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Radioactive waste

From Wikipedia, the free encyclopedia

https://en.wikipedia.org/wiki/Radioactive_waste

Radioactive waste is a type of hazardous waste that contains radioactive material. Radioactive waste is a result of many activities, including nuclear medicine, nuclear research, nuclear power generation, nuclear decommissioning, rare-earth mining, and nuclear weapons reprocessing. The storage and disposal of radioactive waste is regulated by government agencies in order to protect human health and the environment.

Radioactive waste is broadly classified into 3 categories: low-level waste (LLW), such as paper, rags, tools, clothing, which contain small amounts of mostly short-lived radioactivity; intermediate-level waste (ILW), which contains higher amounts of radioactivity and requires some shielding; and high-level waste (HLW), which is highly radioactive and hot due to decay heat, thus requiring cooling and shielding.

In nuclear reprocessing plants about 96% of spent nuclear fuel is recycled back into uranium-based and mixed-oxide (MOX) fuels. The residual 4% is minor actinides and fission products the latter of which are a mixture of stable and quickly decaying (most likely already having decayed in the spent fuel pool) elements, medium lived fission products such as strontium-90 and caesium-137 and finally seven long-lived fission products with half lives in the hundreds of thousands to millions of years. The minor actinides meanwhile are heavy elements other than uranium and plutonium which are created by neutron capture. Their half lives range from years to millions of years and as alpha emitters they are particularly radiotoxic. While there are proposed - and to a much lesser extent current - uses of all those elements, commercial scale reprocessing using the PUREX-process disposes of them as waste together with the fission products. The waste is subsequently converted into a glass-like ceramic for storage in a deep geological repository.

The time radioactive waste must be stored for depends on the type of waste and radioactive isotopes it contains. Short-term approaches to radioactive waste storage have been segregation and storage on the surface or near-surface. Burial in a deep geological repository is a favored solution for long-term storage of high-level waste, while re-use and transmutation are favored solutions for reducing the HLW inventory. Boundaries to recycling of spent nuclear fuel are regulatory and economic as well as the issue of radioactive contamination if chemical separation processes cannot achieve a very high purity. Furthermore, elements may be present in both useful and troublesome isotopes, which would require costly and energy intensive isotope separation for their use - a currently uneconomic prospect.

A summary of the amounts of radioactive waste and management approaches for most developed countries are presented and reviewed periodically as part of a joint convention of the International Atomic Energy Agency (IAEA).

Nature and significance

A quantity of radioactive waste typically consists of a number of radionuclides, which are unstable isotopes of elements that undergo decay and thereby emit ionizing radiation, which is harmful to humans and the environment. Different isotopes emit different types and levels of radiation, which last for different periods of time.

Physics

Decay energy is split among β, neutrino, and γ if any.

Per 65 thermal neutron fissions of ²³⁵U and 35 of ²³⁹Pu.

Has decay energy 380 keV, but its decay product ¹²⁶Sb has decay energy 3.67 MeV.

The radioactivity of all radioactive waste weakens with time. All radionuclides contained in the waste have a half-life — the time it takes for half of the atoms to decay into another nuclide. Eventually, all radioactive waste decays into non-radioactive elements (i.e., stable nuclides). Since radioactive decay follows the half-life rule, the rate of decay is inversely proportional to the duration of decay. In other words, the radiation from a long-lived isotope like iodine-129 will be much less intense than that of a short-lived isotope like iodine-131. The two tables show some of the major radioisotopes, their half-lives, and their radiation yield as a proportion of the yield of fission of uranium-235.

The energy and the type of the ionizing radiation emitted by a radioactive substance are also important factors in determining its threat to humans. The chemical properties of the radioactive element will determine how mobile the substance is and how likely it is to spread into the environment and contaminate humans. This is further complicated by the fact that many radioisotopes do not decay immediately to a stable state but rather to radioactive decay products within a decay chain before ultimately reaching a stable state.

Pharmacokinetics

Exposure to radioactive waste may cause health impacts due to ionizing radiation exposure. In humans, a dose of 1 sievert carries a 5.5% risk of developing cancer, and regulatory agencies assume the risk is linearly proportional to dose even for low doses. Ionizing radiation can cause deletions in chromosomes. If a developing organism such as a fetus is irradiated, it is possible a birth defect may be induced, but it is unlikely this defect will be in a gamete or a gamete-forming cell. The incidence of radiation-induced mutations in humans is small, as in most mammals, because of natural cellular-repair mechanisms, many just now coming to light. These mechanisms range from DNA, mRNA and protein repair, to internal lysosomic digestion of defective proteins, and even induced cell suicide—apoptosis

Depending on the decay mode and the pharmacokinetics of an element (how the body processes it and how quickly), the threat due to exposure to a given activity of a radioisotope will differ. For instance, iodine-131 is a short-lived beta and gamma emitter, but because it concentrates in the thyroid gland, it is more able to cause injury than caesium-137 which, being water soluble, is rapidly excreted through urine. In a similar way, the alpha emitting actinides and radium are considered very harmful as they tend to have long biological half-lives and their radiation has a high relative biological effectiveness, making it far more damaging to tissues per amount of energy deposited. Because of such differences, the rules determining biological injury differ widely according to the radioisotope, time of exposure, and sometimes also the nature of the chemical compound which contains the radioisotope.

Sources

Radioactive waste comes from a number of sources. In countries with nuclear power plants, nuclear armament, or nuclear fuel treatment plants, the majority of waste originates from the nuclear fuel cycle and nuclear weapons reprocessing. Other sources include medical and industrial wastes, as well as naturally occurring radioactive materials (NORM) that can be concentrated as a result of the processing or consumption of coal, oil, and gas, and some minerals, as discussed below.

Nuclear fuel cycle

Front end

Waste from the front end of the nuclear fuel cycle is usually alpha-emitting waste from the extraction of uranium. It often contains radium and its decay products.

Uranium dioxide (UO₂) concentrate from mining is a thousand or so times as radioactive as the granite used in buildings. It is refined from yellowcake (U₃O₈), then converted to uranium hexafluoride gas (UF₆). As a gas, it undergoes enrichment to increase the U-235 content from 0.7% to about 4.4% (LEU). It is then turned into a hard ceramic oxide (UO₂) for assembly as reactor fuel elements.

The main by-product of enrichment is depleted uranium (DU), principally the U-238 isotope, with a U-235 content of ~0.3%. It is stored, either as UF₆ or as U₃O₈. Some is used in applications where its extremely high density makes it valuable such as anti-tank shells, and on at least one occasion even a sailboat keel. It is also used with plutonium for making mixed oxide fuel (MOX) and to dilute, or downblend, highly enriched uranium from weapons stockpiles which is now being redirected to become reactor fuel.

Back end

The back-end of the nuclear fuel cycle, mostly spent fuel rods, contains fission products that emit beta and gamma radiation, and actinides that emit alpha particles, such as uranium-234 (half-life 245 thousand years), neptunium-237 (2.144 million years), plutonium-238 (87.7 years) and americium-241 (432 years), and even sometimes some neutron emitters such as californium (half-life of 898 years for californium-251). These isotopes are formed in nuclear reactors.

It is important to distinguish the processing of uranium to make fuel from the reprocessing of used fuel. Used fuel contains the highly radioactive products of fission (see high-level waste below). Many of these are neutron absorbers, called neutron poisons in this context. These eventually build up to a level where they absorb so many neutrons that the chain reaction stops, even with the control rods completely removed. At that point, the fuel has to be replaced in the reactor with fresh fuel, even though there is still a substantial quantity of uranium-235 and plutonium present. In the United States, this used fuel is usually "stored", while in other countries such as Russia, the United Kingdom, France, Japan, and India, the fuel is reprocessed to remove the fission products, and the fuel can then be re-used. The fission products removed from the fuel are a concentrated form of high-level waste as are the chemicals used in the process. While most countries reprocess the fuel carrying out single plutonium cycles, India is planning multiple plutonium recycling schemes and Russia pursues closed cycle.

Fuel composition and long term radioactivity

Activity of U-233 for three fuel types. In the case of MOX, the U-233 increases for the first 650 thousand years as it is produced by the decay of Np-237 which was created in the reactor by absorption of neutrons by U-235.

Total activity for three fuel types. In region 1, there is radiation from short-lived nuclides, in region 2, from Sr-90 and Cs-137, and on the far right, the decay of Np-237 and U-233.

The use of different fuels in nuclear reactors results in different spent nuclear fuel (SNF) composition, with varying activity curves. The most abundant material being U-238 with other uranium isotopes, other actinides, fission products and activation products.

Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different.

An example of this effect is the use of nuclear fuels with thorium. Th-232 is a fertile material that can undergo a neutron capture reaction and two beta minus decays, resulting in the production of fissile U-233. The SNF of a cycle with thorium will contain U-233. Its radioactive decay will strongly influence the long-term activity curve of the SNF around a million years. A comparison of the activity associated to U-233 for three different SNF types can be seen in the figure on the top right. The burnt fuels are thorium with reactor-grade plutonium (RGPu), thorium with weapons-grade plutonium (WGPu), and Mixed oxide fuel (MOX, no thorium). For RGPu and WGPu, the initial amount of U-233 and its decay around a million years can be seen. This has an effect on the total activity curve of the three fuel types. The initial absence of U-233 and its daughter products in the MOX fuel results in a lower activity in region 3 of the figure on the bottom right, whereas for RGPu and WGPu the curve is maintained higher due to the presence of U-233 that has not fully decayed. Nuclear reprocessing can remove the actinides from the spent fuel so they can be used or destroyed (see Long-lived fission product § Actinides).

Proliferation concerns

Since uranium and plutonium are nuclear weapons materials, there have been proliferation concerns. Ordinarily (in spent nuclear fuel), plutonium is reactor-grade plutonium. In addition to plutonium-239, which is highly suitable for building nuclear weapons, it contains large amounts of undesirable contaminants: plutonium-240, plutonium-241, and plutonium-238. These isotopes are extremely difficult to separate, and more cost-effective ways of obtaining fissile material exist (e.g., uranium enrichment or dedicated plutonium production reactors).

