The object of this container is to store uranium precipitated in Mo-99 production process for long term. It will be used in the Gijang reactor. Uranium precipitates is stored in a container such as a form of filter cake. This container will be stored in the underground concrete storage for about 50 years when considering a storage facility life and safety. It is required high degree of structure integrity when considering decay heat from radioactive precipitated, external impact and etc, because it has to store radioactive materials for long term. In this study, thermal, structural and impact analysis about a container inserted 6 canisters was carried out. In the case of this thermal analysis, the object is to investigate the effect decay heat of uranium precipitate on the structure integrity of a container. The structural analysis was carried out to evaluate a structure integrity of the bottommost portion container among the 9 loaded containers. Finally, the impact analysis was carried out to evaluate the structure integrity of a container when that is dropped from maximum height of 5m during transport. As a result of this study, the structural integrity of this container is satisfactory about thermal, structural and impact.

최근 원전해체, 원전사고 등으로 인한 수계 내 방사성물질 제거 기술로 고분자 소재의 분리막이 사용되고 있지만 고에너지 광선으로 막의 변형 및 파손이 중요한 문제점 중 하나이다. 무기계열의 세라믹막은 강한 내구성을 지니며, 기 존의 한외여과막 수준의 표면 공극 크기를 나노여과막 수준으로 개질할 경우 방사선 물질 제거에 효과적이다. 본 연구에서는 알루미나-지르코니아 나노물질 을 여과코팅 방법으로 세라믹막 표면을 나노여과막 수준으로 개질을 하였고, SEM-EDX, 분획분자량, 오염 전 투수량을 통해 막의 특성 변화를 확인하였다. 제조된 세라믹 나노여과막의 방사선물질 제거 평가로 우라늄(Uranyl nitrate hexahydrate) 2 mg/L 수용액을 사용하였고, ICP-MS 분석결과 40% 제거율을 확인하였다.

In this study, we investigated the unit process parameters in spherical kernel preparation. Nearly perfect spherical microspheres were obtained from the 0.6M of U-concentration in the broth solution, and the microstructure of the kernel appeared the good results in the calcining, reducing, and sintering processes. For good sphericity, high density, suitable microstructure, and no-crack final microspheres, the temperature control range in calcination process was , and the microstructure, the pore structure, and the density of kernel could be controlled in this temperature range. Also, the concentration changes of the ageing solution in aging step were not effective factor in the gelation of the liquid droplets, but the temperature change of the ageing solution was very sensitive for the final ADU gel particles

The effects of thermal treatment conditions on ADU (ammonium diuranate) prepared by SOL-GEL method, so-called GSP (Gel supported precipitation) process, were investigated for kernel preparation. In this study, ADU compound particles were calcined to particles in air and Ar atmospheres, and these particles were reduced and sintered in 4%-/Ar. During the thermal calcining treatment in air, ADU compound was slightly decomposed, and then converted to phases at . At , the phase appeared together with . After sintering of theses particles, the uranium oxide phases were reduced to a stoichiometric . As a result of the calcining treatment in Ar, more reduced-form of uranium oxide was observed than that treated in air atmosphere by XRD analysis. The final phases of these particles were estimated as a mixture of and .

The nano-scale crystallite sizes of uranium oxide powders in simulated spent fuel were measured by the neutron diffraction line broadening method in order to analyze the sintering behavior of the dry process fuel. The mixed and fission product powders were dry-milled in an attritor for 30, 60, and 120 min. The diffraction patterns of the powders were obtained by using the high resolution powder diffractometer in the HANARO research reactor. Diffraction line broadening due to crystallite size was measured using various techniques such as the Stokes' deconvolution, profile fitting methods using Cauchy function, Gaussian function, and Voigt function, and the Warren-Averbach method. The non-uniform strain, stacking fault and twin probability were measured using the information from the diffraction pattern. The realistic crystallite size could be obtained after separation of the contribution from the non-uniform strain, stacking fault and twin.

A study on the electrosorption of uranium ions onto a porous activated carbon fiber (ACF) was performed to treat uraniumcontaining lagoon sludge. The result of the continuous flow-through cell electrosorption experiments showed that the applied negative potential increased the adsorption kinetics and capacity in comparison to the open-circuit potential (OCP) adsorption for uranium ions. Effective U(VI) removal is accomplished when a negative potential is applied to the activated carbon fiber (ACF) electrode. For a feed concentration of 100 mg/L, the concentration of U(VI) in the cell effluent is reduced to less than 1 mg/L. The selective removal of uranium ions from electrolyte was possible by the electrosorption process.

