ditterent process conditions, water quality and analyzing equipment require special treatment ofindividual sample streams.
9.3.2.3.1 Reactor Building Sample Station The Reactor Building Sample Station is located in the Reactor Building. Process samples from the following streams are routed to this panel for analysis:
Reactor water cleanup inlet (hi-temp) a Recctor water citanup inlet (lo-temp) e Reactor water cleanup outlet A m
Reactor water cleanup outlet B a
Control Rod Drive System e
Isolation valves are provided for each reactor water sample line. These valves are operated from the main control room and close automatically upon a LOCA signal.
These valves may be opened for sampling during an accident without removing the LOCA signal.
The Reactor Building Sample Station consists of a sample conditioning rack, constant temperature bath and a chemical fume hood. A continuous purge flow from the selected process stream enters the sample conditioning rack (>500 mL/ min.) and l
through a cooler, if necessary, to reduce temperature to 41 C or lower, then through-one or two flow adjustment valves, depending on inlet pressure. The purge flow is then routed through a chemical fume hood where grab samples can be removed or special measurements made. A small, continuous sample (>100 mL/ min.) is diverted from the main purge flowline through a cooling coillocated in a constant temperature bath, past a temperature gauge, a conductivity cell, flow switch, rotameter and a pressure regulating check valve. The constant temperature bath controls the sample l
temperature to 25'C. A continuous conductivity recorder records the conductivity.
N Process Auxiliaries - Amendment 32 9.3-3
23A6100 Rav. 7 i
ABWR StandardSafetyAnalysis Report 9
Main purge flow and sample flow are in closed lines and are routed through closed drains to the reactor building equipment drain sump.
The Post-Accident Sampling System (PASS) consists of a sample holding rack, sampling rack, sample conditioning rack, local control panel and shielding casks. Samples from the sample conditioning rack, discussed above, are sent to the PASS sample holding rack. A portion of the sample flowis passed through an inline sample vessel. After adequate purging, the sample vessel is isolated and transported to the laboratory. All valves in this operation are operated remotely. The sampling system isolation valves are operated from the main control room and all other valves are operated from the local control panel. After the sample vessel has been isolated and removed, the piping is flushed with demineralized water. The water from purging and flushing is drained to the suppression pool.
The sample holding rack has an enclosure around the sample vessel to contain anyleaks ofliquids or gases. The liquids drain to the radwaste system and the gases go to the reactor building exhaust system.
The PASS isolation valves shall be connected to a reliable source of power that will be available starting at least one hour after a LOCA or ATWS event. The isolation valves shall have Class 1E power and the panels and other equipment shall be powered with i
two offsite power supplies and one onsite power supply.
Gas samples are obtained from a sample line connected to the Containment l
Atmospheric Monitoring System (CAMS). A vacmun pump is provided to transfer the gas sample from a sample holding rack to a sampling rack. The sample is mixed uniformly. In the sampling rack, the gas is passed through and collected in a gas sample holder. After isolation, the gas sample holder is removed and transported to the laboratory for analysis.
The upper limits for activity levels in liquid and gas samples are:
3 Liquid samples 3.70E+10 Bq/cm 3
Gas samples 3.70E+09 Bq/cm l
Means to reduce radiation exposure are provided such as, shielding, remotely operated l
valves, and sample transporting casks. The radiation exposure to any individual shall l
l not be in excess of.05 and.50 Sv to the whole body or extremities, respectively.
Acceptance Criterion II.K.5 of SRP Section 9.3.2 requires the capability of sampling l
liquids of 37.0E+10 Bq/cm. The ABWR design has the capability of sampling liquids of 3
3 3.70E+10 Bq/cm. Sampling will be performed and area radiation measurement will be 9.3-4 Process Auxiliaries - Amendment 35
23A6100Rev 3 ABWR standard safety Analysis aeport a full range ofimportant plant parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded. [I.D.2]
Response
This item is addressed in Subsection 1A.2.3.
l 19A.2.17 Safety System Status Monitoring [ftem (2) (v)]
NRC Position Provide for automatic indication of the bypassed and inoperable status of safety systems.
