Fuel Salt Flow Research Articles

Numerical analysis of dynamic load following response in a natural circulation molten salt power reactor system

The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. An investigation into the power transient behavior of an autonomous load-following, closed-loop, natural circulation MSR system is important to quantifying operational and safety performance under dynamic conditions. This paper presents the results of a STAR-CCM+ and a comparatively simple asymmetric, one-dimensional, numerical model to solve the compound dynamic MSR power behavior subject to flow and temperature reactivity feedback only. Results show that reactor power is affected by fuel salt flow velocity (global) and temperatures (local bulk volume averages) in a coupled, time-delayed manner that results in a unique compound dynamic, closed-loop power feedback mechanism. The 1-D simulation approach opens the possibility of performing inexpensive computations to evaluate time-dependent reactor performance relative to thermo-physical fuel salt limitations. Results also potentially motivate a convincing conclusion that natural circulation MSRs provide a leap in safety and reliability.

Development and verification of a three-dimensional spatial dynamics code for molten salt reactors

Molten Salt Reactor (MSR) has the characteristics of drift effect of delayed neutron precursors (DNP), direct heating of coolant and moderator cooled by fuel salt under most conditions, which bring significant differences from traditional reactors in dynamics calculation. To deal with the unique dynamics characteristics of MSR, a specific three-dimensional spatial dynamics code – TMSR3D has been developed, in which the fourth order standard nodal expansion method and the exponential transformation method are implemented to solve the neutron diffusion equations and the neutron kinetics equations respectively, the parallel multi-channel thermal hydraulics (TH) model is adopted to take into account the feedbacks of local temperature to group constants, and a convection term is introduced to the traditional balance equations for DNP to consider the influence of fuel salt flow. Besides, the least square fitting method is employed to parameterize the few-group homogenized parameters generated by the lattice code DRAGON5. The 3D-LRA core benchmark is first used to verify TMSR3D for solid-fueled reactor problem. Afterwards, the experimental data of the Molten Salt Reactor Experiment (MSRE) collected in the European project MOST are simulated, which includes the transients perturbed by protected fuel pump start-up and coast-down at zero power, and the natural circulation experiment. The numerical results indicate that the developed code is capable of providing a reasonable description for the neutronics-TH coupling phenomena of graphite-moderated channel-type MSRs.

Three dimensional steady‐state neutronics/thermal‐hydraulics coupled simulation for a molten salt reactor moderated by zirconium hydride rods

Zirconium hydride (ZrH) is considered as an alternative moderator in the molten salt reactor (MSR) due to its excellent slowing-down ability, good thermal stability, and high radiation resistance. In the MSR, ZrH moderator is inserted in the form of rod into the fuel salt of the reactor core, and an axial and a transverse fuel salt flow occur between adjacent fuel salt channels. This inter channel flow mixing effect of fuel salt makes the molten salt reactor moderated by zirconium hydride rods (ZrH-MSR) substantially different from other type of reactors in both neutronics and thermal hydraulics. Based on the multi-group neutron diffusion theory and the subchannel thermal-hydraulics model, a steady-state analysis code for ZrH-MSR was developed and verified in this work. Then, a 25 MWth ZrH-MSR proposed by Transatomic Power Corporation was selected to evaluate its steady-state characteristics due to its cost-effective and simplified design. By applying this coupled code, the effects of control rod position, inlet fuel salt flow velocity, and residence time of fuel salt in the external loop on the distributions of the neutron fluxes, delayed neutron precursors (DNPs), and temperature for the 25 MWth ZrH-MSR were analyzed in detail. The simulation results show that the control rod position influences the neutron fluxes greatly, which indirectly affects the distributions of DNPs and core temperature. The inlet fuel salt flow velocity and the residence time of fuel salt in the external loop have little influence on the neutron fluxes of the core but affect the distribution of DNPs significantly. These numerical results can provide valuable information for the research and design of ZrH-MSR.

