• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.3
• /
• pp.161-165
• /
• 2009

To evaluate the efficiency of the decontamination plant for the removal of uranium compounds deposited on the internal surface of $UF_6$ cylinder for its reuse, two demonstration tests of the plant with different ratio of ${Na_2}{CO_3}$ and ${H_2}{O_2}$ were carried out, and each test had 5 steps. The main chemical form removed by the tests was to be identified as ${Na_4}{UO_2}(CO_3)_3$. More than 50% of uranium was removed by water of the first step, and at the following steps the removal amounts were exponentially decreased. On the other hand, the result shows that the injected amount of ${Na_2}{CO_3}$, compared with that of the removed uranium, was stoichiometrically excessed. This suggests that the injected amounts of ${Na_2}{CO_3}$, the generation rate of decontaminated waste, and the decontamination steps could be reduced by a process optimization of the plant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.2
• /
• pp.161-170
• /
• 2006

PFC(Perfluorocarbon) decontamination process is one of best methods to remove hot particulate adhered on the inner surface of hot cell and surface of equipment in hot cell. It was necessary to develop a filtration equipment to reuse the PFC waste solution generated on PFC decontamination due to the high cost of PFC solution and for minimization of the volume of second waste solution. The filtration equipment was developed to remove hot particulate in PFC waste solution. It was made suitable size and weight in consideration of hot cell gate and crane. And it has wheels for easy movement. Flux of the filtration equipment decreased with particulate concentration increase. It consists of pre-filter($1.4{\mu}m$) and final-filter($0.2{\mu}m$) for protection of the flux decrease along filtration time. It treatment capacity of waste solution is 0.2 L/min.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.32-39
• /
• 2005

When PFC(Perfluorocarbonate) decontamination technology is applied to removal of radioactive contaminated particulate adhered at surface during the operation of nuclear research facilities, it is necessary to develop a filtration equipment to reuse of PFC solution due to high price, also to minimize the volume of second wastewater. Contaminated characteristics of hot particulate was investigated and a filtration process was presented to remove suspended radioactive particulate from PFC decontamination wastewater generated on PFC decontamination. The range of size of hot particulate adhered at the surface of research facilities measured by SEM was $0.1{\sim}10{\mu}m$. Hot particulate of more than $2{\mu}m$ in PFC contamination wastewater was removed by first filter and then hot particulate of more than $0.2{\mu}m$ was removed by second filter. Results of filter experiments showed that filtration efficiency of PVDF(Poly vinylidene fluoride), PP(Polypropylene), Ceramic filter was $95{\sim}97\%$. A ceramic filter showed a higher filtration efficiency with a little low permeate volume. Also, a ceramic of inorganic compound could be broken easily on experiment and has a high price but was highly stable at radioactivity in comparison of PVDF and PP of a macromolecule which generate $H_2$ gas in alpha radioactivity environment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.3
• /
• pp.153-160
• /
• 2009

Vertical electrokintic-flushing remediation equipment was developed for the remediation of a radioactive soil near nuclear facilities. An optimum reagent was selected to decontaminate the radioactive soil near nuclear facilities with the developed vertical electrokintic-flushing remediation equipment, and the optimum remediation conditions were established to obtain a higher remediation efficiency. Namely, acetic acid was selected as an optimum reagent due to its higher remediation efficiency. When the electrokinetic remediation and the electrokinetic-flushing remediation results were compared, the removal efficiency of 4.6% and the soil waste solution volume of 1.5 times were increased in the electrokinetic remediation. When the potential gradient within an electrokinetic soil cell was increased by two times (4.0 V/cm), the removal efficiencies of $Co^{2+}$ and $Cs^+$ were increased by about 4.3%($Co^{2+}$ : 98.9%, $Cs^+$ : 96.7%). Also, when the reagent concentration was increased from 0.01M to 0.05M, the removal efficiency of $Co^{2+}$ was increased but that of $Cs^+$ was decreased. Therefore, the optimum remediation conditions were that the acetic concentration was $0.01M{\sim}0.05M$, the potential gredient was 4 V/cm, the injection of reagent 2.4ml/g, and the remediation period was 20days.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.6 no.1
• /
• pp.1-9
• /
• 2008