j
[I.D.3]
Response
The ABWR Standard Plant design fully complies with Regulatory Guide 1.47 (Subsection 7.1.2.10.2). The automatic indication of bypassed and inoperable status of safety systems is, therefore, inherent in the design. Details on human factors are not l
addressed specifically, however, will be addressed by the COL applicant during the conduct of the HSI design implementation process described in Section 18.E.1.
l 19A.2.18 Reactor Coolant System Vents [ Item (2) (vi)]
NRC Position Provide the capability of high point venting of noncondensible gases from the reactor coolant system, and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability oflossef<oolant accident or an unacceptable challenge to containment integrity. [II.B.1]
Response
This issue is addressed in Subsection lA.2.5.
l 19A.2.19 Plant Shielding to Provide Access to Vital Areas and Protect Safoty Equipment f or Post-Accident Operation [ Item (2) (vii)]
NRC Position Perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain TID 14844 source term radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety equipment from the radiation emironment. [II.B.2]
Response
l This item is addressed in Subsection 1A.2.6.
1 19A-7
. Response to CP/ML Rule 10 CFR 60.34tf)- Amendment 33
23A6100 Rw. 7 ABWR StandardSafety Analysis Report 9
19A.2.20 Post-Accident Sampling [ltem (2) (viii)]
NRC Position Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain TID 14844 source term radioactive materials without radiation exposures to any individual exceeding 0.05 Sv to the whole-body or 0.50 Sv to the extremities. Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, iodines and cesiums, and non-volatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations. [II.B.3]
Response
This item is addressed in Subsection I A.2.7.
19A.2.21 Hydrogen Control System Preliminary Design [ item (2) (ix)]
NRC Position Provide a system for hydrogen control that can safely accommodate hydrogen generated by the equivalent of a 100% fuel-clad metal-water reaction. Preliminag design information on the tentatively preferred system option of those being evaluated in paragraph (1) (xii) of 10 CFR 50.34(O is sufficient at the construction permit stage.
The hydrogen control system and associated systems shall provide, with reasonable assurance, that: [II.B.8]
(1) Uniformly distributed hydrogen concentrations in the containment do not exceed 10% during and following an accident that releases an equivalent amount of hydrogen as would be generated from a 100% fuel clad metal-water reaction, or that the post-accident atmosphere will not support hydrogen combustion.
(2) Combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.
(3) Equipment necessay for achieving and maintaining safe shutdown of the plant and maintaining containment integrity will perform its safety function during and after being exposed to the emironmental conditions attendant with the release of hydrogen generated by the equivalent of a 100% fuel-clad metal water reaction including the emironmental conditions created by activation of the hydrogen control system.
(4) If the method chosen for hydrogen controlis a post-accident inerting system, inadvertent actuation of the system can be safely accommodated during plant operation.
19A-8 Response to CP/ML Rule to CFR 50.3Mf)- Amendmert 35
=-
23A6100 Rev. 7 ABWR standardsafetyAnalysis Report Safety issues Index (Continued)
NRC SSAR Title Priority Subsection I.C.8 P;fot-Monitoring of Selected Emergency Procedures for Resolved COL App.
Near-Term Operating License Applicants I.D.1 Control Room Design Reviews Resolved 1 A.2.2 1.D.2 Plant Safety Parameter Display Console Resolved 1 A.2.3 1.D.3 Safety System Status Monitoring Medium 19A.2.17 1.D.5(2) Plant Status and Post-Accident Monitoring Resolved 198.2.65 1.D.5(3) On-Line Reactor Survel!!ance System Near Res.
19B.2.66 1.F.2(2) Include QA Personnel in Review and Approval of Plant Resolved 19A.2.43 Procedures 1.F.2(3) include OA Personnelin All Design, Construction,installa-Resolved 19A.2.43 tion, Testing, and Operation Activities 1.F.2(6) Increase the Size of Licensees' QA Staff Resolved 19A.2.43 1.F.2(9) Clarify Organizational Reporting Levels for the QA Organi-Resolved 19A.2.43 zation 1.G.1 Training Requirements Resolved 1 A.2.4 l.G.2 Scope of Test Program Resolved 19B.2.67 11.B.1 Reacter Coolant System Vents Resolved 1 A.2 5 COL App.
