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{{#Wiki_filter:25A5675AA Revision 7 October 2019 ABWR Design Control Document Tier 1 Copyright 1994, 2010, 2016, 2019 GE-Hitachi Nuclear Energy Americas LLC All Rights Reserved
{{#Wiki_filter:}}
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please read carefully The design, engineering, and other information contained in this document is furnished by GE-Hitachi Nuclear Energy Americas LLC (GEH) for the purpose of supporting the GEH Certification Renewal Application to the United States Nuclear Regulatory Commission (NRC) for renewal of the certification of the ABWR nuclear plant design pursuant to Title 10 Code of Federal Regulations (10 CFR) Part 52.
The only undertakings of GEH with respect to information in this document are contained in the contracts between GEH and its customers or participating utilities, and nothing contained in this document shall be construed as changing that contract. The use of this information by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
No use of or right to copy any of this information contained in this document, other than by the NRC and its contractors in support of GEH's application, is authorized except by contract with GEH, as noted above. The information provided in this document is part of and dependent upon a larger set of knowledge, technology, and intellectual property rights pertaining to standardized, nuclear powered, electric generating facilities that utilize the design certification, as designed and certified to U.S. Codes, Standards, and Regulations by GEH, and referred to as the ABWR nuclear power plant design. Without access and a GEH grant of rights to that larger set of knowledge, technology, and intellectual property rights, this document is not practically or rightfully usable by others, except by the NRC or through contractual agreements with Combined License Applicants and Licensees or customers and participating utilities as set forth in the previous paragraph.
Copyright 2019, GE-Hitachi Nuclear Energy Americas LLC, All Rights Reserved Table of Contents                                                                                                  i
 
25A5675AA Revision 7 ABWR                                                                                    Design Control Document/Tier 1 Tier 1 Table of Contents 1.0 Introduction................................................................................................................... 1.0-1 1.1    Definitions ........................................................................................................ 1.1-1 1.2    General Provisions ............................................................................................ 1.2-1 2.0 Certified Design for ABWR Systems ........................................................................... 2.0-1 2.1    Nuclear Steam Supply Systems 2.1.1        Reactor Pressure Vessel System ..................................................... 2.1-1 2.1.2        Nuclear Boiler System .................................................................. 2.1-13 2.1.3        Reactor Recirculation System....................................................... 2.1-32 2.2    Control and Instrument Systems 2.2.1        Rod Control and Information System ............................................. 2.2-1 2.2.2        Control Rod Drive System.............................................................. 2.2-6 2.2.3        Feedwater Control System............................................................ 2.2-12 2.2.4        Standby Liquid Control System.................................................... 2.2-16 2.2.5        Neutron Monitoring System ......................................................... 2.2-24 2.2.6        Remote Shutdown System ............................................................ 2.2-30 2.2.7        Reactor Protection System............................................................ 2.2-36 2.2.8        Recirculation Flow Control System.............................................. 2.2-43 2.2.9        Automatic Power Regulator System............................................. 2.2-49 2.2.10        Steam Bypass and Pressure Control System................................. 2.2-52 2.2.11        Process Computer System ............................................................ 2.2-55 2.2.12        Refueling Platform Control Computer ......................................... 2.2-57 2.2.13        CRD Removal Machine Control Computer.................................. 2.2-58 2.3    Radiation Monitoring Systems 2.3.1        Process Radiation Monitoring System............................................ 2.3-1 2.3.2        Area Radiation Monitoring System ................................................ 2.3-8 2.3.3        Containment Atmospheric Monitoring System ............................ 2.3-10 2.4    Core Cooling Systems 2.4.1        Residual Heat Removal System...................................................... 2.4-1 2.4.2        High Pressure Core Flooder System ............................................. 2.4-20 2.4.3        Leak Detection and Isolation System ........................................... 2.4-32 2.4.4        Reactor Core Isolation Cooling System........................................ 2.4-39 2.5    Reactor Servicing Equipment 2.5.1        Fuel Servicing Equipment .............................................................. 2.5-1 2.5.2        Miscellaneous Servicing Equipment .............................................. 2.5-2 2.5.3        Reactor Pressure Vessel Servicing Equipment ............................... 2.5-3 2.5.4        RPV Internal Servicing Equipment ................................................ 2.5-4 2.5.5        Refueling Equipment ...................................................................... 2.5-5 2.5.6        Fuel Storage Facility....................................................................... 2.5-7 ii                                                                                                                    Table of Contents
 
25A5675AA Revision 7 ABWR                                                                              Design Control Document/Tier 1 Table of Contents (Continued) 2.5.7    Under-Vessel Servicing Equipment................................................ 2.5-9 2.5.8    CRD Maintenance Facility ........................................................... 2.5-10 2.5.9    Internal Pump Maintenance Facility............................................. 2.5-11 2.5.10    Fuel Cask Cleaning Facility.......................................................... 2.5-12 2.5.11    Plant Start-up Test Equipment ...................................................... 2.5-13 2.5.12    Inservice Inspection Equipment.................................................... 2.5-14 2.6    Reactor Auxiliary Systems 2.6.1    Reactor Water Cleanup System ...................................................... 2.6-1 2.6.2    Fuel Pool Cooling and Cleanup System ......................................... 2.6-6 2.6.3    Suppression Pool Cleanup System ................................................. 2.6-9 2.7    Control Panels 2.7.1    Main Control Room Panels............................................................. 2.7-1 2.7.2    Radioactive Waste Control Panels.................................................. 2.7-7 2.7.3    Local Control Panels....................................................................... 2.7-8 2.7.4    Instrument Racks .......................................................................... 2.7-10 2.7.5    Multiplexing System..................................................................... 2.7-11 2.7.6    Local Control Boxes ..................................................................... 2.7-16 2.8    Nuclear Fuel 2.8.1    Nuclear Fuel.................................................................................... 2.8-1 2.8.2    Fuel Channel ................................................................................... 2.8-2 2.8.3    Control Rod..................................................................................... 2.8-3 2.8.4    Loose Parts Monitoring System...................................................... 2.8-4 2.9    Radioactive Waste System 2.9.1    Radwaste System ............................................................................ 2.9-1 2.10    Power Cycle Systems 2.10.1    Turbine Main Steam System......................................................... 2.10-1 2.10.2    Condensate Feedwater and Condensate Air Extraction System ... 2.10-5 2.10.3    Heater Drain and Vent System ................................................... 2.10-11 2.10.4    Condensate Purification System ................................................. 2.10-12 2.10.5    Condensate Filter Facility ........................................................... 2.10-15 2.10.6    Condensate Demineralizer .......................................................... 2.10-16 2.10.7    Main Turbine .............................................................................. 2.10-17 2.10.8    Turbine Control System.............................................................. 2.10-21 2.10.9    Turbine Gland Seal System ........................................................ 2.10-22 2.10.10  Turbine Lubricating Oil System ................................................. 2.10-25 2.10.11  Moisture Separator Heater .......................................................... 2.10-26 2.10.12  Extraction System ....................................................................... 2.10-27 2.10.13  Turbine Bypass System .............................................................. 2.10-28 2.10.14  Reactor Feedwater Pump Driver................................................. 2.10-30 2.10.15  Turbine Auxiliary Steam System................................................ 2.10-31 Table of Contents                                                                                                                iii
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 Table of Contents (Continued) 2.10.16    Generator .................................................................................... 2.10-32 2.10.17    Hydrogen Gas Cooling System................................................... 2.10-33 2.10.18    Generator Cooling System.......................................................... 2.10-34 2.10.19    Generator Sealing Oil System..................................................... 2.10-35 2.10.20    Exciter ......................................................................................... 2.10-36 2.10.21    Main Condenser .......................................................................... 2.10-37 2.10.22    Off-Gas System........................................................................... 2.10-39 2.10.23    Circulating Water System ........................................................... 2.10-42 2.10.24    Condenser Cleanup Facility........................................................ 2.10-45 2.11 Station Auxiliary Systems 2.11.1    Makeup Water (Purified) System ................................................. 2.11-1 2.11.2    Makeup Water (Condensate) System ........................................... 2.11-3 2.11.3    Reactor Building Cooling Water System...................................... 2.11-6 2.11.4    Turbine Building Cooling Water System ................................... 2.11-19 2.11.5    HVAC Normal Cooling Water System ...................................... 2.11-22 2.11.6    HVAC Emergency Cooling Water System ................................ 2.11-26 2.11.7    Oxygen Injection System............................................................ 2.11-32 2.11.8    This section not used................................................................... 2.11-33 2.11.9    Reactor Service Water System ................................................... 2.11-34 2.11.10    Turbine Service Water System ................................................... 2.11-40 2.11.11    Station Service Air System ......................................................... 2.11-43 2.11.12    Instrument Air System ................................................................ 2.11-46 2.11.13    High Pressure Nitrogen Gas Supply System .............................. 2.11-49 2.11.14    Heating Steam and Condensate Water Return System ............... 2.11-54 2.11.15    House Boiler ............................................................................... 2.11-55 2.11.16    Hot Water Heating System ......................................................... 2.11-56 2.11.17    Hydrogen Water Chemistry System ........................................... 2.11-57 2.11.18    Zinc Injection System ................................................................. 2.11-58 2.11.19    Breathing Air System.................................................................. 2.11-59 2.11.20    Sampling System ........................................................................ 2.11-60 2.11.21    Freeze Protection System............................................................ 2.11-62 2.11.22    Iron Injection System.................................................................. 2.11-63 2.11.23    Potable and Sanitary Water System............................................ 2.11-64 2.12 Station Electrical Systems 2.12.1    Electrical Power Distribution System........................................... 2.12-1 2.12.2    Unit Auxiliary Transformer ........................................................ 2.12-17 2.12.3    Isolated Phase Bus ...................................................................... 2.12-18 2.12.4    Nonsegregated Phase Bus ........................................................... 2.12-19 2.12.5    Metal Clad Switchgear................................................................ 2.12-20 2.12.6    Power Center............................................................................... 2.12-21 2.12.7    Motor Control Center.................................................................. 2.12-22 2.12.8    Raceway System ......................................................................... 2.12-23 iv                                                                                                          Table of Contents
 
25A5675AA Revision 7 ABWR                                                                                    Design Control Document/Tier 1 Table of Contents (Continued) 2.12.9    Grounding Wire .......................................................................... 2.12-24 2.12.10    Electrical Wiring Penetration...................................................... 2.12-25 2.12.11    Combustion Turbine Generator .................................................. 2.12-27 2.12.12    Direct Current Power Supply...................................................... 2.12-29 2.12.13    Emergency Diesel Generator System ......................................... 2.12-38 2.12.14    Vital AC Power Supply .............................................................. 2.12-43 2.12.15    Instrument and Control Power Supply........................................ 2.12-50 2.12.16    Communication System .............................................................. 2.12-56 2.12.17    Lighting and Servicing Power Supply ........................................ 2.12-58 2.13    Power Transmission 2.13.1    Reserve Auxiliary Transformer .................................................... 2.13-1 2.14    Containment and Environmental Control Systems 2.14.1    Primary Containment System ....................................................... 2.14-1 2.14.2    Containment Internal Structures ................................................... 2.14-7 2.14.3    Reactor Pressure Vessel Pedestal ................................................. 2.14-8 2.14.4    Standby Gas Treatment System .................................................... 2.14-9 2.14.5    PCV Pressure and Leak Testing Facility .................................... 2.14-14 2.14.6    Atmospheric Control System...................................................... 2.14-15 2.14.7    Drywell Cooling System............................................................. 2.14-20 2.14.8    Flammability Control System ..................................................... 2.14-23 2.14.9    Suppression Pool Temperature Monitoring System ................... 2.14-28 2.15    Structures and Servicing Systems 2.15.1    Foundation Work .......................................................................... 2.15-1 2.15.2    Turbine Pedestal ........................................................................... 2.15-2 2.15.3    Cranes and Hoists ......................................................................... 2.15-3 2.15.4    Elevators ....................................................................................... 2.15-5 2.15.5    Heating, Ventilating and Air Conditioning Systems .................... 2.15-6 2.15.6    Fire Protection System................................................................ 2.15-48 2.15.7    Floor Leakage Detection System ................................................ 2.15-53 2.15.8    Vacuum Sweep System .............................................................. 2.15-54 2.15.9    Decontamination System ............................................................ 2.15-55 2.15.10    Reactor Building ......................................................................... 2.15-56 2.15.11    Turbine Building......................................................................... 2.15-76 2.15.12    Control Building ......................................................................... 2.15-78 2.15.13    Radwaste Building ...................................................................... 2.15-91 2.15.14    Service Building ......................................................................... 2.15-93 2.15.15    Control Building Annex.............................................................. 2.15-95 2.16    Yard Structures and Equipment 2.16.1    Stack.............................................................................................. 2.16-1 2.16.2    Oil Storage and Transfer System .................................................. 2.16-2 Table of Contents                                                                                                                    v
 
25A5675AA Revision 7 ABWR                                                                                      Design Control Document/Tier 1 Table of Contents (Continued) 2.17    Emergency Planning 2.17.1      Emergency Response Facilities .................................................... 2.17-1 3.0  Additional Certified Design Material ........................................................................... 3.0-1 3.1    Human Factors Engineering ............................................................................. 3.1-1 3.2    Radiation Protection ......................................................................................... 3.2-1 3.3    Piping Design.................................................................................................... 3.3-1 3.4    Instrumentation and Control ............................................................................. 3.4-1 3.5    Initial Test Program .......................................................................................... 3.5-1 3.6    Design Reliability Assurance Program............................................................. 3.6-1 4.0  Interface Requirements ................................................................................................. 4.0-1 4.1    Ultimate Heat Sink............................................................................................ 4.1-1 4.2    Offsite Power System ....................................................................................... 4.2-1 4.3    Makeup Water Preparation System .................................................................. 4.3-1 4.4    Potable and Sanitary Water System.................................................................. 4.4-1 4.5    Reactor Service Water System ......................................................................... 4.5-1 4.6    Turbine Service Water System ......................................................................... 4.6-1 4.7    Communication System .................................................................................... 4.7-1 4.8    Site Security ...................................................................................................... 4.8-1 4.9    Circulating Water System ................................................................................. 4.9-1 4.10    Heating, Ventilating and Air Conditioning System........................................ 4.10-1 5.0  Site Parameters ............................................................................................................. 5.0-1 Appendix A    Legend for Figures..............................................................................Appendix A-1 Appendix B    Abbreviations and Acronyms Used in the ABWR Certified Design Material ...............................................................................................Appendix B-1 Appendix C    Conversion to ASME Standard Units .................................................Appendix C-1 vi                                                                                                                    Table of Contents
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 1.0 Introduction This document provides the certified design material for the Advanced Boiling Water Reactor (ABWR); U.S. NRC Docket No. 52-001.
Introduction                                                                                        1.0-1
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 1.1 Definitions The following definitions apply to terms used in the Design Descriptions and associated ITAAC:
Acceptance Criteria means the performance, physical condition, or analysis results for a structure, system, or component that demonstrates the Design Commitment is met.
Analysis means the calculation, mathematical computation, or engineering or technical evaluation. Engineering or technical evaluations could include, but are not limited to, comparisons with operating experience or design of similar structures, systems, or components.
As-built means the physical properties of the structure, system, or component following the completion of its installation or construction activities at its final location at the plant site.
ASME Code means Section III of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code, unless specifically stated otherwise. Some Tier 1 content in the ABWR DCD specifies that structures, systems, and components be designed and constructed in accordance with ASME Code Section III requirements. When this language is used, it indicates that the Tier 1 requirements related to that content will be met by satisfying the edition and addenda of the ASME Boiler and Pressure Vessel Code, Section III as specified in the DCD and as incorporated by reference in 10 CFR 50.55a subject to the conditions listed in 10 CFR 50.55a, or in accordance with alternatives authorized by the NRC pursuant to 10 CFR 50.55a.
Basic Configuration (for a Building)--- means the arrangement of the building features (e.g.,
floors, ceilings, walls, basemat and doorways) and of the structures, systems, or components within, as specified in the building Design Description.
Basic Configuration (for a System)---- means the functional arrangement of structures, systems, and components specified in the Design Description; and verifications for that system as specified in Section 1.2.
Containment means the Primary Containment System, unless explicitly stated otherwise.
Design Commitment means that portion of the Design Description that is verified by ITAAC.
Design Description means that portion of the design that is certified.
Division (for electrical systems/equipment) is the designation applied to a given safety-related system or set of components which are physically, electrically, and functionally independent from other redundant sets of components.
Division (for mechanical systems/equipment) is the designation applied to a specific set of safety-related components within a system.
Definitions                                                                                                    1.1-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 Inspect or Inspection means visual observations, physical examinations, or review of records based on visual observation or physical examination that compare the structure, system, or component condition to one or more Design Commitments. Examples include walkdowns, configuration checks, measurements of dimensions, and non-destructive examinations.
Test means the actuation or operation, or establishment of specified conditions, to evaluate the performance or integrity of as-built structures, systems, or components, unless explicitly stated otherwise.
Type Test means a test on one or more sample components of the same type and manufacturer to qualify other components of that same type and manufacturer. A type test is not necessarily a test of the as-built structures, systems, or components.
1.1-2                                                                                        Definitions
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 1.2 General Provisions The following general provisions are applicable to the Design Descriptions and associated ITAAC:
Verifications for Basic Configuration for Systems Verifications for Basic Configuration of systems include and are limited to inspection of the system functional arrangement and the following inspections, tests, and analyses:
(1)    Inspections, including non-destructive examination (NDE), of the as-built, pressure boundary welds for ASME Code Class 1, 2, or 3 components identified in the Design Description to demonstrate that the requirements of ASME Code Section III for the quality of pressure boundary welds are met.
(2)    Type tests, analyses, or a combination of type tests and analyses of the Seismic Category I mechanical and electrical equipment (including connected instrumentation and controls) identified in the Design Description to demonstrate that the as-built equipment, including associated anchorage, is qualified to withstand design basis dynamic loads without loss of its safety function.
(3)    Type tests, or type tests and analyses, of the Class 1E electrical equipment identified in the Design Description (or on accompanying figures) to demonstrate that it is qualified to withstand the environmental conditions that would exist during and following a design basis accident without loss of its safety function for the time needed to be functional. These environmental conditions, as applicable to the bounding design basis accident(s), are as follows: expected time-dependent temperature and pressure profiles, humidity, chemical effects, radiation, aging, submergence, and their synergistic effects which have a significant effect on equipment performance. As used in this paragraph, the term Class 1E electrical equipment constitutes the equipment itself, connected instrumentation and controls, connected electrical components (such as cabling, wiring, and terminations), and the lubricants necessary to support performance of the safety functions of the Class 1E electrical components identified in the Design Description, to the extent such equipment is not located in a mild environment during or following a design basis accident.
Electrical equipment environmental qualification shall be demonstrated through analysis of the environmental conditions that would exist in the location of the equipment during and following a design basis accident and through a determination General Provisions                                                                                          1.2-1
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 that the equipment is qualified to withstand those conditions for the time needed to be functional. This determination may be demonstrated by:
(a)    Type testing of an identical item of equipment under identical or similar conditions with a supporting analysis to show that the equipment is qualified; or (b)    type testing of a similar item of equipment under identical or similar conditions with a supporting analysis to show that the equipment is qualified; or (c)    experience with identical or similar equipment under identical or similar conditions with supporting analysis to show that the equipment is qualified; or (d)    analysis in combination with partial type test data that supports the analytical assumptions and conclusions to show that the equipment is qualified.
(4)    Tests or type tests of active safety-related motor-operated valves (MOVs) identified in the Design Description to demonstrate that the MOVs are qualified to perform their safety functions under design basis differential pressure, system pressure, fluid temperature, ambient temperature, minimum voltage, and minimum and/or maximum stroke times.
Treatment of Individual Items The absence of any discussion or depiction of an item in the Design Description or accompanying figures shall not be construed as prohibiting a licensee from utilizing such an item, unless it would prevent an item from performing its safety functions as discussed or depicted in the Design Description or accompanying figures.
When the term operate, operates, or operation is used with respect to an item discussed in the Acceptance Criteria, it refers to the actuation and running of the item. When the term exist, exists, or existence is used with respect to an item discussed in the Acceptance Criteria, it means that the item is present and meets the Design Description.
Implementation of ITAAC Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) are provided in tables with the following three-column format:
Inspections, Tests, Design Commitment                  Analyses                            Acceptance Criteria Each Design Commitment in the left-hand column of the ITAAC tables has an associated Inspections, Tests, or Analyses (ITA) requirement specified in the middle column of the tables.
The identification of a separate ITA entry for each Design Commitment shall not be construed to require that separate inspections, tests, or analyses must be performed for each Design Commitment. Instead, the activities associated with more than one ITA entry may be combined, 1.2-2                                                                                      General Provisions
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 and a single inspection, test, or analysis may be sufficient to implement more than one ITA entry.
An ITA may be performed by the licensee of the plant, or by its authorized vendors, contractors, or consultants. Furthermore, an ITA may be performed by more than a single individual or group, may be implemented through discrete activities separated by time, and may be performed at any time prior to fuel load (including before issuance of the Combined Operating License for those ITAAC that do not necessarily pertain to as-installed equipment).
Additionally, ITA may be performed as part of the activities that are required to be performed under 10CFR Part 50 (including, for example, the Quality Assurance (QA) program required under Appendix B to Part 50); therefore, an ITA need not be performed as a separate or discrete activity.
Discussion of Matters Related to Operations In some cases, the Design Descriptions in this document refer to matters that relate to operation, such as normal valve or breaker alignment during normal operation modes. Such discussions are provided solely to place the Design Description provisions in context (e.g., to explain automatic features for opening or closing valves or breakers upon off-normal conditions). Such discussions shall not be construed as requiring operators during operation to take any particular action (e.g., to maintain valves or breakers in a particular position during normal operation).
Interpretation of Figures In many but not all cases, the Design Descriptions in Section 2 include one or more figures, which may represent a functional diagram, general structural representation, or other general illustration. For I&C systems, the figures also represent aspects of the relevant logic of the system or part of the system. Unless specified explicitly, these Figures are not indicative of the scale, location, dimensions, shape, or spatial relationships of as-built structures, systems, or components. In particular, the as-built attributes of structures, systems, and components may vary from the attributes depicted on these figures, provided that those safety functions discussed in the Design Description pertaining to the figure are not adversely affected.
Rated Reactor Core Thermal Power The rated reactor core thermal power for the ABWR is 3926 Mwt.
General Provisions                                                                                            1.2-3
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.0 Certified Design for ABWR Systems This section provides the certified design material for each of the ABWR systems that is either fully or partially within the scope of the Certified Design.
Certified Design for ABWR Systems                                                                        2.0-1
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 2.1.1 Reactor Pressure Vessel System Design Description The Reactor Pressure Vessel (RPV) System consists of (1) the RPV and its appurtenances, supports and insulation, excluding the Loose Parts Monitoring System, and (2) the reactor internal components enclosed by the vessel, excluding the core (fuel assemblies, control rods, in-core nuclear instrumentation and neutron sources), reactor internal pumps (RIPs), and control rod drives (CRDs). The RPV System is located in the primary containment.
The reactor coolant pressure boundary (RCPB) portion of the RPV and its appurtenances (referred to in this section as the RPV pressure boundary) act as a radioactive material barrier during plant operation.
Certain reactor internals support the core, flood the core during a loss-of-coolant accident (LOCA) and support safety-related instrumentation. Other RPV internals direct coolant flow, separate steam, hold material surveillance specimens, and support instrumentation utilized for plant operation.
The RPV System provides guidance and support for the CRDs. It also distributes sodium pentaborate solution when injected from the Standby Liquid Control (SLC) System.
The RPV System restrains the CRD to prevent ejection of the control rod connected with the CRD in the event of a failure of the RCPB associated with the CRD housing weld. A restraint system is also provided for each RIP in order to prevent the RIP from becoming a missile in the event of a failure of the RCPB associated with the RIP casing weld.
The RPV System is shown on Figures 2.1.1a, 2.1.1b and 2.1.1c; key dimensions and the acceptable variations in these dimensions are presented in Table 2.1.1a. The RPV System parameters (break areas) used in LOCA analyses are identified in Table 2.1.1b. The principal design parameters for the RPV System are listed in Table 2.1.1c.
Reactor Pressure Vessel, Appurtenances, Supports and Insulation The RPV, as shown in Figures 2.1.1a and 2.1.1b, is a vertical, cylindrical vessel of welded construction with removable top head and head closure bolting and seals. The main body of the installed RPV has a cylindrical shell, flange, bottom head, RIP casings, penetrations (including inserted housings), brackets, nozzles, and the shroud support, which has a pump deck forming the partition between the RIP suction and discharge. The shroud support is an assembly consisting of a short vertical cylindrical shell, a horizontal annular pump deck plate and vertical shroud support legs.
The CRD housings are inserted through and welded to the CRD penetrations in the reactor vessel bottom head. The CRDs are mounted into the CRD housings. The in-core housings are inserted through and connected to the bottom head.
Reactor Pressure Vessel System                                                                              2.1-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 For an RPV System that requires to be instrumented for flow-induced vibration (FIV) testing, a flanged nozzle is provided in the top head for bolting of the flange associated with the test instrumentation.
The integral reactor vessel skirt supports the vessel on the Reactor Pressure Vessel Pedestal.
The vessel skirt does not have openings connecting the upper and lower drywell regions.
Anchor bolts extend from the pedestal through the flange of the skirt. RPV stabilizers are provided in the upper portion of the RPV to resist horizontal loads. Lateral supports for the CRD housings and in-core housings are provided.
A restraint system is provided to prevent a RIP from being a missile in case of a postulated failure in the casing weld with the bottom head penetration. The restraint system is connected to the lugs on the RPV bottom head and the RIP motor cover.
The RPV insulation is supported from the reactor shield wall surrounding the vessel. Insulation for the upper head and flange is supported by a steel frame independent of the vessel and piping.
The RPV pressure boundary and the supports (RPV skirt, stabilizer and CRD housing/in-core housing lateral supports) are classified as Seismic Category I. These components are ASME Code Class 1 vessel and supports, respectively. The shroud support and a portion of the CRD housings inside the RPV are classified as Seismic Category I and ASME Code Class CS structures.
The following ASME Code Section II materials (or their equivalents) are used in the RPV pressure boundary: SA-533, Type B, Class 1 (plate); SA-508, Class 3 (forging); SA-508, Class 1 (forging); SB-166 (UNS N06600, bar); SB-167 (UNS N06600, seamless pipe); SB-564 (UNS N06600, forging); SA-182 or SA-336, Grade/Class F316L (maximum carbon 0.020%, forging) or F316 (maximum carbon 0.020% and nitrogen from 0.060 to 0.120%, forging); and SA-540, Grade B23 or B24 (bolting).
A stainless steel weld overlay is applied to the interior of the RPV cylindrical shell and the steam outlet nozzles. Other nozzles and the RIP motor casings do not have cladding. The bottom head is clad with Ni-Cr-Fe alloy. The RIP penetrations are clad with Ni-Cr-Fe alloy or, alternatively, stainless steel.
The materials of the low alloy plates and forging used in construction of the RPV pressure boundary are melted using vacuum degassing to fine grain practice and are supplied in quenched and tempered condition.
Electroslag welding is not applied for the RPV pressure boundary welds. Preheat and interpass temperatures employed for welding of the RPV pressure boundary low alloy steel meet or exceed the values given in ASME Code Section III, Appendix D. Post-weld heat treatment at 593°C minimum is applied to these low-alloy steel welds.
2.1-2                                                                      Reactor Pressure Vessel System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 The RPV pressure boundary welds are given an ultrasonic examination in addition to the radiographic examination performed during fabrication. The ultrasonic examination method, including calibration, instrumentation, scanning sensitivity, and coverage, is based on the requirements imposed by ASME Code Section XI, Appendix I. Acceptance standards also meet the requirements of ASME Code Section XI.
The fracture toughness tests of the RPV pressure boundary ferritic materials, weld metal and heat-affected zone (HAZ) are performed in accordance with the requirements for ASME Code Section III, Class 1 vessel. Both longitudinal and transverse specimens are used to determine the minimum upper-shelf energy (USE) level of the core beltline materials. The minimum USE level for base and weld metal in the core beltline is initially at least 10.4 kgm. Separate, unirradiated baseline specimens are used to determine the transition temperature curve of the core beltline base materials, weld metal, and HAZ.
For the RPV material surveillance program, Charpy V-notch and tensile specimens are manufactured from the same material used in the reactor beltline region. To represent those RPV pressure boundary welds that are in the beltline region, Charpy V-notch specimens of weld metal and HAZ material, and tensile specimens of weld metal are manufactured from sample welds. The specimen capsules contain the specimens and temperature monitors. The surveillance specimen holders having brackets welded to the vessel cladding in the core beltline region are provided to hold four specimen capsules and a neutron dosimeter.
Reactor Pressure Vessel Internals The major reactor internal components in the RPV System are:
(1)    Core Support Structures:
Shroud, shroud support and a portion of CRD housings inside the RPV (both integral to the RPV), core plate, top guide, fuel supports (orificed fuel supports and peripheral fuel supports), and control rod guide tubes (CRGTs). The core support structures are classified as Seismic Category I and ASME Code Class CS structures.
(2)    Other Reactor Internals:
(a)    Feedwater spargers, shutdown cooling (SDC) and low pressure core flooder (LPFL) spargers for the Residual Heat Removal (RHR) System, high pressure core flooder (HPCF) spargers and couplings, and a portion of the in-core housings inside the RPV and in-core guide tubes (ICGTs) with stabilizers.
These components are classified as Seismic Category I and safety-related.
(b)    Surveillance specimen holders, shroud head and steam separators assembly and the steam dryer assembly. These components are classified as non-safety-related.
Reactor Pressure Vessel System                                                                                2.1-3
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 A general assembly of these reactor internal components is shown in Figures 2.1.1a, 2.1.1b, and 2.1.1c.
The shroud support, shroud, and top guide make up a cylindrical assembly that provides a partition to separate the upward flow of coolant through the core from the downward recirculation flow. This partition separates the core region from the downcomer annulus.
The core plate consists of a circular plate with round openings and is stiffened with a rim and beam structure. The core plate provides lateral support and guidance for the CRGTs, ICGTs, peripheral fuel supports, and startup neutron sources. The last two items are also supported vertically by the core plate.
The top guide consists of a circular plate with square openings for fuel assemblies and with a cylindrical side forming an upper shroud extension. Each opening provides lateral support and guidance for four fuel assemblies or, in the case of peripheral fuel, less than four fuel assemblies. Holes are provided in the bottom, where the sides of the openings intersect, to anchor the in-core instrumentation detectors and startup neutron sources.
The fuel supports are of two types: (1) peripheral and (2) orificed. The peripheral fuel supports are located at the outer edge of the active core and are not adjacent to control rods. Each peripheral fuel support supports one peripheral fuel assembly and has an orifice to provide coolant flow to the fuel assembly. Each orificed fuel support holds four fuel assemblies and has four orifices to provide coolant flow distribution to each fuel assembly. The control rods pass through cruciform openings in the center of the orificed fuel supports. This locates the four fuel assemblies surrounding a control rod.
The CRGTs pass through holes in the core plate, have four holes under the core plate and rest on top of the CRD housings. Each CRGT guides the lower end of a control rod and supports an orificed fuel support such that the orifices of the orificed fuel support align with the holes in the CRGT for coolant flow. The lower end of the CRGT is supported by the CRD housing, which, in turn, transmits the weight of the guide tube, fuel supports, and fuel assemblies to the reactor vessel bottom head.
The CRGT base is provided with a device for coupling the CRD with it. The CRD is restrained from ejection, in the case of failure of the weld between a CRD housing and CRD penetration, by the coupling of the CRD with the CRGT base; in this event, the flange at the top of the guide tube contacts the core plate and acts to restrain the ejection. The coupling will also prevent ejection if the housing fails beneath the weld; in this event, the guide tube remains supported on the intact upper housing.
There are six feedwater spargers, three for each of the two feedwater lines. Each sparger is connected to an RPV feedwater nozzle at the double thermal sleeve fitted with the safe end (straight piece) of the nozzle. Feedwater flow enters the middle of the spargers, which are located above the RPV downcomer annulus, and is discharged inward.
2.1-4                                                                        Reactor Pressure Vessel System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Two spargers are provided for two loops of the RHR System; both spargers function as SDC and LPFL spargers. Each sparger is connected to a thermal sleeve fitted with the safe end of each SDC and LPFL inlet nozzle.
Two HPCF spargers with couplings are provided for the two loops of the HPCF System to direct high pressure coolant flow to the upper end of the core during emergency core cooling.
One of the HPCF spargers also distributes sodium pentaborate solution when injected from the SLC System via the connecting HPCF line. The spargers are located inside the cylindrical portion of the top guide. Each sparger is connected via an HPCF coupling to a thermal sleeve fitted with the safe end of each HPCF inlet nozzle.
The ICGTs house the in-core neutron flux monitoring instrumentation assemblies, pass through holes in the core plate, and rest on top of the in-core housings. Two levels of stabilizer latticework give lateral support to the ICGTs. The ICGT stabilizers are connected to either the shroud or the shroud support.
The surveillance specimen holders having brackets welded to the vessel cladding in the core beltline region are provided to hold the specimen capsules and a neutron dosimeter.
The shroud head and steam separators assembly includes the connecting standpipes and forms the top of the core discharge mixture plenum. The steam dryer assembly removes moisture from the wet steam leaving the steam separators. The extracted moisture flows down the dryer vanes to the collecting troughs, then flows through tubes into the downcomer annulus.
Cobalt-base material is only used for hard surfacing of areas in HPCF coupling. The wrought austenitic stainless steel used for the RPV internals is limited to a maximum of 0.02% carbon content. Stainless steel materials are supplied in solution heat-treated condition. Furnace sensitized stainless steel material is not used. Electroslag welding is not applied for structural welds of stainless steel.
The RPV internals are designed to withstand the effects of FIV.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.1.1d provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Pressure Vessel System.
Reactor Pressure Vessel System                                                                              2.1-5
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 VIBRATION INSTRUMENTATION NOZZLE CLOSURE HEAD STEAM FLOW RESTRICTOR ELEV. H STEAM DRYER STEAM OUTLET (4)
RPV STABILIZER STEAM SEPARATOR ELEV. G RHR SDC AND LPFL                                        SHROUD HEAD SPARGER (2)
FEEDWATER SPARGER (6)
ELEV. J HPCF INLET (2)
RHR SDC OUTLET (3)
HPCF COUPLING (2)
ELEV. F HPCF SPARGER (2)
CORE PLATE TOP GUIDE L
SURVEILLANCE                          K            SHROUD SPECIMEN HOLDER B          A                      PERIPHERAL FUEL ORIFICED FUEL SUPPORT                                          SUPPORT ELEV. E RPV SUPPORT SKIRT                                                          ELEV. D P
ANCHOR BOLT ICGT AND STABILIZER CRGT N
DRAIN NOZZLE                                      SHROUD SUPPORT ELEV. C RIP RESTRAINT                                      RIP CASING (10)
(TYPICAL)
CRD HOUSING CRD HOUSING/                                      IN-CORE HOUSING IN-CORE HOUSING LATERAL SUPPORTS Figure 2.1.1a Reactor Pressure Vessel System Key Features 2.1-6                                                              Reactor Pressure Vessel System
 
25A5675AA Revision 7 ABWR                                                  Design Control Document/Tier 1 RPV WALL RIP PENETRATION N
SHROUD M                            SUPPORT LEG Figure 2.1.1b Pump Penetration and Shroud Support Leg Arrangement Reactor Pressure Vessel System                                                  2.1-7
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 NOTES:
: 1. THE ARRANGEMENT IS SHOWN FOR QUARTER CORE ONLY. ROTATIONAL SYMMETRY APPLIES. THE REACTOR INTERNALS ACCOMMODATE THE SHOWN CORE ARRANGEMENT; THE CORE IS NOT A PART OF THE RPV SYSTEM.
1 2
FUEL ASSEMBLY 3
4 CONTROL ROD 5
6 7
8 9
10 11 12 13 14 15 16 17 1    2    3    4  5    6    7    8    9    10    11 12 13  14  15    16    17 Figure 2.1.1c Core Arrangement 2.1-8                                                                Reactor Pressure Vessel System
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 Table 2.1.1a Key Dimensions of RPV System Components and Acceptable Variations Dimension/      Nominal    Acceptable Elevation      Value        Variation Description                                                (Figure 2.1.1a)    (mm)        (mm)
RPV inside diameter (inside cladding)                            A          7112.0        +/-51.0 RPV wall thickness in beltline (without cladding)                B          174.0      +20.0/-4.0 RPV bottom head inside invert, Elevation                          C              0.0    Reference 0.0 RPV support skirt bottom, Elevation                              D          3250.0        +/-75.0 Core plate support/Top of shroud middle flange,                  E          4695.2        +/-15.0 Elevation Top guide support/Top of shroud top flange, Elevation            F          9351.2        +/-20.0 RPV stabilizer connection, Elevation                              G          13,766.0      +/-20.0 Top of RPV flange, Elevation                                      H          17,703.0      +/-65.0 RHR SDC/CUW Outlet Nozzle, Elevation                              J          10,921.0      +/-40.0 Shroud outside diameter                                          K          5600.7        +/-25.0 Shroud wall thickness                                            L            57.2        +/-10.0 Shroud support legs (Fig. 2.1.1b)                                MxN      662.0x153.0  +/-20.0 for M
                                                                                          +/-10.0 for N Control rod guide tube outside diameter                          P          273.05        +/-5.0 Reactor Pressure Vessel System                                                                      2.1-9
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 Table 2.1.1b RPV System Parameters Used in LOCA Analyses Postulated Break Area, Line                                            Inspection Location                                mm2 Steamline              Flow restrictor throat diameter in a steam outlet nozzle.                98,480 Feedwater              Inside diameters of flow nozzles on the spargers of                      83,890 a feedwater line for the total flow area.
RHR Injection          Inside diameters of flow nozzles on an SDC and LPFL                      20,530 sparger for the total flow area.
High Pressure Core      Inside diameters of flow nozzles on an HPCF sparger                        9200 Flooder                for the total flow area.
RHR Shutdown            Inside diameter of an RHR SDC outlet nozzle at the safe end              79,150 Cooling                weld.
Drain                  Inside diameter of the bottom head hole for the drain                      2030 outlet nozzle, near the inside surface of the head and below the hole chamfer.
Note: The areas calculated from the inspections shall not exceed the postulated break areas by 5%.
2.1-10                                                                          Reactor Pressure Vessel System
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 Table 2.1.1c Principal Design Parameters for RPV System Description                                      Value RCPB design pressure (MPaG)                      8.62 RCPB design temperature (°C)                      302 Number of fuel assemblies                        872 Number of control rods                            205 Number of internal pumps                          10 Reactor Pressure Vessel System                                                        2.1-11
 
ABWR 2.1-12 Table 2.1.1d Reactor Pressure Vessel System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the RPV System        1. Inspections of the as-built RPV System will  1. The RPV System conforms with the basic is as defined as Section 2.1.1.                    be conducted.                                    configuration defined in Section 2.1.1.
: 2. The RPV pressure boundary defined in      2. Inspections of the ASME Code required              2. An ASME Code Certified Stress Report Section 2.1.1 is designed to meet the ASME    documents will be conducted.                          exists for the RPV pressure boundary Code Class 1 vessel requirements.                                                                    components.
: 3. The ASME Code components of the RPV              3. A hydrostatic test will be conducted on those 3. The results of the hydrostatic test of the System retain their pressure boundary              code components of the RPV System                ASME Code components of the RPV System integrity under internal pressure that will be      required to be hydrostatically tested by the    conform with the requirements in the ASME experienced during service.                        ASME Code.                                      Code, Section III.
: 4. The materials selection and materials testing 4. Inspections of the as-built RPV System will      4. The RPV System conforms with the 25A5675AA Revision 7 requirements for the RPV System are as          be conducted.                                      materials selection and materials testing defined in Section 2.1.1.                                                                            requirements defined in Section 2.1.1.
: 5. The fabrication process and examination          5. Inspections of the as-built RPV System will  5. The RPV System conforms with the process requirements for the RPV System            be conducted.                                    fabrication process and examination process are as defined in Section 2.1.1.                                                                    requirements defined in Section 2.1.1.
: 6. The material surveillance commitments for        6. Inspections of the as-built RPV System will  6. The material surveillance program for the the reactor pressure vessel core beltline          be conducted for implementation of the          reactor pressure vessel core beltline materials are as defined in Section 2.1.1.          material surveillance commitments.              materials conforms with the commitments defined in Section 2.1.1.
Design Control Document/Tier 1
: 7. The RPV internals withstand the effects of      7. A vibration type test will be conducted on the 7. A vibration type test report exists and FIV.                                                prototype RPV internals of an ABWR.              concludes that the prototype RPV internals have no damage or loose parts as a result of Reactor Pressure Vessel System A flow test and post-test inspections will be conducted on the as-built RPV internals.          the vibration type test.
The as-built RPV internals have no damage or loose parts.
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.1.2 Nuclear Boiler System Design Description General System Description The primary functions of the Nuclear Boiler System (NBS) are:
(1)    Deliver steam from the Reactor Pressure Vessel (RPV) to the Main Steam (MS)System.
(2)    Provide containment isolation of the main steamlines (MSLs) and the feedwater (FW) lines.
(3)    Deliver feedwater from the Condensate, Feedwater, and Condensate Air Extraction (CFCAE) System to the RPV.
(4)    Provide overpressure protection of the reactor coolant pressure boundary (RCPB).
(5)    Provide automatic depressurization of the RPV in the event of a loss- of-coolant accident (LOCA) where the RPV does not depressurize rapidly and the high pressure makeup systems fail to adequately maintain the water level in the RPV.
(6)    Provide instrumentation to monitor the drywell pressure and RPV pressure, metal temperature, and water level.
Figures 2.1.2a, 2.1.2b, 2.1.2c, 2.1.2d, and 2.1.2e show the basic system configuration and scope. Figure 2.1.2f shows the NBS control interfaces.
The NBS equipment shown on Figures 2.1.2a, 2.1.2b, 2.1.2c, 2.1.2d, and 2.1.2e is classified as safety-related except for the non-safety-related part of the MSL drains, equipment associated with the power actuated relief mode of the SRVs, the SRV discharge pipe temperature sensors, and the non-safety-related instruments shown on Figure 2.1.2e.
Main Steam Lines The MSLs direct steam from the RPV to the MS System. The NBS contains only the portion of the MSLs from their connection to the RPV to the boundary with the MS System, which occurs at the seismic interface located downstream of the outboard main steam isolation valves (MSIVs). Figures 2.1.2a and 2.1.2b show the general configuration of the MSLs and the MSL drain lines. The MSL drain lines provide a flow path for the MSIV leakage during an accident.
The combined volume of the steamlines, from the RPV to the main steam turbine stop valves and turbine bypass valves, is greater than or equal to 113.2 m3.
Each MSL has a flow limiter. The MSL flow limiter consists of a flow restricting venturi which is located in each RPV MSL outlet nozzle. The restrictor limits the coolant blowdown rate from Nuclear Boiler System                                                                                    2.1-13
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 the RPV, in the event that a MSL break occurs outside the containment, to a flow rate equal to or less than 200% of rated steam flow at 7.07 MPaG upstream pressure. The throat diameter of each MSL flow limiter is less than or equal to 355 mm.
The pneumatic-operated valve in the MSL drain line shown in Figure 2.1.2.b opens, if either electric power to the valves actuating solenoid is lost, or pneumatic pressure to the valve is lost.
The MSLs and the MSL drain lines are located in the drywell and the steam tunnel.
Main Steam Isolation Valves Two isolation valves are located in a horizontal run of each of the four main steamlines; one valve is inside of the drywell, and the other is near the outside of the primary containment pressure boundary.
The MSIV closing time is equal to or greater than 3 seconds and less than or equal to 4.5 seconds when N2 or air is admitted to the MSIV actuator. The MSIVs are capable of closing within 3 to 4.5 seconds under differential pressure, fluid flow and temperature conditions.
When all the MSIVs are closed, the combined leakage through the MSIVs for all four MSLs is less than or equal to 66.1 liters per minute at standard temperature (20°C) and pressure (one atmosphere absolute pressure) with the differential pressure across the MSIV equal to, or greater than 0.17 MPa.
The MSIVs primary actuation mechanism for opening and closing is pneumatic. Springs close the MSIV if pneumatic pressure to the MSIV actuator is lost.
Feedwater Lines The FW lines direct feedwater from the CFCAE System to the RPV. The NBS contains only the portion of the FW lines from the seismic interface located upstream of the motor-operated valves (MOVs) to their connections to the RPV. Figure 2.1.2c shows the portion of the FW lines within the NBS.
Isolation of each FW line is accomplished by two containment isolation valves consisting of one check valve inside the drywell and one positive closing check valve outside the containment. The FW line isolation check valves are qualified to withstand a FW line break outside the primary containment. The FW line upstream of the outboard isolation valve contains an MOV and a seismic interface restraint.
Safety/Relief Valves The safety/relief valves (SRVs) are located on the MSLs between the RPV and the inboard MSIV. These valves protect against overpressurization of the RCPB. Figures 2.1.2a, 2.1.2b and 2.1.2d show the general configuration of the SRVs and the SRV discharge lines.
The rated capacity of the SRVs is sufficient to prevent a rise in pressure within the RPV of more than 110% of the design pressure (9.48 MPaG) for design basis events.
2.1-14                                                                                  Nuclear Boiler System
 
25A5675AA Revision 7 ABWR                                                                                  Design Control Document/Tier 1 The SRV discharge lines are sized so that critical flow conditions occur through the valve. Each SRV has its own discharge line. The SRV discharge lines terminate at quenchers located below the surface of the suppression pool.
The SRVs provide three main protection functions:
(1)  Overpressure safety operation: The valves function as spring-loaded safety valves and open to prevent RCPB overpressurization. The valves are self-actuated by inlet steam pressure.
The following table identifies the SRV spring set pressures and flow capacities. The opening time for the SRVs, from the time the pressure exceeds the valve set pressure to the time the valve is fully open, is less than or equal to 0.3 seconds.
Set Pressures and Capacities ASME Rated Nameplate Spring          Capacity at 103%
Number*            Set Pressure          Spring Set Pressure        Used For SRVs            of Valves            (MPaG)                  (kg/h each)              ADS J, P                  2                  7.92                    395,000 B, G, M, S            4                  7.99                    399,000 D, E, K, U            4                  8.06                    402,000 C, H, N, T            4                  8.13                    406,000                X A, F, L, R            4                  8.20                    409,000                X
* Eight of the SRVs serve in the automatic depressurization system function.
Spring set pressure tolerances as permitted by the ASME Boiler and Pressure Vessel Code, Section III.
Minimum capacity per the ASME Boiler and Pressure Vessel Code, Section III.
(2)  Overpressure relief operation: The valves are opened using a pneumatic actuator upon receipt of an automatic or manually initiated signal.
For overpressure relief valve operation (power-actuated mode), reactor vessel pressure sensors generate a high pressure trip signal which is used to initiate opening the SRVs. Valve opening is initiated when an electrical signal is received at the solenoid valve associated with power actuated relief (Figure 2.1.2d). The SRV relief mode opening time from the receipt of signal at the valve actuator to the full ASME lift position is less than or equal to 0.25 seconds when the SRV inlet pressure is at or above 6.89 MPaG.
Nuclear Boiler System                                                                                                  2.1-15
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 The SRV pneumatic operator is so arranged that, if it malfunctions, it does not prevent the SRV from opening when steam inlet pressure reaches the spring lift setpoint. Each SRV is provided with its own pneumatic accumulator and inlet check valve for power actuated relief as shown in Figure 2.1.2d.
The SRVs are either DC powered, or powered from uninterruptible AC.
(3)    Automatic depressurization system (ADS) operation: The ADS valves open automatically or manually in the power actuated mode when required during a loss-of-coolant accident (LOCA). Eight of the eighteen SRVs are designated as ADS valves and are capable of operating from either ADS LOCA logic or overpressure relief logic signals. The above table identifies the ADS SRVs.
The ADS accumulator capacity can open the SRV with the drywell pressure at design pressure following failure of the pneumatic supply to the accumulator.
The SRVs can be operated individually in the power-actuated mode by remote manual switches located in the main control room. They are provided with position sensors which provide positive indication of SRV disk/stem position.
Automatic Depressurization System As shown in Figure 2.1.2f, the NBS channel measurements are provided for the Safety System Logic and Control (SSLC) for signal processing, setpoint comparisons, and generating trip signals. Except for the pump running permissive, the SSLC uses a two-out-of-four voting logic for ADS initiation. The ADS logic is automatically initiated when a low reactor water level signal is present. If the RPV low water level signal is present concurrently with high drywell pressure signal, both the main ADS timer (less than or equal to 29 seconds) and the high drywell pressure bypass timer (less than or equal to 8 minutes) are initiated. Absent a concurrent high drywell pressure signal, only the ADS high drywell pressure bypass timer is initiated. Upon the time out of the ADS high drywell pressure bypass timer, concurrent with RPV low water level signal, the main ADS timer is initiated, if not already initiated. The main timer continues to completion and times out only in the continued presence of an RPV low water level signal.
Upon time out of the main ADS timer, concurrent with positive indication by pump discharge pressure of at least one RHR or one HPCF pump running, the ADS function is initiated.
Signals from all four divisions for low reactor water level and high drywell pressure and Division I control logic signal actuate one set of pilots, and sensors from all four divisions for low reactor water and high drywell pressure and Division II control logic signal actuate the second set of pilots, either of which initiates the opening of the ADS SRVs.
ADS initiation is accomplished by redundant trip channels arranged in two divisionally separated logics that control two separate solenoid-operated pneumatic pilots on each ADS SRV. Either pilot can operate the ADS valve. These pilots control the pneumatic pressure 2.1-16                                                                                Nuclear Boiler System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 applied by the accumulators and the High Pressure Nitrogen Gas Supply (HPIN) System. The DC power for the logic is obtained from the SSLC Divisions I and II.
For anticipated transient without scram (ATWS) mitigation, the ADS has an automatic and manual inhibit of the automatic ADS initiation. Automatic initiation of ADS is inhibited unless there is a coincident low reactor water level signal and an average power range monitors (APRMs) ATWS permissive signal from the Neutron Monitoring System. There are main control room switches for the manual inhibit of automatic initiation of ADS.
The ADS can also be initiated manually. On a manual initiation signal, concurrent with positive indication of at least one RHR or one HPCF pump is running, the ADS function is initiated.
NBS Instrumentation The NBS contains the instrument lines and instrumentation for monitoring the reactor pressure and water level. For drywell pressure, turbine inlet pressure, main condenser vacuum, and RPV metal temperature, the NBS contains the sensors. Figure 2.1.2e shows the drywell pressure and RPV instrumentation in the NBS.
The mechanical portion of each division of the safety-related NBS instrumentation located in the Reactor Building is physically separated from the other divisions.
The reactor vessel outside surface (metal) temperatures are measured at the head flange and the bottom head locations.
Figure 2.1.2e shows the water level instrumentation. The instruments that sense the water level are differential pressure devices calibrated for specific RPV pressure and temperature conditions. Instrument zero for the RPV water level ranges is the top of the active fuel. The RPV water level instrumentation considers the effects of dissolved non-condensable gasses in the RPV water level instrumentation lines.
With the exception of turbine inlet pressure sensor and main condenser vacuum sensor located in the Turbine Building, the NBS instrumentation is located in the drywell, the steam tunnel and the Reactor Building.
Other Provisions The NBS equipment identified as safety-related is classified as Seismic Category I except for the American Society of Mechanical Engineers (ASME) Class 3 equipment shown on Figure 2.1.2c. The non-safety-related section of the feedwater lines between the seismic interface restraint and the motor-operated valves shown in Figure 2.1.2c is classified as Seismic Category I. The MSL drain lines from the MSLs to the Main Condenser are seismically analyzed to withstand the Safe Shutdown Earthquake (SSE).
Figures 2.1.2a, 2.1.2b, 2.1.2c, 2.1.2d and 2.1.2e show the ASME Boiler and Pressure Vessel Code classes.
Nuclear Boiler System                                                                                        2.1-17
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 The divisional equipment in the NBS is powered from its respective Class 1E divisions as shown in Figures 2.1.2b, 2.1.2d, and 2.1.2e. In the NBS, independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
The NBS has the following displays and controls in the main control room:
(1)  Parameter displays for the instruments shown on Figures 2.1.2b and 2.1.2e. This includes the reactor vessel pressure, reactor vessel water level, drywell pressure, main condenser vacuum, and turbine inlet pressure.
(2)  Controls and status indication for the active safety-related components shown on Figures 2.1.2b, 2.1.2c (excluding the inboard FW line check valves, and the ASME Boiler and Pressure Vessel Code Class 2 check valves), and 2.1.2d.
(3)  Manual system level initiation capability for the ADS.
(4)  Manual capability to inhibit automatic initiation of the ADS.
NBS components with displays and control interfaces with the Remote Shutdown System (RSS) are shown on Figures 2.1.2a and 2.1.2e.
The safety-related electrical equipment (including instrumentation and controls) shown on Figures 2.1.2b, 2.1.2c, 2.1.2d, and 2.1.2e located in the containment, steam tunnel and Reactor Building, is qualified for a harsh environment.
The MOVs shown on Figure 2.1.2b (except for the ASME Boiler and Pressure Vessel Code Class 2 MOV) have an active safety-related function to close, and perform this function under differential pressure, fluid flow, and temperature conditions.
The check valves (CVs) shown on Figures 2.1.2c and 2.1.2d (ADS pneumatic CVs only) have the safety--related functions to open, close, or both open and close under system pressure, fluid flow, and temperature conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.1.2 provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria, which will be undertaken for the NBS.
2.1-18                                                                                Nuclear Boiler System
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 Figure 2.1.2a NBS Safety/Relief Valves and Steamline Nuclear Boiler System                                                                  2.1-19
 
ABWR 2.1-20 1 3 SRV (TYPICAL)                                              OUTBOARD MAIN STEAM ISOLATION VALVE INBOARD MAIN                                                  NBS MS SYSTEM T    STEAM ISOLATION                      P 2        TO MAIN VALVE                                                                    STEAM TURBINE MAIN STEAM                                                      1 2                          SEISMIC REACTOR      LINE A                    P                                                                  INTERFACE FROM VESSEL                                                                                                      RESTRAINT FROM                                              STEAMLINES STEAMLINES                                        B,C &D          P B,C &D                                                                  DIV I 1
RPV NBS NOTE 2 2 NNS DIV II            DIV I M                  M 25A5675AA Revision 7 DRAIN LINE M
DRAIN LINE 1    NNS NOTE 3 WETWELL NBS          NNS MAIN CONDENSER Design Control Document/Tier 1 NOTE 1        SUPPRESSION POOL                                                                TO MAIN CONDENSER NOTES:
: 1. AT MINIMUM LEVEL
: 2. MAY BE PNEUMATIC
: 3. THE PIPING PRESSURE WELDS IN THE WETWELL AIRSPACE SHALL BE EXAMINED USING ASME CODE CLASS 2 REQUIREMENTS.
Nuclear Boiler System
: 4. EACH MSIV HAS CLASS 1E POSITION SWITCHES WHICH RECEIVE POWER FROM ITS RESPECTIVE CLASS 1E DIVISION.
: 5. EACH SRV HAS POSITION SENSORS.
: 6. THE RCIC SYSTEM STEAM SUPPLY COMES FROM MAIN STEAM LINE B UPSTREAM OF THE INBOARD MSIV.
Figure 2.1.2b NBS Steamline
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 CFCAE NBS          NNS        SEISMIC INTERFACE RESTRAINT NNS M                  2                    M CRD            CUW 3 NBS            3 NBS 3                                                                    3 2                                                                    2 2                                                    2 RCIC NBS                      2                      NBS RHR P                    1                    P CONTAINMENT WALL NBS RPV            RPV NBS 1                      1 RPV RPV NBS 1          REACTOR            1 NBS VESSEL 1                    1 NBS RPV            RPV NBS Figure 2.1.2c NBS Feedwater Line Nuclear Boiler System                                                                              2.1-21
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 HPIN NBS                                      NBS HPIN 3                                            3 RELIEF ACCUMULATOR                                      ADS ACCUMULATOR P
FROM POWER    FROM        FROM ACTUATED      ADS DIV. I  ADS DIV.
RELIEF LOGIC  LOGIC      II LOGIC ADS SRV HPIN  NBS 3
RELIEF ACCUMULATOR P
FROM POWER ACTUATED RELIEF LOGIC NON-ADS SRV Figure 2.1.2d NBS Safety/Relief Valve Pneumatic Lines 2.1-22                                                                      Nuclear Boiler System
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 Figure 2.1.2e NBS Drywell Pressure and Reactor Vessel Instrumentation Nuclear Boiler System                                                          2.1-23
 
ABWR 2.1-24 LOCAL AREA                                        MAIN CONTROL ROOM                                    LOCAL AREA Plant Sensors                                                                                        Device Actuators NMS            NBS              NBS            NBS Manual Inhibit    Manual ADS Manual APRMs                        of Auto ADS        Valve Initiation Initiation      Controls SSLC PROCESSING NBS LOGIC                              NBS Automatic and Manual System Initiation 25A5675AA Revision 7 Drywell Pressure                                    - Sensor Channel Trip Decision NBS Reactor Water Level                                    - System Coincidence Trip Decision Reactor Vessel Pressure
                                                                                      - Control and Interlock Logic
                                                                                      - Division-of-Sensors Bypass NBS  Manual Valve Actuation
                                                                                      - Calibration, Self-Diagnosis RHR Pump Discharge Pressure HPCF Pump Discharge Pressure Design Control Document/Tier 1 Notes:
Nuclear Boiler System
: 1. See Section 3.4, Figure 3.4b for SSLC processing.
Figure 2.1.2f Nuclear Boiler System Control Interface Diagram
 
ABWR Nuclear Boiler System Table 2.1.2 Nuclear Boiler System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the NBS is shown 1. Inspections will be conducted for the NBS        1. The as-built NBS conforms with the basic in Figures 2.1.2a, 2.1.2b, 2.1.2c, 2.1.2d,    System.                                            configuration shown in Figures 2.1.2a, 2.1.2e, and 2.1.2f.                                                                                2.1.2b, 2.1.2c, 2.1.2d, 2.1.2e, and 2.1.2f.
: 2. The ASME Code components of the NBS            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the NBS required to be        ASME Code components of the NBS integrity under internal pressures that will be    hydrostatically tested by the ASME Code.        conform with the requirements in the ASME experienced during service.                                                                        Code, Section III
: 3. The combined volume of the four main        3. Analyses will be performed using as-built      3. The combined steamline volume is greater steamlines (MSLs) and branch lines from the    dimensions of the steamlines to determine          than or equal to 113.2 m3.
RPV to the main steam turbine stop valves      the combined steamline volume.
and turbine bypass valves is greater than or 25A5675AA Revision 7 equal to 113.2 m3.
: 4. The throat diameter of each MSL flow limiter 4. Inspections of the as-built MSL flow limiters  4. The throat diameter of each MSL flow limiter is less than or equal to 355 mm.                will be conducted.                                is less than or equal to 355 mm.
: 5. The pneumatic-operated valve in the MSL        5. Tests will be conducted on the as-built MSL  5. The MSL pneumatic drain line valve shown drain line shown in Figure 2.1.2b opens if        drain valve.                                    in Figure 2.1.2b opens when either electric either electric power to the valve actuating                                                      power to the valve actuating solenoid is lost, solenoid is lost, or pneumatic pressure to the                                                    or pneumatic pressure to the valve is lost.
valve is lost.
: 6. MSIV closing time is equal to or greater than 6.                                                6. The MSIV closing time is equal to or greater Design Control Document/Tier 1 3 seconds and less than or equal to 4.5          a. Tests of the as-built MSIV will be            than 3 and less than or equal to 4.5 seconds.
seconds when N2 or air is admitted into the        conducted under preoperational MSIV actuator. The MSIVs are capable of differential pressure, fluid flow, and closing within 3 to 4.5 seconds under temperature conditions.
differential pressure, fluid flow and temperature conditions.                          b. Tests, or type tests, of an MSIV will be conducted under design basis differential pressure, fluid flow and temperature conditions.
2.1-25
 
Table 2.1.2 Nuclear Boiler System (Continued)
ABWR 2.1-26 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                                Acceptance Criteria
: 7. When all MSIVs are closed, the combined        7. Test and analysis will be conducted on the            7. When all MSIVs are closed, the combined leakage through the MSIVs for all four MSLs        as-built MSIVs to determine the leakage.                  leakage through the MSIVs for all four MSLs is less than or equal to 66.1 liters per minute                                                              is less than or equal to 66.1 liters per minute at standard temperature (20°C) and                                                                          at standard temperature (20°C) and pressure (one atmosphere absolute                                                                            pressure (one atmosphere absolute pressure) with the differential pressure                                                                    pressure) with the differential pressure across the MSIV equal to, or greater than,                                                                  across the MSIV equal to, or greater than, 0.17 MPa.                                                                                                    0.17 MPa.
: 8. Springs close the MSIV if pneumatic                8. Tests will be conducted on the as-built MSIV. 8. The MSIV closes when pneumatic pressure pressure to the MSIV actuator is lost.                                                                is removed from the MSIV actuator.
: 9.                                                    9.                                                    9.
25A5675AA Revision 7
: a. The SRV spring set pressure and flow              a. Analysis and tests (at a test facility) will        a. The SRVs have the capacities and set capacities are given in Section 2.1.2.                be conducted in accordance with the                    pressures shown on Section 2.1.2. The The opening time for the SRVs from the                ASME Code.                                            opening time for the SRVs from the time time the pressure exceeds the valve set                                                                      the pressure exceeds the valve set pressure to the time the valve is fully                                                                      pressure to the time the valve is fully open, is less than or equal to 0.3                                                                          open is less than or equal to 0.3 seconds.                                                                                                    seconds.
: b. The SRV relief mode opening time from              b. Tests of the SRVs will be conducted at a            b. The SRV relief mode opening time from Design Control Document/Tier 1 the receipt of signal at the valve actuator          test facility.                                        the receipt of signal at the valve actuator to the full ASME lift position is less than                                                                  to the full ASME lift position is less than or equal to 0.25 seconds when the SRV                                                                        or equal to 0.25 seconds when the SRV inlet pressure is at or above 6.89 MPaG.                                                                    inlet pressure is at or above 6.89 MPaG.
Nuclear Boiler System
 
Table 2.1.2 Nuclear Boiler System (Continued)
ABWR Nuclear Boiler System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 10. The ADS accumulator can open the SRV        10. An analysis and/or type test will be        10. Either:
with the drywell pressure at design pressure    performed to demonstrate the capacity of the    a. The SRV ADS accumulators have the following failure of the pneumatic supply to    SRV ADS accumulators.                                capacity to lift the stem of the SRVs to the accumulator.
the full open position one time with the drywell pressure at, or above the drywell design pressure, or
: b. The SRV ADS accumulators have the capacity to lift the stem of the SRVs to the full open position five times with the drywell at atmospheric pressure, and an analysis that shows that five SRV lifts at 25A5675AA Revision 7 atmospheric pressure demonstrates the capability to open one time with the drywell at the drywell design pressure.
: 11. For overpressure relief valve operation,    11. Tests will be conducted on the power          11. The valve solenoid receives an initiation reactor vessel pressure sensors generate a      actuated relief logic using simulated input      signal.
high pressure trip signal which is used to      signal to cause trip conditions.
initiate opening of the SRVs.
Design Control Document/Tier 1 2.1-27
 
Table 2.1.2 Nuclear Boiler System (Continued)
ABWR 2.1-28 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 12. The ADS logic is automatically initiated when 12. Tests will be conducted using simulated input 12.
a low reactor water level signal is present.      signals for each NBS process variable to
: a. Upon receipt of a low water level signal, cause trip conditions in two, three, and four concurrent with a high drywell pressure instrument channels of the same process signal, at the input to the ADS initiation variable associated with each of the two logic, the following occurs:
ADS logic divisions.
(1) The main ADS timer initiates and continues to time out in the continued presence of the RPV low water level signal. The time delay for the main ADS timer is less than or equal to 29 seconds.
25A5675AA Revision 7 (2) Upon time out of the main ADS timer, a concurrent signal that represents positive indication of at least one RHR or HPCF pump running, an ADS actuation signal is generated to the associated ADS valve solenoids.
Design Control Document/Tier 1 Nuclear Boiler System
 
Table 2.1.2 Nuclear Boiler System (Continued)
ABWR Nuclear Boiler System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                            Acceptance Criteria
: 12. (Continued)                              12. (Continued)                                            b. Upon receipt of a low water level signal, in the absence of a high drywell pressure signal, at the input to the ADS initiation logic, the following occurs:
(1) The ADS high drywell pressure bypass timer initiates. The time delay for the ADS high drywell pressure bypass timer is less than or equal to 8 minutes.
(2) Upon time out of the ADS high drywell pressure bypass timer, 25A5675AA Revision 7 concurrent with an RPV low water level signal, the main ADS timer initiates and continues to time out in the continued presence of the RPV low water level signal.
(3) Upon time out of the main ADS timer, concurrent with a pump discharge pressure signal that represents positive indication of at least one RHR or HPCF pump Design Control Document/Tier 1 running, an ADS actuation signal is generated to the associated ADS valve solenoids.
: 13. For ATWS mitigation, the ADS has an      13.                                                  13.
automatic and manual inhibit of the
: a. The tests defined in item 12a will be            a. ADS actuation does not occur.
automatic ADS initiation.
conducted with a simulated APRM ATWS permissive signal present.
: b. The test defined in 12a will be conducted        b. ADS actuation does not occur.
with the ADS manual inhibit device set to 2.1-29 inhibit.
 
Table 2.1.2 Nuclear Boiler System (Continued)
ABWR 2.1-30 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                              Acceptance Criteria
: 14. The ADS can be initiated manually.              14. Tests will be conducted by initiating each      14. Upon receipt of a manual initiation signal, an ADS division manually, concurrent with a            ADS actuation signal is generated to the simulated RHR or HPCF pump running                  associated ADS valve solenoids.
signal.
: 15. The RPV water level instrumentation            15. Analyses of the as-built RPV water level        15. An analysis output exists which concludes considers the effects of dissolved non-            instrumentation will be performed using              that the RPV water level instrumentation condensable gasses in the RPV water                available test data and/or operating                considers the effects of dissolved non-instrument lines.                                  experience.                                          condensable gasses in the RPV water level instrument lines.
: 16. The mechanical portion of each division of      16. Inspections of the as-built NBS                  16. The mechanical portion of each NBS the safety-related NBS instrumentation              instrumentation will be conducted.                  instrumentation division is physically 25A5675AA Revision 7 located in the Reactor Building is physically                                                            separated from the other divisions by separated from the other divisions.                                                                      structural and/or fire barriers.
: 17. The MSL drain lines from the MSLs to the        17. An inspection of the stress report containing 17. A stress report exists. This report documents main condenser are seismically analyzed to          the dynamic analysis of the piping will be        that a dynamic seismic analysis has been withstand the SSE.                                  conducted.                                        performed.
: 18. The divisional equipment in the NBS is          18.                                                  18.
powered from its respective Class 1E
: a. Tests will be performed in the NBS by            a. The test signal exists only in the Class divisions as shown in Figures 2.1.2b, 2.1.2d providing a test signal in only one Class            1E division under test in the NBS.
and 2.1.2e. In the NBS, independence is 1E division at a time.
Design Control Document/Tier 1 provided between Class 1E divisions, and between Class 1E divisions and non-Class              b. Inspection of the as-installed Class 1E          b. Physical separation or electrical isolation 1E equipment.                                            divisions in the NBS will be performed.              exists between Class 1E divisions in the NBS. Physical separation or electrical isolation exists between Class 1E divisions and non-Class 1E equipment.
Nuclear Boiler System
: 19. Main control room displays and controls        19. Inspections will be performed on the main        19. Displays and controls exist or can be provided for the NBS are as defined in              control room displays and controls for the          retrieved in the main control room as defined Section 2.1.2.                                      NBS.                                                in Section 2.1.2.
: 20. RSS displays and controls provided for the      20. Inspections will be performed on the RSS        20. Displays and controls exist on the RSS as NBS are defined in Section 2.1.2.                  displays and controls for the NBS.                  defined in Section 2.1.2.
 
Table 2.1.2 Nuclear Boiler System (Continued)
ABWR Nuclear Boiler System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 21. MOVs designated in Section 2.1.2 as having 21. Tests of installed valves for closing will be    21. Upon receipt of an actuating signal, each an active safety function will close under    conducted under preoperational differential          MOV closes.
differential pressure, fluid flow, and        pressure, fluid flow, and temperature temperature conditions.                        conditions.
: 22. The CVs designated in Section 2.1.2 as      22. Tests of the installed valves for opening,      22. Based on the direction of the differential having an active safety-related function        closing, or both opening and closing, will be      pressure across the valve, each CV opens, open, close, or both open and closes, under      conducted under system preoperational              closes, or both opens and closes, depending system pressure, fluid flow, and temperature    pressure, fluid flow, and temperature              upon the valves safety function.
conditions.                                      conditions.
25A5675AA Revision 7 Design Control Document/Tier 1 2.1-31
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.1.3 Reactor Recirculation System Design Description The Reactor Recirculation System (RRS) is an arrangement of 10 variable speed reactor internal pumps (RIP) with motors mounted in the bottom of the RPV. The RRS circulates coolant through the reactor core at variable flow rates. The motor cooling heat exchangers are located inside the RPV pedestal adjacent to the RIP motors. Figure 2.1.3 shows the basic system configuration and scope.
Individual RIPs and motors provide at least 6912 m3/h flow with a total developed head (TDH) of at least 32.6m with 10 RIPs operating and 8291 m3/h with a TDH of at least 35.8m with 9 RIPs operating, with water at 278°C and 7.25 MPa or less. The individual RIPs, and motors have a dry rotating inertia of not less than 17.5 kgm2 and not more than 26.5 kgm2.
Figure 2.1.3 shows the ASME Code class for the RRS piping and components. The motor cover and its nuts and bolts are classified as safety-related, Seismic Category I, ASME Code Class 1 components. The remainder of the system is classified as non-safety-related.
The RIP motor cooling is provided by an auxiliary impeller mounted on the bottom of the motor rotor, which circulates water through the RIP motor and its cooling heat exchanger. The heat exchangers are cooled by the Reactor Building Cooling Water System (RCW).
Each RIP includes an anti-rotation-device (ARD) which prevents reverse RIP motor rotation by reverse flow-induced torque of equal to or less than 7.55 kNm when there is no motor power.
RIP maintenance during reactor shutdown requires a temporary plug to be installed in the RIP diffuser when the RIP impeller, shaft and motor are temporarily removed. The temporary RIP diffuser plug cannot be removed unless the RIP motor housing bottom cover is in place.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.1.3 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the RRS.
2.1-32                                                                            Reactor Recirculation System
 
ABWR Reactor Recirculation System REACTOR SHROUD VESSEL SUPPORT PUMP RPV                                  DECK RIP MOTOR RIP MOTOR COOLING RIPS                    COOLING PIPING 25A5675AA Revision 7 HEAT EXCHANGER HEAT EXCHANGER                        2 RIP    RPV RRS MOTOR HOUSING Design Control Document/Tier 1 ARD RPV                                            2 RRS RRS 1                                            RCW NOTES:
: 1. THE MOTOR COVER, COVER BOLTS AND NUTS                                MOTOR        2 ARE ASME CODE CLASS 1 COMPONENTS.                                    COVER    RPV RRS
: 2. RRS SCOPE IS RIP PUMP AND MOTOR, RIP MOTOR COVER, COVER BOLTS AND NUTS, RIP MOTOR COOLING HEAT EXCHANGER                                  RCW AND PIPING, AND ANTI-ROTATION DEVICE.
Figure 2.1.3 Reactor Recirculation System 2.1-33
 
ABWR 2.1-34 Table 2.1.3 Reactor Recirculation System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the RRS is shown 1. Inspections of the as-built system will be      1. The as-built RRS conforms with the basic on Figure 2.1.3.                              conducted.                                          configuration shown in Figure 2.1.3.
: 2. The ASME components of the RRS retain          2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the their pressure integrity under internal            Code components of the RRS required to be        ASME components of the RRS conform with pressures that will be experienced during          hydrostatically tested by the ASME Code.        the requirements in the ASME Code, Section service.                                                                                            III.
: 3. Individual RIPs and motors provide at least 3. Tests will be conducted on the individual RIP 3. Individual RIPs and motors provide at least 6912 m3/h flow with a total developed head    in a test facility which includes a calibrated  6912 m3/h flow with a total developed head (TDH) of at least 32.6m with water at least    flow element and a RIP section which is          (TDH) of at least 32.6m with water at least 278°C and 7.25 MPa or less, during 10 RIPs    geometrically the same as the RPV bottom        278°C and 7.25 MPa or less, during 10 RIPs operation. During 9 RIPs operation, the        plenum region including the RIP differential    operation. During 9 RIPs operation, the 25A5675AA Revision 7 individual RIP provides at least 8291 m3/h    pressure measurement taps. The RIP              individual RIP provides at least 8291 m3/h with a TDH of at least 35.8m at the same      performance data will be obtained for rated      with a TDH of at least 35.8m at the same temperature and pressure conditions.          reactor conditions and minimum to rated RIP      temperature and pressure conditions.
speed.
: 4. The individual RIPs and motors have a dry      4. Tests will be conducted on a RIP and motor  4. RIP and motor dry rotating inertia is  17.5 rotating inertia of  17.5 and  26.5 kgm2.      rotating assembly in a test facility.          and  26.5 kgm2.
: 5. Each RIP includes an ARD which prevents    5. Tests will be conducted on each ARD in a        5. Each ARD prevents RIP motor rotation in the reverse RIP motor rotation by reverse flow    test facility.                                      reverse direction with a reverse torque of induced torque of  7.55 kNm when there is                                                        7.55kNm.
Design Control Document/Tier 1 no motor power.
: 6. The temporary RIP diffuser plug cannot be      6. Tests of a RIP diffuser plug will be conducted 6. The temporary RIP diffuser plug cannot be removed unless the RIP motor housing              in a test facility by simulating conditions      removed unless the RIP motor housing Reactor Recirculation System bottom cover is in place.                          associated with plug removal with the motor      bottom cover is in place.
housing bottom cover removed.
 
25A5675AA Revision 7 ABWR                                                                            Design Control Document/Tier 1 2.2.1 Rod Control and Information System Design Description The Rod Control and Information System (RCIS) controls and monitors positioning of the control rods in the reactor by the fine motion control rod drive (FMCRD) units of the Control Rod Drive (CRD) System. The RCIS controls rod position to accomplish power changes in the reactor core and to achieve compliance with fuel thermal limits, core thermal-hydraulic stability limits and required FMCRD movements following reactor scram and anticipated transients without scram (ATWS) events.
The RCIS consists of redundant microprocessor-based controllers* and the equipment required to monitor and control the FMCRD. The RCIS can operate in either manual, semi-automatic or automatic control mode and has the control interfaces shown on Figure 2.2.1.
The RCIS is classified as non-safety-related.
The RCIS provides the following:
(1)    A rod worth minimizer which uses control rod position signals to enforce preestablished sequences for control rod movement when the reactor power (neutron flux) is below the low power setpoint by issuing a control rod block signal when an out of sequence control rod movement is attempted.
(2)    An automated thermal limit monitor (ATLM) which uses control rod position signals, neutron flux signals, and fuel operating thermal limits to enforce fuel thermal limits when the reactor power is above the low power setpoint and the plant is in automatic operation.
(3)    A selected control rod run-in function which uses a signal from the Recirculation Flow Control (RFC) System to insert selected control rods into the core.
(4)    An automatic control rod run-in which uses a scram-follow signal from the Reactor Protection System (RPS) to insert all control rods into the core.
(5)    An alternate rod insertion (ARI) function which uses signals from the RFC System to insert all control rods into the core.
* Except for controllers associated with individual FMCRDs.
Rod Control and Information System                                                                              2.2-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 (6)  An automatic control rod withdrawal block in response to:
(a)    A signal from the Neutron Monitoring System (NMS) multi-channel rod block monitor (MRBM), at above the low power setpoint (LPSP), or (b)    A signal from the CRD System FMCRD hollow piston/ball nut separation switches (withdrawal block applies only to separated control rod), or (c)    A signal from the RPS Mode Switch, when in Refuel Mode, that only permits the two control rods associated with the same hydraulic control unit (HCU) being withdrawn from the core at any time.
(7)  A permissive signal to the Refueling Equipment to prevent hoisting a fuel bundle over the reactor pressure vessel unless all control rods are inserted.
(8)  A runback signal to adjustable speed drives (ASD) of RFC System when RCIS initiates signals to insert all control rods.
The RCIS equipment is located in the Reactor Building and Control Building.
The RCIS is powered by two non-Class 1E uninterruptible power supplies.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.1 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the RCIS.
2.2-2                                                                    Rod Control and Information System
 
ABWR Rod Control and Information System MANUAL CONTROL PLANT INPUT SIGNALS
                                                    - FUEL OPERATING PCS      THERMAL LIMITS RFC RCIS    - CONTROL ROD POSITION                                    - RUNBACK NMS    - NEUTRON FLUX
                                                    - MRBM SIGNAL 25A5675AA Revision 7
                                                    --SELECTED SELECTEDCONTROL CONTROLROD ROD                  RCIS            RCIS RFC RFC      RUN-IN SIGNAL                      CONTROLLER          - ROD BLOCK AND CONTROL
                                                    --ARI ARISIGNAL SIGNAL                                                ROD INSERTION RPS    - SCRAM-FOLLOW SCRAM-FOLLOW    SIGNAL SIGNAL RPS    - REACTOR MODE SWITCH REFUEL MODE SIGNAL REFUELING EQUIPMENT
                                                                                                              - ALL CONTROL RODS IN SIGNAL Design Control Document/Tier 1 CRDS CRDS SEPARATION
                                                  - SEPARATION SWITCHES SWITCHES RPS    - REACTIVITY INSERTION/
SCRAM-FOLLOW    SIGNAL APR      WITHDRAWAL REQUEST SIGNALS NOTE:
: 1. INTERCONNECTIONS MAY BE FIBER-OPTIC OR METALLIC.
2.2-3                                              Figure 2.2.1 Rod Control and Information System Control Interface Diagram
 
ABWR 2.2-4                                                                            Table 2.2.1 Rod Control and Information System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspection, Tests, Analyses                          Acceptance Criteria
: 1. The equipment comprising the RCIS is            1. Inspections of the as-built system will be  1. The as-built RCIS conforms with the defined in Section 2.2.1.                          conducted.                                      description in Section 2.2.1.
: 2. The RCIS consists of redundant              2. Tests will be performed by simulating failure  2. There is no loss of RCIS output upon loss of microprocessor based controllers (except for    of each operating RCIS controller.                any one controller.
controllers associated with individual FMCRDs).
: 3. The RCIS provides a rod worth minimizer        3. Tests will be conducted on the RCIS using    3. A control rod block signal occurs when an which uses control rod position signals to        simulated control rod position signals, and      out-of-sequence control rod movement is enforce preestablished sequences for              simulated neutron flux signals.                  simulated and when reactor power is below control rod movement when the reactor                                                              the low power setpoint.
power (neutron flux) is below the low power 25A5675AA Revision 7 setpoint by issuing a control rod block signal when an out of sequence control rod movement is attempted.
: 4. The RCIS provides an ATLM which uses            4. Tests will be conducted on the RCIS using    4. A control rod block signal occurs upon control rod position signals, neutron flux        simulated control rod position signals,        simulation of a control rod movement which signals, and fuel operating thermal limits to      neutron flux signals, and fuel operating        would cause fuel thermal limits to be enforce fuel thermal limits when the reactor      thermal limits.                                approached.
power is above the low power setpoint and the plant is in automatic operation.
Design Control Document/Tier 1
: 5. The RCIS provides a selected control rod        5. Tests will be conducted on the RCIS using    5. A control rod insertion signal occurs for run-in function which uses a signal from the      simulated control rod run-in signal from RFC    those positions assigned to this function Rod Control and Information System RFC System to insert selected control rods        System.                                        upon receipt of a simulated signal from the into the core.                                                                                    RFC System.
: 6. The RCIS provides an automatic control rod 6. Tests will be conducted on the RCIS using a 6. A control rod run-in signal occurs upon run-in which uses a scram-follow signal from  simulated scram-follow signal from the RPS. receipt of a simulated scram-follow signal.
the RPS to insert all control rods into the core.
 
Table 2.2.1 Rod Control and Information System (Continued)
ABWR Rod Control and Information System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspection, Tests, Analyses                            Acceptance Criteria
: 7. The RCIS provides an ARI function which        7. Tests will be conducted on the RCIS using  7. A control rod insertion signal occurs upon uses signals from the RFC System to insert        simulated ARI signals from the RFC System. receipt of a simulated ARI signal.
all control rods into the core.
: 8. The RCIS provides an automatic control rod 8. Tests will be conducted on the RCIS using          8. A control rod withdrawal block signal occurs withdrawal block in response to:              simulated signals from the NMS MRBM at                upon receipt of simulated signals from:
: a. A signal from the NMS MRBM at above        above low power setpoint; and from the
: a. NMS MRBM at above the low power the low power setpoint.                  FMCRD separation switches; and from setpoint, control rods of the same HCU and Refuel
: b. A signal from the CRD System FMCRD        Mode position of RPS Mode Switch.                      b. FMCRD separation switches (withdrawal hollow piston/ball nut separation                                                                    block is only applicable to separated switches (withdrawal block applies only                                                              control rod),
to separated control rod).
25A5675AA Revision 7
: c. An attempt to withdraw a control rod,
: c. A signal from the RPS Mode Switch                                                                    when the RPS mode switch is in Refuel when in Refuel Mode that only permits                                                                Mode and the two control rods the two control rods associated with the                                                              associated with the same HCU are same HCU being withdrawn from the                                                                    withdrawn.
core at anytime.
: 9. The RCIS provides a permissive signal to        9. Tests will be conducted on the RCIS using      9. A permissive signal to the Refueling the Refueling Equipment to prevent hoisting        simulated rod position information.              Equipment occurs only when the simulated a fuel bundle over the reactor pressure                                                              signals indicate that all control rods are vessel unless all control rods are inserted.                                                        inserted. No signal occurs when any rod is Design Control Document/Tier 1 signalled as not inserted.
: 10. The RCIS provides a runback signal to RFC 10. Tests will be conducted on the RCIS using          10. RFC System ASD runback signals occur System ASDs when RCIS initiates signals to    simulated control rods insertion signals.              upon receipt of simulated signals to insert all insert all control rods.                                                                            control rods.
: 11. The RCIS is powered by two non-Class 1E        11. Tests will be performed on the as-built RCIS 11. The test signal exists in only one control uninterruptible supplies.                          by providing a test signal in only one non-      channel at a time.
Class 1E uninterruptible power supply at a time.
2.2-5
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.2.2 Control Rod Drive System Design Description The Control Rod Drive (CRD) System controls changes in core reactivity during power operation by movement and positioning of the neutron absorbing control rods within the core in fine increments in response to control signals from the Rod Control and Information System (RCIS). The CRD System provides rapid control rod insertion (scram) in response to manual or automatic signals from the Reactor Protection System (RPS). Figure 2.2.2 shows the basic system configuration and scope.
The CRD System consists of three major elements: (1) the electro-hydraulic fine motion control rod drive (FMCRD) mechanisms, (2) the hydraulic control unit (HCU) assemblies, and (3) the control rod drive hydraulic system (CRDHS). The FMCRDs provide electric-motor-driven positioning for normal insertion and withdrawal of the control rods and hydraulic-powered rapid control rod insertion (scram) for abnormal operating conditions. Simultaneous with scram, the FMCRDs also provide electric-motor driven run-in of control rods as a path to rod insertion that is diverse from the hydraulic-powered scram. The hydraulic power required for scram is provided by high pressure water stored in the individual HCUs. An HCU can scram two FMCRDs. It also provides the flow path for purge water to the associated drives during normal operation. The CRDHS supplies pressurized water for charging the HCU scram accumulators and purging to the FMCRDs.
There are 205 FMCRDs mounted in housings welded into the reactor vessel bottom head. The FMCRD has a movable hollow piston tube that is coupled at its upper end, inside the reactor vessel, to the bottom of a control rod. The FMCRD can move the control rod up or down over its entire range, by a ball nut and ball screw driven at a speed of 30 mm/s +/-10% by the electric stepper motor. In response to a scram signal, the piston inserts the control rod into the core hydraulically using stored energy in the HCU scram accumulator. The scram water is introduced into the drive through a scram inlet connection on the FMCRD housing, and is then discharged directly into the reactor vessel via clearances between FMCRD parts. The average scram times of all FMCRDs with the reactor pressure as measured at the vessel bottom below 7.48 MPaG are:
Percent Insertion          Time (s) 10                    0.42 40                    1.00 60                    1.44 100                    2.80 2.2-6                                                                                Control Rod Drive System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 These times are measured starting from loss of signal to the scram solenoid pilot valves in the HCUs.
The FMCRD has an electro-mechanical brake with a minimum holding torque of 49 Nm on the motor drive shaft and a ball check valve at the point of connection with the scram inlet line.
Two redundant and separate switches in the FMCRD detect separation of the hollow piston from the ball nut.
There are 103 HCUs, each of which provides water stored in a pre-charged accumulator for scramming two FMCRDs. Figure 2.2.2 shows the major HCU components. The accumulator is connected to its associated FMCRDs by a hydraulic line that includes a scram valve held closed by pressurized control air. To cause a scram, the RPS provides a signal to de-energize the scram solenoid pilot valve (SSPV)that vents the control air from the scram valve, which then opens by spring action. Loss of either electrical power to the SSPV or loss of control air pressure causes scram. A pressure switch detects low accumulator gas pressure and actuates an alarm in the main control room.
The CRD System also provides alternate rod insertion (ARI) as a means of actuating hydraulic scram when an anticipated transient without scram (ATWS) condition exists. Following receipt of an ARI signal, solenoid valves on the scram air header open to reduce pressure in the header, allowing the HCU scram valves to open. The control rod drives then insert the control rods hydraulically.
The CRDHS has pumps, valves, filters, instrumentation, and piping to supply pressurized water for charging the HCUs and purging the FMCRDs.
The CRD System components classified as safety-related are: the HCU components required for scram; the FMCRD components required for scram; the scram inlet piping; the FMCRD reactor coolant primary pressure boundary components; the FMCRD brake and ball check valve; the internal drive housing support; the FMCRD separation switches; and the HCU charging water header pressure instrumentation.
The CRD System components classified as Seismic Category I are: the HCU components required for scram; the FMCRD components required for scram; the scram inlet piping; the FMCRD reactor coolant primary pressure boundary components; the FMCRD brake and ball check valve; the internal drive housing support; the FMCRD separation switches; and the HCU charging water header pressure instrumentation.
Figure 2.2.2 shows the ASME Code class for the CRD System piping and components.
The CRD System is located in the Reactor Building. The FMCRDs are mounted to the reactor vessel bottom head inside primary containment. The HCUs and CRDHS equipment are located in the Reactor Building at the basemat elevation.
Control Rod Drive System                                                                                    2.2-7
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 Each of the four divisional HCU charging header pressure sensors are powered from their respective divisional Class 1E power supply. Independence is provided between the Class 1E divisions for these sensors and also between the Class 1E divisions and non-Class 1E equipment.
For the FMCRD separation switches, independence is provided between the Class 1E divisions and also between the Class 1E divisions and non-Class 1E equipment.
For their preferred source of power, the FMCRDs are collectively powered from one Class 1E division; for their alternate source of power, they are collectively powered from one non-Class 1E Plant Investment Protection (PIP) bus.
The hydraulic portion of the CRD System which performs the scram function is physically separated from and independent of the Standby Liquid Control System.
The CRD System has the following alarms, displays, and controls in the main control room:
(1)    Alarms for separation of the hollow piston from the ball-nut and low HCU accumulator gas pressure.
(2)    Parameter displays for the instruments shown in Figure 2.2.2.
(3)    Controls and status indication for the CRD pumps and flow control valves shown on Figure 2.2.2.
(4)    Status indication for the scram valve position.
The following CRD System safety-related electrical equipment are located in either the Reactor Building or primary containment and are qualified for a harsh environment: the HCU charging header pressure instrumentation, the scram solenoid pilot valves, and FMCRD separation switches.
The check valves (CVs) shown inside the HCU boundary on Figure 2.2.2 and the FMCRD ball check valves have active safety-related functions to close under system pressure, fluid flow, and temperature conditions.
The piping and components of the CRD pump suction supply, which extends from the CRD System interfaces with the Condensate Feedwater and Air Extraction (CFCAE) System and Makeup Water (Condensate) (MUWC) System to the inlet connections of the CRD pumps, are designed for 2.82 MPaG for intersystem loss-of-coolant accident (ISLOCA) conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.2 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the CRD System.
2.2-8                                                                            Control Rod Drive System
 
ABWR Control Rod Drive System TO OTHER HCUs ARI VALVES HCU SSPV IA CRD                                                      EXHAUST NNS P
TO OTHER                                                        CRD 2 HCUs                                                          RPV              RPV 1 CRD EXHAUST                  SCRAM P VALVE FMCRD CFCAE                      DIVISION I    P                      NNS 2 MUWC NNS  CRD      NNS CRD H2 O 25A5675AA Revision 7 DIVISION II    P N2 DIVISION III  P 2            CRD 2 NNS          RPV              RPV DIVISION IV    P 1 CRD SUCTION F          F  FILTERS P                                            FMCRD PRIMARY F        FE                                            CONTAINMENT NNS Design Control Document/Tier 1 MUWC CRD                                                        P F                                          TO OTHER HCUs FLOW CONTROL DRIVE WATER              VALVES CRD PUMPS                        FILTERS NOTES:
: 1. THERE ARE A TOTAL OF 205 FMCRDS AND 103 HCUs.
: 2. THE SSPV FUNCTION IS REPRESENTED BY A SEPARATE SOLENOID VALVE AND A PNEUMATIC VALVE; IN ACTUAL APPLICATION, THEY MAY BE COMBINED INTO A SINGLE VALVE ASSEMBLY THAT IS FUNCTIONALLY EQUIVALENT.
Figure 2.2.2 Control Rod Drive System 2.2-9
 
ABWR 2.2-10 Table 2.2.2 Control Rod Drive System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the CRD System        1. Inspections of the as-built system will be  1. The as-built CRD System conforms with the is as shown on Figure 2.2.2.                        conducted.                                      basic configuration shown on Figure 2.2.2.
: 2. The ASME Code components of the CRD            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              code components of the CRD System                ASME Code components of the CRD integrity under internal pressures that will be    required to be hydrostatically tested by the    System conform with the requirements in the experienced during service.                        ASME Code.                                      ASME Code, Section III.
: 3. The FMCRD can move the control rod up or 3. Tests will be conducted on each installed            3. Each control rod moves up and down over its down over its entire range by a ball nut and FMCRD.                                                  entire range at a speed of 30 mm/s +/-10%.
ball screw driven at a speed of 30 mm/s                                                              The time to insert each control rod from full-
                              +/-10% by the electric stepper motor.                                                                  out to full-in is  135 seconds when driven by the electric stepper motor.
25A5675AA Revision 7
: 4. The average scram times of all FMCRDs        4. Tests will be conducted on each installed    4. The average scram times of all FMCRDs with the reactor pressure as measured at the    HCU and its associated FMCRD. The results        with the reactor pressure as measured at the vessel bottom below 7.48 MPaG are:              of the tests performed at low reactor            vessel bottom below 7.48 MPaG are:
Percent Insertion    Time (s)              pressure  will be extrapolated to the Design Percent Insertion    Time (s) 10              0.42              Commitment pressure (7.48 MPaG).
10              0.42 40              1.00                                                                          40              1.00 60              1.44                                                                          60              1.44 100                2.80                                                                        100                2.80 Design Control Document/Tier 1 These times are measured starting from loss                                                          These times are measured starting from loss of signal to the scram solenoid pilot valves in                                                      of signal to the scram solenoid pilot valves in the HCU.                                                                                              the HCU
: 5. The FMCRD has an electro-mechanical              5. Tests of each FMCRD brake will be            5. The FMCRD electro-mechanical brake has a Control Rod Drive System brake with a minimum holding torque of 49            conducted in a test facility.                  minimum holding torque of 49 Nm on the Nm on the motor drive shaft.                                                                        motor drive shaft.
: 6. Two redundant and separate switches in the 6. Tests of each as-built FMCRD will be                6. Both switches in each FMCRD detect FMCRD detect separation of the hollow        conducted.                                            separation of the hollow piston from the ball piston from the ball nut.                                                                            nut.
 
