Parametric Lattice Study of a Graphite-Moderated Molten Salt Reactor

2014 ◽  
Author(s):  
Boris A. Hombourger ◽  
Jiři Křepel ◽  
Konstantin Mikityuk ◽  
Andreas Pautz
Keyword(s):  
Molten Salt ◽  
Unit Cell ◽  
Fuel Composition ◽  
Molten Salt Reactor ◽  
Closed Cycle ◽  
The Core ◽  
Channel Radius ◽  
Infinite Lattice

This article illustrates the influence of heterogeneity in an infinite lattice of a Molten Salt Reactor moderated by graphite. For a complete description of heterogeneity in a 2D lattice, two variables are needed; in this study the salt share in the unit cell and the channel radius are used. The equilibrium Thorium-based closed-cycle fuel composition is systematically derived for each chosen combination of points, and results such as kinf and the actinide vector composition are calculated. Results show that the heterogeneity effect can indeed be important for optimization of the core design of moderated molten salt reactors.

Primary Shielding Design for an Optimized Molten Salt Reactor

2013 ◽  
Author(s):  
Zhihong Zhang ◽  
Xiaobin Xia ◽  
Jianhua Wang ◽  
Changyuan Li
Keyword(s):  
Molten Salt ◽  
Dose Rate ◽  
Gamma Dose ◽  
Acceptable Level ◽  
Gamma Dose Rate ◽  
Molten Salt Reactor ◽  
Thermal Shield ◽  
Oak Ridge ◽  
The Core ◽  
Shielding Design

Molten salt reactor (MSR) system, a candidate of the Generation IV reactors, has inherent safety, on-line refueling and good neutron economy as typical advantages. An optimized MSR is developed by changing the size of fuel channel and the graphite-to-molten salt volume radio, based on the Molten-Salt Reactor Experiment (MSRE), which was originally developed at the Oak Ridge National Laboratory (ORNL). In this paper, shielding calculations for the optimized MSR are presented. The goal of this study is to determine the necessary shielding to decrease the neutron and gamma dose rate to the acceptable level according to national regulations. The operating temperature of the optimized MSR is designed in the range of 500 °C–700 °C, heat removal is also considered in the shielding design. The shielding calculations are carried out by using Monte Carlo method. The shielding system of the optimized MSR consists of 7 zones: the core, the core can, the reactor vessel, the thermal shield, the reactor cell containment, the shield tank and the concrete wall. The combinations of shielding materials in the thermal shield were evaluated. The thermal shield filled with carbon steel balls and circulating water gets an excellent shielding performance and heat removing effects. The neutron spectra and dose distributions, as well as the energy deposition over different shields have been analyzed. The total neutron dose rate outside the thermal shield is attenuated by a factor of about 104, and the gamma dose rate by a factor of about 103. These results show that the shielding design could low dose rate to an acceptable level outside the shielding and far below dose limit required.

On Voltage-Sensitive Managing the Redox-Potential of MSR Fuel Composition

2008 ◽  
Author(s):  
Pavel N. Alekseev ◽  
Alexander L. Shimkevich
Keyword(s):  
Redox Potential ◽  
Solid Electrolyte ◽  
Molten Salt ◽  
Galvanic Cell ◽  
Fuel Composition ◽  
Molten Salt Reactor ◽  
Exchange Processes ◽  
Cation Conductivity ◽  
The Given

The principles for optimal managing a composition of base solutions for the molten-salt reactor are formulated here for ensuring the given properties and exchange processes as a selective extracting of salt components. The correction of melt properties can be carried out by means of impurity additives parallel with the forced and controllable variation of reduction-oxidation (redox) potential of the non-stoichiometric salts. The accent is done on a possible application of the potentiometer for monitoring and managing of the properties of MSR fuel compositions. For this, one can use the precision methods of e.m.f and the coulomb-metric titration of sodium (lithium) in a galvanic cell upon the base of Na+(Li+)-β″-Al2O3 solid electrolyte with cation conductivity.

