• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1451-1456
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• 2017

Finding technical issues associated with equipment scale-up is an important subject for the investigation of pyroprocessing. In this respect, electrolytic reduction of 1 kg $UO_2$, a unit process of pyroprocessing, was conducted using graphite as an anode material to figure out the scale-up issues of the C anode-based system at pilot scale. The graphite anode can transfer a current that is 6-7 times higher than that of a conventional Pt anode with the same reactor, showing the superiority of the graphite anode. $UO_2$ pellets were turned into metallic U during the reaction. However, several problems were discovered after the experiments, such as reaction instability by reduced effective anode area (induced by the existence of $Cl_2$ around anode and anode consumption), relatively low metal conversion rate, and corrosion of the reactor. These issues should be overcome for the scale-up of the electrolytic reducer using the C anode.

• 전기화학회지
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• 제18권3호
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• pp.121-129
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• 2015

사용후핵연료 재활용을 위한 파이로프로세싱의 전해환원 공정에서는 $Li_2O-LiCl$ 용융염을 전해질로 사용하며 금속산화물 형태의 사용후핵연료를 음극, 백금을 양극으로 사용하여 금속전환체를 제조한다. 따라서, 음극에서는 금속산화물이 금속으로 전환되는 환원반응으로 인해 산소 이온이 생성되고, 양극에서는 그 산소이온이 산소 가스가 되는 산화반응이 발생한다. $650^{\circ}C$의 운전 온도에서 발생하는 양극의 산소 가스로 인한 금속 재질 장치의 부식을 막기 위해 양극을 둘러싸는 슈라우드(shroud)를 사용해 산소 가스를 포집하여 전해질로의 확산을 막는 동시에 장치 외부로 배출되도록 한다. 기존에는 슈라우드 자체의 부식과 산소 가스의 염 내 확산을 방지하기 위하여 세라믹을 사용하였으나 비다공성 재질로 인해 산소 이온의 백금 표면으로의 이동 경로를 제한하여 공정의 속도를 좌우하는 전류 크기를 낮춘다는 문제점이 있었다. 이러한 문제를 극복하기 위하여 스테인레스 스틸 mesh로 구성된 다공성 슈라우드의 사용이 수 그램 규모 실험을 통해 제안된 바 있다. 본 연구에서는 킬로그램 규모의 우라늄산화물 전해환원 운전을 통해 다공성 슈라우드의 안정성을 확인 하고자 하였다. 음극의 우라늄산화물로는 크기 1~4 mm, 밀도 $10.30{\sim}10.41g/cm^3$의 파쇄 펠렛 1 kg이 사용되었으며, 백금 전극과 다공성 슈라우드가 포함된 양극 모듈을 사용하였다. 전해환원 종료후 음극에서 우라늄 금속이 성공적으로 얻어졌으며, 백금 양극 및 다공성 슈라우드도 손상 없이 안정하게 사용되었다. $650^{\circ}C$에서의 LiCl의 점도와 동일한 물과 에틸렌글리콜의 혼합물에서 산소 가스를 주입하여 확인 결과 산소 버블이 다공성 슈라우드 외부로 유출되는 것은 관찰되지 않았다.

• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1733-1739
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• 2017

Molecular dynamics simulations were performed to investigate the uniaxial tensile properties of nanocrystalline $U_{0.5}Th_{0.5}O_2$ solid solution with the Born-Mayer-Huggins potential. The results indicated that the elastic modulus increased linearly with the density relative to a single crystal, but decreased with increasing temperature. The simulated nanocrystalline $U_{0.5}Th_{0.5}O_2$ exhibited a breakdown in the Halle-Petch relation with mean grain size varying from 3.0 nm to 18.0 nm. Moreover, the elastic modulus of $U_{1-y}Th_yO_2$ solid solutions with different content of thorium at 300 K was also studied and the results accorded well with the experimental data available in the literature. In addition, the fracture mode of nanocrystalline $U_{0.5}Th_{0.5}O_2$ was inclined to be ductile because the fracture behavior was preceded by some moderate amount of plastic deformation, which is different from what has been seen earlier in simulations of pure $UO_2$.