INFLUENCE OF FUEL FLOW RATE VARIATION ON MOLTEN SALT REACTOR PERFORMANCE

Molten Salt Reactors are Gen-IV reactors that use liquid fuel. Fluid fuel allows continuous removal of fission gases as well as batch fuel reprocessing. With these control mechanisms the system can be sustained within the desired operating temperature range and required power output. These methods rely on the presence of a chemical processing plant on-site that adds complexity. This also creates a risk of processing plant unavailability due to faults, emergency downtime or maintenance. The work considers variation of fuel salt flow rate in Molten Salt Reactors as a means of controlling reactor operation without using reprocessing. The analysis is performed using the Molten Salt Fast Reactor as an example. An extended version of the SERPENT Monte-Carlo transport code coupled with OpenFOAM generic platform were used for capturing delayed neutron drift, decay heat, gaseous fission product removal, calculating fuel salt velocity vectors and the fuel temperature distribution. The two models were coupled via a script that accounted for reactivity insertion between time steps and the changes caused in the fission power. Results confirm that, while operating at constant power, the difference between fuel inlet and outlet temperatures increase as the flow rate decreases. Burnup analysis has shown that while the average fuel temperature continues to reduce with time, the difference between inlet and outlet temperatures can be controlled by varying the flow rate while maintaining constant power. Finally, the variation in the fuel flow rate has been shown to extend the reactor operating time with no insertion of additional fissile inventory.

Development of a steady state analysis code for molten salt reactor based on nodal expansion method

The unique feature of molten salt reactor (MSR) is that heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture which acts as both fuel and coolant. The fission energy from fissile fuel is released immediately into the fuel salt in the core, and the delayed neutron precursors (DNP) have a loss due to fuel circulation in the primary loop, which both make MSR be characterized by the strong interplay between the neutronics and thermal hydraulics (TH) and imply that the software developed for traditional solid-fueled reactors is not applicable to MSR. In this study, a new coupled neutronics and TH code named TMSR3D is developed for evaluating the steady-state characteristics of liquid-fueled MSR. In this code, the fourth order standard nodal expansion method (NEM) is implemented to solve the neutron diffusion equations, and the parallel multi-channel model is adopted to take into account the feedbacks of local temperature to group constants. Besides, a convection term is introduced into the traditional balance equations for DNP to consider the influence of fuel salt flow. Some experiment data of Molten Salt Reactor Experiment (MSRE) are chosen to verify the code, and the core characteristics including the distributions of neutron fluxes, DNP and temperature under the rated condition are simulated and discussed. Furthermore, the impacts of the inlet mass flow rate, inlet fuel temperature and the fuel salt residence time out of the core are analyzed in detail. The numerical results indicate that the TMSR3D code can provide a reliable description of the neutronics-TH coupling phenomena and of the steady state analysis for graphite-moderated channel-type MSRs.

A Review of Circulating Voids and Gaseous Fission Product Behavior in Molten Salt Reactors

Abstract Molten salt reactors (MSRs) are a family of advanced, generation-IV, nuclear reactors that use a circulating molten alkali fluoride fuel salt as both a primary working fluid and a fuel matrix. Since the fuel salt is liquid, gases become entrained in the fuel salt flow, and these entrained gases are called circulating voids. This paper reviews some aspects of circulating voids on reactor behavior. A formal definition of the circulating void fraction is given. A time-line of Oak Ridge reports on circulating void behavior is presented. Methods to measure the circulating void fraction are described. Foam induced by gas entrained in the fuel salt is discussed. The discovery of circulating voids in the molten salt reactor experiment (MSRE) is detailed. Aspects related to bubble size and interfacial area are reviewed. The Laplace pressure is examined. The void and pressure coefficients of reactivity are examined. Some requirements for a bubble model for xenon behavior analysis are proposed. The effects of bubble surface mobility are examined. Finally, sources of gas production in the fuel salt are investigated.

