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Status Report – ABWR (GEH, HGNE and Toshiba) USA and Japan DATE (2019/11/12)

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Status Report – ABWR (GEH, HGNE and Toshiba) USA and Japan DATE (2019/11/12) Status Report – ABWR (GEH, HGNE and Toshiba) USA and Japan DATE (2019/11/12) This reactor design is an evolution from the previous ABWR, of which 4 units are operating in Japan [https://pris.iaea.org/PRIS/CountryStatistics/CountryDetails.aspx?current=JP], 5 units are under construction in Japan and Taiwan [https://pris.iaea.org/PRIS/CountryStatistics/CountryDetails.aspx?current=TW], and 4 units with a projected start of construction in 2030s. The reference plant is [Kashiwazaki Kariwa Unit No.7 https://pris.iaea.org/PRIS/CountryStatistics/ReactorDetails.aspx?current=384] and has a net power output of 1,315 MWe. INTRODUCTION Indicate which booklet(s): [ 〆 ] Large WCR [ ] SMR [ ] FR The design of the advanced boiling water reactor (ABWR) represents a complete design for a nominal 1350 MWe power plant. The inclusion of such features as reactor internal pumps, fine motion control rod drives, multiplexed digital fiber-optic control systems, and an advanced control room are examples of the type of advancements over previous designs that have been incorporated to meet the ABWR objectives. The ABWR design objectives include: 60 year plant life from full power operating license date, 87% or greater plant availability, less than one unplanned scram per year, 24 month refueling interval, personnel radiation exposure limit of 100 man-rem/year, core damage frequency of less than 10-5/ reactor year, limiting significant release frequency to 10-6/reactor year, and reduced radwaste generation. The principal design criteria governing the ABWR standard plant encompass two basic categories of requirements: those related to either a power generation function or a safety related function. General power generation design criteria The plant is designed to produce electricity from a turbine generator unit using steam generated in the reactor. Heat removal systems are designed with sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational conditions and abnormal operational transients. Backup heat removal systems are designed to remove decay heat generated in the core under circumstances where the normal heat removal systems become inoperative. The capacity of such systems is adequate to prevent fuel cladding damage. The fuel cladding, in conjunction with other plant systems, is designed to retain its integrity so that the consequences of any equipment failures are within acceptable limits throughout the range of normal operational conditions and abnormal operational transients for the design life of the fuel. Control equipment is designed to allow the reactor to respond automatically to load changes and abnormal operational transients. Reactor power level is manually controllable. Interlocks or other automatic equipment are designed as backup to procedural control to avoid conditions requiring the functioning of safety related systems or engineered safety features. General safety design criteria The plant is designed, fabricated, erected and operated in such a way that the release of radioactive material to the environment does not exceed the limits and guideline values of applicable government regulations pertaining to the release of radioactive materials for normal operations, for abnormal transients and for accidents. The reactor core is designed so that its nuclear characteristics counteract a power transient. The reactor is designed so that there is no tendency for divergent oscillation of any operating characteristics considering the interaction of the reactor with other appropriate plant systems. Safety related systems and engineered safety features function to ensure that no damage to the reactor coolant pressure boundary results from internal pressures caused by abnormal operational transients and accidents. Where positive, precise action is immediately required in response to abnormal operational transients and accidents, such action is automatic and requires no decision or manipulation of controls by plant operations personnel. The design of safety related systems, components and structures includes allowances for natural environmental disturbances such as earthquakes, floods, and storms at the plant site. Standby electrical power sources have sufficient capacity to power all safety-related systems requiring electrical power concurrently. Standby electrical power sources are designed to allow prompt reactor shutdown and removal of decay heat under circumstances where normal auxiliary power is not available. A containment is provided that completely encloses the reactor systems, drywell, and pressure suppression “wetwell” chambers. The containment employs the pressure suppression