High-level waste is full of highly radioactive fission products, most of which are relatively short-lived. This is a concern since if the waste is stored, perhaps in deep geological storage, over many years the fission products decay, decreasing the radioactivity of the waste and making the plutonium easier to access. The undesirable contaminant Pu-240 decays faster than the Pu-239, and thus the quality of the bomb material increases with time (although its quantity decreases during that time as well). Thus, some have argued, as time passes, these deep storage areas have the potential to become "plutonium mines", from which material for nuclear weapons can be acquired with relatively little difficulty. Critics of the latter idea have pointed out the difficulty of recovering useful material from sealed deep storage areas makes other methods preferable. Specifically, high radioactivity and heat (80 °C in surrounding rock) greatly increase the difficulty of mining a storage area, and the enrichment methods required have high capital costs.

Pu-239 decays to U-235 which is suitable for weapons and which has a very long half-life (roughly 10⁹ years). Thus plutonium may decay and leave uranium-235. However, modern reactors are only moderately enriched with U-235 relative to U-238, so the U-238 continues to serve as a denaturation agent for any U-235 produced by plutonium decay.

One solution to this problem is to recycle the plutonium and use it as a fuel e.g. in fast reactors. In pyrometallurgical fast reactors, the separated plutonium and uranium are contaminated by actinides and cannot be used for nuclear weapons.

Nuclear weapons decommissioning

Waste from nuclear weapons decommissioning is unlikely to contain much beta or gamma activity other than tritium and americium. It is more likely to contain alpha-emitting actinides such as Pu-239 which is a fissile material used in bombs, plus some material with much higher specific activities, such as Pu-238 or Po.

In the past the neutron trigger for an atomic bomb tended to be beryllium and a high activity alpha emitter such as polonium; an alternative to polonium is Pu-238. For reasons of national security, details of the design of modern bombs are normally not released to the open literature.

Some designs might contain a radioisotope thermoelectric generator using Pu-238 to provide a long-lasting source of electrical power for the electronics in the device.

It is likely that the fissile material of an old bomb which is due for refitting will contain decay products of the plutonium isotopes used in it, these are likely to include U-236 from Pu-240 impurities, plus some U-235 from decay of the Pu-239; due to the relatively long half-life of these Pu isotopes, these wastes from radioactive decay of bomb core material would be very small, and in any case, far less dangerous (even in terms of simple radioactivity) than the Pu-239 itself.

The beta decay of Pu-241 forms Am-241; the in-growth of americium is likely to be a greater problem than the decay of Pu-239 and Pu-240 as the americium is a gamma emitter (increasing external-exposure to workers) and is an alpha emitter which can cause the generation of heat. The plutonium could be separated from the americium by several different processes; these would include pyrochemical processes and aqueous/organic solvent extraction. A truncated PUREX type extraction process would be one possible method of making the separation. Naturally occurring uranium is not fissile because it contains 99.3% of U-238 and only 0.7% of U-235.

Legacy waste

Due to historic activities typically related to the radium industry, uranium mining, and military programs, numerous sites contain or are contaminated with radioactivity. In the United States alone, the Department of Energy states there are "millions of gallons of radioactive waste" as well as "thousands of tons of spent nuclear fuel and material" and also "huge quantities of contaminated soil and water." Despite copious quantities of waste, the DOE has stated a goal of cleaning all presently contaminated sites successfully by 2025. The Fernald, Ohio site for example had "31 million pounds of uranium product", "2.5 billion pounds of waste", "2.75 million cubic yards of contaminated soil and debris", and a "223 acre portion of the underlying Great Miami Aquifer had uranium levels above drinking standards." The United States has at least 108 sites designated as areas that are contaminated and unusable, sometimes many thousands of acres. DOE wishes to clean or mitigate many or all by 2025, using the recently developed method of geomelting, however the task can be difficult and it acknowledges that some may never be completely remediated. In just one of these 108 larger designations, Oak Ridge National Laboratory, there were for example at least "167 known contaminant release sites" in one of the three subdivisions of the 37,000-acre (150 km²) site. Some of the U.S. sites were smaller in nature, however, cleanup issues were simpler to address, and DOE has successfully completed cleanup, or at least closure, of several sites.

Medicine

Radioactive medical waste tends to contain beta particle and gamma ray emitters. It can be divided into two main classes. In diagnostic nuclear medicine a number of short-lived gamma emitters such as technetium-99m are used. Many of these can be disposed of by leaving it to decay for a short time before disposal as normal waste. Other isotopes used in medicine, with half-lives in parentheses, include:

Y-90, used for treating lymphoma (2.7 days)
I-131, used for thyroid function tests and for treating thyroid cancer (8.0 days)
Sr-89, used for treating bone cancer, intravenous injection (52 days)
Ir-192, used for brachytherapy (74 days)
Co-60, used for brachytherapy and external radiotherapy (5.3 years)
Cs-137, used for brachytherapy and external radiotherapy (30 years)
Tc-99, product of the decay of Technetium-99m (221,000 years)

Industry

Industrial source waste can contain alpha, beta, neutron or gamma emitters. Gamma emitters are used in radiography while neutron emitting sources are used in a range of applications, such as oil well logging.

Naturally occurring radioactive material

Substances containing natural radioactivity are known as NORM (naturally occurring radioactive material). After human processing that exposes or concentrates this natural radioactivity (such as mining bringing coal to the surface or burning it to produce concentrated ash), it becomes technologically enhanced naturally occurring radioactive material (TENORM). A lot of this waste is alpha particle-emitting matter from the decay chains of uranium and thorium. The main source of radiation in the human body is potassium-40 (⁴⁰K), typically 17 milligrams in the body at a time and 0.4 milligrams/day intake. Most rocks, especially granite, have a low level of radioactivity due to the potassium-40, thorium and uranium contained.

Usually ranging from 1 millisievert (mSv) to 13 mSv annually depending on location, average radiation exposure from natural radioisotopes is 2.0 mSv per person a year worldwide. This makes up the majority of typical total dosage (with mean annual exposure from other sources amounting to 0.6 mSv from medical tests averaged over the whole populace, 0.4 mSv from cosmic rays, 0.005 mSv from the legacy of past atmospheric nuclear testing, 0.005 mSv occupational exposure, 0.002 mSv from the Chernobyl disaster, and 0.0002 mSv from the nuclear fuel cycle).

TENORM is not regulated as restrictively as nuclear reactor waste, though there are no significant differences in the radiological risks of these materials.

Coal

Coal contains a small amount of radioactive uranium, barium, thorium, and potassium, but, in the case of pure coal, this is significantly less than the average concentration of those elements in the Earth's crust. The surrounding strata, if shale or mudstone, often contain slightly more than average and this may also be reflected in the ash content of 'dirty' coals. The more active ash minerals become concentrated in the fly ash precisely because they do not burn well. The radioactivity of fly ash is about the same as black shale and is less than phosphate rocks, but is more of a concern because a small amount of the fly ash ends up in the atmosphere where it can be inhaled. According to U.S. National Council on Radiation Protection and Measurements (NCRP) reports, population exposure from 1000-MWe power plants amounts to 490 person-rem/year for coal power plants, 100 times as great as nuclear power plants (4.8 person-rem/year). The exposure from the complete nuclear fuel cycle from mining to waste disposal is 136 person-rem/year; the corresponding value for coal use from mining to waste disposal is "probably unknown".

Oil and gas

Residues from the oil and gas industry often contain radium and its decay products. The sulfate scale from an oil well can be very radium rich, while the water, oil, and gas from a well often contain radon. The radon decays to form solid radioisotopes which form coatings on the inside of pipework. In an oil processing plant, the area of the plant where propane is processed is often one of the more contaminated areas of the plant as radon has a similar boiling point to propane.

Radioactive elements are an industrial problem in some oil wells where workers operating in direct contact with the crude oil and brine can be actually exposed to doses having negative health effects. Due to the relatively high concentration of these elements in the brine, its disposal is also a technological challenge. In the United States, the brine is however exempt from the dangerous waste regulations and can be disposed of regardless of radioactive or toxic substances content since the 1980s.

Rare-earth mining

Due to natural occurrence of radioactive elements such as thorium and radium in rare-earth ore, mining operations also result in production of waste and mineral deposits that are slightly radioactive.

Classification

Classification of radioactive waste varies by country. The IAEA, which publishes the Radioactive Waste Safety Standards (RADWASS), also plays a significant role. The proportion of various types of waste generated in the UK:

94% – low-level waste (LLW)
~6% – intermediate-level waste (ILW)
<1% – high-level waste (HLW)

Mill tailings

Uranium tailings are waste by-product materials left over from the rough processing of uranium-bearing ore. They are not significantly radioactive. Mill tailings are sometimes referred to as 11(e)2 wastes, from the section of the Atomic Energy Act of 1946 that defines them. Uranium mill tailings typically also contain chemically hazardous heavy metal such as lead and arsenic. Vast mounds of uranium mill tailings are left at many old mining sites, especially in Colorado, New Mexico, and Utah.

Although mill tailings are not very radioactive, they have long half-lives. Mill tailings often contain radium, thorium and trace amounts of uranium.

Low-level waste

Low-level waste (LLW) is generated from hospitals and industry, as well as the nuclear fuel cycle. Low-level wastes include paper, rags, tools, clothing, filters, and other materials which contain small amounts of mostly short-lived radioactivity. Materials that originate from any region of an Active Area are commonly designated as LLW as a precautionary measure even if there is only a remote possibility of being contaminated with radioactive materials. Such LLW typically exhibits no higher radioactivity than one would expect from the same material disposed of in a non-active area, such as a normal office block. Example LLW includes wiping rags, mops, medical tubes, laboratory animal carcasses, and more. LLW waste makes 94% of all radioactive waste volume in the UK most of it disposed of in Cumbria first in landfill style trenches, and now using grouted metal containers that are stacked in concrete vaults. A new site in the north of Scotland is the Dounreay site which is prepared to withstand a 4m tsunami.

Some high-activity LLW requires shielding during handling and transport but most LLW is suitable for shallow land burial. To reduce its volume, it is often compacted or incinerated before disposal. Low-level waste is divided into four classes: class A, class B, class C, and Greater Than Class C (GTCC).