Multiphysics analysis of the MSFR helium bubbling system: A comparison between neutron diffusion, SP3 neutron transport and Monte Carlo approaches

The Molten Salt Fast Reactor is a fast-spectrum molten salt reactor under development in the framework of the European H2020 SAMOFAR Project (http://samofar.eu/). Among the design peculiarities, this circulating fuel reactor features a helium bubbling system aimed at removing on-line gaseous fission products, and metallic particles as well. From a modelling point of view, the presence of helium bubbles in the core needs to be assessed both from a neutronics and thermal-hydraulics point of view. In this paper, the attention is paid to the first aspect, analysing the void reactivity effect induced by the bubbles by means of a Monte Carlo, an SP3 neutron transport and a neutron diffusion approach. Since the distribution of helium bubbles is not uniform in the core but strongly depends on the fuel salt flow conditions, as well as on the location of their injection and extraction, a coupling scheme between Serpent 2 and OpenFOAM codes is adopted for the Monte Carlo analysis. In this way, the Monte Carlo code Serpent 2 is provided with a real bubble distribution calculated by means of a two-phase Euler-Euler solver implemented in OpenFOAM. The outcome puts in evidence the difference arising when a uniform and a real distribution of bubbles is considered in a Monte Carlo calculation. These results are also compared with an SP3 neutron transport and a neutron diffusion solver implemented in OpenFOAM, highlighting the difference among the three neutronics approaches herein adopted.

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in steady state condition

MSR (molten salt reactor) uses liquid molten salt as the coolant and fuel solvent, making it the only liquid reactor among the six generation IV reactor types. As a liquid reactor the physical properties of the reactor are significantly influenced by the fuel salt flow therefore conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code and applied it to investigate the thermal-hydraulic characteristics of the core in steady state condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The temperature distribution, neutron fluxes and delayed neutron precursors distribution of the core in steady state conditions was studied, and the result analyzed when inlet temperature and velocity were changed. From simulation it was found that the inlet temperature has little influence to neutron distribution however inlet velocity affects the delayed neutron distribution in steady state condition. The results provide some valuable information in design and research of this kind of reactor.

Modelling and analysis of the MSFR transient behaviour

Molten Salt Reactors (MSRs) were conceived at the early stages of nuclear energy in view of the favourable features fostered by a liquid fuel. They were developed as graphite-moderated thorium-fuelled breeder reactors till the seventies, when studies on this reactor concept were mostly abandoned in favour of the liquid–metal fast breeder concepts. A decade ago, the MSRs were included among the six GEN-IV systems and a core optimization process allowing for the GEN-IV main objectives led toward a fast-spectrum MSR concept (MSFR – Molten Salt Fast Reactor). Albeit advantageous in terms of U-233 breeding and/or radio-active waste burning, the new concept lacks the notable know-how available for the thermal-spectrum MSR technology. The present paper preliminarily investigates the MSR dynamics, based on the conceptual MSFR design currently available. A primary objective is to benchmark two different models of the MSFR primary circuit, both of them including a detailed and fully-coupled (node-wise) representation of turbulent fuel-salt flow, neutron diffusion, and delayed-neutron precursor diffusion and convection. A good agreement is generally observed between the adopted models, though some discrepancies exist in the temperature-field, with ensuing mild impacts on the reactor dynamics. The performed analyses are also used for a preliminary characterization of the MSFR steady-state and accidental transient response. Some points of enhancement needed in the MSFR conceptual design are identified, mainly related to in-core velocity and temperature fields. The reactor response following major accidental transient initiators suggests a generally benign behaviour of this reactor concept.