Table 2.2.2 Control Rod Drive System (Continued)
ABWR Control Rod Drive System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 7. Following receipt of an ARI signal, solenoid    7. Tests will be conducted on the as-built ARI      7. Following receipt of a simulated ARI signal, valves on the scram air header open to            valves using a simulated actuation signal.          solenoid valves on the scram air header reduce pressure in the header, allowing the                                                            open to reduce pressure in the header, HCU scram valves to open.                                                                              allowing the HCU scram valves to open.
: 8. Each of the four divisional HCU charging        8.                                                  8.
header pressure sensors are powered from
: a. Tests will be conducted on the as-built          a. The test signal exists only in the Class their respective divisional Class 1E power charging water header sensors by                    1E Division under test.
supply. For the four HCU charging water providing a test signal in only one Class header pressure sensors, independence is 1E division at a time.
provided between Class 1E divisions, and between Class 1E divisions and non-Class            b. Inspections of the as-installed charging        b. Physical separation or electrical isolation 1E equipment.                                          water header sensor Class 1E divisions              exists between Class 1E divisions.
25A5675AA Revision 7 will be conducted.                                  Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 9. For the FMCRD separation switches,        9. Inspections of the as-installed Class 1E              9. In the CRD System, physical separation or independence is provided between the Class    divisions in the CRD System will be                      electrical isolation exists between Class 1E 1E divisions and also between the Class 1E    performed.                                              divisions. Physical separation or electrical divisions and non-Class 1E equipment.                                                                  isolation exists between Class 1E divisions and non-Class 1E equipment.
: 10. For their preferred source of power, the      10. Inspections of the as-built CRD System will      10. For their preferred source of power, the Design Control Document/Tier 1 FMCRDs are collectively powered from one          be conducted.                                        FMCRD motors are collectively powered Class 1E division; for their alternate source                                                          from one Class 1E division; for their alternate of power, they are collectively powered from                                                          source of power, they are collectively one non-Class 1E PIP bus.                                                                              powered from one non-Class 1E PIP bus.
: 11. Main control room alarms, displays and        11. Inspections will be performed on the main      11. Alarms, displays and controls exist or can be controls provided for the CRD System are          control room alarms, displays and controls          retrieved in the main control room as defined defined in Section 2.2.2.                          for the CRD System.                                in Section 2.2.2.
: 12. CVs designated in Section 2.2.2 as having    12. Tests of installed valves for closing will be    12. Each CV closes.
an active safety-related function close under    conducted under system preoperational system pressure, fluid flow, and temperature      pressure, fluid flow, and temperature 2.2-11                        conditions.                                      conditions.
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.2.3 Feedwater Control System Design Description The Feedwater Control (FDWC) System controls the flow of feedwater into the reactor pressure vessel (RPV) to maintain the water level in the vessel during plant operation. The FDWC System consists of redundant, microprocessor-based controllers, and flow sensors for main steamlines and feedwater lines, as shown in the control interface diagram in Figure 2.2.3.
The FDWC digital controllers determine narrow range level signal using three reactor level measurement inputs from the NBS. Sensor signals are transmitted to the FDWC digital controllers by the Non-Essential Multiplexing System (NEMS).
The steam flow in each of four main steamlines is sensed at the RPV nozzle venturis. Sensor signals are transmitted to the FDWC System digital controllers by the NEMS. These measurements are processed in the digital controllers to calculate the total steam flow out of the vessel.
Feedwater flow is sensed at a flow element in each of the two feedwater lines. Sensor signals are transmitted to the FDWC digital controllers by the NEMS. These measurements are processed in the digital controllers to calculate the total feedwater flow into the vessel.
The FDWC System is classified as non-safety-related.
The FDWC System operates in either manual, automatic single-element or automatic three-element control modes. At low feedwater flow, the FDWC System utilizes only water level measurement in automatic single-element control mode. At higher flow rates, the FDWC System in three-element control mode uses water level, steam flow, and feedwater flow measurements for water level control.
The FDWC System monitors reactor water level signals and, if a high RPV water level setpoint is reached, sends trip signals to the Turbine Control System and to the Condensate, Feedwater and Condensate Air Extraction (CFCAE) System. If a low RPV water level setpoint is reached, the FDWC System sends trip signals to the Recirculation Flow Control (RFC) System.
If the FDWC System receives an anticipated transient without scram (ATWS) trip signal from the Safety System Logic and Control (SSLC), the FDWC System issues signals to runback feedwater flow.
Each channel of the FDWC System is powered by separate non-Class 1E uninterruptible power supplies.
The total feedwater flow is displayed on the main control panel. The FDWC System operating mode is selectable from the main control room. The FDWC System microprocessors are located in the Control Building.
2.2-12                                                                              Feedwater Control System
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 Digital controllers used for the FDWC System are redundant, with diagnostic capabilities that identify and isolate failure of level input signals.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.3 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Feedwater Control System.
Feedwater Control System                                                                                2.2-13
 
ABWR 2.2-14 FROM SSLC  ATWS FEEDWATER FLOW RUNBACK TRIP HIGH (TURBINE CONTROL SYSTEM/CONDENSATE, 3-ELEM FEEDWATER, AND CONDENSATE                                            ACTUATOR LEVEL                                                            M/A        VOTER AIR EXTRACTION SYSTEM)                                                DEMAND TRIPS                                                                                SIGNAL LOW (RFC SYSTEM)                        1-ELEM OTHER CHANNEL  (CF&CAE) 3-ELEM ACTUATOR WATER                                                                                      M/A        VOTER FROM NBS                                                                                                                    DEMAND LEVEL 1-ELEM                      SIGNAL (CF&CAE)
FDWC STEAM                                                                                            OTHER CHANNEL FLOW 3-ELEM ACTUATOR 25A5675AA Revision 7 M/A        VOTER DEMAND SIGNAL 1-ELEM FDWC LEVEL                    LEVEL                                                                                  (CF&CAE)
FLOW                                          OTHER CHANNEL SETPOINT                CONTROLLER            CONTROLLER FDWC FEEDWATER FLOW Design Control Document/Tier 1 1-ELEM = SINGLE-ELEMENT MODE DIAGRAM REPRESENTS ONE CHANNEL                                3-ELEM = THREE-ELEMENT MODE M/A    = MANUAL / AUTO STATION Feedwater Control System NOTE:
: 1. INTERCONNECTIONS MAY BE FIBER-OPTIC OR METALLIC.
Figure 2.2.3 Feedwater Control System Control Interface Diagram
 
ABWR Feedwater Control System Table 2.2.3 Feedwater Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The equipment comprising the FDWC                1.Inspections of the as-built system will be      1. The as-built FDWC System conforms with System is defined in Section 2.2.3.                  conducted.                                        the description in Section 2.2.3.
: 2. The FDWC System controls the flow of            2. A test will be performed by simulating a        2. A signal to increase feedwater flow occurs.
feedwater into the RPV.                            decreasing reactor level signal.
: 3. The FDWC System monitors reactor water        3. Tests will be performed on the FDWC              3. When a high RPV water level setpoint is level signals and, if a high RPV water level      System, using simulated RPV water level              reached, trip signals are sent to the Turbine setpoint is reached, sends trip signals to the    signals.                                            Control System and CFCAE System.
Turbine Control System and to the CFCAE When a low RPV water level setpoint is System.
reached, a trip signal is sent to the RFC If a low RPV water level setpoint is reached,                                                          System.
25A5675AA Revision 7 the FDWC System sends trip signals to the RFC System.
: 4. If the FDWC System receives an ATWS trip 4. Tests will be performed on the FDWC        4. When an ATWS trip signal is received, the signal from the SSLC, FDWC issues signals  System, using a simulated ATWS trip signal. FDWC System issues feedwater runback to runback feedwater flow.                                                                signals.
: 5. The FDWC System digital controllers are          5. Tests will be performed by providing a test    5. The test signal exists in only one digital powered by separate non-Class 1E                    signal in only one non-Class 1E                    control channel at a time.
uninterruptible power supplies.                    uninterruptible power supply at a time.
: 6. Main control room controls and displays          6. Inspections will be performed on the main      6. Controls and displays exist or can be Design Control Document/Tier 1 provided for the FDWC System are defined            control room controls and displays for the        retrieved in the main control room as defined in Section 2.2.3.                                  FDWC System.                                      in Section 2.2.3.
: 7. Digital controllers used for the FDWC            7. Tests will be performed by simulating failure  7. There is no loss of FDWC System output System are redundant.                              of each operating FDWC System digital              upon loss of any one digital controller.
controller.
: 8. Digital controllers used for the FDWC            8. Tests will be performed by simulating level    8. There is no loss of FDWC System output System have diagnostic capabilities that            input signal failures to the FDWC System          upon loss of any one level input signal.
identify and isolate failure of level input        digital controllers.
signals.
2.2-15
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.2.4 Standby Liquid Control System The Standby Liquid Control (SLC) System injects neutron absorbing poison into the reactor using a boron solution, thus providing the safety-related function of backup reactor shutdown capability independent of the normal reactivity control system based on insertion of control rods into the core. The SLC System is designed to bring the reactor from full power to a subcritical condition without control rod movement, at any time in a core cycle, and at design basis conditions with the reactor in the most reactive xenon-free state. The SLC System operates over a range of reactor pressure conditions which bound the elevated pressures associated with an anticipated transient without scram (ATWS). Figure 2.2.4 shows the basic system configuration and scope.
The SLC System consists of a boron solution storage tank, two positive displacement pumps, two motor-operated injection valves which are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged through the high pressure core flooder (HPCF) Division B subsystem sparger.
The SLC System uses a dissolved solution of sodium pentaborate as the neutron-absorbing poison. This solution is held in the storage tank which has a heater to maintain solution temperature above the saturation temperature. The heater has automatic actuation and automatic shutoff.
A test tank and associated piping and valves permit testing of the SLC System during plant operation. The tank is supplied with demineralized water, which is pumped in either a closed loop or is injected into the reactor.
Key SLC System equipment performance requirements are:
(1)    Pump flow (minimum)                  378 L/min with both pumps operating 189 L/min with one pump operating (2)    Maximum reactor pressure              8.72 MPaA (for injection)
(3)    Pumpable volume in storage tank      23.1 m3 (minimum)
The SLC System can be manually initiated from the main control room. Each of the two divisions is controlled by a separate switch. When it is manually initiated to inject a liquid neutron absorber into the reactor, the following devices and actions are initiated by each divisional switch:
(1)    The specified division injection valve is opened.
2.2-16                                                                          Standby Liquid Control System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 (2)    The specified division storage tank discharge valve is opened.
(3)    The specified division injection pump is started.
(4)    The reactor water cleanup isolation valves are closed.
Both divisions of the SLC System are automatically initiated during an ATWS condition by safety system and logic control (SSLC) logic. With the storage tank at minimum level and both pumps operating, the system is designed to inject the minimum required boron solution.
Each SLC System pump has an interlock which prevents operation if both the test tank outlet valve and the pump suction valve are closed.
The SLC System provides borated water to the reactor core to compensate for the various reactivity effects. These effects are xenon decay, elimination of steam voids, changing water density due to the reduction in water temperature, Doppler effect in uranium, changes in neutron leakage, and changes in control rod worth. To meet this objective, it is necessary to inject a quantity of boron which produces a minimum concentration of 850 parts per million (ppm) by weight of natural boron in the reactor core at 20°C. To allow for potential leakage and imperfect mixing in the reactor system, an additional approximately 25% (220 ppm) is added to the above requirement, resulting in a total requirement of greater than or equal to 1070 ppm.
The required concentration is thus achieved in a mass of water equal to the sum of the mass of water in the RPV at normal water level (equal to or less than 455 x 103 kg) plus the mass of water in the RPV shutdown cooling piping (equal to or less than 130 x 103 kg). The quantity of boron solution contained in the storage tank above the pump suction shutoff level provides the required concentration of 1070 ppm when injected into the reactor.
The SLC System pumps have sufficient net positive suction head (NPSH) available at the pump. The SLC System pumps are designed to produce discharge pressure to inject the solution into the reactor when the reactor is at pressure conditions corresponding to the system relief valve (10.79 MPaG), which is above peak ATWS pressure in the RPV.
SLC System components required for RPV injection are classified as Seismic Category I.
Figure 2.2.4 shows the ASME Code class for the SLC System piping and components.
The SLC System is located in the Reactor Building. The storage tank, test water tank, the two positive displacement pumps, and associated valving are located in the secondary containment on the floor elevation below the operating floor.
Each of the two SLC System divisions is powered from the respective Class 1E division as shown on Figure 2.2.4. The power supplied to one motor-operated injection valve, suction valve, and injection pump is powered from Division I. The power supply to the other motor-operated injection valve, suction valve, and injection pump is powered from Division II. In the Standby Liquid Control System                                                                              2.2-17
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 SLC system, independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
The SLC System has the following displays, controls and alarms in the main control room:
* Alarms for storage tank temperature and level.
* Parameter displays for the instruments shown on Figure 2.2.4.
* Controls and status indication for the pumps, injection valves, and suction valves.
* A manual system initiation switch for each division.
The motor-operated valves (MOVs) shown on Figure 2.2.4 have an active safety-related function and perform this function under differential pressure, fluid flow and temperature conditions.
The check valves (CVs) shown on Figure 2.2.4 have active safety-related functions to open, close, or both open and close under system pressure, fluid flow, and temperature conditions.
The SLC System is physically separated from and independent of the hydraulic portion of the Control Rod Drive (CRD) System.
The piping and components on the suction side of the pumps up to and including the suction valves and the test loop up to the test tank inlet valve have a design pressure of 2.82 MPaG for intersystem LOCA (ISLOCA) conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.4 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the SLC System.
2.2-18                                                                          Standby Liquid Control System
 
ABWR Standby Liquid Control System VENT PRIMARY                                                                  NNS CONTAINMENT STORAGE                  2 TANK H
T          A L
H L          A L
SAM SLC 2                    HEATER NOTE 1      NOTE 2          SUCTION VALVES HPCF-B 25A5675AA Revision 7 M                        M 1 SLC NOTE 1                                      NOTE 2 PUMPS INJECTION VALVES OUTLET 1    M          M        1                                                VALVE Design Control Document/Tier 1 2  NOTE 1    NOTE 2      2 2
NNS INLET NOTES:                                                                                              VALVE
: 1. POWERED BY CLASS 1E DIVISION I.                                                        P        2
: 2. POWERED BY CLASS 1E DIVISION II.                                                                NNS TEST TANK 2.2-19                                                                Figure 2.2.4 Standby Liquid Control System
 
                                /
ABWR 2.2-20 Table 2.2.4 Standby Liquid Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                              Acceptance Criteria
: 1. The basic configuration of the SLC System is 1. Inspections of the as-built system will be            1. The as-built SLC System conforms with the shown in Figure 2.2.4.                          conducted.                                              basic configuration shown in Figure 2.2.4.
: 2. The ASME Code components of the SLC            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the SLC System that          ASME Code components of the SLC System integrity under internal pressures that will be    are required to be hydrostatically tested by    conform with the requirements in the ASME experienced during service.                        the ASME Code.                                  Code, Section III.
: 3.                                                3.                                                    3.
: a. A test tank and associated piping and          a. Tests will be conducted on each division          a.
valves permit testing of the SLC System            of the as-built SLC System using during plant operation. The tank is                installed controls, power supplies and 25A5675AA Revision 7 supplied with demineralized water, which          other auxiliaries. The following tests will is pumped in either a closed loop or is            be conducted:
injected into the reactor.
(1) Demineralized water will be pumped                (1) Demineralized water is pumped with against a pressure greater than or                    a flow rate greater than or equal to equal to 8.72 MPaA in a closed loop                    189 L/min in the closed loop.
on the test tank.
(2) Demineralized water will be injected              (2) Demineralized water is injected from from the test tank into the reactor.                  the test tank into the reactor.
: b. The SLC System delivers at least                b. Tests will be conducted on the as-built            b. The SLC System injects greater than or Design Control Document/Tier 1 378 L/min of solution with both pumps              SLC System using installed controls,                  equal to 378 L/min into the reactor with operating when the reactor pressure is            power supplies and other auxiliaries.                both pumps running against a discharge less than or equal to 8.72 MPaA.                  Demineralized water will be injected                  pressure of greater than or equal to 8.72 from the storage tank into the reactor                MPaA.
Standby Liquid Control System with both pumps running against a discharge pressure of greater than or equal to 8.72 MPaA.
 
Table 2.2.4 Standby Liquid Control System (Continued)
ABWR Standby Liquid Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: c. The SLC System delivers at least 189        c. Tests will be conducted on the as-built    c. The SLC System injects greater than or L/min of solution with either pump                SLC System using installed controls,            equal to 189 L/min into the reactor with operating when the reactor pressure is            power supplies and other auxiliaries.          either pump running against a discharge less than or equal to 8.72 MPaA.                  Demineralized water will be injected            pressure greater than or equal to 8.72 from the storage tank into the reactor          MPaA.
with one pump running against a discharge pressure of greater than or equal to 8.72 MPaA.
: d. The SLC System can be manually                d. Tests will be conducted on the as-built      d. Each division of the SLC System initiates initiated from the main control room.            SLC System using the manual initiation          when the manual initiation switch for that switch.                                        division is actuated.
25A5675AA Revision 7
: e. Both divisions of the SLC System are          e. Tests will be conducted on the as-built      e. Upon receipt of a simulated ATWS automatically initiated during an ATWS.          SLC System using simulated ATWS                signal, both divisions of SLC signals.                                        automatically initiate.
: f. Each SLC System pump has an interlock        f. Tests will be conducted on each SLC        f. Each SLC System pump is prevented which prevents operation if both the test        System pump start logic using simulated        from operating unless signals indicative tank outlet valve and the pump suction            valve position signals                          of one of the following conditions exist:
valve are closed.                                                                                (1) A suction path from the storage tank is available (the pump suction valve is fully open).
Design Control Document/Tier 1 (2) A suction path from the test tank is available (the test tank outlet valve is fully open).
2.2-21
 
Table 2.2.4 Standby Liquid Control System (Continued)
ABWR 2.2-22 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                        Acceptance Criteria
: g. The performance of the SLC System is            g. The as-built dimensions will be used in a    g.
based on the following plant parameters:            volumetric analysis to calculate the (1) Storage tank pumpable volume is              volumes listed below:
greater than or equal to 23.1 m3.              (1) Minimum Storage tank pumpable              (1) Storage tank pumpable volume is volume.                                        greater than or equal to 23.1 m3.
(2) RPV water inventory is less than or equal to 455 x 103 kg at normal                (2) RPV water inventory at normal water        (2) RPV water inventory is less than or water level and 20°C.                              level and 20°C.                                equal to 455 x 103 kg at 20°C.
(3) RHR shutdown cooling system inventory is less than or equal to 130          (3) RHR shutdown cooling system                (3) RHR shutdown cooling system x 103 kg at 20°C.                                  water inventory at 20°C.                        inventory is less than or equal to 130 x 103 kg at 20°C.
25A5675AA Revision 7
: h. The SLC pumps have sufficient NPSH.              h. Tests will be conducted on the as-built      h. The available NPSH exceeds the NPSH SLC System by injecting demineralized          required as demonstrated by the SLC water using both SLC System pumps              System injecting greater than or equal to from the storage tank to the RPV with          378 liters/minute.
the storage tank at the low level (pump trip level) and a temperature of greater than or equal to 43°C.
: i. The SLC System pump relief valves              i. Shop or field tests will be conducted      i. The SLC System pump relief valves open when the inlet pressure to the                using the SLC System pump to                    open when the inlet pressure to the Design Control Document/Tier 1 valve equals or exceeds the setpoint                determine the relief valve setpoint.            valve equals or exceeds 10.76 MPaG.
(10.76 MPaG).
Standby Liquid Control System
 
Table 2.2.4 Standby Liquid Control System (Continued)
ABWR Standby Liquid Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 4. In the SLC System, independence is            4.                                                  4.
provided between Class 1E divisions, and          a. Tests will be conducted on the SLC              a. The test signal exists only in the Class also between Class 1E divisions and non-              System by providing a test signal in only          1E Division under test in the SLC Class 1E equipment.
one Class 1E Division at a time.                    System.
: b. Inspection of the as-built SLC System            b. In the SLC System, physical separation will be performed.                                  or electrical isolation exists between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 5. Main control room alarms, displays, and      5. Inspections will be performed on the main        5. Alarms, displays, and controls exist or can 25A5675AA Revision 7 controls provided for the SLC System are        control room alarms, displays, and controls        be retrieved in the main control room as defined in Section 2.2.4.                        for the SLC System.                                defined in Section 2.2.4.
: 6. MOVs designated in Section 2.2.4 as having 6. Tests of the installed valves for opening will      6. Upon receipt of the actuating signal, each an active safety-related function open under  be conducted under preoperational                      MOV opens.
system pressure, fluid flow, and temperature  differential pressure, fluid flow, and conditions.                                  temperature conditions.
: 7. The CVs designated in Section 2.2.4 as        7. Tests of the installed valves for opening,      7. Based on the direction of the differential having an active safety-related function        closing, or both opening and closing, will be      pressure across the valve, each CV opens, open, close, or both open and close under        conducted under system preoperational              closes, or both opens and closes, depending Design Control Document/Tier 1 system pressure, fluid flow, temperature        pressure, fluid flow, and temperature              upon the valves safety function.
conditions.                                      conditions.
2.2-23
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.2.5 Neutron Monitoring System Design Description The Neutron Monitoring System (NMS) is a neutron monitoring and protection system. The functions of the system are to:
(1)  Monitor the thermal neutron flux in the reactor core.
(2)  Provide trip signals to the Reactor Protection System (RPS).
(3)  Provide power information to the operator and plant control systems.
The startup range neutron monitor (SRNM), the local power range monitor (LPRM), and the average power range monitor (APRM) are classified as Class 1E safety-related. The automated incore instrument calibration system and the multi-channel rod block monitor (MRBM) are classified as non-safety-related.
The SRNM monitors neutron flux from the source range to 15% of the rated power. The SRNM has ten SRNM channels, each with one detector, which are distributed throughout the reactor core and assigned to four divisions. The SRNM detector is a fixed in-core sensor. Detector cables are separated according to different divisional assignment, connected to their designated preamplifiers located in the Reactor Building, and then transmitted to signal processing electronic units in the Control Building.
The LPRM monitors local neutron flux in the power range up to 125% of the rated power, and overlaps with part of the SRNM range. LPRM detector assemblies are provided and are distributed in the core, with four sensors per each LPRM assembly, to monitor local neutron flux level throughout the core. The LPRM assembly also contains space for automated in-core calibration detector. The LPRM detector outputs are connected to the APRM signal conditioning units in the Control Building, where the signals are processed and amplified.
LPRM detector signals are divided and assigned to four APRM channels corresponding to four divisions. LPRM signals in each APRM channel are summed and averaged to form an APRM signal which represents the core average power.
The Oscillation Power Range Monitor (OPRM) is part of the APRM. Each OPRM receives the identical LPRM signals from the corresponding APRM channel as inputs, and forms many OPRM cells to monitor the neutron flux behavior of all regions of the core. The LPRM signals assigned to each cell are summed and averaged to provide an OPRM signal for this cell. The OPRM trip protection algorithm detects thermal hydraulic instability and provides trip output to the RPS if the trip setpoint is exceeded. The OPRM bypass is controlled by the bypass of the APRM channel it resides with.
The automated in-core instrument calibration system provides local power information at various core locations that correspond to LPRM locations. The automated in-core instrument 2.2-24                                                                            Neutron Monitoring System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 calibration system uses its own set of in-core detectors for local power measurement and provides local power information for three-dimension core power determination and for the calibration of the LPRMs. The measured data are sent to the Process Computer System for such calculation and LPRM calibration.
The MRBM uses LPRM signals to detect local power change during the rod withdrawal. If the averaged LPRM signal exceeds a preset rod block setpoint, a control rod block demand is issued.
Figure 2.2.5 shows the configuration of each NMS division.
Each of the four divisions of the SRNM, LPRM and APRM instruments is powered by its respective divisional Class 1E power supplies. In the NMS outside the primary containment, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The SRNM and APRM trip signal outputs are in four divisions. The SRNM trip and the APRM trip logic are independent from each other. The SRNM generates a high neutron flux trip or a short period trip signal. Any single SRNM channel trip causes a trip in its division. The APRM can generate a high neutron flux trip, a simulated thermal power (STP) trip signal, a rapid core flow decrease trip signal, or a core power oscillation trip signal. The NMS provides these trip signals to the Reactor Protection System (RPS).
The SRNM and APRM are fail-safe in the event of loss of electrical power to any division of their logic.
The NMS bypass function is performed within the NMS. Within the NMS, the bypass functions of the SRNM and the APRM are separate and independent from each other. The SRNM channels are grouped into three bypass groups. Individual SRNM channels can be bypassed. At any one time, up to three SRNM channels can be bypassed. At any one time, only one APRM channel can be bypassed. A bypassed SRNM channel or a bypassed APRM channel does not cause a trip output sent to the RPS.
The NMS provides SRNM flux permissive signal to the Standby Liquid Control (SLC) System and feedwater runback logic within Safety System Logic and Control (SSLC) and an APRM flux permissive signal to the Nuclear Boiler System (NBS) logic within SSLC as part of the anticipated transient without scram (ATWS) logic. The SRNM and APRM flux permissive signals from the NMS indicate when the reactor power level is above or below the setpoint in order to allow or disallow the initiation of ATWS mitigation features.
The NMS has the following displays and controls in the main control room:
(1)    SRNM, LPRM, and APRM neutron flux displays.
(2)    Trip and bypass status displays.
Neutron Monitoring System                                                                                2.2-25
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 (3)  Bypass control devices.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.5 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the NMS.
2.2-26                                                                            Neutron Monitoring System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 DETECTOR        DETECTOR                          DETECTOR RFC CORE PLATE DIFFERENTIAL PRESSURE SENSOR (NOTE 2)
SRNM PREAMP LPRM/APRM (INCLUDES OPRM)
SRNM MRBM                              IN-CORE INSTRUMENT CALIBRATION SYSTEM NMS BOUNDARY MAIN                          RFC            PROCESS RCIS            SSLC                          RCIS CONTROL                        SYSTEM          COMPUTER (ROD BLOCK)        (RPS TRIP)                      (FLUX)
ROOM                          (FLUX)          SYSTEM DISPLAY SSLC                                          SSLC (SLC SYSTEM AND                                  (NBS /ADS)
FEEDWATER RUNBACK)                                  (NOTE 3)
(NOTE 3)
NOTES:
: 1. DIAGRAM REPRESENTS ONE OF FOUR NMS DIVISIONS (MRBM IS A DUAL CHANNEL SYSTEM.
THERE IS ONLY ONE IN-CORE INSTRUMENT CALIBRATION SYSTEM).
: 2. USED FOR RAPID CORE FLOW DECREASE TRIP.
: 3. SRNM AND APRM ATWS PERMISSIVE SIGNALS TO SSLC.
: 4. INTERCONNECTIONS MAY BE FIBER-OPTIC OR METALLIC.
Figure 2.2.5 Neutron Monitoring System Neutron Monitoring System                                                                                  2.2-27
 
ABWR 2.2-28 Table 2.2.5 Neutron Monitoring System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The equipment comprising the NMS is          1. Inspection of the as-built system will be  1. The as-built NMS conforms with the defined in Section 2.2.5.                      conducted.                                    description in Section 2.2.5.
: 2. The OPRM trip protection algorithm detects 2. Tests will be conducted on OPRM using        2. A trip signal to the RPS is generated when thermal hydraulic instability and provides trip simulated LPRM input signals.                  the simulated LPRM signals cause the output to the RPS if the trip setpoint is                                                      OPRM signal to exceed the trip setpoint.
exceeded.
: 3. The MRBM uses LPRM signals to detect        3. Tests will be conducted on MRBM using      3. A control rod block demand signal is issued local power change during the rod              simulated LPRM input signals.                  when the simulated averaged LPRM signal withdrawal. If the averaged LPRM signal                                                        exceeds the preset rod block setpoint.
exceeds a preset rod block setpoint, a control rod block demand is issued.
25A5675AA Revision 7
: 4. Each of the four divisions of the SRNM,  4.                                              4.
LPRM and APRM instruments is powered by
: a. Tests will be performed on the NMS by        a. The test signal exists only in the Class its respective divisional Class 1E power providing a test signal to only one Class      1E division under test in the NMS.
supplies. In the NMS independence is 1E division at a time.
provided between Class 1E divisions, and between Class 1E divisions and non-Class    b. Inspection of the as-installed Class 1E      b. In the NMS, physical separation or 1E equipment.                                  divisions in the NMS  will be performed.      electrical isolation exists between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E Design Control Document/Tier 1 equipment.
: 5. The SRNM generates a high neutron flux trip 5. Tests will be conducted on the SRNM using    5. Trip signals are generated when the or a short period trip signal. Any single      simulated neutron flux and period signals.      simulated input signals exceed trip setpoints.
SRNM channel trip causes a trip in the                                                        Any single SRNM channel trip causes a trip Neutron Monitoring System division.                                                                                      in its division.
: 6. The APRM can generate high neutron flux      6. Tests will be conducted on the APRM using  6. Trip signals are generated when the trip trip, a STP trip signal, a rapid core flow      simulated neutron flux, and core plate        setpoints for high neutron flux, a high STP, a decrease trip signal, or a core power          differential pressure signals.                rapid core flow decrease, and a core power oscillation trip signal.                                                                      oscillation are exceeded.
 
Table 2.2.5 Neutron Monitoring System (Continued)
ABWR Neutron Monitoring System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 7. The SRNM and APRM are fail-safe in the          7. Tests will be conducted on the SRNM and      7. Upon loss of electrical power to one division event of loss of electrical power to any            APRM by disconnecting electrical power to        of either the SRNM or APRM a trip signal is division of their logic.                            one division of logic at a time.                generated in that division.
: 8. Within the NMS, the bypass functions of the      8. Inspections and tests will conducted on the  8. Within the NMS, the bypass functions of the SRNM and the APRM are separate and                  SRNM and APRM bypass functions.                  SRNM and the APRM are separate and independent from each other. The SRNM                                                                independent from each other. The SRNM channels are grouped into three bypass                                                              channels are grouped into three bypass groups. Individual SRNM channels can be                                                              groups. Individual SRNM channels can be bypassed. At any one time, up to three                                                              bypassed. At any one time, up to three SRNM channels can be bypassed. At any                                                                SRNM channels can be bypassed. At any one time, only one APRM channel can be                                                              one time, only one APRM channel can be bypassed.                                                                                            bypassed.
25A5675AA Revision 7
: 9. A bypassed SRNM channel or a bypassed            9. Tests will be conducted on the SRNM and      9. No trip output signal is sent to the RPS, APRM channel does not cause a trip output          APRM bypassed channels using simulated          when a simulated input signal is provided to sent to the RPS.                                    input signals.                                  a bypassed SRNM or a bypassed APRM channel.
: 10. The SRNM and APRM flux permissive              10. Test will be conducted using simulated      10. The SRNM and APRM flux permissive signals from the NMS indicate when the              SRNM and APRM flux signals.                      signals from the NMS indicate when the reactor power level is above or below the                                                            reactor power level is above or below the setpoint in order to allow or disallow the                                                          setpoint in order to allow or disallow the initiation of ATWS mitigation features.                                                              initiation of ATWS mitigation features.
Design Control Document/Tier 1
: 11. Main control room displays and controls        11. Inspections will be performed on the main    11. Displays and controls exist or can be provided for the NMS are as defined in              control room displays and controls for the      retrieved in the main control room as defined Section 2.2.5.                                      NMS.                                            in Section 2.2.5.
2.2-29
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 2.2.6 Remote Shutdown System Design Description The Remote Shutdown System (RSS) provides remote manual control of safety-related systems to bring the reactor to hot shutdown and subsequent cold shutdown conditions from outside the main control room (MCR). Figure 2.2.6 shows the basic system configuration and scope.
The RSS has two divisional panels and associated controls and indicators for interfacing with the following systems:
(1)  Residual Heat Removal (RHR) System (2)  High Pressure Core Flooder (HPCF) System (3)  Nuclear Boiler System (NBS)
(4)  Reactor Service Water (RSW) System (5)  Reactor Building Cooling Water (RCW) System (6)  Electrical Power Distribution (EPD) System (7)  Atmospheric Control (AC) System (8)  Emergency Diesel Generator (DG)
(9)  Make-up Water System (Condensate), (MUWC)
(10) Flammability Control System (FCS)
(11) Suppression Pool Temperature Monitoring (SPTM) System (12) High Pressure Nitrogen Gas Supply System RSS controls and indicators are hard-wired direct to the interfacing components and sensors.
The RSS is classified as a Class 1E safety-related system.
Operation of transfer switches on the RSS panel overrides and isolates the controls from the MCR and transfers control to the RSS. Transfer switch actuation causes alarms in the MCR.
Indications required for plant shutdown are provided on the RSS panels as shown on Figure 2.2.6.
2.2-30                                                                            Remote Shutdown System
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 RSS Division A has the following automatic controls and interlocks for RHR System Division A. RSS Division B has the following automatic controls and interlocks for RHR System Division B and HPCF System Division B:
(1)  RHR minimum flow valve A(B) is commanded open upon receipt of a signal indicating low RHR flow and high RHR pump discharge pressure. The valve is commanded closed upon receipt of a RHR high flow signal.
(2)  RHR pump A (B) is prevented from starting and commanded to stop unless position signals exist which indicate that the valves in the suction piping are fully open.
(3)  RHR injection valve A(B) is prevented from opening and commanded closed when reactor vessel pressure is above a setpoint.
(4)  RHR shutdown cooling suction valve A is prevented from opening unless S/P return valve A and S/P suction valve A are both fully closed.
(5)  RHR shutdown cooling suction valve B is prevented from opening unless S/P return valve B, suppressing pool suction valve B, drywell spray valve B, and wetwell spray valve B are all fully closed.
(6)  RHR shutdown cooling isolation suction valves A(B) are prevented from opening and commanded closed when reactor vessel pressure is above a setpoint.
(7)  HPCF minimum flow valve B is commanded open upon receipt of a signal indicating low HPCF flow and high HPCF pump discharge pressure. The valve is commanded closed upon receipt of a high HPCF flow signal.
(8)  HPCF pump B is prevented from starting and commanded to stop unless position signals exist which indicate that the valves in the suction piping are fully open.
Each of the two RSS divisions is powered from its respective Class 1E division. In the RSS, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The RSS panels are located in the Reactor Building remote from the MCR.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.6 provides a definition of the visual inspections, tests and/or analyses, together with associated acceptance criteria, which will be undertaken for the RSS.
Remote Shutdown System                                                                                    2.2-31
 
ABWR 2.2-32                              - RHR FLOW
                                    - HX. INLET TEMPERATURE
                                    - HX. OUTLET TEMPERATURE                          - RX. WATER LEVEL                                                          MEDIUM VOLTAGE
                                    - PUMP DISCHARGE PRESSURE                          - RX. PRESSURE                                  - CST LEVEL                  BUS VOLTAGE
                                                                  - HPCF FLOW                                          DRYWELL                                                    HPIN A, B
                                                                  - PUMP DISCHARGE                                    PRESSURE                                                    PRESSURE
                                                                                                      - S/P PRESSURE                                              - S/P LEVEL                - RCW FLOW DIVISION I & II TEMPERATURE                                                                              RUN/STOP RHR-A, B          HPCF-B              NBS-A, B      SPTM-A, B                      MUWC-A, B    RCW-A, B                                  DG-A, B AC-A, B                                  EPD INDICATION RSS PANELS 25A5675AA Revision 7
                                                                                                      - CONTROL & TRANSFER SWITCHES
                                                                                                      - VALVE POSITION INDICATION
                                                                                                      - PUMP STOP-RUN INDICATION CONTROL
                                                                                                      - PLANT PARAMETERS INDICATORS Design Control Document/Tier 1 RHR-A, B            EPD              NBS-A, B          HPCF-B          RCW-A, B          RSW-A, B                          FCS-B DIVISION I & II Remote Shutdown System NOTES:
: 1. RSS PANELS A AND B INTERFACE WITH SYSTEM IN DIVISIONS A AND B ( I AND II), RESPECTIVELY.
Figure 2.2.6 Remote Shutdown System
 
ABWR Remote Shutdown System Table 2.2.6 Remote Shutdown System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The equipment comprising the RSS is            1. Inspections of the as-built system will be    1. The as-built RSS conforms with the defined in Section 2.2.6.                          conducted.                                      description in Section 2.2.6.
: 2. Operation of transfer switches on the RSS      2. Tests will be conducted on each as-built RSS 2. Operation of transfer switches on the RSS panel overrides and isolates the controls          division by placing the transfer switches in    panel overrides and isolates the controls from the MCR and transfers control to the          the RSS position. Continuity tests will then    from the MCR and transfers control to the RSS.                                              be conducted between RSS control devices        RSS.
and interfacing equipment. Additional tests will be conducted to attempt actuation of the interfacing equipment from the MCR.
: 3. Transfer switch actuation causes alarms in      3. Tests will be conducted on each as-built RSS 3. Transfer switch actuation causes alarms in the MCR.                                          division by placing the transfer switch in the  the MCR.
25A5675AA Revision 7 RSS position.
: 4. RSS Division A has the following automatic 4.                                                  4.
controls and interlocks for RHR System Division A. RSS Division B has the following automatic controls and interlocks for RHR System Division B and HPCF System Division B:
: a. RHR minimum flow valve A(B) is                  a. Tests will be conducted on the RSS          a. RHR minimum flow valve receives an commanded open upon receipt of a                    using simulated RHR System flow and            open signal when low flow and high Design Control Document/Tier 1 signal indicating low RHR flow and high            pump discharge pressure signals.                discharge pressure signals are RHR pump discharge pressure. The                                                                    simulated. This valve receives a close valve is commanded closed upon receipt                                                              signal when a high flow signal is of a RHR high flow signal.                                                                          simulated.
: b. RHR pump A(B) is prevented from                  b. Tests will be conducted on the RSS          b. RHR pump receives a start signal when starting and commanded to stop unless              using simulated valve position signals.        simulated signals indicate a suction path position signals exist which indicate that                                                          is fully open. A stop signal is received the valves in the suction piping are fully                                                          when simulated signals indicate absence open.                                                                                              of a fully open suction path.
2.2-33
 
Table 2.2.6 Remote Shutdown System (Continued)
ABWR 2.2-34 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria
: c. RHR injection valve A(B) is prevented      c. Tests will be conducted on the RSS        c. RHR injection valve receives an open from opening and commanded closed              using simulated reactor vessel pressure        signal when a low reactor vessel when reactor vessel pressure is above a        signals                                        pressure signal is simulated. When a setpoint.                                                                                      high reactor vessel pressure signal is simulated, the open signal is removed and a close signal is received.
: d. RHR shutdown cooling suction valve A is      d. Tests will be conducted on the RSS          d. RHR shutdown cooling suction valve A prevented from opening unless S/P              using simulated valve position signals.        receives an open signal only when return valve A and S/P suction valve A                                                          simulated signals indicate that S/P are both fully closed.                                                                          suction and return valves are both fully closed.
25A5675AA Revision 7
: e. RHR shutdown cooling suction valve B is      e. Tests will be conducted on the RSS          e. RHR shutdown cooling suction valve B prevented from opening unless S/P              using simulated valve position signals.        receives an open signal only when return valve B, S/P suction valve B,                                                            simulated valve-fully-closed signals are drywell spray valve B, and wetwell spray                                                        present.
valve B are all fully closed.
: f. RHR shutdown cooling isolation suction    f. Tests will be conducted on the RSS        f. RHR shutdown cooling isolation suction valves A(B) are prevented from opening          using simulated reactor vessel pressure        valves receives an open signal only and commanded closed when reactor              signals.                                        when the simulated reactor vessel vessel pressure is above a setpoint.                                                            pressure signal is below a setpoint. The valves receive a close signal when the Design Control Document/Tier 1 simulated signal indicates reactor vessel pressure is above a setpoint.
: g. HPCF minimum flow valve B is                g. Tests will be conducted on the RSS          g. HPCF minimum flow valve receives an commanded open upon receipt of a                using simulated HPCF System flow and            open signal when low flow and high Remote Shutdown System signal indicating low HPCF flow and high        pump discharge pressure signals.                discharge pressure signals are HPCF pump discharge pressure. The                                                              simulated. This valve receives a close valve is commanded closed upon receipt                                                          signal when a high flow signal is of a high HPCF flow signal.                                                                    simulated.
 
Table 2.2.6 Remote Shutdown System (Continued)
ABWR Remote Shutdown System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                          Acceptance Criteria
: h. HPCF pump B is prevented from starting      h. Tests will be conducted on the RSS            h. HPCF pump is permitted to start when and commanded to stop unless position          using simulated valve position signals.          simulated signals indicate a suction path signals exist which indicate that the                                                            is fully open. A stop signal is received valves in the suction piping are fully                                                            when simulated signals indicate absence open.                                                                                            of a fully open suction path.
: 5. Each of two RSS divisions is powered from 5.                                                5.
its respective Class 1E division. In the RSS,
: a. Tests will be performed on the RSS by            a. The test signal exists only in the Class independence is provided between Class 1E providing a test signal in only one Class          1E division under test in the RSS.
divisions, and between Class 1E divisions 1E division at a time.
and non-Class 1E equipment.                                                                      b. In the RSS, physical separation or
: b. Inspection of the as-built Class 1E                electrical isolation exists between Class divisions in the RSS will be performed.            1E divisions. Physical separation or 25A5675AA Revision 7 electrical isolation exists between these Class 1E division and non-Class 1E equipment.
Design Control Document/Tier 1 2.2-35
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.2.7 Reactor Protection System Design Description The Reactor Protection System (RPS) is an instrumentation and control system and its purpose is to initiate reactor scram whenever RPS logic requirements for scram initiation are satisfied.
As shown in Figure 2.2.7a, the RPS interfaces with the Neutron Monitoring System (NMS),
Nuclear Boiler System (NBS), Control Rod Drive (CRD) System, Rod Control and Information System (RCIS), Recirculation Flow Control (RFC) System, Suppression Pool Temperature Monitoring System (SPTM), and the Essential Multiplexing System (EMS). Figure 2.2.7a also depicts the implementation of RPS logic within the Safety System Logic and Control (SSLC).
The RPS has four divisions. Figure 2.2.7b shows the RPS divisional aspects and the signal flow paths from sensors to scram pilot valve solenoids. Equipment within an RPS division consists of sensors (transducers or switches), multiplexers, digital trip modules (DTM), trip logic unit (TLU), output logic unit (OLU), and load drivers (LD). The LDs are only in Divisions II and III.
The RPS is classified as a Class 1E safety-related system.
The RPS consists of logic and circuitry for initiation of both automatic and manual scrams. The automatic scram function is comprised of four independent divisions of sensor instrument channels, hardware/software based logic, and two independent divisions of actuating devices.
Automatic scram is initiated whenever a scram condition is detected by two or more automatic divisions of RPS logic. For automatic scram, the sensor input signals to the RPS originate either from the RPSs own sensors or other systems sensors. For determination of the existence of an automatic scram condition, within each automatic scram channel of the RPS, the DTM of a given RPS channel compares the monitored process variable with the stored setpoint in its memory and issues a trip signal if the monitored process variable exceeds the setpoint. The DTM then sends the trip signal to the TLU of its own channel and the TLUs of the other three channels of RPS, where two-out-of-four voting is performed (see Figure 2.2.7b).
In the case of high suppression pool average temperature trip and inboard/outboard MSIV closure signals, the SPTM module of SSLC and NBS provide their divisional trip signals directly to the corresponding divisional RPS DTM. However, in the case of the NMS, the four channels of the NMS each provide their trip signals to each RPS divisional TLU. A list of conditions that can cause automatic reactor scram is provided below. The name of the system that provides the sensor input signal or the trip signal is shown in brackets.
(1)    Turbine Stop Valves Closure at above 40% power levels [RPS]
(2)    Low Turbine Control Valves Oil Pressure (Fast Closure) at above 40% power levels
[RPS]
(3)    NMS Trips [Discrete trip signals to RPS TLUs]
2.2-36                                                                              Reactor Protection System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 (4)  High Reactor Pressure [NBS]
(5)  Low Reactor Water Level [NBS]
(6)  High Drywell Pressure [NBS]
(7)  Main Steamline Isolation [NBS discrete signals to RPS DTMs]
(8)  Low Control Rod Drive Accumulator Charging Header Pressure [CRD]
(9)  High Suppression Pool Average Temperature [SPTM Module of SSLC trip signals to RPS DTMs]
The TLUs provide their trip signals to their divisional OLUs which are used to control the solid-state LDs that control the Class 1E AC power to the scram solenoids, and relays that control DC power to back-up scram valves. For automatic scram initiation, the TLU trip signals cause the LDs to interrupt Class 1E AC power to the scram solenoids (fail-safe logic), cause the back-up scram relays to supply DC power to back-up scram solenoids, and provide scram follow signals to the RCIS. Each division of RPS controls eight LDs. The LDs are arranged to switch AC power to the scram solenoids in a two-out-of-four format. That is, reactor scram will occur only if two or more divisions of the RPS provide trip signals to their associated LDs.
Manual scram function, which is separate and independent from automatic scram logic, is implemented in Divisions II and III of the RPS. For manual scram initiation, two manual scram push buttons of the RPS must be simultaneously depressed. When manual scram is initiated, the RPS, through manual scram switches, interrupts Class 1E AC power to the scram solenoids, connects divisional Class 1E DC power to back-up scram solenoids, and provides scram follow signals to RCIS. The RPS logic seals in the scram signals and permits reset of scram logic after a time delay of at least 10 seconds.
RPS initiates a reactor internal pump (RIP) trip on receipt of either a turbine stop valve closure or a low turbine control valve oil pressure signal when the reactor power is above 40% (from a turbine first stage pressure signal).
The RPS design is single-failure-proof and redundant. Also, the RPS design is fail-safe in the event of loss of electrical power to one division of RPS logic.
Each of the four RPS divisional logic and associated sensors are powered from their respective divisional Class 1E power supply. In the RPS, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
As shown on Figure 2.2.7a, the RPS has manual divisional trip switches, reactor mode switch, manual scram switches, and scram reset switches for manual controls. Divisional trip displays, and scram solenoids electrical power status lights are also provided. These RPS controls and displays are provided in the main control room. Fail safe RPS sensors are turbine control valve Reactor Protection System                                                                                  2.2-37
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 oil pressure switches, turbine stop valve position switches, and turbine first-stage pressure sensors. These sensors are located in the Turbine Building.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.7 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be performed for the RPS.
2.2-38                                                                            Reactor Protection System
 
ABWR Reactor Protection System LOCAL AREA                                          MAIN CONTROL ROOM                                                LOCAL AREA Plant Sensors                                                                                                        Device Actuators NMS  Trips      RPS              RPS              RPS            RPS Reactor    Manual Div. Trip      Manual        Manual Mode              with            Scram          Scram Switch      seal-in and reset      A              B RPS Manual Scram Reset CRD HCUs CRD  CRD HCU Accumulator                                                SSLC PROCESSING Charging Header Pressure 25A5675AA Revision 7 RPS LOGIC                                  Scram Pilot Valve Solenoid Load Drivers
                                                                                                  - Sensor Channel Trip Decision                            (Div. II, III AC Power)
Drywell Pressure                                              - System Coincidence Trip Decision Reactor Water Level                                          - Control and Interlock Logic NBS  Reactor Vessel Pressure                                      - Manual Division Trip
                                                                                                  - Division-of-Sensors Bypass                              Actuators for Scram Air MSIV Position Switches                                        - Division Maintenance Bypass                              Header Dump Valves
                                                                                                  - Calibration, Self-Diagnosis                            (Div. II, III DC Power)
Main Turbine Stop Valve Position RPS  Turbine Control Valve Oil Pressure Turbine 1st Stage Pressure                                          SPTM Suppression Pool                                      Initiate Scram-Follow RCIS (Control Rod Run-In)
Average Temperature Trip Design Control Document/Tier 1 RFC  Reactor Internal Pump Trip NOTES:
: 1. Diagram represents one of four divisions.
Figure 2.2.7a Reactor Protection System Control Interface Diagram 2.2-39
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 Division I          Division II          Division III        Division IV Sensor              Sensor              Sensor              Sensor A                    B                  C                    D Division I Raceway Division II Raceway Division III Raceway Division IV Raceway DTM                DTM                  DTM                  DTM A                    B                  C                    D TLU A              TLU B              TLU C                TLU D OLU A              OLU B              OLU C                OLU D RPS-                                            RPS-            RPS-            RPS-Group 1                                          Group 2        Group 3          Group 4 Raceways                                        Raceways        Raceways        Raceways Same            Same            Same LDs                LDs Group 1 (A)        Group 1 (B) as              as              as Group 1          Group 1          Group 1 Div. II  Div. III (Manual trip or test logic interfaces not shown)
RPS CRD Solenoid A    Solenoid B Figure 2.2.7b Reactor Protection System 2.2-40                                                                                  Reactor Protection System
 
ABWR Reactor Protection System Table 2.2.7 Reactor Protection System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The Equipment comprising the RPS is              1. Inspection of the as-built system will be      1. The as-built RPS conforms with the defined in Section 2.2.7.                          conducted.                                        description in Section 2.2.7.
: 2. RPS logic uses four independent sensor          2. Tests will be conducted using simulated input 2. The RPS LDs change their states to interrupt instrument channels of each process                signals for each process variable to cause      electrical power to scram solenoids. RPS variable described in Section 2.2.7 for its        trip conditions in two, three, and four          back-up scram relays close and RCIS relays automatic scram function.                          instrument channels of the same process          close to provide signals to RCIS.
variable of the RPS.
: 3. For manual scram initiation two manual          3. Tests will be conducted by depressing the      3. When manual scram push-button A is scram push buttons of the RPS must be              scram push button A, the B scram push-            depressed Division II AC power to A scram simultaneously depressed.                          button, and both.                                solenoids is interrupted. When scram push button B is depressed Division III AC power 25A5675AA Revision 7 to B scram solenoids is interrupted. When both A & B scram push buttons are depressed reactor scram occurs, RPS back-up scram relays close to energize the solenoids of scram air header dump valves and RCIS relays close to provide signals to the RCIS.
: 4. The RPS logic seals in the scram signal, and 4. Tests will be conducted by attempting to          4. During the 10 second time period after scram permits reset of scram logic after a time      reset RPS scram circuitry during the 10              initiation, reset does not occur.
Design Control Document/Tier 1 delay of at least 10 seconds.                  seconds time period after scram initiation.
: 5. RPS initiates an RIP trip on receipt of either  5. Test will be conducted on the as-built RPS    5. The RPS initiates an RIP trip on receipt of a turbine stop valve closure or a low turbine      using simulated turbine stop valve position,      either a simulated signals indicating turbine control valve oil pressure signal when              turbine control valve oil pressure and turbine    stop valve closure or low control valve oil reactor power is above 40% (from a turbine          first stage pressure signals.                    pressure when reactor power is above 40%.
first stage signal).
: 6. RPS design is fail-safe in the event of loss of 6. Tests will be conducted by disconnecting        6. Upon loss of electrical power to one division electrical power to one division of RPS logic. electrical power to one division of RPS logic      of RPS logic, the LDs of that division change at a time.                                        their state to interrupt electrical power to scram solenoids.
2.2-41
 
Table 2.2.7 Reactor Protection System (Continued)
ABWR 2.2-42 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 7. Each of the four RPS divisional logic and    7.                                              7.
associated sensors are powered from their
: a. Tests will be conducted on the as-built        a. The test signal exists only in the Class respective divisional Class 1E power supply.
RPS by providing a test signal to only            1E division under test in the RPS.
In the RPS, independence is provided one Class 1E division at a time.
between Class 1E divisions, and between Class 1E divisions and non-Class 1E            b. Inspection of the as-installed Class 1E        b. In the RPS physical separation or equipment.                                        divisions in the RPS will be performed.          electrical isolation exists between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and Non-Class 1E equipment.
: 8. Main control room displays and controls      8. Inspections will be performed on the main    8. Displays and controls exist or can be 25A5675AA Revision 7 provided for the RPS are as defined in          control room displays and controls for the      retrieved in the main control room as defined Section 2.2.7.                                  RPS.                                            in Section 2.2.7.
Design Control Document/Tier 1 Reactor Protection System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 2.2.8 Recirculation Flow Control System Design Description The Recirculation Flow Control (RFC) System controls reactor power by controlling the recirculation flow rate through the reactor core. This is achieved by modulating the recirculation internal pump (RIP) speeds using voltage and frequency modulation of adjustable speed drive (ASD) outputs.
The RFC System consists of redundant microprocessor-based controllers, adjustable speed drives, and motor generator (MG) sets. There are two MG sets, each of which supplies three of the ten ASDs which power the ten RIPs. The other four ASDs receive power directly from the power supply bus. No more than three RIPs are connected to any one power supply bus.
The RFC System operates in either manual or automatic control modes and has the control interfaces shown on Figure 2.2.8.
Except for the core plate differential pressure sensors provided for the Neutron Monitoring System (NMS), the RFC System is classified as non-safety-related. The four core plate differential pressure sensors for the NMS are classified as Class 1E safety-related.
The RFC System has the logic to generate the following signals to mitigate an anticipated transient without scram (ATWS) event:
(1)    A signal to open the alternate rod insertion (ARI) valves in the Control Rod Drive (CRD) System on a high reactor vessel pressure signal, a low reactor water level signal, or a manual RFC System signal.
(2)    A signal to the Rod Control and Information System (RCIS) to initiate electrical insertion of all control rods on a high reactor vessel pressure signal, a low reactor water level signal, or a manual control rod insertion signal.
(3)    A signal to trip the four RIPs not connected to MG sets on either a high reactor vessel pressure signal or a low reactor water level signal (the latter is not an ATWS mitigation feature).
(4)    A signal to trip the six RIPs connected to MG sets on a low reactor water level signal.
Three of the six RIPs are tripped after a preset time delay.
(5)    A manual RFC System signal to Safety System Logic and Control (SSLC) to initiate the Standby Liquid Control (SLC) System and to initiate Feedwater Control (FDWC)
System runback of feedwater flow.
Recirculation Flow Control System                                                                            2.2-43
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 The RFC System logic issues a signal to the RCIS for selected control rod run-in (SCRRI) to provide stability control when the following conditions occur:
(1)  Two or more RIPs are tripped, and (2)  The reactor power is at or above the preset level, and (3)  Core flow is at or below the preset level.
The RFC System has the logic to generate the following protective signals:
(1)  A signal to reduce all RIP speed on receipt of a signal from the RCIS that an all-rod insertion condition exists (which includes conditions of high reactor vessel pressure, low reactor vessel water level or manual RFC System initiation).
(2)  A signal to trip four RIPs when Reactor Protection System (RPS) provides an RIP trip signal.
When the RIP MG sets power supply breakers open, the MG sets are capable of holding the connected RIPs at their original speeds for at least one second and, after 1 second, assure the speed is at or above a speed coastdown curve defined by a rate of speed decrease of 10% per second for an additional two seconds.
Each channel of the RFC System controller is powered by separate non-Class 1E uninterruptible power supplies. Each of the four safety-related RFC System core plate differential pressure sensors is powered from its respective divisional Class 1E power supply.
In the RFC System, independence is provided between the Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The RFC System digital controllers are located in the Control Building. The ASDs and core plate differential pressure sensors are located in the Reactor Building.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.8 provides a definition of the inspections, tests, and/or analyses, together with the associated acceptance criteria, which will be undertaken for the RFC System.
2.2-44                                                                      Recirculation Flow Control System
 
ABWR Recirculation Flow Control System MANUAL CONTROLS FOR ATWS MITIGATION (ARI,SLC,RCIS,FDWC)
AND RIP/ASD CONTROL PLANT SENSORS                                                              RCIS    RUNBACK RPS        RIP TRIP SPEED DEMAND      ASD RIP AND PUMP TRIP FDWC    LOW REACTOR                                              SIGNALS      (TYPICAL OF 10)    (TYPICAL OF 10)
SYSTEM  WATER LEVEL 25A5675AA Revision 7 RFC SYSTEM SB&PC    REACTOR VESSEL                    CONTROLLER                              ARI SIGNAL SYSTEM  PRESSURE                                                                  TO CRD SYSTEM SLC INITIATION LOW REACTOR                                                              FEEDWATER NBS                                                                                RUNBACK WATER LEVEL (IN SSLC)
CONTROL ROD NEUTRON FLUX/                                                            INSERTION NMS      POWER                                                                      SIGNAL TO RCIS Design Control Document/Tier 1 PLANT INPUT SIGNALS CORE PLATE DIFFERENTIAL                                                                NMS PRESSURE (NOTE 2)
NOTE:
: 1. INTERCONNECTIONS MAY BE FIBER-OPTIC OR METALLIC.
: 2. ONE SENSOR ASSIGNED TO EACH OF FOUR CLASS 1E DIVISIONS.
2.2-45                                                  Figure 2.2.8 Recirculation Flow System Control Interface Diagram
 
ABWR 2.2-46 Table 2.2.8 Recirculation Flow Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Test, Analyses                            Acceptance Criteria
: 1. The equipment comprising the RFC System          1. Inspections of the as-built system will be      1. The as-built RFC System conforms with the is defined in Section 2.2.8.                        conducted.                                        description in Section 2.2.8.
: 2. RFC System consists of redundant                2. Tests will be conducted by simulating failure  2. There is no loss of RFC System output upon microprocessor based controllers.                  of each operating RFC System controller.          loss of any one controller.
: 3. The RFC System has the following logic to        3. Tests will be conducted on the as-built RFC    3. The RFC System logic issues the following mitigate an ATWS event:                            System using simulated reactor vessel              signals to mitigate an ATWS event:
: a. A signal to open the ARI valves of the          pressure, reactor water level, and RFC            a. A signal to open the ARI valves of the CRD System on a high reactor vessel              System manual signals.
CRD System upon receipt of a simulated pressure signal, a low reactor water level                                                            high reactor vessel pressure signal, a signal, or a manual RFC System signal.                                                                simulated low reactor water level signal, 25A5675AA Revision 7
: b. A signal to the RCIS to initiate electrical                                                            or a simulated manual RFC System insertion of all control rods on a high                                                                signal.
reactor vessel pressure signal, a low                                                              b. A signal to the RCIS to initiate electrical reactor water level signal, or a manual                                                                insertion of all control rods upon receipt control rod insertion signal.                                                                          of a simulated high reactor vessel pressure signal, a simulated low reactor water level signal, or a simulated manual control rod insertion signal.
Design Control Document/Tier 1 Recirculation Flow Control System
 
Table 2.2.8 Recirculation Flow Control System (Continued)
ABWR Recirculation Flow Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Test, Analyses                              Acceptance Criteria
: 3. (continued)                                      3. (continued)                                  3. (continued)
: c. A signal to trip the four RIPs not                                                              c. A signal to trip the four RIPs not connected to MG sets on either a high                                                                connected to MG sets upon receipt of reactor vessel pressure signal, or a low                                                            either a simulated high reactor vessel reactor water level signal.                                                                          pressure signal, or a simulated low
: d. A signal to trip the six RIPs connected to                                                            reactor water level signal.
MG sets on a low reactor water level                                                              d. A signal to trip the six RIPs connected to signal. Three of the six RIPs are tripped                                                            MG sets upon receipt of a simulated low ater a preset time delay.                                                                            reactor water level signal. Three of the six RIPs trip after a preset time delay.
: e. A manual RFC System signal to SSLC to 25A5675AA Revision 7 initiate the SLC System and to initiate                                                          e. A signal to initiate the SLC System and FDWC System runback of feedwater                                                                    to initiate FDWC System runback of flow.                                                                                                feedwater flow upon receipt of a simulated manual RFC System signal to SSLC.
: 4. The RFC System logic issues a signal to the 4. Tests will be conducted on the as-built RFC        4. The RFC System logic issues signal to the RCIS for SCRRI to provide stability control    System using simulated two RIPs tripped,              RCIS for SCRRI upon receipt of simulated when the following conditions occur:          reactor power, and core flow signals.                signals for:
: a. Two or more RIPs are tripped, and                                                                a. Two or more RIPs are tripped, and Design Control Document/Tier 1
: b. The reactor power is at or above a                                                                b. The reactor power is at or above a preset level, and                                                                                    preset level, and
: c. Core flow is at or below a preset level.                                                        c. Core flow is at or below a preset level.
: 5. The RFC System logic generates the              5. Tests will be conducted on the as-built RFC  5. The RFC System logic issues the following following protective signals:                      System using simulated reactor water level,      protective signals:
: a. A signal to reduce all RIP speed on            all-rod insertion and trip signals
: a. A signal to reduce all RIP speed upon receipt of a signal from the RCIS that an                                                            receipt of a simulated all-rod insertion all-rod insertion condition exists.                                                                  signal.
: b. A signal to trip four RIPs when RPS                                                              b. A signal to trip four RIPs on receipt of a 2.2-47 provides an RIP trip signal.                                                                        simulated trip signal.
 
Table 2.2.8 Recirculation Flow Control System (Continued)
ABWR 2.2-48 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Test, Analyses                            Acceptance Criteria
: 6. After the power supply breakers open to the    6. Tests will be conducted at a test facility with 6. After the power supply breakers open to the MG set, the MG sets are capable of holding        an M/G set and three associated ASDs using        MG set, the ASD output frequency remains the connected RIPs at their original speeds      simulated full load characteristics of the RIPs    within 1% of the original output frequency for for at least one second and, after 1 second,      and disconnecting power to M/G sets while          at least one second, and then for an assure the speed is at or above a speed          operating at full speeds, or analyses will be      additional two seconds the ASD output coastdown curve defined by a rate of speed        performed to demonstrate applicability of          frequency shall be equal to or greater than a decrease of 10% per second for an                prior tests and test results to the as-built      curve defined by a rate of frequency additional two seconds.                          RFC System MG sets and ASDs.                      decrease of 10% per second.
: 7. Each channel of the RFC System digital      7. Tests will be performed by providing a test      7. The test signals exist in only one digital controller is powered by separate non-Class    signal in only one uninterruptible power            control channel at a time.
1E uninterruptible power supplies.            supply at a time.
25A5675AA Revision 7
: 8. Each of the four RFC System core plate        8.                                              8.
differential pressure sensors is powered from
: a. Tests will be performed on the RFC          a. The test signal exists only in the Class its respective divisional Class 1E power System by providing a test signal in only      1E division under test in the RFC supply. In the RFC System, independence is one Class 1E division at a time.                System.
provided between Class 1E divisions, and between Class 1E divisions and non-Class        b. Inspection of the as-built Class 1E          b. In the RFC System, physical separation 1E equipment.                                      divisions in the RFC System will be            or electrical isolation exists between performed.                                      Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class Design Control Document/Tier 1 1E equipment.
Recirculation Flow Control System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.2.9 Automatic Power Regulator System Design Description The Automatic Power Regulator (APR) System controls reactor power during reactor startup, power generation, and reactor shutdown by commands, either directly or indirectly, to change rod positions, or to change reactor recirculation flow or load setpoint. The APR System consists of redundant digital controllers and has the interfaces shown in the control interface diagram on Figure 2.2.9.
The APR System is classified as non-safety-related.
The APR System operates in either manual or automatic control mode. The system control logic is performed by redundant, digital controllers. The digital controller receives inputs from interfacing system via the non-essential multiplexing system (NEMS). It performs power control calculations and provides system outputs to the NEMS.
The APR System digital controllers are located in the Control Building.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.9 provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria, which will be undertaken for the APR System.
Automatic Power Regulator System                                                                          2.2-49
 
ABWR 2.2-50 MANUAL CONTROL TURBINE CONTROL SYSTEM 25A5675AA Revision 7 ROD CONTROL AND PLANT APR                  INFORMATION INPUT SIGNALS                            SYSTEM                    SYSTEM RECIRCULATION FLOW CONTROL SYSTEM Design Control Document/Tier 1 Automatic Power Regulator System NOTE:
: 1. INTERCONNECTIONS MAY BE FIBER-OPTIC OR METALLIC.
Figure 2.2.9 Automatic Power Regulator System Control Interface Diagram
 
ABWR Automatic Power Regulator System Table 2.2.9 Automatic Power Regulator System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The equipment comprising the APR System      1. Inspections of the as-built system will be      1. The as-built APR System conforms with is defined in Section 2.2.9.                    conducted.                                        description in Section 2.2.9.
: 2. The system control logic is performed by      2. Tests will be performed by simulating failure  2. There is no loss of APR System output upon redundant digital controllers.                  of each operating APR System digital              loss of any one digital controller.
controller.
25A5675AA Revision 7 Design Control Document/Tier 1 2.2-51
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.2.10 Steam Bypass and Pressure Control System Design Description The Steam Bypass and Pressure Control (SB&PC) System controls the reactor pressure during reactor startup, power generation, and reactor shutdown by control of the turbine bypass valves and signals to the Turbine Control System which controls the turbine control valves. The SB&PC System consists of redundant digital controllers and has the interfaces shown in the control interface diagram on Figure 2.2.10.
The SB&PC System is classified as non-safety-related.
The SB&PC System operates in either manual or automatic control modes. The system control calculations and logic are performed by redundant digital controllers.
The SB&PC System digital controllers are located in the Control Building.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.10 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the SB&PC System.
2.2-52                                                                Steam Bypass and Pressure Control System
 
ABWR Steam Bypass and Pressure Control System MANUAL CONTROL TURBINE CONTROL 25A5675AA Revision 7 SYSTEM PLANT INPUT                        SB&PC                  TURBINE SIGNALS                          SYSTEM                  BYPASS SYSTEM RECIRCULATION FLOW CONTROL Design Control Document/Tier 1 SYSTEM NOTE:
: 1. INTERCONNECTIONS MAY BE FIBER-OPTIC OR METALLIC.
Figure 2.2.10 Steam Bypass and Pressure Control System Control Interface Diagram 2.2-53
 
ABWR 2.2-54 Table 2.2.10 Steam Bypass and Pressure Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The equipment comprising the SB&PC          1. Inspections of the as-built system will be      1. The as-built SB&PC System conforms with System is defined in Section 2.2.10.            conducted.                                        the description in Section 2.2.10.
: 2. The SB&PC System consists of redundant      2. Tests will be performed by simulating failure  2. There is no loss of SB&PC System output digital controllers.                            of each operating SB&PC System digital            upon loss of any one digital controller.
controller.
: 3. The SB&PC System controls the reactor      3. A test will be conducted by simulating an        3. Signals to decrease the reactor pressure pressure during reactor startup, power        increasing reactor pressure signal.                occur for the turbine bypass valves and the generation, and reactor shutdown by control                                                        Turbine Control System.
of the turbine bypass valves and signals to the Turbine Control System which controls the turbine control valves.
25A5675AA Revision 7 Steam Bypass and Pressure Control System                                                                                                                                                      Design Control Document/Tier 1
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.2.11 Process Computer System Design Description The Process Computer System (PCS) consists of redundant digital central processing units and associated peripheral equipment and is classified as a non-safety-related system.
The PCS performs local power range monitor (LPRM) calibrations and calculations of fuel operating thermal limits data which it provides to the automated thermal limit monitor (ATLM) function of the Rod Control & Information System (RCIS) for the purpose of updating rod block setpoints.
The PCS functions also as a top-level controller which monitors the overall plant conditions, issues control commands and adjusts setpoints of lower level controllers to support automation of the normal plant startup, shutdown and power range operations. In the event that abnormal conditions develop in the plant during operations in the automatic mode, the PCS automatically reverts to the manual mode of operation.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.2.11 provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria, which will be undertaken for the PCS.
Process Computer System                                                                                2.2-55
 
ABWR 2.2-56 Table 2.2.11 Process Computer System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The equipment comprising the PCS is            1. Inspections of the as-built system will be    1. The as-built PCS conforms with the defined in Section 2.2.11.                        conducted.                                      description in Section 2.2.11.
: 2. The PCS provides LPRM calibration and fuel 2. Tests of the as-built PCS will be conducted        2. LPRM calibration and fuel thermal limits data operating thermal limits data to the ATLM    using simulated plant input signals.                  are received by the ATLM function of the function of the RCIS.                                                                              RCIS.
: 3. In the event that abnormal conditions          3. Tests of the as-built PCS will be conducted  3. Upon receipt of the abnormal plant input develop in the plant during operations in the      using simulated abnormal plant input            signals, the PCS automatically reverts to the automatic mode, the PCS automatically              signals, while the PCS is in the automatic      manual operating mode.
reverts to the manual operating mode.              operating mode.
25A5675AA Revision 7 Design Control Document/Tier 1 Process Computer System
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.2.12 Refueling Platform Control Computer No entry for this system.
Refueling Platform Control Computer                                                    2.2-57
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.2.13 CRD Removal Machine Control Computer No entry for this system.
2.2-58                                                  CRD Removal Machine Control Computer
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 2.3.1 Process Radiation Monitoring System Design Description The Process Radiation Monitoring (PRM) System measures and displays radioactivity levels in process and effluent gaseous and liquid streams, initiates protective actions, and activates alarms in the main control room (MCR)on high radiation signals. The PRM System provides radiological monitoring during plant operation and following an accident. PRM System equipment consists of radiation sensors, radiation process monitors, and effluent samplers. The PRM System consists of independent subsystems each of which contains between one and four monitoring channels. Figure 2.3.1 shows the PRM System control interfaces. As shown on Figure 2.3.1, the PRM System safety-related channel trip signals are provided as inputs to the Safety System Logic and Control (SSLC) for generation of protective action signals.
Portions of the PRM System are classified as Class 1E safety-related (items 1 through 5 below);
the remainder are classified as non-safety-related.
The PRM System provides the following monitoring functions:
(1)    Main Steam Line (MSL) Tunnel Area (4 channels)
The MSL tunnel area is monitored for gamma radioactivity in the steam flow to the turbine. Protective action signals are automatically initiated when any two out of four channels trip.
(2)    Reactor Building Heating, Ventilating and Air Condition (HVAC) Exhaust (4 channels)
The air vent exhaust from the secondary containment is monitored for gamma radioactivity. Protective action signals are automatically initiated when any two out of four channels trip.
(3)    Fuel Handling Area Ventilation Exhaust (4 channels)
The air vent exhaust from the fuel handling area is monitored for gamma radioactivity. Protective action signals are automatically initiated when any two out of four channels trip.
(4)    Control Building Intake Air Supply (4 channels per intake)
The air supply intake to the Control Building is monitored for gamma radioactivity.
Protective action signals are automatically initiated when any two out of four channels trip.
Process Radiation Monitoring System                                                                            2.3-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 (5)    Drywell Sump Liquid Discharge (1 channel per sump)
The liquid waste discharged from each of the drywell LCW and HCW sumps to the Radwaste Building is monitored for gamma radioactivity. A protective action signal is automatically initiated when a channel trips.
(6)    Off-Gas Post-Treatment Discharge (2 channels)
The off-gas discharge from the charcoal vault to the stack is sampled and monitored for airborne radioactivity. Protective action signals are automatically initiated when both channels trip.
(7)    Plant Stack Discharge (2 channels)
The ventilation and the gaseous discharge from the plant stack is sampled and monitored for airborne radioactivity. An alarm is initiated when the detected radiation level exceeds the trip setpoint.
(8)    Radwaste Liquid Discharge (1 channel)
The radwaste liquid discharged from the plant is sampled and monitored for gamma radioactivity. A protective action signal is automatically initiated when the channel trips.
(9)    Intersystem Radiation Leakage (3 channels)
Reactor coolant leakage into the Reactor Building Cooling Water (RCW) System is monitored for gamma radioactivity. One channel is provided for each RCW System division. An alarm is initiated when the detected radiation level exceeds the trip setpoint.
(10) Turbine Gland Seal Condenser Exhaust (1 channel)
The exhaust discharged from the turbine gland seal condenser is monitored for gamma radioactivity. An alarm is initiated when the detected radiation level exceeds the trip setpoint.
Each safety-related PRM System radiation monitoring channel is powered from its respective divisional Class 1E power source. In the PRM System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The PRM System radiation sensors and the effluent samplers are installed locally in the plant, while the radiation process monitors are located in the Control Building.
2.3-2                                                                    Process Radiation Monitoring System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 The PRM System has the following alarms and displays in the MCR:
(1)    Displays of radiation levels.
(2)    Channel trip status.
(3)    Plant stack discharge, intersystem leakage, and turbine gland seal condenser exhaust radiation alarms.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.3.1 provides a definition of the inspections, tests and/or analyses, together with the associated acceptance criteria, which will be undertaken for the Process Radiation Monitoring System.
Process Radiation Monitoring System                                                                        2.3-3
 
ABWR 2.3-4 LOCAL AREA                                  MAIN CONTROL ROOM                            LOCAL AREA Plant Sensors                                                                            Device Actuators PRM                SSLC PROCESSING NOTE 2
                                                                          - Channel MSL Tunnel Area Radiation      Trip Decisions PRM SYSTEM LOGIC Control Building Air Intake Supply Radiation                                                                    CRHA Protective Action Signal
                                                                                            - System Coincidence Trip Decision
                                                                                            - Division-of-Sensors Bypass        HVAC Reactor Building HVAC Exhaust Radiation PRM 25A5675AA Revision 7 Fuel Handling Area Ventilation Exhaust Radiation Drywell Sump LCW Discharge Radiation LDS  Control Logic            LDS  Protective Action Signal Drywell Sump HCW Discharge Radiation Design Control Document/Tier 1 Process Radiation Monitoring System Notes:
: 1. Diagram represents one of four PRM System divisions.
: 2. See Section 3.4, Figure 3.4b for SSLC processing.
Figure 2.3.1 Process Radiation Monitoring System Control Interface Diagram
 
ABWR Process Radiation Monitoring System Table 2.3.1 Process Radiation Monitoring System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The equipment comprising the PRM System 1. Inspection of the as-built system will be          1. The as-built PRM System conforms with the is defined in Section 2.3.1.              conducted.                                            description in Section 2.3.1.
: 2. The MSL tunnel area is monitored for        2. Tests will be conducted using simulated        2. Protective action signals are automatically gamma radioactivity in the steam flow to the    signals to cause trip conditions.                initiated when any two out of four channels turbine. Protective action signals are                                                            trip.
automatically initiated when any two out of four channels trip.
: 3. The air vent exhaust from the secondary        3. Tests will be conducted using simulated    3. Protective action signals are automatically containment is monitored for gamma                signals to cause trip conditions.              initiated when any two out of four channels radioactivity. Protective action signals are                                                      trip.
automatically initiated when any two out of 25A5675AA Revision 7 four channels trip.
: 4. The air vent exhaust from the fuel handling    4. Tests will be conducted using simulated    4. Protective action signals are automatically area is monitored for gamma radioactivity.        signals to cause trip conditions.              initiated when any two out of four channels Protective action signals are automatically                                                      trip.
initiated when any two out of four channels trip.
: 5. The air supply intake to the Control Building  5. Tests will be conducted using simulated    5. Protective action signals are automatically is monitored for gamma radioactivity.              signals to cause trip conditions.              initiated when any two out of four channels Protective action signals are automatically                                                      trip.
Design Control Document/Tier 1 initiated when any two out of four channels trip.
: 6. The liquid waste discharged from each of the 6. Tests will be conducted on each drywell      6. A protective action signal is automatically LCW and HCW drywell sumps to the                sump using a simulated signal to cause a trip    initiated when a channel trips.
Radwaste Building is monitored for gamma        condition.
radioactivity. A protective action signal is automatically initiated when a channel trips.
2.3-5
 
Table 2.3.1 Process Radiation Monitoring System (Continued)
ABWR 2.3-6 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 7. The off-gas discharge from the charcoal      7. Tests will be conducted using simulated    7. Protective action signals are generated vault to the stack is sampled and monitored      signals to cause trip conditions.              when both channels trip.
for airborne radioactivity. Protective action signals are automatically initiated when both channels trip.
: 8. The ventilation and the gaseous discharge      8. Tests will be conducted using simulated    8. An alarm is initiated when the detected from the plant stack is sampled and              signals to cause trip conditions.            radiation level exceeds the trip setpoint.
monitored for airborne radioactivity. An alarm is initiated when the detected radiation level exceeds the trip setpoint.
: 9. The radwaste liquid discharged from the        9. Tests will be conducted using simulated    9. A protective action signal is automatically 25A5675AA Revision 7 plant is sampled and monitored for gamma          signals to cause trip conditions.            initiated when the channel trips.
radioactivity. A protective action signal is automatically initiated when the channel trips.
: 10. Reactor coolant leakage into the Reactor      10. Tests will be conducted using simulated  10. An alarm is initiated when the detected Building Cooling Water (RCW) System is            signals to cause trip conditions.            radiation level exceeds the trip setpoint.
monitored for gamma radioactivity. One channel is provided for each RCW division.
An alarm is initiated when the detected Design Control Document/Tier 1 radiation level exceeds the trip setpoint.
: 11. The exhaust discharged from the turbine      11. Tests will be conducted using simulated  11. An alarm is initiated when the detected Process Radiation Monitoring System gland seal condenser is monitored for            signals to cause trip conditions.            radiation level exceeds the trip setpoint.
gamma radioactivity. An alarm is initiated when the detected radiation level exceeds the trip setpoint.
 
Table 2.3.1 Process Radiation Monitoring System (Continued)
ABWR Process Radiation Monitoring System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                          Acceptance Criteria
: 12. Each safety-related PRM System radiation    12.                                              12.
monitoring channel is powered from its
: a. Tests will be performed on the PRM            a. The test signal exists only in the Class respective divisional Class 1E power source.
System by providing a test signal to only        1E division under test in the PRM In the PRM System, independence is one Class 1E division at a time.                System.
provided between Class 1E divisions, and between Class 1E divisions and non-Class        b. Inspection of the as-built Class 1E          b. In the PRM System, physical separation 1E equipment.                                      divisions in the PRM System will be              or electrical isolation exists between performed.                                      Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 13. Main control room alarms and displays      13. Inspection will be performed on the main      13. Alarms and displays exist or can be retrieved 25A5675AA Revision 7 provided for the PRM System are as defined    control room PRM System alarms and                in the main control room as defined in in Section 2.3.1.                              displays.                                        Section 2.3.1.
Design Control Document/Tier 1 2.3-7
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.3.2 Area Radiation Monitoring System Design Description The Area Radiation Monitoring (ARM) System measures the gamma radiation levels at assigned locations within the plant, displays the measurements in the main control room, and activates alarms when the detected radiation levels exceed preset limits.
The ARM System is a multiple channel instrumentation system consisting of radiation monitors, their associated detectors, and local audible alarms. Each ARM channel monitors the radiation level in its assigned area, and initiates a main control room (MCR) alarm and a local alarm (if provided) when the radiation level exceeds a preset limit.
The ARM System is classified as non-safety-related.
The ARM System radiation sensors and the audible warning alarms are installed locally in the plant, while the radiation monitors are located in the Control Building.
The ARM System has the following alarms and displays in the MCR:
(1)    Displays of radiation levels.
(2)    Channel trip status.
(3)    Alarms.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.3.2 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Area Radiation Monitoring System.
2.3-8                                                                          Area Radiation Monitoring System
 
ABWR Area Radiation Monitoring System Table 2.3.2 Area Radiation Monitoring System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                      Acceptance Criteria
: 1. The equipment comprising the ARM System 1. Inspection of the as-built system will be      1. The as-built ARM System conforms with the is defined in Section 2.3.2.              conducted.                                        description in Section 2.3.2.
: 2. Each ARM channel monitors radiation level 2. Tests will be conducted using simulated      2. The MCR alarm and local audible alarm (if in its assigned area, and initiates a MCR    signals for each channel.                      provided) are initiated when the simulated alarm and a local audible alarm (if provided)                                                radiation level exceeds a preset limit.
when the radiation level exceeds a preset limit.
: 3. MCR alarms and displays provided for the    3. Inspections will be performed on the MCR  3. Alarms and displays exist or can be retrieved ARM System are as defined in Section 2.3.2. alarms and displays for the ARM System.      in the MCR as defined in Section 2.3.2.
25A5675AA Revision 7 Design Control Document/Tier 1 2.3-9
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.3.3 Containment Atmospheric Monitoring System Design Description The Containment Atmospheric Monitoring System (CAMS) is used for post-accident monitoring of the primary containment. The purpose of the CAMS is to:
(1)  Provide information on combustible levels of oxygen and hydrogen in the primary containment.
(2)  Detect and measure the radiation level within the primary containment during and following an accident.
(3)  Detect and measure the hydrogen concentration within the primary containment during and following an accident.
The system monitors the atmospheric conditions in the drywell and in the suppression chamber for radiation levels and for hydrogen and oxygen gas concentration levels, displays the measurements in the main control room (MCR), and activates alarms in the MCR upon detection of high levels of radiation and/or gas concentrations.
The CAMS consists of two independent divisions and each division is composed of two radiation channels and oxygen/hydrogen gas monitoring equipment.
The CAMS is classified as a Class 1E safety-related system.
Operation of each CAMS division can be activated manually or automatically during a post-accident condition by a signal indicating a high drywell pressure or a low reactor water level.
One radiation channel of each CAMS division monitors the radiation level in the drywell and the other channel monitors the radiation level in the suppression chamber.
The oxygen/hydrogen monitoring equipment of each CAMS division analyzes the hydrogen and oxygen gas concentration levels in the drywell or in the suppression chamber and provides separate gas concentration displays in the MCR.
Each CAMS division is powered from its respective divisional Class 1E power source. In the CAMS, independence is provided between the Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Both CAMS divisions are located in the Reactor Building, except for the radiation and the gas process monitors, which are located in the Control Building.
The CAMS has the following alarms, displays, and controls in the MCR:
(1)  Displays of radiation, hydrogen and oxygen levels.
2.3-10                                                              Containment Atmospheric Monitoring System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 (2)    Alarms for radiation levels, and for hydrogen and oxygen gas concentration levels.
(3)    Manual system level initiation for each CAMS division.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.3.3 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Containment Atmospheric Monitoring System.
Containment Atmospheric Monitoring System                                                                2.3-11
 
ABWR 2.3-12 Table 2.3.3 Containment Atmospheric Monitoring System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The equipment comprising the CAMS is        1. Inspection of the as-built system will be    1. The as-built CAMS conforms with the defined in Section 2.3.3.                      conducted.                                      description in Section 2.3.3.
: 2. Operation of each CAMS division can be      2. Tests of each division of the as-built CAMS 2. Each CAMS division is activated upon activated manually by the operator or          will be conducted using manual controls and    receipt of the test signals.
automatically.                                  simulated automatic initiation signals.
: 3. Each CAMS division is powered from its      3.                                              3.
respective divisional Class 1E power source.
: a. Tests will be performed on the CAMS by      a. The test signal exists only in the Class In the CAMS, independence is provided providing a test signal to only one Class      1E division under test in the CAMS.
between Class 1E divisions, and between 1E division at a time.
Class 1E divisions and non-Class 1E equipment.                                      b. Inspection of the as-built Class 1E          b. In the CAMS, physical separation or 25A5675AA Revision 7 divisions in the CAMs will be performed.        electrical isolation exists between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 4. Main control room alarms, displays and      4. Inspections will be performed on the main    4. Alarms, displays and controls exist or can be controls provided for the CAMS are as          control room alarms, displays and controls      retrieved in the main control room as defined defined in Section 2.3.3.                      for the CAMS.                                  in Section 2.3.3.
Containment Atmospheric Monitoring System                                                                                                                                                      Design Control Document/Tier 1
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.4.1 Residual Heat Removal System Design Description The Residual Heat Removal (RHR) System has three separate divisions. The major functions of the RHR System are:
(1)  Containment heat removal.
(2)  Reactor decay heat removal.
(3)  Emergency reactor vessel level makeup and (4)  Augmented fuel pool cooling.
Figures 2.4.1a, 2.4.1b, and 2.4.1c show the basic system configuration and scope. Figure 2.4.1d shows the RHR System control interfaces.
Except for the non-ASME Code components of the alternating current (AC) power source independent water addition feature (Figures 2.4.1b and 2.4.1c), the entire RHR System shown on Figures 2.4.1a, 2.4.1b, and 2.4.1c is classified as safety-related.
The RHR System operates in the following modes:
(1)  Low pressure core flooder (LPFL) (Divisions A, B, and C)
(2)  Suppression pool cooling (Divisions A, B, and C)
(3)  Wetwell spray (Divisions B, and C)
(4)  Drywell spray (Divisions B, and C)
(5)  Shutdown cooling (Divisions A, B, and C)
(6)  Augmented fuel pool cooling, and fuel pool makeup (Divisions B, and C)
(7)  AC power source independent water addition (Divisions B and C)
(8)  Full flow test (Divisions A, B, and C)
(9)  Minimum flow bypass (Divisions A, B, and C)
Low Pressure Core Flooder Mode As shown on Figure 2.4.1d, the RHR System channel measurements are provided to the Safety System Logic and Control (SSLC) for signal processing, setpoint comparisons, and generating trip signals. The RHR System is automatically initiated when either a high drywell pressure or low reactor water level condition exists (i.e., LOCA signal). A RHR initiation signal is provided Residual Heat Removal System                                                                                2.4-1
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 to the systems as identified on Figure 2.4.1d. The SSLC processors use a two-out-of-four voting logic for RHR System initiation. Each RHR division can also be initiated manually (LPFL mode).
Following receipt of an initiation signal, the RHR System automatically initiates and operates in the LPFL mode to provide emergency makeup to the reactor vessel. The initiation signal starts the pumps, which run in the minimum flow mode until the reactor depressurizes to less than the pumps developed head pressure. A low reactor pressure permissive signal occurs above the pumps developed head pressure, which signals the injection valve to open. As the injection valve opens, the reactor pressure is contained by the testable check valve until the reactor pressure becomes less than the pumps developed head pressure of the minimum flow mode, at which time injection flow begins. This sequence satisfies the response requirements for all potential LOCA pipe breaks when the injection valve opens within 36 seconds after receiving the low reactor pressure permissive signal. The LPFL injection flow for each division begins when the reactor vessel pressure is no less than 1.55 MPa above the drywell pressure.
When the reactor vessel pressure is no less than 0.275 MPa greater than the drywell pressure, the LPFL injection flow for each division is 954 m3/h minimum. The LPFL mode is accomplished by all three divisions of the RHR System by transferring water from the suppression pool to the reactor pressure vessel (RPV), via the RHR heat exchangers. The system automatically aligns to the LPFL mode of operation from the test mode, the suppression pool cooling, or wetwell spray modes upon receipt of an initiation signal. The wetwell spray mode is applicable for Divisions B or C. If a drywell spray valve is open in Division B or C, that RHR division automatically aligns to the LPFL mode in response to the injection valve beginning to open. The RPV injection valve in each division requires a low reactor pressure permissive signal to open, and closes automatically on receipt of a high reactor vessel pressure signal.
Suppression Pool Cooling Mode The suppression pool cooling mode of the RHR System limits the long-term post-LOCA temperature of the suppression pool, and limits the long-term peak temperatures and pressures within the wetwell and drywell regions of the containment. In this mode, the RHR System circulates water through the RHR heat exchangers and returns it directly to the suppression pool. This mode is manually initiated by control of individual system components. In the suppression pool cooling mode, the total heat removal capacity between the RHR and ultimate heat sink is no less than 0.371 MJ/s°C for each division. 0.371 MJ/s°C is the limiting heat removal capacity of all the RHR modes. The heat removal path is the RHR heat exchanger, the Reactor Building Cooling Water (RCW) System, and the Reactor Service Water (RSW)
System. In the suppression pool cooling mode, the RHR tube side heat exchanger (Hx) flow rate is 954 m3/h minimum per division. The RHR pumps have sufficient net positive suction head (NPSH) available at the pump. Suction from the suppression pool is the limiting NPSH condition of all the RHR modes.
2.4-2                                                                        Residual Heat Removal System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Containment Spray Mode The containment spray mode of the RHR System is available in Divisions B and C, and consists of the wetwell spray and drywell spray operating together. In this mode, the RHR System pumps suppression pool water to a single wetwell spray header and single drywell spray header through the associated RHR heat exchanger. The containment spray mode of the RHR System is initiated manually by control of individual system components. The drywell spray inlet valves can only be opened if a high drywell pressure condition exists and if the injection valves are fully closed. The wetwell spray flow rate for either Division B or C is no less than 114 m3/h.
Shutdown Cooling Mode In the shutdown cooling mode of operation, the RHR System removes decay heat from the reactor core, and is used to achieve and maintain a cold shutdown condition by removing decay and sensible heat from the core and reactor vessel. This mode reduces reactor pressure and temperature to cold shutdown conditions. In this mode, each division takes suction from the RPV via its dedicated suction line, pumps the water through its respective heat exchanger tubes, and returns the cooled water to the RPV. Two divisions (B and C) discharge water back to the RPV via dedicated spargers, while the third division (A) utilizes the vessel spargers of one of the two feedwater lines (FW-A). Shutdown cooling is initiated manually once the RPV has been depressurized below the system low pressure permissive. In any division, the shutdown cooling suction valve cannot be opened unless the following valves in that division are closed:
(1)    Suppression pool suction valve (2)    Suppression pool return valve (3)    Drywell spray valves (4)    Wetwell spray valve Each shutdown cooling suction valve automatically closes on low reactor water level. The low pressure portions of the shutdown cooling piping are protected from high reactor pressure by automatic closure of the shutdown cooling suction valves on a high reactor vessel pressure. The shutdown cooling flow rate for any division is no less than 954 m3/h.
Augmented Fuel Pool Cooling and Fuel Pool Makeup The augmented fuel pool cooling mode of the RHR System (Divisions B and C) can supplement the Fuel Pool Cooling (FPC) System as follows: (1) directly cooling the fuel pool by circulation fuel pool water through the RHR heat exchanger and returning it to the fuel pool; and (2) while providing shutdown cooling during refueling operations, return the cooled RHR shutdown cooling flow to the fuel pool. Also, this mode provides for fuel pool emergency makeup capability by permitting the RHR pumps (Divisions B and C) to transfer suppression pool water to the fuel pool. This mode is accomplished manually by control of individual system components. In the augmented fuel pool cooling mode, the RHR tube side heat exchanger flow rate for Division B or C is no less than 350 m3/h.
Residual Heat Removal System                                                                                2.4-3
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 AC Independent Water Addition Mode Divisions B and C of the RHR System will also function in an AC independent water addition mode. This mode provides a means of injecting emergency makeup water to the reactor by cross connecting the Reactor Building Fire Protection (FP) System header, or alternately utilizing additional sources of water from external connections just outside the Reactor Building. This makes the mode independent of the normal safety-related AC power distribution network. This mode is accomplished by manually opening two in-series valves on the cross-connection piping just upstream of the tie-in to the normal RHR piping. This is accomplished by local manual action at the valves. Fire Protection System water can be directed to either the RPV, the wetwell or drywell spray sparger, or the spent fuel pool by local manual opening of the Division B or C RHR injection valve, the Division B or C wetwell spray valve, the two Division B or C drywell spray valves, or the two Division B or C valves to the Fuel Pool Cooling and Clean Up System (FPC), respectively. Local manual as used in this paragraph means manually operating the valves at the valves.
Full Flow Test Mode Each division of the RHR System has a full flow test mode to permit pump flow testing during plant operation. In this mode, the system is essentially operated in the suppression pool cooling mode, drawing suction from and discharging back to the suppression pool.
Minimum Flow Bypass Mode Each division of the RHR System has a minimum flow bypass mode that assures there is always flow in the RHR pumps when they are operating. This is accomplished by monitoring pump discharge flow, and opening a minimum flow valve to the suppression pool when flow falls below the minimum value. The minimum flow valve closes when the pump flow exceeds the minimum value. Minimum flow bypass operation is automatic based on a flow signal opening the minimum flow valve when the flow is low, with a concurrent high pump discharge pressure signal.
Other Provisions The RHR System is classified as Seismic Category I. Figures 2.4.1a, 2.4.1b, and 2.4.1c show the ASME Code Class for the RHR System. The RHR System is located in the Reactor Building.
Each of the three divisions is powered from the Class 1E division as shown on Figures 2.4.1a, 2.4.1b, 2.4.1c. In the RHR System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Outside the primary containment, each mechanical division of the RHR System (Divisions A, B, and C) is physically separated from the other divisions.
The RHR System has the following displays and controls in the main control room:
(1)    Parameter displays for the instruments shown on Figures 2.4.1a, 2.4.1b, and 2.4.1c.
2.4-4                                                                        Residual Heat Removal System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 (2)  Controls and status indication for the active safety-related components shown on Figures 2.4.1a, 2.4.1b, and 2.4.1c.
(3)  Manual system level initiation capability for the following modes:
(a)    LPFL initiation (b)    Standby (c)    Shutdown cooling (d)    Suppression pool cooling (e)    Drywell spray RHR System components with displays and control interfaces with the Remote Shutdown System (RSS) are shown on Figures 2.4.1a and 2.4.1b.
The safety-related electrical equipment shown on Figures 2.4.1a, 2.4.1b, and 2.4.1c located inside the primary containment and the Reactor Building is qualified for a harsh environment.
The motor-operated valves shown on Figures 2.4.1a, 2.4.1b, and 2.4.1c have active safety-related functions and perform these functions to open, close, or both open and close, under differential pressure, fluid flow, and temperature conditions.
The check valves (CVs) shown on Figures 2.4.1a, 2.4.1b, and 2.4.1c have safety-related functions to open, close, or both open and close under system pressure, fluid flow, and temperature conditions.
The RHR System main pumps are interlocked to prevent starting with a closed suction path.
Each RHR loop has a continuously running jockey pump to maintain the system piping continuously filled with water. The jocky pump is stopped by a RHR initiation signal or may be stopped or started manually.
The piping and components outside the shutdown cooling suction line containment isolation valves and outside the suppression pool containment isolation valves, and upstream of the suction side of the pump with all its branches have a design pressure of 2.82 MPaG for intersystem LOCA (ISLOCA) conditions. Refer to Figures 2.4.1a, 2.4.1b, and 2.4.1c. For RHR-A, the upgraded branch lines from the main pump suction include the path to and including the suppression pool suction valve, the path to the shutdown cooling outboard containment isolation valve, and the path to the jockey pumps discharge check valve including the jockey pumps bypass return line. For RHR-B and C, the upgraded branch lines include all the paths listed for RHR-A plus the supplemental fuel pool cooling suction path from the Fuel Pool Cooling System (including the RHR isolation valve) that connects to the shutdown cooling suction line, titled From FPC. The upgraded lines also include the pipelines and valves that are part of the AC independent water addition mode that extend from the noncode boundary Residual Heat Removal System                                                                            2.4-5
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 indicated by NNS to the external connection outside the reactor building and to the Fire Protection System interfaces indicated by FP.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.4.1 provides a definition of the inspections, test and/or analyses, together with associated acceptance criteria, which will be undertaken for the RHR System.
2.4-6                                                                        Residual Heat Removal System
 
ABWR Residual Heat Removal System FEEDWATER A PRIMARY CONTAINMENT RPV RHR                                                                                                  NBS 1                                                                                                  2 RHR P
1  2 R          R                                                                    R M M          M RCIC 2 RHR R
25A5675AA Revision 7 M
R R                            M                  M    R                          FE JOCKEY PUMP M                            R S/P      R M                    R HX S
P    T                M          T    R MAIN PUMP          R    R                  R Design Control Document/Tier 1 RCW-A RCW-A          2 RHR RHR 2 RCW                                      RCW NOTES:
: 1. ALL ELECTRICAL POWER LOADS FOR THE CLASS 1E COMPONENTS SHOWN ON THIS FIGURE ARE POWERED FROM CLASS 1E DIVISION I EXCEPT FOR THE OUTBOARD CONTAINMENT ISOLATION VALVE OF THE SHUTDOWN COOLING SUCTION LINE, WHICH IS DIVISION II.
Figure 2.4.1a Residual Heat Removal System (RHR-A) 2.4-7
 
25A5675AA Revision 7 ABWR                          Design Control Document/Tier 1 Figure 2.4.1b Residual Heat Removal System (RHR-B) 2.4-8                                Residual Heat Removal System
 
ABWR Residual Heat Removal System REACTOR BUILDING EXTERNAL CONNECTION FP RHR FROM                                                      NNS PRIMARY                                                FPC RHR CONTAINMENT                              FPC                                        FROM FP 2
FPC M                M DRYWELL SPRAY                                        2 RHR SPARGER                      M        M                TO FPC NNS 1 2                                                                            2 P        M P
25A5675AA Revision 7 12 FCS RHR 2
M          M            HPCF-C                    TO FCS RHR      2                                                                  FE M
M 1
RPV RHR M                      M                          FE M        JOCKEY PUMP WETWELL SPRAY              S/P                                                                  M Design Control Document/Tier 1 SPARGER                      M HX S
P    T                            T MAIN PUMP RHR 2      RCW-C      RCW-C      2 RHR RCW                                  RCW NOTES:
: 1. ALL ELECTRICAL POWER LOADS FOR THE CLASS 1E COMPONENTS SHOWN ON THIS FIGURE ARE POWERED FROM CLASS 1E DIVISION III EXCEPT FOR THE OUTBOARD CONTAINMENT ISOLATION VALVE OF THE SHUTDOWN COOLING SUCTION LINE, WHICH IS DIVISION I.
: 2. DRYWELL AND WETWELL SPRAY SPRAGERS ARE COMMON TO DIVISIONS B AND C.
2.4-9                                                  Figure 2.4.1c Residual Heat Removal System (RHR-C)
 
ABWR 2.4-10 LOCAL AREA                                MAIN CONTROL ROOM                                LOCAL AREA Plant Sensors                                                                              Device Actuators RHR                      RHR Manual Manual System          Pump and Valve Initiation            Controls SSLC PROCESSING RHR    Automatic and Manual System Initiation and Control Drywell Pressure NBS Reactor Water Level                                          RHR LOGIC Reactor Vessel Pressure
                                                                                      - Sensor Channel Trip Decision 25A5675AA Revision 7
                                                                                      - System Coincidence Trip Decision          RHR    Manual Pump and Valve Control
                                                                                      - Control and Interlock Logic
                                                                                      - Division-of-Sensors Bypass Pump Flow Pump Discharge Pressure                            - Calibration, Self-Diagnosis Drywell Spray Valve Position, Div. B & C DG    DG Start Division A Only RHR Injection Valve Position Wetwell Spray Valve Position, Div. B & C.
Shutdown Cooling Suction Valve Position Suppression Pool Suction Valve Position                  LOCA                Pump Discharge Suppression Pool Return Valve Position            RCW                NBS Signal              Pressure Design Control Document/Tier 1 CAMS LOCA Signal Residual Heat Removal System Notes:
: 1. Diagram represents one of three RHR divisions, except as noted.
: 2. See Section 3.4, Figure 3.4b for SSLC Processing.
Figure 2.4.1d Residual Heat Removal System Control Interface Diagram
 
ABWR Residual Heat Removal System Table 2.4.1 Residual Heat Removal System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the RHR System is 1. Inspections of the as-built system will be          1. The as-built RHR System conforms with the shown in Figures 2.4.1a, 2.4.1b, 2.4.1c, and    conducted.                                              basic configuration shown in Figures 2.4.1a, 2.4.1d.                                                                                                2.4.1b, 2.4.1c, and 2.4.1d.
: 2. The ASME Code components of the RHR            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the RHR System that          ASME Code components of the RHR System integrity under internal pressures that will be    are required to be hydrostatically tested by    conform with the requirements in the ASME experienced during service.                        the ASME Code.                                  Code, Section III.
: 3.                                                3.                                                  3.
: a. The RHR System is automatically                a. Tests will be conducted using simulated          a. Each division of the RHR System initiated in the LPFL mode when either a          input signals for each process variable to          receives an initiation signal.
25A5675AA Revision 7 high drywell pressure or a low reactor            cause trip conditions in two, three, and water level condition exists.                      four instrument channels of the same process variable.
: b. Each RHR division can be initiated              b. Tests will be conducted by initiating each        b. Each division of the RHR System manually (LPFL mode).                              division manually.                                  receives an initiation signal.
Design Control Document/Tier 1 2.4-11
 
Table 2.4.1 Residual Heat Removal System (Continued)
ABWR 2.4-12 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 3. continued                                        3. continued                                          3. continued
: c. Following receipt of an initiation signal,        c. Tests will be conducted on each RHR                c. Upon receipt of a simulated initiation the RHR System automatically initiates              division using a simulated initiation signal          signal, the following occurs:
and operates in the LPFL mode to                    and a simulated low reactor pressure (1) The RHR pump receives a signal to provide emergency makeup to the                      permissive signal.
start.
reactor vessel.
(2) The RPV injection valve receives a signal to open provided a low reactor pressure permissive signal is present, and the valve opens within 36 seconds after receiving the low reactor pressure permissive signal.
25A5675AA Revision 7 (3) The suppression pool return valve receives a signal to close.
(4) The wetwell spray valve receives a signal to close (Divisions B and C only).
: d. The LPFL injection flow for each division        d. Tests will be conducted on the as-built        d. The converted RHR flow satisfies the begins when the RPV dome pressure is                RHR System in the RHR LPFL mode.                  following:
no less than 1.55 MPa above the drywell              Analyses will be performed to convert the The LPFL injection flow for each division pressure.                                            test results to the conditions of the begins when the RPV dome pressure is Design Control Document/Tier 1 Design Commitment.
When the RPV dome pressure is no less                                                                  no less than 1.55 MPa above the drywell than 0.275 MPa greater than the drywell                                                                pressure.
pressure, the LPFL injection flow for each                                                              When the RPV dome pressure is no less Residual Heat Removal System division is 954 m3/h minimum.                                                                          than 0.275 MPa greater than the drywell pressure, the LPFL injection flow for each division is 954 m3/h minimum.
 
