Molten Salt Breeder Reactor Research Articles

Hydrodynamics coupled circulating-fuel reactor equations in comoving frame and their analytical solutions for molten salt reactors

Effects of relative motion between moving fuel and neutrons are usually not considered in the analysis of Circulating Fuel Reactors (CFRs). In this study, we formulate neutron transport equation for CFRs in a hydrodynamic representation in terms of velocities relative to moving fuel. Using the P1 approximation in the comoving transport equation, the diffusion equation for CFRs is obtained. Mass transport of precursors is considered in the formulations. Comoving frame CFRs equations are analytically solved for critical slab problem and a closed form criticality condition is obtained. Comoving representation has introduced corrections to Eulerian cross sections arising from the acceleration term. Hydrodynamics coupling has posed density modifications to cross sections. A parametric study of corrective terms is carried out to calculate effects of these corrections on a generic MSR and on Molten Salt Breeder Reactor (MSBR).

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.

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.

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.

Optimal rule based fuzzy-PI controller for core power control of nuclear reactor

In this paper, an optimal rule based fuzzy proportional integral controller is proposed to control the core power of the molten salt breeder reactor (MSBR) model. Firstly, a nonlinear MSBR model is considered. It is then linearized using the perturbation linearizing approach, and the transfer function of the linearized MSBR system is established. Secondly, an optimal Fuzzy-PI control structure is formulated where a PI controller is set as the principal controller for the MSBR model and a fuzzy control system is used to adjust the PI controller parameters during any disturbance in the MSBR. The rule base system of the fuzzy controller is optimized with a reshaped class topper optimization algorithm. Finally, the optimized Fuzzy-PI controller tests under a variety of conditions. Furthermore, the Nyquist stability approach is used to analyze the stability of the closed loop MSBR model.

Modeling Flow Blockage Accident in Graphite-Moderated Molten Salt Reactors

Fuel channel flow blockage is a postulated accident scenario in graphite-moderated molten salt reactors (MSRs), caused by debris partially obstructing the flow in a hole inside a graphite moderator brick. This obstruction produces an increase in the fuel salt bulk temperatures in the blocked channel, resulting in a significant fraction of the deposited fission power being conducted through graphite from the blocked channel to the surrounding unblocked channels. In the blocked channel, a combination of forced convection and natural convection takes place, the latter being dominant at very strong flow rate reductions. At full blockage (with the fragment completely obstructing the flow), because of the poor thermal conductivity of fuel salt, correct estimation of the natural convection impact on heat transfer between the channel bulk and the walls is fundamental to determine whether boiling conditions are reached in the blocked channel. The proposed assessing approach, developed in collaboration with ThorCon in the frame of the ThorCon reactor core design and optimization, is presented in this work. Results from a case study based on the molten salt breeder reactor (MSBR) design and operating conditions are presented and discussed. These analyses show that with respect to full blockage accidents in other reactor types, the consequences of full channel blockage accidents in graphite-moderated MSRs are milder. For graphite-moderated MSRs, the main mechanical limits during the accident arise in thermal stresses in the graphite brick. Graphite thermal and mechanical properties, namely, bulk modulus, thermal expansion coefficient, ultimate strength, and thermal conductivity, are severely impacted by neutron irradiation through the life of the graphite. All these effects are taken into account in the current analysis and are discussed in the present work.

Improvement of a deterministic fuel management code using artificial neural network for liquid-fueled molten salt reactor

The fuel management code of liquid-fueled molten salt reactor (MSR) relies on the coupling of the neutronics code with depletion code. The deterministic method can improve the efficiency of neutronics calculation. However, updating the microscopic cross-sections in depletion calculation is still time-consuming using Monte Carlo code. A surrogate model based on artificial neural network (ANN) is developed to update the cross-sections online. Then a new fuel management code named ThorNEMFM-ANN was developed by integrating the surrogate model into the deterministic fuel management code. The surrogate model and ThorNEMFM-ANN code were validated with the molten salt breeder reactor (MSBR) and molten salt fast reactor (MSFR) benchmarks. Furthermore, the ThorNEMFM-ANN was applied to the fuel cycle analysis of a molten chloride salt fast reactor (MCSFR) for the demonstration of the applicability. The numerical results indicate that the surrogate model is feasible and effective in the fuel cycle simulation of liquid-fueled MSR.

