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).

New Exploration on TMSR: Redesign of the TMSR Lattice

2017 ◽  
Author(s):  
Jinkun Zhao ◽  
Shengyi Si ◽  
Qichang Chen ◽  
Hua Bei
Keyword(s):  
Molten Salt ◽  
Lattice Parameters ◽  
Power Coefficient ◽  
Fuel Composition ◽  
Molten Salt Reactor ◽  
Negative Power ◽  
Flowing Liquid ◽  
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 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, TMSR (Thorium Molten Salt Reactor) 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 P/D ratio (lattice pitch to channel diameter) are redesigned. In this paper, new composition of fuel salt without BeF2, which is also recommend for Molten Salt Fast Reactor (MSFR), is employed instead of LiF-BeF2-ThF4-UF4 adopted in the design of single fluid MSBR. The new fuel composition makes TMSR to benefit from the increased solubility for actinides (e.g. Th4, UF4). Moreover, due to the decent slowing-down power and neutron multiplication effect by (n,2n) reaction of beryllium, BeO is employed as moderator to improve neutron economy instead of graphite. To avoid corrosion on the one hand, Ceramic cladding (e.g. SiC) is introduced to separate the flowing liquid fuel and fixed solid moderator. More importantly, ceramic cladding is capable of maintaining a stable flow channel and supporting the core structure on the other hand. Concerning neutron spectrum, P/D ratio is an important parameter indicating the volume fraction of fuel in the lattice. In order to obtain a suitable spectrum for better breeding and safety features, lattice size and P/D ratio have been optimized for TMSR. Furthermore, since online reprocessing capability and refueling control are key parameters influencing depletion behavior which concerns the sustainability of the reactor system, these issues are also discussed in this paper. Simulation of the redesigned TMSR system is performed to evaluate the outcomes of the lattice parameters optimization. SONG/TANG-MSR codes system is applied in the simulation, which is independently developed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI). A traditional core model with LiF-BeF2-ThF4-UF4 fuel and graphite moderator is also evaluated by the codes for reference. Thanks to the optimized lattice parameters and as consequences of the redesigned lattice, TMSR has achieved a high breeding ratio close to 1.13. With a proper reprocessing and refueling strategy, the doubling time of TMSR can be shortened to about 15 years. Meanwhile a negative power coefficient is still maintained. Based on this lattice design, TMSR will have excellent performance on safety and sustainability.

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.

Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor

2015 ◽  
Vol 281 ◽  
pp. 114-120 ◽  
Author(s):  
C.Y. Zou ◽  
X.Z. Cai ◽  
D.Z. Jiang ◽  
C.G. Yu ◽  
X.X. Li ◽  
...  
Keyword(s):  
Molten Salt ◽  
Temperature Coefficient ◽  
Molten Salt Reactor ◽  
Breeding Ratio

A Review of Molten Salt Reactor Xenon Analysis Literature

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
T. J. Price ◽  
O. Chvala
Keyword(s):  
Molten Salt ◽  
Void Fraction ◽  
Experimental Work ◽  
National Laboratory ◽  
A Priori ◽  
Molten Salt Reactor ◽  
Oak Ridge ◽  
Sensitive Dependence ◽  
Oak Ridge National Laboratory ◽  
Work Done

Abstract This paper presents a review of xenon analyses literature related to molten salt reactors (MSRs). A brief primer of reactor xenon theory is presented for fluid fueled reactors. A review of xenon analysis literature is presented for both the work done by the Oak Ridge National Laboratory, and the later work in academia. A review of experimental work is presented. The paper concludes with describing some of the difficulties in establishing a priori xenon models and includes a commentary on the sensitive dependence of the molten salt reactor xenon behavior on the circulating void fraction.

scholarly journals Batch-Scale Hydrofluorination of Li27BeF4 to Support Molten Salt Reactor Development

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Brian C. Kelleher ◽  
Kieran P. Dolan ◽  
Paul Brooks ◽  
Mark H. Anderson ◽  
Kumar Sridharan
Keyword(s):  
Molten Salt ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
The United States ◽  
Hydraulic Testing ◽  
Molten Salt Reactor ◽  
Fluoride Salt ◽  
Oak Ridge ◽  
University Of Wisconsin ◽  
Analysis Process

Li 2 BeF 4 , or flibe, is the primary candidate coolant for the fluoride-salt-cooled high-temperature nuclear reactor (FHR). Kilogram quantities of pure flibe are required for repeatable corrosion tests of modern reactor materials. This paper details fluoride salt purification by the hydrofluorination–hydrogen process, which was used to regenerate 57.4 kg of flibe originating from the secondary loop of the molten salt reactor experiment (MSRE) at Oak Ridge National Laboratory (ORNL). Additionally, it expounds upon necessary handling precautions required to produce high-quality flibe and includes technological advancements which ease the purification and analysis process. Flibe batches produced at the University of Wisconsin are the largest since the MSRE program, enabling new corrosion, radiation, and thermal hydraulic testing around the United States.

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.

Sign in / Sign up

Export Citation Format

Share Document