• 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.

• 대한화학회지
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• 제56권1호
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• pp.62-67
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• 2012

A series of homo binuclear complexs of the type $[M_2(L/L^')(NO_3)n].mH_2O$, [where $M=U{O_2}^{2+},\;Th^{4+},\;ZrO^{2+}$] and $[(VO)_2(L/L^')(SO_4)_2]{\cdot}2H_2O$, L=1,5,6,9,12,15,16,20 octaaza-7,813,14-tetraphenyl-2,4,17,19-tetramethyl-1,4,6,8,12,14,16,19-docosaoctene (OTTDO) or L'=10:11;21:22-dibenzo-1,5,6,9,12,15,16,20-octaaza-7,813,14-tetraphenyl-2,4,17,19-tetramethyl-1,4,6,8,12,14,16,19-docosaoctene (DOTTOT), n=4 for $U{O_2}^{2+}$, $ZrO^{2+}$ n=8 for $Th^{4+}$ m=1,2,3 respectively, have been synthesized in template method from ethylenediamine/orthophenylene diamine, benzil monohydrazone and acetyl acetone and characterized on the basis of elemental analysis, thermal analysis, molar conductivity, magnetic moment, electronic, infrared, $^1H$-NMR studies. The results indicate that the VO(IV) ion is penta co-ordinated yielding paramagnetic complexes; $UO_2(VI)$, ZrO(IV) ions are hexa co-ordinated where as Th(IV) ion is octa co-ordinated yielding diamagnetic complexes of above composition. The fungi toxicity of the ZrO(IV) and VO(IV) complexes against some fungal pathogen has been studied.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• 분석과학
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• 제17권4호
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• pp.355-362
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• 2004

하나로 원자로에서 조사된 최대 선출력이 121 kW/m이고, 63 at%의 평균 연소도를 갖는 $U_3Si-Al$ 원심 분무 고출력 핵연료를 EPMA를 이용하여 파단면 관찰 및 반응층에 대한 핵분열 생성물을 분석 하였다. 조사된 고출력 $U_3Si-Al$ 핵연료를 EPMA로 화학 조성을 분석하기 위해 선행조건은 방사능 허용 한도가 $3{\times}10^{10}Bq$ 이하로 제한되는 EPMA 기기에 부합 될 수 있게 시험 시편을 최소화 하기 위한 작업이다. 시험 조건에 부합될 수 있는 시편의 제조를 위해 핵연료 천공 장치를 제작하였으며, 천공 장치를 사용하여 ${\Phi}1.57{\times}2mm$의 크기를 갖는 시료를 만들었다. 천공 된 시료를 파단 시편과 연마 시편으로 제조하여 파단면의 관찰 및 반응층(Inter-reaction layer)과 산화층에 대한 EPMA 분석을 수행하였다. 두께가 $16{\mu}m$인 반응층에 대한 평균값은 $UO_2$를 표준 시편으로 calibration한 경우의 조성은 $U_{2.84}$ Si $Al_{14}$ 이였으며, 시험 시편으로 calibration한 경우의 조성은 $U_{3.24}$ Si $Al_{14.1}$ 였다. 또한 반응층에서 핵분열 생성물의 조성을 분석하였으며, 반응층에서의 금속 석출물(metallic precipitates)의 생성은 확인할 수 없었다. 시험 시편의 산화층 조성은 $Ai_2O_3$ 임을 확인했다.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.499-507
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• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

• 한국표면공학회지
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• 제52권6호
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• pp.321-325
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• 2019

Triso-type coating layers of silicon carbide and graphite on UO₂ paticulate nuclear fuel were prepared by using fluidized bed type chemical vapor deposition and self-propagating high temperature synthesis methods to make a coated nuclear fuel of a power plant for hydrogen mass-production. The source and carrier gases were the mixture of methyltrichlorosilane and propane, and inert argon. Chemical analysis and microstructure observation showed that the coated layers were inner graphite, middle silicon carbide and outer graphite. The elastic modulus and nano-hardness of the silicon carbide layer were 503 [GPa] and 36 [GPa], respectively.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.248-257
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• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.