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For the GE ABWR Fact Sheet, please click here.

For an overview of the GE ABWR, please click here.

In the News: GE ABWR

Advanced Boiling Water Reactor (ABWR)

• Generation III boiling water reactor
• Designed by General Electric and is currently offered by the alliance of General Electric and Hitachi
• Generates electrical power by using steam to power a turbine connected to a generator, the steam is boiled from water using heat generated by fission reactions within nuclear fuel

Major areas of improvement include:

• The addition of reactor internal pumps (RIPs) to the bottom of the RPV (reactor pressure vessel) - 10 in total - which achieve improved performance while eliminating large-diameter and complex piping structures at the bottom of the RPV (e.g. the recirculation loop found in earlier BWR models). Only the RIP motor is located outside of the RPV in the ABWR. According to the Tier 1 Design Control Document (which is the officially certified Nuclear Regulatory Commission document generally describing the design of the plant), each RIP has a capacity of 6912 m³/h at nominal capacity and several can be turned off with the reactor at capacity.
• The control rod adjustment capabilities have been supplemented with the addition of the electro-hydraulic Fine Motion Control Rod Drive (FMCRD), allowing for fine position adjustment, while not losing the reliability or redundancy of traditional hydraulic systems which are designed to accomplish rapid shutdown in 2.80 seconds from receipt of an initiating signal, or ARI in a greater but still insignificant time period. The FMCRD also improves defense-in-depth in the event of primary hydraulic and ARI contingencies.
• A fully digital Reactor Protection System (with redundant digital backups as well as redundant manual backups) ensures a high level of reliability and simplification for safety condition detection and response. Standard BWR half plus half (2 out of 4) rapid shutdown logic ensures that nuisance rapid shutdowns are not triggered by single instrument failures. RPS can trigger ARI (alternate rod insertion), FMCRD rod run-in, as well as SLCS (standby liquid control system) actuation in the event these capabilities and systems are necessary.
• Fully digital reactor controls (with redundant digital backup and redundant manual backups) allow the control room to easily and rapidly control plant operations and processes. Separate, redundant critical and non-critical digital multiplexing buses allow for reliability and diversity of instrumentation and control.
• In particular, the reactor can both "fly on autopilot" and also "take off and land on autopilot" or go critical and ascend to power using automatic systems only and do a standard shutdown using automatic systems only. Of course, human operators remain essential to reactor control, but much of the busy-work of bringing the reactor to power and descending from power can be automated at operator discretion.

• The Reactor Water Cleanup System has been enhanced to ensure prompt and complete removal of soluble neutron absorbers injected by the SLCS in an anticipated transient without scram (ATWS) contingency. This decreases operator reticence to utilize the SLCS prior to using other channels to mitigate an ATWS. Indeed, the SLCS is now able to be automatically actuated by the RPS if necessary in the ABWR, as prompt cleanup of soluble neutron absorbers can be achieved.
• The Emergency Core Cooling System (ECCS) has been improved in many areas, providing a very high level of defense-in-depth against accidents, contingencies, and incidents.
  • The overall system has been divided up into 3 divisions; each division is capable - by itself - of reacting to the maximally contingent Limiting Fault/Design Basis Accident (DBA) and terminating the accident prior to core uncovery, even in the event of loss of offsite power and loss of proper feedwater. Previous BWRs had 2 divisions, and uncovery (but no core damage) was predicted to occur for a short time in the event of a severe accident, prior to ECCS response.
  • Eighteen SORVs (safety overpressure relief valves), ten of which are part of the ADS (automatic depressurization system), ensure that RPV overpressure events are quickly mitigated, and that if necessary, that the reactor can be depressurized rapidly to a level where low pressure core flooder (LPCF, the high-capacity mode of the residual heat removal system, which replaces the LPCI and LPCS in previous BWR models) can be used.
  • Further, LPCF can inject against much higher RPV pressures, providing an increased level of safety in the event of intermediate-sized breaks, which could be small enough to result in slow natural depressurization but could be large enough to result in high pressure corespray/coolant injection systems' capacities for response being overwhelmed by the size of the break.
  • Though the Class 1E (life safety critical) power bus is still powered by 3 highly-reliable emergency diesel generators that are safety rated, an additional Plant Investment Protection power bus using a combustion gas turbine is located on-site to generate electricity to provide defense in depth against station blackout contingencies as well as to power important but non-safety critical systems in the event of a loss of offsite power, as well as to start the plant in the event grid black start is needed. Additional diesel firewater pumps may be tied into the plant's service water system too, to enhance cooling capabilities.
  • Though one division of the ECCS does not have high pressure flood (HPCF) capacities, there exists a steam-driven, safety-rated reactor core isolation cooling (RCIC) turbopump outside of the 3 primary ECCS divisions, that is high-pressure rated and has extensive battery backup for its instrumentation and control systems, ensuring cooling is maintained even in the event of a full station blackout with failure of all 3 emergency diesel generators, the combustion gas turbine, primary battery backup, and the diesel firewater pumps.
  • There exists an extremely thick basalt fiber reinforced concrete (BiMAC) pad under the RPV that will both catch and hold any heated fluids that might fall on that pad in extraordinarily contingent situations. In addition, there are several valves within the weir wall (the wall separating the wetwell from the drywell) that are squib-actuated and can perform an orderly flood of the BiMAC pad using the wetwell's water supply, ensuring cooling of that area even with the failure of standard mitigatory systems (e.g. overhead flood capabilities).
• The containment has been significantly improved over old BWR types. Like the old types, it is of the pressure suppression type, designed to handle evolved steam in the event of a transient, incident, or accident by routing the steam using pipes that go into a pool of water, called the wetwell (or torus), the low temperature of which will condense the steam back into liquid water. This will keep pressure low. Notably, the typical ABWR containment has numerous hardened layers between the interior of the primary containment and the outer shield wall, and is cubical in shape. One major enhancement is that the reactor has a standard safe shutdown earthquake acceleration of .2 G (slightly less than 2 m/s²); further, it is designed to withstand a tornado of Old Fujita Scale 6, with > 320 mph wind). Seismic hardening is possible in earthquake-prone areas and has been done at the Lungmen facility in Taiwan which has been hardened up .3 G (slightly less than 3 m/s²) in any direction.
• The ABWR is designed for a lifetime of at least 60 years, though operation beyond that 60 year point will certainly be possible unless safety limits within the expensive to replace reactor pressure vessel is reached. The comparatively simple design of the ABWR also means that no expensive steam generators need to be replaced, either, decreasing total cost of operation.
• According to GE, only after at least 30 million years does the CDP of the ABWR reach 50% (e.g. 3E-7), better than both the AP1000 and the EPR.