Optimal membership function based fuzzy proportional–integral controller for power control of molten salt breeder reactor core

The aim of this paper is to design an optimal membership function (MF)-based fuzzy PI (proportional integral) controller to control the core power of a nuclear reactor. The molten salt breeder reactor (MSBR) is extensively utilized for studying primary power management of nuclear reactors. Many challenges are associated with such systems, including mathematical modelling errors, parametric uncertainties, and external disturbances. Moreover, the liquid fuel used in MSBR systems makes it more challenging to design a suitable controller for the core relative power control of such systems. The conventional PI controller may not perform efficiently under such uncertainties in the MSBR system. Different types of perturbations can be handled by a fuzzy controller, if the coordination among fuzzy rules and membership functions of fuzzy variables is done accurately. The idea of this work is to optimize the settings (range and scale) of fuzzy MFs in a fuzzy base controller. The fuzzy controller will not be used directly in the system; rather, it will be used to tune a PI controller for the system where proper expert knowledge may not be available. A nonlinear dynamic acceleration coefficients-based class topper optimization (NDAC-CTO) algorithm is also developed to optimize the fuzzy membership functions. For load tracking with step disruption, the power control of the MSBR core is investigated. With the proposed controller, a significant improvement of 70 to 80 % in the settling time of the power profile of the MSBR system is achieved in comparison with the existing results.

Thorium as an abundant source of nuclear energy and challenges in separation science

Abstract Today about 440 nuclear power plants, with total installed capacity of 390 GW(e) are in operation worldwide generating around 10% of global electricity which are largely fuelled by enriched uranium oxide [Nuclear Power in the World Today. https://world-nuclear.org/information-library/current-and-future-generation/nuclear-power-in-the-world-today.aspx]. Thorium is 3–4 times more abundant than uranium and needs to be exploited by countries with limited stock of uranium. Historically, there have been several attempts to develop Th based reactors, but none has reached commercial scale. In recent years, High Temperature Reactor based on thorium has gained prominence for production of hydrogen with long term goal of complete carbon neutrality. However, unlike natural uranium, which contains ∼0.7% fissile 235U isotope, natural thorium does not contain any ‘fissile’ material and is made up exclusively of the ‘fertile’ 232Th which can be converted to ‘fissile’ 233U, thereby enlarging the fissile material resources. However, there is a need to develop robust closed fuel cycle to address to the challenges of high gamma dose due to the presence of decay products of 232U. It is necessary to gain more experience with promising THOREX process to achieve the desired recovery and D.F. of 233U from the irradiated 232Th. There is also scope to have a close look at some of the alternative extractants and emerging separation techniques for the reprocessing of spent Th based fuels. In view of the distinct advantages of non aqueous reprocessing over aqueous reprocessing, there is need to intensify the efforts to develop the former on a commercial scale. Molten Salt Breeder Reactor (MSBR) is particularly promising in this context. There is a need to investigate Th as energy amplifier under Accelerator Driven Sub-critical System (ADSS) which has potential to burn long lived radio nuclides, considered as a threat to environment over million of years.

