Hitachi-GE UK ABWR Concept Design

Transcription

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.

2 Form10/00 DISCLAIMERS Proprietary Information This document contains proprietary information of Hitachi-GE Nuclear Energy, Ltd. (Hitachi-GE), its suppliers and subcontractors. This document and the information it contains shall not, in whole or in part, be used for any purpose other than for the Generic Design Assessment (GDA) of Hitachi-GE s. This notice shall be included on any complete or partial reproduction of this document or the information it contains. Copyright No part of this document may be reproduced in any form, without the prior written permission of Hitachi-GE Nuclear Energy Ltd. Copyright (C) 2013 Hitachi-GE Nuclear Energy, Ltd. All Rights Reserved.

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 ( 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

24 2.3. Continuous Construction The design stage of nuclear power plants require an overall coordination of a broad range of engineering tasks, including conceptual design, layout design, equipment logistics plan, shielding plan, as well as the plant construction, operation and maintenance plan. Schedule management, workforce management and QA/QC management are also important during each task phase. In order to perform these tasks efficiently, Hitachi has developed an Advanced Integrated CAE System to actualise high-quality and efficient works. This system works based on not only the plant engineering database but also accumulated experiences and management know-how of previous projects. It is also enhanced day by day through the actual projects as our core in-house engineering system. To improve the construction period, as well as safety and quality, Hitachi has continuously improved its construction technologies since the first Nuclear Power Plant (NPP) construction in the 1970s. Now, Hitachi has 4 main construction strategies, which are: (1) On-site work volume reduction (2) On-site manpower Levelling (3) Improvement of On-site productivity (4) Improvement of On-site support work efficiency These strategies contribute to Hitachi s excellent (Safe, Quality, On-Schedule and On-Budget) execution of NPP projects. In Japan, the construction of nuclear power plants (NPP) has been uninterrupted. The most recently constructed plants, known as ABWRs (advanced boiling water reactors) are of the Generation III + type, which are the most advanced and safest plants operating today. These ABWRs were built and completed On-Budget and On-Schedule, and Hitachi has been involved in all of the ABWRs constructed in Japan. Since the first BWR Nuclear Power Plant (NPP) construction in the 1970s, Hitachi has developed and successfully applied advanced construction technologies. Success of NPP construction depends on how well the overall project can be phased, considering (and controlling) all engineering, procurement/manufacturing and construction aspects. With the development of functional 3D-CAD engineering environment and streamlined design-to-manufacturing / construction systems, detailed 4. Conclusion 23

25 construction planning and management technologies have been continuously improved with the evolution of Hitachi s construction management philosophies. Hitachi has developed its 4 main construction strategies which are summarised below: (1) Broader application of large module/block construction method with VHL for On-site work volume reduction This method utilizes a very heavy-lift crane (VHL) to lift and install large-scale modules and blocks, which can be constructed either at a site-based or an offsite module shop. Hitachi has employed this method since the early 1980s. During the design, a Computer Aided Engineering (CAE) system with special features dedicated to module engineering is utilized to optimize construction. (2) Open-top and Parallel Construction method / Floor Packaging for On-site manpower Levelling In this method, major components to be installed in the area are carried in prior to the ceiling work. Then, after the curing of concrete in the ceilings, mechanical/electrical installation work can proceed in parallel with the upper level of building construction. Therefore, it enables not only a reduction of the manpower peak, but also a shortening of the construction schedule. In addition, a new concept, named the Floor Packaging method, allows progressive hydro-static pressure testing of piping floor-by-floor, which reduces the maximum workload further. (3) Front-Loaded Construction Engineering / Detailed Schedule Management for Improved On-site productivity To introduce and implement the above strategies as planned, construction-oriented engineering is critical. For this purpose, Hitachi conducts front-loaded construction engineering with the fully integrated 3D-CAE system and a detailed schedule management system, which improves the quality of plant engineering to achieve on-schedule construction. (4) Development and Introduction of an integrated construction management system (ICMS) for Improvement of On-site support work efficiency During the construction period, countless equipment and components need to be managed. Detailed planning takes place well before actual work commences to ensure on-time delivery of products and documents. This advanced site construction management system has been developed since The system enables not only the ability to share engineering information and documents, but also to store computerized construction records and monitor real-time construction progress. 2. Advanced Boiling Water Reactor (ABWR) 24

