Nuclear meltdown

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Three Mile Island Nuclear Generating Station consisted of two pressurized water reactors manufactured by Babcock & Wilcox each inside its own containment building and connected cooling towers. TMI-2, which suffered a partial meltdown causing severe fuel damage, is in the background.

A nuclear meltdown is a term for a severe nuclear reactor accident. This can occur when a nuclear power plant system or component failure causes the reactor core to cease being properly controlled and cooled to the extent that the sealed nuclear fuel assemblies – which contain the uranium or plutonium and highly radioactive fission products – begin to overheat and melt. A meltdown is considered very serious because of the possibility that the reactor containment will be defeated, thus releasing the core's highly radioactive and toxic elements into the atmosphere and environment. From an engineering perspective, a meltdown is likely to cause serious damage to the reactor, and possibly total destruction.

Several nuclear meltdowns of differing severity have occurred, from localized core damage to complete destruction of the reactor core. In some cases this has required extensive repairs or decommissioning of a nuclear reactor. In the most extreme cases, such as the Chernobyl disaster, deaths have resulted and the near-permanent civilian evacuation of a large area was required.

A nuclear explosion does not result from a nuclear meltdown because, by design, the geometry and composition of the reactor core do not permit the special conditions necessary for a nuclear explosion. However, the conditions that cause a meltdown may cause a non-nuclear explosion. For example, several power excursion accidents have caused coolant to rapidly over-pressurize, resulting in a steam explosion.

Contents

[edit] Causes

In some reactor types, the fuel assemblies in the core can melt due to the result of heat not being removed from the core. A nuclear reactor does not have to remain critical for a nuclear meltdown to occur because decay heat continues to heat the reactor fuel assemblies after the reactor has shut down, though it decreases significantly with time to the point where natural convection within the coolant combined with heat radiation from the RPV (and reradiation of heat from the RPV to the containment) will be sufficient to keep the core in a permanently steady state. This occurs after a period of days to weeks after control rods are reinserted into the reactor.

Meltdown, as this is called, is an extremely severe incident that may result from several factors, including a loss of pressure control accident, a loss of coolant accident (LOCA), an uncontrolled power excursion (not applicable to light water reactors), or a fire within the reactor core (not applicable to light water reactors). Failures in instrumentation and process control systems may amplify or even cause a series of events resulting in loss of cooling, though contemporary improvements in this area, and the philosophy of extreme conservativism in Western reactor design, known as defense in depth greatly reduce these possibilities.

Except in certain types of former Soviet reactors, such as the RBMK, of Chernobyl infamy, which were built without proper containment buildings, a core meltdown will not, by itself, result in the release of radioactivity to the environment due to the core being contained by 1.2 - 2.4 m (4 ft - 8 ft) of pre-stressed, steel-reinforced concrete, assuring minimal radioactive release in nearly any concievable circumstance.