Real-Time Inventory Change Detection for the Protactinium Decay Inventory of the Molten Salt Breeder Reactor

This work presents a novel monitoring method for detecting material loss from the decay inventory of the molten salt breeder reactor (MSBR) by monitoring for changes to the system dynamics using an isotopic ratio. The isotopic masses in the decay inventory of a MSBR were simulated under several material loss scenarios. In each case, the ratio of 231Pa to 233Pa served as a sensitive and lasting indicator of material loss. This isotope ratio quickly decreased outside the normal range after a material loss, and the ratio remained depressed for several years after the loss. The dynamics of this ratio were driven by the periodic batch discard from the decay inventory every 220 days, which was specified in the MSBR design to periodically remove fission product buildup. For this method, isotopic ratios were found to be rapid and enduring indicators of inventory change if they comprise a pair with a short half-life (e.g., 233Pa) and a long half-life (e.g., 231Pa) relative to the effective half-life induced by the driving system process (e.g., the batch discard cycle). Using such an isotope pair enabled a method to monitor for changes to the effective half-life of the system and by extension changes to the system inputs and outputs.

Transmutation and Breeding Performance Analysis of Molten Chloride Salt Fast Reactor Using a Fuel Management Code with Nodal Expansion Method

The transmutation of transuranic (TRU) elements produced by pressurized water reactors (PWRs) can effectively reduce their radioactive hazards. The molten chloride salt fast reactor (MCSFR) is a type of liquid-fueled molten salt reactor (MSR) using fuel in the form of molten chloride salts. The MCSFR utilizing a fast neutron spectrum and high actinide fraction is considered to be a potential reactor type for TRU transmutation. An online refueling and reprocessing scenario is the unique feature of liquid-fueled MSRs. On account of this characteristic, a new fuel management code named ThorNEMFM with a nodal expansion method (NEM) was developed and validated with the molten salt breeder reactor (MSBR) and the molten salt fast reactor (MSFR) benchmarks. Then, the transmutation and breeding performances of the MCSFR were simulated and analyzed with the ThorNEMFM code. The MCSFR adopts TRU elements as initial fissile loads and online feeding fissile materials. The results show that the transmutation ratio of TRU elements in the MCSFR can reach 50%, and the breeding ratio can reach 1.359. Moreover, the MCSFR has low radiotoxicity due to lower buildup of fission products (FPs).

Neutronics and fuel cycle modeling of a thorium-fueled two-fluid molten salt iso-breeder reactor

This paper presents reactor physics modeling to quantify behavior of a thorium-fueled molten salt iso-breeder reactor (MSiBR) system similar to the two-fluid molten salt breeder reactor (MSBR) documented in ORNL-4528. A Python framework was developed to allow for a flexible, fully parameterized modeling strategy for analysis of the fuel cycle and core neutronics. Modeling was performed in Serpent-2 to assess the MSiBR design and performance. The framework can be used for direct evaluation of reactor design parameters such as the fuel and blanket salt fraction, lattice geometry, and reactor core size. A full core model of the MSiBR system is presented. Temperature feedback coefficients of reactivity for the fuel salt, blanket salt, and graphite are calculated. It is observed that the present model demonstrates passive and inherent self-regulation of reactivity. The developed framework is also used to demonstrate a methodology for evaluating two-fluid reactors operating with a thorium fuel cycle. This work is intended to provide a baseline for future research into the development of two-fluid MSR designs.

Determination of the uranium elemental concentration in molten salt fuel using laser-induced breakdown spectroscopy with partial least squares-artificial neural network hybrid models

Accurate uranium (U) composition analysis of the molten salt nuclear fuel mixture is a necessary step in nuclear fuel quality control. Laser-induced breakdown spectroscopy (LIBS) is an optical emission spectrometry technique for elemental composition. The non-linear phenomena in LIBS signal are the drawback of this technique, especially for high atomic number (Z) element in a complex matrix. To overcome this limitation, a hybrid chemometrics model of the partial least squares-artificial neural network (PLS-ANN) was used to quantify the U composition in the salt fuel mixture of Molten salt breeder reactor. This work shows the optimisation of several parameters, viz., spectra pre-treatment procedure, factor number of PLS for ANN input, number of hidden neurons and layers in ANN, activation function, and number of ANN iterations. The collective advantages of data dimension reduction from PLS and the nonlinear processing ability from ANN, improved the accuracy of LIBS quantitative analysis for U from ≈7% using simple PLS to ≈4% precision using a hybrid model.

