Breeder Reactor Research Articles

Macroscopic transport characteristics in packed bed with a multi-physics inversion strategy for solid breeding blanket of HCCB TBM

The solid breeding blanket with typical packed bed structure is a key component for tritium breeding and thermal conversion in future fusion reactors. The pore scale model for multi-physical field solution of solid breeding blanket is fine but very expensive. The simplified packed-bed scale model is more favored in engineering especially for the system-level calculation of the blanket, but it needs to be inputted appropriate macroscopic transport parameters. In this work, the multi-physics calculations of blanket in Chinese HCCB TBM (Helium-Cooled Ceramic-Breeder Test Blanket Module) at pore scale and packed-bed scale were performed. It was found that the packed-bed scale model with traditional macroscopic transport parameters is not accurate enough, especially the prediction deviation of tritium concentration distribution could reach about 25 %. Hence, a novel multi-physics inversion strategy was proposed to predict the macroscopic transport parameters of the packed bed under the multi-field coupling conditions. The inversion results show that the convection of helium significantly enhances the heat and tritium diffusion in the packed bed, and the effective thermal conductivity and effective diffusivity could be increased by 18 % and 46 %, respectively. Then based on the results by multi-physics inversion strategy, the modified effective thermal conductivity and effective diffusivity correlation models for solid breeding blanket packed bed are constructed. With the modified correlation models, the relative prediction deviations of pressure drop, temperature distribution and concentration distribution are within 2.5 %, 0.6 % and 10 % respectively at different inlet velocities (0.05 m/s ∼ 0.25 m/s) and different inlet temperatures (300 K ∼ 900 K). The study in this paper would assist the thermal-hydraulics analysis of the solid breeding blanket in the fusion reactor.

Study on hot isostatic pressing conditions of ARAA for fabrication of the breeding blanket first wall

The helium-cooled concept with solid breeding materials is one of the candidates for demonstration fusion power reactor (DEMO) breeding blanket. The first wall (FW) is a core component of the breeding blanket. The FW should keep the structural integrity during the normal plasma operation which is subject to high surface heat flux, neutron wall load, and high coolant pressure simultaneously. Various fabrication methods have been studied for the development of the FW fabrication technologies. This study focuses on FW fabrication technologies with hot isostatic pressing (HIP) bonding which is considered as the most promising method for achieving such complex shape like FW having internal cooling channels. We investigated the HIP bonding conditions using Advance Reduced Activation Alloy, ARAA, which is a reduced activation ferritic-martensitic steel under development in Korea. Temperature, holding time and pressure were considered as the parameters to optimize the HIP bonding conditions. The microstructures around the HIP joints changed with respect to the HIP bonding temperature. The amount of oxide particles around the HIP joints varied depending on the degassing conditions. The tensile strength of the HIP joint was not greatly affected by the presence of the HIP bonding line and degassing conditions. On the other hand, it has been confirmed that the Charpy absorbed energy of the HIP joint was significantly affected by degassing conditions. An increase in oxide particles around the HIP joints occurred when degassing temperature and vacuum level were low, and as the quantity of oxide particles increased, the Charpy absorbed energy of the HIP joint decreased substantially.

Atmospheric dispersion of radioactive materials for radiological risk assessment in case of hypothetical sodium cooled fast reactor accident

The latest advancements in nuclear power, namely the Generation IV (Gen IV) reactor design, have been successfully chosen in numerous countries. These cutting-edge reactor designs boast superior safety and increased fuel efficiency compared to their predecessors. Among the Gen IV designs is the sodium-cooled fast breeder reactor, and to substantiate the safety assertions, HYSPLIT (Hybrid Single-Particle Lagrangian Integrated Trajectory) dispersion modeling evolved by the National Oceanic and Atmospheric Administration (NOAA) will be employed to scrutinize the potential consequences of an accident. In this context, the analysis focuses on the 1250 MWt/500 MWe sodium-cooled fast breeder reactor (SFR) and its partial fuel melt accident, considering the crucial radioactive reduction factor. Despite existing studies on the consequences of SFR accidents, these studies need to pay more attention to the significance of the radiation reduction factor. This factor is paramount in accurately calculating dose outcomes, as not all radioactive materials can disperse freely within structures, thereby reducing inhalation and cloud shine doses. Furthermore, this investigation underscores the impact of meteorological data and dispersion model input parameters, such as release heights of radioactive materials and measured air concentration heights, on the results of atmospheric dispersion models. Hence, uncertainties in dispersion models arise due to the variability in these factors. The HYSPLIT dispersion model estimates a cumulative dose of 1.65 rem (0.0165 Sv) for a partial fuel melt scenario. The research reveals a maximum dose of approximately 1.8 rem (0.018 Sv) within a 24-hour period, indicating compliance with NRC limitations for Indian fast reactor accident consequences.

