UO2 Fuels Research Articles

Phase-field simulations of intragranular fission gas bubble evolution in UO2 under post-irradiation thermal annealing

Fission gas bubbles are one of the evolving microstructures that affect thermal mechanical properties, such as thermal conductivity, gas release, volume swelling, and cracking, in operating nuclear fuels. Therefore, fundamental understanding of gas bubble evolution kinetics is essential to predict the thermodynamic property and performance changes of fuels. In this work, a generic phase-field model was developed to describe the evolution kinetics of intragranular fission gas bubbles in UO2 fuels under post-irradiation thermal annealing conditions. Free energy functional and model parameters are evaluated from atomistic simulations and experiments. Critical nucleus size of gas bubbles and gas bubble evolution were simulated. A linear relationship between logarithmic bubble number density and logarithmic mean bubble diameter was predicted, which is in good agreement with experimental data.

Study of Plutonium Recycling Scheme with Uranium Multi-recycling System

We have considered a fuel cycle scheme for multi-recycling of plutonium recovered from spent UO2 fuels of LWRs. Plutonium is to be partially loaded into the next LWRs in the form of MOX fuels with UO2 fuels made by re- enrichment of uranium recovered from spent fuels of LWRs. We consider to recycle plutonium in combination with uranium multi-recycling system previously reported, which can provide uranium fuels for not only fully UO2 fueled PWRs but also partially MOX fueled PWRs. The results of the process system analysis suggest that the proposed scheme makes it possible to recycle plutonium in succession.

Effects of matrix composition on instant release fractions from high burn-up nuclear fuel

ABSTRACTThe release of radionuclides from spent nuclear fuel in contact with water is controlled by two processes – the dissolution of the UO2grains and the rapid release of fission products segregated either to the gap between the fuel and the cladding or to the UO2grain boundaries. The rapid release is often referred to as the Instant Release Fraction (IRF) and is of interest for the safety assessment of geological repositories for spent fuel due to the potential dose contribution.Previous studies have shown that the instant release fraction can be correlated to the fission gas release (FGR) from the spent fuel. Studies comparing results from samples in the form of pellets, fragments, powders and a fuel rodlet have shown that the sample preparation has a significant impact on the instant release, indicating that the differentiation between gap release and grain boundary release should be further explored.Today, there are trends towards power uprates, longer fuel cycles and increasing burn-up putting additional requirements on the nuclear fuel. These requirements are met by the development of new fuel types, such as UO2fuels containing dopants or additives. The additives and dopants affect fuel properties such as grain size and fission gas release. In the present study we have performed experimental leaching studies using two high burnup fuels with and without additives/dopants and compared the fuel types with respect to their instant release behavior. The results of the leaching of the samples for the 3 initial contact periods; 1, 7 and 23 days are reported here.

Lattice Parameter Changes for Spent UO2 Fuels With and Without Gd

Lattice Parameter Changes for Spent UO2 Fuels With and Without Gd

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

The temperature measurements of mixed oxide (MOX) and UO2 fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO2 fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel.

A high burnup model developed for the DIONISIO code

A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO2 fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in 235U.In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2 fuels in the MALIBU program

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO2 assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 × 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 × 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 × 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO2 sample of 72 GWd/t.

Impact assessment of upscattering on resonance calculation using improved ultrafine energy group method

Thermal upscattering has considerable impact on the neutronic analysis of light water reactors. This phenomenon is more significant for reactors with advanced fuels having high plutonium content: Mixed Oxide (MOX) fuels. In this paper, a sensitivity analysis for the impact of upscattering boundary energy on the integral results of MOX and UO2 fuels in light water reactors is performed using WIMSD-5B code. For investigating the impact on the effective cross sections, a spectrum calculation code Microtrans is developed which employs an improved ultrafine group method. Moreover, the impact on the effective cross sections of important resonances of fuel nuclides in thermal and epithermal energy ranges is studied in detail. Calculation results show that changing the upsacttering boundary from 0.625eV to 4eV leads to more than 1000pcm decrease in Keff in MOX fuels due to large resonances of Pu isotopes in thermal energy range, while this impact is about 120pcm for UO2 fuels. In addition, thermal upscattering has significant impact on the effective cross sections below 10eV.

