Fast Reactor Research Articles

A science‐based mixed oxide property model for developing advanced oxide nuclear fuels

AbstractHerein, a science‐based uranium and plutonium mixed oxide (MOX) property model is derived for determining the properties of MOX fuel and analyzing their performance as functions of Pu content, minor‐actinide content, oxygen‐to‐metal ratio, and temperature. The property model is constructed by evaluating the effect of phonons and electronic defects on the heat capacity and thermal conductivity of MOX fuels. The effect of phonons was evaluated based on experimental datasets related to lattice parameters, thermal expansion, and sound speeds. Moreover, the effect of electronic defects was determined by analyzing oxygen‐potential data based on defect chemistry. Furthermore, the model evaluated the effect of the Bredig transition on the thermal properties of MOX fuel by analyzing the irradiation test results. The derived property model is applied to the performance code to analyze fast reactor fuel pins.

Spanish research related to SMRs projects

Small modular reactors (SMRs) are advanced nuclear reactors with a power capacity of up to 300 MW(e) per unit. SMRs encompass a variety of reactor technologies including light water reactors, high temperature gas reactors, molten salt reactors, liquid metal cooled fast reactors, and heat pipe technology-based reactors. The research and design of these diverse SMR types require a broad set of technological capabilities related to nuclear engineering and safety. Within this context, this article attempts to assess the current state of research and technological progress achieved by the Spanish research groups and companies. The results reveal a significant level of maturity among these groups and companies in various domains such as neutronic analysis, thermal hydraulic analysis, the improvement of models related to severe accident scenarios and an active involvement in the design of novel SMR.

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.

Transient behavior of a molten salt fast reactor under two-phase flow conditions with helium bubbling

In the present study, transient behaviors of a chloride molten salt fast reactor without control rods under two-phase flow conditions are analyzed assuming a bypass of the safety protection system. The two-phase flow in the core is caused by helium bubbles injected to remove the gases of the fission product. The void fraction under the rated condition is assumed to be as low as 0.2 %. However, as the fuel flow rate decreases, the void fraction increases inversely proportional to the flow rate and must be considered in any transient analysis. Therefore, analyses with a point reactor kinetics and thermal-hydraulics coupling model have been performed to account for void flow using the CFD code. Comparison between the transient under two-phase flow and the transient under single-phase flow in which the effect of reactivity due to voids is taken into account with the CFD analysis, shows almost the same results, although there are slight differences due to the difference in fuel volume corresponding to the void fraction. When analyzing transients with a system code that has difficulty handling two-component two-phase flow, the void effect has been considered using the above-mentioned single-phase flow analysis. In this way, the entire plant is analyzed using the RELAP5-3D code and the core inlet conditions are passed to the FLUENT code to analyze the transient behavior. The transients analyzed are UTOP, USBO, ULOFF, and ULOHS under the assumption of bypassing the safety protection system. In addition, a transient assuming the loss of helium bubble supply under the decay heat power conditions is analyzed and discussed. The results of the analysis are compared with the results of the same event without voids analyzed using RELAP5-3D. In the case of single-phase flow, the reactor power changes mainly due to temperature feedback. Although the two types of transients have similar appearance, the breakdown of reactivity is quite different. Relating the comparison between the calculated result using the system code and CFD code shows that for the short time behavior until the fuel flow rate shifts to low flow rate, the results of the CFD code, which can analyze the behavior of the helium voids, are more appropriate. However, there are no major problems in the same transient calculation using RELAP5-3D, which calculates the entire plant system.

Improved correlations of the fuel/cladding liquid penetration rate with the out-of-pile transient database

Utilizing data from the Out-of-Pile Transient Database (OPTD), we developed advanced correlations for the cladding liquid penetration rate in metallic fuels used in sodium-cooled fast reactors (SFRs). Based on the available OPTD data, our research highlights the key role of Pu concentration in influencing the liquid penetration behavior, showing different threshold fuel-cladding interaction temperatures, with burnup playing a substantial role, particularly in fuels with a 19% Pu concentration. The penetration rate exhibits a parabolic, diffusion-limited behavior, thereby facilitating predictions of maximum liquid penetration depth over time. These correlations are significantly more accurate, under a variety of conditions, compared to existing single-parameter models, and enhance our understanding of SFR metallic fuel behavior, promoting safer and more efficient reactor operations.