Overview of Fourth Generation Reactors Studied at the Universidade Federal de Minas Gerais (Brazil)

This paper presents the results of the studies about the Fourth Generation Reactors developed at DENU-UFMG, such as the Very High Temperature Reactor (VHTR), Molten Salt Breeder Reactor (MSBR), Sodium Fast Reactor (SFR) and Gas-cooled Fast Reactor (GFR). These studies evaluate neutronic parameters of the nuclear system at steady state and during the burnup using computer models developed at DENU-UFMG using SCALE 6.0 (KENO-VI/ORIGENS), MCNPX 2.6.0 and MCNP5. The results show that the adopted methodologies confirm the capabilities of the codes to simulate specific situations in steady state or transient operating conditions.

Evaluation of Reaction Rate Distribution Measurement in the Prototype Fast Reactor Monju

In the prototype fast breeder reactor Monju, reaction rate distributions of the fission reaction rates of 239Pu, 235U, and 238U and the capture reaction rate of 238U were measured using activation foils during its system startup test. The measurements in the core and radial blanket regions were evaluated in detail, and their reliability and usefulness as the validation data for fast reactor neutronics design methodologies were examined through a comparison with calculations. The reaction rate data measured in Monju were confirmed all reliable and useful as the validation data. The fission reactions of 239Pu, 235U, and 238U can be validated with an accuracy of a few percent in the core and blanket regions. The capture reaction of 238U in the core region also can be validated with a similar accuracy, whereas a precise calculation of the foil cross section is necessary to consider the resonance shielding effects of the surrounding fuel pins and a foil.

Numerical Simulation of the Effect of Molten Salt Channel Number on Temperature Distribution in the Molten Salt Core Reactor

A critical aspect in the design of a nuclear reactor that needs to be considered is thermal analysis. This parameter relates to molten salt circulation, which serves as a fuel, and coolant in Molten Salt Reactor (MSR) core reactors. This study simulates the temperature distribution performance of the MSR core by considering the effect of the number of molten salt channels with inlet temperature variations. Knowing the temperature distribution will predict the characteristics of heat transfer in the form of natural convection for the survival of the reactor. The reactor used a Molten Salt Breeder Reactor (MSBR) design developed by Oak Ridge National Laboratory (ORNL). The Ansys software was used to simulate a 3D model under steady-state conditions. The geometry was divided into two regions: solid (graphite material) and fluid (molten salt). The thickness of the moderator and radius of the molten salt channel were 0.0508 and 0.0208 m, respectively. The moderator height was 1.98 m. The number of molten salt channels varied up to four. The results showed a temperature difference of 9 to 15 K when the channel increased from one to four at 839 K inlet temperature.

Neutronic Design Modification of Passive Compact-Molten Salt Reactor

Passive compact molten salt reactor (PCMSR) is a design concept of a molten salt reactor (MSR) currently under development in Universitas Gadjah Mada, Indonesia. It is designed as a thermal breeder reactor using thorium fuel cycle. However, our previous study shows that the original PCMSR design was incorrectly modelled, primarily overestimating its thorium breeding capability. To improve PCMSR neutronic design, we modified the core configuration by the addition of radial fuel channel layers previously nonexistent in original PCMSR core design in various configurations. Neutronic parameters of modified PCMSR geometries in the beginning of life (BOL) were simulated using MCNP6.2 radiation transport code with ENDF/B-VII.0 library. All variations of fuel layer addition show improvement in both temperature coefficient of reactivity (TCR) and breeding ratio (BR), with TCR values became more negative and BR values are larger than unity, ensuring proper breeding capability. Configuration Inner Core-Outer Blanket (IC-OB) achieves the largest BR and lowest doubling time (DT), whilst its TCR is an improvement from the original design. Therefore, IC-OB fuel layer configuration can be applied to redesign the original PCMSR and used in various design optimization scenarios.

