The Advanced Boiling Water Reactor (ABWR) is a Generation III boiling water reactor. The ABWR was designed by General Electric and is currently offered by the alliance of General Electric and Hitachi. The ABWR 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.

Boiling water reactors (BWRs) are the second most common form of light water reactor with a design that is simpler, far more intuitive, and less costly to build, but less compact and slightly less efficient than their pressurized water reactor (PWR) brothers. Their supercritical water reactor cousins are still only theoretical at this point. The ABWR is the present state of the art in boiling water reactors, and is the first Generation III reactor design to be fully built, with several reactors completed successfully and operating, on time and under budget, in the Japan, with several more nearing completion in that nation, and several other ABWRs nearing completion in Taiwan. Several ABWRs are on order in the United States, as well, with the first licensing application for new nuclear reactors in over a decade being filed for the South Texas Project with two ABWRs in 2007.

The standard ABWR plant design has a net output of about 1350 MWe (3926 MWth), however General Electric Hitachi (GEH) also offers a design with greater electrical output. It has also been certified as a final design in final form by the U.S. Nuclear Regulatory Commission, meaning that its performance, efficiency, output, and safety have already been verified, and far fewer steps have to be gone through to build it rather than a non-certified design.

Overview of the design

While GE Hitachi's ESBWR represents a somewhat major departure from the present standard BWR design with an emphasis on full passive nuclear safety and an entirely different approach to power regulation (natural circulation rather than circulation pumps), the ABWR represents a more evolutionary route for the BWR family, with numerous changes and improvements to the standard BWR design.

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.

The RPV and NS³ have significant improvements, such as the substitution of Internal recirculation pumps improve reliability and performance, eliminating complex Internal recirculation pumps inside of the reactor pressure vessel (RPV) are a major improvement over previous GE reactor plant designs (BWR/6 and prior). These pumps are powered by wet-rotor motors with the housings connected to the bottom of the RPV and eliminating large diameter external recirculation pipes that are possible leakage paths. Construction costs are also reduced. The 10 internal recirculation pumps are located at the bottom of the annulus downcomer region (i.e., between the core shroud and the inside surface of the RPV).

Even though BWRs can operate using only the available natural recirculation thermal pumping head without forced recirculation flow, forced flow is desirable in order to increase the available output from the reactor and as a convenient method to change the reactor output by changing the flow.

Prior to the ABWR, all large commercial nuclear steam supply systems provided by GE from the BWR/3 through the BWR/6 designs used jet pump recirculation systems. These systems have two large recirculation pumps (each up to 9000 Hp) located outside of the reactor pressure vessel (RPV). Each external recirculation pump takes a suction from the bottom of the annulus downcomer region through a large diameter nozzle and discharges through multiple jet pumps inside of the RPV in the annulus downcomer region. There is one nozzle per jet pump for the discharge back into the RPV and the external headers supplying these nozzles. Isolation valves are provided for each of the two external recirculation pumps. In the event of a pipe rupture close to the RPV, those isolation valves will be ineffective and the top region of the reactor may not be covered with water. With all of the jet pumps intact after this design basis accident (DBA)a minimum of two thirds (2/3) of the core will remain covered in water. Calculations indicate that fuel failure would be averted by "steam cooling" wherein the boiling of water in the lower core region will produce mixed quality steam that will absorb heat from the upper core region.

Consequently, internal recirculation pumps eliminate all of the jet pumps (typically 10), all of the external piping, the isolation valves and the large diameter nozzles that penetrated the RPV and needed to suction water from and return it to the RPV. This design therefore reduces the worst leak below the core region to effectively equivalent to a 2 inch diameter leak. The conventional BWR3-BWR6 product line has an analogous potential leak of 24 or more inches in diameter. A major benefit of this design is that it greatly reduces the flow capacity required of the emergency core cooling systems (ECCS). In the event of a fuel failure, a specially constructed basaltic floor with passive cooling features with terminate the flow of corrium before it breaches primary containment.

The first reactors to use internal recirculation pumps were designed by ASEA-Atom (now Westinghouse Electric Company by way of mergers and buyouts, which is owned by Toshiba) and built in Sweden. These plants have operated very successfully for many years.

The internal pumps reduce the required pumping power for the same flow to about half that required with the jet pump system with external recirculation loops. Thus, in addition to the safety and cost improvements due to eliminating the piping, the overall plant thermal efficiency is increased. Eliminating the external recirculation piping also reduces occupational radiation exposure to personnel during maintenance.

A nice operational feature in the ABWR design is electric fine motion control rod drives, first used in the BWRs of AEG (later Kraftwerk Union AG, now AREVA). Older BWRs use a hydraulic locking piston system to move the control rods in six-inch increments. Additionally the fine motion control rod design greatly enhances positive actual control rod position and similarly reduces the risk of a control rod drive accident to the point that no velocity limiter is required at the base of the cruciform control rod blades.

The ABWR is fully automated in response to a loss-of-coolant accident (LOCA), and operator action is not required for 3 days. After 3 days the operators must replenish ECCS water supplies. These and other improvements make the plant significantly safer than previous reactors.

, four ABWRs were in operation in Japan: Kashiwazaki-Kariwa units 6 and 7, which opened in 1996 and 1997, Hamaoka unit 5, opened 2004 having started construction in  2000, and Shika 2 commenced commercial operations on March 15, 2006. Another two, identical to the Kashiwazaki-Kariwa reactors, were nearing completion at Lungmen in Taiwan, and one more (Shimane Nuclear Power Plant 3) had just commenced construction in Japan, with major siteworks to start in 2008 and completion in 2011. Plans for at least six other ABWRs in Japan have been postponed, cancelled, or converted to other reactor types, but three of these (Higashidōri 1 and 2 and Ohma) were still listed as on order by the utilities, with completion dates of 2012 or later.

Several ABWRs are proposed for construction in the United States under the Nuclear Power 2010 Program. However these proposals face fierce competition from more recent designs such as the ESBWR (Economic Simplified BWR, a generation III+ reactor also from GE) and the AP1000 (Advanced, Passive, 1000MWe, from Westinghouse). These designs take passive safety features even further than the ABWR does, as do more revolutionary designs such as the pebble bed modular reactor. However, the US market incentive for construction of an ABWR is that the Nuclear Regulatory Commission (NRC) approved the ABWR design in 1997 and construction would have a smaller regulatory burden for approval; hence ABWRs could be constructed faster than other designs pending approval. There are no ESBWR design reactors in service world wide and the ESBWR design is pending approval by the NRC. The ESBWR is a natural circulation plant with features to be resolved such as the power oscillations expected to the local power induced thermal hydraulic instabilities during initial startup.

On June 19, 2006 NRG Energy filed a Letter Of Intent with the Nuclear Regulatory Commission to build two 1358 MWe ABWRs at the South Texas Project site. [159572] On September 25, 2007, NRG Energy and CPS Energy submitted a Construction and Operations License (COL) request for these plants with the NRC. NRG Energy is a merchant generator and CPS Energy is the nation's largest municipally owned utility.

References and Notes

See also

• Nuclear power
• Nuclear safety in the U.S.
• Economics of new nuclear power plants
• Pressurized water reactor
• Advanced Heavy Water Reactor

External links

• GE Energy's official page for the ABWR
• Engineered Safety Features in the ABWR
• The 1-Week Course of World Nuclear University held in Ankara, Turkey on September 15-19, 2008
• The Energy Information Administration page for new reactor designs.
• ABWR Plant Economics and Project Schedule (old)