A Novel Molten Salt Reactor Concept to Implement the Multi-Step Time-Scheduled Transmutation Strategy

2002 ◽  
Author(s):  
Gyula Csom ◽  
Sandor Feher ◽  
Mate Szieberth
Keyword(s):  
Molten Salt ◽  
Liquid Fuel ◽  
Primary Circuit ◽  
Chemical Processing ◽  
Molten Salt Reactor ◽  
The Core ◽  
Continuous Feed ◽  
Fuel Mixtures ◽  
Continuous Chemical ◽  
Optimal Implementation

Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutational capabilities are studied. The goal is the development of a transmutational technique along with a device implementing it, which yield higher transmutational efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named “multi-region” MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a “conventional” MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on.

Steady Thermal Hydraulics Analysis for a Graphite-Moderated Channel Type Molten Salt Reactor

2008 ◽  
Author(s):  
Dalin Zhang ◽  
Suizheng Qiu
Keyword(s):  
Molten Salt ◽  
Flow Distribution ◽  
Axial Direction ◽  
Theoretical Models ◽  
Outer Wall ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
The Core ◽  
Fuel Salt ◽  
Channel Type

The Molten Salt Reactor (MSR) is one of the six GENIV systems capable of breading and burning. In this paper, a graphite-moderated channel type MSR was selected for conceptual research. For this MSR, a ternary system of 0.15LiF-0.58NaF-0.27BeF2 was proposed as the reactor fuel solvent, coolant and also moderator simultaneously with ca.1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. 169 hexagonal graphite elements, each with a central fuel channel, are arranged in the core symmetrically by 30° angles. The theoretical models of the thermal hydraulics under steady condition are conducted in one-twelfth of the core and calculated by the numerical method. The DRAGON code is adopted to calculate the axial and radial power factors. The flow and heat transfer models in the fuel salt and graphite are founded basing on the fundamental mass, momentum and energy equations. The calculated results show the detailed mass flow distribution in the core; and the temperature of the fuel salt, inner and outer wall in the calculated elements along the axial direction are also obtained.

New Exploration on TMSR: Redesign of the TMSR Lattice

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
J. K. Zhao ◽  
S. Y. Si ◽  
Q. C. Chen ◽  
H. Bei
Keyword(s):  
Molten Salt ◽  
Lattice Parameters ◽  
Power Coefficient ◽  
Fuel Composition ◽  
Molten Salt Reactor ◽  
Negative Power ◽  
Oak Ridge ◽  
Lattice Design ◽  
Lattice Size ◽  
Breeding Ratio

Molten salt reactor (MSR) has been recognized as one of the next-generation nuclear power systems. Most MSR concepts are the variants evolved from the Oak Ridge National Laboratory (ORNL's) molten-salt breeder reactor (MSBR), which employs molten-salt as both fuel and coolant, and normally graphite is used as moderator. Many evaluations have revealed that such concepts have low breeding ratio and might present positive power coefficient. Facing these impediments, thorium molten salt reactor (TMSR) with redesigned lattice is proposed in this paper. Based on comprehensive investigation and screening, important lattice parameters including molten salt fuel composition, solid moderator material, lattice size, structure and lattice pitch to channel diameter (P/D) ratio are redesigned. In this paper, a fuel composition without BeF2 is adopted to increase the solubility for actinides (ThF4, UF4), and BeO is introduced as moderator to improve neutron economy. Moreover, lattice size and structure with cladding to separate fuel and moderator were also optimized. With these lattice parameters, TMSR has a high breeding ratio close to 1.14 and a short doubling time about 15 years. Meanwhile, a negative power coefficient is maintained. Based on this lattice design, TMSR can have excellent performance of safety and sustainability. SONG/TANG-MSR codes system is applied in the simulation, which is independently developed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI).