26 These strategies contribute to Hitachi s excellent (Safe, Quality, On-Schedule and On-Budget) execution of NPP projects. 2. Advanced Boiling Water Reactor (ABWR) 25

27 3. Proposal of With each new generation of BWRs, the goal has been to simplify the design and improve operations, including safety improvement for both workers and the public. The UK-ABWR inherited a technologically rich legacy of design, development and operating experience from which to provide a plant that minimises radiological exposure to workers and to the public, and to minimise radwaste to meet the as low as reasonably practicable (ALARP) and so far as is reasonably practicable (SFAIRP) principles. The following sub-sections describe the design approaches that were used to accomplish those goals. In addition, it is described that the concept of UK-ABWR will satisfy specific safety-requirements of international regulatory bodies such as WENRA and IAEA and those of UK ONR which are explained in the UK ONR Safety Assessment Principles (SAPs). For the purposes of this discussion, the words SFAIRP, ALARP and as low as reasonably achievable (ALARA) are used interchangeably. 3.1 Plant Employees Safety Target Historically, operating BWRs have had higher worker occupational radiation exposures than PWRs. Improvements in plant operations and lower forced outage rates have improved all LWRs and reduced the gap between PWRs and BWRs. Starting with ABWRs, much design attention was paid to reducing occupational exposure, based on operating experience. Since commencement of operations in , the two ABWR units at Kashiwazaki-Kariwa in Japan have averaged about 360 man-msv per reactor per year. The factors involved in this improvement are described below Plant Employees Safety Target - Normal Operation The ABWR combines advanced facility design features and administrative procedures designed to keep the occupational radiation exposure to personnel as low as reasonably achievable (ALARA). During the design phase, layout, shielding, ventilation and monitoring instrument designs are integrated with traffic, security and access control. Operating plant results are continuously integrated during the design phase. Clean and controlled access areas are separated. Reduction in the plant personnel radiation exposure is achieved by (1) minimizing the necessity and amount of personnel time spent in radiation areas, and (2) minimizing radiation levels in routinely occupied plant areas in the vicinity of plant equipment which are expected to require personnel attention. 2. Advanced Boiling Water Reactor (ABWR) 26

28 Minimization of the time required by plant operators in high radiation areas is achieved through design improvement relative to the current operating BWRs and features such as described in these examples: The elimination of external recirculation loops by adopting reactor internal pump system has allowed the reduction in worker dose related to RRS inspection. Worker dose for the inside of the PCV is estimated to decrease dramatically compared to BWR-5, as shown in Figure BWR-5 ABWR MS Nozzle floor RRS Riser floor RRS Motor floor RRS Pump floor SRV floor Platform RHR Nozzle floor Diaphragm floor RIP Motor floor Platform Dose rate (msv/h) Dose rate (msv/h) Figure Improvement of Reactor Recirculation System (RRS) Equipment is designed to facilitate maintenance. The Fine Motion Control Rod Drives (FMCRDs) require greatly reduced maintenance compared to the locking piston drives used in previous BWRs. The FMCRD scram water is discharged directly into the reactor vessel, which allows the elimination of certain components required by the Locking Piston CRDs (LPCRD), such as the scram discharge valve, hydraulic discharge lines and the scram discharge volume. Internal shootout support structure is provided to prevent the inadvertent ejection of the drive and control rod in the event of the CRD housing failure. This replaces the external beam support structures used in previous BWRs. The FMCRD purge flow during normal operation precludes entrance of reactor water and crud into the drive. All of these features of the FMCRD reduce personnel radiation exposure compared with LPCRD used in previous BWRs. Permanently installed monorails and cranes, semi-automated removal tools and nearby location of maintenance areas reduce duration of personnel time spent in proximity to higher dose rate areas. 2. Advanced Boiling Water Reactor (ABWR) 27