Table 2.4.1 Residual Heat Removal System (Continued)
ABWR Residual Heat Removal System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                  Inspections, Tests, Analyses                              Acceptance Criteria
: 3. continued                                            3. continued                                          3. continued
: e. The system automatically aligns to the                e. Tests will be conducted on each RHR                e. Each division automatically aligns to the LPFL mode of operation from the test                    division using simulated LPFL initiation              LPFL mode of operation from the test mode, the suppression pool cooling or                    signals.                                              mode, the suppression pool cooling or wetwell spray modes upon receipt of an                                                                        wetwell spray modes upon receiving an initiation signal.                                                                                            initiation signal. The wetwell spray mode is applicable for Divisions B or C.
: f. If a drywell spray valve is open in Division        f. Tests will be conducted on RHR Division          f. Drywell spray valves in a division close B or C, that RHR division automatically                  B and C drywell spray mode using a                    on receipt of injection valve not fully reverts to the LPFL mode in response to                  simulated injection valve opening signal.            closed signal in that division.
the injection valve beginning to open.
25A5675AA Revision 7
: g. The RPV injection valve in each division              g. Tests will be conducted on the injection          g. The RPV injection valve in each division requires a low reactor vessel pressure                  valves in each RHR division using a                  requires a low reactor vessel pressure permissive signal to open and closes                    simulated reactor vessel pressure signal.            permissive signal to open and closes automatically on receipt of a high reactor                                                                    automatically on receipt of a high reactor vessel pressure signal.                                                                                        vessel pressure signal.
: 4.                                                      4.                                                    4.
: a. In the suppression pool cooling mode,                a. Inspections and analyses will be                  a. In the suppression pool cooling mode, the total heat removal capacity                          performed to determine the heat                      the total heat removal capacity requirement between the RHR System                      exchanger's effective heat removal                    requirements between the RHR System Design Control Document/Tier 1 and ultimate heat sink is no less than                  capacity, for each division.                          and ultimate heat sink is no less than 0.371 MJ/s°C for each division.                                                                              0.371 MJ/s°C for each division.
: b. In the suppression pool cooling mode,                b. Tests will be performed on each RHR                b. In the suppression pool cooling mode, the RHR tube side heat exchanger flow                    division.                                            the RHR tube side heat exchanger flow rate is 954 m3/h minimum, per division.                                                                        rate is 954 m3/h minimum, per division.
2.4-13
 
Table 2.4.1 Residual Heat Removal System (Continued)
ABWR 2.4-14 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                            Acceptance Criteria
: 4. continued                                4. continued                                        4. continued
: c. The RHR pumps have sufficient NPSH.      c. Inspections, tests and analyses will be          c. The available NPSH exceeds the performed upon the as -built RHR                    required NPSH required by the pumps.
System. Inspections of the as-built                Test result/report confirms that the RHR system will be performed to obtain piping          valves, RHR pumps and RHR heat system dimensions and other necessary              exchangers perform their intended information. The required NPSH of                  functions during post-LOCA operation for procured pumps will be determined by an            a minimum of 30 days.
inspection of the vendor specifications.
The analysis will consider the effects of:
                                                                                    - Pressure losses for pump inlet piping and components.
25A5675AA Revision 7
                                                                                    - Suction from the suppression pool with water level at the minimum value.
                                                                                    - Analytically derived values for blockage of pump suction strainers based upon the as-built system.
                                                                                    - Design basis debris loading of pumped fluid under conditions ranging from normal operating to design basis accident conditions.
Design Control Document/Tier 1
                                                                                    - Design basis fluid temperature (100°C).
                                                                                    - Containment at atmospheric pressure.
Residual Heat Removal System
                                                                                    - Confirm vertical and horizontal separation between the SRV Quencher and RHR Suction Strainer.
 
Table 2.4.1 Residual Heat Removal System (Continued)
ABWR Residual Heat Removal System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                  Inspections, Tests, Analyses                            Acceptance Criteria
: 5.                                                      5.                                                  5.
: a. The drywell spray inlet valves can only be          a. Tests will be performed of the drywell            a. The two in-series drywell spray valves opened if a high drywell pressure exists                spray valve interlock logic using                    are blocked from being open and if the injection valves are fully closed.          simulated drywell pressure and valve                simultaneously unless signals indicative position signals.                                    of the following conditions exist concurrently:
(1) Drywell pressure is high.
(2) The RPV injection valve is fully closed.
(3) The shutdown cooling suction valve 25A5675AA Revision 7 is fully closed.
The drywell spray valves will automatically close if signals indicative of the following condition exists:
(1) The RPV injection valve is not fully closed.
: b. The wetwell spray flow rate for either              b. Tests will be conducted on Divisions B            b. RHR Division B provides wetwell spray Division B or C is no less than 114 m3/h.              and C in the wetwell spray mode.                    flow greater than or equal to 114 m3/h.
RHR Division C provides wetwell spray Design Control Document/Tier 1 flow greater than or equal to 114 m3/h.
: 6.                                                      6.                                                  6.
: a. Shutdown cooling is initiated manually              a. Tests will be conducted on the RHR                a. The RHR shutdown mode operates when once the RPV has been depressurized                    shutdown cooling mode for manual                    reactor vessel pressure is below system below the system low pressure                          initiation, using simulated reactor vessel          low pressure permissive. The RHR permissive.                                            pressure signals.                                    shutdown mode is not manually initiated when reactor vessel pressure is not less than the low pressure permissive.
2.4-15
 
Table 2.4.1 Residual Heat Removal System (Continued)
ABWR 2.4-16 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                            Acceptance Criteria
: 6. continued                                        6. continued                                        6. continued
: b. In any division, the shutdown cooling            b. Tests will be conducted on each RHR              b. In any division, the shutdown cooling suction valve cannot be opened unless                division to open the shutdown cooling              suction valve cannot be opened unless the following valves in that division are            suction valve.                                      the following valves in that division are closed:                                                                                                  closed:
Suppression pool suction valve.                                                                          Suppression pool suction valve.
Suppression pool return valve.                                                                          Suppression pool return valve.
Drywell spray valves.                                                                                    Drywell spray valves.
Wetwell spray valve.                                                                                    Wetwell spray valve.
: c. Each shutdown cooling suction valve            c. Tests will be conducted on each RHR            c. Each shutdown cooling suction valve automatically closes on low reactor water            division using a simulated reactor water            automatically closes on low reactor water 25A5675AA Revision 7 level.                                              level signal.                                      level.
: d. The low pressure portions of the                  d. Tests will be conducted on the shutdown          d. The shutdown cooling suction valves shutdown cooling piping are protected                cooling suction valves in each RHR                  close when the RHR System receives a from high reactor pressure by automatic              division using a simulated reactor vessel          simulated high reactor vessel pressure closure of the shutdown cooling suction              pressure signal.                                    signal.
valves on a high reactor vessel pressure signal.
: e. In the shutdown cooling mode, the RHR            e. In the shutdown cooling mode, system            e. The RHR heat exchangers tube side flow tube side heat exchanger flow rate is                functional tests will be performed to              rate is greater than or equal to 954 m3/h.
greater than or equal to 954 m3 /h.
Design Control Document/Tier 1 determine system flow rate through each            Heat exchanger removal capacity in this heat exchanger. Inspections and analy-              mode is bounded by suppression pool ses shall be performed to verify that the          cooling requirements.
shutdown cooling mode is bounded by Residual Heat Removal System suppression pool cooling requirements.
: 7. In the augmented fuel pool cooling mode, the 7. Tests will be performed to determine system 7.            The RHR tube side heat exchanger flow rate RHR tube side heat exchanger flow rate for      flow rate through each heat exchanger in the              is greater than or equal to 350 m3/h in the Divisions B or C is no less than 350 m3/h      augmented fuel pool cooling mode. Inspec-                augmented fuel pool cooling mode. Heat (heat exchanger heat removal capacity in this  tions and analyses shall be performed to ver-            exchanger heat removal capacity in this mode is bounded by suppression pool cool-      ify that the augmented fuel pool cooling mode            mode is bounded by suppression pool cool-ing requirements).                              is bounded by suppression pool cooling                    ing requirements.
requirements.
 
Table 2.4.1 Residual Heat Removal System (Continued)
ABWR Residual Heat Removal System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 8.                                              8.                                                8.
: a. Each division of the RHR has a minimum        a. Tests will be conducted on the pump            a. The pump minimum flow valve receives a flow bypass mode that assures there is          minimum flow valve interlock logic using          signal to open when signals indicative of always flow in the RHR pumps when they          simulated pressure and flow signals.              the following conditions exist are operating.                                                                                      concurrently:
(1) Pump discharge pressure is high when the pump starts.
(2) Pump flow is low.
The pump minimum flow valve receives a signal to close when a signal indicative of 25A5675AA Revision 7 the following condition exists:
(1) Pump flow exceeds minimum value.
: b. Each division of the RHR System has a        b. Tests and analyses will b e conducted on        b. The available minimum flow exceeds the minimum flow bypass mode that assures            each division of the RHR System in the            required minimum flow.
there is always flow in the RHR pumps            minimum flow mode. The tests will when they are operating.                        quantify pump flow and compare with pump required minimum flow.
: 9. Each of the three RHR divisions is powered 9.                                              9.
from the Class 1E division as shown on
: a. Tests will be performed on the RHR          a. The test signal exists only in the Class 1E Design Control Document/Tier 1 Figures 2.4.1a, 2.4.1b and 2.4.1c. In the RHR System by providing a test signal to only      division under test in the RHR System.
System, independence is provided between one Class 1E division at a time.
Class 1E divisions, and between Class 1E divisions and non-Class 1E equipment.
: b. Inspection of the as-installed Class 1E        b. In the RHR System, physical separation divisions in the RHR System will be                or electrical isolation exists between performed.                                        Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
2.4-17
 
Table 2.4.1 Residual Heat Removal System (Continued)
ABWR 2.4-18 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                              Acceptance Criteria
: 10. Each mechanical division of the RHR System 10. Inspections of the as-built RHR System will              10. Each mechanical division of the RHR System (Divisions A, B, C) is physically separated    be performed.                                                is physically separated from other mechanical from the other divisions.                                                                                  divisions of RHR System by structural and/or fire barriers with the exception of components inside primary containment.
: 11. Main control room displays and controls          11. Inspections will be performed on the main        11. Displays and controls exist or can be provided for RHR System are defined in                control room displays and controls for the            retrieved in the main control room as defined Section 2.4.1.                                        RHR System.                                          in Section 2.4.1.
: 12. RSS displays and controls provided for the  12. Inspections will be performed on the RSS                12. Displays and controls exist on the RSS as RHR System are as defined in Section 2.4.1. displays and controls for the RHR System.                  defined in Section 2.4.1.
: 13.                                                  13.                                                  13.
25A5675AA Revision 7
: a. MOVs designated in Section 2.4.1 as                a. Tests of installed valves for opening,            a. Upon receipt of the actuating signal, each having an active safety function open,                closing or both opening and closing, will            MOV opens, closes, or both opens and close, or both open and close under                  be conducted under preoperational                    closes, depending upon the valves differential pressure, fluid flow, and                differential pressure, fluid flow, and                safety functions.
temperature conditions.                              temperature conditions.
: b. Check valves (CVs) designated in                  b. Tests of installed valves for opening,            b. Based on the direction of the differential Section 2.4.1 as having an active safety-            closing, or both opening and closing, will            pressure across the valve, each CV related function open, close, or both open            be conducted under system                            opens, closes, or both opens and closes, and close, under system pressure, fluid              preoperational pressure, fluid flow, and              depending upon the valves safety Design Control Document/Tier 1 flow, and temperature conditions.                    temperature conditions.                              functions.
Residual Heat Removal System
 
Table 2.4.1 Residual Heat Removal System (Continued)
ABWR Residual Heat Removal System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 14. The RHR System main pumps are                  14. Tests will be conducted on the RHR pump      14. Each RHR System pump is prevented from interlocked to prevent starting with a closed      start logic using simulated valve position      starting unless signals indicative of one of the suction path.                                      signals.                                        following conditions exists:
: a. A suction path from the suppression pool is available. (The suppression pool suction valve is fully open.)
: b. A suction path from the RPV via the shutdown cooling suction line is available. (The shutdown cooling suction valve and inboard and outboard isolation valves are all fully open.)
25A5675AA Revision 7 Design Control Document/Tier 1 2.4-19
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.4.2 High Pressure Core Flooder System Design Description The High Pressure Core Flooder (HPCF) System is comprised of two separate divisions. The function of the HPCF System is to provide emergency makeup water to the reactor vessel for transient or loss-of-coolant accident (LOCA) events. Each HPCF division consists of a pump, piping, valves and controls and can utilize either of two water sources, the condensate storage tank (CST) or the suppression pool (S/P). The primary source of suction water supply is from the CST. The S/P water is the secondary source of supply. Figure 2.4.2a shows the basic system configuration and scope. Figure 2.4.2b shows the HPCF System control interfaces.
The HPCF System is classified as safety-related.
The HPCF System operates in the following modes:
(1)  High pressure flooder.
(2)  Full flow test.
(3)  Minimum flow bypass.
High Pressure Flooder Mode As shown on Figure 2.4.2b, the HPCF System channel measurements are provided to the Safety System Logic and Control (SSLC) for signal processing, setpoint comparisons, and generating trip signals. The HPCF System is automatically initiated in the high pressure flooder mode when either a high drywell pressure signal or low reactor water level signal exists. Both divisions of the HPCF System are actuated at a reactor water level below the RCIC actuation level. The SSLC System processors use a two-out-of-four voting logic for system initiation and shutdown. Manual HPCF System initiation can also be performed.
Following receipt of an initiation signal, the HPCF System automatically initiates and operates in the high pressure flooder mode to provide water to the core region of the reactor. The pumps are motor-driven centrifugal pumps that provide flow as a function of reactor vessel pressure.
The flow in each division is not less than a value corresponding to a straight line between a flow of 182 m3/h at a differential pressure of 8.12 MPa and a flow of 727 m3/h at a differential pressure of 0.69 MPa. The HPCF System has the capability to deliver at least 50% of these flow rates with 171°C water at the pump suction. The differential pressure values represent the difference between the reactor vessel pressure and the pressure of the air space of the source water for the pump. System flow into the reactor vessel is achieved within 16 seconds of receipt of an initiation signal and power available at the emergency busses.
The HPCF pumps have sufficient net positive suction head (NPSH) available at the pumps.
2.4-20                                                                      High Pressure Core Flooder System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 During this mode, pump suction is from the CST. Automatic transfer of pump suction from the CST to the S/P occurs when a low CST water level or high suppression pool water level signal exists. The CST and suppression pool water level signals are processed through the SSLC two-out-of-four voting logic to initiate suction transfer.
When a high water level signal in the reactor pressure vessel exists, the reactor vessel injection valve is automatically closed. When the low reactor water level initiation signal recurs, the injection valve automatically re-opens to reestablish HPCF flow.
Full Flow Test Mode Each division of the HPCF System has a full flow test mode to permit testing during plant operation. In this mode, water is taken from the suppression pool and returned to the suppression pool via the test return line. The injection valve is kept closed to prevent any vessel injection during the test.
If a system initiation signal occurs during the full flow test mode, each division of the HPCF System automatically aligns to the high pressure flooder mode.
Minimum Flow Bypass Mode Each division of the HPCF System has a minimum flow bypass mode that assures there is always flow in the HPCF pumps when they are operating. This is accomplished automatically by monitoring pump discharge flow, and opening a minimum flow valve to the suppression pool when flow falls below the minimum value. The minimum flow valve closes when the pump flow exceeds the minimum value. Minimum flow bypass operation is automatic based on a flow signal opening the minimum flow valve when the flow is low, with a concurrent high pump discharge pressure signal.
Other Provisions The HPCF System is classified as Seismic Category I. Figure 2.4.2a shows the ASME Code Class for the HPCF System. The HPCF System is located both inside the primary containment and within the Reactor Building.
Each of the two HPCF divisions is powered from the respective Class 1E division as shown on Figure 2.4.2a. In the HPCF System, independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
Outside the primary containment, except for piping from the CST, each mechanical division of the HPCF System (Divisions B and C) is physically separated from the other division. Outside the primary containment, except for piping from the CST, both HPCF divisions are physically separated from the Reactor Core Isolation Cooling (RCIC) System.
The HPCF System has the following displays and controls in the main control room:
(1)    Parameter displays for the instruments shown on Figure 2.4.2a.
High Pressure Core Flooder System                                                                          2.4-21
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 (2)    Controls and status indication for the active safety-related components shown on Figure 2.4.2a.
(3)    Manual system level initiation capability for the high pressure flooder mode.
HPCF System components with displays and control interfaces with the Remote Shutdown System (RSS) are shown on Figure 2.4.2a The safety-related electrical equipment shown on Figure 2.4.2a located inside the primary containment and in the Reactor Building is qualified for a harsh environment.
The motor-operated valves (MOVs) shown on Figure 2.4.2a have active safety-related functions to open, close, or both open and close, and perform these functions under differential pressure, fluid flow, and temperature conditions.
The check valves (CVs) shown on Figure 2.4.2a have safety-related functions to open, close, or both open and close under system pressure, fluid flow, and temperature conditions.
The HPCF System pumps have interlocks which prevent operation if both suction valves are closed.
The HPCF System suction piping and components from the pump suction valves to the pump inlet have a design pressure of 2.82 MPaG for intersystem LOCA (ISLOCA) conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.4.2 provides a definition of the inspections, test and/or analyses, together with associated acceptance criteria, which will be undertaken for the HPCF System.
2.4-22                                                                    High Pressure Core Flooder System
 
ABWR High Pressure Core Flooder System R                                                      PRIMARY CONTAINMENT FROM                                FE        2 CONDENSATE MUWC HPCF STORAGE TANK                                                                              2 1                HPCF RPV 1
MUWC R
2 HPCF                                                      M          P TO HPCF-C        TO HPCF-B                                                                  HPCF 1 TO S/P VIA SLC R                          RHR TEST M R          RETURN LINE OF M
SAME DIVISION 25A5675AA Revision 7 RCIC HPCF 2                                                              2 HPCF RHR R
P                                                      R P                              M        S/P R
S MAIN PUMP SPCU HPCF 2
Design Control Document/Tier 1 R
M NOTES:
: 1. DIVISION B SHOWN, DIVISION C IDENTICAL EXCEPT INTERFACE CONNECTIONS WITH RSS AND SLC ON DIVISION B ONLY.
: 2. ALL ELECTRICAL POWER LOADS FOR THE CLASS 1E COMPONENTS SHOWN ON THIS FIGURE ARE POWERED FROM CLASS 1E DIVISION II (DIVISION C POWERED FROM CLASS 1E DIVISION III).
2.4-23 Figure 2.4.2a High Pressure Core Flooder System
 
ABWR 2.4-24 LOCAL AREA                                      MAIN CONTROL ROOM                          LOCAL AREA Plant Sensors                                                                              Device Actuators HPCF                  HPCF Manual Manual System            Pump and Valve Initiation              Controls System Flow Pump Discharge Pressure                                  SSLC PROCESSING HPCF Pump Suction Pressure Condensate Storage Tank Suction Valve Position Suppression Pool Suction Valve Position HPCF LOGIC HPCF Automatic  and Manual System
                                                                                                  - Sensor Channel Trip Decision            Initiation and Control 25A5675AA Revision 7 Drywell Pressure                                      - System Coincidence Trip Decision NBS    Reactor Water Level                                  - Control and Interlock Logic
                                                                                                  - Division-of-Sensors Bypass
                                                                                                  - Calibration, Self-Diagnosis        HPCF Manual Pump and Valve Actuation Condensate Storage MUWC Tank Water Level DG    Initiation Signal in Same Division NBS    Pump Discharge Pressure ACS    Suppression Pool Water Level Design Control Document/Tier 1 High Pressure Core Flooder System Notes:
: 1. Diagram represents one of two HPCF divisions.
: 2. See Section 3.4, Figure 3.4b for SSLC Processing.
: 3. HPCF Division C has manual initiation and monitoring capability which is diverse from SSLC (see Section 3.4).
Figure 2.4.2b High Pressure Core Flooder System Control Interface Diagram
 
ABWR High Pressure Core Flooder System Table 2.4.2 High Pressure Core Flooder System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                  Inspections, Tests, Analyses                              Acceptance Criteria
: 1. The basic configuration of the HPCF System 1. Inspections of the as-built system will be                1. The as-built HPCF System conforms with the is as shown on Figures 2.4.2a and 2.4.2b. conducted.                                                    basic configuration shown on Figures 2.4.2a and 2.4.2b.
: 2. The ASME Code components of the HPCF 2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary          Code components of the HPCF System          ASME Code components of the HPCF integrity under internal pressures that will be required to be hydrostatically tested by the System conform with the requirements in the experienced during service.                    ASME Code.                                  ASME Code, Section III.
: 3.                                                    3.                                                  3.
: a. The HPCF System is automatically                    a. Tests will be conducted using simulated          a. Each division of the HPCF System initiated in the high pressure flooder                input signals for each process variable to          receives an initiation signal.
25A5675AA Revision 7 mode when either a high drywell pressure              cause trip conditions in two, three, and signal or a low reactor water level signal            four instrument channels of the same exists.                                                process variable.
: b. Manual HPCF System initiation can be                b. Tests will be conducted by manually              b. Each division of the HPCF System performed in the high pressure flooder                initiating each HPCF division.                      receives an initiation signal.
mode.
: c. Following receipt of an initiation signal,        c. Tests will be conducted on each HPCF            c. Upon receipt of a simulated initiation the HPCF System automatically initiates                division using a simulated initiation                signal, the following occurs:
and operates in the high pressure flooder              signal.                                              -  The HPCF pump receives a signal to Design Control Document/Tier 1 mode to provide water to the core region start.
of the reactor.
                                                                                                                                                          -  The RPV injection valve receives a signal to open.
                                                                                                                                                          -  The condensate storage tank suction valve receives a signal to open.
                                                                                                                                                          -  The test line return valve receives a signal to close.
2.4-25
 
Table 2.4.2 High Pressure Core Flooder System (Continued)
ABWR 2.4-26 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 3. continued                                        3. continued                                        3. continued
: d. The HPCF System flow in each division is        d. Tests will be conducted on each division          d. The converted HPCF flow satisfies the not less than a value corresponding to a            of the as-built HPCF System in the HPCF              following:
straight line between a flow of 182 m3/h            high pressure flooder mode. Analyses will The HPCF System flow in each division is at a differential pressure of 8.12 MPa and          be performed to convert the test results to 3                                                                                      not less than a value corresponding to a a flow of 727 m /h at a differential                the conditions of the Design straight line between a flow of 182 m3/h pressure of 0.69 MPa.                              Commitment.                                          at a differential pressure of 8.12 MPa and a flow of 727 m3/h at a differential pressure of 0.69 MPa.
: e. The HPCF System has the capability to            e. Analyses will be performed of the as-built        e. The HPCF System has the capability to deliver at least 50% of the flow rates in          HPCF System to assess the system flow                deliver at least 50% of the flow rates in 25A5675AA Revision 7 item 3d with 171°C water at the pump                capability with 171°C water at the pump              item 3d with 171°C water at the pump suction.                                            suction.                                            suction.
: f. System flow into the reactor vessel is        f. Tests will be conducted on each HPCF            f. The HPCF System flow is achieved within achieved within 16 seconds of receipt of            division using simulated initiation signals.        16 seconds of receipt of a simulated an initiation signal and power available at                                                              initiation signal.
the emergency busses.
Design Control Document/Tier 1 High Pressure Core Flooder System
 
Table 2.4.2 High Pressure Core Flooder System (Continued)
ABWR High Pressure Core Flooder System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                          Acceptance Criteria
: 3. continued                              3. continued                                        3. continued
: g. The HPCF pumps have sufficient NPSH    g. Inspections, tests and analyses will be          g. The available NPSH exceeds the available at the pumps.                    performed upon the as-built system.                required NPSH required by the pumps.
Inspections of the as-built system will be          Test result/report confirms that the HPCF performed to obtain piping system                  valves and HPCF pumps perform their dimensions and other necessary                      intended functions during post-LOCA information. The required NPSH of                  operation for a minimum of 30 days.
procured pumps will be determined by an inspection of the vendor specifications.
The analysis will consider the effects of:
                                                                                      - Pressure losses for pump inlet piping and components.
25A5675AA Revision 7
                                                                                      - Suction from the suppression pool with water level at the minimum value.
                                                                                      - Analytically derived values for blockage of pump suction strainers based upon the as-built system.
                                                                                      - Design basis debris loading of pumped fluid under conditions ranging from normal operating to design basis accident conditions.
                                                                                      - Confirm vertical and horizontal Design Control Document/Tier 1 separation between the SRV Quencher and HPCF Suction Strainer.
                                                                                      - Design basis fluid temperature (100°C).
                                                                                      - Containment at atmospheric pressure.
2.4-27
 
Table 2.4.2 High Pressure Core Flooder System (Continued)
ABWR 2.4-28 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                          Acceptance Criteria
: 3. continued                                    3. continued                                        3. continued
: h. Automatic transfer of pump suction from      h. Tests will be conducted on each HPCF            h. HPCF System receives suction transfer the CST to the suppression pool occurs          division using simulated input signals for          initiation signal.
when a low CST water level or high              each process variable to cause trip suppression pool water level signal            conditions in two, three, and four exists.                                        instrument channels of the same process variable.
: i. Following receipt of a suction transfer          i. Test will be conducted on each HPCF          i. Upon receipt of a simulated suction initiation signal, the HPCF System                    division using simulated suction transfer        transfer initiation signal, the following automatically switches pump suction.                  initiation signals.                              occurs:
                                                                                                                                                    -  Suppression pool suction valve 25A5675AA Revision 7 opens.
                                                                                                                                                    -  CST suction valve closes.
: j. When a high water level signal in the            j. Tests will be conducted on each HPCF        j. The HPCF System receives a signal to reactor pressure vessel exists, the                  division using simulated high reactor            close the reactor vessel injection valve.
reactor vessel injection valve is                    water level signals to cause trip automatically closed.                                conditions in two, three, and four instrument channels of water level variable.
: k. Following receipt of an injection valve          k. Tests will be conducted on each HPCF        k. Upon receipt of a simulated injection Design Control Document/Tier 1 closure signal, the HPCF System                      division using a simulated injection valve        valve closure signal, the reactor vessel automatically closes the vessel injection            closure signal.                                  injection valve closes.
High Pressure Core Flooder System valve.
: l. Following HPCF System injection valve            l. Tests will be conducted on each HPCF        l. Upon receipt of a simulated low reactor closure on a high reactor water level                division using a simulated low reactor            water level signal, the vessel injection signal, when the low water level initiation          water level signal.                              valve opens.
signal recurs, the vessel injection valve automatically re-opens to re-establish HPCF flow.
 
Table 2.4.2 High Pressure Core Flooder System (Continued)
ABWR High Pressure Core Flooder System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 3. continued                                          3. continued                                        3. continued
: m. Each division of the HPCF System has a            m. Tests will be conducted on each as-built        m. Water is pumped at a flow rate of not less full flow test mode to permit testing during          HPCF division, using installed controls,            than 182 m3/h in the test flow mode.
plant operation.                                      power supplies and other auxiliaries.
Water will be pumped in the test flow mode with system head equivalent to a pressure differential of at least 8.12 MPa between the RPV and the air space of the source water for the pump.
: n. If a system initiation signal occurs during      n. Tests will be performed on each HPCF            n. Upon receipt of a simulated initiation the full flow test mode, each division of          division using simulated initiation signals.        signal, each HPCF division automatically the HPCF System automatically aligns to                                                                aligns to the high pressure flooder mode 25A5675AA Revision 7 the high pressure flooder mode.                                                                        of operation from the test mode.
: o. Each division of the HPCF System has a          o. Tests will be conducted on the pump              o. The pump minimum flow valve receives a minimum flow bypass mode that assures              minimum flow valve interlock logic using            signal to open when signals indicative of there is always flow in the HPCF pumps              simulated pressure and flow signals.                the following conditions exist when they are operating.                                                                                concurrently:
                                                                                                                                                    -  Pump discharge pressure is high when the pump starts and,
                                                                                                                                                    -  Pump flow is low.
Design Control Document/Tier 1 The pump minimum flow valve receives a signal to close when a signal indicative of the following condition exists:
                                                                                                                                                    -  Pump flow exceeds the minimum value.
: p. Each division of the HPCF System has a          p. Tests and analyses will be conducted on          p. The available minimum flow exceeds the minimum flow bypass mode that assures              each division of the HPCF System in the            required minimum flow.
there is always flow in the HPCF pumps              minimum flow mode. The tests will when they are operating.                            quantify pump flow and compare with pump required minimum flow.
2.4-29
 
Table 2.4.2 High Pressure Core Flooder System (Continued)
ABWR 2.4-30 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 4. Each of the two HPCF divisions is powered    4.                                                4.
from the respective Class 1E division as
: a. Tests will be performed on the HPCF            a. The test signal exists only in the Class 1E shown on Figure 2.4.2a. In the HPCF System by providing a test signal in one          division under test in the HPCF System.
System, independence is provided between Class 1E division at a time.
Class 1E divisions, and between Class 1E                                                            b. In the HPCF System, physical separation divisions and non-Class 1E equipment.            b. Inspection of the as-built Class 1E                or electrical isolation exists between divisions in the HPCF System will be              Class 1E divisions. Physical separation performed.                                        or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 5. Outside the primary containment, except for 5. Inspections of the as-built HPCF System will 5. Outside the primary containment, except for piping from the CST, each mechanical          be performed.                                  piping from the CST, each mechanical 25A5675AA Revision 7 division of the HPCF System (Divisions B and                                                  division of the HPCF System is physically C) is physically separated from the other                                                      separated from the other mechanical division division. Except for piping from the CST, both                                                of the HPCF System, and both HPCF HPCF divisions are physically separated from                                                  divisions are separated from the RCIC the RCIC System.                                                                              System by structural and/or fire barriers.
: 6. Main control room displays and controls    6. Inspections will be performed on the main        6. Displays and controls exist or can be provided for the HPCF System are as defined    control room displays and controls for the          retrieved in the main control room as defined in Section 2.4.2.                              HPCF System.                                        in Section 2.4.2.
: 7. RSS displays and controls provided for the  7. Inspections will be performed on the RSS        7. Displays and controls exist on the RSS as Design Control Document/Tier 1 HPCF System are as defined in Section          displays and controls for the HPCF System.        defined in Section 2.4.2.
2.4.2.
High Pressure Core Flooder System
: 8. MOVs designated in Section 2.4.2 as having 8. Tests of installed valves for opening, closing    8. Upon receipt of the actuation signal, each an active safety-related function open, close, or both opening and closing, will be                MOV opens, closes, or both opens and or both open and close under differential      conducted under preoperational differential        closes, depending upon the valves safety pressure, fluid flow, and temperature          pressure, fluid flow, and temperature              functions. The following valve opens in the conditions.                                    conditions.                                        following time limit:
Valve                    Time (s)
Injection valve            16 open
 
Table 2.4.2 High Pressure Core Flooder System (Continued)
ABWR High Pressure Core Flooder System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 9. CVs designated in Section 2.4.2 as having an 9. Tests of installed valves for opening, closing, 9. Based on the direction of the differential active safety-related function open, close, or  or both opening and closing, will be              pressure across the valve, each CV opens, both open and close, under system pressure,    conducted under system preoperational              closes, or both opens and closes, depending fluid flow, and temperature conditions.        pressure, fluid flow, and temperature              upon the valves safety functions.
conditions.
: 10. The HPCF System pumps have interlocks          10. Tests will be conducted on each HPCF      10. Each HPCF System pump is prevented from which prevent operation if both suction valves    System pump start logic using simulated      operating unless signals indicative of one of are closed.                                        valve position signals.                      the following conditions exists:
: a. A suction path from the S/P is available (the S/P suction valve is fully open).
25A5675AA Revision 7
: b. A suction path from the condensate storage tank is available (the CST suction valve is fully open).
Design Control Document/Tier 1 2.4-31
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.4.3 Leak Detection and Isolation System Design Description The Leak Detection and Isolation System (LDS) is a control and instrumentation system whose function is to detect and monitor leakage from the reactor coolant pressure boundary and initiate isolation of the leakage source. The system is designed to initiate automatic isolation of the process lines that penetrate the containment by closing the isolation valves. The functions of the LDS include: isolation of the main steamlines, the primary and secondary containment, and individual system process lines; activation of the Standby Gas Treatment System (SGTS);
monitoring of leakages inside and outside the primary containment; and providing the monitored leakage parameters in the main control room.
The LDS is classified as a Class 1E safety-related system.
The LDS logic design uses four instrument channels to monitor each leakage parameter that initiates an isolation function on a two-out-of-four channel trip.
As shown on Figure 2.4.3, the LDS safety-related channel measurements are provided as inputs to the Safety System Logic and Control (SSLC) for signal processing, setpoint comparisons, and generation of the trip signals that initiate the isolation functions. The LDS isolation logic consists of safety-related sensors, redundant instrument channels and logic processors that initiate the automatic isolation functions. Once isolation is initiated, the logic seals in the isolation signal, and operator action is required to reset the logic to its normal state.
The following primary and secondary containment isolation and automatic control functions are provided by the LDS using four instrument channels to monitor leakage:
(1)    Closure of the main steamline isolation valves (MSIVs) and main steamline (MSL) drain valves on a signal indicating low reactor water level, high main steamline flow in any main steamline, high ambient temperature in the MSL tunnel area or in the Turbine Building along the MSLs, low main condenser vacuum, or low steam inlet pressure to the main turbine.
(2)    Isolation of the Reactor Water Cleanup (CUW) System process lines on a signal indicating low reactor water level, high ambient MSL tunnel area temperature, high mass differential flow, high ambient temperature in the CUW areas, or when the Standby Liquid Control (SLC) System is activated.
(3)    Initiation of the SGTS on a signal indicating high drywell pressure, low reactor water level, high radiation in the secondary containment or high radiation in the fuel handling area.
2.4-32                                                                        Leak Detection and Isolation System
 
25A5675AA Revision 7 ABWR                                                                            Design Control Document/Tier 1 (4)      Isolation of Reactor Building Heating, Ventilating and Air Conditioning (HVAC)
System on a signal indicating high drywell pressure, low reactor water level, high radiation in the secondary containment or high radiation in the fuel handling area.
(5)      Isolation of containment purge and vent lines on a signal indicating high drywell pressure, low reactor water level, high radiation in the secondary containment or high radiation in the fuel handling area.
(6)      Isolation of the Reactor Building Cooling Water (RCW) System and of the HVAC Normal Cooling Water (HNCW) System lines on a signal indicating high drywell pressure or low reactor water level.
(7)      Isolation of the Residual Heat Removal (RHR) System shutdown cooling system loops on a signal indicating high reactor pressure or low reactor water level. Also, each RHR shutdown cooling division is individually isolated on a signal indicating high ambient temperature in its respective equipment area.
(8)      Isolation of the Reactor Core Isolation Cooling (RCIC) System steamline to the RCIC turbine on a signal indicating high steam flow in the RCIC line, low steam pressure in the RCIC line, high RCIC turbine exhaust pressure, or high ambient temperature in the RCIC equipment area.
(9)      Isolation of the Suppression Pool Cleanup (SPCU) System on a signal indicating high drywell pressure or low reactor water level.
(10) Isolation of the Flammability Control System (FCS) on a signal indicating high drywell pressure or low reactor water level.
(11) Isolation of the drywell sump low conductivity waste (LCW) and high conductivity waste (HCW) discharge lines on a signal indicating high drywell pressure or low reactor water level. Also, each discharge line is individually isolated on a signal indicating high radioactivity in the discharged liquid waste; only one channel is used for this function.
(12) Isolation of the LDS fission products monitor drywell sample and return lines on a signal indicating high drywell pressure or low reactor water level.
(13) The LDS provides to the neutron monitoring system a signal indicating a high drywell pressure or low reactor water level.
Separate manual controls in the control room are provided in LDS design for logic reset, MSIV operational control, MSIV partial closure tests, and for manual isolation of primary and secondary containment.
Leak Detection and Isolation System                                                                          2.4-33
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 Each MSIV has three pilot solenoid valves; two are used for operational control and the third is used to test the MSIV for partial closure. Each MSIV pilot solenoid valve is controlled separately by the LDS as follows:
(1)  Two of the three pilot solenoid valves of the MSIV are each provided with four divisional control signals to open the valve. MSIV closure occurs on loss of any two of the four divisional signals.
(2)  The third MSIV pilot solenoid valve is provided with one-out-of-two manual control signals to test the MSIV for partial closure. Division I or III manual signal is used to close the outboard MSIV, while Division II or IV manual signal is used to close the inboard MSIV.
Except for MSIVs, the LDS provides three separate divisional isolation signals (Divisions I, II and III) for automatic closure of the primary and secondary containment isolation valves. Each LDS divisional isolation signal initiates closure of the isolation valves that are assigned in the same division.
The LDS design includes the following manual controls for separate isolation of the RCIC System, and closure of the MSIVs and the primary and secondary containment isolation valves:
(1)  Four MSIV isolation switches-one per Divisions I, II, III, and IV.
Closure of all the MSIVs requires the actuation of any two of the four divisional MSIV isolation switches.
(2)  Three primary and secondary containment isolation switches-one per Divisions I, II and III.
Each isolation switch closes its respective divisional isolation valves in the primary and secondary containment, except for the MSIVs and RCIC.
(3)  Two RCIC isolation switches-one per Divisions I and II.
Either isolation switch isolates the steamline to the RCIC turbine and causes turbine trip. Division I switch closes the inboard, while Division II switch closes the outboard isolation valves.
Manual reset controls are provided at the divisional level to initialize the logic and to reset the logic after isolation has cleared. Separate reset functions are provided in the LDS logic design for the MSIVs, the RCIC, and the containment isolation.
The LDS design uses redundant channels and is fail-safe in the event of loss of electrical power to one division of LDS logic.
2.4-34                                                                    Leak Detection and Isolation System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Each of the four LDS divisional logic channels and associated sensors is powered from its respective divisional Class 1E power supply. In the LDS, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The LDS sensors are located in the Reactor Building and Turbine Building; the logic processors are located in the Control Building.
The LDS has the following displays and controls in the main control room:
(1)      Parameter displays for LDS plant sensors defined on Figure 2.4.3.
(2)      LDS manual controls as described above.
(3)      LDS divisional trip status.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.4.3 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the LDS.
Leak Detection and Isolation System                                                                      2.4-35
 
ABWR 2.4-36 LOCAL AREA                                              MAIN CONTROL ROOM CONTROLS                                                  LOCAL AREA PLANT SENSORS                                                                                                                    DEVICE ACTUATORS Reactor Bldg HVAC Exhaust Radiation Fuel Handling Area Vent Exhaust Radiation PRM Drywell Sump LCW MSIVs          Containment          RCIC Discharge Radiation
* Manual Isolation
* PCV Isolation
* RCIC Isolation Drywell Sump HCW
* MSIV Test
* Logic Reset
* Logic Reset Discharge Radiation
* Logic Reset SLC  SLC Initiation Signal RHR Area Temperature RCIC Area Temperature RCIC Steam Line Pressure SSLC PROCESSING - NOTE 2 25A5675AA Revision 7 RCIC Steam Line Flow LDS  CUW Mass Differential Flow NBS    MSIV Closure CUW Area Temperature                                                                  LDS LOGIC MSL Tunnel Area Amb. Temperature MSL A, B, C, D Steam Flow                                                - Sensor Channel Trip Decision                            RHR MSL Turbine Area Temperature                                              - System Coincidence Trip Decision                        CUW
                                                                                                                      - Control and Interlock Logic                            RCW
                                                                                                                      - Manual Division Trip (MSIV only)                        FCS      Primary HNCW      Containment
                                                                                                                      - Division-of-Sensors Bypass RCIC RCIC Turbine Exhaust Pressure                                                                                                        SPCU      Isolations
                                                                                                                      - Calibration, Self-Diagnosis AC LDS RW Drywell Pressure Reactor Water Level NBS  Reactor Vessel Pressure                                                                                                              R/B Secondary Containment Main Condenser Vacuum                                                                                                              HVAC Isolation Design Control Document/Tier 1 MSL Turbine Inlet Pressure MSIV Position Indications SGTS System Initiation Leak Detection and Isolation System RCIC System Isolation Notes:
: 1. Diagram represents one of four LDS divisions.
: 2. See Section 3.4, Figure 3.4b for SSLC processing.
Figure 2.4.3 Leak Detection and Isolation System Interface Diagram
 
ABWR Leak Detection and Isolation System Table 2.4.3 Leak Detection & Isolation System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                              Acceptance Criteria
: 1. The equipment comprising the LDS is              1. Inspection of the as-built system will be          1. The as-built LDS conforms with the defined in Section 2.4.3.                            conducted.                                            description in Section 2.4.3.
: 2. LDS logic uses four independent sensor            2. Tests will be conducted using simulated input 2. Isolation signal is initiated when at least any instrument channels of each process                  signals for each process variable to cause      two out of four channels have tripped.
variable described in Section 2.4.3 for its          trip conditions in two, three, and four automatic control and isolation functions.          instrument channels of the same process variable.
: 3. Each MSIV can be subjected to a partial          3. Tests will be conducted by actuating each          3. When the test switch is actuated, the MSIV closure test from the main control room.            MSIV test switch.                                    partially closes and then reopens automatically.
25A5675AA Revision 7
: 4. LDS provides separate manual controls in          4. Tests will be performed on the as-built            4. Upon manual actuation, the following actions the main control room for MSIV closure, for          system as follows:                                    occur:
isolation of the primary and secondary
: a. Simultaneously actuate any two of the              a. Closure of all the MSIVs occurs only containment, and for isolation of the RCIC four MSIV isolation switches to close all            when any two out of four switches are System.
MSIVs.                                                actuated.
: b. Actuate each RCIC isolation switch                b. Isolation of the RCIC System occurs (Divisions I and II) to isolate the RCIC              when Division I switch closes the inboard System.                                              or Division II switch closes the outboard isolation valves.
: c. Actuate each primary and secondary Design Control Document/Tier 1 containment isolation switch (Divisions I,      c. Each divisional primary and secondary II and III) to isolate the containment.              containment isolation switch closes only its respective containment isolation valves.
: 5. Manual reset controls are provided to            5. Tests will be performed using the LDS reset        5. The logic circuitry resets for LDS operation.
perform reset functions as described in              controls.
Section 2.4.3.
: 6. LDS design is fail-safe in the event of loss of 6. Tests will be conducted by disconnecting            6. Upon loss of electrical power to one division electrical power to one division of LDS logic. electrical power to one division of LDS logic          of LDS logic, the affected LDS divisional at a time.                                              channel trips.
2.4-37
 
Table 2.4.3 Leak Detection & Isolation System (Continued)
ABWR 2.4-38 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 7. Each of the four LDS divisional logic        7.                                                7.
channels and associated sensors is powered
: a. Tests will be performed on the LDS by            a. The test signal exists only in the Class from its respective divisional Class 1E power providing a test signal to only one Class          1E division under test in the LDS.
supply. In the LDS, independence is 1E division at a time.
provided between Class 1E divisions, and between Class 1E divisions and non-Class        b. Inspection of the as-installed Class 1E          b. In the LDS, physical separation or 1E equipment.                                      divisions in the LDS will be performed.            electrical isolation exists between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 8. Main control room displays and controls      8. Inspections will be performed on the main      8. Displays and controls exist or can be 25A5675AA Revision 7 provided for the LDS are as defined in          control room displays and controls for the        retrieved in the main control room as defined Section 2.4.3.                                  LDS.                                              in Section 2.4.3.
Design Control Document/Tier 1 Leak Detection and Isolation System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.4.4 Reactor Core Isolation Cooling System Design Description The Reactor Core Isolation Cooling (RCIC) System consists of a turbine, pump, piping, valves, controls and instrumentation. The RCIC turbine is driven by the steam from the reactor pressure vessel (RPV) which then drives the RCIC pump. The function of the RCIC System is to provide makeup water to the RPV.
The RCIC steam supply to the turbine branches off one of the main steamlines inside containment upstream of the inboard MSIV and exhausts to the suppression pool (S/P). The primary source of RCIC pump suction is the Condensate Storage Tank (CST). The suppression pool is the secondary source of RCIC pump suction. Figure 2.4.4a shows the basic system configuration and scope. Figure 2.4.4b shows RCIC System control interfaces.
The RCIC System shown on Figure 2.4.4a is classified as safety-related.
The RCIC System operates in the following modes:
(1)    RPV water makeup.
(2)    Full flow test.
(3)    Minimum flow bypass.
RPV Water Makeup Mode As shown on Figure 2.4.4b, the RCIC System channel measurements are provided to the Safety System Logic and Control (SSLC) System for signal processing, setpoint comparisons, and generating trip signals. The RCIC System is automatically initiated when either a high drywell pressure or low reactor water level condition exists. RCIC System is actuated at a reactor water level higher than the High Pressure Core Flooder (HPCF) system actuation level. The SSLC processors use a two-out-of-four voting logic for system initiation and shutdown. Manual RCIC System initiation can be performed from the main control room (MCR). The RCIC System can be started by local operation of RCIC System components outside the MCR.
The RCIC System automatically shuts down when a high reactor water level condition exists.
Following RCIC shutdown on high reactor water level signal, the RCIC System automatically restarts to provide RPV water makeup, if the low reactor water level initiation signal recurs.
During this mode, the primary source pump suction is the CST. Automatic transfer of pump suction from the CST to the S/P occurs when a low CST water level or a high suppression pool water level signal exists. This transfer can be manually overridden from the MCR. The CST and S/P water level signals are processed through SSLCs two-out-of four voting logic to initiate suction transfer.
Reactor Core Isolation Cooling System                                                                    2.4-39
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 In the RPV water makeup mode, the RCIC pump delivers a flow rate of at least 182 m3/h against a maximum differential pressure (between the RPV and the suction source) of 8.12 MPa. This flow rate is achieved within 29 seconds of receipt of the system initiation signal. The RCIC pump has sufficient net positive suction head (NPSH) available at the pump.
The RCIC System operates for a period of at least 2 hours under conditions of no AC power availability and no other simultaneous failures, accidents or other design basis conditions.
The RCIC system is capable of injecting sufficient water to the vessel to maintain core cooling with suction aligned to the suppression pool and a suction temperature of 121°C (250°F) during beyond design basis events (e.g. Extended Station Blackout).
Full Flow Test Mode The RCIC System has a full flow test mode to permit pump flow testing during plant operation.
During the test, water is pumped from the suppression pool and returned to the suppression pool via the test return line. The vessel injection valve is kept closed.
If a system initiation signal occurs during the full flow test mode, the RCIC System automatically aligns to the RPV water makeup mode.
Minimum Flow Bypass Mode The RCIC System has a minimum flow bypass mode that assures there is always flow in the RCIC pump when it is operating. This is accomplished automatically by monitoring pump discharge flow, and opening a minimum flow valve to the suppression pool when flow falls below minimum value. The minimum flow valve closes when the pump flow exceeds the minimum value. Minimum flow bypass operation is automatic based on a flow signal opening the minimum flow valve when the flow is low, with a concurrent high pump discharge pressure signal.
Other Provisions The RCIC System shown on Figure 2.4.4a is classified as Seismic Category I. Figure 2.4.4a shows the ASME Code class for the RCIC System. The RCIC System is located inside primary containment and in the Reactor Building.
As shown on Figure 2.2.4a, the RCIC System components are powered from Class 1E Division I, except for the steam supply outboard containment isolation valve, which is powered from Class 1E Division II. All RCIC System components shown on Figure 2.2.4a except the inboard containment isolation valves are powered from DC sources. In the RCIC System, independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
Outside the primary containment, except for the piping from the CST, the RCIC System shown on Figure 2.4.4a is physically separated from the two divisions of the High Pressure Core Flooder (HPCF) System.
2.4-40                                                                    Reactor Core Isolation Cooling System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 The RCIC System has the following displays and controls in the main control room (MCR):
(1)    Parameter displays for the instruments shown on Figure 2.4.4a.
(2)    Controls and status indication for the active safety-related components shown on Figure 2.4.4a.
(3)    Manual system level initiation capability for RPV water makeup mode.
(4)    Manual override of the automatic CST to S/P suction transfer.
The safety-related electrical components (including instrumentation and control) shown on Figure 2.4.4a located inside primary containment and in the Reactor Building are qualified for a harsh environment.
The motor-operated valves (MOVs) shown on Figure 2.4.4a have active safety-related functions to open, close, or both open and close, and performs these functions under differential pressure, fluid flow, and temperature conditions.
The check valves (CVs) shown on Figure 2.4.4a have active safety-related functions to open, close, or both open and close under system pressure, fluid flow, and temperature conditions.
The RCIC turbine is tripped if a low pump suction pressure condition is present.
The following RCIC System components:
(1)    Piping and components from the pump suction MOVs up to the pump inlet, (2)    Barometric condenser and associated equipment have a design pressure of 2.82 MPaG for intersystem loss-of-coolant accident (ISLOCA) conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.4.4 provides a definition of the inspections, test and/or analyses, together with associated acceptance criteria, which will be undertaken for the RCIC System.
Reactor Core Isolation Cooling System                                                                      2.4-41
 
ABWR 2.4-42                                                                              PRIMARY MUWC        RCIC CONTAINMENT 2
MAIN STEAM M
NBS LINE RPV    1 RCIC                                                                                                    TO FEEDWATER M                    M 2 NBS    LINE RCIC 1 2          M            M FE M
P 2
M                                          NNS M
25A5675AA Revision 7 Note 1      TURBINE M                                        NOTE 1 NNS P
2 TO RHR A TEST RETURN 2 RHR    LINE M                                                      M RCIC SUPPRESSION POOL          S                                                                                    FROM CST (MUWC)
Design Control Document/Tier 1 2
Reactor Core Isolation Cooling System RCIC  HPCF NOTES:
: 1. RCIC TURBINE IS NOT COVERED BY ASME CODE SECTION III BUT IS DESIGNED, FABRICATED, AND INSTALLED TO SAFETY-RELATED STANDARDS AND IS CONSISTENT WITH THE ASME CODE SECTION III REQUIREMENTS. THE TURBINE INCLUDES THE TURBINE TRIP AND THROTTLE VALVE.
: 2. ALL RCIC SYSTEM COMPONENTS SHOWN ON THIS FIGURE ARE POWERED FROM CLASS 1E DIVISION I EXCEPT FOR THE OUTBOARD CONTAINMENT ISOLATION VALVE () WHICH IS CLASS 1E DIVISION II.
: 3. ALL RCIC SYSTEM COMPONENTS SHOWN ON THIS FIGURE EXCEPT THE INBOARD CONTAINMENT ISOLATION VALVES, ARE POWERED FROM DC SOURCES.
Figure 2.4.4a Reactor Core Isolation Cooling System
 
ABWR Reactor Core Isolation Cooling System LOCAL AREA                                      MAIN CONTROL ROOM                        LOCAL AREA Plant Sensors                                                                          Device Actuators RCIC                RCIC Manual Manual System          Pump and Valve Initiation            Controls System Flow Pump Discharge Pressure RCIC    Pump Suction Pressure CST Suction Valve Position                    SSLC PROCESSING S/P Suction Valve Position 25A5675AA Revision 7 RCIC SYSTEM LOGIC                    RCIC Automatic  and Manual System Drywell Pressure                                                                      Initiation and Control NBS    Reactor Water Level                      - Sensor Channel Trip Decision
                                                                                            - System Coincidence Trip Decision
                                                                                            - Control and Interlock Logic
                                                                                            - Division-of-Sensors Bypass                Manual Pump and Valve RCIC Actuation
                                                                                            - Calibration, Self-Diagnosis MUWC CST Water Level ACS    S/P Water Level Design Control Document/Tier 1 Notes:
: 1. See Section 3.4, Figure 3.4b for SSLC Processing.
: 2. The inboard steam supply isolation valve has diverse actuation and status indication. See Section 3.4.
Figure 2.4.4b Reactor Core Isolation Cooling System Control Interface Diagram 2.4-43
 
ABWR 2.4-44 Table 2.4.4 Reactor Core Isolation Cooling System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                              Acceptance Criteria
: 1. The basic configuration of the RCIC System          1. Inspections of the as-built system will be        1. The as-built RCIC System conforms with the is as shown on Figures 2.4.4a and 2.4.4b.              conducted.                                            basic configuration shown on Figures 2.4.4a and 2.4.4b.
: 2. The ASME Code components of the RCIC            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the RCIC System              ASME Code components of the RCIC integrity under internal pressures that will be    required to be hydrostatically tested by the    System conform with the requirements in the experienced during service.                        ASME Code.                                      ASME Code Section III.
: 3.                                                    3.                                                    3.
: a. The RCIC System is automatically                    a. Tests will be conducted using simulated            a. The RCIC System receives an initiation initiated in the RPV water makeup mode                input signals for each process variable to            signal.
25A5675AA Revision 7 when either a high drywell pressure or a              cause trip conditions in two, three, and low reactor water level condition exists.              four instrument channels of the same process variable.
: b. Manual RCIC System initiation can be                b. Tests will be conducted by manually                b. The RCIC System receives an initiation performed.                                            initiating RCIC System.                              signal.
: c. Following receipt of an initiation signal,        c. Tests will be conducted on the RCIC              c. Upon receipt of a simulated initiation the RCIC System automatically initiates                System using simulated initiation signal.            signal, the following occurs:
and operates in the RPV water makeup (1) Steam supply bypass valve receives mode.
open signal.
Design Control Document/Tier 1 (2) Test return valves receive close Reactor Core Isolation Cooling System signal.
(3) CST suction valve receives open signal.
(4) Injection valve receives open signal after a 10-second time delay.
(5) Steam admission valve receives open signal after a 10-second time delay.
 
Table 2.4.4 Reactor Core Isolation Cooling System (Continued)
ABWR Reactor Core Isolation Cooling System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                        Acceptance Criteria
: 3. continued                                      3. continued                                      3. continued
: d. The RCIC System automatically shuts            d. Tests will be conducted using simulated        d. RCIC System receives shutdown signal.
down when a high reactor water level              high reactor water level signals to cause condition exists.                                  trip conditions in two, three, and four instrument channels of water level variable.
: e. Following receipt of shutdown signal, the        e. Tests will be conducted on RCIC System      e. Upon receipt of simulated shutdown RCIC System automatically terminates                using simulated shutdown signal.                signals, the following occurs:
the RPV water makeup mode.
(1) Steam supply bypass valve receives close signal.
(2) RCIC initiation logic resets.
25A5675AA Revision 7 (3) Injection valve receives close signal.
(4) Steam admission valve receives close signal.
: f. Following RCIC shutdown on high reactor        f. Tests will be conducted using simulated    f. Upon receipt of simulated low reactor water level signal, the RCIC System                low reactor water level signals.                water level signals, the following occurs:
automatically restarts to provide RPV (1) Steam supply bypass valve receives water makeup if low reactor water level open signal.
signal recurs.
(2) Test return valves receive close Design Control Document/Tier 1 signal.
(3) CST suction valve receives open signal.
(4) Injection valve receives open signal after a 10 second time delay.
(5) Steam admission valve receives open signal after a 10 second time delay.
2.4-45
 
Table 2.4.4 Reactor Core Isolation Cooling System (Continued)
ABWR 2.4-46 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                              Acceptance Criteria
: 3. continued                                    3. continued                                        3. continued
: g. The RCIC System automatically initiates      g. Tests will be conducted using simulated          g. The RCIC System receives suction suction transfer from the CST to the            input signals for each process variable to          transfer initiation signal.
suppression pool when either a low CST          cause trip conditions in two, three, and water level or a high suppression pool          four instrument channels of the same water level exists.                            process variable.
: h. Following receipt of suction transfer          h. Tests will be conducted using simulated          h. Upon receipt of simulated suction transfer initiation signal, the RCIC System                suction transfer initiation signals.                initiation signals, the following occurs:
automatically switches pump suction.
(1) Suppression pool suction valve This transfer can be manually overridden opens.
from the MCR.
(2) CST suction valve closes. The 25A5675AA Revision 7 suction transfer can be manually overridden from the MCR.
: i. In the RPV water makeup mode, the            i. Tests will be conducted in a test facility on  i.
RCIC pump delivers a flow rate of at least        the RCIC System pump and turbine.
(1) The RCIC pump delivers a flow rate 182 m3/h against a maximum differential of at least 182 m3/h against a pressure (between the RPV and the maximum differential pressure pump suction) of 8.12 MPa.
(between the RPV and the pump suction) of 8.12 MPa.
(2) The RCIC turbine delivers the speed Design Control Document/Tier 1 and torque required by the pump at Reactor Core Isolation Cooling System the above conditions.
 
Table 2.4.4 Reactor Core Isolation Cooling System (Continued)
ABWR Reactor Core Isolation Cooling System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                            Acceptance Criteria
: 3. continued                                3. continued                                        3. continued
: j. The RCIC System pump has sufficient      j. Inspections, tests, and analyses will be        j. The available NPSH exceeds the NPSH.                                        performed based upon the as-built                  required NPSH required by the pump.
system. Inspections of the as-built                Test result/report confirms that the RCIC system will be performed to obtain piping          valves and RCIC pumps perform their system dimensions and other necessary              intended functions during post-LOCA information. The required NPSH of                  operation for a minimum of 12 hours.
procured pumps will be determined by an inspection of the vendor specifications.
The analysis will consider the effects of:
(1) Pressure losses for pump inlet piping and components.
25A5675AA Revision 7 (2) Suction from suppression pool with water level at the minimum value.
(3) Analytically derived values for blockage of pump suction strainers based upon the as-built system.
(4) Design basis debris loading of pumped fluid under conditions ranging from normal operating to design basis accident conditions.
Design Control Document/Tier 1 (5) Design basis fluid temperature (77°C).
(6) Containment at atmospheric pressure.
(7) Confirm vertical and horizontal separation between the SRV Quencher and RCIC Suction Strainer.
2.4-47
 
Table 2.4.4 Reactor Core Isolation Cooling System (Continued)
ABWR 2.4-48 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 3. continued                                        3. continued                                    3. continued
: k. The RCIC System operates for a period            k. Inspections and analyses of the as-built      k. The RCIC System can operate for a of at least 2 hours under conditions of no          RCIC and supporting systems will be              period of at least 2 hours under AC power availability and no other                  performed to determine RCIC capability.          conditions of no AC power availability and simultaneous failures, accidents, or other                                                          no other simultaneous failures, accidents, design basis conditions.                                                                            or other design basis conditions.
: l. The RCIC can be started by local              l. Tests will be conducted locally on RCIC      l. RCIC System components required for operation of the RCIC System                      System components required for system              system operation can be actuated locally.
components outside the MCR.                        operation.
: 4. If a system initiation signal occurs during the 4. Test will be conducted using simulated          4. The RCIC System automatically aligns to full flow test mode, the RCIC System              initiation signals.                                RPV water makeup mode from test mode 25A5675AA Revision 7 automatically aligns to the RPV water                                                                upon receipt of an initiation signal.
makeup mode.
: 5. The RCIC System has a minimum flow              5. Tests will be conducted on the pump            5. The pump minimum flow valve receives a bypass mode that assures there is always          minimum flow valve interlock logic using          signal to open when signals indicative of the flow in the RCIC pump when it is operating.        simulated pressure and flow signals.              following conditions exist concurrently:
: a. Pump discharge pressure is high when the pump starts.
: b. Pump flow is low.
Design Control Document/Tier 1 The pump minimum flow valve receives a signal to close when a signal indicative of the Reactor Core Isolation Cooling System following condition exists:
: a. Pump flow exceeds minimum value.
 
Table 2.4.4 Reactor Core Isolation Cooling System (Continued)
ABWR Reactor Core Isolation Cooling System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria
: 6. As shown on Figure 2.4.4a, the RCIC System 6.                                              6.
components are powered from Class 1E
: a. Tests will be performed in the RCIC          a. The test signal exists only in the Class 1E division I, except for the steam supply System by providing a test signal in only      division under test in the RCIC System.
outboard containment isolation valve which is one Class 1E division at a time.
powered from Class 1E division II. All RCIC                                                  b. In the RCIC System physical separation System components shown on Figure 2.4.4a      b. Inspections of the as-built Class 1E            or electrical isolation exists between except the inboard containment isolation        divisions in the RCIC System will be            Class 1E divisions in the RCIC System.
valves are powered from DC sources. In the      performed.                                      Physical separation or electrical isolation RCIC System, independence is provided                                                            exists between Class 1E divisions and between Class 1E Divisions and between                                                          non-Class 1E equipment.
Class 1E Divisions and non-Class 1E equipment.
25A5675AA Revision 7
: 7. Outside the primary containment, except for 7. Inspections of the as-installed RCIC System 7. Outside the primary containment, except for the piping from the CST, the RCIC System      will be performed.                            the piping from the CST, the RCIC System shown on Figure 2.4.4a, is physically                                                        shown on Figure 2.4.4a, is physically separated from the two divisions of the HPCF                                                  separated from the two divisions of the HPCF System.                                                                                      System by structural and/or fire barriers.
: 8. Main control room displays and controls      8. Inspections will be performed on the main    8. Displays and controls exist or can be provided for RCIC System are as defined in      control room displays and controls for the      retrieved in the main control room as defined Section 2.4.4.                                  RCIC System.                                    in Section 2.4.4.
Design Control Document/Tier 1 2.4-49
 
Table 2.4.4 Reactor Core Isolation Cooling System (Continued)
ABWR 2.4-50 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                                Acceptance Criteria
: 9.                                                  9.                                                  9.
: a. MOVs designated in Section 2.4.4 as              a. Tests of installed valves for opening,            a. Upon receipt of the actuating signal, each having active safety-related function              closing, or both opening and closing will            MOVs opens, closes, or both opens and open, close or both open and close under            be conducted under pre-operational                  closes, depending upon the valves differential pressure, fluid flow, and              differential pressure, fluid flow, and              safety functions. The following valves temperature conditions.                            temperature conditions.                              open, or close, in the following time limits:
Valve              Time Steam Supply        < 30 s Close Containment 25A5675AA Revision 7 Isolation Valves Injection Valve    < 15 s Open
: b. CVs designated in Section 2.4.4 as              b. Tests of installed valves for opening,            b. Based on the direction of the differential having an active safety-related function            closing, or both opening and closing, will          pressure across the valve, each CV open, close, or both open and close,                be conducted under system                            opens, closes, or both opens and closes, under system pressure, fluid flow, and              preoperational pressure, fluid flow, and            depending upon the valves safety temperature conditions.                            temperature conditions.                              functions.
Design Control Document/Tier 1
: 10. The RCIC turbine is tripped if low suction      10. Test will be conducted using a simulated low 10. The turbine trip and throttle valve receives a pressure condition is present.                      suction pressure signal.                        trip signal.
Reactor Core Isolation Cooling System
: 11. The RCIC system has the capability of        11. Analyses will be performed of the as-built  11. The RCIC system is capable of injecting injecting sufficient water to the vessel to      RCIC System to assess the system capability    sufficient water to the vessel to maintain core maintain core cooling with suction aligned to    with 121&deg;C water at the pump suction.          cooling with suction aligned to the the suppression pool and a suction                                                                suppression pool and a suction temperature temperature of 121&deg;C (250&deg;F) during beyond                                                        of 121&deg;C (250&deg;F) during beyond design basis design basis events (e.g. Extended Station                                                        events (e.g. Extended Station Blackout).
Blackout).
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.5.1 Fuel Servicing Equipment No entry for this system.
Fuel Servicing Equipment                                                                2.5-1
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.5.2 Miscellaneous Servicing Equipment No entry for this system.
2.5-2                                                        Miscellaneous Servicing Equipment
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 2.5.3 Reactor Pressure Vessel Servicing Equipment No entry for this system.
Reactor Pressure Vessel Servicing Equipment                                                2.5-3
 
25A5675AA Revision 7 ABWR                                                  Design Control Document/Tier 1 2.5.4 RPV Internal Servicing Equipment No entry for this system.
2.5-4                                                        RPV Internal Servicing Equipment
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.5.5 Refueling Equipment Design Description The Reactor Building is supplied with a refueling machine for fuel movement and servicing plus an auxiliary platform for servicing operations from the vessel flange level.
The refueling machine is a gantry crane, which spans the reactor vessel and the storage pools on bedded tracks in the refueling floor. A telescoping mast and grapple suspended from a trolley system is used to lift and orient fuel bundles for placement in the core and/or storage racks. Two auxiliary hoists, one main and one auxiliary monorail trolley-mounted, are provided for in-core servicing. Control of the machine is from an operator station on the refueling floor.
The refueling machine is classified as non-safety-related.
A position indicating system and travel limit computer are provided to locate the grapple over the vessel core and prevent collision with pool obstacles. The mast grapple has a redundant load path so that no single component failure results in a fuel bundle drop. Interlocks on the machine:
(1) prevent hoisting a fuel bundle over the vessel unless an all-control-rod-in permissive is present; (2) limit vertical travel of the fuel grapple to provide shielding over the grappled fuel during transit; (3) prevent lifting of fuel without grapple hook engagement and load engagement.
The refueling machine is classified as Seismic Category I.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.5.5 provides a definition of the inspection, test, and/or analyses, together with associated acceptance criteria, which will be undertaken for the refueling machine. No entries are proposed for the auxiliary platform.
Refueling Equipment                                                                                          2.5-5
 
ABWR 2.5-6                                                                          Table 2.5.5 Refueling Equipment Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the refueling          1. Inspections of the as-built refueling machine 1. The as-built refueling machine conforms with machine is described in Section 2.5.5.                will be conducted.                              the basic configuration described in Section 2.5.5.
: 2. Interlocks on the machine:                        2. Tests will be conducted on the as-built      2. Interlocks on the machine:
: a. Prevent hoisting a fuel bundle over the          refueling machine using simulated signals        a. Prevent hoisting a fuel bundle over the vessel unless an all-control-rod-in              and loads.                                          vessel unless an all-control-rod-in permissive is present.                                                                                permissive is present.
: b. Limit vertical travel of the fuel grapple to                                                      b. Limit vertical travel of the fuel grapple to provide shielding over the grappled fuel                                                              provide shielding over the grappled fuel during transit.                                                                                      during transit.
25A5675AA Revision 7
: c. Prevent lifting of fuel without grapple                                                          c. Prevent lifting of fuel without grapple hook engagement and load                                                                              hook engagement and load engagement.                                                                                          engagement.
Design Control Document/Tier 1 Refueling Equipment
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.5.6 Fuel Storage Facility Design Description The Fuel Storage Facility provides storage racks for the temporary and long-term storage of new and spent fuel and associated equipment. The fuel storage racks are designed to prevent inadvertent criticality.
The racks are classified as non-safety-related.
Racks provide storage for new fuel and spent fuel in the spent fuel storage pool in the Reactor Building. The racks are top loading, with fuel bail extended above the rack. The fuel storage racks have a minimum storage capacity of 270% of the reactor core, which is equivalent to a minimum of 2354 fuel storage positions. The fuel storage racks maintain a subcriticality of at least 5% k under normal and abnormal conditions. The rack arrangement prevents accidental insertion of fuel assemblies between adjacent racks and allows flow to prevent the water from exceeding 100&deg;C.
The racks are classified as Seismic Category I.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.5.6 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the fuel storage racks.
Fuel Storage Facility                                                                                    2.5-7
 
ABWR 2.5-8                                                                          Table 2.5.6 Fuel Storage Facility Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the new and spent 1. Inspections of the as-built system will be          1. The as-built new and spent fuel storage fuel racks is described in Section 2.5.6.      conducted                                              racks conform with the basic configuration described in Section 2.5.6.
: 2. The new and spent fuel racks maintain a        2. Analyses will be performed to determine the      2. An analysis report exists which concludes subcriticality of at least 5% k under dry or      keff of the as-built new and spent fuel racks.      that the new and spent fuel racks have a flooded conditions.                                                                                    subcriticality of at least 5%k under dry or flooded conditions.
: 3. The rack arrangement prevents accidental        3. Inspections of the as-built new and spent fuel 3. The rack arrangement prevents accidental insertion of fuel assemblies between              racks will be performed.                          insertion of fuel assemblies between adjacent racks.                                                                                      adjacent racks.
25A5675AA Revision 7
: 4. The rack arrangement allows flow to prevent 4. An analysis of the as-built spent fuel rack will 4. An analysis report exists which concludes the water from exceeding 100&deg;C.                be performed to determine the maximum              that the rack arrangement allows flow to water temperature.                                  prevent the water from exceeding 100&deg;C.
Design Control Document/Tier 1 Fuel Storage Facility
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.5.7 Under-Vessel Servicing Equipment No entry for this system.
Under-Vessel Servicing Equipment                                                      2.5-9
 
25A5675AA Revision 7 ABWR                                                  Design Control Document/Tier 1 2.5.8 CRD Maintenance Facility No entry for this system.
2.5-10                                                            CRD Maintenance Facility
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.5.9 Internal Pump Maintenance Facility No entry for this system.
Internal Pump Maintenance Facility                                                    2.5-11
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.5.10 Fuel Cask Cleaning Facility No entry for this system.
2.5-12                                                            Fuel Cask Cleaning Facility
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.5.11 Plant Start-up Test Equipment No entry for this system.
Plant Start-up Test Equipment                                                          2.5-13
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.5.12 Inservice Inspection Equipment No entry for this system.
2.5-14                                                          Inservice Inspection Equipment
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.6.1 Reactor Water Cleanup System Design Description The Reactor Water Cleanup (CUW) System removes particulate and dissolved impurities from the reactor coolant by circulating a portion of the reactor coolant through a filter-demineralizer.
The CUW System removes excess coolant from the reactor system during startup, shutdown and hot standby. The excess water is directed to the radwaste or main condenser. The CUW System also provides processed water to the reactor head spray nozzle for Reactor Pressure Vessel (RPV) cooldown.
The CUW System reduces RPV temperature gradients by maintaining circulation in the bottom head of the RPV during periods when the reactor internal pumps are unavailable.
Figure 2.6.1 shows the basic CUW System configuration and scope. Except for the primary containment penetration and isolation valves, the CUW System is classified as non-safety-related. The major portion of the system is located outside of the primary containment in the Reactor Building.
CUW System piping and components from the RPV, out to and including the outboard isolation valves, are part of the reactor coolant pressure boundary and are classified as Seismic Category I. The remainder of the piping system is classified as non-Seismic Category I. Figure 2.6.1 shows the ASME Code class for the CUW system components The inboard containment isolation valve is powered from Class 1E Division II, and the outboard containment isolation valves are powered from Class 1E Division I. In the CUW System, independence is provided between the Class 1E divisions, and between the Class 1E divisions and non-Class 1E equipment.
The main control room has control and open/close status indication for the containment isolation valves.
The safety-related electrical equipment, located in the primary containment and Reactor Building is qualified for a harsh environment.
The motor-operated valves (MOVs) for containment isolation shown in Figure 2.6.1 have active safety-related functions to close and perform this function under differential pressure, fluid flow and temperature conditions.
The check valves (CVs) shown on Figure 2.6.1, have active safety-related function to close under system pressure, fluid flow, and temperature conditions.
The CUW suction line is provided with a flow restrictor which provides flow restricting and flow monitoring functions. Maximum throat diameter is 135 mm.
Reactor Water Cleanup System                                                                                2.6-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 The CUW piping and components downstream of the blowdown valve leading towards the Radwaste System shown on Figure 2.6.1 have a design pressure of 2.82 MPaG for intersystem loss-of-coolant accident (ISLOCA) conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.6.1 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the CUW System.
2.6-2                                                                        Reactor Water Cleanup System
 
ABWR Reactor Water Cleanup System RPV CUW                                      M    NOTE 1 1
INSIDE 1 3    STEAM TUNNEL          M          FE RHR LOOP B                        FEEDWATER RPV                                              NBS CUW RHR  NOTE 2    NOTE 1        NOTE 1  3 RCW 1 CUW    M          M            M                                  CUW 3 RHX NRHX RPV                                              1 3 FE                                                          CUW 3 CUW 1 RCW 25A5675AA Revision 7 CONTAINMENT NRHX      RCW 3 CUW CUW 3 RCW NNS 3 FILTER NNS    M                          DEMINERALIZERS RW CUW Design Control Document/Tier 1 FE M
3                                        M MC CUW NOTES:
: 1. THE INBOARD CONTAINMENT ISOLATIONVALVE IS POWERED FROM CLASS 1E DIVISION II AND THE OUTBOARD ISOLATION VALVES ARE POWERED FROM CLASS 1E DIVISION I.
: 2. NOT CONSIDERED A CONTAINMENT ISOLATION VALVE; CLASSIFIED AS NON-SAFETY-RELATED.
2.6-3 Figure 2.6.1 Reactor Water Cleanup System
 
ABWR 2.6-4                                                                          Table 2.6.1 Reactor Water Cleanup System Inspections, Tests, Analyses and Acceptance Criteria Design Commitments                              Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration for the CUW System 1. Inspection of the as-built system will be            1. The as-built CUW System conforms with the is as shown in Figure 2.6.1.                  conducted.                                              basic configuration shown in Figure 2.6.1.
: 2. The ASME Code components of the CUW            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the CUW System                ASME Code components of the CUW integrity under internal pressures that will be    required to be hydrostatically tested by the    System conform with the requirements in the experienced during service.                        ASME Code.                                      ASME Code, Section III.
: 3. The inboard containment isolation valve is    3.                                                  3.
powered from Class 1E Division II, and the
: a. Tests will be performed on the CUW              a. The test signal exists only in the Class outboard containment isolation valves are System by providing a test signal in only          1E division under test in the CUW powered from Class 1E Division I. In the one Class 1E division at a time.                    System.
CUW System, independence is provided 25A5675AA Revision 7 between Class 1E divisions, and between            b. Inspections of the as-installed Class 1E        b. In the CUW System, physical separation Class 1E divisions and non-Class 1E                    divisions in the CUW System will be                or electrical isolation exists between equipment.                                            performed.                                          Class 1E divisions. Physical separation or electrical isolation exists between Class 1E divisions and non-Class 1E equipment.
: 4. Main control room displays and controls        4. Inspections will be performed on the main        4. Displays and controls exist or can be provided for CUW System are as defined in        control room displays and controls for the          retrieved in main control room as defined in Section 2.6.1.                                    CUW System.                                        Section 2.6.1.
Design Control Document/Tier 1 Reactor Water Cleanup System
 
Table 2.6.1 Reactor Water Cleanup System (Continued)
ABWR Reactor Water Cleanup System Inspections, Tests, Analyses and Acceptance Criteria Design Commitments                              Inspections, Tests, Analyses                                Acceptance Criteria
: 5.                                                  5.                                                  5.
: a. MOVs designated in Section 2.6.1 as              a. Tests of installed valves for closing will        a. Upon receipt of the actuation signal each having an active safety- related function          be conducted under preoperational                    MOV closes. The following valves close close under differential pressure and              differential pressure, fluid flow, and              in the following time limits:
fluid flow and temperature conditions.              temperature conditions.
Valve                        Time (s)
Suction line inboard          30 Close containment isolation valve 25A5675AA Revision 7 Suction line outboard        30 Close containment isolation valve
: b. CVs designated in Section 2.6.1 as              b. Tests of installed valves for closing will        b. Each CV closes.
having an active safety-related function            be conducted under system pre-close under system pressure, fluid flow,            operational pressure, fluid flow, and and temperature conditions.                        temperature conditions.
: 6. Maximum throat diameter of the CUW              6. Inspections will be performed on the CUW          6. Maximum throat diameter of the CUW suction line flow restrictor is 135 mm.            suction line flow restrictor throat diameter.        suction line flow restrictor is 135 mm.
Design Control Document/Tier 1
: 7. RPV Head Spray line will have a high point      7. Inspections will be performed on the as built 7. RPV Head Spray line will have a high point vent line with the proper slope to prevent          CUW piping to confirm proper elevation and      vent line with the proper slope to prevent buildup of Hydrogen Gas during operation.          slope.                                          buildup of Hydrogen Gas during operation.
2.6-5
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.6.2 Fuel Pool Cooling and Cleanup System Design Description The Fuel Pool Cooling and Cleanup (FPC) System (Figure 2.6.2) removes decay heat generated by the spent fuel assemblies in the spent fuel storage pool. The system also maintains the water quality and monitors and maintains the water level above the spent fuel in the spent fuel storage pool. Figure 2.6.2 shows the basic FPC System configuration and scope.
The FPC System is classified non-safety-related, except for piping connections and valves for safety-related fuel pool makeup and supplemental cooling by the Residual Heat Removal (RHR) System.
The safety-related makeup water source for the spent fuel storage pool is provided by the RHR System, which pumps suppression pool water to the FPC System.
The spent fuel storage pool has no piping connections (inlet, outlet, drains or other piping) located below a point 3m above the top of active fuel located in the spent fuel storage racks.
The FPC System components, with the exception of the filter/demineralizer unit, are classified as Seismic Category I. Figure 2.6.2 shows the ASME Code class for the FPC System piping and components.
The FPC System is located in the Reactor Building.
The FPC System has non-safety parameter displays in the main control room for instruments shown on Figure 2.6.2 (for example, narrow range water level). In addition, two safety-related spent fuel pool wide range level instruments are provided. Indication of the spent fuel pool level is provided in the Main Control Room (MCR) as well as in another appropriate area that is accessible post-accident.
The check valves (CVs) shown on Figure 2.6.2 have active safety-related functions to open, close, or both open and close under system pressure, fluid flow, and temperature conditions.
The piping and components of the FPC System at the suction side of the RHR System from the upstream isolation valve have a design pressure of 2.82 MPaG for intersystem LOCA (ISLOCA) conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.6.2 provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria, which will be undertaken for the FPC System.
2.6-6                                                                    Fuel Pool Cooling and Cleanup System
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 Figure 2.6.2 Fuel Pool Cooling and Cleanup System Fuel Pool Cooling and Cleanup System                                                                    2.6-7
 
ABWR 2.6-8                                                                            Table 2.6.2 Fuel Pool Cooling and Cleanup System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the FPC System is 1. Inspection of the as-built system will be        1. The as-built FPC System conforms with the as shown on Figure 2.6.2.                      conducted.                                          basic configuration shown on Figure 2.6.2.
: 2. The ASME Code components of the FPC            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the FPC System                ASME Code components of the FPC System integrity under internal pressures that will be    required to be hydrostatically tested by the    conform with requirements in the ASME experienced during service.                        ASME Code.                                      Code, Section III.
: 3. The safety-related makeup water source for 3. Tests will conducted on the as-built FPC and 3. The combined RHR System and FPC the spent fuel storage pool is provided by the RHR Systems by aligning the systems so        System operation transfers water from RHR System, which pumps suppression            that the RHR System draws water from the      suppression pool to the spent fuel storage pool water to the FPC System.                  suppression pool and discharges into the      pool.
spent fuel storage pool.
25A5675AA Revision 7
: 4. The spent fuel storage pool has no piping      4. Inspections of the as-built spent fuel storage 4. The spent fuel storage pool has no piping connections (inlet, outlet, drains or other      pool will be conducted.                          connections (inlet, outlet, drains or other piping) located below a point 3m above the                                                          piping) located below a point 3m above the top of active fuel located in the spent fuel                                                        top of active fuel located in the spent fuel storage racks.                                                                                      storage racks.
: 5. Non-safety main control room displays      5. Inspections will be performed on the          5. Displays exist or can be retrieved in the main provided for the FPC System are as defined    non-safety main control room displays for the    control room as defined in Section 2.6.2.
in Section 2.6.2.                            FPC System.
Design Control Document/Tier 1
: 6. CVs designated in Section 2.6.2 as having      6. Tests of installed valves for opening, closing, 6. Based on the direction of the differential Fuel Pool Cooling and Cleanup System an active safety-related function open, close,    or both opening and closing, will be              pressure across the valve, each CV opens, or open and close, under system pressure,        conducted under system preoperational              close, or both opens and closes depending fluid flow, and temperature conditions.          pressure, fluid flow, and temperature              upon the valves safety functions.
conditions.
: 7. The safety related displays provided for the 7. Inspections will be performed of the safety      7. Displays exist or can be retrieved in both the FPC System spent fuel pool wide range          related FPC system displays in both the            main control room and an alternate location.
water level are as described in Section 2.6.2. main control room and at an alternate location.
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.6.3 Suppression Pool Cleanup System Design Description The Suppression Pool Cleanup (SPCU) System removes particulates and dissolved impurities from the suppression pool by circulating suppression pool water through the Fuel Pool Cooling (FPC) System water treatment equipment. The SPCU System also provides a source of makeup water to the spent fuel storage pool and the Reactor Building Cooling Water (RCW) System surge tanks using either the suppression pool or condensate storage tank (CST) water via the High Pressure Core Flooder (HPCF) System supply piping. Figure 2.6.3 shows the basic system configuration and scope.
Except for the primary containment penetration and isolation valves, the SPCU System is classified as non-safety-related.
The SPCU System piping and components, as shown on Figure 2.6.3, are classified as Seismic Category I. Figure 2.6.3 shows ASME Code class for the SPCU System piping and components.
The SPCU System is located outside the primary containment in the Reactor Building.
The inboard containment isolation valves are powered from Class 1E Division II, and the outboard containment isolation valve is powered from Class 1E Division I. In the SPCU System, independence is provided between the Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
The main control room has control and open/close status indication for the containment isolation valves.
The safety-related electrical equipment located outside the primary containment in the Reactor Building is qualified for a harsh environment.
The motor-operated valves (MOVs) for containment isolation, shown on Figure 2.6.3 have active safety-related function to close and perform this function under differential pressure, fluid flow, and temperature conditions.
The check valve (CV) for containment isolation shown on Figure 2.6.3, has active safety-related function to close under system pressure, fluid flow, and temperature conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.6.3 provides definition of inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the SPCU System.
Suppression Pool Cleanup System                                                                          2.6-9
 
ABWR 2.6-10                                                                                            TO SPENT FUEL STORAGE POOL        FPC SPCU 3
PRIMARY CONTAINMENT                                  TO RCW SURGE TANKS RCW SPCU 3
M                                            P FROM FPC FILTER-2 3                        SPCU FPC DEMINERALIZER 3
25A5675AA Revision 7 SUPPRESSION POOL 2 3 M          M P    TO FPC Design Control Document/Tier 1 M                                          FILTER-DEMINERALIZER FROM CST (MUWC)                                        SPCU FPC Suppression Pool Cleanup System HPCF SPCU                                3 3
NOTES:
: 1. THE INBOARD CONTAINMENT ISOLATION VALVES ARE POWERED FROM CLASS 1E DIVISION II AND THE OUTBOARD CONTAINMENT ISOLATION VALVE IS POWERED FROM CLASS 1E DIVISION I.
Figure 2.6.3 Suppression Pool Cleanup System
 
ABWR Suppression Pool Cleanup System Table 2.6.3 Suppression Pool Cleanup System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                                Acceptance Criteria
: 1. The basic configuration of the SPCU System 1. Inspections of the as-built system will be                1. The as-built SPCU System conforms with the is as shown on Figure 2.6.3.                  conducted.                                                    basic configuration shown on Figure 2.6.3.
: 2. The ASME Code components of the SPCU 2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary          Code components of the SPCU System          ASME Code components of the SPCU integrity under internal pressures that will be required to be hydrostatically tested by the System conform with the requirements in the experienced during service.                    ASME Code.                                  ASME Code Section III.
: 3. The inboard containment isolation valves are 3.                                              3.
powered from Class 1E Division II, and the
: a. Tests will be conducted in the SPCU          a. The test signal exists only in the Class outboard containment isolation valve is System by providing a test signal in only      1E division under test in the SPCU powered from Class 1E Division I. In the one Class 1E division at a time.                System.
SPCU System, independence is provided 25A5675AA Revision 7 between Class 1E divisions, and between        b. Inspections of the as-built Class 1E        b. In the SPCU System, physical Class 1E divisions and non-Class 1E                divisions in the SPCU System will be            separation or electrical isolation exists equipment.                                        performed.                                      between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 4. Main control room displays and controls          4. Inspections will be conducted on the main            4. Displays and controls exist or can be provided for the SPCU System are as                control room displays and controls for the              retrieved in the main control room as defined defined in Section 2.6.3.                          SPCU System.                                            in Section 2.6.3.
Design Control Document/Tier 1
: 5.                                                  5.                                                      5.
: a. MOVs designated in Section 2.6.3 as              a. Tests of installed valves for closing will          a. Upon receipt of the actuating signal, having an active safety-related function            be conducted under preoperational                      each MOV closes.
close under differential pressure, fluid            differential pressure, fluid flow, and flow, and temperature conditions.                  temperature conditions.
: b. The CV designated in Section 2.6.3 as            b. Tests of the installed valve for closing will        b. The CV closes.
having an active safety-related function            be conducted under system closes under system pressure, fluid flow,          preoperational pressure, fluid flow, and and temperature conditions.                        temperature conditions.
2.6-11
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.7.1 Main Control Room Panels Design Description The Main Control Room Panels (MCRP) consist of the main control console, the large display panel, the supervisors console, the auxiliary or back panels and their respective internal wiring.
The MCRP locates and configures the alarms displays and controls for plant systems. Those parts of the MCRP that contain Class 1E equipment are classified as Seismic Category I.
Non-Class 1E and divisional Class 1E control and instrument power is provided for the MCRP.
Independence is provided between Class 1E divisions and also between the Class 1E divisions and non-Class 1E equipment.
The MCRP has, as a minimum, the fixed alarms, displays, and controls shown on Table 2.7.1a.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.7.1a provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria, which will be undertaken for the MCRP.
Main Control Room Panels                                                                                    2.7-1
 
ABWR 2.7-2                                        Table 2.7.1a        Main Control Room Panels Fixed Position Alarms, Displays and Controls A. Fixed Position Controls Manual Scram Initiation Switch (A)              DG (A) Start Switch                  Div. I Manual/Auto Main Steamline Isolation Reset Switch Manual Scram Initiation Switch (B)              DG (B) Start Switch                  Div. II Manual/Auto Main Steamline Isolation Reset Switch Reactor Mode Switch                            DG (C) Start Switch                  Div. III Manual/Auto Main Steamline Isolation Reset Switch Div. I Main Steamline Manual                    RCIC System Standby Mode Initiation  Div. IV Manual/Auto Main Steamline Isolation Isolation Switch                                Switch                                Reset Switch Div. II Main Steamline Manual                  Condensate Pump Standby Mode          Primary Containment Div. I Isolation Reset Isolation Switch                                Initiation Switches                  Switch Div. III Main Steamline Manual                  Reactor Feedpump Standby Mode        Primary Containment Div. II Isolation Reset 25A5675AA Revision 7 Isolation Switch                                Initiation Switches                  Switch Div. IV Main Steamline Manual                  Condensate Pump Startup Mode          Primary Containment Div. III Isolation Reset Isolation Switch                                Initiation Switches                  Switch Primary Containment Div. I Manual              Reactor Feedpump Startup Mode        RHR (A) Shutdown Cooling Mode Initiation Isolation Switch                                Initiation Switches                  Switch Primary Containment Div. II Manual              SLC (A) Pump Control Switch          RHR (B) Shutdown Cooling Mode Initiation Isolation Switch                                                                      Switch Primary Containment Div. III Manual            SLC (B) Pump Control Switch          RHR (C) Shutdown Cooling Mode Initiation Design Control Document/Tier 1 Isolation Switch                                                                      Switch RCIC Initiation Switch                          ADS (A) Inhibit Switch                ARI (A) Manual Initiation Switch HPCF (B) Initiation Switch                      ADS (B) Inhibit Switch                ARI (B) Manual Initiation Switch Main Control Room Panels HPCF (C) Initiation Switch                      RHR (A) Standby Mode Switch          Recirculation Runback Initiation Switch (A)
RHR (A) Initiation Switch                      RHR (B) Standby Mode Switch          Recirculation Runback Initiation Switch (B)
RHR (B) Initiation Switch                      RHR (C) Standby Mode Switch          RIP Start/Stop Control Switch (10)
RHR (C) Initiation Switch                      Main Steam Isolation Valve            ARI (A) Logic Reset Switch Control Switch (8)
 
Table 2.7.1a        Main Control Room Panels Fixed Position Alarms, Displays and Controls (Continued)
ABWR Main Control Room Panels A. Fixed Position Controls (Continued)
ARI (B) Logic Reset Switch                    RHR (A) Suppression Pool Cooling Mode      Div. II ADS Manual ADS Channel 2 Initiation Initiation Switch                          Switch CRD Charging Water Pressure Low Scram        RHR (B) Suppression Pool Cooling Mode      RCIC Div. I Isolation Logic Reset Switch Bypass Switch (A)                            Initiation Switch CRD Charging Water Pressure Low Scram        RHR (C) Suppression Pool Cooling Mode      RCIC Div. II Isolation Logic Reset Switch Bypass Switch (B)                            Initiation Switch CRD Charging Water Pressure Low Scram        RHR (B) Primary Containment Vessel Spray    RCIC Inboard Isolation Control Switch Bypass Switch (C)                            Mode Initiation Switch CRD Charging Water Pressure Low Scram        RHR (C) Primary Containment Vessel Spray    RCIC Outboard Isolation Control Switch Bypass Switch (D)                            Mode Initiation Switch Manual Scram Reset Switch                    SGTS (B) Initiation Switch                  Fire Protection System Motor Pump Control 25A5675AA Revision 7 Switch RPS Div. I Trip Reset Switch                  SGTS (C) Initiation Switch                  Fire Protection System Diesel Pump Control Switch RPS Div. II Trip Reset Switch                Div. I Manual ADS Channel 1                FCS (B) Control Switch Initiation Switch RPS Div. III Trip Reset Switch                Div. I Manual ADS Channel 2                FCS (C) Control Switch Initiation Switch RPS Div. IV Trip Reset Switch                Div. II Manual ADS Channel 1 Design Control Document/Tier 1 Initiation Switch 2.7-3
 
ABWR 2.7-4                                Table 2.7.1a        Main Control Room Panels Fixed Position Alarms, Displays and Controls (Continued)
B. Fixed Position Displays RPV Water Level                              RCIC Flow                            SRV Positions RCIC Turbine Speed                            RCIC Injection Valve Status          Suppression Pool Level Wetwell Pressure                              HPCF (B) Injection Valve Status      Main Steamline Flow Suppression Pool Bulk Average                HPCF (C) Injection valve status      SLC Boron Tank Water Level Temperature HPCF (B) Flow                                RHR (A) Flow                          Recirculation Pump Speeds HPCF (C) Flow                                RHR (A) Injection Valve Status        Average Drywell Temperature RPV Pressure                                  RHR (B) Flow                          Wetwell Hydrogen Concentration Level Drywell Pressure                              RHR (B) Injection Valve Status        Drywell Hydrogen Concentration Level 25A5675AA Revision 7 Reactor Power Level,                          RHR (C) Flow                          Drywell Oxygen Concentration (Neutron Flux, APRM)
Reactor Power Level (SRNM)                    RHR (C) Injection Valve Status        Wetwell Oxygen Concentration Reactor Thermal Power                        Emergency Diesel Generator (A)        FCS (B) Operating Status Operating Status MSIV Position Status (Inboard And            Emergency Diesel Generator (B)        FCS (C) Operating Status Outboard Valves)                              Operating Status Reactor Mode Switch Mode Indications          Emergency Diesel Generator (C)        Main Stack Radiation Level Design Control Document/Tier 1 Operating Status Main Steamline Radiation                      Primary Containment Water Level      Time Scram Solenoid Lights (8) Status              Condensate Storage Tank Water Level  Drywell Radiation Level Manual Scram Switch (A) Indicating            SLC Pump (A) Discharge Pressure      Wetwell Radiation Level Main Control Room Panels Light Status Manual Scram Switch (B) Indicating            SLC Pump (B) Discharge Pressure Light Status RPV Isolation Status Display                  Main Condenser Pressure
 
ABWR Main Control Room Panels Table 2.7.1a    Main Control Room Panels Fixed Position Alarms, Displays and Controls (Continued)
C. Fixed Position Alarms
* Indicated RPV Water Level Abnormal            RPV Water Level Low (ECCS Initiation)      CAMS H2/O2 Level High RPV Water Level Low (Scram Level)            Control Rod Not Inserted To/Beyond MSBWP    CAMS (A) System Abnormal RPV Pressure High                            RPV Water Level High                        CAMS (B) System Abnormal Drywell Pressure High                        Fire Protection System Status              Reactor Building P Low Neutron Flux High-High                        ADS (A) Logic Initiated                    Area Temperature High Neutron Monitoring System Inoperative        ADS (B) Logic Initiated                    Area HVAC T High MSIV Closure                                  SRV Open                                    R/B HVAC Exhaust Radiation High CRD Charging Water Pressure Low              Main Steam Line Flow High                  Reactor Building Area Radiation High Rapid Core Flow Decrease                      HPIN (A) System Status                      Reactor Building Floor Drain Sump Water Level 25A5675AA Revision 7 High-High Main Turbine Trip                            HPIN (B) System Status                      R/B HVAC System Status Main Generator Trip                          Leak Detection Isolation                    Stack Radioactivity High Main Steam Line Radiation High                RWCU System Status                          RCW Radioactivity High Reactor Scram                                Reactor Period Short                        Radwaste Effluent Radioactivity High RPV Low Level Isolation Incomplete (Scram    ADS Div. I Inhibited/Auto Out Of Service    Turbine Building Ventilation System (TBVS)
Water Level)                                                                              Status Design Control Document/Tier 1 RPV Low Level Isolation Incomplete (ATWS      ADS Div. II Inhibited/Auto Out Of Service  Radiation Monitor High Scram Level)
RPV Low Level/Drywell Pressure High          Suppression Pool Bulk Average Temperature  RCIC System Status Isolation Incomplete                          High RPV Water Level Low (ATWS Scram Level)        Drywell Average Temperature High            HPCF (B) System Status RPV Water Level Low (HPCF Initiation Level)  Suppression Pool Water Level High/Low      HPCF (C) System Status
* Functional Definitions 2.7-5
 
