1 Form10/00 Document ID : GA Document Number : XE-GD-0135 Revision Number : A Generic Design Assessment Hitachi-GE Concept Design Hitachi-GE Nuclear Energy, Ltd.
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3 Table of Contents 1. Introduction Advanced Boiling Water Reactor (ABWR) Simple & Reliable Nuclear-Power Generation System Evolutional Designs Continuous Construction Proposal of Plant Employees Safety Target Plant Employees Safety Target - Normal Operation Plant Employees Safety Target Accidents Public Safety Target Public Safety Target - Normal Operation Public Safety Target Accidents Green Design Consideration Fuel Design Development for Efficiency Radioactive Waste Management Principles Spent Fuel Management Principles CO 2 Reduction Compliance with Specific Requirements Nuclear Quality Assurance Programme Non-Nuclear Quality Assurance Programme Conclusion Table of Contents
4 Acronyms Abbreviations and Acronyms Description ABWR AC ADS ALARA ALARP AM ARI ASME ATWS CAE CCF CDF CFR COPS CRDA CUW D/G DBA DiD DW ECCS EDG FCS FEPC FHA FLSS FMCRD FP FPC GDA GS-R-3 Advanced Boiling Water Reactor Alternating Current Automatic Depressurization system As Low As Reasonably Achievable As Low As Reasonably Practicable Accident Management Alternate Rod Insertion American Society of Mechanical Engineers Anticipated Transients Without Scram Computer Aided Engineering Common Caused Failure Core Damage Frequency Code of Federal Regulations Containment over pressure protection system Control Rod Drop Accident Reactor Water Clean - Up System Diesel Generator Design Basis Accident Defence in Depth Dry Well Emergency Core Cooling Systems Emergency Diesel Generator Flammability Control System The Federation of Electric Power Companies of Japan Fuel Handling Accident Flooder System of Specific Safety Facility Fine Motion Control Rod Drive Fire Protection Fuel Pool Cooling and Cleanup System Generic Design Assessment IAEA Safety Standards Series GS-R-3 Acronyms
5 Acronyms(Contd.) Abbreviations and Acronyms Description HCU Hydraulic Control Unit HPCF High Pressure Core Flooder system IAEA International Atomic Energy Agency ICMS Integrated Construction Management System IRM Intermediate Range Monitor ISO International Organization for Standardization JEAC Japan Electric Association Code JEAG Japan Electric Association Guide JNES Japan Nuclear Energy Safety Organization KK-6 Kashiwazaki-Kariwa Nuclear Power Station Unit 6 KK-7 Kashiwazaki-Kariwa Nuclear Power Station Unit 7 LDF Low Drywell Flooder LOCA Loss of Coolant Accident LPFL Low Pressure Flooder system LUHS Loss of Ultimate Heat Sink LWMS Liquid Waste Management System MG Motor-Generator MSLBA Main Steam Line Break Accident MUWC Make Up Water System (Condensate) NPP Nuclear Power Plant OGRA Off Gas System Rupture Accident PCT Peak Cladding Temperature PCV Primary Containment Vessel PRA Probabilistic Risk Assessment PSA Probabilistic Safety Analysis PWR Pressurized Water Reactor QA/QC Quality Assurance / Quality Control QMS Quality Management System RC&IS Rod Control & Information System RCCV Reinforced Concrete Containment Vessel RCIC Reactor Core Isolation Cooling System Acronyms
6 Acronyms(Contd.) Abbreviations and Acronyms Description RFC RHR RIP RPS RPT RPV RRS RSW S/P SAP SB&PC SBO SCCRI SFAIRP SGTS SLC SPCU SRI SRM SRNM SRV SSC TEDE TEPCO TIP VHL WENRA 3D/CAD Recirculation Flow Control System Residual Heat Removal System Reactor Internal Pump Reactor Protection System Recirculation Pump Trip Reactor Pressure Vessel Reactor Recirculation System Reactor Building Service Water System Suppression Pool Safety Assessment Principles Steam Bypass and Pressure Control System Station Blackout Selected Control Rod Run-In So Far As Is Reasonably Practicable Standby Gas Treatment System Standby Liquid Control System Suppression Pool Cleanup System Selected Rod Insertion Source Range Monitor Source Range Neutron Monitors Safety Relief Valve Structures, Systems and Components Total Effective Dose Equivalent Tokyo Electric Power Company, Incorporated Traversing In-core Probe Very Heavy Lift Crane Western Europian Nuclear Regulators Association 3D Computer Aided Design Acronyms