II.B.2 Plant Shielding to Provide Access to Vital Areas and Protect Safety Resolved 1 A.2.6 Equipment for Post-Accident Operation 11.B.3 Post-Accident Sampling Resolved 1A.2.7 II.B.4 Training for Mitigating Core Damage Resolved COL App.
II.B.8 Rulemaking Proceeding on Degraded Core Accidenta Resolved 19A.2.1 19A.2.2 19A.2.43 19A.2.45 ll.D.1 Testing Requirements Resolved 1 A.2.9 II.D.3 Relief and Safety Wlve Position Indication Resolved 1 A.2.1C ll.E.4.1 Dedicated Penetrations Resolved 1 A.2.13 II.E.4.2 Isolation Dependability Resolved 1 A.2.14 '
li.E.*.4 Purging Recolved 19A.2.27 II.E.6.1 Test Adequacy Study Resolved 19B.2.68 COL App.
II.F.1 Additional Accident Monitoring instrumentation Resolved 1 A.2.15 ll.F.2 Identif' cation of and Recovery from Conditions Leading to inade-Resolved 1 A.2.16
{
quate Core Cwling
\\
ll.F.3 Instruments for Monitoring Accident Conditions Resolved 1 A.2. i7 II.J.4.1 Revise Deficiency Reporting Requirements Resolved COL App.
Resolution of Applicable Unresolved Safety issues and Generic Safety Issues - Amendment 3S 198-5
=
r--
23A6100 Rev. 4 ABWR Standard SafetyAnalysis Report O
Safety Issues Index (Continued)
NRC SSAR Title Priority Subsection ll.K.1(5) Safety-Related Valve Position Description Resolved 1 A.2.18 18.8.7 l'.K.1(10) Review and Modify Procedures for Removing Safety-Resolved 1 A.3.2 Related Systems from Service ll.K.1(13) Propose Technical Specifications Changes Reflecting Resolved 19B.2.69 Implementation of All Bulletin Items ll.K.1(22) Describe Automatic and Manual Actions for Proper Resolved 1 A.2.20 Functioning of Auxiliary Heat Removal Systems When FW System Not Operable ll.K.1(23) Describe Uses and Types of RV LevelIndication for Resolved 1 A.2.21 Automatic and Manual Initiation Safety Systems ILK.3(3) Report Safety and Relief Valve Failures Promptly and Resolved 1 A.3.4 Challenges Annually ll.K.3(11) Cont;ol Use of PORV Supplied by Control Components, Resolved 198.2.70 inc. Until Further Review Complete ll.K.3(13) Separation of HPCI and RCIC Sysem initiation Levels Resolved 1 A.2.22 II.K.3(15) Modify Break Detection Logic to Prevent Spurious Resolved 1 A.2.23 Isolation of HPCI and RCIC Systems COL App.
II.K.3(16) Reduction of Challenges and Failures of Relief Valves-Resolved 1 A.2.24 Feasibility Study and System Modification ll.K.3(17) Report and Outage of ECC Systems-Licensee Report Resolved 1 A.2.25 and Technical Specification Changes ll.K.3(18) Modification of ADS Logic-Feasibility Study and Resolved 1 A.2.26 Modification for increased Diversity for Some Event Sequences ll.K.3(21) Restart of Core Spray and LPCI Systems on Low Level-Resolved 1 A.2.27 Design and Modification ll.K.3(22) Automatic Switchover of RCIC System Suction-Verify Resolved 1 A.2.28 Procedures and Modify Design ll.K.3(24) Confirm Adequacy of Space Cooling for HPCI and RCIC Resolved 1 A.2.29 Systems ll K.3.(25) Effect of Loss of AC Power on Pump Seals Resolved 1 A.2.30 f
II.K.3(27) Provide Common Reference Level for Vessel Level Resolved 1 A.2.21 instrumentation ll.K.3(28) Study and Verify Qualification of Accumulators on ADS Resolved 1 A.2.31 Valves ll.K.3(30) Revised Small-Break LOCA Methods to Show Resolved 1 A.2.32 Compliance with 10 CFR 50, Appendix K 19B-6 Resolution of Applicable Unresolved Safety Issues and Generic Safety Issues - Amendment 34
l 23A6100 Rev. 7 ABWR standardsafetyAnalysis Report performance during an accident. The purpose of this issue is to improve the accuracy of measurement of airborne iodine concentrations.
