ABWR passes UK GDA step 2, but full PSA needed 3 September 2014 3 September 2014

A pre-construction design review of the Hitachi-GE ABWR in the UK has proceeded to the next phase. Following a preliminary phase that ended in January 2014, the reactor design completed an initial technical assessment in August. But there remains lots more work to do before final approval, tentatively scheduled for December 2017.

"At the end of our Step 2 assessments we have not identified any fundamental safety or security issues that might prevent issue of a DAC or that would need to be addressed in order to acquire one," wrote the UK regulator, the Office for Nuclear Regulation. The reactor is also being reviewed by the UK Environment Agency, which said it has not found any 'obviously unacceptable' issues so far.

The next step of the process is a year-long analysis of the design at the systems level and primarily focuses on the safety arguments. The final step is the detailed assessment phase and focuses on the evidence provided by Hitachi-GE in the safety analysis, on a sampling basis, and typically takes around 28 months.

Despite the positive assesment, several technical issues did come up during the step 2 review.

First was an issue about common-cause failure in the reactor's instrumentation and control systems, an issue that has also affected the UK EPR and Westinghouse AP1000 designs that have previously passed through the process. ONR reports: "During the early stages of ONR's assessment, a potential shortfall in the diversity between the safety system and logic control (SSLC) platform technology and other control systems was identified as a regulatory concern. Following extensive engagement with the RP, it has committed to modify the technology of the SSLC to be diverse from other control systems for the UK ABWR. It has also agreed to enhance the isolation of its SSLC from the other control systems, and also to provide additional isolation of the plant computer control system from more general nuclear power station computer networks through the use of one way data diodes."

Second, ONR has raised concerns about Hitachi-GE's probabilistic safety assessment work for the reactor design. ONR reports: "The bounding CDF estimate could result in risk figures that would not meet ONR's expectations for new reactors when compared against the numerical targets in the SAPs. Although Hitachi-GE has indicated that this evaluation is conservative, the analyses provided are simplified and appear to be incomplete. At this point ONR do not have sufficient information to properly understand the risk profile for the UK ABWR, as this requires a full scope, modern standards PSA."

As a result, ONR said that it would request Hitachi-GE to develop a detailed PSA programme and submit the PSA models, data, supporting analysis and accompanying documentation throughout Steps 3 and 4.

In the report, ONR published an interesting history of the ABWR and an overview of its design.

"The development of the Advanced Boiling Water Reactor (ABWR) began in 1978, and was first adopted in the construction of the Kashiwazaki-Kariwa Nuclear Power Station Unit 6 and Unit 7, which commenced commercial operation in 1996 and 1997. At full power, a single ABWR reactor produces around 1350MWe of electricity.

"The BWR uses demineralised water as a coolant and neutron moderator. Heat is produced by nuclear fission in the reactor core, and this causes the cooling water to boil, producing steam. The steam is directly used to drive a turbine, after which the steam is cooled in a condenser and converted back to water. This water is then returned to the reactor core, completing the loop. The cooling water is maintained at about 7.5 MPa, or slightly lower, so that it boils in the core at about 285°C. This is fundamentally different from other reactor types which do not use this direct cycle of steam to drive the turbine generators.

"The basic layout of an ABWR comprises a reactor building, control building and turbine building, the configuration of which is site dependant, however they are located immediately adjacent to each other.

"The major part of the reactor building is the Reinforced Concrete Containment Vessel (RCCV), which contains the Reactor Pressure Vessel (RPV). The RCCV is a steel lined reinforced concrete structure, cylindrical in nature, 36m tall, 29m in diameter and with 2m thick walls. It has two principal functions, to contain pressure and prevent leakage. The reinforced concrete handles the functions of pressure containment and shielding, and the liner handles the function of leakage prevention. The RCCV is divided into a drywell and a suppression chamber by the diaphragm floor and the RPV pedestal. The suppression chamber contains the suppression pool and an air space. Vapour flows, which could be generated by a Loss of Coolant Accident (LOCA), flow from the drywell space to the suppression pool through horizontal vent pipes embedded into the RPV pedestal, where the steam is condensed.

