• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1208-1212
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• 2020

Knowledge on fuel enthalpy and its temperature derivative, the heat capacity, are important quantities in determination of fuel behavior in normal reactor operation and reactor transients. The aim of this study is to compare the heat capacity of oxide and nitrite fuels by using Einstein-Debye approximation. A simple analytical expression was performed to calculate the heat capacity of fuels. To test the validity and reliability, the calculated formulas were compared to published results for various nuclear fuels including UO₂, ThO₂, PuO₂ and UN. Calculated formulas yielded results in consistent with literature.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.377-384
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• 2013

During the last three decades, South Korean nuclear power plants have discharged about 5,950 tons of spent fuel and the maximum burn-up reached 55 GWd/MTU in 2002. This study was performed to support the development of Korean dry spent fuel storage alternatives. First, we chose V5H-$17{\times}17$ and KSFA-$16{\times}16$ as representative domestic spent fuels, considering current accumulation and the future generation of the spent fuels. Examination reveals that their average burn-ups have already increased from 33 to 51 GWd/MTU and from 34.8 to 48.5 GWd/MTU, respectively. Evaluation of the fuel characteristics shows that at the average burn-up of 42 GWd/MTU, the oxide thickness, hydrogen content, and hoop stress ranged from $30{\sim}60{\mu}m$, 250 ~ 500 ppm, and 50 ~ 75 MPa, respectively. But when burn-up exceeds 55 GWd/MTU, those characteristics can increase up to 100 ${\mu}m$, 800 ppm, and 120 MPa, respectively, depending on the power history. These results demonstrate that most Korean spent nuclear fuels are expected to remain within safe bounds during long-term dry storage, however, the excessive hoop stress and hydrogen concentration may trigger the degradation of the spent fuel integrity early during the long-term dry storage in the case of high burn-up spent fuels exceeding 45 GWd/MTU.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.255-266
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• 2021

Solid-state mechanochemical reduction combined with subsequent melting consolidation was suggested as a technical option for the oxide reduction in pyroprocessing. Ni ingot was produced from NiO as a starting material through this technique while Li metal was used as a reducing agent. To determine the technical feasibility of this approach for pyroprocessing, which handles spent nuclear fuels, thermodynamic calculations of the phase stabilities of various metal oxides of U and other fission elements were made when several alkaline and alkali-earth metals were used as reducing agents. This technique is expected to be beneficial, not only for oxide reduction but also for other unit processes involved in pyroprocessing.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4393-4411
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• 2022

In recent years, beryllium oxide has been widely utilized in multiple compact nuclear reactors as the neutron moderator, the neutron reflector or the matrix material with dispersed nuclear fuels due to its prominent properties. In the past 70 years, beryllium oxide has been studied extensively, but rarely been systematically organized. This article provides a systematic review of the application history, thermal properties, mechanical properties, corrosion behavior and fabrication methods of beryllium oxide. Data from previous literature are extracted and sorted out, and all of these original data are attached as the supplementary material, so that subsequent researchers can utilize this paper as a database for beryllium oxide research in reactor design or simulation analysis, etc. In addition, this review article also attempts to point out the insufficiency of research on beryllium oxide, and the possible key research areas about beryllium oxide in the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.193-207
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• 2022

Thermodynamically, TRUO_x, REO_x, and SrO_x can be chlorinated using ammonium chloride (NH₄Cl) as a chlorinating agent, whereas uranium oxides (U₃O₈ and UO₂) remain in the oxide form. In the preliminary experiments of this study, U₃O₈ and CeO₂ are reacted separately with NH₄Cl at 623 K in a sealed reactor. CeO₂ is highly reactive with NH₄Cl and becomes chlorinated into CeCl₃. The chlorination yield ranges from 96% to 100%. By contrast, U₃O₈ remains as UO₂ even after chlorination. We produced U/REO_x- and U/SrO_x-simulated fuels to understand the chlorination characteristics of the oxide compounds. Each simulated fuel is chlorinated with NH₄Cl, and the products are dissolved in LiCl-KCl salt to separate the oxide compounds from the chloride salt. The oxide compounds precipitate at the bottom. The precipitate and salt phases are sampled and analyzed via X-ray diffraction, scanning electron microscope-energy dispersive spectroscopy, and inductively coupled plasma-optical emission spectroscopy. The analysis results indicate that REO_x and SrO_x can be easily chlorinated from the simulated fuels; however, only a few of U oxide phases is chlorinated, particularly from the U/SrO_x-simulated fuels.

