• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.12 no.2
• /
• pp.121-133
• /
• 2014

If the spent fuels or the high-level radioactive wastes can be disposed of in the depth of 3~5 km and more stable rock formation, it has several advantages. For example, (1)significant fluid flow through basement rock is prevented, in part, by low permeability, poorly connected transport pathways, and (2)overburden self-sealing. (3)Deep fluids also resist vertical movement because they are density stratified and reducing conditions will sharply limit solubility of most dose-critical radionuclides at the depth. Finally, (4) high ionic strengths of deep fluids will prevent colloidal transport. Therefore, as an alternative disposal concept to the deep geological disposal concept(DGD), very deep borehole disposal(DBD) technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, for the preliminary applicability analyses of the DBD system for the spent fuels or high level wastes, the DBD concepts which have been developed by some countries according to the rapid advance in the development of drilling technology were reviewed. To do this, the general concept of DBD system was checked and the study cases of foreign countries were described and analyzed. These results will be used as an input for the analyses of applicability for DBD in Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.12 no.2
• /
• pp.107-119
• /
• 2014

The worldwide Status-of-Art survey for the in-situ experiments of the engineered barrier system for HLW underground disposal was performed as a preliminary action for the design of the in-situ demonstration at KURT. Some nations, which have executed or is ongoing the in-situ experiments at their underground research facilities, were summarized in this review. The demonstration projects reviewed were TBT/Sweden-France, LOT/Sweden, HE-E/Switzerland, PRACLAY/Belgium, FEBEX/Spain, HORONOBE/Japan, and BCE/Canada. The investigated items for the projects were mainly their purposes, constitutional structures, test conditions, monitoring parameters and the measuring tools, and test results. In this review, the hardware design and the assembling of the test system were more concentrated rather than their experimental results, because the purpose of this review is to achieve the necessary information for the practical design of the in-situ experiment to be installed at KURT. A mid scale in-situ demonstration of EBS at KURT, that is called IN-DBES, will be launched right after the completion the expanding project of KURT in 2015. It is hoped that the structural design, installing methods, hardware equipments required in the establishment of IN-DEBS will be referred on this review.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.18 no.1
• /
• pp.73-81
• /
• 2020

Strontium (⁹⁰Sr) and nickel (⁵⁹Ni) have been considered as key radionuclides in the safety assessment of radioactive waste disposal. Through various efforts to impede the migration of radioactive nuclides underground, it has been established that some minerals generated from the corrosion of the waste containers have a positive chemical interaction with these radionuclides. Among these minerals we selected mackinawite (FeS), an iron and sulfur compound, and performed a sorption experiment for the Sr and Ni in FeS under anoxic and alkaline conditions by reflecting deep underground environments. The effects of pH on sorption were likewise investigated in the pH range of 8 ~ 12. As a result, it was found that strontium failed to exhibit a good sorption capacity in a weak alkaline range, while nickel showed a noticeably higher sorption affinity over the entire experimental pH range. Moreover, we determined that as the pH increased in the solution, the distribution coefficients (K_d) were increased for both nuclides, which reflects when an alkalinity increses, the surface of the mineral charges much negatively by detaching the hydrogen or cations on the mineral surface. Thus, it can be concluded that the cationic nuclides of Sr and Ni can attach easily to the mineral under strong alkalinity.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.18 no.1
• /
• pp.31-41
• /
• 2020

We studied the anisotropic shrinkage and deformation characteristics of large size sintered bodies in the manufacturing of glass-ceramic composite wasteform. We used uranium-bearing waste, generated from the treatment of spent uranium catalyst. Sintered specimens were prepared in several forms, comprising a circular disk, and a quarter disk in several diameters of up to 40 cm. Regardless of form or size, the sintered bodies had high isotropic shrinkage when they were fabricated using green bodies prepared at 60 MPa. The average anisotropy rate and average shrinkage rate were 1.6%, and 37.4%, respectively. We confirmed that the glass-ceramic composite wasteform in a large scale disk-type for packing in a 200 L drum could be fabricated with a tolerable anisotropy shrinkage. This has resulted in a significant reduction in the volume of radioactive waste to be disposed of with highly stable wasteform.

• Korean Journal of Materials Research
• /
• v.8 no.12
• /
• pp.1099-1109
• /
• 1998

To investigate the effect of V and Sb on the corrosion behavior of Zr- based alloys, corrosion tests were performed on 6 kinds of Zr alloys in an autoclave at $360^{\circ}C$ for 100 days. The transition of the corrosion rate occurred in the sample containing 0.1wt.%V after 10 days but did not occur in the samples containing 0.2wt.%V and 0.4wt.%V. The corrosion resistance of V containing alloys increased with increasing V contents from 0.1 to 0.4wt.% and the alloys containing 0.4wt.%V showed the best corrosion resistance. In the ternary alloys containing 0.1wt.%Sb and 0.4wt.%Sb, the corrosion rate increased significantly from the short exposure time. It was observed that the optimal Sb content for corrosion resistance was 0.2wt.%. The size and volume fraction of precipitates increased with increasing V and Sb contents. The superior corrosion resistance was observed in the Zr alloy having precipitate size of 0.11-0.13$\mu\textrm{m}$. From the result of corrosion behavior and the obserbation of precipitates, the optimal size of the precipitate appear to control the electron conduction in the cathodic reaction and play an important role in maintaining a stable oxide microstructure.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.1
• /
• pp.41-48
• /
• 2010

