• 대한화학회지
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• 제27권4호
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• pp.279-286
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• 1983

몇가지의 알파입자를 방출하는 핵종, 즉 악티늄족 원소들, $^{207}Bi$ and $^{210}Po$을 전해석출하는 장치를 만들었다. 스텐레스 원판으로된 환원전극에 이 동위원소들을 석출했으며(석출부분 직경=18mm), 산화전극으로는 백금선을 썼다. 전해질로 염화암모늄을 쓰고, 초기 pH=4, 염소이온농도=0.6M이하, 용액부피=15ml로 하여 1.5암페어(전류밀도=0.59A/$cm^2$)의 전류를 100분간 흘려주어 98.3%의 석출 회수율과 ${\pm}$0.7%의 재현도를 얻었다. 석출된 시료의 알파스펙트럼을 측정한 결과 에너지 분리도로서 $^{210}Po=18.3keV, ^{234}U=21.8keV$ 및 $^{239}Pu=36.0keV$인 반치전폭(full width at half maximum)을 얻었다. 국산 천연우라늄(충북·괴산) 시료를 전해석출하여 그의 알파스펙트럼을 구한 결과 $^{238}U\;:\;^{234}U\;=\;:\;6.1{\times}10^{-5}$을 얻었으며 $1.8{\sim}10^{13} neutrons/cm^2{\cdot}sec$인 중성자속으로 144일 동안 쪼여준 238U 시료를 전해석출하여 그의 알파스펙트럼을 구한 결과 $^{238}U\;:\;^{239}Pu\;:\;241Am\;=\;100\;:\;0.0263\;:\;5.20{\times}10^{-5}$을 얻었다. 조사시료중의 $^{238}U$에 대한 본실험의 정량결과는 고체형광측정법 및 질량 스펙트럼법에 의한 결과들과 상대오차 1.6% 이내에서 일치하였으며, $^{239}Pu$의 경우는 음이온교환분리-알파스펙트럼 측정 및 삼불화테노일아세톤(thenoyltrifluoroacetone)을 쓴 용매추출-알파스펙트럼 측정에 의한 정량결과들과 상대오차 ${\pm}$4.0%이내에서 일치하였다.

• 한국수소및신에너지학회논문집
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• 제24권4호
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• pp.295-301
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• 2013

Long-term global energy-demand growth is expected to increase driven by strong energy-demand growth from developing countries. Fusion power offers the prospect of an almost inexhaustible source of energy for future generations, even though it also presents so far insurmountable scientific and engineering challenges. One of the challenges is safe handling of hydrogen isotopes. Metal hydrides such as depleted uranium hydride or ZrCo hydride are used as a storage medium for hydrogen isotopes reversibly. The metal hydrides bind with hydrogen very strongly. In this paper, we carried out a modeling and simulation work for absorption/desorption of hydrogen by ZrCo in a horizontal annulus cylinder bed. A comprehensive mathematical description of a metal hydride hydrogen storage vessel was developed. This model was calibrated against experimental data obtained from our experimental system containing ZrCo metal hydride. The model was capable of predicting the performance of the bed for not only both the storage and delivery processes but also heat transfer operations. This model should thus be very useful for the design and development of the next generation of metal hydride hydrogen isotope storage systems.

• 방사성폐기물학회지
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• 제7권3호
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• pp.167-174
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• 2009

EBR-II사용후핵연료의 파이로건식처리공정에 의해 발생된 우라늄의 순도에 대한 포괄적인 분석을 수행하였다. 분석 결과를 미국 아이다호 국립연구소 및 한국원자력 연구원의 협력과제 하에서 한국과 미국의 저준위 폐기물 기준으로 비교하였다. 미국의 저준위 폐기물 기준은 우라늄 등위원소를 포함하지 않으나, 한국의 경우는 포함하는 것으로 조사되었다. 분석 결과 EBR-II 우라늄 생성물 내에서 저준위 기준을 초과하는 유일한 알파 핵종은 우라늄 동위원소가 아니라 Pu-239였다. 생성물 내의 Pu 오염은 개량된 염증류공정을 통한 예비실험 결과 획기적으로 줄일 수 있음을 알 수 있었으며, 보다 공정을 개선시킨다면 제안된 기술을 이용하여 미국의 저준위 기준을 만족시킬 수 있을 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.47-58
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• 2015

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. Thefsensitivity of cooling time before prior to pyro-processing was studied. As the cooling time sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to ${\leq}20%$ in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

• 분석과학
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• 제17권4호
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• pp.302-307
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• 2004

환경시료중의 극미량의 악티늄계 동위원소들을 분석하기는 무척 어렵다. 이들 원소들은 개별 분리하는 작업이 필요하며, 알파분광법으로 분석한 어떤 핵종들은 검출감도도 높은 편이다. 이런 극미량의 악티늄계 동위원소들을 분석하기 위해 용매추출법이 결합된 TRU-Spec 이온교환수지와 음이온 교환수지를 사용하여 악티늄계 원소들을 분리한 후 알파분광법으로 검출하였다. 그리고 U과 Th의 검출한계를 낮추기 위해 중성자방사화분석법을 적용하였다. 중성자방사화분석법을 적용하기 위한 바탕물질로 고순도 V foil을 사용하여 검출감도를 10배 향상시킬 수 있었으며, 이 분석법을 표준시료인 NIST-4354, IAEA-368 퇴적물 시료에 적용한 결과 표준값과 10% 이내에서 잘 일치하였다.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.79-88
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• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.875-883
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• 2015

