• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.293-305
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• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.9-27
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• 2021

As an example of research activities in decontamination for decommissioning, new data are presented on the options for corrosion layer dissolution during the decommissioning decontamination, or persulfate regeneration for decontamination solutions re-use. For the management of spent decontamination solutions, new method based on solvent extraction of radionuclides into ionic liquid followed by electrodeposition of the radionuclides has been developed. Fields of applications of composite inorganic-organic absorbers or solid extractants with polyacrylonitrile (PAN) binding matrix for the treatment of liquid radioactive waste are reviewed; a method for americium separation from the boric acid containing NPP evaporator concentrates based on the TODGA-PAN material is discussed in more detail. Performance of a model of radionuclide transport, developed and implemented within the GoldSim programming environment, for the safety studies of the LLW/ILW repository is demonstrated on the specific case of the Richard repository (Czech Republic). Continuation and even broadening of these activities are expected in connection with the approaching end of the lifespan of the first blocks of the Czech NPPs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.55-64
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• 2003

In this study, the maintenance process by the servo manipulator has been developed for the parts of the equipment, which we unable to reach out by the Master-Slave Manipulator(MSM) in a hot cell. To do this, a virtual mock-up is implemented using the iか prototyping technology. Using this mock-up, the workspace of the manipulators in the hot cell and the operator's view through the wall-mounted lead glass have been analyzed. In addition, the path planning of the servo manipulator using the collision detection function of the virtual mock-up has been established. From these, the maintenance process for the parts of the equipment, which are located at the outside of the MSM's workspace using the servo manipulator has been proposed and verified through the graphic simulation. It is revealed that the proposed remote maintenance process of the equipment can effectively be used in the real hot cell operation. It is also believed that the implemented virtual mock-up of the hot cell can effectively be applied in analyzing the various hot cell operation and enhancing the reliability and safety in a hot cell remote handling for the spent fuel management.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.491-505
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• 2018

As an alternative to deep geological disposal technology, which is considered as a reference concept, the domestic applicability of deep borehole disposal technology for high level radioactive waste, including spent fuel, has been preliminarily evaluated. Usually, the environment of deep borehole disposal, at a depth of 3 to 5 km, has more stable geological and geo-hydrological conditions. For this purpose, the characteristics of rock distribution in the domestic area were analyzed and drilling and investigation technologies for deep boreholes with large diameter were evaluated. Based on the results of these analyses, design criteria and requirements for the deep borehole disposal system were reviewed, and preliminary reference concept for a deep borehole disposal system, including disposal container and sealing system meeting the criteria and requirements, was developed. Subsequently, various performance assessments, including thermal stability analysis of the system and simulation of the disposal process, were performed in a 3D graphic disposal environment. With these analysis results, the preliminary evaluation of the domestic applicability of the deep borehole disposal system was performed from various points of view. In summary, due to disposal depth and simplicity, the deep borehole disposal system should bring many safety and economic benefits. However, to reduce uncertainty and to obtain the assent of the regulatory authority, an in-situ demonstration of this technology should be carried out. The current results can be used as input to establish a national high-level radioactive waste management policy. In addition, they may be provided as basic information necessary for stakeholders interested in deep borehole disposal technology.

• 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 Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.397-410
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• 2018

When considering the long-term safety assessment of spent-nuclear fuel management, americium is one of the most radio-toxic actinides. Although spectroscopic methods are widely used for the study of actinide chemistry, application of those methods to americium chemistry has been limited. Herein, we purified $^{241}Am$ to obtain a highly pure stock solution required for spectroscopic studies. Quantitative and qualitative analyses of purified $^{241}Am$ were carried out using liquid scintillation counting, and gamma and alpha radiation spectrometry. Highly sensitive absorption spectrometry coupled with a liquid waveguide capillary cell and time-resolved laser fluorescence spectroscopy were employed for the study of Am(III) hydrolysis and oxalate (Ox) complexation. $Am^{3+}$ ions under acidic conditions exhibit maximum absorbance at 503 nm, with a molar absorption coefficient of $424{\pm}8cm^{-1}{\cdot}M^{-1}$. $Am(OH)_3(s)$ colloidal particles formed under near neutral pH conditions were identified by monitoring the absorbance at around 506-507 nm. The formation of ${Am(Ox)_3}^{3-}$ was detected by red-shifts of the absorption and luminescence spectra of 4 and 5 nm, respectively. In addition, considerable enhancements of the luminescence intensities were observed. The luminescence lifetime of ${Am(Ox)_3}^{3-}$ increased from 23 to 56 ns, which indicates that approximately six water molecules are replaced by carboxylate ligands in the inner-sphere of the Am(III). These results suggest that ${Am(Ox)_3}^{3-}$ is formed through the bidentate coordination of the oxalate ligands.