Explicit model predictive controller for power control of molten salt breeder reactor core

The molten salt breeder reactor (MSBR) uses liquid molten salt as fuel. The modeling of the control system for power control of the MSBR core is difficult due to the fuel fluidity. To solve this problem, an explicit model predictive control technique is adopted. First, the nonlinear modeling of the MSBR core is presented. Then, a linearized single-input-single-output model is established from the nonlinear MSBR core model. As the model predictive controller (MPC) has good control effect in tracking problems, an explicit MPC (EMPC) is designed for MSBR to control the power in core. The control of core power in the MSBR is studied for examples of load-following. Results of this study show that the power of MSBR core can be well-controlled by adopting EMPC.

Reactor cell neutron dose for the molten salt breeder reactor conceptual design

The private sector’s recent interest in the active development of molten salt reactors has led to the need to develop and test advanced modeling and simulation tools to analyze various advanced reactor types under numerous conditions. This paper discusses the effort undertaken to model the Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR) design using ORNL’s Shift Monte Carlo code. The MSBR model integrates a Monte Carlo N-Particle (MCNP) MSBR core model with an MCNP model that was generated from a CAD model of the external components and the reactor building, which was subsequently run in Shift. This paper focuses on development of the fully integrated model and its use in performing neutron transport calculations in the reactor cell area. This model is intended to aid in understanding radiological dose conditions during operation, as well as the iron dpa rates in the reactor vessel. The neutron biological dose rates and flux calculated in the reactor cell are much higher in the MSBR than in typical light-water reactors. The implications of these results and future work are also discussed in this paper.

Internal model control based proportional-integral controller with class topper optimization for power control of molten salt breeder reactor core

Internal model control (IMC) method is based on mathematical process model to design the controller. In IMC method, the approximated object model is parallelized with the actual model. The inverse of invertible part of the approximated dynamic model and a low pass filter is considered in the proposed controller design. As an IMC approach has an intuitive design with simple structure and it does not need an accurate model of the object, to explore its application in power control, molten salt breeder reactor (MSBR) core is chosen. A proportional-integral (PI) controller based on IMC technology is designed to control core power of MSBR. The filter constant of IMC technology is optimized with Class topper optimization (CTO) algorithm which improves robustness of the control scheme. Power control of MSBR core is studied for load tracking with step disturbance. This study shows that core power is controlled by CTO based IMC-PI controller.

Neutronic Evaluation of MSBR System Using MCNP Code

The concept of Molten Salt Reactor use Th to breed fissile 233U, where an initial source of fissile material needs to be provided. However, there is no available 233U and so; the fissile fuel supply is one of the unresolved problems. Thus, it is necessary to use existing fissile materials such as 235U or Pu to produce 233U. Current studies analyze the fuel transition from 235U/Th or Pu/Th to 233U/Th and, in this context, the present work evaluates the criticality and the neutron flux of MSBR (Molten Salt Breeder Reactor) considering the fuel: (i) mix of Th and enriched U; (ii) the combination of Th and reprocessed Pu; and (iii) matrix of reprocessed Pu/minor actinides (MAs) and Th. The goal is to verify which of these fuels can be used as initial fissile supply. The MSBR core was simulated by MCNPX 2.6.0 code and the criticality model presents similar behavior of previous studies. The results show that reprocessed fuels could have a potential to be used as initial fissile supply, but these fuels present a neutron flux profile less flattens than traditional 233U/Th. It is possible that a new distribution of fuel elements may improve this profile and future simulations will be performed to evaluate this behavior. The uranium, must has high enrichment value to be used as initial seed. Other studies need be performed to evaluates the uranium enrichment and the U/Th ratio that produces similar core criticality to traditional fuel.