Numerical analysis of liquid sodium spreading on sloped floor surface under pool fire scenario in SFR systems

The accidental leaks of coolant from the heat transport circuit of Sodium-cooled Fast Reactor (SFR) can create fire events due to the reactivity of sodium with atmospheric oxygen, which can pose threat to safe operation of the reactor. The leaked sodium eventually reaches the cell floor, spreads gradually, and forms a burning pool. The thermal consequences of pool fire depend on its surface area, which is determined by the spreading process of sodium on the floor surface. In this context, numerical investigation has been carried out on the spreading behavior of burning sodium pool on the plane/sloped cell floor by developing a 3-D numerical model based on the viscous-gravity current equation of the shallow layer model. The governing equations of the model have been solved numerically by implementing the Finite Volume Method. The model developed has been validated using the experimental results published on liquid pool spreading under burning and non-burning conditions on plane and sloped floors for different leak rates. Using this model, numerical simulations have been carried out on sodium pool spreading under various leak scenarios envisaged in SFR cells based on the specific burning characteristics of sodium pool fire. The time-varying pool size and shape, pool height profile, and maximum pool area have been estimated for a wide range of sodium leak rates and typical floor slopes of interest to nuclear power plants. These predictions contribute towards the numerical evaluation of sodium pool fire scenarios pertaining to the safety of SFR.

MPOGen: A Python package to prepare few-group homogenized cross sections for core calculations by APOLLO3®

Since 2008, CEA has been developing a new generation of multi-purpose code called APOLLO3®. APOLLO3® integrates lattice calculation and core simulations into a single software package designed for Light Water Reactors (LWR) and Fast Reactors analysis. This paper introduces MPOGen, an object-oriented module written in Python3, developed at CEA to automatize the generation of input files for the preparation of few-group homogenized cross section libraries with the new lattice part of APOLLO3®. The module drives APOLLO3® through the computation of few-group homogenized cross sections (Schneider et al., 2016). It significantly expedites the input set-up process, resulting in substantial time savings for engineers.MPOGen still targets market of LWR, as PWR and VVER. The package supports an arbitrary number of parameter space. In this work, six parameters are used to span Cartesian and non-Cartesian grids. MPOGen also supports custom geometries generated by graphical interfaces (GUI), notably Silène and ALAMOS (Tomatis et al., 2022).

Superior corrosion resistance of a slurry FeAl coating on 316LN stainless steel in 550 ℃ liquid lead-bismuth eutectic

An intermetallic Fe-Al coating was successfully produced on 316LN austenite stainless steel by slurry Aluminizing. The coating greatly enhanced the corrosion resistance of substrate in LBE (lead-bismuth eutectic) containing low (10−7 wt%) and high (1.8 ×10−3 wt%) oxygen concentrations at 550 ℃ and the coating was mostly intact after 1500 h exposure test. The high corrosion resistance results from the formation of a uniform layer of alumina oxide on the coating which can be achieved when the dissolved oxygen is above 10−18 wt%. This FeAl coating maybe a promising option for protecting structure material in the fourth LBE-cooled fast reactor.

Analysis method for aerosol deposition behavior under sodium spray fire

Sodium spray fire is a significant containment-related safety concern in sodium-cooled fast reactors (SFR). In a highly-energetic accident such as a core disruptive accident (CDA), the sodium spray fire not only poses a threat to containment integrity but also causes the generation and release of sodium aerosol source terms with strong radioactivity and chemical hazards. In our method, the sodium aerosol generation model, log-normal size distribution model, agglomeration models, and deposition models were applied. In particular, different turbulent energy dissipation rates were considered in the agglomeration models according to the different thermal-hydraulic states of containment and confinement, and the simplified Crump model and Beresnev models have been selected to describe the gravity-diffusion and the thermophoresis deposition behavior respectively, due to the specific conditions of sodium spray fires. Simulation studies have been divided into two categories based on a potential design concept: HTCA, AB5, and AB6 experiments were introduced to describe the drastic thermodynamic state variations caused by spray fires in the confinement, while ATF and AB7 experiments characterized a relatively stable environment similar to the containment. From the simulation results, the analysis method can be considered reasonable with respect to experimental values. Moreover, calculations have been performed for a hypothetical sodium spray fire caused by a rotator plug failure case in a paper-based European SFR.