  • In a loss of pressure control accident, the pressure of the confined coolant falls below specification without the means to restore it. In some cases this may reduce the heat transfer efficiency (when using an inert gas as a coolant) and in others may form an insulating 'bubble' of steam surrounding the fuel assemblies (for pressurized water reactors). In the latter case, due to localized heating of the steam 'bubble' due to decay heat, the pressure required to collapse the steam 'bubble' may exceed reactor design specifications until the reactor has had time to cool down. (This event is less likely to occur in boiling water reactors, where the core may be deliberately depressurized so that the Emergency Core Cooling System may be turned on).
  • In a loss of coolant accident, either the physical loss of coolant (which is typically deionized water, an inert gas, or liquid sodium) or the loss of a method to ensure a sufficient flow rate of the coolant occurs. A loss of coolant accident and a loss of pressure control accident are closely related in some reactors. In a pressurized water reactor, a loss of coolant accident can also cause a steam 'bubble' to form in the core due to excessive heating of stalled coolant or by the subsequent loss of pressure control accident caused by a rapid loss of coolant.
  • In an uncontrolled power excursion accident, a sudden power spike in the reactor exceeds reactor design specifications due to a sudden increase in reactor reactivity. An uncontrolled power excursion occurs due to significantly altering a parameter that affects the exponential rate of a nuclear chain reaction (examples include ejecting a control rod or significantly altering the nuclear characteristics of the moderator, such as by rapid cooling). In extreme cases the reactor may proceed to a condition known as prompt critical. This is especially a problem in reactors that have a positive void coefficient of reactivity, such as former Soviet RBMKs, a positive temperature coefficient, or can trap certain deleterious fission products within their fuel or moderators, such as former Soviet RBMKs; the Chernobyl disaster was caused by this condition. Western light water reactors are not subject to uncontrolled power excursions because loss of coolant decreases, rather than increases, core reactivity; "transients," as power fluctuations are called, are limited in LWRs to linear increases in reactivity that will rapidly decrease with time (approximately 125% - 150% of maximum thermal power for a few milliseconds in worst-case scenarios).
  • Core-based fires may also severely endanger the core and potentially cause the fuel assemblies to melt. A fire inside a reactor may be caused by an air addition to certain non-naval military or non-Western nuclear reactors (as it is possible for graphite to ignite inside the reactor core given oxygen) resulting in the uncontrolled heating of the coolant or moderator of the reactor. Without taking proper precautions Wigner energy may accumulate which will greatly increase the severity of the fire (for example, during the UK military's Windscale fire). Western light water reactors, by design, do not have flammable cores or moderators and are not subject to core fires.
  • Byzantine faults and cascading failures within instrumentation and control systems may cause severe problems in reactor operation, potentially leading to core damage. For example, a failure of an instrument to report liquid levels correctly may logarithmically amplify a minor problem, like a stuck-open relief valve; another example would be a fire within a cable-tray that so severely deranges the control pathways to essential machinery that the reactor is unable to be cooled using normal channels. This has been the route that the two emergencies within civil nuclear power in the West occurred. The Browns Ferry fire saw a fire start within a cable spreading room below the reactor control room. The cables were damaged, and reactor remote control was lost for several hours; however, the core was not damaged because plant personnel manually activated cooling systems. (Modifications including backup cable pathways and a secondary control room for safe shutdown have been installed in all Western plants since that time.) The Three Mile Island accident was caused by a stuck-open power operated pressure relief valve combined with a deceptive water level gauge that caused reactor operators to respond in a technically correct but practically wrong fashion to the contingency, which resulted in core damage. (Modifications to respond to this have included enhanced training for reactor operators, better instrumentation design, and redundant instrumentation pathways.)

[edit] Sequence of events

TMI-2 Core End-State Configuration

What happens when reactor fuel melts depends upon reactor design, and is the subject of conjecture and some actual experience (see below).

Before the core of a nuclear reactor can melt, a number of events/failures must already have happened. Once the core melts, it will almost certainly destroy the fuel bundles and internal structures of the reactor vessel (although it may not penetrate the reactor vessel). (Note that nearly half of the core at Three Mile Island melted but the molten debris [called "melt"] still stayed within the reactor vessel.) If the melt drops into a pool of water (for example, coolant or moderator), a steam explosion called a Fuel-Coolant Interaction (FCI) is likely. If air is available any exposed flammable substances will probably burn fiercely, but the liquid nature of the molten core poses special problems.

In the worst case scenario, the above-ground containment would fail at an early stage, (due to say an FCI within the reactor vessel, ejecting part of the vessel as a missile - this was the 'alpha-mode' failure of the 1975 Rasmussen (WASH-1400) study), or there could be a large hydrogen explosion or some other over-pressure event. Such an event could scatter urania-aerosol and volatile fission-products directly into the atmosphere. However, these events are considered essentially incredible in modern 'large-dry' containments. (The WASH-1400 report was replaced by better-based newer studies, and now the Nuclear Regulatory Commission has disavowed them all and is preparing the over-arching State-of-the-Art Reactor Consequence Analyses [SOARCA] study - see the Disclaimer in NUREG-1150.)