Intermediate-level waste

Intermediate-level waste (ILW) contains higher amounts of radioactivity compared to low-level waste. It generally requires shielding, but not cooling. Intermediate-level wastes includes resins, chemical sludge and metal nuclear fuel cladding, as well as contaminated materials from reactor decommissioning. It may be solidified in concrete or bitumen or mixed with silica sand and vitrified for disposal. As a general rule, short-lived waste (mainly non-fuel materials from reactors) is buried in shallow repositories, while long-lived waste (from fuel and fuel reprocessing) is deposited in geological repository. Regulations in the United States do not define this category of waste; the term is used in Europe and elsewhere. ILW makes 6% of all radioactive waste volume in the UK.

High-level waste

High-level waste (HLW) is produced by nuclear reactors and the reprocessing of nuclear fuel. The exact definition of HLW differs internationally. After a nuclear fuel rod serves one fuel cycle and is removed from the core, it is considered HLW. Spent fuel rods contain mostly uranium with fission products and transuranic elements generated in the reactor core. Spent fuel is highly radioactive and often hot. HLW accounts for over 95% of the total radioactivity produced in the process of nuclear electricity generation but it contributes to less than 1% of volume of all radioactive waste produced in the UK. Overall, the 60-year-long nuclear program in the UK up until 2019 produced 2150 m³ of HLW.

The radioactive waste from spent fuel rods consists primarily of cesium-137 and strontium-90, but it may also include plutonium, which can be considered transuranic waste. The half-lives of these radioactive elements can differ quite extremely. Some elements, such as cesium-137 and strontium-90 have half-lives of approximately 30 years. Meanwhile, plutonium has a half-life that can stretch to as long as 24,000 years.

The amount of HLW worldwide is currently increasing by about 12,000 tonnes every year. A 1000-megawatt nuclear power plant produces about 27 t of spent nuclear fuel (unreprocessed) every year. For comparison, the amount of ash produced by coal power plants in the United States alone is estimated at 130,000,000 t per year and fly ash is estimated to release 100 times more radiation than an equivalent nuclear power plant.

In 2010, it was estimated that about 250,000 t of nuclear HLW were stored globally. This does not include amounts that have escaped into the environment from accidents or tests. Japan is estimated to hold 17,000 t of HLW in storage in 2015. As of 2019, the United States has over 90,000 t of HLW. HLW have been shipped to other countries to be stored or reprocessed and, in some cases, shipped back as active fuel.

The ongoing controversy over high-level radioactive waste disposal is a major constraint on the nuclear power's global expansion. Most scientists agree that the main proposed long-term solution is deep geological burial, either in a mine or a deep borehole. As of 2019 no dedicated civilian high-level nuclear waste site is operational as small amounts of HLW did not justify the investment before. Finland is in the advanced stage of the construction of the Onkalo spent nuclear fuel repository, which is planned to open in 2025 at 400–450 m depth. France is in the planning phase for a 500 m deep Cigeo facility in Bure. Sweden is planning a site in Forsmark. Canada plans a 680 m deep facility near Lake Huron in Ontario. The Republic of Korea plans to open a site around 2028. The site in Sweden enjoys 80% support from local residents as of 2020.

The Morris Operation in Grundy County, Illinois, is currently the only de facto high-level radioactive waste storage site in the United States.

Transuranic waste

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Transuranic waste (TRUW) as defined by U.S. regulations is, without regard to form or origin, waste that is contaminated with alpha-emitting transuranic radionuclides with half-lives greater than 20 years and concentrations greater than 100 nCi/g (3.7 MBq/kg), excluding high-level waste. Elements that have an atomic number greater than uranium are called transuranic ("beyond uranium"). Because of their long half-lives, TRUW is disposed of more cautiously than either low- or intermediate-level waste. In the United States, it arises mainly from nuclear weapons production, and consists of clothing, tools, rags, residues, debris, and other items contaminated with small amounts of radioactive elements (mainly plutonium).

Under U.S. law, transuranic waste is further categorized into "contact-handled" (CH) and "remote-handled" (RH) on the basis of the radiation dose rate measured at the surface of the waste container. CH TRUW has a surface dose rate not greater than 200 mrem per hour (2 mSv/h), whereas RH TRUW has a surface dose rate of 200 mrem/h (2 mSv/h) or greater. CH TRUW does not have the very high radioactivity of high-level waste, nor its high heat generation, but RH TRUW can be highly radioactive, with surface dose rates up to 1,000,000 mrem/h (10,000 mSv/h). The United States currently disposes of TRUW generated from military facilities at the Waste Isolation Pilot Plant (WIPP) in a deep salt formation in New Mexico.

Prevention

A future way to reduce waste accumulation is to phase out current reactors in favor of Generation IV reactors, which output less waste per power generated. Fast reactors such as BN-800 in Russia are also able to consume MOX fuel that is manufactured from recycled spent fuel from traditional reactors.

The UK's Nuclear Decommissioning Authority published a position paper in 2014 on the progress on approaches to the management of separated plutonium, which summarises the conclusions of the work that NDA shared with UK government.

Management

Of particular concern in nuclear waste management are two long-lived fission products, Tc-99 (half-life 220,000 years) and I-129 (half-life 15.7 million years), which dominate spent fuel radioactivity after a few thousand years. The most troublesome transuranic elements in spent fuel are Np-237 (half-life two million years) and Pu-239 (half-life 24,000 years). Nuclear waste requires sophisticated treatment and management to successfully isolate it from interacting with the biosphere. This usually necessitates treatment, followed by a long-term management strategy involving storage, disposal or transformation of the waste into a non-toxic form. Governments around the world are considering a range of waste management and disposal options, though there has been limited progress toward long-term waste management solutions.

In the second half of the 20th century, several methods of disposal of radioactive waste were investigated by nuclear nations, which are :

"Long-term above-ground storage", not implemented.
"Disposal in outer space" (for instance, inside the Sun), not implemented—as it would be currently too expensive.
"Deep borehole disposal", not implemented.
"Rock melting", not implemented.
"Disposal at subduction zones", not implemented.
Ocean disposal, by the USSR, the United Kingdom, Switzerland, the United States, Belgium, France, the Netherlands, Japan, Sweden, Russia, Germany, Italy and South Korea (1954–93). This is no longer permitted by international agreements.
"Sub-seabed disposal", not implemented, not permitted by international agreements.
"Disposal in ice sheets", rejected in Antarctic Treaty
"Deep well injection", by USSR and USA.
Nuclear transmutation, using lasers to cause beta decay to convert the unstable atoms to those with shorter half-lives.

In the United States, waste management policy completely broke down with the ending of work on the incomplete Yucca Mountain Repository. At present there are 70 nuclear power plant sites where spent fuel is stored. A Blue Ribbon Commission was appointed by President Obama to look into future options for this and future waste. A deep geological repository seems to be favored. 2018 Nobel Prize for Physics-winner Gérard Mourou has proposed using Chirped pulse amplification to generate high-energy and low-duration laser pulses to transmute highly radioactive material (contained in a target) to significantly reduce its half-life, from thousands of years to only a few minutes.

Initial treatment

Vitrification

Long-term storage of radioactive waste requires the stabilization of the waste into a form that will neither react nor degrade for extended periods. It is theorized that one way to do this might be through vitrification. Currently at Sellafield the high-level waste (PUREX first cycle raffinate) is mixed with sugar and then calcined. Calcination involves passing the waste through a heated, rotating tube. The purposes of calcination are to evaporate the water from the waste and de-nitrate the fission products to assist the stability of the glass produced.

The 'calcine' generated is fed continuously into an induction heated furnace with fragmented glass. The resulting glass is a new substance in which the waste products are bonded into the glass matrix when it solidifies. As a melt, this product is poured into stainless steel cylindrical containers ("cylinders") in a batch process. When cooled, the fluid solidifies ("vitrifies") into the glass. After being formed, the glass is highly resistant to water.

After filling a cylinder, a seal is welded onto the cylinder head. The cylinder is then washed. After being inspected for external contamination, the steel cylinder is stored, usually in an underground repository. In this form, the waste products are expected to be immobilized for thousands of years.

The glass inside a cylinder is usually a black glossy substance. All this work (in the United Kingdom) is done using hot cell systems. Sugar is added to control the ruthenium chemistry and to stop the formation of the volatile RuO₄ containing radioactive ruthenium isotopes. In the West, the glass is normally a borosilicate glass (similar to Pyrex), while in the former Soviet Union it is normal to use a phosphate glass. The amount of fission products in the glass must be limited because some (palladium, the other Pt group metals, and tellurium) tend to form metallic phases which separate from the glass. Bulk vitrification uses electrodes to melt soil and wastes, which are then buried underground. In Germany a vitrification plant is in use; this is treating the waste from a small demonstration reprocessing plant which has since been closed down.

Phosphate ceramics

Vitrification is not the only way to stabilize the waste into a form that will not react or degrade for extended periods. Immobilization via direct incorporation into a phosphate-based crystalline ceramic host is also used. The diverse chemistry of phosphate ceramics under various conditions demonstrates a versatile material that can withstand chemical, thermal, and radioactive degradation over time. The properties of phosphates, particularly ceramic phosphates, of stability over a wide pH range, low porosity, and minimization of secondary waste introduces possibilities for new waste immobilization techniques.

Ion exchange

It is common for medium active wastes in the nuclear industry to be treated with ion exchange or other means to concentrate the radioactivity into a small volume. The much less radioactive bulk (after treatment) is often then discharged. For instance, it is possible to use a ferric hydroxide floc to remove radioactive metals from aqueous mixtures. After the radioisotopes are absorbed onto the ferric hydroxide, the resulting sludge can be placed in a metal drum before being mixed with cement to form a solid waste form. In order to get better long-term performance (mechanical stability) from such forms, they may be made from a mixture of fly ash, or blast furnace slag, and portland cement, instead of normal concrete (made with portland cement, gravel and sand).

