Prospects for fission gas extraction from in-reactor UO2 nuclear fuel elements

Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO₂ as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10⁻⁶ cm³/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed.     

An Apparatus for Continuously Measuring Radon-222 Exhalation from Ground

High burnup MOX and UO₂ test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple and reirradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO₂ test rods reached about 84GWd/tHM and 72GWd/t, respectively. Fuel temperature was measured continuously during the re-irradiation tests. Thermal conductivity change in high burnup fuel was evaluated from the results of comparison between the measured fuel temperature and the data calculated by using the fuel analysis code FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellets is comparable to that of UO₂ fuel pellets in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atoms gradually diminishes with increasing burnup because other factors that affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in a high burnup region.     

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

A model for the release of fission gas from irradiated UO₂ fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO₂ fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m⁻¹, burnups between 10 and 300 MW h(kg U)⁻¹, and power histories including constant, high-to-low and low-to-high power periods.The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO₂, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.     

The release of fission products from nuclear fuel during irradiation by both lattice and grain boundary diffusion

An analysis has been made to describe the steady state release of radioactive fission products from nuclear fuel during irradiation, by both lattice and grain-boundary diffusion. The analysis is appropriate for calculating the release of volatile fission products from regions of fuel pins prior to the development of gross interlinked porosity. In general terms, the dependence of fractional release on decay constant, temperature and solute atom-grain boundary interaction energy has been investigated. The analysis is applied to a specific irradiation experiment where the release of the rare gases, ¹³³I, ¹³¹I, and ¹³²Te were measured from both small and very large grain UO₂ samples.     

Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO₂ fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO₂-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO₂), composite (UO₂-10%volBeO), silicide (U₃Si₂) and mixed oxide ((Th_0.9,U_0.1)O₂) fuel types under normal operation conditions. Compared to UO₂, it was found that the UO₂-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U₃Si₂ fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th_0.9,U_0.1)O₂ fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO₂ fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.     

In-pile Xe diffusion coefficient in UO2 determined from the modeling of intragranular bubble growth and destruction under irradiation

Intragranular bubbles grow in the nuclear fuel by diffusion and precipitation of fission gases, mainly xenon; and are ultimately destroyed, under irradiation, by fission fragments. This article will attempt to determine the in-pile bubble distributions taking into account the evolution of the concentration profile around a bubble during its growth and the destruction process by fission fragments. From these distributions a relation between the bubble mean radius and the diffusion coefficient of xenon can be established, allowing the determination, from experimental measurements of intragranular bubble sizes, of the in-pile Xe diffusion coefficient in UO₂. The estimated activation energy (0.9 eV) is about one order of magnitude lower than the widely used value of 3.9 eV determined from out-of-pile experiments. This effect can be attributed to the presence of point defects created by the irradiation.     

Modeling of fission gas swelling in LWR UO2 fuel under irradiation

An understanding of the behavior of fission gas in uranium dioxide (UO₂) fuel is necessary for the prediction of the performance of fuel rods under irradiation. A mechanistic model for matrix swelling by the fission gas in LWR UO₂ fuel is presented. The model takes into account intragranular and intergranular fission gas bubbles behavior as a function of irradiation time, temperature, fission rate and burn-up. The intragranular bubbles are assumed to be nucleated along the track of fission fragments, which play the dual role of creator and destroyer of intragranular bubbles. The intergranular bubble nuclei is produced until such time that a gas atom is more likely to be captured by an existing nucleus than to meet another gas atom and form a new nucleus. The capability of this model was validated by a comparison with the measured data of fission gas behavior such as intragranular bubble size, bubble density and total fuel swelling. It was found that the calculated intragranular bubble size and density are in reasonable agreement with the measured results in a broad range of average fuel burn-ups 6–83 GW d/tU. Especially, the model correctly predicts the fuel swelling up to a burn-up of about 70 GW d/tU.     

Compression and Decompression Waves in Steam-Water System

The ratio of the γactivities per fission from fission products of ²³⁹Pu and ²³⁵U, and its time dependence were measured by double fission chamber technique. The γ-activity from the fission products of ²³⁹Pu fission was lower than the corresponding activity relevant to ²³⁵U fissions. The ratio varied with the cooling time allowed after irradition.

