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.

Physical Design on PMNSR With LEU Core

2017 ◽  
Author(s):  
Wu Xiaobo ◽  
Peng Dan ◽  
Hong Jingyan ◽  
Lu Jin ◽  
Hao Qian
Keyword(s):  
Neutron Flux ◽  
Physical Design ◽  
Second Phase ◽  
Core Element ◽  
Control Rod ◽  
Power Test ◽  
Calculation Results ◽  
Leading Position ◽  
High Enrichment ◽  
Fuel Elements

Prototype Miniature Neutron source Reactor (PMNSR) is a pool-tank type research reactor,applying high enrichment Uranium as fuel, light water as moderator and coolant, beryllium as reflector. Recently, in order to prevent nuclear proliferation, PMNSR carried out low enrichment uranium (LEU) core conversion, and the enrichment of U-235 decreased from 90% to under 20%. Research on PMNSR with LEU core mainly includes theory design, zero power experiment, core replacement. The physical design of PMNSR with LEU is the main part of theory design, which plays a great role in LEU conversion. At the first stage of LEU conversion, it performs preliminary physical calculation and analysis concerning U-235 fuel enrichment, and the number of critical fuel elements, the reactivity worth of control rod, the reactivity worth of top beryllium reflector, the neutron flux of inter-irradiation tube are calculated, which provides important data for the fuel elements design, fabrication, zero power test safety analysis and experiment for LEU conversion. In the second phase, it conducts the result verification on zero power test and preliminary physical design and a preliminary error analysis resulted from it thereof. More over, it modifies input file of LEU conversion, optimizes core element loading deployment, the reactivity worth of central control rod, the neutron flux rate of inner radiation site, offering statistics for the replacement and start-up experiments. In the last period, grounded on the counting abnormal analysis in loading, it explains the reasons with calculation results, completing PMNSR LEU conversion. PMNSR physical design takes the leading position in LEU conversion. It supplies reference data to ensure completion of PMNSR conversion and lays a theoretical foundation for Ghana and Nigeria MNSR LEU core conversion.

Design of U3Si2-Al Plate-Type Fuel Element for China Advanced Research Reactor

2010 ◽  
Author(s):  
Aimin Zhang ◽  
Yalun Kang
Keyword(s):  
Fuel Element ◽  
Neutron Flux ◽  
Research Reactor ◽  
High Heat ◽  
Reactor Power ◽  
High Heat Flux ◽  
Water Tank ◽  
Enriched Uranium ◽  
Fuel Elements ◽  
Plate Type

China Advanced Research Reactor (CARR), which will be critical in China Institute of Atomic Energy (CIAE) in 2010, is a multipurpose, high neutron flux and tank-type (inverse neutron trap) reactor with compact core. Its nominal reactor power is 60MW and the maximum thermal neutron flux is about 8.0×1014n/cm2·s in heavy water tank. It has a cylindrical core having a diameter of about 450mm and a height of 850mm. The CARR’s core consists of seventeen plate-type standard fuel elements and four follower fuel elements, initially loaded with 10.97 kg of 235U. The fuel element has been designed with U3S2-Al dispersion containing 235U of (19.75±0.20)wt.% low enriched uranium (LEU) and having a density of 4.3gU/cm3. The aluminum alloy is used as the cladding. There are twenty-one and seventeen fuel plates in the standard and follower fuel element, respectively. There are specific requirements for design of the fuel element and strict limitation for the operation parameters due to the high heat flux and high velocity of coolant in CARR. Irradiation test of fuel element had been carried out at fuel element power of 3.1±20%MW at Russia MIR reactor. Average burnup of fuel element is up to 40%. This paper deals with the detailed design of fuel element for CARR, out-pile and in-pile test projects, including selection of fuel and structure material, description of element structure, miniplates and fuel element irradiation experiment, measurement of properties of fuel plate, fabrication of fuel element and test results.

