• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.3 no.3
• /
• pp.249-256
• /
• 2005

Thermal decomposition and stabilization characteristics of the solid cake after the dissolution of nitrate of the lagoon sludge was investigated. Most of the nitrates were dissolved in the water and removed to the filtrate, but small amount of nitrates, calcium carbonate and uranium were remained in the solid cake. The solid cake was thermally decomposed in the muffle furnace at $900^{\circ}C$ for 5 hours. Uranium, which is in the lagoon 1, was stabilized with $NaNO_3$ decomposition to $Na_{2}O{\cdot}2UO_3$ form. For the lagoon 2, it is confirmed that CaO, which was created by thermal decomposition of the $Ca(NO_3)_2$ and $CaCO_3$, was transferred to $Ca(OH)_2$ in the air with water. Because it is known that $Ca(OH)_2$ is stable material, further additives did not need to the stabilization of the thermal decomposition of the lagoons.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05b
• /
• pp.37-40
• /
• 1998

Fe량을 0 - 0.4 wt.%까지 변화시켜 Zr-Fe 이원계 합금을 진공아크용해로 제조한 후 1050 $^{\circ}C$에서 30분 균질화 처리 및 700 $^{\circ}C$에서의 열간압연, 그리고 2회의 걸친 냉간압연 및 열처리를 통하여 판재로 제조되었다. 중간 열처리 시 열처리 온도 및 시간을 조절하여 최종 냉간가공에 앞서 각 시편의 결정립 크기 및 경도 값이 같도록 조절하였다. 최종 냉간가공 시 냉간가공량을 60%로 동일하게 조절하였고, 최종열처리 시 열처리 온도 및 시간을 300-750 $^{\circ}C$, 5-3000분으로 각각 변화시켰다. 재결정 정도는 미세조직 관찰 및 경도 측정으로 평가되었으며, 석출물의 구조, 분포 및 형상 등도 TEM으로 분석되었다. 0.1 wt.% 정도까지의 Fe 첨가는 Fe를 첨가하지 않은 순수지르코늄에 비하여 지르코늄입자의 빠른 성장을 야기해, 조대한 재결정 지르코늄 입자들이 나타났다. 그러나 Fe 첨가량이 0.1wt.%이상 첨가되면, Zr$_3$Fe 석출물에 의한 입자성장 억제효과로 지르코늄 입자의 크기는 오히려 작아졌다. 결론적으로, Fe의 첨가는 지르코늄의 확산을 가속시켰다는 것을 보여준다. 한편, 750 $^{\circ}C$에서 열처리 시 이차 재결정현상으로 지르코늄 입자가 비정상적으로 매우 커졌으며, 동시에 annealing twining 현상이 관찰되었다. 이러한 annealing twining 현상은, 입자성장속도가 임계치 이상으로 갑자기 커진, Zr$_3$Fe 석출물이 거의 없는 합금에서만 나타났다. 이 결과를 토대로 annealing twining 현상은 입자의 빠른 이동이 필요 조건이라는 결론을 도출하였다. .Ar-4vol.%H$_2$ 분위기보다 H$_2$분위기에서 소결했을 때 밀도가 더 높았다. 그러나, 결정립은 $UO_2$와 $UO_2$-Li$_2$O의 경우, 수소분위기에서 소결했을 때, (U,Ce)O$_2$와 (U,Ce)O$_2$-Li$_2$O에서는 Ar-4vol.%H$_2$분위기에서 소결했을 때 더욱 성장하였다.설명해 줄 수 있다. 넷째, 불규칙적이며 종잡기 힘들고 단편적인 것으로만 보이던 중간언어도 일정한 체계 속에서 변화한다는 사실을 알 수 있다. 다섯째, 종전의 오류 분석에서는 지나치게 모국어의 영향만 강조하고 다른 요인들에 대해서는 다분히 추상적인 언급으로 끝났지만 이 분석을 통 해서 배경어, 목표어, 특히 중간규칙의 역할이 괄목할 만한 것임을 가시적으로 관찰할 수 있 다. 이와 같은 오류분석 방법은 학습자의 모국어 및 관련 외국어의 음운규칙만 알면 어느 학습대상 외국어에라도 적용할 수 있는 보편성을 지니는 것으로 사료된다.없다. 그렇다면 겹의문사를 [-wh]의리를 지 닌 의문사의 병렬로 분석할 수 없다. 예를 들어 누구누구를 [주구-이-ν가] [누구누구-이- ν가]로부터 생성되었다고 볼 수 없다. 그러므로 [-wh] 겹의문사는 복수 의미를 지닐 수 없 다. 그러면 단수 의미는 어떻게 생성되는가\ulcorner 본 논문에서는 표면적 형태에도 불구하고 [-wh]의미의 겹의문사는 병렬적 관계의 합성어가 아니라 내부구조를 지니지 않은 단순한 단어(minimal $X^{0}$ elements)로 가정한다. 즉, [+wh] 의미의 겹의문사는 동일한 구성요 소를 지닌 병렬적 합성어

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.3 no.1
• /
• pp.67-75
• /
• 2005