The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing $^{99}Mo$. In this method, the medical isotope production system itself is used to extract $^{99}Mo$ or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of $^{99}Mo$ by irradiating targets. In this study, the neutronic performance and $^{99}Mo$, $^{89}Sr$, and $^{131}I$ production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ~1,500 Ci/wk (~325 6-day Ci) of $^{99}Mo$ at the end of a cycle.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.645-650
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• 2017

This technical note demonstrates the feasibility of using laser ablation inductively coupled plasma mass spectrometry for the characterization of U-7Mo/Ale5Si dispersion fuel. Our measurements show 5.0% Relative Standard Deviation (RSD) for the reproducibility of measured $^{98}Mo/^{238}U$ ratios in fuel particles from spot analysis, and 3.4% RSD for $^{98}Mo/^{238}U$ ratios in a NIST-SRM 612 glass standard. Line scanning allows for the distinction of U-7Mo fuel particles from the Al-5Si matrix. Each mass spectrum peak indicates the presence of U-7Mo fuel particles, and the time width of each peak corresponds to the size of that fuel particle. The size of the fuel particles is estimated from the time width of the mass spectrum peak for $^{98}Mo$ by considering the scan rate used during the line scan. This preliminary application clearly demonstrates that laser ablation inductively coupled plasma mass spectrometry can directly identify isotope ratios and sizes of the fuel particles in U-Mo/Al dispersion fuel. Once optimized further, this instrument will be a powerful tool for investigating irradiated dispersion fuels in terms of fission product distributions in fuel matrices, and the changes in fuel particle size or shape after irradiation.

• 한국재료학회:학술대회논문집
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• 한국재료학회 2011년도 춘계학술발표대회
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• pp.15-15
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• 2011

As you know, boron compounds, borax ($Na_2B_4O_5(OH)_4{\cdot}8H_2O$) etc. were known thousands of years ago. As for natural boron, it has two naturally occurring and stable isotopes, boron 11 ($^{11}B$) and boron 10 ($^{10}B$). The neutron absorption $^{10}B$ is included about 19~20% with 80~81% $^{11}B$. Boron is similar to carbon in its capability to form stable covalently bonded molecular networks. The mass difference results in a wide range of ${\beta}$ values between the $^{11}B$ and $^{10}B$. The $^{10}B$ isotope, stable with 5 neutrons is excellent at capturing thermal neutrons. For example, it is possible to decrease a thermal neutron required for the nuclear reaction of uranium 235 ($^{235}U$). If $^{10}B$ absorbs a neutron ($^1n$), it will change to $^7Li+^1{\alpha}$ (${\alpha}$ ray, like $^4He$) with prompt ${\gamma}$ ray from $^{11}B$ $^{11}B$ (equation 1). $$^{10}B+^1n\;{\rightarrow}\;^{11}B\;{\rightarrow}\; prompt \;{\gamma}\;ray (478 keV), \;^7Li+4{\alpha}\;(4He)\;\;\;\;{\cdots}\; (1)$$ If about 1% boron is added to stainless steel, it is known that a neutron shielding effect will be 3 times the boron free steel. Enriched boron or $^{10}B$ is used in both radiation shielding and in boron neutron capture therapy. Then, $^{10}B$ is used for reactivity control and in emergency shutdown systems in nuclear reactors. Furthermore, boron carbide, $B_4C$, is used as the charge of a nuclear fission reaction control rod material and neutron cover material for nuclear reactors. The $B_4C$ powder of natural B composition is used as a charge of a control material of a boiling water reactor (BWR) which occupies commercial power reactors in nuclear power generation. The $B_4C$ sintered body which adjusted $^{10}B$ concentration is used as a charge of a control material of the fast breeder reactor (FBR) currently developed aiming at establishment of a nuclear fuel cycle. In this study for new boron compound, silicon boride ceramics for capturing thermal neutrons, preparation and characterization of both silicon tetraboride ($SiB_4$) and silicon hexaboride ($SiB_6$) and ceramics produced by sintering were investigated in order to determine the suitability of this material for nuclear power generation. The relative density increased with increasing sintering temperature. With a sintering temperature of 1,923 K, a sintered body having a relative density of more than 99% was obtained. The Vickers hardness increased with increasing sintering temperature. The best result was a Vickers hardness of 28 GPa for the $SiB_6$ sintered at 1,923K for 1 h. The high temperature Vickers hardness of the $SiB_6$ sintered body changed from 28 to 12 GPa in the temperature range of room temperature to 1,273 K. The thermal conductivity of the SiB6 sintered body changed from 9.1 to 2.4 W/mK in the range of room temperature to 1,273 K.