• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.1
• /
• pp.83-91
• /
• 2018

While using an electrokinetic method to analyze the characteristics of cement solidification of radioactive wastes from decontaminated uranium soil and concrete, the compressive strength, pH, electrical conductivity, irradiation effects, and volume expansion were measured for the solidified cement specimens. The workability of cement solidified from radioactive waste was about 170-190%. After the solidified cement was irradiated, the compressive strength decreased by about 15%, but met the criteria ($34kgf{\cdot}cm^{-2}$) of KORAD (Korea Radioactive Waste Agent). According to the results of SEM-EDS for solidified cement, the aluminum phase was well combined with cement, while the calcium phase was separated from cement. The volume of solidified cement in radioactive wastes was dependent on the waste-to-cement ratio and the amount of water, and increased by about 30% under the conditions used in this study. Therefore, it was concluded that permanent disposal of electrokinetically decontaminated radioactive wastes is appropriate.

• Journal of the Korean Chemical Society
• /
• v.44 no.6
• /
• pp.526-532
• /
• 2000

Cation exchange separation and inductively coupled plasma atomic emission spectrometric(ICP-AES) determination of ruthenium in HCl solutions were studied to quantitatively determine ruthenium in spent nuclear fuels. Ruthenium-bearing samples were dissolved with the mixed acid solution(9 : 1 mole ratio, HCl-HNO$_3$) using an acid digestion bomb. Based on the absorption spectra and ion exchange behaviour of ruthenium in hydrochloric acid media, its possible chemical species were discussed. On a cation exchange column (0.7 ${\times}$ 8.0 cm) packed with AG 50W ${\times}$ 8(100~200 mesh) and equilibrated with 0.5 M HCl, ruthenium was eluated with 0.5 M HCl while uranium was retained on the column. The established separation method was applied to a simulated spent nuclear fuel and resulted in the recovery of 98.5% with a relative standard deviation of 0.7%.

• Nuclear Engineering and Technology
• /
• v.28 no.2
• /
• pp.118-128
• /
• 1996

The lattice parameters of stoichiometric $UO_2$ and $U_{1-y}Er_{y}O_2$ in the range of y=0.01 to y =0.33 were determined with use of X-ray diffraction data. Oxygen potentials have been measured by means of a thermogravimetric method in the range of 1200~$1500^{\circ}C$ and $10^{-14}$ $\leq$ $Po_2$ $\leq$ $10^{-3}$ for pure $UO_2$ and $U_{1-y}Er_{y}O_{2{\pm}x}$ solid solutions with y=0.02, y=0.06 and y=0.20, respectively. Their oxygen partial pressures were maintained by controlling $CO_2$/CO mixture atmosphere, and the $Po_2$ values corresponding to x of $U_{1-y}Er_{y}O_{2{\pm}x}$ solid solutions were measured with an electrolyte oxygen sensor. The lattice parameter decreases linearly with an increase in the erbium content. The change of the lattice parameter can be expressed in a linear equation of y as a($\AA$) =5.4695-0.220y for 0 $\leq$y$\leq$0.33. The experimental coefficient of y -0.220 in $U_{1-y}Er_{y}O_2$ was an intermediate value between the calculated values -0.273 and -0.156 in the case of $U^{5+}$ and $U^{6+}$, respectively. The (equation omitted) has been found to undergo abrupt increase in the range of -360 to -270 kJ/mole for y=0.06 and -320 to -220 H/mole for y=0.20, respectively, in the temperature range of 1200-$1500^{\circ}C$. (equation omitted) increases with erbium content, but the effect of the dopant for x =0.01 is less significant than that for stoichiometry. The oxygen potentials for $UO_2$ and $U_{0.98}Er_{0.02}O_{2+x}$ can be approximately represented by the $U^{5+}$/$U^{4+}$ model but those for y$\geq$ 0.06 in $U_{1-y}Er_{y}O_{2{\pm}x}$ solid solutions cannot be interpreted by the mean uranium valence model.

• Analytical Science and Technology
• /
• v.14 no.4
• /
• pp.317-323
• /
• 2001

Dissolution procedures of Monsanto catalyst which has been used to produce acrylronitrile by ammoxidation of propylene have been studied. Optimum dissolution condition of the catalyst supported on silica was obtained by microwave digestion system with mixed of HCl, HF and $H_2O_2$. When a safety device was activated by increased pressure in microwave vessel, Bi, Fe, Mo, Sb and U were not volatilized even though silica was volatilized as $SiF_4$. Quantification results by this method were $SiO_2$ $50.5{\pm}0.4%$, $Sb_2O_3$ $29.6{\pm}0.6%$, $UO_2$ $10.2{\pm}0.1%$, $Fe_2O_3$ $6.1{\pm}0.1%$, $MoO_3$ $0.73{\pm}0.01%$ and $Bi_2O_3$ $0.49{\pm}0.01%$ by ICP-AES and the relative error was within ${\pm}10%$ except bismuth.