j Acceptance Criteria Airborne iodine concentrations must be accurately determined throughout the plant under accident conditions.
Resolution Item III.D.3.3(1) which concerns in-plant radiation monitoring is resolved in Subsection 12.3.4 which also references each area detectorlocation on the plantlayout i
drawings for each building (Figures 12.3-56 through 12.3-73) as well as the specific area radiation channels for each building, the detector map location, the channel sensitivity range, and the local alarm areas (Tables 12.3-3 through 12.3-7).
References 196.2.72-1 NUREG-0660, NRC Action Plan Developed as a Result of the TMI-2 Accident, l
U.S. NRC, May 1980.
19B.2.12-2 NUREG-0737, Claripcation of TMI Action Plan Requirements, U.S. NRC, November 1980.
V 19B.2.73 Ill.D.3.3(2): Set Criteria Requiring Licensees to Evaluate Need for Additional Survey Equipment issue NUREG-0660 (Reference 19B.2.73-1) is a guideline to improve nuclear power plant worker radiation protection to allow workers to take effective action to control the course and consequences of an accident, as well as to keep exposures as low as reasonably achievable (AIARA) during normal operation and accidents.
i Acceptance Criteria This issue required the NRR to set criteria requiring licensees to evaluate in their plants the need for additional survey equipment and radiation monitors in vital areas and requiring, as necessary, installation of area monitors with remote readout. The NRR evaluated the need to specify the minimum types and quantities of portable monitoring instrumentation, including very high dose rate survey instruments. Operating reactors were reviewed for conformance with Standard Review Plan (SRP) Section 12.3.4, Area Radiation and Airborne Radioactivity MonitoringInstrumentation. The NRR revised the SRP Sections 12.5 and 12.3.4 to incorporate additional monitor requirement criteria.
Resolution Item III.D.3.3(2) which concerns licensees evaluate the need for additional radiation
[_
survey equipment is resolved in Subsection 12.3.4. This item also concerned the need
\\
to specify the minimum types and quantities of portable monitoring instnunentation, including very high dose rate survey instruments. As noted in Subsections 12.5.2, i
Resolution of Applicable Unresolved Safety Issues and Generic Safety Issues - Amendment 35 198-123
23A6900 Rev. 7 i
ABWR StandardSafety Analysis Report i
0 19A.2.39 and 19A.S.5, COL applicants will provide the portable instruments in operating reactors that accurately measure radio-iodine concentration in plant areas under accident conditions.
Refercnces 19B.2.73-1 NUREG-0660, NRC Action Plan Developed as a Result of the Thil-2 Accident, l
U.S. NRC, May 1980.
198.2.73-2 NUREG-0737, Clarification of Tbil Action Plan Requirements, U.S. NRC, November 1980.
198.3 COL License information 198.3.1 COL Applicant Safety issues The COL applicant shall provide resolutions for the issues identified as COL applicant in the Safety Issues Index consistant with the documentation format discussed in Subsection 19B.I.l.
19B.3.2 Testing of isolators As established in Section 7A.3, the COL applicantis required to establish a test program for fiber optic-type isolators used between safety-related and non-safety-related systems.
If other types ofisolators are used (those subject to electricalleakage due to maximum credible electrical faults), the COL applicant shall implement the required testing, inspection, and replacement isolators when needed (See Subsection 19B.2.53).