Radiochemical and chemical characterization of fuel, salt, and deposit from the electrorefining of irradiated U-6 wt% Zr in hot cells

Abstract Metal fuels are considered as the promising candidates for future fast breeder reactors. Pyro-chemical reprocessing is the ideal method for reprocessing spent metallic fuels due to the inherent process advantages. Electrorefining run was demonstrated in a hot cell facility with irradiated U-6 wt% Zr alloy at 500 °C using LiCl–KCl eutectic melt. In order to understand the behavior of the actinides and various fission products during high-temperature electrolysis, various process streams, viz., irradiated metal alloy fuel, the eutectic salt, and the cathode deposit were analyzed for the uranium, plutonium, and other fission product contents. Various methods employed for characterizing the process streams and the behaviors of some of the fission products during the electrolysis process are highlighted. The major gamma emitting radionuclides present in the irradiated fuel were 106Ru, 125Sb, 134Cs, 137Cs, 144Ce, and 154Eu. During electrorefining, cesium, cerium and europium were oxidized and dissolved in the molten media, whereas ruthenium and antimony remained in the anode basket. A minor contamination of zirconium was found in the cathode deposit.

Understanding the Subsurface Microstructure and Thermal Behavior of Model Oxide Dispersion Strengthened Alloys Through FIB_SEM and TEM

AbstractA potential but an underexplored application of FIB_SEM is its ability to image the subsurface microstructure and capability for an associated chemical analysis. In this article, agglomerated microstructures of two model oxide dispersion strengthened alloy systems, consisting of ZrO2 and Y2O3 dispersions, are evaluated to understand its elevated temperature behaviors. The systems under evaluation are relevant as candidate nuclear structural materials for next-generation fast breeder reactors, and FIB_SEM technique is effectively used along with TEM and EDS for a comprehensive understanding of the material microstructure, and the results are discussed. Distinct microstructures are observed for the two systems with a difference in crystallite size distribution and presence of micron-sized dispersoids in Y2O3. The varied behavior of dispersoids is discussed in terms of its pre-annealing microstructures, and the superiority of ZrO2 over Y2O3 as a dispersoid for ODS alloys is emphasized.

핵융합로 증식 블랑켓 냉각 충돌제트의 유동 및 열전달 특성 연구

핵융합로 증식 블랑켓 냉각 충돌제트의 유동 및 열전달 특성 연구

A comparative study on deep learning models for condition monitoring of advanced reactor piping systems

Advanced nuclear reactors offer innovative applications due to their portability, reliability, resiliency, and high capacity factors. To operate them on a wider scale, reducing maintenance life-cycle costs while ensuring their integrity is essential. Autonomous operations in advanced nuclear reactors using augmented Digital Twin (DT) technology can serve as a cost-effective solution by increasing awareness about the system’s health. A key component of nuclear DT frameworks is the condition monitoring of safety systems, such as piping-equipment systems, which involves acquiring and monitoring the plant’s sensor data. This research proposes a condition monitoring methodology utilizing deep learning algorithms, such as multilayer perceptions (MLP) and convolutional neural networks (CNNs), to detect degradation and its severity in nuclear piping-equipment systems. Sensor signals are processed to obtain the power spectral density and the Short-Time Fourier transform, and feature extraction methodologies are proposed to develop degradation-sensitive data repositories. The performance of MLP, one-dimensional (1D) CNN, and 2D CNN within the proposed condition monitoring framework is compared using a finite element model of a 3D piping system subjected to seismic loads as the application case study. Various approaches, such as dropout, k-Fold validation, regularization, and early stopping of training the network, are investigated to avoid overfitting the models to the input sensor data. The predictive capability and computational capacity of the deep learning algorithms are also compared to detect degradation in the Z-pipe system of the Experimental Breeder Reactor II (EBRII). The Z-pipe system is subjected to harmonic excitations that represent normal operating loads, such as pump-induced vibrations. The findings of the study indicate that the proposed artificial intelligence (AI)-driven condition monitoring framework demonstrates superior prediction accuracies with a 2D CNN, whereas the MLP exhibits higher computational efficiency.