29 The materials used in the main coolant system consist mainly of austenitic stainless steel, Nibased alloys, carbon steel and low alloy steel components. The use of Co-based alloys and cobalt content of the alloys used in core region, as well as final feedwater heater tubes, is minimised to reduce the creation of activation products and associated potential for gamma radiation. Equipment and piping are designed to reduce accumulation of radioactive materials. Piping is constructed of seamless pipe wherever possible, and filters and demineralisers are backwashed and flushed prior to maintenance. Leakage from equipment is piped to sumps and floor drains. Radioactive isotopes in the primary coolant are limited by the use of CUW and the Condensate demineraliser. Clean purge water is continuously supplied to the FMCRDs to keep the equipment from accumulating radioactive contaminants. The purge water is also supplied into the reactor through the gap between shaft and stretch tube of each RIP for avoiding radioactive contamination inside the RIP motor-case. The retractable Source Range Monitor (SRM) and Intermediate Range Monitor (IRM) neutron detectors have been replaced with fixed in-core Source Range Neutron Monitors (SRNMs). Material selection takes ALARA requirements into consideration. In addition to minimising radiation levels and time spent in radiation areas, other features are incorporated to further reduce personnel radiation exposure. Radiation zones are established in all areas of the plant as a function of both the access requirements and radiation sources in that area. Operating activities, inspection requirements of equipment, maintenance activities and abnormal operating conditions are considered in determining the appropriate zoning for a given area. Extensive consideration is given to implementation of radiation shielding. The primary objective of radiation shielding is to protect operating personnel and the general public from radiation emanating from the reactor, power conversion systems, radwaste process systems and auxiliary systems. Radiation shielding is also designed to limit the radiation exposure of critical components within specified limits to assure that their performance and design life are maintained. For all areas potentially having airborne radioactivity, the ventilation systems are designed such that during normal and maintenance operations, airflow progresses from an area of low potential 2. Advanced Boiling Water Reactor (ABWR) 28

30 contamination to an area of higher potential contamination. This is achieved by keeping specific zones at higher or lower pressure, as required, in respect to their adjacent compartments. The maximum individual annual worker dose is controlled by the Plant Operator s administrative procedures. In Japan, it is a common practice to limit maximum doses to 20mSv, which is 2/5 of the regulatory annual limit 50 msv and 1/5 of the regulatory 5-year total limit of 100 msv. In the UK, regulatory limits apply for normal operation and for accidents for both workers, groups of workers and members of the public. The normal operational limit for radiation workers (those expected to be exposed to elevated levels of radiation as part of their employment) is set at 20mSv per year with an objective to limit annual dose to 1mSv. These limits are defined as Basic Safety Limit (BSL) and Basic Safety Objective (BSO) and are connected with the ALARP concept. A licensee who operates a plant is expected to make much greater effort (time and money) to reduce exposure if it is at a level near the BSL than for exposures at a level near BSO. Below BSO, no further action is justified to reduce dose further. Limits for non-radiation workers or for groups of workers are lower. The plant personnel radiation exposure will also depend on plant operation conditions such as the outage time, the fuel cycle, man-hour shift requirements, and so on. The plant personnel radiation exposure target of UKABWR would be decided by the adoption of the abovementioned various exposure reduction measures based on ABWR experiences and the plant conditions. Radiation exposure of ABWR over several years compared to BWR (including ABWR) and PWR is shown on Figure as reference. Figure Trend of Radiation exposure *Source: Operational Status of Nuclear Facilities in Japan, JNES 2. Advanced Boiling Water Reactor (ABWR) 29

