Fast Reactor Mixed Oxide Research Articles

Nanoindentation testing of high burn-up fast reactor mixed oxide fuel

Continued research into the mechanical properties of nuclear fuel is necessary to improve modeling of the pellet-cladding mechanical interactions that occur in nuclear reactors during operation. Small scale mechanical testing with focused ion beam equipment can provide a means to study the mechanical properties of irradiated nuclear fuels by measuring their mechanical properties on minute volumes of material. In this work, room temperature nanoindentation measured the hardness and Young's modulus of fast reactor mixed oxide (MOX) fuel irradiated to 183 GWd/tHM. The hardness of the MOX fuel had a measured value of 11.2 ± 0.4 GPa. The Young's modulus of the MOX fuel had a value of 158 ± 6 GPa. The measured hardness and Young's modulus are discussed in the context of the available literature data. In addition, pathways toward elevated temperature testing and other mechanical properties are proposed.

Modelling and assessment of thermal conductivity and melting behaviour of MOX fuel for fast reactor applications

Thermal conductivity and melting temperature of nuclear fuel are essential for analysing its performance under irradiation, since they determine the fuel temperature profile and the melting safety margin, respectively. A starting literature review of data and correlations revealed that most models implemented in state-of-the-art fuel performance codes (FPCs) describe the evolution of thermal conductivity and melting temperature of Light Water Reactor (LWR) MOX (uranium-plutonium mixed oxide) fuels, in limited ranges of operation and without considering the complete set of fundamental dependencies (i.e., fuel temperature, burn-up, plutonium content, stoichiometry, and porosity). Since innovative Generation IV nuclear reactor concepts (e.g., ALFRED, ASTRID, MYRRHA) employ MOX fuel to be irradiated in Fast Reactor (FR) conditions, codes need to be extended and validated for application to design and safety analyses on fast reactor MOX fuel. The aim of this work is to overcome the current modelling and code limitations, providing fuel performance codes with suitable correlations to describe the evolution under irradiation of fast reactor MOX fuel thermal conductivity and melting temperature. The new correlations have been obtained by a statistically assessed fit of the most recent and reliable experimental data. The resulting laws are grounded on a physical basis and account for a wider set of effects on MOX thermal properties (fuel temperature, burn-up, deviation from stoichiometry, plutonium content, porosity), providing clear ranges of applicability for each parameter considered. As a first test series, the new correlations have been implemented in the TRANSURANUS fuel performance code, compared to state-of-the-art correlations, and assessed against integral data from the HEDL P-19 fast reactor irradiation experiment. The integral validation provides promising results, pointing out a satisfactory agreement with the experimental data, meaning that the new models can be efficiently applied in engineering fuel performance codes.

Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products

This work presents a comparison between the microstructural evolution of three annular fast-reactor mixed-oxide (MOX) fuel pellets irradiated to varying burnups at the Fast Flux Test Facility (FFTF). Fuel pellets irradiated to 3.4%, 13.7%, and 21.3% fissions per initial metal atom (FIMA) were examined using scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) techniques. The cross-section of the low burnup pellet displayed minor structural changes, but the central annulus of the pellets at 13.7% and 21.3% FIMA shrank from their starting size. The high burnup fuel pellet featured streaking and porosity migration associated with columnar grain growth. The radial fission product distribution in each of the pellets had a higher number density of metallic particles >5 μm in diameter near the fuel centerline. Solid fission products in the fuel-cladding gap were observed in the low and intermediate burnup pellets. The low burnup sample showed minor accumulation of Ba in the gap, while the volatile Cs was primarily observed at the pellet surface. The intermediate burnup pellet displayed a porous mixture of fission products, consistent with the joint-oxide gain (JOG) that has been previously observed in fast-reactor MOX fuel pellets.