The size and main ingredient of hot particulate generated during the nuclide experiment in hot cells of nuclear facilities were 0.5300 $\mu$m and UO$\_2$. A cyclone filter equipment which consists of a cyclone and Bag/HEPA filter was devised to remove hot particulate generated during the nuclide experiment in hot cells of nuclear facilities. The experimental conditions to maximize the collection efficiency of hot particulate were suggested through experiments done with the cyclone filter equipment. With the large size of simulated particulate, the collection efficiency of the particulate was high. When the size of simulated particulate was more than 5 $\mu$m, the collection efficiency of the particulate was more than $80\%$ and when the size of simulated particulate was less than 1.0 urn, the collection efficiency decreased by less than $70\%$. If the inflow velocity of simulated particulate was increased, the collection efficiency of the particulate was also increased. When the inflow velocity of simulated particulate was more than 12m/sec, the collection efficiency was higher than $70\%$, but after 17 m/sec inflow velocity, no change observed. The collection efficiency of the simulated particulate can be enhanced with the length of vortex finder inside the chamber. With the length of vortex finder, 7.2cm, the observed collection efficiency of the particulate was the maximum. Moreover, when the sub-cone was attached under the cyclone, the collection efficiency of cyclone increased $2\%$. It was found that effect by attachment of sub-cone was not serious.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.335-343
• /
• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with $3.2\%$ of enrichment had been irradiated up to 35,000 MWd/MTU(reference data). The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, chemical analysis (destructive method) has been used but it mattes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

• Applied Microscopy
• /
• v.35 no.3
• /
• pp.113-119
• /
• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with 3.2% of enrichment had been irradiated up to 35,000 MWd/MTU. The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, destructive method analysis has been used but it makes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

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

• Proceedings of the KSME Conference
• /
• 2001.06e
• /
• pp.51-56
• /
• 2001

Vapor explosion is one of the most important problems encountered in severe accident management of nuclear power plants. In spite of many efforts, a lot of questions still remain for the fundamental understanding of vapor explosion phenomena. Therefore, KAERI launched a real material experiment called TROI using 20 kg of UO2 and ZrO2 to investigate the vapor explosion phenomena. In addition, a small-scale experiment with molten-tin/water system was performed to quantify the characteristics of vapor explosion and to understand the phenomenology of vapor explosion. A number of instruments were used to measure the physical change occurring during the vapor explosion. In this experiment, the vapor explosion generated by molten fuel water interaction is visualized using high speed camera and the pressure behavior accompanying the explosion is investigated.

• Nuclear Engineering and Technology
• /
• v.50 no.1
• /
• pp.145-150
• /
• 2018

A new methodology, the crack-spallation model, has been developed to analyze gas-solid reactions dominated by crack growth inside of the solid reactant and spallation phenomena. The new model physically represents three processes of the reaction progress: (1) diffusion of gas reactant through pores; (2) growth of product particle in pores; and (3) crack and spallation of solid reactant. The validation of this method has been conducted by comparison of results obtained in an experiment for oxidation of $UO_2$ and the shrinking core model. The reaction progress evaluated by the crack-spallation model shows better agreement with the experimental data than that evaluated by the shrinking core model. To understand the trigger point during the reaction progress, a detailed analysis has been conducted. A parametric study also has been performed to determine mass diffusivities of the gas reactant and volume increase constants of the product particles. This method can be appropriately applied to the gas-solid reaction based on the crack and spallation phenomena such as the voloxidation process.

• Proceedings of the Korean Powder Metallurgy Institute Conference
• /
• 2006.09a
• /
• pp.187-188
• /
• 2006

In contrast with the Finite Element Method, the Discrete Element Method (DEM) takes explicitly into account the particulate nature of powders. DEM exhibits some drawbacks and many advantages. Simulations can be computationally expensive and they are only able to represent a volume element. However, these simulations have the great advantage of providing a wealth of information at the microstructural level. Here we demonstrate that the method is well suited for modelling, in coordination with FEM, the compaction of ceramic $UO_2$ particles that have been aggregated. Aggregates of individual ceramic crystallites that are strongly bonded together are represented by porous spheres.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.112-116
• /
• 1996

신형경수로의 대안으로서 가압경수로의 단점을 보완하고, 가압중수로의 장점을 채택한 중수감속 경수로의 핵적 개념설계를 제안하였다. 냉각재와 감속재가 서로 다른 채넬을 통해 흐르는 기존 가압중수로의 Pressure-Tube 설계의 장점을 채택하여, 냉각재는 경수를 감속재는 중수를 사용하는 중수감속 가압경수로(DPWR, Deuterium-moderated PWR)의 설계 타당성을 검토하였다. 기본적으로 CANDU의 system설계를 Proven Technology로서 가능한 많이 채택하고, CANFLEX 핵연료 설계도 기존 연구 결과로서 최대한 활용하였다. 월성 2,3,4호기 FSAR의 사양을 그대로 사용하여 기존 중수로의 37봉 핵연료 다발을 6$\times$6 직각 배열 등가 핵연료집합체로 재구성한 후, SEU $UO_2$ 핵연료에 대해 HELIOS코드를 사용하여 핵적 특성을 검토하였다. 냉각재 온도계수가 음의 안전성을 갖고 있으며, 기존 중수로보다 연소도가 훨씬 큰 원자로가 설계될 수 있음을 확인하였다. 또한 발전소 이용률의 증대, 사용후 핵연료 발생량의 감소를 기대할 수 있었다.