31 Plant Employees Safety Target Accidents The layout and shielding considerations that are required to bring exposures ALARP for normal operation also benefit the plant workers during accidents. The limiting evaluations were done for the DBA, because US regulations require radiological evaluations to include consideration of accident sequences in which the release of a significant fraction of the core inventory of fission products to the containment is postulated. The only permanently manned on-site location after accidents is the Main Control Room. Calculated exposure to workers located in the control room after the postulated DBA is under 100 msv (10 rem) TEDE(Total Effective Dose Equivalent). More details will be available in Dose Assessment for STEP2 of the GDA process, which will also consider other DBAs Public Safety TargetPublic Safety Target - Normal Operation ALARP considerations also went into design for minimizing off-site liquid and airborne releases during normal operation. Significant efforts have been expended to minimize both the radioactivity generated within the UK-ABWR and to remove the radioactivity from gaseous waste and liquid waste generated in the UK-ABWR The design of the UK-ABWR includes a gaseous waste treatment system which collects, conveys, and discharges gaseous radioactive waste. The system includes processes and technologies to remove certain radionuclides prior to discharge, such as filters. The main gaseous treatment system includes hold up columns that are filled with charcoal. The purpose of these charcoal absorbers is to retain the fission products for the defined period during which they undergo radioactive decay, reducing the amount of radioactivity released into the environment The design of the UK-ABWR includes a liquid waste treatment system that collects, conveys and discharges aqueous radioactive wastes. The liquid treatment system is designed to recycle treated liquid waste as much as possible (except for detergent liquid waste such as laundry drain to reduce the amount of radioactivity being released into the environment. There may be times, however, when some liquid discharges are necessary (due to capacity limits for on-site storage or unsuitability for re-use). Where this occurs and when practicable, the aqueous radioactive waste liquid is treated prior to its disposal to the environment. The treatment system includes filtration, demineralisation and evaporation processes to remove certain materials and radionuclide. 2. Advanced Boiling Water Reactor (ABWR) 30

32 For all discharges of radioactive wastes into the environment a full demonstration that the systems in place within the UK-ABWR are the Best Available Techniques (BAT) will be provided in the subsequent Hitachi-GE submissions as part of the UK Generic Design Assessment (GDA). These assessments will show that gaseous discharges and liquid discharges have been reduced to a very low level within the UK-ABWR. This will be supported by prospective radiological dose assessments as required by the Environment Agency s P&ID. As an example, Figure is provided, which shows the annual dose during normal operation for Japan ABWR. 50 Dose Limit: 50 μsv Annual Dose (μsv/y) External Exposure from Gaseous Effluents Internal Exposure by Inhalation Internal Exposure from Farm Products Internal Exposure from Sea Products 0 Evaluated Figure ABWR Example of Evaluated Annual Dose during normal operation *Source: Application for Approval of Alteration in Reactor Establishment Permit for Shika Nuclear Power Plant (Construction of Unit No. 2) Public Safety Target Accidents Most of the large body of regulation and many of the plant features are devoted to protecting the public against radiation releases from accidents. In addressing this topic in the context of ALARP, we will first look at the results from Design Basis analysis, then look at the results from a Level 3 Probabilistic Safety Analysis (PSA) which ties together the best-estimate analysis of releases together with their probability of occurrence. The dose criteria in Japanese regulation are summarized in table 5 compared with IAEA and UK. 2. Advanced Boiling Water Reactor (ABWR) 31

33 ABWR Example of Radiological Consequences for DBAs and Radiological Consequences at Site Boundary for Hypothetical accident has large margin for this criteria as shown on Figure and Figure Further discussion will be provided in Fault Study during STEP2 of GDA, as compared with UK criteria described in Table Table Dose Criteria in Japanese and UK regulations maximum doses to individuals (Public) UK criteria Japan IAEA BSL BSO 1mSv/y 20mSv/y 1mSv/y 1mSv/y normal operation 50μSv/y 10mSv/y 0.5mSv/y - Accident condition Site evaluation 5mSv/y (major accident) Whole Body:0.25Sv Thyroid:1.5Sv (hypothetical accident) Whole Body:0.25Sv Thyroid:3Sv 1mSv/y (>10-3 /y) 0.01mSv/y (>10-3 /y) 10mSv/y 0.01mSv/y ( /y) ( /y) 100mSv/y 0.01mSv/y (<10-4 /y) (<10-4 /y) Effective Dose equivalent (msv) 1.0E E E E E E E E-05 Dose Limit: 5 msv Large Margin OGRA MSLBA FHA LOCA CRDA OGRA : Off Gas System Rupture Accident 2. Advanced Boiling Water Reactor (ABWR) 32