Numerical analysis for a molten salt reactor in the presence of localized perturbations

Molten salt reactors (MSRs) have seen a marked resurgence of interest over the past few decades, highlighted by their inclusion as one of the six Generation IV reactor types. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this paper, the attention is focused on the behaviors of an MSR in the presence of localized perturbations caused by fissile precipitates and gas bubbles. A neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors, and the group constants dependent on the temperature are calculated by the code DRAGON. In addition, the k-epsilon turbulent model is adopted to establish the flow and heat transfer. The thermo-hydraulic and neutronic models are coupled through the temperature, heat source and velocity. The effects of the localized perturbation on the distributions of power, temperature, neutron fluxes and delayed neutron precursors are obtained and discussed in detail. The results provide some valuable information for the research and design of this new generation reactor.

Development of a steady state analysis code for a molten salt reactor

The molten salt reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for transmutation, and production of electricity, hydrogen and fissile fuels. In this study, a single-liquid-fueled MSR is designed for conceptual research, in which no solid material is present in the core as moderator, except for the external reflector. The fuel salt flow makes the MSR neutronics different from that of conventional reactors using solid fuels, and couples the flow and heat transfer strongly. Therefore, it is necessary to study the core characteristics with due attention to the coupling among flow, heat transfer and neutronics. The standard turbulent model is adopted to establish the flow and heat transfer model, while the diffusion theory is used for the neutronics model, which consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six groups of delayed neutron precursors. These two models which are coupled through the temperature and heat source are coded in a microcomputer program. The distributions of the velocity, temperature, neutron fluxes, and delayed neutron precursors under the rated condition are obtained. In addition, the effects of the inflow temperature, inflow velocity, and the fuel salt residence time out of the core are discussed in detail. The results provide some valuable information for the research and design of the new generation molten salt reactors.

Analysis on the neutron kinetics for a molten salt reactor

The neutron kinetics of the molten salt reactor is significantly influenced by the fuel salt flow, which leads to the analysis methods for the conventional reactors using solid fuels not being applicable for the molten salt reactors. In this study, a neutron kinetic model considering the fuel salt flow is established based on the neutron diffusion theory, which consists of two-group neutron diffusion equations for the fast and thermal neutron fluxes and six-group balance equations for delayed neutron precursors. The temperature feedback on the neutron kinetics is considered by introducing a heat transfer model in the core, in which the group constants which are dependent on the temperature are calculated by the code DRAGON. The mathematical equations are discretized and numerically calculated by developing a code, in which the fully implicit scheme is adopted for the time-dependent terms, and the power law scheme is for the convection–diffusion terms. The neutron kinetics is conducted during three transient conditions including the rods drop transient, the pump coastdown transient and the inlet temperature drop transient. The relative power changes and the distributions of the temperature, neutron fluxes and delayed neutron precursors under these three different transient conditions are obtained in the study. The results provide some valuable information for the research and design of this new generation reactor.

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt**Supported by National Nature Science Foundation of China (10575079)

The Molten Salt Reactor (MSR), one of the `Generation IV' concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition.

Transient characteristics of small molten salt reactor during blockage accident

AbstractThis paper reports the results from a transient core analysis of a small molten salt reactor (MSR) when a duct blockage accident occurred. The focus of this study is a numerical model employed in order to consider the interaction among fuel salt flow, heat transfer, and nuclear reactions. The numerical model comprises continuity and momentum conservation equations for fuel salt flow, two‐group neutron diffusion equations for fast and thermal neutron fluxes, transport equations for six‐group delayed neutron precursors, and energy conservation equations for fuel salt and graphite moderators. The analysis results show the following: (1) the effect of the self‐control performance of the MSR on the effective multiplication factor and thermal power output of the reactor after the blockage accident is insignificant, (2) fuel salt and graphite moderator temperatures increase drastically but locally at the blockage area and its surroundings, (3) the highest fuel salt temperature after the blockage accident is 1,363 K; this value is lower than the boiling point of fuel salt and the melting temperature of the reactor vessel, (4) the change in the distributions of fast and thermal neutron fluxes after the blockage accident when compared with the distributions at the rated condition is very slight, and (5) delayed neutron precursors, especially the first delayed neutron precursor, accumulate at the blockage area due to its large decay constant. These results imply that the safety of the MSR is assured in the case of a blockage accident. © 2006 Wiley Periodicals, Inc. Heat Trans Asian Res, 35(6): 434–450, 2006; Published online in Wiley InterScience (www.interscience.wiley.com). DOI 10.1002/htj.20123

