Behavior of nuclear fuel during a reactor accident
How nuclear fuel behaves during a reactor accident
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This page describes how uranium dioxide nuclear fuel and its zirconium alloy cladding behave under reactor accident conditions — principally loss-of-coolant accidents (LOCAs) and reactivity-initiated accidents (RIAs). The article also addresses relevant in-reactor behaviour during normal nuclear reactor operation, because phenomena that develop cumulatively over the reactor lifetime — such as cladding hydriding, fuel swelling, and waterside corrosion — are the primary factors that determine how fuel responds when an accident occurs. Research in this field is often expensive to conduct and has therefore frequently been carried out on a collaborative international basis, 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 the cladding can swell during reactor operation. The cladding encases the fuel to form a fuel pin and is subject to mechanical deformation. The gap between the fuel and the cladding is normally filled with helium gas to improve thermal contact. During irradiation, the inventory of gas inside the fuel pin increases as noble gases — principally krypton and xenon — accumulate as fission products. If a Loss-of-coolant accident (LOCA), such as the Three Mile Island accident, or a Reactivity Initiated Accident (RIA), such as the Chernobyl disaster or SL-1, occurs, the temperature of this trapped gas rises. Because the fuel pin is sealed, pressure increases in accordance with the ideal gas law (PV = nRT), and the cladding may deform and burst. Determining the precise moment and mechanism of failure is complex; rigorous burst criteria must simultaneously account for thermal creep, non-isothermal phase transformations, and rapid oxidation.[1]
Both corrosion and irradiation alter the properties of the zirconium alloy commonly used as cladding, rendering it progressively more brittle; experiments conducted on unirradiated zirconium alloy tubes can therefore be misleading. Hydrogen uptake from waterside corrosion further complicates failure mechanics: increasing hydrogen concentration in the cladding systematically lowers the burst stress and burst temperature of Zircaloy-4, and can induce novel multi-location ruptures under LOCA conditions that conventional single-location burst models do not predict.[2] The thermal creep behaviour of cladding is also sensitive to dissolved hydrogen: hydrogen in solid solution accelerates creep rates, while precipitated hydrides impede dislocation motion, producing complex non-monotonic effects on cladding deformation under sustained load.[3]
According to one study,[4] the following difference between the cladding failure modes of unused and used fuel was observed. Unirradiated fuel rods, pressurized and subjected to a simulated RIA transient in a special reactor at the Japanese Nuclear Safety Research Reactor (NSRR), failed late in the transient by ductile ballooning when the cladding temperature was high, producing a burst opening. In contrast, used fuel (irradiated to 61 GW·d/tonne uranium) failed early in the transient through brittle fracture in the form of a longitudinal crack. Complementary tube burst tests confirmed that hydrided zirconium tubing is weaker and bursts at lower internal pressure.[5]
The common failure process of fuel cladding in water-cooled reactors involves a transition to film boiling followed by ignition of the zirconium in steam. The effects of the resulting hot hydrogen reaction-product flow on the fuel pellets and on the fuel bundle wall are illustrated in the accompanying image.
Fuel
Nuclear fuel pellets swell during use as a result of fission gas formation and radiation damage to the crystal lattice. Fission gases accumulate in a central void that develops progressively as burnup increases; as this void forms, the originally cylindrical pellet fractures into pieces. The outward expansion of the pellet causes pellet-cladding interaction, imposing mechanical stresses on the cladding from within. Guidance on the analysis of fuel pin swelling is available in NASA technical reports.[6]
Fission gas release
As fuel is degraded or heated, volatile fission products trapped within the uranium dioxide matrix may be liberated. A report on the release of 85Kr, 106Ru, and 137Cs from uranium in the presence of air has been written; for example, see Colle et al.[7]
In that study, uranium dioxide was found to convert to U3O8 in air over approximately 300–500 °C. The conversion did not begin immediately; during an initial induction period, a U3O7 layer was observed on the uranium dioxide surface, after which the sample gained mass as oxidation proceeded. Approximately 3–8% of krypton-85 was released during the oxidation, while far smaller fractions of ruthenium (0.5%) and caesium (2.6 × 10−3%) were liberated.[8]
Heat transfer between cladding and coolant
In a water-cooled power reactor — or in a water-filled spent fuel pool (SFP) — an understanding of heat transfer from the cladding surface to the surrounding water is important for analyzing reactivity initiated accidents. In a French study, a metal pipe immersed in water under representative PWR and SFP conditions was electrically heated to simulate nuclear heat generation within a fuel pin, with thermocouples monitoring the pipe temperature. Under simulated PWR conditions, water at 280 °C and 15 MPa entered the annular test section (outer pipe 14.2 mm diameter; inner test pipe 9.5 mm outer diameter and 600 mm long) at approximately 4 m·s−1, and the cladding was subjected to heating rates of 2,200–4,900 °C·s−1.