ABWR 2.7-6                                                                    Table 2.7.1b Main Control Room Panels Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. Equipment comprising the MCRP, as defined 1. Inspections of the as-built system will be      1. The as-built MCRP conforms with the in Section 2.7.1, is available in the MCR. conducted.                                        description in Section 2.7.1.
: 2. Non-Class 1E and divisional Class 1E        2.                                              2.
control and instrument power is provided for
: a. Tests will be conducted on the MCRP by      a. The test signal exists only in Class 1E the MCRP. In the MCRP, independence is providing a test signal to only one Class      division under test in the MCRP.
provided between Class 1E divisions, and 1E division at a time.
between Class 1E divisions and non-Class 1E equipment.                                  b. Inspections of the as-built Class 1E        b. In the MCRP, physical separation or divisions in the MCRP will be conducted.        electrical isolation exists between Class 1E divisions. Physical separation or 25A5675AA Revision 7 electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
Design Control Document/Tier 1 Main Control Room Panels
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 2.7.2 Radioactive Waste Control Panels No entry. Covered in Section 2.9.1.
Radioactive Waste Control Panels                                                            2.7-7
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.7.3 Local Control Panels Design Description The Local Control Panels (LCP) consist of safety-related and non-safety-related local panels, control boxes, instrument racks and their respective internal wiring. LCPs function as protective housings and support structures for electrical and electronic equipment and facilitate local control operation.
LCPs that support safety-related equipment are classified as safety-related and Seismic Category I. Safety-related LCPs are located in Seismic Category I structures and in their divisional areas.
Safety-related LCPs are powered from their respective Class 1E divisions. Independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
LCPs which are located in areas designated as harsh environment areas are qualified for harsh environments.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.7.3 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the LCP.
2.7-8                                                                                    Local Control Panels
 
ABWR Local Control Panels Table 2.7.3 Local Control Panels Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the LCPs is      1. Inspections of the as-built system will be    1. The as-built LCPs conform with the basic described in Section 2.7.3.                    conducted.                                      configuration described in Section 2.7.3.
: 2. Safety-related LCPs are powered from their 2.                                                2.
respective Class 1E divisions. Independence
: a. Tests will be conducted in the LCPs by          a. A test signal exists in only the Class 1E is provided between Class 1E divisions and providing a test signal to only one Class          division under test in the LCPs.
between Class 1E divisions and non-Class 1E division at a time.
1E equipment.
: b. Inspections of the as-built Class 1E            b. In the LCPs, physical separation or divisions in the LCPs will be conducted.            electrical isolation exists between as-built Class 1E divisions. Physical separation or electrical isolation exists 25A5675AA Revision 7 between these Class 1E divisions and non-Class 1E equipment.
Design Control Document/Tier 1 2.7-9
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.7.4 Instrument Racks No entry. Covered in Section 2.7.3.
2.7-10                                                                          Instrument Racks
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.7.5 Multiplexing System Design Description The Multiplexing System consists of the essential multiplexing system (EMS) and the non-essential multiplexing system (NEMS).
Essential Multiplexing System The EMS provides distributed data acquisition and control networks that support the control and monitoring of the plant protection and safety systems. The EMS comprises electrical devices and circuitry, including remote multiplexing units (RMUs), transmission lines, and control room multiplexing units (CMUs) that acquire data from remote process sensors and discrete devices located within the plant, and then multiplex the data to Safety System Logic and Control (SSLC) equipment in the main control room area. SSLC operates on the input signals according to the required system logic functions, and transmits multiplexed control signals to RMUs outside of the main control room. The RMUs distribute the signals to the final actuators of the supported systems driven equipment. In addition to SSLC, the EMS also supports the data acquisition and transmission of other safety-related signals used for display and recording.
The EMS is classified as a Class 1E safety-related system.
There are four divisions of EMS equipment, with no direct interconnections among divisions.
Each division of equipment has independent control of data acquisition and multiplexing.
System timing is asynchronous among the four divisions, so that timing and clock signals in any one division only influence data transmission functions within that division. EMS uses a deterministic communications protocol; i.e., sensor signals and control data are guaranteed network access on an equal basis without interference from other signals or network traffic.
Class 1E analog and discrete sensors of the plant safety systems are connected to RMUs outside the main control room. These RMUs perform signal conditioning, analog-to-digital conversion for continuous process inputs, change-of-state detection for discrete inputs, and data message formatting prior to signal transmission. The RMUs are limited to acquisition of sensor data and the output of control signals. Trip decisions and other control logic functions are performed in SSLC processors in the main control room area. The RMUs transmit serial, time-multiplexed data streams representing the identity and status of the plant variables to the CMUs. Each division of EMS has two transmission lines interconnecting the RMUs and CMUs of that division.
The CMUs demultiplex the data and condition the signals for use in either the controllers of SSLC or in monitoring systems. After the input data is processed in SSLC, the resulting trip logic decisions are transmitted (for engineered safety features (ESF) functions only) as a serial, time-multiplexed data stream to the CMUs, which acquire the data and transmit it via EMS to Multiplexing System                                                                                        2.7-11
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 RMUs in the local areas, where the digital data is converted to signals for actuation of control devices.
Data communications from EMS to non-safety-related systems or devices for control or display purposes use an isolating transmission medium and buffering devices. Data cannot be transmitted from the non-safety-related side to EMS.
The EMS features automatic self-test and automatically reconfigures after detecting failure of one channel (either a cable break or device failure) within a division. The system returns to normal operation after reconfiguration with no interruption of data communication. If an RMU or CMU fails, that unit is automatically removed from service. Self-test runs continuously and faults are indicated in the main control room. Loss of data communications in a division of EMS does not cause transient or erroneous data to occur at system outputs.
Each of the four EMS divisions is powered from its respective divisions Class 1E DC division.
Independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
The EMS is located in the Reactor Building and the Control Building.
EMS has the following alarms and displays in the main control room:
(1)    Inoperative indication for each RMU and CMU.
(2)    Channel availability (Channel 1 or 2) for each EMS division.
(3)    Display and control of data transmission parameters and off-line self-test functions.
Non-Essential Multiplexing System The NEMS provides data communications for non-safety-related plant functions. NEMS acquires non-safety-related data from process sensors and discrete devices located throughout the plant and transmits these signals to the non-safety-related control systems for control function processing. Equipment status data is transmitted to operator control panels for monitoring alarm annunciation and to the plant computer systems for data recording and displays. The NEMS also transmits processed, non-safety-related, control signals to actuator circuits to activate valves, motor drives, alarms, monitors and indicators of the interfacing systems. The electrical devices of the NEMS consist of remote multiplexing units (RMUs),
transmission lines, and control room multiplexing units (CMUs).
NEMS is redundant to the same number of channels as the supported systems that require multiplexing. Thus, a portion of NEMS is dual redundant if a supported system is dual redundant, but is triply redundant if a supported system has that level of redundancy.
The NEMS is classified as non-safety-related, and is powered from non-Class 1E UPS.
2.7-12                                                                                  Multiplexing System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Inspections, Tests, Analyses and Acceptance Criteria Table 2.7.5 provides a definition of the visual inspections, tests and analyses, together with associated acceptance criteria, which will be undertaken for the EMS and NEMS.
Multiplexing System                                                                                        2.7-13
 
ABWR 2.7-14 Table 2.7.5 Essential Multiplexing System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The equipment comprising the Multiplexing      1. Inspection of the as-built EMS and NEMS        1. The as-built EMS and NEMS conform with System is defined in Section 2.7.5.                will be conducted.                                the description in Section 2.7.5.
: 2. EMS uses a deterministic communications        2. Tests of the EMS communications protocol      2. EMS uses a deterministic communications protocol.                                          will be conducted in a test facility.            protocol.
: 3. Data communications from EMS to non-        3. Tests on the EMS data communications will        3. EMS communications only permits data safety-related systems or devices uses an      be conducted in a test facility.                    transfer from the EMS to the non-safety-isolating transmission medium and buffering                                                          related systems or devices. Control or timing devices. Data cannot be transmitted from the                                                        signals are not exchanged between EMS non-safety-related side to EMS.                                                                      and non-safety-related systems or devices.
: 4. The EMS features automatic self-test and        4. Tests will be conducted on each as-built      4. There is no loss of EMS data communication 25A5675AA Revision 7 automatically reconfigures after detecting        EMS division by individually simulating the      as a result of the fault. Fault occurrence is failure of one channel (either a cable break      following, while simultaneously transmitting      displayed in the main control room.
or device failure) within a division. The          and monitoring test data streams:
system returns to normal operation after
: a. Single cable break.
reconfiguration with no interruption of data communication.                                      b. Loss of one RMU.
: c. Loss of one CMU.
: 5. Loss of data communications in a division of 5. Tests will be performed in one division of  5. Data communication is lost without EMS does not cause transient or erroneous      EMS at a time. While simulated input signals    generation of transient or erroneous signals.
data to occur at system outputs.                are being transmitted cable segments in Design Control Document/Tier 1 redundant paths will be disconnected and EMS outputs monitored.
Multiplexing System
 
Table 2.7.5 Essential Multiplexing System (Continued)
ABWR Multiplexing System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 6. Each of four EMS divisions is powered from  6.                                                  6.
its respective divisions Class 1E DC
: a. Tests will be performed on EMS by                a. The test signal exists only in the Class division. In the EMS, independence is providing a test signal in only one Class          1E division under test in the EMS.
provided between Class 1E divisions, and 1E division at a time.
between Class 1E divisions and non-Class 1E equipment.                                    b. Inspection of the as-installed Class 1E          b. In the EMS, physical separation or divisions in the EMS will be performed.            electrical isolation exists between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 7. Main control room alarms and displays        7. Inspections will be performed on the main        7. Alarms and displays exist or can be retrieved 25A5675AA Revision 7 provided for the EMS are as defined in          control room alarms and displays for the            in the main control room as defined in Section 2.7.5.                                  EMS.                                                Section 2.7.5.
Design Control Document/Tier 1 2.7-15
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.7.6 Local Control Boxes No entry. Covered in Section 2.7.3.
2.7-16                                                                      Local Control Boxes
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.8.1 Nuclear Fuel Design Description The fuel assembly is designed to ensure that possible fuel damage would not result in the release of radioactive materials in excess of prescribed limits. The fuel assembly is comprised of the fuel bundle, channel and channel fastener. The fuel bundle is comprised of fuel rods, water rods, fuel rods containing burnable neutron absorber, spacers, springs and assembly end fittings.
The following is a summary of the principal design requirements which must be met by the fuel and is evaluated using methods and criteria to assure that:
(1)  Fuel rod failure is predicted to not occur as a result of normal operation and anticipated operational occurrences.
(2)  Control rod insertion will not be prevented as a result of normal operation, anticipated operational occurrences or postulated accident.
(3)  The number of fuel rod failures will not be underestimated for postulated accidents.
(4)  Coolability will be maintained for all design basis events, including seismic and LOCA events.
(5)  Specified acceptable fuel design limits (thermal and mechanical design limits) will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
(6)  In the power operating ranges, the prompt inherent nuclear feedback characteristics will tend to compensate for a rapid increase in reactivity.
(7)  The reactor core and associated coolant, control and protection systems will be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.
Nuclear Fuel                                                                                              2.8-1
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.8.2 Fuel Channel Design Description The fuel channels are zirconium-based (or equivalent) and preclude cross-flow in the core region. These channels form the flow path for bundle coolant flow, provide surfaces for control rod guidance, provide structural stiffness to the bundle during lateral loadings, transmit seismic loadings to the top guide and fuel support castings, and provide a heat sink during loss-of-coolant accident (LOCA).
The following is a summary of the principal design criteria which are met by the fuel channels:
(1)  During any design basis events including the mechanical loading from safe shutdown earthquake event combined with LOCA event, fuel channel damage will not be so severe as to prevent control rod insertion when it is required.
(2)  Coolability will be maintained for all design basis events.
(3)  Channel bowing will not cause specified acceptable fuel design limits to be exceeded during normal operation and anticipated operational occurrences.
2.8-2                                                                                            Fuel Channel
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.8.3 Control Rod Design Description Control rods in the reactor perform the functions of power distribution shaping, reactivity control, and scram reactivity insertion for safety shutdown response and have the following design features:
(1)    A cruciform cross-sectional envelope shape.
(2)    A coupling at the bottom for attachment to the control rod drive.
(3)    Contain neutron absorbing materials.
The following is a summary of the principal design criteria which are met by the control rod:
(1)    The control rod stresses, strains, and cumulative fatigue will be evaluated to not exceed the ultimate stress or strain of the material.
(2)    The control rod will be evaluated to be capable of insertion into the core during design basis modes of operation including safe shutdown earthquake event combined with LOCA event.
(3)    The material of the control rod will be compatible with the reactor environment.
(4)    The reactivity worth of the control rods will be included in the plant core analyses, and will provide, under conditions of normal operation (including anticipated operational occurrences), appropriate margin for malfunctions such as two stuck rods (associated with a given accumulator), or accidental control rod withdrawal, without exceeding specified acceptable fuel design limits.
Control Rod                                                                                              2.8-3
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.8.4 Loose Parts Monitoring System Design Description The Loose Parts Monitoring System (LPMS) monitors the reactor pressure vessel (RPV) for indications of loose metallic parts within the reactor pressure vessel. The LPMS detects structure borne sound that can indicate the presence of loose parts impacting against the reactor pressure vessel and internals. The system alarms when sensor signal characteristics exceeds preset limits.
The LPMS consists of sensors, cables, signal conditioning equipment, alarming monitors, signal analysis and data acquisition equipment. The LPMS processes signals from multiple sensors mounted on the external surfaces of the reactor coolant pressure boundary. The LPMS is classified as non-safety-related.
The LPMS has provisions for both automatic and manual startup of data acquistion equipment with automatic activation in the event the preset alert level is reached or exceeded. The system also initiates an alarm in the main control room when an alert condition is reached.
The LPMS electronic components located inside the primary containment perform their function following all seismic events which do not require plant shutdown.
Inspections, Tests, Analyses and Acceptance Criteria Tables 2.8.4 provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria, which will be undertaken for LPMS.
2.8-4                                                                            Loose Parts Monitoring System
 
ABWR Loose Parts Monitoring System Table 2.8.4 Loose Parts Monitoring System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. Equipment comprising the LPMS is defined      1. Inspection of the as-built system will be    1. The as-built LPMS conforms with the in Section 2.8.4.                                conducted.                                      description in Section 2.8.4.
: 2. The LPMS monitors the RPV for indication of 2. Tests will be conducted on the as-built          2. The LPMS sensitivity, without the loose metallic parts.                          LPMS.                                              background noise associated with plant operation, is such that it can detect a metallic loose part that weighs from 0.11 kg to 13.6 kg and impacts with a maximum kinetic energy of 0.68 joules on the inside surface of the RPV within 0.91m of a sensor.
: 3. Main control room alarms provided for the      3. Inspections will be performed on the main    3. Alarms exist or can be retrieved in the main LPMS are defined in Section 2.8.4.                control room alarms for the LPMS.                control room as defined in Section 2.8.4.
25A5675AA Revision 7
: 4. The LPMS electronic components located        4. Analyses will be performed or tests will be  4. An analysis or test report exists which inside the primary containment perform their      conducted on the seismic capability of the      concludes that the LPMS electronic function following all seismic events which do    LPMS electronic components located in the        components located inside the primary not require plant shutdown.                      primary containment.                            containment perform their function following all seismic events which do not require plant shutdown.
Design Control Document/Tier 1 2.8-5
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.9.1 Radwaste System Design Description The Radwaste (RW) System consists of a liquid waste system, a solid waste system and a radioactive drain transfer system . The liquid waste system includes primary containment penetrations, and inboard and outboard motor-operated isolation valves for the high conductivity and low conductivity waste drains from the lower drywell. The liquid waste system collects, treats, monitors, and either recycles treated radioactive liquid wastes within the plant or discharges them to the environs. The solid waste system sorts, processes, monitors and packages processed solid radwastes for shipment to an offsite disposal facility.
The RW System is classified as non-safety-related with the exception of the primary containment isolation function.
The primary containment penetrations and isolation valves are classified as Seismic Category I and ASME Code Class 2. The back flow check valves in the emergency core cooling system (ECCS) equipment room sumps are classified as Seismic Category I.
The RW System processing equipment is located in the Radwaste Building.
The inboard containment isolation valves are powered from Class 1E Division II, and the outboard isolation valves are powered from Class 1E Division I. In the RW System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The main control room has control and open/close status indications for the primary containment isolation valves.
The safety-related electrical equipment that provides containment isolation, located in the primary containment and the Reactor Building, is qualified for a harsh environment.
The primary containment isolation motor-operated valves (MOVs) have active safety-related function to close and perform these functions under differential pressure, fluid flow, and temperature conditions.
The liquid waste system has one discharge line which has a radiation monitor. Discharge flow is terminated on receipt of a high radiation signal from this monitor.
The radioactive drain transfer system in each divisional area of the ECCS pump rooms and the Control Building are physically separated from drains in the other divisions. Figures 2.9.1a and 2.9.1b show the basic system configuration and scope.
Radwaste System                                                                                            2.9-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 Inspections, Tests, Analyses and Acceptance Criteria Table 2.9.1 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Radwaste System.
2.9-2                                                                                    Radwaste System
 
25A5675AA Revision 7 ABWR                                                      Design Control Document/Tier 1 DIVISION A                      DIVISION B                    DIVISION C ROOMS WITH                      ROOMS WITH                    ROOMS WITH FLOOR DRAINS                      FLOOR DRAINS                  FLOOR DRAINS DIVISION A                      DIVISION B                    DIVISION C SUMP                            SUMP                          SUMP AND PUMP                        AND PUMP                      AND PUMP CONTROL BUILDING NOTES:
: 1. THE SYSTEM HAS NO VALVES, PUMPS, OR OTHER ACTIVE COMPONENTS IN THE DRAINAGE PATHS.
Figure 2.9.1a Radioactive Drain Transfer System Radwaste System                                                                      2.9-3
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 DIVISION A                      DIVISION B                          DIVISION C ECCS PUMP                        ECCS PUMP                            ECCS PUMP ROOMS                            ROOMS                                ROOMS DIVISION A                        DIVISION B                          DIVISION C SUMP                            SUMP                                SUMP AND PUMPS                        AND PUMPS                            AND PUMPS SECONDARY CONTAINMENTECCS AREAS DIVISION A            NON-DIVISIONAL                  DIVISION B                DIVISION C ROOMS WITH              ROOMS WITH                  ROOMS WITH                ROOMS WITH FLOOR DRAINS            FLOOR DRAINS                FLOOR DRAINS              FLOOR DRAINS SUMP                                                  SUMP AND PUMPS                                            AND PUMPS SECONDARY CONTAINMENT - OTHER AREAS NOTES:
: 1. THE SYSTEM HAS NO VALVES, PUMPS, OR OTHER ACTIVE COMPONENTS IN THE DRAINAGE PATHS.
Figure 2.9.1b Radioactive Drain Transfer System 2.9-4                                                                                    Radwaste System
 
ABWR Radwaste System Table 2.9.1 Radwaste System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration for the RW System is 1. Inspection of the as-built system will be      1. The as-built RW System conforms with the described in Section 2.9.1.                    conducted.                                        basic configuration described in Section 2.9.1.
: 2. The ASME Code components of the RW              2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the RW System                ASME Code components of the RW System integrity under internal pressures that will be    required to be hydrostatically tested by the    conform with the requirements in the ASME experienced during service.                        ASME Code.                                      Code, Section III.
: 3. The inboard containment isolation valves are 3.                                              3.
powered from Class 1E Division II, and the
: a. Tests will be performed on the RW            a. The test signal exists only in the Class outboard isolation valves are powered from System by providing a test signal in only      1E division under test in the RW System.
Class 1E Division I. In the RW System, 25A5675AA Revision 7 one Class 1E division at a time.
independence is provided between Class 1E                                                      b. In the RW System, physical separation divisions and non-Class 1E equipment.          b. Inspection of the as-installed Class 1E        or electrical isolation exists between divisions in the RW System will be              Class 1E divisions. Physical separation performed.                                      or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 4. Main control room displays and controls      4. Inspections will be performed on the main    4. Displays and controls exist or can be provided for the RW System are as defined        control room displays and controls for the      retrieved in the main control room as defined in Section 2.9.1.                                RW System.                                      in Section 2.9.1.
Design Control Document/Tier 1
: 5. MOVs designated in Section 2.9.1 as having 5. Tests of installed valves for closing will be    5. Upon receipt of the actuating signal, each an active safety-related function close under conducted under preoperational differential        MOV closes.
differential pressure, fluid flow, and        pressure, fluid flow, and temperature temperature conditions.                      conditions.
: 6. The liquid waste system has one discharge 6. Tests will be conducted on the as-built liquid    6. The discharge flow terminates upon receipt line which has a radiation monitor. Discharge waste system using a simulated high                of a simulated high radiation signal.
flow is terminated on receipt of a high      radiation signal.
radiation signal from this monitor.
2.9-5
 
Table 2.9.1 Radwaste System (Continued)
ABWR 2.9-6 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 7. The radioactive drain transfer system in each 7. Tests will be conducted on the as-built      7. No interconnection exist (i.e. no water divisional area of the ECCS pump rooms and      system by individually pressuring each          leakage in to other divisions not being the Control Building are physically separated    divisional area drains with water and          tested).
from drains in the other divisions.              observing other divisional area drains for interdivisional leakage.
25A5675AA Revision 7 Design Control Document/Tier 1 Radwaste System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.10.1 Turbine Main Steam System Design Description The Turbine Main Steam (MS) System, as shown in Figure 2.10.1, supplies steam generated in the reactor to the turbine, steam auxiliaries and turbine bypass valves. The MS boundaries are shown in Figure 2.10.1. The MS System does not include the seismic interface restraint nor main turbine stop or bypass valves.
The MS System:
(1)  Accommodates operational stresses such as internal pressure and dynamic loads without failures.
(2)  Provides a seismically analyzed fission product leakage path to the main condenser.
(3)  Has suitable access to permit in-service testing and inspections.
(4)  Closes the steam auxiliary (SA) valve(s) on a main steam isolation valve (MSIV) isolation signal. These valves fail closed on loss of electrical power to the valve actuating solenoid or on loss of pneumatic pressure.
The MS System main steam piping consists of four lines from the seismic interface restraint to the main turbine stop valves. The header arrangement upstream of the turbine stop valves allows the valves to be tested on-line and also supplies steam to the power cycle auxiliaries.
The MS System is classified as non-safety-related. However, the MS System is analyzed, fabricated and examined to ASME Code Class 2 requirements, and classified as non-Seismic Category I. Inservice inspection shall be performed in accordance with ASME Section XI requirements for Code Class 2 piping. ASME authorized nuclear inspector and ASME Code stamping is not required.
MS piping, including the steam auxiliary valve(s), from the seismic interface restraint to the main stop and main turbine bypass valves is analyzed to demonstrate structural integrity under safe shutdown earthquake (SSE) loading conditions.
The MS System is located in the steam tunnel and Turbine Building.
Turbine Main Steam System                                                                                  2.10-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.1 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the MS System.
2.10-2                                                                          Turbine Main Steam System
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 NBS    MS                                      MS OTHERS SYSTEM                              SYSTEM 2                                          2 NOTE 1                                NOTE 1 P
TO STEAM AUXILIARIES NOTE 2 SEISMIC INTERFACE                                                          MAIN TURBINE RESTRAINT                                                          STOP VALVES TO FROM MAIN NBS TURBINE TO TURBINE BYPASS        MAIN VALVES        CONDENSER NOTES:
: 1. AS MODIFIED PER DESIGN DESCRIPTION.
: 2. MULTIPLE LINES MAY BE USED.
ISOLATION PROVISIONS ARE REQUIRED FOR EACH STEAM AUXILIARY LINE.
Figure 2.10.1 Turbine Main Steam System Turbine Main Steam System                                                                    2.10-3
 
ABWR 2.10-4 Table 2.10.1 Main Steam System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the MS System is 1. Inspections of the as-built system will be      1. The as-built MS System conforms with the as shown on Figure 2.10.1.                    conducted.                                          basic configuration shown in Figure 2.10.1.
: 2. The ASME Code components of the MS              2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the MS System                ASME Code components of the MS System integrity under internal pressures that will be    required to be hydrostatically tested by the    conform with the requirements in the ASME experienced during service.                        ASME Code.                                      Code, Section III.
: 3. Upon receipt of an MSIV closure signal, the  3. Using simulated MSIV closure signals tests    3. The SA valve(s) close(s) following receipt of SA valve(s) close(s).                            will be performed on the SA valves.              a simulated MSIV closure signal.
: 4. The SA valve(s) fail(s) closed on loss of    4. Test will be performed on SA valves.          4. The SA valve(s) close(s) on loss of electrical electrical power to the valve actuating                                                            power to the valve actuating solenoid or on 25A5675AA Revision 7 solenoid or on loss of pneumatic pressure.                                                        loss of pneumatic pressure.
The pneumatically operated SA valve(s) close(s) when either electrical power to the valve actuating solenoid is lost or pneumatic pressure to the valve(s) is lost.
: 5. MS piping, including the SA valve(s) from the 5. A seismic analysis of the as-built MS piping  5. An analysis report exists which concludes seismic interface restraint to the main stop    and SA valve(s) will be performed.                that the as-built MS piping and SA valve(s) and main turbine bypass valves are analyzed                                                        can withstand a SSE without loss of to demonstrate structural integrity under SSE                                                      structural integrity.
loading conditions.
Design Control Document/Tier 1 Turbine Main Steam System
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.10.2 Condensate Feedwater and Condensate Air Extraction System The Condensate Feedwater and Condensate Air Extraction (CFCAE) System consists of two subsystems: the Condensate and Feedwater System (CFS) and the Main Condenser Evacuation System (MCES).
Design Description Condensate and Feedwater System The function of the CFS is to receive condensate from the condenser hotwells, supply condensate to the Condensate Purification System (CPS), and deliver feedwater to the reactor.
Condensate is pumped from the main condenser hotwell by the condensate pumps, passes through the low pressure feedwater heaters to the feedwater pumps, and then is pumped through the high pressure heaters to the reactor. Figure 2.10.2a shows the basic system configuration.
The CFS boundaries extend from the main condenser outlet to (but not including) the seismic interface restraint outside the containment.
The CFS is classified as non-safety-related.
The CFS is controlled by signals from the Feedwater Control System.
The CFS is located in the steam tunnel and Turbine Building.
The CFS has parameter displays for the instruments shown on Figure 2.10.2a in the main control room.
Main Condenser Evacuation System The MCES removes the hydrogen and oxygen produced by the radiolysis of water in the reactor, and other power cycle noncondensable gases. The system exhausts the gases to the Off-Gas System (OGS) during plant operation, and to the Turbine Building compartment exhaust system at the beginning of each startup. The MCES consists of redundant steam jet air ejector (SJAE) units for power plant operation, and a mechanical vacuum pump for use during startup. Figure 2.10.2b shows the basic system configuration.
The MCES is classified as non-safety-related.
The MCES is located in the Turbine Building.
Steam supply to the SJAE provides dilution of the hydrogen and prevents the offgas from reaching the flammable limit of hydrogen. When the steam flow drops below the setpoint for stream dilution, the Off-Gas System is isolated.
The vacuum pump is tripped and its discharge valve is closed upon receiving a main steamline high radiation signal.
Condensate Feedwater and Condensate Air Extraction System                                              2.10-5
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 The MCES has the following displays in the main control room:
(1)  Parameter displays for the instruments shown on Figure 2.10.2b.
(2)  Status indication for the vacuum pump and SJAE discharge valves.
Inspections, Tests, Analyses and Acceptance Criteria Tables 2.10.2a and 2.10.2b provide a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the CFCAE System, respectively.
2.10-6                                              Condensate Feedwater and Condensate Air Extraction System
 
ABWR Condensate Feedwater and Condensate Air Extraction System MC NNS CFS P                                              CONDENSATE NBS CFS                              FE    P                            PUMPS NNS TO REACTOR SEISMIC    HIGH INTERFACE  PRESSURE            FEEDWATER        LOW RESTRAINT  HEATERS              PUMPS        PRESSURE      CPS 25A5675AA Revision 7 HEATERS NNS CFCAE CRD Design Control Document/Tier 1 NOTES:
: 1. RELIEF VALVE DISCHARGE AND VENTS ARE CHANNELED THROUGH CLOSED SYSTEMS.
: 2. FEEDWATER AND CONDENSATE PUMP REDUNDANCY IS PROVIDED.
Figure 2.10.2a Condensate and Feedwater System 2.10-7
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 MCAE HVAC (TURBINE BUILDING NNS        COMPARTMENT EXHAUST)
M M
VACUUM PUMP                          TURBINE BUILDING COMPARTMENT EXHAUST M        STEAM JET TO OFF-GAS EJECTOR A SYSTEM NNS FROM MAIN                                          MCAE OGS CONDENSER NNS                M          FE MC MCAE DILUTION NNS STEAM AS MCAE M
TO OFF-GAS EJECTOR B SYSTEM NNS MCAE OGS M          FE DILUTION STEAM      NNS AS MCAE Figure 2.10.2b Main Condenser Evacuation System 2.10-8                                          Condensate Feedwater and Condensate Air Extraction System
 
ABWR Condensate Feedwater and Condensate Air Extraction System Table 2.10.2a        Condensate and Feedwater System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the CFS is as      1. Inspections of the as-built CFS will be      1. The as-built CFS conforms with the basic shown on Figure 2.10.2a.                          conducted.                                      configuration shown in Figure 2.10.2a.
: 2. The CFS is controlled by signals from the      2. Tests of the as-built CFS will be conducted  2. The CFS starts and operates in response to Feedwater Control System.                        using simulated input signals.                  the simulated signals.
: 3. Main control room displays provided for the    3. Inspections will be performed on the main    3. Displays exist or can be retrieved in the main CFS are as defined in Section 2.10.2.            control room displays for the CFS.              control room as defined in Section 2.10.2.
25A5675AA Revision 7 Design Control Document/Tier 1 2.10-9
 
ABWR 2.10-10 Table 2.10.2b Main Condenser Evacuation System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the MCES is as    1. Inspections of the as-built MCES will be    1. The as-built MCES conforms with the basic shown on Figure 2.10.2b.                        conducted.                                      configuration shown in Figure 2.10.2b.
: 2. When the steam flow drops below the          2. Tests will be conducted on the as-built MCES 2. The SJAE suction valves close on receipt of setpoint for steam dilution, the Off-Gas        using simulated signals for steam flow.        a simulated low flow signal.
System is isolated.
: 3. The vacuum pump is tripped and its            3. Tests will be conducted on the as-built MCES 3. The vacuum pump trips and the discharge discharge valve is closed upon receiving a      using simulated signals for radiation in the    valve closes upon receipt of a simulated high main steamline high radiation signal.            main steamlines.                                radiation signal.
: 4. Main control room displays provided for the  4. Inspections will be performed on the main    4. Displays exist or can be retrieved in the main MCES are as defined in Section 2.10.2.          control room displays for the MCES.            control room as defined in Section 2.10.2.
25A5675AA Revision 7 Condensate Feedwater and Condensate Air Extraction System Design Control Document/Tier 1
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.10.3 Heater Drain and Vent System No entry for this system.
Heater Drain and Vent System                                                        2.10-11
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.10.4 Condensate Purification System Design Description The Condensate Purification System (CPS) purifies and treats the condensate, using filtration to remove insoluble solids, and ion exchange demineralizer to remove soluble solids. The CPS consists of full flow high efficiency particulate filters followed by full flow deep bed demineralizers. Figure 2.10.4 shows the basic system configuration.
The CPS is classified as non-safety-related.
The CPS is located in the Turbine Building.
The CPS has alarms and display for effluent conductivity in the main control room.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.4 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the CPS.
2.10-12                                                                          Condensate Purification System
 
ABWR Condensate Purification System CFS  CPS                                      CPS CFS NNS                                      NNS 25A5675AA Revision 7 TO CONDENSATE LOW PRESSURE              DEMINERALIZER            FILTER FLOW FW HEATERS CM Design Control Document/Tier 1 Figure 2.10.4 Condensate Purification System 2.10-13
 
ABWR 2.10-14 Table 2.10.4 Condensate Purification System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the CPS is as      1. Inspections of the as-built System will be    1. The as-built CPS conforms with the basic shown on Figure 2.10.4.                          conducted.                                      configuration shown in Figure 2.10.4.
: 2. Main control room alarm and display          2. Inspections will be performed on the main    2. Alarm and display exist or can be retrieved in provided for the CPS are as defined in          control room alarm and display for the CPS.      the main control room as defined in Section Section 2.10.4.                                                                                  2.10.4.
25A5675AA Revision 7 Design Control Document/Tier 1 Condensate Purification System
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 2.10.5 Condensate Filter Facility No entry. Covered in Section 2.10.4.
Condensate Filter Facility                                                                2.10-15
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.10.6 Condensate Demineralizer No entry. Covered in Section 2.10.4.
2.10-16                                                                Condensate Demineralizer
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.10.7 Main Turbine Design Description The Main Turbine (MT) uses the energy in steam from the reactor to drive the plant generator.
The major turbine components are:
(1)  A high pressure section.
(2)  An intermediate section (between high pressure and low pressure sections).
(3)  Low pressure sections.
The major fluid system boundaries are:
(1)  Turbine Main Steam 2.10.1.
(2)  Main Condenser 2.10.21.
(3)  Turbine Gland Seal 2.10.9.
(4)  Extraction System 2.10.12.
The MT is classified as non-safety-related.
The MT has the following features that prevent overspeed:
(1)  Main turbine stop valves (MTSV)/Control valves (CV) [MTSVs trip/CVs trip and modulate].
(2)  Combined intermediate valves (CIVs) consist of intercept valves (IVs) and intercept stop valves (ISVs) [IVs trip and modulate/ISVs trip].
(3)  Extraction line non-return valves (trip).
(4)  Redundant valve closure mechanisms (i.e., fast acting solenoid valves and emergency trip fluid system).
(5)  Redundant normal speed control.
Three levels of signals to MT valves (i.e., normal speed control/overspeed trip/backup overspeed trip).
Main Turbine                                                                                      2.10-17
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 Overspeed trip occurs as follows:
Overspeed Condition                                Protective Action (1)  Exceeds normal speed                Normal speed control signals the CVs and IVs to control setpoint.                  close.
(2)  Exceeds overspeed trip setpoint. Overspeed trip signals MTSVs, CVs, IVs, ISVs, and extraction line non-return valves to close.
(3)  Exceeds backup overspeed            Backup overspeed trip signals MTSVs, CVs, IVs, trip setpoint.                      ISVs, and extraction line non-return valves to close.
The turbine MTSV closes in 0.1 seconds or greater. The turbine CV trip closure is 0.08 seconds or greater. In the modulating mode, the full stroke servo-closure of the turbine CV is 2.5 seconds or greater.
The MT System has the following alarms and displays in the main control room:
(1)    Overspeed alarm.
(2)    Parameter displays for turbine speed and inlet steam pressure.
The main turbine stop valves are analysed to demonstrate structural integrity under safe shutdown earthquake (SSE) loading conditions.
The MT is located within the Turbine Building. The axis of the turbine and generator is orientated within the Turbine Building to be inline with the Reactor and Control Buildings.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.7 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the MT System.
2.10-18                                                                                      Main Turbine
 
Main Turbine ABWR Table 2.10.7 Main Turbine System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                    Acceptance Criteria
: 1. The basic configuration of the MT System is 1. Inspection of the as-built MT will be      1. The as-built MT conforms with the basic as described in Section 2.10.7.                conducted.                                    configuration described in Section 2.10.7.
: 2. MT System overspeed protective actions are 2. Tests will be conducted on the as-built MT  2. The following protective actions occur:
as defined in Section 2.10.7.                System using simulated overspeed signals.
Overspeed            Protective Action Condition
: a. Exceeds          Normal speed normal          control signals the speed control    CVs and IVs to setpoint.        close.
25A5675AA Revision 7
: b. Exceeds          Overspeed trip overspeed        signals MTSVs, trip setpoint. CVs, ISVs, IVs, and extraction line non-return valves to close.
: c. Exceeds          Backup overspeed backup          trip signals MTSVs, overspeed        CVs, ISVs, IVs, and trip setpoint. extraction line non-Design Control Document/Tier 1 return valves to close.
: 3. The turbine MTSV closes in 0.10 seconds or 3. Tests will be conducted on the as-built      3. The turbine MTSV closes in 0.10 seconds or greater.                                      turbine MTSV.                                  greater
: 4. The turbine CV trip closure is 0.08 seconds  4. Tests will be conducted on the as-built  4. The turbine CV trip closure is 0.08 seconds or greater.                                      turbine CV.                                  or greater.
2.10-19
 
Table 2.10.7 Main Turbine System (Continued)
ABWR 2.10-20 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 5. In the modulating mode, the full stroke servo 5. Tests will be conduced on the as-built turbine 5. In the modulating mode, the full stroke servo closure of the turbine CV is 2.5 seconds or      CV.                                              closure of the turbine CV is 2.5 seconds or greater.                                                                                          greater.
: 6. Main control room alarms and displays        6. Inspections will be performed on the main    6. Alarms and displays exist or can be retrieved provided for the MT are as defined in Section    control room alarms and displays for the MT. in the main control room as defined in 2.10.7.                                                                                          Section 2.10.7.
: 7. The axis of the turbine and generator is      7. Inspections will be conducted of the as-built  7. The axis of the turbine and generator is oriented within the Turbine Building to be        turbine and generator.                            oriented within the Turbine Building to be in inline with the Reactor and Control Buildings.                                                      line with the Reactor and Control Buildings.
: 8. The MTSVs are analysed to demonstrate          8. A seismic analysis of the as-built MTSVs will 8. An analysis report exists which concludes structural integrity under SSE loading            be performed.                                    that the as-built MTSVs can withstand an 25A5675AA Revision 7 conditions.                                                                                        SSE without the loss of structural integrity.
Design Control Document/Tier 1 Main Turbine
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 2.10.8 Turbine Control System No entry. Covered in Section 2.10.7.
Turbine Control System                                                                    2.10-21
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.10.9 Turbine Gland Seal System Design Description The Turbine Gland Seal (TGS) System prevents the escape of radioactive steam from the turbine shaft casing penetrations and valve stems and prevents air inleakage through subatmospheric turbine glands. Figure 2.10.9 shows the basic system configuration.
The TGS System consists of a sealing steam pressure regulator, steam seal header and a gland seal condenser (GSC) with two full capacity exhaust blowers and associated piping, valves and instrumentation.
The TGS System is bounded by the Main Turbine and the Turbine Bypass System. The TGS System receives steam from either the Turbine Main Steam System, the feedwater heater drain tank vent header or auxiliary steam sources. The exhaust blowers discharge to the Turbine Building compartment exhaust system.
The TGS System is classified as non-safety-related.
The TGS System is located in the Turbine Building.
The TGS System has displays for gland seal condenser and steam seal header pressure in the main control room.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.9 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the TGS System.
2.10-22                                                                            Turbine Gland Seal System
 
ABWR Turbine Gland Seal System TURBINE BUILDING HVAC (TURBINE BUILDING                        COMPARTMENT COMPARTMENT EXHAUST)                          EXHAUST TGS    NNS EXHAUST BLOWERS MAIN TURBINE STOP AND CONTROL VALVES                                                                GSC MAIN STEAM                                                                                                          CROSSAROUND STEAM NOTE 1            COMBINED NNS    TO CROSSAROUND                                                      INTERMEDIATE MS TGS            STEAM                                                      VALVES TO MC TO MC 25A5675AA Revision 7 NNS TGS MC TO GSC                    TO FW HEATER TO MC                      TO MC CFCAE TGS TO FW HEATER NNS TO GSC            TO GSC                  TO GSC TGS AS NNS      SOURCES OF GLAND TO CROSSAROUND                                                                          LOW HIGH                                                                    SEAL STEAM STEAM                                                                        PRESSURE PRESSURE Design Control Document/Tier 1 TO MC                                TURBINE                              TURBINE(S)              P        MAIN STEAM BYPASS                                                                                          FW HEATER VALVES                                                                                          DRAIN TANK STEAM SEAL HEADER TURBINE P
MC TGS                                                                                                      AUXILIARY NOTE 3 NNS                                                                                                    STEAM NOTES;
: 1. STEAM PATH BETWEEN HIGH AND LOW PRESSURE TURBINES.
: 2. DELETED
: 3. TYPICAL FOR INTERFACES WITH MC.
2.10-23 Figure 2.10.9 Turbine Gland Seal System
 
ABWR 2.10-24 Table 2.10.9 Turbine Gland Seal System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the TGS System    1. Inspections of the as-built system will be  1. The as-built TGS System conforms with the is as shown on Figure 2.10.9.                    conducted.                                      basic configuration shown on Figure 2.10.9.
: 2. Main control room displays provided for the  2. Inspections will be performed on the main    2. Displays exist or can be retrieved in the main TGS System are as defined in Section            control room displays for the TGS System.      control room as defined in Section 2.10.9.
2.10.9.
25A5675AA Revision 7 Design Control Document/Tier 1 Turbine Gland Seal System
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.10.10 Turbine Lubricating Oil System No entry for this system.
Turbine Lubricating Oil System                                                        2.10-25
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.10.11 Moisture Separator Heater No entry for this system.
2.10-26                                                            Moisture Separator Heater
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.10.12 Extraction System No entry for this system.
Extraction System                                                                    2.10-27
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.10.13 Turbine Bypass System Design Description The Turbine Bypass System (TBS) discharges main steam directly to the condenser. The TBS is bounded by the Turbine Main Steam System and the Main Condenser.
The TBS is classified as non-safety-related.
The TBS consists of a valve chest that is connected to the main steamlines upstream of the main turbine stop valves, and dump lines that connect each regulating valve outlet to the condenser shell.
The turbine bypass valves are opened by a signal from the Steam Bypass and Pressure Control System.
The turbine bypass valves open upon turbine trip or generator load rejection, automatically trip closed whenever the vacuum in the condenser falls below a preset value, and fail closed on loss of electrical power or hydraulic system pressure.
The TBS is analyzed to demonstrate structural integrity under the safe shutdown earthquake (SSE) loading conditions.
The TBS is located in the Turbine Building.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.13 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the TBS.
2.10-28                                                                              Turbine Bypass System
 
ABWR Turbine Bypass System Table 2.10.13 Turbine Bypass System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration for the TBS is        1. Inspections of the as-built TBS will be      1. The as-built TBS conforms with the basic described in Section 2.10.13 and Turbine        conducted.                                      configuration of Section 2.10.13 and Turbine Main Steam System, Figure 2.10.1.                                                                Main Steam System, Figure 2.10.1.
: 2. The turbine bypass valves are opened by a 2. Tests will be conducted using a simulated        2. Turbine bypass valves open upon receipt of signal from the Steam Bypass and Pressure    signal.                                              simulated signal from the Steam Bypass and Control System.                                                                                  Pressure Control System.
: 3. The TBS is analysed to demonstrate            3. A seismic analysis of the as-built TBS will be 3. An analysis report exists which concludes structural integrity under SSE loading          performed.                                        that the as-built TBS can withstand a SSE conditions.                                                                                        without loss of structural integrity.
25A5675AA Revision 7 Design Control Document/Tier 1 2.10-29
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.10.14 Reactor Feedwater Pump Driver No entry. Covered in Section 2.10.2.
2.10-30                                                            Reactor Feedwater Pump Driver
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.10.15 Turbine Auxiliary Steam System No entry for this system.
Turbine Auxiliary Steam System                                                        2.10-31
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.10.16 Generator No entry for this system.
2.10-32                                                                      Generator
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.10.17 Hydrogen Gas Cooling System No entry for this system.
Hydrogen Gas Cooling System                                                        2.10-33
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.10.18 Generator Cooling System No entry for this system.
2.10-34                                                            Generator Cooling System
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.10.19 Generator Sealing Oil System No entry for this system.
Generator Sealing Oil System                                                        2.10-35
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.10.20 Exciter No entry for this system.
2.10-36                                                                        Exciter
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.10.21 Main Condenser Design Description The Main Condenser (MC) condenses and deaerates the exhaust steam from the main turbine (MT) and provides a heat sink for the Turbine Bypass (TB) System. The MC is also a collection point for other steam cycle drains and vents.
The MC hotwell provides a holdup volume for main steam isolation valve (MSIV) fission product leakage.
The MC is classified as non-safety-related and non-seismic Category I. The supports and anchors for the MC are designed to withstand a safe shutdown earthquake (SSE).
The MC is located in the Turbine Building (T/B).
The MC tubes are made from corrosion-resistant material. The MC operates at a vacuum; consequently, leakage is into the shell side of the MC. Circulating water leakage from the tubes to the condenser is detected by measuring the conductivity of sample water extracted beneath the tube bundles. In addition, a conductivity monitor is located at the discharge of the condensate pumps, and alarms are provided in the main control room.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.21 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the MC.
Main Condenser                                                                                        2.10-37
 
ABWR 2.10-38 Table 2.10.21 Main Condenser Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The supports and anchors for the MC are        1. An analysis of the ability of the as-built  1. An analysis report exists which concludes designed to withstand a safe shutdown            condenser supports and anchors to              that the as-built main condenser supports earthquake.                                      withstand a safe shutdown earthquake will      and anchors are able to withstand a safe be performed.                                  shutdown earthquake.
: 2. A conductivity monitor is located at the      2. The as-built system will be inspected.      2. A conductivity monitor exists at the discharge discharge of the condensate pumps.                                                                of the condensate pumps.
: 3. Main control room alarms provided for the      3. Inspections will be performed on the main  3. Alarms exist in the main control room as main condenser are as defined in Section          control room alarms for the main condenser. defined in Section 2.10.21.
2.10.21.
25A5675AA Revision 7 Design Control Document/Tier 1 Main Condenser
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.10.22 Off-Gas System Design Description The Off-Gas System (OGS) treats the gas exhausted from the main turbine condensers to control the release of gaseous radioactivity discharged to the plant environment.
The OGS has redundant hydrogen/oxygen recombiners to reduce process gas volume and noble gas adsorption beds to provide radionuclide retention/decay. A high efficiency particulate air (HEPA) filter is also provided. Figure 2.10.22 shows the basic system configuration.
Radiation levels in the OGS discharge stream are monitored (two channels). A main control room alarm and automatic OGS isolation are initiated when the radiation level exceeds setpoints.
The system pressure boundary of the OGS (including the hydrogen analyzers) is capable of withstanding an internal hydrogen explosion.
The adsorption beds and their support structure do not collapse under seismic loads corresponding to the safe shutdown earthquake (SSE) ground accelerations.
The OGS is classified as non-safety-related.
The OGS is located in the Turbine Building.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.22 provides a definition of the inspection, tests and/or analyses, together with associated criteria, which will be undertaken for the OGS.
Off-Gas System                                                                                          2.10-39
 
ABWR 2.10-40 H        H A        A H        H A        A HE      HE MCES OGS 25A5675AA Revision 7 NNS                                                                              TO PLANT RECOMBINER                                RE      RE                    ENVIRONMENT SJAE A TRAIN A P
ADSORBTION                      HEPA BEDS                        FILTER MCES OGS NNS Design Control Document/Tier 1 RECOMBINER SJAE B TRAIN B Off-Gas System                              Figure 2.10.22 Off-Gas System
 
ABWR Off-Gas System Table 2.10.22 Off-Gas System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the OGS is as        1. Inspections of the as-built system will be    1. The as-built OGS conforms with the basic shown on Figure 2.10.22.                            conducted.                                        configuration shown in Figure 2.10.22.
: 2. The OGS pressure-retaining components            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic tests retain their integrity under internal pressure      pressure-retaining components of the OGS.        demonstrate that the pressure-retaining that will be experienced during service.                                                            components of the OGS can retain their integrity under internal pressure that will be experienced during service.
: 3. Automatic OGS isolation is initiated when        3. Tests will be conducted on the as-built OGS    3. OGS automatically isolates when the radiation levels in the discharge stream            using a simulated radiation signal.              simulated signal exceeds the setpoint.
exceed the setpoint.
25A5675AA Revision 7
: 4. Main control room alarm provided for the        4. Inspections will be conducted on the main      4. Alarm exists in the main control room as OGS is as defined in Subsection 2.10.22.            control room alarm for the OGS.                  defined in Section 2.10.22.
: 5. The adsorption beds and their support            5. A seismic analysis of the adsorption beds      5. A structure analysis report exists which structures do not collapse under seismic            and their support structures will be              concludes that collapse of the adsorption loads corresponding to the SSE ground              performed.                                        beds and their support structures do not accelerations.                                                                                        occur.
: 6. The system pressure boundary of the OGS          6. A hydrostatic test of the OGS pressure        6. The OGS pressure boundary retains its is capable of withstanding an internal              boundary will be conducted in the plant with      integrity under the test conditions.
hydrogen explosion.                                test pressures equal to or greater than 1.5 Design Control Document/Tier 1 times design pressure.
2.10-41
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.10.23 Circulating Water System Design Description The Circulating Water (CW) System provides a supply of cooling water to the Main Condenser to remove the heat rejected by the turbine cycle and auxiliary systems and transport it to the power cycle heat sink. The parts of the CW System that are in the Turbine Building are within the Certified Design. Those parts of the system that are outside the Turbine Building are not in the Certified Design. Figure 2.10.23 shows the system basic configuration and scope of the CW System within the Certified Design.
The CW System is classified as non-safety-related.
For the CW System, condenser area water level sensors are provided. A high water level signal causes an alarm in the main control room (MCR). A high-high water level signal closes the condenser valves in the CW System.
The CW System motor operated valve position indications are provided in the main control room (MCR).
Interface Requirements The parts of the CW System (including the power cycle heat sink) which are not within the Certified Design shall meet the following requirements:
(1)  Design features shall be provided to limit flooding in the Turbine Building.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.10.23 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the parts of the CW Systems within the Certified Design.
2.10-42                                                                              Circulating Water System
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 TURBINE BUILDING CW        SITE NNS        SPECIFIC SCOPE M
CONDENSER M
NOTES:
: 1. MULTIPLE LINES MAY BE USED.
: 2. CONDENSER ISOLATION PROVISIONS ARE REQUIRED FOR EACH LINE.
Figure 2.10.23 Circulating Water System Circulating Water System                                                                2.10-43
 
ABWR 2.10-44 Table 2.10.23 Circulating Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. A basic configuration for the CW System is  1. Inspections of the as-built system will be  1. The as-built CW System conforms with the as shown on Figure 2.10.23.                    conducted.                                      basic configuration shown on Figure 2.10.23.
: 2. The circulating water condenser valves are  2. Testing of the as-built CW System will be    2. The circulating water condenser valves are closed in the event of a system isolation      performed using simulated signals.              closed in the event of a system isolation signal from the condenser area level                                                            signal from the condenser area level switches.                                                                                      switches.
: 3. MCR alarms and displays provided for the    3. Inspections will be performed on the MCR    3. Alarms and displays exist or can be retrieved CW System are as defined in Section            alarms for the CW System.                      in the MCR as defined in Section 2.10.23.
2.10.23.
25A5675AA Revision 7 Design Control Document/Tier 1 Circulating Water System
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.10.24 Condenser Cleanup Facility No entry for this system.
Condenser Cleanup Facility                                                          2.10-45
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.11.1 Makeup Water (Purified) System Design Description The Makeup Water (Purified) (MUWP) System is a distribution system with components located throughout the plant. The MUWP provides demineralized makeup water to the condensate storage tank, the surge tanks which are shared by the Reactor Building Cooling Water System and Heating, Ventilation, and Air Conditioning Emergency Cooling Water System and other plant systems.
The MUWP System consists of distribution piping and valves. Makeup water is supplied to the system by the Makeup Water Preparation System.
The MUWP System is classified as non-safety-related with the exception of the primary containment isolation function which is safety-related. The primary containment pipe penetration and isolation valves are classified as Seismic Category I and ASME Code Class 2.
The outboard containment isolation valve is a manual valve locked closed during standby, hot standby and power operation. The inboard containment isolation valve is a check valve (CV) that has an active safety-related function to close under system pressure, fluid flow, and temperature conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.1 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the MUWP System.
Makeup Water (Purified) System                                                                          2.11-1
 
ABWR 2.11-2 Table 2.11.1 Makeup Water (Purified) System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the safety-related 1. Inspections of the as-built safety-related    1. The as-built safety-related portion of the portion of the MUWP System is as described      portions of the MUWP System will be              MUWP System conforms with the basic in Section 2.11.1.                              conducted.                                      configuration described in Section 2.11.1.
: 2. The CV designated in Section 2.11.1 as        2. Tests of the installed valve for closing will be 2. The CV closes.
having an active safety-related function        conducted under system preoperational closes under system pressure, fluid flow, and    pressure, fluid flow, and temperature temperature conditions.                          conditions.
25A5675AA Revision 7 Design Control Document/Tier 1 Makeup Water (Purified) System
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.11.2 Makeup Water (Condensate) System Design Description The Makeup Water (Condensate) (MUWC) System is a distribution system with components located throughout the plant. Figure 2.11.2 shows the basic system configuration and scope.
Except for the level sensors and associated piping, the MUWC System is classified as non-safety-related.
The level sensors and associated piping are classified as Seismic Category I. Figure 2.11.2 shows the ASME Code class for the MUWC System piping and components.
The level instruments are located in the Reactor Building; the condensate storage tank (CST) and pump(s) are located outside the Reactor Building.
Each of the four MUWC System water level sensors is powered from the respective divisional Class 1E power supply. In the MUWC System, independence is provided between the Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The MUWC System has displays for CST water level in the main control room.
MUWC System components with display interfaces with the Remote Shutdown System (RSS) are shown on Figure 2.11.2.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.2 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the MUWC System.
Makeup Water (Condensate) System                                                                        2.11-3
 
ABWR 2.11-4 CRD MUWC              MUWC MUWP NNS                NNS MUWC RW/B NNS T/B R
NOTE 1 & 2  L                      CST 2 NNS R/B 25A5675AA Revision 7 2
CRD MUWC RCIC NNS MUWC HPCF HPCF Makeup Water (Condensate) System Design Control Document/Tier 1 HPCF NOTES:
: 1. ONE SENSOR ASSIGNED TO EACH OF FOUR CLASS 1E DIVISIONS.
: 2. RSS INTERFACE IS FOR DIVISION I AND DIVISION II LEVEL SENSOR.
Figure 2.11.2 Makeup Water (Condensate) System
 
ABWR Makeup Water (Condensate) System Table 2.11.2 Makeup Water (Condensate) (MUWC) System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the MUWC          1. Inspections of the as-built system will be      1. The as-built MUWC System conforms with System is as shown on Figure 2.11.2.            conducted.                                          the basic configuration on Figure 2.11.2.
: 2. The ASME Code components of the MUWC 2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary          Code components of the MUWC System          ASME Code components of the MUWC integrity under internal pressures that will be required to be hydrostatically tested by the System conform with the requirements in the experienced during service.                    ASME Code.                                  ASME Code, Section III.
: 3. Each of the four MUWC System water level      3.                                                  3.
sensors is powered from the respective
: a. Tests will be performed on the MUWC              a. The test signal exists only in the Class divisional Class 1E power supply. In the System by providing a test signal in only          1E division under test in the MUWC MUWC System, independence is provided one Class 1E division at a time.                    System.
between Class 1E divisions, and between 25A5675AA Revision 7 Class 1E divisions and non-Class 1E                b. Inspections of the as-built Class 1E            b. In the MUWC System, physical equipment.                                            divisions in the MUWC System will be                separation or electrical isolation exists performed.                                          between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 4. Main control room displays provided for the  4. Inspections will be performed on the main  4. Displays exist or can be retrieved in the main MUWC System are as defined in Section            control room displays for the MUWC System. control room as defined in Section 2.11.2.
2.11.2 Design Control Document/Tier 1
: 5. RSS displays provided for the MUWC            5. Inspections will be performed on the RSS        5. Displays exist on the RSS as defined in System are as defined in Section 2.11.2.        displays for the MUWC System.                      Section 2.11.2.
2.11-5
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.11.3 Reactor Building Cooling Water System Design Description The Reactor Building Cooling Water (RCW) System distributes cooling water through three physically separated and electrically independent divisions. The system removes heat from plant auxiliaries and transfers it to the Ultimate Heat Sink (UHS) via the Reactor Service Water (RSW) System. The RCW System removes heat from emergency core cooling equipment, including the emergency diesel generators (DGs) during a safe reactor shutdown cooling function. RCW System configurations are shown in Figures 2.11.3a, 2.11.3b, and 2.11.3c.
Figure 2.11.3d shows the RCW System control interfaces. All components cooled by the RCW System are parts of other systems and are not part of the RCW System. Each RCW division includes two pumps which circulate cooling water through the equipment cooled by the RCW System and through three heat exchangers which transfer the RCW heat to the UHS via the RSW System.
The RCW System performs a safe reactor shutdown cooling function following either a loss-of-coolant accident(LOCA) or a loss-of-preferred-power (LOPP)or both. Assuming a single active failure in any mechanical or electrical division or RCW support system, which disables any one of the three RCW divisions, the other two divisions perform safe reactor shutdown cooling.
Tables 2.11.3a, 2.11.3b, and 2.11.3c show which equipment receives RCW flow during various plant operating and emergency conditions. The tables also indicate how many heat exchangers are in service under each condition.
The RCW System is classified as safety-related except for those portions as shown on Figures 2.11.3a, 2.11.3b, and 2.11.3c as non-nuclear safety.
The RCW System responses to a LOCA signal are the following:
(1)  Starts any standby RCW pumps.
(2)  Opens any closed standby RCW heat exchanger outlet valves.
(3)  Opens all Residual Heat Removal (RHR) System heat exchanger cooling water outlet valves.
(4)  Closes all RCW containment isolation valves.
(5)  Closes valves to the following non-safety-related components (to Reactor Water Cleanup System (CUW) and reactor internal pump (RIP) MG sets).
(6)  Opens the RCW water temperature pneumatic control valves (located downstream of RCW heat exchangers) and closes the RCW heat exchanger bypass valves.
2.11-6                                                                    Reactor Building Cooling Water System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 (7)    Overrides the RCW pump trip signal from low surge tank and low stand pipe level.
Safety-related valves separate the safety-related portions of the RCW System from the non-safety-related portions of the system. The separation valves to the non-safety-related RCW System are automatically or remote-manually operated, and their positions are indicated in the main control room.
Component design parameters are:
Division A/B              Division C Discharge flow rate (per pump)                    1420 m3/h                1237 m3/h Heat exchanger design basis heat removal          47.73 GJ/h              44.38 GJ/h capacities:(per heat exchanger)
These heat removal capabilities include a 20% margin above the minimum required for design basis accident conditions. Consequently, plant operation is acceptable with heat exchanger capacities greater than or equal to 80% of these values.
Figures 2.11.3a, 2.11.3b, and 2.11.3c show the ASME Code Class for the RCW System piping and components. The safety-related portions of the RCW divisions are classified as Seismic Category I. The piping to the fuel pool cooling (FPC) system heat exchangers and room coolers are classified as Seismic Category I.
The RCW pumps and heat exchangers are located in the lower floors of the Control Building.
The equipment cooled by the RCW divisions are located in the Control Building, Reactor Building, Turbine Building, and Radwaste Building, (Figures 2.11.3a, 2.11.3b, and 2.11.3c).
Each of the three RCW divisions is powered from its respective Class 1E division as shown in Figures 2.11.3a, 2.11.3b, and 2.11.3c. In the RCW System, independence is provided between the Class 1E divisions and also between the Class 1E divisions and non-Class 1E equipment.
The safety-related portion of each mechanical division of the RCW System (Divisions A, B, C) is physically separated from the safety-related portions of the other divisions.
The RCW System has the following displays and controls in the main control room:
(1)    Parameter displays for instruments shown on Figures 2.11.3a, 2.11.3b, and 2.11.3c.
(2)    Controls and status displays for the RCW active safety-related components shown on Figures 2.11.3a, 2.11.3b, and 2.11.3c.
The RCW System components with displays and control interfaces with the Remote Shutdown System (RSS) are identified in Figures 2.11.3a and 2.1.3b.
Reactor Building Cooling Water System                                                                  2.11-7
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 The safety-related electrical equipment shown on Figures 2.11.3a, 2.11.3b, and 2.11.3c, located in the Reactor Building, is qualified for a harsh environment.
The motor-operated valves (MOVs) shown on Figures 2.11.3a, 2.11.3b, and 2.11.3c have active safety-related functions to open, close, or both open and close, and perform these functions under differential pressure, fluid flow, and temperature conditions.
The check valves (CVs) shown on Figures 2.11.3a, 2.11.3b, and 2.11.3c have active safety-related functions to open, close, or both open and close under system pressure, fluid flow, and temperature conditions.
A separate surge tank of at least 16m3 is provided for each RCW division. Each surge tank is shared with the corresponding division of the HVAC Emergency Cooling Water (HECW)
System. Makeup water is provided for the surge tank by the Makeup Water (Purified) (MUWP)
System by an automatic or main control room signal. Low water level signals in the surge tanks do the following (in order of decreasing level):
(1)  Lowopens the MUWP makeup water valve.
(2)  Low-Low closes the pneumatic and motor-operated valves which stop flow to the non-safety-related components.
The Suppression Pool Cleanup (SPCU) System provides a backup surge tank water supply.
The pneumatic-operated valves shown in Figures 2.11.3a, 2.11.3b, and 2.11.3c fail as follows in the event that either electric power to the valve-actuating solenoid is lost or pneumatic pressure to the valve is lost: RCW makeup valves from the MUWP fail open, RCW water temperature control valves fail open, RCW heat exchanger bypass valves fail closed, and the safety-related/non-safety-related separation valve fails closed.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.3d provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria, which will be undertaken for the RCW System.
2.11-8                                                                  Reactor Building Cooling Water System
 
25A5675AA Revision 7 ABWR                                                                                      Design Control Document/Tier 1 M      MUWP      RCW RCW    OTHERS                            OTHERS RCW                                          3 3                                                  3                                      P SPCU RCW FE                                                          T    M R        3 RHR HX (Reactor Building)
L M R                SURGE TANK DG HX                                                  (Reactor Building)      L (Reactor Building) 3 RCW M                                                                M                                          HECW Fuel Pool Cooling HX and Room Coolers (Reactor Building)                                            L                  TO 3 NNS                                                        NNS 3                                    HECW OTHER (SAFETY-RELATED) HXs (Reactor and Control Building)
OTHERS RCW RCW      OTHERS                                    NNS P    FE    M R NNS                                                              M CRD AND CUW PUMPS (Reactor Building) 3 NNS                                                      NNS 3 M
CUW HX
( Reactor Building)
M NON-SAFETY-RELATED HXs
( Control Building)
R                                              NON-SAFETY-RELATED HXs FE                                                      (Turbine Building)
M
* M
* M T                                            DRYWELL EQUIPMENT COOLERS (Reactor Building)
NNS 2          2 NNS                                                NNS 2                2 NNS R M P
P RCW HX                                                            R FROM          (Control Building)      3 RCW RSW                                        RSW R M                                  TO RSW                                    RCW PUMP (Control Building)
RCW HX (Control Building)        3 RCW FROM                                      RSW RSW R M                                  TO RSW R
RCW HX (Control Building)        3 RCW                                      RCW PUMP FROM                                        RSW                                    (Control Building)
RSW TO RSW NOTES:
                                                                                            = PRIMARY CONTAINMENT
: 1. ALL ELECTRICAL POWER LOADS FROM THE CLASS 1E COMPONENTS SHOWN ON THIS FIGURE ARE POWERED FROM CLASS 1E DIVISION I EXCEPT FOR THE
* OUTBOARD CONTAINMENT ISOLATION VALVE, WHICH IS POWERED FROM DIVISION II.
Figure 2.11.3a Reactor Building Cooling Water System (RCW-A)
Reactor Building Cooling Water System                                                                                              2.11-9
 
25A5675AA Revision 7 ABWR                                                                                      Design Control Document/Tier 1 M        MUWP RCW 3
OTHERS RCW RCW    OTHERS                                        3 3                                                                                        P RCW FE                                                            T    M R SPCU 3
RHR HX (Reactor Building)
L M R                SURGE TANK (Reactor Building)    L DG HX (Reactor Building) 3 RCW M                                                                                                              HECW M
                          `
Fuel Pool Cooling HX and Room Coolers (Reactor Building)                                          L 3 NNS                                                            NNS 3 OTHER (SAFETY-RELATED) HXs (Reactor and Control Building)
RCW    OTHERS                          OTHERS  RCW NNS                                              NNS P      FE    M                                                                    M R
CUW PUMP (Reactor Building) 3 NNS                                                        NNS 3 M
CUW Hx
( Reactor Building)
M NON-SAFETY-RELATED HXs R                                        (Control, Reactor, and Turbine Building)
FE M
* M
* M T
DRYWELL EQUIPMENT COOLERS 2 NNS                    ( Reactor Building)
NNS 2                                                                      NNS 2              2 NNS R    M P                                                            P RCW HX                                                        R (Control Building)        3 RCW FROM                                        RSW RSW R M                                    TO RSW                                  RCW PUMP (Control Building)
RCW HX (Control Building)      3 RCW FROM                                        RSW R M RSW                                TO RSW R
RCW HX (Control Building)        3 RCW FROM                                        RSW RSW                                                                      RCW PUMP TO RSW                              (Control Building)
NOTES:
: 1. THIS DIVISION IS POWERED FROM CLASS 1E DIVISION II, EXCEPT FOR THE CONTAINMENT OUTBOARD ISOLATION VALVE, WHICH IS POWERED FROM DIVISION III.
                                                                                              *  =PRIMARY CONTAINMENT Figure 2.11.3b Reactor Building Cooling Water System (RCW-B) 2.11-10                                                                                      Reactor Building Cooling Water System
 
25A5675AA Revision 7 ABWR                                                                                  Design Control Document/Tier 1 M
MUWP RCW 3
RCW OTHERS                                OTHERS RCW                SPCU RCW              P 3                                                3                    3 FE                                                          T        M RHR HX                                                                    L (Reactor Building)
SURGE TANK (Reactor Building)      L M
3 RCW DG HX                                                                      HECW (Reactor Building)
TO L                HECW OTHER (SAFETY-RELATED) HXs (Reactor and Control Building)
RCW      OTHERS                            OTHERS  RCW FE            NNS                                                  NNS P        M                                                                          M CRD PUMP 3
(Reactor Building)                  NNS    3 NNS M
NON-SAFETY-RELATED HXs (Radwaste & Turbine Building)
NON-SAFETY-RELATED HXs (Turbine Building)
FE T
M P                                                              P RCW HX FROM          (Control Building)        3 RCW RSW                                        RSW M                                TO RSW                                      RCW PUMP (Control Building)
RCW HX (Control Building)        3 RCW FROM RSW                                        RSW M                                  TO RSW RCW HX (Control Building)        3 RCW FROM RSW                                        RSW                          RCW PUMP TO RSW                                  (Control Building)
NOTES:
: 1. ALL ELECTRICAL POWER LOADS FOR THE CLASS 1E COMPONENTS SHOWN ON THIS FIGURE ARE POWERED FROM CLASS 1E DIVISION III.
Figure 2.11.3c Reactor Building Cooling Water System (RCW-C)
Reactor Building Cooling Water System                                                                                    2.11-11
 
ABWR 2.11-12 LOCAL AREA                                  MAIN CONTROL ROOM                                LOCAL AREA Plant Sensors                                                                                Device Actuators RCW Manual Pump and Valve Controls SSLC PROCESSING RCW SYSTEM LOGIC                                Automatic:
                                                                                                - Sensor Channel Trip Decision                      - LOCA Alignment 25A5675AA Revision 7
                                                                                                - System Coincidence Trip Decision            RCW - Surge Tank Level Control RCW Surge Tank Level                        - Control and Interlock Logic
                                                                                                - Division-of-Sensors Bypass                        - Stop Flow to Non-Safety-Related
                                                                                                - Calibration, Self-Diagnosis                        Components RSW  LOCA Signal          RHR    LOCA Signal RCW Manual Pump and Valve Actuation Reactor Building Cooling Water System                                                                                                                                                    Design Control Document/Tier 1 Notes:
: 1. Diagram represents one of three RCW divisions.
: 2. See Section 3.4, Figure 3.4b for SSLC processing.
Figure 2.11.3d Reactor Building Cooling Water System Control Interface Diagram
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 Table 2.11.3a          Reactor Building Cooling Water Cooling Loads Division A Hot Standby Normal                        (loss of Operating                        AC      Emergency Operating Mode/Components*                        Conditions Shutdown            Power)        (LOCA)
RCW/RSW Heat Exchangers In Service                      2            3              3            3 SAFETY-RELATED Emergency Diesel Generator A RHR Heat Exchanger A Others (safety-related)f NON-SAFETY-RELATED CUW Heat Exchanger FPC Heat Exchanger A**
Inside Drywell Others (non-safety-related)
* Some of these cooling loads are serviced by only one or two RCW divisions. These components may be reassigned to other RCW divisions if redundancy and divisional alignment of supported and supporting systems is maintained and the design basis cooling capacity of the RCW divisions is assured.
Equipment does not receive RCW in this mode.
Equipment receives RCW in this mode.
f HECW refrigerators, room coolers (RHR, RCIC, CAMS), RHR motor bearing and seal coolers, and CAMS cooler.
    ** Includes FPC room cooler.
Reactor Building Cooling Water System                                                                    2.11-13
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Table 2.11.3b Reactor Building Cooling Water Cooling Loads Division B Hot Standby Normal                      (loss of Operating                        AC          Emergency Operating Mode/Components*                  Conditions    Shutdown        Power)          (LOCA)
RCW/RSW Heat Exchangers In                        2              3            3                3 Service SAFETY-RELATED Emergency Diesel Generator B RHR Heat Exchanger B Others (safety-related)f NON-SAFETY-RELATED RWCU Heat Exchanger FPC Heat Exchanger B**
Inside Drywell Others (non-safety-related)
* Some of these cooling loads are serviced by only one or two RCW divisions. These components may be reassigned to other RCW divisions if redundancy and divisional alignment of supported and supporting systems is maintained and the design basis cooling capacity of the RCW divisions is assured.
Equipment does not receive RCW in this mode.
Equipment receives RCW in this mode.
f HECW refrigerators, room coolers (RHR, HPCF, SGTS, FCS, CAMS), RHR and HPCF motor bearing and seal coolers, and CAMS cooler.
        ** Includes FPC room cooler.
2.11-14                                                                        Reactor Building Cooling Water System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 Table 2.11.3c Reactor Building Cooling Water Cooling Loads Division C Hot Standby Normal                      (loss of Operating                        AC      Emergency Operating Mode/Components*                          Conditions Shutdown            Power)        (LOCA)
RCW/RSW Heat Exchangers In Service                        2              3            3            3 SAFETY-RELATED Emergency Diesel Generator C RHR Heat Exchanger C Others (safety-related)f NON-SAFETY-RELATED Others (Non-safety-related)
* Some of these cooling loads are serviced by only one or two RCW divisions. These components may be reassigned to other RCW divisions if redundancy and divisional alignment of supported and supporting systems is maintained and the design basis cooling capacity of the RCW divisions is assured.
Equipment does not receive RCW in this mode.
Equipment receives RCW in this mode.
f HECW refrigerators; SGTS and FCS room coolers; room coolers, motor bearing coolers, and mechanical seal coolers for RHR and HPCF.
Reactor Building Cooling Water System                                                                        2.11-15
 
ABWR 2.11-16 Table 2.11.3d Reactor Building Cooling Water (RCW) System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the RCW System 1. Inspections of the as-built system will be          1. The as-built RCW System conforms with the is as shown on Figures 2.11.3a, 2.11.3b and  conducted.                                            basic configuration shown in Figures 2.11.3c.                                                                                            2.11.3a, 2.11.3b and 2.11.3c.
: 2. The ASME Code components of the RCW            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the RCW System                ASME Code components of the RCW integrity under internal pressures that will be    required to be hydrostatically tested by the    System conform with the requirements in the experienced during service.                        ASME Code.                                      ASME Code, Section III.
: 3. The RCW System responses to a LOCA              3. Using simulated LOCA signals, tests will be  3. Upon receipt of simulated LOCA signals, the signal are as specified in Section 2.11.3.        performed for the RCW System.                    responses of the RCW System are as specified in Section 2.11.3.
25A5675AA Revision 7
: 4. The RCW pump flow capacities and the          4. An analysis of the as-built RCW System will 4. The estimated heat removal capacities of the RCW heat exchanger heat removal                  be performed. Tests will be performed of the  as-built RCW System divisions exceed the capacities are as specified in Section 2.11.3. flow capacities of the installed RCW pumps. estimated heat removal requirements of the Inspections and analyses will be performed    components cooled by the RCW System to estimate the heat removal capacities of    divisions during LOCA conditions.
the RCW heat exchangers. Inspections and analyses will be performed to estimate the heat removal requirements of the as-built components which are cooled by the RCW System during LOCA conditions.
Design Control Document/Tier 1
: 5. Each of the three RCW divisions is powered 5.                                              5.
Reactor Building Cooling Water System from its respective Class 1E division as
: a. Tests will be performed on the RCW          a. The test signal exists only in the Class shown in Figures 2.11.3a, 2.11.3b, and System by providing a test signal in only      1E division under test in the RCW 2.11.3c. In the RCW System, independence one Class 1E division at a time.                System.
is provided between the Class 1E divisions and also between the Class 1E divisions and  b. Inspections of the as-installed Class 1E    b. Physical separation or electrical isolation non-Class 1E equipment.                          Divisions in the RCW System will be            exists between Class 1E divisions in the performed.                                      RCW System. Physical separation or electrical isolation exists between Class 1E divisions and non-Class 1E equipment.
 
Table 2.11.3d Reactor Building Cooling Water (RCW) System (Continued)
ABWR Reactor Building Cooling Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 6. The safety-related portion of each              6. Inspections of the as-built RCW System will      6. The safety-related portions of each mechanical division of the RCW System              be performed.                                      mechanical division of the RCW System is (Divisions A, B,C) is physically separated                                                            physically separated from the safety-related from the safety-related portions of the other                                                          portions of the other mechanical divisions of divisions.                                                                                            the RCW System.
: 7. Main control room displays and controls    7. Inspections will be performed on the main            7. Displays and controls exist or can be provided for the RCW System are as defined    control room displays and controls for the              retrieved in the main control room as defined in Section 2.11.3.                            RCW System.                                              in Section 2.11.3.
: 8. RSS displays and controls provided for the      8. Inspections will be performed on the RSS        8. Displays and controls exist on the RSS as RCW system are as defined in Section              displays and controls for the RCW System.          defined in Section 2.11.3.
2.11.3.
25A5675AA Revision 7
: 9. MOVs designated in Section 2.11.3 as            9. Tests of installed valves for opening and        9. Upon receipt of the actuation signal, each having an active safety-related function will      closing, will be conducted under pre-              MOV opens, closes, or both opens and open, close, or both open and close under          operational differential pressure, fluid flow,      closes, depending upon the valves safety differential pressures, fluid flow, and            and temperature conditions.                        functions.
temperature conditions.
: 10. CVs, designated in Section 2.11.3 as having 10. Tests of installed valves for opening, closing, 10. Based on the direction of the differential an active safety-related function, open,        or both opening and closing, will be                pressure across the valve, each CV opens, close, or both open and close under system      conducted under system preoperational              closes, or both opens and closes, depending pressure, fluid flow, and temperature          pressure, fluid flow, and temperature              upon the valves safety function.
Design Control Document/Tier 1 conditions.                                    conditions.
: 11. The pneumatic-operated valves shown in          11. Tests will be performed on the as-built valves 11. The pneumatic actuated valves listed below Figures 2.11.3a, 2.11.3b, and 2.11.3c fail as      by initiating loss of pneumatic pressure and      fail as desired when either electric power to follows in the event that either electric power    power to the actuating solenoids.                  the valve actuating solenoid is lost or to the valve actuating solenoid is lost or                                                            pneumatic pressure to the valve is lost:
pneumatic pressure to the valve is lost:                                                              MUWP makeup water valves fail open, RCW MUWP makeup valves fail open, RCW water                                                                water temperature control valves fail open, temperature control valves fail open, RCW                                                              RCW heat exchanger bypass valves fail heat exchanger bypass valves fail closed,                                                              closed, and the safety-related/non-safety-and the safety-related/non-safety-related                                                              related separation valves fail closed.
2.11-17 separation valves fail closed.
 
Table 2.11.3d Reactor Building Cooling Water (RCW) System (Continued)
ABWR 2.11-18 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 12. A surge tank with a capacity of greater than  12. Inspection and a volume calculation using  12. The capacity of the surge tanks is greater or equal to 16 m3 is provided for each RCW        as-built dimensions will be performed.          than or equal to 16 m3.
division.
: 13. A low surge tank water level signal opens the 13. Tests will be performed on the as-built      13. The MUWP makeup valve opens and MUWP makeup valve and closes the                  equipment.                                      pneumatic and motor-operated valves which pneumatic and motor-operated valves which                                                          stop flow to the non-safety-related stop flow to the non-safety-related                                                                components close upon receipt of a low components.                                                                                        surge tank water level signal.
25A5675AA Revision 7 Reactor Building Cooling Water System                                                                                                                                                      Design Control Document/Tier 1
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.11.4 Turbine Building Cooling Water System Design Description The Turbine Building Cooling Water (TCW) System removes heat from the auxiliary equipment in the Turbine Building and rejects this heat to the Turbine Service Water (TSW)
System. Figure 2.11.4 shows the basic system configuration and scope.
The TCW System is classified as a non-safety-related.
The TCW System is located inside the Turbine Building.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.4 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the TCW System.
Turbine Building Cooling Water System                                                                    2.11-19
 
ABWR 2.11-20 HNCW TCW  TCW MUWP NNS  NNS SURGE TANK 25A5675AA Revision 7 TSW TCW NNS      HEAT EXCHANGER TSW TCW NNS      HEAT            TO FROM                                                                              TURBINE BUILDING EXCHANGER TURBINE BUILDING                                                                  AUXILIARY EQUIPMENT AUXILIARY EQUIPMENT Turbine Building Cooling Water System                                                                                                          Design Control Document/Tier 1 TSW TCW NNS      HEAT EXCHANGER Figure 2.11.4 Turbine Building Cooling Water System
 
ABWR Turbine Building Cooling Water System Table 2.11.4 Turbine Building Cooling Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                    Acceptance Criteria
: 1. The basic configuration for the TCW System 1. Inspection of the as-built system will be  1. The as-built TCW System conforms with the is as shown on Figure 2.11.4.                conducted.                                    basic configuration shown on Figure 2.11.4.
25A5675AA Revision 7 Design Control Document/Tier 1 2.11-21
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.11.5 HVAC Normal Cooling Water System Design Description The Heating Ventilating and Air Conditioning (HVAC) Normal Cooling Water (HNCW)
System delivers chilled water to the Drywell Cooling System and to non-safety-related fan coil units of building HVAC systems. Figure 2.11.5 shows the basic system configuration and scope.
The HNCW System is classified as non-safety-related with the exception of the primary containment isolation function.
The HNCW System pumps and refrigerators are located in the Turbine Building.
The primary containment penetrations and isolation valves are classified as Seismic Category I, and ASME Code Class 2.
The inboard containment isolation valves is powered from Class 1E Division II, and the outboard isolation valves are powered from Class 1E Division I. In the HNCW System, independence is provided is between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The main control room has control and open/close status indication for the primary containment isolation valves.
The safety-related electrical equipment that provides primary containment isolation and is located in the primary containment and the Reactor Building is qualified for a harsh environment.
The primary containment isolation motor-operated valves (MOVs) shown on Figure 2.11.5 have active safety-related function to close and perform this function under differential pressure, fluid flow, and temperature conditions.
The check valve (CV) for containment isolation shown on Figure 2.11.5 has an active safety-related function to close under system pressure, fluid flow, and temperature conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.5 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the HNCW System.
2.11-22                                                                    HVAC Normal Cooling Water System
 
25A5675AA Revision 7 ABWR                                                      Design Control Document/Tier 1 Figure 2.11.5 HVAC Normal Cooling Water System HVAC Normal Cooling Water System                                                  2.11-23
 
ABWR 2.11-24 Table 2.11.5 HVAC Normal Cooling Water (HNCW) System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the HNCW              1. Inspections of the as-built system will be      1. The as-built HNCW System conforms with System is as shown on Figure 2.11.5.                conducted.                                        the basic configuration shown in Figure 2.11.5.
: 2. The ASME Code components of the HNCW 2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the retain their pressure boundary integrity under Code components of the HNCW System          ASME Code components of the HNCW internal pressures that will be experienced    required to be hydrostatically tested by the System conform with the requirements in the during service.                                ASME Code.                                  ASME Code, Section III.
: 3. The inboard containment isolation valves is 3.                                              3.
powered from Class 1E Division II, and the
: a. Tests will be performed on the HNCW          a. The test signal exists only in the Class outboard isolation valves are powered from System by providing a test signal in only      1E division under test in the HNCW Class 1E Division I. In the HNCW System, 25A5675AA Revision 7 one Class 1E division at a time.                System.
independence is provided between Class 1E divisions, and between Class 1E divisions and non-Class 1E equipment.
: b. Inspection of the as-installed Class 1E        b. In the HNCW System, physical divisions in the HNCW System will be              separation or electrical isolation exists performed.                                        between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
Design Control Document/Tier 1
: 4. Main control room displays and controls    4. Inspections will be performed on the main            4. Displays and controls exist or can be provided for HNCW System are as defined in    control room displays and controls for the              retrieved in main control room as defined in HVAC Normal Cooling Water System Section 2.11.5.                              HNCW System.                                            Section 2.11.5.
: 5. MOVs designated in Section 2.11.5 as            5. Tests of installed valves for closing will be  5. Upon receipt of the actuating signal, each having an active safety-related function,          conducted under preoperational differential        MOV closes.
close under differential pressure, fluid flow,      pressure, fluid flow, and temperature and temperature conditions.                        conditions.
 
Table 2.11.5 HVAC Normal Cooling Water (HNCW) System ABWR HVAC Normal Cooling Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 6. The CV designated in Section 2.11.5 as        6. Tests of the installed valve for closing will be 6. The CV closes.
having an active safety-related function        conducted under system preoperational closes under system pressure, fluid flow, and    pressure, fluid flow, and temperature temperature conditions.                          conditions.
25A5675AA Revision 7 Design Control Document/Tier 1 2.11-25
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.11.6 HVAC Emergency Cooling Water System Design Description The Heating Ventilating and Air Conditioning (HVAC) Emergency Cooling Water (HECW)
System delivers chilled water to the:
(1)  Control Room Habitability Area HVAC System.
(2)  Control Building Safety-Related Equipment Area HVAC System.
(3)  Reactor Building HVAC System (safety-related electrical equipment HVAC).
Figures 2.11.6a and 2.11.6b show the basic system configuration and scope.
The HECW System is classified as safety-related except for the chemical addition tank and associated piping and valves.
The HECW System is manually initiated.
Each HECW System refrigerator unit has a capacity of not less than 2.43 GJ/h. In Division A, the refrigerator unit on standby automatically starts if the other refrigerator unit is stopped. In Divisions B and C, any refrigerator unit on standby automatically starts if any of the other refrigerator units in Division B or C is stopped.
Safety-related portions of the HECW System are classified as Seismic Category I. Figures 2.11.6a and 2.11.6b show the ASME Code class for the HECW System piping and components.
The HECW System pumps and refrigerator units are located in the Control Building.
Each of the three HECW System divisions is powered from the respective Class 1E divisions as shown on Figures 2.11.6a and 2.11.6b. In the HECW System, independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
Except for the connections to the chemical addition tanks, each mechanical division of the HECW System (Divisions A, B, C) is physically separated from the other divisions.
The HECW System has the following main control room (MCR) displays and controls:
(1)  Control and status indications for the refrigerator units and pumps shown on Figure 2.11.6a and 2.11.6b.
(2)  Parameter displays for instruments shown on Figures 2.11.6a and 2.11.6b.
The check valves (CVs) shown on Figures 2.11.6a and 2.11.6b have active safety-related functions to open, close, or both open and close under system pressure, fluid flow, and temperature conditions.
2.11-26                                                                  HVAC Emergency Cooling Water System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 The pneumatic-operated valves shown in Figures 2.11.6a and 2.11.6b fail as follows in the event that either electric power to the valve-actuating solenoid is lost or pneumatic pressure to the valve is lost: the differential pressure control valves fail closed, and the flow control valves to the cooling coils fail open.
To address the beyond-design-basis event of a postulated aircraft impact, design features provide mechanical cross connects (HECW-Division A), along with electrical power manual alignments from Division I to MCR HVAC (Div III / C) fans and components.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.6 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the HECW System.
HVAC Emergency Cooling Water System                                                                      2.11-27
 
25A5675AA Revision 7 ABWR                                                                              Design Control Document/Tier 1 HECW HVAC                    HVAC HECW FROM SURGE TANK            3                            3 P
RCW                            REACTOR BUILDING HVAC HECW 3                          SYSTEM (SAFETY-RELATED ELECTRICAL EQUIPMENT HVAC)
P P
CONTROL BUILDING SAFETY-RELATED EQUIPMENT AREA HVAC SYSTEM P
dP RETURN FROM CRHA AREA HVAC SYSTEM C                3 NNS  TO CHEMICAL                        NNS 3 ADDITION TANK                                          SUPPLY TO CRHA AREA HVAC SYSTEM C T
REFRIGERATOR 3 HECW RCW REFRIGERATOR 3 HECW RCW NOTES:
: 1. DIVISION A IS POWERED FROM CLASS 1E, DIVISION I.
Figure 2.11.6a HVAC Emergency Cooling Water System (HECW-A) 2.11-28                                                                            HVAC Emergency Cooling Water System
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 HECW HVAC                      HVAC HECW FROM SURGE TANK                3                                3 P
RCW                                REACTOR BUILDING HVAC HECW 3                            SYSTEM (SAFETY-RELATED ELECTRICAL EQUIPMENT HVAC)
RETURN TO HECW A FROM CRHA AREA HVAC SYSTEM C P
CONTROL ROOM HABITABILITY AREA HVAC SYSTEM SUPPLY FROM HECW A TO P    CRHA AREA CONTROL BUILDING                HVAC SYSTEM C SAFETY-RELATED EQUIPMENT AREA HVAC SYSTEM P
dP 3 NNS                    TO CHEMICAL          NNS 3 ADDITION TANK T
REFRIGERATOR 3 HECW RCW REFRIGERATOR 3 HECW RCW NOTES:
: 1. THIS FIGURE SHOWS ONE OF TWO SIMILAR DIVISIONS.
DIVISIONS B AND C ARE POWERED FROM CLASS 1E, DIVISION II AND III, RESPECTIVELY.
Figure 2.11.6b HVAC Emergency Cooling Water System (HECW-B and C)
HVAC Emergency Cooling Water System                                                              2.11-29
 
ABWR 2.11-30 Table 2.11.6 HVAC Emergency Cooling Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration for the HECW          1. Visual inspections of the as-built system    1. The as-built configuration of the HECW System is shown on Figures 2.11.6a and            configuration will be conducted.                System is in accordance with 2.11.6b.                                                                                          Figures 2.11.6a and 2.11.6b.
: 2. The ASME Code components of the HECW          2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their integrity under internal      Code components of the HECW System              ASME Code components of the HECW pressures that will be experienced during        required to be hydrostatically tested by the    System conform with the requirements in the service.                                          ASME Code.                                      ASME Code, Section III.
: 3. Each HEWC System refrigerator unit has a      3. Type tests will be conducted on an as-built  3. Each HEWC System refrigerator unit has a capacity of not less than 2.43 GJ/h.              HECW System refrigerator units at a test        capacity of not less than 2.43 GJ/h.
facility.
25A5675AA Revision 7
: 4. In Division A, the refrigerator unit on standby 4. Tests will be conducted on each as-built    4. In Division A, the refrigerator unit on standby automatically starts if the other refrigerator    HECW System refrigerator unit in Divisions      automatically starts upon receipt of a unit is stopped. In Divisions B and C, any        A, B and C, using simulated signals            simulated signal indicating that the other refrigerator unit on standby automatically        indicating another refrigerator unit is        refrigerator unit is stopped. In Divisions B starts if any of the other refrigerator units in  stopped.                                        and C, the refrigerator unit on standby Divisions B or C is stopped.                                                                      automatically starts upon receipt of a simulated signal indicating that any of the other refrigerator units in Divisions B or C is stopped.
: 5. Each of the three HECW System divisions is 5.                                              5.
HVAC Emergency Cooling Water System                                                                                                                                                          Design Control Document/Tier 1 powered from the respective Class 1E
: a. Tests will be performed on the HECW          a. The test signal exists only in the Class divisions as shown on Figures 2.11.6a and System by providing a test signal in only      1E division under test in the HECW 2.11.6b. In the HECW System, one Class 1E division at a time.                System.
independence is provided between Class 1E divisions, and between Class 1E divisions    b. Inspections of the as-built Class 1E        b. In the HECW System, physical and non-Class 1E equipment.                      divisions in the HECW System will be            separation or electrical isolation exists performed.                                      between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
 
Table 2.11.6 HVAC Emergency Cooling Water System (Continued)
ABWR HVAC Emergency Cooling Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 6. Except for the connections to the chemical    6. Inspections of the as-built HECW System will 6. Each mechanical division of the HECW addition tank, each mechanical division of        be conducted.                                  System is physically separated from the the HECW System (Divisions A, B, C) is                                                            other mechanical divisions of the HECW physically separated from the other divisions.                                                    System by structural and/or fire barriers, with the exception connections to the chemical addition tank.
: 7. Main control room displays and controls        7. Inspections will be performed on the main      7. Displays and controls exist or can be provided for the HECW System are as              control room displays and controls for the        retrieved in the main control room as defined defined in Section 2.11.6.                        HECW System.                                      in Section 2.11.6.
: 8. CVs designated in Section 2.11.6 as having 8. Tests of installed valves for opening, closing, 8. Based on the direction of the differential an active safety-related function open, close, or both opening and closing, will be              pressure across the valve, each CV opens, 25A5675AA Revision 7 or both open and close under system            conducted under system preoperational            closes, or both opens and closes, depending pressure, fluid flow, and temperature          pressure, fluid flow, and temperature            upon the valves safety functions.
conditions.                                    conditions.
: 9. The pneumatic-operated valves shown in          9. Tests will be performed on the as-built valves 9. The pneumatic actuated valves listed below Figures 2.11.6a and 2.11.6b fail as follows in    by initiating loss of pneumatic pressure and      fail as specified when either electric power to the event that either electric power to the        power to the actuating solenoids.                the valve actuating solenoid is lost or valve actuating solenoid is lost or pneumatic                                                        pneumatic pressure to the valve is lost: the pressure to the valve is lost: the differential                                                      differential pressure control valves fail pressure control valves fail closed, and the                                                        closed, and the flow control valves to the flow control valves to the cooling coils fail                                                        cooling coils fail open.
Design Control Document/Tier 1 open.
: 10. Design features provide emergency            10. Inspections of the as-built design features for 10. Design features provide mechanical cross mechanical cross connects (HECW-Division          mechanical cross connects and electrical            connects (HECW-Division A), along with A), along with power manual alignments from      power manual capability will be conducted.          electrical power manual alignments from Division I to MCR HVAC (Div III / C) fans and                                                        Division I to MCR HVAC (Div III / C) fans and components.                                                                                          components.
2.11-31
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.11.7 Oxygen Injection System No entry for this system.
2.11-32                                                            Oxygen Injection System
 
25A5675AA Revision 7 ABWR                                              Design Control Document/Tier 1 2.11.8 This section not used.
This section not used.                                                    2.11-33
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.11.9 Reactor Service Water System Design Description The Reactor Service Water (RSW) System removes heat from the Reactor Building Cooling Water (RCW) System and rejects this heat to the Ultimate Heat Sink (UHS). The portions of the RSW System that are in the Control Building are within the Certified Design. Those portions of the RSW System that are outside the Control Building are not in the Certified Design. Figure 2.11.9a shows the basic system configuration and scope within the Certified Design. Figure 2.11.9b shows the RSW System control interfaces.
The RSW System provides cooling water flow to either two or three of the RCW System heat exchangers in each division. On a loss-of-coolant accident and/or loss of preferred power (LOCA and/or LOPP) signal, any closed valves for standby heat exchangers are automatically opened and cooling flow is provided to all three heat exchangers in each division.
For each division of the RSW System, the heat exchanger inlet and outlet valves close upon receipt of a signal indicating Control Building flooding in that division.
The RSW System is classified as Seismic Category I and ASME Code Section III, Class 3 and consists of three separate safety-related divisions.
Each of the three RSW divisions is powered by its respective Class 1E division. In the RSW System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment. Each mechanical division of the RCW system (Divisions A, B, C) is physically separated from the other divisions.
The RSW System has the following main control room (MCR) displays and controls: control and status displays for the valves shown on Figure 2.11.9a. The RSW System components with status displays and control interfaces with the Remote Shutdown System (RSS) are identified in Figure 2.11.9a.
The motor-operated valves (MOVs) shown on Figure 2.11.9a all have active safety-related functions to open and close under differential pressure and fluid flow conditions.
Interface Requirements Part of the RSW System that are not within the Certified Design shall meet the following requirements:
(1)    Design features shall be provided to limit the maximum flood height to 5.0 meters in each RCW heat exchanger room.
(2)    The design shall have three divisions which are physically separated. For any structure(s) housing RSW System components, there shall be inter-divisional boundaries (including walls, floors, doors and penetrations) that have three-hour fire 2.11-34                                                                          Reactor Service Water System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 rating. In addition, there shall be inter-divisional flood control features which preclude flooding from occuring in more than one division. Each division shall be powered by its respective Class 1E division. Each division shall be capable of removing the design heat capacity (as specified in Section 2.11.3) of the RCW heat exchangers in its division.
(3)    Upon receipt of a loss-of-coolant (LOCA) signal, components in standby mode shall start and/or align to the operating mode.
(4)    RSW System Divisions A and B shall have control interfaces with the Remote Shutdown System (RSS) as required to support RSW operation during RSS design basis conditions.
(5)    If required by the elevation relationships between the UHS and the RSW System components in the Control Building (C/B), the RSW System shall have antisiphon capability to prevent a C/B flood after an RSW System break and after the RSW System pumps have been stopped.
(6)    RSW System pumps in any division shall be tripped on receipt of a signal indicating flooding in that division of the C/B basement area.
(7)    Any tunnel structures used to route RSW System piping to the Control Building shall be classified as Seismic Category I. Tunnel flooding due to site flood conditions shall be precluded.
(8)    The site specific design of RSW demonstrates that at least one division has adequate cooling capacity following postulated aircraft impact strike locations on the RSW. At least one division of RSW is physically separated from the other two divisions by 50 meters or greater horizontal distance. The RSW pump houses are separated by 50 meters or greater horizontal distance from the R/B and C/B. A strike on one of the RSW pump houses does not prevent the operation of safe shutdown equipment located in the R/B and C/B.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.9 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the portions of the RSW System within the Certified Design.
Reactor Service Water System                                                                                2.11-35
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 CONTROL BUILDING RCW RSW 3
M RCW HEAT                                          M EXCHANGER M
RCW M
HEAT EXCHANGER M
RCW HEAT                                          M EXCHANGER 3 SITE RSW SPECIFIC SCOPE NOTES:
: 1. THIS FIGURE SHOWS ONE OF THREE SIMILAR DIVISIONS. ALL ELECTRICAL POWER LOADS FOR THE COMPONENT IN DIVISIONS A, B, AND C ARE POWERED FROM DIVISIONS I, II, AND III, RESPECTIVELY.
: 2. VALVES SHOWN ABOVE IN DIVISIONS A AND B HAVE CONTROLS AND OPEN/CLOSE STATUS DISPLAY ON THE REMOTE SHUTDOWN PANEL.
Figure 2.11.9a Reactor Service Water System 2.11-36                                                                        Reactor Service Water System
 
ABWR Reactor Service Water System LOCAL AREA                            MAIN CONTROL ROOM                          LOCAL AREA Plant Sensors                                                                  Device Actuators RSW Manual Pump and Valve Controls SSLC PROCESSING RSW SYSTEM LOGIC Automatic:
25A5675AA Revision 7
                                                                                      - Sensor Channel Trip Decision
                                                                                      - System Coincidence Trip Decision  RSW  - LOCA Alignment C/B    Basement Water Level              - Control and Interlock Logic
                                                                                      - Division-of-Sensors Bypass              - Flood Control
                                                                                      - Calibration, Self-Diagnosis RCW LOCA Signal              RSW  Manual Pump and Valve Control Design Control Document/Tier 1 Notes:
: 1. Diagram represents one of three RSW divisions.
: 2. See Section 3.4, Figure 3.4b for SSLC Processing.
2.11-37                                                                  Figure 2.11.9b Reactor Service Water
 
ABWR 2.11-38 Table 2.11.9 Reactor Service Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the RSW System    1. Inspections of the as-built system will be    1. The as-built RSW System conforms with the is as shown on Figure 2.11.9.                    conducted.                                      basic configuration shown in Figure 2.11.9.
: 2. The ASME Code components of the RSW            2. A hydrostatic test will be conducted on those 2. The results of the hydrostatic test of the System retain their pressure boundary              Code components of the RSW System                ASME Code components of the RSW integrity under internal pressures that will be    required to be hydrostatically tested by the    System conform with the requirements in the experienced during service.                        ASME Code.                                      ASME Code, Section III.
: 3. On a LOCA and/or LOPP signal, any closed 3. Using simulated LOCA and/or LOPP signals, 3. Upon receipt of simulated LOCA and/or valves for standby heat exchangers are      tests will be performed on standby heat      LOPP signals, the standby heat exchanger automatically opened.                      exchanger inlet and outlet valves.          inlet and outlet valves open.
: 4. For each division of RSW, the heat              4. Using simulated signals, tests will be      4. The heat exchanger inlet and outlet valves 25A5675AA Revision 7 exchanger inlet and outlet valves close upon      conducted on the heat exchanger inlet and      close upon receipt of a signal indicating receipt of a signal indicating Control Building    outlet valves.                                Control Building flooding in that division.
flooding in that division.
: 5. Each of the three RSW divisions is powered 5.                                              5.
by its respective Class 1E division. In the
: a. Tests will be performed on the RSW          a. The test signal exists only in the Class RSW System, independence is provided System by providing a test signal in only      1E Division under test in the RSW between Class 1E divisions, and between one Class 1E division at a time.                System.
Class 1E divisions and non-Class 1E equipment.                                    b. Inspections of the as-installed Class 1E    b. Physical separation or electrical isolation divisions in the RSW System will be            exists between Class 1E divisions in the Design Control Document/Tier 1 performed.                                      RSW System. Physical separation or electrical isolation exists between Class 1E divisions and non-Class 1E equipment.
Reactor Service Water System
: 6. Each mechanical division of the RSW          6. Inspections of the as-built system will be    6. Each mechanical division of the RSW System (Divisions A, B, C) is physically        performed.                                      System is physically separated from other separated.                                                                                        mechanical divisions of the RSW System by structural and/or fire barriers.
: 7. MCR displays and controls provided for the    7. Inspections will be performed on the MCR      7. Displays and controls exist or can be RSW System are as defined in Section            displays and controls for the RSW System.        retrieved in the MCR as defined in Section 2.11.9.                                                                                          2.11.9.
 
