• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3704-3710
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• 2021

The characteristics of the external radiation on the surface of the casks for spent nuclear fuel (SNF) storage by dry method are investigated for the case when the spatial distribution of SNF in the basket changes due to the destruction of the fuel rod claddings. The surface areas are determined, where the changes in fluxes of neutrons, produced by ²⁴⁴Cm actinide, and γ-quanta, produced by long-lived isotopes, are maximum in the result of the decrease in the height of the SNF area. Concrete (VSC-24) and metal (SC-21) casks are considered as examples. The procedure of periodic measurement of the dose rate of neutrons or γ-quanta at the specified points of the cask surface is proposed for identifying the fuel rod cladding destruction. Under normal operation, the decrease in the dose rate produced by neutrons as the function of SNF storage duration is determined by the half-life of ²⁴⁴Cm, and for γ-quanta - by the half-lives of long-lived SNF isotopes. Consequently, a stepwise change in the dose rate of neutrons or γ-quanta, detected by the measurements, as compared to the previous one, would indicate the destruction of the fuel rod claddings.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.363-367
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• 2003

Much attention has been given to an electrochemical reduction process for converting uranium oxide to uranium metal in molten salt. The process has the versatility of being adopted for reducing other actinide and rare-earth metals from their oxides. Using the metal oxide to be reduced as a integrated cathode designed originally and inert conductors as anodes, oxygen anions are removed from the cathode and oxidized at the surface of the anodes in a molten salt cell. However, the electrochemical properties of alkali and alkali-earth metal oxides in molten salt have not been investigated thoroughly, which made the process incomplete when it is considered as a unit process in a back-end fuel cycle. It is well known that cesium and strontium Isotopes in spent fuel are main contributors for head load. The properties of cesium, strontium, and barium oxides such as the dissolution rates and reduction potentials in molten LiC1 dissolving $Li_2O$ are examined.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1013-1021
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• 2020

A combination of simultaneous thermal analysis, evolved gas analysis and non-ambient XRD techniques was used to characterise and investigate the conversion reactions of ammonium uranates into uranium oxides. Two solid phases of the ternary system NH₃ - UO₃ - H₂O were synthesised under specified conditions. Microspheres prepared by the sol-gel method via internal gelation were identified as 3UO₃·2NH₃·4H₂O, whereas the product of a typical ammonium diuranate precipitation reaction was associated to the composition 3UO₃·NH₃·5H₂O. The thermal decomposition profile of both compounds in air feature distinct reaction steps towards the conversion to U₃O₈, owing to the successive release of water and ammonia molecules. Both compounds are converted into α-U₃O₈ above 550 ℃, but the crystallographic transition occurs differently. In compound 3UO₃·NH₃·5H₂O (ADU) the transformation occurs via the crystalline β-UO₃ phase, whereas in compound 3UO₃·2NH₃·4H₂O (microspheres) an amorphous UO₃ intermediate was observed. The new insights obtained on these uranate systems improve the information base for designing and synthesising minor actinide-containing target materials in future applications.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1373-1380
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• 2019

Besides promising implications as fertile nuclear materials, thorium carbonitrides are of great interest owing to their peculiar physical and chemical properties, such as high density, high melting point, good thermal conductivity. This paper reports first-principles simulation results on the structural, electronic and magnetic properties of cubic thorium carbonitrides $ThC_xN_{(1-x)}$ (X = 0.03125, 0.0625, 0.09375, 0.125, 0.15625) employing formalism of density-functional-theory. For the simulation of physical properties, we incorporated full-potential linearized augmented plane-wave (FPLAPW) method while the exchange-correlation potential terms in Kohn-Sham Equation (KSE) are treated within Generalized-Gradient-Approximation (GGA) in conjunction with Perdew-Bruke-Ernzerhof (PBE) correction. The structural parameters were calculated by fitting total energy into the Murnaghan's equation of state. The lattice constants, bulk moduli, total energy, electronic band structure and spin magnetic moments of the compounds show dependence on the C/N concentration ratio. The electronic and magnetic properties have revealed non-magnetic but metallic character of the compounds. The main contribution to density of states at the Fermi level stems from the comparable spectral intensity of Th (6d+5f) and (C+N) 2p states. In comparison with spin magnetic moments of ThSb and ThBi calculated earlier with LDA+U approach, we observed an enhancement in the spin magnetic moments after carbon-doping into ThN monopnictide.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.141-154
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• 2015

