• Proceedings of the IEEK Conference
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• 2001.10a
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• pp.179-182
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• 2001

The chloride volatilization method for the recovery of zirconium and removal of uranium from zirconium containing metallic wastes formed in spent fuel reprocessing was studied using the simulated alloy waste, i.e. the mixture of Zr foil and UO$_2$/U$_3$O$_{8}$ powder. When the simulated waste was heated to react with chlorine gas at 350- l00$0^{\circ}C$, the zirconium metal changed to volatile ZrCl$_4$showing high volatility ratio (Vzr) of 99%. The amount of volatilized uranium increases at higher temperatures causing lowering of decontamination factor (DF) of uranium. This is thought to be caused by the chlorination of UO$_2$ with ZrCl$_4$vapor. The highest DF value of 12.5 was obtained when the reaction temperature was 35$0^{\circ}C$. Addition of 10 vol.% oxygen gas into chlorine gas was effective for suppressing the volatilization of uranium, while the volatilization ratio of zirconium was decreased to 68% with the addition of 20 vol.% oxygen. In the case of the mixture of Zr foil and U$_3$O$_{8}$, the V value of uranium showed minimum (44%) at 40$0^{\circ}C$ with chlorine gas giving the highest DF value 24.3. When the 10 vol.% oxygen was added to chlorine gas, the V value of zirconium decreased to 82% at $600^{\circ}C$, but almost all the uranium volatilized (Vu=99%), which may be caused by the formation of volatile uranium chlorides under oxidative atmosphere.ere.

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

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.671-675
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• 2003

In accordance with the KRR-1 & 2 decommissioning project, the decontamination and dismantling activities of the KRR-2 auxiliary facilities, radioisotope production facilities, were completed from Aug 2001 to Dec 2002. The auxiliary facilities were composed of the concrete hot-cell, lead hot-cells and several laboratories for the radioisotope production. The dismantling objects are home hoods, experimental desks, sinks, and contaminated inner facilities. For the purpose of the safe decommissioning activity, the method statements and working procedures were set up. The manpower of the total 20,933 man-hour was required and several dismantling equipments were also. The maximum surface contamination is: 9.24 Bq/$\textrm{cm}^2$ in removable contamination and 350,000 cpm in fixed contamination. The total amount of 62.146 Ton was raised as dismantled waste with kinds of the concretes, wood, steels, etc. The collective dose was evaluated as 0.33 mam-mSv during this period.

• Nuclear Engineering and Technology
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• v.9 no.2
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• pp.75-81
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• 1977

This experiment has been carried out to determine the pH dependent cation exchange capacity concerning the sorption phenomenon of long-lived radionuclides contained in low-level liquid radioactive waste on various clay minerals. The pH dependent cation exchange capacity determined by Sawhney's method are used to the analysis of sorption phenomenon. About 70 percent of the total cation exchange capacity is contributed by the pH dependent CEC due to the negative charge originated naturally in clays in case of clinoptilolite, vermiculite and sodalite. It is sugested in this test that the high neutral salt CEC, that is, highly charged clays would show good fixation yield. The removal of radionuclides at the pH range more than pH 9 is considered the hydroxide precipitation of metal ion rather than the cation exchange. The Na-clay prepared by the method of successive isomorphic substitution with electrolyte showed a considerable improvement in removal efficiency for the decontamination.

• Journal of Soil and Groundwater Environment
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• v.8 no.1
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• pp.17-26
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• 2003

The sorption experiment of cobalt was performed after the TRIGA soil was intentionally contaminated with cobalt was found that the sorption equilibrium coeficiency of soil decontamination was high when the ratio of soil mass to the volume of citric acid becomes 1:5 The TRIGA soil contaminated with 0.01 M, 0.001 M, and 0.0001 M of cobalt solution were decontaiminated with 0.01 M citric acid. The cobalt concentrtion in the wastewater were measured to be correspondingly 36.0, 14.0, 1.5 ppm. The results of wastewater recycling experiment by chemical precipitation method revealed that corresponding cobalt removal efficiency were 97% 88%. It was shown that the removal efficiency decreases as the cobalt concentration in the wastewater decreases. During the decontamination experiment, a lot of NaOH had to be added, and the volume of final solid waste reached almost 10% of that of the contaminated soil. The result of wastewater recyling experiment by ion exchange resin meted rethod revealed that to more the strong acid resins are used, the higher the cobalt removal efficiency becomes and the cobalt removal efficiency becomes and the lower the pH of recycling wastewater become. In order to obtain more than 95% removal efficiency, more than 0.625 g of strong acid resin was necessary in each of 3 experiments. There was an unexpected problem that a lot of strong acid resin waste was produced which amounts to 9.2% (volume) of the contaminated soil.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.419-427
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• 2019