34 MSLBA: Main Steam Line Break Accident FHA : Fuel Handling Accident LOCA : Loss of Coolant Accident CRDA : Control Rod Drop Accident Figure ABWR Example of Radiological Consequences for DBAs 1.0E E E+02 Dose Limit: 3000 msv LOCA MSLBA Dose Limit: 250 msv Dose (msv) 1.0E E E E E-03 Tyroid Dose Whole Body Dose Figure ABWR example of Radiological Consequences at Site Boundary for Hypothetical accident 3.3. Green Design Consideration Fuel Design Development for Efficiency BWR s flexible core design enables high fuel economy such as high burn-up, which results in a reduced amount of spent fuel, and uranium saving. As described in Section 2.1, BWRs use spectrum shift operation. Also, BWRs have a larger core size, the flexibility from which improves neutron efficiency by using a low leakage fuel loading pattern (e.g. low reactivity fuels at core periphery) and low leakage enrichment distribution in fuel assembly (e.g. low enrichment at both lower and upper end). Among continuous developments of BWR fuel, the extended fuel rod array has enlarged the thermal margin, and central water rods have optimized uranium to the water ratio. Through the above features, ABWR achieves high burn-up and uranium saving. Regarding uranium saving, ABWR can save 15% of natural uranium compared with PWRs (Table 2.1-1), which means smaller mining operation consequences such as carbon emissions and mining wastes. 2. Advanced Boiling Water Reactor (ABWR) 33

35 Radioactive Waste Management Principles The fundamental principles behind the management of radioactive waste in the UK are that doses to individuals are to be kept as low as reasonably practicable, with economic and social factors being taken into account by the optimization of radiation protection. The Environment Agency requires optimization of radiation protection with respect to the public to be achieved through the use of the best available techniques. For this purpose, the following claims are raised: Eliminate or Reduce the Generation of Radioactive Waste Minimise the Radioactivity in Radioactive Waste Disposed to the Environment Minimise the Volume of Radioactive Waste Disposed to Other Premises Optimise the Disposal Routes for Waste Transferred to Other Premises Minimise the Impact on the Environment and Members of the Public from Radioactive Waste that is disposed to the Environment To eliminate or reduce the generation of radioactive waste, Best Available Techniques are applied, such as Water Quality Control, Suitable Material Selection, etc. To minimise the radioactivity in radioactive waste disposed to other premises, Best Available Techniques such as Off-gas Hold-up system, etc., are applied Spent Fuel Management Principles The UK Government s current assumption is that spent fuel from new nuclear power stations will be used on a once-through basis, with it being disposed of in a retrievable form as waste rather than being reprocessed. The management of spent fuel will be the responsibility of the operator, who will be required to provide appropriate interim storage for the spent fuel likely to arise during a power station s lifetime, until a suitable disposal repository becomes available. Reference: - Great Britain. Parliament. House of Commons. Department for Business Enterprise and Regulatory Reform, Meeting the Energy Challenge - A White Paper on Nuclear Power. London: The Stationery Office CO 2 Reduction Restraint of global warming by the reduction of CO 2 emission is considered the main influence that the nuclear power plant (NPP) operation gives to the environment. The load factor of the nuclear 2. Advanced Boiling Water Reactor (ABWR) 34