Table 2.11.9 Reactor Service Water System (Continued)
ABWR Reactor Service Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 8. RSS displays and controls provided for the    8. Inspections will be performed on the RSS    8. Indications and controls exist on the RSS as RSW System are as defined in Section              displays and controls for the RSW System.      defined in Section 2.11.9.
2.11.9.
: 9. MOVs designated in Section 2.11.9 as          9. Tests of installed valves, for opening and    9. Upon receipt of the actuating signal, each having an active safety-related function open    closing will be conducted under                  MOV opens and closes, depending on the and close under differential pressure, fluid    preoperational differential pressure, fluid      valves safety function.
flow, and temperature conditions.                flow, and temperature conditions.
25A5675AA Revision 7 Design Control Document/Tier 1 2.11-39
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.11.10 Turbine Service Water System Design Description The Turbine Service Water (TSW) System removes heat from the Turbine Building Cooling Water (TCW) System and rejects this heat to the power cycle heat sink which is part of the Circulating Water System. The portions of the TSW System that are in the Turbine Building are within the Certified Design. Those portions of the TSW System that are outside the Turbine Building are not in the Certified Design. Figure 2.11.10 shows the basic system configuration and scope of the portion within the Certified Design.
The TSW System is classified as non-safety-related.
Interface Requirements The portions of the TSW System which are not part of the Certified Design shall meet the following requirement:
* None identified for this system.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.10 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken, for the portions of the TSW System within the Certified Design.
2.11-40                                                                          Turbine Service Water System
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 TURBINE BUILDING TCW  TSW NNS TCW HEAT EXCHANGER TCW HEAT EXCHANGER TCW HEAT EXCHANGER NNS SITE TSW SPECIFIC SCOPE Figure 2.11.10 Turbine Service Water System Turbine Service Water System                                                                2.11-41
 
ABWR 2.11-42 Table 2.11.10 Turbine Service Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the TSW System  1. Inspections of the as-built system will be  1. The as-built TSW System conforms with the is as shown on Figure 2.11.10.                conducted.                                      basic configuration shown on Figure 2.11.10.
25A5675AA Revision 7 Design Control Document/Tier 1 Turbine Service Water System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.11.11 Station Service Air System Design Description The Station Service Air (SA)System consists of two air compressing trains, an air receiver tank, two trains of filters, piping, valves, controls and instrumentation. Figure 2.11.11 shows basic SA System configuration and scope.
The SA System provides compressed air for general plant use. The SA System also provides backup to the Instrument Air (IA) System in the event that IA System pressure is lost.
Except for the containment penetration and isolation valves, the SA System is classified as non-safety-related.
The containment penetration and isolation valves are classified as Seismic Category I. Figure 2.11.11 shows the ASME Code class for the SA System components.
The check valve (CV) for containment isolation shown on Figure 2.11.11 has an active safety-related function to close under system pressure, fluid flow, and temperature conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.11 provides a definition of inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the SA System.
Station Service Air System                                                                              2.11-43
 
ABWR 2.11-44                                                              PRIMARY CONTAINMENT NNS  2                  2  NNS TO AIR HOSE                      TO REACTOR BUILDING, CONNECTIONS                        CONTROL BUILDING, AND RADWASTE BUILDING 25A5675AA Revision 7 IA SA NNS TURBINE BUILDING EQUIPMENT F
Design Control Document/Tier 1 COMPRESSOR F
Station Service Air System COMPRESSOR                        AIR RECEIVER TANK TURBINE BUILDING Figure 2.11.11 Station Service Air System
 
ABWR Station Service Air System Table 2.11.11 Station Service Air System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the SA System is    1. Inspections of the as-built system will be    1. The as-built SA System conforms with the as shown on Figure 2.11.11.                        conducted.                                      basic configuration shown on Figure 2.11.11.
: 2. The ASME Code components of the SA              2. A pressure test will be conducted on those    2. The results of the pressure test of the ASME System retain their pressure boundary              Code components of the SA System                Code components of the SA System conform integrity under internal pressures that will be    required to be pressure tested by the ASME      with the requirements in ASME Code Section experienced during service.                        Code.                                            III.
: 3. The CV designated in Section 2.11.11 as        3. Tests of the installed valve for closing will be 3. The CV closes.
having an active safety-related function          conducted under system preoperational closes, under system pressure, fluid flow,        pressure, fluid flow, and temperature and temperature conditions.                        conditions.
25A5675AA Revision 7 Design Control Document/Tier 1 2.11-45
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.11.12 Instrument Air System Design Description The Instrument Air (IA) System consists of two air compressing trains, an air receiver tank, two drying trains, piping, valves, controls and instrumentation. Figure 2.11.12 shows the basic IA System configuration and scope.
The IA System provides compressed air for pneumatic equipment, valves, controls and instrumentation outside the primary containment.
The IA System distribution piping penetrates the primary containment. During plant operation, this line is supplied with nitrogen by the High Pressure Nitrogen Gas Supply (HPIN) System.
In the event that HPIN System pressure is lost, the IA System provides air backup by remote manual alignment of IA System.
Except for the containment penetration and isolation valves, the IA System is classified as non-safety-related.
The IA containment penetration and isolation valves are classified as Seismic Category I.
Figure 2.11.12 shows the ASME Code class for the IA System piping and components.
The IA System containment isolation valve is powered from Class 1E Division I. In the IA System, independence is provided between the Class 1E division and non-Class 1E equipment.
The main control room has controls and open/close status indication for the containment isolation valve.
The safety-related electrical equipment that provides containment isolation and is located outside primary containment in the Reactor Building is qualified for a harsh environment.
The motor-operated valve (MOV) shown on Figure 2.11.12 has an active safety-related function to close and perform this function under differential pressure, fluid flow, and temperature conditions.
The check valve (CV) for containment isolation shown on Figure 2.11.12 has an active safety-related function to close under system pressure, fluid flow, and temperature conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.12 provides a definition of inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the IA System.
2.11-46                                                                                  Instrument Air System
 
ABWR Instrument Air System PRIMARY CONTAINMENT P
REACTOR BUILDING EQUIPMENT                      CONTROL          SERVICE      RADWASTE BUILDING        BUILDING      BUILDING M
NOTE 1              EQUIPMENT        EQUIPMENT    EQUIPMENT PRIMARY CONTAINMENT EQUIPMENT                                                  NNS NNS 2                                            IA HPIN 2  NNS REACTOR BUILDING 25A5675AA Revision 7 NNS SA IA TURBINE BUILDING                                        P EQUIPMENT F
AIR DRYER COMPRESSOR Design Control Document/Tier 1 F
AIR RECEIVER COMPRESSOR AIR DRYER TURBINE BUILDING NOTES:
: 1. CONTAINMENT ISOLATION VALVE IS POWERED FROM CLASS 1E DIVISION I.
2.11-47 Figure 2.11.12 Instrument Air System
 
ABWR 2.11-48 Table 2.11.12 Instrument Air System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                                Acceptance Criteria
: 1. The basic configuration of the IA System is      1. Inspections of the as-built IA System will be        1. The as-built IA System conforms with the shown on Figure 2.11.12.                            conducted.                                              basic configuration shown on Figure 2.11.12.
: 2. The ASME Code components of the IA              2. A pressure test will be conducted on those 2. The results of the pressure test of the ASME System retain their pressure boundary              Code components of the IA System required    Code components of the IA System conform integrity under internal pressures that will be    to be pressure tested by the ASME Code.      with the requirements in ASME Code Section experienced during service.                                                                      III.
: 3. The IA System containment isolation valve is 3.                                            3.
powered from Class 1E Division I. In the IA
: a. Tests will be performed on the IA System    a. The test signal exists in the IA System System, independence is provided between by providing a test signal in only one        only when the signal is applied to the the Class 1E division and non-Class 1E Class 1E division at a time.                  division associated with the IA System.
equipment.
25A5675AA Revision 7
: b. In the IA System, physical separation or
: b. Inspection of the as-installed Class 1E        electrical isolation exists between the division in the IA System will be              Class 1E division and non-Class 1E performed.                                    equipment.
: 4. Main control room displays and controls      4. Inspections will be performed on the main    4. Displays and controls exist or can be provided for the IA System are as defined in    control room displays and controls for the IA    retrieved in the main control room as defined Section 2.11.12.                                System.                                          in Section 2.11.12.
: 5.                                                  5.                                                      5.
: a. The MOV designated in Section 2.11.12            a. Tests of the installed valve for closing will        a. Upon receipt of the actuating signal the Design Control Document/Tier 1 as having an active safety-related                  be conducted under preoperational                      MOV closes.
function closes under differential                  differential pressure, fluid flow, and pressure, fluid flow, and temperature              temperature conditions.
conditions.
: b. The CV designated in Section 2.11.12 as          b. Tests of installed valve for closing will be        b. The CV closes.
Instrument Air System having an active safety-related function            conducted under system preoperational closes under system pressure, fluid flow,          pressure, fluid flow, and temperature and temperature conditions.                        conditions.
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.11.13 High Pressure Nitrogen Gas Supply System Design Description The High Pressure Nitrogen Gas Supply (HPIN) System provides nitrogen to pneumatic equipment inside the primary containment. Figure 2.11.13 shows the basic HPIN System configuration and scope.
The HPIN System consists of:
(1)  Two divisional systems (Divisions A and B) which are supplied from bottled nitrogen supplies. These systems can supply nitrogen to the automatic depressurization system (ADS) accumulators on the safety/relief valves (SRVs).
(2)  A non-divisional system that is supplied from the Atmospheric Control (AC) System.
This system can supply nitrogen to the non-ADS and ADS accumulators on the SRVs.
The two divisional systems and the containment penetrations and isolation valves on the non-divisional system are classified as safety-related.
During operation, all SRV accumulators are supplied from the non-divisional system. If the pressure sensor in either of the safety-related systems indicates low pressure, the valve between that system and the non-divisional system closes and the supply valve to the bottled nitrogen supply in that division opens. If the pressure sensor in the non-divisional system indicates a low pressure, the valves between the non-divisional and the divisional systems close.
The capacity of the bottled nitrogen supply in each HPIN division maintains the ADS valves in that division in an open condition for a period of at least seven days following a design basis accident.
The two divisional systems and the containment penetration and isolation valves in the non-divisional system are classified as Seismic Category I. Figure 2.11.13 shows the ASME Code class for the HPIN System piping and components.
Except for the isolation valves and distribution piping inside the primary containment, the HPIN System is located in the Reactor Building.
Each of the two HPIN divisions is powered from the respective Class 1E division as shown on Figure 2.11.13. In the HPIN System, independence is provided between the Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Outside the primary containment and except for the interconnection through the non-divisional system, each mechanical division (Divisions A and B) is physically separated from the other division.
High Pressure Nitrogen Gas Supply System                                                                  2.11-49
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 The HPIN System has the following displays and controls in the main control room:
(1)    Parameter displays for the sensors shown on Figure 2.11.13.
(2)    Control and status indication for the active safety-related components shown on Figure 2.11.13.
The HPIN System has Pressure indication for Division I and Division II on the Remote Shutdown Panel.
The safety-related electrical equipment shown on Figure 2.11.13 located in the Reactor Building is qualified for a harsh environment.
The motor-operated valves (MOVs) shown on Figure 2.11.13 have active safety-related functions to open, close, or both open and close, and perform these functions under differential pressure, fluid flow, and temperature conditions.
The check valves (CVs) shown on Figure 2.11.13, have active safety-related functions to both open and close under system pressure, fluid flow, and temperature conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.13 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the HPIN System.
2.11-50                                                              High Pressure Nitrogen Gas Supply System
 
ABWR High Pressure Nitrogen Gas Supply System INSIDE PRIMARY CONTAINMENT M                                  P        M FROM BOTTLED TO NITROGEN SUPPLY ADS 3    2        2 3              SRVs DIVISION B 3    M NNS IA NNS HPIN        P            M 25A5675AA Revision 7 TO NON-ADS NNS                                                          SRVs AC  HPIN NNS      2    2  3 NNS 3  M M                                  P        M                            TO ADS Design Control Document/Tier 1 FROM BOTTLED SRVs NITROGEN SUPPLY 3
3    2        2  3 HPIN NBS DIVISION A NOTES:
: 1. HPIN SYSTEM DIVISION-A IS POWERED FROM CLASS 1E DIVISION I, AND HPIN SYSTEM DIVISION-B IS POWERED FROM CLASS 1E DIVISION II.
THE CONTAINMENT ISOLATION VALVE OF NON-DIVISIONAL PORTION OF THE HPIN SYSTEM IS POWERED FROM CLASS 1E DIVISION I.
2.11-51                                                              Figure 2.11.13 High Pressure Nitrogen Gas Supply System
 
ABWR 2.11-52 Table 2.11.13 High Pressure Nitrogen Gas Supply System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the HPIN System    1. Inspections of the as-built system will be    1. The as-built HPIN System conforms with the is as shown on Figure 2.11.13.                    conducted.                                        basic configuration shown on Figure 2.11.13.
: 2. The ASME Code components of the HPIN            2. A pressure test will be conducted on those    2. The results of the pressure test of the ASME System retain their pressure boundary              Code components of the HPIN System              Code components of the HPIN System integrity under internal pressures that will be    required to be pressure tested by the ASME      conform with the requirements in ASME experienced during service.                        Code.                                            Code Section III.
: 3. If the pressure sensor in either of the safety- 3. Tests will be conducted on each division of  3. If the pressure sensor in either of the safety-related systems indicates low pressure, the        the as-built HPIN System using simulated        related systems indicates low pressure, the valve between that system and the non-            pressure signals.                                valve between that system and the non-divisional system closes and the supply                                                            divisional system closes and the supply valve to the bottled nitrogen supply in that                                                        valve to the bottled nitrogen supply in that 25A5675AA Revision 7 division opens.                                                                                    division opens.
: 4. If the pressure sensor in the non-divisional  4. Tests will be conducted on the as-built HPIN 4. If the pressure sensor in the non-divisional system indicates a low pressure, the valves      System using simulated pressure signals.        system indicates a low pressure, the valves between the non-divisional and the divisional                                                    between the non-divisional and the divisional systems close.                                                                                  systems close.
: 5. The capacity of the bottled nitrogen in each 5. Analyses of the installed HPIN will be          5. The capacity of the bottled nitrogen in each HPIN division maintains the ADS valves in      performed. The analyses will consider              HPIN division maintains the ADS valves in that division in an open condition for a period nitrogen leakage from the ADS actuators            that division in an open condition for a period of at least seven days following a design      when maintaining the ADS valves open.              of at least seven days following a design High Pressure Nitrogen Gas Supply System                                                                                                                                                            Design Control Document/Tier 1 basis accident.                                Leakage from HPIN components when the              basis accident.
system is in this mode will also be considered. The analyses will compare the total storage capacity in each division with the total leakage that occurs in a seven day period.
 
Table 2.11.13 High Pressure Nitrogen Gas Supply System (Continued)
ABWR High Pressure Nitrogen Gas Supply System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 6. Each of the two HPIN divisions is powered      6.                                                  6.
from the respective Class 1E division as
: a. Tests will be performed in the HPIN              a. The test signal exists only in the Class shown on Figure 2.11.13. In the HPIN System by providing a test signal in only            1E division under test in the HPIN System, independence is provided between one Class 1E division at a time.                    System.
Class 1E divisions, and between Class 1E divisions and non-Class 1E equipment.                b. Inspections of the as-installed Class 1E          b. In the HPIN System, physical separation divisions in the HPIN System will be                or electrical isolation exists between performed.                                          Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 7. Outside the primary containment and except 7. Inspections of the as-built HPIN System will          7. Outside the primary containment and except 25A5675AA Revision 7 for the interconnection through the non-      be conducted.                                            for the interconnection through the non-divisional system, each mechanical division                                                            divisional system, each mechanical division (Divisions A and B) of the HPIN System is                                                              (Divisions A and B) of the HPIN System is physically separated from the other division.                                                          physically separated from the other division by structural and/or fire barriers.
: 8. Main control room displays and controls    8. Inspections will be performed on the main            8. Displays and controls exist or can be provided for the HPIN System are as defined    control room displays and controls for the              retrieved in the main control room as defined in Section 2.11.13.                            HPIN System.                                            in Section 2.11.13.
: 9.                                                9.                                                  9.
Design Control Document/Tier 1
: a. MOVs designated in Section 2.11.13 as          a. Tests of installed valves for opening,            a. Upon receipt of the actuating signal, having an active safety-related function          closing, or both opening and closing will            each MOV opens, closes, or both opens open, close, or both open and close                be conducted under preoperational                    and closes, depending upon the valves under differential pressure, fluid flow,          differential pressure, fluid flow, and              safety functions.
and temperature conditions.                        temperature conditions.
: b. CVs designated in Section 2.11.13 as            b. Tests of installed valves for both opening        b. Based on the direction of the differential having an active safety-related function          and closing, will be conducted under                pressure across the valve, each CV both both open and close, under system                  system pre-operational pressure, fluid              opens and closes.
pressure, fluid flow, and temperature              flow, and temperature conditions.
2.11-53 conditions.
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 2.11.14 Heating Steam and Condensate Water Return System No entry for this system.
2.11-54                                                Heating Steam and Condensate Water Return System
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.11.15 House Boiler No entry for this system.
House Boiler                                                                        2.11-55
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.11.16 Hot Water Heating System No entry for this system.
2.11-56                                                            Hot Water Heating System
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.11.17 Hydrogen Water Chemistry System No entry for this system.
Hydrogen Water Chemistry System                                                    2.11-57
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.11.18 Zinc Injection System No entry for this system.
2.11-58                                                                Zinc Injection System
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.11.19 Breathing Air System No entry for this system.
Breathing Air System                                                                  2.11-59
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.11.20 Sampling System Design Description The Sampling (SAM) System obtains samples from systems throughout the plant. A part of the SAM System is a post-accident sampling system (PASS). The PASS takes post-accident gas samples from the primary containment and reactor coolant samples for analysis. The PASS collects samples during and after an accident and is shielded and remotely operated.
The PASS collects reactor coolant samples for measurement of boron and radionuclides (noble gases, iodines, cesiums and non-volatile isotopes).
The SAM System and PASS are classified as non-safety-related.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.20 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the SAM System.
2.11-60                                                                                    Sampling System
 
ABWR Sampling System Table 2.11.20 Sampling System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the SAM System  1. Inspections of the as-built system will be  1. The as-built SAM System conforms with the is as described in Section 2.11.20.            conducted.                                      basic configuration described in Section 2.11.20.
: 2. The PASS collects samples of containment    2. A test of the as-built PASS will be conducted 2. Containment gas and reactor coolant gases and reactor coolant.                    to obtain samples.                              samples are collected by the PASS.
25A5675AA Revision 7 Design Control Document/Tier 1 2.11-61
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.11.21 Freeze Protection System No entry for this system.
2.11-62                                                            Freeze Protection System
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.11.22 Iron Injection System No entry for this system.
Iron Injection System                                                                  2.11-63
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.11.23 Potable and Sanitary Water System Design Description The Potable and Sanitary Water (PSW) System provides water to the Reactor Building, Control Building, Turbine Building, Radwaste Building and Service Building and collects liquid sanitary wastes and entrained solids and conveys them to a sewage facility and then to a site discharge structure. Nonradioactive drain subsystems throughout the plant collect nonradioactive waste water and convey it to the site discharge structure. Water is supplied to the PSW System by the Makeup Water Preparation System.
Those parts of the PSW System that are within the Reactor Building, Control Building, Turbine Building, Radwaste Building and Service Building are within the Certified Design. Those parts of the PSW System that are outside these buildings are not within the scope of the Certified Design.
The PSW System is classified as non-safety-related.
The PSW System has no interconnections with radioactive systems having the potential for transferring radioactive materials into the PSW System.
Interface Requirements The portions of the PSW System which are not part of the Certified Design shall meet the following requirement:
* None for this system.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.23 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the portions of the PSW System within the Certified Design.
2.11-64                                                                      Potable and Sanitary Water System
 
ABWR Potable and Sanitary Water System Table 2.11.23 Potable and Sanitary Water System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration for the PSW System 1. Inspections of the as-built system will be        1. The as-built PSW System conforms with the is as described in Section 2.11.23.          conducted.                                            basic configuration described in Section 2.11.23.
: 2. The PSW System has no interconnections          2. Tests will be conducted on the as-built      2. No water inleakage from the radioactive with radioactive systems having the potential      nonradioactive drain system by pressurizing      drains in to the PSW System is observed.
for transferring radioactive materials into the    radioactive floor drains with water and PSW System.                                        observing the nonradioactive drains for evidence of inleakage from the radioactive floor drains.
25A5675AA Revision 7 Design Control Document/Tier 1 2.11-65
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.12.1 Electrical Power Distribution System Design Description The AC Electrical Power Distribution (EPD) System consists of the transmission network (TN), the plant switching stations, the Main Power Transformer (MPT), the Unit Auxiliary Transformers (UAT), the Reserve Auxiliary Transformer(s) (RAT(s)), the plant main generator (PMG) output circuit breaker, the medium voltage metal-clad (M/C) switchgear, the low voltage power center (P/C) switchgear, and the motor control centers (MCCs). The distribution system also includes the power, instrumentation and control cables and bus ducts to the distribution system loads, and the protection equipment provided to protect the distribution system equipment. The EPD System within the scope of the Certified Design starts at the low voltage terminals of the MPT and the low voltage terminals of the RAT(s) and ends at the distribution system loads. Interface requirements for the TN, plant switching stations, MPT, and RAT(s) are specified below.
The plant EPD System can be supplied power from multiple power sources; these are independent transmission lines from the TN, the PMG, and the combustion turbine generator (CTG). In addition, the EPD System can be supplied from three onsite Class 1E Standby Power Sources (Emergency Diesel Generators (DGs)). The Class 1E portion of the EPD System is shown in Figure 2.12.1.
During plant power operation, the PMG supplies power through the PMG output circuit breaker through the MPT to the TN, and to the UATs. When the PMG output circuit breaker is open, power is backfed from the TN through the MPT to the UATs.
The UATs can supply power to the non-Class 1E load groups of medium voltage M/C power generation (PG) and plant investment protection (PIP) switchgear, and to the three Class 1E divisions (Division I, II, and III) of medium voltage M/C switchgear.
The RAT(s) can supply power to the non-Class 1E load groups of medium voltage M/C PG and PIP switchgear, and to the three Class 1E divisions (Division I, II, and III) of medium voltage M/C switchgear.
Non-Class 1E load groups of medium voltage M/C switchgear are supplied power from a UAT with an alternate power supply from a RAT. In addition, the non-Class 1E medium voltage M/C switchgear can be supplied power from the CTG.
Class 1E medium voltage M/C switchgear are supplied power directly (not through any bus supplying non-Class 1E loads) from at least a UAT or a RAT. Class 1E medium voltage M/C switchgear can also be supplied power from their own dedicated Class 1E DG or from the non-Class 1E CTG.
The UATs are sized to supply their load requirements, during design operating modes, of their respective Class 1E divisions and non-Class 1E load groups. UATs are separated from the Electrical Power Distribution System                                                                    2.12-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 RAT(s). In addition, UATs are provided with their own oil pit, drain, fire deluge system, grounding, and lightning protection system.
The PMG, its output circuit breaker, and UAT power feeders are separated from the RAT(s) power feeders. The PMG, its output circuit breaker, and UAT instrumentation and control circuits, are separated from the RAT(s) instrumentation and control circuits.
The MPT and its switching station instrumentation and control circuits, from the switchyard(s) to the main control room (MCR), are separated from the RAT(s) and its switching station instrumentation and control circuits.
The medium voltage M/C switchgear and low voltage P/C switchgear, with their respective transformers, and the low voltage MCCs are sized to supply their load requirements. M/C and P/C switchgear, with their respective transformers, and MCCs are rated to withstand fault currents for the time required to clear the fault from the power source. The PMG output circuit breaker, and power feeder and load circuit breakers for the M/C and P/C switchgear, and MCCs are sized to supply their load requirements and are rated to interrupt fault currents.
Class 1E equipment is protected from degraded voltage conditions.
EPD System interrupting devices (circuit breakers and fuses) are coordinated so that the circuit interrupter closest to the fault opens before other devices.
Instrumentation and control power for the Class 1E divisional medium voltage M/C switchgear and low voltage P/C switchgear is supplied from the Class 1E DC power system in the same division.
The PMG output circuit breaker is equipped with redundant trip devices which are supplied from separate, non-Class 1E DC power systems.
EPD System cables and bus ducts are sized to supply their load requirements and are rated to withstand fault currents for the time required to clear the fault from its power source.
For the EPD System, Class 1E power is supplied by three independent Class 1E divisions.
Independence is maintained between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
The only non-Class 1E loads connected to the Class 1E EPD System are the Fine Motion Control Rod Drives (FMCRDs) and the associated AC standby lighting system.
External (to Reactor Building) connections are provided to all three 1E Reactor Building 480 VAC Power Centers for portable External diesel generators. The connectors are isolated from the Power Centers by normally open 1E breakers.
There are no automatic connections between Class 1E divisions.
2.12-2                                                                      Electrical Power Distribution System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 Class 1E medium voltage M/C switchgear and low voltage P/C switchgear and MCCs are identified according to their Class 1E division. Class 1E M/C and P/C switchgear and MCCs are located in Seismic Category I structures, and in their respective divisional areas.
Class 1E EPD System cables and raceways are identified according to their Class 1E division.
Class 1E divisional cables are routed in Seismic Category I structures and in their respective divisional raceways.
Harmonic Distortion waveforms do not prevent Class 1E equipment from performing their safety functions.
The EPD System supplies an operating voltage at the terminals of the Class 1E utilization equipment that is within the utilization equipment's voltage tolerance limits.
An electrical grounding system is provided for (1) instrumentation, control, and computer systems, (2) electrical equipment (switchgear, distribution panels, transformers, and motors) and (3) mechanical equipment (fuel and chemical tanks). Lightning protection systems are provided for buildings and for structures and transformers located outside of the buildings. Each grounding system and lightning protection system is separately grounded to the plant grounding grid.
The EPD System has the following alarms, displays and controls in the MCR:
(1)      Alarms for degraded voltage on Class 1E medium voltage M/C switchgear.
(2)      Parameter displays for PMG output voltage, amperes, watts, vars, and frequency.
(3)      Parameter displays for EPD System medium voltage M/C switchgear bus voltages and feeder and load amperes.
(4)      Controls for the PMG output circuit breaker, medium voltage M/C switchgear feeder circuit breakers, load circuit breakers from the medium voltage M/C switchgear to their respective low voltage P/C switchgear, and low voltage feeder circuit breakers to the low voltage P/C switchgear.
(5)      Status indication for the PMG output circuit breaker and the medium voltage M/C switchgear circuit breakers.
The EDP System has the following displays and controls at the Remote Shutdown System (RSS):
(1)      Parameter displays for the bus voltages on the Class 1E Divisions I and II medium voltage M/C switchgear.
Electrical Power Distribution System                                                                        2.12-3
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 (2)    Controls and status indication for the UAT, RAT(s), CTG and DG Class 1E feeder circuit breakers to the Division I and II medium voltage M/C switchgear, the load circuit breakers from the Class 1E Division I and II medium voltage M/C switchgear to their respective low voltage P/C switchgear, and the low voltage feeder circuit breakers to the Class 1E Division I and II low voltage P/C switchgear.
Class 1E equipment is classified as Seismic Category I.
Class 1E equipment which is located in areas designated as harsh environment areas is qualified for harsh environments.
Monitoring of the normal and alternate power feeds on the high voltage side of the UAT and RAT using micro-processor based protective relays to detect open phase conditions, whether one, two, or three phases, with or without accompanying ground fault.
All three phases of the nonsafety-related MPTs, UATs or RAT, on both the primary (high) and secondary (low) sides shall be monitored for under voltage, open phase, and ground faults by nonsafety-related micro-processor based protective relays. When an under voltage, open phase or ground fault is detected in any combination of one, two, or three phases by the designated MPT, UAT, or RAT protective relay, the protective relay shall send an alarm via the nonsafety-related alarm system to the Main Control Room.
Electric power to Class 1E safety-related busses is provided through a two feeder circuit breakers in series, one nonsafety-related and the other safety-related. This design ensures that the nonsafety-related protective relays shall, as appropriate, trip or fast transfer the plant nonsafety-related medium voltage busses and open the nonsafety-related circuit breaker feeds to the safety-related busses.
The Class 1E safety-related micro-processor based bus protective relay controlling the safety-related circuit breaker feeding nonsafety-related UAT Normal Preferred Power (NPP) power to the safety-related bus will automatically separate the safety-related bus from the nonsafety-related bus fed by the UAT NPP power source with detection of OPC or ground faults. The safety-related bus protective relay will then fast transfer the safety-related bus to the nonsafety-related RAT Alternative Preferred Power (APP) power source. The safety-related bus protective relay controlling the safety-related circuit breaker feeding nonsafety-related RAT APP power to the safety-related bus will automatically separate the safety-related bus from the nonsafety-related bus fed by the RAT power source with detection of OPC or ground faults. The safety-related bus protective relay will then fast transfer the safety-related bus to the UAT NPP power source.
If, as a result of either the nonsafety-related or safety-related micro-processor based protective relaying action, both UAT NPP and RAT APP power feeds have been separated from the safety-related bus, then the safety-related bus protective relays and safety-related sequencing logic will: 1) open the safety-related circuit breakers feeding the safety-related bus, 2) shed the 2.12-4                                                                      Electrical Power Distribution System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 safety-related loads on the safety-related bus, 3) start the safety-related emergency diesel generators, and 4) sequence on the safety-related loads.
There are three independent safety-related medium voltage busses which provide electric power to the safety-related equipment divisions. The safety-related protective relays and safety-related sequencing logic on each of the three safety-related busses are independent of those on the other safety-related busses. Therefore, the design conforms to the IEEE Std 603 single failure criterion.
Interface Requirements The portions of the EPD System which are not part of the Certified Design shall meet the following requirements:
The offsite system shall consist of a minimum of two independant offsite transmission circuits from the TN.
Voltage variations of the offsite TN during steady state operation shall not cause voltage variations at the loads of more than plus or minus 10% of the loads nominal ratings.
The normal steady state frequency of the offsite TN shall be within plus or minus 2 hertz of 60 hertz during recoverable periods of system instability.
The offsite transmission circuits from the TN through and including the main step-up power transformers and RAT(s) shall be sized to supply their load requirements, during all design operating modes, of their respective Class 1E divisions and non-Class 1E load groups.
The impedances of the main step-up power transformers and RAT(s) shall be compatible with the interrupting capability of the plants circuit interrupting devices.
The independence of offsite transmission power, instrumentation, and control circuits shall be compatible with the portion of the offsite transmission power, instrumentation, and control circuits within GEs design scope.
Instrumentation and control system loads shall be compatible with the capacity and capability design requirements of DC systems within GEs design scope.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.1 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the EPD System.
Electrical Power Distribution System                                                                        2.12-5
 
25A5675AA Revision 7 ABWR                            Design Control Document/Tier 1 Figure 2.12.1 Class 1E Electrical Power Distribution System 2.12-6                              Electrical Power Distribution System
 
ABWR Electrical Power Distribution System Table 2.12.1 Electric Power Distribution System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration for the EPD System    1. Inspection of the as-built system will be    1. The as-built EPD System conforms with the is described in Section 2.12.1.                  conducted.                                      basic configuration described in Section 2.12.1.
: 2. UATs are sized to supply their load          2. Analyses for the as-built UATs to determine    2. Analyses for as-built UATs exist and requirements, during design operating            their load requirements will be performed.        conclude that UAT capacity, as determined modes, of their respective Class 1E divisions                                                      by its nameplate rating, exceeds its analyzed and non-Class 1E load groups.                                                                      load requirements, during design operating modes, for its Class 1E division and non-Class 1E load group.
: 3. UATs are separated from the RAT(s).            3. Inspections of the as-built UATs will be      3. As-built UATs are separated from the RAT(s) conducted.                                      by a minimum of 15.24m.
25A5675AA Revision 7
: 4. UATs are provided with their own oil pit,      4. Inspections of the as-built UATs will be      4. As-built UATs are provided with their own oil drain, fire deluge system, grounding, and        conducted.                                      pit, drain, fire deluge system, grounding, and lightning protection systems.                                                                      lightning protection systems.
: 5. The PMG and its output circuit breaker is    5. Inspections for the as-built PMG, the PMG      5. As-built PMG and its output circuit breaker is separated from the RAT(s) power feeders.        output circuit breaker, the RAT(s) and their      separated from the RAT(s) power feeders by The PMG and its output circuit breaker          respective instrumentation and control            a minimum of 15.24m, or by walls or floors.
instrument and control circuits are separated    circuits will be conducted.                      The PMG and its output circuit breaker from the RAT(s) instrumentation and control                                                        instrument and control circuits are separated circuits.                                                                                          from the RAT(s) instrumentation and control Design Control Document/Tier 1 circuits by a minimum of 15.24m, or by walls or floors outside the MCR, and are separated by routing the circuits in separate raceways inside the MCR.
2.12-7
 
Table 2.12.1 Electric Power Distribution System (Continued)
ABWR 2.12-8 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 6. UATs power feeders, and instrumentation        6. Inspections for the as-built UATs and RAT(s) 6. As-built UAT power feeders are separated and control circuits are separated from the        power feeders, and instrumentation and          from the RAT(s) power feeders by a RAT(s) output power feeders, and                  control circuits will be conducted.            minimum of 15.24m, or by walls or floors, instrumentation and control circuits.                                                              except at the switchgear, where they are routed to opposite ends of the medium voltage M/C switchgear. As-built UAT instrumentation and control circuits, are separated from the RAT(s) instrumentation and control circuits by a minimum of 15.24m, or by walls or floors, except as follows: a) at the non-Class 1E DC power sources, where they are routed in separate raceways, b) 25A5675AA Revision 7 inside the MCR, where they are separated by routing the circuits in separate raceways, and c) at the switchgear, where they are routed to opposite ends of the medium voltage M/C switchgear and routed in separate raceways inside the switchgear.
: 7. The MPT and its switching station              7. Inspections for the as-built MPT and RAT(s)    7. As-built MPT and its switching station instrumentation and control circuits are          and their respective switching station            instrumentation and control circuits, from the separated from the RAT(s) and its switching        instrumentation and control circuits will be      switchyard(s) to the MCR, are separated station instrumentation and control circuits.      conducted.                                        from the RAT(s) and its switching station Design Control Document/Tier 1 instrumentation and control circuits by a minimum of 15.24m, or by walls or floors.
Electrical Power Distribution System MPT and its switching station instrumentation and control circuits, inside the MCR, are separated from the RAT(s) and its switching station instrumentation and control circuits by routing the circuits in separate raceways.
 
Table 2.12.1 Electric Power Distribution System (Continued)
ABWR Electrical Power Distribution System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 8. Medium voltage M/C switchgear, low voltage 8. Analyses for the as-built EPD System to              8. Analyses for the as-built EPD System exist P/C switchgear, with their respective          determine load requirements will be                  and conclude that the capacities of the Class transformers, and MCCs, and their              performed.                                            1E switchgear, P/C transformers, MCCs, and respective switchgear and MCC feeder and                                                              their respective feeder and load circuit load circuit breakers are sized to supply their                                                      breakers, as determined by their nameplate load requirements.                                                                                    ratings, exceed their analyzed load requirements.
: 9.                                                  9.                                                9.
: a. Medium voltage M/C switchgear, low              a. Analyses for the as-built EPD System to        a. Analyses for the as-built EPD System voltage P/C switchgear, with their                  determine fault currents will be                  exist and conclude that the Class 1E respective transformers, and MCCs, are              performed.                                        switchgear, with their respective 25A5675AA Revision 7 rated to withstand fault currents for the                                                            transformers, and MCC, current
: b. Analyses for the as-built EPD System to time required to clear the fault from its                                                            capacities exceed their analyzed fault determine fault currents will be power source.                                                                                        currents for the time required, as performed.
determined by the circuit interrupting
: b. The PMG output circuit breaker, medium device coordination analyses, to clear voltage M/C switchgear, low voltage P/C the fault from its power source.
switchgear and MCC feeder and load circuit breakers are rated to interrupt                                                            b. Analyses for the as-built EPD System fault currents                                                                                        exist and conclude that the analyzed fault currents do not exceed the PMG output circuit breaker, and M/C, P/C Design Control Document/Tier 1 switchgear, and MCC feeder and load circuit breakers interrupt capacities, as determined by their nameplating ratings.
2.12-9
 
Table 2.12.1 Electric Power Distribution System (Continued)
ABWR 2.12-10 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 10. Class 1E equipment is protected from          10.                                                10.
degraded voltage conditions.                        a. Analyses for the as-built EPD System to        a. Analyses for the as-built EPD System determine the trip conditions for                  exist and conclude that the Class 1E degraded voltage conditions will be                preferred offsite power feeder breakers performed.                                        to the Class 1E M/C switchgear will trip before Class 1E loads experience
: b. Tests for each as-built Class 1E M/C degraded voltage conditions exceeding switchgear will be conducted by those voltage conditions for which the providing a simulated degraded voltage Class 1E equipment is qualified.
signal.
: b. As-built Class 1E feeder breakers from preferred offsite power to the Class 1E 25A5675AA Revision 7 M/C switchgear trip when a degraded voltage condition exists.
: 11. EPD System interrupting devices (circuit      11. Analyses for the as-built EPD System to        11. Analyses for the as-built EPD System exist breakers and fuses) are coordinated so that        determine circuit interrupting device              and conclude that the analyzed circuit the circuit interrupter closest to the fault      coordination will be performed.                    interrupter closest to the fault will open opens before other devices.                                                                          before other devices.
: 12. Instrumentation and control power for Class 12. Tests of the as-built Class 1E medium and        12. A test signal exists in only the Class 1E 1E divisional medium voltage M/C                low voltage switchgear will be conducted by          division under test.
switchgear and low voltage P/C switchgear is    providing a test signal in only one Class 1E Design Control Document/Tier 1 supplied from the Class 1E DC power            division at a time.
system in the same division.
Electrical Power Distribution System
: 13. The PMG output circuit breaker is equipped    13. Tests of the as-built PMG output circuit      13. A test signal exists in only the circuit under with redundant trip devices which are              breaker will be conducted by providing a test    test.
supplied from separate non-Class 1E DC            signal in only one trip circuit at a time.
power systems.
: 14. EPD System cables and bus ducts are sized 14. Analyses for the as-built EPD System cables 14. Analyses for the as-built EPD System exist to supply their load requirements.            and bus ducts will be performed.                and conclude that cable and bus duct capacities, as determined by cable and bus duct ratings, exceed their analyzed load requirements.
 
Table 2.12.1 Electric Power Distribution System (Continued)
ABWR Electrical Power Distribution System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 15. EPD System cables and bus ducts are rated 15. Analyses for the as-built EPD System to              15. Analyses for the as-built EPD System exist to withstand fault currents for the time      determine fault currents will be performed.              and conclude that cables and bus ducts will required to clear its fault from its power                                                              withstand the analyzed fault currents for the source.                                                                                                time required, as determined by the circuit interrupting device coordination analyses, to clear the analyzed faults from their power sources.
: 16. For the EPD System, Class 1E power is          16.                                                16.
supplied by three independent Class 1E
: a. Tests on the as-built EPD System will be        a. A test signal exists in only the Class 1E divisions. Independence is maintained conducted by providing a test signal in            division under test in the EPD System.
between Class 1E divisions, and between only one Class 1E division at a time.
Class 1E divisions and non-Class 1E                                                                      b. In the EPD System, physical separation 25A5675AA Revision 7 equipment.                                            b. Inspections of the as-built EPD System              or electrical isolation exists between Class 1E divisions will be conducted.              Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 17. Class 1E medium voltage M/C switchgear          17. Inspections of the as-built EPD System          17. As-built Class 1E M/C and P/C switchgear, and low voltage P/C switchgear and MCCs            Class 1E M/C and P/C switchgear and                and MCCs are identified according to their are identified according to their Class 1E          MCCs will be conducted.                            Class 1E division.
division.
Design Control Document/Tier 1
: 18. Class 1E M/C and P/C switchgear and            18. Inspections of the as-built Class 1E M/C and 18. As-built Class 1E M/C and P/C switchgear, MCCs are located in Seismic Category I              P/C switchgear and MCCs will be conducted.      and MCCs are located in Seismic Category I structures and in their respective divisional                                                        structures and in their respective divisional areas.                                                                                              areas.
: 19. Class 1E EPD System cables and raceways 19. Inspections of the as-built Class 1E EPD                19. As-built Class 1E EPD System cables and are identified according to their Class 1E  System cables and raceways will be                          raceways are identified according to their division.                                  conducted.                                                  Class 1E division.
: 20. Class 1E divisional cables are routed in        20. Inspection of the as-built Class 1E EPD        20. As-built Class 1E divisional cables are routed Seismic Category I structures and in their          System divisional cables and raceways will          in Seismic Category I structures and in their 2.12-11 respective divisional raceways.                    be conducted.                                      respective divisional raceways.
 
Table 2.12.1 Electric Power Distribution System (Continued)
ABWR 2.12-12 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 21. Harmonic Distortion waveforms do not      21. Analyses for the as-built EPD System to            21. Analyses for the as-built EPD System exist prevent Class 1E equipment from performing    determine harmonic distortions will be                and conclude that harmonic distortion their safety functions.                        performed.                                            waveforms do not exceed 5% voltage distortion on the Class 1E EPD System.
: 22. The EPD System supplies an operating          22.                                                22.
voltage at the terminals of the Class 1E                                                                a. Analyses for the as-built EPD System
: a. Analyses for the as-built EPD System to utilization equipment that is within the                                                                  exist and conclude that the analyzed determine voltage drops will be utilization equipment's voltage tolerance                                                                  operating voltage supplied at the performed.
limits.                                                                                                    terminals of the Class 1E utilization
: b. Tests of the as-built Class 1E EPD Sys-equipment is within the utilization tem will be conducted by operating con-equipment's voltage tolerance limits, as nected Class 1E loads at their analyzed 25A5675AA Revision 7 determined by their nameplate ratings.
minimum voltage.
: b. Connected Class 1E loads operate at their analyzed minmum voltage, as determined by the voltage drop analyses.
: 23. An electrical grounding system is provided    23. Inspections of the as-built EPD System plant 23. The as-built EDP System instrumentation, for (1) instrumentation, control, and              Grounding and Lightning Protection Systems      control, and computer grounding system, computer systems, (2) electrical equipment        will be conducted.                              electrical equipment and mechanical (switchgear, distribution panels,                                                                  equipment grounding system, and lightning Design Control Document/Tier 1 transformers, and motors) and (3)                                                                  protection systems provided for buildings mechanical equipment (fuel and chemical                                                            and for structures and transformers located Electrical Power Distribution System tanks). Lightning protection systems are                                                            outside of the buildings are separately provided for buildings and for structures and                                                      grounded to the plant grounding grid.
transformers located outside of the buildings.
Each grounding system and lightning protection system is separately grounded to the plant grounding grid.
: 24. MCR alarms, displays and controls provided 24. Inspections will be conducted on the MCR          24. Displays and controls exist or can be for the EPD System are as defined in Section  alarms, displays and controls for the EPD              retrieved in the MCR as defined in Section 2.12.1.                                        System.                                                2.12.1.
 
Table 2.12.1 Electric Power Distribution System (Continued)
ABWR Electrical Power Distribution System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                            Acceptance Criteria
: 25. RSS displays and controls provided for the      25. Inspections will be conducted on the as-built 25. Displays and controls exist or can be EPD System are as defined in Section                RSS displays and controls for the EPD            retrieved on the RSS as defined in Section 2.12.1.                                              System.                                          2.12.1.
: 26. DELETED                                          26. DELETED                                          26. DELETED
: 27. DELETED                                          27. DELETED                                          27. DELETED
: 28. Each MPT, UAT and RAT nonsafety-related          28. A test will be performed on each as-built    28. Using simulated signals in any combination micro-processor based protective relay(s),          MPT, UAT and RAT nonsafety-related                of the three phases, at the designated upon detecting:                                      protective relay(s), using simulated fault or    nonsafety-related protective relay setpoints phase loss signals in any combination of the      in any combination of phase fault or loss, the open phase conditions (OPC) three phases, to demonstrate that, at the        as-built MPT, UAT and RAT protective phase to phase faults                            designated protective relay setpoints:            relay(s) will initiate:
25A5675AA Revision 7 ground faults                                    a. alarms will be sent to the Main Control          a. alarms in the Main Control Room and, as ground to phase faults                              Room and, as appropriate,                          appropriate, in any combination of the three phases on            b. either trip or fast transfer the nonsafety-      b. either trip or fast transfer the nonsafety-the primary [high] or secondary [low] side of          related electrical busses to an alternate          related electrical busses to an alternate the transformer will:                                  power source.                                      power source.
: a. alarm to the Main Control Room and, as appropriate,
: b. either trip or fast transfer the nonsafety-Design Control Document/Tier 1 related electrical busses to an alternate electric power source.
2.12-13
 
Table 2.12.1 Electric Power Distribution System (Continued)
ABWR 2.12-14 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 29. The nonsafety-related electric power feeder 29. A test will be performed on the as-built      29. Using simulated signals, at the designated circuit breakers to each safety-related bus    nonsafety-related protective relays, using        nonsafety-related protective relay setpoints, shall be monitored for an unbalanced phase      simulated signals, to demonstrate that, at the    each UPC is:
condition (UPC), the source of which is        designated protective relay setpoints, each
: a. detected, provided by nonsafety-related offsite electric  UPC is:
power source. The monitoring nonsafety-                                                            b. an alarm is sent to the Main Control
: a. detected, related micro-processor based protective                                                              Room, and relays shall:                                  b. an alarm is sent to the Main Control
: c. a trip signal is sent to open the Room, and
: a. detect the UPC,                                                                                    nonsafety-related UAT NPP and RAT
: c. a trip signal is sent to open the                  APP power feeder breakers to the Class
: b. send an alarm via the alarm system to nonsafety-related UAT NPP and RAT                  1 E safety-related bus.
the Main Control Room, and APP power feeder breakers to the Class 25A5675AA Revision 7
: c. send a trip signal to open the nonsafety-        1E safety-related bus.
related offsite electric power feeder circuit breakers to the Class 1E safety-related bus.
Design Control Document/Tier 1 Electrical Power Distribution System
 
Table 2.12.1 Electric Power Distribution System (Continued)
ABWR Electrical Power Distribution System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                              Acceptance Criteria
: 30. Class 1E safety-related micro-processor          30. A test will be performed on each as-built        30. Using simulated signals, at the designated based protective relays located on the              Class 1E safety-related protective relay,            Class 1E safety-related protective relay medium voltage safety-related electric power        using simulated signals in any combination          setpoints, in any combination of phase fault busses control the safety-related normal and        of fault or phase loss conditions, to                or loss conditions, the as-built safety-related alternative electric power feeder circuit            demonstrate that, at the designated safety-          protective relays will perform the following breakers to ensure that the safety-related          related protective relay setpoints, the              functions:
busses are protected against any                    appropriate safety-related protective relays
: a. If an alternate nonsafety-related power combination or fault or phase loss conditions.      will perform the following functions:
source is available, then the safety-Safety-related protective relays will perform
: a. If an alternate nonsafety-related power                related protective relays controlling the the following functions:
source is available, then the safety-                safety-related feeder circuit breakers will
: a. If an alternate nonsafety-related power                related protective relays controlling the            appropriately trip or fast transfer power source is available, then the safety-                safety-related feeder circuit breakers will          between incoming alternate nonsafety-25A5675AA Revision 7 related protective relays controlling the            appropriately trip or fast transfer power            related power sources, or safety-related feeder circuit breakers will          between incoming alternate nonsafety-
: b. If no alternate nonsafety-related power appropriately trip or fast transfer power            related power sources, or source is available, then the safety-between incoming alternate nonsafety-
: b. If no alternate nonsafety-related power                related protective relays controlling the related power sources, or source is available, then the safety-                appropriate safety-related circuit
: b. If no alternate nonsafety-related power                related logic and safety-related                    breakers will:
source is available, then the safety-                protective relays controlling the (1) isolate the safety-related bus, related logic and safety-related                    appropriate safety-related circuit protective relays controlling the                    breakers will:                                      (2) shed the safety-related loads to Design Control Document/Tier 1 appropriate safety-related circuit                                                                            protect the safety-related equipment, (1) isolate the safety-related bus, breakers will:                                                                                            (3) start the safety-related emergency (2) shed the safety-related loads to (1) isolate the safety-related bus,                                                                          diesel generator, and protect the safety-related equipment, (2) shed safety-related loads to protect                                                                  (4) sequence on the safety-related (3) start the safety-related emergency the safety-related equipment,                                                                            loads.
diesel generator, and (3) start the safety-related emergency (4) sequence on the safety-related diesel generator, and loads.
(4) sequence on the safety-related loads.
2.12-15
 
Table 2.12.1 Electric Power Distribution System (Continued)
ABWR 2.12-16 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 31. There is no data communication from the        31. An inspection will be performed on each as-      31. An inspection demonstrates that there is no nonsafety-related to the Class 1E safety-          built safety-related and nonsafety-related          data communication from the nonsafety-related micro-processor based protective          protective relay(s) to verify that there is no      related to the safety-related protective relays.                                            data communication from the nonsafety-              relays.
related to the safety-related protective relays.
25A5675AA Revision 7 Design Control Document/Tier 1 Electrical Power Distribution System
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 2.12.2 Unit Auxiliary Transformer No entry. Covered in Section 2.12.1.
Unit Auxiliary Transformer                                                                  2.12-17
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 2.12.3 Isolated Phase Bus No entry. Covered in Section 2.12.1.
2.12-18                                                                      Isolated Phase Bus
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 2.12.4 Nonsegregated Phase Bus No entry. Covered in Section 2.12.1.
Nonsegregated Phase Bus                                                                  2.12-19
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.12.5 Metal Clad Switchgear No entry. Covered in Section 2.12.1.
2.12-20                                                                    Metal Clad Switchgear
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 2.12.6 Power Center No entry. Covered in Section 2.12.1.
Power Center                                                                            2.12-21
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.12.7 Motor Control Center No entry. Covered in Section 2.12.1.
2.12-22                                                                      Motor Control Center
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 2.12.8 Raceway System No entry. Covered in Section 2.12.1.
Raceway System                                                                          2.12-23
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.12.9 Grounding Wire No entry. Covered in Section 2.12.1.
2.12-24                                                                        Grounding Wire
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 2.12.10 Electrical Wiring Penetration Design Description Electrical penetrations are provided for electrical cables passing through the primary containment.
Electrical penetrations are classified as safety-related.
Electrical penetrations are protected against currents that are greater than their continuous current rating.
Electrical penetrations are classified as Seismic Category I.
Divisional electrical penetrations only contain cables of one Class 1E division. Independence is provided between divisional electrical penetrations and also between divisional electrical penetrations and penetrations containing non-Class 1E cables.
Electrical penetrations are qualified for a harsh environment.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.10 provides a definition of the inspections, tests, and/or analyses, together with the associated acceptance criteria, which will be undertaken for the Electrical Wiring Penetrations.
Electrical Wiring Penetration                                                                              2.12-25
 
ABWR 2.12-26 Table 2.12.10 Electrical Wiring Penetration Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                              Acceptance Criteria
: 1. The basic configuration of the Electrical        1. Inspections of the as-built Electrical Wiring    1. The as-built Electrical Wiring Penetration Wiring Penetration is described in Section          Penetration will be conducted.                      conforms with the basic configuration 2.12.10.                                                                                                described in Section 2.12.10.
: 2. Electrical penetrations are protected against 2. Analyses for the as-built electrical                2. Analyses for the as-built electrical currents that are greater than their            penetrations and protective features will be          penetrations and protective features exist continuous current ratings.                      performed.                                            and conclude either 1) that the maximum current of the circuits does not exceed the continuous current rating of the penetration, or 2) that the circuits have redundant protective devices in series and that the redundant protection devices are 25A5675AA Revision 7 coordinated with the penetration's rated short circuit thermal capacity data and prevent current from exceeding the continuous current rating of the electrical penetrations.
: 3. Divisional electrical penetrations only contain 3. Inspections of the as-built divisional electrical 3. As-built divisional electrical penetrations only cables of one Class 1E division.                  penetrations will be conducted.                      contain cables of one Class 1E division.
: 4. Independence is provided between divisional 4. Inspections of the as-built electrical                4. Physical separation exists between as-built electrical penetrations and between            penetrations will be conducted.                          divisional electrical penetrations. Physical divisional electrical penetrations and                                                                  separation exists between these divisional Design Control Document/Tier 1 penetrations containing non-Class 1E                                                                    electrical penetrations and penetrations cables.                                                                                                containing non-Class 1E cables.
Electrical Wiring Penetration
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.12.11 Combustion Turbine Generator Design Description The Combustion Turbine Generator (CTG) is a self-contained unit with its own supporting auxiliary systems. The CTG functions as an alternate AC power source.
The CTG is classified as non-safety-related.
The CTG can supply power to the non-Class 1E plant investment protection (PIP) busses or to the Class 1E divisional busses. The CTG capacity to supply power is at least as large as the capacity of an emergency diesel generator (DG). The CTG is located outside the Reactor Building.
The CTG has the following displays and controls in the main control room (MCR):
(1)    Parameter displays for the CTG output voltage, amperes, kVA, and frequency.
(2)    Controls for manually initiating the CTG.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.11 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the CTG.
Combustion Turbine Generator                                                                          2.12-27
 
ABWR 2.12-28 Table 2.12.11 Combustion Turbine Generator Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the CTG is        1. Inspections of the as-built CTG will be      1. The as-built CTG conforms with the basic described in Section 2.12.11.                  conducted.                                      configuration described in Section 2.12.11.
: 2. The CTG can supply power to the non-Class 2. Tests on the as-built CTG will be conducted      2. The as-built CTG can supply power to the 1E busses or to the Class 1E divisional      by connecting the CTG to the non-Class 1E          non-Class 1E PIP busses or to the Class 1E busses.                                      PIP busses and to the Class 1E divisional          divisional busses.
busses.
: 3. The CTG capacity to supply power is at least 3. Inspections of the as-built CTG and DGs will 3. The as-built CTG capacity to supply power is as large as the capacity of a DG.              be conducted.                                  at least as large as the capacity of a DG, as determined by the CTG and DG nameplate ratings.
25A5675AA Revision 7
: 4. MCR displays and controls provided for the  4. Inspections will be conducted on the MCR      4. Displays and controls exist or can be CTG are as defined in Section 2.12.11.          displays and controls for the CTG.              retrieved in the MCR as defined in Section 2.12.11.
Design Control Document/Tier 1 Combustion Turbine Generator
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 2.12.12 Direct Current Power Supply Design Description The Direct Current Power Supply consists of Class 1E and non-Class 1E batteries, battery chargers, and their respective direct current (DC) distribution panels, motor control centers (MCC), power, and instrumentation and control cables to the distribution system loads. The DC distribution system also includes the protection equipment provided to protect the DC distribution equipment. The Class 1E Direct Current Power Supply and its connections to the Electrical Power Distribution (EPD) System are shown on Figure 2.12.12.
The Class 1E DC electrical power distribution system consists of four Class 1E divisions (Divisions I, II, III, and IV) of batteries with their respective DC electrical distribution panels, DC MCCs, if provided for motor loads, and battery chargers. The Class 1E DC distribution system provides DC power to Class 1E DC equipment and instrumentation and control circuits.
The non-Class 1E DC electrical power distribution system consists of non-Class 1E batteries with their respective DC electrical distribution panels, DC MCC, if provided for motor loads, and battery chargers. The non-Class 1E DC distribution system provides DC power to non-Class 1E DC equipment and instrumentation and control circuits.
Except for Division IV, each Class 1E divisional (Divisions I,II, and III) battery is provided with a normal battery charger supplied alternating current (AC) power from a MCC in the same Class 1E division as the battery. The Division IV normal battery charger is supplied AC power from a Division II MCC. There are no automatic connections between Class 1E divisions.
Interlocks are provided to prevent manual paralleling between Class 1E divisions.
Each Class 1E battery is sized to supply its design loads, at the end-of-installed-life, for a minimum of 2 hours without recharging.
Each Class 1E normal battery charger is sized to supply its respective Class 1E division's normal steady-state loads while charging its respective Class 1E battery.
The Class 1E battery, and battery charger circuit breakers, and DC distribution panels, MCCs, and their circuit breakers and fuses are sized to supply their load requirements. The Class 1E battery, battery charger, and DC distribution panels, and MCCs are rated to withstand fault currents for the time required to clear the fault from its power source. Circuit breakers and fuses in Class 1E battery, battery charger, DC distribution panel, and MCC circuits are rated to interrupt fault currents.
Class 1E DC electrical distribution system circuit interrupting devices (circuit breakers and fuses) are coordinated so that the circuit interrupter closest to the fault opens before other devices.
Direct Current Power Supply                                                                                  2.12-29
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 Class 1E DC electrical distribution system cables are sized to supply their load requirements and are rated to withstand fault currents for the time required to clear the fault from its power source.
The Class 1E DC electrical distribution system supplies an operating voltage at the terminals of the Class 1E utilization equipment that is within the utilization equipment's voltage tolerance limits.
Each Class 1E battery is located in a Seismic Category I structure and in its respective divisional battery room.
Class 1E DC distribution panels and MCCs are identified according to their Class 1E division and are located in Seismic Category I structures and in their respective divisional areas.
Class 1E DC distribution system cables and raceways are identified according to their Class 1E division. Class 1E divisional cables are routed in Seismic Category I structures and in their respective divisional raceways.
For the Class 1E DC electrical distribution system, independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
The only non-Class 1E load connected to the Class 1E DC electrical power distribution system is the associated DC emergency lighting system.
The Class 1E DC power supply has the following alarms and displays in the main control room (MCR):
(1)  Alarms for battery ground detection.
(2)  Parameter Displays for battery voltage and amperes.
(3)  Status indication for battery circuit breaker/disconnect position.
Class 1E equipment is classified as Seismic Category I.
Class 1E equipment which is located in areas designated as harsh environment areas is qualified for harsh environments.
2.12-30                                                                            Direct Current Power Supply
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.12 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Direct Current Power Supply.
Direct Current Power Supply                                                                              2.12-31
 
ABWR 2.12-32                        DIV II                                            DIV I  DIV IV                                  DIV III BATTERY                                          BATTERY BATTERY                                  BATTERY DIV II                    DIV I                          DIV II          DIV III AC MCC                    AC MCC                        AC MCC          AC MCC NORMAL                    NORMAL                        NORMAL            NORMAL BATTERY                    BATTERY                      BATTERY          BATTERY CHARGER                    CHARGER                      CHARGER          CHARGER 25A5675AA Revision 7 DC                    DC                                DC            DC DISTRIBUTION          DISTRIBUTION                      DISTRIBUTION  DISTRIBUTION PANEL                  PANEL                            PANEL          PANEL Design Control Document/Tier 1 LOCAL        CVCF        DCMCC  LOCAL      CVCF        LOCAL        CVCF      LOCAL        CVCF DISTR        POWER                DISTR      POWER        DISTR        POWER      DISTR        POWER Direct Current Power Supply PANEL        SUPPLY              PANEL      SUPPLY      PANEL        SUPPLY    PANEL        SUPPLY (TYP)                            (TYP)                  (TYP)                  (TYP)
DIV II DC                    DIV I DC                      DIV IV DC            DIV III DC Figure 2.12.12 Direct Current Power Supply (Class 1E)
 
ABWR Direct Current Power Supply Table 2.12.12 Direct Current Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the Direct Current 1. Inspections of the as-built system will be        1. The as-built Direct Current Power Supply Power Supply is described in Section            conducted.                                          conforms with the basic configuration 2.12.12.                                                                                              described in Section 2.12.12.
: 2. Except for Division IV, each Class 1E          2. Inspections of the as-built Class 1E Direct    2. Each as-built Class 1E divisional (Divisions divisional (Divisions I,II, and III) battery is    Current Power Supply will be conducted.            I,II, and III) battery is provided with a normal provided with a normal battery charger                                                                battery charger supplied AC power from a supplied AC power from a MCC in the same                                                              MCC in the same Class 1E division as the Class 1E division as the battery. The Division                                                        battery. The Division IV normal battery IV normal battery charger is supplied AC                                                              charger is supplied AC power from a Division power from a Division II MCC.                                                                        II MCC.
: 3. Interlocks are provided to prevent manual      3. Tests of the as-built Class 1E interlocks will  3. The as-built Class 1E interlocks prevent 25A5675AA Revision 7 paralleling between Class 1E divisions.          be conducted by attempting to close each            paralleling between Class 1E divisions. The interlocked pair of breakers.                      connections between Class 1E divisions are manual only.
: 4. Each Class 1E battery is sized to supply its 4.                                                4.
design loads, at the end-of-installed-life, for a
: a. Analyses for the as-built Class 1E          a. Analyses for the as-built Class 1E minimum of 2 hours without recharging.
batteries to determine battery capacities      batteries exist and conclude that each will be performed based on the design          Class 1E battery has the capacity, as duty cycle for each battery.                    determined by the as-built battery rating, to supply its analyzed design loads, at Design Control Document/Tier 1 the end-of-installed-life, for a minimum of 2 hours without recharging.
: b. Tests of each as-built class 1E battery          b. The capacity of each as-built Class 1E will be conducted by simulating loads              battery equals or exceeds the analyzed which envelope the analyzed battery                battery design duty cycle capacity.
design duty cycle.
2.12-33
 
Table 2.12.12 Direct Current Power Supply (Continued)
ABWR 2.12-34 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 5. Each Class 1E normal battery charger is        5. Tests of each as-built Class 1E normal        5. Each as-built Class 1E normal battery sized to supply its respective Class 1E            battery charger will be conducted by              charger can supply its respective Class 1E division's normal steady state loads while        supplying its respective Class 1E division's      division's normal steady state loads while charging its respective Class 1E battery.          normal steady state loads while charging its      charging its respective Class 1E battery.
respective Class 1E battery.
: 6. The Class 1E DC battery and battery charger 6. Analyses for the as-built Class 1E DC        6. Analyses for the as-built Class 1E DC circuit breakers, and DC distribution panels,  electrical distribution system to determine    electrical distribution system exist and MCCs, and their circuit breakers and fuses,    the capacities of the battery and battery      conclude that the capacities of Class 1E are sized to supply their load requirements. charger circuit breakers and DC distribution    battery and battery charger circuit breakers, panels, MCCs, and their circuit breakers and    and DC distribution panels, MCCs, and their fuses, will be performed.                      circuit breakers and fuses, as determined by their nameplate ratings, exceed their 25A5675AA Revision 7 analyzed load requirements.
Design Control Document/Tier 1 Direct Current Power Supply
 
Table 2.12.12 Direct Current Power Supply (Continued)
ABWR Direct Current Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 7.                                                  7.                                                7.
: a. The Class 1E battery, battery chargers,          a. Analyses for the as-built Class 1E DC          a. Analyses for the as-built Class 1E DC and DC distribution panels, and MCCs                electrical distribution system to                electrical distribution system exist and are rated to withstand fault currents for          determine fault currents will be                  conclude that the capacities of as-built the time required to clear the fault from          performed.                                        Class 1E battery, battery charger, DC its power source.                                                                                    distribution panel, and MCC current capacities exceed their analyzed fault currents for the time required, as determined by the circuit interrupting device coordination analyses, to clear the fault from its power source.
25A5675AA Revision 7
: b. Circuit breakers and fuses in Class 1E          b. Analyses for the as-built Class 1E DC          b. Analyses for the as-built Class 1E DC battery, battery charger, DC distribution          electrical distribution system to                electrical distribution system exist and panel, and MCC circuits are rated to                determine fault currents will be                  conclude that the analyzed fault currents interrupt fault currents.                          performed.                                        do not exceed the interrupt capacity of circuit breakers and fuses in the battery, battery charger, DC distribution panel, and MCC circuit, as determined by their nameplate ratings.
: 8. Class 1E DC electrical distribution system    8. Analyses for the as-built Class 1E DC            8. Analyses for the as-built Class 1E DC circuit interrupting devices (circuit breakers    electrical distribution system to determine        electrical distribution system circuit Design Control Document/Tier 1 and fuses) are coordinated so that the circuit    circuit interrupting device coordination will be    interrupting devices exist and conclude that interrupter closest to the fault opens before    performed.                                          the analyzed circuit interrupter closest to the other devices.                                                                                        fault will open before other devices.
: 9. Class 1E DC electrical distribution system      9. Analyses for the as-built Class 1E DC          9. Analyses for the as-built Class 1E DC cables are sized to supply their load              electrical distribution system cables to          electrical distribution system cables exist and requirements.                                      determine their load requirements will be        conclude that the Class 1E DC electrical performed.                                        distribution system cable capacities, as determined by cable ratings, exceed their analyzed load requirements.
2.12-35
 
Table 2.12.12 Direct Current Power Supply (Continued)
ABWR 2.12-36 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                            Acceptance Criteria
: 10. Class 1E DC electrical distribution system        10. Analyses for the as-built Class 1E DC          10. Analyses for the as-built Class 1E DC cables are rated to withstand fault currents          electrical distribution system to determine        electrical distribution system exist and for the time required to clear the fault from its    fault currents will be performed.                  conclude that the Class 1E DC electrical power source.                                                                                            distribution system cables will withstand the analyzed fault currents for the time required, as determined by the circuit interrupting device coordination analyses, to clear the fault from its power source.
: 11. The Class 1E DC electrical distribution          11.                                                11.
system supplies an operating voltage at the
: a. Analyses for the as-built Class 1E DC            a. Analyses for the as-built Class 1E DC terminals of the Class 1E utilization electrical distribution system to                  electrical distribution system exist and equipment that is within the utilization 25A5675AA Revision 7 determine system voltage drops will be              conclude that the analyzed operating equipment's voltage tolerance limits.
performed.                                          voltage supplied at the terminals of the Class 1E utilization equipment is within the utilization equipment's voltage tolerance limits, as determined by their nameplate ratings.
: b. Tests of the as-built Class 1E DC system        b. Connected as-built Class 1E loads will be conducted by operating                      operate at less than or equal to the connected Class 1E loads at less than or            minimum allowable battery voltage and equal to the minimum allowable battery              at greater than or equal to the maximum Design Control Document/Tier 1 voltage and at greater than or equal to            battery charging voltage.
the maximum battery charging voltage.
Direct Current Power Supply
: 12. Each Class 1E battery is located in a Seismic 12. Inspections of the as-built Class 1E batteries 12. Each as-built Class 1E battery is located in a Category I structure and in its respective        will be conducted.                                Seismic Category I structure and in its divisional battery room.                                                                            respective divisional battery room.
 
Table 2.12.12 Direct Current Power Supply (Continued)
ABWR Direct Current Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                              Acceptance Criteria
: 13. Class 1E DC distribution panels, and MCCs 13. Inspections of the as-built Class 1E DC              13. As-built DC distribution panels and MCCs are identified according to their Class 1E      distribution panels and MCCs will be                  are identified according to their Class 1E division and are located in Seismic Category    conducted.                                            division and are located in Seismic Category I structures and in their respective divisional                                                        I structures and in their respective divisional areas.                                                                                                areas.
: 14. Class 1E DC distribution system cables and 14. Inspections of the as-built Class 1E DC      14. As-built Class 1E DC distribution system raceways are identified according to their    distribution system cables and raceways will    cables and raceways are identified according Class 1E division. Class 1E divisional cables  be conducted.                                    to their Class 1E division. Class 1E divisional are routed in Seismic Category I structures                                                    cables are routed in Seismic Category I and in their respective divisional raceways.                                                    structures and in their respective divisional raceways.
25A5675AA Revision 7
: 15. For the Class 1E DC electrical distribution  15.                                                  15.
system, independence is provided between
: a. Tests will be conducted on the as-built          a. A test signal exists in only the Class 1E Class 1E divisions, and between Class 1E DC electrical distribution system by                division under test in the DC electrical divisions and non-Class 1E equipment.                  providing a test signal in only one Class            distribution system.
1E division at a time.
: b. Inspections of the as-built DC electrical        b. In the as-built DC electrical distribution distribution system will be conducted.              system, physical separation or electrical isolation exists between Class 1E divisions. Physical separation or Design Control Document/Tier 1 electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 16. MCR alarms and displays provided for the      16. Inspections will be conducted on the alarms      16. Alarms and displays exist or can be retrieved Direct Current Power Supply are as defined        and displays for the Direct Current Power            in the MCR as defined in Section 2.12.12.
in Section 2.12.12.                              Supply.
2.12-37
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.12.13 Emergency Diesel Generator System Design Description The Emergency Diesel Generator (DG) System consists of three diesel engines and their respective combustion air intake system, starting air system, fuel oil system (from the day tank to the engine), lubricating oil system, engine jacket cooling water system, engine exhaust system and silencer, governor system, and generator with its excitation and voltage regulation systems.
The three DGs are classified as Class 1E, safety-related and supply standby AC power to their respective Class 1E Electrical Power Distribution (EPD) System divisions (Divisions I, II, and III). The DG connections to the EPD System are shown on Figure 2.12.1.
The DGs are sized to supply their load demand following a loss-of-coolant accident (LOCA).
The DG air start receiver tanks are sized to provide five DG starts without recharging their tanks.
A loss of preferred power (LOPP) signal (bus under-voltage) from an EPD System medium voltage divisional bus automatically starts its respective DG, and initiates automatic load shedding and connection of the DG to its divisional bus. A DG automatically connects to its respective bus when DG required voltage and frequency conditions are established and required motor loads are tripped. After a DG connects to its respective bus, the non-accident loads are automatically sequenced onto the bus.
LOCA signals from the Residual Heat Removal (RHR) (Division I) and High Pressure Core Flooder (HPCF) (Divisions II and III) systems automatically start their respective divisional DG. After starting, the DGs remain in a standby mode (i.e. running at required voltage and frequency, but not connected to their busses), unless a LOPP signal exists. When LOCA and LOPP signals exist, load shedding occurs and required motor loads are tripped, the DG automatically connects to its respective divisional bus. After a DG connects to its respective bus, the LOCA loads are automatically sequenced onto the bus.
A manual start signal from the main control room (MCR) or from the local control station in the DG area starts a DG. After starting, the DG remains in a standby mode, unless a LOPP signal exists.
DGs start, attain required voltage and frequency, and are ready to load in  20 seconds after receiving an automatic or manual start signal.
When a DG is operating in parallel (test mode) with offsite power, a loss of the offsite power source used for testing or a LOCA signal overrides the test mode by disconnecting the DG from its respective divisional bus.
2.12-38                                                                    Emergency Diesel Generator System
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 The DG units are classified Seismic Category I and DG auxiliary systems are classified Seismic Category I, ASME Code Class 3, and Class 1E, and are located in their respective divisional areas in the Reactor Building. The DG combustion air intakes are located above the maximum flood level. The DG combustion air intakes are separated from DG exhaust ducts. Class 1E DG unit auxiliary systems are supplied electrical power from the same Class 1E division as the DG unit. Independence is provided between Class 1E divisions and also between Class 1E divisions and non-Class 1E equipment. Each divisional DG (Divisions I, II, and III) with its auxiliary systems is physically separated from the other divisions.
The DG System has the following displays and controls in the MCR.
(1)    Parameter displays for the DG output voltage, amperes, watts, vars, frequency, and engine speed.
(2)    Controls for manually starting and stopping the DG units.
The DG System has displays at the Remote Shutdown System (RSS) for DG run and stop indication.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.13 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Emergency Diesel Generator System.
Emergency Diesel Generator System                                                                      2.12-39
 
ABWR 2.12-40 Table 2.12.13 Emergency Diesel Generator System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the DG System is 1. Inspection of the as-built system will be      1. The as-built DG System conforms with the described in Section 2.12.13.                  conducted.                                        basic configuration described in Section 2.12.13.
: 2. The DGs are sized to supply their load      2. Analyses to determine DG load demand,          2. Analyses for the as-built DG systems exist demand following a LOCA.                        based on the as-built DG load profile, will be    and conclude that the DG System capacities performed.                                        exceed, as determined by their nameplate ratings, their load demand following a LOCA.
: 3. DG air start receiver tanks have capacity for 3. Tests on the as-built DG Systems will be    3. As-built DGs start five times without five DG starts without recharging their tanks. conducted by starting the DGs five times.      recharging their air start receiver tanks.
: 4. A LOPP signal (bus under-voltage) from an 4. Tests on the as-built DG Systems will be        4. As-built DGs automatically start on receiving 25A5675AA Revision 7 EPD System medium voltage divisional bus      conducted by providing a simulated LOPP            a LOPP signal and attain a voltage and automatically starts its respective DG, and  signal.                                            frequency in  20 seconds which assures an initiates automatic load shedding and                                                            operating voltage and frequency at the connection of the DG to its divisional bus. A                                                    terminals of the Class1E utilization DG automatically connects to its respective                                                      equipment that is within the tolerance limits bus when DG required voltage and                                                                of the utilization equipment, automatically frequency conditions are established and                                                        connect to their respective divisional bus, required motor loads are tripped. After a DG                                                    after required motor loads are tripped, and connects to its respective bus, the non-                                                        sequence their non-accident loads onto the accident loads are automatically sequenced                                                      bus.
onto the bus.
Design Control Document/Tier 1
: 5. LOCA signals from the RHR (Division I) and 5. Tests on the as-built DG Systems will be        5. As-built DGs automatically start on receiving Emergency Diesel Generator System HPCF (Divisions II and III) System              conducted by providing a simulated LOCA          a LOCA signal and attain a voltage and automatically start their respective divisional signal, without a LOPP signal.                  frequency in  20 seconds which assures an DG. After starting, the DGs remain in a                                                          operating voltage and frequency at the standby mode (i.e. running at required                                                          terminals of the Class 1E utilization voltage and frequency, but not connected to                                                      equipment that is within the tolerance limits their busses), unless a LOPP signal exists.                                                      of the utilization equipment, and remain in the standby mode.
 
Table 2.12.13 Emergency Diesel Generator System (Continued)
ABWR Emergency Diesel Generator System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 6. When LOCA and LOPP signals exist, load        6. Tests on the as-built DG Systems will be  6. In the as-built DG Systems, when LOCA and shedding occurs, and required motor loads        conducted by providing simulated LOCA and    LOPP signals exist, the DG automatically are tripped, the DG automatically connects to    LOPP signals.                                connects to its respective divisional bus. The its respective divisional bus. After a DG                                                    automatic load sequence begins at  20 connects to its respective bus, the LOCA                                                      seconds. Following application of each load, loads are automatically sequenced onto the                                                    the bus voltage does not drop more than bus.                                                                                          25% measured at the bus. Frequency is restored to within 2% of nominal, and voltage is restored to within 10% of nominal within 60% of each load sequence time interval.
The HPCF and RHR pump motor loads are sequenced on to the bus in  36 seconds for 25A5675AA Revision 7 design basis events.
: 7. A manual start signal from the MCR or from 7. Tests on the as-built DG Systems will be      7. As-built DGs automatically start on receiving the local control station in the DG area starts conducted by providing a manual start signal    a manual start signal from the MCR or from a DG. After starting, the DG remains in a      from the MCR and from the local control        the local control station and attain a voltage standby mode (i.e. running at required          station, without a LOPP signal.                and frequency in  20 seconds which voltage and frequency, but not connected to                                                    assures an operating voltage and frequency its bus), unless a LOPP signal exists.                                                          at the terminals of the Class 1E utilization equipment that is within the tolerance limits of the utilization equipment and remain in the standby mode.
Design Control Document/Tier 1
: 8. When a DG is operating in parallel (test        8. Tests on the as-built DG Systems will be  8. When the as-built DG Systems are operating mode) with offsite power, a loss of the offsite    conducted by providing simulated loss of      in the test mode with offsite power and a loss power source used for testing or a LOCA            offsite power and LOCA signals while          of offsite power or a LOCA signal is received, signal overrides the test mode by                  operating the DGs in the test mode.          DGs automatically disconnect from their disconnecting the DG from its respective                                                        respective divisional buses.
divisional bus.
2.12-41
 
Table 2.12.13 Emergency Diesel Generator System (Continued)
ABWR 2.12-42 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                          Acceptance Criteria
: 9. In the DG system, Class 1E DG unit auxiliary 9.                                            9.
systems are supplied electrical power from
: a. Tests will be conducted in the as-built    a. The test signal exists in only the Class the same Class 1E division as the DG unit.
DG Systems by providing a test signal in      1E division under test in the DG System.
Independence is provided between Class 1E only one Class 1E division at a time.
divisions and between Class 1E divisions                                                      b. In the DG systems, physical separation and non-Class 1E equipment.                    b. Inspections of the as-built Class 1E          or electrical isolation exists between divisions in the DG systems will be            Class 1E divisions. Physical separation conducted.                                    or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 10. Each divisional DG (Divisions I, II, and III)      10. Inspections of the as-built DG Systems will  10. Each DG with its auxiliary systems is with its auxiliary systems is physically              be conducted.                                    physically separated from the other divisions 25A5675AA Revision 7 separated from the other divisions.                                                                      by structural and/or fire barriers.
: 11. MCR displays and controls provided for the 11. Inspections will be conducted on the MCR              11. Displays and controls exist or can be DG System are as defined in Section 2.12.13    displays and controls for the as-built DG                retrieved in the MCR as defined in Section Systems.                                                  2.12.13.
: 12. RSS displays provided for the DG System            12. Inspections will be conducted on the RSS      12. Displays exist or can be retrieved on the are as defined in Section 2.12.13                      displays for the as-built DG Systems.            RSS as defined in Section 2.12.13.
Emergency Diesel Generator System Design Control Document/Tier 1
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.12.14 Vital AC Power Supply Design Description The Vital AC Power Supply consists of Class 1E and non-Class 1E uninterruptible power supplies, and their respective alternating current (AC) distribution panels, power, and instrumentation and control cables to the distribution system loads. The AC distribution system also includes the protection equipment provided to protect the AC distribution equipment. The Class 1E Vital AC Power Supply connections to the Electrical Power Distribution (EPD)
System and the Direct Current Power Supply are shown on Figure 2.12.14.
The Class 1E Vital AC Power Supply consists of four divisions (Division I, II, III, and IV) of uninterruptible power supplies with their respective distribution panels. Each Class 1E power supply provides uninterruptible, regulated AC power to Class 1E circuits which require continuity of power during a loss of preferred power (LOPP). Each Class 1E Vital AC Power Supply is a constant voltage constant frequency (CVCF) inverter power supply unit.
The non-Class 1E Vital AC Power Supply consists of uninterruptible power supplies with their respective distribution panels. Each non-Class 1E power supply provides uninterruptible, regulated AC power to non-Class 1E circuits which require continuity of power during a LOPP.
Each non-Class 1E Vital AC Power Supply is a CVCF inverter power supply unit.
Each Class 1E CVCF unit has three input power sources. Except for the Division IV CVCF unit, the normal power to each Class 1E CVCF unit is supplied from an AC motor control center (MCC) in the same Class 1E division as the CVCF unit. The Division IV Class 1E CVCF unit is supplied AC power from a Division II AC MCC. The backup power for each Class 1E CVCF unit is supplied from the direct current (DC) battery in the same Class 1E division as the CVCF unit. In addition, each Class 1E CVCF unit contains an alternate power supply. The alternate power supply is supplied power from the same AC power source as the normal power supply.
Each Class 1E CVCF normal and backup power supply is synchronized, in both frequency and phase, with its alternate power supply and maintains continuity of power during transfer from the inverter to the alternate supply. Automatic transfer between each Class 1E CVCF unit's three power sources is provided. Manual transfer between each Class 1E CVCF unit power source is also provided.
Each Class 1E CVCF unit is sized to provide output power to its respective distribution panel loads. There are no automatic connections between Class 1E divisions.
Class 1E CVCF units and their respective distribution panels are identified according to their Class 1E division and are located in Seismic Category I structures and in their respective divisional areas. Independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
Vital AC Power Supply                                                                                    2.12-43
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 Class 1E Vital AC Power Supply system distribution panels and their circuit breakers and fuses are sized to supply their load requirements. Distribution panels are rated to withstand fault currents for the time required to clear the fault from its power source. Circuit breakers and fuses are rated to interrupt fault currents.
Class 1E Vital AC Power Supply system interrupting devices (circuit breakers and fuses) are coordinated so that the circuit interrupter closest to the fault opens before other devices.
Class 1E Vital AC Power Supply system cables are sized to supply their load requirements and are rated to withstand fault currents for the time required to clear the fault from its power source.
The Class 1E Vital AC Power Supply system supplies an operating voltage at the terminals of the Class 1E utilization equipment that is within the utilization equipments voltage tolerance limits.
Class 1E Vital AC Power Supply system cables and raceways are identified according to their Class 1E division. Class 1E divisional cables are routed in Seismic Category I structures and in their respective divisional raceways.
The Class 1E Vital AC Power Supply has alarms for high and low CVCF unit output voltage and frequency in the main control room (MCR).
Class 1E equipment is classified as Seismic Category I.
Class 1E equipment which is located in areas designated as harsh environment areas is qualified for harsh environments.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.14 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Vital AC Power Supply.
2.12-44                                                                                  Vital AC Power Supply
 
25A5675AA Revision 7 ABWR                                                  Design Control Document/Tier 1 Figure 2.12.14 Vital AC Power Supply Vital AC Power Supply                                                          2.12-45
 
ABWR 2.12-46 Table 2.12.14 Vital AC Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the Vital AC      1. Inspections of the as-built system will be    1. The as-built Vital AC Power Supply conforms Power Supply is described in Section            conducted.                                      with the basic configuration described in 2.12.14.                                                                                          Section 2.12.14.
: 2. Each Class 1E CVCF unit has three input      2. Inspections of the as-built Class 1E Vital AC 2. Each as-built CVCF unit has three input power sources. Except for the Division IV      Power Supply system will be conducted.          power sources. Except for the Division IV CVCF unit, the normal power to each Class                                                        CVCF unit, the normal power to each CVCF 1E CVCF unit is supplied from an AC MCC in                                                      unit is supplied from an AC MCC in the same the same Class 1E division as the CVCF                                                          Class 1E division as the CVCF unit. The unit. The Division IV Class 1E CVCF unit is                                                      Division IV CVCF unit is supplied AC power supplied AC power from a Division II AC                                                          from a Division II AC MCC. The backup MCC. The backup power for each Class 1E                                                          power for each CVCF unit is supplied from 25A5675AA Revision 7 CVCF unit is supplied from the DC battery in                                                    the DC battery in the same Class 1E division the same Class 1E division as the CVCF                                                          as the CVCF unit. In addition, each Class 1E unit. In addition, each Class 1E CVCF unit                                                      CVCF unit contains an alternate power contains an alternate power supply. The                                                          supply. The alternate power supply is alternate power supply is supplied power                                                        supplied power from the same AC power from the same AC power source as the                                                            source as the normal power supply.
normal power supply.
: 3. Automatic transfer between each Class 1E    3. Tests on each as-built Class 1E CVCF unit      3. Each as-built Class 1E CVCF unit CVCF unit's three power sources is provided    will be conducted by providing a test signal in    automatically and manually transfers and maintains continuity of power during      one power source at a time. A test of the          between the unit's three power sources and Design Control Document/Tier 1 transfer from the inverter to the alternate    manual transfer will also be conducted.            maintains continuity of power during transfer supply. Manual transfer between each Class                                                        from the inverter to the alternate supply.
1E CVCF unit power source is also provided.
: 4. Each Class 1E CVCF unit is sized to provide 4. Analyses for each as-built Class 1E CVCF        4. Analyses for each as-built Class 1E CVCF output power to its respective distribution    unit to determine the power requirements of        unit exist and conclude that each CVCF Vital AC Power Supply panel loads.                                  its loads will be performed.                      units capacity, as determined by its nameplate rating, exceeds its analyzed load requirements.
 
Table 2.12.14 Vital AC Power Supply (Continued)
ABWR Vital AC Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 5. Class 1E CVCF units and their respective        5. Inspections of the as-built Class 1E CVCF        5. The as-built Class 1E CVCF units and their distribution panels are identified according to    units and their respective distribution panels      respective distribution panels are identified their Class 1E division and are located in        will be conducted.                                  according to their Class 1E division and are Seismic Category I structures and in their                                                            located in Seismic Category I structures and respective divisional areas.                                                                          in their respective divisional areas.
: 6. In the Vital AC Power Supply, independence 6.                                                      6.
is provided between Class 1E divisions, and
: a. Tests on the Vital AC Power Supply will                a. A test signal exists only in the Class 1E between Class 1E divisions and non-Class be conducted by providing a test signal                    division under test in the Vital AC Power 1E equipment.
in only one Class 1E division at a time.                  Supply.
: b. Inspections of the as-built Class 1E              b. In the Vital AC Power Supply, physical divisions in the Vital AC Power Supply              separation or electrical isolation exists 25A5675AA Revision 7 will be conducted.                                  between the Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 7. Class 1E Vital AC Power Supply system            7. Analyses for the as-built distribution panels  7. Analyses for the as-built Class 1E Vital AC distribution panels and their respective circuit    and their respective circuit breakers and          Power Supply system distribution panels and breakers and fuses are sized to supply their        fuses to determine their load requirements        their respective circuit breakers and fuses load requirements.                                  will be performed.                                exist and conclude that the capacities of the distribution panels, circuit breakers, and Design Control Document/Tier 1 fuses exceed, as determined by their nameplate ratings, their analyzed load requirements.
2.12-47
 
Table 2.12.14 Vital AC Power Supply (Continued)
ABWR 2.12-48 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                          Acceptance Criteria
: 8. Class 1E Vital AC Power Supply system              8. Analyses for the as-built Class 1E            8. Analyses for the as-built Class 1E Vital AC distribution panels are rated to withstand            distribution system to determine fault            Power Supply system distribution panels fault currents for the time required to clear        currents will be performed.                      exist and conclude that the current capacities the fault from its power source.                                                                        of the distribution panels, exceed their analyzed fault currents for the time required, as determined by the circuit interrupting device coordination analyses, to clear the fault from its power source.
: 9. Class 1E Vital AC Power Supply system              9. Analyses for the as-built Class 1E            9. Analyses for the as-built Class 1E Vital AC distribution panel circuit breakers and fuses        distribution system to determine fault            Power Supply distribution system exist and are rated to interrupt fault currents.                currents will be performed.                      conclude that the analyzed fault currents do not exceed the distribution system circuit 25A5675AA Revision 7 breakers and fuses interrupt capabilities, as determined by their nameplate ratings.
: 10. Class 1E Vital AC Power Supply system            10. Analyses for the as-built Class 1E            10. Analyses for the as-built Class 1E Vital AC interrupting devices (circuit breakers and            distribution system to determine circuit          Power Supply system circuit interrupting fuses) are coordinated so that the circuit            interrupting device coordination will be          devices (circuit breakers and fuses) interrupter closest to the fault opens before        performed.                                        coordination exist and conclude that the other devices.                                                                                          analyzed circuit interrupter closest to the fault will open before other devices.
: 11. Class 1E Vital AC Power Supply system            11. Analyses for the as-built Class 1E            11. Analyses for the as-built Class 1E Vital AC Design Control Document/Tier 1 cables are sized to supply their load                distribution system cables to determine their    Power Supply system cables exist and requirements.                                        load requirements will be performed.              conclude that the capacities of the distribution system cables exceed, as determined by their cable ratings, their analyzed load requirements.
Vital AC Power Supply
 
Table 2.12.14 Vital AC Power Supply (Continued)
ABWR Vital AC Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 12. Class 1E Vital AC Power Supply system            12. Analyses for the as-built Class 1E          12. Analyses for the as-built Class 1E Vital AC cables are rated to withstand fault currents          distribution system to determine fault          Power Supply system cables exist and for the time required to clear the fault from its    currents will be performed.                    conclude that the distribution system cable power source.                                                                                        current capacities exceed their analyzed fault currents for the time required, as determined by the circuit interrupting device coordination analyses, to clear the fault from its power source.
: 13. The Class 1E Vital AC Power Supply system 13. Analyses for the as-built Class 1E Vital AC        13. Analyses for the as-built Class 1E Vital AC supplies an operating voltage at the          Power Supply system to determine voltage                Power Supply system exist and conclude terminals of the Class 1E utilization        drops will be performed.                                that the analyzed operating voltage supplied equipment that is within the utilization                                                              at the terminals of the Class 1E utilization 25A5675AA Revision 7 equipments voltage tolerance limits.                                                                equipment is within the utilization equipments voltage tolerance limits, as determined by their nameplate ratings.
: 14. Class 1E Vital AC Power Supply system          14. Inspections of the as-built Class 1E Vital AC 14. As-built Class 1E Vital AC Power Supply cables and raceways are identified according        Power Supply system cables and raceways          system cables and raceways are identified to their Class 1E division. Class 1E divisional    will be conducted.                                according to their Class 1E division. Class cables are routed in Seismic Category I                                                              1E divisional cables are routed in Seismic structures and in their respective divisional                                                        Category I structures and in their respective raceways.                                                                                            divisional raceways.
Design Control Document/Tier 1
: 15. MCR alarms provided for the Class 1E Vital      15. Inspections will be conducted on the MCR      15. Alarms exist or can be retrieved in the MCR AC Power Supply are as defined in Section          alarms for the as-built Class 1E Vital AC        as defined in Section 2.12.14.
2.12.14.                                            Power Supply.
2.12-49
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.12.15 Instrument and Control Power Supply Design Description The Instrument and Control Power Supply consists of Class 1E and non-Class 1E interruptible power supplies and their respective alternating current (AC) distribution panels, power, and instrumentation and control cables to the distribution system loads. The AC distribution system also includes the protection equipment provided to protect the AC distribution equipment. The Class 1E Instrument and Control Power Supply connections to the Electrical Power Distribution (EPD) System are shown on Figure 2.12.15.
The Class 1E Instrument and Control Power Supply consists of three divisions (Division I, II, and III) of interruptible power supplies with their respective distribution panels. Each Class 1E power supply provides interruptible, regulated AC power to Class 1E circuits which do not require continuity of power during a loss of preferred power (LOPP).
The non-Class 1E Instrument and Control Power Supply consists of an interruptible power supply with its respective distribution panel. The non-Class 1E power supply provides interruptible, regulated AC power to non-Class 1E circuits which do not require continuity of power during a LOPP.
Each Class 1E Instrument and Control Power Supply is a voltage regulating device. The power to each Class 1E Instrument and Control Power Supply voltage regulating device is supplied from an AC MCC in the same Class 1E division as the device.
Each Class 1E Instrument and Control Power Supply is sized to provide output power to its respective distribution panel loads.There are no automatic connections between Class 1E divisions.
Class 1E Instrument and Control Power Supplies and their respective distribution panels are identified according to their Class 1E division and are located in Seismic Category I structures and in their respective divisional areas. Independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
Class 1E Instrument and Control Power Supply system distribution panels and their circuit breakers and fuses are sized to supply their load requirements. Distribution panels are rated to withstand fault currents for the time required to clear the fault from its power source. Circuit breakers and fuses are rated to interrupt fault currents.
Class 1E Instrument and Control Power Supply system interrupting devices (circuit breakers and fuses) are coordinated so that the circuit interrupter closest to the fault opens before other devices.
2.12-50                                                                      Instrument and Control Power Supply
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Class 1E Instrument and Control Power Supply system cables are sized to supply their load requirements and are rated to withstand fault currents for the time required to clear the fault from its power source.
The Class 1E Instrument and Control Power Supply system supplies an operating voltage at the terminals of the Class 1E utilization equipment that is within the utilization equipments voltage tolerance limits.
Class 1E Instrument and Control Power Supply system cables and raceways are identified according to their Class 1E division. Class 1E divisional cables are routed in Seismic Category I structures and in their respective divisional raceways.
Class 1E equipment is classified as Seismic Category I.
Class 1E equipment which is located in areas designated as harsh environment areas is qualified for harsh environments.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.15 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Instrument and Control Power Supply.
Instrument and Control Power Supply                                                                      2.12-51
 
25A5675AA Revision 7 ABWR                                                      Design Control Document/Tier 1 CLASS 1E INSTRUMENT AND CONTROL POWER SUPPLY CLASS 1E AC MCC VOLTAGE REGULATING DEVICE REACTOR                    CONTROL BUILDING                  BUILDING TYPICAL OF 3 1 PER DIVISION (DIVISIONS I, II, III)
Figure 2.12.15 Instrument and Control Power Supply 2.12-52                                                        Instrument and Control Power Supply
 
ABWR Instrument and Control Power Supply Table 2.12.15 Instrument and Control Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the Instrument      1. Inspections of the as-built system will be      1. The as-built Instrument and Control Power and Control Power Supply is described in          conducted.                                          Supply conforms with the basic configuration Section 2.12.15.                                                                                      described in Section 2.12.15.
: 2. The power to each Class 1E Instrument and 2. Inspections of the as-built Class 1E        2. The power to each as-built Class 1E Control Power Supply voltage regulating      Instrument and Control Power Supply will be    Instrument and Control Power Supply device is supplied from an AC MCC in the    conducted.                                    voltage regulating device is supplied from an same Class 1E division as the device.                                                      AC MCC in the same Class 1E division as the device.
: 3. Each Class 1E AC Instrument and Control        3. Analyses for each as-built Class 1E              3. Analyses for each as-built Class 1E Power Supply is sized to provide output          Instrument and Control Power Supply to              Instrument and Control Power Supply exist power to its respective distribution panel        determine the power requirements of its            and conclude that each Instrument and 25A5675AA Revision 7 loads.                                            loads will be performed.                            Control Power Supply capacity, as determined by its nameplate rating, exceeds its analyzed load requirements.
: 4. Class 1E Instrument and Control Power          4. Inspections of the as-built Class 1E            4. The as-built Class 1E Instrument and Control Supplies and their respective distribution        Instrument and Control Power Supplies and          Power Supplies and their respective panels are identified according to their Class    their respective distribution panels will be        distribution panels are identified according to 1E division and are located in Seismic            conducted.                                          their Class 1E division and are located in Category I structures and in their respective                                                        Seismic Category I structures and in their divisional areas.                                                                                    respective divisional areas.
Design Control Document/Tier 1
: 5. In the Instrumentation and Control Power      5.                                                  5.
Supply, independence is provided between
: a. Tests on the Instrumentation and Control        a. A test signal exists only in the Class 1E Class 1E divisions, and between Class 1E Power Supply will be conducted by                  division under test in the Instrumentation divisions and non-Class 1E equipment.                  providing a test signal in only one Class          and Control Power Supply.
1E division at a time.
: b. In the Instrumentation and Control
: b. Inspections of the as-built Class 1E                Power Supply, physical separation or divisions in the Instrumentation and                electrical isolation exists between the Control Power Supply will be conducted.            Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 2.12-53                                                                                                                                            1E equipment.
 
