• 에너지공학
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• 제25권1호
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• pp.153-162
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• 2016

요르단 수출용 연구용 원자로(JRTR, Jordan Research and Training Reactor)에 사용되는 판형핵연료(Plate-type-fuel)의 부수로를 모의하는 좁은 사각 유로에서 하향류 유동 조건의 임계열유속(CHF, Critical Heat Flux) 실험 연구가 수행되었다. 실험은 연구용 원자로 인허가 요건을 만족하는 유동 조건에서 수행되었으며 크기가 다른 두 가지 시험부를 이용하였다. 두 시험부는 각각 피션몰리(Fission moly) 우라늄 타겟의 부수로와 판형핵연료의 부수로를 모의하며 모두 원형과 같은 크기로 제작되었다. 각 시험부에 대해 임계열유속 실험이 수행되었으며 이를 통해 임계열유속에 영향을 주는 변수를 분석하였다. 그리고 실험결과를 기존의 임계열유속 모델과 비교하였으며 모델의 예측 정확도를 제시하였다. 이를 통해 좁은 사각 유로 내 하향류 유동 조건에서의 임계열유속 예측에 대한 기존 모델의 적용성을 분석하였다.

• 한국환경과학회지
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• 제6권4호
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• pp.399-407
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• 1997

The pyrolysis reactions of atomic hydrogen with chloroform were studied In a 4 cm 1.6. tubular flow reactor with low flow velocity 1518 cm/sec and a 2.6 cm 1.4. tubular flow reactor with high flow velocity (1227 cm/sec). The hydrogen atom concentration was measured by chemiluminescence titration with nitrogen dioxide, and the chloroform concentrations were determined using a gas chromatography. The chloroform conversion efficiency depended on both the chloroform flow rate and linear flow velocity, but 416 not depend on the flow rate of hydrogen atom. A computer model was employed to estimate a rate constant for the initial reaction of atomic hydrogen with chloroform. The model consisted of a scheme for chloroform-hydrogen atom reaction, Runge-Kutta 4th-order method for Integration of first-order differential equations describing the time dependence of the concentrations of various chemical species, and Rosenbrock method for optimization to match model and experimental results. The scheme for chloroform-hydrogen atom reaction Included 22 elementary reactions. The rate constant estimated using the data obtained from the 2.6 cm 1.4. reactor was to be 8.1 $\times$ $10^{-14}$ $cm^3$/molecule-sec and 3.8 $\times$ $10^{-15}$ cms/molecule-sec, and the deviations of computer model from experimental results were 9% and 12% , for the each reaction time of 0.028 sec and 0.072 sec, respectively.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.929-944
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• 2009

The present study is concerned with the extension of the Effective Convectivity Model (ECM) to the phase-change problem to simulate the dynamics of the melt pool formation in a Light Water Reactor (LWR) lower plenum during hypothetical severe accident progression. The ECM uses heat transfer characteristic velocities to describe turbulent natural convection of a melt pool. The simple approach of the ECM method allows implementing different models of the characteristic velocity in a mushy zone for non-eutectic mixtures. The Phase-change ECM (PECM) was examined using three models of the characteristic velocities in a mushy zone and its performance was compared. The PECM was validated using a dual-tier approach, namely validations against existing experimental data (the SIMECO experiment) and validations against results obtained from Computational Fluid Dynamics (CFD) simulations. The results predicted by the PECM implementing the linear dependency of mushy-zone characteristic velocity on fluid fraction are well agreed with the experimental correlation and CFD simulation results. The PECM was applied to simulation of melt pool formation heat transfer in a Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lower plenum. The study suggests that the PECM is an adequate and effective tool to compute the dynamics of core melt pool formation.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3580-3596
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• 2021

An integral test facility, PILLAR, was commissioned, aiming to provide valuable experimental results which can be referenced by system and component designers and used for the performance demonstration of liquid-metal-cooled, passive small modular reactors (SMRs) toward their licensing. The setup was conceptualized by a scaling analysis which allows the vertical arrangements to be conserved from its prototypic reactor, scaled uniformly in the radial direction achieving a flow area reduction of 1/200. Its final design includes several heater rods which simulate the reactor core, and a single heat exchanger representing the steam generators in the prototype. The system behaviors were characterized by its data acquisition system implementing various instruments. In this paper, we present not only a detailed description of the facility components, but also selected experimental results of both steady-state and transient cases. The obtained steady-state test results were utilized for the benchmark of a system code, achieving a capability of accurate simulations with ±3% of maximum deviations. It was followed by qualitative comparisons on the transient test results which indicate that the integral system behaviors in passive LBE-cooled systems are able to be predicted by the code.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Korean Membrane Journal
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• 제1권1호
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• pp.1-7
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• 1999

