Uranium-plutonium Mixed Oxide Research Articles

Thermal conductivity measurement of uranium–plutonium mixed oxide doped with Nd/Sm as simulated fission products

The thermal conductivities of near-stoichiometric (U,Pu,Am)O2 doped with Nd and/or Sm, which are major fission products (FPs) generated during irradiation of uranium–plutonium mixed oxides (MOX) fuels, as simulated fission products are evaluated from 1073 to 1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneities are prepared using ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the phonon transport model λ=(A+BT)−1. The dependences of the coefficients A and B on the Nd/Sm content (CNd and CSm, respectively) are evaluated as: A(mK/W)=1.70×10−2+0.93CNd+1.20CSm, B(m/W)=2.39×10−4.

Validating Geant4 with Monte Carlo simulations in the context of nuclear disarmament verification

In this paper neutron transport simulation results of experimental configurations used during the 2019 IPNDV measurement campaign at the Belgian Nuclear Research Centre SCK CEN in Mol are presented. As these are a good basis for benchmarking simulations, results of the Monte-Carlo simulation codes Geant4, SCALE/KENO-VI, MCNP and openMC are compared with the aim of validating Geant4. With fissile material present as plutonium–uranium mixed oxide, (α,n) emissions are included in the simulations. A Geant4 extension to calculate neutron multiplication factors keff was developed and applied to the IPNDV configurations and a set of OECD/NEA benchmark experiments. Neutron fluxes of the IPNDV configurations through a reference volume are simulated. The Geant4 (α,n) simulation toolkit SaG4n is compared to the SCALE code ORIGEN and small deviations are observed. Geant4 calculations of keff for criticality benchmark experiments deviate in the mean by ＋0.4%, for MCNP and openMC deviations are ≤6%, for KENO-VI ≤8%. Geant4’s total neutron fluxes through a reference volume next to the IPNDV configurations agree within a margin of ±5% with MCNP and openMC and −14 % with KENO-VI. Considerable differences between all codes were observed for thermal neutron scattering but Geant4 results are still within the other codes’ variability. Differences to Geant4 for epithermal energies can be addressed to a not yet included “Doppler Broadening Rejection Correction”. By comparing Geant4 simulations of a complex set-up with various reference codes and demonstrating its good performance as well as some discrepancies, this study contributes to the validation of Geant4 neutron physics in fissile material systems and for nuclear disarmament verification simulations.

Monitoring and simulation of the fuel irradiation behavior in nuclear reactors based on phononic crystal structure

We have presented in the current work a novel idea for simulating the irradiation behaviors of the nuclear fuel pellets in nuclear reactors by using a one-dimensional defective phononic crystal (1D-DPnC) design was presented. The transmission spectra of the incident mechanical waves were considered basic data for expressing the characteristics of different nuclear fuel-pellets. Herein, the density, sound speed, and Young’s modulus of the fuel-pellets represent the key parameters that are influenced by the irradiation behaviors of these pallets. Mixed plutonium–uranium oxide (MOX) nuclear fuel is considered the main fuel in the present study. In addition, a comparison is performed for this fuel with other types of nuclear fuels. Moreover, the mechanical properties of these MOX-pellets are dependent on the porosity, the ratio of oxygen-to-metal (O/M), and the plutonium (Pu-content). The theoretical treatments depend on the transfers matrix method to compute the transmission spectra through the 1D-DPnC. The numerical findings provided that the MOX-pellet has the highest performance compared to other fuel pellets and with sensitivity equal to 59.388 × 103 Hz s/m. It was also reported that the effects of the percentage of the O/M and Pu- content in MOX had a minor effect in a comparison with the impact of porosity. The theoretical simulation agreed extremely with the experimental data reported for these nuclear fuels. Because of the close relationship between sound speed and density, this sensor can be utilized to monitor the porosity, O/M, Pu-content, and density of fuel-pellets as a quick and non-destructive evaluation technique in a nuclear fuel fabrication laboratory. This article has proven theoretically that MOX fuel produced from nuclear waste of uranium dioxide and plutonium dioxide gives excellent results compared to other types of nuclear fuels, and this agrees with experimental researches. Thus, it may contribute in preserving the environment from nuclear waste, and this can be considered a novel kind of purification of environmental pollution treatment.

