Future Fast Reactors Research Articles

Gas Entrainment Simulation for Fast Reactors using Freesurface Lattice Boltzmann Method

In fast reactor designs, preventing gas entrainment (GE) caused by free-surface vortices is crucial. Numerical analysis is typically used to identify scenarios with and without GE and speedup the design process. However, conventional numerical methods based on incompressible Navier-Stokes solvers are often inefficient and inaccurate. To address this, we explore the Lattice Boltzmann Method (LBM) as an alternative, known for its computational efficiency and ability to simulate complex flows. In this study, we employ freesurface LBM to accelerate GE analysis, greatly reducing computational costs. Our LBM simulations show good agreement with experimental data, suggesting a promising approach for faster analysis in future fast reactor designs.

A fine pore-preserved deep neural network for porosity analytics of a high burnup U-10Zr metallic fuel

U-10 wt.% Zr (U-10Zr) metallic fuel is the leading candidate for next-generation sodium-cooled fast reactors. Porosity is one of the most important factors that impacts the performance of U-10Zr metallic fuel. The pores generated by the fission gas accumulation can lead to changes in thermal conductivity, fuel swelling, Fuel-Cladding Chemical Interaction (FCCI) and Fuel-Cladding Mechanical Interaction (FCMI). Therefore, it is crucial to accurately segment and analyze porosity to understand the U-10Zr fuel system to design future fast reactors. To address the above issues, we introduce a workflow to process and analyze multi-source Scanning Electron Microscope (SEM) image data. Moreover, an encoder-decoder-based, deep fully convolutional network is proposed to segment pores accurately by integrating the residual unit and the densely-connected units. Two SEM 250 × field of view image datasets with different formats are utilized to evaluate the new proposed model’s performance. Sufficient comparison results demonstrate that our method quantitatively outperforms two popular deep fully convolutional networks. Furthermore, we conducted experiments on the third SEM 2500 × field of view image dataset, and the transfer learning results show the potential capability to transfer the knowledge from low-magnification images to high-magnification images. Finally, we use a pre-trained network to predict the pores of SEM images in the whole cross-sectional image and obtain quantitative porosity analysis. Our findings will guide the SEM microscopy data collection efficiently, provide a mechanistic understanding of the U-10Zr fuel system and bridge the gap between advanced characterization to fuel system design.

Comparison of Zirconium Redistribution in BISON EBR-II Models Using FIPD and IMIS Databases with Experimental Post Irradiation Examination

Metallic fuels have seen increased interest for future sodium fast reactors due to their material properties: high thermal conductivities and advantageous neutronic properties allow for greater fission densities. One drawback to typical metallic fuels is zirconium redistribution, which impacts this advantageous material and its neutronic properties. Unfortunately, the processes behind zirconium migration behavior are understood using first principles, so before these fuels are implemented in future fast reactors, characterization and fuel qualification regimes must be completed. These activities can be supported through the use of robust modeling using the most accurate empirical models currently available to fuel researchers around the world. The tool that allows researchers to model this complex coupled thermo-mechanical behavior and nuclear properties is BISON. Additionally, BISON model parameters need to be compared against PIE measurements. The current work utilizes two fuel pins from EBR-II experiment X441 to optimize various model parameters, including porosity correction factor, thermal conductivity, phase transition temperature, and diffusion coefficient multipliers, before implementing the final model for seven fuel pins with differing characteristics. To properly evaluate the BISON simulations, the results are compared to PIE metallography data for each fuel pin, to ensure the zirconium redistribution is properly reflected in the simulation results. Six out of seven analyzed fuel pins demonstrate good agreement between the metallography images and BISON results, showing alignment of the Zr-rich, Zr-depleted, and moderately Zr-enriched zones at various axial heights along the fuel pins. Further work is needed to refine the model parameters for general pin use.

