• Journal of Internet of Things and Convergence
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• v.6 no.1
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• pp.1-8
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• 2020

Following the advanced fuel cycle initiative (AFCI) promotions in the United States, the analytic proposition for global fuel cycle initiative (GFCI) has been investigated using dynamical simulations. The political and economic aspects are considered simultaneously due to the particular characteristics of the nuclear materials. The spent nuclear fuels (SNFs) are treated as the reprocessing by the nuclear non-proliferation treaty (NPT) exemption nations and the NPT excluded nations. Otherwise, the pyroprocessing and repository can be done without NPT restriction. In addition, the international trade is considered as the economic aspect where the energy production is a key issue of the GFCI. The dynamical simulations have been done until 2050. The result of the International Trade shows the gradually increasing shape. Additionally, the Nuclear Power Plant Operation shows the increasing by stepwise shape.

• Nuclear industry
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• v.19 no.6 s.196
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• pp.56-60
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• 1999

원자력 정책에서 안전성과 운영 실적 환경$\cdot$보전$\cdot$경제성 등은 매우 중요한 인자이다. 핵주기의 선택은 에너지 정책, 연료의 다양성, 공급의 안정과 관련된 모든 사회적$\cdot$환경적 영향에 있어 매우 중요하다. 특히 원전의 고준위 방사성 폐기물인 사용후 핵연료 관리는 높은 방사선 준위뿐만 아니라 장기적인 관리 기간이 소요되는 어려운 사업이다. 본 연구는 사용후 핵연료 관리 방안인 재처리와 직접 처분의 비용 분석, 안전성, 대국민용인 측면을 살펴보았다. 직접 처분이 재처리에 비해 약 $7{\%}$ 정도의 경제성이 있고, 직접 처분의 사용후 핵연료는 재처리 폐기물보다 높은 위험도를 갖는다. 대국민 용인 측면서는 두가지 처리 방법 모두 받아들여지지 않는다. 결론적으로, 사용후 핵연료 관리는 모든 사회 $\cdot$환경적 영향과 경제성을 고려한 핵주기 정책과 병행하여 지속적인 기술 개발을 통한 안전성 확보가 필요하다.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3580-3585
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• 2022

The thermal stability of carbohydrazide, hydrazine nitrate, acetohydroxamic acid in nitric acid solutions has been studied at atmospheric pressure and above atmospheric pressure. The volumes of gaseous products of thermolysis and the maximum rate of gas evolution have been determined at atmospheric pressure. It has been shown that, despite the high rate of gas evolution and large volumes of evolved gases, the conditions for the development of autocatalytic oxidation are not created. Exothermic processes are observed in a closed vessel in the temperature range of 50-250 ℃. With an increase in the concentration of nitric acid, the temperatures of the onset of exothermic effects for all mixtures decrease, and the values of the total thermal effects of reactions increase, to the greatest extent for solutions with carbohydrazide.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.653-660
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• 2022

The interaction between UN and CdCl₂ in the LiCl-KCl molten eutectic was studied at 1023 K. The chlorination was monitored by sampling and recording the redox potential of the medium. At 1023 K the chlorination of UN with cadmium chloride in the molten LiCl-KCl eutectic proceeds completely and results in the formation of uranium chlorides. The melts of the LiCl-KCl-UCl₃ or LiCl-KCl-UCl₄ compositions can be obtained by the end of experiment depending on the presence of metallic cadmium in the reaction zone. The higher the concentration of the chlorinating agent, the faster the reaction rate. At [CdCl₂]/[UN] = 1.65 (10% excess) the reaction proceeds to completion in about 7.5 h. At [CdCl₂]/[UN] = 7 the complete chlorination takes 2.5-3 h.

• Nuclear Engineering and Technology
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• v.8 no.2
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• pp.69-76
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• 1976

This experiment has been carried out for the removal of long-lived radioactive-nuclides (Sr-90, Ru-106, Cs-137 and Ce-144) contained in the low-level radioactive effluents from the spent fuel reprocessing plant and nuclear power plant, in order to determine the decontaminability of various chemical coagulants and domestic clay mineral (montmorillonite). Phosphate process showed prominent efficiency for the removal of Ce-144, and lime-soda process did good removal efficiency for Sr-90. About Cs-137 copper-ferrocyanide process is much desirable. In phosphate or lime-soda process, most favorable removal efficiency was obtained at more than pH 11. The montmorillonite treated with sodium chloride showed a considerable improvement in the removal of the radioactive-nuclides. By a combined chemicals-montmorillionite process, the radioactive-nuclides could be more effectively removed than by the only chemicals process.