concept. A safety envelope is provided that basically encloses the containment, with the exception of the areas above the containment top slab and drywell head. The containment and safety envelope, in conjunction with other safety related features, limit radiological effects of design basis accidents to less than the prescribed acceptable limits. The reactor building surrounds the containment/safety envelope and serves as a secondary containment. Provisions are made for removing energy from the containment as necessary to maintain the integrity of the containment system following accidents that release energy to the containment. Emergency core cooling is designed to limit fuel cladding temperature to less than the limits of the U.S. NRC Regulation 10CFR50.46 (2200F or 1204C) in the event of a design basis loss of coolant accident (LOCA). The emergency core cooling is designed for continuity of core cooling over the complete range of postulated break sizes in the reactor coolant pressure boundary piping. Emergency core cooling is initiated automatically when required regardless of the availability of off site power supplies and the normal generating system of the plant. The control room is shielded against radiation so that continued occupancy under design basis accident conditions is possible. In the event that the control room becomes uninhabitable, it is possible to bring the reactor from power range operation to cold shutdown conditions by utilizing alternative controls and equipment that are available outside the control room. Fuel handling and storage facilities are designed to prevent inadvertent criticality and to maintain shielding and cooling of spent fuel as necessary to meet operating and off-site dose constraints. FIG. 1.1. ABWR - steam cycle Reactor Units in PRIS (if applicable): Country Site Name No. of Units (operating / under construction Japan Kashiwazaki-Kariwa No. 6 & 7 / Operating Japan Homaoka No. 5 / Operating Taiwan Lungmen No. 1 & 2/ under construction Japan Shika No. 2 / Operating Japan Shimane No.3 / under construction Japan Ohma (no number) / under construction Japan Higashidori No.1 / under construction Table 1: ARIS Category Fields (see also Spreadsheet “Categories”) for Booklet ARIS Category Input Select from Current/Intended Purpose Commercial – Commercial – Electric/Non-electric, Electric Prototype/FOAK, Demonstration, Experimental Main Intended Application Baseload Baseload, Dispatchable, Off- (once commercial) grid/Remote, Mobile/Propulsion, Non-electric (specify) Reference Location On Coast or On Coast, Inland, Below-Ground, Inland Floating-Fixed, Marine-Mobile, Submerged-Fixed (Other-specify) Reference Site Design Single Unit or Single Unit, Dual Unit, Multiple Unit (reactor units per site) Dual Unit (# units) Reactor Core Size (1 core) Large (>3000 Small (<1000 MWth), MWth) Medium (1000-3000 MWth), Large (>3000 MWth) Reactor Type BWR PWR, BWR, HWR, SCWR, GCR, GFR, SFR, LFR, MSR, ADS Core Coolant H2O H2O, D2O, He, CO2, Na, Pb, PbBi, Molten Salts, (Other-specify) Neutron Moderator H2O H2O, D2O, Graphite, None, (Other- specify) NSSS Layout Direct-cycle Loop-type (# loops), Direct-cycle, Semi-integral, Integral, Pool-type Primary Circulation Forced (10 Forced (10 Internal pumps), Natural Internal pumps) Thermodynamic Cycle Rankine Rankine, Brayton, Combined-Cycle (direct/indirect) Secondary Side Fluid N.A. H2O, He, CO2, Na, Pb, PbBi, Molten Salts, (Other-specify) Fuel Form Fuel Fuel Assembly/Bundle, Coated Assembly/Bundle Sphere, Plate, Prismatic, Contained Liquid, Liquid Fuel/Coolant Fuel Lattice Shape Square Square, Hexagonal, Triangular, Cylindrical, Spherical, Other, n/a Rods/Pins per Fuel 9x9 or 10x10 #, n/a Assembly/Bundle Fuel Material Type Oxide Oxide, Nitride, Carbide, Metal, Molten Salt, (Other-specify) Design Status Detailed Conceptual, Detailed, Final (with secure suppliers) Licensing Status In Operation (4 DCR, GDR, PSAR, FSAR, Design units) Licensed (in Country), Under Construction (# units), In Operation (4 units) Table 2: ARIS Parameter Fields (see also Spreadsheet “Data”) for Booklet ARIS Parameter Value Units or Examples Plant Infrastructure Design Life years 60 Lifetime Capacity Factor %, defined as Lifetime MWe-yrs 87< delivered / (MWe capacity * Design Life), incl. outages Major Planned Outages 30-60 days every # days every # months (specify purpose, 12-24 months including refuelling) Operation / Maintenance # Staff in Operation / Maintenance Crew / Human Resources during Normal Operation Reference Site Design n Units/Modules 1 or 2 Capacity to Electric Grid MWe (net to grid) 1,350 Non-electric Capacity e.g. MWth heat at x ºC, m3/day N.A. desalinated water, kg/day hydrogen, etc. In-House Plant Consumption MWe 40-60 Plant Footprint m2 (rectangular building envelope) Site Footprint m2 (fenced area) Emergency Planning Zone km (radius)
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