7 1. Introduction Hitachi developed the advanced boiling water reactor (ABWR) with development concepts of enhanced safety, higher operability, reduced dose equivalent, performance and enhanced cost efficiency. The first ABWRs are the Unit 6 and Unit 7 of Kashiwazaki-Kariwa Nuclear Power Station (KK-6 and KK-7) of Tokyo Electric Power Company, Inc (TEPCO). Hitachi-GE Nuclear Energy, Ltd. (hereafter referred to as Hitachi-GE) has already constructed four ABWRs within Japan. Although the standard Japanese ABWR design in Japan has been established since the completion of KK-6 and 7, Hitachi-GE expects that, during Generic Design Assessment (GDA), the design may need some design changes to deal specifically with UK requirements. However, KK-6 and 7 with further improvements and optimisation incorporated in Ohma 1, Shimane 3 and Shika 2 will be used as reference design in the. The reference design will be confirmed in the Design Reference Point. This document provides a high level design description which defines the design concept of as one of the Step-1 submissions for the GDA. This concept also includes the applicable safety targets to. Some of the numerical information included is based on documents already in the public and may need to be revised on completion of UKABWR-specific studies. 2. Advanced Boiling Water Reactor (ABWR) 6
8 2. Advanced Boiling Water Reactor (ABWR) Hitachi developed the ABWR in 1985, in collaboration with various international partners and support from power companies with experience in BWR plant operation. The main technological features employed are as follows: (1) Large scale, highly efficient plant (2) Highly economical reactor core (3) Reactor coolant recirculation system driven by internal pumps (4) Advanced control rod drive mechanism (5) Overall digital control and instrumentation (6) Reinforced concrete containment vessel These features constitute a highly-functional, enhanced-safety nuclear reactor system, with a compact, easy-to-operate, and efficient turbine of excellent performance. For information on the ABWR, please see the ABWR General Description document which will be published on the UK GDA website. For more detailed general information of ABWR system design features, ABWR Design Control Documents (Revision 4, March 1997) (DCD) are available on the US Nuclear Regulatory Commission website (http://www.nrc.gov/reactors/new-reactors/design-cert/abwr.html). The key specifications of ABWR are shown below. Table 2-1 Key specifications of ABWR Output Item Plant Output Reactor Thermal Output Specification 1,350 MWe class 3,926 MWt Reactor rated pressure 7.07MPa Reactor Fuel Assemblies 872 Core Reactor Equipment Control Rods Recirculation System Control Rod Drive 205 rods Internal pump method Hydraulic / electric motor drive method Primary Containment Vessel ECCS / PCV cooling System Residual Heat Removal System Reinforced concrete with built-in liner 3 divisions 3 divisions Turbine Turbine (final blade length) 52 inches 2. Advanced Boiling Water Reactor (ABWR) 7
9 System Moisture Separation Method Reheat type 2.1. Simple & Reliable Nuclear-Power Generation System One of the world s most common types of nuclear power generating plants, boiling water reactors, are characterized by a system wherein steam generated inside the reactor is directly passed to the turbine to simplify the process and equipment. Since the introduction of the boiling water reactor technology from General Electric in the 1960s, Hitachi has participated in the design, development and construction of over 20 nuclear power plants within Japan. BWRs have a simple direct-cycle configuration in which the generated steam is supplied to the turbine directly. The ABWR, which was developed primarily in Japan and the USA, represents the most advanced example of this type of reactor. This use of a direct-cycle configuration allows highly efficient generation of electric power at a much lower reactor operating pressure. A high efficiency turbine system is adopted based on the direct-cycle features, including the use of a 52-inch long blade for the last stage of the turbine, a two-stage moisture separator re-heater, and a heater drain pump-up system connected to the condensate system. Thermal efficiency is enhanced through the use of this system and potential for leakage of radiation is minimised. From the viewpoint of safety in particular, the operating pressure of the reactor is less than half that of a pressurized water reactor (PWR), another type of light water reactor. The feature of the BWR power generation system is utilized in the design of the safety equipment, because it is easy to inject water directly into the reactor, and therefore the basic approach to achieving safety is to provide a number of different alternative methods for water injection. The high burn-up fuel is especially important with regards to reducing the spent fuel and saving uranium resources. ABWR can reduce uranium consumption by using the characteristics of a BWR. BWRs use spectrum shift operation whereby the core flow rate is progressively reduced from the start to the middle of the fuel cycle, to promote the build-up of the fissionable isotope plutonium-239 (Pu 239). Then the core flow rate is increased again toward the end of the fuel cycle to enhance the fission of the said Pu 239, which increases the energy produced per unit weight of uranium. ABWR can save 15% natural uranium compared with PWRs, as described in Table Advanced Boiling Water Reactor (ABWR) 8
