• Nuclear Engineering and Technology
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• 제56권1호
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• pp.283-291
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• 2024

Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of <10^-6/ry. However, in case of such accidents, the mixture of the molten fuel and structural materials interacts with sodium. This phenomenon is known as Molten Fuel Coolant Interaction (MFCI) and results in fragmentation of the melt due to various instabilities. The fragmented particles settle as a debris bed on the core catcher at the bottom of the reactor vessel, and continue to generate decay heat. Characteristics of the debris particles play a vital role in heat transfer from the bed and need thorough investigation. The size, shape, and physical state of the debris depend on the associated fragmentation mechanism, superheating of the melt, and sodium temperature. Experiments have been conducted by releasing simulated corium, a molten mixture of alumina and iron generated by the aluminothermy process at ~2400 ℃ into liquid sodium, to study the fragmentation phenomena. After the experiment, the fragmented debris was retrieved and the particle size distribution was determined by sieve analysis. The debris was subjected to microscopic investigation for obtaining morphological characteristics. Based on the characteristics of debris, an attempt has been made to assess of fragmentation mechanism of simulated corium in sodium.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4048-4056
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• 2023

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

• 방사성폐기물학회지
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• 제16권2호
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• pp.195-202
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• 2018

원칙적으로, 지층 처분은 고준위 방사성 폐기물의 최종 처분을 위한 안전한 방법으로 간주된다. 그러나 사용후핵연료에 함유된 $^{99}Tc$ 및 $^{129}I$와 같은 일부 장수명 핵분열 생성물은 지하 환경에서 흡수성이 적은 음이온 핵종으로 이동성이 매우 크며 수백 keV 범위의 베타선 방출로 생태계에 피폭선량을 야기시킬 수 있다. 따라서 이 두 핵종을 효율적으로 분리하여 방사능으로 유해하지 않은 핵종으로 전환할 수 있다면 처분 안정성에 긍정적인 영향을 줄 수 있다. 이를 위한 하나의 방법은 이 두 가지 핵종을 원자로에서 수명이 짧은 핵종 또는 안정적인 핵종으로 변환하는 것이다. 이를 위해 두 핵종을 태우는 데 어느 원자로 유형이 더 효율적인지 평가하는 것이 필요하다. 본 연구에서는 경수로(PWR), 중수로(CANDU) 및 고속로(SFR, MET-1000)의 $^{99}Tc$ 및 $^{129}I$의 핵 변환 시뮬레이션 결과를 비교하고 고찰하였다.

• 전기학회논문지
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• 제62권4호
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• pp.437-443
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• 2013

This paper was developed a monitoring and control system to use reactive power control algorithm. This algorithm could be improved voltage stability in power system. This method was controlled the voltage for stability improvement, effective usage of reactive power, and the increase of the power quality. PMS(Power Management System) has been calculate voltage sensitivity, and control reactive power compensation device. The voltage control was used to the FACTS, MSC/MSR(Mechanically Switched Capacitors/Reactors), and tap of transformer in power system. The reactive power devices in power system were control by voltage sensitivity ranking of each bus. Also, to secure momentary reactive power, it had been controlled as the rest of reactive power in the each bus. In here, reactive power has been MSC/MSR. The simulation result, First control was voltage control as fast response control of FACTS. Second control was voltage control through the necessary reactive power calculation as slow response control of MSR/MSR. Third control was secured momentary reactive reserve power. This control was method by cooperative control between FACTS and MSR/MSC. Therefore, the proposed algorithm was had been secured the suitable reactive reserve power in power system.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.296-307
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• 2020

Permanent magnet sodium flowmeter (PMSF) have been used to measure the sodium flow in fast breeder reactors. Due to the effects of irradiation, thermal cycling, time lapse, etc., the magnetic flux density of the PMSF will decrease after being used in the reactor for a period of time. Therefore, it must be calibrated regularly. But some flowmeters that immersed in sodium cannot be removed for an off-line calibration, so the on-line calibration is required. However, the best online calibration accuracy of PMSF using cross-correlation analysis method was 2.0-level without considering the repeatability. In order to further improve this work, the operational principle of the transducer in PMSF is analyzed and the design principle of the transducer is proposed. The transducers were tested on the sodium flow loop to collect the experimental data. The signal characteristics are analyzed from the time and frequency domains, respectively. The cross-correlation analysis method based on biased estimation is adopted to obtain the flow rate. The verification experimental results showed that the measurement accuracy is 1.0-level when the flow velocity is above 0.5 m/s, and the measurement accuracy is 3.0-level when the flow velocity is in the range of 0.2 m/s to 0.5 m/s.