Modeling and Validation Studies on Solidification of Sodium Nitrate and LiCl‐KCl

AbstractTernary alloy fuel will be used as driver fuel for future fast reactors. Reprocessing of this spent alloy fuels will be carried out by pyrochemical reprocessing which is based on molten salt electrorefining at high temperature. During electrorefining, fission products get accumulated in the electrolyte along with some actinides in the form of their chlorides. The presence of fission products in the salt will add to the radioactivity levels and hence have to be stripped off. Melt crystallization is pursued to remove fission products from contaminated salt and to reuse the purified salt for electrorefining. Numerical modeling and experimental studies on tracking the phases during solidification of molten sodium nitrate salt were carried out. The numerical predictions closely matched with the experimental results. A resistance measurement technique was developed to accurately track the solidification interface so as to validate the numerical predictions.

Hydraulic design optimization of spherical header for future fast reactors of India

The liquid sodium as coolant in a typical pool type Liquid Metal cooled Fast Breeder Reactor (LMFBR), flows through primary sodium pump, spherical header and finally enters the grid plate, which acts as an inlet plenum to heat-generating sub-assemblies. The reduction of pressure drop in primary circuit would reduce the operating cost of the system. Spherical header, a vital component in primary circuit suffers major flow non-uniformities resulting in higher pressure drop. Reduction of the pressure drop by uniformization of sodium flow in spherical header is considered. To this end, a 3D computational fluid dynamics (CFD) analysis is carried out to investigate the hydrodynamics of flow, inside the spherical header. Towards this a relationship between pressure loss coefficient and geometrical dimensions, viz., diameter and height of the cone, inside the header are established in order to minimize the pressure drop. The variation of pressure loss coefficient is estimated for header without and with the internals. The static pressure loss coefficient inside the spherical header without an internal cone is found to be 5.59. With a suitable and optimized geometrical dimension of the cone as internal flow guiding device, the loss coefficient is reduced to 0.53. The loss coefficient predicted for the spherical header is validated against the experimental estimates. Consequently, the simplified design of the flow guiding device is achieved by removing baffles.

About the enthalpy of formation of the UZr2 compound

The U-Zr phase diagram has been the subject of many recent studies, particularly for the development of metallic fuels for future fast reactors. The so-called δ-UZr2 phase is part of this renewed interest as it is likely to be formed during irradiation of U-Zr10wt% fuel. Although this phase has been the subject of several structural studies, there is a single experimental determination of its enthalpy of formation by dissolution calorimetry. Most of the Calphad and DFT models of this phase rely on this value to support their validation. In this paper, we propose a reinterpretation of this measurement which tends to show that the original interpretation of the authors was probably wrong. Our reinterpretation leads to a new endothermic estimate of the standard enthalpy of formation of δ-UZr2 which is discussed in relation to the other data of the U-Zr system. The need for new heat capacity and formation enthalpy measurements is highlighted. The present work does not invalidate the most recent CALPHAD descriptions.

Simulation Studies of the Sodium Hydraulic Behavior to Design a Mockup for Eddy Current Flowmeter Testing

Abstract The Eddy current flowmeter sensor plays an important part in the instrumentation of the future sodium fast reactors, which will allow when located above the core to detect plugging (partial or total) of a subassembly and when located on the primary pump to measure core pressure drop and core flowrate. This article describes the predesign phase of a mockup for Eddy current flowmeter tests and qualifications under sodium conditions for the detection of a partial or full plugging of a fuel subassembly. These tests will be the first of their kind, as the Eddy current flowmeter had never been qualified at nominal conditions of a reactor core outlet (sodium temperature, velocity, and output flow).

Demonstration of Aqueous Reprocessing of U-Zr and U-Pu-Zr Metallic Alloy Fuels Using an Ejector Mixer-settler with Tri-n-Butyl Phosphate (TBP) as the Extractant