O 198-124 Resolution of Applicable Unresolved Safetyissues and Generic SafetyIssues - Amendment 3S
23A6100 Rev. 5 ABWR standardsafetyAnstrsisneport Table 19E.2-1 Potential Suppression Pool Bypass Lines Pathway Basis For Exclusion Number Size (mm)
Isolation (See Description of Lines From To (1 in. = 25.4 mm)
Valves Notes)
Main Steam Line Drain 1
RPV ST 550 CK,CK Reactor Inst. Lines 37 RPV RB 6
CK, MA 1
HPCF Discharge 2
RPV RB 200 CK, MO HPCF Equalizing 2
RPV RB 20 MO, MO HPCF Suction 2
'2 Supp Pool Instrumentation 6
CK 2
SLC Injection 1
RPV RB 40 CK,CK RCIC Steam Supply 1
RCIC Discharge 1
RCIC Min. Flow 1
RCIC Suction 1
SP RB-200 MO 2
RCIC Turbine Exhaust 1
RCIC Turb. Exh Vac Bkr 1
RCIC Vac Pump Discharge 1
RHR LPFL Discharge 2
RPV RB 250 CK, MO RHR Equalizing Lines 2
RPV RB 20 MO, MO RHR Wetwell Spray 2
WW RB 100 MO 2,4 RHR Drywell Spray 2
l CUW Suction
CUW Head Spray Line i
MO 1
CUW Instrument Lines 4
CK j
Post Accident Sampling 4
<1 CK,CK 1
(
RIP Cooling Water 4
LDS Instruments 9
CK Deterministic Analysis of Plant Performance - Amendment 35 19E.2-159
23A6100 Rev. 7 ABWR StandardSafetyAnalysis Report O
Table 19E.2-1 Potential Suppression Pool Bypass Lines (Continued)
Pathway Basis For Exclusion Number Size (mrn)
Isolation (See Description of Lines From To (1 in. = 25.4 mm)
Valves Notes)
LDS Instruments 9
CK SPCU Suction 1
SPCU Return 1
l Cont. Atmosphere Monitor 6
Drywell Sump Drains 2
DW RB 100 MO, MO HVCW/RBCW Supply 4
HVCW/DWCW Return 4
DW Exhaust /SGTS 1
Wetwell Vent to SGTS 1
DW Purge 1
- nerting Makeup 1
DW WW 50 AO,AO WW Inerting/ Purge 1
Instrument Air (and Service 2
Alr)
SRV Pneumatic Supply 3
Flammability Control 2
DW RB 100 (AO, MO) 3 ADS /SRV Discharge 8
RPV WW 300 RV ACS Supply 2
DW WW 550 AO, AO WW/DW Vacuum Breaker 8
DW WW 500 CK Miscellaneous Leakage 1
Access Tunnels 2
NOTES:
Legends and Acronyms Pathway Source (From)
Termination (To)
RPV Reactor Pressure Vessel WW Wetwell DW Drywell RB Reactor Building SP Suppression Pool WW Wetwell ST Steam Tunnel 19E.2-160 Deterministic Analysis of Plant Performance - Amendment 35
m 23A6100 Rev. 7 ABWR standardsafetyAnalysis Report
!\\
Table 19E.2-21 Summary of Bypass Probabilities Figures 19E.2-19a Flow Split Bypass Probability Bypass Bypass to Pathway Fraction -
Equation Probability Fraction 19E.219k Lines from the RPV Main Steam 6.7E-1 4*P1 *(P3
- P4+P5) 1.6E-6 1.1 E-6 A
Main Steam Leakage 2.2E-5 4
- P2*(P3
- P4+ P5) 1.1 E-2 2.5E-7 A
Feedwater 5.2E-1 2*P9'P9'P15 2.4E-8 1.3E-8 B
Reactor Inst. Lines 3.1 E-5 30*P13*P9 6.0E-5 1.9E-9 D
HPCF Discharge 1.1 E-1 2*P9'P10*P14 1.3E-7 1.5E-8*