Synroc

The Australian Synroc (synthetic rock) is a more sophisticated way to immobilize such waste, and this process may eventually come into commercial use for civil wastes (it is currently being developed for U.S. military wastes). Synroc was invented by Ted Ringwood, a geochemist at the Australian National University. The Synroc contains pyrochlore and cryptomelane type minerals. The original form of Synroc (Synroc C) was designed for the liquid high-level waste (PUREX raffinate) from a light-water reactor. The main minerals in this Synroc are hollandite (BaAl₂Ti₆O₁₆), zirconolite (CaZrTi₂O₇) and perovskite (CaTiO₃). The zirconolite and perovskite are hosts for the actinides. The strontium and barium will be fixed in the perovskite. The caesium will be fixed in the hollandite. A Synroc waste treatment facility began construction in 2018 at ANSTO.

Long-term management

The time frame in question when dealing with radioactive waste ranges from 10,000 to 1,000,000 years, according to studies based on the effect of estimated radiation doses. Researchers suggest that forecasts of health detriment for such periods should be examined critically.Practical studies only consider up to 100 years as far as effective planning and cost evaluations are concerned. Long term behavior of radioactive wastes remains a subject for ongoing research projects in geoforecasting.

Remediation

Algae has shown selectivity for strontium in studies, where most plants used in bioremediation have not shown selectivity between calcium and strontium, often becoming saturated with calcium, which is present in greater quantities in nuclear waste. Strontium-90 with a half life around 30 years, is classified as high-level waste.

Researchers have looked at the bioaccumulation of strontium by Scenedesmus spinosus (algae) in simulated wastewater. The study claims a highly selective biosorption capacity for strontium of S. spinosus, suggesting that it may be appropriate for use of nuclear wastewater. A study of the pond alga Closterium moniliferum using non-radioactive strontium found that varying the ratio of barium to strontium in water improved strontium selectivity.

Above-ground disposal

Dry cask storage typically involves taking waste from a spent fuel pool and sealing it (along with an inert gas) in a steel cylinder, which is placed in a concrete cylinder which acts as a radiation shield. It is a relatively inexpensive method which can be done at a central facility or adjacent to the source reactor. The waste can be easily retrieved for reprocessing.

Geologic disposal

The process of selecting appropriate deep final repositories for high-level waste and spent fuel is now underway in several countries with the first expected to be commissioned sometime after 2010. The basic concept is to locate a large, stable geologic formation and use mining technology to excavate a tunnel, or large-bore tunnel boring machines (similar to those used to drill the Channel Tunnel from England to France) to drill a shaft 500 metres (1,600 ft) to 1,000 metres (3,300 ft) below the surface where rooms or vaults can be excavated for disposal of high-level radioactive waste. The goal is to permanently isolate nuclear waste from the human environment. Many people remain uncomfortable with the immediate stewardship cessation of this disposal system, suggesting perpetual management and monitoring would be more prudent.

Because some radioactive species have half-lives longer than one million years, even very low container leakage and radionuclide migration rates must be taken into account. Moreover, it may require more than one half-life until some nuclear materials lose enough radioactivity to cease being lethal to living things. A 1983 review of the Swedish radioactive waste disposal program by the National Academy of Sciences found that country's estimate of several hundred thousand years—perhaps up to one million years—being necessary for waste isolation "fully justified."

The proposed land-based subductive waste disposal method disposes of nuclear waste in a subduction zone accessed from land and therefore is not prohibited by international agreement. This method has been described as the most viable means of disposing of radioactive waste, and as the state-of-the-art as of 2001 in nuclear waste disposal technology.

Another approach termed Remix & Return would blend high-level waste with uranium mine and mill tailings down to the level of the original radioactivity of the uranium ore, then replace it in inactive uranium mines. This approach has the merits of providing jobs for miners who would double as disposal staff, and of facilitating a cradle-to-grave cycle for radioactive materials, but would be inappropriate for spent reactor fuel in the absence of reprocessing, due to the presence of highly toxic radioactive elements such as plutonium within it.

Deep borehole disposal is the concept of disposing of high-level radioactive waste from nuclear reactors in extremely deep boreholes. Deep borehole disposal seeks to place the waste as much as 5 kilometres (3.1 mi) beneath the surface of the Earth and relies primarily on the immense natural geological barrier to confine the waste safely and permanently so that it should never pose a threat to the environment. The Earth's crust contains 120 trillion tons of thorium and 40 trillion tons of uranium (primarily at relatively trace concentrations of parts per million each adding up over the crust's 3 × 10¹⁹ ton mass), among other natural radioisotopes. Since the fraction of nuclides decaying per unit of time is inversely proportional to an isotope's half-life, the relative radioactivity of the lesser amount of human-produced radioisotopes (thousands of tons instead of trillions of tons) would diminish once the isotopes with far shorter half-lives than the bulk of natural radioisotopes decayed.

In January 2013, Cumbria county council rejected UK central government proposals to start work on an underground storage dump for nuclear waste near to the Lake District National Park. "For any host community, there will be a substantial community benefits package and worth hundreds of millions of pounds" said Ed Davey, Energy Secretary, but nonetheless, the local elected body voted 7–3 against research continuing, after hearing evidence from independent geologists that "the fractured strata of the county was impossible to entrust with such dangerous material and a hazard lasting millennia."

Horizontal drillhole disposal describes proposals to drill over one km vertically, and two km horizontally in the earth's crust, for the purpose of disposing of high-level waste forms such as spent nuclear fuel, Caesium-137, or Strontium-90. After the emplacement and the retrievability period, drillholes would be backfilled and sealed. A series of tests of the technology were carried out in November 2018 and then again publicly in January 2019 by a U.S. based private company. The test demonstrated the emplacement of a test-canister in a horizontal drillhole and retrieval of the same canister. There was no actual high-level waste used in this test.

European Commission Joint Research Centre report of 2021 (see above) concluded:

Management of radioactive waste and its safe and secure disposal is a necessary step in the lifecycle of all applications of nuclear science and technology (nuclear energy, research, industry, education, medical, and others). Radioactive waste is therefore generated in practically every country, the largest contribution coming from the nuclear energy lifecycle in countries operating nuclear power plants. Presently, there is broad scientific and technical consensus that disposal of high-level, long-lived radioactive waste in deep geologic formations is, at the state of today’s knowledge, considered as an appropriate and safe means of isolating it from the biosphere for very long time scales.

Ocean floor disposal

From 1946 through 1993, thirteen countries used ocean disposal or ocean dumping as a method to dispose of nuclear/radioactive waste with an approximation of 200,000 tons sourcing mainly from the medical, research and nuclear industry.

Ocean floor disposal of radioactive waste has been suggested by the finding that deep waters in the North Atlantic Ocean do not present an exchange with shallow waters for about 140 years based on oxygen content data recorded over a period of 25 years. They include burial beneath a stable abyssal plain, burial in a subduction zone that would slowly carry the waste downward into the Earth's mantle, and burial beneath a remote natural or human-made island. While these approaches all have merit and would facilitate an international solution to the problem of disposal of radioactive waste, they would require an amendment of the Law of the Sea.

Nuclear submarines have been lost and these vessels reactors must also be counted in the amount of radioactive waste deposited at sea.

Article 1 (Definitions), 7., of the 1996 Protocol to the Convention on the Prevention of Marine Pollution by Dumping of Wastes and Other Matter, (the London Dumping Convention) states:

""Sea" means all marine waters other than the internal waters of States, as well as the seabed and the subsoil thereof; it does not include sub-seabed repositories accessed only from land."

Transmutation

There have been proposals for reactors that consume nuclear waste and transmute it to other, less-harmful or shorter-lived, nuclear waste. In particular, the integral fast reactor was a proposed nuclear reactor with a nuclear fuel cycle that produced no transuranic waste and, in fact, could consume transuranic waste. It proceeded as far as large-scale tests but was eventually canceled by the U.S. Government. Another approach, considered safer but requiring more development, is to dedicate subcritical reactors to the transmutation of the left-over transuranic elements.

An isotope that is found in nuclear waste and that represents a concern in terms of proliferation is Pu-239. The large stock of plutonium is a result of its production inside uranium-fueled reactors and of the reprocessing of weapons-grade plutonium during the weapons program. An option for getting rid of this plutonium is to use it as a fuel in a traditional light-water reactor (LWR). Several fuel types with differing plutonium destruction efficiencies are under study.

Transmutation was banned in the United States in April 1977 by President Carter due to the danger of plutonium proliferation, but President Reagan rescinded the ban in 1981. Due to economic losses and risks, the construction of reprocessing plants during this time did not resume. Due to high energy demand, work on the method has continued in the EU. This has resulted in a practical nuclear research reactor called Myrrha in which transmutation is possible. Additionally, a new research program called ACTINET has been started in the EU to make transmutation possible on a large, industrial scale. According to President Bush's Global Nuclear Energy Partnership (GNEP) of 2007, the United States is actively promoting research on transmutation technologies needed to markedly reduce the problem of nuclear waste treatment.

There have also been theoretical studies involving the use of fusion reactors as so-called "actinide burners" where a fusion reactor plasma such as in a tokamak, could be "doped" with a small amount of the "minor" transuranic atoms which would be transmuted (meaning fissioned in the actinide case) to lighter elements upon their successive bombardment by the very high energy neutrons produced by the fusion of deuterium and tritium in the reactor. A study at MIT found that only 2 or 3 fusion reactors with parameters similar to that of the International Thermonuclear Experimental Reactor (ITER) could transmute the entire annual minor actinide production from all of the light-water reactors presently operating in the United States fleet while simultaneously generating approximately 1 gigawatt of power from each reactor.

Re-use

Spent nuclear fuel contains abundant fertile uranium and traces of fissile materials. Methods such as the PUREX process can be used to remove useful actinides for the production of active nuclear fuel.

Another option is to find applications for the isotopes in nuclear waste so as to re-use them. Already, caesium-137, strontium-90 and a few other isotopes are extracted for certain industrial applications such as food irradiation and radioisotope thermoelectric generators. While re-use does not eliminate the need to manage radioisotopes, it can reduce the quantity of waste produced.

The Nuclear Assisted Hydrocarbon Production Method, Canadian patent application 2,659,302, is a method for the temporary or permanent storage of nuclear waste materials comprising the placing of waste materials into one or more repositories or boreholes constructed into an unconventional oil formation. The thermal flux of the waste materials fractures the formation and alters the chemical and/or physical properties of hydrocarbon material within the subterranean formation to allow removal of the altered material. A mixture of hydrocarbons, hydrogen, and/or other formation fluids is produced from the formation. The radioactivity of high-level radioactive waste affords proliferation resistance to plutonium placed in the periphery of the repository or the deepest portion of a borehole.