ABSTRACT U-Zr and U-Pu-Zr metallic alloy fuels have been employed in fast reactors in the early days and these fuels have been considered for use in future fast reactors in India. It is proposed to develop an aqueous method based on the PUREX process for the reprocessing of spent metallic fuels. In the present study, an ejector mixer-settler was employed to demonstrate the feasibility of aqueous reprocessing of U-Zr and U-Pu-Zr metallic alloy fuels with tri-n-Butyl Phosphate (TBP) as the extractant. Continuous counter-current mixer settler runs were separately carried out with U-Zr and U-Pu-Zr feed solutions using 1.1 M TBP/n-DD solvent. Quantitative extraction of U and U + Pu was achieved within five stages leaving zirconium in the raffinate in the case of U-Zr and U-Pu-Zr feed solutions, respectively. Quantitative back extraction of uranium from loaded organic phase was achieved using 0.01 M HNO3 in the case of U-Zr mixer-settler runs. Dual strip solutions such as 2 M and 0.01 M HNO3 were used for the quantitative stripping of U and Pu from the loaded organic phase in the case of U-Pu-Zr mixer-settler runs. The proposed flow sheets demonstrate that the aqueous reprocessing of metallic fuels using a PUREX-based method is feasible.

Design of core catchers for sodium cooled FBRs – Challenges

The whole core meltdown scenario in sodium cooled fast reactors is considered under design extension criteria and has a very low probability of occurrence. To mitigate such a hypothetical severe accident in fast breeder reactors, a core catcher has been provided to accommodate the core debris within the primary containment boundary. In sodium cooled fast breeder reactor (FBR), an in-vessel core catcher, is provided to receive and disperse the fuel debris arising out of core meltdown during Hypothetical Core Disruptive Accident (HCDA). The core catcher prevents hot debris from reaching and damaging the main vessel. It also enables adequate heat transfer by natural convection to keep debris in stable conditions. The present study focuses on design of core catcher with sacrificial barriers and techniques to improve natural cooling of debris on the core catcher for FBRs. Studies are carried out towards the selection of suitable sacrificial material for core catcher, numerical analysis on multilayer and thermal-hydraulic analysis on multi jets core catcher concepts for selection of best suitable technique for debris bed coolability. The core catcher concept with sacrificial barrier and cooling pipes is found to possess enhanced coolability of debris bed. Experimental studies on compatibility of sacrificial ceramic material with sodium indicate strong dependence on microstructure. It has also been found that the multilayer core catcher with MgO as delay bed of 50 mm thickness is adequate to reduce the bottom temperature of core catcher below the design safety limit. Judicious mix of all these findings is expected to have unique design to accommodate whole core meltdown in future fast reactors.

Long-term exposure of MgAl2O4 and Y2O3 thermal barrier coatings in molten sodium

Thermal barrier coatings (TBCs) of magnesium aluminate spinel (MgAl2O4) and yttria (Y2O3) were investigated for long term compatibility with molten sodium for applications in sacrificial core catcher of Future Fast Reactors (FFRs). Thermal barrier ceramic coatings were deposited over austenitic grade stainless steel SS 316LN (flat and cylindrical samples), with an intermediate NiCrAlY bondcoat, using Atmospheric Plasma Spray (APS) technique. The coated samples were exposed to high purity molten sodium at 400 °C under an inert argon atmosphere for 500 h and 1000 h. Sodium exposed samples were subjected to post-cleaning by vacuum distillation followed by chemical cleaning for effective removal of sodium from the specimens. Visual examinations revealed discoloration in the Na exposed samples. Microstructural characterization of Na exposed samples revealed densification in the ceramic topcoats. X-ray diffraction analysis on the topcoats did not reveal any phase change or formation of any corrosion product upon interaction with sodium. Tensile adhesion test on cylindrical samples, as per ASTM C633, revealed ~62% and ~33% drop in the bonding strength after 1000 h of sodium exposure for spinel and yttria coatings, respectively. The mechanical properties were correlated to the microstructural characteristics of the TBC system before and after Na exposure.

Pressure induced renormalization of thermal conductivity of Uranium Nitride: A first principles study