As cladding temperature rose, heat transfer initially increased due to nucleate boiling at the surface. When the heat flux exceeded the critical heat flux, a boiling crisis occurred: the cladding became too hot for sustained nucleate boiling, the surface dried out, and the rate of heat transfer dropped. Subsequent further heating restored boiling, but now in the film boiling regime.[9]
Hydriding and waterside corrosion
As a nuclear fuel assembly accumulates burnup, radiation alters not only the fuel pellets but the cladding material itself. Zirconium reacts chemically with the surrounding water coolant, forming a protective oxide layer on the exterior of the cladding. In PWRs, this oxide typically consumes approximately one-fifth of the cladding wall thickness; the corrosion layer is thinner in BWRs. The governing reaction is:
- Zr + 2 H2O → ZrO2 + 2 H2 (g)
Hydriding occurs when hydrogen produced by this reaction precipitates as hydride phases within the zirconium, embrittling the cladding. These hydride bands form in circumferential rings through the cladding wall. As the growing inventory of fission products exerts hoop stress on the cladding, the hydride bands reduce its capacity to accommodate strain without fracture. This hydrogen absorption fundamentally alters cladding failure mechanics during a LOCA, significantly lowering the burst stress threshold and causing novel multi-location ruptures that can compromise the fuel barrier sooner than conventional single-rupture models anticipate.[10] To overcome the limitations of traditional burst criterion models — which assume symmetric tube deformation and cannot account for azimuthal temperature gradients or the mixed α+β phase region — data-driven approaches using deep neural networks have been applied and have demonstrated substantially lower prediction errors for burst temperature, burst stress, and burst strain compared with analytical burst criteria.[11]
The material limitations imposed by hydriding are among the factors that restrict the achievable burnup of nuclear fuel in a reactor. CRUD (Chalk River Unidentified Deposits), first identified at Chalk River Laboratories, accumulates on the exterior of the cladding as burnup increases.
When a nuclear fuel assembly is prepared for on-site dry storage, it is dried and placed in a spent nuclear fuel shipping cask alongside scores of other assemblies. After removal from the reactor and a cooling period in the spent fuel pool, the hydrides within the cladding reorient from a circumferential arrangement — aligned with the direction of hoop stress — to a radial arrangement pointing outward from the fuel pin. This reorientation leaves the cladding in a fragile state: if a cask were to fall during transport, the weakened cladding could fracture and release spent fuel pellets into the interior of the cask.
Corrosion on the inside of the cladding
Zirconium alloys are susceptible to stress corrosion cracking when exposed to iodine,[12] which is produced as a fission product and, depending on the composition of the fuel, may escape from the pellet into the pellet-cladding gap.[13] Experiments on pressurized zircaloy-4 tubing have confirmed that iodine significantly accelerates the rate of stress-corrosion crack initiation and propagation.[14]
Graphite-moderated reactors
In carbon dioxide-cooled, graphite-moderated reactors such as Magnox and AGR power stations, a significant corrosion reaction occurs between carbon dioxide and graphite, producing two molecules of carbon monoxide per molecule of CO2:
- CO2 + C → 2 CO
This reaction is one of the processes that limits the operational lifetime of such reactors.
Water-cooled reactors
Corrosion
In water-cooled reactors, radiolysis of the coolant by radiation generates hydrogen peroxide and oxygen, which can cause stress corrosion cracking of metallic components including fuel cladding and structural pipework. To mitigate this, hydrazine and hydrogen are injected into the primary cooling circuit of BWRs and PWRs as corrosion inhibitors to adjust the redox chemistry of the system.[15]
Thermal stresses during emergency core cooling
During a loss-of-coolant accident (LOCA), cladding surface temperatures may reach 800–1,400 K. Before emergency coolant water is reintroduced, the hot cladding is exposed to steam, during which zirconium is progressively oxidized. The initial oxidation product is an oxygen-enriched α-phase, Zr(O); continued oxidation produces zirconia (ZrO2). The longer the cladding remains at elevated temperature in steam, the more brittle it becomes.