Mechanical and electrical performance of glass fiber reinforced plastic insulation for cryogenic application in fusion magnet irradiated in fast breeder reactor

In a superconducting (SC) magnet fusion reactor, the insulation material used has to withstand significant fluxes of neutrons and neutron-induced gamma radiation. Epoxy-based glass fiber-reinforced plastic (GFRP) composites are the insulation system for fusion magnets and are highly demanding. This insulation system requires excellent mechanical and electrical performance under mixed neutron and gamma radiation at cryogenic temperatures. The insulation material developed consists of boron-free S-glass fiber impregnated with a two-component modified Diglycidyl ether of Bisphenol-A (DGEBA) epoxy resin. An anhydride-cured DGEBA resin system is not suitable for cryogenic temperatures. Therefore, to overcome this thermal stress issue in cryogenic components of SC magnets, a high-toughness, flexible, aromatic amine-cured epoxy resin system was developed for the insulation system. In order to assess the radiation resistance for mechanical and electrical properties, the composite samples were irradiated in the fast breeder test reactor (FBTR) at IGCAR, Kalpakkam, India. The irradiation experiment was carried out for a neutron fluence of 1.0E+21 n/m2 and a gamma dose of 0.4 MGy at a full reactor power of 32 MWt. The tensile strength, interlaminar shear strength, and breakdown strength measurements were investigated before and after irradiation at room temperature and 77 K. The test results indicate that no significant degradation was observed in the properties of the irradiated samples up to the above-mentioned radiation doses.

Evaluation of Specific Wear Coefficient of Sliding Contact Pairs through Tribological Experiments

In fuel handling machine of Prototype Fast Breeder Reactor, colomony-5 is extensively used for hardfacing of rubbing surfaces. To avoid the challenges associated with hardfacing of the components, it is proposed to eliminate hardfacing by using a forged component of a suitable material. 440C steel is a candidate material for this application. Hence, experiments were conducted to evaluate the tribological properties of 440C steel under various conditions seen during the operation. The tests were carried out using pin on disc tribometer by varying the load from 10-20 N and sliding speed from 5-30 mm/s. Friction and specific wear coefficients were evaluated from the experiments. Besides, Global Incremental Wear Model (GIWM) was used for prediction of wear behaviour of the steel.

Performance evaluation and qualification of hard coatings used in absorber rod drive mechanism of Indian fast breeder reactor

Prevention of surface damage at various guiding and bearing locations assumes high importance in ensuring reliable operation of Absorber Rod Drive Mechanisms (ARDM). Contact at an unintended location during any of the operating conditions, including seismic conditions, may cause unacceptable damage, reducing the availability of the ARDMs. Based on international and in-house experiences, the materials and surface coatings for mating surfaces and guiding locations were selected in the ARDMs of Indian Fast Breeder Reactors (FBR).Various mechanical and metallurgical investigations followed by integral assembly tests with full-scale components on hard face coatings were done to evaluate the performance of the hard coatings and to qualify them for reactor operation. In spite of detailed investigations, surface damage-related problems were faced during the initial stages of integral assembly testing, which severely delayed the qualification testing. Key lessons learned from the experimental program are that a holistic approach involving complete kinematics is necessary in the design of the systems to prevent unwanted surface damage-related issues. All these details are discussed in the present paper.From the qualification testing, it can be concluded that the performance of hard coatings for the ARDMs of Indian FBRs is satisfactory. The material and design selections made have formed the basis for the given applications in future FBRs in India.

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.

Validation of Pronghorn’s subchannel code using EBR-II shutdown heat removal tests: SHRT-17 and SHRT-45R

Pronghorn-Subchannel, referred to as Pronghorn-SC throughout this document, is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Pronghorn-SC was initially designed to model flows in water-cooled, square-lattice, bare subassemblies. Its capability has been extended to model flows in liquid-metal-cooled, triangular-lattice, wire-wrapped subassemblies. To ensure the accuracy of Pronghorn-SC in predicting the behavior of liquid sodium-cooled reactors, the code was validated by comparing calculations with experimental data, obtained from the experimental breeder reactor (EBR-II), Shutdown Heat Removal Tests (SHRT) 17 and 45R. The steady-state calculation at the beginning of the transients was validated using temperature measurements taken at different axial elevations in the instrumented subassembly XX09. The validation exercise was performed in successive stages. First, a comparison between the measured temperature profiles and standalone Pronghorn-SC simulations using a uniform pin power profile was made. The pin power profile was then refined using a Serpent-2 model of the reactor core. Finally, the radial temperature profile was further corrected considering the inter-assembly heat transfer. A Pronghorn Finite-Volume (FV) thermal hydraulic simulation of XX09 and its six neighboring subassemblies; calculated the heat flux on the inner duct surface of the XX09 subassembly to inform the Pronghorn-SC model. Last, a transient validation calculated the peak temperature evolution during the SHRT tests.