Steady State Analysis of Small Molten Salt Reactor (Effect of Fuel Salt Flow on Reactor Characteristics)

The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thorium-uranium fuel cycle. The feature of the MSR is that fuel salt flows inside the reactor during the nuclear fission reaction. In the previous study, the authors developed numerical model with which to simulate the effects of fuel salt flow on the reactor characteristics. In this study, we apply the model to the steady-state analysis of a small MSR system and estimate the effects of fuel flow. The model consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, transport equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and the graphite moderator. The following results are obtained: (1) in the rated operation condition, the peaks of the neutron fluxes slightly move toward the bottom from the center of the reactor and the delayed neutron precursors are significantly carried by the fuel salt flow, and (2) the extension of residence time in the external-loop system and the rise of the fuel inflow temperature show weak negative reactivity effects, which decrease the neutron multiplication factor of the small MSR system.

Transient Response of Small Molten Salt Reactor at Duct Blockage Accident

This paper performed transient core analysis of a small Molten Salt Reactor (MSR) at the time of a duct blockage accident. The numerical model employed in this study consists of continuity and momentum conservation equations for fuel salt flow, two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The analysis shows that (1) the effective multiplication factor and reactor power after the blockage accident hardly change because of the self-control performance of the MSR, (2) fuel salt and graphite moderator temperatures rise at the blockage point and its vicinity, drastically but locally, (3) the highest temperature after the blockage accident is 1 363 K, very lower than the boiling point of fuel salt and melt point of reactor vessel, (4) fast and thermal neutron fluxes distributions after the blockage accident hardly change, and (5) delayed neutron precursors accumulate at the blockage point, especially 1st delayed neutron precursor due to is large decay constant. These results lead that the safety of MSR is assured in the blockage accident.

Transient Core Characteristics of Small Molten Salt Reactor Coupling Problem between Heat Transfer/flow and Nuclear Fission Reaction

This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k0, putting the fission reaction rate back on the rated state.

小型溶融塩炉の非定常炉心特性 : 閉塞事故時の核分裂反応-燃料塩流動の相互干渉(オーガナイズドセッション2 高温エネルギー変換)

This paper performed the transient core analysis of a small molten salt reactor in the case of blockage accident. The emphasis is that the numerical model developed in this paper takes into account the interaction between fission reaction and fuel salt flow. The model consists of two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis in the case of blockade accident are as follows ; (1) neutron multiplication factor hardly changes, (2) the outlet temperature of fuel salt decreases 10 K, and (3) fuel salt and graphite temperatures largely increase at the blockage point, but lower than fuel salt boiling temperature and the molten temperature of the reactor vessel.

1157.流れと温度分布を考慮した溶融塩炉の定常炉心特性

The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thoriumuranium fuel cycle. The notable feature of the MSR is that fuel salt flows the inside of the reactor accompanying nuclear fission reaction. In the previous study, the authors had developed numerical model to simulate the effects of the fuel salt flow on reactor characteristics and estimated the basic effects in a simplified system. This paper applies the model to the steady state analysis of the MSR system and simulates the effects of not only fuel flow but also fuel salt inflow temperature and residence time in the 1st loop system. The model consists of two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The following results are obtained: (1) the fuel salt flow affects the distributions of the delayed neutron precursors, especially long-lived one, and (2) the extension of residence time in the 1st loop system and the rise of fuel inflow temperature shows negative reactivity effects, decreasing neutron multiplication factor.