본 논문에서는 고준위 핵폐기물의 지하 처분 시 사용되는 핵폐기물 처분장치의 기본 구조설계에 필요한 처분장치내의 핵 폐기물다발들을 지지하는 내부 삽입물의 구조형상과 재원 또 처분장치의 화학적 부식을 방지하기 위해 외곽에 설치하는 외곽쉘과 위아래 덮개의 두께를 결정하기 위하여 처분장치 구조물에 대한 선형정적 구조해석을 수행하였다. 해석 대상 처분장치는 가압경수로와 중수로의 핵폐기물 처분장치를 사용하였다. 일반적으로 핵폐기물 처분장치는 지하수백 미터에 위치하는 화강암 등의 암반 내에 설치하게 되는데 이 때 지하수의 침수에 의한 지하수압 및 처분장치 외곽에 완충장치로 설치하는 벤토나이트 버퍼의 팽윤압을 견디어 내야 한다. 따라서 이와 같은 압력의 변화에 따른 처분장치 구조물에 발생하는 응력 및 변형 등을 알기 위해서는 처분장치 구조물에 대한 구조해석을 수행해야 된다. 이를 위하여 본 논문에서는 처분장치에 대하여 선형정적 구조해석을 수행하였다.

The densification and grain growth mechanisms of in and in have been investigated. Uranium dioxide powder compacts were sintered at 1 in or at 110 in for various times from 0.5 h to 16 h. The grain size and density of the specimens were measured. From the measured data, the mechanisms of the densification and grain growth were determined by use of available kinetic equations which express the relations between densification and grain growth. In both atmospheres, it has been found that the densification was controlled by the lattice diffusion and the grain growth by the surface diffusion of atoms around pores. It appears that the surface diffusivity as well as the lattice diffusivity increase considerably with the increase in O/U ratio in the specimen.

사염화우라늄 제조를 위해 염소가스와 탄소를 이용한 이산화우라늄의 염소화반응에 대하여 연구하였다. 이론적측면에서 열화학적 자료를 이용한 계산을 통하여 일어날 수 있는 반응들을 확인하였으며, 염소화반응이 진행되는 동안 초래될 현상에 대하여 검토하였다. 실험결과로 부터 반응온도, 반응시간 및 질소가스 주입비율이 사염화우라늄 제조에 미치는 영향을 정량적으로 평가하였다. 순수한 이산화우라늄을 사용한 사염화우라늄 제조공정에서의 적절한 반응시간과 반응온도는 각각 약 2시간과 500˚C-700˚C범위였으며, 질소가스의 적정 주입량은 염소가스의 약 50%로 나타났다.

사염화우라늄을 제조하기 위한 가장 효율적인 반응계는 이산화우라늄, 염소가스와 탄소분말이다. 여러 가지 실험변수 가운데 이산화우라늄의 염소화반응에 사용된 염소가스 주입량과 탄소의 양이 사염화우라늄 제조에 미치는 영향에 관하여 연구하였다. 각각의 실험변수들에 대한 전화율과 휘발률 계산을 통해 효율적인 반응을 위한 적정 염소가스 주입량과 탄소의 양을 구하였고, 이산화우라늄의 증가함에 따라 직접접촉에 의한 기체-고체반응에서는 전화율과 휘발률은 증가했으나 이후 과량을 첨가함에 따라 감소하였고, 용융염내의 기체-액체반응에서는 전화율의 미미한 증가와 휘발률의 감소를 확인하였가. 염소주입량이 증가함에 따라 전화율과 휘발률이 증가했으며, 과량의 염소가수 주입시 고이온가 염화물의 생성량이 증가하였다.

