Corrosion of zircaloy in water results in the release of hydrogen. A portion of the hydrogen released, ranging from about 5% to 20%, diffuses through the oxide layer and into the metal. This causes embrittlement of the base metal that can lead to cladding failure. The mechanism of hydrogen embrittlement is discussed in Module 2, Properties of Metals. The zirconium alloy zircaloy-2, which has been used extensively as a fuel-rod cladding, is subject to hydrogen embrittlement, especially in the vicinity of surface defects. The alloy zircaloy-4 is, however, less susceptible to embrittlement. As with metals in general, irradiation decreases the ductility and increases the embrittlement of zirconium and the zircaloys. The magnitude of the radiation effect depends upon the neutron spectrum, fluence, temperature, and microstructure (or texture) of the material. Different fabrication processes yield products with different textures; therefore, the radiation embrittlement of zircaloy is dependent on its fabrication history.

Irradiation at high temperatures can lead to brittle fracture of stainless steels used as cladding in fast liquid metal breeder reactors. The effects of irradiation on metals is discussed in more detail in a later chapter of this module.

Effects on Fuel Due to Swelling and Core BurgjW

One of the requirements of a good fuel is to be resistant to radiation damage that can lead to dimensional changes (for example, by swelling, cracking, or creep). Early reactors and some older gas-cooled reactors used unalloyed uranium as the fuel. When unalloyed uranium is irradiated, dimensional changes occur that present drawbacks to its use as a fuel. The effects are of two types: 1) dimensional instability without appreciable change in density observed at temperatures below about 450C, and 2) swelling, accompanied by a decrease in density, which becomes important above 450C. Other reactors use ceramic fuels, with uranium dioxide being the most common, have the advantages of high-temperature stability and adequate resistance to radiation. Uranium dioxide (UO₂) has the ability to retain a large proportion of the fission gases, provided the temperature does not exceed about 1000C. Other oxide fuels have similar qualities.

Even though fission product swelling is less with oxide fuels, this irradiation-induced volume increase has been observed in UO₂ and mixed-oxide fuels for a number of years. This swelling of the fuel has generally been attributed to both gaseous fission-product bubble formation and the accumulation of solid fission products. Swelling can cause excessive pressure on the cladding, which could lead to fuel element cladding failure. Swelling also becomes a consideration on the lifetime of the fuel element by helping to determine the physical and mechanical changes resulting from irradiation and high temperature in the fuel and the cladding. Fuel element life or core burnup, which indicates the useful lifetime of the fuel in a reactor, is also determined by the decrease in reactivity due to the decrease in fissile material and the accumulation of fissionproduct poisons. Under operating conditions, fuel pellets undergo marked structural changes as a result of the high internal temperatures and the large temperature gradients. Thermal stresses lead to radial cracks and grain structure changes. These structural changes tend to increase with the specific power and burnup of the fuel.

Summary

The important information in this chapter is summarized below.

Nuclear Reactor Core Problems Summary

Fuel Pellet-Cladding Interaction (PCI)

PCI may lead to cladding failure and subsequent release of fission products into the reactor coolant.

Expansion of the fuel pellets due to high internal temperatures, cracking due to thermal stresses, and irradiation-induced swelling may lead to contact of the fuel with the cladding.

Design features to counteract PCI include:

An increase in the cladding thickness

An increase in the clad-pellet gap, with pressurization to obviate cladding collapse

The introduction of a layer of graphite or other lubricant between the fuel and the cladding

Operational limitations to reduce PCI

Plant procedures limit the maximum permissible rate at which power may be increased to lessen the effect of PCI.

Fuel Densification

Densification, which is the reverse of swelling, is a result of irradiation. Such behavior can cause the fuel material to contract and lead to irregularities in the thermal power generation.

Three principle effects:

An increase in the linear heat generation rate by an amount directly proportional to the decrease in pellet length

An increased local neutron flux and a local power spike in the axial gaps in the fuel column

A decrease in the clearance gap heat conductance between the pellets and the cladding. This decrease in heat transmission capability will increase the energy stored in the fuel pellet and will cause an increased fuel temperature.

To minimize these effects on power plant operation, limits are established on the power level rate of change and the maximum cladding temperature (1200C) allowable during a loss of coolant accident.

Fuel Cladding Embrittlement

Embrittlement is caused by hydrogen diffusing into the metal. Cladding embrittlement can lead to cladding failure.

Zircaloy-4 and different fabrication processes are used to minimize the effect of hydrogen embrittlement.

Fuel Burnup and Fission Product Swelling

High fuel burnup rate can cause the reactor to be refueled earlier than designed. Swelling can cause excessive pressure on the cladding, which could lead to fuel element cladding failure.

Operational maximum and minimum coolant flow limitations help prevent extensive fuel element damage.