It has not been determined to what extent a molten mass can melt through a structure (although that was tested in the Loss-of-Fluid-Test Reactor described in Test Area North's fact sheet[1]). The molten reactor core could penetrate the reactor vessel and the containment structure and burn down (via a melt-concrete interaction) to groundwater (this has not happened at any meltdown to date: see China Syndrome). A water moderated reactor would go non-critical as soon as the water boiled away (with a fast reactor it is possible that the molten mass might mix with any material it melts, diluting itself down to a non-critical state). In the Chernobyl accident, the fuel became non-critical when it melted and flowed away from the graphite moderator - however, it took considerable time to cool. If hot uranium dioxide is combined with iron(II) oxide a eutectic is formed which may cause the fuel to become more mobile than it would otherwise be.[2]

It should be noted that the molten core of Chernobyl (that part that didn't vaporize in the fire) flowed in a channel created by the structure of its reactor building (e.g., walls and stairways) and froze in place before a core-concrete interaction could happen. In the basement of the reactor at Chernobyl, a large "elephant's foot" of congealed core material was found. Furthermore, the time delay and the lack of a direct path to the atmosphere (such as a containment building is designed to provide) would work to significantly ameliorate the radiological release. Any steam-explosions/FCI which occurred would probably work mainly to increase cooling of the core-debris. However, if the basement of the reactor building were penetrated the groundwater itself would likely be severely contaminated, and its flow could carry the contamination far afield.

In the best case scenario, the reactor vessel would hold the molten material (as at Three Mile Island), limiting most of the damage to the reactor itself. The American Nuclear Society has said "despite melting of about one-third of the fuel, the reactor vessel itself maintained its integrity and contained the damaged fuel".[3] However the Three Mile Island example also illustrates the difficulty in predicting such behavior: the reactor vessel was not built for, and not expected to remain intact with, the temperatures it experienced when it the core melted, but possibly because some of the melted material collected at the bottom of the vessel and cooled early on in the accident, it created a resistant shell against further pressure and heat. Such a possibility was not predicted by the engineers who designed the reactor and would not necessarily occur under duplicate conditions, but was largely seen as instrumental in the preservation of the reactor vessel's integrity. (However, it should be noted that the reactor vessel was inside a containment building, as in all non-Soviet nuclear plants, so a failure of the reactor vessel would not automatically mean that radioactive material would be released into the environment.)

All non-Soviet nuclear power plants are designed with Emergency Core Cooling Systems, some active and some passive and automatic. CANDU reactors, for example, are designed with at least one, and generally two, large low-temperature and low-pressure water reservoirs around its fuel/coolant channels. The first is the bulk heavy-water moderator (a separate system from the coolant), and the second is the light-water-filled shield tank. It has been shown that even under severe loss-of-coolant conditions these backup heat sinks are sufficient to prevent either the fuel meltdown in the first place (using the moderator heat sink), or the breaching of the core vessel should the moderator eventually boil off (using the shield tank heat sink). [Allen et al.]

[edit] Prevention, Suppression, and Containment of Core Damage Events in Former Soviet Reactors

Former Soviet RBMKs, however, found only in Russia and the CIS, do not have containment buildings and also have ECCS systems that are considered grossly insufficient by Western standards. As such, it might be possible to stop a loss of coolant event prior to core damage occuring, but once core damage begins, massive radioactive release is assured.

[edit] Former Soviet (present Russian & ROW) VVERs

The VVER is a former Soviet-origin pressurized light water reactor that is far more inherently stable and inherently safe than the former Soviet RBMK. This is because it uses light water as a moderator (rather than graphite), has well understood operating characteristics, and has a negative void coefficient of reactivity. In addition, some have been built with more than marginal containments, some have quality ECCS systems, and some have been upgraded to international standards of control and instrumentation. Present generations of VVERs (the VVER-1000) are built to Western-equivalent levels of instrumentation, control, and containment systems.