Ductility is quantified using a ring-compression test in which a cladding ring is compressed diametrically at a constant displacement rate (2 mm·min−1) until the first crack appears. The elongation between the point of peak load and the point at which the load has declined to 80% of the crack-initiation load is recorded as L0.8 (in mm). A larger L0.8 value indicates a more ductile sample.
In experiments in which zirconium is heated in steam to 1,473 K and then slow-cooled to 1,173 K before quenching in water, increasing the hold time at 1,473 K progressively reduces the L0.8 value, confirming the embrittling effect of prolonged steam exposure.[16] Reactor-design-specific predictive models are essential to calculate precise burst temperatures and strains for given heating rates; for example, a dedicated burst criterion has been developed for the fuel cladding used in Indian Pressurized Heavy Water Reactors (PHWRs).[17]
Aging of steels
Irradiation degrades the mechanical properties of structural steels used in reactor components. SS316, for example, becomes less ductile and less tough under irradiation, and its susceptibility to creep and stress corrosion cracking increases.[18]
Cracking and overheating of the fuel
As a fuel pellet heats up, the central region expands more than the rim owing to the radial temperature gradient. The resulting thermal stress causes the pellet to crack in a star-shaped pattern that radiates from the center to the edge. This cracking increases the total surface area of the fuel, which in turn accelerates fission product release — both during accident conditions and during the longer-term storage of spent fuel.
The temperature of the fuel as a function of distance x from the centre of the pellet is:
- Tx = TRim + ρ (rpellet² – x²) / (4 Kf)
where ρ is the power density (W·m−3) and Kf is the thermal conductivity of the fuel. A PhD thesis on this subject has been published by a student at the Royal Institute of Technology in Stockholm.[19]
Model calculations using this equation, assuming a BWR rim temperature of 200 °C and a power density of 250 W·m−3, for pellets of three different diameters are shown below. In practice, oxide fuel pellets are typically about 10 mm in diameter.
Temperature profile for a 20 mm diameter fuel pellet with a power density of 250 W per cubic metre. The central temperature varies substantially between fuel types.
Temperature profile for a 26 mm diameter fuel pellet with a power density of 250 W per cubic metre.
Temperature profile for a 32 mm diameter fuel pellet with a power density of 250 W per cubic metre.
The effect of power density on centreline temperature is shown below for 20 mm pellets at two higher power levels. For all pellet sizes — and most markedly for uranium dioxide, which has a relatively low thermal conductivity — a limit must be set on the power density to prevent centreline melting. The melting point of uranium dioxide is approximately 3,300 K.[20]
Temperature profile for a 20 mm diameter fuel pellet with a power density of 500 W per cubic metre. The calculated centreline temperature for uranium oxide exceeds its melting point.
Temperature profile for a 20 mm diameter fuel pellet with a power density of 1,000 W per cubic metre. Fuel types other than uranium dioxide are not compromised at this scale.
Loss of volatile fission products from pellets
Heating of fuel pellets can drive volatile fission products from the pellet core into the gap between the cladding and the fuel. If xenon leaves the pellet rapidly, the concentrations of 134Cs and 137Cs in the gap increase. Should the zircaloy cladding tube subsequently fail, a greater release of radioactive caesium results. Because the two caesium isotopes arise from different decay chains, they occur at different locations within a fuel pin.
Volatile iodine and xenon isotopes can diffuse out of the pellet on a timescale of minutes and accumulate in the pellet-cladding gap, where xenon may decay to long-lived caesium isotopes before or after reaching the gap.