• Nuclear industry
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• v.4 no.6 s.22
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• pp.19-23
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• 1984

현재 우리나라에는 3기의 원전이 운전중에 있으며 6기가 건설중에 있다. 즉, 2기의 가압경수로(PWR)의 1기의 가압중수로(HWR)가 운전중에 있으며 건설중인 것은 모두 PWR이다. 현재 추진되고 있는 원전계획을 수행하면, 1990년에는 사용후핵연료가 년간 약200톤씩 방출될 것이며, 2000년에는 500톤 정도가 방출되게 된다. 이와 같은 핵사용후연료는 현재 소내저장하고 있으나 '90년대 중반 부터는 소내저장용량이 한계에 달하게 될 것으로 전망되고 있다. 따라서 본교에서의 이와 같은 소내저장 한계성에 대처할 수 있는 가능한 방안을 검토하고, 이를 경제성 측면에서 분석하고자 하였다. 사용후핵연료의 관리방안에 대한 경제성 분석을 위해서는 장래의 원전계획, 원자로형 및 핵연료주기방식 등에 대한 여러 가지 가정이 필요하게 된다. 원전계획은 정부에서 발표한 $\ulcorner$5차5개년 수정계획$\lrcorner$에 의거하여 원전시설용량은 현재의 2GWe에서 2000년에는 22GWe로, 2025년에는 44GWe로 늘어나는 것으로 보았다. 이와 같은 원전계획을 바탕으로 6가지 핵연료주기에 관한 시나리오를 설정하였다. 즉, 사용후핵연료를 비순환방식으로 운영하는 2가지 경우, 순환방식으로 운영되는 3가지 경우 그리고 FBR에 활용하는 1가지의 경우에 대하여 검토하였다. 사용후핵연료의 관리방식에 따른 장기적인 안목에서의 경제성 분석은 핵연료주기비용 뿐만아니라 원전의 투자비도 함께 분석하는 것이 합리적이며, 따라서 본교에서는 계획기간 동안의 6가지 시나리로에 따른 원전 및 핵연료주기에 관한 총 투자비를 비교하였고, 1982년 가격으로 현가화한 단가도 비교${\cdot}$검토 하였다. 이와 같은 6가지 시나리오에 대한 경제성을 비교해 본 결과, 핵연료주기선택의 경제성평가에 큰 영향을 주는 핵연료주기요소는 재처리비, 재처리시 방출되는 폐기물의 처리${\cdot}$처분비 그리고 사용후핵연료 저장방식으로 판명되었으며 6가지 시나리오에 대한 경제성 비교평가 결과, 다음과 같은 결론을 얻었다. 단기적인 안목에서는 소내저장용량을 확장하는 방안이 가장 바람직하며, 중기적인 관점에서는 소외집중저장설비가 활발히 수행되는 시점에서는 사용후핵연료를 재처리하여 재활용하는 방안도 강구되어야 할 것이다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.455-472
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• 2018

It is important to study how to manage dry storage casks of spent nuclear fuels (SNF), because wet storage spaces for SNF will shortly be at full capacity in the Republic of Korea. The US has operated a dry storage cask system for several decades, and has carried out significant studies into how to successfully manage dry storage cask for SNF. This type of expertise and experience is currently lacking in the Republic of Korea. The degradation of dry casks is an important issue that must be considered. In particular, chloride-induced stress corrosion cracking (CISCC) is known to lead to the release of radioisotopes from canisters. The U.S. Department of Energy, U.S. Nuclear Regulatory Commission, and the Electric Power Research Institute have undertaken research into the CISCC mechanism. In addition, Sandia National Laboratories (SNL) has extensively researched CISCC and how to manage it in dry storage canisters. In this review paper, the probabilistic model proposed by the SNL is analyzed and, based on this model, US-based CISCC research is reviewed in detail. This paper will inform the management of dry cask storage of SNF from light water reactors in austenite stainless steel canisters in the Republic of Korea.

• Korean Chemical Engineering Research
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• v.51 no.6
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• pp.700-708
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• 2013

Pyroprocessing has been developed for the purpose of resolving the current spent nuclear fuel management issue and enhancing the recycle of valuable resources. Pyroprocessing has been developed with the dry technologies which are performed under high temperature conditions excluding any aqueous processes. Pyro-processes which are based on the electrochemical principles require pretreatment processes and a voloxidation process is considered as a pretreatment step for an electrolytic reduction process. Various kinds of gas conditions are applicable to the voloxidation process and the understanding of Cs behavior during the process is of importance for the analyses of waste characteristics and heat load on the overall pyroprocessing. In this study, the changes of chemical compounds with the gas conditions were calculated by analyzing gas-solid reaction behavior based on the chemical equilibria on a Cs-Te-O system. $Cs_2TeO_3$ and $Cs_2TeO_4$ were selected after a Tpp diagram analysis and it was confirmed that they are relatively stable under oxidizing atmospheres while it was shown that Cs and Te would be removed by volatilization under reducing atmosphere at a high temperature. This work provided basic data for predicting Cs behavior during the voloxidation process at which compounds are chemically distributed as the first stage in the pyroprocessing and it is expected that the results would be used for setting up material balances and related purposes.