• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.115-122
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• 2010

In this study, the removal of Cs and Tc from a simulated fission products (FP) solution which were co-dissolved with U during the oxidative-dissolution of spent fuel in a mixed carbonate solution of $(Na_2CO_3-NaHCO_3)-H_2O_2$ was investigated by using a selective precipitation method. As Cs and Tc might cause an unstable behavior due to the high decay heat emission of Cs as well as the fast migration of Tc when disposed of underground, it is one of the important issues to removal them in views of the increase of disposal safety. The precipitation of Cs and Re (as a surrogate for Tc) was examined by introducing sodium tetraphenylborate (NaTPB) and tetraphenylphosponium chloride (TPPCl), respectively. Precipitation of Cs by NaTPB and that of Re by TPPCl were completed within 5 minutes. Their precipitation rates were not influenced so much by the temperature and stirring speed even if they were increased by up to $50^{\circ}C$ and 1,000 rpm. However, the pH of the solution was found to have a great influence on the precipitation with NaTPB and TPPCl. Since Mo tends to co-precipitate with Re at a lower pH, especially, it was effective that a selective precipitation of Re by TPPCl was carried out at pH of above 9 without co-precipitation of Mo and Re. Over 99% of Cs was precipitated when the ratio of [NaTPB]/[Cs]>1 and more than 99% of Re, likewise, was precipitated when the ratio of [TPPCl]/[Re]>1.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.195-202
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• 2018

As a rule, geological disposal is considered a safe method for final disposal of high-level radioactive waste. However, some long-lived fission products like $^{99}Tc$ and $^{129}I$ contained in spent nuclear fuel are highly mobile as less sorbing anionic species in the subsurface environment and can mainly cause exposure dose to the ecosystem by emission of beta rays in the hundreds of keV range. Therefore, if these two nuclides can be separated and converted with high efficiency into radioactively unharmful nuclides, this would have a positive effect on disposal safety. One candidate method is to transmute these two nuclides in nuclear reactors into short-lived nuclides or into stable nuclides. For this purpose, it is necessary to evaluate which reactor type is more efficient in burning these two nuclides. In this study, the simulation results of nuclear transmutation of $^{99}Tc$ and $^{129}I$ in light water reactor (PWR), heavy water reactor (CANDU) and fast neutron reactor (SFR, MET-1000) are compared and discussed.

• Analytical Science and Technology
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• v.21 no.4
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• pp.326-331
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• 2008

Distribution coefficients (Kd) of uranyl ion onto divinylbenzene amidoxime resins were measured in sodium carbonate solution and the Kd values were increased up to about 70 as the resin bead size was decreased. At a condition of 0.0044 M $Na_2CO_3$, the adsorption capacity for uranium was $3.4{\mu}mole$ U/g-resin. The Kd values in the 0.5 M $Na_2CO_3-NaHCO_3$ solution, ranging from pH 9 to pH 11, revealed that they were increased as the pH increased and revealed lower values than those in the pure sodium carbonate solution. The amidoxime resins were characterized by FTIR-ATR showing the absorption bands of the amidoxime functional groups. A species of the uranyltricarbonate complex, $UO_2(CO_3)_3^{-4}$, was confirmed by UV-Vis spectroscopy, revealing four absorption peaks between 400 and 500 nm. Uranium was separated from some fission products by a column operation. However, most of the uranium and fission products were eluted before an adsorption and only a small amount of uranium was adsorbed onto the resin due to the low capacity of the resin.

• KIEAE Journal
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• v.8 no.5
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• pp.55-60
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• 2008

The purpose of this study is to increase the use of Hwangto and examine the strength according to what it is compounded with. Hwangto-concrete containing Hwanto without cement nor organic chemical products were compared to the traditional cement concrete through some durability experiments. We expect to gain more knowledge on the potentials of Hwangto-concrete as an architectural source. 1) As Hwangto binder amount rises, the value of slump increases too. The reason is that the increase of the quantity of cement causes the increase of the amount of material and the decrease of the amount of aggregate. 2) When the mixed component into Hwangto-concrete remains at 2%, the compress strength is generally dispersed high along the per unit fission, in case the amount of which is at $400(g/m^3)$. The highest compress strength is 39MPa. It means that it can be applied to common structures and we need to conduct a basic property test to ensure the strength and fluidness. 3) Hwangto-concrete is expected to be highly used in the ocean structure and chemical industry because it has better resistance to sulfuric acid and to hydrochloric acid than the cement-concrete has. The result of this study is as follows. It is expected that Hwangto-concrete will be widely applied and further research on its durability and tests for its basic substantial characteristics based on future component added to it.

• Journal of Radiation Protection and Research
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• v.10 no.1
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• pp.3-13
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• 1985

The sorption behavior of some typical fission products such as Cs-137, long-lived radionuclide; Ce-144, rare-earth element; and Co-60, corrosion product on zeolite A, zeolite F-9 (faujasite) and amorphous zeolite was determined with the salt concentrations, 0.01 M- to 2.0 M- nitric acid and ammonium nitrate, and the shaking time, 15 minutes interval from 15 minute to 90 minute. Kd values were obtained through the batch experiment. In conclusion, the optimal conditions for isolation and removal of the typical radionuclides are as following: zeolite, amorphous zeolite; concentration, $0.01\;M-HNO_3\;and\;0.1\;M-NH_4NO_3$; pH4; shaking time, one hour; the most effective species, Cs-137.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1998.05a
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• pp.235-242
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• 1998

