Design Feasible Area on Water Cooled Thorium Breeder Reactor in Equilibrium States

2006 ◽  
Author(s):  
Sidik Permana ◽  
Naoyuki Takaki ◽  
Hiroshi Sekimoto
Keyword(s):  
Molten Salt ◽  
Equilibrium States ◽  
Fissile Material ◽  
Fuel System ◽  
Thermal Spectrum ◽  
Thermal Region ◽  
Reactivity Coefficient ◽  
Η Value ◽  
Fuel Material

Thorium as supplied fuel has good candidate for fuel material if it is converted into fissile material 233U which shows superior characteristics in the thermal region. The Shippingport reactor used 233U-Th fuel system, and the molten salt breeder reactor (MSBR) project showed that breeding is possible in a thermal spectrum. In the present study, feasibility of water cooled thorium breeder reactor is investigated. The key properties such as flux, η value, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The results show the feasibility of breeding for different MFR and burn-ups. The required 233U enrichment is about 2%–9% as charge fuel. The lower MFR and the higher enrichment of 233U are preferable to improve the average burn-up; however the design feasible window is shrunk. This core shows the design feasible window especially in relation to MFR with negative void reactivity coefficient.

Development and Verification of Dynamics Code for Molten Salt Reactors

2004 ◽  
Author(s):  
Jiri Krepel ◽  
Ulrich Grundmann ◽  
Ulrich Rohde
Keyword(s):  
Molten Salt ◽  
Liquid Fuel ◽  
Expansion Method ◽  
Primary Circuit ◽  
Fuel System ◽  
Delayed Neutrons ◽  
Flow Acceleration ◽  
Fuel Flow ◽  
Reactor Dynamics ◽  
Reactivity Increase

To perform transient analysis for Molten Salt Reactors (MSR), the reactor dynamics code DYN3D developed in FZR was modified for MSR applications. The MSR as a liquid fuel system can serve as a thorium breeder and also as an actinide burner. The specifics of the reactor dynamics of MSR consist in the fact, that there is direct influence of the fuel velocity to the reactivity, which is caused by the delayed neutrons precursors drift. This drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit. This leads to a reactivity loss due to the fuel flow acceleration or to the reactivity increase in the case of deceleration. For the first analyses, a 1D modified version DYN1D-MSR of the code has been developed. By means of the DYN1D-MSR, several transients typical for the liquid fuel system were analyzed. Transients due to the overcooling of fuel at the core inlet, due to the reactivity insertion, and the fuel pump trip have been considered. The results of all transient studies have shown that the dynamic behavior of MSR is stable when the coefficients of thermal feedback are negative. For studying space-dependent effects like e.g. local blockages of fuel channels, a 3D code version DYN3D-MSR will be developed. The nodal expansion method used in DYN3D for hexagonal fuel element geometry of VVER can be applied considering MSR design with hexagonal graphite channels.

Adjoint‐based sensitivity analysis of circulating liquid fuel system for the multiphysics model of molten salt reactor

2020 ◽  
Vol 44 (5) ◽  
pp. 3934-3953
Author(s):  
Yeong Shin Jeong ◽  
Eric Cervi ◽  
Antonio Cammi ◽  
Hisashi Ninokata ◽  
In Cheol Bang
Keyword(s):  
Sensitivity Analysis ◽  
Molten Salt ◽  
Liquid Fuel ◽  
Molten Salt Reactor ◽  
Fuel System

scholarly journals MATRIX EXPONENTIAL METHODS FOR PARALLEL COMPUTING OF ISOTOPIC DEPLETION AND SPECIES TRANSPORT FOR MOLTEN SALT REACTOR ANALYSIS

2021 ◽  
Vol 247 ◽  
pp. 06047
Author(s):  
Zack Taylor ◽  
Benjamin Collins ◽  
Ivan Maldonado
Keyword(s):  
Differential Equations ◽  
Molten Salt ◽  
Fission Products ◽  
Matrix Exponential ◽  
Fissile Material ◽  
Species Transport ◽  
The Matrix ◽  
Exponential Method ◽  
Reactor Loop ◽  
Exponential Methods

Matrix exponential methods have long been utilized for isotopic depletion in nuclear fuel calculations. In this paper we discuss the development of such methods in addition to species transport for liquid fueled molten salt reactors (MSRs). Conventional nuclear reactors work with fixed fuel assemblies in which fission products and fissile material do not transport throughout the core. Liquid fueled molten salt reactors work in a much different way, allowing for material to transport throughout the primary reactor loop. Because of this, fission product transport must be taken into account. The set of partial differential equations that apply are discretized into systems of first order ordinary differential equations (ODEs). The exact solution to the set of ODEs is herein being estimated using the matrix exponential method known as the Chebychev Rational Approximation Method (CRAM).

scholarly journals Candidate Core Designs for the Transformational Challenge Reactor

2021 ◽  
Vol 2 (1) ◽  
pp. 74-85
Author(s):  
Brian J. Ade ◽  
Benjamin R. Betzler ◽  
Aaron J. Wysocki ◽  
Michael S. Greenwood ◽  
Phillip C. Chesser ◽  
...  
Keyword(s):  
Reactor Core ◽  
Fissile Material ◽  
Design Concepts ◽  
Thermal Spectrum ◽  
Triso Fuel ◽  
Ceramic Fuel ◽  
Near Term ◽  
Fuel Particles ◽  
Fuel Elements ◽  
Spectrum Design