• Journal of Energy Engineering
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• v.13 no.4
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• pp.259-266
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• 2004

Specific heat mechanism of oxide fuel is contributed by lattice vibration, dilatation, conduction electron and defect and excess specific heat. Model of oxide fuel for specific heat consists of specific heat at constant pressure term, dilatation specific heat term and defect specific heat term. In this study experimental and published data on the specific heats of oxide nuclear fuels have been reviewed and analyzed to recommend the best fitting model. The oxide fuels considered in this paper were UO$_2$, mixed (U, Pu) oxides and spent fuel. The specific heat data of spent fuel has been replaced by that of simulated fuel.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.863-868
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• 2014

The evaluation of thermal properties of actinide oxide fuels is a problem of high importance for the development of new generation reactors. In the present study, an expression obtained for n-dimensional Debye functions is used to derive a simple analytical expression for the specific heat capacity of nuclear fuels. To test the validity and reliability of this expression, the analytical expression is applied to $UO_2$, $NpO_2$, $ThO_2$, and $PuO_2$. It is seen that the formula was in agreement with the experimental and theoretical results reported in the literature.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.997-1001
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• 2016

Conductive ceramics are being developed to replace current Pt anodes in the electrolytic reduction of spent oxide fuels in pyroprocessing. While several conductive ceramics have shown promising electrochemical properties in small-scale experiments, their long-term stabilities have not yet been investigated. In this study, the chemical stability of conductive $La_{0.33}Sr_{0.67}MnO_3$ in $LiCl-Li_2O$ molten salt at $650^{\circ}C$ was investigated to examine its feasibility as an anode material. Dissolution of Sr at the anode surface led to structural collapse, thereby indicating that the lifetime of the $La_{0.33}Sr_{0.67}MnO_3$ anode is limited. The dissolution rate of Sr is likely to be influenced by the local environment around Sr in the perovskite framework.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1398-1405
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• 2019

We use a molecular dynamics simulation to explore thermal transport in oxide nuclear fuels with point defects. The effect of vacancy and substitutional defects on the thermal conductivity of plutonium dioxide and uranium dioxide is investigated. It is found that the thermal conductivities of these fuels are reduced significantly by the presence of small amount of vacancy defects; 0.1% oxygen vacancy reduces the thermal conductivity of plutonium dioxide by more than 10%. The missing of larger atoms has a more detrimental impact on the thermal conductivity of actinide oxides. In uranium dioxide, for example, 0.1% uranium vacancies decrease the thermal conductivity by 24.6% while the same concentration of oxygen vacancies decreases the thermal conductivity by 19.4%. However, uranium substitution has a minimal effect on the thermal conductivity; 1.0% uranium substitution decreases the thermal conductivity of plutonium dioxide only by 1.5%.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1893-1908
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• 2021

In light of the importance of helium production in influencing the behaviour of fast reactor fuels, in this work we present a burn-up module with the objective to calculate the production of helium in both in-pile and out-of-pile conditions tracking the evolution of 23 alpha-decaying actinides. This burn-up module relies on average microscopic cross-section look-up tables generated via SERPENT high-fidelity calculations and involves the solution of the system of Bateman equations for the selected set of actinide nuclides. The results of the burn-up module are verified in terms of evolution of actinide and helium concentrations by comparing them with the high-fidelity ones from SERPENT, considering two representative test cases of (U,Pu)O₂ fuel in fast reactor conditions. In addition, a code-to-code comparison is made with the independent state-of-the-art module TUBRNP (implemented in the TRANSURANUS fuel performance code) for the same test cases. The herein presented burn-up module is available in the SCIANTIX code, designed for coupling with fuel performance codes.