Boiling water reactor safety systems

In the limiting case of an ATWS (Anticipated Transient Without Scram) derangement, high neutron power levels (~ 200%) can occur for less than a second, after which actuation of SRVs will cause the pressure to rapidly drop off.

[6] ECCS systems accomplish this by maintaining reactor pressure vessel (RPV) cooling water level, or if that is impossible, by directly flooding the core with coolant.

This allows HPCI to remove steam from the reactor and slowly depressurize it without the need for operating the safety or relief valves.

This minimizes the number of times the relief valves need to operate, and reduces the potential for one sticking open and causing a small LOCA.

During a station blackout (where all off-site power is lost and the diesel generators fail) the RCIC system may be "black started" with no AC and manually activated.

ADS then confirms with the Low Alarm Water Level, verifies at least 1 low-pressure cooling pump is operating, and starts a 105-second timer.

Versioning note: In ABWRs and (E)SBWRs, there are additional water spray systems to cool the drywell and the suppression pool.

Versioning note: ABWRs replace LPCI with low-pressure core flooder (LPCF), which operates using similar principles.

If Level 1 is not resubmerged within 50 seconds after the countdown started, DPVS fires and rapidly vents steam contained within the reactor pressure vessel into the drywell.

(In fact, both the ESBWR and the ABWR are designed so that even in the maximum feasible contingency, the core never loses its layer of water coolant.)

The water within the RPV will boil into steam from the decay heat, and natural convection will cause it to travel upwards into the drywell, into piping assemblies in the ceiling that will take the steam to four large heat exchangers – the Passive Containment Cooling System (PCCS) – located above the drywell – in deep pools of water.

At this point, all that needs to happen is for the pools that cool the PCCS heat exchangers to be refilled, which is a comparatively trivial operation, doable with a portable fire pump and hoses.

The standby liquid control system is designed to deliver the equivalent of 86 gpm of 13% by weight sodium pentaborate solution into a 251-inch BWR reactor vessel.

[10] SLCS, in combination with the alternate rod insertion system, the automatic recirculation pump trip and manual operator actions to reduce water level in the core will ensure that the reactor vessel does not exceed its ASME code limits, the fuel does not suffer core damaging instabilities, and the containment does not fail due to overpressure during high power scram failure.

The SLCS can also be injected during a LOCA or a fuel cladding failure to adjust the ph of the reactor coolant that has spilled, preventing the release of some radioactive materials.

There are five levels of shielding: If every possible measure standing between safe operation and core damage fails, the containment can be sealed indefinitely, and it will prevent any substantial release of radiation to the environment from occurring in nearly any circumstance.

The containment isolation system is responsible for automatically closing these valves to prevent the release of radioactive material and is an important part of a plant's safety analysis.

An example of parameters which are monitored by the isolation system include containment pressure, acoustic or thermal leak detection, differential flow, high steam or coolant flow, low reactor water level, or high radiation readings in the containment building or ventilation system.

This provides redundancy as well as making the system immune to the single failure of any inboard or outboard valve operator or isolation signal.

Within less than a second from power outage, auxiliary batteries and compressed air supplies are starting the Emergency Diesel Generators.

This happens after a few seconds, as the approximately 200,000 L/min (3,300 L/s, 52,500 US gal/min, 875 US gal/s) of water these systems release begin to cool first the top of the core, with LPCI deluging the fuel rods, and CS suppressing the generated steam until at approximately T+100 seconds, all of the fuel is now subject to deluge and the last remaining hot-spots at the bottom of the core are now being cooled.

The peak temperature that was attained was 900 °C (1,650 °F) (well below the maximum of 1,200 °C (2,190 °F) established by the NRC) at the bottom of the core, which was the last hot spot to be affected by the water deluge.

The core is cooled rapidly and completely, and following cooling to a reasonable temperature, below that consistent with the generation of steam, CS is shut down and LPCI is decreased in volume to a level consistent with maintenance of a steady-state temperature among the fuel rods, which will drop over a period of days due to the decrease in fission-product decay heat within the core.

After a few days of LPCI, decay heat will have sufficiently abated to the point that defueling of the reactor is able to commence with a degree of caution.

The ABWR and ESBWR, the most recent models of the BWR, are not vulnerable to anything like this incident in the first place, as they have no liquid penetrations (pipes) lower than several feet above the waterline of the core, and thus, the reactor pressure vessel holds in water much like a deep swimming pool in the event of a feedwater line break or a steam line break.

Fuel rod uncovery will briefly take place, but maximum temperature will only reach 600 °C (1,112 °F), far below the NRC safety limit.

Prior to the Fukushima Daiichi disaster, no incident approaching the DBA or even a LBLOCA in severity had occurred with a BWR [citation needed].

The most severe incident that had previously occurred with a BWR was in 1975 due to a fire caused by extremely flammable urethane foam installed in the place of fireproofing materials at the Browns Ferry Nuclear Power Plant; for a short time, the control room's monitoring equipment was cut off from the reactor, but the reactor shut down successfully, and, as of 2009, is still producing power for the Tennessee Valley Authority, having sustained no damage to systems within the containment.

General Electric defended the design of the reactor, stating that the station blackout caused by the 2011 Tōhoku earthquake and tsunami was a "beyond-design-basis" event which led to Fukushima I nuclear accidents.

As an emergency measure, operators resorted to using firetrucks and salvaged car batteries to inject seawater into the drywell to cool the reactors, but only achieved intermittent success and three cores overheated.