Advanced boiling water reactor

Boiling water reactors (BWRs) are the second most common form of light water reactor with a direct cycle design that uses fewer large steam supply components than the pressurized water reactor (PWR), which employs an indirect cycle. 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 complete and operating. The first reactors were built on time and under budget in Japan, with others under construction there and in Taiwan. ABWRs were on order in the United States, including two reactors at the South Texas Project site. The projects in both Taiwan and US are reported to be over-budget.

The standard ABWR plant design has a net electrical output of about 1.35 GW, generated from about 3926 MW of thermal power.

Overview of the design

The ABWR represents an evolutionary route for the BWR family, with numerous changes and improvements to previous BWR designs.

Major areas of improvement include:

• The addition of reactor internal pumps (RIP) mounted on the bottom of the reactor pressure vessel (RPV) – 10 in total – which achieve improved performance while eliminating large recirculation pumps in containment and associated large-diameter and complex piping interfaces with 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 nominal capacity of 6912 m³/h.
• The control rod adjustment capabilities have been supplemented with the addition of an electro-hydraulic Fine Motion Control Rod Drive (FMCRD), allowing for fine position adjustment using an electrical motor, while not losing the reliability or redundancy of traditional hydraulic systems which are designed to accomplish rapid shutdown in 2.80 s from receipt of an initiating signal, or ARI (alternate rod insertion) 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 (RPS) (with redundant digital backups as well as redundant manual backups) ensures a high level of reliability and simplification for safety condition detection and response. This system initiates rapid hydraulic insertion of control rods for shutdown (known as SCRAM by nuclear engineers) when needed. Two-out-of-four per parameter rapid shutdown logic ensures that nuisance rapid shutdowns are not triggered by single instrument failures. RPS can also trigger ARI, FMCRD rod run-in to shut down the nuclear chain reaction. The standby liquid control system (SLCS) actuation is provided as diverse logic in the unlikely event of an Anticipated Transient Without Scram.
• Fully digital reactor controls (with redundant digital backup and redundant manual (analog) backups) allow the control room to easily and rapidly control plant operations and processes. Separate redundant safety and non-safety related digital multiplexing buses allow for reliability and diversity of instrumentation and control.
  • In particular, the reactor is automated for startup (i.e., initiate the nuclear chain reaction and ascent to power) and for standard shutdown using automatic systems only. Of course, human operators remain essential to reactor control and supervision, but much of the busy-work of bringing the reactor to power and descending from power can be automated at operator discretion.
• 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), eight 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 (safety-related) power bus is still powered by 3 highly-reliable emergency diesel generators that are safety related, 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.
  • 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 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 basaltic reinforced concrete pad under the RPV that will both catch and hold any heated core melt that might fall on that pad in extraordinarily contingent situations. In addition, there are several fusible links within the wall separating the wetwell from the lower drywell that flood the pad using the wetwell's water supply, ensuring cooling of that area even with the failure of standard mitigation systems.
• The containment has been significantly improved over the conventional Mark I type. Like the conventional Mark I type, 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 enclosed in the wetwell (or torus in the case of the Mark I), the low temperature of which will condense the steam back into liquid water. This will keep containment 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 .3G; further, it is designed to withstand a tornado with >320 mph wind speed. Seismic hardening is possible in earthquake-prone areas and has been done at the Lungmen facility in Taiwan which has been hardened up 0.4 g in any direction. The containment is inerted with nitrogen before operation to prevent fires, and can be deinerted after reactor shutdown for maintenance.
• The ABWR is designed for a lifetime of at least 60 years. 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 GEH's Probabilistic Risk Assessment, a core damage event would occur no more often than once in six million years as the core damage frequency (CDF) of the ABWR is 1.6×10⁻⁷, second in lowest CDF probability to the ESBWR.

The RPV and Nuclear Steam Supply System (NSSS) have significant improvements, such as the substitution of RIPs, eliminating conventional external recirculation piping loops and pumps in the containment that in turn drive jet pumps producing forced flow in the RPV. RIPs provide significant improvements related to reliability, performance and maintenance, including a reduction in occupational radiation exposure related to containment activities during maintenance outages. 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. 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). Consequently, internal recirculation pumps eliminate all of the jet pumps in the RPV, all of the large external recirculation loop pumps and 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 (51 mm) 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 ECCS.

The first reactors to use internal recirculation pumps were designed by ASEA-Atom (now Westinghouse Electric Company by way of mergers and buyouts, which was 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.

An 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. The electric 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.