Genesis of 137Cs
Fission yields calculated for 235U with thermal neutrons (0.0253 eV) using data from the chart of the nuclides.[21]
Genesis of 134Cs
Unlike 137Cs, no 134Cs is produced directly by fission. Its precursor is stable 133Cs, which arises from the decay of longer-lived xenon and iodine isotopes. The 133Cs must then undergo neutron activation in the reactor core to form 134Cs; direct formation via decay of 134Xe is not possible because 134Xe is stable. Because of this distinct formation pathway, the physical distribution of 134Cs within a fuel pin differs from that of 137Cs.
| Element | Isotope | Decay mode | Half-life | Direct fission yield |
|---|---|---|---|---|
| In | 133 | β | 0.18 s | 0.00% |
| Sn | 133 | β | 1.45 s | 0.07% |
| Sb | 133 | β | 2.5 min | 1.11% |
| Te | 133m | β (82.5%) | 55.4 min | 0.49% |
| Te | 133 | β | 12.5 min | 0.15% |
| I | 133 | β | 20.8 h | 1.22% |
| Xe | 133 | β | 5.2 days | 0.00% |
| Cs | 133 | — | stable (undergoes neutron activation in core) | 0.00% |
| Cs | 134 | β | 2.1 years | 6.4 × 10−6% |
Fission yields calculated for 235U with thermal neutrons (0.0253 eV) using data from the chart of the nuclides.[21]
Post-irradiation examination: a representative study
In a representative post-irradiation examination (PIE) study, 20%-enriched uranium dispersed in three different inert matrices was examined to determine the physical location of different fission product isotopes and chemical elements:
- A solid solution of urania in yttria-stabilized zirconia (YSZ), with a Y:Zr atomic ratio of 1:4.
- Urania particles in an inert matrix composed of a mixture of YSZ and spinel (MgAl2O4).
- Urania particles dispersed in an inert matrix of YSZ and alumina.
The three fuels differed in their retention of fission xenon: the YSZ solid solution retained 97% of 133Xe, the YSZ/spinel composite retained 94%, and the YSZ/alumina composite retained only 76%. The long-lived 133Xe can diffuse slowly out of pellets and, if it does so before undergoing neutron activation, is not converted to 134Cs. In contrast, the shorter-lived 137Xe was more strongly retained — 99%, 98%, and 95% in the three fuels respectively. The 137Cs concentration was lowest in the pellet core and highest at the rim, reflecting the outward migration of its xenon precursor; the less volatile 106Ru was distributed more uniformly throughout the pellet.[22]
A subsequent study on urania–YSZ particles dispersed in alumina, irradiated to 105 GW·d·m−3,[23] used SEM imaging of the alumina–fuel particle interface to show that fission products were well confined within the fuel, with little migration into the surrounding alumina matrix. Neodymium was distributed uniformly throughout the fuel; caesium was nearly homogeneous but showed slightly elevated concentrations near xenon bubble sites. Much of the xenon resided in bubbles, while ruthenium was concentrated as nanoparticles that were not consistently colocated with the xenon bubbles.
Release of fission products into coolant: Three Mile Island scenario
At Three Mile Island, a recently SCRAMmed core was starved of cooling water; the resulting decay heat caused the core to dry out and the fuel to be damaged. Attempts were subsequently made to restore core cooling with water. According to the International Atomic Energy Agency, for a 3,000 MW(t) PWR, normal coolant activity levels and the levels expected following two types of accident-induced fuel damage — gap release (release of activity from the pellet-cladding gap) and core melt (significant fuel melting followed by water recovery) — are as follows:[24]
Chernobyl release
The volatility of individual elements strongly governs the fraction of their radioactive inventory that is released during a severe nuclear accident. At Chernobyl, essentially all of the core's xenon and iodine was released, while refractory elements such as zirconium were released in only small fractions. The physical and chemical forms of the release included gases, aerosols, and finely fragmented solid fuel. According to the OECD NEA ten-year assessment of Chernobyl,[25] elements in the form of gases, volatile compounds, or semi-volatile compounds (such as CsI) were released readily, while less volatile elements that form solid solutions with the fuel remained largely inside the reactor. Ruthenium is especially mobile when nuclear fuel is heated in air;[26] in the ionizing radiation environment of spent fuel and in the presence of oxygen, radiolysis can generate volatile ruthenium(VIII) oxide (boiling point approximately 40 °C (104 °F)), which is a powerful oxidiser and reacts with virtually any hydrocarbon present. This mobility has been of concern in PUREX reprocessing, most recently in the context of the airborne radioactivity increase in Europe in autumn 2017.