Pd-ceramic composite membranes and catalytic membrane reactors(CMR) have been studied for hydrogen and its isotopes (deuterium and tritium) purification and recovery in the fusion reactor fuel cycle. Particularly a closed-loop process has been studied for recovering tritium from tritiated water by means of a CMR in which the water gas shift reaction takes place. The development of the techniques for coating micro-porous ceramic tubes with Pd and Pd/Ag thin layers is described : P composite membranes have been produced by electroless deposition (Pd/Ag film of 10-20 $\mu$m) and rolling of thin metal sheets (Pd and Pd/Ag membranes of 50-70 $\mu$m). Experimental results of the electroless membranes have shown a not complete hydrogen selectivity because of the presence of some defects(micro-holes) in the metallic thin layer. Conversely the rolled thin Pd and Pd/ag membranes have separated hydrogen from the other gases with a complete selectivity giving rise to a slightly larger (about a factor 1.7) mass transfer resistance with respect to the electroless membranes. Experimental tests have confirmed the good performances of the rolled membranes in terms of chemical stability over several weeks of operation. Therefore these rolled membranes and CMR are adequate for applications in the fusion reactor fuel cycle as well as in the industrial processes where high pure hydrogen is required (i.e. hydrocarbon reforming for fuel cell)

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2984-2996
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• 2023

A 100 kW thermal power pool-type light water reactor and Pu(Be) as a fast neutron source were used to determine the appropriate carrier for irradiating boron-containing samples with neutron beams. The tested materials (carriers) were subjected to neutron beams in the reactor's tangential channel. The geometrical arrangement of experimental facilities relative to the neutron beam trajectory, as well as the effect of sample thickness on the count rate, were investigated. The majority of the detectable charged particles emitted by the neutron beam's interaction with tested materials and the detector's detecting layer are protons (recoiled hydrogen) and particles generated in nuclear reactions (protons and alpha particles), respectively. Stopping and Range of Ions in Matter (SRIM) software was used to do theoretical calculations for the range of expected released particles in various materials, including human tissue. The results of measurement and calculation are in good agreement. According to experiments and theoretical calculations, the number of protons emitted by tissue-like materials may commit a dose comparable to that of boron capture reactions. Furthermore, the range of protons is significantly larger than that of alpha particles, which most probably changes dose distribution in healthy cells surrounding the tumor, which is undesirable in the BNCT approach.

• 전력전자학회논문지
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• 제4권5호
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• pp.483-490
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• 1999

본 논문은 전동기 과전압 억제용으로 사용되는 인버터 출력 리액터의 최적 설계 방식을 제시하고 있다. 전력선의 길이가 상대적으로 짧은 엘리베이터 구동 시스템의 인버터 출력단에 리액터를 사용하는 경우, 전동기 선간의과 전압 특성을 과 전압 동작 주파수인 상당한 고 주파수 대역에서의 출력 리액터와 전동기 특성에 의해 좌우된다. 따라서, 고 주파수 대역에서의 출력 리액터 및 전동기의 동작 특성을 분석하고, 출력 리액터의 필요 파라미터를 추출하였다. 이를 리액터의 설계 변수하고, 리액터의 고 주파수 특성을 고주파수 특성을 고려하여, 새로운 구조의 출력 리액터 설계방식을 제안하였다. 통상의 출력 리액터와 제안된 방식의 출력 리액터를 용량 15kW급의 엘리베이터 유도 전동기 구동 시스템에서 비교 실험하여, 제안된 방식의 효과를 검증하였다.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.537-556
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• 2005

A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1355-1364
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• 2019

A recently published nuclear forensics methodology for source discrimination of separated weapons-grade plutonium utilizes intra-element isotope ratios and a maximum likelihood formulation to identify the most likely source reactor-type, fuel burnup and time since irradiation of unknown material. Sensitivity studies performed here on the effects of random measurement error and the uncertainty in intra-element isotope ratio values show that different intra-element isotope ratios have disproportionate contributions to the determination of the reactor parameters. The methodology is robust to individual errors in measured intra-element isotope ratio values and even more so for uniform systematic errors due to competing effects on the predictions from the selected intra-element isotope ratios suite. For a unique sample-model pair, simulation uncertainties of up to 28% are acceptable without impeding successful source-reactor discrimination. However, for a generic sample with multiple plausible sources within the reactor library, uncertainties of 7% or less may be required. The results confirm the critical role of accurate reactor core physics, fuel burnup simulations and experimental measurements in the proposed methodology where increased simulation uncertainty is found to significantly affect the capability to discriminate between the reactors in the library.