Breeder reactors can run on U-238 and transuranic elements, which comprise the majority of spent fuel radioactivity in the 1,000–100,000-year time span.

Space disposal

Space disposal is attractive because it removes nuclear waste from the planet. It has significant disadvantages, such as the potential for catastrophic failure of a launch vehicle, which could spread radioactive material into the atmosphere and around the world. A high number of launches would be required because no individual rocket would be able to carry very much of the material relative to the total amount that needs to be disposed of. This makes the proposal impractical economically and increases the risk of one or more launch failures. To further complicate matters, international agreements on the regulation of such a program would need to be established. Costs and inadequate reliability of modern rocket launch systems for space disposal has been one of the motives for interest in non-rocket spacelaunch systems such as mass drivers, space elevators, and other proposals.

National management plans

Sweden and Finland are furthest along in committing to a particular disposal technology, while many others reprocess spent fuel or contract with France or Great Britain to do it, taking back the resulting plutonium and high-level waste. "An increasing backlog of plutonium from reprocessing is developing in many countries... It is doubtful that reprocessing makes economic sense in the present environment of cheap uranium."

In many European countries (e.g., Britain, Finland, the Netherlands, Sweden, and Switzerland) the risk or dose limit for a member of the public exposed to radiation from a future high-level nuclear waste facility is considerably more stringent than that suggested by the International Commission on Radiation Protection or proposed in the United States. European limits are often more stringent than the standard suggested in 1990 by the International Commission on Radiation Protection by a factor of 20, and more stringent by a factor of ten than the standard proposed by the U.S. Environmental Protection Agency (EPA) for Yucca Mountain nuclear waste repository for the first 10,000 years after closure.

The U.S. EPA's proposed standard for greater than 10,000 years is 250 times more permissive than the European limit. The U.S. EPA proposed a legal limit of a maximum of 3.5 millisieverts (350 millirem) each annually to local individuals after 10,000 years, which would be up to several percent of the exposure currently received by some populations in the highest natural background regions on Earth, though the U.S. United States Department of Energy (DOE) predicted that received dose would be much below that limit. Over a timeframe of thousands of years, after the most active short half-life radioisotopes decayed, burying U.S. nuclear waste would increase the radioactivity in the top 2000 feet of rock and soil in the United States (10 million km²) by approximately 1 part in 10 million over the cumulative amount of natural radioisotopes in such a volume, but the vicinity of the site would have a far higher concentration of artificial radioisotopes underground than such an average.

Mongolia

After serious opposition about plans and negotiations between Mongolia with Japan and the United States of America to build nuclear-waste facilities in Mongolia, Mongolia stopped all negotiations in September 2011. These negotiations had started after U.S. Deputy Secretary of Energy Daniel Poneman visited Mongolia in September 2010. Talks took place in Washington, D.C. between officials of Japan, the United States, and Mongolia in February 2011. After this the United Arab Emirates (UAE), which wanted to buy nuclear fuel from Mongolia, joined in the negotiations. The talks were kept secret and, although the Mainichi Daily News reported on them in May, Mongolia officially denied the existence of these negotiations. However, alarmed by this news, Mongolian citizens protested against the plans and demanded the government withdraw the plans and disclose information. The Mongolian President Tsakhiagiin Elbegdorj issued a presidential order on September 13 banning all negotiations with foreign governments or international organizations on nuclear-waste storage plans in Mongolia. The Mongolian government has accused the newspaper of distributing false claims around the world. After the presidential order, the Mongolian president fired the individual who was supposedly involved in these conversations.

Illegal dumping

Authorities in Italy are investigating a 'Ndrangheta mafia clan accused of trafficking and illegally dumping nuclear waste. According to a whistleblower, a manager of the Italy's state energy research agency Enea paid the clan to get rid of 600 drums of toxic and radioactive waste from Italy, Switzerland, France, Germany, and the United States, with Somalia as the destination, where the waste was buried after buying off local politicians. Former employees of Enea are suspected of paying the criminals to take waste off their hands in the 1980s and 1990s. Shipments to Somalia continued into the 1990s, while the 'Ndrangheta clan also blew up shiploads of waste, including radioactive hospital waste, sending them to the sea bed off the Calabrian coast. According to the environmental group Legambiente, former members of the 'Ndrangheta have said that they were paid to sink ships with radioactive material for the last 20 years.

In 2008, Afghan authorities accused Pakistan of illegally dumping nuclear waste in the southern parts of Afghanistan when the Taliban were in power between 1996 and 2001. The Pakistani government denied the allegation.

Accidents

A few incidents have occurred when radioactive material was disposed of improperly, shielding during transport was defective, or when it was simply abandoned or even stolen from a waste store. In the Soviet Union, waste stored in Lake Karachay was blown over the area during a dust storm after the lake had partly dried out. At Maxey Flat, a low-level radioactive waste facility located in Kentucky, containment trenches covered with dirt, instead of steel or cement, collapsed under heavy rainfall into the trenches and filled with water. The water that invaded the trenches became radioactive and had to be disposed of at the Maxey Flat facility itself. In other cases of radioactive waste accidents, lakes or ponds with radioactive waste accidentally overflowed into the rivers during exceptional storms. In Italy, several radioactive waste deposits let material flow into river water, thus contaminating water for domestic use. In France in the summer of 2008, numerous incidents happened: in one, at the Areva plant in Tricastin, it was reported that, during a draining operation, liquid containing untreated uranium overflowed out of a faulty tank and about 75 kg of the radioactive material seeped into the ground and, from there, into two rivers nearby; in another case, over 100 staff were contaminated with low doses of radiation. There are ongoing concerns around the deterioration of the nuclear waste site on the Enewetak Atoll of the Marshall Islands and a potential radioactive spill.

Scavenging of abandoned radioactive material has been the cause of several other cases of radiation exposure, mostly in developing nations, which may have less regulation of dangerous substances (and sometimes less general education about radioactivity and its hazards) and a market for scavenged goods and scrap metal. The scavengers and those who buy the material are almost always unaware that the material is radioactive and it is selected for its aesthetics or scrap value. Irresponsibility on the part of the radioactive material's owners, usually a hospital, university, or military, and the absence of regulation concerning radioactive waste, or a lack of enforcement of such regulations, have been significant factors in radiation exposures. For an example of an accident involving radioactive scrap originating from a hospital see the Goiânia accident.

Transportation accidents involving spent nuclear fuel from power plants are unlikely to have serious consequences due to the strength of the spent nuclear fuel shipping casks.

On 15 December 2011, top government spokesman Osamu Fujimura of the Japanese government admitted that nuclear substances were found in the waste of Japanese nuclear facilities. Although Japan did commit itself in 1977 to these inspections in the safeguard agreement with the IAEA, the reports were kept secret for the inspectors of the International Atomic Energy Agency. Japan did start discussions with the IAEA about the large quantities of enriched uranium and plutonium that were discovered in nuclear waste cleared away by Japanese nuclear operators. At the press conference Fujimura said: "Based on investigations so far, most nuclear substances have been properly managed as waste, and from that perspective, there is no problem in safety management," but according to him, the matter was at that moment still being investigated.

Associated hazard warning signs

The trefoil symbol used to indicate ionizing radiation.
2007 ISO radioactivity danger symbol intended for IAEA Category 1, 2 and 3 sources defined as dangerous sources capable of death or serious injury.
The dangerous goods transport classification sign for radioactive materials

Boiling water reactor safety systems

From Wikipedia, the free encyclopedia

Boiling water reactor safety systems are nuclear safety systems constructed within boiling water reactors in order to prevent or mitigate environmental and health hazards in the event of accident or natural disaster.

Like the pressurized water reactor, the BWR reactor core continues to produce heat from radioactive decay after the fission reactions have stopped, making a core damage incident possible in the event that all safety systems have failed and the core does not receive coolant. Also like the pressurized water reactor, a boiling water reactor has a negative void coefficient, that is, the neutron (and the thermal) output of the reactor decreases as the proportion of steam to liquid water increases inside the reactor.

However, unlike a pressurized water reactor which contains no steam in the reactor core, a sudden increase in BWR steam pressure (caused, for example, by the actuation of the main steam isolation valve (MSIV) from the reactor) will result in a sudden decrease in the proportion of steam to liquid water inside the reactor. The increased ratio of water to steam will lead to increased neutron moderation, which in turn will cause an increase in the power output of the reactor. This type of event is referred to as a "pressure transient".

Safety systems

The BWR is specifically designed to respond to pressure transients, having a "pressure suppression" type of design which vents overpressure using safety-relief valves to below the surface of a pool of liquid water within the containment, known as the "wetwell", "torus" or "suppression pool". All BWRs utilize a number of safety/relief valves for overpressure, up to 7 of these are a part of the Automatic Depressurization System (ADS) and 18 safety overpressure relief valves on ABWR models, only a few of which have to function to stop the pressure rise of a transient. In addition, the reactor will already have rapidly shut down before the transient affects the RPV (as described in the Reactor Protection System section below.)

Because of this effect in BWRs, operating components and safety systems are designed with the intention that no credible scenario can cause a pressure and power increase that exceeds the systems' capability to quickly shut down the reactor before damage to the fuel or to components containing the reactor coolant can occur. 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. Neutronic power will fall to far below nominal power (the range of 30% with the cessation of circulation, and thus, void clearance) even before ARI or SLCS actuation occurs. Thermal power will be barely affected.

In the event of a contingency that disables all of the safety systems, each reactor is surrounded by a containment building consisting of 1.2–2.4 m (3.9–7.9 ft) of steel-reinforced, pre-stressed concrete designed to seal off the reactor from the environment.

