• 전력전자학회논문지
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• 제26권1호
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• pp.25-31
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• 2021

This paper describes the design of a power cell-based pulsed power modulator with fast rise times. The pulse-generating section of the pulse power modulator is a series stack of power cells. Each power cell is composed of a storage capacitor, a pulse switch, and a bypass diode. When the pulse switches are turned on, the capacitors are connected in series and the sum of voltages is applied to the load. For output pulses with fast rise times, an IGBT with fast turn-on characteristics is adopted as a pulse switch and the optimized gate driving method is used. Pspice simulation is performed to account for the gate driving method. A 10 kV, 12-power cell-based pulsed power modulator is tested under resistive load and plasma reactor load. The rise times of output pulses less than 20 ns are confirmed, showing that the pulsed power modulator can be effectively applied to pulsed power applications with fast rise times.

• 한국환경과학회지
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• 제20권8호
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• pp.929-942
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• 2011

A novel conductivity technique was developed to detect penetration depth of the vessel fluid into the feedpipe. For a given reactor geometry, critical agitator speeds were experimentally determined at the onset of feedpipe backmixing using Rushton 6 bladed disk turbine (6BD) and high efficiency axial flow type 3 bladed (HE-3) impellers. The ratio of the feedpipe velocity to the critical agitator speed ($v_f/v_t$) was constant for either laminar or turbulent feedpipe flow regimes. Compared to the results of fast competitive reaction, feedpipe backmixing had to penetrate at least one feedpipe diameter into the feedpipe to significantly influence the yield of the side product. However, higher $v_f/v_t$ than that for L/d = 0 (position at the feedpipe end) of the conductivity technique is recommended to completely eliminate feedpipe backmixing in conservative design criteria. The conductivity technique was successful in all feedpipe flow conditions of laminar, transitional and turbulent flow regimes.

• 한국압력기기공학회 논문집
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• 제10권1호
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• pp.107-112
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• 2014

The Code of Federal Regulations, Title 10, Part 50, Appendix H requires surveillance program for reactor pressure vessel(RPV) that the peak neutron fluence at the end of the design life of the vessel will exceed $1.0E+17n/cm^2$ (E>1.0MeV). 2D/1D Synthesis method based on DORT 3.1 transport calculation code has been widely used to determine fast neutron(E>1.0MeV) fluence exposure to RPV in the beltline region. RAPTOR-M3G(RApid Parallel Transport Of Radiation-Multiple 3D Geometries) performing full 3D transport calculation was developed by Westinghouse and KRIST(Korea Reactor Integrity Surveillance Technology) and applied for the evaluations of In-Vessel and Ex-Vessel neutron dosimetry. The reaction rates from measurement and calculation were compared and the results show good agreements each other.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.115-125
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• 2016

The prototype SFR has currently been under design by KAERI. The size of its major components is much larger than that of APR1400 and high temperature materials are applied for it. The increased size of components and those specific materials effect on material procurement, manufacturing process and fabrication facilities. The manufacturing methods are studied for Reactor Vessel/Guard Vessel, Control Rod Drive Mechanism, Heat Exchanger, Primary Pump, Reactor Vessel Internals, Steam Generator and In-Vessel Transfer Machine. The proper manufacturing methods are suggested for each component including side forging technology for ultra large forgings of Reactor Vessel to minimize the weld seams on which In-service Inspection should be conducted.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.376-385
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• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.368-375
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• 2016

A sodiume-water reaction takes place when high-pressured water vapor leaks into sodium through a tiny defect on the surface of the heat transfer tube in a steam generator of the sodium-cooled fast reactor. The sodiume-water reaction brings deterioration of the mechanical strength of the heat transfer tube at the initial leakage site. As a result, it damages the crack itself, which may eventually enlarge into a larger opening. This self-enlargement is called "self-wastage phenomenon." In this study, a simulant experiment was proposed to reproduce the self-enlargement of a crack and to evaluate the mechanism of the self-wastage. The damage on the surface of the crack was simulated by making the neutralization reaction with hydrochloric acid solution and sodium hydroxide solution. A numerical investigation was carried out to validate the feasibility of the approach and to determine experimental conditions. From the computation results, it is observed that when 5M HCl is injected into 5M of NaOH with 0.05 m/s inlet velocity, the temperature at the surface near the crack increased over 319.26 K. The computational results show that the self-wastage phenomenon is capable of being reproduced by the simulant experiment.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.185-190
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• 1996

RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the ＊SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4084-4094
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• 2022

The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. The first cycle of FAST testing was completed and four rodlets were removed for the initial post irradiation examination (PIE). The rodlet design and irradiation conditions were evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % fissions per initial metal atoms [%FIMA]), control rodlets, and a helium-bonded annular rodlet (4.7 %FIMA). Nondestructive PIE has been completed and includes visual inspection, neutron radiography and gamma scanning of the FAST capsules and rodlets. Radiography confirmed the integrity of the experiments, revealed that the annulus in the annular fuel was filled at a modest burnup (4.7 %FIMA), and indicated potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior, except for cesium in the He-bonded annular fuel. Future destructive PIE will be necessary to fully interpret the effects of accelerated irradiation on U-Zr metallic fuel behavior.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1874-1889
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• 2022

In this study, we evaluate the thermal-hydraulic performance and economics of Printed Circuit Heat Exchanger (PCHE) according to the channel types and associated shape variables for the design of recuperators with Sodium-cooled Fast Reactors (SFRs). To perform the evaluations with variables such as the Reynolds number, channel types, tube diameter, and shape variables, a code for the heat exchanger is developed and verified through a comparison with experimental results. Based on the code, the volume and pressure drop are calculated, and an economic assessment is conducted. The zigzag type, which has bending angle of 80° and a tube diameter of 1.9 mm, is the most economical channel type in a SFR using CO₂ as the working fluid. For a SFR using N₂, we recommend the airfoil type with vertical and horizontal numbers of 1.6 and 1.1, respectively. The airfoil type is superior when the mass flow rate is large because the operating cost changes significantly. When the mass flow rate is small, volume is a more important design parameter, therefore, the zigzag type is suitable. In addition, we conduct a sensitivity analysis based on the production cost of the PCHE to identify changes in optimal channel types.

• JSTS:Journal of Semiconductor Technology and Science
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• 제5권3호
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• pp.149-158
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• 2005

This paper describes design, fabrication, and application of the silicon based temperature controllable micro reactor. In order to achieve fast temperature variation and low energy consumption, reaction chamber of the micro reactor was thermally isolated by etching the highly conductive silicon around the reaction chamber. Compared with the model not having thermally isolated structure, the thermally isolated micro reactor showed enhanced thermal performances such as fast temperature variation and low energy consumption. The performance enhancements of the micro reactor due to etched holes were verified by thermal experiment and numerical analysis. Regarding to 42 percents reduction of the thermal mass achieved by the etched holes, approximately 4 times faster thermal variation and 5 times smaller energy consumption were acquired. The total size of the fabricated micro reactor was $37{\times}30{\times}1mm^{3}$. Microchannel and reaction chamber were formed on the silicon substrate. The openings of channel and chamber were covered by the glass substrate. The Pt electrodes for heater and sensor are fabricated on the backside of silicon substrate below the reaction chamber. The dimension of channel cross section was $200{\times}100{\mu}m^{2}$. The volume of reaction chamber was $4{\mu}l$. The temperature of the micro reactor was controlled and measured simultaneously with NI DAQ PCI-MIO-16E-l board and LabVIEW program. Finally, the fabricated micro reactor and the temperature control system were applied to the thermal denaturation and the trypsin digestion of protein. BSA(bovine serum albumin) was chosen for the test sample. It was successfully shown that BSA was successfully denatured at $75^{\circ}C$ for 1 min and digested by trypsin at $37^{\circ}C$ for 10 min.