• Journal of Energy Engineering
• /
• v.13 no.4
• /
• pp.242-258
• /
• 2004

The duplex burnable poison absorbers concept was suggested by Korea Atomic Energy Research Institute. This BP rod is composed of inner region of natural U-Gd$_2$O$_3$ and outer shell of enriched UO$_2$-Er$_2$O$_3$. It is expected that this burnable absorber has same reactivity control capability with gadolinia burnable absorber used in extened fuel cycle. In order to evaluate the nuclear feasibility of duplex BPs, the nuclear design characteristics were compared with that of four types of burnable absorbers; gadolinia, erbia, IFBA, dysprosia duplex BP on 24 months fuel cycle for Korean Standard Nuclear Power plants. According to the evaluation results of nuclear characteristics, the duplex BPs were better than other BPs on k-infinitives, reactivity holddown worth (RHW), pin power peaking and moderator temperature coefficient (MTC). The possibility of nuclear core design was also confirmed based on the optimized fuel assemblies which were searched for a sensitivity analysis. Characteristics of core design with duplex BPs was compared with that of reference core with gadolinia BPs for cycle length, power peaking and MTC. The duplex BP core had a little longer cycle length by 4 to 7 days because of increased amount of fissile in enriched uranium at the outer shell of duplex BP In case of power peaking F$\_$Q/ of duplex BP core was reduced from 1.5773 to 1.5335. MTC was also less -0.48 pcm/C than that of reference core. Finally, evaluation of fuel cycle economy was performed for the manufacturing feasibility test and fuel cost evaluation with duplex BPs. Fuel cycle economy of duplex BP core almost was equivalent with that of gadolinia BP core.

• Korean Journal of Legislative Studies
• /
• v.23 no.1
• /
• pp.187-226
• /
• 2017

North Korea discourse is doubtful. A considerable portion is distorted under political objectives, group identity, and interests. Surely, there are facts based on North Korea's conducts. Apparent deceptions commonly exist as well though. Korean media does not endeavor to set the records straight and there are no revision towards mislead information. This is substantially dangerous as it can misjudge North Korean policies, beget national antipathy, and interferes with rational and constructive policy making. This study stems from such concerns and takes such cases as HEU(Highly Enriched Uranium) suspicion of 2002, dispute covering BDA(Banco Delta Asia)'s counterfeiting, and the abandonment of the Geneva Agreed Framework into consideration. The first part concentrates on fathoming the truth of the three cases. References from US government, academia, think tanks, media were inquired with an addition of secondary material from Korea and China. Secondly it examines whether domestic news properly reflects the precedent facts along the process of discovery. The cause and solution suggested by domestic media were organized and inductively reconstituted to frames. The last study questions the structural factors that reproduces suspicion analogs. Today's dangers facing Korean society are essentially not natural but artificial. This research hopes to foster peace by analyzing related discourses that are infamous to reinterpret reality.

• Economic and Environmental Geology
• /
• v.56 no.5
• /
• pp.603-618
• /
• 2023

This study focused on evaluating the suitability of the WRK (waste repository Korea) bentonite as a buffer material in the SNF (spent nuclear fuel) repository. The U (uranium) adsorption/desorption characteristics and the adsorption mechanisms of the WRK bentonite were presented through various analyses, adsorption/desorption experiments, and kinetic adsorption modeling at various pH conditions. Mineralogical and structural analyses supported that the major mineral of the WRK bentonite is the Ca-montmorillonite having the great possibility for the U adsorption. From results of the U adsorption/desorption experiments (intial U concentration: 1 mg/L) for the WRK bentonite, despite the low ratio of the WRK bentonite/U (2 g/L), high U adsorption efficiency (>74%) and low U desorption rate (<14%) were acquired at pH 5, 6, 10, and 11 in solution, supporting that the WRK bentonite can be used as the buffer material preventing the U migration in the SNF repository. Relatively low U adsorption efficiency (<45%) for the WRK bentonite was acquired at pH 3 and 7 because the U exists as various species in solution depending on pH and thus its U adsorption mechanisms are different due to the U speciation. Based on experimental results and previous studies, the main U adsorption mechanisms of the WRK bentonite were understood in viewpoint of the chemical adsorption. At the acid conditions (＜pH 3), the U is apt to adsorb as forms of UO₂²⁺, mainly due to the ionic bond with Si-O or Al-O(OH) present on the WRK bentonite rather than the ion exchange with Ca²⁺ among layers of the WRK bentonite, showing the relatively low U adsorption efficiency. At the alkaline conditions (>pH 7), the U could be adsorbed in the form of anionic U-hydroxy complexes (UO₂(OH)₃^-, UO₂(OH)₄^2-, (UO₂)₃(OH)₇^-, etc.), mainly by bonding with oxygen (O^-) from Si-O or Al-O(OH) on the WRK bentonite or by co-precipitation in the form of hydroxide, showing the high U adsorption. At pH 7, the relatively low U adsorption efficiency (42%) was acquired in this study and it was due to the existence of the U-carbonates in solution, having relatively high solubility than other U species. The U adsorption efficiency of the WRK bentonite can be increased by maintaining a neutral or highly alkaline condition because of the formation of U-hydroxyl complexes rather than the uranyl ion (UO₂²⁺) in solution,and by restraining the formation of U-carbonate complexes in solution.

