scholarly journals Neutronic Analysis of LEU-started Molten Chloride Fast Reactor without Fuel Reprocessing

2021 ◽  
Vol 4 (1) ◽  
pp. 8
Author(s):  
R. Andika Putra Dwijayanto ◽  
Andang Widi Harto
Keyword(s):  
Fast Reactor ◽  
Molar Fraction ◽  
Chloride Salt ◽  
Inherent Safety ◽  
Delayed Neutron ◽  
Core Zone ◽  
Molten Chloride ◽  
Enriched Uranium ◽  
Fuel Reprocessing ◽  
Coefficient Of Reactivity

One of the rarely explored molten salt reactor (MSR) designs is the molten chloride fast reactor (MCFR). This MSR design employs chloride salt instead of fluoride and operated in a fast spectrum. MCFR brings all the advantages of an MSR including breeding whilst being able to burn plutonium and minor actinides efficiently. Since not many countries have access to civilian plutonium, MCFR can also be started using low-enriched uranium (LEU). This study is an initial neutronic analysis of an MCFR using LEU as its startup fuel. Parameters analyzed are conversion ratio (CR) and its neutronic safety, namely effective delayed neutron fraction (βeff), temperature coefficient of reactivity (TCR), and void coefficient of reactivity (VCR). The core is divided into Core Zone and Blanket Zone. The fuel composition of NaCl-UCl3 with a molar fraction ratio of 60:40 and 50:50 is used in Core Zone and Blanket Zone, respectively. The neutronic calculation is performed using MCNP6 code with ENDF/B-VII library. For reference geometry, CR is valued at 0.9298, βeff at 0.00731, TCR at -19.8 pcm/°C, and average VCR at -154.31 pcm/void%. Thereby, the MCFR fulfills inherent safety criteria. Although its value is remarkably high, CR can be further optimized by modifying the separator and reflector material.

scholarly journals INVESTIGATION ON INHERENT SAFETY OF ONE FLUID-MOLTEN SALT REACTOR (OF-MSR) WITH VARIOUS STARTING FUEL

2020 ◽  
Vol 22 (2) ◽  
pp. 54
Author(s):  
R. Andika Putra Dwijayanto ◽  
Dedy Prasetyo Hermawan
Keyword(s):  
Molten Salt ◽  
Temperature Coefficient ◽  
Fuel Cycle ◽  
Minor Actinides ◽  
Molten Salt Reactor ◽  
Inherent Safety ◽  
Safety Aspects ◽  
Safe Option ◽  
Enriched Uranium ◽  
Coefficient Of Reactivity

Molten salt reactor (MSR) is often associated with thorium fuel cycle, thanks to its excellent neutron economy and online reprocessing capability. However, since 233U, the fissile used in pure thorium fuel cycle, is not commercially available, the MSR must be started with other fissile nuclides. Different fissile yields different inherent safety characteristics, and thus must be assessed accordingly. This paper investigates the inherent safety aspects of one fluid MSR (OF-MSR) using various fissile fuel, namely low-enriched uranium (LEU), reactor grade plutonium (RGPu), and reactor grade plutonium + minor actinides (PuMA). The calculation was performed using MCNPX2.6.0 programme with ENDF/B-VII library. Parameters assessed are temperature coefficient of reactivity (TCR) and void coefficient of reactivity (VCR). The result shows that TCR for LEU, RGPu, and PuMA are -3.13 pcm, -2.02 pcm and -1.79 pcm, respectively. Meanwhile, the VCR is negative only for LEU, whilst RGPu and PuMA suffer from positive void reactivity. Therefore, for the OF-MSR design used in this study, LEU is the only safe option as OF-MSR starting fuel.Keywords: MSR, Temperature coefficient of reactivity, Void coefficient of reactivity, Low enriched uranium, Reactor grade plutonium, Minor actinides

scholarly journals Th-U Breeding Performances in an Optimized Molten Chloride Salt Fast Reactor

2020 ◽  
pp. 1-18
Author(s):  
He Liaoyuan ◽  
Xia Shaopeng ◽  
Chen Jingen ◽  
Liu Guimin ◽  
Wu Jianhui ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Chloride Salt ◽  
Molten Chloride

Effect of $$^{{\textbf{37}}}{\textbf{Cl}}$$37Cl enrichment on neutrons in a molten chloride salt fast reactor

2020 ◽  
Vol 31 (3) ◽  
Author(s):  
Liao-Yuan He ◽  
Guang-Chao Li ◽  
Shao-Peng Xia ◽  
Jin-Gen Chen ◽  
Yang Zou ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Chloride Salt ◽  
Molten Chloride