Multi-scale structural investigation of uranium-plutonium mixed oxides (U1-yPuy)O2-x with high plutonium content

Uranium-Plutonium mixed oxides U1-yPuyO2-x are considered as the reference fuels for Sodium-cooled Fast neutron Reactors (SFRs). Nevertheless, there is a lack of experimental data on their thermal properties for a Pu content above 0.45. The first step to fill this gap is to manufacture samples with high plutonium contents, which is the purpose of this work. Here, the manufacturing of U1-yPuyO2 samples with y = 0.60, 0.65 and 0.70 by a co-milling powder metallurgy process is detailed. Relying on an optimization of each manufacturing step (blending, co-milling, sieving, pelletizing and sintering), in order to obtain dense, monophasic, stoichiometric and homogenous materials.A multi-scale characterization strategy was used to determine sample microstructural properties including density measurements, Thermal-Ionization Mass Spectrometer (TIMS), ceramography, Electron Probe Micro Analyses, µ-Raman Spectroscopy and X-Ray Diffraction. The results obtained show that dense (> 95% TD), homogenous (chemical distribution of cations and oxygen), monophasic and stoichiometric (O/M = 2.00) samples have been achieved. These properties correspond to those needed to optimize the thermo-physical properties measurements. This also shows that the coupling of multi-scale characterizations is mandatory to evaluate the structural and microstructural properties of this kind of material.

Semi-supervised generative approach to chemical disorder: application to point-defect formation in uranium-plutonium mixed oxides.

Chemical disorder has a major impact on the characterization of the atomic-scale properties of highly complex chemical compounds, such as the properties of point defects. Due to the vast amount of possible atomic configurations, the study of such properties becomes intractable if treated with direct sampling. In this work, we propose an alternative approach, in which samples are selected based on the local atomic composition around the defect, and the defect formation energy is obtained as a function of this local composition with a reduced computational cost. We apply this approach to (U, Pu)O2 nuclear fuels. The formation-energy distribution is computed using machine-learning generative methods, and used to investigate the impact of chemical disorder and the range of influence of local composition on the defect properties. The predicted distributions are then used to calculate the concentration of thermal defects. This approach allows for the first time for the computation of the latter property with a physically meaningful exploration of the configuration space, and opens the way to a more efficient determination of physico-chemical properties in other chemically-disordered compounds such as high-entropy alloys.

Biodegradable Methane Sulfonic Acid-Based Nonaqueous Dissolution, Estimation, and Recovery: Toward Development of a Simplified Scheme for Plutonium-Bearing Fuel Matrices

Designing actinide separation methodologies, which simplify process scheme and minimize radioactive waste generation, form the crux for enhancing the sustainability of the nuclear industry. The present study is a step toward development of a universal and simplified scheme for the bulk assay of plutonium and uranium as a part of chemical quality control of plutonium-bearing fuel matrices. Dissolution, estimation, and recovery form the key steps for the same. Plutonium oxide, plutonium–uranium mixed oxide, and plutonium–uranium mixed carbide were quantitatively dissolved in the biodegradable organic acid, viz., methane sulfonic acid (MSA), by heating under an infrared lamp. Usage of corrosive hydrofluoric acid or refluxing conditions were avoided. Stabilization of a single oxidation state, viz., +(IV) and +(VI) for plutonium and uranium in MSA, respectively, was confirmed by ultraviolet–visible spectrophotometry and electrometry. Electrometric studies also exhibited sharp and well separated reduction peaks for U and Pu in MSA, revealing promising possibility of simultaneous voltammetric determination of U and Pu on commercially available glassy carbon electrode. The redox energetic diffusion coefficient (D0) and heterogenous electron transfer kinetic parameters (αn and k0) of this novel system, viz., U and Pu in MSA, were obtained using cyclic voltammetry. Single stage recovery by oxalate precipitation yielding Pu removal of >98% was also demonstrated. The findings reported herein reveal the scope for prospective applications in the front end as well as back end of the nuclear fuel cycle.

Cation interdiffusion in uranium–plutonium mixed oxide fuels: Where are we now?

The diffusion phenomena in uranium–plutonium mixed oxides U1−yPuyO2dictate the physicochemical properties of mixed oxides (MOX) nuclear fuel throughout manufacturing, irradiation, and storage. More precisely, it is paramount to estimate the cation interdiffusion insofar as it dovetails with the actinide redistribution during sintering and under irradiation. This paper draws a critical review of the existing experimental data of U and Pu interdiffusion coefficients in MOX fuel.