However, the containment building does not protect the fuel during the whole fuel cycle. Most importantly, the spent fuel resides long periods of time outside the primary containment. A typical spent fuel storage pool can hold roughly five times the fuel in the core. Since reloads typically discharge one third of a core, much of the spent fuel stored in the pool will have had considerable decay time. But if the pool were to be drained of water, the discharged fuel from the previous two refuelings would still be "fresh" enough to melt under decay heat. However, the zircaloy cladding of this fuel could be ignited during the heatup. The resulting fire would probably spread to most or all of the fuel in the pool. The heat of combustion, in combination with decay heat, would probably drive "borderline aged" fuel into a molten condition. Moreover, if the fire becomes oxygen-starved (quite probable for a fire located in the bottom of a pit such as this), the hot zirconium would rob oxygen from the uranium dioxide fuel, forming a liquid mixture of metallic uranium, zirconium, oxidized zirconium, and dissolved uranium dioxide. This would cause a release of fission products from the fuel matrix quite comparable to that of molten fuel. In addition, although confined, BWR spent fuel pools are almost always located outside of the primary containment. Generation of hydrogen during the process would probably result in an explosion, damaging the secondary containment building. Thus, release to the atmosphere is more likely than for comparable accidents involving the reactor core.

Reactor Protection System (RPS)

The Reactor Protection System (RPS) is a system, computerized in later BWR models, that is designed to automatically, rapidly, and completely shut down and make safe the Nuclear Steam Supply System (NSSS – the reactor pressure vessel, pumps, and water/steam piping within the containment) if some event occurs that could result in the reactor entering an unsafe operating condition. In addition, the RPS can automatically spin up the Emergency Core Cooling System (ECCS) upon detection of several signals. It does not require human intervention to operate. However, the reactor operators can override parts of the RPS if necessary. If an operator recognizes a deteriorating condition, and knows an automatic safety system will activate, they are trained to pre-emptively activate the safety system.

If the reactor is at power or ascending to power (i.e. if the reactor is supercritical; the control rods are withdrawn to the point where the reactor generates more neutrons than it absorbs), there are safety-related contingencies that may arise that necessitate a rapid shutdown of the reactor, or, in Western nuclear parlance, a "SCRAM". The SCRAM is a manually triggered or automatically triggered rapid insertion of all control rods into the reactor, which will take the reactor to decay heat power levels within tens of seconds. Since ≈ 0.6% of neutrons are emitted from fission products ("delayed" neutrons), which are born seconds or minutes after fission, all fission can not be terminated instantaneously, but the fuel soon returns to decay heat power levels. Manual SCRAMs may be initiated by the reactor operators, while automatic SCRAMs are initiated upon:

Turbine stop-valve or turbine control-valve closure.
1. If turbine protection systems detect a significant anomaly, admission of steam is halted. Reactor rapid shutdown is in anticipation of a pressure transient that could increase reactivity.
2. Generator load rejection will also cause closure of turbine valves and trip RPS.
3. This trip is only active above approximately 1/3 reactor power. Below this amount, the bypass steam system is capable of controlling reactor pressure without causing a reactivity transient in the core.
Loss of off-site power (LOOP)
1. During normal operation, the reactor protection system (RPS) is powered by off-site power
  1. Loss of off-site power would open all relays in the RPS, causing all rapid shutdown signals to come in redundantly.
  2. would also cause MSIV to close since RPS is fail-safe; plant assumes a main steam break is coincident with loss of off-site power.
Neutron monitor trips – the purpose of these trips is to ensure an even increase in neutron and thermal power during startup.
1. Source-range monitor (SRM) or intermediate-range monitor (IRM) upscale:
  1. The SRM, used during instrument calibration, pre-critical, and early non-thermal criticality, and the IRM, used during ascension to power, middle/late non-thermal, and early or middle thermal stages, both have trips built in that prevent rapid decreases in reactor period when reactor is intensely reactive (e.g. when no voids exist, water is cold, and water is dense) without positive operator confirmation that such decreases in period are their intention. Prior to trips occurring, rod movement blocks will be activated to ensure operator vigilance if preset levels are marginally exceeded.
2. Average power range monitor (APRM) upscale:
  1. Prevents reactor from exceeding pre-set neutron power level maxima during operation or relative maxima prior to positive operator confirmation of end of startup by transition of reactor state into "Run".
3. Average power range monitor / coolant flow thermal trip:
  1. Prevents reactor from exceeding variable power levels without sufficient coolant flow for that level being present.
4. Oscillation Power Range Monitor
  1. Prevents reactor power from rapidly oscillating during low flow high power conditions.
Low reactor water level:
1. Loss of coolant contingency (LOCA)
2. Loss of proper feedwater (LOFW)
3. Protects the turbine from excessive moisture carryover if water level is below the steam separator and steam dryer stack.
High water level (in BWR6 plants)
1. Prevents flooding of the main steam lines and protects turbine equipment.
2. Limits the rate of cold water addition to the vessel, thus limiting reactor power increase during over-feed transients.
High drywell (primary containment) pressure
1. Indicative of potential loss of coolant contingency
2. Also initiates ECCS systems to prepare for core injection once the injection permissives are cleared.
Main steam isolation valve closure (MSIV)
1. Protects from pressure transient in the core causing a reactivity transient
2. Only triggers for each channel when the valve is greater than 8% closed
3. One valve may be closed without initiating a reactor trip.
High RPV pressure:
1. Indicative of MSIV closure.
2. Decreases reactivity to compensate for boiling void collapse due to high pressure.
3. Prevents pressure relief valves from opening.
4. Serves as a backup for several other trips, like turbine trip.
Low RPV pressure:
1. Indicative of a line break in the steam tunnel or other location which does not trigger high drywell pressure
2. Bypassed when the reactor is not in Run mode to allow for pressurization and cooldown without an automatic scram signal
Seismic event
1. Generally only plants in high seismic areas have this trip enabled.
Scram Discharge Volume High
1. In the event that the scram hydraulic discharge volume begins to fill up, this will scram the reactor prior to the volume filling. This prevents hydraulic lock, which could prevent the control rods from inserting. This is to prevent an ATWS (Anticipated Transient Without Scram).

Emergency core-cooling system (ECCS)

Diagram of a generic BWR reactor pressure vessel

While the reactor protection system is designed to shut down the reactor, ECCS is designed to maintain adequate core cooling. The ECCS is a set of interrelated safety systems that are designed to protect the fuel within the reactor pressure vessel, which is referred to as the "reactor core", from overheating. The five criteria for ECCS are to prevent peak fuel cladding temperature from exceeding 2200 °F (1204 °C), prevent more than 17% oxidation of the fuel cladding, prevent more than 1% of the maximum theoretical hydrogen generation due the zircalloy metal-water reaction, maintain a coolable geometry, and allow for long-term cooling. 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.

These systems are of three major types:

High-pressure systems: These are designed to protect the core by injecting large quantities of water into it to prevent the fuel from being uncovered by a decreasing water level. Generally used in cases with stuck-open safety valves, small breaks of auxiliary pipes, and particularly violent transients caused by turbine trip and main steam isolation valve closure. If the water level cannot be maintained with high-pressure systems alone (the water level still is falling below a preset point with the high-pressure systems working full-bore), the next set of systems responds.
Depressurization systems: These systems are designed to maintain reactor pressure within safety limits. Additionally, if reactor water level cannot be maintained with high-pressure coolant systems alone, the depressurization system can reduce reactor pressure to a level at which the low-pressure coolant systems can function.
Low-pressure systems: These systems are designed to function after the depressurization systems function. They have large capacities compared to the high-pressure systems and are supplied by multiple, redundant power sources. They will maintain any maintainable water level, and, in the event of a large pipe break of the worst type below the core that leads to temporary fuel rod "uncovery", to rapidly mitigate that state prior to the fuel heating to the point where core damage could occur.

High-pressure coolant injection system (HPCI)

The high-pressure coolant injection system is the first line of defense in the emergency core cooling system. HPCI is designed to inject substantial quantities of water into the reactor while it is at high pressure so as to prevent the activation of the automatic depressurization, core spray, and low-pressure coolant injection systems. HPCI is powered by steam from the reactor, and takes approximately 10 seconds to spin up from an initiating signal, and can deliver approximately 19,000 L/min (5,000 US gal/min) to the core at any core pressure above 6.8 atm (690 kPa, 100 psi). This is usually enough to keep water levels sufficient to avoid automatic depressurization except in a major contingency, such as a large break in the makeup water line. HPCI is also able to be run in "pressure control mode", where the HPCI turbine is run without pumping water to the reactor vessel. 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.

Versioning note: Some BWR/5s and the BWR/6 replace the steam-turbine driven HPCI pump with the AC-powered high-pressure core spray (HPCS); ABWR replaces HPCI with high-pressure core flooder (HPCF), a mode of the RCIC system, as described below. (E)SBWR does not have an equivalent system as it primarily uses passive safety cooling systems, though ESBWR does offer an alternative active high-pressure injection method using an operating mode of the Control Rod Drive System (CRDS) to supplement the passive system.

Isolation Condenser (IC)

Some reactors, including some BWR/2 and BWR/3 plants, and the (E)SBWR series of reactors, have a passive system called the Isolation Condenser. This is a heat exchanger located above containment in a pool of water open to atmosphere. When activated, decay heat boils steam, which is drawn into the heat exchanger and condensed; then it falls by weight of gravity back into the reactor. This process keeps the cooling water in the reactor, making it unnecessary to use powered feedwater pumps. The water in the open pool slowly boils off, venting clean steam to the atmosphere. This makes it unnecessary to run mechanical systems to remove heat. Periodically, the pool must be refilled, a simple task for a fire truck. The (E)SBWR reactors provide three days' supply of water in the pool. Some older reactors also have IC systems, including Fukushima Dai-ichi reactor 1, however their water pools may not be as large.

Under normal conditions, the IC system is not activated, but the top of the IC condenser is connected to the reactor's steam lines through an open valve. The IC automatically starts on low water level or high steam pressure indications. Once it starts, steam enters the IC condenser and condenses until it is filled with water. When the IC system is activated, a valve at the bottom of the IC condenser is opened which connects to a lower area of the reactor. The water falls to the reactor by gravity, allowing the condenser to fill with steam, which then condenses. This cycle runs continuously until the bottom valve is closed.