In general, conventional criticality analysis for spent fuel transport/storage systems have been performed based on the assumption of fresh fuel concerning the potential uncertainties from number density calculations of actinide nuclides and fission products in spent fuel. However, these evaluation methods cause financial losses due to an excessive criticality margin. In order to overcome this disadvantage, many studies have recently been conducted to design and commercialize a transportation and storage cask applied to the Burnup Credit (BUC). This study conducted an assessment to ensure criticality safety for reactor operating parameters, axial burn-up profiles and misload accident conditions, which are the factors that are likely to affect criticality safety when the BUC is applied to the dual-purpose cask under development at the KOrea RADioactive waste agency (KORAD). As a result, it was found that criticality resulting from specific power, changed substantially and relied on conditions of low enrichment and high burn-up. Considering the end effect in the case of high burn-up produced a positive-definite result. In particular, the increment of maximum effective multiplication factors due to misloading was 0.18467, confirming that misload is a factor that must be taken into account when applying the BUC. The results of this study may therefore be utilized as references in developing technologies to apply the BUC to domestic models and operational procedures or preventing any misload accidents during the process of spent fuel loading.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.123-132
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• 2007

This study was performed to evaluate the co- and mutual separation for Am, Cm and RE elements from the simulated multi-component solution equivalent to real HLW level by a Zr-DEHPA(di-(2-ethylhexyl) phosphoric acid containing Zirconium)/$NDD(n-dodecane)-HNO_3$ extraction system. Zr-DEHPA was self-synthesized and the optimal condition of (15g/L Zr-1M DEHPA)/NDD-1M $HNO_3$ was selected taking into consideration of prevention of the third phase, and effects of concentration of DEHPA, nitric acid and impregnant amount of Zr on the co-extraction of Am, Cm and RE. In that condition, the extraction yields were 81% (Am), 85% (Cm), more than 80% (RE elements), 98% (Mo), 85% (Fe), 98% (U), 73% (Np), and less than 5% (other elements) so that the system developed for the co-extraction of Am-Cm/RE was proved to be available. For that, however, U, Np, Mo and Fe was elucidated to have to be removed in advance, and Zr inducing the third phase formation was found to be practically excluded. The co-extracted Am-Cm/RE were sequentially separated in an order of Am-Cm (stripping agent : 0.05 M DTPA-1M Lactic acid of pH 3.6)${\rightarrow}RE$ (stripping agent : 5M $HNO_3$), and then their separation factors were evaluated. At above conditions, Am of 65.4%, Cm of 63.9%, RE (except for Y) of more than 85% were stripped.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.213-218
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• 2009

Electrowinning process recovers uranium with actinide elements from spent fuels and is a key step in the Pyroprocessing because of proliferation resistance. An analysis of heat transfer of the Electrowinning cell was conducted to develop basic tool for designing engineering-scale Electrowinner. For the calculation of the heat transfer, ANSYS CFX commercial code was adapted. As a result of the calculation, the vertical Heating Zone length had great effect upon temperature of LiCl-KCl eutectic salt. To maintain constant temperature in the salt, the Heating Zone length should be three times longer than the height of the salt. However, the argon and salt temperatures were barely affected by the Cooling Zone length. The temperature under the Cell cover was mainly influenced by the number of the cooling plates. When the cooling plates were installed more than the number of 5, temperature under the cover was maintained below $250^{\circ}C$. These temperature properties had similar tendency toward the temperature of the Cell which was measured from experiments, Simulated heat transfer information from this study could be used to design engineering-scale Electrowinner.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.163-174
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• 2006

The $Ph{\acute{e}}bus$ FP project is an international reactor safety project. Its main objective is to study the release, transport and retention of fission products in a severe accident of a light water reactor (LWR). The FPT4 test was performed with a fuel debris bed geometry, to look at late phase core degradation and the releases of low volatile fission products and actinides. Post Test Analyses results indicate that releases of noble gases (Xe, Kr) and high-volatile fission products (Cs, I) were nearly complete and comparable to those obtained during $Ph{\acute{e}}bus$ tests performed with a fuel bundle geometry (FPT1, FPT2). Volatile fission products such as Mo, Te, Rb, Sb were released significantly as in previous tests. Ba integral release was greater than that observed during FPT1. Release of Ru was comparable to that observed during FPT1 and FPT2. As in other $Ph{\acute{e}}bus$ tests, the Ru distribution suggests Ru volatilization followed by fast redeposition in the fuelled section. The similar release fraction for all lanthanides and fuel elements suggests the released fuel particles deposited onto the plenum surfaces. A blockage by molten material induced a steam by-pass which may explain some of the low releases. The revaporisation testing under different atmospheres (pure steam, $H_2/N_2$ and steam /$H_2$) and up to $1000^{\circ}C$ was performed on samples from the first upper plenum. These showed high releases of Cs for all the atmospheres tested. However, different kinetics of revaporisation were observed depending on the gas composition and temperature. Besides Cs, significant revaporisations of other elements were observed: e.g. Ag under reducing conditions, Cd and Sn in steam-containing atmospheres. Revaporisation of small amounts of fuel was also observed in pure steam atmosphere.