Heat Transfer Simulation of the Fuel Transport Cask for Spherical Fuel Elements in Molten Salt Reactor

2016 ◽  
Author(s):  
Yang Liu ◽  
Jun Wang
Keyword(s):  
Heat Transfer ◽  
Molten Salt ◽  
Operating Temperature ◽  
Thermal Design ◽  
Radioactive Material ◽  
Molten Salt Reactor ◽  
Temperature Ranges ◽  
Water Reactors ◽  
Fuel Elements ◽  
Safe Operating

Fuel transport is an indispensable task for nuclear power plants. For pressurized water reactors (PWR) and boiling water reactors (BWR), many research projects have been completed for designing and testing the transport casks for fresh fuel as well as spent fuel [1–3]. To ensure the safety of nuclear fuel during the transportation, many aspects should be analyzed and examined for the casks with fuel inside, such as heat transfer and temperature calculation, radiation protection, nonproliferation issues, etc. The transport cask discussed in this paper is especially for new spherical fuel elements, which should be designed in accordance with the stipulations in the GB11806 Regulations for the Safe Transport of Radioactive Material [4]. The Transport Cask for spherical fuel elements used in molten salt reactor (MSR) should be designed in accordance with the safety standards for transport of radioactive material. It is necessary to evaluate the thermal performance of the transport cask separately in normal transport condition and in accident transient. The MSR fuel sphere elements cask is in a circular cylinder shape and composed of inner container and outer shell cask. The objective of the thermal analysis of the cask under hypothetical accident conditions is to demonstrate that the cask containment boundary structural components are maintained within their safe operating temperature ranges. The heat transfer process (conduction, convection, and radiation) is simulated by ANSYS-CFX in this paper and it is demonstrated that the components of cask are maintained in safe operating temperature ranges. The calculation results are below limit temperatures, indicating that the thermal design of the cask could meet the Standard Regulations. The result is also compared with the fire test, which shows the calculation model is conservative and rational.

Real-time wide-range neutron flux monitor for thorium-based molten salt reactor

2018 ◽  
Vol 29 (8) ◽  
Author(s):  
Xiang Zhou ◽  
Zi-Hao Liu ◽  
Chao Chen ◽  
Guo-Qing Huang ◽  
Ze-Jie Yin
Keyword(s):  
Neutron Flux ◽  
Molten Salt ◽  
Real Time ◽  
Flux Monitor ◽  
Molten Salt Reactor ◽  
Neutron Flux Monitor ◽  
Wide Range

scholarly journals The pulsed subcritical amplifier of neutron flux driven by high-intensity neutron generator

2019 ◽  
Vol 34 (1) ◽  
pp. 1-12
Author(s):  
Vladimir Babenko ◽  
Volodymyr Pavlovych ◽  
Volodymyr Gulik
Keyword(s):  
Neutron Flux ◽  
Heterogeneous Systems ◽  
High Energy ◽  
Energy Applications ◽  
Nuclear Models ◽  
In Series ◽  
Initial Source ◽  
High Level ◽  
Energy Amplification ◽  
Subcritical Reactor

The subcritical reactor driven by external neutron source could apply as useful instrument for modern nuclear energy applications requiring high-level irradiation of different materials by the high-energy and high-intense neutron flux (e. g., nuclear waste transmutation, radiopharmaceutical production, etc.). The propagation of neutron pulses through the subcritical nuclear system was considered in the present paper. Simple homogeneous subcritical systems and a model of two-zone subcritical reactor were computationally investigated using Monte Carlo MCNP4c transport code. The propagation of one initial neutron pulse and series of one hundred neutron pulses through the presented subcritical nuclear models were simulated. In this study, the neutron multiplication factor, the neutron flux, the energy amplification factor, the total energy of neutrons in initial pulse, etc. were obtained and analyzed. The presented calculations have shown that the considered pulse subcritical systems can be successfully used as effective amplifiers of neutron flux from the initial source. The modeling results indicate that there is an achievement of a stable, high level of neutron flux caused by the accumulation of delayed neutrons from previous pulses in series of one hundred pulses for both homogeneous and heterogeneous systems.

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).

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