However, even with these positive developments, certain older VVER models raise a high level of concern, especially the VVER-440 V230. This is due to the fact that these were built with extremely marginal "confinements" (not a containment building), have inadequate instrumentation and control systems, and have marginal ECCS systems. Still, marginal ECCS systems can be backfitted and instrumentation and control systems can be retrofitted. During the 1970s, Finland built 2 VVER-440 V230 models, however, the Finns built them to Western standards with a full containment, world-class instrumentation and control standards, and major ECCS enhancements. The Bulgarians also had a bunch of VVER-440 V230 models, but they opted to shut them down upon joining the EU rather than backfit them, and are instead building new VVER-1000 models. Many non-EU states maintain V230 models, including Russia and the CIS. Many of these states - rather than abandoning the reactors entirely - have opted to upgrade the ECCS, develop standard procedures, install proper instrumentation and control systems, and strengthen the confinement buildings. Though confinements cannot be transformed into containments, the risk of a limiting fault resulting in core damage can be greatly reduced.

The VVER-440 V213 model was built to the first set of Soviet nuclear safety standards. It possesses a modest containment building, and the ECCS systems, though not completely to Western standards, are reasonably comprehensive. Many VVER-440 V213 models possessed by former Soviet bloc countries have been upgraded to fully automated Western-style instrumentation and control systems, improving safety to Western levels for accident prevention - but not for accident containment, which is of a modest level compared to Western plants. These reactors are regarded as "safe enough" by Western standards to continue operation without major modifications, though most owners have performed major modifications to bring them up to generally equivalent levels of nuclear safety.

The VVER-1000 type has a definitely adequate Western-style containment, the ECCS is robust by Western standards, and instrumentation and control has been markedly improved to Western levels.

[edit] Comparability analysis

It may safely be assumed that with RBMKs of any type, any limiting fault followed by partial or total ECCS failure or failure to SCRAM when indicated will result in core damage and radioactive release to the environment.

The following assumptions may be made about the VVER reactors:

  • VVER-440 V230 models WITHOUT substantial upgrades: Assume that limiting fault (LBLOCA) will result in core damage if ECCS suffers any degredation in performance, delayed activation, or failures. Assume that radioactive release to environment is assured if RPV is breached.
  • VVER-440 V230 models WITH substantial upgrades: Assume that limiting fault (LBLOCA) is less likely to result in core damage than unmodified V230; in particular, ECCS will have sufficient capacity to respond to limiting faults with some redundancy. Confinement strengthening may prevent radioactive release in some core damage scenarios.
  • VVER-440 V230 Finnish models: Assume will perform at level of Generation II Western PWR.
  • VVER-440 V213: Assume that limiting fault (LBLOCA) will successfully be responded to by ECCS, and that reserve capacity does exist for ECCS; this will prevent core damage in most circumstances. If core damage does occur, assume that - depending on severity - radioactive release to the environment could take place with stock containment.
  • VVER-1000: Make assumptions based on "newer" (post-1985) Generation II reactors.

[edit] Prevention, Suppression, and Containment of Core Damage Events in Western Reactors

Within the design of Western reactors, a great deal of work goes into the prevention of a serious core damage event. If such an event were to occur, three different physical processes will provide additional time to the plant operators between the start of the accident (the loss of cooling) and the escape of molten corium into the containment (a full meltdown):