Mechanical properties and XRD studies of silicon carbide inert matrix fuel fabricated by a low temperature polymer precursor route

The mechanical properties of silicon carbide (SiC) inert matrix fuel (IMF) pellets fabricated by a low temperature (1050°C) polymer precursor route were evaluated at room temperature. The Vickers hardness was mainly related to the chemical bonding strength between the amorphous SiC phase and the β-SiC particles. The biaxial fracture strength with pre-notch and fracture toughness were found to be mostly controlled by the pellet density. The maximum Vickers hardness, biaxial fracture strength with pre-notch and fracture toughness achieved were 5.6GPa, 201MPa and 2.9MPam1/2 respectively. These values appear to be superior to the reference MOX or UO2 fuels. Excellent thermal shock resistance for the fabricated SiC IMF was proven and the values were compared to conventional UO2 pellets. XRD studies showed that ceria (PuO2 surrogate) chemically reacted with the polymer precursor during sintering, forming cerium oxysilicate. Whether PuO2 will chemically react in a similar manner remains unclear.

In-situ X-ray diffraction study of phase transformations in the Am–O system

In the frame of minor actinides recycling, americium can be transmuted by adding it in UO2 or (U, Pu)O2 fuels. Americium oxides exhibiting a higher oxygen potential than U or Pu oxides, its addition alters the fuel properties. To comprehend its influence, a thorough knowledge of the Am–O phase equilibria diagram and of thermal expansion behavior is of main interest. Due to americium scarcity and high radiotoxicity, few experimental reports on this topic are available. Here we present in-situ high-temperature XRD results on the reduction from AmO2 to Am2O3. We show that fluorite (Fm-3m) AmO2 is reduced to cubic (Ia-3) C′-type Am2O3+δ, and then into hexagonal (P63/mmc) A-type Am2O3, which remains stable up to 1840K. We also demonstrate the transitional existence of the monoclinic (C2/m) B-type Am2O3. At last, we describe, for the first time, the thermal expansion behavior of the hexagonal Am2O3 between room temperature and 1840K.

Research at the CEA in the field of safety in 2nd and 3rd generation light water reactors

Abstract The research programs at the CEA in the field of safety in nuclear reactors are carried out in a framework of international partnerships. Their purpose is to develop studies on: – The methods allowing for the determination of earthquake hazards and their consequences; – The behaviour of fuel in an accident situation; – The comprehension of deflagration and detonation phenomena of hydrogen and the search for effective prevention methods involving an explosion risk; – The cooling of corium in order to stop its progression in and outside the vessel thereby reducing the risk of perforating the basemat; – The behaviour of the different fission product families according to their volatility for the UO2 and MOX fuels.

Dopant solubility and lattice contraction in gadolinia and gadolinia–chromia doped UO2 fuels

Gadolinia doped UO2 fuel is widely used as burnable neutron absorber in Light Water Reactors to reduce power peaking and excess reactivity during the first reactor cycle of fresh fuel assemblies. The thermal conductivity of gadolinia doped fuel is substantially lower than that of standard UO2. To maintain safety margins later in life, some design or operating restrictions can be defined, for example to compensate higher fission gas release levels. Development of large grain U/Gd fuel by suitable doping, e.g. Cr2O3, could offer a solution to such restrictions, but solid state information about the double doped (U1−x−yGdxCry)O2 system is very scarce. In the present paper, we present X-ray diffraction and microstructure results of standard U/Gd fuel and chromia doped U/Gd fuel manufactured by powder metallurgy. The dissolution of chromium in (U1−xGdx)O2 as a function of Gd content, the role of free UO2 and the lattice contraction at different Gd and Cr doping levels of (U1−x−yGdxCry)O2 is studied both for the single doped and double doped system. On the basis of lattice contraction and precise measurements of the composition of the solid solution phases, the evolution of theoretical density with dopant concentration is derived.

Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications

ABSTRACTThe feasibility of the fabrication of tungsten based nuclear fuel cermets via Spark Plasma Sintering (SPS) is investigated in this work. CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar properties of these materials. This study shows that after a short time sintering, greater than 90 % density can be achieved, which is suitable to possess good strength as well as the ability to contain fission products. The mechanical properties and the densities of the samples are also investigated as functions of the applied pressures during the sintering.