10 Table Comparison of Fuel Economy (PWR data source: Evaluation of Impacts of Operating Cycle Length Extension on typical PWR Plant, FEPC, 2009) 2.2. Evolutional Designs ABWR is the only design which applies advanced light water technology classified as Generation III or III+, and has been in commercial operation for several years as shown on Figure Figure Definition of Reactor generation and ABWR status (Source: GenIVInternational Forum The BWR design has passed through a series of evolutionary changes and achieved significant technological evolution with the current generation of the ABWR. The major key features of the ABWR design are as follows: 2. Advanced Boiling Water Reactor (ABWR) 9
11 Improved safety and reliability Low cost and a short construction period A simpler and more robust design A longer fuel operation cycle The key technological advancements of the ABWR are as follows: (1) Reinforced Concrete Containment Vessel (RCCV) ABWR plant adopts the pressure suppression type containment vessel. In an accident condition such as large break Loss of Coolant Accident (LOCA), steam released from the break is condensed by the large volume of suppression pool water. Moreover, an inert PCV atmosphere is maintained by introduction of Nitrogen to prevent hydrogen combustion in the event of an accident. The RCCV is a cylindrical shaped vessel made of reinforced concrete with a steel liner, designed to be the primary containment vessel for the ABWR plant. The RCCV has also been made significantly smaller than the previous type of containment (Mark III), which has been possible due to the elimination of recirculation piping. This has also reduced the size of the reactor building. The construction period, as well as the wall material volume, has been reduced by introduction of the RCCV design. (2) Reactor Internal Pump (RIP) For the recirculation of coolant, 10 RIPs are arranged at the bottom of the Reactor Pressure Vessel (RPV) instead of conventional Primary Loop Recirculation pumps (External recirculation system). With this new design, no external recirculation piping or pumps are required in the lower portion of the RPV. This simplifies the structure of the RPV and removes a potential radiation source. As a result, work efficiency is enhanced and radiation exposure in maintenance work is reduced. Also, by eliminating the large pipes connected to the lower portion of the RPV, the possibility of coolant level dropping below the Top of Active Fuel (TAF) level is minimised even in the event of a LOCA. This design has improved the overall nuclear safety aspects of the ABWR. General information relating to the Reactor Recirculation System can be found in ABWR DCD rev.4 chapter (3) Fine Motion Control Rod Drive (FMCRD) An FMCRD has been developed for high performance ABWR plant operation. The control rods are electrically controlled by motor-driven screws during normal operation, which enables 2. Advanced Boiling Water Reactor (ABWR) 10
12 the operator to control the reactor power precisely by small movements of the Control Rods. Reactor Power can be easily controlled by the fine motion control rods combined with changing the core flow rate, as described in Figure Figure Reactor Power control by FMCRD and RIP This feature has shortened the start-up duration required to reach the rated power. In an emergency situation, the Hydraulic Control Unit (HCU), which uses pressurized water, reliably allows the Control Rods to be rapidly inserted to shut down the reactor (scram). In a scram situation, the Control Rods are hydraulically inserted rapidly and the electric motor driven screw insertion follows on providing a back-up or defence in depth to the hydraulic insertion. For general information of Functional Design of Reactivity Control System with FMCRD can be found in ABWR DCD rev.4 chapter 4.6. (4) Digital control and instrumentation Digital monitoring control system of ABWR would adopt the increased utilization of multiple technologies and fault tolerance improvement technologies, as well as the use of redundant fibre optic systems for data transmission in the creation of hierarchical information networks, as the following: Large-Scale Display Board facilitates sharing of information The overall plant status would be supplied as shared information Warnings are displayed using hierarchies for improved identification Expanded Automation reduces Load on Operator 2. Advanced Boiling Water Reactor (ABWR) 11