  1. The time required for the water to boil away (coolant, moderator). In the event of a LOCA, LWRs and CANDUs are designed to automatically SCRAM (a SCRAM being the immediate and full insertion of all control rods) and spin up the ECCS. This greatly reduces reactor thermal power (but does not remove it completely); this delays core "uncovery", which is defined as the point when the fuel rods are no longer covered by coolant and can begin to heat up.
  2. The time required for the fuel to melt. After the water has boiled, then the time required for the fuel to reach its melting point will be dictated by the heat input due to decay of fission products, the heat capacity of the fuel and the melting point of the fuel. In the worst cases involving Generation II LWRs, between 5 to 10 minutes are required for the fuel to heat beyond the critical fuel surface temperature, 1100oC or 2200oF. This fuel melting temperature is conservative - but is the point beyond which "all warranties are void". Assuming the ECCS can be activated within the 5 to 10 minutes prior to the excession of this critical temperature, the reactor will return to stability without core damage. The ECCS automatically spins up upon SCRAM, so worst-case scenarios predict approximatly 40 seconds from event initiation to ECCS activation if at least part of the ECCS is functional. As the ECCS has multiple, redundant backup systems, it is highly unlikely that it will completely fail.
    1. However, if the ECCS cannot be fully or partially activated, then events continue - and the next event, after these 5 to 10 minutes are up, is significant fuel failure. Fuel failure occurs when, due to the temperature, the zircalloy fuel sheathing of the fuel rods cracks and releases fission products. This will be detected by a massive rise in radioactivity within the RPV and the primary coolant piping due to the release of fission products. If the coolant loop is breached, as in a LOCA, radiation levels will rise to high levels within the primary containment as fission products are released into the containment.
    2. Once again, if the ECCS can be fully or partially activated before the accident progresses, the chain of events may be stopped. But otherwise, events will progress. Some time will pass between fuel failure and corium formation; design conservatism, however, puts the moment of corium formation as beginning at the point of fuel failure. Once corium has formed, it's safe to assume the reactor is a total loss.
  3. The time required for the molten fuel to breach the primary pressure boundary. This consists of the time required for the molten metal of the core (the corium) to breach the primary pressure boundary (in light water reactors this is the pressure vessel; in CANDU and RBMK reactors this is the array of pressurized fuel channels) will depend on temperatures and boundary materials. Whether or not the fuel remains critical in the conditions inside the damaged core or beyond will play a significant role. Time estimates indicate in a worst-case Western LWR event with complete loss of the ECCS, there remains between 30 and 150 minutes from corium formation prior to RPV breach. Even partial ECCS activation can delay this significantly, and provide time for the remainder of the ECCS to be brought back online. It's highly unlikely that the staff of a Western LWR will be completely unable to restore at least part of the ECCS prior to the RPV being breached.

If the RPV is breached, at least in Western plants, there is an airtight containment building consisting of pre-stressed, steel-reinforced concrete 1.2 - 2.4 meters thick that stands between the molten corium and the outside world. Though radiation would be at a lethal level within the primary containment, doses outside of it would be insignificant. Further, modern containments are designed - or have been retrofitted - to allow for the orderly release of pressurized gasses that may be generated in an event without releasing radionucleides. (This is done by piping a pressure release valve to a series of activated carbon and HEPA filters that are designed to trap any radionucleoides in the event that pressure release from the containment becomes necessary.) Hydrogen/oxygen recombiners also are installed within the containment to prevent any combination of gasses from building up within that could deflagrate and threaten containment integrity.

This assures that even with a molten core cooling within the containment building, there is almost no possibility of any offsite dose of significance to local citizens; for example, in the Three Mile Island event in 1979, a theoretical person standing at the plant property line during the entire event would have received a dose of approximately 2 millisieverts (200 millirem), between a chest X-ray's and a CT scan's worth of radiation. This was due to outgassing by an uncontrolled system that, today, would have been backfitted with activated carbon and HEPA filters to prevent radionuclide release.

Thus, if all else fails, the containment can be sealed and abandoned in place with release of extremely limited offsite radioactivity. Pressure management will have to be observed carefully, at least in the near term and responded to as indicated. After a number of years for fission products to decay - probably taking a decade or two - the containment can be reopened for decontamination and demolition.

