Diffusion Coefficient of Fission Gas in Uranium Dioxide Powder Formed by Carbon Dioxide Oxidation of Uranium

In connection with a program to study the behavior of punctured fuel elements for the Tokai Atomic Power Reactor, the diffusion coefficient of fission gas in uranium oxide powder formed by CO₂ oxidation of U was determined by post-irradiation experiment, in which the fractional release of fission gas during isothermal heating of the powder was measured. The U was oxidized at 600° and 700°C, and in both cases the O/U ratio of the oxides, measured gravimetrically, was 2.0. The diffussion coefficients in the oxide powder formed by oxidation at 600°C were found to be 1.4× 10-²⁰, 1.3×10-¹⁹, 1.1×10¹⁸ and 1.0×10-¹⁷, cm²sec-¹, respectively at 450°, 550°, 650° and 750°C, and in the oxide powder formed at 700°C, 7.4×10-¹⁹ and 3.6×10-¹⁶cm²sec-¹ at 600° and 700°C, respectively. Activation energies calculated for the two oxide powders were comparatively low.     

The radial distribution of fission gases and other fission products in irradiated PWR fuel

The radial distribution of fission gas (xenon) and other fission products (cesium, ruthenium, cerium) has been measured on UO₂ fuel pellets irradiated in commercial pressurized water reactors to burnups between 13.23 and 48.26 GWd/tU. Fission gas release occurs from the pellet center, and at temperatures $\text{< 1300° C}$ is confined to the region of grain growth. The maximum fractional release measured at the center ranges from 20% to 30%. Only at high burnup (48.26 GWd/tU) an additional release of cesium has been observed. This is considered as evidence for an increase in fission product release at higher burnups. At fuel center line temperature $\text{> 1500° C}$ a high fission gas release is accompanied by a high cesium release. The local release starts at the onset of fission gas bubbles precipitating on grain boundaries and saturates in the center of the pellet at a fractional release value of about 90%.     

Studies of Ceramic Fuels with Use of Fission Gas Release Loop, (IV)

In-pile release of fission gas from sintered UC pellets in the presence of 8–230 ppm of water vapor in the He sweep gas was measured over the temperature range of 160°–1,000°C. A very complex release behavior was observed and the mechanisms of release were deduced from the manner in which the release depended on the decay constant. It was established that the release of short-lived fission gases during irradiation was controlled mainly by pseudo-recoil, while chemical reaction between UC and water vapor, as well as knock-out, appeared to contribute much more significantly in the case of the longer-lived fission gases. The release of fission gas after reactor shutdown was shown to be governed by the UC-H₂O reaction. The ratio of the release due to this reaction in reference to the total release was found to be dependent not only on the concentration of the water vapor but also on the amount present of the accumulated reaction products. Also, a discussion is given on the inordinately high release of ^135mXe observed at 600°C.     

Fission Gas Release from Nuclear Fuels, (II)

The release behavior of fission gases in U-metal, UO₂ and uranium carbides, irradiated at a relatively low temperature (below 100°C) to low dosage, was studied by out-of-pile experiments.

It was found that fission gas (¹³³Xe) released from a specimen by fission fragment recoil is mostly captured in the wall of the irradiating capsule or in the capsule support material.

The amount of fission gas released into the void space of the capsule is proportional to the surface area and to the fuel burn-up, and is controlled by a knock-out release mechanism. The number of U atoms considered to take part in the knock-out mechanism by evaporation or displacement due to the intrusion of a recoil fission fragment, is estimated to be 1.4×10⁵~2.7×10⁵ atoms for U-metal and 5×10⁴~10×10⁴ atoms for UO₂ and uranium carbides.     

Effects of Void Migration Process on Fission-Gas Release

The release rate of fission-gas from U0₂ was continuously measured during irradiation in the Hitachi Training Reactor. The U0₂ specimen was heated electrically in in-core assemblies by tungsten heaters, either arranged axially transversing the specimen (producing radial temperature gradient) or cylindrically outside the specimen (uniform heating). In a case of the axially heated annular U0₂ pellets with radial temperature gradient, the rate of fission-gas release increased with time under constant temperature and flux. This experimental result was explained by considering as fission gas release mechanism a combination of diffusion and pore migration processes. On the other hand, when U0₂ pellets were heated isothermally up to 2,000°C, the fission-gas release rate could be explained in terms of diffusion process alone. No effect of equiaxial grain growth was observed on the rate of fission-gas release.     