Table 2.12.15 Instrument and Control Power Supply (Continued)
ABWR 2.12-54 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 6. Class 1E Instrument and Control Power            6. Analyses for the as-built distribution panels  6. Analyses for the as-built Class 1E Instrument Supply system distribution panels and their          and their respective circuit breakers and          and Control Power Supply system respective circuit breakers and fuses are            fuses to determine their load requirements        distribution panels and their respective circuit sized to supply their load requirements.            will be performed.                                breakers and fuses exist and conclude that the capacities of the distribution panels, circuit breakers, and fuses exceed, as determined by their nameplate ratings, their analyzed load requirements.
: 7. Class 1E Instrument and Control Power            7. Analyses for the as-built Class 1E              7. Analyses for the as-built Class 1E Instrument Supply system distribution panels are rated          distribution system to determine fault            and Control Power Supply system to withstand fault currents for the time            currents will be performed.                        distribution panels exist and conclude that required to clear the fault from its power                                                              the current capacities of the distribution 25A5675AA Revision 7 source.                                                                                                panels exceed their analyzed fault currents for the time required, as determined by the circuit interrupting device coordination analyses, to clear the fault from its power source.
: 8. Class 1E Instrument and Control Power          8. Analyses for the as-built Class 1E                8. Analyses for the as-built Class 1E Instrument Supply system distribution panel circuit          distribution system to determine fault              and Control Power Supply distribution breakers and fuses are rated to interrupt fault    currents will be performed.                          system exist and conclude that the analyzed currents.                                                                                              fault currents do not exceed the distribution Design Control Document/Tier 1 system circuit breakers and fuses interrupt capabilities, as determined by their Instrument and Control Power Supply nameplate ratings.
: 9. Class 1E Instrument and Control Power            9. Analyses for the as-built Class 1E              9. Analyses for the as-built Class 1E Instrument Supply system interrupting devices (circuit          distribution system to determine circuit          and Control Power Supply system circuit breakers and fuses) are coordinated so that          interrupting device coordination will be          interrupting devices (circuit breakers and the circuit interrupter closest the fault opens      performed.                                        fuses) coordination exist and conclude that before other devices.                                                                                  the analyzed circuit interrupter closest to the fault will open before other devices.
 
Table 2.12.15 Instrument and Control Power Supply (Continued)
ABWR Instrument and Control Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 10. Class 1E Instrument and Control Power        10. Analyses for the as-built Class 1E            10. Analyses for the as-built Class 1E Instrument Supply system cables are sized to supply          distribution system cables to determine their    and Control Power Supply system cables their load requirements.                          load requirements will be performed.              exist and conclude that the capacities of the distribution system cables exceed, as determined by their cable ratings, their analyzed load requirements.
: 11. Class 1E Instrument and Control Power        11. Analyses for the as-built Class 1E            11. Analyses for the as-built Class 1E Instrument Supply system cables are rated to withstand      distribution system to determine fault            and Control Power Supply system cables fault currents for the time required to clear    currents will be performed.                      exist and conclude that the distribution the fault from its power source.                                                                    system cable current capacities exceed their analyzed fault currents for the time required, as determined by the circuit interrupting 25A5675AA Revision 7 device coordination analyses, to clear the fault from its power source.
: 12. The Class 1E Instrument and Control Power 12. Analyses for the as-built Class 1E Instrument 12. Analyses for the as-built Class 1E Instrument Supply system supplies an operating voltage  and Control Power Supply system to                and Control Power Supply system exist and at the terminals of the Class 1E utilization  determine voltage drops will be performed.        conclude that the analyzed operating voltage equipment that is within the utilization                                                        supplied at the terminals of the Class 1E equipments voltage tolerance limits.                                                          utilization equipment is within the utilization equipments voltage tolerance limits, as determined by their nameplate ratings.
Design Control Document/Tier 1
: 13. Class 1E Instrument and Control Power        13. Inspections of the as-built Class 1E          13. As-built Class 1E Instrument and Control Supply system cables and raceways are            Instrument and Control Power Supply              Power Supply system cables and raceways identified according to their Class 1E            system cables and raceways will be                are identified according to their Class 1E division. Class 1E divisional cables are          conducted.                                        division. Class 1E divisional cables are routed in Seismic Category I structures and                                                        routed in Seismic Category I structures and in their respective divisional raceways.                                                            in their respective divisional raceways.
2.12-55
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.12.16 Communication System Design Description The parts of the plant Communication System within the Certified Design consist of a power-actuated paging and broadcasting system and a separate sound-powered telephone system. The parts of the Communication System associated with off-site communications are not within the Certified Design.
The power-actuated paging system provides intraplant station to station communications and area broadcasting in buildings and outside areas. The system consists of at least two channels, with one channel allowing access from the plant telephone system. Each channel is provided with an amplifier and a distribution frame. Handsets and speakers are provided. The power-actuated paging system is powered from plant power supply and is backed by its own battery.
The sound-powered communication system consists of a main communication patch panel, a set of communication stations and a system of cables and jacks. This system provides communication capability between the main control room (MCR), Remote Shutdown System (RSS) panel, electrical equipment area and diesel generator areas. The patch panel is located outside the MCR. The sound-powered communication system does not require any electrical power source for its operation.
The plant Communication System is classified as non-safety related.
Interface Requirements The parts of the Communication System which are not within the Certified Design shall meet the following requirements:
An emergency communication system for off-site communication shall be provided.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.16 provides a definition of the inspections, tests, and/or analyses, together with the associated acceptance criteria, which will be undertaken for the Communication System.
2.12-56                                                                              Communication System
 
ABWR Communication System Table 2.12.16 Communication System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the plant        1. Inspections of the as-built plant          1. The as-built plant Communication System Communication System is described in          Communication System will be conducted.        conforms with the basic configuration Section 2.12.16.                                                                              described in Section 2.12.16.
: 2. The power actuated paging system provides 2. Tests of the as-built power actuated paging  2. The power actuated paging system provides intraplant, station to station communications system will be conducted.                      intraplant, station to station communications and area broadcasting in buildings and                                                        and area broadcasting in buildings and outside areas.                                                                                outside areas.
: 3. The sound-powered communications system 3. Tests of the as-built sound-powered            3. The sound-powered communications system provides communication capability between  communications system will be conducted.          provides communication capability between the main control room, remote shutdown                                                        the main control room, remote shutdown panel, electrical equipment area and the                                                      panel, electrical equipment area and the 25A5675AA Revision 7 diesel generator areas.                                                                      diesel generator areas.
Design Control Document/Tier 1 2.12-57
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.12.17 Lighting and Servicing Power Supply Design Description The Lighting and Servicing Power Supply (LSPS) consists of multiple lighting systems and a non-Class 1E service power supply system. The non-Class 1E service power supply system supplies power to non-Class 1E loads which are not required for plant power operation.
There are four lighting systems: the normal alternating current (AC) lighting system, the standby AC lighting system, the emergency direct current (DC) lighting system, and the guide lamp lighting system.
The normal AC lighting system provides lighting needed for operation, inspection, and repairs during normal plant operation in areas containing non-safety related equipment. The normal lighting system is part of the plants non-safety-related systems and is supplied by the non-Class 1E power system buses.
The AC standby lighting system is comprised of the non-Class 1E AC standby lighting system and the associated AC standby lighting system. The non-Class 1E AC standby lighting system serves both safety-related and non-safety-related areas and their passageways and stairwells and is powered by the plant investment protection (PIP) busses. The associated AC standby lighting system serves the safety-related divisional areas and the passageways and stairwells leading to the divisional areas.
Each division of associated AC standby lighting is supplied power from its respective Class 1E division (Division I, II, and III). The associated AC standby lighting in the main control room (MCR) is supplied from divisions II and III. The associated AC standby lighting in the division IV battery room and other division IV instrumentation and control areas is supplied from division II.
The DC emergency lighting system is comprised of the non-Class 1E DC emergency lighting system and the associated DC emergency lighting system. The DC emergency lighting system provides DC backup lighting, when AC lighting is lost, until the normal or standby lighting systems are energized. The non-Class 1E DC emergency lighting system supplies the lighting needed in plant areas containing non-safety-related equipment and is supplied by the non-Class 1E DC system. The associated DC emergency lighting system supplies the lighting needed in plant areas containing safety-related equipment.
Each division of associated DC emergency lighting is supplied by power from its respective Class 1E division (Divisions I, II, III, and IV).The associated DC emergency lighting in the MCR is supplied from divisions II and III.
The guide lamp light system serves stairways, exit routes, and major control areas (MCR and Remote Shutdown System (RSS) areas). Each Class 1E guide lamp unit is a self-contained battery pack unit containing a rechargeable battery with a minimum 8-hour capacity. The Class 2.12-58                                                                      Lighting and Servicing Power Supply
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 1E guide lamp units are supplied AC power from the same power source that supplies the associated AC standby lighting system in the area in which they are located. The non-Class 1E guide lamp units in non-safety-related plant areas are supplied power by the non-Class 1E system.
Lighting circuits, excluding lighting fixtures, that are connected to a Class 1E power source are identified as associated circuits and are treated as Class 1E circuits. In the LSPS, independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
Class 1E or associated lighting distribution system equipment is identified according to its Class 1E division and is located in Seismic Category I structures, and in its respective divisional areas.
Class 1E or associated lighting system cables and raceways are identified according to their Class 1E division. Class 1E or associated lighting system cables are routed in their respective divisional raceways and in Seismic Category I structures. Associated DC emergency lighting system cables are not routed with any other cables and are specifically identified as DC lighting.
Class 1E equipment is classified as Seismic Category I.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.12.17 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Lighting and Servicing Power Supply.
Lighting and Servicing Power Supply                                                                        2.12-59
 
ABWR 2.12-60 Table 2.12.17 Lighting and Servicing Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the LSPS is          1. Inspections of the as-built system will be  1. The as-built LSPS conforms with the basic described in Section 2.12.17.                      conducted.                                      configuration described in Section 2.12.17.
: 2. Each division of associated AC standby        2. Tests on the associated AC standby lighting 2. The as-built associated AC standby lighting lighting is supplied power from its respective    will be conducted by providing a test signal in is supplied power only from its respective Class 1E division.                                only one Class 1E division at a time.          Class 1E division.
: 3. The associated AC standby lighting in the      3. Tests on the associated AC standby lighting 3. The as-built associated AC standby lighting MCR is supplied from Divisions II and III.        will be conducted by providing a test signal in in the MCR is supplied from Divisions II and only one Class 1E division at a time.          III.
: 4. The associated AC standby lighting in the      4. Tests on the associated AC standby lighting 4. The as-built associated AC standby lighting Division IV battery room and other Division        will be conducted by providing a test signal in in the Division IV battery room and other 25A5675AA Revision 7 IV instrumentation and control areas is            only one Class 1E division at a time.          Division IV instrumentation and control areas supplied from Division II.                                                                        is supplied from Division II.
: 5. Each division of associated DC emergency 5. Tests on the associated DC emergency              5. The as-built associated DC emergency lighting is supplied power from its respective lighting will be conducted by providing a test    lighting is supplied power from its respective Class 1E division.                            signal in only one Class 1E division at a time. Class 1E division.
: 6. The associated DC emergency lighting in the 6. Tests on the associated DC emergency            6. The as-built associated DC emergency MCR is supplied from Divisions II and III. lighting will be conducted by providing a test    lighting in the MCR is supplied from Divisions signal in only one Class 1E division at a time. II and III.
Lighting and Servicing Power Supply Design Control Document/Tier 1
 
Table 2.12.17 Lighting and Servicing Power Supply (Continued)
ABWR Lighting and Servicing Power Supply Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                            Acceptance Criteria
: 7. Each Class 1E guide lamp unit is a self-    7.                                              7.
contained, battery pack unit containing a
: a. Inspections of the as-built Class 1E        a. The Class 1E guide lamp units are self-rechargeable battery with a minimum 8-hour guide lamp units will be conducted.            contained, battery pack units containing capacity. The Class 1E guide lamp units are a rechargeable battery with a minimum supplied AC power from the same power 8-hour capacity.
source that supplies the associated AC standby lighting system in the area in which    b. Tests on the as-built Class 1E guide        b. The Class 1E guide lamp units are they are located.                                  lamp units will be conducted by providing      supplied AC power from the same power a test signal in only one Class 1E              source that supplies the associated AC division at a time.                            standby lighting system in the area in which it is located. The Class 1E guide lamp units are turned on when the 25A5675AA Revision 7 associated AC standby lighting system in the area in which they are located is lost.
: 8. Lighting circuits, excluding lighting fixtures,  8. Inspections of the associated lighting circuits 8. The as-built associated lighting circuits are that are connected to a Class 1E power              will be conducted.                                identified as associated circuits and treated source are identified as associated circuits                                                            as Class 1E circuits.
and treated as Class 1E circuits.
: 9. In the LSPS, independence is provided            9.                                                  9.
between Class 1E divisions, and between
: a. Tests on the LSPS will be conducted by          a. A test signal exists in only the Class 1E Class 1E divisions and non-Class 1E                      providing a test signal in only one Class          division under test in the LSPS.
Design Control Document/Tier 1 equipment.                                                1E division at a time.
: b. Inspections of the as-built Class 1E            b. In the LSPS, physical separation or divisions in the LSPS will be conducted.            electrical isolation exists between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
2.12-61
 
Table 2.12.17 Lighting and Servicing Power Supply (Continued)
ABWR 2.12-62 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 10. Class 1E or associated lighting distribution  10. Inspections of the as-built Class 1E and    10. The as-built Class 1E and associated lighting system equipment is identified according to        associated lighting systems will be              distribution system equipment is identified its Class 1E division and is located in            conducted.                                      according to its Class 1E division and is Seismic Category I structures, and in its                                                          located in Seismic Category I structures, and respective divisional areas (except for                                                            in its respective divisional areas (except for features in design commitment No. 3, 4 and                                                          features in design commitment No. 3, 4 and 6).                                                                                                6).
: 11. Class 1E or associated lighting system        11. Inspections of the as-built Class 1E and    11. The as-built Class 1E and associated lighting cables and raceways, are identified                associated lighting system cables and            system cables and raceways are identified according to their Class 1E division.              raceways will be conducted.                      according to their Class 1E division.
: 12. Class 1E or associated lighting system      12. Inspections of the as-built Class 1E and        12. The as-built Class 1E and associated lighting 25A5675AA Revision 7 cables are routed in their respective          associated lighting system cables and              system cables are routed in their respective divisional raceways and in Seismic Category    raceways will be conducted.                        divisional raceways and in Seismic Category I structures.                                                                                      I structures.
: 13. Associated DC emergency lighting system        13. Inspections of the as-built associated DC  13. Associated DC emergency lighting system cables are not routed with any other cables        emergency lighting system cables will be        cables are not routed with any other cables and are specifically identified as DC lighting. conducted.                                      and are specifically identified as DC lighting.
Lighting and Servicing Power Supply Design Control Document/Tier 1
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 2.13.1 Reserve Auxiliary Transformer No entry. Covered in Section 2.12.1.
Reserve Auxiliary Transformer                                                              2.13-1
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 2.14.1 Primary Containment System Design Description The Primary Containment System (PCS) encompasses:
(1)    A reinforced concrete containment vessel (RCCV) with an internal steel liner. The structure includes various penetrations, equipment hatches and personnel access locks. This structure provides an essentially leaktight barrier against the uncontrolled release of radioactivity to the environment as long as postulated accident conditions require.
(2)    Structures inside the primary containment which partition the containment into drywell and wetwell regions, provide equipment support, radiation protection, and components for operation of the ABWR pressure suppression containment.
Figure 2.14.1 shows the basic configuration and scope.
The steel-lined reinforced concrete containment structure supported by a reinforced concrete basemat provides the primary containment pressure barrier of the RCCV and is classified as ASME Code Section III. The reactor pressure vessel (RPV) support pedestal and a diaphragm floor partition the containment volume into drywell and wetwell regions. The RPV support pedestal is a double shell steel structure filled with concrete. The diaphragm floor is a reinforced concrete structure. Other major internal structures within the containment are the reactor shield wall, lower drywell personnel and equipment access tunnels and the drywell equipment and piping support structure (DEPSS). These internal structures are steel fabrications.
Penetrations through the containment pressure boundary include the drywell head closure, equipment hatches to both upper and lower drywell regions, personnel locks into upper and lower drywells, a combined personnel access and equipment hatch into the wetwell, and piping and electrical penetration sleeves. These pressure boundary appurtenances are steel structures classified as ASME Code Section III, Division 1, Class MC. Furthermore, the drywell head closure thickness is equal to or greater than 31.7 mm.
The containment design pressure is 309.9 kPaG. The design temperatures for the drywell and the wetwell are 171&deg;C and 104&deg;C, respectively. The maximum calculated pressures and temperatures for the design basis accident are less than these design conditions. The primary containment pressure boundary including penetrations and isolation valves, has a leak rate equal to or less than 0.5% per day (excluding MSIV leakage) of the containment gas mass at the maximum calculated containment pressure for the design basis accident.
The reinforced concrete diaphragm floor, separating the upper drywell and the wetwell gas spaces, has a steel liner plate on the underside. The design differential pressure of the diaphragm floor between drywell and wetwell is 172.6 kPa in the downward direction.
Primary Containment System                                                                                    2.14-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 The RPV pedestal forms the lower drywell region and consists of a cylindrical double shell composite steel structure. It is anchored to the basemat and supports the RPV through a support ring girder. The pedestal also supports the reactor shield wall. The pedestal consists of two concentric steel cylinders joined together radially by vertical steel diaphragms and filled with concrete. The pressure suppression venting paths are an integral part of the pedestal structure, which includes (1) the ducts which interconnect the lower and upper drywell regions, (2) the vertical downcomers from the interconnecting ducts to the horizontal vents, and (3) the horizontal vents that direct steam into the suppression pool. The horizontal vents consist of 30 pipes uniformly spaced around the perimeter of the pedestal in ten stacks of three each. The total horizontal vent area is greater or equal to 11.55 m2. The distance from the pedestal containing these horizontal vents to the outer suppression pool wall is greater than 7.4m. All HVAC ducts, cabling and piping between the upper and lower drywells are routed through the interconnecting ducts.
Vacuum relief between the drywell volumes and the wetwell gas space is provided by vacuum breaker valves on piping sleeves penetrating the pedestal wall. Eight normally closed swing check valves with a total flow area of at least 1.53 m2 are provided. Each vacuum breaker has two position indication switches that provide position indication and an alarm in the main control room (MCR). The positon switches have adequate sensitivity to detect the allowable suppression pool (S/P) bypass capability of the containment.
The water volume in the suppression pool including the vents is equal to or greater than 3,580 cubic meters. The safety relief valve (SRV) discharge lines terminate in standard X type quenchers. The horizontal center line of the safety relief valve discharge line (SRVDL) quencher arms are located at or below the elevation of the center layer of horizontal vents in the suppression pool. The quenchers are placed in the suppression pool in two radial rings. Eighteen of 10 equally spaced locations in each radial ring have quenchers installed.
Water return paths connect the region within the pedestal to the vertical downcomers and horizontal vent paths. The lower drywell floor is provided with corium protection fill of at least 1.5 meters thickness and a minimum 79 m2 area clear of obstructions to debris spreading. The corium protection fill contains less than 4% of calcium carbonate material by weight. Sumps imbedded in the concrete are protected by corium shields. Thermally activated flooding valves are also located in this region.
The following PCS components are classified as Seismic Category I; the reinforced concrete containment structure, the drywell head, equipment hatches to both upper and lower drywell regions, personnel locks into upper and lower drywells, the combined personnel access and equipment hatch into the wetwell, the basemat, the reactor pedestal, the reactor shield wall, the DEPSS, and containment piping and electrical penetration sleeves.
The containment internal structures designated Seismic Category I, are designed and constructed to accommodate the dynamic and static load conditions and load combinations 2.14-2                                                                          Primary Containment System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 associated with the containment design basis accident. The loads to be applied to these structures are associated with:
(1)    Live loads, dead loads, temperature effects and building vibration loads from normal plant operation.
(2)    Earthquakes loads from safe shutdown earthquake.
(3)    Blowdown pressures and temperature from design basis loss-of-coolant accidents.
(4)    Hydrodynamic loads and structural vibrations resulting from steam discharges into the suppression pool.
(5)    Reaction forces on structures resulting from pipe break jets or fluid impacts.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.1 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Primary Containment System.
Primary Containment System                                                                                2.14-3
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 STEEL DRYWELL HEAD CLOSURE 31.7 mm THICK NOTE 5      TEN DRYWELL INTERCONNECTING DUCTS 11.3m2 TOTAL 3F NOTE 4 PERSONNEL AIR LOCK                                                                                    EQUIPMENT REACTOR                              DEPSS NOTE 3                                                                                      HATCH SHIELD WALL NOTE 3 CONCRETE FILLED                                                                      2F A
CONTINUOUS            RPV DIAPHRAGM            STEEL LINER FLOOR                IN DRYWELL                                                      NOTE 5 1F EIGHT WETWELL TO                                                          ACCESS HATCH STEEL LINER            NOTE 3 DRYWELL PLATE ACCESS              VACUUM TUNNEL              BREAKERS                                            B                            B1F WITH                1.53m2                            D EQUIPMENT            TOTAL                TEN VERTICAL                                        ACCESS HATCH                                    DOWNCOMERS                                          TUNNEL WITH NOTE 3                                    11.3m2 TOTAL                                      PERSONNEL WATER RETURN PATHS                WATER                        AIR LOCK CONCRETE FILLED        LEVEL                        NOTE 3 REACTOR PEDESTAL          C                                  B2F REINFORCED            CONTINUOUS        TEN THERMALLY              THIRTY CONCRETE              STEEL LINER                                  HORIZONTAL ACTIVATED CONTAINMENT          IN WETWELL                                    VENTS FLOODER VALVES              11.55m2 VESSEL TOTAL B3F CORIUM                          NOTE 5 NOTE 2                      PROTECTION FILL BASEMAT NOTE 1 AREAS:                  NOTES:
A. UPPER DRYWELL        1. CORIUM PROTECTION FILL DEPTH IS 1.5m AND      4. R/B FLOOR DESIGNATIONS, NOT B. WETWELL GAS SPACE      UNOBSTRUCTED SURFACE AREA IS  79m2.              PART OF PCS; SHOWN HERE FOR
: 2. RPV PEDESTAL IN THE LOWER DRYWELL                REFERENCE ONLY.
C. SUPPRESSION POOL REGION HAS 1.64m CONCRETE THICKNESS          5. LINES BETWEEN (a) RCCV WALL D. LOWER DRYWELL          (EXCEPT ADJACENT TO VERTICAL                      AND R/B FLOOR SLABS, (b) RCCV DOWNCOMERS).                                      TOP SLAB AND R/B POOL GATES AND (c) RCCV BASEMAT AND R/B
: 3. PERSONNEL AND EQUIPMENT ACCESS POINTS ARE AT VARIOUS CIRCUMFERENTIAL                    BASEMAT SHOW CODE BOUNDARY LOCATIONS.                                        OF RCCV (ASME CODE SECTION III DIVISION 2 STRUCTURE).
Figure 2.14.1 Primary Containment System 2.14-4                                                                                    Primary Containment System
 
ABWR Primary Containment System Table 2.14.1 Primary Containment System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the PCS is as      1. Inspections of the as-built system will be    1. The as-built PCS conforms with the basic shown on Figure 2.14.1.                          conducted.                                        configuration shown on Figure 2.14.1.
: 2. The primary containment pressure boundary 2. Inspections of ASME Code required                  2. An ASME Code Certified Stress Report defined in Section 2.14.1 is designed to meet documents will be conducted.                          exists for the pressure boundary ASME Code, Section III requirements.                                                                components.
: 3. The ASME Code pressure boundary                3. A structural integrity test (SIT) will be      3. The results of the SIT of the pressure components of the PCS retain their integrity      conducted on the pressure boundary                boundary components conform with the under internal pressures that will be            components of the PCS per ASME Code              requirements of the ASME Code.
experienced during service.                      requirements.
: 4. The maximum calculated pressures and          4. Analyses of the design basis accident will be 4. The maximum calculated pressures and 25A5675AA Revision 7 temperatures for the design basis accident        performed using as-built PCS data.              temperatures are less than design are less than design conditions.                                                                  conditions.
: 5. The primary containment pressure boundary 5. An integrated leak rate test of the primary        5. The primary containment pressure boundary including penetrations and isolation valves    containment will be conducted.                      including penetrations and isolation valves has a leak rate equal to or less than 0.5% per                                                      has a leak rate equal to or less than 0.5% per day (excluding MSIV leakage) of                                                                    day (excluding MSIV leakage) of containment gas mass at the maximum                                                                containment gas mass at the maximum calculated containment pressure for the                                                            calculated containment pressure for the design basis accident.                                                                              design basis accident.
Design Control Document/Tier 1
: 6. The design differential pressure of the        6. An SIT will be conducted of the diaphragm      6. An SIT report exists concluding that the diaphragm floor between the drywell and          floor with the drywell pressure greater than      diaphragm floor is able to withstand the wetwell is 172.6 kPa in the downward              wetwell pressures by 1.0 times the design        design differential pressure.
direction.                                        differential pressure.
: 7. The horizontal vent system consists of 30      7. Inspection of the installed horizontal vent    7. Confirmation that horizontal vent system is vents configured as described in Section          system will be conducted.                        configured as described in Section 2.14.1.
2.14.1.
: 8. MCR displays and alarms provided for the      8. Inspections will be performed on the MCR      8. Displays and alarms exist or can be retrieved PCS are as defined in Section 2.14.1.            displays and alarms for the PCS.                  in the MCR as defined in Section 2.14.1.
2.14-5
 
Table 2.14.1 Primary Containment System (Continued)
ABWR 2.14-6 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 9. The vacuum breaker position switches have 9. Analysis of the as-built vacuum breakers will 9. The vacuum breaker position switches have adequate sensitivity to detect the allowable be performed. These analyses will determine      adequate sensitivity to detect the allowable S/P bypass capability of the containment. the maximum vacuum breaker flow area            S/P bypass capability of the containment.
(drywell-to-wetwell) which could exist undetected by the as-installed position switches. The loss coefficients associated with the flow area will be evaluated on the basis of the drywell-to-wetwell flow path geometric details. The flow area and loss coefficients will be combined into an overall drywell- to- wetwell A K factor which will be compared to the allowable value.
25A5675AA Revision 7
: 10. The water volume in the suppression pool        10. Analyses of the as-built PCS will be        10. The water volume in the suppression pool including the vents is equal to or greater than    performed.                                      including the vents is equal to or greater than 3580 m3.                                                                                            3580 m3.
: 11. The SRVDL quencher arms are located at or 11. Inspection of the installed SRVDL quenchers 11. The SRVDL quenchers are located within the below the elevation of the center layer of      will be conducted.                            suppression pool as described in Section horizontal vents in the suppression pool. The                                                2.14.1.
quenchers are placed in the suppression pool in two radial ring. Eighteen of 10 equally spaced locations in each radial ring have Design Control Document/Tier 1 quenchers installed.
: 12. The corium protection fill contains less than 12. Tests will be performed on corium protection 12. Corium protection fill contains less than 4%
4% of calcium carbonate material by weight.      fill materials to determine the calcium          of calcium carbonate material by weight.
carbonate content in a test facility Primary Containment System
: 13. Lower drywell imbedded sumps are              13. Inspections of the lower drywell sump corium 13. Lower drywell imbedded sumps are protected by corium shields.                      protection shields will be performed.            protected by corium shields.
: 14. The containment internal structures are able 14. A structural analysis will be performed which 14. A structural analysis report exists which to withstand the structural design basis loads  reconciles the as-built data with structural      concludes that the as-built internal structures as defined in Section 2.14.1.                    design as defined in Section 2.14.1.              are able to withstand the design basis loads as defined in Section 2.14.1.
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 2.14.2 Containment Internal Structures No entry. Covered in Section 2.14.1.
Containment Internal Structures                                                            2.14-7
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.14.3 Reactor Pressure Vessel Pedestal No entry. Covered in Section 2.14.1.
2.14-8                                                            Reactor Pressure Vessel Pedestal
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.14.4 Standby Gas Treatment System Design Description The Standby Gas Treatment System (SGTS) is used to filter the gaseous effluent from either the primary or secondary containment. The purpose of the SGTS is to limit the discharge of radioactivity to the environment on receipt of a signal from the Leak Detection System (LDS).
SGTS consists of two redundant divisions. Figure 2.14.4 shows the basic system configuration and scope.
The SGTS is classified as safety-related.
Each division of the SGTS (except cooling fan and associated damper) is automatically initiated by signals from the LDS. Each SGTS division can be manually initiated from Main Control Room (MCR).
The SGTS maintains a negative pressure of 6.4 mm water gauge or greater in the secondary containment relative to the outdoor atmosphere within 20 minutes when the secondary containment is isolated. Each SGTS process fan capacity is at least 6800 m3/h (21&deg;C and 1 atmosphere abs.) with the secondary containment not isolated. The absorber efficiency for removal of all forms of iodine (elemental, organic, particulate, and hydrogen iodide) from the influent stream is at least 99%.
After SGTS initiation, each cooling fan starts automatically when a signal indicates that the process fan in that division is not operating.
The SGTS has four safety-related differential pressure sensors for monitoring secondary containment pressure with respect to ambient pressure outside. One sensor is located on each of the four sides of the Reactor Building.
The SGTS is classified as Seismic Category I.
The SGTS is located in the Reactor Building.
The SGTS Division B is powered from Class 1E Division II, except for the cooling fan and associated damper, which is powered by Class 1E Division III. The SGTS Division C is powered from Class 1E Division III, except for the cooling fan and associated damper, which is powered by Class 1E Division II. Each of the four differential pressure sensors is powered from its respective Class 1E division. In the SGTS, independence is provided between Class 1E divisions and also between the Class 1E divisions and non-Class 1E equipment.
Except for the common connection to the plant stack, each mechanical division of the SGTS (Divisions B and C) is physically separated from the other division.
Standby Gas Treatment System                                                                            2.14-9
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 The SGTS has the following displays and controls in the main control room:
(1)  Parameter displays for the instruments shown on Figure 2.14.4.
(2)  Controls and status indication for the active safety-related components shown on Figure 2.14.4.
(3)  Manual system level initiation capability.
The safety-related electrical equipment is shown on Figure 2.14.4 and located in the Reactor Building is qualified for a harsh environment.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.4 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, that will be undertaken for the SGTS.
2.14-10                                                                        Standby Gas Treatment System
 
ABWR Standby Gas Treatment System FROM PRIMARY CONTAINMENT DIVISION B            PROCESS M          STACK FAN ACS                                                                            SGTS FILTER TRAIN SGTS M                FE FROM COOLING      M SECONDARY                                                                      FAN CONTAINMENT 25A5675AA Revision 7 FROM PRIMARY CONTAINMENT DIVISION C            PROCESS M
ACS                                                      FAN FILTER TRAIN SGTS M                FE FROM SECONDARY                                                                    COOLING      M Design Control Document/Tier 1 CONTAINMENT                                                                    FAN SIDE 1              SIDE 2                  SIDE 3              SIDE 4 REACTOR BUILDING dP                    dP                  dP                  dP 2.14-11 Figure 2.14.4 Standby Gas Treatment System
 
ABWR 2.14-12 Table 2.14.4 Standby Gas Treatment System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                              Acceptance Criteria
: 1. The basic configuration of the SGTS is as        1. Inspections of the as-built system will be          1. The as-built SGTS conforms with the basic shown on Figure 2.14.4.                            conducted.                                            configuration shown on Figure 2.14.4.
: 2. Each division of the SGTS (except cooling  2. Tests will be conducted on each as-built                2. The process fan starts and dampers open to fan and associated damper) is automatically    SGTS division using simulated initiation                    allow process flow.
initiated by signals from the LDS.            signals.
: 3. Each SGTS division can be manually              3. Tests will be conducted by initiating each          3. Each division of the SGTS receives an initiated from the MCR.                            division manually.                                    initiation signal.
: 4.                                                  4.                                                    4.
: a. The SGTS maintains a negative                    a. Tests will be conducted on each as-built            a. The SGTS maintains a negative pressure of 6.35 mm water gauge or                  SGTS division.                                        pressure of 6.35 mm water gauge or 25A5675AA Revision 7 greater in the secondary containment                                                                      greater in the secondary containment relative to the outdoor atmosphere within                                                                  relative to the outdoor atmosphere within 20 minutes when the secondary                                                                              20 minutes when the secondary containment is isolated.                                                                                  containment is isolated.
: b. Each SGTS process fan capacity is at            b. Tests will be conducted on each as-built            b. Each SGTS process fan capacity is at least 6800 m3/h (at 21&deg;C, 1 atmosphere              SGTS division.                                        least 6800 m3/h (at 21&deg;C, 1 atmosphere abs.) with the secondary containment                                                                      abs.) with the secondary containment not isolated.                                                                                              not isolated.
: 5. After SGTS initiation, each cooling fan starts 5. Tests will be conducted on each division              5. The cooling fan starts automatically when a Design Control Document/Tier 1 automatically when a signal indicates that        using signals indicating that the process fan            signal indicates that the process fan is not the process fan in that division is not          is not operating.                                        operating.
operating.
Standby Gas Treatment System
: 6. Each filter train will have at least 99%        6.                                                    6. Each filter train will have at least 99%
removal efficiency for all forms of iodine                                                                removal efficiency for all forms of iodine
: a. Tests will be conducted on each as-built (elemental, organic, particulate and                                                                      (elemental, organic, particulate and filter train.
hydrogen iodide).                                                                                          hydrogen iodide).
: b. Tests in test facility will be conducted the iodine absorbing material.
 
Table 2.14.4 Standby Gas Treatment System (Continued)
ABWR Standby Gas Treatment System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 7. The SGTS Division B is powered from Class 7.                                                7.
1E Division II, except for the cooling fan and
: a. Tests will be performed on the SGTS by      a. The test signal exists only in the Class associated damper, which is powered by providing a test signal in only one Class      1E division under test in the SGTS.
Class 1E Division III. The SGTS Division C is 1E division at a time.
powered from Class 1E Division III, except for the cooling fan and associated damper,    b. Inspections of the as-built Class 1E        b. In the SGTS, physical separation or which is powered by Class 1E Division II.        divisions in the SGTS will be performed.        electrical isolation exists between Class Each of the four differential pressure sensors                                                    1E divisions. Physical separation or is powered from its respective Class 1E                                                          electrical isolation exists between Class division. In the SGTS, independence is                                                            1E divisions and non-Class 1E provided between Class 1E divisions and                                                          equipment.
also between the Class 1E divisions and 25A5675AA Revision 7 non-Class 1E equipment.
: 8. Except for the common connection to the        8. Inspections of the as-built the SGTS will be  8. Each mechanical division of the SGTS is plant stack, each mechanical division of the      performed.                                        physically separated from other mechanical SGTS (Divisions B and C) is physically                                                              division of the SGTS by structure and/or fire separated from the other division.                                                                  barriers.
: 9. MCR displays and controls provided for the    9. Inspections will be performed on the MCR      9. Displays and controls exist or can be SGTS are as defined in Section 2.14.4.            displays and controls for the SGTS.              retrieved in the MCR as defined in Section 2.14.4.
Design Control Document/Tier 1 2.14-13
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.14.5 PCV Pressure and Leak Testing Facility No entry for this system.
2.14-14                                                  PCV Pressure and Leak Testing Facility
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.14.6 Atmospheric Control System Design Description The Atmospheric Control (AC) System consists of a nitrogen supply, injection lines, exhaust lines, bleed line, valves, controls, and instrumentation. The AC System also has the containment overpressure protection system (COPS). Figure 2.14.6 shows the basic system configuration and scope.
The AC System is capable of providing an inert atmosphere within the primary containment.
Except for the primary containment penetrations, isolation valves, and suppression pool level sensors, the AC System is classified as non-safety-related.
The outer rupture disk of the COPS has a rupture differential pressure of less than 0.03 MPa.
The inner rupture disk of the COPS is selected such that the COPS has an actuation pressure of 0.72 MPa (absolute) +/-5%. The COPS has the capacity to allow at least 28 kg/s steam flow when the containment is at the actuation pressure of the system.
The AC System primary containment penetrations, isolation valves, and suppression pool level sensors are classified as Seismic Category I. Figure 2.14.6 shows the ASME Code class for the AC System piping and components.
AC System components are located in the Reactor Building, except for the nitrogen supply.
Figure 2.14.6 shows the Class 1E divisional power assignments for the AC System components. In the AC System, independence is provided between the Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The main control room has control and open/close status indication for the containment isolation valves.
AC System components with display interfaces with the Remote Shutdown System (RSS) are shown on the Figure 2.14.6.
The safety-related electrical equipment located in the Reactor Building is qualified for a harsh environment.
The COPS pneumatic actuated valves shown on Figure 2.14.6 have active safety-related functions to both open and close, and perform these functions against a pressure of 0.72 MPa (absolute) +/-5% and under fluid flow and temperature conditions.
The two valves in the containment overpressure protection system fail open on loss of pneumatic pressure or loss of electrical power to the valve actuating solenoid. The other Atmospheric Control System                                                                              2.14-15
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 pneumatic valves shown on Figure 2.14.6 fail closed on loss of pneumatic pressure or loss of electrical power to the valve actuating solenoids.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.6 provides a definition of the inspections, tests and/or analyses, together with associated criteria, which will be undertaken for the AC System.
2.14-16                                                                          Atmospheric Control System
 
ABWR Atmospheric Control System PLANT STACK    OUTER AC 3    RUPTURE DISK HPIN TO REACTOR BLDG                                            P                                      FROM NNS AC              NITROGEN HVAC EXHAUST SUPPLY HVAC SGTS AC 2      P                                              2 NNS AC 2        P P
P AC SGTS 2
P P            P 25A5675AA Revision 7 DRYWELL FROM 2 NNS INNER      3                                                                                                      NITROGEN RUPTURE                          P                                                                                SUPPLY 2                                                                          P DISK                      BLEED LINE VALVE P                                                            P            P S/P              S/P FROM REACTOR BLDG HVAC L                                          DIVISION I - R        2 HVAC L
AC Design Control Document/Tier 1 P          P      DIV. III L                                      L  DIVISION II - R DIV. IV CONTAINMENT OVERPRESSURE PROTECTION SYSTEM NOTES:
: 1. INBOARD CONTAINMENT ISOLATION VALVES ARE POWERED FROM CLASS 1E DIVISION II OUTBOARD CONTAINMENT ISOLATION VALVES ARE POWERED FROM CLASS 1E DIVISION I EXCEPT AS NOTED WITH  , WHICH IS POWERED FROM CLASS 1E DIVISION III.
: 2. THE COPS FLOW PATH FROM THE PRIMARY CONTAINMENT TO THE PLANT STACK HAS NO VALVES OTHER THAN THOSE SHOWN ON THE FIGURE.
: 3. THE ACS HAS PROVISIONS FOR SUPPLYING NITROGEN TO THE COPS PIPING BETWEEN THE INNER AND OUTER RUPTURE DISKS.
Figure 2.14.6 Atmospheric Control System 2.14-17
 
ABWR 2.14-18 Table 2.14.6 Atmospheric Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the AC System is      1. Inspections of the as-built AC System will be 1. The as-built AC System conforms with the as shown on Figure 2.14.6.                          conducted.                                      basic configuration shown on Figure 2.14.6.
: 2. The ASME Code components of the AC              2. A pressure test will be conducted on those 2. The results of the pressure test of the ASME System retain their pressure boundary              Code components of the AC System              Code components of the AC System integrity under internal pressures that will be    required to be pressure tested by the ASME    conform with the requirements in ASME experienced during service.                        Code.                                        Code Section III.
: 3. The outer rupture disk of the COPS has a        3. Tests will be conducted in a test facility to  3. The outer rupture disk of the COPS has a rupture differential pressure of less than 0.03    determine rupture disk bursts conditions.          rupture differential pressure of less than 0.03 MPa. The inner rupture disk of the COPS is                                                            MPa. The inner rupture disk of the COPS is selected such that the COPS has an                                                                    selected such that the COPS has an actuation pressure of 0.72 MPa (absolute)                                                            actuation pressure of 0.72 MPa (absolute) 25A5675AA Revision 7
                                +/-5%.                                                                                                  +/-5%.
: 4. The COPS has the capacity to allow at least 4. Analyses of the steam flow rate will be        4. The COPS has the capacity to allow at least 28 kg/s steam flow when the containment is    conducted for as-built system. These              28 kg/s steam flow when the containment is at the actuation pressure of the system.      analyses will consider compressible steam          at the actuation pressure of the system.
flow and the as-built system loss coefficients.
: 5. Figure 2.14.6 shows the Class 1E divisional 5.                                              5.
power assignments for the AC System
: a. Tests will be performed in the AC            a. The test signal exists only in the Class components. In the AC System, System by providing a test signal in only      1E division under test in the AC System.
independence is provided between Class 1E Design Control Document/Tier 1 one Class 1E division at a time.
divisions, and between Class 1E divisions                                                      b. In the AC System physical separation or and non-Class 1E equipment.                    b. Inspections of the as-installed Class 1E        electrical isolation exists between Class divisions in the AC System will be              1E divisions. Physical separation or performed.                                      electrical isolation exists between these Atmospheric Control System Class 1E divisions and non-Class 1E equipment.
: 6. Main control room displays and controls      6. Inspections will be performed on the main    6. Displays and controls exist or can be provided for the AC System are as defined in    control room displays and controls for the AC    retrieved in the main control room as defined Section 2.14.6.                                System.                                          in Section 2.14.6.
: 7. RSS displays provided for the AC System          7. Inspections will be performed on the RSS      7. Displays exist on the RSS as defined in are as defined in Section 2.14.6.                  displays for the AC System.                      Section 2.14.6.
 
Table 2.14.6 Atmospheric Control System (Continued)
ABWR Atmospheric Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 8. The COPS pneumatic actuated valves            8. Tests will be conducted in a test facility for  8. Upon receipt of an actuating signal, each shown on Figure 2.14.6 have active safety-      both opening and closing under differential        valve both opens and closes.
related functions to both open and close, and    pressure, fluid flow and temperature perform these functions against a pressure      conditions.
of 0.72 MPa (absolute) +/-5% and under fluid flow and temperature conditions.
: 9. The two valves in the containment          9. Tests will be conducted on the as-built AC        9. The two valves in the containment overpressure protection system fail open on    System pneumatic valves.                              overpressure protection system fail open on loss of pneumatic pressure or loss of                                                                loss of pneumatic pressure or loss of electrical power to the valve actuating                                                              electrical power to the valve actuating solenoid. The other pneumatic valves shown                                                          solenoid. The other pneumatic valves shown on Figure 2.14.6 fail closed on loss of                                                              on Figure 2.14.6 fail closed on loss of 25A5675AA Revision 7 pneumatic pressure or loss of electrical                                                            pneumatic pressure or loss of electrical power to the valve actuating solenoids.                                                              power to the valve actuating solenoids.
Design Control Document/Tier 1 2.14-19
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.14.7 Drywell Cooling System Design Description The Drywell Cooling (DWC) System circulates the drywell atmosphere through coolers, thus maintaining its temperature during plant operation. Figure 2.14.7 shows the basic system configuration and scope.
The DWC System consists of three fan coil units and two chilled water units. Each fan coil unit consists of a cooling coil and a fan. These units are cooled by the Reactor Building Cooling Water (RCW) System. Each chilled water unit consists of a cooling coil only. These units are cooled by the Heating Ventilating and Air Conditioning Normal Cooling (HNCW) System.
The DWC System is classified as non-safety-related.
The DWC System is located inside the drywell.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.7 provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria, which will be undertaken for the DWC System.
2.14-20                                                                                Drywell Cooling System
 
ABWR Drywell Cooling System TO CHILLED                          TO CHILLED                              TO CHILLED WATER UNIT                          WATER UNIT                              WATER UNIT RCW                                    RCW                                    RCW DWC NNS                                DWC NNS                                DWC NNS FROM                                    FROM                                FROM DRYWELL            C/C                                    C/C                                      C/C DRYWELL                            DRYWELL ATMOSPHERE                              ATMOSPHERE                          ATMOSPHERE FAN COIL UNIT                          FAN COIL UNIT                            FAN COIL UNIT 25A5675AA Revision 7 HNCW                                                                            HNCW DWC NNS                                                                        DWC NNS Design Control Document/Tier 1 FROM FAN                              TO                                    FROM FAN                              TO C/C                                                                            C/C COIL UNITS                            DRYWELL                              COIL UNITS                            DRYWELL CHILLED WATER UNIT                                                              CHILLED WATER UNIT 2.14-21 Figure 2.14.7 Drywell Cooling System
 
ABWR 2.14-22 Table 2.14.7 Drywell Cooling System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                    Acceptance Criteria
: 1. The basic configuration of the DWC System  1. Inspections of the as-built system      1. The as-built DWC System conforms with the is as shown on Figure 2.14.7.                  configuration will be conducted.            basic configuration shown in Figure 2.14.7.
25A5675AA Revision 7 Design Control Document/Tier 1 Drywell Cooling System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.14.8 Flammability Control System Design Description The Flammability Control System (FCS) is provided to control the potential buildup of hydrogen and oxygen in the containment from radiolysis of water after a design basis loss-of-coolant accident (LOCA). The system consists of two independent and redundant hydrogen and oxygen recombiners. Cooling water required for operation of the system after a LOCA is taken from the Residual Heat Removal (RHR) System. Figure 2.14.8 shows the basic system configuration and scope.
The FCS is classified as safety-related.
After a LOCA, the system can be manually actuated from the main control room if high oxygen concentrations exist in the primary containment. Each recombiner removes gas from the drywell, recombines the oxygen with hydrogen, and returns the gas mixture, along with the condensate to the wetwell.
The system is classified as Seismic Category I. Figure 2.14.8 shows ASME Code class for the FCS piping and components.
The FCS is located in the Reactor Building.
Each of the two FCS divisions is powered from the respective Class 1E division as shown on Figure 2.14.8. In the FCS, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Each mechanical division of the FCS (Divisions B and C) is physically separated from the other division.
The FCS has the following displays and controls in the main control room:
(1)    Controls and status indication for the valves shown on Figure 2.14.8.
(2)    Controls and status indication for the recombiner unit.
FCS components with display and control interfaces with the Remote Shutdown System (RSS) is shown on Figure 2.14.8.
The safety-related electrical equipment shown on Figure 2.14.8, and included in the recombiner units, is qualified for a harsh environment.
The motor operated valves (MOVs) shown on Figure 2.14.8 and active safety-related MOVs in the recombiners, if any, have active safety-related functions to both open and close, and perform these functions under differential pressure, fluid flow, and temperature conditions.
Flammability Control System                                                                            2.14-23
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 The check valves (CVs) shown on Figure 2.14.8 have active safety-related functions to both open and closer under system pressure, fluid flow, and temperature conditions.
The pneumatic valves shown on Figure 2.14.8 fail to the closed position in the event of loss of pneumatic pressure or loss of electrical power to the valve actuating solenoids.
Inspections, Tests, Analyses and Acceptance CriteriaI Table 2.14.8 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the FCS.
2.14-24                                                                          Flammability Control System
 
ABWR Flammability Control System FCS B PCS                              PCS FCS C 2                                      2 RECOMBINER                                                                                        RECOMBINER UNIT B                                    DRYWELL                                                UNIT C P      M                      RPV                          M      P P      M                                          M      P WETWELL FCS B PCS                              PCS FCS C                                                              25A5675AA Revision 7 R                                2                                      2 2                                                                                                    2 RHR FCS B                                                                                            FCS C RHR Design Control Document/Tier 1 NOTE:
: 1. CLASS 1E ELECTRICAL POWER FOR FCS UNIT B IS SUPPLIED FROM DIVISION II EXCEPT FOR THE PNEUMATIC ISOLATION VALVE DUAL SOLENOIDS, WHICH IS DIVISIONS I AND III. UNIT C IS SUPPLIED FROM DIVISION III EXCEPT FOR THE OUTBOARD PNEUMATIC ISOLATION VALVE DUAL SOLENOIDS, WHICH ARE DIVISION I AND II.
2.14-25 Figure 2.14.8 Flammability Control System
 
ABWR 2.14-26 Table 2.14.8 Flammability Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration for the FCS is as      1. Inspections of the as-built system will be  1. The as-built FCS conforms with the basic shown on Figure 2.14.8.                          conducted.                                      configuration shown on Figure 2.14.8.
: 2. The ASME Code components of the FCS            2. A pressure test will be conducted on those 2. The results of the pressure test of the ASME retain their pressure boundary integrity under    Code components of the FCS required to be    code components of the FCS conform with internal pressures that will be experienced      pressure tested by the ASME code.            the requirements in the ASME Code, Section during service.                                                                                III.
: 3. Each of the two FCS divisions is powered  3.                                                  3.
from the respective Class 1E division as
: a. Tests will be performed in the FCS by              a. The test signal exists only in the Class shown on Figure 2.14.8. In the FCS, providing a test signal in only one Class              1E division under test in the FCS.
independence is provided between Class 1E 1E division at a time.
divisions, and between Class 1E divisions 25A5675AA Revision 7 and non-Class 1E equipment.                  b. Inspection of the as-installed Class 1E            b. Physical separation or electrical isolation divisions in the FCS will be performed.                exists between Class 1E divisions in the FCS. Physical separation or electrical isolation exists between Class 1E divisions and non-Class 1E equipment in the FCS.
: 4. Each mechanical division of the FCS          4. Inspections of the as-built FCS will be      4. Each mechanical division of the FCS is (Divisions B, C) is physically separated from    conducted.                                      physically separated from the other the other divisions.                                                                              mechanical divisions of FCS by structural and/or fire barriers.
Design Control Document/Tier 1
: 5. Main control room displays and controls        5. Inspections will be performed on the main    5. Displays and controls exist or can be provided for the FCS are as defined in            control room displays and controls for the      retrieved in the main control room as defined Section 2.14.8.                                  FCS.                                            in Section 2.14.8.
Flammability Control System
: 6. RSS display and control provided for the      6. Inspections will be performed on the RSS    6. Display and control exists on the RSS as FCS are as defined in Section 2.14.8.            display and control for the FCS.                defined in Section 2.14.8.
: 7. MOVs designated in Section 2.14.8 as          7. Tests of installed valves for both opening and 7. Upon receipt of the actuating signal, each having an active safety-related function open    closing will be conducted under                  MOV both opens and closes, depending on and close under differential pressure and        preoperational differential pressure, fluid      the valves safety function.
fluid flow and temperature conditions.          flow, and temperature conditions.
 
Table 2.14.8 Flammability Control System (Continued)
ABWR Flammability Control System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 8. CVs designated in Section 2.14.8 as having    8. Tests of installed valves for both opening and 8. Based on the direction of the differential an active safety-related function open and        closing will be conducted under                  pressure across the valve, each CV opens or close under system pressure, fluid flow, and      preoperational system pressure, fluid flow,      closes depending upon the valves safety temperature conditions.                          and temperature conditions.                      functions.
: 9. The pneumatic valves shown on Figure          9. Tests will be conducted on the as-built FCS  9. The pneumatic valves shown on Figure 2.14.8 fail close in the event of loss of        pneumatic valves.                                2.14.8 fail close in the event of loss of pneumatic pressure or loss of electrical                                                          pneumatic pressure or loss of electrical power to the valve actuating solenoid.                                                            power to the valve actuating solenoid.
25A5675AA Revision 7 Design Control Document/Tier 1 2.14-27
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.14.9 Suppression Pool Temperature Monitoring System Design Description The Suppression Pool Temperature Monitoring (SPTM) System monitors the suppression pool water temperature and provides signals for initiation of automatic scram on high suppression pool temperature. Figure 2.14.9 shows the SPTM System control interfaces.
The SPTM System is classified as a Class 1E safety-related system and consists of four Class 1E divisions (Division I, II, III, and IV) of temperature sensors and their respective logic processors.
The SPTM System temperature sensors are located in the suppression pool. There are four divisions of temperature sensors in each quadrant of the suppression pool.
In each SPTM System division, the suppression pool average temperature is calculated by corresponding divisional logic processors of Safety System Logic and Control (SSLC) using output signals from SPTM temperature sensors. In each SSLC SPTM division, a suppression pool average temperature trip signal is generated by the logic processor and sent to the Reactor Protection System (RPS) when the calculated divisional average temperature exceeds the high suppression pool average temperature setpoint.
Each of the four SPTM System divisional logic is powered from its respective divisional Class 1E power supply. Independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
The SPTM System temperature sensors are located in the suppression pool; the SPTM System logic processors are located in the Control Building.
The SPTM System has parameter displays for suppression pool temperatures in the main control room (MCR).
The SPTM System provides Division I and II suppression pool temperature displays to the Remote Shutdown System (RSS).
Inspections, Tests, Analyses and Acceptance Criteria Table 2.14.9 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the SPTM System.
2.14-28                                                          Suppression Pool Temperature Monitoring System
 
ABWR Suppression Pool Temperature Monitoring System LOCAL AREA                                    MAIN CONTROL ROOM                        LOCAL AREA Plant Sensors                                                                          Device Actuators (None)
SSLC PROCESSING SPTM SYSTEM LOGIC 25A5675AA Revision 7
                                                                                                              - Suppression Pool Average Temperature Trip SPTM Suppression Pool Temperature                          - Calibration, Self-Diagnostics RPS Suppression Pool Average Temperature Trip Design Control Document/Tier 1 NOTES:
: 1. Diagram represents one of four divisions.
: 2. See Section 3.4, Figure 3.4b for SSLC processing.
Figure 2.14.9 Suppression Pool Temperature Monitoring System Control Interface Diagram 2.14-29
 
ABWR 2.14-30 Table 2.14.9 Suppression Pool Temperature Monitoring System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The equipment comprising the SPTM            1. Inspection of the as-built system will be        1. The as-built SPTM System conforms with the System is defined in Section 2.14.9.            conducted.                                          description in Section 2.14.9.
: 2. In each SPTM System division, the          2. Tests will be conducted in each division of        2. In each SPTM System division, a high suppression pool average temperature is        the SPTM System using simulated                      suppression pool average temperature trip calculated by the divisional SSLC logic        temperature sensor signals.                          signal is generated by the SSLC logic processors using output signals from the                                                            processor and sent to the RPS when the temperature sensors. In each SPTM System                                                            calculated divisional average temperature division, a high suppression pool average                                                            exceeds the high suppression pool average temperature trip signal is generated by the                                                          temperature setpoint.
SSLC logic processor and sent to the RPS when the respective calculated divisional 25A5675AA Revision 7 average temperature exceeds the high suppression pool average temperature setpoint.
: 3. Each of the four SPTM System divisional      3.                                                  3.
logics is powered from its respective
: a. Tests will be performed on the SPTM              a. A test signal exists only in the Class 1E divisional Class 1E power supply. In the System by providing a test signal in only          division under test in the SPTM System.
SPTM System, independence is provided one Class 1E division at a time.
between Class 1E divisions, and between                                                                b. In the SPTM System, physical Suppression Pool Temperature Monitoring System Class 1E divisions and non-Class 1E                b. Inspections of the as-built Class 1E                separation or electrical separation exists equipment.                                            divisions in the SPTM System will be                between Class 1E divisions. Physical Design Control Document/Tier 1 performed.                                          separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 4. MCR displays provided for the SPTM            4. Inspections will be conducted on the MCR        4. Displays exist or can be retrieved in the MCR System are as defined in Section 2.14.9.        displays for the SPTM System.                      as defined in Section 2.14.9.
: 5. RSS displays provided for the SPTM System 5. Inspections will be conducted on the RSS            5. Displays exist on the RSS as defined in are as defined in Section 2.14.9.            displays for the SPTM System.                          Section 2.14.9.
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 2.15.1 Foundation Work No entry. Covered in Section 2.15.10.
Foundation Work                                                                          2.15-1
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.15.2 Turbine Pedestal No entry. Covered in Section 2.15.11.
2.15-2                                                                        Turbine Pedestal
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.15.3 Cranes and Hoists Design Description Cranes and Hoists are used for maintenance and refueling tasks.
During refueling/servicing, the Reactor Building (R/B) crane handles the shield plugs, drywell and reactor vessel heads, and the steam dryer/separators. The minimum crane coverage includes the R/B refueling floor laydown area, and the R/B equipment storage pit. During plant operation, the crane handles new fuel shipping containers and the spent fuel shipping casks. For these activities, the minimum crane coverage includes the R/B equipment hatches, and the spent fuel cask loading and washdown pits.
The upper drywell hoists are used during outages to service valves and equipment inside the upper drywell.
The lower drywell hoists service valves and equipment inside the lower drywell during outages.
The Cranes and Hoists are classified as non-safety-related.
The R/B crane is interlocked to prevent movement of heavy loads over the spent fuel storage portion of the spent fuel storage pool. The hoisting and braking system of the R/B crane are redundant.
The R/B crane has a lifting capacity greater than or equal to the heaviest expected load.
The upper drywell hoists and lower drywell hoists are classified as Seismic Category I.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.15.3 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Cranes and Hoists.
Cranes and Hoists                                                                                        2.15-3
 
ABWR 2.15-4 Table 2.15.3 Cranes and Hoists Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the Cranes and    1. Inspection of the as-built system will be    1. The as-built Cranes and Hoists System Hoists System is described in Section 2.15.3. conducted.                                      conforms with the description in Section 2.15.3
: 2. The R/B crane is interlocked to prevent    2. Tests will be conducted of the as-built R/B    2. The R/B crane interlock prevents the movement of heavy loads over the spent fuel    crane movement using a heavy load.                carrying of a load greater than one fuel storage portion of the spent fuel storage                                                        assembly and its associated handling pool.                                                                                            devices over the spent fuel storage portion of the spent fuel storage pool.
: 3. The R/B crane has a lifting capacity greater  3. Analyses will be performed to determine the 3. The rated load for the as-built R/B crane than or equal to the heaviest expected load.      heaviest expected load. Load tests of the as-  equals or exceeds the heaviest expected built R/B crane will be conducted.            load. The R/B crane carries:
25A5675AA Revision 7
: a. A static load at 125% of rated load.
: b. An operational load at 100% of rated load.
Design Control Document/Tier 1 Cranes and Hoists
 
25A5675AA Revision 7 ABWR                                      Design Control Document/Tier 1 2.15.4 Elevators No entry for this system.
Elevators                                                          2.15-5
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 2.15.5 Heating, Ventilating and Air Conditioning Systems Design Description Control Room Habitability Area HVAC System The Control Room Habitability Area (CRHA) Heating, Ventilating and Air Conditioning (HVAC) System provides a controlled environment for personnel comfort and safety, and for the operation of equipment in the main control area envelope (MCAE). The system consists of two (redundant) divisions. Each division consists of an air conditioning unit with two supply fans, two exhaust fans, and an emergency filtration unit with two circulating fans. The emergency filtration unit will have at least 99% removal efficiency for all forms of iodine (elemental, organic, particulate, and hydrogen iodide) from the influent system.
Toxic gas monitors may be required in the outside air intakes of the CRHA HVAC System; these sensors are not in the Certified Design.
Figure 2.15.5a shows the basic configuration and scope for the CRHA HVAC System.
The CRHA HVAC System is classified as safety-related.
The CRHA HVAC System operates in the following modes:
(1)    Normal operating.
(2)    High radiation.
(3)    Outside smoke.
(4)    Smoke removal.
Normal Operating Mode In the normal operating mode, one air conditioning unit, one supply fan, and one exhaust fan operate in each division. The exhaust fan automatically starts when the supply fan is started.
The MCAE is maintained at a minimum pressure of 3.2 mm water gauge above the outside atmosphere.
High Radiation Mode On receipt of a Process Radiation Monitoring (PRM) System signal for high radiation in the outside air intake of the operating division, the normal outside air intake dampers close, the exhaust air dampers close, the exhaust fan stops, the minimum outside air intake dampers open, and one fan of the emergency filtration unit starts.
In the high radiation mode, a positive pressure of at least 3.2 mm water gauge is maintained in the MCAE relative to the outside atmosphere. Each emergency filtration unit treats a 2.15-6                                                            Heating, Ventilating and Air Conditioning Systems
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 mixture of MCAE recirculated air and outside makeup air to maintain the positive pressure with not more than 3400 m3 per hour (@ one atmosphere absolute pressure, 0&deg;C) of outside air.
The redundant division of the CRHA HVAC System starts on a low flow signal from the operating emergency filtration unit. The redundant division is connected to an outside air intake, which is separated from the other intake by a minimum of 50m.
Outside Smoke Mode When smoke detection sensors in the operating outside air intake detect smoke, a signal will initiate MCAE air recirculation by isolating the outside air intake, closing the exhaust damper and stopping the exhaust fan.
Smoke Removal Mode The smoke removal mode is manually initiated by closing the recirculation damper and starting both exhaust fans at high speed in conjunction with a supply fan.
The remaining discussion in this section is not mode-specific and applies (unless stated otherwise) to the entire CRHA HVAC System.
MCAE temperature is maintained between 21&deg;C and 26&deg;C, with a relative humidity between 10% and 60%, except when in the smoke removal mode.
The CRHA HVAC System is classified as Seismic Category I. The CRHA HVAC System is located in the Control Building.
Each of the two CRHA HVAC System divisions, with the exception of the motor operated isolation dampers, is powered from the respective Class 1E division as shown on Figure 2.15.5a. Each pair of motor operated isolation dampers in series is powered from two independent Class 1E divisions (one damper is powered from Class 1E division II and the other damper is from Class 1E division III). In the CRHA HVAC System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Each mechanical division of the CRHA HVAC System (Divisions B and C) is physically separated from the other division, except for the common ducts in the MCAE.
To address the beyond-design-basis event of a large aircraft crash, design features provide mechanical cross connects (HECW-Division A), along with electrical power manual alignments from Division I to MCR HVAC (Div III / C) fans and components.
Fire dampers with fusible links in HVAC duct work close under air flow conditions.
Heating, Ventilating and Air Conditioning Systems                                                          2.15-7
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 The CRHA HVAC System has the following displays and controls in the main control room:
(1)    Controls and status indication for the active safety-related components shown on Figure 2.15.5a.
(2)    Parameter displays for the instruments shown on Figure 2.15.5a, except for the smoke detectors.
Interface Requirements Toxic gas monitors will be located in the outside air intakes of the CRHA HVAC System, if the site is adjacent to toxic gas sources with the potential for releases of significance to plant operating personnel in the MCAE. These monitors should have the following requirements:
(1)    Be located in the outside air intakes of each division of the CRHA HVAC System.
(2)    Be capable of detecting toxic gas concentrations at which personnel protective actions must be initiated.
Control Building Safety-Related Equipment Area HVAC System The Control Building Safety-Related Equipment Area (CBSREA)HVAC System provides a controlled temperature environment for the operation of equipment in the Control Building, excluding the MCAE. The system also limits hydrogen concentration in the battery rooms. The CBSREA HVAC System consists of three independent safety-related divisions, each serving a designated area. Each division consists of an air conditioning unit with two supply fans, and two exhaust fans.
The CBSREA HVAC System also ventilates rooms that contain non-safety-related equipment and provides supplemental cooling in these rooms using non-safety-related fan coil units (FCUs).
The basic system configuration and scope for the CBSREA HVAC System is shown on Figures 2.15.5b, 2.15.5c and 2.15.5d.
The CBSREA HVAC System is classified as safety-related except for the FCUs.
The CBSREA HVAC System operates in the following modes:
(1)    Normal operating mode, including accident conditions.
(2)    Smoke removal mode.
Normal Operating Mode In the normal operating mode, one air conditioning unit, one supply fan, and one exhaust fan of each division operate. The exhaust fan automatically starts when the supply fan is started.
2.15-8                                                          Heating, Ventilating and Air Conditioning Systems
 
25A5675AA Revision 7 ABWR                                                                            Design Control Document/Tier 1 In the areas served by the CBSREA HVAC System, the temperature is maintained below 40&deg;C.
Hydrogen concentration is maintained at less than 2% by volume in the battery rooms.
Smoke Removal Mode The smoke removal mode is manually initiated by closing the recirculation damper, and starting both exhaust fans in conjunction with a supply fan to allow outside air purging of the affected Control Building area. The normal operating mode is used to remove smoke from the battery rooms.
The remaining discussion in this section is not mode-specific and applies (unless stated otherwise) to the entire CBSREA HVAC System.
The CBSREA HVAC System is classified as Seismic Category I, except for the non-safety-related fan coil units. The CBSREA HVAC System is located in the Control Building.
Each of the three CBSREA HVAC System divisions is powered from the respective Class 1E division as shown on Figures 2.15.5b, 2.15.5c and 2.15.5d. In the CBSREA HVAC System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Each mechanical division of the CBSREA HVAC System (Divisions A, B and C) is physically separated from the other divisions. CBSREA HVAC System Division B duct penetrations of Division IV firewalls are provided with fire dampers.
Fire dampers with fusible links in HVAC duct work close under air flow conditions.
The CBSREA HVAC System has the following displays and controls in the main control room:
(1)    Controls and status indication for the active safety-related components shown on Figures 2.15.5b, 2.15.5c and 2.15.5d.
(2)    Parameter displays for the instruments shown on Figures 2.15.5b, 2.15.5c and 2.15.5d.
Reactor Building HVAC System The Reactor Building (R/B) HVAC System provides a controlled environment for the operation of equipment in the Reactor Building.
The Reactor Building HVAC System consists of three independent safety-related divisions.
Each division is composed of the following systems:
(1)    R/B Safety-Related Equipment HVAC System.
(2)    R/B Safety-Related Electrical Equipment HVAC System.
Heating, Ventilating and Air Conditioning Systems                                                          2.15-9
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 (3)  R/B Safety-Related Diesel Generator HVAC System.
The Reactor Building HVAC System includes the following non-safety-related systems:
(1)  R/B Secondary Containment HVAC System.
(2)  R/B Primary Containment Supply/Exhaust System.
(3)  R/B Main Steam Tunnel HVAC System.
(4)  R/B Non-Safety-Related Equipment HVAC System.
(5)  R/B Reactor Internal Pump (RIP) Adjustable Speed Drive (ASD) Control Panel HVAC System R/B Safety-Related Equipment HVAC System The R/B Safety-Related Equipment HVAC System provides cooling of safety-related equipment areas, and consists of independent fan coil units. Figure 2.15.5e shows the basic system configuration and scope.
The R/B Safety-Related Equipment HVAC System is classified as safety-related.
The Residual Heat Removal (RHR) System, High Pressure Core Flooder (HPCF) System and Reactor Core Isolation Cooling (RCIC) System pump room FCUs are automatically initiated upon startup of their respective room process pump. The Containment Atmospheric Monitoring System (CAMS) and Standby Gas Treatment System (SGTS) room FCUs are automatically initiated upon isolation of the Reactor Building Secondary Containment HVAC System.The Flammability Control System (FCS) room FCUs are also initiated upon a manual FCS start signal.
The temperature in the safety-related equipment areas is maintained below 40&deg;C, except for the RHR, HPCF, and RCIC pump rooms, which are maintained below 66&deg;C during pump operation.
The R/B Safety-Related Equipment HVAC System is classified as Seismic Category I. The R/B Safety-Related Equipment HVAC System is located in the Reactor Building.
Each of the three divisions of the R/B Safety-Related Equipment HVAC System is powered from the respective Class 1E division as shown on Figure 2.15.5e. In the R/B Safety-Related Equipment HVAC System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Each mechanical division (Divisions A, B, C) of the R/B Safety-Related Equipment HVAC System is physically separated from the other divisions.
2.15-10                                                        Heating, Ventilating and Air Conditioning Systems
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 The R/B Safety-Related Equipment HVAC System has the following displays and controls in the main control room:
(1)    Controls and status indication for the FCUs shown on Figure 2.15.5e.
The safety-related electrical equipment shown on Figure 2.15.5e located in the Reactor Building is qualified for a harsh environment.
R/B Safety-Related Electrical Equipment HVAC System The R/B Safety-Related Electrical Equipment HVAC System provides cooling of safety-related electrical equipment areas, and consists of three independent divisions. Each division consists of an air conditioning unit with two supply fans, and two exhaust fans. Figures 2.15.5f, 2.15.5g, and 2.15.5h show the basic system configuration and scope.
The R/B Safety-Related Electrical Equipment HVAC System is classified as safety-related.
Normal Operating Mode In the normal operating mode, the air conditioning unit, one supply fan, and one exhaust fan of each division operate. The exhaust fan automatically starts when the supply fan is started.
In the areas served by the R/B Safety-Related Electrical Equipment HVAC System temperature is maintained below 40&deg;C, except in the diesel generator (DG) engine rooms during DG operation.
Smoke Removal Mode The smoke removal mode is manually initiated by closing the recirculation damper, stopping the exhaust fan, opening the exhaust fan bypass damper to allow outside air purging of the affected area, and starting the smoke removal fan in conjunction with the supply fan. The normal operating mode is used to remove smoke from the DG day tank rooms.
The R/B Safety-Related Electrical Equipment HVAC System is classified as Seismic Category I. The R/B Safety-Related Electrical Equipment HVAC System is located in the Reactor Building.
Each of the three divisions of the R/B Safety-Related Electrical Equipment HVAC System is powered from the respective Class 1E division as shown on Figures 2.15.5f, 2.15.5g, and 2.15.5h. In the R/B Safety-Related Electrical Equipment HVAC System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Each mechanical division of the R/B Safety-Related Electrical Equipment HVAC System (Divisions A, B, C) is physically separated from the other divisions.
Heating, Ventilating and Air Conditioning Systems                                                        2.15-11
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 Fire dampers with fusible links in HVAC duct work close under air flow conditions.
The R/B Safety-Related Electrical Equipment HVAC System has the following displays and controls in the main control rooms:
(1)    Controls and status indication for the active safety-related components shown on Figures 2.15.5f, 2.15.5g, and 2.15.5h.
(2)    Parameter displays for the instruments shown on Figures 2.15.5f, 2.15.5g and 2.15.5h.
R/B Safety-Related Diesel Generator HVAC System The R/B Safety-Related DG HVAC System provides ventilation for the DG rooms when the DGs operate, and consists of three independent divisions. Each division consists of a filter unit and two supply fans. Figure 2.15.5i shows the basic system configuration and scope.
The R/B Safety-Related DG HVAC System is classified as safety-related.
On receipt of a DG start signal, both DG supply fans start. When the DG is operating, the R/B Safety-Related DG HVAC System and the R/B Safety-Related Electrical Equipment HVAC System maintain the temperature below 50&deg;C.
The R/B Safety-Related DG HVAC System is classified as Seismic Category I. The R/B Safety-Related DG HVAC System is located in the Reactor Building.
Each of the three divisions of the R/B Safety-Related DG HVAC System is powered from the respective Class 1E division as shown on Figure 2.15.5i. In the R/B Safety-Related DG HVAC System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
Each mechanical division of the R/B Safety-Related DG HVAC System (Divisions A, B, C) is physically separated from the other divisions.
The R/B Safety-Related DG HVAC System has the following displays and controls in the main control room:
(1)    Controls and status indication for the active safety-related components shown on Figure 2.15.5i.
R/B Secondary Containment HVAC System The R/B Secondary Containment HVAC System provides heating and cooling for the secondary containment. Figure 2.15.5j shows the basic system configuration and scope.
Except for the secondary containment isolation dampers, the R/B Secondary Containment HVAC System is classified as non-safety-related.
2.15-12                                                        Heating, Ventilating and Air Conditioning Systems
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 Normal Operating Mode In the normal operating mode, two supply fans and two exhaust fans operate. The supply fans operate only when the exhaust fans are operating.
The R/B Secondary Containment HVAC System maintains a negative pressure in the secondary containment relative to the outside atmosphere.
The R/B Secondary Containment HVAC System isolation dampers are closed upon receipt of an isolation signal from the Leak Detection System (LDS) or a signal indicating loss of secondary containment supply and exhaust fans.
Smoke Removal Mode The smoke removal mode is manually initiated by starting the standby exhaust and supply fans, opening the exhaust filter unit bypass dampers, and partially closing exhaust dampers for divisions not affected by fire.
The R/B Secondary Containment HVAC System penetrations of secondary containment and isolation dampers are classified as Seismic Category I. The R/B Secondary Containment HVAC System is located in the Reactor Building, except for some of the R/B secondary containment HVAC supply and exhaust air components which are located in the Turbine Building.
Each R/B Secondary Containment HVAC System isolation damper requiring electrical power is powered from the Class 1E division, as shown on Figure 2.15.5j. In the R/B Secondary Containment HVAC System, independence is provided between Class 1E divisions, and also between Class 1E divisions and non-Class 1E equipment.
Fire dampers with fusible links in HVAC duct work close under air flow conditions.
The R/B Secondary Containment HVAC System has the following displays and controls in the main control room:
(1)    Control and status indication for the active components shown on Figure 2.15.5j.
(2)    Parameter displays for the instruments shown on Figure 2.15.5j.
The exhaust duct secondary containment isolation dampers are located in the secondary containment and qualified for a harsh environment.
The pneumatically-operated secondary containment isolation dampers, shown on Figure 2.15.5j, fail to the closed position in the event of loss of pneumatic pressure or loss of electrical power to the valve actuating solenoids.
Heating, Ventilating and Air Conditioning Systems                                                            2.15-13
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 R/B Primary Containment Supply/Exhaust System The R/B Primary Containment Supply/Exhaust System removes inert atmosphere and provides air for primary containment prior to personnel entry, and consists of a supply fan, a filter unit, and an exhaust fan as shown on Figure 2.15.5j.
The R/B Primary Containment Supply/Exhaust System is classified as non-safety-related. The R/B Primary Containment Supply/Exhaust System is located in the secondary containment R/B Main Steam Tunnel HVAC System The R/B Main Steam Tunnel HVAC System provides cooling to the main steam tunnel and consists of two FCUs. Each FCU has two fans. The FCUs are started manually.
The R/B Main Steam Tunnel HVAC System is classified as non-safety-related. The R/B Main Steam Tunnel HVAC System is located in the Reactor Building.
R/B Non-Safety-Related Equipment HVAC System The R/B Non-Safety-Related Equipment HVAC System provides cooling to the non-safety-related equipment rooms. There are six air handling units in the system. Each consists of a cooling coil, fan(s), and filter, as required.
The R/B Non-Safety-Related Equipment HVAC System is classified as non-safety-related, and is located in the Reactor Building.
Reactor Internal Pump ASD HVAC System The Reactor Internal Pump ASD HVAC System provides cooling to the RIP ASD power panels. The system consists of a two recirculating air conditioning units with cooling coils and four supply fans.
The RIP ASD HVAC System is classified as non-safety-related, and is located in the Reactor Building.
Turbine Island HVAC System The Turbine Island HVAC System provides heating, cooling, and ventilation for the Turbine Island. The Turbine Island HVAC System consists of the following non-safety-related systems.
(1)  Turbine Building (T/B) HVAC System.
(2)  Electrical Building (E/B) HVAC System.
Turbine Building (T/B) HVAC System The T/B HVAC System provides cooling and ventilation for the Turbine Building. The T/B HVAC System consists of:
(1)  T/B supply system with an air conditioning unit and three supply fans.
2.15-14                                                        Heating, Ventilating and Air Conditioning Systems
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 (2)    T/B exhaust system with three exhaust fans.
(3)    T/B compartment exhaust system with two exhaust fans.
(4)    T/B lube oil area exhaust system with two fans.
(5)    T/B unit coolers and electric unit heaters.
The T/B HVAC System is classified as non-safety-related. The T/B HVAC System is located in the Turbine Building.
Electrical Building (E/B) HVAC System The E/B HVAC System provides cooling and ventilation for the electrical equipment rooms.
The system consists of two air conditioning units, supply fans, two exhaust fans, unit coolers and electric unit heaters.
The E/B HVAC System is classified as non-safety-related. The E/B HVAC System is located in the Electrical Building of the Turbine Island.
Radwaste Building HVAC System The Radwaste Building HVAC System provides a controlled environment for personnel comfort and safety for the Radwaste Building areas. The system consists of:
(1)    An air conditioning unit and two supply fans for the Radwaste Building control room (2)    An air conditioning unit with, two supply fans, and three exhaust fans for the process areas of the Radwaste Building.
The Radwaste Building HVAC System is classified as non-safety-related, and is located in the Radwaste Building.
Service Building HVAC System The Service Building (S/B) HVAC System provides controlled environment for personnel comfort in the S/B.
The S/B HVAC System consists of two non-safety-related systems:
(1)    Clean Area HVAC System.
(2)    Controlled Area HVAC System.
The S/B HVAC System is classified as non-safety-related, and is located in the Service Building.
Heating, Ventilating and Air Conditioning Systems                                                            2.15-15
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 Clean Area HVAC System The Clean Area HVAC System provides a controlled environment for personnel comfort and safety in the Clean Area for the duration of a design basis accident. The system consists of an air conditioning unit with two supply fans, two exhaust fans, and an emergency filtration unit with two circulating fans. The emergency filtration unit has at least 95% removal efficiency for all forms of iodine (elemental, organic, particulate, and hydrogen iodide) from the influent system.
Toxic gas monitors may be required in the outside air intake of the Clean Area HVAC System; these sensors are not in the Certified Design.
The Clean Area HVAC System is classified as non-safety-related. The Clean Area HVAC System is located in the S/B. The Clean Area HVAC System of the S/B serves the Technical Support Center (TSC) the Operational Support Center (OSC) and other clean areas inside the S/B.
On receipt of a signal from the TSC or main control room (MCR), the normal air intake damper closes, the minimum outside air intake damper opens and the ventilation air for the Clean Area is routed through the emergency filtration unit.
In the high radiation mode, a positive pressure is maintained in the Clean Area relative to the outside atmosphere.
Interface Requirements Toxic gas monitors will be located in the outside air intakes of the Clean Area HVAC System, if the site is adjacent to toxic gas sources with the potential for releases of significance to plant operating personnel in the Clean Area. These monitors shall have the following requirements:
(1)    Be located in the outside air intake of the Clean Area HVAC System.
(2)    Be capable of detecting toxic gas concentrations at which personnel protective actions must be initiated.
Controlled Area HVAC System The Controlled Area HVAC System serves the controlled access area, excluding the clean areas, and it consists of two exhaust fans. The Controlled Area HVAC System obtains its supply air from the Clean Area HVAC System. The Controlled Area HVAC System is located in the Service Building.
Inspections, Tests, Analyses and Acceptance Criteria For portions of the CRHA HVAC system within the Certified Design, Table 2.15.5a provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the CRHA HVAC Systems.
2.15-16                                                            Heating, Ventilating and Air Conditioning Systems
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Table 2.15.5b provides a definition of the inspections, tests and/or analyses, together with associated acceptance criteria which will be under taken for the Control Building Safety-Related Equipment Area HVAC System.
Table 2.15.5c provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Building Safety-Related Equipment HVAC System.
Table 2.15.5d provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Building Safety-Related Electrical Equipment HVAC System.
Table 2.15.5e provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Building Safety-Related DG HVAC System.
Table 2.15.5f provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Building Secondary Containment HVAC System.
Table 2.15.5g provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Building Primary Containment Supply/Exhaust System.
Table 2.15.5h provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Building Main Steam Tunnel HVAC System.
Table 2.15.5i provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Building Non-Safety-Related Equipment HVAC System.
Table 2.15.5j provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Reactor Internal Pump ASD HVAC System.
Table 2.15.5k provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Turbine Island HVAC System.
Table 2.15.5l provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Radwaste Building HVAC System.
Heating, Ventilating and Air Conditioning Systems                                                        2.15-17
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 Table 2.15.5m provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Service Building HVAC System.
2.15-18                                                      Heating, Ventilating and Air Conditioning Systems
 
ABWR Heating, Ventilating and Air Conditioning Systems TORNADO OUTSIDE                                                              MISSILE C/B                                                                  BARRIER (3)
TORNADO MISSILE                                                            dP BARRIER (3)
CIRCULATING FANS M                          M              EMERGENCY FILTRATION                                                      TD          TD UNIT M                            M                    F FE TD      SD MCAE C/B C/B MCAE MCAE SUPPLY MCAE RETURN RECIRCULATION M      ME      T DAMPER M        M 25A5675AA Revision 7 MCAE RETURN C/B MCAE (5)
MCAE EXHAUST EXHAUST FANS MCAE SUPPLY        MCAE C/B HECW      (4)
M    M (5)
Design Control Document/Tier 1 (4)
HECW NOTES:                                          AIR CONDITIONING UNIT SUPPLY FANS
: 1. THIS FIGURE SHOWS ONE OF TWO DIVISIONS. ELECTRICAL POWER                  3. TORNADO MISSILE BARRIERS ALSO RESIST DESIGN BASIS HURRICANE MISSILES.
LOADS FOR THE COMPONENTS OF DIVISION B EXCEPT MOTOR OPERATED ISOLATION DAMPERS ARE POWERED FROM CLASS 1E DIVISION II. ELECTRICAL      4. HECW "A" IS DESIGNED TO BE CROSS CONNECTED TO SUPPORT MCR AREA POWER LOADS FOR THE COMPONENTS OF DIVISION C EXCEPT MOTOR                    HVAC "C" MANUAL MODE OPERATION UNDER POSTULATED AIRCRAFT IMPACT SCENARIOS.
OPERATED ISOLATION DAMPERS ARE POWERED FROM CLASS 1E DIVISION III.        5. MCR AREA HVAC "C" IS DESIGNED TO BE POWERED FROM DIVISION I UNDER MANUAL
: 2. EACH PAIR OF MOTOR OPERATED ISOLATION DAMPERS HAS ONE DAMPER                MODE OPERATION FOR POSULATED AIRCRAFT IMPACT SCENARIOS.
POWERED FROM CLASS 1E DIVISION II AND THE OTHER DAMPER FROM CLASS 1E DIVISION III 2.15-19 Figure 2.15.5a Control Room Habitability Area HVAC System
 
ABWR 2.15-20 M RECIRCULATION DAMPER TORNADO MISSILE BARRIER (3)
TD HECW T
T HVAC EQUIP DIV A RCW PUMP AND HX SUPPLY FANS              DIV A HECW 25A5675AA Revision 7 AIR CONDITIONING UNIT                          ELEC EQUIP DIV I HECW CHILLER DIV A ELEC EQUIP NON-DIV FCU                      TD Heating, Ventilating and Air Conditioning Systems BATTERY DIV I Design Control Document/Tier 1 NOTES:                                                            BATTERY NON-DIV
: 1. CLASS 1E ELECTRICAL SUPPLY AND EXHAUST FANS SHOWN ARE POWERED FROM CLASS 1E DIVISION I.
EXHAUST FANS  TORNADO MISSIL E
: 2. FCU COOLING WATER SUPPLIED BY THE HNCW SYSTEM.                                              BARRIER (3)
: 3. TORNADO MISSILE BARRIERS ALSO RESIST DESIGN BASIS HURRICANE MISSILES.
Figure 2.15.5b Control Building Safety-Related Equipment Area HVAC System (Division A)
 
ABWR Heating, Ventilating and Air Conditioning Systems RECIRCULATION M DAMPER TORNADO MISSILE BARRIER (3)
TD HECW T
T HVAC EQUIP DIV B RCW PUMP AND HX SUPPLY FANS              DIV B HECW 25A5675AA Revision 7 AIR CONDITIONING UNIT                          ELEC EQUIP DIV II ELEC EQUIP DIV IV HECW CHILLER DIV B TD BATTERY DIV II Design Control Document/Tier 1 NOTES:
BATTERY
: 1. CLASS 1E ELECTRICAL LOADS SHOWN ARE POWERED                      DIV IV FROM CLASS 1E DIVISION II.
: 2. DIVISION B DUCT PENETRATIONS OF DIVISION IV FIREWALLS ARE PROVIDED WITH FIRE DAMPERS.                                  EXHAUST FANS  TORNADO MISSIL E BARRIER (3)
: 3. TORNADO MISSILE BARRIERS ALSO RESIST DESIGN BASIS HURRICANE MISSILES.
2.15-21 Figure 2.15.5c Control Building Safety-Related Equipment Area HVAC System (Division B)
 
ABWR 2.15-22 RECIRCULATION M DAMPER TORNADO MISSILE BARRIER (3)
TD HECW T
HVAC EQUIP DIV C RCW PUMP AND HX SUPPLY FANS                  DIV C HECW 25A5675AA Revision 7 AIR CONDITIONING UNIT                              ELEC EQUIP DIV III HECW CHILLER DIV C ELEC EQUIP NON-DIV FCU TD Heating, Ventilating and Air Conditioning Systems BATTERY DIV III Design Control Document/Tier 1 NOTES:
: 1. CLASS 1E ELECTRICAL LOADS SHOWN ARE POWERED FROM CLASS 1E DIVISION III.
: 2. FCU COOLING WATER SUPPLIED BY THE HNCW SYSTEM.                                                                EXHAUST FANS  TORNADO MISSIL E BARRIER (3)
: 3. TORNADO MISSILE BARRIERS ALSO RESIST DESIGN BASIS HURRICANE MISSILES.
Figure 2.15.5d Control Building Safety-Related Equipment Area HVAC System (Division C)
 
ABWR Heating, Ventilating and Air Conditioning Systems DIVISION A                  DIVISION B                DIVISION C RHR-A                      RHR-B                      RHR-C FCU                        FCU                        FCU RCIC-A                      HPCF-B                    HPCF-C FCU                        FCU                        FCU CAMS-A                      CAMS-B                      FCS-C FCU                        FCU                        FCU SGTS-B 25A5675AA Revision 7 SGTS-C FCU FCU FCS-B FCU NOTES:
Design Control Document/Tier 1
: 1. FCU COOLING WATER IS SUPPLIED BY THE RCW SYSTEM.
: 2. NORMAL VENTILATION AND SMOKE REMOVAL IS PROVIDED BY THE R/B SECONDARY CONTAINMENT HVAC SYSTEM.
: 3. ELECTRICAL POWER LOADS FROM DIVISIONS A, B, AND C ARE POWERED FROM CLASS 1E DIVISIONS I, II, AND III, RESPECTIVELY.
Figure 2.15.5e Reactor Building Safety-Related Equipment HVAC System 2.15-23
 
ABWR 2.15-24 M DAMPER                                                    M DAMPER TORNADO MISSILE BARRIER (2)
TD FMCRD PANEL HECW                                              ROOM A      SMOKE REMOVAL T                          FAN RSS PANEL ROOM A ELEC EQUIP ROOM SUPPLY FANS                    DIV I 25A5675AA Revision 7 HECW AIR CONDITIONING UNIT                              HVAC EQUIP ROOMS DIV A DG-A MCC AREA DG-A                            TD ENGINE Heating, Ventilating and Air Conditioning Systems ROOM DAY TANK Design Control Document/Tier 1 ROOM NOTES:
: 1. CLASS 1E ELECTRICAL LOADS SHOWN                                                EXHAUST FANS      TORNADO MISSIL E ARE POWERED BY DIVISION I.                                                                        BARRIER (2)
: 2. TORNADO MISSILE BARRIERS ALSO RESIST DESIGN BASIS HURRICANE MISSILES.
Figure 2.15.5f Reactor Building Safety-Related Electrical Equipment HVAC System (Division A)
 
ABWR Heating, Ventilating and Air Conditioning Systems DAMPER M                                                            M DAMPER TORNADO MISSILE BARRIER (2)
TD HECW                                            FMCRD PANEL    SMOKE ROOM B      REMOVAL T                          FAN ELEC EQUIP ROOM DIV IV RSS PANEL SUPPLY FANS                  ROOM B HECW AIR CONDITIONING                                ELEC EQUIP UNIT                                        ROOM DIV II 25A5675AA Revision 7 RIP ASD RM B HVAC EQUIP ROOMS DIV B DG-B MCC AREA DG-B                            TD Design Control Document/Tier 1 ENGINE ROOM DAY TANK NOTES:                                                              ROOM
: 1. CLASS 1E ELECTRICAL LOADS SHOWN ARE POWERED BY DIVISION II.
: 2. TORNADO MISSILE BARRIERS ALSO RESIST DESIGN BASIS HURRICANE MISSILES.          EXHAUST FANS      TORNADO MISSIL E BARRIER (2) 2.15-25 Figure 2.15.5g Reactor Building Safety-Related Electrical Equipment HVAC System (Division B)
 
ABWR 2.15-26 M  DAMPER                                                  M  DAMPER TORNADO MISSILE BARRIER (2)
TD HECW                                                          SMOKE REMOVAL T                          FAN ELEC EQUIP ROOM DIV III RIP ASD SUPPLY FANS                PANEL RM A 25A5675AA Revision 7 HECW AIR CONDITIONING UNIT                              HVAC EQUIP ROOMS DIV C DG-C MCC AREA DG-C                              TD ENGINE Heating, Ventilating and Air Conditioning Systems ROOM DAY TANK Design Control Document/Tier 1 ROOM NOTES:
: 1. CLASS 1E ELECTRICAL LOADS SHOWN                                                EXHAUST FANS        TORNADO MISSIL E ARE POWERED BY DIVISION III.                                                                          BARRIER (2)
: 2. TORNADO MISSILE BARRIERS ALSO RESIST DESIGN BASIS HURRICANE MISSILES.
Figure 2.15.5h Reactor Building Safety-Related Electrical Equipment HVAC System (Division C)
 
ABWR Heating, Ventilating and Air Conditioning Systems TORNADO MISSILE TORNADO MISSILE BARRIER (2)
BARRIER (2)
DIESEL    EXHAUST AIR 25A5675AA Revision 7 GENERATOR F
ROOM TD TD                  FILTER            SUPPLY UNIT              FANS NOTES:
: 1. THIS FIGURE SHOWS ONE OF THREE IDENTICAL DIVISIONS.
Design Control Document/Tier 1 ELECTRICAL POWER LOADS FOR DIVISIONS A, B, AND C ARE POWERED FROM CLASS 1E DIVISIONS I, II, AND III, RESPECTIVELY.
: 2. TORNADO MISSILE BARRIERS ALSO RESIST DESIGN BASIS HURRICANE MISSILES.
2.15-27 Figure 2.15.5i Reactor Building Safety-Related Diesel Generator HVAC System
 
ABWR 2.15-28 OUTSIDE R/B dP R/B SECONDARY CONTAINMENT                                    M T/B M
M T
T            HNCW FE    NOTE 1 NOTE 1 TD P        P                            HVAC ACS F
F PRIMARY CONTAINMENT                          DIVISION I SUPPLY FAN                            AREA HNCW 25A5675AA Revision 7 SUPPLY FANS                                                    PRIMARY CONTAINMENT EXHAUST FAN DIVISION II AREA STACK                                                                HVAC ACS HVAC                        F                                                                                              DIVISION III AREA M
FE Heating, Ventilating and Air Conditioning Systems F
Design Control Document/Tier 1 M
NOTE 1  NOTE 1          M      M P      P                                  M  M F
M    M EXHAUST FANS  M NOTES:
: 1. THE OUTBOARD ISOLATION DAMPER                                    R/B    SECONDARY CONTAINMENT SOLENOID VALVES ARE POWERED BY CLASS 1E DIVISION I. THE INBOARD                    T/B ISOLATION DAMPER SOLENOID VALVES ARE POWERED BY CLASS 1E DIVISION II.
Figure 2.15.5j Reactor Building Secondary Containment HVAC System
 
ABWR Heating, Ventilating and Air Conditioning Systems Table 2.15.5a          Control Room Habitability Area HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the CRHA HVAC          1. Inspections of the as-built system will be    1. The as-built CRHA HVAC System conforms System is as shown on Figure 2.15.5a.                conducted.                                      with the basic configuration shown on Figure 2.15.5a.
: 2. The emergency filtration unit have at least    2.                                                  2. The emergency filtration unit iodine removal 99% removal efficiency for all forms of iodine    a. Test will be conducted on each as-built          efficiency is at least 99%.
(elemental organic, particulate, and                emergency filtration unit.
hydrogen iodide).
: b. Tests in a test facility will be conducted on the iodine absorber material.
: 3. The exhaust fan automatically starts when          3. Tests will be conducted on each division of  3. The exhaust fan automatically starts when the supply fan is started.                            the CRHA HVAC System by starting the            the supply fan is started.
25A5675AA Revision 7 supply fan.
: 4. The MCAE is maintained at a minimum                4. Tests will be conducted on the as-built CRHA 4. The MCAE is maintained at a minimum pressure of 3.2 mm water gauge above the              HVAC System in the normal mode of              pressure of 3.2 mm water gauge above the outside atmosphere.                                  operation.                                      outside atmosphere.
: 5.                                                    5.                                              5.
: a. On receipt of a PRM System signal for              a. Tests will be conducted on each CRHA          a. Upon receipt of a simulated initiation high radiation in the outside air intake of          HVAC System division using a simulated          signal the following occurs:
the operating division, the normal                    initiation signal.
(1) Normal outside air intake dampers outside air intake dampers close, the Design Control Document/Tier 1 are closed.
exhaust air dampers close, the exhaust fan stops, the minimum outside air intake                                                              (2) Exhaust air dampers are closed.
dampers open, and one fan of the                                                                        (3) Exhaust fan is stopped.
emergency filtration unit starts.
(4) Minimum outside air intake dampers are opened.
(5) Emergency filtration unit fan is started.
2.15-29
 
Table 2.15.5a      Control Room Habitability Area HVAC System (Continued)
ABWR 2.15-30 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                            Acceptance Criteria
: b. In the high radiation mode, positive          b. Tests will be conducted on each division        b. The MCAE is maintained at a positive pressure of at least 3.2 mm water gauge          of the as-built CRHA HVAC System in                pressure of at least 3.2 mm water gauge is maintained in the MCAE relative to the        the high radiation mode.                            relative to the outside atmosphere with outside atmosphere. Each emergency                                                                  outside makeup air of not more than filtration unit treats a mixture of MCAE                                                            3400 m3/h (@ one atmosphere absolute recirculated air and outside makeup air                                                              pressure, 0&deg;C).
to maintain the positive pressure with not more than 3400 m3/h (@ one atmosphere absolute pressure, 0&deg;C) of outside air.
: c. The redundant division of the CRHA          c. Tests will be conducted on each division      c. The redundant division of the CRHA HVAC System starts on a low flow signal          of the as-built CRHA HVAC System                    HVAC System starts on a low flow signal 25A5675AA Revision 7 from the operating emergency filtration          using simulated low flow signals.                  from the operating emergency filtration unit.                                                                                                unit.
: d. The redundant division of the CRHA            d. Inspections will be conducted on the            d. The CRHA HVAC System outside air HVAC System is connected to an outside          CRHA HVAC System.                                  intakes are at least 50m apart.
air intake which is separated from the other by a minimum of 50m.
: 6. When smoke detection sensors in the            6. Tests will be conducted on each CRHA          6. Upon receipt of a simulated initiation signal Heating, Ventilating and Air Conditioning Systems operating outside air intake detects smoke, a    HVAC System division using a simulated            the following occurs:
signal will initiate MCAE air recirculation by    smoke signal.
Design Control Document/Tier 1
: a. Outside air intake dampers are closed.
isolating the outside air intake, closing the exhaust damper, and stopping the exhaust                                                            b. Exhaust air dampers are closed.
fan.                                                                                                c. Exhaust fan is stopped.
 