Numerical analysis of sodium-argon two-phase flow in narrow channels under severe accident conditions in pool type SFR

A two-phase flow model for sodium-argon mixture flow through the reactor vessel head penetrations under the energetic core disruptive accident in a sodium fast reactor (SFR) is presented. The model evaluates the total sodium mass in sodium-argon mixture released to the reactor containment building (RCB) through penetrations. The driving pressure at the penetration entrance is given as input. The two-phase frictional and local pressure drops, inertial and piezo-metric heads are accounted for in the model. Depending on the argon volume fraction, either a homogenous or separated flow approach is adopted to evaluate the mixture flow. Reactor-scale analysis show that the presence of argon in vessel head penetrations can significantly mitigate the total sodium release mass during the slug impact phase. Also, the two-phase frictional pressure drop plays a mitigating role and needs to be included in the model formulation to avoid overestimation of the sodium release mass. Fluid-structure interaction effects such as vessel head bending and hold-down bolt elongation have negligible influence on sodium release mass as long as the deformations are within the elastic limit. The model is conceived to be part of an integrated code system to evaluate the sodium fire and the maximum RCB pressure under severe accident conditions in a pool-type SFR.

Validation of GAMMA+ code for SFR application using FFTF loss-of-flow-without-scram test results

The GAMMA+ code developed by the Korea Atomic Energy Research Institute (KAERI) for the transient analysis of gas-cooled reactors was significantly improved to accommodate the characteristics of sodium-cooled fast reactors (SFRs). In this paper, to validate GAMMA+, the experimental data from loss-of-flow-without-scram (LOFWOS) test #13 in the Fast Flux Test Facility (FFTF) were used as an integral effect test result for the SFR. A combination of one-dimensional modeling and simplified two-dimensional modeling was employed to model the thermal-fluidic behavior of the FFTF, with the latter applied exclusively to the outlet plenum. A point kinetics model and various core reactivity feedback models were coupled with the thermal-fluidic model to analyze the unprotected loss-of-coolant-flow accident in the FFTF. The comparison of the FFTF test data and GAMMA+ simulation results showed good agreement, as the maximum outlet temperature of the hottest Proximity Instrumented Open Test Assembly (PIOTA) Row 2 was calculated with an accuracy of –2.1°C. The two-dimensional modeling improved the accuracy of predicting the sodium temperature in the primary hot leg and enhanced the understanding of the thermal stratification in the outlet plenum. This study showed that GAMMA+ can contribute to the development of safer and more efficient SFRs as a safety analysis tool.

Experimental and numerical assessment of helium bubble lift during natural circulation for passive molten salt fast reactor

To remove insoluble fission products, which could possibly cause reactor instability and significantly reduce heat transfer efficiency from primary system of molten salt reactor, a helium bubbling method is employed into a passive molten salt fast reactor. In this regard, two-phase flow behavior of molten salt and helium bubbles was investigated experimentally because the helium bubbles highly affect the circulation performance of working fluid owing to an additional drag force. As the helium flow rate is controlled, the change of key thermal-hydraulic parameters was analyzed through a two-phase experiment. Simultaneously, to assess the applicability of numerical model for the analysis of two-phase flow behavior, the numerical calculation was performed using the OpenFOAM 9.0 code. The accuracy of the numerical analysis code was evaluated by comparing it with the experimental data. Generally, numerical results showed a good agreement with the experiment. However, at the high helium injection rates, the prediction capability for void fraction of helium bubbles was relatively low. This study suggests that the multiphaseEulerFoam solver in OpenFOAM code is effective for predicting the helium bubbling but there exists a room for further improvement by incorporating the appropriate drag flux model and the population balance equation.

Neutronic evaluation of a small lead-cooled nuclear reactor as an actinides burner

This study employs the Monte Carlo N-Particle code, version 6 (MCNP6), to simulate ELECTRA and investigate its neutronic parameters and the transmutation of spent fuel over a 15-year burnup period, as referenced. The primary objective is to evaluate actinide burning in a low-power fast reactor. The study consists of three phases: 1) validating Model 1 for MCNP6 code with results obtained by SERPENT code from previous studies, 2) conducting neutronic analysis of three ELECTRA models (1, 2, and 3) with varying levels of detail, and 3) assessing burnup using the most detailed ELECTRA MCNP6 model (Models 3). The neutronic calculations obtained from the MCNP6 code align with previous studies. Notably, including all core components in the simulation yields significant disparities in criticality and neutron flux compared to a simpler model. Despite its small size and low power, ELECTRA, as a research reactor, exhibits essential characteristics for effectively burning plutonium from Light Water Reactor (LWR) spent fuel. Therefore, ELECTRA contributes to developing LFRs and supports the GIF’s research goals for reactor development.