C HPCF Equalizing Line 1.0E-3 2*P10*P11*P13 6.7E-8 6.7E-11
- C SLC injection 3.0E-3 1*P9'P13 3.6E-7 1.1 E-9 B
RCIC Steam Supply 6.9E-2 1*P8'P14 5.2E-9 3.6E-10 E
LPFL Discharge 1.7E-1 2*P9'P10*P15 6.7E-8 1.1 E-8*
C LPFL Equalizing Line 1.0E-3 2*P10*P11*P13 6.7E-8 6.7E-11*
C CUW Suction 1.2E-1 1*P8"P14 5.2E-9 6.2E-10 E
CUW Inst Lines 3.1 E-5 4*P13*P9 8.1 E-6 2.5E-10 D
Post Acc Sampling 1.0E-3 4*P8"P11 3.6E-7 3.7E-10 J
LDS Instruments 3.1 E-5 9'P13*P9 1.8E-5 5.7E-10 D
SRV Discharge 6.9E-2 8*P14 1.3E-4 8.8E-6 K
Lines from the Drywell l
Cont Atmos Monitor 8.9E-4 6*P8'P13 2.6E-6 2.3E-9 J
LDS Samples 1.7E-3 2*P8'P11 1.6E-7 2.6E-10 J
Drywell Sump Drain 3.0E-2 2*P8'P13 1.6E-7 4.7E-9 J
DW Purge 5.4E-1 1*P6*P11 1.1 E-6 6.2E-7 I
Inerting Makeup 1.2E-2 1*P6 7.4E-4 8.9E-6 I
ACS Supply 7.5E-1 2*P12*P6 1.5E-6 1.1 E-6 H
t WW-DW Vac Bkr 2.6E-1 8'P9 6.7E-2 1.7E-2 G
Grand Total excluding vacuum breaker 2.1 E-5 Goal 8.4E-4 g
(
- These lines may be excluded for station blackout events.
g t Addressed on Containment Event Trees.
Deterministic Analysis of Plant Performance - Amendment 35 19E.2175
23A6100 Rw. 4 1
ABWR Standard Safety Analysis Report O
Table 19E.2-22 NUREG/CR-4551 Grand Gulf APET Events by Category Event Number' Description Plant Damage State Grouping Events 1
initiating Event Type 2
Station Blackout 3
DC Power Availability 4
S/RV Fails to Reclose l
5 HPCI Failure 6
RCIC Failure Initially 7
CRD Injection Failure 8
Condensate System Failure 9
LPCS/LPCI Systems Failure 10 RHR Failure 11 Service Water /LPCI Crosstie Failure 12 Fire Protection Crosstie Failure 13 Containment Spray Failure l
14 Vessel Depressurization l
15 Time Core Damage 20 Plant Damage State Summary Structural Capacity / Initial Containment Status 16 Containment Isolation (Pre-existing Leakage) 17 Extent of Pool Bypass initially 18 Containment Capacity (Quasi-static / Dynamic Loading) 19 Drywell Capacity (Quasi-static / Dynamic Loading) l Systems Behavior / Operator Actions 21 Ignitors Turned On Before Core Damage l
22 Containment Vented Before Core Damage 23 SRV Vacuum Breakers Stick Open 26 RV Pressure During Core Damage 27 Status of Hydrogen Ignitors Before Vessel Breach l
28 RV injection Restored During Core Damage 30 Containment Spray Status 53 Upper Pool Dump 19E.2176 Deterministic Analysis of Plant Performance - Amendment 34 l
l
ABWR CERTIFIED DESIGN MATERIAL O
Revision 6-Page Change Instruction The following pages have been changed, please make the changes in your copy of the Certified Design Material. Pages are listed below as page pairs (front and back).
REMOVE PAGE NO.
ADD PAGE NO.
