• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.147-157
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• 2004

The production of $^{14}C$ occurs in the Moderator(MOD), Primary Heat Transport System (PHTS), Annulus Gas System(AGS) and Fuel in the CANDU reactor. Among the four systems, The MOD system is the largest contributor to $^{14}C$ production(approximately 94.8%). $^{14}C$ is distributed of $^{14}CO_2$, $H_2^{14}CO_3$, $H^{14}{CO_3}^-$ and $^{14}{CO_3}^{2-}$ species as a function of the pH of water. Of these species, $H_2^{14}CO_3$ and $H^{14}{CO_3}^-$ form are predominant because the pH of MOD system is ＞ 5. In this paper, adsorption-desorption characteristics of bicarbonate ion (${HCO_3}^-$) by IRN 150 resin was investigated. ${HCO_3}^-$ ion existed in neutral condition(app. pH 7)was reacted with ion exchange resin (IRN-150) and saturated with it. Then $NaNO_3$ and $Na_3PO_4$ solutions selected as extraction materials were used to make an investigation into feasibility of ${HCO_3}^-$ extraction from resin saturated with ${HCO_3}^-$. Desorption of $CO^{2+}$ and $Cs^+$ ion by $Na^+$ ion was not occurred, and desorption of ${HCO_3}^-$ ion by ${NO_3}^-$ and ${PO_4}^{3-}$ was occurred slowly. Also, the status of ion exchange which is used in Wolsong NPPs and generation of spent resin yearly were surveyed.

• Tunnel and Underground Space
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• v.11 no.3
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• pp.257-269
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• 2001

Numerical simulation is performed to understand the structural behavior of an underground radwaste repository, assumed to be located at the depth of 500 m, in a granitic rock mats, in which a fault intersects the roof of the repository cavern. Two dimensional universal distinct element code, UDEC is used in the analysis. The numerical model includes a granitic rock mass, a canister with PWR spent fuels surrounded by the compacted bentonite inside the deposition hole, and the mixed bentonite backfilled in the rest of the space within the repository cavern. The structural behavior of three different cases, each case with a fault of an angle of $33^{\circ},\;45^{\circ},\;and\;58^{\circ}$ passing through the cavern roof-wall intersection, has been compared. And then fro the case with the $45^{\circ}$ fault, the hydro-mechanical, thermo-mechanical, and thermo-hydro-mechanical interaction behavior have been studied. The effect of the time-dependent decaying heat, from the radioactive materials in PWR spent fuels, on the repository and its surroundings has been studied. The groundwater table is assumed to be located 10m below the ground surface, and a steady state flow algorithm is used.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.345-352
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• 2006

The cemented and paraffin wastes form which are incorporated the concentrated wastes, the cemented waste form which is incorporated the spent ion-exchange resins, and the miscellaneous waste(decontamination paper) were irradiated up to $10^{+8}$ rads at $5.43{\times}10^{+5}$ rads/hr with Co-60(72,023.9 Ci) as an external irradiation source. As a result, the radiolysis gases such as $H_2,\;CH_4,\;N_2,\;C_2H_6,\;O_2,\;CO\;and\;CO_2$, were measured in all the wastes. The major gas which was generated in all the wastes was hydrogen($H_2$). The volume of the generated gases showed a difference from $0.029{\sim}0.788\;cm^3.atm/1.1g$ according to the type of wastes, and more was generated in the cemented waste form incorporated a spent ion-exchange resin than in the other wastes. More hydrogen($H_2$) gas was generated in the decontamination paper waste than in the other wastes, and the G($H_2$) value was 0.12.

• Tunnel and Underground Space
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• v.29 no.4
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• pp.215-229
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• 2019

This technical note describes the current status of Finnish radioactive waste disposal project which started to construct the repository for spent nuclear waste for the first time in the world. Finland started operating nuclear power plant in 1977 and is currently operating four nuclear power plants. After detailed site surveys started in 1993, Olkiluoto was finally selected by the parliament of Finland as the site for geological disposal in 2001 followed by a construction license in 2015. If the operating license is approved by the government in the 2020s, it would be the world's first case of geological disposal. In ONKALO, a site-specific underground research facility at the site of Olkiluoto, various studies were conducted to verify the safety of the repository. Finland uses the KBS-3 disposal concept, and Korea considers a similar disposal concept because of similar rock formations. The entire process in Finland including the operation status of intermediate and low-level waste disposal, site investigation and selection stages, and the latest rock mechanics and hydrogeological studies in ONKALO are presented. Suggestions for the radioactive waste disposal in Korea is given based on the Finnish case.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.179-189
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• 2014

Pyroprocessing has been developed for the purpose of resolving the current spent nuclear fuel management issue and enhancing the recycle of valuable resources. An electrolytic reduction of the pyroprocessing is a process to reduce oxides into metals using LiCl as an electrolyte and requires a post-treatment process due to the inclusion of residual salt in porous metal products. A vacuum distillation has been adopted for various molten salt systems and could be applied to the post-treatment process of the electrolytic reduction. The residual salt in the metal products includes LiCl, alkali chlorides, and alkaline earth chlorides. In this paper, vapor pressures of chlorides have been estimated and the composition changes on the residual liquid during the vacuum distillation process have been calculated. A model combining a material balance and vapor-liquid equilibrium relations has been proposed under a constant vapor discharging flow rate and liquid composition changes have been calculated using the vapor pressures with respect to a dimensionless time. The behaviors have been compared with temperature and molten salt composition changes to simulate the process condition variation. The distillation of the residual salt has been dominated by LiCl which is the main component of the salt and CsCl of which vapor pressure is higher than that of LiCl would be readily removed. RbCl exhibits similar vapor pressure with LiCl and maintains its composition. However, $SrCl_2$ and $BaCl_2$ of which vapor pressures are much lower than that of LiCl are concentrated with time and expected to be possibly precipitated during the distillation when the initial compositions are increased.