Analysis of MERCI Decay Heat Measurement for PWR UO2 Fuel Rod

Decay heat measurements, called the MERCI experiment, were conducted at Commissariat à l'Energie Atomique (CEA)/Saclay to characterize accurately residual power at short cooling time and verify its prediction by decay code and nuclear data. The MOSAÏC calorimeter, developed and patented by CEA/Grenoble (DTN/SE2T), enables measurement of the decay heat released by a pressurized water reactor (PWR) fuel rod sample between 200 and 4 W within a precision of 1%. The MERCI experiment included three phases. At first, a UO2 fuel rod sample was irradiated in the CEA/Saclay experimental reactor OSIRIS. The burnup achieved at the end of irradiation was [approximately]3.5 GWd/tonne. The second phase was the transfer of the fuel rod sample from its irradiation location to a hot cell, to be inserted inside the MOSAÏC calorimeter. It took 26 min to carry out the transfer. Finally, decay heat released by the PWR sample was measured from 27 min to 42 days after shutdown. Postirradiation examinations were performed to measure concentrations of some heavy nuclei (U, Pu) and fission products (Cs, Nd). The decay heat was predicted using a calculation scheme based on the PEPIN2 depletion code, the TRIPOLI-4 Monte Carlo code, and the JEFF3.1.1 nuclear data file. The MERCI experiment analysis shows that the discrepancy between the calculated and the experimental decay heat values is included between -10% at 27 min and +6% at 12 h, 30 min after shutdown. From 4 up to 42 days of cooling time, the difference between calculation and measurement is about ±1%, i.e., experimental uncertainty. The MERCI experiment represents a significant contribution for code validation; the time range above 105 s has not been validated previously.

Nucleation and growth of intragranular defect and insoluble atom clusters in nuclear oxide fuels

Uranium and plutonium oxides are subjected to high levels of radiation damage due to the slowing of fission fragments. In addition the composition of the material evolves over time as a result of fission events. Rare gases which constitute an abundant class of fission products are particularly insoluble and therefore tend either to be released from the fuel or form small nanometre size clusters. Bubbles are liable to grow and become trapping sites for migrating defects or other insoluble atoms. Interactions between migrating atoms, defects and existing clusters will determine the rate and extent to which clusters grow. Because the transfer of gas from within the grain to the grain boundaries is thought of as being the rate limiting process for fission gas release, a review of phenomena occurring on the sub-grain scale is carried out. The microstructural modifications induced by neutron irradiations of UO2 fuels are discussed with an emphasis on their relation to fission gas release. Based mainly on TEM studies, the phenomena which are usually taken into account in fission gas behaviour models are looked at and the limitations of these models outlined. More recent experimental and modelling approaches involving ion-irradiation experiments and atomic scale modelling are presented. It is shown that combining these approaches may lead, despite the complexity inherent to the system, to a better understanding of basic radiation induced microstructural changes, clustering events, and rare gas behaviour.

Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

Critical experiments of two cores each loaded with fresh 5 × 5 test PWR-type fuel rods of 235U enrichment of 3.8 wt% or irradiated 5 × 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors k eff's of the critical cores were about −1:2%Δk for the diffusion calculations (JENDL-3.2), −0:5%Δk for the transport calculations (JENDL-3.3), and −0:5 and 0.1%Δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was −2:35 ± 0:07Δk/kk′. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel.

Measurement and Analysis of Reactivity Worth of 241Am Sample in Water-Moderated Low-Enriched UO2 Fuel Lattices at TCA

The reactivity worths of 22.82 grams of 241Am oxide sample were measured and theoretically analyzed in water-moderated UO2 fuel lattices in seven cores of the Tank-Type Critical Assembly (TCA) at the Japan Atomic Energy Agency for an integral test of 241Am nuclear data. These cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The sample reactivity worth was measured with an uncertainty of 2.1% or less. The theoretical analysis was performed using the JENDL-3.3 nuclear data by a Monte Carlo calculation method. Ratios of calculation to experiment (C/Es) of the reactivity worth were between 0.91 and 0.97, and showed no apparent dependence on the neutron spectrum. In addition, sensitivity analysis based on the deterministic calculation method was carried out to obtain the impact of changing the 241Am capture cross section on the sample reactivity worth. The result of this analysis showed that the C/E could be significantly improved by almost uniformly increasing the 241Am capture cross section of JENDL-3.3 by 25–30%.