Preliminary Study on Production of Xenon-133 from Neutron-Irradiated Uranium Metal and Oxides by Oxidation

A method was developed for recovering the fission product ¹³³Xe from several kinds of reactor-irradiated U targets, including Al-U alloy, metallic U, and uranium oxides.

In order to observe the release of ¹³³Xe from U compounds at high temperatures, irradiated targets were heated at 500°~1,000°C in vacuum or under gas flow. The liberated ¹³³Xe was trapped on charcoal beds, and the release rate of ¹³³Xe from the compounds was determined by measuring the activity accumulating in the traps.

More than 90% of the ¹³³Xe was liberated from the alloy upon melting and from metal and uranium oxide upon oxidation.

The isolated Xe was purified by a system embodying cold traps and cryogenic distillation.

The final products were sealed in ampoules. They proved to possess radiochemical purity exceeding 99.9%, and less than 1 μ/ampoule (1 ml) of non-radioactive gaseous contaminants.     

Oxidation of Uranium by Partially Decomposed Carbon Dioxide

The oxidation rate of U rods by partially decomposed CO₂ was measured thermogravimetrically and compared with that by undecomposed CO₂, in an attempt to study the effect of ionizing radiation on the compatibility of U with CO₂. The decomposition was by silent discharge. Measurement was performed for cases of 0.29%, 0.44% and 0.68% decomposed CO₂, at 500° and 600°C. It was found that the oxidation rate rises almost linearly with increasing percentage of decomposition of the CO₂. At 500°C the oxidation rate with 0.68% decomposed CO₂ was eighteen times higher than with undecomposed CO₂, and five times higher at 600°C. Such increase was related to O₂ content in the decomposed gas.     

Oxidation and Gas Release Behavior of UC,U(C,N) and UN at High Temperature

A study was made of the oxidizing behavior at high temperature (800°–1,800°C) in vacuum of UC, UN and U(C,N) samples containing added oxygen in excess amounts, through observations of gas release, X-ray diffraction analysis, and microphotography.

The oxidation in vacuum of UC and U(C,N) was found to proceed above 1,200°C by stepwise reactions from one temperature interval to the next, the process differing however according to the chemical state of the oxygen present in the samples. In the temperature range below 1,200° C, the UC and U(C,N) samples reacted violently with the free oxygen present in dissolved state, to form UO₂. Between 1,200° and 1,400° C the UO₂ thus produced reacted with the UC or U(C,N), forming solid solutions of U(C,O) and U(C,N,O) respectively: Above 1,600°C, these solid solutions gradually decomposed back into UC and U(C,N), and U. In all stages of oxidation, large amounts of CO—and N₂ in the case of U(C,N)—evolved from the samples as reaction products. In the case of UN, no reaction was observed below 1,200°C, and only oxidized above that temperature to form UO₂ and N₂ by the action of the dissolved oxygen present.

These results indicate that in the case of UC and U(C,N), the quantity of gases evolving from the oxidation is dictated by the total amount of oxygen contained in the samples, while that from UN is dependent on the amount of molecular oxygen alone.     

The mechanisms of in-pile fission gas release from UO2

In-pile release mechanisms of fission gas from UO₂ at low temperatures were studied. The release of ¹³³Xe, ¹³⁵Xe, ¹³⁸Xe, ^85mKr, ⁸⁸Kr and ⁸⁷Kr from a sintered UO₂ pellet was measured at temperatures ranging from 250 to 930°C using a graphite specimen holder. The release from the holder, in which a fraction of fission gas was recoil-implanted, was subtracted to obtain the net release from the UO₂ pellet. Knock-out release from the UO₂ was measured directly, and it was found that it was not the main release mechanism, at least not for short-lived nuclides. A ‘pseudo-recoil’ release model is proposed to explain the low temperature release under irradiation. In the model, some of the defects produced by fission fragments act as short-lived carriers for fission gas.     