13 Automatic operations, including control rod operation, reduces operator workload to primarily overall plant monitoring operations Integrated Digital Control System Integrated Digital Control System contributes to improve reliability and ease of maintenance. (5) Safety Systems Probabilistic Risk Assessment (PRA) of ABWR/BWRs and PWRs in point of safety divisions is shown on Figure Figure Probabilistic Risk Assessment (PRA) of ABWR/BWRs and PWRs The safety level for both PWR and BWR designs is improved according to the number of divisions compared to the IAEA safety target. ABWR successfully improves Core Damage Frequency (CDF) by enhancing the BWR inherent safety features. The features of the BWR power generation system are utilized in the design of the safety equipment. The BWR uses direct-cycle operation at a relatively low pressure and has a large water inventory and steam buffer in the RPV. These features lead to slow behaviour in the occurrence of pipe rupture, as well as easy pressure reduction, thus making it easy to directly inject water into the reactor. The basic approach to achieve safety for BWRs is, therefore, to provide multiple and diverse methods for water injection. Furthermore, with ABWRs, there are no longer large-diameter nozzles below the core region of the 2. Advanced Boiling Water Reactor (ABWR) 12
14 reactor pressure vessel, and the water level decreases slowly even in the event of limiting LOCA. All three sections of the emergency cooling system have a high pressure injection system in addition to a low pressure injection system. This ensures core flooding can be maintained and safety preserved in the event of a LOCA. The residual heat removal system is also divided into three divisions. With this type of system, according to the results of a probabilistic safety assessment, core melt frequency will be less compared to conventional BWRs, and thus safety shall be enhanced. The Emergency Core Cooling System (ECCS) injection network of the ABWR is comprised of multiple systems in the configuration shown in Figure 2.2-4: a RCIC, two HPCFs, and three LPFLs. The ADS assists the injection network under certain conditions. ABWR has multi diverse defence layers of reactor water level, as shown on Figure Further information relating to the ECCS can be found in ABWR DCD rev.4 chapter 6.3. Figure ABWR safety system configuration 2. Advanced Boiling Water Reactor (ABWR) 13
15 Figure Multi diverse defence layers of reactor water level The primary purpose of the HPCF is to maintain the reactor vessel coolant inventory after small break LOCAs which do not depressurize the reactor vessel. The primary purpose of the LPFL is to provide coolant inventory makeup and core cooling during large break LOCAs, and to provide containment cooling. The LPFL can also be used to provide coolant inventory make up following a small break LOCA if HPCF is not available, by first reducing reactor pressure using the ADS. The ADS utilizes a number of the reactor safety relief valves (SRVs) to reduce reactor pressure during small break LOCAs in the event of HPCF failure. HPCFs and LPFLs are motor-driven type pumps with emergency power supplies, and can operate even during an off-site power loss. The RCIC System injects water into the reactor pressure vessel (RPV) using a high pressure pump driven by a steam turbine. The RCIC steam supply line branches off one of the main steam lines leaving the RPV and goes to the RCIC turbine. As RCIC is a steam-turbine-driven type pump it can operate even after total loss of AC power, that is, a Station Black-Out (SBO) thereby giving ABWR complete diversity of ECCS as well as redundancy. These systems, as ECCS network, start operating automatically in the event of LOCA upon the reactor low water level signal in the following sequence (RCIC : L2 or L1.5, HPCF : L1.5, LPFL : L1) or on the drywell high pressure signal, which indicates a LOCA, in order to maintain the core covered with water. Japanese licensing regulations specify that the accident sequence resulting from a failure of the HPCF pipework is analysed as this is a limiting case LOCA with greatest potential to lead to core uncovery and fuel damage. LOCA analysis of HPCF line break is shown in Figure It has been shown that no 2. Advanced Boiling Water Reactor (ABWR) 14
16 core uncover criteria would be achieved and the peak cladding temperature (PCT) satisfies the design criteria (PCT<1200 o C) with a large margin. Figure2.2-6 LOCA analysis (for example analysis of actual Japan ABWR) The design criteria can be satisfied by any system in the ECCS except RCIC. The LOCA analysis based on only activating LPFL following RPV depressurisation with ADS, is shown in Figure and again shows PCT satisfies the PCT design criteria by a large margin despite water level dropping momentarily below TAF. The safety system configuration of the ABWR almost satisfies the N+2 criteria, in which one division is assumed to be in a testing/maintenance state, while the other division is assumed to suffer a single failure occurrence. The exception to N+2, as mentioned above, occurs if Division B and C are not operational (i.e., unavailable because of failure or maintenance of EDG), leaving only Division A which has RCIC and LPFL/RHR(A). If in addition to the failure of ECCS (B) and (C) divisions, a LOCA break occurs in LPFL/RHR(A) (i.e. in the feedwater line that RHR(A) interfaces with for RPV injection), the steam turbine driven RCIC cannot be credited for long-term LOCA makeup. This is because the RCIC pump is turbine driven and during a LOCA, the steam available from the RPV does not provide sufficient power to pump the quantity of water required to compensate for the LOCA. Therefore, some minor design change may be required to enable to accomplish full N+2 criteria compliance. 2. Advanced Boiling Water Reactor (ABWR) 15