This ratio was applied to power distribution measurements by γ-scanning method in multi-region cores composed of PuO₂-UO₂ and UO₂ fuels. To obtain the relative power, the measured γ-activities from the fission products in the fuel rods were corrected for the difference between the γ-activities per fission from the fission products.     

Effect of helium on the microstructure of spent fuel in a repository: An operational approach

In a repository, the release of radionuclides from spent fuel rods will strongly depend on the pellet microstructure existing when water comes into contact with the spent fuel surface, i.e. after 10,000 years of disposal. During this period, a large quantity of He atoms is produced by α-disintegrations of actinides in the spent fuel. A conservative model is proposed here to evaluate the consequences of He on the spent fuel microstructure. According to the solubility and diffusion properties of He under repository conditions, two scenarios are considered: He atoms can be trapped in fission gas bubbles or form new bubbles. In spite of the conservative assumptions of the model, the calculated values of bubble or pore pressure are much lower than critical values derived from rupture criteria. No evolution of the microstructure of the spent UO₂ fuel is thus expected before the breaching of the canister.     

Finite element analysis of local overheating within plutonium enriched UO2 fuel rods caused by PuO2 islands

Within natural UO₂ fuel elements enriched with plutonium, this last material should form PuO₂ solid solutions inside the UO₂ pellets, in a wide range of concentrations. If the solutions are obtained by mechanical mixing of the oxides, PuO₂ islands are formed in the UO₂ matrix. These islands may be the source of several problems in the fuel behaviour, the most important being the overheating of the matrix in the neighbourhood of the particles. It is caused by the large fission cross section of plutonium compared with that of uranium.A detailed study of the thermal effects produced by PuO₂ particles in the UO₂ matrix and the cladding is then important for the specification of their permissible size. A portion of the fuel rods with spherical particles in the most significant places was studied. In order to obtain the dimensionless overheating of the fuel and cladding produced by the presence of those particles, the spacial distribution of temperature was calculated, solving the stationary and linear bidimensional equation of heat conducting using a finite element code. Several geometrical variables and material properties have been taken as dimensionless parameters. A satisfactory convergence of the numerical results to an asymptotic limit with a wellknown exact solution, has been obtained.     

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.     

On the role of grain boundary diffusion in fission gas release

It is generally believed that thermal fission gas release from LWR fuel occurs mainly via interconnected grain boundary bubbles. Grain boundary diffusion is not considered to be a significant mechanism. We investigated this supposition by two methods; first, by assessing the distance a gas atom can migrate in a grain boundary containing perfectly absorbing traps. For areal number densities and fractional coverages by the traps observed in fuel irradiated to burnups exceeding ∼20 MWd/kg, gas atoms will be trapped after a migration distance equal to the size of a grain or less. This supports the supposition for medium-to-high burnups. However, the above-mentioned model is inapplicable for trace-irradiated specimens. In our second analysis, we examined Xe release from trace-irradiated UO₂. The measurements indicated that the liberation involves more than only lattice diffusion at the specimen surface, and that the data are consistent with sequential lattice and grain boundary diffusion unimpeded by intergranular traps. The analysis also provided rough estimates of the grain boundary diffusion coefficient in UO₂.     

Advanced Doped UO2 Pellets in LWR Applications

The nuclear industry strives to reduce the fuel cycle cost, enhance flexibility and improve the reliability of operation. This can be done by both increasing the fuel weight and optimizing rod internal properties that affect operational margins. Further, there is focus on reducing the consequences of fuel failures. To meet these demands Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO₂ fuel containing additions of chromium and aluminium oxides. This paper presents results from the extensive investigation program which covered examinations of doped and reference standard pellets both in the manufactured and irradiated states.

The additives facilitate pellet densification during sintering and enlarge the pellet grain size. The final manufactured doped pellets reach about 0.5% higher density within a shorter sintering time and a five fold larger grain size compared with standard UO₂ fuel pellets. The physical properties of the pellets, including heat capacity, thermal expansion coefficient, melting temperature, thermal diffusivity, have been investigated and differences between the doped and standard UO₂ pellets are small.