• Journal of Radiation Protection and Research
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• v.9 no.1
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• pp.33-42
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• 1984

A great variety of nuclear gamma rays emitted from fission and activation products of spent nuclear fuel contains much information that can be elicited without affecting the integrity of the fuel elements. But the extraction of such information from the complex spectrum is difficult and requires computer codes. In the present work, a versatile code 'CAERI' was developed which locates peaks and calculates their areas for X-rays as well as gamma rays using elegant features of some widely used programs for gamma-ray peak fitting. 'CAERI' coded in FORTRAN used infinite series approximation more accurate than other workers various, simple, piecewise series approximations for evaluations of the Voigt function which represents the X-ray peak with non-negligible natural line width. 'CAERI' can handle even a complex multiplet consisting of peaks from X-rays and gamma rays in arbitrary mixture, which one often encounters in the isotopic analysis of heavy elements such as U and Pu. The results of the fitting performed on the test spectra of $^{177m}\;Lu\;{\gamma}-ray\;and\;^{235}U\;K_{\alpha}$X-ray show good agreement with those by previous workers.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.57-65
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• 2005

This study suggested a new method for the solidification of molten salt waste generated from the electro-metallurgical process in the spent fuel treatment. Using binary material system, sodium silicate and phosphoric acid, metal chlorides were converted into metal phosphate in the micro-reaction module formed by SiO$_{2} particles. The volatile element in the reaction module would little vaporized below 1100$^{circ}$C After the gel product was mixed with borosilicate glass powder and thermally treated at 1000$^{circ}$C, li exists as Li$_{3}$PO$_4$ separated from glass phase and, Cs and Sr would be incorporated into an amorphous phase from XRD analysis. In case of the addition of ZrCl$_{4}$ to the binary system, the gel products were transformed into NZP structure considered as an prospective ceramic waste form after heat-treatment above 700 $^{circ}$C. From these results, the gel-route pretreatment can be considered as an effective approach to the solidincation of molten salt waste by the confirmed process or waste form and this also would be an alternative method on the ANL method using zeolites in USA by the confirmation of its chemical durability as an future work.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.309-322
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• 2007

This study deals with the irradiation effects on some selected components which are being used in an Advanced Spent Fuel Conditioning Process (ACP). Irradiation test components have a higher priority from the aspect of their reliability because their degradation or failure is able to critically affect the performance of an ACP equipment. Components that we chose for the irradiation tests were the AC servo motor, potentiometer, thermocouples, accelerometer and CCD camera. ACP facility has a number of AC servo motors to move the joints of a manipulator and to operate process equipment. Potentiometers are used for a measurement of several joint angles in a manipulator. Thermocouples are used for a temperature measurement in an electrolytic reduction reactor, a vol-oxidation reactor and a molten salt transfer line. An accelerometer is installed in a slitting machine to forecast an incipient failure during a slitting process. A small CCD camera is used for an in-situ vision monitoring between ACP campaigns. We made use of a gamma-irradiation facility with cobalt-60 source for an irradiation test on the above components because gamma rays from among various radioactive rays are the most significant for electric, electronic and robotic components. Irradiation tests were carried out for enough long time for total doses to be over expected threshold values. Other components except the CCD camera showed a very high radiation hardening characteristic. Characteristic changes at different total doses were investigated and threshold values to warrant at least their performance without a deterioration were evaluated as a result of the irradiation tests.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.55-61
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• 2013

The recovery of Zr from Zircaloy-4 (Zry-4) cladding hulls using a chlorination method was demonstrated for complete conversion of Zr into $ZrCl_4$. A chlorination reaction was performed by reacting Zry-4 hulls for 8 h under a 70 cc/min $Cl_2$ + 70 cc/min Ar flow at $380^{\circ}C$. The initial weight of the reactant (51.7 g) decreased to 0.49 g after 8 h of operation, which is only 0.95wt% of the initial weight. The weight of the total reaction products was 121.7 g with a high Zr purity of 99.80wt%. Fe and Sn were identified as major (0.18wt%) and minor (0.02wt%) impurities of the reaction products, respectively. It was also shown that Zr exhibited a high recovery ratio of 96.95wt% with a relatively small experimental loss of 2.34wt%. Observation of the reaction residues revealed that the chlorination reaction was dominant along the longitudinal direction, and surface oxide layers remained as reaction residues. The high purity and recovery ratio of Zr proposed the feasibility of the chlorination technique as an effective hull waste treatment method.

• Journal of the Korean Society for Marine Environment & Energy
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• v.10 no.2
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• pp.112-117
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• 2007

This study aims to develop an alternative energy like oil made from marine organic waste by marine products waste, spent fishing nets. There are already many commercial examples and case studies based on the petroleum industry-refuse plastic or refuse tire, however, it is rare that a research developing alternative energy from food waste and organic waste. Therefore, this study investigated the oil made from thermal decomposition under the high temperature and high pressure condition, and examined the possibility for commercial use by testing its own characteristics. A bio-oil from thermal decomposition at $250^{\circ}C$ and 40 atm was hard to remove impurities because of its high viscosity, showed lower caloric value than heavy oil, and generated various gases which were not appropriate for the use of fuel. It is noticeable that thermal decomposition was occurred at $250{\pm}5^{\circ}C$ using steam pressure, which much lower compared to the existing method of thermal decomposition, more than $500^{\circ}C$. Since the high viscosity of bio-oil, it is necessary a further study to use as liquid fuel.