Some work on TRISO fuel heated in air has also been published.[27]
Table of chemical data
| Element | Gas | Metal | Oxide | Solid solution | Radioisotopes | Release at Chernobyl[25] | T for 10% release from UO2 | T for 10% release from U3O8 |
|---|---|---|---|---|---|---|---|---|
| Br | Yes | — | — | — | — | — | — | — |
| Kr | Yes | — | — | — | 85Kr | 100% | — | — |
| Rb | Yes | — | Yes | — | — | — | — | — |
| Sr | — | — | Yes | Yes | 89Sr, 90Sr | 4–6% | 1,950 K | — |
| Y | — | — | — | Yes | — | 3.5% | — | — |
| Zr | — | — | Yes | Yes | 93Zr, 95Zr | 3.5% | 2,600 K | — |
| Nb | — | — | Yes | — | — | — | — | — |
| Mo | — | Yes | Yes | — | 99Mo | >3.5% | — | 1,200 K |
| Tc | — | Yes | — | — | 99Tc | — | — | 1,300 K |
| Ru | — | Yes | — | — | 103Ru, 106Ru | >3.5% | — | — |
| Rh | — | Yes | — | — | — | — | — | — |
| Pd | — | Yes | — | — | — | — | — | — |
| Ag | — | Yes | — | — | — | — | — | — |
| Cd | — | Yes | — | — | — | — | — | — |
| In | — | Yes | — | — | — | — | — | — |
| Sn | — | Yes | — | — | — | — | — | — |
| Sb | — | Yes | — | — | — | — | — | — |
| Te | Yes | Yes | Yes | Yes | 132Te | 25–60% | 1,400 K | 1,200 K |
| I | Yes | — | — | — | 131I | 50–60% | 1,300 K | 1,100 K |
| Xe | Yes | — | — | — | 133Xe | 100% | 1,450 K | — |
| Cs | Yes | — | Yes | — | 134Cs, 137Cs | 20–40% | 1,300 K | 1,200–1,300 K |
| Ba | — | — | Yes | Yes | 140Ba | 4–6% | 1,850 K | 1,300 K |
| La | — | — | — | Yes | — | 3.5% | 2,300 K | — |
| Ce | — | — | — | Yes | 141Ce, 144Ce | 3.5% | 2,300 K | — |
| Pr | — | — | — | Yes | — | 3.5% | 2,300 K | — |
| Nd | — | — | — | Yes | — | 3.5% | 2,300 K | — |
The releases of fission products and uranium from uranium dioxide (from spent BWR fuel, burnup was 65 GWd t−1) which was heated in a Knudsen cell has been repeated.[7] Fuel was heated in the Knudsen cell both with and without preoxidation in oxygen at c 650 K. It was found even for the noble gases that a high temperature was required to liberate them from the uranium oxide solid. For unoxidized fuel 2300 K was required to release 10% of the uranium while oxidized fuel only requires 1700 K to release 10% of the uranium.
According to the report on Chernobyl used in the above table 3.5% of the following isotopes in the core were released 239Np, 238Pu, 239Pu, 240Pu, 241Pu and 242Cm.
Degradation of the whole fuel element
Water and zirconium can react violently at 1200 °C, at the same temperature the zircaloy cladding can react with uranium dioxide to form zirconium oxide and a uranium/zirconium alloy melt.[29]
PHEBUS
In France a facility exists in which a fuel melting incident can be made to happen under strictly controlled conditions.[30][31] In the PHEBUS research program fuels have been allowed to heat up to temperatures in excess of the normal operating temperatures, the fuel in question is in a special channel which is in a toroidal nuclear reactor. The nuclear reactor is used as a driver core to irradiate the test fuel. While the reactor is cooled as normal by its own cooling system the test fuel has its own cooling system, which is fitted with filters and equipment to study the release of radioactivity from the damaged fuel. Already the release of radioisotopes from fuel under different conditions has been studied. After the fuel has been used in the experiment it is subject to a detailed examination (PIE), In the 2004 annual report from the ITU some results of the PIE on PHEBUS (FPT2) fuel are reported in section 3.6.[32][33]
LOFT
The Loss of Fluid Tests (LOFT) were an early attempt to scope the response of real nuclear fuel to conditions under a loss-of-coolant accident, funded by USNRC. The facility was built at Idaho National Laboratory, and was essentially a scale-model of a commercial PWR. ('Power/volume scaling' was used between the LOFT model, with a 50MWth core, and a commercial plant of 3000MWth).
