This page describes how uranium dioxide nuclear fuel behaves during both normal nuclear reactor operation and under reactor accident conditions, such as overheating. Work in this area is often very expensive to conduct, and so has often been performed on a collaborative basis between groups of countries, usually under the aegis of the Organisation for Economic Co-operation and Development's Committee on the Safety of Nuclear Installations (CSNI).
Swelling
Cladding Both the fuel and cladding can swell. Cladding covers the fuel to form a fuel pin and can be deformed. It is normal to fill the gap between the fuel and the cladding with helium gas to permit better thermal contact between the fuel and the cladding. During use the amount of gas inside the fuel pin can increase because of the formation of noble gases (krypton and xenon) by the fission process. If a Loss-of-coolant accident (LOCA) (e.g. Three Mile Island) or a Reactivity Initiated Accident (RIA) (e.g. Chernobyl or SL-1) occurs then the temperature of this gas can increase. As the fuel pin is sealed the pressure of the gas will increase (PV = nRT) and it is possible to deform and burst the cladding. It has been noticed that both corrosion and irradiation can alter the properties of the zirconium alloy commonly used as cladding, making it brittle. As a result, the experiments using unirradiated zirconium alloy tubes can be misleading.
According to one paper the following difference between the cladding failure mode of unused and used fuel was seen.
Unirradiated fuel rods were pressurized before being placed in a special reactor at the Japanese Nuclear Safety Research Reactor (NSRR) where they were subjected to a simulated RIA transient. These rods failed after ballooning late in the transient when the cladding temperature was high. The failure of the cladding in these tests was ductile, and it was a burst opening.
The used fuel (61 GW days/tonne of uranium) failed early in the transient with a brittle fracture which was a longitudinal crack.
It was found that hydrided zirconium tube is weaker and the bursting pressure is lower.
The common failure process of fuel in the water-cooled reactors is a transition to film boiling and subsequent ignition of zirconium cladding in the steam. The effects of the intense hot hydrogen reaction product flow on the fuel pellets and on the bundle's wall well represented on the sidebar picture.
Fuel The nuclear fuel can swell during use, this is because of effects such as fission gas formation in the fuel and the damage which occurs to the lattice of the solid. The fission gases accumulate in the void that forms in the center of a fuel pellet as burnup increases. As the void forms, the once-cylindrical pellet degrades into pieces. The swelling of the fuel pellet can cause pellet-cladding interaction when it thermally expands to the inside of the cladding tubing. The swollen fuel pellet imposes mechanical stresses upon the cladding. A document on the subject of the swelling of the fuel can be downloaded from the NASA web site.
Fission gas release As the fuel is degraded or heated the more volatile fission products which are trapped within the uranium dioxide may become free. For example, see Colle et al..
A report on the release of 85Kr, 106Ru and 137Cs from uranium when air is present has been written. It was found that uranium dioxide was converted to U3O8 between about 300 and 500 °C in air. They report that this process requires some time to start, after the induction time the sample gains mass. The authors report that a layer of U3O7 was present on the uranium dioxide surface during this induction time. They report that 3 to 8% of the krypton-85 was released, and that much less of the ruthenium (0.5%) and caesium (2.6 × 10−3%) occurred during the oxidation of the uranium dioxide.P. Wood and G.H. Bannister, CEGB report
Heat transfer between the cladding and the water In a water-cooled power reactor (or in a water-filled spent fuel pool, SFP), if a power surge occurs as a result of a reactivity initiated accident, an understanding of the transfer of heat from the surface of the cladding to the water is very useful. In a French study, metal pipe immersed in water (both under typical PWR and SFP conditions), was electrically heated to simulate the generation of heat within a fuel pin by nuclear processes. The temperature of the pipe was monitored by thermocouples and for the tests conducted under PWR conditions the water entering the larger pipe (14.2 mm diameter) holding the test metal pipe (9.5 mm outside diameter and 600 mm long) was at 280 °C and 15 MPa. The water was flowing past the inner pipe at circa 4 ms−1 and the cladding was subjected to heating at 2200 to 4900 °C s−1 to simulate an RIA. It was found that as the temperature of the cladding increased the rate of heat transfer from the surface of the cladding increased at first as the water boiled at nucleation sites. When the heat flux is greater than the critical heat flux a boiling crisis occurs. This occurs as the temperature of the fuel cladding surface increases so that the surface of the metal was too hot (surface dries out) for nucleation boiling. When the surface dries out the rate of heat transfer decreases, after a further increase in the temperature of the metal surface the boiling resumes but it is now film boiling.
