Nuclear power is a very dangerous way to boil water
As I stated before, this is a BWR. Hence the primary coolant loop does boil. This is inherently more dangerous than a modern PWR where the primary coolant is not allowed to boil. In a BWR you have to actively cool the core after shutdown, something that is not the case with a PWR. If these stations were PWR there would be no problem.
Yes, boiling water in your primary coolant is dangerous but nowadays we don't do it. Japan in particular has had a big thing for BWRs and since 1994 the only BWRs opened in the world have been in Japan:http://en.wikipedia.org/wiki/List_of_BWRs
BWR does however have some advantages too:
- The reactor vessel and associated components operate at a substantially lower pressure (about 75 times atmospheric pressure) compared to a PWR (about 158 times atmospheric pressure).
- Pressure vessel is subject to significantly less irradiation compared to a PWR, and so does not become as brittle with age.
- Operates at a lower nuclear fuel temperature.
- Fewer components due to no steam generators and no pressurizer vessel. (Older BWRs have external recirculation loops, but even this piping is eliminated in modern BWRs, such as the ABWR.)
- Lower risk (probability) of a rupture causing loss of coolant compared to a PWR, and lower risk of core damage should such a rupture occur. This is due to fewer pipes, fewer large diameter pipes, fewer welds and no steam generator tubes.
- NRC assessments of limiting fault potentials indicate if such a fault occurred, the average BWR would be less likely to sustain core damage than the average PWR due to the robustness and redundancy of the Emergency Core Cooling System (ECCS).
- Unlike PWRs, BWRs have at least a few steam-turbine driven ECCS systems that can be directly operated by steam produced after a reactor shutdown, and require no electrical power. This results in less dependence on emergency diesel generators—of which there are four—in any event.
- Measuring the water level in the pressure vessel is the same for both normal and emergency operations, which results in easy and intuitive assessment of emergency conditions.
- Can operate at lower core power density levels using natural circulation without forced flow.
- A BWR may be designed to operate using only natural circulation so that recirculation pumps are eliminated entirely. (The new ESBWR design uses natural circulation.)
- BWRs do not use boric acid to control fission burn-up, leading to less possibility of corrosion within the reactor vessel and piping. (Corrosion from boric acid must be carefully monitored in PWRs; it has been demonstrated that reactor vessel head corrosion can occur if the reactor vessel head is not properly maintained. See Davis-Besse. Since BWRs do not utilize boric acid, these contingencies are eliminated.)
- BWRs generally have N-2 redundancy on their major safety-related systems, which normally consist of four "trains" of components. This generally means that up to two of the four components of a safety system can fail and the system will still perform if called upon.
- Due to their single major vendor (GE/Hitachi), the current fleet of BWRs have predictable, uniform designs that, while not completely standardized, generally are very similar to one another. The ABWR/ESBWR designs are completely standardized. Lack of standardization remains a problem with PWRs, as, at least in the United States, there are three design families represented among the current PWR fleet (Combustion Engineering, Westinghouse, and Babcock & Wilcox), within these families, there are quite divergent designs.
- Additional families of PWRs are being introduced. For example, Mitsubishi's APWR, Areva's US-EPR, and Westinghouse's AP1000/AP600 will add diversity and complexity to an already diverse crowd, and possibly cause customers seeking stability and predictability to seek other designs, such as the BWR.
- BWRs are overrepresented in imports, if the importing nation doesn't have a nuclear navy (PWRs are favored by nuclear naval states due to their compact, high-power design used on nuclear-powered vessels; since naval reactors are generally not exported, they cause national skill to be developed in PWR design, construction, and operation), or special national aspirations (special national aspirations lead to a marked preference for the CANDU reactor type due to special features of that type). This may be due to the fact that BWRs are ideally suited for peaceful uses like power generation, process/industrial/district heating, and desalinization, due to low cost, simplicity, and safety focus, which come at the expense of larger size and slightly lower thermal efficiency.
Sweden is standardized mainly on BWRs.
Germany has a large number of BWRs, which are overrepresented in their national fleet compared to the US.
Mexico's only two reactors are BWRs.
Japan experimented with both PWRs and BWRs, but most builds as of late have been of BWRs, specifically ABWRs.
In the CEGB open competition in the early 1960s for a standard design for UK 2nd-generation power reactors, the PWR didn't even make it to the final round, which was a showdown between the BWR (preferred for its easily understood design as well as for being predictable and "boring") and the AGCR, a uniquely British design; the indigenous design won, possibly on technical merits, possibly due to the proximity of a general election.
- Complex calculations for managing consumption of nuclear fuel during operation due to "two phase (water and steam) fluid flow" in the upper part of the core. This requires more instrumentation in the reactor core. The innovation of computers, however, makes this less of an issue.
- Much larger pressure vessel than for a PWR of similar power, with correspondingly higher cost. (However, the overall cost is reduced because a modern BWR has no main steam generators and associated piping.)
- Contamination of the turbine by short-lived activation products. This means that shielding and access control around the steam turbine are required during normal operations due to the radiation levels arising from the steam entering directly from the reactor core. This is a moderately minor concern, as most of the radiation flux is due to Nitrogen-16, which has a half-life measured in seconds, allowing the turbine chamber to be entered into within minutes of shutdown.
- Though the present fleet of BWRs are said to be less likely to suffer core damage from the "1 in 100,000 reactor-year" limiting fault than the present fleet of PWRs are (due to increased ECCS robustness and redundancy) there have been concerns raised about the pressure containment ability of the as-built, unmodified Mark I containment – that such may be insufficient to contain pressures generated by a limiting fault combined with complete ECCS failure that results in extremely severe core damage. In this double failure scenario, assumed to be extremely unlikely prior to the Fukushima I nuclear accidents, an unmodified Mark I containment can allow some degree of radioactive release to occur. This is supposed to be mitigated by the modification of the Mark I containment; namely, the addition of an outgas stack system that, if containment pressure exceeds critical setpoints, is supposed to allow the orderly discharge of pressurizing gases after the gases pass through activated carbon filters designed to trap radionuclides.
- A BWR requires active cooling for a period of several hours to several days following shutdown, depending on its power history. Full insertion of BWRs control rods safely shuts down the primary nuclear reaction. However, radioactive decay of the fission products in the fuel will continue to actively generate decay heat at a gradually decreasing rate, requiring pumping of cooling water for an initial period to prevent overheating of the fuel. If active cooling fails during this post-shutdown period, the reactor can still overheat to a temperature high enough that zirconium in the fuel cladding will react with water and steam, producing hydrogen gas. In this event there is a high danger of hydrogen explosions, threatening structural damage to the reactor and/or associated safety systems and/or the exposure of highly radioactive spent fuel rods that may be stored in the reactor building (approx 15 tons of fuel is replenished each year to maintain normal BWR operation) as happened with the Fukushima I nuclear accidents.
- Control rods are inserted from below for current BWR designs. There are two available hydraulic power sources that can drive the control rods into the core for a BWR under emergency conditions. There is a dedicated high pressure hydraulic accumulator and also the pressure inside of the reactor pressure vessel available to each control rod. Either the dedicated accumulator (one per rod) or reactor pressure is capable of fully inserting each rod. Most other reactor types use top entry control rods that are held up in the withdrawn position by electromagnets, causing them to fall into the reactor by gravity if power is lost.
I do however doubt that the Japenese installations have the level of redundancy that BWRs elsewhere have.
More efficient still (and safer) is the AGR design by some margin.