• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.724-730
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• 2018

Chromium and chromium nitride were selected as potential barriers to prevent fuel-clad chemical interaction (FCCI) between the cladding and the fuel material. In this study, ferritic/martensitic HT-9 steel and misch metal were used to simulate the reaction between the cladding and fuel fission product, respectively. Radio frequency magnetron sputtering was used to deposit Cr and CrN films onto the cladding, and the gas flow rates of argon and nitrogen were fixed at certain values for each sample to control the deposition rate and the crystal structure of the films. The samples were heated for 24 h at 933 K through the diffusion couple test, and considerable amount of interdiffusion (max. thickness: $550{\mu}m$) occurred at the interface between HT-9 and misch metal when the argon and nitrogen were used individually. The elemental contents of misch metal were detected at the HT-9 through energy dispersive X-ray spectroscopy due to the interdiffusion. However, the specimens that were sputtered by mixed gases (Ar and $N_2$) exhibited excellent resistance to FCCI. The thickness of these CrN films were only $4{\mu}m$, but these films effectively prevented the FCCI due to their high adhesion strength (frictional force ${\geq}1,200{\mu}m$) and dense columnar microstructures.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.653-660
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• 2022

The interaction between UN and CdCl₂ in the LiCl-KCl molten eutectic was studied at 1023 K. The chlorination was monitored by sampling and recording the redox potential of the medium. At 1023 K the chlorination of UN with cadmium chloride in the molten LiCl-KCl eutectic proceeds completely and results in the formation of uranium chlorides. The melts of the LiCl-KCl-UCl₃ or LiCl-KCl-UCl₄ compositions can be obtained by the end of experiment depending on the presence of metallic cadmium in the reaction zone. The higher the concentration of the chlorinating agent, the faster the reaction rate. At [CdCl₂]/[UN] = 1.65 (10% excess) the reaction proceeds to completion in about 7.5 h. At [CdCl₂]/[UN] = 7 the complete chlorination takes 2.5-3 h.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.423-429
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• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.237-245
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• 2018

Potential air ingress scenarios during accidents in nuclear reactors or spent fuel pools have raised the question of the influence of air, especially of nitrogen, on the oxidation of zirconium alloys, which are used as fuel cladding tubes and other structure materials. In this context, the reaction of zirconium with nitrogen-containing atmospheres and the formation of zirconium nitride play an important role in understanding the oxidation mechanism. This article presents the results of analysis of the interaction of the oxygen-preloaded niobium-bearing alloy $M5^{(R)}$ with nitrogen over a wide range of temperatures ($800-1400^{\circ}C$) and oxygen contents in the metal alloy (1-7 wt.%). A strongly increasing nitriding rate with rising oxygen content in the metal was found. The highest reaction rates were measured for the saturated ${\alpha}-Zr(O)$, as it exists at the metal-oxide interface, at $1300^{\circ}C$. The temperature maximum of the reaction rate was approximately 100 K higher than for Zircaloy-4, already investigated in a previous study. The article presents results of thermogravimetric experiments as well as posttest examinations by optical microscopy, scanning electron microscopy (SEM), and microprobe elemental analyses. Furthermore, a comparison with results obtained with Zircaloy-4 will be made.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.123-134
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• 2020

The interaction between UN and CdCl₂ in the LiCl-KCl molten eutectic was studied at 773 K. The reaction was controlled by sampling the melt, as well as by analysis of the resulting precipitate. The process was shown to proceed according to several parallel reactions. The summary reaction was determined to have two stages: a fast one and a slow one. The 19-53% UN → UCl₃ conversion was obtained for the molar ratio of CdCl₂/UN = 1.22-14.9. The rest of UN converts into the precipitate of complex composition (UNCl + U₂N₃ + U₄N₇ + UN₂). The increase in the CdCl₂/UN molar ratio from 1.22 to 14.9 resulted in the decrease in duration of the first "fast" stage of the process from 18 h to 1 h.

• Composites Research
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• v.36 no.6
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• pp.435-440
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• 2023

This study aimed to enhance the thermal conductivity of B₄C/Al composite materials, commonly used in transport/storage containers for spent nuclear fuel, by incorporating both boron carbide (B₄C) and cubic boron nitride(cBN) as reinforcing agents in an aluminum (Al) matrix. The composite materials were successfully manufactured through a stir casting process and practical neutron-absorbing materials were obtained by rolling the fabricated composite ingot. The evaluation of the thermal conductivity of the fabricated composites was carried out because thermal conductivity is critical for neutron absorbing materials. The thermal conductivity measurement results indicated an approximately 3% increase in thermal conductivity under the same volume fraction when compared to composite materials using only B₄C particles. Through neutron absorption cross-sectional area calculations, it was confirmed that the neutron absorption capability decreased to a negligible level. Based on the findings of this study, new design approaches for neutron absorption materials are proposed, contributing to the development of high-performance transport/storage containers.