Nuclear Meltdown - Light Water Reactors

Light Water Reactors

Before the core of a light water nuclear reactor can be damaged, two precursor events must have already occurred:

• A limiting fault (or a set of compounded emergency conditions) that leads to the failure of heat removal within the core (the loss of cooling). Low water level uncovers the core, allowing it to heat up.
• Failure of the Emergency Core Cooling System (ECCS). The ECCS is designed to rapidly cool the core and make it safe in the event of the maximum fault (the design basis accident) that nuclear regulators and plant engineers could imagine. There are at least two copies of the ECCS built for every reactor. Each division (copy) 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 were to occur, and a complete failure of all ECCS divisions were to occur, both Kuan, et al and Haskin, et al describe six stages between the start of the limiting fault (the loss of cooling) and the potential escape of molten corium into the containment (a so-called "full meltdown"):

1. Uncovering of the Core – In the event of a transient, upset, emergency, or limiting fault, LWRs 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 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 (0.5 °F/s) and 1 °C/s (1.8 °F/s) (3)."
3. Fuel ballooning and bursting – "In less than half an hour, the peak core temperature would reach 1,100 K (1,520 °F). 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 1,500 K (2,240 °F), 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 1,500 K (2,240 °F), the power from oxidation exceeds that from decay heat (4,5) 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 1,700 K (2,600 °F), 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 1,700 K (2,600 °F), 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. Release of molten core materials into 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 (FCI) to substantially stress or breach the primary pressure boundary when the corium relocates to the lower plenum of the reactor pressure vessel ("RPV"). 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 2,200 to 3,200 K (3,500 to 5,300 °F), its fall into liquid water at 550 to 600 K (530 to 620 °F) 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 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 said "despite melting of about one-third of the fuel, the reactor vessel itself maintained its integrity and contained the damaged fuel."