TOC-v/si TOC-v/si 2.11.3 - 5 thru 8 2.11.3 - 5 thru 8 3.2-13,14 3.2-13,14 3.6-1,2 OV i
vs MR72in@lM25%gl (1)
A
4 O
CDM MODIFICATION PAGES i
25AS447 Rev. 6 ABWR certisedoesiga usterial C
t 3.0 Additional Certified Design Matei al 3.1 Human Factors Engineering 3.2 Radiation Protection 3.3 Piping Design 3.4 Instrumentation and Control 3.5 Initial Test Program 3.6 Design Reliability Assurance Program 4.0 Interface Requirements 4.1 Ultimate Heat Sink 4.2 Offsite Power System (2.12.1) 4.3 Makeup Water Preparation System 4.4 Potable and Sanitary Water System (2.11.23) 4.5 Reactor Senice Water System (2.11.9) 4.6 Turbine Senice Water System (2.11.10) 4.7 Communication System (2.12.16) 4.8 Site Security l
4.9 Circulating Water Eptem (2.10.23) 4.10 Heating, Ventilating and Air Conditioning (2.15.5) 5.0
. Site Parameters l
Appendices Appendix A Legend For Figures Appendix B Abbreviations and Acronyms
]
Appendix C -
Conversion to ASME Standard Units 1
- Underlined sections -Title only, no entry for design certification.
Table of Contents VM
25AS447 Rev. 6 ABWR certifiedoesign Material fm k
I v
MUWP RCW l
RCW OTHERS OTHERS RCW 3
P SPCU RCW 3
-]
~
3 (Reactor Building) 1 W
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J e-i M R SURGE TANK l
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L l
(Reactor Building)
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NNS 3
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~ "
~
(Reactor Building)
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NON SAF E T D HXs r_______
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(%tml Buddmo) 3RCW RCW PUMP
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FROM I
(Control Building)
RSW M TO RSW>RSW NOTES:
- 1. ALL ELECTRICAL POWER LOADS FROM THE CLASS 1E COMPONENTS SHOWN
j']
ON THIS FIGURE ARE POWERED FROM CLASS 1E DIVISION I EXCEPT FOR THE j
OUTBOARD CONTAINMENT ISOLATION VALVE. WHICH IS POWERED FROM DMSION 11.
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Figure 2.11.3a Reactor Building Cooling Water System (RCW-A)
Reactor Building Cooling Water System 2.11.3-5 l
l i
25AS447 Rev. 6 ABWR certinescesignuaterias O
MUWP RCW OTHERS RCW RCW OTHERS 3
3 SPCU RCW FE T
R 3
(Reactor Building) l
-_J L
I l
SURGE TANK r-- - - - - - - 1 R
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(Reactor Building)
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HECW M
_ _ _ _ _ _ _j V
_ y Fuel Pool Cooling HX and Room Coolers L
(Reactor Building) u 5
d 3 NNS NNS 3 I
_ _ _ _ _ _ _,I t OTHER (SAFETY RELATED) HXs i'
l (Reactor and Control Building)
RCW OTHERS OTHERS RCW NNS NNS FE R
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q CUW PUMP
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l h
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RCW HX u
(Cont oi Buildino) 3 RCW R
TO RSW R
RCW HX C
W l
(Control BuildinC) 3RCW M
SW RCW PUMP TO RSW (Control Building) l NOTES:
- 1. THIS DIVISION IS POWERED FROM CLASS 1E DIVISION 11,
- = PRIMARY CONTAINMENT EXCEPT FOR THE CONTAINMENT OUTBOARD ISOLATION V ALVE, WHICH IS POWERED FROM DIVISION til Figure 2.11.3b Reactor Building Cooling Water System (RCW-B) l l
2.11.3-6 Reactor Building Cooling Water System
25AS447 Rev. 6
.ABWR certisedoesiga nterial s'
V MUWP RCW 3
RCW OTHERS OTHERS RCW SPCU RCW P
3 3
3 FE l
RHRN
]T M
[ _ _ <a ~< a" mat _
g,R;e =,, g i
--9
~
3RCW i
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HECW OTHER (SAFETY RELATED) HXs (Reactor and Control Building) l RCW OTHERS OTHERS RCW NNS NNS FE CRD PUMP "l NNS I (Reactor Building)
~
1 NNS l 3 3
g
[
l
______d NON-SAFETY-RELATED HXs y,
(Radwaste & Turbine Building)
L J
l I
L________j
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p_d NON-SAFETY-RELATED HXs (Turbine Building)
L IP m-a RCW HX i
FROM (Control Building) 3RCW C
RSW -->J RSW TO RSW RCW PUMP (Control Building)
RCW HX
" pqoy (Control Building) 3RCW RSW _ -->J l
gSW TO RSW i
RCW HX C
