• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.105-115
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• 2012

Necessity of demonstration test to evaluate the structural integrity of a basket for accident conditions arose during license approval procedure for the WSPP's dry storage facility named MACSTOR/KN-400. A drop test facility for demonstration was constructed in KAERI site and demonstration tests for basket drop were conducted. As the upper welding region of a loaded basket was collided with a dropped basket during the drop test, the welding in this region was fractured and leakage happened after the drop test. The enhancement of basket design was needed since the existing basket design was not able to satisfy the performance requirement. The directions for design modification were determined and six enhanced designs were derived based on these directions. Structural analyses and specimen tests for each enhanced design were conducted. By evaluating structural analysis results and test results, one among six enhanced designs was decided as a final design for revision. The final design was the one to reduce the height of central post of a basket and to decrease the impact velocity with a dropped basket. Test basket models were fabricated with accordance with the final enhanced design. Additional demonstration test was performed for this test model and all the performance requirements were satisfied.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.221-227
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• 2007

Fenton's reagent is applied to directly decompose the ion-exchange resins, IRN-78 and the mixed resin with IRN-77. The newly applied procedures is to dry the resin first and the catalyst solution is completely absorbed into the resin, then a limited dose of $H_2O_2$ is introduced for an effective reaction between the reagents within the resin. As a characteristic on the decomposition of IRN-78, the resin mixture should be heated to $40^{\circ}C$ to induce the initial reaction and lag time is also needed for about 20 minutes until the main reaction occurs. The effectiveness of the decomposition is investigated using $CuSO_4,\;Cu(NO_3)_2\;and\;FeSO_4$ as a catalyst and the decomposition rate is compared depending on the concentration of each catalyst and the amount of $H_2O_2$. The most effective catalyst was found to be $FeSO_4$ for IRN-78 alone and the mixed resin with IRN-77, and $FeSO_4$ showed a special effect that the reaction was initiated without heating and a lag time. Furthermore, the optimum concentration of the catalyst for each resin and the mixed one is suggested in the view point of the amount of $H_2O_2$ needed and the stability of the decomposition reaction.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.57-63
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• 2010

Layer crystallization process was tested for the separation(or concentration) of cesium and strontium fission products in a LiCl waste salt generated from an electrolytic reduction process of a spent oxide fuel. In a crystallization process, impurities (CsCl and $SrCl_2$) are concentrated in a small fraction of the LiCl salt by the solubility difference between the melt phase and the crystal phase. Based on the phase diagram of LiCl-CsCl-$SrCl_2$ system, the separation possibility by using crystallization was determined and the molten salt temperature profile during layer crystallization operation was predicted by using mathematical calculation. In the layer crystallization process, the crystal growth rate strongly affects the crystal structure and therefore the separation efficiency. In the conditions of about 20-25 l/min cooling air flow rate and less than 0.2g/min/$cm^2$ crystal flux, the separation efficiency of both CsCl and $SrCl_2$ showed about 90% by the layer crystallization process, assuming a LiCl salt reuse rate of 90wt%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.59-64
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• 2018

In case of high contents of rare earths in the LiCl-KCl salt, it is not easy to recover U and TRU metals as a usable resource form from LiCl-KCl eutectic salts generated from the pyroprocessing of spent nuclear fuel. In this study, a conversion of $UCl_3$ into an oxide form using $K_2CO_3$ and an electrodeposition of $NdCl_3$ into a metal form in $LiCl-KCl-UCl_3-NdCl_3$ system were conducted to resolve the problem. Before conducting the conversion, experimental conditions for the conversion were determined by performing a thermodynamic equilibrium calculation. In this study, almost all of $UCl_3$ disappeared in the LiCl-KCl salt when the injection of $K_2CO_3$ reached theoretical equivalent for the conversion, and then $NdCl_3$ was effectively electrodeposited as a metal form using liquid zinc cathode. After that, the LiCl-KCl salt became transparent, and uranium oxides were precipitated to the bottom of the LiCl-KCl salt. These results will be utilized in designing a process to separate U and rare earths in LiCl-KCl salt.