The high thermal conductivity of Uranium nitride (UN) makes it one of the best fuels for future fast neutron reactors. Combining density functional perturbation theory and semi classical Boltzman approach, we obtained the anharmonic properties of phonons and electrons, quantitatively. The electrical and thermal properties such as electrical conductivity (σ), electron thermal conductivity (Ke), lattice thermal conductivity (KL) and Seebeck coefficient (S) under various external hydrostatic pressures have been calculated as an effective controlling probe. We showed that the application of high compressive hydrostatic pressure increases the electrical and thermal conductivity due to the significant coupling within the deformation potential. Furthermore, this enhancement is addressed by the weakly correlated nature of U5f electron that treats as the itinerant carriers. The large contribution of U5f states at Fermi level is accompanied by the Wiedemann-Franz coefficient (L) and reaches to the value of free electron model, exhibits that this model is much applicable at extreme pressures. It is found that Seebeck coefficient depends on the temperature and pressure due to the disparate contribution of electron-phonon coupling at higher temperature. For this complex material, figure of merit is only determined by S2 parameter. Our results suggest that the external pressure can be exploited as a significant probe to control KL/Ke ratio for Uranium Nitride based structures on identifying materials with higher figure of merit efficiency than available at present.

Development of a Selective Americium Separation Process Using H4TPAEN as Water-Soluble Stripping Agent

ABSTRACTRecycling americium from spent nuclear fuel is considered as an important option for a future sustainable nuclear fuel cycle. The PUREX solvent extraction process allows to separate uranium and plutonium from a solution of spent fuel dissolved in nitric acid. In a second step, it would be desirable to separate americium from fission products, and especially lanthanides, but also from curium in order to further transmute americium in future fast neutron reactors. For this purpose, we report the investigation of a new solvent extraction process based on the use of TODGA as an extractant in the organic phase and H4TPAEN as a selective Am(III) stripping agent in the aqueous phase. With this promising system it is possible to co-extract Am, Cm and lanthanides at high acidity and then selectively strip americium from the loaded organic phase at pH around 1 with H4TPAEN. Batch extraction data were acquired to evaluate best conditions to develop a liquid-liquid extraction flow sheet. The influence of several parameters like concentration of ligand, cations and temperature on Am(III)/Cm(III) and Am/lanthanides separation was evaluated. Especially, the ability of H4TPAEN complexing agent to strip macro-concentrations of Am was demonstrated.

Measurement of Th232(n,γ) reaction cross sections in the neutron energy range of 11–19 MeV

The cross sections for the $^{232}\mathrm{Th}(n,\ensuremath{\gamma})$ reaction have been measured in the neutron energy range of 11--19 MeV by using activation and offline $\ensuremath{\gamma}$-ray spectrometric techniques. The neutrons of desired energy were obtained by the $^{7}\mathrm{Li}(p,n)$ reaction using a proton beam of 13--21 MeV from the 14UD BARC-TIFR Pelletron facility at Mumbai, India. The $^{232}\mathrm{Th}(n,\ensuremath{\gamma})$ reaction cross sections were calculated using the computer code talys-1.8. The uncertainties in the measurement have been studied using covariance analysis of the experimental data. The results from the present work have been compared with the evaluated data of ENDF/B-VII.1 and JENDL-4.0 as well as the theoretically calculated values based on talys-1.8, and were found to be in good agreement. The outcome of the present work is important for the development of future fast reactors and accelerator driven subcritical systems (ADSs).

Development of a novel ZrO2 dispersion strengthened 9Cr ferritic steel: Characterization of milled powder and subsequent annealing behavior

A novel ODS ferritic steel with ZrO2 dispersoids is developed as a candidate structural material for future fast reactor applications. Encouraged by our earlier results demonstrating the retention of crystallinity by ZrO2 dispersoids during milling, the 9Cr ODS steel powder has been synthesized under optimized ball milling conditions and subsequently characterized. The analysis of milled powder showed fine dispersions of ZrO2 with a size range of 2–10nm. Upon annealing at 1223K, precipitation of non-stoichiometric FeO was observed along with fine ZrO2 dispersoids. All the constituent phases are identified using Synchrotron XRD, Conventional and HRTEM. An excellent thermal stability has been manifested by the dispersoids as it retained its fine structure during the annealing treatment with ~95% of dispersoids measuring <11nm. Additionally, ~95% of the FeO precipitates are shown to be <16nm in size, indicating itself as a strong precipitation strengthener to the material. This two-fold strengthening by ZrO2 dispersoids and FeO precipitates, along with nanostructured matrix is believed to play the major role in the high-temperature strength of the material being developed and it could lead to the development of a new generation ODS steel.

Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III) using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV) operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR). For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU), though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment configuration, U5-FRs appear to provide the best solution for using uranium, even if the economic impact of this consumption during the first years of operation is significant.

High-precision Measurement of the 238U(n,γ) Cross Section with the Total Absorption Calorimeter (TAC) at n_TOF, CERN

The neutron capture cross section of 238U is fundamental to the design and operation of current reactors and future fast nuclear reactors, and thus must be measured to a high level of accuracy. An experiment has been performed at the CERN n_TOF facility using a 4π Total Absorption Calorimeter (TAC) to measure the capture cross section in the resolved resonance region between 1 eV and 25 keV. A preliminary analysis of the TAC data is presented with particular emphasis to the experimental background in this energy region of interest.

Complete determination of neutron yield from 62 MeV protons on9Be for the design of a low – power ADS

Within the European Partitioning & Transmutation research programs, infrastructures specifically dedicated to the study of fundamental reactor physics of future fast neutron-based reactors are very important. In this respect, an Accelerator Driven System low-power prototype, based on a 70 MeV proton beam impinging on a thick Beryllium converter, was recently proposed and designed within the INFN-E project. The world data on neutron yield from Be target are scarce in this proton energy range. This lack of data calls for a dedicated measurement which was performed at INFN Laboratori Nazionali del Sud, covering a wide angular range, from 0 to 150 degrees, and an almost complete neutron energy interval, from thermal up to the beam energy. In this contribution the results are discussed together with the description of the proposed ADS facility.

Developing Procedures for Fracture Toughness Tests in Lead-bismuth Eutectic

T91 ferritic-martensitic steel is one of the candidate materials to future fast reactors where lead-bismuth eutectic (LBE) is chosen as the coolant. The interaction of the steel with LBE coolant should be investigated by means of mechanical tests and detailed microstructural analysis. Quantitative analysis requires fracture toughness tests in LBE. However tests in LBE are not easily applied with conventional methods due to the conductive nature of LBE. In addition, there are no standards or guidelines for fracture toughness tests in LBE. In this work fracture toughness tests in inert environment and LBE are compared and technical problems encountered during tests in LBE are discussed.

Measurement of Cross-sections of Yttrium (n,xn) Threshold Reactions by Means of Gamma Spectroscopy

Neutron activation and gamma spectrometry are usable also for the determination of cross-sections of different neutron reactions. We have studied the cross-section of yttrium (n,xn) threshold reactions using quasi mono-energetic neutron source based on the reaction on 7Li target at Nuclear Physics Institute of ASCR in Rez. Yttrium (n,xn) threshold reactions are suitable candidates for fast neutron field measurement by activation detectors. Fast neutron field monitoring is necessary already today at a wide range of accelerator facilities and will gain on importance in future fast reactors of generation IV, accelerator transmutation systems or fusion reactors. The knowledge of the cross-sections is crucial for such purpose. Unfortunately, the cross-section is sufficiently known only for 89Y(n,2n)88Y reaction. For higher orders of reactions there are almost no experimental data. Special attention was paid to the 89Y(n,3n)87Y reaction. The nuclei are produced, both in the ground state with half-life 79.8hours and in the isomeric state with half-life 13.38hours. The isomer decays mainly through the gamma transition to the ground state, the beta decay of the excited state is negligible within our accuracy. The cross-sections of both 87Y productions were analyzed.

Role of Microstructure on Fracture Behavior of T91 Steel in Presence of Liquid Sodium

The sensitivity to liquid sodium embrittlement (LME) of T91 martensitic steel, one of the selected structural materials for future sodium fast reactors has been investigated. The study took into account the role of microstructure. Small punch tests and three points bending tests were carried out in a purified and controlled atmosphere. Precipitation state and dislocations structure resulting from a tempering at 550° C provoked LME of the T91 steel between 200 and 550 °C. Secondary Ions Mass Spectroscopy investigation suggested that sodium penetration at prior austenitic grains boundaries promoted by plastic deformation occurred and caused brittle crack initiation. Brittle cracks propagated in sodium preferentially along martensitic laths-boundaries. J integral calculations confirmed a drop in toughness of T91 tempered at 550°C by liquid sodium up to 80 %.