Effect of cationic chemical disorder on defect formation energies in uranium–plutonium mixed oxides

At the atomic scale, uranium–plutonium mixed oxides (U,Pu)O2 are characterized by cationic chemical disorder, which entails that U and Pu cations are randomly distributed on the cation sublattice. In the present work, we study the impact of disorder on point defect formation energies in (U,Pu)O2 using interatomic-potential and density functional theory (DFT + U) calculations. We focus on bound Schottky defects (BSD) that are among the most stable defects in these oxides. As a first step, we estimate the distance RD around the BSD up to which the local chemical environment significantly affects their formation energy. To this end, we propose an original procedure in which the formation energy is computed for several supercells at varying levels of disorder. We conclude that the first three cation shells around the BSD have a non-negligible influence on their formation energy (RD≃7.0Å). We apply then a systematic approach to compute the BSD formation energies for all the possible cation configurations on the first and second nearest neighbor shells around the BSD. We show that the formation energy can range in an interval of 0.97 eV, depending on the relative amount of U and Pu neighboring cations. Based on these results, we propose an interaction model that describes the effect of nominal and local composition on the BSD formation energy. Finally, the DFT + U benchmark calculations show a satisfactory agreement for configurations characterized by a U-rich local environment and a larger mismatch in the case of a Pu-rich one. In summary, this work provides valuable insights on the properties of BSD defects in (U,Pu)O2 and can represent a valid strategy to study point defect properties in disordered compounds.

Multiphysics thermo-mechanical behavior modeling for annular uranium-plutonium mixed oxide fuels in a lead-cooled fast reactor

Understanding and predicting nuclear fuel behavior is the cornerstone for fuel design and optimization, and the safe and economic operation of nuclear reactors. Due to the excellent neutronic economy and safe operational characteristics of lead/lead-bismuth-based liquid metal coolant, lead-cooled fast reactor (LFR) has received much attention and has been regarded as one of the top candidate reactors for Generation IV nuclear power plants. Mixed oxide (MOX) fuel is beneficial to improve fuel cycle utilization and consume weapons-grade plutonium, among which hollow MOX fuel has excellent heat transfer advantages, which can further improve fuel economy and safety, and has good development prospects. In this work, an annular uranium-plutonium mixed oxide (MOX) fuel operating in a liquid lead/lead-bismuth cooled fast reactor is modeled and simulated to predict its behavior under transient and steady-state operation. This model has several modules working in a fully coupled approach based on COMSOL Multiphysics, such as heat transfer, fuel burnup, oxygen redistribution, plutonium redistribution, porosity evolution, fission gas release, JOG growth, and grain size evolution. The modeling results were benchmarked with the existing codes and experimental data, and the results were in satisfactory agreement. The parameters analysis was carried out in this work. Consistent with the solid MOX fuel, the O/M ratio (or deviation from the stoichiometry of oxygen) significantly affects temperature evolution, fission gas migration and release behavior, and plutonium redistribution. Linear power also has a significant influence on fuel performance.

Determination of uranium and plutonium in mixed oxide samples by X-ray methods and their comparison with potentiometry

Uranium and plutonium in nuclear fuels are generally determined by conventional potentiometric methods which generates large quantities of radioactive waste. In the present study, K-edge densitometry (KED) and Energy dispersive X-ray fluorescence (EDXRF) have been examined for the determination of uranium and plutonium in solutions as an alternative to potentiometric methods. In the present study, uranium plutonium mixed oxide (MOX) pellets were dissolved in nitric acid medium containing small quantities of HF. Uranium and plutonium in MOX solutions were determined by KED and EDXRF methods and the results are compared with the one obtained by conventional potentiometric methods. The relative standard deviation (RSD) of the analytical results for U and Pu using KED were found to be better than ±1% and ± 0.9% respectively. Similarly, RSD for U and Pu by EDXRF were found to be better than ±1.6% and ± 0.8%, respectively. The absolute difference between KED and EDXRF with respect to potentiometric data for uranium and plutonium were calculated and it was found to be marginal indicating that KED and EDXRF methods can be employed for the routine analysis of U and Pu in solutions. X-ray methods employed in the present study are non-consumptive, direct and offer precise determination of U and Pu and generate no waste.

Application of the SCIANTIX fission gas behaviour module to the integral pin performance in sodium fast reactor irradiation conditions

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their “pre-INSPYRE” versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the “post-INSPYRE” code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

A scenario study on the transition to a closed nuclear fuel cycle using the nuclear energy system modelling application package (NESAPP)

The paper presents the results of a case study on evaluating performance and sustainability metrics for Russian nuclear energy deployment scenarios with thermal and sodium-cooled fast reactors in a closed nuclear fuel cycle. Ten possible scenarios are considered which differ in the shares of thermal and sodium-cooled fast reactors, including options involving the use of mixed uranium-plutonium oxide fuel in thermal reactors. The evolution of the following performance and sustainability metrics is estimated for the period from 2020 to 2100 based on the considered assumptions: annual and cumulative uranium consumption, needs for uranium enrichment capacities, fuel fabrication and reprocessing capacities, spent fuel stocks, radioactive wastes, amounts of plutonium in the nuclear fuel cycle, amounts of accumulated depleted uranium, and the levelised electricity generation cost. The results show that the sustainability of the Russian nuclear energy system can be significantly enhanced through the intensive deployment of sodium-cooled fast reactors and the transition to a closed nuclear fuel cycle. The authors have highlighted some issues for further considerations, which will lead to more rigorous conclusions regarding the preferred options for the development of the national nuclear energy system.