아미드옥심기와 복합재료 섬유흡착제를 제조하였고 해수로부터 우라늄이온의 분리 특성을 조사하였다. 흡착량은 흡착시간이 증가함에 따라 증가하였고 An:TEGMA:DVB의 몰비가 1:0.1:0.003인 수지가 pH 8 부근에서 최대 흡착능을 나타내었다. 또한 흡착량은 CFA에 첨가한 흡착제의 양이 증가함에 따라 증가하였으며 1시간 까지 선형적으로 증가하였고, 25˚C에서 최대흡착량을 나타내었다. 한편 Ca, Mg 이온은 흡-탈착 cycle이 반복될수록 증가하였으며 그양은 각각 0.3, 0.9mmole/g-Ads로 우라늄 이온의 그것보다 매우 낮았다. 흡착된 우라늄 이온의 탈착은 흡착제의 종류에 관계없이 약 30분 이내에 거의 100% 탈착되었다.

해수로 부터 우라늄 분리를 위한 아미드옥심형 섬유복합재료 흡착제를 제조하였고, IR, 팽윤도 실험, CHN 원소분석, SEM 및 흡착능 실험을 통하여 그 특성을 알아보았다. AN-TEGMA 및 AN-TEGMA-DVB 공중합체의 팽윤율과 수율은 가교제의 함량이 증가할 수록 감소하였으며, 수율은 AN-TEGMA 공중합체의 경우 85-92%였고 AN-TEGMA-DVB 공중합체는 82-88%였다. 다공도도가교체의 함량이 증가할 수록 감소하였으며 AN-TEGMA-DVB 공중합체가 AN-TEGMA 공중합체보다 작았다. 또한 전자현미경 관찰 결과 제조한 섬유 복합재료 흡착제의 표면에 흡착제가 고루 분포되어 있는 것을 확인하였고, 흡착제의 최적 첨가량은 40wt%이었다. 섬유복합재료 흡착제의 우라늄 흡착량은 pH 8 부근에서 최대 였으며, 해수의 pH가 8.3임을 감안할 때 해수 우라늄 분리에 적합한 소재 임을 알 수 있었다.

Two kinds of powders and dispersed nuclear fuel meats have been prepared by conventional comminution process and a newly developed rotating disk atomization process. In contrast to angular shape and broad size distribution of the conventionally processed powder, the atomized powder was spherical and showed narrow size distribution. For the atomized powder, the heat treatment time for the formation of by a peritectoid reaction was reduced to about one tenth, thanks to microstructure refinement by rapid cooling of about 5104 K/s. The extruding pressure of atomized powder and Al powder mixture was lower than that of comminuted and Al powder mixture. The elongation of the atomization processed fuel meats was much higher than that of the comminution processed fuel meats and remained over 10% up to 80wt.% of powder fraction in the fuel meats. It appears therefore that the loading density of in fuel meat can be increased by using atomized powder. The atomized spherical particles were randomly distributed, while the comminuted particles with angular and longish shape were considerably aligned along the extrusion direction. Along the transverse direction of the extraction the electrical conductivity of the atomization processed fuel meats was appreciably higher than that of comminution processed fuel meats. This tendency became pronounced as content increased. Because the thermal conduction which is believed to be proportioned to the electrical conduction in the nuclear fuel meats occurs in radial direction, the atomization processed fuel can be better used in research reactors where high thermal conductivity is required.

The measurements of uranium with nuclear fission track technique on the Holocene carbonate components and submarine cements in South Florida, U.S.A. and the Bahamas have allowed not only characteristic uranium concentrations but also spatial distribution. Relatively high uranium concentrations were found in coral skeletons (2.5 ppm). ooids (2.8 ppm), and peloids (3.2 ppm) whereas most of the modern calcareous organisms contain low uranium concentrations. Varied uranium concentrations were found in submarine cements; more than 3 ppm in acicular aragonite, 2 to 3 ppm in micritic Mg-calcite in inter- and intraparticle pores, and 0.7 to 2.8 ppm in micirtic envelopes. Heterogeneous distributions of uranium were quite common in both skeletons and inorganic carbonates. Marine organisms seem to discriminate against uranium while they are alive and thereby they contain low uranium concentrations whereas inorganic carbonate components incorporate uranium in equilibrium with seawater and thereby the contain high uranium concentration. In incorporation of uramiun into carbonate componets physiology and mineralogy seem to be important in organism whereas minerablogy and CO₂ content of seawater are thought to be important in inorganic components. Characteristic uranium concentrations and spatial distribution pattern in modern carbonates suggest that uranium can be used as a powerful diagenentic indicator in studying ancient carbonate rocks. This study reveals that the fission track technique is an advantageous tool in studying petrography