• 한국환경보건학회지
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• 제31권2호
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• pp.99-106
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• 2005

The removals of organic matter, nitrogen and phosphate in wastewater were investigated with Sequencing Batch Reactor (SBR). Glucose and sodium acetate were Used for organic carbon source so as to know nutrient removal efficiency in proportion to MLSS concentration. In the case of glucose, the COD removal rate was $74\%,\;41\%\;and\;66\%$ in MLSS 5000, 3000 and 1000, respectively. On equal terms, the BOD was $57\%,\;21\%\;and\;38\%$, the T-N was $24\%,\;13\%\;and\;44\%$, and the T-P was $12\%,\;21\%\;and\;33\%$. As a result, the removal rate of organic materials showed the finest remove when MLSS was 5000, but the nutrient removal rate appeared as was best when MLSS was 1000. In the case of sodium acetate, the COD removal rate was $83\%,\;81\%\;and\;86\%$ in MLSS 5000, 3000 and 1000, respectively. On equal terms, the BOD was appeared by $76\%,\;82\%\;and\;92\%$, the T-N $57\%,\;42\%\;and\;78\%$, and the T-P $48\%,\;52\%\;and\;38\%$. As a result, organic and T-N removal rates were best when MLSS was 1000. But, the T-P removal rates were best when MLSS was 3000. Glucose was shown fast removal in reaction beginning, but screened by more efficient thing though sodium acetate removes organic matter, nitrogen and phosphate. Form of floc was ideal in all reactors regardless of carbon source and MLSS concentration. And its diameter was about $200\~500{\mu}m$.

• Nuclear Engineering and Technology
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• 제13권1호
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• pp.22-35
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• 1981

우리나라의 장기 우라늄 원광누적 소요량을 원자력 발전 계획모형, 원자로형 투입 방안 및 가능 핵주기에 따라 추정하였다. 투입 가능 모형은 가압경수로, 중수로 및 고속증식로로 선정하였으며, 가능 대체 핵주기로서는 가압경수로의 경우에, U자체 재순환 주기, U 및 Pu 자체 재순환 주기, 연소도 증가에 의한 개량 핵주기를 고려하였다. 또 U 자체 재순환이 가능한 경우에 대해서, 재처리 후 저장된 핵분열성 Pu 누적량을 계산하였으며, 이에 따라 고속증식로의 도입가능시기를 추정하였다. 우라늄 원광 누적 소요량의 최대치는 전세계 우라늄 원광 소요량의 약 4∼5%를 차지할 것으로 추정되었으며, 원자력 발전 계획 모형 상한의 경우에는 U 자체 재순환이 1990년부터 이루어질때, 2000년까지 1200MWe급 고속증식로 2기가 도입가능할 것으로 추정되었다.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.581-591
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• 2017

Influences of hold time and stress ratio on cyclic creep properties of Grade 91 steel were systemically investigated using a wide range of cyclic creep tests, which were performed with hold times (HTs) of 1 minute, 3 minutes, 5 minutes, 10 minutes, 20 minutes, and 30 minutes and stress ratios (R) of 0.5, 0.8, 0.85, 0.90, and 0.95 under tension loading cycles at $600^{\circ}C$. Under the influence of HT, the rupture time increased to HT = 5 minutes at R = 0.90 and R = 0.95, but there was no influence at R = 0.50, 0.80, and 0.85. The creep rate was constant regardless of an increase in the HT, except for the case of HT = 5 minutes at R = 0.90 and R = 0.95. Under the influence of stress ratio, the rupture time increased with an increase in the stress ratio, but the creep rate decreased. The cyclic creep led to a reduction in the rupture time and an acceleration in the creep rate compared with the case of monotonic creep. Cyclic creep was found to depend dominantly on the stress ratio rather than on the HT. Fracture surfaces displayed transgranular fractures resulting from microvoid coalescence, and the amount of microvoids increased with an increase in the stress ratio. Enhanced coarsening of the precipitates in the cyclic creep test specimens was found under all conditions.

• 대한금속재료학회지
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• 제49권12호
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• pp.937-944
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• 2011

Studies have been carried out in order to reduce fuel-cladding chemical interaction (FCCI) behavior of metallic fuel in sodium-cooled fast reactors (SFR) using an electroplating technique. A $20{\mu}m$ thick Cr layer has been plated by the electrochemical method in the Sargent bath over the HT9 (12Cr-1Mo) clad material and diffusion couple tests of the U-10Zr metallic fuel as well as the rare earth alloy (70Ce-29La) have been conducted. The results show that the Cr plating can prevent FCCI behavior along the fuel-clad interface. However, cracks developed through the thickness during plating, which resulted in the migration of some fuel constituents. Variation of bath temperature, application of pulse current, and post heat treatment have been conducted to control such cracks. We found out that some conditions like the pulse current and the post heat treatment enhanced the layer property by reducing the internal cracks and improving the diffusion couple test.

• 대한금속재료학회지
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• 제49권7호
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• pp.549-556
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• 2011

This study aimed to evaluate the performance of diffusion barriers in order to prevent fuel-cladding chemical interaction (FCCI) between the metallic fuels and the cladding materials, a potential hazard for nuclear fuel in sodium-cooled fast reactors. In order to prevent FCCI, Zr or V metal is deposited on the ferritic-martensitic stainless steel surface by physical vapor deposition with a thickness up to $5{\mu}m$. The diffusion couple tests using uranium alloy (U-10Zr) and a rare earth metal such as Ce-La alloy and Nd were performed at temperatures between 660~800$^{\circ}C$. Microstructural analysis using SEM was carried out over the coupled specimen. The results show that significant interdiffusion and an associated eutectic reaction ocurred in the specimen without a diffusion barrier. However, with the exception of the local dissolution of the Zr layer in the Ce-La alloy, the specimens deposited with Zr and V exhibited superior eutectic resistance to the uranium alloy and rare earth metal.