17 Figure LOCAanalysis (Minimum ECCS System Requirement) (6) The Defence in Depth concept including lessons learned from the Fukushima-daiichi nuclear power plant (NPP) accident To accomplish a higher level of nuclear safety, additional design changes are being developed to take into consideration the Fukushima-daiichi NPP accident caused by the earthquake and subsequent tsunamis on March 11, ABWR safety features are based on the Defence in Depth (DiD) concept. The DiD is a common concept for safety in many areas and is sometimes called a belt and braces approach. The concept works by providing additional ways of achieving the safety functions using different systems (layers of protection) even though the primary safety systems are reliable enough to provide the required level of protection. IAEA shows five levels of the DiD (Table 2.2-1). ABWRs are compliant with the international criteria. Levels 2 and 3 in the criteria are achieved by having well-designed Safety Systems to deliver the safety functions. The ABWR safety systems have been assessed and found to achieve the lowest core damage frequency (CDF) in the Generation III + reactor plant groups. The systems to achieve the Level 4 DiD already ensure good response to accidents, however this response will be enhanced based on lessons learned from the Fukushima-daiichi NPP accident. 2. Advanced Boiling Water Reactor (ABWR) 16
18 The method by which control of reactivity is achieved with high reliability and redundancy is shown in Figure The Reactor Protection System (RPS) provides a high reliability actuation signal which has 2 out of 4 voting logic. In addition ARI(Alternative Rod Insertion) and RPT(Recirculation Pump Trip) systems act to insert control rods and reduce recirculation flow by tripping RIPs (which adds additional negative reactivity), by different actuation signals from those of RPS. The Control Rod insertion is achieved by diverse means using the HCU for hydraulic insertion of the control rods with the electric-motor driven screw insertion of control rods acting as a follow up in case the hydraulic insertion fails. Finally, the SLC (Stand by Liquid Control System) acts as a diverse reactivity control system with the ability to shut down the reactor from full power operation to cold shutdown conditions and to maintain this state by injecting neutron absorbing solution into the core in the unlikely event that control rod insertion is not available. An outline of the DiD systems to ensure core cooling is shown in Figure Safety systems are established based on design conditions with sufficient safety margins, so as to deal with all conceivable severe accident scenarios. The ABWR design includes the possibility to use the Make-up water Condensate system (MUWC) and Fire protection system (FP) as alternative injection systems for accident management. (see the mark A on Fuigure2.2-9) Lessons learned from the accident at Fukushima Daiichi Nuclear Power Station indicate the need to consider the potential for sitewide damage caused by beyond design basis external hazards (e.g. flooding) and management under severe plant conditions. The most important approach for the beyond design basis external hazard is to provide diversified methods of coolant injection into RPV. This is assisted because the BWR uses direct-cycle operation at lower pressure making it relatively easy to inject water directly into the reactor. To reduce risk of core damage, and to deal with events beyond the accident scenarios considered in the design, various accident management equipment, including multi-use portable equipment for injecting water into the reactor or removing heat generated in the reactor, will be provided. (see the mark B on Fuigure2.2-9) In addition to the above measures, the backup building concept will be equipped with an alternative power supply and reactor water injection function to mitigate consequences in the case of a large degree of damage to the reactor building. The backup building is located separately from the reactor building. Locating the backup water injection equipment away from the reactor building provides a redundant, diverse and segregated water injection method. It should also be useful for functions such as providing a frontline base during emergencies and a secure and protected storage facility for mobile equipment. (see the mark C on Fuigure2.2-9 and outline of the backup building shown on Fuigure ) 2. Advanced Boiling Water Reactor (ABWR) 17