Uranium deposits from an electrorefining process contain about 30% salt. In order to recover pure uranium and transform it into an ingot, the salts have to be removed from the uranium deposits. Major process variables for the salt distillation process of the uranium deposits are hold temperature and vacuum pressure. Effects of the variables on the salt removal efficiency were studied in the previous study[1]. By applying the Hertz-Langmuir relation to the salt evaporation of the uranium deposits, the evaporation coefficients were obtained at the various conditions. The operational conditions for achieving above 99% salt removal were deduced. The salt distilled uranium deposits tend to form the eutectic melt with iron, nickel, chromium for structural material of salt evaporator. In this study, we investigated the hold temperature limitation in order to prevent the formation of the eutetic melt between urnaium and other metals. The reactions between the uranium metal and stainless steel were tested at various conditions. And for enhancing the evaporation rate of the salt and the efficient recovery of the distilled salt, the thermal analysis of the salt distiller was conducted by using commercial CFX software. From the thermal analysis, the effect of Ar gas flow on the evaporation of the salt was studied.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.1
• /
• pp.1-7
• /
• 2010

In the present work, an electrowinning process in the LiCl-KCl/Cd system is considered to model and analyze the electrotransport of the actinide and rare-earth elements. A simple dynamic modeling of this process was performed by taking into account the material balances and diffusion-controlled electrochemical reactions in a diffusion boundary layer at an electrode interface between the molten salt electrolyte and liquid cadmium cathode. The proposed modeling approach was based on the half-cell reduction reactions of metal chloride occurring on the cathode. This model demonstrated a capability for the prediction of the concentration behaviors, a faradic current of each element and an electrochemical potential as function of the time up to the corresponding electrotransport satisfying a given applied current based on a galvanostatic electrolysis. The results of selected case studies including five elements (U, Pu, Am, La, Nd) system are shown, and a preliminary simulation is carried out to show how the model can be used to understand the electrochemical characteristics and provide better information for developing an advanced electrowinner.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.1
• /
• pp.65-70
• /
• 2010

The distillation rate on LiCl-KCl eutectic salt under different vacuums from 0.5-50 mmHg was first investigated by using both a non-isothermal and a isothermal thermogravimetric (TG) analysis. Based on the non-isothermal TG data, distillation rate equations as a function of the temperature could be derived. Calculated flux by these model flux equations was in agreement with the distillation rate obtained from isothermal TG analysis. A distillation rate of $10^{-4}-10^{-5}$ mole $cm^{-2}sec^{-1}$ is obtainable at temperatures less than 1300K and vacuums of 0.5-50 mmHg. About a 99% salt distillation efficiency was obtained after an hour at a temperature above 1150 K under 50 mmHg in a small scale distillation test system. An increase in the vaporizing surface area is relatively effective for removing residual salt in the remaining particles, when compared to that for the vaporizing time. Over 99.95% of total distillation efficiency was obtained for a 1-h distillation operation by increasing the inner surface area from $4.52cm^2$ to $12.56cm^2$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.12 no.1
• /
• pp.69-77
• /
• 2014

Unsaturated hydraulic conductivity of the bentonite-buffer was evaluated using the relative humidity data. The method for calculating unsaturated hydraulic conductivity was deduced from the general analytical equation representing the movement of water in unsaturated media, which was applied to the experimental results of water infiltration tests for identifying the behavior of unsaturated hydraulic conductivity according to the water saturation. Unlike the saturated condition, the hydraulic gradient and water flux were irregularly changed, and the unsaturated hydraulic conductivity was increased with increasing the experimental time. Swelling of bentonite grains due to the water absorption increased the volume and size of pore within bentonite, resulting in the increase of water velocity and unsaturated hydraulic conductivity. This result suggested the necessity of further investigation on the correlation between the swelling degree of bentonite-buffer and unsaturated hydraulic conductivity. The method used in this study can be useful technique for evaluating long-term hydraulic performance of bentonite-buffer in the radioactive waste disposal system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.12 no.1
• /
• pp.59-68
• /
• 2014

The public's access to the disposal facilities should be restricted during the institutional control period. Even after the institutional control period, disposal facilities should be designed to protect radiologically against inadvertent human intruders. This study is to assess the effective dose equivalent to the inadvertent intruder after the institutional control period thorough the GENII. The disposal unit was allocated with different kind of radioactive waste and the effects of the radiation dose to inadvertent intruder were evaluated in accordance with the institutional control period. As a result, even though there is no institutional control period, all were satisfied with the regulatory guide, except for the disposal unit with only spent filter. However, the disposal unit with only spent filter was satisfied with the regulatory guide after the institutional control period of 300 years. But the disposal unit with spent filter mixed with dry active waste could shorten the institutional control period. So the institutional control period can be reduced through the mixing the other waste with spent filter in disposal unit. Therefore, establishing an appropriate plan for the disposal unit with spent filter and other radioactive waste will be effective for radiological safety and reduction of the institutional control period, rather than increasing the institutional control period and spending costs for the maintenance and conservation for the disposal unit with only spent filter.