The fission gas bubble distribution in a mixed oxide fast reactor fuel pin

The fission gas bubble distribution has been studied in a mixed oxide fast reactor fuel pin irradiated in DIDO MTR to 2.8% burn-up at centre and surface temperatures of 2000 and 1000°C. The intragranular fission gas bubbles are very small (<6 nm diameter) and this is a consequence of the high re-solution rate at fast reactor ratings. The bubbles nucleate heterogeneously and linear arrays of bubbles, due to nucleation on fission tracks, are observed up to irradiation temperatures of 1900°C. At 1980°C ~4% of the fission gas produced is present in intragranular bubbles. There is no definite evidence for gas bubble mobility or coalescence. Apart from any effects of columnar grain growth fission gas release in fast reactor fuel pins seems to occur predominantly by the diffusion of single gas atoms, at least up to irradiation temperatures of 2000°C.     

Studies of Ceramic Fuels with the Use of Fission Gas Release Loop, (I)

Fission gas release from a UO²-graphite mixture was studied during irradiation with the use of the Fission Gas Release Loop in the JRR-3 reactor. The release rates of fission krypton and xenon increased proportionally with neutron flux (6×10¹⁰–6×10¹² n/cm²·sec) and exponentially with temperature (400°–1,000°C). A burst of fission gas was observed when the specimen was abruptly heated to a higher temperature. These results can be explained by a mechanism whereby fission gas is trapped in defects created in graphite by fission fragments and released through annealing of the defects.     

Fission xenon release from lightly irradiated (Th,U)O2 powders

The release of ¹³³Xe from (Th,U)O₂ was studied at a low fission density by using a post-irradiation technique. The uranium concentrations of the specimens ranged from 0.15 to 20 mol% U0₂. Heating curves of the release gave almost the same pattern in shape, while the total release, which was the combination of an in-pile release and a post-irradiation release up to 1000°C, increased with uranium concentration except for the nominally pure ThO₂. Effects of preparation conditions of specimens such as atmosphere, temperature and stoichiometry were also studied and found to be minor. Possible release mechanisms were discussed.     

Effects of Annealing under High Static Pressure on Precipitation Behavior of Fission Gas Bubbles in Irradiated UO2

Annealing experiments were carried out on irradiated UO₂ in argon gas under high pressure (600 and 1,000 kg/cm²) as well as atmospheric, at temperatures of 1,400°–1,600°C. The effects of high external pressure on the behavior of fission gas bubbles in the irradiated UO₂ were studied by comparing replica electron micrographs of fractured surfaces of specimens annealed under different temperatures and pressures. The results indicate that high pressures such as above 600 kg/cm² can be effective in surpressing the growth of fission gas bubbles in both intergranular and intragranular zones, and in inhibiting the joining together of intergranular bubbles to form direct passages for fission gas release.     

The migration of intragranular fission gas bubbles in irradiated uranium dioxide

The mobility of intragranular fission gas bubbles in uranium dioxide, irradiated at 1600–1800°C, has been studied following isothermal annealing at temperatures below 1600°C. The intragranular fission gas bubbles, average diameter approximately 2 nm, are virtually immobile at temperatures below 1500°C. The bubbles have clean surfaces with no solid fission product contamination and are faceted to the highest observed irradiation temperature of 1800°C. This bubble faceting is believed to be a major cause of bubble immobility. In fuel operating below 1500°C the predominant mechanism allowing the growth of intergranular bubbles and the subsequent gas release must be the diffusion of dissolved gas atoms rather than the movement of entire intragranular bubbles.     

Investigation of swelling of U-Pu mixed carbide

Mixed carbide fuel samples irradiated in various types of capsules were investigated with respect to fuel swelling and fission gas behaviour. The irradiations were carried out in the FR 2 reactor in Karlsruhe at temperatures between 300 and 1750°C up to 5.5% burnup. The swelling was evaluated by immersion density measurements in carbon tetrachloride. The fission gas determinations were carried out by measuring the released gas and by measuring the retained fission gas.The swelling rate of mixed carbide is a strong function of temperature. At temperatures below 1000°C it is between 1 and 1.5% per % burnup. At temperatures above 1000°C the swelling rate increases with temperature. It is about 3% per % burnup at 1300°C and about 12% per % burnup at 1750°C. The swelling rate at high temperatures decreases with increasing burnup due to a saturation of the fission gas bubble porosity.     