Numerical Investigation of Safety System Parameters in Molten Salt Reactor: Wall Effect on Freeze Valve Opening Time

2021 ◽  
Author(s):  
Muhammad Ilham ◽  
Indarta Kuncoro Aji ◽  
Tomio Okawa
Keyword(s):  
Molten Salt ◽  
Power Plants ◽  
Melting Process ◽  
Wall Effect ◽  
Fourth Generation ◽  
Reactor Wall ◽  
Molten Salt Reactor ◽  
Oak Ridge ◽  
Valve System ◽  
The Core

Abstract Molten Salt Reactor (MSRs) is one of the fourth generation Nuclear Power Plants with better capabilities and potentialities compared to previous generation, the enthusiasm for molten fuel reactor has been increasing around the world. MSRs has passive safety where if the core is overheating cause by accident event, the liquid salt fuel was required to be moved to the safety drain tank underneath the core vessel by gravity force. During this occasion, the freeze valve (FV) that formed in the pipe located between the core and drain tank must be melt out promptly to prevent the vessel to reach it is melting point. In this paper, we conduct on thermal analysis of the freeze valve at the solidification and melting process based on finite elements methods. The enthalpy-porosity method adopted by ANSYS Fluent was used to simulated the designed system at specified condition. The Oak Ridge National Laboratory of Molten Salt Reactor Experiment Freeze valve system was used as a references for parameters investigation. Using pipe wall thickness of 5 mm, 10 mm, and 15 mm to examined the wall effect to thermal properties of the designed freeze valve. The wall pipe for FV systems material was also investigate in order to examine its effect to the opening time. Further, the temperature distributions of the valve system were obtained and analyzed. It was found that the wall effect has significant impact to the solidification and melting process.

Breeding Properties Study on High Power Thorium Molten Salt Reactor

2017 ◽  
Author(s):  
Yafen Liu ◽  
Rui Guo ◽  
Xiangzhou Cai ◽  
Rui Yan ◽  
Yang Zou ◽  
...  
Keyword(s):  
Molten Salt ◽  
High Power ◽  
Liquid Fuel ◽  
Molten Salt Reactor ◽  
The Core ◽  
Optimized Model ◽  
Core Conditions ◽  
Thorium Molten Salt Reactor ◽  
Reactivity Insertion ◽  
Two Fluid

The Molten Salt Reactor (MSR), the only one using liquid fuel in the six types ‘Generation IV’ reactors, is very different from reactors in operation now and has initiated very extensive interests all over the world. This paper is primarily aimed at investigating the breeding characteristics of high power (1000 MWe) Thorium Molten Salt Reactor (TMSR) based on the two-fluid Molten Salt Breeder Reactor (MSBR) with superior breeding performance. We explored the optimized structure to be a thorium based molten salt breeder reactor with different core conditions and different post-processing programs, and finally got the breeding ratio of 1.065 in our TMSR model. At last we analyzed the transient security of our optimized model with results show that the temperature coefficient of core is −3 pcm/K and a 2000 pcm reactivity insertion can be successfully absorbed by the core if the insertion time is more than or equal to 5 seconds and the core behaves safely.

Numerical Research on Steady Coupling of Neutronics and Thermal-Hydraulics for a Molten Salt Reactor

2008 ◽  
Author(s):  
Dalin Zhang ◽  
Changliang Liu ◽  
Libo Qian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Thermal Neutron ◽  
Molten Salt ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Steady Condition ◽  
Delayed Neutron ◽  
The Core ◽  
Production Of Hydrogen ◽  
Fuel Salt ◽  
Salt Flow