Application of In Situ X-ray Absorption Spectroscopy to Study Dilute Chromium Ions in a Molten Chloride Salt

2019 ◽  
Author(s):  
Jisue Moon ◽  
Carter Abney ◽  
Dmitriy Dolzhnikov ◽  
James M. Kurley ◽  
Kevin A. Beyer ◽  
...  
Keyword(s):  
Absorption Spectroscopy ◽  
Local Structure ◽  
Chloride Salt ◽  
X Ray ◽  
Molten Chloride ◽  
Stable Species ◽  
Higher Temperature ◽  
Cr Concentration ◽  
X Ray Absorption

The local structure of dilute CrCl<sub>3</sub> in a molten MgCl<sub>2</sub>:KCl salt was investigated by <i>in situ</i> x-ray absorption spectroscopy (XAS) at temperatures from room temperature to 800<sup>o</sup>C. This constitutes the first experiment where dilute Cr speciation is explored in a molten chloride salt, ostensibly due to the compounding challenges arising from a low Cr concentration in a matrix of heavy absorbers at extreme temperatures. CrCl<sub>3</sub> was confirmed to be the stable species between 200 and 500<sup>o</sup>C, while mobility of metal ions at higher temperature (>700<sup>o</sup>C) prevented confirmation of the local structure.

scholarly journals Thermal and Transport Properties of Molten Chloride Salts with Polarization Effect on Microstructure

Energies ◽  
2021 ◽  
Vol 14 (3) ◽  
pp. 746
Author(s):  
Jianfeng Lu ◽  
Senfeng Yang ◽  
Gechuanqi Pan ◽  
Jing Ding ◽  
Shule Liu ◽  
...  
Keyword(s):  
Heat Capacity ◽  
Polarization Effect ◽  
Chloride Salt ◽  
Coulombic Interaction ◽  
Molten Chloride ◽  
Self Diffusion ◽  
Local Structures ◽  
Property Data ◽  
Chloride Salts ◽  
And Storage

Molten chloride salt is recognized as a promising heat transfer and storage medium in concentrating solar power in recent years, but there is a serious lack for thermal property data of molten chloride salts. In this work, local structures and thermal properties for molten chloride salt—including NaCl, MgCl2, and ZnCl2—were precisely simulated by Born–Mayer–Huggins (BMH) potential in a rigid ion model (RIM) and a polarizable ion model (PIM). Compared with experimental data, distances between cations, densities, and heat capacities of molten chloride slats calculated from PIM agree remarkably better than those from RIM. The polarization effect brings an extra contribution to screen large repulsive Coulombic interaction of cation–cation, and then it makes shorter distance between cations, larger density and lower heat capacity. For NaCl, MgCl2, and ZnCl2, PIM simulation deviations of distances between cations are respectively 3.8%, 3.7%, and 0.3%. The deviations of density and heat capacity for NaCl between PIM simulation and experiments are only 0.6% and 2.2%, and those for MgCl2 and ZnCl2 are 0.7–10.7%. As the temperature rises, the distance between cations increases and the structure turns into loose state, so the density and thermal conductivity decrease, while the ionic self-diffusion coefficient increases, which also agree well with the experimental results.

Prospects for fast reactors

1987 ◽  
Vol 92 (1-2) ◽  
pp. 57-71
Author(s):  
C. W. Blumfield
Keyword(s):  
Fast Reactor ◽  
High Energy ◽  
High Energy Density ◽  
Fast Reactors ◽  
Advantages And Disadvantages ◽  
Reactor Technology ◽  
European Collaboration ◽  
Fast Reactor Fuel ◽  
Fuel Reprocessing ◽  
Planned Activities

SynopsisThe background to recent major advances in the construction and operation of fast reactors is outlined with particular reference to the Dounreay Prototype Fast Reactor. The advantages and disadvantages of sodium as a coolant of the high energy density assembly are discussed and an account given of the consequences of a leak and the precautions taken against this eventuality. Attention is drawn to the safety aspects of the system. The economics of the plans for fuel reprocessing are explained and an account given of the progress in the fabrication of fast reactor fuel pins. Finally the environmental impact of present and planned activities on the Dounreay site is presented in the context of participation in the European Collaboration on Fast Reactor Technology and attention drawn to the importance of the planning inquiry in progress at Dounreay.

Delayed Neutron Data for Fast Reactor Analysis

1973 ◽  
Vol 50 (3) ◽  
pp. 208-215 ◽  
Author(s):  
J. E. Cahalan ◽  
K. O. Ott
Keyword(s):  
Fast Reactor ◽  
Delayed Neutron ◽  
Neutron Data
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