Preliminary safety analysis of the TRANSMUTEX sub-critical reactor using the GeN-Foam multi-physics solver

This work presents a preliminary safety analysis of the accelerator-driven system under development at TRANSMUTEX. It is a sub-critical lead-cooled fast reactor with a nominal power of 300 MWth having the following features: (1) an inert-matrix fuel; (2) a keff = 0.974; (3) a heavy metal spallation target; and (4) a 4 MW proton accelerator. The safety analyses are based on the GeN-Foam multi-physics code, whose point-kinetics sub-solver has been extended for the analysis of sub-critical reactor configurations. In the paper, we first present the extension of the code and its verification against analytical results. Then, we present the results of our preliminary safety analysis, including in particular the simulation of various unprotected (double-fault) accidental transients.

Study on bubble dynamics in sodium using three-dimensional MPS method

Sodium-cooled fast reactor (SFR) is one of the most promising Generation IV reactors. Design and safety evaluation of a SFR requires sufficient knowledge about gas bubble behavior in sodium pool. It is challenging to perform numerical study on bubble dynamics in sodium pool due to the discontinuity caused by the differences of physical properties. In this study, a three-dimensional (3D) Moving Particle Semi-implicit (MPS) method is presented featuring high-order accurate scheme and Riemann-based stable multiphase model. Numerical discontinuity caused by large density, large viscosity ratio and surface tension can be handled by the proposed method. To verify the proposed method, two cases from the bubble diagram of Grace (1973) is simulated and rising velocities and shapes of rising gas bubbles are compared, both are agreeing very well. Then co-axial coalescence of gas bubbles are calculated and compared with existing experimental and numerical results. Finally bubble rising behavior in sodium is analyzed using the developed 3D MPS method.

Heat transfer in the argon space of pool-type sodium-cooled fast reactor: Numerical simulation and experimental validation

Liquid sodium serves as the coolant in the pool-type sodium-cooled fast reactor. The primary penetration equipment on the primary side is immersed in a sodium pool. Argon gas is employed as the covering gas above liquid sodium. To determine the temperature distribution of the roof slab of the main vessel for strength analysis, it is necessary to analyze the heat transfer characteristics. Owing to the large volume of the original reactor, it is difficult to establish a full-scale experimental model. Consequently, a scaled-down experiment is required to obtain the temperature distribution of the roof slab and argon space. In addition, the complex heat transfer characteristics within the roof slab and argon space make it difficult to maintain similarities between the experimental and original facilities. Numerical simulations are a useful approach for extrapolating an experiment to the original reactor. In this study, a scaled-down heat transfer experimental facility for an argon space (HEFA) was established based on the similarity principle, and a numerical model was set up according to the HEFA. The validity of the numerical methods was confirmed by comparing the experimental and numerical results for the HEFA. Based on similarity theory, the temperature distribution in the HEFA was extrapolated to that in the original reactor. A numerical model was established based on the original reactor. By comparing the extrapolated temperature distribution and numerical distribution of the original reactor, it was found that they were in good agreement. Therefore, the validity of the extrapolated temperature distribution was confirmed.

Contributions of MYRRHA to the European Fusion Energy Roadmap

One of the main objectives of the EUROfusion Roadmap to Fusion Energy during Horizon Europe is to finalize the conceptual design (thus preparing the engineering design phase) and initiate the construction of a fusion spectrum neutron source (IFMIF-DONES). The Belgian Nuclear Research Center is currently developing the MYRRHA research facilities, which include a subcritical accelerator-driven system ignited by a 600 MeV superconducting linear proton accelerator coupled to a 100 MeV proton injector. The implementation of the multipurpose fast spectrum reactor MYRRHA will proceed in phases, with phase I, covering 2019–2027, including the deployment of the 100 MeV proton accelerator and associated fusion material research facilities. The research on fusion materials on the basis of the 100 MeV proton beam and with hot cell facilities will focus on innovative materials and composites and will be fully complementary to IFMIF-DONES as well to the planned research program on using currently available nuclearized facilities such as BR2 (Belgian Research Reactor 2) and Hot Cells for mechanical, microstructural and chemical analysis. A bespoke fusion irradiation module is also foreseen on the MYRRHA reactor, the construction of which is planned in the third phase of the implementation of the project (i.e. after 2035). Provisions have been made in the core of the MYRRHA reactor to include so-called In-Pile Sections around the central core or at the central core position below the spallation target when operating in subcritical mode. In this contribution, we review the role and technical capabilities that MYRRHA facilities may provide in the framework of the European Fusion Electricity Roadmap and specifically for DEMO Engineering Design, Material's Development and enhancement of Technology Readiness Level. Aspects of education and training, being an integral part of the European nuclear fusion program, are also touched upon. Finally, the synchronization of the timelines for the implementation of MYRRHA and the EU fusion facilities are discussed.