Aiming at one of decisive alternatives for long term aspect of nuclear power concerns, an integral and closed nuclear system, AMBIDEXTER (Advanced Molten-salt Break-even Inherently-safe Dual-mission Experimental and TEst Reactor) concept is under development. The AMBIDEXTER complex essentially comprises two mutually independent loops of the radiation/material transport and the heat/energy conversion, centered at the integrated reactor assembly, which enables one to utilize maximum benefits of nuclear energy under minimum risks of nuclear radiation. And it provides precious radioisotopes and radiation sources from its waste stream. Also the reactor operates at very low level of fission products inventory throughout its lifetime. The nuclear and thermalhydraulic characteristics of the molten TH/$^{233}$ U fuel salt extend the capability of the self-sustaining AMBIDEXTER fuel cycle to enhance resource security and safeguard transparency. The reactor system is consisted of a single component module of the core, heat exchangers and recirculation pumps with neither pipe connections nor active valves in between, which will significantly improve inherent features of nuclear safety. States of the core technologies associated with designing and developing the AMBIDEXTER concept are mostly available in commercialized form and thus demonstration of integral aspects of the concept should be the prime area in future R＆D programs.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.183-192
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• 2010

An electrochemical reduction of a mixture of metal oxides was conducted in a LiCl molten salt containing 3 wt% $Li_2O$ at $650^{\circ}C$. The oxide reduction was carried out by applying a current to an electrolysis cell, and the $Li_2O$ concentration was analyzed during each run. The concentration of $Li_2O$ in the electrolyte bulk phase gradually decreases according to Faraday's law due to a slow diffusion of the $O^{2-}$ ions. A hindrance effect of the unreduced metal oxides was observed for the reduction of the uranium oxide. Cs, Sr, and Ba of high heat-load fission products were diffused into and accumulated in the salt phase as predicted with thermodynamic consideration.

• Journal of The Korean Society of Clinical Toxicology
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• v.9 no.2
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• pp.39-48
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• 2011

In mass casualty situation due to radiological accidents, it is important to start aggressive management with rapid triage decisions. External contamination needs immediate decontamination and internal contamination should be treated with special expertise and equipment to prevent the rapid uptake of radionuclides by target organs. Acute radiation syndrome shows a sequence of events that varies with the severity of the exposure. More severe exposures generally lead to more rapid onset of symptoms and severe clinical findings. After the massive exposure, various systems of the body reflect their severe damages that can lead to death within hours or up to several months. The disease progression has classically been divided into four stages: prodromal, latent, manifest illness, and recovery or death. Three characteristic clusters of symptoms including the hematopoietic syndrome, the gastrointestinal syndrome and the cerebrovascular syndrome are all associated with the acute radiation syndrome. The standard medical management of the patients with a potentially survivable radiation exposure includes good medical, surgical and supportive measures. Specific treatment with cytokines and bone marrow transplantation should be considered. The management of internal contamination is much the same as the treatment of poisoning. The standard decontamination should be applied to reduce uptake, and the chelating agents can be administered to enhance the clearance of radioisotopes. Radioactive iodine ($^{131}I$) as one of the nuclear fission products can increase the incidence of thyroid cancer in children. Potential benefit of potassium iodide prophylaxis is greater especially in neonates, infants and small children.

• Journal of the Korean Ceramic Society
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• v.44 no.10
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• pp.580-584
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• 2007

The ZrC layer instead of SiC layer is a critical and essential layer in TRISO coated fuel particles since it is a protective layer against diffusion of fission products and provides mechanical strength for the fuel particle. In this study, we carried out computational simulation before actual experiment. With these simulation results, Zirconium carbide (ZrC) films were chemically vapor deposited on $ZrO_2$ substrate using zirconium tetrachloride $(ZrCl_4),\;CH_4$ as a source and $H_2$ dilution gas, respectively. The change of input gas ratio was correlated with growth rate and morphology of deposited ZrC films. The growth rate of ZrC films increased as the input gas ratio decreased. The microstructure of ZrC films was changed with input gas ratio; small granular type grain structure was exhibited at the low input gas ratio. Angular type structure of increased grain size was observed at the high input gas ratio.

• Nuclear Engineering and Technology
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• v.38 no.6
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• pp.459-482
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• 2006

A geochemical approach to the long-term safety of waste disposal is discussed in connection with the significance of actinides, which shall deliver the major radioactivity inventory subsequent to the relatively short-term decay of fission products. Every power reactor generates transuranic (TRU) elements: plutonium and minor actinides (Np, Am, Cm), which consist chiefly of long-lived nuclides emitting alpha radiation. The amount of TRU actinides generated in a fuel life period is found to be relatively small (about 1 wt% or less in spent fuel) but their radioactivity persists many hundred thousands years. Geological confinement of waste containing TRU actinides demands, as a result, fundamental knowledge on the geochemical behavior of actinides in the repository environment for a long period of time. Appraisal of the scientific progress in this subject area is the main objective of the present paper. Following the introductory discussion on natural radioactivities, the nuclear fuel cycle is briefly brought up with reference to actinide generation and waste disposal. As the long-term disposal safety concerns inevitably with actinides, the significance of the aquatic actinide chemistry is summarized in two parts: the fundamental properties relevant to their aquatic behavior and the geochemical reactions in nanoscopic scale. The constrained space of writing allows discussion on some examples only, for which topics of the primary concern are selected, e.g. apparent solubility and colloid generation, colloid-facilitated migration, notable speciation of such processes, etc. Discussion is summed up to end with how to make a geochemical approach available for the long-term disposal safety of nuclear waste or for the performance assessment (PA) as known generally.