Fukushima: Background on Reactors

(Updated February 2012)

The Fukushima Daiichi reactors are GE boiling water reactors (BWR) of an early (1960s) design supplied by GE, Toshiba and Hitachi, with what is known as a Mark I containment. Reactors 1-3 came into commercial operation 1971-75. Reactor power is 460 MWe for unit 1, 784 MWe for units 2-5, and 1100 MWe for unit 6. The fuel assemblies are about 4 m long, and there are 400 in unit 1, 548 in units 2-5, and 764 in unit 6. Each assembly has 60 fuel rods containing the uranium oxide fuel within zirconium alloy cladding. Unit 3 has a partial core of mixed-oxide (MOX) fuel (32 MOX assemblies, 516 LEU). They all operate normally at 286°C at core outlet under a pressure of 6930 kPa and with 115-130 kPa pressure in dry containment. The operating pressure is about half that in a PWR. NISA says maximum design base pressure for reactor pressure vessels (RPV) is 8240 kPa at 300°C, and for containment (PCV) is about 500 kPa*.

* NISA gives 430 kPa for unit 1 and 380 kPa for 2-3 at 140°C as 'maximum', apparently gauge pressure, so add 101 for absolute: 530 and 480 kPa. Before venting, unit 1 RPV got to 900 kPa and PCV to 850 kPa early on 12th.

The BWR Mark I has a Primary Containment system comprising a free-standing bulb-shaped drywell of 30 mm steel backed by a reinforced concrete shell, and connected to a torus-shaped wetwell beneath it containing the suppression pool (with 3000 m3 of water in units 2-5). The drywell, also known as the Primary Containment Vessel (PCV), contains the reactor pressure vessel (RPV). For simplicity, we will use the term 'dry containment' here. The water in the suppression pool acts as an energy-absorbing medium in the event of an accident. The wetwell is connected to the dry containment by a system of vents, which discharge under the suppression pool water in the event of high pressure in the dry containment. The function of the primary containment system is to contain the energy released during any loss-of-coolant accident (LOCA) of any size reactor coolant pipe, and to protect the reactor from external assaults. The Japanese version of the Mark I is slightly larger than the original GE version.

During normal operation, the dry containment atmosphere and the wetwell atmosphere are filled with inert nitrogen, and the wetwell water is at ambient temperature. A small amount of hydrogen is routinely formed by radiolytic decay of water, and this is normally dealt with by recombiners in the containment vessel. They would be insufficient for countering major hydrogen formation due to oxidation of zirconium fuel cladding. Apart from this, at low containment pressures hydrogen and other gases are routinely vented through charcoal filters which trap most radionuclides.

If a loss of coolant accident (LOCA) occurs, steam flows from the dry containment (drywell) through a set of vent lines and pipes into the suppression pool, where the steam is condensed. Steam can also be released from the reactor vessel through the safety relief valves and associated piping directly into the suppression pool. Steam will be condensed in the wetwell, but hydrogen and noble gases are not condensable and will pressurise the system, as will steam if the wetwell water is boiling. In this case emergency systems will activate to cool the wetwell, see below. Excess pressure from the wetwell (above 300 kPa) can be vented through the 120 m emission stack via a hardened pipe or into the secondary containment above the reactor service floor of the building. If there has been fuel damage, vented gases will include noble gases (krypton & xenon), iodine and caesium, the latter being scrubbed in some scenarios. Less volatile elements in any fission product release will plate out in the containment. (The later Mark II containments are similar to Mark I, but both are much smaller than the Mark III and those which became standard in PWRs.)

The secondary containment houses the emergency core cooling systems and the spent/ used fuel pool. It is not designed to contain high pressure.

Decay Heat Removal

The primary cooling circuit of the BWR takes steam from above the core, in the reactor pressure vessel, to the turbine in an adjacent building. After driving the turbines it is condensed and the water is returned to the pressure vessel by powerful steam-driven pumps. There are also two powerful jet-pump recirculation systems forcing water down around the reactor core and shroud. When the reactor is shut down, the steam in the main circuit is diverted via a bypass line directly to the condensers, and the heat is dumped there, to the sea. In both situations a steam-driven turbine drives the pumps, at least until the pressure drops to about 450 kPa (50 psig), but condenser function depends on large electrically-driven pumps for the seawater which are not backed up by the diesel generators.