• Korean Chemical Engineering Research
• /
• v.50 no.4
• /
• pp.696-701
• /
• 2012

Non-aqueous processes have been developed for stable management and reuse of spent fuels. Nowadays, a plan for the management of spent fuel is being sought focusing on a non-aqueous process in Korea. Named as pyroprocessing, it includes an electrolytic reduction process using molten salt at high temperature for the spent fuels, which provides metallic product for a following electro-refining process. The electrolytic reduction process utilizes electrochemical reaction producing Li to convert oxides into metals in high temperature LiCl medium. Various kinds of elements in the spent fuels would be distributed in the system according to their respective reactivity with the reductant, Li, and the medium, LiCl. This study elucidates the reactions of the elements to understand the behavior of composite elements on the spent fuels by thermodynamic calculations. Uranium and transuranic are reduced into their metallic forms while rare-earth oxides, except for Eu, are stable against the reaction at a process temperature. This study also covers the tendency of reactions with respect to the temperature and, finally, estimates radioactivity and heat load on the distributed phases based on the reference spent fuel characteristics.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.2
• /
• pp.133-156
• /
• 2013

Natural analogue studies play an important role in the safety case which requires multiple lines of evidence including the safety assessment for the geological disposal of radioactive wastes. In this study, foreign status of natural analogue studies was investigated by summarizing natural analogue results according to the research topics related with repository materials and radionuclide migration and retardation. Main results, issues, and applicability of the foreign natural analogue studies were also analyzed. The results of domestic natural analogue studies were classified into studies using uranium ore bodies, rocks, groundwaters, and archeological artifacts, respectively, and their main results were summarized. There are massive materials for natural analogue studies which have been carried out during last several decades but they have not been actively applied to the safety assessment and safety case development for the radioactive waster disposal. Thus, in this study, applicable methods of natural analogues were summarized and a methodology for improving their applicability was examined. Natural analogue study is apparently necessary to improve and illustrate the reliability of safety assessment for a radioactive waste repository. Therefore, it is necessary to develop a methodology and construct a natural analogue information database for the application of the results from natural analogue studies to safety case development.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
• /
• 2008.04a
• /
• pp.69-75
• /
• 2008

기포반응도는 월성발전소를 비롯한 CANDU형 원자로의 주된 안전성 쟁점사안으로 끊임없이 논의되어 왔다. 이는 설계기준사고가 노심에서 열에너지 불균형이 원인이 되어 기준이상의 핵연료 파손과 방사성물질 누출로 발전할 위험이 있는 사건들로 정의될 때, 사건 진행 과정에 기포반응도 증가는 조기에 운전중단을 실패할 경우 출력폭주로 이어지므로 사건의 결말이 중대사고로 전환될 위험이 크기 때문이다. 본 연구는 공개된 최신 핵자료인 ENDF/B-VII.0를 NJOY.99로 처리한 연속에너지 반응단면적 라이브러리를 구축하고 MCNP-4C에 접속하여 37봉 천연우라늄 핵연료다발의 표준노심격자에 대한 기포반응도를 시뮬레이션하여, 지금까지 각종문헌에 제시된 값들과 비교, 종합하므로 내제된 불확실성을 추정하는 내용이다. ENDF/B-VII.0 기반 MCNP-4C의 CANDU 노심격자 모델은 동일한 핵자료와 핵종농도를 사용한 WIMS-IAEA 모델과 비교할 때, 초기 노심의 임계도 오차 약 3.51mk가 연소 진행에 따라 $7.5\times10^{-4}mk$/MWD/teU의 비율로 감소하는 것으로 나타났다. 또한 MCNP-4C 예측기포반응도는 초기노심에서 기포율 50% 및 100%에 대해 각각 8.38 및 15.96mk, 평형노심에서 7.68 및 14.72mk로 계산된다. 이는 월성 2, 3, 4 FSAR의 초기노심 및 평형노심에서 100% 기포상태에 대한 값, 약15.0 및 10.6mk와 비교할 때, 초기노심은 약 1.0mk 평형노심은 약4, 1mk 보수적이지만, 다른 연구결과들과는 최대오차 ${\pm}1{\sim}2mk$ 이내에서 잘 일치하는 것으로 평가되었다. 본 연구는 CANDU 노심의 기포반응도 불확실성 요인의 규명 및 영향평가를 위한 노력의 일부로서 앞으로 감속재의 붕산농도 변화, 감속재 및 냉각재의 중수 순도 변화, 기기노화에 의한 격자 구조 및 물성 변화, 중성자속 및 출력 분포 불균형, 반응도조절장치의 위치, 등 주요 설계변수의 변화에 대한 반응도영향 분석연구를 계속할 계획이다.