Nuclear meltdown


A nuclear meltdown is a severe nuclear reactor accident that results in core damage from overheating. The term nuclear meltdown is not officially defined by the International Atomic Energy Agency, however it has been defined to mean the accidental melting of the core or fuel of a nuclear reactor, and is in common usage a reference to the core's either complete or partial collapse.
A core meltdown accident occurs when the heat generated by a nuclear reactor exceeds the heat removed by the cooling systems to the point where at least one nuclear fuel element exceeds its melting point. This differs from a fuel element failure, which is not caused by high temperatures. A meltdown may be caused by a loss of coolant, loss of coolant pressure, or low coolant flow rate, or be the result of a criticality excursion in which the reactor's power level exceeds its design limits.
Once the fuel elements of a reactor begin to melt, the fuel cladding has been breached, and the nuclear fuel and fission products within the fuel elements can leach out into the coolant. Subsequent failures can permit these radioisotopes to breach further layers of containment. Superheated steam and hot metal inside the core can lead to fuel–coolant interactions, hydrogen explosions, or steam hammer, any of which could destroy parts of the containment. A meltdown is considered very serious because of the potential for radioactive materials to breach all containment and escape into the environment, resulting in radioactive contamination and fallout, and potentially leading to radiation poisoning of people and animals nearby.

Causes

Nuclear power plants generate electricity by heating fluid via a nuclear reaction to run a generator. If the heat from that reaction is not removed adequately, the fuel assemblies in a reactor core can melt. A core damage incident can occur even after a reactor is shut down because the fuel continues to produce decay heat.
A core damage accident is caused by the loss of sufficient cooling for the nuclear fuel within the reactor core. The reason may be one of several factors, including a loss-of-pressure-control accident, a loss-of-coolant accident, an uncontrolled power excursion. Failures in control systems may cause a series of events resulting in loss of cooling. Contemporary safety principles of defense in depth ensure that multiple layers of safety systems are always present to make such accidents unlikely.
The containment building is the last of several safeguards that prevent the release of radioactivity to the environment. Many commercial reactors are contained within a thick pre-stressed, steel-reinforced, air-tight concrete structure that can withstand hurricane-force winds and severe earthquakes.
  • In a loss-of-coolant accident, either the physical loss of coolant 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 LOCA 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 a loss-of-forced-circulation accident, a gas cooled reactor's circulators fail to circulate the gas coolant within the core, and heat transfer is impeded by this loss of forced circulation, though natural circulation through convection will keep the fuel cool as long as the reactor is not depressurized.
  • 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, and in others may form an insulating "bubble" of steam surrounding the fuel assemblies. 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.. In a depressurization fault, a gas-cooled reactor loses gas pressure within the core, reducing heat transfer efficiency and posing a challenge to the cooling of fuel; as long as at least one gas circulator is available, however, the fuel will be kept cool.

    Light-water reactors (LWRs)