36 power plant affects the emission volume of CO 2 when nuclear power generation is included in the baseline of a country s power supply. This is because when the load factor of an NPP decreases due to an unscheduled outage or a prolonged maintenance period, it is necessary to make up for the power shortage using a different generation station, such as fossil fuel power stations. The ABWR-type NPP has been designed to achieve a very low unscheduled outage rate, due to reliable systems and equipment. In addition, a shortening of inspection outage periods is expected because the apparatus is easier to maintain compared to a conventional BWR. The total emission of CO 2 in Japan will be reduced by approximately 2,500,000 tonnes by a 1% NPP operation rate improvement, according to an estimate by the Japanese Central Res. Inst. of Electric Power Industry. Therefore, adopting ABWR, which achieves such characteristics, enables continuous, reliable supply of power in conjunction with controlling emissions to the environment as much as possible Compliance with Specific Requirements The original ABWR was designed to meet old Japanese regulations. Japanese regulations were strengthened after the Fukushima accident, and Japanese ABWRs are being modified to meet the new regulations. The UKABWR design is based on the original Japanese ABWR, and the design is therefore now under modification to achieve compliance as described in the UK regulation Safety Assessment Principles (SAPs). The design is also under modification to meet the Western European Nuclear Regulators Association s (WENRA) Reactor Safety Reference Levels and Safety of New NPP Designs. SAPs and WENRA are high-level expectations, and therefore detailed design will be developed to reflect the expectations of UK nuclear regulators and achieve levels of safety which satisfy the ALARP criteria. UKABWR will also be designed, constructed and operated in line with other non-nuclear related Laws and Standards. Hitachi-GE continues to communicate with UK nuclear regulators and nuclear industries to understand UK expectations Nuclear Quality Assurance Programme (1) Organisation The QMS based upon JEAC4111 and JEAG4121, equivalent to ISO9001:2008 taking IAEA GS-R-3 into consideration, requires the promotion of a strong nuclear-safety culture, graded QMS processes and controls, independent design verification, and examination and tests by a person or persons other than the person or persons who perform the work. The Contractor will establish a Quality Assurance organisational structure with the Quality Assurance Department reporting directly to senior management, and shall be independent of cost and schedule 2. Advanced Boiling Water Reactor (ABWR) 35

37 pressures. Persons performing Quality Assurance functions will have full authority and organisational freedom to perform their work. With regard to equipment of Safety-Related pressure boundaries, the Contractor will conform to the organisation complying with ASME Sec. III NCA-4134 (ASME NQA-1). (2) Safety-Related Work The Japanese standard QMS for Nuclear Power Plants is established in order to ensure that Quality Assurance activities are managed in a planned and systematic manner. The QMS covers design, manufacturing, construction and nuclear power plant operation, to ensure safe and reliable nuclear facility operation. Specific Quality Assurance controls are established for those Structures, Systems and Components (SSCs) classified as Safety-Related. Design, fabrication and construction and operation of Safety-Related SSCs will be performed in accordance with the Contractor Quality Assurance Programme requirements. The overall programme philosophy is based on the underlying core value that effective organisational Quality Assurance is inseparably linked to the establishment and continual development of a strong nuclear-safety culture. This principle is maintained through the establishment and implementation of processes and procedures that ensure individuals are free to report any problems without fear of retribution. With regard to equipment of Safety-Related pressure boundary, all work will comply with ASME Sec. III NCA-4134 (NQA-1) requirements Non-Nuclear Quality Assurance Programme The Contractor will establish and implement a QMS based on ISO9001:2008 or equivalent for non-nuclear safety related project activities. 2. Advanced Boiling Water Reactor (ABWR) 36

38 4. Conclusion ABWR is the latest design of the BWR family, wherein direct cycle design simplifies the process and equipment. Evolutional designs applied to the ABWR improve the plant s safety, reliability, and efficiency compared to previous BWR designs. Hitachi has also continuously improved BWR and ABWR construction technologies for over 40 years. These improvements minimise radiological exposure to workers and to the public, and also minimise radwaste to meet the as low as reasonably practicable (ALARP). In addition, the inherited a technologically rich legacy of design, development and operating experience which will satisfy SAP and WENRA. Much of the information presented in this report has been discussed with ONR and EA in a series of workshops. Hitachi-GE has noted the comments from these workshops and these will be addressed in due course in GDA. 4. Conclusion 37