Early cycle activities under the Transformational Challenge Reactor (TCR) program focused on analyzing and maturing four reactor core design concepts: two fast-spectrum systems and two thermal-spectrum systems. A rapid, iterative approach has been implemented through which designs can be modified and analyzed and subcomponents can be manufactured in parallel over time frames of weeks rather than months or years. To meet key program initiatives (e.g., timeline, material use), several constraints—including fissile material availability (less than 250 kg of HALEU), component availabilities, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small (less than one cubic meter in volume) cores with near-term viability. The fast-spectrum designs did not meet the fissile material constraint, so the thermal-spectrum systems became the primary design focus. Since significant progress has been made on advanced moderator materials (YHx) under the TCR program, gas-cooled thermal-spectrum systems using less than 250 kg of HALEU that occupy less than 1 m3 are now feasible. The designs for two of these systems have been evolved and matured. In both thermal-spectrum design concepts, bidirectional coolant flow is used. Coolant flows down through YHx moderator elements and is reversed in a bottom manifold and core support structure, and then flows up though or around the fuel elements. The main difference between the two thermal-spectrum design concepts is the fuel elements—one uses traditional UO2 ceramic fuel, and the other uses UN-bearing TRISO fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems are assessed and summarized herein.

scholarly journals Neutronic Evaluation of MSBR System Using MCNP Code

2021 ◽  
Vol 9 (2B) ◽  
Author(s):  
Clarysson Alberto Mello da Silva ◽  
Alana Lima Vieira ◽  
Isabella Resende Magalhães ◽  
Claubia Pereira
Keyword(s):  
Neutron Flux ◽  
Molten Salt ◽  
Fissile Material ◽  
Uranium Enrichment ◽  
Flux Profile ◽  
High Enrichment ◽  
Initial Seed ◽  
Initial Source ◽  
Fuel Elements ◽  
Mcnpx 2.6.0 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.

Analysis of Sustainable Thorium Fuel Utilization in Molten Salt Reactors Starting From Enriched Uranium

2017 ◽  
Author(s):  
Deyang Cui ◽  
Xiangzhou Cai ◽  
Jingen Chen ◽  
Chenggang Yu
Keyword(s):  
Molten Salt ◽  
Fuel Cycle ◽  
Utilization Efficiency ◽  
Fissile Material ◽  
Fuel Utilization ◽  
Light Water ◽  
Waste Production ◽  
Water Reactors ◽  
Enriched Uranium ◽  
Two Stages

Molten salt reactor (MSR), as one of the six systems selected by the Generation IV International Forum (GIF) for future advantaged reactors research and development (R&D), has excellent performances such as high inherent safety, desirable breeding capacity, low radioactive waste production, flexible fuel cycle and non-proliferation. Meanwhile, thorium, as an appealing alternative nuclear fuel to uranium, is more abundant than uranium in the earth’s crust. Realization of thorium fuel cycle in MSRs will greatly contribute to sustainable energy supply for global development. The objective of this paper is to analyze and evaluate thorium fuel utilization in a program in which MSRs are expected to be developed step by step. The program can be described as follows: 1 The first stage is a converter reactor fueled with low enriched uranium. With limited processing based on current chemical partitioning technology and fuel-feeding techniques in the generation-I MSR; 2 The second stage is a 233U production reactor. By using the enriched uranium, it can produce 233U which does not exist in nature; 3 The third stage is a thorium breeding reactor. It is a breeder reactor with Th/233U fuel cycle, and sustainable thorium utilization for energy production is expected to be eventually realized. By employing an in-house developed tool based on SCALE6.1, the performance of MSR fueled with low enriched uranium is firstly assessed. It is found that MSR is attractive regarding conversion ratio when compared with light water reactors. Then we illustrate the feasibility of 233U production in MSR. Enriched uranium with two enrichments are used as driver fuels to start MSR and produce 233U. The results show that 233U production can be achieved and the double time is about 79.1 years for 20% enriched uranium and 28.3 years for 60% enriched uranium. Finally, the performance of MSR based on pure Th/233U fuel cycle is evaluated. It is found that breeding fissile material is possible in MSR and the breeding ratio is desirable (1.049). Comparison of the three-stage MSRs is also conducted and the results indicate that the resource utilization efficiency is much higher in stage-III than that in the first two stages and much less minor actinides is produced in MSR operating on Th/233U fuel cycle than that in traditional light water reactor.

scholarly journals Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step

Nukleonika ◽  
2015 ◽  
Vol 60 (4) ◽  
pp. 907-914 ◽  
Author(s):  
Davide Rodrigues ◽  
Gabriela Durán-Klie ◽  
Sylvie Delpech
Keyword(s):  
Molten Salt ◽  
Fast Reactor ◽  
Liquid Fuel ◽  
Noble Metals ◽  
Fission Products ◽  
Selective Extraction ◽  
Fissile Material ◽  
Metallic Lithium ◽  
Fuel Salt ◽  
Static Conditions

Abstract The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR) concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by contacting LiF-ThF4-UF4-NdF3 with a lab-made Bi-Li pool and for which extraction efficiencies of 0.7% for neodymium and 14.0% for uranium were measured. It was concluded that in static conditions, the extraction is governed by a kinetic limitation and not by the thermodynamic equilibrium.

Macroeconomic Equilibrium Theory in the Context of Methodological Problems of Contemporary Economic Science

2008 ◽  
pp. 77-88
Author(s):  
M. Likhachev
Keyword(s):  
Empirical Analysis ◽  
Equilibrium States ◽  
Equilibrium Theory ◽  
Economic Systems ◽  
Economic Science ◽  
The Real ◽  
Theoretical Status ◽  
Methodological Problems ◽  
Real Stability

The article is devoted to the analysis of methodological problems in using the conception of macroeconomic equilibrium in contemporary economics. The author considers theoretical status and relevance of equilibrium conception and discusses different areas and limits of applicability of the equilibrium theory. Special attention is paid to different epistemological criteria for this theory taking into account both empirical analysis of the real stability of economic systems and the problem of unobservability of equilibrium states.

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