Radionuclide release from research reactor spent fuel

Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO 2 fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in 235U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO 2-fuel (LWR fuel, enrichment in 235U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Jülich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl 2-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl x -Al and U 3Si 2-Al) was studied in 400 mL MgCl 2-rich salt brine in the presence of Fe 2+ under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH) 3(s) and Eu(OH) 3(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu(IV) oxyhydroxides species due to radiolytic effects cannot completely be ruled out. Solution concentrations of U were within the range of the solubility limits of the solid phase U(OH) 4(am). The determined concentrations of U and Am in solution were about one order of magnitude higher for the U 3Si 2-Al fuel sample. Here, the formation of U/Si containing secondary phase components and their influence on radionuclide solubility cannot be ruled out. Results of this work show that the U 3Si 2-Al and UAl x -Al dispersed research reactor spent fuel samples dissolved completely within the test period of 3.5 years in MgCl 2-rich brine in the presence of Fe 2+. In view of final disposal this means that these fuel matrices represent no barrier. The radionuclides will be released instantaneously. Cs (the long-lived isotope 135Cs is of special concern with respect to final disposal) and Sr were classified as mobile radionuclide species. For U, Am, Pu and Eu, a reimmobilization was observed. Sorption is the process which is assumed to be responsible for the reimmobilization of the long-lived actinide Am and the lanthanide Eu. Solution concentrations of U and Pu seem to be controlled by their solubility controlling solid phases.

Measurement and Analysis of Reactivity Worth of 241Am Sample in Water-Moderated Low-Enriched UO2 Fuel Lattices at TCA

The reactivity worths of 22.82 grams of 241Am oxide sample were measured and theoretically analyzed in water-moderated UO2 fuel lattices in seven cores of the Tank-Type Critical Assembly (TCA) at the Japan Atomic Energy Agency for an integral test of 241Am nuclear data. These cores provided a systematic variation in the neutron spectrum between the thermal and resonance energy regions. The sample reactivity worth was measured with an uncertainty of 2.1% or less. The theoretical analysis was performed using the JENDL-3.3 nuclear data by a Monte Carlo calculation method. Ratios of calculation to experiment (C/Es) of the reactivity worth were between 0.91 and 0.97, and showed no apparent dependence on the neutron spectrum. In addition, sensitivity analysis based on the deterministic calculation method was carried out to obtain the impact of changing the 241Am capture cross section on the sample reactivity worth. The result of this analysis showed that the C/E could be significantly improved by almost uniformly increasing the 241Am capture cross section of JENDL-3.3 by 25–30%.

Analysis of Core Physics Experiments on Fresh and Irradiated PWR UO2 Fuels in the REBUS Program

Critical experiments of two cores each loaded with fresh 5 × 5 test PWR-type fuel rods of 235U enrichment of 3.8 wt% or irradiated 5 × 5 test rods of rod average burnup of 55 GWd/t in the REBUS program were analyzed using diffusion, transport, and continuous-energy Monte Carlo calculation codes coupled with nuclear data libraries based on JENDL-3.2 and JENDL-3.3. Biases in effective multiplication factors k eff's of the critical cores were about −1:2%Δk for the diffusion calculations (JENDL-3.2), −0:5%Δk for the transport calculations (JENDL-3.3), and −0:5 and 0.1%Δk for the Monte Carlo calculations (JENDL-3.3 and JENDL-3.2, respectively). The measured core fission rate and Sc- or Co-activation rate distributions were generally well reproduced using the three types of calculation. The burnup reactivity determined using the measured water level reactivity coefficients was −2:35 ± 0:07Δk/kk′. The calculated result of the Monte Carlo calculations agreed with it; however, the diffusion and transport calculations overestimated the absolute value by about 7%, which would be mainly attributed to the errors in the calculation of the reactivity caused by changing the fuel compositions from fresh fuel to irradiated fuel.