Flexible operation and maintenance optimization of aging cyber-physical energy systems by deep reinforcement learning

Cyber-Physical Energy Systems (CPESs) integrate cyber and hardware components to ensure a reliable and safe physical power production and supply. Renewable Energy Sources (RESs) add uncertainty to energy demand that can be dealt with flexible operation (e.g., load-following) of CPES; at the same time, scenarios that could result in severe consequences due to both component stochastic failures and aging of the cyber system of CPES (commonly overlooked) must be accounted for Operation & Maintenance (O&M) planning. In this paper, we make use of Deep Reinforcement Learning (DRL) to search for the optimal O&M strategy that, not only considers the actual system hardware components health conditions and their Remaining Useful Life (RUL), but also the possible accident scenarios caused by the failures and the aging of the hardware and the cyber components, respectively. The novelty of the work lies in embedding the cyber aging model into the CPES model of production planning and failure process; this model is used to help the RL agent, trained with Proximal Policy Optimization (PPO) and Imitation Learning (IL), finding the proper rejuvenation timing for the cyber system accounting for the uncertainty of the cyber system aging process. An application is provided, with regards to the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED).

A multi-index evaluation method of supercritical CO2 Brayton cycle for nuclear power plants design

ABSTRACT This paper investigated four different S-CO2 Brayton cycle layouts for nuclear energy conversion: simple recuperation cycle (SR), recompression cycle (RC), re-heating cycle (RH), and intercooling cycle (IC). We compared these S-CO2 Brayton cycle schemes for nuclear energy conversion using the G1+TOPSIS multi-index evaluation method. The evaluation results of different schemes based on their safety, thermodynamics, techno-economic and compactness are given. The results show that for Generation IV reactors with the same designed thermal power, the thermodynamic performance of the S-CO2 system is better for higher reactor exit temperature. Among the schemes, the gas-cooled fast reactor (GFR)+RC scheme has the highest thermal efficiency (47.4%) and exergy efficiency (56.48%). The GFR+IC scheme has the lowest specific cost (1861.3$/W) and the internal rate of return (24.8%). The re-heating cycle (RH) has worse indexes, but it requires the lowest initial investment cost. The intercooling cycle (IC) has the lowest levelized cost of electricity (0.0134 $/KW•h) coupling to GFR. Considering all indexes of four aspects, the reactor’s performance ranking is MSR>LFR>SFR>GFR, and the S-CO2 system's performance ranking is RC>SR>IC>RH. For Generation IV nuclear energy conversion technologies, the molten salt reactor (MSR)+RC scheme should be given priority, while GFR+RH schemes should be carefully considered.

Investigations of Suction Recirculation in Primary Sodium Pump of Future Indian Fast Breeder Reactors

The design of the next generation of fast breeder reactors has commenced, with the main targets being enhanced safety and improved economy. Nuclear heat generated in the fuel subassembly of fast reactors is removed by circulating sodium through the core using centrifugal pumps. The primary sodium pumps (PSPs) used are large-capacity pumps, and the design of these pumps is different from that of traditional pumps. Though many works have been reported for the performance prediction of centrifugal pumps, most of these works have been carried out in a decoupled way, and only a few works have been reported where the pump is modeled with all the associated geometric structures. Centrifugal pumps are prone to a phenomenon called suction recirculation, which occurs when pumps are operated significantly below the best efficiency point. This suction recirculation has a strong potential to damage the impeller. Correlations given in the literature for the prediction of the onset of recirculation cannot be used for complicated inlet geometries, and three-dimensional computation fluid dynamics (CFD) investigations are most suited for such applications. Many devices have been reported in the literature to reduce the intensity of (or to suppress) suction recirculation. Webs provided in the suction plenum will modify the velocity distribution at the impeller inlet and also can influence suction recirculation. In this work, the centrifugal pump used for primary sodium pumping for fast reactor applications is simulated using CFD techniques in an integrated way. The frozen rotor approach is used to simulate the impeller-diffuser hydraulics. The effect of flow hydraulics in the suction plenum, flow distribution in the standpipe–pump gap, and flow conditions in the pool on the performance characteristics of PSPs are simulated. The flow rate for the onset of suction recirculation is predicted and compared with correlations available in the literature. Simulations are carried out to study the effect of webs on suction recirculation. The effects of the number of webs and the web geometry are also studied.