(Control Building) 3RCW FROM RSW-- ->l Q SW RCW PUMP TO RSW (Control Building)
NOTES:
- 1. ALL ELECTRICAL POWER LOADS FOR THE CLASS 1E COMPONENTS
['N SHOWN ON THIS FIGURE ARE POWERED FROM CLASS 1E DIVISION lit t
)
%J Figure 2.11.3c Reactor Building Cooling Water System (RCW-C)
Reactor Building Cooling Water System 2.11.3 7
9 h
~
c in e
LOCAL AREA MAIN CONTROL ROOM LOCAL AREA Plant Sensors Device Actuators RCW Manual Pump and Valve Controls
'r r
7 SSLC PROCESSING RCW SYSTEM LOGIC Aut matic
- Sensor Channel Trip Dedslon
-LOM Angnmed 4
- System Coincidence Trip Dedslon RCW
-Surge Tank LevelControl RCW Surge Tank Level
- Control and Interlock Logic
- DMalon-of-Sensors Bypass
-Stop Flow to Non-Safety-Related
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- Calibration, Self-Diagnosis Components g
4 JL
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LOCA Signal l RCW Manual Pump and Valve Actuation RSW LOCA S!gnalI IRHR
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- 1. Diagram represents one of three RCW divisions.
g-g
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l
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Figure 2.11.3d Reactor Building Cooling Water System Control interface Diagram g.
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25A5447 Rett. 6 ABWR certisedoesign uaterial O
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2SA5447 Rev. 6 ABWR certisedoesign uaterior pv) i 3.6 Design Reliability Assurance Program Design Description The Design Reliability Assurance Program (D-RAP) is a program that will be performed during the detailed design and equipment specification phase prior to initial fuelload.
The D-RAP evaluates and prioritizes the structures, systems and components (SSCs) in the design, based on their degree of risk significance. The D-RAP will identify the dominant failure modes for the risk-significant SSCs. The D-RAP will also identify the key assumptions and ri.sk insights for the risk-significant SSCs.
The D-RAP scope includes risk-significant SSCs as determined by probabilistic, deterministic, or other methods used for design certification to identify and prioritize risk-significant SSCs.
The D-RAP purpose is to provide reasonable assurance that the plant design proceeds in a manner that is consistentwith the original bases and design assumptions for the risk insights for the risk-significant SSCs.
The D-RAP objectives are to provide reasonable assurance that the plant is designed
(
such that: (1) it is consistent with the assumptions and risk insights for these risk-significant SSCs, (2) the risk-significant SSCs will not degrade to an unacceptable level during their design life, (3) the frequency of transients that challenge these SSCs will be acceptably low, and (4) these SSCs will function reliably when challenged.
Inspections, Tests, Analyses and Acceptance Criteria Table 3.6 provides a definition of the inspections, tests, analyses, and at,sociated acceptance criteria, which will be performed for Advanced Boiling Water Reactor (ABWR)D-RAP.
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Design Reliability Assurance Program 3.6-1
Table 3.6 Design Reliability Assurance Program b
4 03 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria 1.
The Design Reliability Assurance Program 1.
Inspections of the design reliability 1.
(D RAP) includes: scope, purpose, assurance program will be conducted.
Documentation existe that describes ob;jectives; the process used to evaluate a.
the scope, purpose, and obj,ectives of and prioritize the structures, systems and D-RAP used dunng plant des,gn, and i
components (SSCs); and the list of SSCs concludes that the detailed design of designated as risk-significant. For those nsk-s,gnd, cant SSCs,is consistent i
i SSCs designated as risk-significant, the with the D-RAP Design Desenption.
process used to determine dominant failure modes considered industry b.
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