• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4159-4166
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• 2023

Laser decontamination of radioactive surfaces is an innovative technology. Our contribution to improving this technology includes studies on laser beam decontamination with a pulsed laser of an average power of 150 W, equipped with a hand guided working head. Our investigations are focused on metallic surfaces typical in nuclear power plants, such as stainless steel, bright and rusted mild steel, galvanized steel, and painted steel. As typical nuclides of contaminated surfaces we chose Co-60 and Cs-137, the most frequently occurring nuclides in many nuclear plant components; Sr-85 as a representative of Sr-90, the potentially most harmful fission nuclide; and Am-241 as a representative of the minor alpha-radiation emitting actinides. Here, we present our results of decontamination and recovery ratios. Decontamination ratios of 90-100% were achieved on different surfaces.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.47-58
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• 2015

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. Thefsensitivity of cooling time before prior to pyro-processing was studied. As the cooling time sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ${\leq}20%$ in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

• 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.

• Journal of the Mineralogical Society of Korea
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• v.16 no.4
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• pp.307-320
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• 2003

Garnet has been considered as a possible matrix for the immobilization of radioactive actinides. It is expected that Fe­based garnet be able to have the high substitution ability of actinide elements because ionic radius of Fe in tetrahedral site is larger than that of Si of Si­based garnet. Accordingly, we synthesized Fe­garnet with the batch composition of $Ca_{2,5}$C $e_{0.5}$Z $r_2$F $e_3$ $O_{12}$ and $Ca_2$CeZrFeF $e_3$ $O_{12}$ and studied their phase relations and properties. Mixed samples were fabricated in pellet forms under the pressure of 400 kg/$\textrm{cm}^2$ and were sintered in the temperature range of 1100∼140$0^{\circ}C$ in atmospheric conditions. Phase identification and chemical composition of synthesized samples were analyzed by XRD and SEM/EDS. In results, where the compounds were sintered at 130$0^{\circ}C$, we optimally obtained Fe­garnets as the main phase, even though some minor phases like perovskite were included. The compositions of Fe­garnets synthesized from the batch compositions of $Ca_{2,5}$C $e_{0.5}$Z $r_2$F $e_3$ $O_{12}$ and $Ca_2$CeZrFeF $e_3$ $O_{12}$, are $Ca_{2.5­3.2}$C $e_{0.3­0.7}$Z $r_{1.8­2.8}$F $e_{1.9­3.2}$ $O_{12}$ and $Ca_{2.2­2.5}$C $e_{0.8­1.0}$Z $r_{1.3­1.6}$ F $e_{0.4­.07}$ F $e_{3­3.2}$ $O_{12}$, respectively. Ca contents were exceeded and Ce contents were exceeded or depleted in 8­coodinated site, comparing to the initial batch composition. These results were caused by the compensation of the difference of ionic radius between Ca and Ce.