Before the core of a light-water nuclear reactor can be damaged, two precursor events must have already occurred:
  • A limiting fault that leads to the failure of heat removal within the core. Low water level uncovers the core, allowing it to heat up.
  • Failure of the emergency core cooling system. The ECCS is designed to rapidly cool the core and make it safe in the event of the maximum fault that nuclear regulators and plant engineers could imagine. There are at least two copies of the ECCS built for every reactor. Each division of the ECCS is capable, by itself, of responding to the design basis accident. The latest reactors have as many as four divisions of the ECCS. This is the principle of redundancy, or duplication. As long as at least one ECCS division functions, no core damage can occur. Each of the several divisions of the ECCS has several internal "trains" of components. Thus the ECCS divisions themselves have internal redundancy – and can withstand failures of components within them.
The Three Mile Island accident was a compounded group of emergencies that led to core damage. What led to this was an erroneous decision by operators to shut down the ECCS during an emergency condition due to gauge readings that were either incorrect or misinterpreted; this caused another emergency condition that, several hours after the fact, led to core exposure and a core damage incident. If the ECCS had been allowed to function, it would have prevented both exposure and core damage. During the Fukushima incident the emergency cooling system had also been manually shut down several minutes after it started.
If such a limiting fault occurs, and a complete failure of all ECCS divisions also occurs, both Kuan, et al and Haskin, et al describe six stages between the start of the limiting fault and the potential escape of molten corium into the containment :
  1. Uncovering of the core – In the event of a transient, upset, emergency, or limiting fault, LWRs are designed to automatically SCRAM and spin up the ECCS. This greatly reduces reactor thermal power ; this delays core becoming uncovered, which is defined as the point when the fuel rods are no longer covered by coolant and can begin to heat up. As Kuan states: "In a small-break LOCA with no emergency core coolant injection, core uncovery generally begins approximately an hour after the initiation of the break. If the reactor coolant pumps are not running, the upper part of the core will be exposed to a steam environment and heatup of the core will begin. However, if the coolant pumps are running, the core will be cooled by a two-phase mixture of steam and water, and heatup of the fuel rods will be delayed until almost all of the water in the two-phase mixture is vaporized. The TMI-2 accident showed that operation of reactor coolant pumps may be sustained for up to approximately two hours to deliver a two phase mixture that can prevent core heatup."
  2. Pre-damage heat up – "In the absence of a two-phase mixture going through the core or of water addition to the core to compensate water boiloff, the fuel rods in a steam environment will heat up at a rate between 0.3 °C/s and 1 °C/s ."
  3. Fuel ballooning and bursting – "In less than half an hour, the peak core temperature would reach. At this temperature, the zircaloy cladding of the fuel rods may balloon and burst. This is the first stage of core damage. Cladding ballooning may block a substantial portion of the flow area of the core and restrict the flow of coolant. However, complete blockage of the core is unlikely because not all fuel rods balloon at the same axial location. In this case, sufficient water addition can cool the core and stop core damage progression."
  4. Rapid oxidation – "The next stage of core damage, beginning at approximately, is the rapid oxidation of the Zircaloy by steam. In the oxidation process, hydrogen is produced and a large amount of heat is released. Above, the power from oxidation exceeds that from decay heat unless the oxidation rate is limited by the supply of either zircaloy or steam."
  5. Debris bed formation – "When the temperature in the core reaches about, molten control materials will flow to and solidify in the space between the lower parts of the fuel rods where the temperature is comparatively low. Above, the core temperature may escalate in a few minutes to the melting point of zircaloy due to increased oxidation rate. When the oxidized cladding breaks, the molten zircaloy, along with dissolved UO2 would flow downward and freeze in the cooler, lower region of the core. Together with solidified control materials from earlier down-flows, the relocated zircaloy and UO2 would form the lower crust of a developing cohesive debris bed."
  6. Corium relocation to the lower plenum – "In scenarios of small-break LOCAs, there is generally a pool of water in the lower plenum of the vessel at the time of core relocation. The release of molten core materials into the water always generates large amounts of steam. If the molten stream of core materials breaks up rapidly in water, there is also a possibility of a steam explosion. During relocation, any unoxidized zirconium in the molten material may also be oxidized by steam, and in the process hydrogen is produced. Recriticality also may be a concern if the control materials are left behind in the core and the relocated material breaks up in unborated water in the lower plenum."
At the point at which the corium relocates to the lower plenum, Haskin, et al relate that the possibility exists for an incident called a fuel–coolant interaction to substantially stress or breach the primary pressure boundary when the corium relocates to the lower plenum of the reactor pressure vessel.
This is because the lower plenum of the RPV may have a substantial quantity of water - the reactor coolant - in it, and, assuming the primary system has not been depressurized, the water will likely be in the liquid phase, and consequently dense, and at a vastly lower temperature than the corium. Since corium is a liquid metal-ceramic eutectic at temperatures of, its fall into liquid water at may cause an extremely rapid evolution of steam that could cause a sudden extreme overpressure and consequent gross structural failure of the primary system or RPV. Though most modern studies hold that it is physically infeasible, or at least extraordinarily unlikely, Haskin, et al state that there exists a remote possibility of an extremely violent FCI leading to something referred to as an alpha-mode failure, or the gross failure of the RPV itself, and subsequent ejection of the upper plenum of the RPV as a missile against the inside of the containment, which would likely lead to the failure of the containment and release of the fission products of the core to the outside environment without any substantial decay having taken place.
The American Nuclear Society has commented on the TMI-2 accident, that despite melting of about one-third of the fuel, the reactor vessel maintained its integrity and contained the damaged fuel.