"The RPV is a cylindrical steel vessel that contains the core and reactor internals. The RPV consists of a removable hemispherical top head, cylindrical shells, a bottom head, and some nozzles. The RPV is installed vertically on the pedestal inside the containment building. The RPV is around 21 metres in height, 7.4 metres in diameter and with a steel wall thickness of around 17 centimetres. The RPV functions as the pressure retaining barrier to retain light water coolant and as the barrier to isolate radioactive material generated in the core from outside the RPV. The vessel contains the core, steam separator, steam dryer, reactor internal pumps and the control rod arrangements.

"The reactor core is an upright cylinder containing 872 fuel assemblies. A fuel assembly has a square array of fuel rods and a hollow pipe (water rod), where water coolant flows. Fuel assemblies are placed inside a zircaloy channel box. Its functions include forming the coolant flow path and guiding the insertion and withdrawal of control rods between fuel assemblies. Each fuel rod is made out of Uranium Dioxide (UO2) pellets with less than 5wt% Uranium-235 enrichment and a Zircaloy-2 cladding tube; both ends of which have plugs welded on. Its plenum is filled with Helium gas. At the end of a fuel cycle, the spent fuel will contain radioactive fission products which remain confined inside the cladding.

"The control rods are cruciform and are inserted between every 4 fuel assembles. They perform the twin functions of power distribution shaping and reactivity control. The control rods enter the vessel through the bottom dome and are inserted under either electric or hydraulic power.

"During normal operation, steam generated at the reactor is transferred to the turbine facility via four main steam pipes. In order to prevent overpressure of the reactor following a transient in operation or an accident, steam inside the reactor is discharged into the suppression pool by the relief valve function or the safety valve function of the Safety Relief Valves (SRVs). Main steam isolation valves are arranged in pairs (either side of the containment wall) to isolate the reactor in case of fuel failure, Main Steam Line Break Accident (MSLBA) or LOCAs.

"The ABWR also has a standby liquid control system which acts as a second line of criticality control. If required it can inject borated water directly into the RPV, which will bring the core to a sub-critical state and maintain it as it cools.

"The ABWR safety systems include three independent divisions of Emergency Core Cooling Systems (ECCS). Each division of the ECCS has one high pressure and one low pressure make-up system. Two of the three pumps for the high pressure system are electrically driven and one is steam driven whereas for the low pressure system all three are electrically driven and these low pressure pumps also perform the residual heat removal function. For the high pressure system the choice of whether to use steam or electrically driven pumps is dependent on the accident transient. Heat is removed from the reactor during normal shutdown, reactor isolation or loss of cooling accidents via the residual heat removal system. Three individual cooling loops are available with heat exchanges cooled via the reactor cooling water system. There are also diverse pressure control and emergency cooling systems located in the back- building.

"The primary volume of the RCCV is maintained in an inert state by the use of a nitrogen atmosphere, which reduces the likelihood of a combustion if a leak of hydrogen occurs. In addition, there is a flammability control system which uses hydrogen recombiners to control the build up of a mixture of hydrogen and oxygen in the RCCV should a LOCA occur.

"In the event of a loss of off-site electrical power the ABWR has three independent emergency diesel generators to provide power to the essential safety systems. It also has two additional back-up diesel generators supporting a diverse electrical power supply system in the back-up building located away from the main reactor, control and turbine buildings.

"Fuel is inserted and removed from the RPV via the fuelling machine which is located on the operating floor. Spent fuel is placed in the spent fuel pool immediately after removal from the RPV and kept there until it has cooled sufficiently to allow placement into fuel casks. The casks are loaded at the operating floor level and lowered to ground level before dispatch."

Picture: ABWR reactor building cutaway diagram from ONR report