The original intention (1963–1975) was to study only one or two major (large break) LOCA, since these had been the main concern of US 'rule-making' hearings in the late 1960s and early 1970s. These rules had focussed around a rather stylised large-break accident, and a set of criteria (e.g. for extent of fuel-clad oxidation) set out in 'Appendix K' of 10CFR50 (Code of Federal Regulations). Following the accident at Three Mile Island, detailed modelling of much smaller LOCA became of equal concern.
38 LOFT tests were eventually performed and their scope was broadened to study a wide spectrum of breach sizes. These tests were used to help validate a series of computer codes (such as RELAP-4, RELAP-5 and TRAC) then being developed to calculate the thermal-hydraulics of LOCA. The data was shared with other organizations through international partnerships and has been used to validate other codes such as the French code CATHARE[34] or the South Korean code SPACE.[35]
See also
Contact of molten fuel with water and concrete
Water
Extensive work was done from 1970 to 1990 on the possibility of a steam explosion or FCI when molten 'corium' contacted water. Many experiments suggested quite low conversion of thermal to mechanical energy, whereas the theoretical models available appeared to suggest that much higher efficiencies were possible. A NEA/OECD report was written on the subject in 2000 which states that a steam explosion caused by contact of corium with water has four stages.[36]
- Premixing
- As the jet of corium enters the water, it breaks up into droplets. During this stage the thermal contact between the corium and the water is not good because a vapor film surrounds the droplets of corium and this insulates the two from each other. It is possible for this meta-stable state to quench without an explosion or it can trigger in the next step
- Triggering
- A externally or internally generated trigger (such as a pressure wave) causes a collapse of the vapor film between the corium and the water.
- Propagation
- The local increase in pressure due to the increased heating of the water can generate enhanced heat transfer (usually due to rapid fragmentation of the hot fluid within the colder more volatile one) and a greater pressure wave, this process can be self-sustained. (The mechanics of this stage would then be similar to those in a classical ZND detonation wave).
- Expansion
- This process leads to the whole of the water being suddenly heated to boiling. This causes an increase in pressure (in layman's terms, an explosion), which can result in damage to the plant.
Recent work
Work in Japan in 2003 melted uranium dioxide and zirconium dioxide in a crucible before being added to water. The fragmentation of the fuel which results is reported in the Journal of Nuclear Science and Technology.[37]
Concrete
A review of the subject can be read at [38] and work on the subject continues to this day; in Germany at the FZK some work has been done on the effect of thermite on concrete, this is a simulation of the effect of the molten core of a reactor breaking through the bottom of the pressure vessel into the containment building.[39][40][41]
Lava flows from corium
The corium (molten core) will cool and change to a solid with time. It is thought that the solid is weathering with time. The solid can be described as Fuel Containing Mass, it is a mixture of sand, zirconium and uranium dioxide which had been heated at a very high temperature[42] until it has melted. The chemical nature of this FCM has been the subject of some research.[43] The amount of fuel left in this form within the plant has been considered.[44] A silicone polymer has been used to fix the contamination.
The Chernobyl melt was a silicate melt which did contain inclusions of Zr/U phases, molten steel and high uranium zirconium silicate. The lava flow consists of more than one type of material—a brown lava and a porous ceramic material have been found. The uranium to zirconium for different parts of the solid differs a lot, in the brown lava a uranium rich phase with a U:Zr ratio of 19:3 to about 38:10 is found. The uranium poor phase in the brown lava has a U:Zr ratio of about 1:10.[29] It is possible from the examination of the Zr/U phases to know the thermal history of the mixture. It can be shown that before the explosion that in part of the core the temperature was higher than 2000 °C, while in some areas the temperature was over 2400–2600 °C.
Spent fuel corrosion
Uranium dioxide films
Uranium dioxide films can be deposited by reactive sputtering using an argon and oxygen mixture at a low pressure. This has been used to make a layer of the uranium oxide on a gold surface which was then studied with AC impedance spectroscopy.[45]
Noble metal nanoparticles and hydrogen
According to the work of the corrosion electrochemist Shoesmith[46] the nanoparticles of Mo-Tc-Ru-Pd have a strong effect on the corrosion of uranium dioxide fuel. For instance his work suggests that when the hydrogen (H2) concentration is high (due to the anaerobic corrosion of the steel waste can) the oxidation of hydrogen at the nanoparticles will exert a protective effect on the uranium dioxide. This effect can be thought of as an example of protection by a sacrificial anode where instead of a metal anode reacting and dissolving it is the hydrogen gas which is consumed.