This study examined the effect of an electrokinetic-flushing remediation for a soil of a high permeability. The soil was sampled from the site around a research atomic reactor which had high hydro-conductivities due to a high content of sand in the soil. The flow rate of the washing reagent was fast at the beginning but it was reduced as time lapsed. In the case of using citric acid as a washing reagent, the flow rate was fastest, 78.7 ml/day. The removal efficiencies of $Co^{2+}$ and $Cs^+$ from a soil cell with acetic acid were the highest, which were 95.2% and 84.2% respectively. The soil waste-solution volume generated from the electrokinetic remediation was reduced to about 1/20 of that from the soil washing remediation. Meanwhile, the electrokinetic-flushing method enhanced the removal efficiencies of $Co^{2+}$ and $Cs^+$ from the soil by about 6% and 2% respectively, compared to those by the electrokinetic method. Consequently, it was found that the electrokinetic-flushing method was more effective for the remediation of a soil with a high permeability.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.83-89
• /
• 2003

A temperature- and pressure-reducing process is utilized to handle the spent fuel assembly in the post-irradiation examination facility. This process includes three separated unit processes. First one is the decontamination process to clean the spent fuel assembly casks. The second process is the temperature-reducing process to reduce the temperature elevated by decay process in the spent fuel assembly. The third process is the filtration process to remove insoluble particles existed in the casks using filters. Up-to-date technologies as well as practical theories related to the temperature- and pressure-reducing process is reviewed in this report. The test-operation process for various tests and the test results of the temperature- and pressure-reducing process for J-44 and K-23 spent fuel assemblies are also described in detail. This report must be effectively used for the normal operation of the facility with the awareness of unprecedented problems which could occur by continuing operation of the PIE facility.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.2
• /
• pp.203-212
• /
• 2019

Disposal nonconformity of radioactive wastes refers to radioactive wastes that need to be treated, solidified and packaged during operation or decommissioning of NPPs, and are typically exemplified by particulate radioactive wastes with dispersion characteristics. These wastes include the dried powders of concentrated wastes generated in the process of operating NPPs, slurry and sludge, various powdered wastes generated in the decommissioning process (crushed concrete, decontamination sludge, etc.), and fine radioactive soil, which is not easy to decontaminate. As these particulate wastes must be packaged so that they become non-dispersive, they are solidified with solidification agents such as cement and polymer. If they are treated using existing solidification methods, however, the volume of the final wastes will increase. This drawback may increase the disposal cost and reduce the acceptability of disposal sites. Accordingly, to solve these problems, this study investigates the pelletization of particulate radioactive wastes in order to reduce final waste volume.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.1 no.1
• /
• pp.11-23
• /
• 2003

A study on the electrolytic dissolution of SUS-304 and Inconel-600 specimen was carried out in neutral salt electrolyte to evaluate the applicability of electrochemical decontamination process for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant in Korea. Although the best electrolytic dissolution performance for the specimens was observed in a Na2s04 electrolyte, a NaNO$_3$ neutral salt electrolyte, in which about 30% for SUS-304 and the same for Inconel-600 in the weight loss was shown in comparison with that in a Na$_2$SO$_4$ solution, was selected as an electrolyte for the electrochemical decontamination of metallic wastes with the consideration on the surface of system components contacted with nitric acid and the compatibility with lagoon wastes generated during the facility operation. The effects of current density, electrolytic dissolution time, and concentration of NaNO$_3$ on the electrolytic dissolution of the specimens were investigated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as UO$_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO$_3$ solution with the current density or In mA/$\textrm{cm}^2$. it was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO$_3$ neutral salt electrolyte by reducing $\alpha$ and $\beta$ radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.

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

• Applied Chemistry for Engineering
• /
• v.3 no.2
• /
• pp.327-334
• /
• 1992

As a promising method for the volume reduction of the low-level liquid waste, the pervaporation process was studied using a cellulose acetate membrane. Experimental results showed that the pervaporation method, usually applied to separation of organic materials, has a good decontamination effect for the volume reduction of liquid waste and the evaporation rate of water in this process was markedly faster than that of natural evaporation method, a wide-used process for the volume reduction of liquid waste. Depending on the feed solution conditions, the pervaporation characteristics were evaluated by the experimental results and the optimum conditions for preparation of the cellulose acetate membrane were established to increase the pervaporation flux through the membrane.