Reactor core isolation cooling system (RCIC)

The reactor core isolation cooling system is not an emergency core cooling system proper, but it is included because it fulfills an important-to-safety function which can help to cool the reactor in the event of a loss of normal heat sinking capability; or when all electrical power is lost. It has additional functionality in advanced versions of the BWR.

RCIC is an auxiliary feedwater pump meant for emergency use. It is able to inject cooling water into the reactor at high pressures. It injects approximately 2,000 L/min (600 gpm) into the reactor core. It takes less time to start than the HPCI system, approximately 30 seconds from an initiating signal. It has ample capacity to replace the cooling water boiled off by residual decay heat, and can even keep up with small leaks.

The RCIC system operates on high-pressure steam from the reactor itself, and thus is operable with no electric power other than battery power to operate the control valves. Those turn the RCIC on and off as necessary to maintain correct water levels in the reactor. (If run continuously, the RCIC would overfill the reactor and send water down its own steam supply line.) 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. The RCIC system condenses its steam into the reactor suppression pool. The RCIC can make up this water loss, from either of two sources: a makeup water tank located outside containment, or the wetwell itself. RCIC is not designed to maintain reactor water level during a LOCA or other leak. Similar to HPCI, the RCIC turbine can be run in recirculation mode to remove steam from the reactor and help depressurize the reactor.

Versioning note: RCIC and HPCF are integrated in the ABWRs, with HPCF representing the high-capacity mode of RCIC. Older BWRs such as Fukushima Unit 1 and Dresden as well as the new (E)SBWR do not have a RCIC system, and instead have an Isolation Condenser system.

Automatic depressurization system (ADS)

The Automatic depressurization system is not a part of the cooling system proper, but is an essential adjunct to the ECCS. It is designed to activate in the event that there is either a loss of high-pressure cooling to the vessel or if the high-pressure cooling systems cannot maintain the RPV water level. ADS can be manually or automatically initiated. When ADS receives an auto-start signal when water reaches the Low-Low-Low Water Level Alarm setpoint. 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. When the timer expires, or when the manual ADS initiate buttons are pressed, the system rapidly releases pressure from the RPV in the form of steam through pipes that are piped to below the water level in the suppression pool (the torus/wetwell), which is designed to condense the steam released by ADS or other safety valve activation into water), bringing the reactor vessel below 32 atm (3200 kPa, 465 psi), allowing the low-pressure cooling systems (LPCS/LPCI/LPCF/GDCS) to restore reactor water level. During an ADS blowdown, the steam being removed from the reactor is sufficient to ensure adequate core cooling even if the core is uncovered. The water in the reactor will rapidly flash to steam as reactor pressure drops, carrying away the latent heat of vaporization and providing cooling for the entire reactor. Low pressure ECCS systems will re-flood the core prior to the end of the emergency blowdown, ensuring that the core retains adequate cooling during the entire event.

Low-pressure core spray system (LPCS)

The Core Spray system, or Low-Pressure Core Spray system is designed to suppress steam generated by a major contingency and to ensure adequate core cooling for a partially or fully uncovered reactor core. LPCS can deliver up to 48,000 L/min (12,500 US gal/min) of water in a deluge from the top of the core. The core spray system collapses steam voids above the core, aids in reducing reactor pressure when the fuel is uncovered, and, in the event the reactor has a break so large that water level cannot be maintained, core spray is capable of preventing fuel damage by ensuring the fuel is adequately sprayed to remove decay heat. In earlier versions of the BWR (BWR 1 or 2 plants), the LPCS system was the only ECCS, and the core could be adequately cooled by core spray even if it was completely uncovered. Starting with Dresden units 2 and 3, the core spray system was augmented by the HPCI/LPCI systems to provide for both spray cooling and core flooding as methods for ensuring adequate core cooling. For most BWR models, core spray ensures the upper 1/3rd of the core does not exceed 17% cladding oxidation or 1% hydrogen production during a LOCA when used in combination with the LPCI system.

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

Low-pressure coolant injection (LPCI)

Low-pressure coolant injection is the emergency injection mode of the Residual Heat Removal (RHR) system. LPCI can be operated at reactor vessel pressures below 375 psi. LPCI consists of several pumps which are capable of injecting up to 150,000 L/min (40,000 US gal/min) of water into the reactor. Combined with the Core Spray system, the LPCI is designed to rapidly flood the reactor with coolant. The LPCI system was first introduced with Dresden units 2 and 3. The LPCI system can also use the RHR heat exchangers to remove decay heat from the reactor and cool the containment to cold conditions. Early versions of the LPCI system injected through the recirculation loops or into the down comer. Later versions of the BWR moved the injection point directly inside the core shroud to minimize time to reflood the core, substantially reducing the peak temperatures of the reactor during a LOCA.

Versioning note: ABWRs replace LPCI with low-pressure core flooder (LPCF), which operates using similar principles. (E)SBWRs replace LPCI with the DPVS/PCCS/GDCS, as described below.

Depressurization valve system (DPVS) / passive containment cooling system (PCCS) / gravity-driven cooling system (GDCS)

The (E)SBWR has an additional ECCS capacity that is completely passive, quite unique, and significantly improves defense in depth. This system is activated when the water level within the RPV reaches Level 1. At this point, a countdown timer is started.

There are several large depressurization valves located near the top of the reactor pressure vessel. These constitute the DPVS. This is a capability supplemental to the ADS, which is also included on the (E)SBWR. The DPVS consists of eight of these valves, four on main steamlines that vent to the drywell when actuated and four venting directly into the wetwell.

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. This will cause the water within the RPV to gain in volume (due to the drop in pressure) which will increase the water available to cool the core. In addition, depressurization reduces the saturation temperature enhancing the heat removal via phase transition. (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.)

If Level 1 is still not resubmerged within 100 seconds of DPVS actuation, then the GDCS valves fire. The GDCS is a series of very large water tanks located above and to the side of the Reactor Pressure Vessel within the drywell. When these valves fire, the GDCS is directly connected to the RPV. After ~50 more seconds of depressurization, the pressure within the GDCS will equalize with that of the RPV and drywell, and the water of the GDCS will begin flowing into the RPV.

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. The steam will be cooled, and will condense back into liquid water. The liquid water will drain from the heat exchanger back into the GDCS pool, where it can flow back into the RPV to make up for additional water boiled by decay heat. In addition, if the GDCS lines break, the shape of the RPV and the drywell will ensure that a "lake" of liquid water forms that submerges the bottom of the RPV (and the core within).

There is sufficient water to cool the heat exchangers of the PCCS for 72 hours. 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.

Standby liquid control system (SLCS)

The SLCS is a backup to the reactor protection system. In the event that RPS is unable to scram the reactor for any reason, the SLCS will inject a liquid boron solution into the reactor vessel to bring it to a guaranteed shutdown state prior to exceeding any containment or reactor vessel limits. 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. 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 consists of a tank containing borated water as a neutron absorber, protected by explosively-opened valves and redundant pumps, allowing the injection of the borated water into the reactor against any pressure within; the borated water will shut down a reactor and maintain it shut down. 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.

Versioning note: The SLCS is a system that is never meant to be activated unless all other measures have failed. In the BWR/1 – BWR/6, its activation could cause sufficient damage to the plant that it could make the older BWRs inoperable without a complete overhaul. With the arrival of the ABWR and (E)SBWR, operators do not have to be as reluctant about activating the SLCS, as these reactors have a reactor water cleanup system (RWCS) which is designed to remove boron – once the reactor has stabilized, the borated water within the RPV can be filtered through this system to promptly remove the soluble neutron absorbers that it contains and thus avoid damage to the internals of the plant.

Containment system

The ultimate safety system inside and outside of every BWR are the numerous levels of physical shielding that both protect the reactor from the outside world and protect the outside world from the reactor.

There are five levels of shielding:

The fuel rods inside the reactor pressure vessel are coated in thick Zircaloy shielding;
The reactor pressure vessel itself is manufactured out of 6-inch-thick (150 mm) steel, with extremely high temperature, vibration, and corrosion resistant surgical stainless steel grade 316L plate on both the inside and outside;
The primary containment structure is made of steel 1 inch thick;
The secondary containment structure is made of steel-reinforced, pre-stressed concrete 1.2–2.4 meters (3.9–7.9 ft) thick.
The reactor building (the shield wall/missile shield) is also made of steel-reinforced, pre-stressed concrete 0.3 to 1 m (0.98 to 3.28 ft) thick.

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.

Varieties of BWR containments

As illustrated by the descriptions of the systems above, BWRs are quite divergent in design from PWRs. Unlike the PWR, which has generally followed a very predictable external containment design (the stereotypical dome atop a cylinder), BWR containments are varied in external form but their internal distinctiveness is extremely striking in comparison to the PWR. There are five major varieties of BWR containments:

The "premodern" containment (Generation I); spherical in shape, and featuring a steam drum separator, or an out-of-RPV steam separator, and a heat exchanger for low-pressure steam, this containment is now obsolete, and is not used by any operative reactor.

the Mark I containment, consisting of a rectangular steel-reinforced concrete building, along with an additional layer of steel-reinforced concrete surrounding the steel-lined cylindrical drywell and the steel-lined pressure suppression torus below. The Mark I was the earliest type of containment in wide use, and many reactors with Mark Is are still in service today. There have been numerous safety upgrades made over the years to this type of containment, especially to provide for orderly reduction of containment load caused by pressure in a compounded limiting fault. The reactor building of the Mark I generally is in the form of a large rectangular structure of reinforced concrete.

the Mark II containment, similar to the Mark I, but omitting a distinct pressure suppression torus in favor of a cylindrical wetwell below the non-reactor cavity section of the drywell. Both the wetwell and the drywell have a primary containment structure of steel as in the Mark I, as well as the Mark I's layers of steel-reinforced concrete composing the secondary containment between the outer primary containment structure and the outer wall of the reactor building proper. The reactor building of the Mark II generally is in the form of a flat-topped cylinder.
the Mark III containment, generally similar in external shape to the stereotypical PWR, and with some similarities on the inside, at least on a superficial level. For example, rather than having a slab of concrete that staff could walk upon while the reactor was not being refueled covering the top of the primary containment and the RPV directly underneath, the Mark III takes the BWR in a more PWR-like direction by placing a water pool over this slab. Additional changes include abstracting the wetwell into a pressure-suppression pool with a weir wall separating it from the drywell.