Modeling molybdeneum-99 production in molten salt reactors

Molten salt reactors (MSRs) are being investigated as a possible source of 99Mo for medical uses. This paper describes computational modeling being performed to determine the feasibility for use of several MSR concepts in 99Mo production: the Molten Salt Breeder Reactor (MSBR), the Molten Salt Demonstration Reactor (MSDR), and the Molten Salt Reactor Experiment (MSRE). This modeling is accomplished with SCALE/TRITON using a three-mixture system and estimated removal terms. Results showed that the MSBR, MSDR and MSRE were capable of supplying an estimated 35,800,000, 13,700,000, and 132,000 six-day Ci/week of 99Mo, respectively, which is far more than the global demand. A range of operating parameters has been determined in which each reactor type is expected to have sufficient production capability to meet global demand.

Possible impacts of molten salt reactors on the International Monitoring System

Molten salt reactors (MSRs) are gaining support as many countries look for ways to increase power generation and replace aging nuclear energy production facilities. MSRs have inherently safe designs, are scalable in size, can burn transuranic wastes from traditional solid fuel nuclear reactors, can store excess heat in thermal reservoirs for water desalination, and can be used to produce medical isotopes as part of the real-time liquid-fuel recycling process.The ability to remove 135Xe in real time from the fuel improves the power production in an MSR because 135Xe is the most significant neutron-absorbing isotope generated by nuclear fission. Xenon-135, and other radioactive gases, are removed by sparging the fuel with an inert gas while the liquid fuel is recirculated from the reactor inner core through the heat exchangers.Without effective abatement technologies, large amounts of radioactive gas could be released during the sparging process. This work examines the potential impact of radioxenon releases on samplers used by the International Monitoring System (IMS) to detect nuclear explosions. Atmospheric transport simulations from seven hypothetical MSRs on different continents were used to evaluate the holdup time needed before release of radioxenon so IMS samplers would register few detections. Abatement technologies that retain radioxenon isotopes for at least 120 d before their release will be needed to mitigate the impacts from a molten salt breeder reactor used to replace a nuclear power plant. A holdup time of about 150 d is needed to reduce emissions to the average level of current nuclear power plants.

Th-U Breeding Performances in an Optimized Molten Chloride Salt Fast Reactor

The molten chloride salt fast reactor (MCFR) with Th-U fuel cycle is attracting more and more attention because of its excellent performance, such as high solubility of actinides, superior breeding capacity, low waste production, and high inherent safety. First of all, the breeding capability of an MCFR at equilibrium state was optimized by an in-house automated optimization program. Based on an optimized MCFR, an investigation of Th-U cycle performance was performed. Due to the lack of available 233U in nature, transuranium (TRU), low-enriched uranium (LEU), and Pu were employed as the startup fissile materials, and two different transition scenarios, a breeding and burning (B&B) scenario and a pre-breeding and burning (PBB) scenario were studied. The results show that the breeding performance in an MCFR is attractive and that the Th-U fuel transition can be achieved smoothly by using LEU, Pu, or TRU as startup fissile materials. In the B&B scenario, the average net 233U production rate in the LEU-started, TRU-started, and Pu-started modes during 200-year operation is 374, 321, and 323 kg/a, respectively, at the reprocessing rate of 200 L/day. While in the PB&B scenario the average annual 233U production is about 570 kg for all startup cases at a reprocessing rate of 40 L/day, and the corresponding doubling time is about 9.6 years. Besides, the values of the fuel Doppler coefficient and density coefficient are negative, and the total temperature coefficient of reactivity in all scenarios is below −8 pcm/K. In addition, the value of radiotoxicity of the Th-U cycle in an MCFR is lower than that in other molten salt reactors like the molten-salt breeder reactor and the molten fluoride salt reactor due to the lower buildup of fission products and TRU under its hard spectrum.