Spark plasma sintering of (U,Ce)O2 as a MOx nuclear fuel surrogate

Mixed uranium plutonium oxides (MOx) are currently used as nuclear fuel in thermal fission reactors and are the leading candidate for use in solid fuelled fast reactors given a wealth of irradiation data from research and prototype systems. Spark plasma sintering (SPS) has been shown to reduce the sintering temperature and time required to manufacture UO2 fuel pellets and may benefit MOx production through the more efficient retention of volatile minor actinides (MAs) such as AmO2 in the manufacture of these fuels. In this study we have examined SPS as an alternative field assisted sintering route to UO2 and MOx fuel manufacture on surrogate (U0.93,Ce0.07)O2 prepared by a micronized master blend (MIMAS) type process. SPS has been shown to produce pellets of high density, > 95% TD at sintering temperatures of 1573 K and hold time of 5 min, comparable to equivalent pellets produced through conventional sintering at 1973 K for 5 h. The formation of Ce-rich regions has been measured showing the size distribution of Ce-rich particles in SPS Ce-MOx is similar to conventionally sintered Ce-MOx and are appropriate for MOx fuel specifications. However, a significant difference is observed in the Ce content of the Ce-rich particles which remain close to their primary blend composition, as well as having a high degree of heterogeneity within the Ce-rich regions. The effect of the shortened sintering temperatures and times on other microstructural properties are also revealed.

Burnup and neutronic analysis for UOX and MOX fuel assemblies in VVER reactor

The present work examined the feasibility of utilizing Mixed Uranium–Plutonium Oxide (MOX) fuel in the VVER-1000 reactor core. In this study, about 30% of Uranium Oxide (UOX) fuel assemblies of the original core are replaced by MOX fuel assemblies. The effect of MOX fuel on the safety parameters is studied by using the MCNP6 code and the nuclear data library (ENDF/B)-VII.1. The safety parameters examined included the effective multiplication factor, temperature coefficients, void coefficient, boron coefficient, shutdown margin and the kinetic parameters. The excess reactivity, neutron flux and power density distributions and power peaking factor are also examined. All estimated values of reactivity coefficients analyzed for the core contains 30% MOX fuel confirmed to be negative, which guarantees the reactor safety and power stability in a normal state. Also, the effective delayed neutron fraction was lower than that of the full UOX core as well as the neutron lifetime was shorter, this shows that the MOX core responds faster than the UOX core. The maximum power factor for the MOX core was equal to the design limit. The shutdown margin satisfied the design requirement. The results confirmed that the VVER-1000 is able to meet the safety requirement for core designs with MOX loadings of up to 30%.

Assessment of three European fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

Critical analysis of light-water moderation UO2 and uranium-plutonium mixed oxide lattices by continuous-energy Monte Carlo calculation

ABSTRACT Critical analysis of 120 light-water moderation lattices composed of 2.6 wt% UO2 fuel rods or 3.0 wt%-PuO2 uranium-plutonium mixed oxide (MOX) fuel rods was performed using a continuous-energy Monte Carlo calculation code to validate the JENDL-4.0 nuclear data library. As a result, the average neutron multiplication factor (keff ) of the 35 UO2 lattices (hydrogen-to-heavy-metal atomic ratio (H/HM): 4.3–8.7) was 0.99930 with the uncertainty of 0.00005 (1 σ), and that of the 85 MOX lattices (H/HM: 12.0–27.6) was 0.99856 with the uncertainty of 0.00006 (1 σ). The keff values of the UO2 lattices partly showed a decreasing trend with H/HM, while those of the MOX lattices an increasing trend. The keff values of the MOX lattices presented an increasing trend with the measurement dates. Assuming that the trend was relating to the uncertainties in the neutron capture cross-section of 241Am, a simulation to modify the cross-section was performed by increasing the atomic number densities of 241Am in the MOX fuel. It indicated that the neutron-flux-average capture cross-sections of 241Am for the MOX lattices should be increased by 10% to 20% from those calculated by the cross-section in JENDL-4.0.