Hydriding and waterside corrosion As a nuclear fuel bundle increases in burnup (time in reactor), the radiation begins changing not only the fuel pellets inside the cladding, but the cladding material itself. The zirconium chemically reacts to the water flowing around it as coolant, forming a protective oxide on the surface of the cladding. Typically a fifth of the cladding wall will be consumed by oxide in PWRs. There is a smaller corrosion layer thickness in BWRs. The chemical reaction that takes place is:
Zr + 2 H2O → ZrO2 + 2 H2 (g)
Hydriding occurs when the product gas (hydrogen) precipitates out as hydrides within the zirconium. This causes the cladding to become embrittled, instead of ductile. The hydride bands form in rings within the cladding. As the cladding experiences hoop stress from the growing amount of fission products, the hoop stress increases. The material limitations of the cladding is one aspect that limits the amount of burnup nuclear fuel can accumulate in a reactor.
CRUD (Chalk River Unidentified Deposits) was discovered by Chalk River Laboratories. It occurs on the exterior of the clad as burnup is accumulated.
When a nuclear fuel assembly is prepared for onsite storage, it is dried and moved to a spent nuclear fuel shipping cask with scores of other assemblies. Then it sits on a concrete pad for a number of years waiting for an intermediate storage facility or reprocessing. The transportation of radiation-damaged cladding is tricky, because it is so fragile. After being removed from the reactor and cooling down in the spent fuel pool, the hydrides within the cladding of an assembly reorient themselves so that they radially point out from the fuel, rather than circularly in the direction of the hoop stress. This puts the fuel in a situation so that when it is moved to its final resting place, if the cask were to fall, the cladding would be so weak it could break and release the spent fuel pellets inside the cask.
Corrosion on the inside of the cladding Zirconium alloys can undergo stress corrosion cracking when exposed to iodine; the iodine is formed as a fission product which depending on the nature of the fuel can escape from the pellet.Energy Citations Database (ECD) – Document #4681711. Osti.gov (1971-07-01). Retrieved on 2011-03-17. It has been shown that iodine causes the rate of cracking in pressurised zircaloy-4 tubing to increase.
Graphite moderated reactors In the cases of carbon dioxide cooled graphite moderated reactors such as magnox and AGR power reactors an important corrosion reaction is the reaction of a molecule of carbon dioxide with graphite (carbon) to form two molecules of carbon monoxide. This is one of the processes which limits the working life of this type of reactor.
Water-cooled reactors
Corrosion In a water-cooled reactor the action of radiation on the water (radiolysis) forms hydrogen peroxide and oxygen. These can cause stress corrosion cracking of metal parts which include fuel cladding and other pipework. To mitigate this hydrazine and hydrogen are injected into a BWR or PWR primary cooling circuit as corrosion inhibitors to adjust the redox properties of the system. A review of recent developments on this topic has been published.
Thermal stresses upon quenching In a loss-of-coolant accident (LOCA) it is thought that the surface of the cladding could reach a temperature between 800 and 1400 K, and the cladding will be exposed to steam for some time before water is reintroduced into the reactor to cool the fuel. During this time when the hot cladding is exposed to steam some oxidation of the zirconium will occur to form a zirconium oxide which is more zirconium rich than zirconia. This Zr(O) phase is the α-phase, further oxidation forms zirconia. The longer the cladding is exposed to steam the less ductile it will be. One measure of the ductility is to compress a ring along a diameter (at a constant rate of displacement, in this case 2 mm min−1) until the first crack occurs, then the ring will start to fail. The elongation which occurs between when the maximum force is applied and when the mechanical load is declined to 80% of the load required to induce the first crack is the L0.8 value in mm. The more ductile a sample is the greater this L0.8 value will be.
In one experiment the zirconium is heated in steam to 1473 K, the sample is slowly cooled in steam to 1173 K before being quenched in water. As the heating time at 1473 K is increased the zirconium becomes more brittle and the L0.8 value declines.
Aging of steels Irradiation causes the properties of steels to become poorer, for instance SS316 becomes less ductile and less tough. Also creep and stress corrosion cracking become worse. Papers on this effect continue to be published.
Cracking and overheating of the fuel This is due to the fact that as the fuel expands on heating, the core of the pellet expands more than the rim. Because of the thermal stress thus formed the fuel cracks, the cracks tend to go from the center to the edge in a star shaped pattern. A PhD thesis on the subject has been publishedMicrosoft Word – fuelReport.doc. (PDF) . Retrieved on 2011-03-17. by a student at the Royal Institute of Technology in Stockholm (Sweden).
The cracking of the fuel has an effect on the release of radioactivity from fuel both under accident conditions and also when the spent fuel is used as the final disposal form. The cracking increases the surface area of the fuel which increases the rate at which fission products can leave the fuel.
The temperature of the fuel varies as a function of the distance from the center to the rim. At distance x from the center the temperature (Tx) is described by the equation where ρ is the power density (W m−3) and Kf is the thermal conductivity.
Tx = TRim + ρ (rpellet² – x²) (4 Kf)−1
To explain this for a series of fuel pellets being used with a rim temperature of 200 °C (typical for a BWR) with different diameters and power densities of 250 Wm−3 have been modeled using the above equation. These fuel pellets are rather large; it is normal to use oxide pellets which are about 10 mm in diameter.