• Journal of Radiation Protection and Research
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• v.34 no.4
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• pp.190-194
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• 2009

Gamma spectral results for macroalgae samples taken from the environment of Ulchin nuclear power plants in Korea (east coast), showed 766 keV peaks, which were identified as $^{95}Nb$ by several research institutes. After the enhancement of liquid radioactive waste disposal facility at Ulchin NPP site, the $^{95}Nb$ amount in the liquid radioactive waste outflow has drastically reduced, but the expected reduction in $^{95}Nb$ specific activity from environmental samples did not actually show up on gamma spectroscopy. Detailed re-investigation revealed that along with 766 keV peak, other peaks (63, 92 and 1001 keV) from $^{234}Th-^{234}mPa$ decay series were also detected on spectroscopy, and that the measured half lives of the four peaks were very close to known half life of $^{234}Th-^{234}mPa$ decay series, which is 24.1 day. The measured gamma yield ratios of 766 keV peak to 1001 peak were very close to known ratio 0.35 for $^{234}mPa$. It is concluded that 766 keV peaks on gamma spectroscopy of Ulchin NPP environmental samples were mainly from $^{234}mPa$, which is one of naturally occurring radionuclides.

• Journal of Radiation Protection and Research
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• v.26 no.4
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• pp.393-398
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• 2001

The non-combustible RI wastes disposed of in hospital every year emit ${\gamma}$-ray or ${\beta}$-ray but their activities are very low to the extent of background. Development of more simple methods is needed because the conventional detection methods are so ineffective and complex. In this study, to solve this problem, detection method using efficiency curve for ${\gamma}$-ray emitting radioactive wastes measurement is proposed and experimental detection efficiency equation is also determined through HPGe's standard specimen measurement. For ${\beta}$-emitting radioisotopes detection, new measurement method using detection efficiency estimated by Monte Carlo simulation and SBD measurements is also proposed. According to the results of this paper, the unknown activity of low-level radioactive wastes without LSC requiring the preparation of standard sample and measurement for standard source detection efficiency could be determined efficiently and simply about ${\pm}17%$ in errors by using the theoretical detection efficiency and the SBD measurement result.

• Journal of the Mineralogical Society of Korea
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• v.29 no.2
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• pp.59-71
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• 2016

To investigate the sorption characteristics of Cs, which is one of the major isotopes of nuclear waste, on natural zeolite chabazite, XRD, EPMA, EC, pH, and ICP analysis were performed to obtain the informations on chemical composition, cation exchange capacity, sorption kinetics and isotherm of chabazite as well as competitive adsorption with other cations ($Li^+$, $Na^+$, $K^+$, $Rb^+$, $Sr^{2+}$). The chabazite used in this experiment has chemical composition of $Ca_{1.15}Na_{0.99}K_{1.20}Mg_{0.01}Ba_{0.16}Al_{4.79}Si_{7.21}O_{24}$ and its Si/Al ratio and cation exchange capacity (CEC) were 1.50 and 238.1 meq/100 g, respectively. Using the adsorption data at different times and concentrations, pseudo-second order and Freundlich isotherm equation were the most adequate ones for kinetic and isotherm models, indicating that there are multi sorption layers with more than two layers, and the sorption capacity was estimated by the derived constant from those equations. We also observed that equivalent molar fractions of Cs exchanged in chabazite were different depending on the ionic species from competitive ion exchange experiment. The selectivity sequence of Cs in chabazite with other cations in solution was in the order of $Na^+$, $Li^+$, $Sr^{2+}$, $K^+$ and $Rb^+$ which seems to be related to the hydrated diameters of those caions. When the exchange equilibrium relationship of Cs with other cations were plotted by Kielland plot, $Sr^{2+}$ showed the highest selectivity followed by $Na^+$, $Li^+$, $K^+$, $Rb^+$ and Cs showed positive values with all cations. Equilibrium constants from Kielland plot, which can explain thermodynamics and reaction kinetics for ionic exchange condition, suggest that chabazite has a higher preference for Cs in pores when it exists with $Sr^{2+}$ in solution, which is supposed to be due to the different hydration diameters of cations. Our rsults show that the high selectivity of Cs on chabazite can be used for the selective exchange of Cs in the water contaminated by radioactive nuclei.