Table 2.15.5a      Control Room Habitability Area HVAC System (Continued)
ABWR Heating, Ventilating and Air Conditioning Systems Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 7. Each of the two CRHA System divisions is        7.                                                7.
powered from the respective Class 1E
: a. Tests will be performed on the CRHA            a. The test signal exists only in the Class division as shown on Figure 2.15.5a. In the HVAC System by providing a test signal            1E division under test in the CRHA CRHA HVAC System, independence is in only one Class 1E division at a time.          HVAC System.
provided between Class 1E divisions, and between Class 1E divisions and non-Class            b. Inspection of the as-built Class 1E 1E equipment.                                          divisions in the CRHA HVAC System will          b. In the CRHA HVAC System, physical be performed.                                      separation or electrical isolation exists between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class equipment.
25A5675AA Revision 7
: 8. Each mechanical division of the CRHA            8. Inspections of the as-built CRHA HVAC          8. Each mechanical division of the CRHA HVAC System (Division B and C) is                  System will be performed.                          HVAC System is physically separated from physically separated from the other division,                                                        the other mechanical division of the CRHA except for the common ducts in the MCAE.                                                              HVAC System by structural and/or fire barriers.
: 9. Fire dampers with fusible links in HVAC duct 9. Type tests of fire dampers in a test facility will 9. Fire dampers close under system air flow work close under air flow conditions.          be performed for closure under system air            conditions.
flow conditions.
: 10. Main control room displays and controls        10. Inspections will be performed on the main      10. Displays and controls exist or can be Design Control Document/Tier 1 provided for CRHA HVAC System are as              control room displays and controls for the        retrieved in the main control room as defined defined in Section 2.15.5.                        CRHA HVAC System.                                  in Section 2.15.5.
: 11. Design features provide mechanical cross      11. Inspections of the as-built design features for 11. Design features provide mechanical cross connects (HECW-Division A), along with            mechanical cross connects and electrical            connects (HECW-Division A), along with electrical power manual alignments from          power manual capability will be conducted.          electrical power manual alignments from Division I to MCR HVAC (Div III / C) fans and                                                        Division I to MCR HVAC (Div III / C) fans and components.                                                                                          components.
2.15-31
 
ABWR 2.15-32 Table 2.15.5b Control Building Safety-Related Equipment Area HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the CBSREA        1. Inspections of the as-built system will be      1. The as-built CBSREA HVAC System HVAC System is as shown on Figures              conducted.                                        conforms with the basic configuration shown 2.15.5b, 2.15.5c and 2.15.5d.                                                                      on Figures 2.15.5b, 2.15.5c and 2.15.5d.
: 2. The exhaust fan automatically starts when    2. Tests will be conducted on each division of    2. The exhaust fan automatically starts when the supply fan is started.                      the as-built CBSREA HVAC System by                the supply fan is started.
starting the supply fan.
: 3. Hydrogen concentration is maintained at less 3. Flow tests will be conducted on each battery 3. Hydrogen concentration is maintained at less than 2% by volume in the battery rooms.        room served by the CBSREA HVAC System.          than 2% by volume in the battery rooms.
Hydrogen concentration analyses will be performed for each battery room using measured flow rates and maximum expected 25A5675AA Revision 7 battery hydrogen evolution rates.
: 4. Each of the three CBSREA HVAC System        4.                                                4.
divisions is powered from the respective
: a. Tests will be performed on the CBSREA          a. The test signal exists only in the Class Class 1E division as shown on Figures HVAC System by providing a test signal            1E division under test in the CBSREA 2.15.5b, 2.15.5c, and 2.15.5d. In the in only one Class 1E division at a time.          HVAC System.
CBSREA HVAC System, independence is provided between Class 1E divisions, and          b. Inspection of the as-built Class 1E divisions in the CBSREA HVAC System            b. In the CBSREA HVAC System, physical Heating, Ventilating and Air Conditioning Systems between Class 1E divisions and non-Class 1E equipment.                                        will be performed.                                separation or electrical isolation exists Design Control Document/Tier 1 between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 5. Each mechanical division of the CBSREA        5. Inspections of the as-built CBSREA HVAC      5. Each mechanical division of the CBSREA HVAC System (Divisions A, B and C) is            System will be conducted.                        HVAC System is physically separated from physically separated from the other divisions.                                                    the other mechanical divisions of the CBSREA HVAC System by structural and/or fire barriers.
 
Table 2.15.5b Control Building Safety-Related Equipment Area HVAC System (Continued)
ABWR Heating, Ventilating and Air Conditioning Systems Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 6. Fire dampers with fusible links in HVAC duct 6. Type tests of fire dampers in a test facility will 6. Fire dampers close under system air flow work close under air flow conditions.          be performed for closure under system air            conditions.
flow conditions.
: 7. Main control room displays and controls      7. Inspections will be performed on the main      7. Displays and controls exist or can be provided for CBSREA HVAC System are as          control room displays and controls for the        retrieved in the main control room as defined defined in Section 2.15.5.                      CBSREA HVAC System.                              in Section 2.15.5.
25A5675AA Revision 7 Design Control Document/Tier 1 2.15-33
 
ABWR 2.15-34 Table 2.15.5c Reactor Building Safety-Related Equipment HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the R/B Safety-      1. Inspections of the as-built system will be      1. The as-built R/B Safety-Related Equipment Related Equipment HVAC System is as                conducted.                                          HVAC System conforms with the basic shown on Figure 2.15.5e.                                                                              configuration as shown on Figure 2.15.5e.
: 2. The RHR, HPCF, and RCIC pump room              2. Tests will be conducted on each pump room 2. Each pump room FCU starts when a signal FCUs are automatically initiated upon start-      FCU using simulated signals indicating pump  indicates start-up of their respective room up of their respective room process pumps.        start-up.                                    process pump.
: 3. The CAMS and SGTS room FCUs are              3. Tests will be conducted on each as-built    3. The CAMS and SGTS room FCUs are automatically initiated upon isolation of the    safety-related FCUs using simulated signals    automatically initiated upon isolation of the R/B Secondary Containment HVAC System.          indicative isolation of the R/B Secondary      R/B Secondary Containment HVAC System.
Containment HVAC System.
25A5675AA Revision 7
: 4. The FCS room FCUs are initiated upon a          4. Tests will be conducted on each as-built FCS 4. The FCS room FCU starts upon receipt of a manual FCS start signal.                          room FCU using a simulated initiation signal. signal indicating FCS start.
: 5. Each of the three division of the R/B Safety-  5.                                                  5.
Related Equipment HVAC System is
: a. Tests will be performed on the R/B              a. The test signal exists only in the Class powered from the respective Class 1E Safety-Related Equipment HVAC                      1E division under test in the in the R/B division as shown on Figure 2.15.5e. In the System by providing a test signal in only          Safety-Related Equipment HVAC R/B Safety-Related Equipment HVAC one Class 1E division at a time.                    System.
System, independence is provided between Heating, Ventilating and Air Conditioning Systems Class 1E divisions, and between Class 1E            b. Inspection of the as-built Class 1E              b. In the R/B Safety-Related Equipment divisions and non-Class 1E equipment.                  divisions in the R/B Safety-Related                HVAC System, physical separation or Design Control Document/Tier 1 Equipment HVAC System will be                      electrical isolation exists between Class performed.                                          1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-class 1E equipment.
: 6. Each mechanical division (Divisions A, B, C) 6. Inspections of the as-built R/B Safety-            6. Each mechanical division of the R/B Safety-of the R/B Safety-Related Equipment HVAC        Related Equipment HVAC System will be                  Related Equipment HVAC System is System is physically separated from the        conducted.                                            physically separated from the other other divisions.                                                                                      mechanical divisions of the R/B Safety-Related Equipment HVAC System by structural and/or fire barriers.
 
Table 2.15.5c Reactor Building Safety-Related Equipment HVAC System (Continued)
ABWR Heating, Ventilating and Air Conditioning Systems Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                        Acceptance Criteria
: 7. Main control room displays and controls  7. Inspections will be performed on the main    7. Displays and controls exist or can be provided for the R/B Safety-Related          control room displays and controls for the      retrieved in the main control room as defined Equipment HVAC System are as defined in      R/B Safety-Related Equipment HVAC              in Section 2.15.5.
Section 2.15.5.                              System.
25A5675AA Revision 7 Design Control Document/Tier 1 2.15-35
 
ABWR 2.15-36 Table 2.15.5d Reactor Building Safety-Related Electrical Equipment HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the R/B Safety-  1. Inspections of the as-built system will be      1. The as-built R/B Safety-Related Electrical Related Electrical Equipment HVAC System        conducted.                                        Equipment HVAC System conforms with the is as shown on Figures 2.15.5f, 2.15.5g, and                                                      basic configuration shown on Figures 2.15.5h.                                                                                          2.15.5f, 2.15.5g, and 2.15.5h.
: 2. The exhaust fan automatically starts when    2. Tests will be conducted on each division of    2. The exhaust fan automatically starts when the supply fan is started.                      the as-built R/B Safety-Related Electrical        the supply fan is started.
Equipment HVAC System by starting the supply fan.
: 3. Each of the three division of the R/B Safety- 3.                                                3.
Related Electrical Equipment HVAC System
: a. Tests will be performed on the R/B              a. The test signal exists only in the Class is powered from the respective Class 1E 25A5675AA Revision 7 Safety-Related Electrical Equipment                1E division under test in the R/B Safety-division as shown on Figures 2.15.5f, HVAC System by providing a test signal              Related Electrical Equipment HVAC 2.15.5g, and 2.15.5h. In the R/B safety-in only one Class 1E division at a time.            System.
related electrical equipment HVAC system, independence is provided between Class 1E        b. Inspection of the as-built Class 1E divisions, and between Class 1E divisions          divisions in the R/B Safety-Related              b. In the R/B Safety-Related Electrical and non-Class 1E equipment.                        Electrical Equipment HVAC System will              Equipment HVAC System, physical be performed.                                      separation or electrical isolation exists between Class 1E divisions. Physical Heating, Ventilating and Air Conditioning Systems separation or electrical isolation exists between these Class 1E divisions and Design Control Document/Tier 1 non-Class 1E equipment.
: 4. Each mechanical division of the R/B Safety-  4. Inspections of the as-built R/B Safety-        4. Each mechanical division of the R/B Safety-Related Electrical Equipment HVAC System        Related Electrical Equipment HVAC System          Related Electrical Equipment HVAC System (Divisions A, B, and C) is physically            will be conducted.                                is physically separated from the other separated from the other divisions.                                                                mechanical divisions of the R/B Safety-Related Electrical Equipment HVAC System by structural and/or fire barriers.
: 5. Fire dampers with fusible links in HVAC duct 5. Type tests of fire dampers in a test facility will 5. Fire dampers close under system air flow work close under air flow conditions.          be performed for closure under system air            conditions.
flow conditions.
 
Table 2.15.5d Reactor Building Safety-Related Electrical Equipment HVAC System (Continued)
ABWR Heating, Ventilating and Air Conditioning Systems Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria
: 6. Main control room displays and controls      6. Inspections will be performed on the main    6. Displays and controls exist or can be provided for R/B Safety-Related Electrical      control room displays and controls for the      retrieved in the main control room as defined Equipment HVAC System are as defined in        R/B Safety-Related Electrical Equipment        in Section 2.15.5.
Section 2.15.5.                                HVAC System.
25A5675AA Revision 7 Design Control Document/Tier 1 2.15-37
 
ABWR 2.15-38 Table 2.15.5e Reactor Building Safety-Related Diesel Generator HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the R/B Safety-    1. Inspections of the as-built system will be      1. The as-built R/B Safety-Related DG HVAC Related DG HVAC System is as shown on            conducted.                                        System conforms with the basic Figure 2.15.5i.                                                                                    configuration shown on Figure 2.15.5i.
: 2. On receipt of a DG start signal, both DG      2. Tests will be conducted on each division of    2. On receipt of a DG start signal, both DG supply fans start.                              the as-built R/B Safety-Related DG HVAC            supply fans start.
System using a simulated DG start signal.
: 3. Each of the three divisions of the R/B Safety- 3.                                                3.
Related DG HVAC System is powered from
: a. Tests will be performed on the R/B              a. The test signal exists only in the Class the respective Class 1E division as shown on Safety-related DG HVAC System by                    1E division under test in the R/B Safety-Figure 2.15.5i. In the R/B safety-related DG providing a test signal in only one Class          Related DG HVAC System.
HVAC system, independence is provided 25A5675AA Revision 7 1E division at a time.
between Class 1E divisions, and between Class 1E divisions and non-Class 1E              b. Inspection of the as-built Class 1E              b. In the R/B Safety-Related DG HVAC equipment.                                          divisions in the R/B Safety-Related DG              System, physical separation or electrical HVAC System will be performed.                      isolation exists between Class 1E divisions. Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment Heating, Ventilating and Air Conditioning Systems
: 4. Each mechanical division of the R/B Safety- 4. Inspections of the as-built R/B Safety-          4. Each mechanical division of the R/B Safety-Design Control Document/Tier 1 Related DG HVAC System (Divisions A, B        Related DG HVAC System will be                      Related DG HVAC System is physically and C) is physically separated from the other  conducted.                                          separated from the other mechanical divisions.                                                                                          divisions of the R/B Safety-Related DG HVAC System by structural and/or fire barriers.
: 5. Main control room displays and controls      5. Inspections will be performed on the main      5. Displays and controls exist or can be provided for R/B Safety-Related DG HVAC          control room displays and controls for the        retrieved in the main control room as defined System are as defined in Section 2.15.5.        R/B Safety-Related DG HVAC System.                in Section 2.15.5.
 
ABWR Heating, Ventilating and Air Conditioning Systems Table 2.15.5f Reactor Building Secondary Containment HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the R/B      1. Inspections of the as-built system will be          1. The as-built R/B Secondary Containment Secondary Containment HVAC System is as    conducted.                                            HVAC System conforms with the basic shown on Figure 2.15.5j.                                                                          configuration shown on Figure 2.15.5j.
: 2. The R/B Secondary Containment HVAC          2. Tests will be conducted on the R/B              2. The R/B Secondary Containment HVAC System maintains a negative pressure in the    Secondary Containment HVAC System in              System maintains a negative pressure in the secondary containment relative to the          the normal mode of operation.                      secondary containment relative to the outside atmosphere.                                                                              outside atmosphere.
: 3. The R/B Secondary Containment HVAC              3. Tests will be conducted on the R/B          3. Upon receipt of a simulated signal, isolation System isolation dampers are closed upon          Secondary Containment HVAC System              dampers are automatically closed.
receipt of an isolation signal from the LDS, or    using simulated LDS isolation and loss of signal indicating loss of secondary                secondary containment supply and exhaust 25A5675AA Revision 7 containment supply and exhaust fans.              fans signals.
: 4. The smoke removal mode is manually            4. Tests will be conducted in the smoke        4. On manual initiation of smoke removal mode initiated by starting the standby exhaust and    removal mode.                                  the following occurs:
supply fans, operating the exhaust filter unit
: a. The standby exhaust fan starts.
bypass dampers, and partially closing the exhaust dampers for divisions not affected                                                        b. The standby supply fan starts.
by fire.                                                                                          c. The filter unit bypass damper opens.
: d. The exhaust dampers of divisions not affected by fire partially close to a Design Control Document/Tier 1 predetermined position.
: e. The measured air flow rate and the pressure in the ducts are at least equal to the values of the as-built smoke removal analysis.
2.15-39
 
Table 2.15.5f Reactor Building Secondary Containment HVAC System (Continued)
ABWR 2.15-40 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 5. Each R/B Secondary Containment HVAC          5.                                                  5.
System isolation damper requiring electrical
: a. Tests will be performed on the R/B                a. The test signal exists only in the Class power is powered from the Class 1E division, Secondary Containment HVAC System                    1E division under test in the R/B as shown on Figure 2.15.5j. In the R/B by providing a test signal in only one                Secondary Containment HVAC System.
secondary containment HVAC system, Class 1E division at a time.
independence is provided between Class 1E                                                            b. In the R/B Secondary Containment divisions, and between Class 1E divisions      b. Inspection of the as-built Class 1E                  HVAC System, physical separation or and non-Class 1E equipment.                        divisions in the R/B Secondary                        electrical isolation exists between Class Containment HVAC System will be                      1E divisions. Physical separation or performed.                                            electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
25A5675AA Revision 7
: 6. Fire dampers with fusible links in HVAC duct 6. Type tests of fire dampers in a test facility will 6. Fire dampers close under system air flow work close under air flow conditions.          be performed for closure under system air            conditions.
flow conditions.
: 7. Main control room displays and controls    7. Inspections will be performed on the main  7. Displays and controls exist or can be provided for the R/B Secondary Containment    control room displays and controls for the    retrieved in the main control room as defined HVAC System are as defined in Section        R/B Secondary Containment HVAC System.        in Section 2.15.5.
2.15.5.
Heating, Ventilating and Air Conditioning Systems
: 8. The pneumatically-operated secondary            8. Tests will be conducted on the as-built R/B  8. The secondary containment isolation containment isolation dampers, shown on            Secondary Containment HVAC System                dampers shown on Figure 2.15.5j fail to the Design Control Document/Tier 1 Figure 2.15.5j, fail to the closed position in    pneumatic isolation dampers.                    closed position on loss of pneumatic the event of loss of pneumatic pressure or                                                          pressure or loss of electrical power to the loss of electrical power to the valve actuating                                                    valve actuating solenoids.
solenoids.
 
ABWR Heating, Ventilating and Air Conditioning Systems Table 2.15.5g Reactor Building Containment Supply/Exhaust System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                      Acceptance Criteria
: 1. The basic configuration of the R/B Primary 1. Inspections of the as-built system will be  1. The as-built R/B Primary Containment Containment Supply/Exhaust HVAC System        conducted.                                      Supply/Exhaust HVAC System conforms is as described in Section 2.15.5.                                                            with the basic configuration described in Section 2.15.5.
25A5675AA Revision 7 Design Control Document/Tier 1 2.15-41
 
ABWR 2.15-42 Table 2.15.5h Reactor Building Main Steam Tunnel HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                      Acceptance Criteria
: 1. The basic configuration of the R/B Main  1. Inspections of the as-built system will be  1. The as-built R/B Main Steam Tunnel HVAC Steam Tunnel HVAC System is as described    conducted.                                      System conforms with the basic in Section 2.15.5.                                                                          configuration described in Section 2.15.5.
25A5675AA Revision 7 Heating, Ventilating and Air Conditioning Systems                                                                                                                                              Design Control Document/Tier 1
 
ABWR Heating, Ventilating and Air Conditioning Systems Table 2.15.5i Reactor Building Non-Safety-Related HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                      Acceptance Criteria
: 1. The basic configuration of the R/B Non-      1. Inspections of the as-built system will be  1. The as-built R/B Non-Safety-Related Safety-Related HVAC System is as                conducted.                                      Equipment HVAC System conforms with the described in Section 2.15.5.                                                                    basic configuration described in Section 2.15.5.
25A5675AA Revision 7 Design Control Document/Tier 1 2.15-43
 
ABWR 2.15-44 Table 2.15.5j Reactor Internal Pump ASD Control Panel HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the RIP ASD      1. Inspections of the as-built system will be  1. The as-built RIP ASD HVAC System HVAC System is as described in Section        conducted.                                      conforms with the basic configuration 2.15.5.                                                                                        described in Section 2.15.5.
25A5675AA Revision 7 Heating, Ventilating and Air Conditioning Systems                                                                                                                                            Design Control Document/Tier 1
 
ABWR Heating, Ventilating and Air Conditioning Systems Table 2.15.5k Turbine Island HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                      Acceptance Criteria
: 1. The basic configuration of the Turbine Island 1. Inspections of the as-built system will be  1. The as-built Turbine Island HVAC System HVAC System is as described in Section          conducted.                                      conforms with the basic configuration 2.15.5.                                                                                          described in Section 2.15.5.
25A5675AA Revision 7 Design Control Document/Tier 1 2.15-45
 
ABWR 2.15-46 Table 2.15.5l Radwaste Building HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 1. The basic configuration of the Radwaste      1. Inspections of the as-built system will be  1. The as-built Radwaste Building HVAC Building HVAC System is as described in        conducted.                                      System conforms with the basic Section 2.15.5.                                                                                configuration described in Section 2.15.5.
25A5675AA Revision 7 Heating, Ventilating and Air Conditioning Systems                                                                                                                                                  Design Control Document/Tier 1
 
ABWR Heating, Ventilating and Air Conditioning Systems Table 2.15.5m Service Building HVAC System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the S/B HVAC        1. Inspections of the as-built system will be      1. The as-built S/B HVAC System conforms System is as described in Section 2.15.5.          conducted.                                        with the basic configuration described in Section 2.15.5.
: 2. On receipt of a signal from the TSC or MCR, 2. A test of the Clean Area HVAC System will          2. Upon receipt of a simulated isolation signal, the normal air intake damper closes, the      be performed using a simulated isolation              the normal Clean Area air intake damper minimum outside air in take damper opens,      signal for the intake.                                closes, the minimum outside air intake and the ventilation air for the Clean Area is                                                        damper opens, and ventilation for the Clean routed through the emergency filtration unit.                                                        Area is routed through the emergency filtration unit.
: 3. In the high radiation mode, a positive          3. A test will be conducted of the as-built Clean 3. The Clean Area is maintained at a positive pressure is maintained in the Clean Area          Area HVAC System in the simulated high            pressure relative to the outside atmosphere.
25A5675AA Revision 7 relative to the outside atmosphere.                radiation mode.
: 4. The emergency filtration unit for the Clean    4.                                                  4. The emergency filtration unit efficiency is at Area ventilation air has at least 95% removal                                                        least 95%.
: a. Tests will be conducted on each as-built efficiency for all forms of iodine (elemental,      emergency filtration unit.
organic, particulate, and hydrogen iodide).
: b. Tests in a test facility will be conducted of the iodine absorber material.
Design Control Document/Tier 1 2.15-47
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.15.6 Fire Protection System Design Description The Fire Protection System (FPS) detects, alarms and extinguishes fires. Fire detection and alarm systems are provided in all fire areas. The FPS consists of a motor driven pump, a diesel drive pump, sprinkler systems, standpipes and hose reels, and portable extinguishers. The foam systems are also used for special applications. The basic configuration of the FPS water supply system is shown on Figure 2.15.6. The FPS provides fire protection for the Reactor Building, Control Building, Turbine Building, Radwaste Building, and other plant buildings.
Areas covered by sprinklers or foam systems are also covered by the manual hose system. Areas covered only by manual hoses can be reached from at least two hose stations. A hose reel and fire extinguisher are located no greater than 30.5m from any location within the buildings.
The FPS is classified as non-safety-related. The sprinkler systems and the standpipe systems in the Reactor and Control Buildings and portions of the FPS water supply system identified in Figure 2.15.6 remain functional following a safe shutdown earthquake (SSE). These portions of the water supply are separated from the remainder of the system by valves as shown in Figure 2.15.6.
Fresh water is used for the water supply system. Two sources with a minimum capacity of 1140 m3 for each source are provided. A minimum of 456 m3 is reserved for use by the portion of the suppression system used for the Reactor and Control Buildings. Both the diesel driven pump and motor driven pump independently supply a minimum flow of 1893 liters/min at a pressure greater than 448.2 kPa at the most hydraulically remote hose connection in either the Reactor or Control Building. The two fire water pumps provide 5678 liters/min of flow each at a differential pressure of 863 kPa.
Fire water supply connections to Loops B and C of the Residual Heat Removal System piping are provided from the portion of FPS used for the Reactor and Control Buildings. These connections are part of the AC independent water addition mode of the RHR System for reactor vessel injection, wetwell or drywell spray, or spent fuel pool makeup.
Automatic foam water extinguishing systems are provided for the diesel generator rooms and day tank rooms.
Fire detection and alarm systems are supplied with power from a non-Class 1E uninterruptible power supply.
The FPS has the following displays and alarms in the Main Control Room (MCR):
(1)  Detection system fire alarms.
(2)  Status of FPS pumps.
2.15-48                                                                                Fire Protection System
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 A plant fire hazards analysis considers potential fire hazards and assesses the effects of postulated fires on the ability to shutdown the reactor and to maintain the reactor in a safe, cold shutdown condition. Each postulated fire is documented in a Fire Hazards Report.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.15.6 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Fire Protection System.
Fire Protection System                                                                                      2.15-49
 
ABWR 2.15-50 COMPONENTS WITHIN PHANTOM BOX                                                  TO REMAIN FUNCTIONAL FOLLOWING AN SSE                                          REACTOR AND CONTROL BUILDINGS ALTERNATE SUPPLY 1140 m 3 DIESEL DRIVEN PUMP 25A5675AA Revision 7 JOCKEY PUMP NORMAL SUPPLY Design Control Document/Tier 1 3
1140 m MOTOR DRIVEN PUMP TO Fire Protection System TURBINE, RADWASTE, SERVICE AND OTHER PLANT BUILDINGS POST INDICATOR VALVE OR EQUIVALENT Figure 2.15.6 Fire Protection Water Supply System
 
ABWR Fire Protection System Table 2.15.6 Fire Protection System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration for the FPS is        1. Inspections of the as-built FPS will be        1. The as-built configuration of the FPS is in defined in Section 2.15.6                        conducted.                                        accordance with Section 2.15.6.
: 2. Fire detection and alarm systems are          2. Inspection and testing of the as-built        2. The detectors respond to the simulated fire provided in all fire areas.                      detectors will be performed using simulated      conditions.
fire conditions.
: 3. The FPS for the Reactor and Control          3. Tests will be conducted of the as-built FPS. 3. The FPS for the Reactor and Control Buildings supplies a minimum flow of 1893                                                          Buildings supplies a minimum flow of 1893 liters/min at a pressure greater than 448.2                                                        liters/min at a pressure greater than 448.2 kPa at the most hydraulically remote hose                                                          kPa at the most hydraulically remote hose connection in either the Reactor or Control                                                        connection in either the Reactor or Control Building.                                                                                          Building.
25A5675AA Revision 7
: 4. Automatic foam-water extinguishing systems 4. Inspections of the as-built foam-water            4. The automatic foam-water suppression are provided for the diesel generator and day extinguishing systems will be conducted.            systems are present and initiation logic is tank rooms.                                  The automatic logic will be tested using            actuated under simulated fire conditions.
simulated fire conditions.
: 5. The sprinkler systems and the standpipe      5. Seismic analyses of the as-built FPS will be    5. An analysis report exists which concludes systems in the Reactor and Control Buildings    performed.                                        that as-built sprinkler systems and the and the portions of the FPS water supply                                                          standpipe systems in the Reactor and system identified in Figure 2.15.6 remain                                                          Control Buildings and the portions of the FPS functional following an SSE.                                                                      water supply system identified in Figure Design Control Document/Tier 1 2.15.6 remain functional following an SSE.
: 6. The fire detection and alarm systems are      6. Inspections of the as-built FPS will be        6. The FPS is supplied with power from a non-supplied with power from a non-Class 1E          conducted.                                        Class 1E uninterruptible power supply.
uninterruptible power supply.
: 7. MCR alarms and displays provided for the      7. Inspections will be performed on the MCR      7. Alarms and displays exist or can be retrieved FPS are as defined in Section 2.15.6.            alarms, and displays for the FPS.                in the MCR as defined in Section 2.15.6.
: 8. Two fire water supply system pumps provide 8. Tests will be conducted of the as-built FPS      8. Two fire water supply system pumps provide 5678 liters/min of flow each at a differential pumps in a test facility.                          5678 liters/min of flow each at a differential pressure of 863 kPa.                                                                              pressure of 863 kPa.
2.15-51
 
Table 2.15.6 Fire Protection System (Continued)
ABWR 2.15-52 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Acceptance Criteria
: 9. A plant fire hazards analysis considers      9. Inspections of the Fire Hazards Report will  9. A Fire Hazards Report exists for the as-built potential fire hazards and assesses the          be conducted.                                    plant and concludes that for each postulated effects of postulated fires on the ability to                                                    fire, the plant can be shutdown and shutdown the reactor and to maintain the                                                          maintained in a safe, cold shutdown reactor in a safe, cold shutdown condition.                                                      condition.
Each postulated fire is documented in a Fire Hazards Report.
25A5675AA Revision 7 Design Control Document/Tier 1 Fire Protection System
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.15.7 Floor Leakage Detection System No entry for this system.
Floor Leakage Detection System                                                      2.15-53
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.15.8 Vacuum Sweep System No entry for this system.
2.15-54                                                              Vacuum Sweep System
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 2.15.9 Decontamination System No entry for this system.
Decontamination System                                                              2.15-55
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 2.15.10 Reactor Building Design Description The Reactor Building (R/B) is a structure which houses and provides protection and support for the reactor primary systems, the primary containment and much of the plant safety-related equipment. Figures 2.15.10a through 2.15.10o show the basic configuration and scope of the R/B*.
The R/B is constructed of reinforced concrete and structural steel with a steel frame and reinforced concrete roof. The R/B encloses the primary containment. The R/B slabs and fuel pool girders are integrated with the reinforced concrete containment vessel (RCCV). The R/B slabs are supported by columns, shear walls and beams to carry vertical loads to the basemat and transfer horizontal loads through the RCCV and R/B shear walls to the basemat and R/B foundation. The R/B, together with the RCCV and the reactor pedestal, are supported by a common basemat. Inside the RCCV, the basemat is considered part of the Primary Containment System (PCS); outside the RCCV, the basemat is part of the R/B. The top of the R/B basemat is located 20.2m +/- 0.3m below the finished grade elevation.
The R/B is divided into three separate divisional areas for mechanical and electrical equipment and four divisional areas for instrumentation racks. Inter-divisional boundaries have the following features:
(1)  Inter-divisional walls, floors, doors and penetrations, and penetrations in the external R/B walls to connecting tunnels, which have three-hour fire rating.
(2)  Watertight doors in the basement to prevent flooding in one division from propagating to other divisions.
(3)  Divisional walls in the basement are 0.6 meters thick or greater.
Watertight doors on Emergency Core Cooling System rooms have open/close sensors with status indication and alarms in the main control room.
The R/B flooding that results from component failures in any of the R/B divisions does not prevent safe shutdown of the reactor. The basement floor is the collection location point for floods. The building configuration at this elevation is such that even for a flooding event involving release of either the suppression pool or the condensate storage tank (CST) water into the R/B, no more than one division of safety-related equipment is affected. Except for the basement area, safety-related electrical, instrumentation and control equipment is located at least 20 cm above the floor surface.
* The overall building dimensions provided in Figures 2.15.10a through 2.15.10o are provided for information only and are not intended to be part of the certified ABWR information.
2.15-56                                                                                            Reactor Building
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 The R/B is protected against external flood. The following design features are provided:
(1)  External walls below flood level are equal to or greater than 0.6 meters thick to prevent ground water seepage.
(2)  Penetrations in the external walls below flood level are provided with flood protection features.
(3)  A tunnel connects the Radwaste Building, Turbine Building, Control Building and Reactor Building for the liquid radwaste system piping. The penetrations from the tunnel to the Reactor Building are watertight.
The R/B is protected against the pressurization effects associated with postulated rupture of pipes containing high-energy fluid that occur in subcompartments of the R/B.
There are three divisionally separated tunnels for routing Oil Storage and Transfer (OST)
System piping and cable from the fuel oil storage tanks to the R/B. These tunnels are configured so that any fuel oil leakage does not accumulate at the R/B boundary. Tunnel flooding due to site flood conditions is precluded by protecting the entrances against water entry.
The R/B and oil transfer tunnels are classified as Seismic Category I. They are designed and constructed to accommodate the dynamic and static loading conditions associated with the various loads and load combinations which form the structural design basis. The loads are (as applicable)those associated with:
(1)  Natural phenomenawind, floods, tornados (including tornado missiles), hurricane (including hurricane missiles), earthquakes, rain and snow.
(2)  Internal eventsfloods, pipe breaks and missiles.
(3)  Normal plant operationlive loads, dead loads, temperature effects and building vibration loads.
Systems, structures, and components located in the R/B and classified as safety-related are protected against inter-divisional flooding that results from postulated failures in Seismic Category I or non-nuclear safety (NNS) components located in the R/B or from external flooding events. Each postulated flooding event is documented in a Flood Analysis Report which concludes the reactor can be shutdown safely and maintained in a safe, cold shutdown condition without offsite power.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.15.10 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the R/B.
Reactor Building                                                                                          2.15-57
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10a Reactor Building ArrangementSection A-A 2.15-58                                                                                Reactor Building
 
25A5675AA Revision 7 ABWR                                      Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10b Reactor Building ArrangementSection B-B Reactor Building                                                                                      2.15-59
 
25A5675AA Revision 7 ABWR                                  Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10c Reactor Building Arrangement, Floor B3F with Divisional Boundary for FloodElevation -8200 mm 2.15-60                                                                                    Reactor Building
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10d Reactor Building Arrangement, Floor B3F with Divisional Boundary for FireElevation -8200 mm Reactor Building                                                                                2.15-61
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10e Reactor Building ArrangementElevation -5100 mm 2.15-62                                                                                  Reactor Building
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10f Reactor Building Arrangement, Floor B2FElevation -1700 mm Reactor Building                                                                              2.15-63
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10g Reactor Building ArrangementElevation 1500 mm 2.15-64                                                                                  Reactor Building
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10h Reactor Building Arrangement, Floor B1FElevation 4800 mm Reactor Building                                                                              2.15-65
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10i Reactor Building ArrangementElevation 8500 mm 2.15-66                                                                                  Reactor Building
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10j Reactor Building Arrangement, Floor 1FElevation 12300 mm Reactor Building                                                                              2.15-67
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10k Reactor Building Arrangement, Floor 2FElevation 18100 mm 2.15-68                                                                                  Reactor Building
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10l Reactor Building Arrangement, Floor 3FElevation 23500 mm Reactor Building                                                                              2.15-69
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10m Reactor Building ArrangementElevation 27200 mm 2.15-70                                                                                  Reactor Building
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10n Reactor Building Arrangement, Floor 4FElevation 31700 mm Reactor Building                                                                              2.15-71
 
25A5675AA Revision 7 ABWR                  Security-Sensitive Information    Design  Control Document/Tier 1 Withhold from Public Disclosure under 10 CFR 2.390 Figure 2.15.10o Reactor Building ArrangementElevations 34500 mm and 38200 mm 2.15-72                                                                      Reactor Building
 
ABWR Reactor Building Table 2.15.10 Reactor Building Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the R/B is shown      1. Inspections of the as-built structure will be    1. The as-built R/B conforms with the basic on Figures 2.15.10a through 2.15.10o.              conducted.                                          configuration shown in Figures 2.15.10a through 2.15.10o.
: 2. The top of the R/B basemat is located 20.2m 2. Inspections of the as-built structure will be        2. The top of the R/B basemat is located 20.2m
                      +/-0.3m below the finished grade elevation.      conducted.                                              +/-0.3m below the finished grade elevation.
3a. Inter-divisional walls, floors, doors and      3a. Inspections of the as-installed inter-divisional 3a. The as-installed walls, floors, doors and penetrations, and penetrations in the              boundaries and external wall penetrations to        penetrations that form the inter-divisional external R/B walls to connecting tunnels,          connecting tunnels will be conducted.                boundaries and external wall penetrations to have a three-hour fire rating.                                                                          connecting tunnels have a three-hour fire rating.
25A5675AA Revision 7 3b. Steel Roof Trusses supporting the Reactor      3b. Inspections of the as-built steel trusses will  3b. Steel Roof Trusses supporting the Reactor Building roof (el. 49700mm) are fireproofed        be conducted.                                        Building roof (el. 49700mm) are fireproofed and encased with a 3 hr, 5 psid fire retardant                                                          and encased with a 3 hr, 5 psid fire retardant material/system that will not be dislodged by                                                          material/system that will not be dislodged by the postulated aircraft impact overpressure.                                                            the postulated aircraft impact overpressure.
: 4. The R/B has divisional areas with walls and      4. Inspections of the as-built walls and            4. The as-built R/B has walls and watertight watertight doors are as shown on Figures            watertight doors will be conducted.                doors as shown on Figures 2.15.10a through 2.15.10a through 2.15.10o.                                                                              2.15.10o.
: 5. Main control room displays and alarms            5. Inspections will be performed on the main    5. Displays and alarms exist or can be retrieved Design Control Document/Tier 1 provided for the R/B are as defined in              control room displays and alarms for the R/B. in the main control room as defined in Section 2.15.10.                                                                                    Section 2.15.10.
: 6. A flooding event involving release of either 6. Inspections will be conducted of the                6. Penetrations (except for watertight doors) in the suppression pool or the CST water does      divisional boundaries shown on Figure                  the divisional walls are at least 2.5m above not affect more than one division of safety-    2.15.10c.                                              the floor level of -8200 mm.
related equipment.
: 7. Except for the basement area, safety-related 7. Inspections will be conducted of the as-built        7. Except for the basement area, safety-related electrical, instrumentation, and control        equipment.                                              electrical, instrumentation, and control equipment is located at least 20 cm above                                                              equipment is located at least 20 cm above the floor surface.                                                                                      the floor surface.
2.15-73
 
Table 2.15.10 Reactor Building (Continued)
ABWR 2.15-74 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria
: 8. The R/B is protected against external floods    8. Inspections of the as-built structure will be  8.
by having:                                          conducted.                                          a. External walls below flood level are
: a. External walls below flood level that are                                                              equal to or greater than 0.6m thick to equal to or greater than 0.6m thick to                                                                  prevent ground water seepage.
prevent ground water seepage.
: b. Penetrations in the external walls below
: b. Penetrations in the external walls below                                                                flood level are provided with flood flood level provided with flood protection                                                              protection features.
features.
: c. Penetrations from the tunnel to the
: c. Watertight penetrations to the Reactor                                                                  Reactor Building are watertight.
Building from the tunnel that connects the Radwaste Building, Turbine Building 25A5675AA Revision 7 and Reactor Building for the liquid radwaste system piping.
: 9. There are three divisionally separated          9. Inspections of the as-bulit tunnels will be    9. There are three divisionally separated tunnels for routing OST system piping from          conducted.                                        tunnels for routing OST System piping from the fuel storage tanks to the R/B. These                                                              the fuel storage tanks to the R/B. These tunnels are configured so that any fuel oil                                                            tunnels are configured so that any fuel oil leakage does not accumulate at the R/B                                                                leakage does not accumulate at the R/B boundary. Tunnel flooding due to site flood                                                            boundary. Tunnel flooding due to site flood conditions is precluded by protecting the                                                              conditions is precluded by protecting the Design Control Document/Tier 1 entrances against water entry.                                                                        entrances against water entry.
: 10. The R/B and oil transfer tunnels are able to    10. A structural analysis will be performed which 10. A structural analysis report exists which withstand the structural design basis loads        reconciles the as-built data with structural      concludes that the as-built R/B and oil as defined in Section 2.15.10.                      design basis as defined in Section 2.15.10.      tranfer tunnels are able to withstand the structural design basis loads as defined in Section 2.15.10.
Reactor Building
 
Table 2.15.10 Reactor Building (Continued)
ABWR Reactor Building Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 11. Systems, structures and components located 11. Inspections of the Flood Analysis Report and 11. A Flood Analysis Report exists for the as-in the R/B and classified as safety-related    the as-built flood protection features will be  built R/B and concludes that for each are protected against inter-divisional flooding conducted.                                      postulated flooding event, the reactor can be that results from postulated failures in                                                        shutdown safely and maintained in a safe, Seismic Category I or NNS related                                                              cold shutdown condition without offsite components located in the R/B or from                                                          power. The Flood Analysis Report includes external flooding events. Each postulated                                                      the results of inspections of the as-built flood flooding event is documented in a Flood                                                        protection features.
Analysis Report which concludes the reactor can be shutdown safely and maintained in a safe, cold shutdown condition without offsite power.
25A5675AA Revision 7 Design Control Document/Tier 1 2.15-75
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.15.11 Turbine Building Design Description The Turbine Building (T/B) includes the electrical building and houses the main turbine generator and other power conversion cycle equipment and auxiliaries. The T/B is located adjacent to the safety-related Seismic Category I Control Building. With the exception of instrumentation associated with monitoring of condenser pressure, turbine first-stage pressure, turbine control valve oil pressure and stop valve position, there is no safety-related equipment in the T/B. The electrical building houses various plant support systems and equipment such as non-divisional switchgear and chillers.
A tunnel connects the Radwaste Building, Turbine Building, Control Building and Reactor Building for the liquid radwaste system piping. The penetrations from the tunnel to the Turbine Building are watertight and have a three hour fire rating.
Flood conditions in the T/B, except for the electrical building, are prevented from propagating into the Control Building (C/B) via the Service Building. This is achieved by locating the access from the T/B to the S/B at or above grade level and providing a flood control doorway at the access location.
The T/B is not classified as a Seismic Category I structure. However, the building is designed such that damage to safety-related functions does not occur under seismic loads corresponding to the safe shutdown earthquake (SSE) ground acceleration.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.15.11 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Turbine Building.
2.15-76                                                                                      Turbine Building
 
ABWR Turbine Building Table 2.15.11 Turbine Building Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the T/B is        1. Inspections of the as-built structure will be    1. The as-built T/B conforms with the basic described in Section 2.15.11.                    conducted.                                          configuration described in Section 2.15.11.
: 2. The T/B is designed such that damage to      2. A seismic analysis of the as-built T/B will be  2. A structural analysis report exists which safety-related functions does not occur          performed.                                          concludes that under seismic loads under seismic loads corresponding to the                                                            corresponding to the SSE ground SSE ground acceleration.                                                                            acceleration the as-built T/B does not damage safety-related functions.
25A5675AA Revision 7 Design Control Document/Tier 1 2.15-77
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.15.12 Control Building Design Description The Control Building (C/B) is a structure which houses and provides protection and support for plant control and electrical equipment, batteries, portions of the Reactor Building Cooling Water (RCW) System, and C/B heating, ventilating and air conditioning equipment. The C/B is located between the Reactor and Turbine Buildings. Figures 2.15.12a through 2.15.12h show the basic configuration and scope of the C/B.*
The C/B is constructed of reinforced concrete and structural steel. The C/B is a shear-wall structure (which accommodates seismic loads) consisting of the perimeter walls, the steam-tunnel walls and the connected supporting floors. Columns and walls carry vertical loads to the basemat. The top of the C/B basemat is located 20.2m +/-0.3m below the finished grade elevation.
The C/B, except for the main control area envelope, is divided into three separate divisional areas for mechanical and electrical equipment and four divisional areas for instrumentation and control equipment (including batteries). Interdivisional boundaries have the following features:
(1)  Inter-divisional walls, floors, doors and penetrations , and penetrations in the external C/B walls to connecting tunnels, which have three-hour fire rating.
(2)  Watertight doors to prevent flooding in one division from propagating to other divisions.
(3)  Divisional walls in the basement are 0.6m thick or greater.
The main control area envelope is separated from the rest of the C/B by walls, floors, doors and penetrations which have three-hour fire rating.
Watertight doors between flood divisions have open/close sensors with status indication and alarms in the main control room.
The C/B flooding that results from component failures in any of the C/B divisions does not prevent safe shutdown of the reactor. The basement floor is the collection point for floods.
Except for the basement and main control area envelope, safety-related electrical equipment and instrumentation and control equipment is located at least 20 centimeters above the floor surface. Level sensors are located in the basement area of each of the three mechanical divisions. These sensors send signals to the corresponding divisions of the Reactor Service Water (RSW) System indicating flooding in that division of the C/B. These sensors are located no higher than 1500 mm above the C/B basement floor.
* The overall building dimensions provided in Figures 2.15.12a through 2.15.12h are for information only and are not intended to be part of the certified ABWR information.
2.15-78                                                                                              Control Building
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 The basement area level sensors are powered from their respective divisional Class 1E power supply. Independence is provided between the Class 1E divisions for these sensors and also between the Class 1E divisions and non-Class 1E equipment.
To protect the C/B against an external flood the following design features are provided:
(1)    External walls below flood level are equal to or greater than 0.6m thick to prevent ground water seepage.
(2)    Penetrations in the external walls below flood level are provided with flood protection features.
Within the C/B, the steam tunnel has no penetrations from the steam tunnel into other areas of the C/B. The concrete thickness of the steam tunnel walls, floor and ceiling within the C/B is equal to or greater than 1.6m.
The C/B is classified as Seismic Category I. It is designed and constructed to accommodate the dynamic and static loading conditions associated with the various loads and load combinations which form the structural design basis. The loads are those associated with:
(1)    Natural phenomenawind, floods, tornadoes (including tornado missiles), hurricane (including hurricane missiles), earthquakes, rain and snow.
(2)    Internal eventsfloods, pipe breaks and missiles.
(3)    Normal plant operationlive loads, dead loads and temperature effects.
The steam tunnel is protected against pressurization effects that occur in the steam tunnel as a result of postulated rupture of pipes containing high energy fluid.
Systems, structures and components located in the C/B and classified as safety-related are protected against inter-divisional flooding that results from postulated failures in Seismic Category I or non-nuclear safety (NNS) components located in the C/B or from external flooding events. Each postulated flooding event is documented in a Flood Analysis Report which concludes the reactor can be shutdown safely and maintained in a safe, cold shutdown condition without offsite power.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.15.12 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Control Building.
Control Building                                                                                          2.15-79
 
25A5675AA Revision 7 ABWR                  Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 2.15.12a Control Building Arrangement, Section A-A 2.15-80                                                                      Control Building
 
25A5675AA Revision 7 ABWR                          Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 2.15.12b Control Building Arrangement, Section B-B Control Building                                                                            2.15-81
 
25A5675AA Revision 7 ABWR                  Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 2.15.12c Control Building Arrangement, Floor B4FElevation -8200 mm 2.15-82                                                                      Control Building
 
25A5675AA Revision 7 ABWR                          Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 2.15.12d Control Building Arrangement, Floor B3FElevation -2150 mm Control Building                                                                            2.15-83
 
25A5675AA Revision 7 ABWR                  Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 2.15.12e Control Building Arrangement, Floor B2FElevation 3500 mm 2.15-84                                                                      Control Building
 
25A5675AA Revision 7 ABWR                          Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 2.15.12f Control Building Arrangement, Floor B1FElevation 7900 mm Control Building                                                                            2.15-85
 
25A5675AA Revision 7 ABWR                  Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 2.15.12g Control Building Arrangement, Floor 1FElevation 12300 mm 2.15-86                                                                      Control Building
 
25A5675AA Revision 7 ABWR                          Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 2.15.12h Control Building Arrangement, Floor 2FElevation 17150 mm Control Building                                                                            2.15-87
 
ABWR 2.15-88 Table 2.15.12 Control Building Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the C/B is shown    1. Inspections of the as-built structure will be  1. The as-built C/B conforms with the basic on Figures 2.15.12a through 2.15.12h.            conducted.                                        configuration shown on Figures 2.15.12a through 2.15.12h.
: 2. The top of the C/B basemat is located 20.2m 2. Inspections of the as-built structure will be      2. The top of the C/B basemat is located 20.2m
                      +/-0.3m below the finished grade elevation.      conducted.                                            +/-0.3m below the finished grade elevation.
: 3. Inter-divisional walls, floors, doors and      3. Inspections of the as-installed inter-divisional 3. The as-installed walls, floors, doors and penetrations, and penetrations in the            boundaries and external wall penetrations to        penetrations that form the inter-divisional external C/B walls to connecting tunnels,        connecting tunnels will be conducted.              boundaries, and penetrations in the external have a three-hour fire rating.                                                                        C/B walls to connecting tunnels, have a three-hour fire rating.
25A5675AA Revision 7
: 4. The C/B has divisional areas with walls and    4. Inspections of the as-built walls, and doors    4. The as-built C/B has walls and watertight watertight doors as shown on Figures              will be conducted.                                doors as shown on Figures 2.15.12a through 2.15.12a through 2.15.12h.                                                                          2.15.12h.
: 5. The main control area envelope is separated 5. Inspections of the as-built structure will be      5. The as-built C/B has a main control area from the rest of the C/B by walls, floors,    conducted.                                            envelope separated from the rest of the C/B doors and penetrations which have a three-                                                          by walls, floors, doors and penetrations hour fire rating.                                                                                    which have a three-hour fire rating.
: 6. Main control room displays and alarms          6. Inspections will be performed on the main    6. Displays and alarms exist or can be retrieved provided for the C/B are as defined in            control room displays and alarms for the C/B. in the main control room as defined in Design Control Document/Tier 1 Section 2.15.12.                                                                                  Section 2.15.12.
: 7. Except for the basemat and main control        7. Inspections will be conducted of the as-built  7. Except for the basemat and main control area envelope, safety-related electrical          equipment.                                        area envelope, safety-related electrical equipment and instrumentation, and control                                                          equipment and instrumentation, and control equipment is located at least 20 cm above                                                            equipment is located at least 20 cm above the floor surface.                                                                                  the floor surface.
: 8. Level sensors are located in the basement      8. Inspections of the as-built equipment will be  8. Level sensors are located in the basement Control Building area of each of the three mechanical              conducted.                                        area of each of the three mechanical divisions. These sensors are located no                                                              divisions. These sensors are located no higher than 1500 mm above the C/B                                                                    higher than 1500 mm above the C/B basement floor.                                                                                      basement floor.
 
Table 2.15.12 Control Building (Continued)
ABWR Control Building Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                                Inspections, Tests, Analyses                              Acceptance Criteria
: 9. The basement area level sensors are              9.                                                  9.
powered from their respective divisional
: a. Tests will be conducted on the as-built          a. The test signal exists only in the Class Class 1E power supply. Independence is sensors by providing a test signal in only          1E division under test.
provided between the Class 1E divisions for one Class 1E division at a time.
these sensors and also between the Class                                                                    b. Physical separation or electrical isolation 1E divisions and non-Class 1E equipment.              b. Inspections of the as-installed Class 1E            exists between Class 1E divisions.
divisions will be conducted.                        Physical separation or electrical isolation exists between these Class 1E divisions and non-Class 1E equipment.
: 10. The C/B is protected against external floods    10. Inspections of the as-built structure will be    10. The C/B is protected against external floods by having:                                          conducted.                                          by having:
25A5675AA Revision 7
: a. External walls below flood level equal to                                                              a. External walls below flood level equal to or greater than 0.6m thick to prevent                                                                      or greater than 0.6m thick to prevent ground water seepage.                                                                                      ground water seepage.
: b. Penetrations in the external walls below                                                                b. Penetrations in the external walls below flood level provided with flood protection                                                                flood level provided with flood protection features.                                                                                                  features.
: 11. Within the C/B, the steam tunnel has no      11. Inspections of the as-built structure will be      11. Within the C/B, the steam tunnel has no penetrations from the steam tunnel into other    conducted.                                              penetrations from the steam tunnel into other areas of the C/B.                                                                                        areas of the C/B.
Design Control Document/Tier 1
: 12. The concrete thickness of the steam tunnel 12. Inspections of the as-built structure will be          12. The concrete thickness of the steam tunnel walls, floor and ceiling within the C/B is equal conducted.                                              walls, floor and ceiling within the C/B is equal to or greater than 1.6m.                                                                                  to or greater than 1.6m.
: 13. The C/B is able to withstand the structural      13. A structural analysis will be performed which 13. A structural analysis report exists which design basis loads as defined in Section            reconciles the as-built data with structural      concludes that the as-built C/B is able to 2.15.12.                                            design basis as defined in Section 2.15.12.      withstand the structural design basis loads as defined in Section 2.15.12.
2.15-89
 
Table 2.15.12 Control Building (Continued)
ABWR 2.15-90 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 14. Systems, structures and components located 14. Inspections of the Flood Analysis Report and 14. A Flood Analysis Report exists for the as-in the C/B and classified as safety-related    the as-built flood protection features will be  built C/B and concludes that for each are protected against inter-divisional flooding conducted.                                      postulated flooding event, the reactor can be that results from postulated failures in                                                        shutdown safely and maintained in a safe, Seismic Category I or NNS related                                                              cold shutdown condition without offsite components located in the C/B or from                                                          power. The Flood Analysis Report includes external flooding events. Each postulated                                                      the results of inspections of the as-built flood flooding event is documented in a Flood                                                        protection features.
Analysis Report which concludes the reactor can be shutdown safely and maintained in a safe, cold shutdown condition without offsite power.
25A5675AA Revision 7 Design Control Document/Tier 1 Control Building
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.15.13 Radwaste Building Design Description The Radwaste Building (RW/B) is a structure which houses the solid and liquid radwaste treatment systems. The RW/B is classified as non-safety-related.
Flood conditions in the RW/B are prevented from propagating into the Reactor Building and Turbine Building by providing the penetrations in external walls below flood level with flood protection features.
A tunnel connects the Radwaste Building, Turbine Building, Control Building and Reactor Building for the liquid radwaste system piping. The penetrations from the tunnel to the Radwaste Building are watertight.
The external walls of the RW/B below grade and the basemat are classified as Seismic Category I. The exterior walls above grade, the floor slabs, the interior columns, and the roof are classified as non-seismic.
The external walls of the RW/B below grade and the basemat are designed and constructed to accommodate the dynamic and static loading conditions associated with the various loads and load combinations which form the structural design basis. The loads are those associated with:
(1)  Natural phenomenawind, floods, tornados, hurricanes, earthquakes, rain and snow.
(2)  Internal eventfloods.
(3)  Normal plant operationslive loads, dead loads and temperature effects.
The exterior walls above grade, the floor slabs, the interior columns and the roof are designed such that damage to safety-related functions does not occur under seismic loads corresponding to the safe shutdown earthquake (SSE) ground acceleration.
The tunnel connecting the Radwaste Building, Turbine Building, Control Building and Reactor Building is designed such that damage to penetration seals at the interface with safety-related structures does not occur under seismic loads corresponding to the safe shutdown earthquake (SSE) ground acceleration. Flooding of this tunnel during design basis site flood conditions is prevented.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.15.13 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Radwaste Building.
Radwaste Building                                                                                        2.15-91
 
ABWR 2.15-92 Table 2.15.13 Radwaste Building Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria
: 1. The basic configuration of the RW/B is        1. Inspections of the as-built structure will be  1. The as-built RW/B conforms with the basic described in Section 2.15.13.                    conducted.                                        configuration in Section 2.15.13.
: 2. The external walls of the RW/B below grade    2. A structural analysis will be performed which 2. A structural analysis report exists which and the basemat are able to withstand the        reconciles the as-built data with the structural concludes that the as-built RW/B is able to design basis loadings as defined in Section      design basis as defined in Section 2.15.13.      withstand the structural design basis loads 2.15.13.                                                                                          as defined in Section 2.15.13.
: 3. The exterior walls above grade, the floor    3. A seismic analysis will be performed.            3. A structural analysis report exists which slabs, the interior columns and the roof are                                                        concludes that under seismic loads designed such that damage to safety-related                                                        corresponding to the SSE ground functions does not occur under seismic loads                                                        acceleration, the as-built RW/B does not corresponding to the SSE ground                                                                    damage safety-related functions.
25A5675AA Revision 7 acceleration.
: 4. The tunnel connecting the Radwaste            4. A seismic analysis will be performed.          4. A structural analysis report exists which Building, Turbine Building, Control Building,                                                      concludes that under seismic loads and Reactor Building is designed such that                                                          corresponding to the SSE ground damage to penetration seals at the interface                                                        acceleration, the tunnel does not damage with safety-related structures does not occur                                                      penetration seals at the interface with safety-under seismic loads corresponding to the                                                            related structures.
safe shutdown earthquake (SSE) ground acceleration.
Design Control Document/Tier 1 Radwaste Building
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 2.15.14 Service Building Design Description The Service Building (S/B) is a structure which houses the Technical Support Center, Operational Support Center, and the counting room for analyzing post-accident samples. The S/B is classified as non-safety-related. It is located adjacent to the Control Building.
The S/B is not classified as a Seismic Category I structure.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.15.14 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the Service Building.
Service Building                                                                                        2.15-93
 
ABWR 2.15-94 Table 2.15.14 Service Building Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the S/B is      1. Inspections of the as-built structure will be  1. The as-built S/B conforms with the basic described in Section 2.15.14.                conducted.                                        configuration described in Section 2.15.14.
25A5675AA Revision 7 Design Control Document/Tier 1 Service Building
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 2.15.15 Control Building Annex Design Description The control Building Annex (CBA) is a structure, which houses the RIP MG Sets, the RIP MG Set control panels and the RIP MG Set Air Handling Unit. The CBA is located adjacent to the safety-related Seismic Category I Control Building.
The CBA is not classified as a Seismic Category I structure.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.15.15 provides a definition of the inspections, tests, and/or analyses together with associated acceptance criteria, which will be undertaken for the Control Building Annex.
Control Building Annex                                                                                  2.15-95
 
ABWR 2.15-96 Table 2.15.15 Control Building Annex Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the CBA is      1. Inspections of the as-built structure will be  1. The as-built CBA conforms with the basic described in Section 2.15.15.                conducted.                                        configuration described in Section 2.15.15.
25A5675AA Revision 7 Design Control Document/Tier 1 Control Building Annex
 
25A5675AA Revision 7 ABWR                                                    Design Control Document/Tier 1 2.16.1 Stack No entry for this system.
Stack                                                                            2.16-1
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 2.16.2 Oil Storage and Transfer System Design Description The Oil Storage and Transfer (OST) System consists of three independent Emergency Diesel Generator (DG) fuel oil storage and transfer systems with their respective fuel storage tanks, transfer pumps, day tanks, and instrumentation and controls. Figure 2.16.2 shows the basic system configuration and scope.
The three divisions (Divisions I, II, and III) of the OST System provides fuel oil to their respective divisional DGs.
The OST System is classified as safety-related.
Each DG fuel oil storage tank provides a minimum seven (7) day fuel oil supply with its respective DG supplying its maximum lose-of-coolant accident (LOCA) load demand.
DG fuel oil is transferred automatically from the storage tanks to the day tanks by day tank low level signals. Manual control of DG fuel oil transfer is also provided. DG fuel oil is transferred from the storage tanks to the day tanks at a rate which exceeds the DG consumption rates while supplying their maximum LOCA load demand.
Each DG fuel oil day tank provides a minimum four (4) hour fuel oil supply with its respective DGs supplying its maximum LOCA load demand. Fuel oil is transferred from the day tanks to the engine fuel oil pumps by gravity flow.
The OST System, including the DG fuel oil storage and day tanks, is classified as Seismic Category I. Figure 2.16.2 shows the ASME Code class for the OST System.
Each of the three OST System Class 1E divisions is powered from its respective Class 1E division. In the OST System, independence is provided between Class 1E divisions, and also between the Class 1E divisions and non-Class 1E equipment.
The three DG fuel oil storage tanks are separately located underground outside of the Reactor Building. The oil storage tank external equipment is located above the maximum flood level and protected from missiles generated by the environment.
Within the Reactor Building, each mechanical division of the OST System is physically separated from the other divisions.
The OST System has the following displays and controls in the main control room (MCR):
(1)    Parameter displays for DG fuel oil storage tank levels and day tank levels.
(2)    Controls and status indication for DG fuel oil transfer pumps.
2.16-2                                                                          Oil Storage and Transfer System
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Inspections, Tests, Analyses and Acceptance Criteria Table 2.16.2 provides the definition of the inspection, tests, and/or analyses, together with the associated acceptance criteria, which will be undertaken for the OST System.
Oil Storage and Transfer System                                                                            2.16-3
 
ABWR 2.16-4 REACTOR BUILDING VENT DIESEL VENT FUEL OIL L            DAY TANK FILL                  TRANSFER 25A5675AA Revision 7 GRADE    LINE                  PUMP OIL STORAGE DG AND TRANSFER SYSTEM 3 BURIED TO DIESEL DIESEL ENGINE FUEL FUEL PUMP (GRAVITY STORAGE FLOW)
Design Control Document/Tier 1 TANK L
Oil Storage and Transfer System NOTES:
: 1. FIGURE REPRESENTS ONE OF THREE OIL STORAGE AND TRANSFER SYSTEM DIVISIONS.
: 2. EACH OF THE THREE DIVISIONS IS POWERED FROM ITS RESPECTIVE CLASS 1E DIVISION.
Figure 2.16.2 Oil Storage and Transfer System
 
ABWR Oil Storage and Transfer System Table 2.16.2 Oil Storage and Transfer System Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The basic configuration of the OST System      1. Inspections of the as-built system will be    1. The as-built OST system conforms with is as shown on Figure 2.16.2.                      conducted.                                        basic configuration shown on Figure 2.16.2.
: 2. The ASME Code components of the OST            2. A pressure test will be conducted on those    2. The results of the pressure test of the ASME System retain their pressure boundary              code components of the OST System                Code components of the OST System integrity under internal pressures that will be    required to be pressure tested by the ASME        conform with the requirements in the ASME experienced during service.                        Code.                                            Code, Section III.
: 3. Each DG fuel oil storage tank provides a      3. Analyses for the as-built DG fuel oil storage 3. Each as-built DG fuel oil storage tank minimum seven (7) day fuel oil supply with its    tanks to determine the required fuel oil        provides a minimum seven (7) day fuel oil respective DG supplying its maximum LOCA          storage volume based on DG fuel                  supply with its respective DG supplying its load demand.                                      consumption data and LOCA load demand            maximum LOCA load demand.
will be performed. Inspections of the as-built 25A5675AA Revision 7 DG fuel oil storage tanks to determine usable fuel storage volume will be conducted.
: 4. DG fuel oil is transferred automatically from  4. Tests on the as-built DG fuel oil transfer  4. The as-built DG fuel oil transfer system the storage tanks to the day tanks by day          systems will be conducted by simulating day    operation occurs automatically on the day tank low level signals. Manual control of DG      tank low level signals, and by manual          tank low level signals, and when initiated fuel transfer is also provided.                    control.                                      manually.
: 5. DG fuel oil is transferred automatically from  5. Tests on each division of the as-built DG fuel 5. DG fuel oil is transferred automatically from the storage tanks to the day tanks at a rate      oil transfer systems will be conducted by        the storage tanks to the day tanks at a rate which exceeds the DG consumption rates            transferring fuel oil while the DGs are          which exceeds the DG consumption rates Design Control Document/Tier 1 while supplying their maximum LOCA load            supplying their maximum LOCA loads.              while supplying their maximum LOCA load demand.                                                                                              demand.
: 6. Each DG fuel oil day tank provides a          6. Analyses for the as-built DG oil day tanks to 6. Each DG fuel oil day tanks provides a minimum four (4) hour fuel oil supply with its    determine the required fuel oil volume using    minimum four (4) hour fuel oil supply with its respective DG supplying its maximum LOCA          DG fuel consumption data and LOCA load          respective DG supplying its maximum LOCA load demand.                                      demand will be performed. Inspections of the    load demand.
as-built DG fuel oil day tanks to determine usable fuel storage volume will be conducted.
2.16-5
 
Table 2.16.2 Oil Storage and Transfer System (Continued)
ABWR 2.16-6 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria
: 7. The DG fuel oil storage and day tanks are      7. Seismic analyses on the fuel oil storage and 7. Seismic analyses reports exist and conclude classified as Seismic Category I.                day tanks will be performed.                    that the DG fuel oil and storage and day tanks are able to withstand Seismic loads.
: 8. Each of the three OST System divisions is  8.                                            8.
powered from its respective Class 1E
: a. Tests will be conducted in the as-built    a. The test signal exists only in the Class division. In the OST System, independence OST System by providing a test signal in      1E division under test in the OST is provided between the Class 1E divisions, only one Class 1E division at a time.          System.
and between the Class 1E divisions and non-Class 1E equipment.                            b. Inspections of the as-built Class 1E        b. In the OST System, physical separation divisions in the OST System will be            or electrical isolation exists between conducted.                                    Class 1E divisions. Physical separation or electrical isolation exists between 25A5675AA Revision 7 these Class 1E divisions and non-Class 1E equipment.
: 9. Within the Reactor Building, each              9. Inspections of the as-built OST System will  9. Within the Reactor Building, each mechanical division of the OST System is          be conducted.                                    mechanical division of the OST System is physically separated from the other divisions.                                                    physically separated from the other mechanical divisions by structural and/or fire barriers.
: 10. MCR displays and controls provided for the    10. Inspections will be conducted on the MCR    10. Displays and controls exist or can be OST System are as defined in Section              displays and controls for the OST System.        retrieved in the MCR as defined in Section Design Control Document/Tier 1 2.16.2.                                                                                            2.16.2.
Oil Storage and Transfer System
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 2.17.1 Emergency Response Facilities Design Description The Technical Support Center (TSC) and Operational Support Center (OSC) are the only emergency facilities within the scope of the ABWR Standard Plant.
The purpose of the TSC is to provide management and technical support to personnel in the Main Control Room during emergency conditions. The TCS radiological habitability is comparable to the control room habitability under accident conditions. The TSC is non-safety-related and is not Seismic Category I. The TSC is located in the Service Building and has sufficient space to accommodate at least 25 individuals. The TSC has voice communication equipment for communication with the Main Control Room, Emergency Operations Facility, OSC and NRC Headquarters Operation Center.
The TSC has displays for the plant parameters listed in Table 2.7.1a, Item B, Fixed Position Displays.
The purpose of the OSC is to provide an assembly area separate from the Main Control Room and TSC where licensee operations support personnel can report in an emergency. The OSC is non-safety-related and is not Seismic Category I. The OSC is located in the Service Building and has voice communication equipment for communication with the main control room and the TSC.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.17.1 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be performed for the Emergency Response Facilities.
Emergency Response Facilities                                                                          2.17-1
 
ABWR 2.17-2 Table 2.17.1 Emergency Response Facilities Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The TSC and OSC are located in the Service 1. An inspection will be performed for the            1. The TSC and OSC are in different locations Building.                                    location of the TSC and OSC.                          in the Service Building. The TSC is adjacent to the passage from the Service Building to the Control Building.
: 2. The TSC has sufficient space to                2. An inspection will be performed of the floor  2. The TSC has at least 175 m2 of floor space.
accommodate at least 25 individuals.              space in the TSC.
: 3. The TSC has voice communication                3. A test will be performed of the TSC voice      3. The TSC voice communication with the Main equipment for communication with the Main          communication equipment.                          Control Room, Emergency Operations Control Room, Emergency Operations                                                                  Facility, OSC, and NRC Headquarters Facility, OSC, and NRC Headquarters                                                                  Operation Center is audible and intelligible at Operation Center.                                                                                    each location.
25A5675AA Revision 7
: 4. The TSC has displays for the plant              4. An inspection will be conducted on the        4. Displays exist or can be retrieved in the TSC parameters listed in Table 2.7.1a, Item B,        displays for the TSC.                            for the plant parameters listed in Table Fixed Position Displays.                                                                            2.7.1a, Item B.
: 5. The OSC has voice communication                5. A test will be performed of the OSC voice      5. The OSC voice communication with the Main equipment for communication with the main          communication equipment.                          Control Room and TSC is audible and control room and TSC.                                                                                intelligible at each location.
: 6. The TCS has comparable habitability to the      6. An inspection of the as-built TSC habitability 6. The TSC radiological habitability is control room habitability under accident          system will be performed, including a test of    comparable to the control room habitability Design Control Document/Tier 1 conditions.                                        its capabilities.                                under accident conditions such that doses to an individual do not exceed 5 rem whole body radiation exposure or 30 rem thyroid over the 30-day post-accident period.
Emergency Response Facilities
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 3.0 Additional Certified Design Material This section provides additional certified design material for those aspects of the Certified Design that cannot be conveniently covered in the system-by-system information presented in Section 2.0. This additional material addresses plant-wide, multi-system issues; the extent to which the material applies to each of the individual Section 2.0 systems is defined by scope/application discussions in each of the Section 3.0 entries.
Additional Certified Design Material                                                                        3.0-1
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 3.1 Human Factors Engineering Design Description The ABWR certified designs human-system interface (HSI) will be developed, designed, and evaluated based upon a human factors systems analysis and shall reflect human factors principles. The HSI scope applies to the main control room (MCR) and Remote Shutdown System (RSS). Further, within the MCR, the HSI scope includes that area which provides the displays, controls and alarms required for normal, abnormal and emergency plant operations.
The HSI design effort will be directed by a multi-disciplinary HFE Design Team comprised of personnel with expertise in human factors engineering (HFE) and in other technical areas relevant to the HSI design, evaluation and operations. The HFE Design Team shall develop a Program Plan to establish methods for implementing the HSI design through a process of human factor systems analysis as shown in Figure 3.1. Implementation of that process will be as follows:
(1)    A System Functional Requirements Analysis Implementation Plan will be developed which establishes that plant system functional requirements will be analyzed to identify those functions which must be performed to satisfy the objectives of each functional area. System functional requirements analyses, as corrected to account for nonconformances, will be conducted in conformance with the provisions of this plan.
The functional analyses will determine the objectives, performance requirements and constraints of the design, and establish the functions which must be accomplished to meet the objectives and required performance.
(2)    An Allocation of Functions Implementation Plan will be developed to establish methods of allocating functions to personnel, system elements and personnel-system combinations. An analysis of the allocation of system functions, as corrected to account for nonconformances, will be conducted in conformance with the provisions of this Plan.
(3)    A Task Analysis Implementation Plan will be developed to establish methods for conducting the task analysis. The task analysis, as corrected to account for nonconformances, will be conducted in conformance with the provisions of this Plan and will be used to identify the behavioral requirements of the tasks the personnel are required to perform in order to achieve the functions allocated to them. The task analysis will identify the information and control requirements that form the basis for specifying the requirements for the displays, data processing and controls needed to carry out the tasks. The task analysis will also be used to maintain human performance requirements within human capabilities, as an input for developing personnel skill, personnel training, plant procedures and system communication requirements and as an input to the evaluation of established plant operations control room staffing levels.
Human Factors Engineering                                                                                  3.1-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 (4)  A Human-System Interface Design Implementation Plan will be developed to establish methods for applying human engineering principles in the design definition and evaluation of the HSI. HSI design definition and evaluation, as corrected to account for nonconformances, shall be conducted in conformance with the provisions of this plan.
(5)  A Human Factors Verification and Validation Implementation Plan will be developed to establish methods for conducting an evaluation of the HSI design as an integral system using HFE evaluation principles, procedures and criteria. The HSI design, as corrected to account for nonconformances, will be evaluated as an integrated system in conformance with the provisions of this plan.
(6)  The as-built configuration of the MCR and RSS shall be in conformance with the certified and validated MCR and RSS designs.
Inspections, Tests, Analyses and Acceptance Criteria Table 3.1 provides a definition of the instructions, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken to demonstrate compliance with the HFE commitments for the certified design.
3.1-2                                                                            Human Factors Engineering
 
ABWR Human Factors Engineering SYSTEM FUNCTIONAL REQUIREMENTS DEFINITION ALLOCATION OF FUNCTIONS TASK ANALYSIS 25A5675AA Revision 7 HUMAN-SYSTEM                        PROCEDURE DEVELOPMENT INTERFACE DESIGN HUMAN FACTORS Design Control Document/Tier 1 VERIFICATION AND VALIDATION (FEEDBACK)
IMPLEMENTED DESIGN 3.1-3 Figure 3.1 Human-System Interface Design Implementation Process
 
ABWR 3.1-4                                                                              Table 3.1 Human Factors Engineering Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                Design Acceptance Criteria
: 1.                                                  1.                                          1.
: a. A multi-disciplinary HFE Design Team            a. The composition of the HFE Design        a. The HFE design team shall be shall be established and be comprised of            Team shall be reviewed.                    comprised of the following expertise:
personnel with expertise in HFE and in (1) Technical Project Management other technical areas relevant to the HSI design, evaluation and operation.                                                                (2) Systems Engineering (3) Nuclear Engineering (4) Control and Instrumentation Engineering (5) Architect Engineering 25A5675AA Revision 7 (6) Human Factors (7) Plant Operations (8) Computer Systems Engineering (9) Plant Procedure Development (10)Personnel Training
: b. An HFE Program Plan shall be                    b. The HFE Program Plan shall be            b. The HFE Program Plan shall establish:
developed which establishes that the                reviewed.
(1) Methods and criteria for the HSI human-system interfaces shall be Design Control Document/Tier 1 development, design and evaluation developed, designed, and evaluated in accordance with accepted human based upon human factors systems factors practices and principles.
analysis and shall reflect human factors principles. The HSI scope shall apply to                                                        (2) Methods for addressing:
Human Factors Engineering the MCR and RSS.                                                                                    (a) The ability of the operating personnel to accomplish assigned tasks.
(b) Operator workload levels and vigilance.
 
Table 3.1 Human Factors Engineering (Continued)
ABWR Human Factors Engineering Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                Inspections, Tests, Analyses                  Design Acceptance Criteria 1.b. Continued                      1.b. Continued                                1.b. Continued (c) Operating personnel situation awareness.
(d) The operators information processing requirements.
(e) Operator memory requirements.
(f) The potential for operator error.
(3) HSI design and evaluation scope which applies to the MCR and RSS.
25A5675AA Revision 7 The HSI scope shall address normal, abnormal and emergency plant operations, and test and maintenance interfaces that impact the functions of the operations personnel. The HSI scope shall also address the development of operating technical procedures for normal, abnormal and emergency plant operations and the identification of personnel training Design Control Document/Tier 1 needs applicable to the HSI design.
3.1-5
 
Table 3.1 Human Factors Engineering (Continued)
ABWR 3.1-6 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                Inspections, Tests, Analyses                Design Acceptance Criteria 1.b. Continued                      1.b. Continued                                1.b. Continued (4) The HFE Design Team as being responsible for:
(a) The development of HFE plans and procedures.
(b) The oversight and review of HFE design, development, test, and evaluation activities.
(c) The initiation, recommendation, and provision of solutions through designated channels for 25A5675AA Revision 7 problems identified in the implementation of the HFE activities.
(d) Verification of implemention of solutions to problems.
(e) Assurance that HFE activities comply with the HFE plans and procedures.
(f) Phasing of activities.
Design Control Document/Tier 1 (5) The methods for the identification, closure and documentation of human factors issues.
Human Factors Engineering (6) The HSI design configuration control procedures.
(7) The methods for reviewing HSI operating experience.
 
Table 3.1 Human Factors Engineering (Continued)
ABWR Human Factors Engineering Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                    Design Acceptance Criteria
: 2.                                                  2.                                              2.
: a. A System Functional Requirements                a. The System Functional Requirements          a. The System Functional Requirements Analysis Implementation Plan shall be              Analysis Implementation Plan shall be          Analysis Implementation Plan shall developed which establishes that plant              reviewed.                                      establish:
system requirements shall be analyzed (1) Methods and criteria for conducting to identify those functions which must be the System Functional performed to satisfy the objectives of Requirements Analysis in each functional area. System function accordance with accepted human analysis shall determine the objective, factors practices and principles.
performance requirements, and constraints of the design, and establish                                                            (2) That system requirements shall the functions which must be                                                                            define the system functions and 25A5675AA Revision 7 accomplished to meet the objectives and                                                                those system functions shall pro-required performance.                                                                                  vide the basis for determining the associated HSI performance requirements.
(3) That functions critical to safety shall be identified.
(4) That descriptions shall be developed for each of the identified functions and for overall system configuration Design Control Document/Tier 1 design itself. Each function shall be identified and described in terms of inputs (observable parameters which will indicate system status),
functional processing (control process and performance measures required to achieve the function),
functional operations (including detecting signals, measuring information, comparing one measurement with another, 3.1-7                                                                                                                                      processing
 
Table 3.1 Human Factors Engineering (Continued)
ABWR 3.1-8 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                    Design Acceptance Criteria 2.a. Continued                                2.a. Continued                                        information, and acting upon decisions to produce a desired condition or result such as a system or component operation actuation or trip), outputs, feedback (how to determine correct discharge of function), and interface requirements so that subfunctions are related to larger functional elements.
: b. An analysis of system functional            b. The analysis of the system functional        b. The system functional requirements requirements shall be conducted.              requirements shall be reviewed.                analysis, as corrected to account for nonconformances, is conducted in accordance with the requirements of the 25A5675AA Revision 7 Human Factors Engineering Program Plan and the System Functional Requirements Analysis Implementation Plan.
Design Control Document/Tier 1 Human Factors Engineering
 
Table 3.1 Human Factors Engineering (Continued)
ABWR Human Factors Engineering Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                      Design Acceptance Criteria
: 3.                                                3.                                              3.
: a. An Allocation of Function Implementation        a. The Allocation of Function                    a. The Allocation of Function Implemen-Plan shall be developed which                      Implementation Plan shall be reviewed.          tation Plan shall establish:
establishes the methods for allocating (1) The methods and criteria for the functions to personnel, system elements, execution of function allocation in and personnel-system combinations.
accordance with accepted human factors practices and principles.
(2) That aspects of system and functions definition shall be analyzed in terms of resulting human performance requirements based on 25A5675AA Revision 7 the user population.
(3) That the allocation of functions to personnel, system elements, and personnel system combinations shall reflect:
(a) Sensitivity, precision, time, and safety requirements.
(b) Reliability of system performance.
Design Control Document/Tier 1 (c) The number and the necessary skills of the personnel required to operate and maintain the system.
(4) That allocation criteria, rationale, analyses, and procedures shall be documented.
3.1-9
 
Table 3.1 Human Factors Engineering (Continued)
ABWR 3.1-10 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                  Design Acceptance Criteria 3.a. Continued                                    3.a. Continued                                      (5) That analyses shall confirm that the personnel can perform tasks allocated to them while maintaining operator situation awareness, acceptable personnel workload, and personnel vigilance.
: b. A functional allocation analysis shall be      b. The functional allocation analysis shall    b. The functional allocation analysis, as conducted.                                        be reviewed.                                  corrected to account for nonconformances, is conducted in accordance with the requirements of the Human Factors Engineering Program Plan and the Allocation of Functions 25A5675AA Revision 7 Implementation Plan.
Design Control Document/Tier 1 Human Factors Engineering
 
Table 3.1 Human Factors Engineering (Continued)
ABWR Human Factors Engineering Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                    Design Acceptance Criteria
: 4.                                                  4.                                              4.
: a. A Task Analysis Implementation Plan              a. The Task Analysis Implementation Plan        a. The Task Analysis Implementation Plan shall be developed which establishes                shall be reviewed.                              shall establish:
that task analysis shall be conducted and (1) The methods and criteria for conduct used to identify the behavioral of the task analyses in accordance requirements of the tasks the personnel with accepted human factors are required to perform in order to practices and principles.
achieve the functions allocated to them.
The task analysis shall be used to                                                                  (2) The scope of the task analysis which maintain human performance                                                                              shall include operations performed requirements within human capabilities;                                                                at the operator interface in the MCR be used as an input for developing                                                                      and at the RSS. The analyses shall 25A5675AA Revision 7 personnel skill, personnel training, and                                                                be directed to the range of plant system communication requirements                                                                      operating modes, including startup, and as an input to the evaluation of                                                                    normal operations, abnormal established plant operations control                                                                    operations, transient conditions, low room staffing levels; and form the basis                                                                power and shutdown conditions. The for specifying the requirements for the                                                                analyses shall also address operator displays, data processing and controls                                                                  interface operations during periods needed to carry out tasks.                                                                              of maintenance, test and inspection of plant systems and equipment, including HSI equipment.
Design Control Document/Tier 1 (3) That the analysis shall be used to identify which tasks are critical to safety.
3.1-11
 
Table 3.1 Human Factors Engineering (Continued)
ABWR 3.1-12 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                    Design Acceptance Criteria 4.a. Continued                                4.a. Continued                                        (4) That task analysis shall develop narrative descriptions of the personnel activities required for successful completion of the task.
(5) That task analysis shall identify requirements for alarms, displays, data processing, and controls.
(6) That task analysis results shall be made available as input to the personnel training programs.
: b. A task analysis shall be conducted.          b. The task analysis shall be reviewed.        b. The task analysis, as corrected to 25A5675AA Revision 7 account for nonconformances, is conducted in accordance with the requirements of the Human Factors Engineering Program Plan and the Task Analysis Implementation Plan.
Design Control Document/Tier 1 Human Factors Engineering
 
Table 3.1 Human Factors Engineering (Continued)
ABWR Human Factors Engineering Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                  Design Acceptance Criteria
: 5.                                                  5.                                          5.
: a. HSI Design Implementation Plan shall be          a. The HSI Design Implementation Plan        a. The HSI Design Implementation Plan developed which establishes that human              shall be reviewed.                          shall establish:
engineering principles and criteria shall (1) The methods and criteria for HSI be applied in the design definition and design in accordance with accepted evaluation of the HSI.
human factors practices and principles.
(2) That the HSI design shall implement the information and control requirements:
25A5675AA Revision 7 (a) developed through the task analyses, including the displays, controls and alarms necessary for the execution of those tasks identified in the task analyses as being critical tasks and, (b) defined in Table 2.7.1.a.
(3) The methods for comparing the consistency of the HSI human performance, equipment design and Design Control Document/Tier 1 associated workplace factors with that modeled and evaluated in the completed task analysis.
(4) The HSI design criteria and guidance for control room operations during periods of maintenance, test and inspection.
(5) The test and evaluation methods for resolving HFE/HSI design issues.
These test and evaluation methods 3.1-13                                                                                                                                  shall include the criteria to be used in selecting HFE/HSI design and evaluation tools.
 
Table 3.1 Human Factors Engineering (Continued)
ABWR 3.1-14 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                      Design Acceptance Criteria
: b. The HSI design shall be implemented.            b. The HSI design implementation shall be        b. The HSI design implementation and reviewed.                                        analyses, as corrected to account for nonconformances, are conducted in accordance with the requirements of the Human Factors Engineering Program Plan and the HSI Design Implementation Plan,
: 6.                                                  6.                                              6.
: a. A Human Factors Verification and                a. The Human Factors V&V Plan shall be          a. The Human Factors V&V Validation (V&V) Implementation Plan                reviewed.                                        Implementation Plan shall establish:
shall be developed which establishes (1) The methods and criteria for 25A5675AA Revision 7 that the HSI design shall be evaluated as conducting the Human factors V&V an integrated system using HFE in accordance with accepted human evaluation principles, procedures and factors practices and principles.
criteria.
(2) That scope of the evaluations of the integrated HSI shall include:
(a) The HSI (including both the interface of the operator with the HSI equipment hardware and the interface of the operator with Design Control Document/Tier 1 the HSI equipments software driven functions).
(b) The Plant and Emergency Operating Procedures.
Human Factors Engineering (c) The HSI work environment.
 
Table 3.1 Human Factors Engineering (Continued)
ABWR Human Factors Engineering Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                Inspections, Tests, Analyses                  Design Acceptance Criteria 6.a. Continued                        6.a. Continued                                6.a. Continued (3) That evaluations of the HSI equipment shall be conducted to confirm that the controls, displays, and data processing functions identified in the task analyses are provided.
(4) That integration of HSI equipment with each other, with the operating personnel and with the Plant and Emergency Operating Procedures shall be evaluated through the 25A5675AA Revision 7 conduct of dynamic task performance testing. The dynamic task performance tests and evaluations shall have as their objectives:
(a) Confirmation that the identified critical functions can be achieved using the integrated HSI design.
Design Control Document/Tier 1 (b) Confirmation that the HSI design and configuration can be operated using the established MCR staffing levels.
3.1-15
 
Table 3.1 Human Factors Engineering (Continued)
ABWR 3.1-16 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                Inspections, Tests, Analyses                  Design Acceptance Criteria 6.a. Continued                        6.a. Continued                                6.a.(4) Continued (c) Confirmation that the Plant and Emergency Operating Procedures provide direction for completing the identified tasks associated with normal, abnormal and emergency operations.
(d) Confirmation that the time dependent and interactive aspects of the HSI equipment performance allow for task 25A5675AA Revision 7 accomplishment.
(e) Confirmation that the allocation of functions is sufficient to enable task accomplishment.
(5) That dynamic task performance test evaluations shall be conducted over the range of operational conditions and upsets.
Design Control Document/Tier 1 Human Factors Engineering
 
Table 3.1 Human Factors Engineering (Continued)
ABWR Human Factors Engineering Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                Inspections, Tests, Analyses                  Design Acceptance Criteria 6.a. Continued                        6.a. Continued                                6.a. Continued (6) The HFE performance measures to be used as the basis for evaluating the dynamic task performance test results. These performance measures shall address:
(a) Operating crew primary task performance characteristics, such as task times and procedure compliance.
(b) Operating crew errors and error 25A5675AA Revision 7 rates.
(c) Operating crew situation awareness.
(d) Operating crew workload.
(e) Operating crew communications and coordination.
(f) Anthropometry evaluations.
(g) HSI equipment performance Design Control Document/Tier 1 measures.
(7) The methods to confirm that HFE issues identified and documented have been resolved in the integrated HSI design.
(8) The methods and criteria to be used to confirm that critical human tasks, as defined by the task analysis, have been addressed in the integrated HSI design.
3.1-17
 
Table 3.1 Human Factors Engineering (Continued)
ABWR 3.1-18 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                    Design Acceptance Criteria
: b. A human factors engineering analysis of        b. The analyses of the integrated HSI          b. The human factors engineering analysis the integrated HSI design shall be                design shall be reviewed.                      of the HSI design, as corrected to conducted.                                                                                        account for nonconformances, is conducted in accordance with the requirements of the Human Factors Engineering Program Plan and the Human Factors V&V Implementation Plan.
: 7. The as-built configuration of the MCR and      7. Inspections of the as-built MCR and RSS will 7. An as-built evaluation report exists which RSS shall be in conformance with the              be conducted.                                  concludes that the as-built MCR and RSS certified and validated MCR and RSS                                                              conform to the certified and validated MCR designs.                                                                                          and RSS configurations, including layouts, 25A5675AA Revision 7 environmental characteristics, the HSI, alarms, displays and controls.
Design Control Document/Tier 1 Human Factors Engineering
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 3.2 Radiation Protection Design Description The ABWR design provides radiation protection features to keep exposures for both plant personnel and the general public below allowable limits. This section applies to the radiological shielding and ventilation design of the Reactor Building, Turbine Building, Control Building, Service Building, and Radwaste Building.
The plant design provides radiation shielding for rooms, corridors and operating areas commensurate with their occupancy requirements. Shielded cubicles, labyrinth access and provisions for temporary shielding are used to reduce exposure. Under accident conditions, plant shielding designs permit operators to perform required safety functions in vital areas of the plant. A vital area is an area which will or may require occupancy to permit an operator to aid in the mitigation of or recovery from an accident. In addition to protection of operating personnel, the plant design provides radiation shielding to protect the general public.
Plant ventilation systems maintain concentrations of airborne radionuclides at levels consistent with personnel access requirements. In addition, airborne radioactivity monitoring is provided for those normally occupied areas of the plant in which there exists a significant potential (greater than 0.1 per year) for airborne contamination.
Inspections, Tests, Analyses and Acceptance Criteria Tables 3.2a and 3.2b provide a definition of the inspections, tests, and/or analyses, together with associated design acceptance criteria, which will be undertaken for the ABWR plant shielding, ventilation and airborne monitoring equipment.
Radiation Protection                                                                                          3.2-1
 
25A5675AA Revision 7 ABWR                            Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2a Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Section A-A 3.2-2                                                                              Radiation Protection
 
25A5675AA Revision 7 ABWR                                    Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2b Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Section B-B Radiation Protection                                                                                  3.2-3
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2c Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Floor B3FElevation -8200 mm 3.2-4                                                                                Radiation Protection
 
25A5675AA Revision 7 ABWR                                  Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2d Reactor Building Radiation Zone Map for Full Power and Shutdown OperationsElevation -5100 mm Radiation Protection                                                                                3.2-5
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2e Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Floor B2FElevation -1700 mm 3.2-6                                                                                Radiation Protection
 
25A5675AA Revision 7 ABWR                                    Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2f Reactor Building Radiation Zone Map for Full Power and Shutdown OperationsElevation 1500 mm Radiation Protection                                                                                3.2-7
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2g Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Floor B1FElevation 4800 mm 3.2-8                                                                                Radiation Protection
 
25A5675AA Revision 7 ABWR                                  Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2h Reactor Building Radiation Zone Map for Full Power and Shutdown OperationsElevation 8500 mm Radiation Protection                                                                                3.2-9
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2i Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Floor 1FElevation 12300 mm 3.2-10                                                                                Radiation Protection
 
25A5675AA Revision 7 ABWR                                    Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2j Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Floor 2FElevation 18100 mm Radiation Protection                                                                                3.2-11
 
25A5675AA Revision 7 ABWR                                Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2k Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Floor 3FElevation 23500 mm 3.2-12                                                                                Radiation Protection
 
25A5675AA Revision 7 ABWR                                    Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2l Reactor Building Radiation Zone Map for Full Power and Shutdown OperationsElevation 27200 mm Radiation Protection                                                                                3.2-13
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2m Reactor Building Radiation Zone Map for Full Power and Shutdown Operations, Floor 4FElevation 31700mm 3.2-14                                                                                Radiation Protection
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2n Reactor Building Radiation Zone Map for Full Power and Shutdown OperationsElevations 34500 mm and 38200 Radiation Protection                                                                                                                          3.2-15
 
25A5675AA Revision 7 ABWR                Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2o Control Building Radiation Zone Map for Full Power Operations, Section A-A 3.2-16                                                                  Radiation Protection
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2p Control Building Radiation Zone Map for Full Power Operations, Section B-B Radiation Protection                                                                                                                                                          3.2-17
 
25A5675AA Revision 7 ABWR                Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2q Control Building Radiation Zone Map for Full Power Operation, Floor B4FElevation -8200 mm 3.2-18                                                                  Radiation Protection
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2r Control Building Radiation Zone Map for Full Power Operation, Floor B3FElevation -2150 mm Radiation Protection                                                                                                                                                                        3.2-19
 
25A5675AA Revision 7 ABWR                Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2s Control Building Radiation Zone Map for Full Power Operation, Floor B2FElevation 3500 mm 3.2-20                                                                  Radiation Protection
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2t Control Building Radiation Zone Map for Full Power Operation, Floor B1FElevation 7900 mm Radiation Protection                                                                                                                                                                        3.2-21
 
25A5675AA Revision 7 ABWR                Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2u Control Building Radiation Zone Map for Full Power Operation, Floor 1FElevation 12300 mm 3.2-22                                                                  Radiation Protection
 
25A5675AA Revision 7 ABWR                              Security-Sensitive Information    Design Withhold from Public Disclosure under 10 CFR 2.390 Control Document/Tier 1 Figure 3.2v Control Building Radiation Zone Map for Full Power Operation, Floor 2FElevation 17150 mm Radiation Protection                                                                                                                                                                        3.2-23
 
ABWR 3.2-24 Table 3.2a Plant Shielding Design Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                        Design Acceptance Criteria
: 1. The plant design shall provide radiation      1. An analysis of the expected radiation levels 1. Maximum expected radiation dose rates in shielding for rooms, corridors and operating      in each plant area will be performed to verify  each plant area (deep dose equivalent areas commensurate with their occupancy          the adequacy of the shielding design. This      measured at 30 cm from the source of the requirements.                                    analysis shall consider the following:          radiation, not contact dose rates) are no
: a. Confirmatory calculations shall consider    greater than the dose rates specified for the significant radiation sources (greater      following zones, based on the access than 5% contribution) for an area.          requirements of that area for plant operation Radiation source strength in plant          and maintenance.
systems and components will be determined based upon an assumed              Zone    Dose Rate        Access source term of 3,700 MBq/s offgas                      (&#xb5;Sv/h)        Requirements 25A5675AA Revision 7 release rate (after 30 minutes decay), a 11.1 MBq/gram-steam N-16 source term            A          6      Uncontrolled, at the vessel exit nozzle, and a core                              unlimited access.
inventory commensurate with a 4005              B        <10      Controlled, unlimited MWt equilibrium core at 51.6 kW/liter.                              access.
Source terms shall be adjusted for radiological decay and buildup of              C        <50      Controlled, limited activated corrosion and wear products.                              access 20 h/week.
: b. Commonly accepted shielding codes,              D        <250      Controlled, limited using nuclear properties derived from                                access 4 h/week.
Design Control Document/Tier 1 well known references (such as Vitamin C and ANSI/ANS-6.4) shall be used to            E        <1000      Controlled, limited model and evaluate plant radiation                                  access 1 h/week.
environments.
F        1000      Restricted, (1) For non-complex geometries, point                              infrequent access.
kernel shielding codes (such as                                Authorization Radiation Protection QAD or GGG) shall be used.                                      required.
 