The in-reactor performance of the ADOPT pellets has been investigated in pool-side and hotcell Post Irradiation Examinations (PIEs), as well as in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced fission gas release, improved PCI performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. Fuel segments have been exposed to ramp tests and enhanced power steady-state operation in the Studsvik R2 reactor after base-irradiation to above 30 MWd/kgU in a commercial BWR. ADOPT reveals up to 50% lower fission gas release than standard UO₂ pellets. The fuel degradation behaviour has been studied in two oxidizing tests, a thermal-microbalance test and an erosion test under irradiation. The tests show that ADOPT pellets have a reduced rate of fuel washout, as compared to standard UO₂ pellets.

Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA-96 Optima2 reloads in 2005.     

Metal fission product behaviour in high temperature reactors -UO2 coated particle fuel

The migrational behaviour of a number of metallic fission products in the coated UO₂ particle fuel, proposed for HighTemperature Reactors, is described. The derivation of parameters enabling calculations to be made of the release of selected important isotopes is also discussed, particularly with reference to ‘Triso’ coated particles in which the silicon carbide layer is either defective or absent.     

Experimental and thermodynamic evaluation of the melting behavior of irradiated oxide fuels

Onset of melting is an important performance limit for irradiated UO₂ and UO₂-based nuclear reactor fuels. Melting (solidus) temperatures are reasonably well known for starting fuel materials such as UO₂ and (U,Pu)O₂, however the influence of burnup on oxide fuel melting behavior continues to represent an area of considerable uncertainty. In this paper we report the results of a variety of melting temperature measurements on pseudo-binary fuel-fissia mixtures such as UO₂-PuO₂, UO₂-CeO₂, UO₂-BaO, UO₂-SrO, UO₂-BaZrO₃, and UO₂-SrZrO₃. These measurements were performed using the thermal arrest technique on tungsten-encapsulated specimens. Several low melting eutectics, the existence of which had previously been inferred from post-irradiation examinations of high burnup mixed oxide fuels, were characterized in the course of the investigation. Also, an assessment of melting temperature changes in irradiated oxide fuels due to the production and incorporation of soluble oxidic fission products was performed by application of solution theory to the available pseudo-binary phase diagram data. The results of this assessment suggest that depression of oxide fuel solidus temperatures by dissolved fission products is substantially less than that indicated by earlier experimental studies.     

Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

A model for the release of stable fission gases by diffusion from sintered LWR UO₂ fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model.     

Analysis of VIP-BWR reactor core physics experiments on UO2 and MOX mockup fuel assemblies - part 2: analysis with SIMULATE5

Analysis of the three test cores in the VIP-BWR program was performed in a two-dimensional geometrical model with CASMO5 coupled with the JENDL-4.0-based neutron data library, and reported in the previous paper. Following the study, interpretation of the experiments were carried out in a three-dimensional geometrical model with SIMULATE5 for the code validation study. The nuclear libraries for the SIMULATE5 calculations were generated with CASMO5 with the JENDL-4.0-based neutron data library. The effective multiplication factors of the critical cores ranged from 0.9983 to 1.0023 with measurement uncertainties of 0.0003 to 0.0004 (one σ). The root mean squares of (the calculated/the measured-1) for the fission rates at the core-mid plain of all the measured fuel rods were about 3% for the three cores. It was noticed that the calculations underestimated the fission rates of the UO₂ fuel rods and overestimated those of the MOX fuel rods for the test cores loaded with MOX fuel rods, which was consistent with trends in the preceding analysis studies of the VIP-BWR program and other MOX core experiments, and the biases were confirmed in the calculation results of power distributions in MOX-fueled light water reactor cores.     

Development of fuel-model interfaces: Investigations by XPS, TEM, SEM and AFM

The presented work aims to reproducibly prepare UO₂-Pd thin film model systems for spent nuclear fuel in order to further investigate surface reactions of these films under relevant redox conditions. The sputter co-deposition of U and Pd (fission product) in the presence of O₂ results in the homogenous distribution of Pd in a crystalline UO₂ matrix. Heating the films causes the diffusion of film components. Hereby, the formation of ε-particles has to be clarified. First electrochemical studies show the influence of the nobel metal Pd on the redox behaviour of UO₂. With increasing Pd concentration the matrix dissolution is decreased. However, we could demonstrate that blocked oxidation processes are of temporary nature. The passivation of the Pd reactive sites with increasing number of cycles finally induces the approximation of the mixed system to the redox behaviour of the pure UO₂ system.     

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO₂ or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.