Advanced containments; the present models of BWR containments for the ABWR and the ESBWR are harkbacks to the classical Mark I/II style of being quite distinct from the PWR on the outside as well as the inside, though both reactors incorporate the Mark III-ish style of having non-safety-related buildings surrounding or attached to the reactor building, rather than being overtly distinct from it. These containments are also designed to take far more stress than previous containments were, providing advanced safety. In particular, GE regards these containments as being able to withstand a direct hit by a tornado beyond Level 5 on the Old Fujita Scale with winds of 330+ miles per hour. Such a tornado has never been measured on earth. They are also designed to withstand seismic accelerations of .2 G, or nearly 2 meters per second² in any direction.

Containment Isolation System

Many valves passing in and out of the containment are required to be open to operate the facility. During an accident where radioactive material may be released, these valves must shut to prevent the release of radioactive material or the loss of reactor coolant. 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. The isolation system is separated into groups for major system functions. Each group contains its own criteria to trigger an isolation. The isolation system is similar to reactor protection system in that it consists of multiple channels, it is classified as safety-related, and that it requires confirmatory signals from multiple channels to issue an isolation to a system. 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. These isolation signals will lock out all of the valves in the group after closing them and must have all signals cleared before the lockout can be reset.

Isolation valves consist of 2 safety-related valves in series. One is an inboard valve, the other is an outboard valve. The inboard is located inside the containment, and the outboard is located just outside the containment. This provides redundancy as well as making the system immune to the single failure of any inboard or outboard valve operator or isolation signal. When an isolation signal is given to a group, both the inboard and outboard valves stroke closed. Tests of isolation logic must be performed regularly and is a part of each plant's technical specifications. The timing of these valves to stroke closed is a component of each plant's safety analysis and failure to close in the analyzed time is a reportable event.

Examples of isolation groups include the main steamlines, the reactor water cleanup system, the reactor core isolation cooling (RCIC) system, shutdown cooling, and the residual heat removal system. For pipes which inject water into the containment, two safety-related check valves are generally used in lieu of motor operated valves. These valves must be tested regularly as well to ensure they do indeed seal and prevent leakage even against high reactor pressures.

Hydrogen management

During normal plant operations and in normal operating temperatures, the hydrogen generation is not significant. When the nuclear fuel overheats, zirconium in Zircaloy cladding used in fuel rods oxidizes in reaction with steam:

Zr + 2H₂O → ZrO₂ + 2H₂

When mixed with air, hydrogen is flammable, and hydrogen detonation or deflagration may damage the reactor containment. In reactor designs with small containment volumes, such as in Mark I or II containments, the preferred method for managing hydrogen is pre-inerting with inert gas—generally nitrogen—to reduce the oxygen concentration in air below that needed for hydrogen combustion, and the use of thermal recombiners. Pre-inerting is considered impractical with larger containment volumes where thermal recombiners and deliberate ignition are used. Mark III containments have hydrogen igniters and hydrogen mixers which are designed to prevent the buildup of hydrogen through either pre-ignition prior to exceeding the lower explosive limit of 4%, or through recombination with Oxygen to make water.

The safety systems in action: the Design Basis Accident

The Design Basis Accident (DBA) for a nuclear power plant is the most severe possible single accident that the designers of the plant and the regulatory authorities could reasonably expect. It is, also, by definition, the accident the safety systems of the reactor are designed to respond to successfully, even if it occurs when the reactor is in its most vulnerable state. The DBA for the BWR consists of the total rupture of a large coolant pipe in the location that is considered to place the reactor in the most danger of harm—specifically, for older BWRs (BWR/1-BWR/6), the DBA consists of a "guillotine break" in the coolant loop of one of the recirculation jet pumps, which is substantially below the core waterline (LBLOCA, large break loss of coolant accident) combined with loss of feedwater to make up for the water boiled in the reactor (LOFW, loss of proper feedwater), combined with a simultaneous collapse of the regional power grid, resulting in a loss of power to certain reactor emergency systems (LOOP, loss of offsite power). The BWR is designed to shrug this accident off without core damage.

The description of this accident is applicable for the BWR/4.

The immediate result of such a break (call it time T+0) would be a pressurized stream of water well above the boiling point shooting out of the broken pipe into the drywell, which is at atmospheric pressure. As this water stream flashes into steam, due to the decrease in pressure and that it is above the water boiling point at normal atmospheric pressure, the pressure sensors within the drywell will report a pressure increase anomaly within it to the reactor protection system at latest T+0.3. The RPS will interpret this pressure increase signal, correctly, as the sign of a break in a pipe within the drywell. As a result, the RPS immediately initiates a full SCRAM, closes the main steam isolation valve (isolating the containment building), trips the turbines, attempts to begin the spinup of RCIC and HPCI, using residual steam, and starts the diesel pumps for LPCI and CS.

Now let us assume that the power outage hits at T+0.5. The RPS is on a float uninterruptible power supply, so it continues to function; its sensors, however, are not, and thus the RPS assumes that they are all detecting emergency conditions. Within less than a second from power outage, auxiliary batteries and compressed air supplies are starting the Emergency Diesel Generators. Power will be restored by T+25 seconds.

Let us return to the reactor core. Due to the closure of the MSIV (complete by T+2), a wave of backpressure will hit the rapidly depressurizing RPV but this is immaterial, as the depressurization due to the recirculation line break is so rapid and complete that no steam voids will likely collapse to liquid water. HPCI and RCIC will fail due to loss of steam pressure in the general depressurization, but this is again immaterial, as the 2,000 L/min (600 US gal/min) flow rate of RCIC available after T+5 is insufficient to maintain the water level; nor would the 19,000 L/min (5,000 US gal/min) flow of HPCI, available at T+10, be enough to maintain the water level, if it could work without steam. At T+10, the temperature of the reactor core, at approximately 285 °C (545 °F) at and before this point, begins to rise as enough coolant has been lost from the core that voids begin to form in the coolant between the fuel rods and they begin to heat rapidly. By T+12 seconds from the accident start, fuel rod uncovery begins. At approximately T+18 areas in the rods have reached 540 °C (1,004 °F). Some relief comes at T+20 or so, as the negative temperature coefficient and the negative void coefficient slows the rate of temperature increase. T+25 sees power restored; however, LPCI and CS will not be online until T+40.

At T+40, core temperature is at 650 °C (1,202 °F) and rising steadily; CS and LPCI kick in and begins deluging the steam above the core, and then the core itself. First, a large amount of steam still trapped above and within the core has to be knocked down first, or the water will be flashed to steam prior to it hitting the rods. 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. Following defueling, LPCI can be shut down. A long period of physical repairs will be necessary to repair the broken recirculation loop; overhaul the ECCS; diesel pumps; and diesel generators; drain the drywell; fully inspect all reactor systems, bring non-conformal systems up to spec, replace old and worn parts, etc. At the same time, different personnel from the licensee working hand in hand with the NRC will evaluate what the immediate cause of the break was; search for what event led to the immediate cause of the break (the root causes of the accident); and then to analyze the root causes and take corrective actions based on the root causes and immediate causes discovered. This is followed by a period to generally reflect and post-mortem the accident, discuss what procedures worked, what procedures didn't, and if it all happened again, what could have been done better, and what could be done to ensure it doesn't happen again; and to record lessons learned to propagate them to other BWR licensees. When this is accomplished, the reactor can be refueled, resume operations, and begin producing power once more.

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. The BWR 5s and 6s have additional tolerance, deeper water levels, and much faster emergency system reaction times. Fuel rod uncovery will briefly take place, but maximum temperature will only reach 600 °C (1,112 °F), far below the NRC safety limit.

According to a report by the U.S. Nuclear Regulatory Commission into the Fukushima Daiichi nuclear disaster, the March 2011 Tōhoku earthquake and tsunami that caused that disaster was an event "far more severe than the design basis for the Fukushima Daiichi Nuclear Power Plant". The reactors at this plant were BWR 3 and BWR 4 models. Their primary containment vessels had to be flooded with seawater containing boric acid, which will preclude any resumption of operation and was not anticipated in the DBA scenario. In addition, nothing similar to the chemical explosions that occurred at the Fukushima Daiichi plant was anticipated by the DBA.

Prior to the Fukushima Daiichi disaster, no incident approaching the DBA or even a LBLOCA in severity had occurred with a BWR. There had been minor incidents involving the ECCS, but in those circumstances it had performed at or beyond expectations. 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. The fire had nothing to do with the design of the BWR – it could have occurred in any power plant, and the lessons learned from that incident resulted in the creation of a separate backup control station, compartmentalization of the power plant into fire zones and clearly documented sets of equipment which would be available to shut down the reactor plant and maintain it in a safe condition in the event of a worst-case fire in any one fire zone. These changes were retrofitted into every existing US and most Western nuclear power plants and built into new plants from that point forth.

Notable activations of BWR safety systems

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. According to the Nuclear Energy Institute, "Coincident long-term loss of both on-site and off-site power for an extended period of time is a beyond-design-basis event for the primary containment on any operating nuclear power plant".

The reactors shut down as designed after the earthquake. However, the tsunami disabled four of the six sets of switchgear and all but three of the diesel backup generators which operated the emergency cooling systems and pumps. Pumps were designed to circulate hot fluid from the reactor to be cooled in the wetwell, but only units 5 and 6 had any power. Units 1, 2 and 3 reactor cores overheated and melted. Radioactivity was released into the air as fuel rods were damaged due to overheating by exposure to air as water levels fell below safe levels. 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. Reactors 1–3, and by some reports 4 all suffered violent hydrogen explosions March 2011 which damaged or destroyed their top levels or lower suppression level (unit 2).

As emergency measures, helicopters attempted to drop water from the ocean onto the open rooftops. Later water was sprayed from fire engines onto the roof of reactor 3. A concrete pump was used to pump water into the spent fuel pond in unit 4.

According to NISA, the accident released up to 10 petabecquerels of radioactive iodine-131 per hour in the initial days, and up to 630 PBq total, about one eighth the 5200 PBq released at Chernobyl.

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