Still, even though the secondary containment consists of pre-stressed, steel-reinforced concrete between 1.2 - 2.4 meters thick, there is a possibility, however remote, that the containment could be breached after the core damage event occurred. This might take place if:

  1. A >8.0 Richter scale earthquake occurred;
  2. A F6 tornado with 320+ mph winds hit it;
  3. It was attacked with bunker-buster bombs;
  4. A nuclear weapon detonated in the immediate vicinity;
  5. It was struck by an asteroid.

[edit] If the containment is breached

The longer the reactor operators were able to retain the fission products within the core will reduce the size of the radioactive release. This is because the worst isotopes in a fission product mixture are short lived. For example if all the iodine in a core was released one week after criticality was terminated by a SCRAM then the thyroid dose suffered by the population would be lower than if the iodine had escaped the plant one hour after the reactor was scrammed. Even while the Chernobyl accident had dire[quantify] off-site effects, much of the radioactivity remained within the building. If the building were to fail and dust was to be released into the environment then the release of a given mass of fission products which have aged for twenty years would have a smaller effect than the release of the same mass of fission products (in the same chemical and physical form) which had only undergone a short cooling time (such as one hour) after the nuclear reaction has been terminated. However if a nuclear reaction was to occur again within the Chernobyl plant (for instance if rainwater was to collect and act as a moderator) then the new fission products would have a higher specific activity and thus pose a greater threat if they were released. N.B. to prevent a post accident nuclear reaction steps have been taken (such as adding neutron poisons to key parts of the basement).

[edit] Effects

The effects of a nuclear meltdown depend on the safety features designed into a reactor. A modern reactor is designed both to make a meltdown highly unlikely, and to contain one should it occur. In the future passively safe or inherently safe designs will make the possibility exceedingly unlikely.

In a modern reactor, a nuclear meltdown, whether partial or total, should be contained inside the reactor's containment structure. Thus (assuming that no other major disasters occur) while the meltdown will severely damage the reactor itself, possibly contaminating the whole structure with highly radioactive material, a meltdown alone will generally not lead to significant radiation release or danger to the public. The effects are therefore primarily economic[4].

In practice, however, a nuclear meltdown is often part of a larger chain of disasters (although there have been so few meltdowns in the history of nuclear power that there is not a large pool of statistical information from which to draw a credible conclusion as to what "often" happens in such circumstances). For example, in the Chernobyl accident, by the time the core melted, there had already been a large steam explosion and graphite fire and major release of radioactive contamination (as with almost all Soviet reactors, there was no containment structure at Chernobyl).

[edit] Reactor design

Although pressurized water reactors are more susceptible to nuclear meltdown in the absence of active safety measures, this is not a universal feature of civilian nuclear reactors. Much of the research in civilian nuclear reactors is for designs with passive safety features that would be much less susceptible to meltdown, even if all emergency systems failed. For example, pebble bed reactors are designed so that complete loss of coolant for an indefinite period does not result in the reactor overheating. The General Electric ESBWR and Westinghouse AP1000 have passively-activated safety systems. The CANDU reactor has two low-temperature and low-pressure water systems surrounding the fuel (i.e. moderator and shield tank) that act as back-up heat sinks and preclude meltdowns and core-breaching scenarios [Allen et al.].

Fast breeder reactors are more susceptible to meltdown than other reactor types, due to the larger quantity of fissile material and the higher neutron flux inside the reactor core, which makes it more difficult to control the reaction.