Table 3.2a Plant Shielding Design (Continued)
ABWR Radiation Protection Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                      Design Acceptance Criteria (2) For complex geometries, more            Plant layout such that access to higher sophisticated two or three              zones (areas with higher dose rates) is from dimensional transport codes (such        lower zoned areas. Corridors and normal as DORT or TORT) shall be used.          traffic areas are Zone C or less. Control rooms are Zone B or less. Radiation zones for the Reactor Building and Control Building are indicated in Figures 3.2a through 3.2v.
: c. A safety factor shall be applied based upon benchmark comparisons.
: 2. The plant design shall provide shielded        2. Using the methods identified in (1) above,    2. Shielding design of a room including any cubicles, labyrinth access, and space for          radiation levels present in rooms shall be        temporary shielding is such that radiation 25A5675AA Revision 7 temporary shielding to reduce radiation            evaluated for the contribution from adjacent      from adjacent rooms shall contribute no exposure from adjacent rooms.                      rooms.                                            more than a small fraction (10% or less) of the dose rate or less than 0.6 &#xb5;Sv/h whichever is larger, in the room. For this purpose, the drywell shall be considered a room.
: 3. The plant radiation shielding design shall  3. An analysis of the expected high radiation    3. Under accident conditions, radiation permit plant personnel to perform required      levels in each area which will or may require    shielding design allows access to occupancy safety functions in vital areas of the plant    occupancy to permit plant personnel to aid in    and egress from areas required to maintain (including access and egress of these areas)    the mitigation of or recovery from an accident    post-accident safety functions such that Design Control Document/Tier 1 under accident conditions.                      (vital area) shall be performed to verify the    individual personnel radiation doses do not adequacy of the plant shielding design. This      exceed 0.05 Sv to the whole body, or its analysis shall use calculational methods          equivalent, for the duration of the accident consistent with (1.b) above and a radiation      (based on the required frequency of access source term (adjusted for radioactive decay)      to each vital area).
based on the following:
3.2-25
 
Table 3.2a Plant Shielding Design (Continued)
ABWR 3.2-26 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                      Design Acceptance Criteria
: 3. (continued)                                  3. (continued)                                    3. (continued)
: a. Liquid containing systems: 100% of the        For areas requiring continuous occupancy core equilibrium noble gas inventory,        (such as the main control room, technical 50% of the core equilibrium halogen          support center, and emergency operations inventory and 1% of the equilibrium core      support center), design dose rates shall not inventory of the remaining radionuclides      exceed 150 &#xb5;Sv/h (averaged over 30 days).
are assumed to be mixed in the reactor coolant and recirculation liquids recirculated by the Residual Heat Removal (RHR) System, the High Pressure Core Flooder (HPCF) System, and the Reactor Core Isolation Cooling 25A5675AA Revision 7 (RCIC) System.
: b. Gas containing systems: 100% of the core equilibrium noble gas inventory and 25% of the core equilibrium halogen activity are assumed to be mixed in the containment atmosphere. For vapor containing systems (such as the main steam lines), these core inventory fractions are assumed to be contained in the reactor coolant vapor space.
Design Control Document/Tier 1
: 4. The plant design shall provide radiation        4. Using the methods identified in (1) above,  4. As a result of normal operations, the shielding to protect the general public outside    the radiation dose to the maximally exposed    radiation dose from direct and scattered of the controlled area.                            member of the general public outside of the    radiation shine to the maximally exposed controlled area from direct and scattered      member of the public outside of the radiation shine shall be determined.          controlled area is equal to or less than 25
                                                                                                                            &#xb5;Sv/year.
Radiation Protection
 
ABWR Radiation Protection Table 3.2b Ventilation and Airborne Monitoring Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                      Design Acceptance Criteria
: 1. Plant design shall provide for containment of 1. Expected concentrations of airborne            1. Calculation of radioactive airborne airborne radioactive materials and the          radioactive material shall be calculated by      concentration shall demonstrate that:
ventilation system will maintain                radionuclide for normal plant operations and
: a. For normally occupied rooms and areas concentrations of airborne radionuclides at      anticipated operational occurrences for each of the plant (i.e., those areas requiring levels consistent with personnel access          equipment cubicle, corridor, and operating routine access to operate and maintain needs.                                          area requiring personnel access.
the plant), equilibrium concentrations of Calculations shall consider:
airborne radionuclides will be a small
: a. Total ventilation flow rates for each area.        fraction (10% or less) of the occupational concentration limits listed in 10CFR20 Appendix B, January 1994.
: b. Typical leakage characteristics for            b. For rooms that require infrequent access 25A5675AA Revision 7 equipment located in each area.                  (such as for non-routine equipment maintenance), the ventilation system shall be capable of reducing radioactive airborne concentrations to (and maintaining them at) the occupational concentration limits listed in 10CFR20 Appendix B, January 1994,during the periods that occupancy is required.
: c. For rooms where access is not
: c. A radiation source term in each fluid            anticipated to perform scheduled Design Control Document/Tier 1 system based upon an assumed offgas              maintenance or surveillance (such as the rate of 3,700 MBq/s (30 minute decay)            backwash receiving tank room), plant appropriately adjusted for radiological          design shall provide containment and decay and buildup of activated corrosion          ventilation to reduce airborne and wear products.                                contamination spread to other areas of lower contamination.
3.2-27
 
Table 3.2b Ventilation and Airborne Monitoring (Continued)
ABWR 3.2-28 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                        Design Acceptance Criteria
: 2. Airborne radioactivity monitoring shall be      2. An analysis shall be performed to identify the 2. Airborne radioactivity monitoring system provided for those normally occupied areas          plant areas that require airborne radioactivity  shall be installed as defined in this certified of the plant in which there exists a significant    monitoring.                                      design commitment.
potential for airborne contamination (greater than 0.1 per year). The airborne radioactivity system shall:
: a. Have the capability of detecting the time integrated concentrations of the most limiting internal dose particulate and iodine radionuclides in each area equivalent to the occupational concentration limits in 10CFR20, 25A5675AA Revision 7 Appendix B, January 1994, for 10 hours.
: b. Provide a calibrated response, representative of the concentrations within the area (i.e., air sampling monitors in ventilation exhaust streams shall collect an isokinetic sample).
: c. Provide local audible alarms (visual alarms in high noise areas) with variable alarm setpoints, and Design Control Document/Tier 1 readout/annunciation capability.
Radiation Protection
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 3.3 Piping Design Design Description Piping associated with fluid systems is categorized as either nuclear safety-related (i.e., Seismic Category I) or non-nuclear safety (NNS) related (i.e., non-Seismic Category I). The piping shall be designed for a design life of 60 years. Piping systems that must remain functional during and following a safe shutdown earthquake (SSE) are designated as Seismic Category I and are further classified as American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code) Class 1, 2 or 3. The piping design requirements identified in this section encompass piping systems classified as nuclear safety-related and unless otherwise specified in this description, piping systems means nuclear safety-related piping systems. Piping systems and their components are designed and constructed in accordance with the ASME Code requirements identified in the individual system Design Descriptions.
Piping systems shall be designed to meet their ASME Code class and Seismic Category I requirements. The ASME Code Class 1, 2 and 3 piping systems shall be designed to retain their pressure integrity and functional capability under internal design and operating pressures and design basis loads. Piping stresses due to static and dynamic loads shall be combined and calculated in accordance with the ASME Code and shall be shown to be less than the ASME Code allowables for each service level.
For ASME Code Class 1 piping systems, a fatigue analysis shall be performed in accordance with the ASME Code Class 1 piping requirements. Environmental effects shall be included in the fatigue analysis. The Class 1 piping fatigue analysis shall show that the ASME Code Class 1 piping fatigue requirements have been met.
For ASME Code Class 2 and 3 piping systems, piping stress ranges due to thermal expansion shall be calculated in accordance with the ASME Code Class 2 and 3 piping requirements. The piping stress analysis shall show that the ASME Code Class 2 and 3 piping thermal expansion stress range requirements have been met. For the ASME Code Class 2 and 3 piping systems and their components which will be subjected to severe thermal transients, the effects of these transients shall be included in the design.
Feedwater lines shall be designed for thermal stratification loads.
Piping systems shall be designed to minimize the effects of erosion/corrosion.
For those piping systems using ferritic materials as permitted by the design specification, the ferritic materials and fabrication processes shall be selected to ensure that the piping system is not susceptible to brittle fracture under the expected service conditions.
For those piping systems using austenitic stainless steel materials as permitted by the design specification, the stainless steel piping material and fabrication process shall be selected to Piping Design                                                                                                3.3-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 reduce the possibility of cracking during service. Chemical, fabrication, handling, welding, and examination requirements that reduce cracking shall be met.
Piping system supports shall be designed to meet the requirements of ASME Code Subsection NF.
For piping systems, the pipe applied loads on attached equipment shall be calculated and shown to be less than the equipment allowable loads.
Analytical methods and load combinations used for analysis of piping systems shall be referenced or specified in the ASME Code Certified Stress Report. Piping systems and their supports shall be mathematically modeled to provide results for piping system frequencies up to the analysis cutoff frequency. Computer programs used for piping system dynamic analysis shall be benchmarked.
Systems, structures and components that shall be required to be functional during and following an SSE shall be protected against the dynamic effects associated with postulated high energy pipe breaks in Seismic Category I and NNS piping systems. In addition, structures, systems, and components that shall be required to be functional during and following an SSE shall be protected against or qualified to withstand the environmental effects of spraying, flooding, pressure and temperature due to postulated pipe breaks and cracks in Seismic Category I and NNS piping systems. Each postulated piping failure shall be documented in a Pipe Break Analysis Report which concludes the reactor can be shut down safely and maintained in a safe, cold shutdown condition without offsite power. The Pipe Break Analyses Report shall specify the criteria used to postulate breaks and the analytical methods used to perform the pipe break analysis. For postulated pipe breaks, the Pipe Break Analysis Report shall confirm: (1) piping stresses in the containment penetration area shall be within their allowable stress limits, (2) pipe whip restraints and jet shield designs shall be capable of mitigating pipe break loads, (3) loads on safety-related systems, structures and components shall be within their design loads limits, and (4) safety-related piping required to be functional during and following an SSE are protected against or qualified to withstand the environmental conditions that would exist without loss of its safety function for the time needed to be functional. Piping systems that are qualified for leak-before-break design may exclude design features to mitigate the dynamic effects from postulated high energy pipe breaks.
Piping systems shall be designed to provide clearance from structures, systems, and components where necessary for the accomplishment of the structure, system, or component's safety function as specified in the respective structure or system Design Description.
The as-built piping shall be reconciled with the piping design required by this section.
3.3-2                                                                                        Piping Design
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Inspections, Tests, Analyses and Acceptance Criteria Table 3.3 provides a definition of the inspections, tests, analyses, and associated acceptance criteria, which will be performed for Advanced Boiling Water Reactor (ABWR) nuclear safety-related and NNS related piping systems, as specified in this design description. Piping classification information is provided in the individual Certified Design Material (CDM) entry for each of the piping systems. Furthermore, Table 3.3 may be completed on an individual system basis.
Piping Design                                                                                            3.3-3
 
ABWR 3.3-4                                                                      Table 3.3 Piping Design Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                          Acceptance Criteria
: 1. The piping system shall be designed to meet 1. Inspections of ASME Code required                1. An ASME Code Certified Stress Report its ASME Code Class and Seismic Category      documents will be conducted.                        exists for the piping system and concludes I requirements.                                                                                    that the design complies with the The ASME Code Class 1, 2, and 3 piping                                                            requirements of ASME Code, Section III.
system shall be designed to retain its pressure integrity and functional capability under internal design and operating pressures and design basis loads. Piping and piping components shall be designed to show compliance with the requirements of ASME Code Section III.
25A5675AA Revision 7
: 2. Systems, structures, and components, that    2. Inspections of the Pipe Break Analysis          2. A Pipe Break Analysis Report and Leak-are required to be functional during and        Report, or Leak-Before-Break Report (if            Before-Break Report (if applicable) exist for following an SSE, shall be protected against    applicable), will be conducted.                    the as-built plant and concludes that for each or qualified to withstand the dynamic and                                                          postulated piping failure, the reactor can be An inspection of the as-built high and environmental effects associated with                                                              shut down safely and maintained in a safe, moderate energy pipe break mitigation postulated failures in Seismic Category I and                                                      cold shutdown condition without offsite features (including spatial separation) will be NNS piping systems. Each postulated piping                                                          power. The Pipe Break Analysis Report performed.
failure shall be documented in a Pipe Break                                                        includes the results of inspections of high Analysis Report. Piping systems that are                                                            and moderate energy pipe break mitigation qualified for leak-before-break design may                                                          features including spatial separation.
Design Control Document/Tier 1 exclude design features to mitigate the dynamic effects from postulated high energy pipe breaks.
: 3. The as-built piping shall be reconciled with    3. A reconciliation analysis using the as-      3. An as-built stress report exists and the piping design required in Section 3.3.        designed and as-built information will be        concludes that the as-built piping has been performed.                                      reconciled with the design documents used for design analysis. For ASME Code Class piping, the as-built stress report includes the Piping Design ASME Code Certified Stress Report and documentation of the results of the as-built reconciliation analysis.
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 3.4 Instrumentation and Control Introduction Subsection A provides a description of the configuration of safety-related, digital instrumentation and control (I&C) equipment encompassed by Safety System Logic and Control (SSLC). Subsection B contains a description of the hardware and software development process used in the design, testing, and installation of I&C equipment. This includes descriptions of the processes used to establish programs that assess and mitigate the effects of electromagnetic interference, establish setpoints for instrument channels, and ensure the qualification of the installed equipment. Subsection C discusses the diverse features implemented in I&C system design to provide backup support for postulated worst-case common-mode failures of SSLC.
The devices addressed in this section are electronic components of the ABWRs I&C systems.
These components are configured as real-time microcontrollers that use microprocessors and other programmable logic devices to perform data acquisition, data communications, and system logic processing. These components also contain automatic, on-line self-diagnostic features to monitor these tasks and off-line test capability to aid in maintenance and surveillance. The operating programs for these controllers are integrated into the hardware as firmware [software permanently stored in programmable read-only memory (PROM)]. A controllers operating system can permit field adjustment of selected parameters under proper change control. Adjustable parameters are stored in electrically-alterable read-only memory (EAROM) or equivalent.
A. Safety System Logic and Control Design Description Safety-related monitoring and trip logic for the plant protection systems resides in SSLC equipment. SSLC integrates the automatic decision-making and trip logic functions and manual operator initiation functions associated with the safety actions of the safety-related systems.
SSLC generates the protective function signals that activate reactor trip and provide safety-related mitigation of reactor accidents. The relationship between SSLC and systems for plant protection is shown in Figure 3.4a.
SSLC equipment comprises microprocessor-based, software-controlled signal processors that perform signal conditioning, setpoint comparison, trip logic, system initiation and reset, self-test, calibration, and bypass functions. The signal processors associated with a particular safety-related system are an integral part of that system. Functions in common, such as self-test, calibration, bypass control, power supplies and certain switches and indicators, belong to SSLC. However, SSLC is not, by itself, a system; SSLC is the aggregate of signal processors for several safety-related systems. SSLC hardware and software are classified as Class 1E, safety-related.
Instrumentation and Control                                                                                  3.4-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 Sensors used by the safety-related systems can be either analog, such as process control transmitters, or discrete, such as limit switches and other contact closures. While some sensor signals are hardwired directly to the SSLC processors, most sensor signals are transmitted from the instrument racks in the Reactor Building to the SSLC equipment in the Control Building via the Essential Multiplexing System (EMS). Both analog and discrete sensors are connected to remote multiplexing units (RMUs) in local areas, which perform signal conditioning, analog-to-digital conversion for continuous process inputs, change-of-state detection for discrete inputs, and message formatting prior to signal transmission. The RMUs are limited to acquisition of sensor data and the output of control signals. Trip decisions and other control logic functions are performed in SSLC processors in the main control room area.
The basic hardware configuration for one division of SSLC is shown in Figure 3.4b. Each division runs independently i.e., asynchronously) with respect to the other divisions. The following steps describe the processing sequence for incoming sensor signals and outgoing control signals. These steps are performed simultaneously and independently in each of the four divisions:
(1)  The digitized sensor inputs from RMUs are received in the control room at control room multiplexing units (CMUs), which associate sensor signals with their logic processing channel. These sensor signals are decoded by a microprocessor-based function, the Digital Trip Module (DTM). For sensor signals hardwired to the control room, the DTM also performs digitizing and signal conditioning tasks. For each system function, the DTM then compares these inputs to preprogrammed threshold levels (setpoints) for possible trip action. The DTM provides a discrete trip decision for each setpoint comparison.
(2)  For Reactor Protection System (RPS) trip and main steam isolation valve (MSIV) closure functions, trip outputs from the DTM are then compared, using a 2-out-of-4 coincidence logic format, with trip outputs from the DTMs of the other three divisions. The trip outputs are compared in the trip logic unit (TLU), another microprocessor-based device. The logic format for the DTM and TLU is fail-safe (i.e., de-energize-to-operate). Thus, a reactor trip or MSIV closure signal occurs on loss of input signal or power to the DTM, but, because of the 2--out-of-4 logic format in the TLU, a tripped state does not appear at the output of the TLU (for a single division loss of power). Loss of signal or power to a divisions TLU also causes a tripped output state, but the 2-out-of-4 configuration of actuator load drivers prevents de-energization of the pilot valve solenoids.
3.4-2                                                                            Instrumentation and Control
 
25A5675AA Revision 7 ABWR                                                                            Design Control Document/Tier 1 (3)    Trip outputs are sent from the TLU to the RPS and MSIV output logic units (OLUs).
The OLUs use non-microprocessor circuitry to provide a diverse (i.e., not software-based) interface for the following manual functions:
(a)    Manual reactor trip (per division: 2-out-of-4 for completion).
(b)    MSIV closure (per division: 2-out-of-4 for completion).
(c)    MSIV closure (eight individual control switches).
(d)    RPS and MSIV trip reset.
(e)    TLU output bypass The OLUs distribute the automatic and manual trip outputs to the MSIV pilot valve and scram pilot valve actuating devices and provide control of trip seal-in, reset, and TLU output bypass (division-out-of-service bypass). Bypass inhibits automatic trip but has no effect on manual trip. The OLUs also provide a manual test input for de-energizing a division's parallel load drivers (part of the 2-out-of-4 output logic arrangement) so that scram or MSIV closure capability can be confirmed without solenoid de-energization. The OLUs are located external to the TLU so that manual MSIV closure or manual reactor trip (per division) can be performed either when a division's microprocessor logic is bypassed or when failure of sensors or microprocessor logic equipment causes trip to be inhibited.
(4)    Trips are transmitted across divisions for 2-out-of-4 voting via fiber optic data links to preserve signal isolation among divisions. The TLU also receives inputs directly from the trip outputs of the Neutron Monitoring System, manual control switches, and contact closures from limit switches and position switches used for equipment interlocks. In addition, plant sensor signals and contact closures that do not require transmittal to other divisions for 2-out-of-4 trip comparison are provided as inputs directly to the TLU. In this case, the TLU also performs the trip setpoint comparison (DTM) function.
(5)    For Leak Detection and Isolation System (LDS) functions (except MSIV),
emergency core cooling system (ECCS) functions, other safety-related supporting functions, and Electrical Power Distribution System functions such as diesel generator start and load sequencing, logic processing is performed as above, but in DTMs separate from the RPS/MSIV DTMs and in Safety System Logic Units (SLUs). The SLUs are similar to TLUs, but are dual redundant in each processing channel for protection against inadvertent initiation. Dual SLUs both receive the same inputs from the DTM, manual control switch inputs, and contact closures. Both SLU outputs must agree before the final trip actuators are energized. The logic format for the DTM and SLUs is fail-as-is (i.e., energize-to-operate) for ECCS and other safety-related supporting functions. Thus, loss of power or equipment failure does not cause a trip or initiation action. However, containment isolation signals are in Instrumentation and Control                                                                                    3.4-3
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 fail-safe format and cause an isolation signal output on loss of power or signal.
Besides performing 2-out-of-4 voting logic, the SLUs also provide interlock logic functions conforming to the logic diagram requirements of each supported safety system.
As shown in Figure 3.4b, a pair of SLU are located in each of two engineered safety feature (ESF) processing channels, ESF1 and ESF2. ESF1 processes initiation logic for functions which service the reactor vessel at low pressure (e.g. RHR), while ESF2 provides the same support for the vessel at high pressure (e.g. Reactor Core Isolation Cooling (RCIC) System and High Pressure Core Flooder (HPCF) System).
Associated LDS and ESF functions are also allocated to these logic channels.
(6)    For reactor trip or MSIV closure, if a 2-out-of-4 trip condition of sensors is satisfied, all four divisions trip outputs produce a simultaneous coincident trip signal (e.g.,
reactor trip) and transmit the signal through hardwired connections (and isolators where necessary) to load drivers that control the protective action of the actuators.
The load drivers are themselves arranged in a 2-out-of-4 configuration, so that at least two divisions must produce trip outputs for protective action to occur.
(7)    For ESF functions, the trip signals in three divisions are transmitted by the Essential Multiplexing System to the RMUs, where a final 2-out-of-2 logic comparison is made prior to distribution of the control signals to the final actuators. ESF outputs do not exist in Division IV.
The DTM, TLU, and OLUs for RPS and MSIV in each of the four instrumentation divisions are powered from their respective divisional Class 1E AC sources. The DTMs and SLUs for ESF 1 and ESF 2 in Divisions I, II, and III are powered from their respective divisional Class 1E DC sources. In SSLC, independence is provided between Class 1E divisions, and also between Class 1E divisions, and also non-Class 1E equipment.
Bypassing of any single division of sensors (i.e., those sensors whose trip status is confirmed by 2-out-of-4 logic) is accomplished from each divisional SSLC cabinet by means of the manually-operated bypass unit. When such bypass is made, all four divisions of 2-out-of-4 input logic become 2-out-of-3 while the bypass state is maintained. During bypass, if any two of the remaining three divisions reach trip level for any sensed input parameter, then the output logic of all four divisions trips (for RPS and MSIV functions) or the three ECCS divisions initiate the appropriate safety system equipment.
Bypassing of any single division of output trip logic (i.e., taking a logic channel out of service) is also accomplished by means of the bypass unit. This type of bypass is limited to the fail-safe (de-energize-to-operate) reactor trip and MSIV closure functions, since removal of power from energize-to-operate signal processors is sufficient to remove that channel from service.
3.4-4                                                                              Instrumentation and Control
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 When a trip logic output bypass is made, the TLU trip output in a division is inhibited from affecting the output load drivers by maintaining that divisions load drivers in an energized state. Thus, the 2-out-of-4 logic arrangement of output load drivers for the RPS and MSIV functions effectively becomes 2-out-of-3 while the bypass is maintained.
Bypass status is indicated in the main control room until the bypass condition is removed. An electrical interlock rejects attempts to remove more than one SSLC division from service at a time.
ESF1 and ESF2 logic are each processed in two redundant channels within each divisional train of ESF equipment. In order to prevent spurious actuation of ESF equipment, final output signals are voted 2-out-of-2 at the remote multiplexing units by means of series-connected load drivers at the RMU outputs. However, in the event of a failure detected by self-test within either processing channel, a bypass (ESF output channel bypass) is applied automatically (with manual backup) such that the failed channel is removed from service. The remaining channel provides 1-out-of-1 operation to maintain availability during the repair period. Channel failures are alarmed in the main control room. If a failed channel is not automatically bypassed, the operator is able to manually bypass the channel by a hardwired connection from the main control room.
A portion of the anticipated transient without scram (ATWS) mitigation features is provided by SSLC circuitry, with initiating conditions as follows:
(1)    Initiation of automatic Standby Liquid Control System (SLCS) injection: High dome pressure and startup range neutron monitor (SRNM) ATWS permissive for 3 minutes or greater, or low reactor water level and SRNM ATWS permissive for 3 minutes or greater.
(2)    Initiation of feedwater runback: High dome pressure and SRNM ATWS permissive for 2 minutes or greater. Reset permitted only when both signals drop below the setpoints.
These ATWS features are implemented in four divisions of SSLC control circuitry that are functionally independent and diverse from the circuitry used for the Reactor Protection System (Figure 3.4c).
SSLC has the following alarms, displays, and controls in the main control room:
(1)    SSLC signal processor inoperative (INOP).
(2)    SSLC manual controls for bypass as described above.
(3)    Displays for bypass status.
Instrumentation and Control                                                                                3.4-5
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 (4)  Divisional flat display panels that provide display and control capability for manual ESF functions.
(5)  Display and control of calibration and off-line self-test functions.
Inspections, Tests, Analyses and Acceptance Criteria Table 3.4, Items 1 through 6, provides a definition of the inspections, tests and analyses, together with associated acceptance criteria, which will be undertaken for SSLC.
B. I & C Development and Qualification Processes Hardware and Software Development Process The ABWR design uses programmable digital equipment to implement operating functions of instrumentation and control (I&C) systems. The equipment is in the form of embedded controllers (i.e., a control program developed in software is permanently stored in PROM, and thus becomes part of the controllers hardware).
A quality assurance program encompassing software is employed as a controlled process for software development, hardware integration, and final product and system testing. The development process for safety-related hardware and software includes a verification and validation (V&V) program. Non-safety-related hardware and software will be developed using a planned design process similar to the safety-related development program, but with periodic design reviews rather than formal V&V.
System functional performance testing for each system using the software-based controllers discussed herein is addressed in Section 2 system entries.
An overall software development plan establishes the requirements and methodology for software design and development. The plan also defines methods for auditing and testing software during the design, implementation, and integration phases. These phases are part of the software life cycle, a planned development method to ensure the quality of software throughout its period of usage. The relationship between components of the plan and I&C design activities is shown in Figure 3.4d.
As part of the design of software for safety-related applications, the software development plan, at each defined phase of the software life cycle, addresses software requirements that have been defined as safety-critical. Safety-critical is defined as those computer software components (processes, functions, values or computer program states) in which errors (inadvertent or unauthorized occurrence, failure to occur when required, occurrence out of sequence, occurrence in combination with other functions, or erroneous values) can result in a potential hazard or loss of predictability or control of a system. Potential hazards are failure of a safety-related function to occur on demand and spurious occurrence of a safety-related function in an unsafe direction.
3.4-6                                                                                  Instrumentation and Control
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 The overall software development plan comprises the following plans:
(1)    A Software Management Plan (SMP) that establishes standards, conventions and design processes for I&C software.
A SMP shall be instituted which establishes that software for embedded control hardware shall be developed, designed, evaluated, and documented per a design development process that addresses, for safety-related software, software safety issues at each defined life-cycle phase of the software development.
The SMP defines the following software life-cycle phases:
(a)  Planning (b)  Design definition (c)  Software design (d)  Software coding (e)  Integration (f)  Validation (g)  Change control The SMP shall state that the output of each defined life-cycle phase shall be documents that define the current state of that design phase and the design input for the next design phase and the software products are developed using the SMP.
(2)    A Configuration Management Plan (CMP) that establishes the standards and procedures controlling software design and documentation.
A CMP shall be instituted that establishes the methods for maintaining, throughout the software design process, the design documentation, procedures, evaluated software, and the resultant as-installed software.
The CMP addresses:
(a)  Identification of CMP software documentation.
(b)  Management of software change control.
(c)  Control and traceability of software changes.
(d)  Verification of software to design requirements.
(e)  Dedication of commercial software.
(3)    A V&V plan that establishes verification reviews and validation testing procedures.
Instrumentation and Control                                                                                  3.4-7
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 A V&V plan shall be developed which establishes that developed software shall be subjected to structured and documented verification reviews and validation testing, including testing of the software integrated into the target hardware.
The V&V plan addresses:
(a)    Independent design verification.
(b)    Baseline software reviews.
(c)    Testing.
(d)    Procedure for software revisions.
Electromagnetic Compatibility Electromagnetic compatibility (EMC) is the ability of equipment to function properly when subjected to an electromagnetic environment. An EMC compliance plan to confirm the level of immunity to electrical noise is part of the design, installation, and pre-operational testing of I&C equipment.
Electrical and electronic components in the systems listed below are qualified according to the established plan for the anticipated levels of electrical interference at the installed locations of the components:
(1)  Safety System Logic and Control.
(2)  Essential Multiplexing System.
(3)  Non-Essential Multiplexing System.
(4)  Other microprocessor-based, software controlled systems or equipment.
The plan is structured on the basis that EMC of I&C equipment is verified by factory testing and site testing of both individual components and interconnected systems to meet electromagnetic compatibility requirements for protection against the effects of:
(1)  Electromagnetic Interference (EMI).
(2)  Radio Frequency Interference (RFI).
(3)  Electrostatic Discharge (ESD).
(4)  Electrical surge [Surge Withstand Capability (SWC)].
To be able to predict the degree of electromagnetic compatibility of a given equipment design, the following information is developed:
(1)  Characteristics of the sources of electrical noise.
3.4-8                                                                              Instrumentation and Control
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 (2)    Means of transmission of electrical noise.
(3)    Characteristics of the susceptibility of the system.
(4)    Techniques to attenuate electrical noise.
After these characteristics of the equipment are identified, noise susceptibility is tested for four different paths of electrical noise entry:
(1)    Power feed lines.
(2)    Input signal lines.
(3)    Output signal lines.
(4)    Radiated electromagnetic energy.
Instrument Setpoint Methodology Setpoints for initiation of safety-related functions are determined, documented, installed and maintained using a process that establishes a general program for:
(1)    Specifying requirements for documenting the bases for selection of trip setpoints.
(2)    Accounting for instrument inaccuracies, uncertainties, and drift.
(3)    Testing of instrumentation setpoint dynamic response.
(4)    Replacement of setpoint-related instrumentation.
The determination of nominal trip setpoints includes consideration of the following factors:
Design Basis Analytical Limit In the case of setpoints that are directly associated with an abnormal plant transient or accident analyzed in the safety analysis, a design basis analytical limit is established as part of the safety analysis. The design basis analytical limit is the value of the sensed process variable prior to or at the point which a desired action is to be initiated. This limit is set so that associated licensing safety limits are not exceeded, as confirmed by plant design basis performance analysis.
Allowable Value An allowable value is determined from the analytical limit by providing allowances for the specified or expected calibration capability, the accuracy of the instrumentation, and the measurement errors. The allowable value is the limiting value of the sensed process variable at which the trip setpoint may be found during instrument surveillance.
Instrumentation and Control                                                                                      3.4-9
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 Nominal Trip Setpoint The nominal trip setpoint value is calculated from the analytical limit by taking into account instrument drift in addition to the instrument accuracy, calibration capability, and the measurement errors. The nominal trip setpoint value is the limiting value of the sensed process variable at which a trip action will be set to operate at the time of calibration.
Signal Processing Devices in the Instrument Channel Within an instrument channel, there may exist other components or devices that are used to further process the electrical signal provided by the sensor (e.g., analog-to-digital converters, signal conditioners, temperature compensation circuits, and multiplexing and demultiplexing components). The worst-case instrument accuracy, calibration accuracy, and instrument drift contributions of each of these additional signal conversion components are separately or jointly accounted for when determining the characteristics of the entire instrument loop.
Not all parameters have an associated design basis analytical limit (e.g., main steamline radiation monitoring). An allowable value may be defined directly based on plant licensing requirements, previous operating experience or other appropriate criteria. The nominal trip setpoint is then calculated from this allowable value, allowing for instrument drift. Where appropriate, a nominal trip setpoint may be determined directly based on operating experience.
Procedures will be used that provide a method for establishing instrument nominal trip setpoint and allowable value. Because of the general characteristics of the instrumentation and processes involved, two different methods are applied:
(1)  Computational (2)  Historical data The computational method is used when sufficient information is available regarding a dynamic process and the associated instrumentation. The procedure takes into account channel instrument accuracy, calibration accuracy, process measurement accuracy, primary element accuracy, and instrument drift. If the resulting nominal trip setpoint and allowable value are not acceptable when checked to ensure that they will not result in an unacceptable level of trips caused by normal operational transients, then more rigorous statistical evaluation or the use of actual operational data may be considered.
Some setpoint values have been historically established as acceptable, both for regulatory and operational requirements. These setpoints have non-critical functions or are intended to provide trip actions related to gross changes in the process variable. The continued recommendation of these historically accepted setpoint values is another method for establishing nominal trip setpoint and allowable values. This approach is only valid where the governing conditions remain essentially unaltered from those imposed previously and where the historical values have been adequate for their intended functions.
3.4-10                                                                            Instrumentation and Control
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 The setpoint methodology plan requires that activities related to instrument setpoints be documented and stored in retrievable, auditable files.
Equipment Qualification (EQ)
Qualification of safety-related instrumentation and control equipment is implemented by a program that assures this equipment is able to complete its safety-related function under the environmental conditions that exist up to and including the time the equipment has finished performing that function. Qualification specifications consider conditions that exist during normal, abnormal, and design basis accident events in terms of their cumulative effect on equipment performance for the time period up to the end of equipment life.
The material discussed herein identifies an EQ program that addresses the spectrum of design basis environmental conditions that may occur in plant areas where I&C equipment is installed.
Not all safety-related I&C equipment will experience all of these conditions; the intent is that qualification be performed by selecting the conditions applicable to each particular piece of equipment and performing the necessary qualification.
As-built I&C components are environmentally qualified if they can withstand the environmental conditions associated with design basis events without loss of their safety functions for the time needed to be functional. Safety-related I&C components are designed to continue normal operation after loss of HVAC. The environmental conditions are as follows, as applicable to the bounding design basis events: Expected time-dependent temperature and pressure profiles, humidity, chemical effects, radiation, aging, seismic events, submergence, and synergistic effects which have a significant effect on equipment performance.
I&C equipment environmental qualification is demonstrated through analysis of the environmental conditions that would exist in the location of the equipment during and following a design basis accident and through a determination that the equipment is qualified to withstand those conditions for the time needed is functional. This determination may be demonstrated by:
(1)    Type testing of an identical item of equipment under identical or similar conditions with a supporting analysis to show that the equipment to be qualified.
(2)    Type testing of a similar item of equipment with a supporting analysis to show that the equipment is qualified.
(3)    Experience with identical or similar equipment under similar conditions with a supporting analysis to show that the equipment is qualified.
(4)    Analysis in combination with partial type test data that supports the analytical assumptions and conclusions to show that the equipment is qualified.
Instrumentation and Control                                                                                  3.4-11
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 The installed condition of safety-related I&C equipment is assured by a program whose objective is to verify that the installed configuration is bounded by the test configuration and test conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 3.4, Items 7 through 15, provides a definition of the inspections, tests and analyses, together with associated acceptance criteria, which will be used to demonstrate compliance with the above commitments for hardware and software development, electromagnetic compatibility, instrument setpoint methodology, and equipment qualification.
C. Diversity and Defense-in-Depth Considerations Subsection B discusses processes for developing hardware and software qualification programs that will assure a low probability of occurrence of both random and common-mode system failures for the installed ABWR I&C equipment. However, to address the concern that software design faults or other initiating events common to redundant, multi-divisional logic channels could disable significant portions of the plants automatic standby safety functions (the reactor protection system and engineered safety features systems) at the moment when these functions are needed to mitigate an accident, several diverse backup features are provided for the primary automatic logic:
* Manual scram and isolation by the operator in the main control room in response to diverse parameter indications.
* Core makeup water capability from the feedwater system, Control Rod Drive (CRD)
System, and condensate system, which are diverse from SSLC and the EMS.
* Availability of manual high pressure injection capability.
* Long term shutdown capability provided in a conventionally hardwired, 2-division, analog Remote Shutdown System (RSS); local displays of process variables in RSS are continuously powered and so are available for monitoring at any time.
Thus, to maintain protection system defense-in-depth in the presence of a postulated worst-case event (i.e., undetected, 4-division common mode failure of protection system communications or logic processing functions in conjunction with a large break LOCA), diversity is provided in the form of hardwired backup of reactor trip, diverse display of important process parameters, defense-in-depth arrangement of equipment, and other equipment diversity as outlined in the following table:
3.4-12                                                                                Instrumentation and Control
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 Diverse Backup Support for SSLC Equipment Diverse Features of      Functional Diversity      Defense-in-Depth Protection System        in Protection System        Configuration  Equipment Diversity (1) 2-button scram                H (2) Manual division                H trip (3) Reactor mode                  H switch placed in shutdown mode.
(4) Manual MSIV                    H closure (5) ATWS mitigation                D (6) Fail-safe RPS and                                      D fail-as-is ESF in separate processing channels (7) Non-Essential                                          D Multiplexing System (NEMS) independent and diverse from EMS (8) OLUs diverse                                                              H from software-based logic Instrumentation and Control                                                                        3.4-13
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 Diverse Backup Support for SSLC Equipment (Continued)
Diverse Features of    Functional Diversity      Defense-in-Depth Protection System      in Protection System        Configuration  Equipment Diversity (9) Independent                                                                H Displays (a) Reactor water level (b) Reactor water level low alarm (c) Drywell pressure (d) Drywell pressure high alarm (e) Reactor Water Cleanup System (CUW) isolation valve status (f) RCIC stream line isolation valve status (g) HPCF flow (10)Containment                                                                H Isolation (a) CUW line inboard isolation valve (b) RCIC steam line inboard isolation valve manual initiation 3.4-14                                                                      Instrumentation and Control
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 Diverse Backup Support for SSLC Equipment (Continued)
Diverse Features of        Functional Diversity      Defense-in-Depth Protection System          in Protection System        Configuration      Equipment Diversity (11)HPCF manual                                                                    H start in loop C (Division III)
(12)RSS with                                                                        H continuous display of monitored process parameters H = Function hardwired (not multiplexed) from sensor or control switch to actuator; control logic, if needed, is diverse from that of the primary protection system.
D = Function uses logic diverse from primary protection system but is not necessarily hardwired.
Diverse equipment can be in the form of digital devices, digital software-based devices, or non-digital as long as these devices are not subject to the same common mode failure as the primary protection system components.
Inspections, Tests, Analyses and Acceptance Criteria Table 3.4, Item 16, provides a definition of the inspection, tests and analyses, together with associated acceptance criteria, which will be used to demonstrate compliance with the above commitments for diverse backup support SSLC.
Instrumentation and Control                                                                                3.4-15
 
ABWR 3.4-16 SSLC MANUAL CONTROLS PROCESS SENSOR                                                    IN MAIN CONTROL ROOM OR TRIP LOGIC SIGNALS NEUTRON MONITORING SYSTEM PROCESS RADIATION SSLC                                                                                        CRD HCUs MONITORING SYSTEM SAFETY SYSTEM LOGIC AND CONTROL REACTOR PROTECTION                                                                                                REACTOR TRIP        Scram Pilot Valve SYSTEM                                                        (represents one of four divisions)
Solenoid Load Drivers NUCLEAR BOILER                                            PROCESSES SAFETY LOGIC TRIP DECISIONS SYSTEM                                                                                                  BACKUP REACTOR TRIP    Actuators for Scram Air Header Dump Valves LEAK DETECTION AND ISOLATION SYSTEM Initiate Scram-Following RESIDUAL HEAT                                                                                                                            (Control Rod Run-In) via RCIS REMOVAL SYSTEM REACTOR CORE ISOLATION COOLING 25A5675AA Revision 7 SYSTEM HIGH PRESSURE CORE                                                                                                                              MSIV Pilot Valve FLOODER SYSTEM                                                                                                                            Solenoid Load Drivers REACTOR BUILDING                                                                              CONTAINMENT COOLING WATER                                                                                  ISOLATION                          See Note 1 REACTOR SERVICE E
PCV Isolation Valves WATER                            M HVAC EMERGENCY                      S                                                                              E COOLING WATER EMERGENCY            M                    Final Control Elements of DIESEL GENERATOR                                                                              CORE COOLING          S                    Engineered Safety Features ELECTRICAL POWER DISTRIBUTION SYSTEM Design Control Document/Tier 1 STANDBY GAS TREATMENT SYSTEM ATMOSPHERIC                                        Interdivisional CONTROL SYSTEM Signal Transfer HVAC (Safety-Related)                                For 2-out-of-4 Coincidence Logic SUPPRESSION POOL Instrumentation and Control TEMPERATURE MONITORING SYSTEM Notes:
: 1. No PCV isolation trips or ECCS initiation outputs in Division IV.
Figure 3.4a Safety System Logic and Control (SSLC) Control Interface Diagram
 
ABWR Instrumentation and Control DIV. II                                DIV. III MANUAL      MANUAL TO ALARMS, DISPLAYS OR PROCESS COMPUTER                                                                                POWER                                  POWER DIV. I      DIV. I (TYPICAL FOR EACH P UNIT IN CONTROL ROOM)                              ISOLATION MANUAL SCRAM A MANUAL SCRAM B TRIP MULTIPLEXED INPUT Hardwired Outputs                                    MODE SWITCH                      MODE SWITCH                            TO OTHER (not multiplexed)                        RPS        IN SHUTDOWN                      IN SHUTDOWN            c              SCRAM DATA INTERFACE                                                                                          SCRAM DIV. II                    SCRAM DIV. III                          GROUPS OUTPUT                                                                        LD a
TRIPS TO              TRIPS FROM                                    LOGIC                                LD        TO OTHER DIV. II III IV TLUs  DIV. II III IV DTMs                                                                            SCRAM UNIT                                            GROUPS            a                  b LD        LD SENSORS                                                                                                                                        non-P                      c                  d TO OTHER SCRAM                  LD        LD RMU                        CMU            DTM                                  TLU                                              GROUPS                                                        d LD RPS/MSIV                                3/4                              MSIV b
LD            2/4 P                          P P                                  P                                    OUTPUT LOGIC  ...
UNIT                      TO "A" SOLENOIDS                  TO "B" SOLENOIDS DIVISION-OF-SENSORS          TRIP LOGIC OUTPUT                    ...
BYPASS CONTROL              BYPASS CONTROL non-P                        OF GROUP 1                        OF GROUP 1 SCRAM PILOT                      SCRAM PILOT MULTIPLEXED OUTPUT                                      VALVES                            VALVES DATA INTERFACE TRIPS TO TO MSIV PILOT DIV. II III SLUs SENSORS                                                                                                                                                                                                                      VALVE SOLENOID
                                                                                                                                                                                                    ...                                  ...                LOAD DRIVERS RMU DTM                                  SLU 1                                    CMU                                      RMU ESF 1                                                                                                                            ... ...
P                                                                                              2/4 P                                  P P                                  P 25A5675AA Revision 7 INTERLOCK INPUTS TRIPS DIV. II                                                                                          FROM LIMIT SWITCHES                        2/2 FROM DIV. III                                                                                          OR MCC CONTACTS SENSORS DTMs DIV. IV SLU 2 CMU                                      RMU HARDWIRED INPUTS 2/4 P                              P                                  P CONTROL TRIPS TO                                                                                                                                                    OUTPUTS ESF OUTPUT CHANNEL DIV. II III SLUs                                                                                                                                              TO AUTO/MANUAL BYPASS CONTROL ACTUATING DEVICES DTM                                  SLU 3                                    CMU                                      RMU ESF 2                                  2/4 P                                  P P                                  P INTERLOCK INPUTS TRIPS DIV. II                                                                                          FROM LIMIT SWITCHES                        2/2        ...
Design Control Document/Tier 1 FROM DIV. III                                                                                          OR MCC CONTACTS DATA ACQUISITION                                  DTMs DIV. IV SLU 4 CMU                                      RMU EMS                                                                                                                                                                              ... ...
2/4 P                                  P P
NOTES:
: 1. EMS ARRANGEMENT SHOWN IS A SIMPLIFIED EXAMPLE FOR ONE DIVISION. ACTUAL QUANTITY                        SENSOR TRIP DECISION            SYSTEM TRIP DECISION                                CONTROL OUTPUTS AND INTERCONNECTIONS OF RMUs AND CMUs WILL BE DETERMINED WITHIN SCOPE OF EMS DESIGN.
THERE ARE NO SLUs, CMUs, OR RMUs in DIV. IV.
: 2. A. OUTPUTS TO PROCESS COMPUTER, ALARMS OR DISPLAYS ON DEDICATED FIBER OPTIC DATA LINKS.
SSLC                                                                    EMS B. CONTROL SWITCH INPUTS TO SSLC NOT SHOWN.
C. INPUTS FROM NMS AND PRRM NOT SHOWN.
D. INTERDIVISIONAL COMMUNICATIONS USE FIBER OPTIC DATA LINKS.
3.4-17 Figure 3.4b Safety System Logic & Control Block Diagram
 
ABWR 3.4-18 LOCAL AREA                                                  MAIN CONTROL ROOM PLANT SENSORS                                                ATWS LOGIC & CONTROL NMS                                                RFC SRNM                                            Manual Permissive                                    SLC, FDWC ATWS LOGIC PROCESSORS 25A5675AA Revision 7
                                                                                          - Sensor Channel Trip Decision
                                                                                          - System Coincidence Trip Decision                                    SLC
                                                                                          - Control and Interlock Logic                                      INITIATION Reactor Water Level SSLC            - Division-of-Sensors Bypass                                        LOGIC NBS                                            logic Reactor Vessel Pressure              processing equipment FEEDWATER RUNBACK LOGIC SSLC Logic Processing for Other Design Control Document/Tier 1 Safety Systems Instrumentation and Control INTERDIVISIONAL SIGNAL TRANSFER FOR 2-out-of-4 COINCIDENCE LOGIC Notes:
: 1. Diagram represents one of four ATWS divisions.
: 2. Remaining ATWS functions are processed as part of Recirculation Flow Control System logic and Nuclear Boiler System logic.
Figure 3.4c Anticipated Transient Without Scram (ATWS) Control Interface Diagram
 
ABWR Instrumentation and Control HARDWARE AND SOFTWARE                                              SOFTWARE QUALITY ASSURANCE DESIGN ACTIVITIES                                    CONFIGURATION MANAGEMENT PLAN (revision control, change control)
VERIFICATION & VALIDATION        SOFTWARE MANAGEMENT PLAN PLAN                      (Software Life-cycle Phases)
SYSTEM DESIGN SPECIFICATION PLANNING EQUIPMENT DESIGN SPECIFICATION VERIFICATION Baseline Review HARDWARE/SOFTWARE DESIGN                                                                                  DESIGN SPECIFICATION                                                                            DEFINITION VERIFICATION Baseline Review HARDWARE DESIGN                                      SOFTWARE DESIGN SPECIFICATION                                        SPECIFICATION SOFTWARE DESIGN ELECTRICAL WIRING 25A5675AA Revision 7 DIAGRAM VERIFICATION Baseline Review HARDWARE                                            SOFTWARE IMPLEMENTATION                                      IMPLEMENTATION SOFTWARE CODING VERIFICATION INTEGRATION                                      Baseline Review Design Control Document/Tier 1 INTEGRATION I/O CHECK AND SOFTWARE LOGIC CHECK VERIFICATION Baseline Review SIMULATION TEST VALIDATION VALIDATION Baseline Review CHANGE CONTROL (re-entry point into life-cycle phas e determined by change procedure) 3.4-19 Figure 3.4d Integrated Hardware/Software Development Process
 
ABWR 3.4-20 Table 3.4 Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                        Acceptance Criteria Safety System Logic and Control
: 1. The equipment comprising SSLC is defined 1. Inspections of the as-built SSLC will be          1. The as-built SSLC conforms with the in Section 3.4(A). The equipment comprising  conducted.                                          description in Section 3.4(A). Diverse backup diverse backup support functions for SSLC is                                                    support equipment for SSLC conforms with defined in Section 3.4 (C).                                                                      the description in Section 3.4 (C).
: 2. Safety-related monitoring and trip logic for 2. Tests will be performed on as-installed SSLC 2. A test report exists which concludes that the the plant protection systems resides in SSLC    using simulated input signals. System          SSLC design basis performance equipment. SSLC integrates the automatic        outputs will be monitored to determine          requirements are met.
decision-making and trip logic functions and    operability of safety-related functions.
manual operator initiation functions associated with the safety actions of the 25A5675AA Revision 7 safety-related systems. SSLC generates the protective function signals that activate reactor trip and provide safety-related mitigation of reactor accidents.
: 3. The DTM, TLU, and OLUs for RPS and            3.                                            3.
MSIV in each of the four instrumentation
: a. Tests will be performed on SSLC by        a. The test signal exists only in the Class divisions are powered from their respective providing a test signal to the I&C            1E division under test in SSLC.
divisional Class 1E AC sources. The DTMs equipment in only one Class 1E division and SLUs for ESF 1 and ESF 2 in Divisions I, at a time.
Design Control Document/Tier 1 II, and III are powered from their respective divisional Class 1E DC sources, as are the      b. Inspection of the as-installed Class 1E    b. In SSLC, physical separation or ESF DTMs in Division IV. In SSLC,                  divisions in SSLC will be performed.          electrical isolation exists between Class independence is provided between Class 1E                                                        1E divisions. Physical separation or divisions and between Class 1E divisions                                                          electrical isolation exists between these Instrumentation and Control and non-Class 1E equipment.                                                                      Class 1E divisions and non-Class 1E equipment.
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                          Acceptance Criteria Safety System Logic and Control
: 4. SSLC provides the following bypass          4. Tests will be performed on the as-built SSLC 4. Results of bypass tests are as follows:
functions:                                      as follows:
: a. Division-of-sensors bypass                    a(1) Place one division of sensors in            a(1) No trip change occurs at the voted trip
: b. Trip logic output bypass                          bypass. Apply a trip test signal in              output of each TLU and SLU. Bypass place of each sensed parameter that              status is indicated in main control
: c. ESF output channel bypass                        is bypassed. At the same time, apply a            room.
redundant trip signal for each parameter in each other division, one division at a time. Monitor the voted trip output at each TLU and SLU.
25A5675AA Revision 7 Repeat for each division.
a(2) For each division in bypass, attempt        a(2) Each division not bypassed cannot be to place each other division in                  placed in bypass, as indicated at OLU division-of-sensors bypass, one at a              output; bypass status in main control time.                                            room indicates only one division of sensors is bypassed.
Design Control Document/Tier 1 3.4-21
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-22 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                            Acceptance Criteria Safety System Logic and Control
: 4. (continued)                                4. (continued)                                      4. (continued) b(1) Place one division in trip-logic-output        b(1) No trip change occurs at the trip output bypass. Operate manual auto-trip test                of the RPS OLU or MSIV OLU, switch. Monitor the trip output at the              respectively. Bypass status is indicated RPS OLU. Operate manual auto-                        in main control room.
isolation test switch. Monitor the trip output at the MSIV OLU. Repeat for each division.
25A5675AA Revision 7 b(2) For each division in bypass, attempt to          b(2) Each division not bypassed cannot be place the other divisions in trip-logic-              placed in bypass, as indicated at OLU output bypass, one at a time.                        output; bypass status in main control room indicates only one trip logic output is bypassed.
c(1)  Apply common test signal to any one          c(1) Monitored test output signal does not pair of dual-SLU signal inputs. Monitor            change state when power is removed test signal at voted 2-out-of-2 output in          from either SLU. Bypass status and loss RMU area. Remove power from one                    of power to SLU are indicated in main Design Control Document/Tier 1 SLU, restore power, then remove power              control room.
from other SLU. Repeat test for all pairs of dual SLUs in each division.
c(2) Disable auto-bypass circuit in bypass          c(2) Monitored test output signal is lost when Instrumentation and Control unit. Repeat test c(1), but operate                  power is removed from either SLU, but is manual ESF loop bypass switch for                    restored when manual bypass switch is each affected loop.                                  operated. Bypass status, auto-bypass inoperable, and loss of power to SLU are indicated in main control room.
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria Safety System Logic and Control
: 5. A portion of the anticipated transient without5. Tests will be conducted using simulated input 5. Four redundant output signals occur for each scram (ATWS) mitigation features is              signals for the process variables used by the    of the following ATWS mitigating functions provided by SSLC circuitry, with initiating      ATWS logic.                                      (one set in each of the four divisions of conditions as follows:                                                                            ATWS outputs) that lead to initiation of these For feedwater runback logic, reset attempts
: a. Initiation of automatic SLCS injection on    will be made before initiating test signals      functions:
high dome pressure and SRNM ATWS            drop below setpoints.                            a. Initiation of automatic SLCS injection on permissive for 3 minutes or greater, or                                                            high dome pressure and SRNM ATWS low reactor water level and SRNM ATWS                                                              permissive for 3 minutes or greater, or permissive for 3 minutes or greater.                                                              low reactor water level and SRNM ATWS permissive for 3 minutes or greater.
25A5675AA Revision 7
: b. Initiation of feedwater runback on high dome pressure and SRNM ATWS                                                                  b. Initiation of feedwater runback on high permissive for 2 minutes or greater.                                                              dome pressure and SRNM ATWS Reset is permitted only when both                                                                  permissive for 2 minutes or greater.
signals drop below the setpoints.                                                                  Reset is permitted only when both signals drop below the setpoints.
: 6. Main control room alarms, displays and      6. Inspections will be performed on the main      6. Alarms, displays and controls exist or can be controls provided for SSLC are as defined in    control room alarms, displays and controls        retrieved in the main control room as defined Section 3.4.                                    for SSLC                                          in Section 3.4.
Design Control Document/Tier 1 3.4-23
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-24 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                      Acceptance Criteria Hardware/Software Development
: 7. A quality assurance program encompassing 7. The program for quality assurance that      7. A quality assurance program is in place that software is employed as a controlled process encompasses software shall be reviewed.        defines controlled processes for software for software development hardware                                                          development, hardware integration, and final integration, and final product and system                                                  product and system testing. As a minimum, testing.                                                                                    the program requires a Software Management Plan, Configuration Management Plan and Verification and Validation Plan as described in the following items.
: 8. A Software Management Plan (SMP) shall        8. The Software Management Plan shall be  8. The Software Management Plan shall define:
25A5675AA Revision 7 be instituted which establishes that software    reviewed.                                  a. The organization and responsibilities for for embedded control hardware shall be development of the software design; the developed, designed, evaluated, and procedures to be used in the software documented per a design development development; the interrelationships process that addresses, for safety-related between software design activities; and software, software safety issues at each the methods for conducting software defined life-cycle phase of the software safety analyses.
development.
: b. That the software safety analyses to be The SMP shall state that the output of each                                                    conducted for safety-related software Design Control Document/Tier 1 defined life-cycle phase shall be documents                                                    applications shall:
that define the current state of that design                                                    (1) Identify software requirements phase and the design input for the next                                                              having safety-related implications.
design phase.
(2) Document the identified safety-Instrumentation and Control critical software requirements in the software requirements specification for the design.
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                Acceptance Criteria Hardware/Software Development
: 8. (continued)                              8. (continued)                          8b. (continued)
(3) Incorporate into the software design the safety-critical software functions specified in the software requirements specification.
(4) Identify in the coding and test of the developed software, those software modules which are safety-critical.
(5) Evaluate the performance of the 25A5675AA Revision 7 developed safety-critical software modules when operated within the constraints (including the effects of potential unintended functions) imposed by the established system requirements, software design, and computer hardware requirements.
(6) Evaluate software interfaces of safety-critical software modules.
Design Control Document/Tier 1 (7) Perform equipment integration and validation testing that demonstrate that safety-related functions identified in the design input requirements are operational.
3.4-25
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-26 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                  Inspections, Tests, Analyses                Acceptance Criteria Hardware/Software Development
: 8. (continued)                        8. (continued)                        8. (continued)
: c. The software engineering process, which is composed of the following life-cycle phases:
(1) Planning (2) Design Definition (3) Software Design (4) Software Coding 25A5675AA Revision 7 (5) Integration (6) Validation (7) Change control
: d. The Planning phase design activities, which shall address the following system design requirements and software development plans:
(1) Software Management Plan.
Design Control Document/Tier 1 (2) Software Configuration Management Plan.
(3) Verification and Validation Plan.
(4) Equipment design requirements.
Instrumentation and Control (5) Safety analysis of design requirements.
(6) Disposition of design and/or documentation nonconformances identified during this phase.
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                Acceptance Criteria Hardware/Software Development
: 8. (continued)                              8. (continued)                        8. (continued)
: e. The Design Definition phase design activities, which shall address the development of the following implementing equipment design and configuration requirements:
(1) Equipment schematic.
(2) Equipment hardware and software performance specification.
25A5675AA Revision 7 (3) Equipment users manual.
(4) Data communications protocol, including timing analysis and test methods.
(5) Safety analysis of the developed design definition.
(6) Disposition of design and/or documentation nonconformances identified during this phase.
Design Control Document/Tier 1 3.4-27
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-28 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                Acceptance Criteria Hardware/Software Development
: 8. (continued)                              8. (continued)                        8. (continued)
: f. The Software Design phase, which shall address the design of the software architecture and program structure elements, and the definition of software module functions:
(1) Software Design Specification.
(2) Safety analysis of the software design.
25A5675AA Revision 7 (3) Disposition of design and/or documentation nonconformances identified during this phase.
: g. The Software Coding phase, which shall address the following software coding and testing activities of individual software modules:
(1) Software source code.
(2) Software module test reports.
Design Control Document/Tier 1 (3) Safety analysis of the software coding.
(4) Disposition of nonconformances identified in this phases design Instrumentation and Control documentation and test results.
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                  Acceptance Criteria Hardware/Software Development
: 8. (continued)                              8. (continued)                        8. (continued)
: h. The Integration phase, which shall address the following equipment testing activities that evaluate the performance of the software when installed in hardware prototypical of that defined in the Design Definition phase:
(1) Integration test reports.
(2) Safety analysis of the integration test 25A5675AA Revision 7 results.
(3) Disposition of nonconformances identified in this phases design documentation and test results.
: i. The Validation phase, which comprises the development and implementation of the following documented test plans and procedures:
(1) Validation test plans and Design Control Document/Tier 1 procedures.
(2) Validation test reports.
(3) Description of as-tested software.
(4) Safety analysis of the validation test results.
(5) Disposition of nonconformances identified in this phases design documentation and test results 3.4-29
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-30 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                      Acceptance Criteria Hardware/Software Development
: 8. (continued)                                8. (continued)                              8i. (continued)
(6) Software change control procedures.
: j. The Change Control phase, which begins with the completion of validation testing, and addresses changes to previously validated software and the implementation of the established software change control procedures.
25A5675AA Revision 7
: 9. A Configuration Management Plan (CMP)      9. The Configuration Management Plan shall  9. The Configuration Management Plan shall shall be instituted that establishes the      be reviewed.                                define:
methods for maintaining, throughout the
: a. The specific product or system scope to software design process, the design which it is applicable.
documentation, procedures, evaluated software, and the resultant as-installed                                                    b. The organizational responsibilities for software.                                                                                      software configuration management.
: c. Methods to be applied to:
(1) Identify design interfaces.
Design Control Document/Tier 1 (2) Produce software design documentation.
(3) Process changes to design interface documentation and software design documentation.
Instrumentation and Control
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                Acceptance Criteria Hardware/Software Development
: 9. (continued)                              9. (continued)                        9c. (continued)
(4) Process corrective actions to resolve deviations identified in software design and design documentation, including notification to end user of errors discovered in software development tools or other software.
(5) Maintain status of design interface documentation and developed software design documentation.
25A5675AA Revision 7 (6) Designate and control software revision status. Such methods shall require that software code listings present direct indication of the software code revision status.
: d. Methods for, and the sequencing of, reviews to evaluate the compliance of software design activities with the requirements of the CMP.
Design Control Document/Tier 1
: e. The configuration management of tools (such as compilers) and software development procedures.
: f. Methods for the dedication of commercial software for safety-related usage.
3.4-31
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-32 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                            Acceptance Criteria Hardware/Software Development
: 9. (continued)                                    9. (continued)                                      9. (continued)
: g. Methods for tracking error rates during software development, such as the use of software metrics.
: h. The methods for design record collection and retention.
: 10. A Verification and Validation Plan (V&VP)    10. The Verification and Validation Plan shall be  10. The Verification and Validation Plan shall shall be developed which establishes that        reviewed.                                          define:
developed software shall be subjected to
: a. That baseline reviews of the software 25A5675AA Revision 7 structured and documented verification development process are to be reviews and validation testing, including conducted during each phase of the testing of the software integrated into the software development life cycle.
target hardware.
: b. The scope and methods to be used in the baseline reviews to evaluate the implemented design, design documentation, and compliance with the requirements of the Software Management Plan and Configuration Management Plan.
Design Control Document/Tier 1
: c. The requirements for use of commercial software and commercial development tools for safety-related applications and that such use is a controlled and Instrumentation and Control documented procedure.
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                  Acceptance Criteria Hardware/Software Development
: 10. (continued)                              10. (continued)                        10. (continued)
: d. That verification shall be performed as a controlled and documented evaluation of the conformity of the developed design to the documented design requirements at each phase of baseline review.
: e. That validation shall be performed through controlled and documented testing of the developed software as 25A5675AA Revision 7 installed in the target hardware that demonstrates compliance of the software with the software requirements specifications and compliance of the device(s) under test with the system design specifications.
: f. That for safety-related software, verification reviews and validation testing are to be conducted by personnel who are knowledgeable in the technologies Design Control Document/Tier 1 and methods used in the design, but who did not develop the software design to be reviewed and tested.
3.4-33
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-34 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                Acceptance Criteria Hardware/Software Development
: 10. (continued)                              10. (continued)                        10. (continued)
: g. That for safety-related software, design verification reviews shall be conducted as part of the baseline reviews of the design material developed during the Planning through Integration phases of the software development life-cycle (as defined in Criterion 8b, above), and that validation testing shall be conducted as 25A5675AA Revision 7 part of the baseline review of the Validation phase of the software development life-cycle.
: h. That validation testing shall be conducted per a documented test plan and procedure.
: i. That for non-safety-related software development, verification and validation shall be performed through design reviews conducted as part of the Design Control Document/Tier 1 baseline reviews completed at the end of the phases in the software development life cycle. These design reviews shall be performed by personnel knowledgeable Instrumentation and Control in the technologies and methods used in the design development.
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                    Acceptance Criteria Hardware/Software Development
: 10. (continued)                                10. (continued)                            10. (continued)
: j. The products which shall result from the baseline reviews conducted at each phase of the software development life-cycle; and that the defined products of the baseline reviews and the V&V Plan shall be documented and maintained under configuration management.
: k. The methods for identification, closure, 25A5675AA Revision 7 and documentation of design and/or design documentation nonconformances.
: l. That the software development is not complete until the specified V&V activities are complete and design documentation is consistent with the developed software.
: 11. Software development shall be performed in 11. Review software development results. 11. Software development has been completed Design Control Document/Tier 1 accordance with the software management                                                  as defined in the SMP, CMP, and V&V plan.
plan, configuration management plan, and                                                  Noncompliance with the SMP, CMP, and V&V plan.                                                                                V&V plan may occur during implementation of these plans, provided that corrective action is taken for any such noncompliances.
3.4-35
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-36 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                          Acceptance Criteria Electromagnetic Compatibility
: 12. Electrical and electronic components in the 12. The EMC compliance plan will be reviewed. 12. An EMC compliance plan is in place. The systems listed below are qualified for the                                                      plan requires, for each system qualified, anticipated levels of electrical interference at                                                system documentation that includes the installed locations of the components                                                      confirmation of component and system according to an established plan:                                                              testing for the effects of high electrical field
: a. Safety System Logic and Control                                                              conditions and current surges. As a minimum, the following information is
: b. Essential Multiplexing System                                                                documented in a qualification file and subject
: c. Non-Essential Multiplexing System                                                          to audit:
: d. Other microprocessor-based, software                                                          a. Expected performance under test 25A5675AA Revision 7 controlled systems or equipment                                                                  conditions for which normal system operation is to be ensured.
The plan is structured on the basis that electromagnetic compatibility (EMC) of I&C                                                      b. Normal electrical field conditions at the equipment is verified by factory testing and                                                        locations where the equipment must site testing of both individual components                                                          perform as above.
and interconnected systems to meet EMC                                                          c. Testing methods used to qualify the requirements for protection against the                                                              equipment, including:
effects of:
(1) Types of test equipment.
: a. Electromagnetic Interference (EMI)
Design Control Document/Tier 1 (2) Range of normal test conditions.
: b. Radio Frequency Interference (RFI)
(3) Range of abnormal test conditions
: c. Electrostatic Discharge (ESD)                                                                        for expected transient environment.
: d. Electrical surge [Surge Withstand Instrumentation and Control Capability (SWC)]
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                Acceptance Criteria Electromagnetic Compatibility
: 12. (continued)                                12. (continued)                        12.(continued)
(4) Location of testing and exact configuration of tested components and systems, including interconnecting cables, connections to electrical power distribution system, and connections to interfacing devices used during normal plant operation.
: d. Test results that show the component or 25A5675AA Revision 7 system is qualified for its application and remains qualified after being subjected to the range of normal and abnormal test conditions specified above.
The plan establishes separate test regimes for each element of EMC, using the following approaches:
: a. EMI and RFI Protection. An EMC compliance plan for each component or Design Control Document/Tier 1 system identified in the design commitment includes tests to ensure that equipment performs its functions in the presence of the specified EMI/RFI electrical noise environment, including the low range of the EMI spectrum, without equipment damage, spurious actuation, or inhibition of functions.
3.4-37
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-38 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                Acceptance Criteria Electromagnetic Compatibility
: 12. (continued)                                12. (continued)                        12a. (continued)
As part of the pre-operational test program, the EMC compliance plan calls for each system to be subjected to EMI/RFI testing. Tests cover potential EMI and RFI susceptibility over four different paths:
(1) Power feed lines (2) Input signal lines 25A5675AA Revision 7 (3) Output signal lines (4) Radiation The test program includes sensitivity of components identified in the design commitment to radiation from plant communication transmitters and receivers.
: b. ESD Protection. An EMC compliance plan for each component or system Design Control Document/Tier 1 identified in the design commitment includes tests to ensure that equipment performs its functions in the presence of the specified ESD environment without Instrumentation and Control equipment damage, spurious actuation, or inhibition of functions.
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses                Acceptance Criteria Electromagnetic Compatibility
: 12. (continued)                                12. (continued)                        12b. (continued)
The plan is structured on the basis that ESD protection is confirmed by factory tests that determine the susceptibility of instrumentation and control equipment to electrostatic discharges.
The EMC compliance plan includes standards, conventions, design considerations, and test procedures to ensure ESD protection of the plant 25A5675AA Revision 7 instrumentation and control equipment.
The plan requires test documentation confirming that, for each component tested, the following conditions have been met:
(1) No change in output signal status was observed during the test.
(2) The equipment performed its normal functions after the test.
Design Control Document/Tier 1 3.4-39
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-40 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                          Inspections, Tests, Analyses              Acceptance Criteria Electromagnetic Compatibility 12.(continued)                                  12. (continued)                          c. SWC Protection. An EMC compliance plan for each component or system identified in the design commitment includes tests to ensure that equipment performs its functions for the specified SWC environment without equipment damage, spurious actuation, or inhibition of functions.
The EMC compliance plan includes standards, conventions, design 25A5675AA Revision 7 considerations, and test procedures to ensure SWC protection of the plant instrumentation and control equipment.
The plan is structured on the basis that SWC protection is confirmed by factory tests that determine the surge withstand capability of the plant instrumentation and control equipment.
The plan documents the level of Design Control Document/Tier 1 compliance of each system with the grounding and shielding practices of the standards specified under this certified design commitment.
Instrumentation and Control
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                            Acceptance Criteria Setpoint Methodology
: 13. Setpoints for initiation of safety-related    13. Inspections will be performed of the setpoint 13. The setpoint methodology plan is in place.
functions are determined, documented,              methodology plan used to determine,              The plan generates requirements for:
installed and maintained using a process that      document, install, and maintain instrument
: a. Documentation of data, assumptions, establishes a plan for:                            setpoints.
and methods used in the bases for
: a. Specifying requirements for documenting                                                              selection of trip setpoints.
the bases for selection of trip setpoints.
: b. Consideration of instrument channel
: b. Accounting for instrument inaccuracies,                                                              inaccuracies (including those due to uncertainties, and drift.                                                                          analog-to-digital converters, signal
: c. Testing of instrumentation setpoint                                                                  conditioners, temperature compensation 25A5675AA Revision 7 dynamic response.                                                                                  circuits, and multiplexing and demultiplexing components), instrument
: d. Replacement of setpoint-related                                                                      calibration uncertainties, instrument drift, instrumentation.                                                                                    and uncertainties due to environmental The setpoint methodology plan requires that                                                            conditions (temperature, humidity, activities related to instrument setpoints be                                                          pressure, radiation, EMI, power supply documented and stored in retrievable,                                                                  variation), measurement errors, and the auditable files.                                                                                        effect of design basis event transients are included in determining the margin between the trip setpoint and the safety Design Control Document/Tier 1 limit.
: c. The methods used for combining uncertainties.
: d. Use of written procedures for preoperational testing and tests performed to satisfy the Technical Specifications.
3.4-41
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-42 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                            Inspections, Tests, Analyses                          Acceptance Criteria Setpoint Methodology
: 13. (continued)                                  13. (continued)                                  13. (continued)
: e. Documented evaluation of replacement instrumentation which is not identical to the original equipment.
: 14. Qualification of safety-related I&C equipment 14. A review will be conducted of the equipment 14. An I&C equipment qualification program is in is implemented by a program that assures          qualification program.                          place. Documentation for the I&C EQ this equipment is able to complete its safety-                                                    program is recorded in a product qualification related function under the environmental                                                          file that includes a list of safety-related I&C conditions that exist up to and including the                                                    equipment accompanied by the following time the equipment has finished performing                                                        I&C equipment information:
25A5675AA Revision 7 that function. Qualification specifications
: a. Performance specifications under consider conditions that exist during normal, conditions existing during and after abnormal, and design basis accident events design basis accidents. These include in terms of their cumulative effect on voltage, frequency, load, and other equipment performance for the time period electrical characteristics that assure up to the end of equipment life.
specified equipment performance.
: b. Environmental conditions at the location where the equipment is installed. These conditions include:
Design Control Document/Tier 1 (1) Number and /or duration of equipment functional and test cycles/events.
(2) Process fluid conditions (where Instrumentation and Control applicable to the I&C equipment)
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                Acceptance Criteria Equipment Qualification
: 14. (continued)                              14. (continued)                        14. (continued)
(3) Voltage, frequency, load, and other electrical characteristics of the equipment.
(4) Dynamic loads associated with seismic events.
(5) Dynamic loads associated with hydrodynamic conditions.
(6) System transients and other 25A5675AA Revision 7 vibration inducing events.
(7) Pressure, temperature, humidity.
(8) Chemical and radiation environments.
(9) Electromagnetic compatibility (10) Aging.
(11) Submergence (if any).
(12) Consideration of synergistic effects Design Control Document/Tier 1 that have significant effect on equipment performance.
(13) Consideration of margins for unquantified uncertainty.
: c. One (or a combination) of the following testing methods used to qualify the equipment:
3.4-43
 
Table 3.4 Instrumentation and Control (Continued)
ABWR 3.4-44 Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                        Inspections, Tests, Analyses                Acceptance Criteria Equipment Qualification
: 14. (continued)                              14. (continued)                        14.(continued)
(1) Type testing of an identical item of equipment under identical or similar conditions with a supporting analysis to show that the equipment to be qualified is acceptable.
(2) Type testing of a similar item of equipment with a supporting analysis to show that the equipment to be qualified is acceptable.
25A5675AA Revision 7 (3) Experience with identical or similar equipment under similar conditions with a supporting analysis to show that the equipment to be qualified is acceptable.
(4) Analysis in combination with partial type test data that supports the analytical assumptions and conclusions.
Design Control Document/Tier 1
: d. Documented results of the qualification that show the equipment performs its safety function when subjected to the conditions predicted to be present when Instrumentation and Control it must perform its safety function up to the end of its qualified life.
 
Table 3.4 Instrumentation and Control (Continued)
ABWR Instrumentation and Control Inspections, Tests, Analyses and Acceptance Criteria Design Commitment                              Inspections, Tests, Analyses                              Acceptance Criteria Equipment Qualification
: 15. A program exists whose objective is to verify 15. A review will be conducted of the program          15. A program for as-built verification of safety-that the installed configuration of safety-      established for as-built verification of safety-      related I&C equipment is in place and related I&C equipment is bounded by the test      related I&C equipment.                                contains requirements for a documented configuration and test conditions or that an                                                            evaluation that the installed configuration is analysis exists which concludes that any                                                                bounded by the test configuration and differences will not affect the safety function                                                          conditions or that an analysis exists which of the I&C equipment.                                                                                    concludes that any differences will not affect the safety function of the I&C equipment.
: 16. Diversity is provided, as described in Section 16.                                            16.
25A5675AA Revision 7 3.4C, in the form of hardwired backup of
: a. Tests will be performed using simulated      a. Item 5, Section 3.4C: Refer to Item 4 of reactor trip, diverse display of important input signals for items 5, 9, and 11 in        Table 3.4 for results of ATWS tests.
process parameters, defense-in-depth Section 3.4C. For items 9 and 11 only,          Item 9, Section 3.4C: Each independent arrangement of equipment, and equipment turn off power to SSLC equipment in four        display indicates its specified parameter diversity.
divisions.                                      or responds to its specified alarm setpoint as tabulated in Section 3.4C.
Item 11, Section 3.4C: HPCF system initiation signals that duplicate those produced by SSLC are produced at the outputs of the hardwired, diverse signal Design Control Document/Tier 1 path.
: b. Inspection of the as-installed                    b. The features listed as items 1, 2, 3, 4, 6, configuration of items 1,2, 3, 4, 6, 7, 8,          7, 8, 10, and 12 in Section 3.4C, are 10, and 12 in Section 3.4C, will be                  implemented as hardwired, diverse, and performed.                                          independent of SSLC, as specified in the table.
3.4-45
 
25A5675AA Revision 7 ABWR                                                                            Design Control Document/Tier 1 3.5 Initial Test Program Design Description The ABWR Initial Test Program (ITP) is a program that will be conducted following completion of construction and construction-related inspections and tests and extends to commercial operation. The test program will be composed of preoperational and startup test phases. The general objective of the ITP is to confirm that performance of the as-built facility is in compliance with the design characteristics used for SSAR safety evaluations.
The preoperational test phase of the ITP will consist of those test activities conducted prior to fuel loading. Preoperational testing will be conducted to demonstrate proper performance of structures, systems, components, and design features in the assembled plant. Tests will include, as appropriate, logic and interlocks test, control and instrumentation functional tests, equipment functional tests, system operational test, and system vibration and expansion measurements.
The startup test phase of the ITP will begin with fuel loading and extends to commercial operation. The primary objective of the startup phase testing will be to confirm integrated plant performance with the nuclear fuel in the reactor pressure vessel and the plant at various power levels. Startup phase testing will be conducted at five test conditions during power ascension:
open vessel, heatup, low power, mid-power, and high power. The following tests will be conducted during power operation testing:
(1)  Core performance analysis.
(2)  Steady-state testing.
(3)  Control system tuning and demonstration.
(4)  Minor and major transients.
Testing during all phases of the ITP will be conducted using step by step written procedures to control the conduct of each test. Such test procedures will delineate established test methods and applicable acceptance criteria. The test procedures will be developed from preoperational and startup test specifications. Approved test procedures will be made available to the NRC approximately 60 days prior to their intended use for preoperational tests and 60 days prior to scheduled fuel loading for startup phase tests. The preoperational and startup test specifications will also be made available to the NRC. Administratively, the ITP will be controlled in accordance with a startup administrative manual. This manual will contain the administrative requirements that govern the conduct of test program, review, evaluation and approval of test results, and test records retention.
Initial Test Program                                                                                            3.5-1
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 Inspections, Tests, Analyses and Acceptance Criteria This section represents a commitment that combined operating license applicants referencing the certified design will implement an ITP that meets the objectives presented above.
Inspections, tests, analyses and acceptance criteria (ITAAC) aimed at verification of ITP implementation are neither necessary nor required.
3.5-2                                                                                Initial Test Program
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 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 fuel load. 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 risk 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 consistent with 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 associated acceptance criteria, which will be performed for Advanced Boiling Water Reactor (ABWR)D-RAP.
Design Reliability Assurance Program                                                                          3.6-1
 
ABWR 3.6-2                                                                          Table 3.6 Design Reliability Assurance Program 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.
: a. Documentation exists that describes the objectives; the process used to evaluate and scope, purpose, and objectives of D-prioritize the structures, systems and RAP used during plant design, and components (SSCs); and the list of SSCs concludes that the detailed design of designated as risk-significant. For those risk-significant SSCs is consistent with SSCs designated as risk-significant, the the D-RAP Design Description.
process used to determine dominant failure modes considered industry experience,                                                          b. Documentation exists and concludes analytical models, and applicable                                                                that the process (probabilistic, requirements. Also, for those SSCs                                                                deterministic, or other methods) used to designated as risk-significant, the key                                                          evaluate and prioritize the SSCs in the 25A5675AA Revision 7 assumptions and risk insights considered                                                          design is based on the risk- significance operations, maintenance, and monitoring                                                          of the SSCs.
activities.                                                                                    c. A list of SSCs exists that is based on the risk-significance of SSCs.
: d. For those SSCs designated as risk significant:
i)    Documentation exists and concludes that the process to determine dominant failure modes considered Design Control Document/Tier 1 industry experience, analytical Design Reliability Assurance Program models, and applicable requirements.
ii)  Documentation exists and concludes that the key assumptions and risk insights from probabilistic, deterministic, or other methods considered operations, maintenance, and monitoring activities.
 
25A5675AA Revision 7 ABWR                                                                            Design Control Document/Tier 1 4.0 Interface Requirements The interface requirements defined in this section are similar in nature to the Design Commitments identified in the tables of Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) provided in Section 2.0 above. In particular, the following matters are addressed in one or more of the interface requirements: supply of cooling and makeup water, heat removal for water-cooled systems, separation and independance of divisions, Remote Shutdown System (RSS) controls for system operation, seismic capability, criteria for electrical power systems monitors, automatic initiation of system operation, and flood limiting features. Each of these design features/design characteristics is also discussed in one or more of the Design Commitments and their corresponding ITAAC in Section 2.0.
An applicant for a combined operating license (COL) that references the ABWR Certified Design must provide design features or characteristics that comply with the interface requirements for the ABWR design and ITAAC for the site-specific portions of the facility design, in accordance with 10 CFR 52.79 (c). Because the interface requirements for the ABWR design are similar to the ABWR Design Descriptions in Section 2.0, for which ITAAC have been developed, compliance with the interface requirements is verifiable through inspection, testing, or analysis. Therefore,there is justification that a COL applicant will be able to develop ITAAC to verify compliance with the design features or characteristics that implement the interface requirements.
Interface Requirements                                                                                        4.0-1
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 4.1 Ultimate Heat Sink Interface Requirements The Ultimate Heat Sink (UHS) removes the heat load of the Reactor Service Water (RSW)
System during of plant operation. The UHS is not within the Certified Design. The UHS will meet the following requirements:
(1)  Provide cooling water to the RSW System for normal plant operation and to permit safe shutdown and cooldown of the plant and maintain the plant in a safe shutdown condition for design basis events.
(2)  Makeup water for the UHS shall not be required for at least 30 days following a design basis accident.
(3)  Any active safety-related system, structure, or components within the UHS shall have three divisions powered by their respective Class 1E divisions. Each division shall be physically separated and electrically independent of the other divisions.
The site specific design of the UHS demonstrates that sufficient capacity is maintained to support RSW System cooling capacity following postulated aircraft impact strike locations on the UHS. Divisional separation of the RSW components that interface with the UHS is required in accordance with 10 CFR 50.150.
(4)  UHS System Divisions A and B components shall have control interfaces with the Remote Shutdown System (RSS) as required to support UHS operation during RSS design basis conditions.
(5)  Be classified as Seismic Category I.
Ultimate Heat Sink                                                                                        4.1-1
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 4.2 Offsite Power System Interface Requirements No entry. Covered in Section 2.12.1.
Offsite Power System                                                                        4.2-1
 
25A5675AA Revision 7 ABWR                                                                    Design Control Document/Tier 1 4.3 Makeup Water Preparation System Interface Requirements The Makeup Water Preparation (MWP) System provides makeup water to the plant via the Makeup Water (purified) (MUWP) System and the Potable and Sanitary Water System. The MWP System is not within the Certified Design. A site-specific MWP System will be designed for any facility which has adopted the Certified Design to provide demineralized water to the MUWP System.
Makeup Water Preparation System                                                                        4.3-1
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 4.4 Potable and Sanitary Water System Interface Requirements Covered in Section 2.11.23.
Potable and Sanitary Water System                                                        4.4-1
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 4.5 Reactor Service Water System Interface Requirements No entry. Covered in Section 2.11.9.
Reactor Service Water System                                                                4.5-1
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 4.6 Turbine Service Water System Interface Requirements Covered in Section 2.11.10.
Turbine Service Water System                                                              4.6-1
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 4.7 Communication System Interface Requirements Covered in Section 2.12.16.
Communication System                                                                    4.7-1
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 4.8 Site Security Provisions for site security are not within the Certified Design and will be provided by each licensee on a site-specific basis.
Site Security                                                                                            4.8-1
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 4.9 Circulating Water System Covered in Section 2.10.23.
Circulating Water System                                                                  4.9-1
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 4.10 Heating, Ventilating and Air Conditioning System Covered in Section 2.15.5.
Heating, Ventilating and Air Conditioning System                                              4.10-1
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 5.0 Site Parameters This section provides a definition of the site parameters used as the basis for the Certified Design.
Site Parameters                                                                                            5.0-1
 
25A5675AA Revision 7 ABWR                                                                          Design Control Document/Tier 1 Table 5.0 ABWR Site Parameters Maximum Ground Water Level:                                Severe Wind:                  Basic Wind Speed:
61.0 cm below grade                                    177 km/h(1)/197 km/h(2)
Maximum Flood (or Tsunami) Level:                          Extreme Wind 30.5 cm below grade Tornado Precipitation (for Roof Design):
* Maximum tornado wind speed:              483 km/h(11)
(3)
* Maximum rainfall rate:                    49.3 cm/h
* Maximum pressure drop:                  13.827 kPaD
* Maximum snow load:                          2.394 kPa
* Missile spectra:                        Spectrum I(4)
Hurricane
* Maximum hurricane wind speed:            257 km/h(8)
* Missile spectra:                        Spectrum I(4)
Ambient Design Temperature:                                Soil Properties:
1% Exceedance Values
* Minimum static bearing
* Maximum:                              37.8&deg;C dry bulb    capacity:                                    718.20 kPa 25&deg;C wet bulb (coincident)
* Minimum shear wave velocity:                  305 m/s(6) 26.7&deg;C wet bulb (non-coincident)
* Liquefaction potential:                None at plant site
* Minimum:                                    -23.3&deg;C-                                            resulting from site 0% Exceedance Values (Historical Limit)                                                        specific SSE ground
* Maximum:                              46.1&deg;C dry bulb                                                        motion 26.7&deg;C wet bulb (coincident)
* Minimum Dynamic Bearing 27.2&deg;C wet bulb (non-coincident)      Capacity:                                      2700 kPa
* Minimum:                                        -40&deg;C
* Maximum Settlement(9):                              75mm
* Maximum Foundation Angular Exclusion Area Boundary (EAB): An area whose                Distortion:                                      1/750(10) boundary has a Chi/Q less than or equal to 1.37x10-3s/m3.                                            Seismology:
* SSE response spectra:    See Figures 5.0a and 5.0b(7)
Meteorological Dispersion (Chi/Q):
* Maximum 2-hour 95% EAB                  1.37 x 10-3 s/m3
* Maximum 2-hour 95% LPZ                  4.11 x 10-4 s/m3
* Maximum annual average (8760 hour) LPZ                        1.17 x 10-6 s/m3 (1) Fastest-mile (203 km/h 3-second gust); 50-year recurrence interval; value to be utilized for design of non-safety-related structures only.
(2) Fastest-mile (224 km/h 3-second gust); 100-year recurrence interval; value to be utilized for design for safety-related structures only.
(3) Maximum value for 1 hour over 2.6 km2 probable maximum precipitation (PMP) with ratio of 5 minutes to 1 hour PMP of 0.32. Maximum short-term rate: 15.7cm/5 min.
5.0-2                                                                                                    Site Parameters
 
25A5675AA Revision 7 ABWR                                                                                Design Control Document/Tier 1 (4) Spectrum I missiles consist of a massive high kinetic energy missile which deforms on impact, a rigid missile to test penetration resistance, and a small rigid missile of a size sufficient to just pass through any openings in protective barriers. These missiles consists of an 1810 kg automobile, a 130 kg, 20 cm diameter armor piercing artillery shell, and a 2.54 cm diameter solid steel sphere. These missiles have a horizontal tornado missile velocity of 35% of the maximum tornado wind speed and a horizontal hurricane missile velocity of 59% of the maximum hurricane wind speed. These missiles have a vertical tornado missile velocity of 70% of the horizontal tornado missile velocity (with the exception of the solid steel sphere) and a vertical hurricane missile velocity of 26 m/s. The solid steel sphere has a vertical tornado missile velocity of 35% of the maximum tornado wind speed. The automobile missile is considered to impact at all altitudes less than 9.14 m (30 feet) above all plant grade levels within 0.8 km (0.5 mile) of the plant structures. The armor piercing artillery shell and solid steel sphere are considered to impact the full height of the structure. The first two missiles are assumed to impact at normal incidence, the last to impinge upon barrier openings in the most damaging directions.
(5) At foundation level of the reactor and control buildings.
(6) This is the minimum shear wave velocity at low strains after the soil property uncertainties have been applied.
(7) Free-field, at plant grade elevation.
(8) Fastest-mile wind speed. This corresponds to 286.5 km/h 3-second gust wind speed per RG 1.221 measured at 10 m above ground over open terrain.
(9) Settlement is long term (post construction) value.
(10) Angular distortion is defined as the slope between two adjacent columns. Angular distortion is long term (post construction) value.
(11) Maximum tornado wind speed is in fastest 1/4-mile. The corresponding 3-second gust wind speed is 483 km/h.
Site Parameters                                                                                                          5.0-3
 
25A5675AA Revision 7 ABWR                                                                                                                  Design Control Document/Tier 1 CRITICAL DAMPING: 2%, 3%, 4%, 5%, AND 7%
1000 cm 0
10 100 cm                                                              10 10                                                                      g PSEUDOSPRECTAL VELOCITY (cm/s) cm                              1 g
1 10 0.
1      cm g
: 0. 1
: 0.                                                cm 01                                          1 g                                  0. 0 1
: 0.                                                                                cm 00                                                                          00 1                                                                          1 g                                                                0.
0.
00 01 g
0.1 0.1                                1                                    10                                  100 FREQUENCY (Hz)
Figure 5.0a Horizontal Safe Shutdown Earthquake Design Spectra 5.0-4                                                                                                                                                Site Parameters
 
25A5675AA Revision 7 ABWR                                                                                                                Design Control Document/Tier 1 CRITICAL DAMPING: 2%, 3%, 4%, 5%, AND 7%
1000 cm 0
10 100 cm                                                          10 g
PSEUDOSPRECTAL VELOCITY (cm/s) 10 cm                              1 g
1 10 0.
1      cm g
: 0. 1
: 0.                                            cm 01 g                                  0. 01 1
: 0.                                                                            cm 00                                                                      00 1                                                                      1 g                                                            0.
0.
00 01 g
0.1 0.1                                1                                    10                              100 FREQUENCY (Hz)
Figure 5.0b Vertical Safe Shutdown Earthquake Design Spectra Site Parameters                                                                                                                                      5.0-5
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Appendix A Legend for Figures For a number of the systems presented in Section 2, figures depicting the Basic Configuration of the systems have been provided to help facilitate the Design Description. For I&C systems, the figures represent a diagram of significant aspects of the logic of the system. For other systems and buildings, these figures represent a functional diagram, representation, or illustration of design-related information. Unless otherwise specified explicitly, these figures are not necessarily indicative of the scale, location, dimensions, shape, or spatial relationships of as-built structures, systems, and components. In particular, the as-built attributes of structures, systems and components may vary from the attributes depicted on these figures, provided that those safety functions discussed in the Design Description are not adversely affected.
The figures contain information that uses the following conventions:
Mechanical Equipment Line classification:
Figure Designation ASME Code Class 1                                                1 ASME Code Class 2                                                2 ASME Code Class 3                                                3 Legend for Figures                                                                                    Appendix A-1
 
25A5675AA Revision 7 ABWR                                                                Design Control Document/Tier 1 Figure Designation Non-ASME Code/                                            NNS Non-Nuclear Safety Other Line Type:                                          This legend can be used for pneumatic lines when needed for clarity. ASME Code class for such lines is defined on the system figure.
Classification/System Boundaries:
The following is a self-explanatory example of how ASME Code class change and system boundary are identified on the figures:
Instrumentation:
Conductivity monitor                                    CM Differential pressure indicator                        dP Display and/or control interface with RSS              R Flow element                                            FE Hydrogen analyzer                                      HE Level controller                                        LC Level detector                                          L Moisture element                                        ME Pressure element                                        P Radiation element                                      RE Speed detector                                          S Appendix A-2                                                                            Legend for Figures
 
25A5675AA Revision 7 ABWR                                                                                Design Control Document/Tier 1 Temperature element                                                      T Vibration detector                                                        V Note:
Instrumentation should be shown as:(lines connecting the FE instruments do not indicate ASME Code classes or wire type)
Equipment:
Annunciator                          H        Relief valve (H=high, L=low)                    A L
Plug or Ball valve Butterfly valve Check valve                                  Probe Damper                                        Pump Fan, Blower                                  Solenoid Filter                            F          Strainer                  S Flow restrictor                              Three way valve Gate valve                                    Vacuum breaker          VB Globe valve                                  Valve type not specified Main Turbine                                  Water trap              TR Stop Valve Notes: 1. Valves shown do not denote either open or closed position.
: 2. Valves shown without operators may be local manual valves.
: 3. Components shown in phantom are not part of the system on the figure it appears.
Legend for Figures                                                                                        Appendix A-3
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 Valve Operators:
Motor                              M Pneumatic P
Electrical Equipment Cable or conduit Cable connection Connection to bus Circuit Interrupting Device Transformer Appendix A-4                                                                      Legend for Figures
 
25A5675AA Revision 7 ABWR                                                                        Design Control Document/Tier 1 Battery Note 1: Devices shown do not denote either open or closed position.
Note 2: Circuit Interrupting Devices may consist of circuit breakers, fuses or a combination of breakers and fuses.
Building Divisional Barrier                            Door (Note 1 & 3)
(Note 2)
* Door (Note 3)                                  Door (Note 3)
Elevator                                      Grating Floor Grid line identifier                          Grid line intersection (for information only)        RA              (for information only)
Hatch                                          Opening or Legend for Figures                                                                                  Appendix A-5
 
25A5675AA Revision 7 ABWR                                                                      Design Control Document/Tier 1 Removable block                                Secondary wall                        BW                containment barrier for R/B and MCAE for C/B (Note 2), or radiation zone boundary Sliding door                                  Stairway D  U Sump pit                                      Typical floor designation:
B3F-Basement, 3rd floor NOTES:
: 1. Swing of door can be either way.
: 2. Divisional and secondary containment barriers and MCAE are fire barriers unless specified otherwise.
: 3.
* Denotes watertight door.
Control and Instrumentation Cables:                Fiber-optic Metallic Fiber-optic or metallic Appendix A-6                                                                              Legend for Figures
 
25A5675AA Revision 7 ABWR                                                        Design Control Document/Tier 1 Cables not connected Sensor Switch Legend for Figures                                                              Appendix A-7
 
25A5675AA Revision 7 ABWR                                            Design Control Document/Tier 1 Appendix B Abbreviations and Acronyms Used in the ABWR Certified Design Material Abbreviations and Acronyms                                          Appendix B-1
 
25A5675AA Revision 7 ABWR                              Design Control Document/Tier 1 Appendix B-2                              Abbreviations and Acronyms
 
25A5675AA Revision 7 ABWR                                                            Design Control Document/Tier 1 Note: These abbreviations and acronyms apply to the ABWR Certified Design Material. Other documents may use different abbreviations or acronyms.
Abbreviations and Acronyms                                                          Appendix B-3
 
25A5675AA Revision 7 ABWR                                                          Design Control Document/Tier 1 Appendix C Conversion to ASME Standard Units From                      To convert to                      Divide by (1)      Pressure/Stress kilopascal                1 Pound/Square Inch                6.894757 kilopascal                1 Atmosphere (STD)                  101.325 kilopascal                1 Foot of Water (39.2&deg;F)            2.98898 kilopascal                1 Inch of Water (60&deg;F)              0.24884 kilopascal                1 Inch of HG (32&deg;F)                3.38638 (2)      Force/Weight newton                    1 Pound - force                    4.448222 kilogram                  1 Ton (Short)                      907.1847 kilogram                  1 Tons (Long)                      1016.047 (3)      Heat/Energy/Power joule                      1 Btu                              1055.056 joule                      1 Calorie                          4.1868 kilowatt-hour              1 Btu                              0.0002930711 kilowatt                  1 Horsepower(U.K.)                  0.7456999 kilowatt-hour              1 Horsepower-Hour                  0.7456999 kilowatt                  1 Btu/Min                          0.0175725 joule/gram                1 Btu/Pound                        2.326 (4)      Length millimeter                1 Inch                              25.4 centimeter                1 Inch                              2.54 meter                      1 Inch                              0.0254 meter                      1 Foot                              0.3048 centimeter                1 Foot                              30.48 meter                      1 Mile                              1609.344 kilometer                  1 Mile                              1.609344 (5)      Volume liter                      1 Cubic Inch                        0.01638706 cubic centimeter          1 Cubic Inch                        16.38706 Conversion to ASME Standard Units                                                  Appendix C-1
 
25A5675AA Revision 7 ABWR                                                              Design Control Document/Tier 1 From                              To convert to                      Divide by cubic meter                      1 Cubic Foot                        0.02831685 cubic centimeter                  1 Cubic Foot                        28316.85 liter                            1 Cubic Foot                        28.31685 cubic meter                      1 Cubic Yard                        0.7645549 liter                            1 Gallon (US)                      3.785412 cubic centimeter                  1 Gallon (US)                      3785.412 E-03 cubic centimeter            1 Gallon (US)                      3.785412 (6)      Volume Per Unit Time cubic centimeter/s                1 Cubic Foot/Min                    471.9474 cubic meter/h                    1 Cubic Foot/Min                    1.69901 liter/s                          1 Cubic Foot/Min                    0.4719474 cubic meter/s                    1 Cubic Foot/Sec                    0.02831685 E-05 cubic meter/s                1 Gallon/Min (US)                  6.30902 cubic meter/h                    1 Gallon/Min (US)                  0.22712 liter/s (101.325 kPaA,15.56&deg;C)    1 STD CFM (14.696 psia, 60oF)      0.4474 cubic meter/h                    1 STD CFM (14.696 psia, 60oF)      1.608 (101.325 kPaA,15.56&deg;C)
(7)      Velocity centimeter/s                      1 Foot/Sec                          30.48 centimeter/s                      1 Foot/Min                          0.508 meter/s                          1 Foot/Min                          0.00508 meter/min                        1 Foot/Min                          0.3048 centimeter/s                      1 Inches/Sec                        2.54 (8)      Area square centimeter                1 Square Inch                      6.4516 E-04 square meter                1 Square Inch                      6.4516 square centimeter                1 Square Foot                      929.0304 E-02 square meter                1 Square Foot                      9.290304 (9)      Torque newton-meter                      1 Foot Pound                        1.355818 (10)    Mass Per Unit Time kilogram/s                        1 Pound/Sec                        0.4535924 Appendix C-2                                                          Conversion to ASME Standard Units
 
25A5675AA Revision 7 ABWR                                                                  Design Control Document/Tier 1 From                                  To Convert to                    Divide by kilogram/min                          1 Pound/Min                      0.4535924 kilogram/h                            1 Pound/Min                      27.215544 (11)    Mass Per Unit Volume kilogram/cubic meter                  1 Pound/Cubic Inch                27679.90 kilogram/cubic meter                  1 Pound/Cubic Foot                16.01846 kilogram/cubic centimeter            1 Pound/Cubic Inch                0.0276799 liter/s                              1 Gallon/Min                      0.0630902 (12)    Dynamic Viscosity Pa*s                                  1 Pound-Sec/Sq Ft                47.88026 (13)    Specific Heat/Heat Transfer joule/kilogram kelvin                1 Btu/Pound-Deg F                4186.8 watt/square meter kelvin              1 Btu/Hr-Sq Ft-Deg F              5.678263 watt/square meter kelvin              1 Btu/Sec-Sq Ft-Deg F            2.044175E+4 watt/square meter                    1 Btu/Hr-Sq Ft                    3.154591 (14)    Temperature degree celsius                        Degrees Fahrenheit                T&deg;F = T&deg;Cx1.8+32 degree C increment                    1 Degree F Increment              0.555556 (15)    Electricity coulomb                              1 ampere hour                    3600 siemens/meter                        1 mho/centimeter                  100 (16)    Light candels/square meter                  1 candela/square inch            1550.003 lux                                  1 footcandle                      10.76391 (17)    Radiation megabequerel                          1 curie                          37,000 gray                                  1 rad                            0.01 sievert                              1 rem                            0.01 Note:
Rounding of Caculated values per Appendix C of ANSI/IEEE Std. 268.
Conversion to ASME Standard Units                                                          Appendix C-3}}

Latest revision as of 03:08, 23 December 2024

GE-Hitachi ABWR Design Control Document Tier 1 & 2, Rev. 7 - Tier 1
ML20007E325
Person / Time
Site: 05200001, 05200045
Issue date: 12/20/2019
From: Michelle Catts
GE-Hitachi Nuclear Energy Americas
To:
Office of Nuclear Reactor Regulation
Shea J
References
GEHITACHIABWR, GEHITACHIABWR.SUBMISSION.8, 25A5675.P, 25A5675.P.7
Download: ML20007E325 (656)
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