TEM characterization of high burnup fast-reactor MOX fuel

This work examines the radial evolution of a high burnup fast reactor mixed-oxide (MOX) fuel pellet irradiated at the Fast Flux Test Facility (FFTF) using scanning (SEM) and transmission electron microscopy (TEM). At 21.3% fissions per initial metal atoms (FIMA), the sample analyzed in this work is likely among the highest burnup oxide fuel samples ever evaluated using TEM. Initial SEM examination focused on the characterization of the fuel microstructure and solid fission product behavior as a function of radial position. The fuel pellet underwent extensive restructuring as demonstrated by columnar grain formation near the central void. Oxide fission product phases were confined to near the fuel centerline and decomposed from perovskite BaZrO3 to the fluorite ZrO2 type crystal structure. Metallic five metal precipitates (FMPs) were observed at all radial positions and the Pd-rich metallic phase increased in area fraction moving toward the pellet surface. TEM examination focused on the determination of dislocation density ranging from the fuel pellet center to the pellet surface. Dislocation density was highest in the cooler regions of the fuel pellet, with the outer edge experiencing grain refinement and high defect concentrations consistent with the rim structure. The results of the paper highlight the variable dislocation behavior as a function of radial position in high burnup MOX fuels, as well as describe the largely unexplored change in crystal structure of the oxide solid fission product phase.

Radial microstructural evolution in low burnup fast reactor MOX fuel

Examination of irradiated fast reactor mixed-oxide (MOX) fuel was conducted to analyze microstructural behavior and defect evolution as a function of radial position. The fuel analyzed in this work was irradiated to 3.4% fissions per initial metal atom (FIMA) in the Fast Flux Test Facility. Scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) were used to analyze microstructural features and fission product secondary phases. A dual-beam focused ion beam (FIB)/SEM was used to prepare several samples along the fuel radius for transmission electron microscopy (TEM) defect analysis. The microstructure near the center of the fuel pellet showed separated grain boundaries accompanied by fission gas bubbles. The fuel pellet did not reach the threshold linear heating rate (LHR) needed for restructuring, formation of the columnar or equiaxed regions was not observed. Metallic fission products aggregated in pores along grain boundaries near the pellet center, but no metallic precipitates were visible in the outer region. Moreover, no appreciable amount of insoluble perovskite phase was formed. The dislocation loop density is highest near the fuel central void but decreases past the mid-point of the fuel radius likely due to the self-irradiation effects. Dislocation lines follow an inverse trend, with the lower density observed near the hottest regions and increasing in the pellet outer radius. These results indicate that the thermal gradient in MOX fuels heavily influences the localized fission product morphology and defect behavior.

Coupled computer code study on irradiation performance of a fast reactor mixed oxide fuel element with an emphasis on the fission product cesium behavior

A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup for discussion on the axial distribution of Cs, fuel-to-cladding mechanical interaction owing to the Cs-fuel-reactions and the Cs-Mo-O compound formed in the fuel-to-cladding gap by comparing the calculated results with experimental data obtained from post irradiation examinations.

Direct burn-up determination of fast reactor mixed oxide (MOX) fuel by preferential evaporation of interfering elements

Spent fuel of uranium–plutonium mixed oxide (MOX) from sodium cooled Fast Breeder Test Reactor (FBTR) was analyzed for at.% burn-up by preferential evaporation method. A sequential pattern of analysis of fission monitor Nd and heavy elements, U and Pu provided an un-interfered isotopic composition. Concentrations of individual elements were determined by isotopic dilution mass spectrometry. The proposed method provides at.% burn-up with an uncertainty of about 4% (compared to ASTM method), is less time consuming, does not involve any chemical separation, reduction in radioactive waste and substantial reduction in the radiation exposure to analyst.

Fully Coupled Modeling of Burnup-Dependent (U1−y , Pu y )O2−x Mixed Oxide Fast Reactor Fuel Performance

During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1−yPuy)O2−x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance. However, it is not that significant compared to the deviation from stoichiometry of oxygen due to the similar material properties of UO2 and PuO2.

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet–cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and 137Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with half of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO 2 spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle.

Effects of Nuclear Data Library on BFS and ZPPR keff Analysis Results

The data-processing system CONSYST/ABBN coupled with the ABBN-93 nuclear data library (NDL) was used in the analysis of BFS-62 and ZPPR JUPITER experiment series fast reactor mixed oxide cores, applying the Japan Nuclear Cycle Development Institute (JNC), Japan, core calculation code CITATION-FBR. The FFCP cell code was used to take into account the spatial cell heterogeneity and resonance effects based on the subgroup approach.The NDL effect has been studied by comparing the results calculated using the ABBN-93 nuclear data with the former ones obtained at JNC based on the JENDL-3.2 NDL. Calculation analysis results for the keff parameter for four BFS-62 cores as well as for three ZPPR JUPITER experiment series cores (ZPPR-9, ZPPR-13A, and ZPPR-17A) have been obtained. The estimated uncertainty in the keff values caused by possible errors in calculation of the applied corrections was estimated to be ˜0.3% in the case of the ZPPR cores and 0.2% for the BFS-62 cores. The NDL effect determined by applying the ABBN-93 nuclear data in the JNC calculation route for the keff parameter appeared to be ˜0.3% for the ZPPR and BFS-62 cores with plutonium. As for BFS-62 uranium-loaded cores, the NDL effect was ˜0.1%. Sensitivity analysis was applied, and it has shown that the main contributors to the NDL effect are uranium, plutonium, and iron cross sections.This work is closely related to the JNC-Institute of Physics and Power Engineering, Russian Federation, collaboration on experimental investigation of excess weapon plutonium disposition in the BN-600 reactor using the BFS facility.