Thermal conductivity-based hydrogen in argon detector (HAD): Design, development of instrumentation, and performance evaluation.

The Hydrogen in Argon Detectors (HADs) working on the principle of thermal conductivity difference between argon (reference) and argon + H2 (sample) in the cover gas of a sodium-cooled fast breeder reactor using the Wheatstone bridge circuit. The output of HAD is very sensitive to the gas flow rate and to the variation in ambient temperature. The RMSnoise of current was brought within ±1 µA by a constant current source across the bridge. The temperature correction factor fed in the processor brought down the RMSnoise in the HAD signal within ±2.4 mV than the uncorrected one having RMSnoise of ±230 mV. The filtered noise of HAD facilitates enhancing the detection limit of HAD down to 30 ppm (3σ) of hydrogen in argon.

Numerical Simulation of the Effect of Fusion Welding Processes on Residual Stress in Stainless Steel Pipe Weld Joints Used in Fast Reactors

The main structural material used in the building of Fast Breeder Reactors (FBR) is stainless steel of type 316L(N). Finite Element Model (FEM)-based simulations were carried out to understand the effect of various welding processes on the residual stresses induced in 316L(N) stainless steel primary sodium-carrying pipes with an inner diameter of 208 mm and a wall thickness of 5.6 mm. The welding processes considered include multi-pass Tungsten Inert Gas (TIG) welding, A-TIG, laser, hybrid laser-TIG, and Hybrid laser-Metal Inert Gas (MIG). First, single or hybrid heat sources are chosen based on the welding process, and then weld bead profiles obtained during simulation are made to match with that of the actual weld bead profiles obtained during experiments. Then, the optimised heat source is employed for the thermal analysis. The thermal analysis output is then successively linked and provided as input to the mechanical analysis, which uses an isotropic hardening model to predict hoop and axial residual stresses in pipe weld joints. The results show that the A-TIG weld joint exhibited lower hoop stress (165 MPa at inner diameter and -2 MPa at outer diameter) in comparison with that of the other weld joints. The lower residual stress is attributed due to the straight-sided weld bead, lower peak temperature, and slower cooling rate of the A-TIG welding. Therefore, A-TIG welding is recommended among the chosen fusion welding processes for welding of 316L(N) stainless steel primary sodium carrying pipes to minimise the residual stresses and mitigate the premature failure of the pipe weld joints.

Radial basis function neural network controller for power control of molten salt breeder reactor of nuclear power plant

The molten salt breeder reactor (MSBR) basically uses molten salt as fuel. Due to molten salt, it is difficult to control power of the MSBR core. To solve this problem, a neural network control scheme is adopted for the same system. Firstly, a nonlinear MSBR model is taken and after linearizing a single input single output (SISO) system is constructed out of it. Radial basis function neural network (RBFNN) has strong tolerance to input noise, online learning ability and good generalization. Therefore, RBFNN controller is used for controlling the core power of the MSBR. The conventional RBFNN dominates in the neural active region. Compact RBF networks has been designed by using newly developed algorithm. By using Lyapunov stability criterion, it has proved that the variation of core power is bounded and it may be comparable to any traditional neural networks or fuzzy interference system. Results of the study reflect the efficiency of RBFNN controller for controlling power of MSBR core.

A finite volume method based multi-group neutron diffusion model for space-time nuclear reactor kinetics problems in RZ geometry

The transient analysis of Fast Breeder Reactor (FBR) requires the development of high fidelity space-time nuclear reactor kinetics codes for the realistic prediction of spatial and temporal behavior of neutron flux and fission power along with the feedback effects. So, a space-time nuclear reactor kinetics model has been developed for solving the coupled time dependent multi-group neutron diffusion equations and the delayed neutron precursor equations in axi-symmetric cylindrical (r-z) geometry. The numerical model employs an orthogonal mesh finite volume method for spatial discretization and implicit method for time discretization. A computer code named MANDIR-KRZ (Multi-group Analysis of Neutron DIffusion in Reactor for Kinetics problems in RZ geometry) has been developed and employed to reconstruct the keff, transient neutron flux and power results of different benchmark problems. Mathematical formulation of the numerical model and results of the code for Dodds and FBR transient benchmark problems are discussed in this paper.