• Journal of the Korean Chemical Society
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• v.27 no.4
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• pp.279-286
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• 1983

An apparatus was made for the electrodeposition of alpha emitting actinide nuclides, $^{207}Bi$ and $^{210}Po$. The electrodeposition was made on a polished stainless steel plate cathode. The anode was made of platinum wire and to stir the solution. With the ammonium chloride as electrolyte initial pH = 4, chloride concentration = 0.6M and solution volume = 15ml, a current of 1.5 ampere(current density = 0.59A/$cm^2$) was flowed for 100 minutes for the quantitative recovery of electrodeposition and on average recovery of 98.3% was obtained within ${\pm}$0.7% uncertainty. Alpha spectrometry of the electrodeposited sample showed alpha peaks from $^{210}Po, ^{234}U$ and $^{239}Pu$ having energy resolution (FWHM) of 18.3, 21.8 and 36.0 keV respectively. The electrodeposition and alpha spectrometry for a natural uranium sample of domestic origin gave $^{238}U : ^{234}U = 1 : 6.1{\times}10^{-5}$ and for a neutron-irradiated uranium sample did $^{238}U : ^{239}Pu : ^{241}Am = 100 : 0.0263 : 5.20{times}10^{-5}$. The result of $^{238}U$ determination in the irradiated sample by electrodeposition-alpha spectrometry was in accord within ${\pm}1.6%$ of relative error with the results of solid fluorimetry and mass spectrometry. For $^{239}Pu$ the result of electrodeposition-alpha spectrometry was in accord within ${\pm}$4.0% of relative error with the results of anion exchange separation and the thenoyltrifluoroacetone(TTA) extraction both followed by alpha spectrometries.

• Economic and Environmental Geology
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• v.38 no.2 s.171
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• pp.187-196
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• 2005

Structural sites which cations can occupy in garnet structure are centers of the tetrahedron, octahedron, and distorted cube sharing edges with the tetrahedron and octahedron. Among them, the size of cation occuping at tetrahedral site (the center of tetrahedron) is closely related with the size of a unit cell of garnet. Accordingly, garnet containing iron with relative large ionic radii in tetrahedral site can be considered as a promising matrix for the immobilization of the elements with large ionic radii, such as actinides in radioactive wastes. We synthesized several garnets with the batch composition of $Ca_{1.5}GdCe_{0.5}ZrFeFe_3O_{12}$, and studied their properties and phase relations under various conditions. Mixed samples were fabricated in a pellet form under a pressure of $200{\~}400{\cal}kg/{\cal}cm^2$ and were sintered in the temperature range of $1100\~1400^{\circ}C$ in air and under oxygen atmospheres. Phase identification and chemical analysis of synthesized samples were conducted by XRD and SEM/EDS. In results, garnet was obtained as the main phase at $1300^{\circ}C$, an optimum condition in this system, even though some minor phases like perovskite and unknown phase were included. The compositions of garnet and perovskite synthesized from the batch composition of $Ca_{1.5}GdCe_{0.5}ZrFeFe_3O_{12}$ were ranged $[Ca_{l.2-1.8}Gd_{0.9-1.4}Ce_{0.3-0.5}]^{VIII}[Zr_{0.8-1.3}Fe_{0.7-1.2}]^{VI}[Fe_{2.9-3.1}]^{IV}O_{12}$ and $Ca_{0.1-0.5}Gd_{0.0-0.8}Ce_{0.1-0.5}\;Zr_{0.0-0.2}Fe_{0.9-1.1}O_3$, respectively. Ca content was exceeded and Ce content was depleted in the 8-coordinated site, comparing to the initial batch composition. This phenomena was closely related to the content of Zr and Fe in the 6-coordinated site.