19 The countermeasures for safety functions in various plants accident conditions for ABWR are shown in Table as an example. The analysis of the robustness of the UKABWR design against extreme events, considered in response to the Great East Japan Earthquake which occurred on March 2011 and the specific designs to mitigate such extreme beyond design basis events are described in STEP1a C3a document (Resilience of Design against Fukushima type Events). 2. Advanced Boiling Water Reactor (ABWR) 18
20 Table The Defence in Depth of IAEA definition Levels of the Defence in Depth Level 1 Level 2 Level 3 Objectives Prevention of abnormal operation and failures [Shut down the reactor safety] Control of abnormal operations and detection of failures [Core cooling and Maintaining containment function] Control of accident within the design basis Example Means of ABWR Anti-Earthquake Measures Feedwater/Level Control system RPS (Reactor Protection System) ECCS Primary Containment vessel Secondary Containment SGTS, FCS Level 4 Control of severe plant conditions Inert PCV Severe accident measures Post-Fukushima enhancement Level 5 Mitigation of radiological consequences Exclusion distance Off-site emergency response SLC pump 3.SLC Reactivity control by boron injection 2.ARI+RPT Control rod insertion by different actuation signal from that of RPS 1.RPS Control rod insertion by high reliability actuation signal by RPS (2 out of 4) Enough shutdown margin Figure Control of reactivity with high reliability and redundancy 2. Advanced Boiling Water Reactor (ABWR) 19
21 R/B A B C Pump A B Diesel Driven FP MUWC Mobile Pump DW CST Accident management by risk analysis Fukushima Lessons learned Accident Management for large damage of R/B RCIC Pump s C FLSS Air-cooled DG ECCS for DBA Backup Building Figure Outline of the countermeasures for core cooling Flooder System of Specific Safety Facility Figure Outline of Backup building 2. Advanced Boiling Water Reactor (ABWR) 20
22 Table Example of the countermeasures for safety functions in various plants accident conditions (1/2) Safety Functions Plant Conditions Normal Operation Transient Design Basis Accident Beyond DBA (include Core Damage condition) Reactivity -RFC -RPT -Hydraulic CR Insertion -Scram followed by FMCRD Control -RC&IS -SCRRI and/or SRI (Scram) -ARI -SB&PC -SLC Reactor Cooling -Reactor Feedwater Pump -etc. -RCIC -RIP-MG Set -HPCF -ADS/LPFL -Enhancement of Buildings Water- Tightness -Alternative Water Injection (MUWC, Diesel driven FP Pump) - FLSS (in Backup building) -Pumper Truck -RPV Depressurisation Enhancement Residual Heat Removal -Main Condenser -RHR Shut down Cooling Mode -etc. -RHR Shut down Cooling Mode -RHR Suppression Pool Cooling Mode -Enhancement of Buildings Water- Tightness - Alternate Heat Exchange Facility -PCV Venting with AM and/or COPS (Feed and Bleed) -Transportable Nitrogen Gas Injection System Containment Preservation and Cooling -DWC -DWC -Isolation Valves -RHR Containment Spray Mode -Enhancement of Buildings Water- Tightness - Alternate Nitrogen Injection System -Pumper Truck -Lower DW injection with AM and/or LDF -PCV Head Cooling 4. Conclusion 21
23 Table Example of the countermeasures for safety functions in various plants accident conditions (2/2) Safety Functions Plant Conditions Normal Operation Transient Design Basis Accident Beyond DBA Spent Fuel Cooling -FPC -Feed from MUWC -Supplemental Feed from SPCU -Supplemental Feed from RHR -Enhancement of Buildings Water- Tightness -Supplemental Feed from SPCU -Diesel-Driven FP Pump -Alternative Water Injection -Pumper truck DC Source -Thyristor -Thyristor -8-hours Duration Batteries + Charging by EDG -Enhancement of Buildings Water- Tightness -8-hours Duration Batteries + Charging by Alternative AC Source (Backup Building) -Transportable Battery AC Source Notes: -Main Generator -Main Generator or Auxiliary Transformer and Grid COPS: Containment over pressure protection system FLSS: Flooder System of Specific Safety Facility LDF: Low Drywell Flooder RFC: Reactor Recirculation Flow Control System SRI: Selected Rod Insertion DWC: Drywell Cooling System SPCU: Suppression Pool Clean-up Water System FP: Fire Protection System -7-days Operable EDG -Alternative AC Source (air cooled DG in Backup building) -Enhancement of Buildings Water- Tightness -Mobile Power Supply RC&IC: Rod Control and Information System SCRRI: Selected Control Rod Run In FPC: Fuel Pool Cooling and Cleanup System EDG:Emergency Diesel Generator 2. Advanced Boiling Water Reactor (ABWR) 22
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