The Molten Salt Reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for production of electricity, actinide burning, production of hydrogen, and production of fissile fuels. In this paper, a single-liquid-fueled MSR was selected for conceptual research. For this MSR, a ternary system of 15%LiF-58%NaF-27%BeF2 was proposed as the reactor fuel solvent, coolant and also moderator with ca. 1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. The fuel salt flow makes the MSR different from the conventional reactors using solid fissile materials, and makes the neutronics and thermal-hydraulic coupled strongly, which plays the important role in the research of reactor safety analysis. Therefore, it’s necessary to study the coupling of neutronics and thermal-hydraulic. The theoretical models of neutronics and thermal-hydraulics under steady condition were conducted and calculated by numerical method in this paper. 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 flow effect. The thermal-hydraulic model was founded on the base of the fundamental conservation laws: the mass, momentum and energy conservation equations. These two models were coupled through the temperature and heat source. The spatial discretization of the above models is based on the finite volume method (FVM), and the thermal-hydraulic equations are computed by SIMPLER algorithm with domain extension method on the staggered grid system. The distribution of neutron fluxes, the distribution of the temperature and velocity and the distribution of the delayed neutron precursors in the core were obtained. The numerical calculated results show that, the fuel salt flow has little effect to the distribution of fast and thermal neutron fluxes and effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition, and the flow could remove the heat generated by the neutron reactions easily to ensure the reactor safe. The obtained results serve some valuable information for the research and design of this new generation reactor.

Molten Salt Fast Reactor Blanket Design and Proliferation Resistance Assessment

2014 ◽  
Author(s):  
Valentyn Bykov ◽  
Jiři Křepel ◽  
Andreas Pautz
Keyword(s):  
Molten Salt ◽  
Positive Impact ◽  
Reactor Wall ◽  
Molten Salt Reactor ◽  
Radiological Hazard ◽  
Breeding Performance ◽  
Interested Party ◽  
Plutonium Isotopes ◽  
The Core ◽  
Two Fluid

Molten Salt Reactor (MSR) designs are frequently accompanied by a blanket salt. This way the irradiation of the outer reactor wall will be strongly reduced. On the other hand, the barrier between the core and blanket will undergo higher irradiation and it will be necessary to replace it several times during the reactor lifetime. Furthermore, this blanket salt will also have a positive impact on neutron economy by improving the breeding performance. In this paper a blanket of a generic two fluid molten salt reactor utilizing fast thorium-uranium cycle was investigated. This was done by tracking the evolution of uranium, neptunium and plutonium isotopes with burnup, which was then influenced by removal of uranium from the blanket. A significant reduction in the production of minor actinides was observed. The uranium vector removed from the core was then investigated for proliferation resistance, using NUREC proliferation resistance metric and comparison with other weapon designs. The evaluation concluded that while the presence of U-232 increases radiological hazard associated with this uranium, thereby erecting a radiological barrier, it cannot be treated as “self-protecting” based on IAEA and NRC standards, requiring 1 Sv/h at 1m dose rate. Moreover ideas on how an interested party could reduce this radiological hazard were discussed.

Breeding Properties Study on High-Power Thorium Molten Salt Reactor

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Yafen Liu ◽  
Rui Guo ◽  
Xiangzhou Cai ◽  
Rui Yan ◽  
Yang Zou ◽  
...  
Keyword(s):  
Molten Salt ◽  
High Power ◽  
Liquid Fuel ◽  
Molten Salt Reactor ◽  
The Core ◽  
Optimized Model ◽  
Core Conditions ◽  
Thorium Molten Salt Reactor ◽  
Reactivity Insertion ◽  
Two Fluid

Molten salt reactor (MSR), the only one using liquid fuel in the six types “Generation IV” reactors, is very different from reactors in operation now and has initiated very extensive interests all over the world. This paper is primarily aimed at investigating the breeding characteristics of high-power thorium molten salt reactor (TMSR) based on the two-fluid molten salt breeder reactor (MSBR) with a superior breeding performance. We explored the optimized structure to be a thorium-based molten salt breeder reactor with different core conditions and different postprocessing programs, and finally got the breeding ratio of 1.065 in our TMSR model. At last we analyzed the transient security of our optimized model with results show that the temperature coefficient of core is −3 pcm/K and a 2000 pcm reactivity insertion can be successfully absorbed by the core if the insertion time is more than or equal to 5 s and the core behaves safely.

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