Accidental fires are widely acknowledged to be risk factors that can contribute to a nuclear meltdown. It is for this reason that circuit integrity measures are used for the electrical wiring that runs between control rooms and reactors. Ideally, a reactor is equipped with two "shutdown trains" or two sets of wires so that if one should fail, the other can be used to shut down the reactor. This common procedure became the subject of controversy during the Thermo-Lag scandal, when whistleblower Gerald W. Brown notified the NRC that the fire testing used to qualify Thermo-Lag was inadequate, meaning the fire-resistance rating thought to exist was in fact much lower, which meant that the majority of NRC licensees did not have operable protection of its safe shutdown wiring. Similar criticisms were leveled by US Congressman Ed Markey at the use of combustible silicone foam as firestops. The problem did not occur in German plants as operators must follow not just the directives of their federal regulators but are also required to follow the local building code, which makes product certification mandatory. Bounding in US and Canadian plants is not based on product certification. The Canadian disclosures by Gerald W. Brown revealed that Canadian plants also used unbounded silicone foam and Elastaseal based on indefensible test reports. The safe shutdown trains, typically consisting of wiring inside of cable trays used single-sided "fireproofing", consisting of sheet metal and proprietary intumescent sheets, for three dimensional cable trays. The disclosures were made public by the Canadian Broadcasting Corporation's "The National" program, which caused the proceedings of the Select Committee on Ontario Hydro Nuclear Affairs to take place. Still, to this date, neither the NRC, nor the Canadian Nuclear Safety Commission require product certification, which is mandatory for civilian construction.

[edit] Other theoretical consequences of a nuclear meltdown

If the reactor core becomes too hot, it might melt through the reactor vessel (although this has not happened to date) and the floor of the reactor chamber and descend until it becomes diluted by surrounding material and cooled enough to no longer melt through the material underneath, or until it hits groundwater. This type of nuclear meltdown is known as a China Syndrome. Note that a nuclear explosion does not happen in a nuclear meltdown due to the low fissility of the radioactive components. However, a steam explosion may occur if it hits water.

The geometry and presence of the coolant has a twin role, and both cools the reactor as well as slowing down emitted neutrons. The latter role is crucial to maintaining the chain-reaction, and so even without coolant the molten core is designed to be unable to form an uncontrolled critical mass (a recriticality). However, the molten reactor core will continue generating enough heat through unmoderated radioactive decay ('decay heat') to maintain or even increase its temperature.

[edit] Meltdowns that have occurred

A number of Russian nuclear submarines have experienced nuclear meltdowns. The only known large scale nuclear meltdowns at civilian nuclear power plants were in the Chernobyl disaster at Chernobyl Nuclear Power Plant, Ukraine, in 1986, and the Three Mile Island accident at Three Mile Island, Pennsylvania, USA, in 1979, although there have been partial core meltdowns at:

Not all of these were caused by a loss of coolant and in several cases (the Chernobyl disaster and the Windscale fire, for example) the meltdown was not the most severe problem.

[edit] See also

[edit] References

  • Rasmussen N. (editor) (1975) Reactor Safety Study WASH-1400, USNRC
  • P.J. Allen, J.Q. Howieson, H.S. Shapiro, J.T. Rogers, P. Mostert and R.W. van Otterloo, "Summary of CANDU 6 Probabilistic Safety Assessment Study Results", Nuclear Safety, Vol 31 No 2 Ap-Jn 1990.
  1. ^ Test Area North
  2. ^ S.V. Bechta, E.V. Krushinov, V.I. Almjashev, S.A. Vitol, L.P. Mezentseva, Yu.B. Petrov, D.B. Lopukh, V.B. Khabensky, M. Barrachin, S. Hellmann, K. Froment, M. Fisher, W. Tromm, D. Bottomley, F. Defoort and V.V. Gusarov, Journal of Nuclear Materials, 2007, 362, 46
  3. ^ ANS : Public Information : Resources : Special Topics : History at Three Mile Island : What Happened and What Didn't in the TMI-2 Accident
  4. ^ Partial Fuel Meltdown Events
  5. ^ Page 300, Radioactivity, Ionizing Radiation and Nuclear Energy, Jiŕí Hála and James D. Navratil, Published by Konvoj (Brno) 2003, ISBN 807302053X

[edit] External links

Retrieved from "http://en.wikipedia.org/wiki/Nuclear_meltdown"
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