In shutdown mode at low pressure, the Residual Heat Removal (RHR) system then operates in a secondary circuit (RHR is connected into the two jet-pump recirculation circuits), driven by smaller electric pumps, and circulates water from the pressure vessel to RHR heat exchangers which dump the heat to the sea, using external electric pumps in the secondary circuit. This RHR system is fully supported by the diesel generators. Unit 1 had an isolation condenser (IC) for passive core cooling, with reactor steam going to an external condenser, and it needed only DC battery power to operate. Units 2-5 have a Reactor Core Isolation Cooling (RCIC) actuated automatically which can provide make-up water to the reactor vessel (without any heat removal circuit). It is driven by a small steam turbine using steam from decay heat, injecting water from a condensate storage tank or the suppression pool and controlled by the DC battery system. The RCIC systems played a helpful role in units 2 & 3 until the suppression pool water boiled, to 11am on 12th in unit 3, and to 2pm on 14th in unit 2.

Then there is an Emergency Core Cooling System (ECCS) as further back-up for loss of coolant. It has high-pressure and low-pressure elements. The high pressure coolant injection (HPCI) system in units 1-3 has pumps powered by steam turbines which are deigned to work over a wide pressure range. The HPCI draws water from the large torus suppression chamber beneath the reactor as well as a water storage tank, and requires only DC battery power. For use below about 700 kPa, there is also a Low-Pressure Coolant Injection (LPCI) mode through the RHR system but utilising suppression pool water, and a core spray system, all electrically-driven. All ECCS sub-systems require some power to operate valves etc, and the battery back-up to generators may provide this.

Beyond these original systems, Tepco in 1990s installed provision for water injection via the fire extinguisher system through the RHR system (injecting via the jet-pump nozzles) as part of it Severe Accident Management (SAM) countermeasures. Air-cooled diesel generators were installed at Daiichi 2, 4 & 6 - the last being the only one to survive the tsunami.

The Fukushima reactors have much of their switchgear on the ground floor in the turbine buildings rather than elevated, as at some similar US plants. Also they have control rooms with analogue instrumentation typical of the period, so not only did many instruments fail, but data could not be downloaded and accessed remotely to assist diagnosis and remedial action.


A frequently-voiced concern during the first week of the accident was regarding fuel meltdown. This starts to occur if the fuel itself reached 2500°C (or more, up to 2800°C, depending on make-up). At this point, the fuel rods slump within the assemblies. Conceivably, the “corium” (a mixture of molten cladding, fuel, and structural steel) drops to the bottom of the reactor vessel. If the hot fuel or cladding is exposed to cooling water en route, it may solidify and fracture, falling to the bottom of the reactor vessel. Given that the melting point of the steel reactor vessel is about 1500°C, there is an obvious possibility on the corium penetrating the steel if it remains hot enough.* In any case the BWR pressure vessels have numerous penetrations at the bottom for control rods and instrumentation, so any corium, to the extent it remained molten, would possibly shower into the bottom of the drywell containment. The whole fuel melt scenario is much more probable with a sudden major loss of coolant when the reactor is at full power than in the Fukushima situation, at least beyond the first few days. Before fuel melting, cladding cracks at about 1200°C, its oxidation begins at about 1300°C (releasing hydrogen from steam) and the zirconium cladding melts at about 1850°C and reacts with uranium oxide to form a molten eutectic, which would release volatile fission products such as caesium. These temperatures are quite possible days after shutdown in the absence of cooling.

* In the 1979 US Three Mile Island accident, 19 tonnes of corium reached the bottom of the pressure vessel without causing any apparent damage, after about half the core melted. Metallurgical examination suggests that the 127 mm thick pressure vessel steel glowed red-hot for an hour.

Oxidation of the zirconium cladding in the presence of steam produces hydrogen exothermically, with 5.8 MJ/kg of Zr from this exacerbating the fuel decay heat problem. There is some potential for this to become self-sustaining at high temperatures, giving rise to a zirconium cladding fire with a burn front along the axis of the fuel rods. Such a fire is possible in a spent fuel pond following major loss of coolant from leakage or boiling.

You may also be interested in