The application of ion crystallization technology was considered as a way to increase the operating efficiency and improve the operating performance of a liquid waste management system (LWMS) in the Advanced Power Reactor 1400 (APR 1400). Although ion crystallization technology has not been practically applied to Nuclear Power Plants (NPPs) until now, a previous experimental study demonstrated that it is possible to selectively remove at least 95% of various nuclide ions present in the liquid radioactive waste of NPPs. We reviewed the possibility of applying ion crystallization technology to the existing LWMS by applying the nuclide removal rate of ion crystallization technology and prepared a way to improve the existing LWMS in the APR 1400. Furthermore, we determined the optimized application location of ion crystallization technology in the existing LWMS by considering decontamination characteristics of the ion crystallization technology and the existing LWMS design features and operating experiences. The application of ion crystallization technology to the liquid waste collection tank, where liquid radioactive materials are collected, will have the least impact on the existing design while providing the greatest improvement. It is expected that the application of ion crystallization technology to the current APR 1400 or new NPPs would increase the operating efficiency of the LWMS and result in an improvement of system performance.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.407-417
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• 2004

This study is to verify radiation detection method by using $\alpha$-spectroscopy and ${\gamma}$-spectroscopy for concretes and components which will be generated during the decommissioning of the uranium conversion plant. Components and inside walls of the building were contaminated with natural uranium materials. Some parts of the stainless steel pipes and concretes of the walls were sampled and analyzed their alpha and gamma activities respectively. Alpha and gamma activities are well matched each other in the range of high activity region to 0.01 Bq/g and gamma activities are over estimated comparing alpha activities corresponded in below 0.005 Bq/g region for the natural uranium of AUC sample. The $^{238}U$ originated from natural products of conversion process could be distinguished by measuring $^{214}Pb$ or $^{214}Bi$ and $^{234}Th$ or $^{234m}Pa$. Uranium contaminations mainly are in the wall surface of the plant. Decontamination process of generating wastes which can be reached tp background level gamma activities measured by gamma spectroscopy can also be used to conservative assessment data.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.676-680
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• 2003

Dissolution of magnetite powders by 0.05 M citric acid was investigated at $50^{\circ}C$. All the tests were performed in the pH range between 2.0 to 5.0, which was adjusted using nitric acid or sodium hydroxide. Concentration of each of the dissociated chemical species of citric acid under various solution pHs was calculated using the ionization constants. Variation of zeta potential of magnetite with pH changes was also investigated. The dissolution reaction was explained by comparing the concentration of the dissociated chemical species of citric acid with the zeta potential. Longer than 3 h of induction time was required to dissolve the magnetite. The dissolution behaviour of magnetite was well described by the equation. The physical meaning of each parameter was explained successfully from the model equation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.153-164
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• 2014

Currently, there are many programs which are now being developed or already developed to predict radionuclide and corrosion product at the stage of designing NPP. However, since there are not many developments in evaluating quantity of activation corrosion products occurring when disassembling a nuclear power plant there exist some difficulties in calculating accurately. In order to evaluate activation products inventory for the research of effect of neutron activation in the reactor vessel, component of nuclear reactor and adjacent structures, it should be evaluated by using operation history of nuclear reactor, material composition of structure and average neutron flux in every field representing fixed structure of nuclear reactor. In this study, CORA, PACTOLE, CRUDSIM, CREAT and ACE codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor which is used at the stage of designing. As a future study, the accuracy in calculating the quantity of product corrosion can be increase by finding out the possibility of use and improvement for evaluation of the decontamination.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.171-178
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• 2014

After the expiration of operating license of a plant, all infrastructures within the plant must be safely dismantled to the point that it no longer requires measures for radiation protection. Despite the fact that Kori 1 and Wolsong 1 are close to the expiration of their operating license, sufficient technologies for radiological characterization, decontamination and dismantling is still under development. The purpose of this study is to develop one of methods for radiological inventory assessment on measuring object by using direct measure of large component with In-Situ measurement technique. Radiological inventory was assessed by analyzing nuclide using portable gamma spectroscopy without dismantling reactor head, and the result of direct measurement was supplemented by performing indirect measurement. Radiochemical analysis were performed on surface contamination samples as well. During the study, radiological inventory of reactor vessel calculated expanding the result. Based on the result and the radioactivity variation of each radionuclides time frame for decommissioning can be decided. Thus, it is expected that during the decommissioning of plants, the result of this study will contribute to the reduction of radiation exposure to workers.