Development of Mixed-Oxide Fuel Manufacture in the United Kingdom and the Influence of Fuel Characteristics on Irradiation Performance

The fabrication of mixed uranium-plutonium oxide fuel for the U.K. Fast Breeder Reactor Programme is discussed, and the experience gained by the United Kingdom Atomic Energy Authority and British Nuclear Fuels pic over the past 25 yr is used to identify aspects of the manufacturing routes that influence irradiation performance and subsequent recycle. The design of a small-scale production plant for fast reactor mixed-oxide (MOX) fuels is discussed, and its conversion to produce 8 tonne/yr (heavy metal) thermal MOX fuel is outlined.

Fast reactor fuel structure evolution under irradiation / Brennstoffstrukturänderung in schnellen Brutreaktoren unter Bestrahlung.

Under steady-state and transient operational conditions, fast reactor mixed oxide fuel is subjected to a thermal state which leads to significant changes in the fuel. Two items have been identified to influence significantly the fuel temperature distribution: fuel densification and relocation via vaporization/condensation. In the experiments DUELL and POTOM, especially designed for the investigation of fuel restructuring and power-to-melt conditions, these effects have clearly been elaborated. Fuel restructuring and related phenomena like fuel densification, cracking, crack healing and relocation have a significant impact on fuel behaviour. Experimental results and calculations, carried out with the fuel pin behaviour code SATURN-S, give a consistent picture of the time and temperature dependent fuel structure evolution

Evaluation of the fission gas release behavior from fast reactor mixed oxide fuel based on local concentration measurement of retained xenon

The radial distribution of retained xenon concentration in the mixed oxide fuel pellets irradiated in JOYO MK-1 and Rapsodie up to a burnup of 100 GWd/t were measured using electron probe microanalyser (EPMA) and ion probe microanalyser (IMA). For the irradiated fuel up to a burnup level of 50 GWd/t at a linear power less than 300 W/cm, the local release of xenon occurred at the densified region just outside the equiaxed grain region in which fission gas almost escaped from the pellet. Beyond 50 GWd/t under around 450 W/cm, xenon release was accompanied by submicron size intragranular bubble formation just inside the unrestructured region. In the unrestructured region, the retained fission gas was found to be released at the burnups beyond 50 GWd/t, which is related to the saturation concentration of fission gas within grains. The fission gas release rates in unrestructured and restructured regions were correlated with burnup up to 100 GWd/t.

Plutonium redistribution in fast reactor mixed oxide fuel pins

Some redistribution effects of uranium and plutonium, caused by thermal diffusion and evaporation-condensation processes in mixed oxide fuels, are discussed by means of autoradiographs of sections of fuel pins irradiated in the fast flux of the RAPSODIE reactor. The change in the stoichiometric state as a function of burnup and the radial distribution of oxygen are described and their influence on the redistribution processes is discussed. A model and suitable data are given to calculate redistribution effects on the basis of thermal diffusion in fast reactor fuels. In fuel pins with power ratings of 500 W/cm and 600 W/cm the enrichment of plutonium around the central cavity produces an increase in the central temperature of about 100°C and 250° C, respectively.

Fast Reactor Fuel Performance Model Development

AbstractThe status of fast reactor mixed-oxide fuel performance model development, involving the OLYMPUS and CYGRO-F codes, is reviewed. Input information in several critical areas is examined including: the swelling, irradiation creep, and ductility of stainless-steel cladding; and the swelling and plasticity of mixed-oxide fuel. The predictions from the two codes are illustrated by parametric studies and application to high fluence/burnup ratio fuel rods of interest to the LMFBR program. The analytical studies indicate the major effects of stainless-steel swelling on fuel rod performance and the relationship between fuel density and cladding creep strain.