A technique for uranium and plutonium determination using coulometric potentiostatic facility UPK-19 for analysis of mixed-oxide fuel

The content of uranium and plutonium is the main characteristic of mixed uranium-plutonium oxide fuel, which is strictly controlled and has a very narrow range of the permissible values. We focused on developing a technique for measuring mass fractions of uranium and plutonium by controlled potential coulometry using a coulometric unit UPK-19 in set with a R-40Kh potentiostat-galvanostat. Under conditions of sealed enclosures, a special design of the support stand which minimized the effect of fluctuations in ambient conditions on the signal stability was developed. Optimal conditions for coulometric determination of plutonium and uranium mass fractions were specified. The sulfuric acid solution with a molar concentration of 0.5 mol/dm3 was used as a medium. Lead ions were introduced into the background electrolyte to decrease the minimum voltage of hydrogen reduction to –190 mV. The addition of aluminum nitride reduced the effect of fluoride ions participating as a catalyst in dissolving MOX fuel samples, and the interfering effect of nitrite ions was eliminated by introducing a sulfamic acid solution into the cell. The total content of uranium and plutonium was determined by evaluation of the amount of electricity consumed at the stage of uranium and plutonium co-oxidation. Plutonium content was measured at the potentials, at which uranium remains in the stable state, which makes it possible to subtract the contribution of plutonium oxidation current from the total oxidation current. The error characteristics of the developed measurement technique were evaluated using the standard sample method and the real MOX fuel pellets. The error limits match the requirements set out in the specifications for MOX fuel. The technique for measuring mass fractions of uranium and plutonium in uranium-plutonium oxide nuclear fuel was certified. The relative measurement error of the mass fraction of plutonium and uranium was ±0.0070 and ±0.0095, respectively. The relative error of the ratio of the plutonium mass fraction to the sum of mass fractions of uranium and plutonium was ±0.0085.

Modelling of plutonium diffusion in (U,Pu)O2±x mixed oxide

The modelling of the thermo-kinetic properties of uranium-plutonium mixed oxide (MOX) is of utmost importance for optimizing its synthesis and for predicting its behaviour in Fast Breeder Reactors. Despite the stakes and likely because of experimental issues, little or no experimental data are available for the entire MOX system. We circumvent here the difficulties by developing a mobility database for plutonium using the DICTRA code. A well-established model of MOX formalized within the Compound Energy Formalism ensures the thermodynamic description. Rationalisation of the mobility parameters combined with the use of both cBΩ model and the few experimental data lead to a full and comprehensive description of plutonium self-diffusion in MOX for any plutonium content, O/M ratio and temperature. Additionally, we show that the observed plateau of the self-diffusion as a function of the oxygen to metal ratio (O/M) is related to the constant Pu3+ fraction for very low O/M ratio. Moreover, the observed minimum close to O/M = 2 is found for the lowest mobility of Pu4+.

CALCULATION OF 2-DIMENSIONAL PWR MOX/UO2 CORE BENCHMARK OECD NEA 6048 WITH SRAC CODE

The mixed uranium-plutonium oxide fuel (MOX/UO2) is an interesting fuel for future power reactors. This is due to the large amount of plutonium that can be processed from spent fuel of nuclear plants or from plutonium weapons. MOX/UO2 fuel is very flexible to be applied in thermal reactors such as PWR and it is more economical than UO2 fuel. However, due to the different nature of neutron interactions of MOX in PWR, it will change the reactor core design parameters and also its safety characteristic. The purpose of this study is to determine the accuracy of SRAC2006 code system in generation of cross-sections and calculation of reactor core design parameters such as criticality, reactivity of control rods and radial power distribution. In this study, PWR MOX/UO2 Core Transient Benchmark is used to verify the code that models a MOX/UO2 fueled core. SRAC-CITATION result is different from DeCART by 0.339% from. SRAC-CITATION result of single rod worth in All Rods Out (ARO) conditions are quite good with a maximum difference of 6.34% compared to BARS code and 4.74% compared to PARCS code. In All Rods In (ARI) condition, SRAC-CITATION results compared to the PARCS code is quite good where the maximum difference is 9.72%, but compared to BARS code, it spikes up to 33.24% at maximum difference. In the other case, overall radial power density results are quite good compared to the reference. Its maximum deviation from DeCART code is 5.325% in ARO condition and 6.234% in ARI condition. Based on the results of these calculations, SRAC code system can be used to generate cross-section and to calculate some neutronic parameters. Hence, it can be used to evaluate the neutronic parameters of the MOX/UO2 PWR core design.Keywords: MOX/UO2 fuel, Criticality, Power peaking factor, SRAC2006

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.