• Journal of Powder Materials
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• v.23 no.5
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• pp.384-390
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• 2016

Tri-isotropic (TRISO) coatings on zirconia surrogate beads are deposited using a fluidized-bed vapor deposition (FB-CVD) method. The silicon carbide layer is particularly important among the coated layers because it acts as a miniature pressure vessel and a diffusion barrier to gaseous and metallic fission products in the TRISO-coated particles. In this study, we obtain a nearly stoichiometric composition in the SiC layer coated at $1400^{\circ}C$, $1500^{\circ}C$, and $1400^{\circ}C$ with 20 vol.% methyltrichlorosilane (MTS), However, the composition of the SiC layer coated at $1300-1350^{\circ}C$ shows a difference from the stoichiometric ratio (1:1). The density decreases remarkably with decreasing SiC deposition temperature because of the nanosized pores. The high density of the SiC layer (${\geq}3.19g/cm^2$) easily obtained at $1500^{\circ}C$ and $1400^{\circ}C$ with 20 vol.% MTS did not change at an annealing temperature of $1900^{\circ}C$, simulating the reactor operating temperature. The evaluation of the mechanical properties is limited because of the inaccurate values of hardness and Young's modulus measured by the nano-indentation method.

• Analytical Science and Technology
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• v.16 no.2
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• pp.110-116
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• 2003

A study has been done on the separation of iodide from spent pressurized water reactor (PWR) fuels and its quantitative determination using ion chromatography. Spent PWR fuels were dissolved with mixed acid of nitric and hydrochloric acids (80 : 20 molL%) which can oxidize iodide to iodate to prevent it from be vaporized. After reducing ${IO_3}^-$ ­to $I_2$ in 2.5 M $HNO_3$ with $NH_2OH{\cdot}HCl$, Iodine was selectively separated from actinides and all other fission products with carbontetrachloride and back-extracted with 0.1 M $NaHSO_3$. Recovered iodide was determined using the ion chromatograph of which the column was installed in a glove box for the analysis of radioactive materials. In practice, spent PWR fuel with 42,000~44,000 MWd/MtU was analyzed and its quantity was compared to that calculated by burnup code, ORIGEN2. The agreement was achieved with a deviation of -8.3~-0.5% from the ORIGEN 2 data, $324.5{\sim}343.6{\mu}g/g$.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.277-286
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• 1989

Destructive methods are used for the turnup determination of an irradiated PWR fuel. One of the methods includes U, Pu, Nd-148 and Nd-(145+146) determination by an isotope dilution mass spectrometry using triple spikes (U-233, Pu-242 and Nd-150). The method involves two sequential ion exchange resin separation procedures. Pu is eluted from the first anion exchange resin column (Dowex AG 1$\times$8) with 12 M HCl-0.1 M HI mixed solution, followed by U elution with 0.1 M HCl. Nd is isolated from other fission products on the second anion exchange resin column (Dowex AG 1$\times$4) with a nitric acid-methanol eluent. Each fraction is analysed by thermal ionization mass spectrometry. The difference between Nd-148 and Nd-(145+146) method is found with an average 2.07%. The results are compared with those by the heavy element method using U and Pu isotopes and by the destructive y-spectrometric measurement of Cs-137. The dependences of isotope composition of U and Pu on burn-up, and correlation between those isotopes are illustrated graphically.

• Applied Chemistry for Engineering
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• v.29 no.2
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• pp.127-137
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• 2018

Radioactive contamination has become an important environmental concern after the accident occurred in Fukushima Daiichi Nuclear Power Plants. $^{134}Cs$ and $^{137}Cs$ are the major fission products and they are main problems in radioactive contamination. Huge amounts of Cs were released during the Fukushima Daiichi Nuclear Power Plants accident and as a result of this incident, many researchers focused on the development of adsorbents for decontamination of radiotoxic cesium. This review will critically evaluate recent advances in the preparation of Prussian blue and its analogue compounds, which are promising materials for cesium removal. Furthermore, this review will discuss recent studies on the cesium adsorption using different types of clay and clay based adsorbents and summarize various types of newly developed Cs adsorbents reported in recent years.

• Journal of Powder Materials
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• v.21 no.6
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• pp.434-440
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• 2014

Silicon carbide(SiC) layer is particularly important tri-isotropic (TRISO) coating layers because it acts as a miniature pressure vessel and a diffusion barrier to gaseous and metallic fission products in the TRISO coated particle. The high temperature deposition of SiC layer normally performed at $1500-1650^{\circ}C$ has a negative effect on the property of IPyC layer by increasing its anisotropy. To investigate the feasibility of lower temperature SiC deposition, the influence of deposition temperature on the property of SiC layer are examined in this study. While the SiC layer coated at $1500^{\circ}C$ obtains nearly stoichiometric composition, the composition of the SiC layer coated at $1300-1400^{\circ}C$ shows discrepancy from stoichiometric ratio(1:1). $3-7{\mu}m$ grain size of SiC layer coated at $1500^{\circ}C$ is decreased to sub-micrometer (< $1{\mu}m$) $-2{\mu}m$ grain size when coated at $1400^{\circ}C$, and further decreased to nano grain size when coated at $1300-1350^{\circ}C$. Moreover, the high density of SiC layer (${\geq}3.19g/cm^3$) which is easily obtained at $1500^{\circ}C$ coating is difficult to achieve at lower temperature owing to nano size pores. the density is remarkably decreased with decreasing SiC deposition temperature.

• Analytical Science and Technology
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• v.16 no.6
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• pp.431-437
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• 2003

Metalization yield of uranium oxide to uranium metal from lithium reduction process of spent pressurized water reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metalization yield was measured. Metalization yield of the solid part was 90.7~95.9 wt%, and the powder being 77.8~71.5 wt% individually. Oxidation behaviour of the quartemary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At $600{\sim}700^{\circ}C$, weight increments of alloy of Mo, Ru, Rh and Pd was 0.40~0.55 wt%. Phase change on the surface of the alloy was started at $750^{\circ}C$. In particular, Mo was rapidly oxidized and then the alloy lost 0.76~25.22 wt% in weight.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.63-69
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• 2012

The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of $^{57}Co$ and $^{57}Ni$ were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.675-682
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• 2013

The sequential separation process, composed of an oxygen sparging process for separating lanthanides and a zone freezing process for separating Group I and II fission products, was evaluated and tested with a surrogate eutectic waste salt generated from pyroprocessing of used metal nuclear fuel. During the oxygen sparging process, the used lanthanide chlorides (Y, Ce, Pr and Nd) were converted into their sat-insoluble precipitates, over 99.5% at $800^{\circ}C$; however, Group I (Cs) and II (Sr) chlorides were not converted but remained within the eutectic salt bed. In the next process, zone freezing, both precipitation of lanthanide precipitates and concentration of Group I/II elements were preformed. The separation efficiency of Cs and Sr increased with a decrease in the crucible moving speed, and there was little effect of crucible moving speed on the separation efficiency of Cs and Sr in the range of a 3.7 - 4.8 mm/hr. When assuming a 60% eutectic salt reuse rate, over 90% separation efficiency of Cs and Sr is possible, but when increasing the eutectic salt reuse rate to 80%, a separation efficiency of about 82 - 86 % for Cs and Sr was estimated.

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.31 no.2 s.257
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• pp.277-284
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• 2007

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of the support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the support plate. Such solutions are developed based on three dimensional finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.