• Journal of the military operations research society of Korea
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• v.16 no.2
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• pp.14-28
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• 1990

The objective of this study is to propose how to acquire through international cooperation the sensitive nuclear technology, so called reprocessing technology. In spite of the need to reuse spent fuel, the transfer of the sensitive technology has been tightly controlled by the nuclear advanced countries due to the fear of nuclear proliferation and, in fact, it would be impossible to secure it by the economic means. In this regard, as a means of acquiring the sensitive nuclear technology, this study proposes the following; 1) President's declaration concerning the peaceful uses of nuclear energy, 2) the establishment and maintenance of national basis through inter-ministerial cooperation, 3) as a confidence building measure, the efforts to strengthen our role in the international nuclear community, and 4) the establishment of the synthetic feedback system to efficiently coordinate. In line with those stated above, this study suggests that it be necessary to invest consistently for developing new technologies and cultivating human resources. Furthermore, this study proposes the necessity to resolve the problems lying ahead by the national consensus achieved through the discussions among the public concerning the sensitive nuclear technology.

• Journal of the Korean Society of Radiology
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• v.15 no.1
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• pp.37-43
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• 2021

Republic of Korea has many nuclear facilities in the country, and Democratic People's Republic of Korea(North Korea) locates in the surrounding country. Therefore, it is necessary to construct the target facility's nuclear forensic data in a preemptive response to the changing international situation. For this reason, this study suggests "signature" materials used to understand the origins and sources of nuclear and other radioactive materials, taking into account domestic nuclear facilities and the nuclear fuel cycle. In domestic, pressurized light water reactors and pressurized heavy water reactors are in operation, and enriched and natural uranium are used as fuels. In the front-end fuel cycle, the signature materials can be nature uranium and UF6 in the uranium enrichment process. The domestic back-end fuel cycle adopts a non-circulating cycle excluding the reprocessing process, and the primary signature material is spent nuclear fuel. According to IAEA recommendation, the importance of these materials as the signature and characteristic contents are suggested in this study. To prove the integrity of nuclear material and build a national nuclear forensics library, it is necessary to grasp the signature material and acquire the characteristic data considering the domestic nuclear facilities and the nuclear fuel cycle.

• 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.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.188-198
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• 2021

A modern approach to nuclear energy involves reprocessing like transmutations of spent nuclear fuel products to reduce their radiotoxicity and time needed for their storage. For this purpose, they are immobilized in inert matrices made of zirconia and can be "burned" in fast neutron reactor or Accelerator Driven System. These matrices in spherical form can be obtained by sol-gel process. The paper presents a method of microspheres fabrication based on the combined Complex Sol-Gel Process and double extraction process consisting in the preparation of zirconium-ascorbate sol and simultaneous extraction of water and nitrates. The procedure allows obtaining gel microspheres with a diameter of 50 ㎛, which after heat treatment are processed into the final product. The synthesis of zirconia microspheres with Yttrium by internal gelation process is well known for over a decade now. However, the explanation and characterization of synthesis of such material by extraction of water process is rarely found. Parameters such as: pH, viscosity, shape, sphericity and crystal structure have been determined for synthesized products and semi-products. In addition, preliminary research consisting in irradiation of the obtained materials in fast and thermal neutron flux was carried out. The obtained results are presented and described in this work.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1644-1650
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• 2023

In spent fuel reprocessing, UO₂(DBP)₂ (U-DBP) can be deposited in stainless steel equipment. U-DBP must be removed by dissolution and the process must not cause corrosion to stainless steel. This study was conducted to find the best scheme for dissolution. U-DBP was manufactured by the titrimetric sedimentation method. The effects of different factors on the dissolution of U-DBP were investigated. For example, solid-liquid ratio, hydrazine carbonate solutions with different mass components, mixed solutions containing different concentrations of H₂O₂, and different carbonates. The results indicated that U-DBP does not have a regular crystal morphology. With the increase of the solid-liquid ratio and the mass fraction of hydrazine carbonate, the concentration of U(VI) at the dissolution equilibrium increases gradually. The addition of H₂O₂ has a great promotion effect on the dissolution. However, when the concentration of H₂O₂ is greater than 0.5 M, the dissolution solution may have an erosive effect on the stainless steel. (NH₄)₂CO₃ can increase the dissolution capacity of dissolved U-DBP, but it may also accelerate the corrosion of stainless steel.