Recoil release of fission products from nuclear fuel

An analytical approximation is developed for calculating recoil release from nuclear fuel into gas filled interspaces. This expression is evaluated for a number of interspace geometries and shown to be generally accurate to within about 10% by comparison with numerical calculations. The results are applied to situations of physical interest and it is demonstrated that recoil can be important when modelling fission product release from low temperature CAGR pin failures. Furthermore, recoil can contribute significantly in experiments on low temperature fission product release, particularly where oxidation enhancement of this release is measured by exposing the fuel to CO₂. The calculations presented here are one way of allowing for this, other methods are suggested.     

The mechanisms of fission gas release from (Th,U)O2

The releases of xenon from three (Th, U)O₂ specimens with different U contents were measured over a wide range of fission dose from $\text{2.9 × 10}{}^{19}$ to $\text{2.2 × 10}{}^{22}$ fissions m^?3 by using a post-irradiation technique. The releases were found to decrease with dose and to level off at higher doses. Measurements of the changes in lattice parameter and specific surface area of the same specimens enabled one to conclude that the decrease in release originates in the trapping of xenon by the vacancies and vacancy clusters induced by fission fragments. And the release mechanisms of fission gas were proposed based on the proper evaluation of the observation on radiation damage and recovery in oxide fuel.     

Post-irradiation studies on knock-out and pseudo-recoil releases of fission products from fissioning UO2

By using post-irradiation techniques, in-pile releases of ¹³³Xe, ^85mKr, ⁸⁸Kr, ⁸⁷Kr and ¹³⁸Xe from UO₂ fissioning at low temperatures below about 200° C are studied: these are analyzed into a time-dependent knock-out and time-independent pseudo-recoil releases. For the latter, a “self knock-out” mechanism is proposed: when a fission fragment loses thoroughly its energy near the UO₂ surface and stops there, it will knock out the surface substances and accordingly the fragment (i.e. the fission product) will be released. The effective thickness of the layer where the self knock-out occurs is found to be ~7Å. As for the knock-out release, the following is estimated from its dependence on various factors: the knock-out release of fission products occurs from the surface layer with the effective thickness of ~20Å: the shape of UO₂ matrix knocked out by one fission fragment passing through the surface is equivalent to a cylinder ~32Å diameter by ~27Å thick, (i.e. the knock-out coefficient for UO₂ is ~660 uranium atoms per knock-out event). On the basis of the above estimations, the conclusions derived from the past in-pile studies of fission gas releases are evaluated.     

Artificial neural network modeling for fission gas release in LWR UO2 fuel under RIA conditions

A fission gas release (FGR) model was developed by using an artificial neural network method to predict fission gas release in UO₂ fuel under reactivity initiated accident (RIA) conditions. Based on the test data obtained in the CABRI test reactor and nuclear safety research reactor, the model takes into account the effect of the five parameters: pellet average burnup, peak fuel enthalpy, the ratio of peak fuel enthalpy to pulse width, fission gas release during base-irradiation, and grain size of a fuel pellet. The parametric study of the model, producing a physically reasonable trend of FGR for each parameter, shows that the pellet average burnup and the ratio of peak fuel enthalpy to pulse width are two of the most important parameters. Depending on the combination of input values for the five parameters, the application of the model to a fuel rod under typical RIA conditions of light water reactor produces 1.7-14.0% of FGR for the pellet average burnup ranging from 20 to 70 MW d/kg U.     

Fission product release in high-burn-up UO2 oxidized to U3O8

Results of oxidation experiments on high-burn-up UO₂ are presented where fission-product vaporisation and release rates have been measured by on-line mass spectrometry as a function of time/temperature during thermal annealing treatments in a Knudsen cell under controlled oxygen atmosphere. Fractional release curves of fission gas and other less volatile fission products in the temperature range 800-2000 K were obtained from BWR fuel samples of 65 GWd t⁻¹ burn-up and oxidized to U₃O₈ at low temperature. The diffusion enthalpy of gaseous fission products and helium in different structures of U₃O₈ was determined.