• Journal of the Korean Magnetics Society
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• v.4 no.4
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• pp.303-306
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• 1994

Fundamental characteristics of giant magnetostrictive alloy $Terfenol-D(Tb_{0.3}Dy_{0.7}Fe_{1.9~1.95})$ are rreasured and discussed on the application for actuators. The magnetostriction is measured by laser displacement rreasuring system and the applied compressive stress is measured by load cell. Magnetostrictions increased as the applied compressive stresses increased. When the stress is 7 MPa, the magnetostriction is 1000 ppm at 1500 Oe. As the stresses iocreased from 0 to 14 MPa, the magnetic fields for saturating the magnetostriction also increased. The temperature increased during the experiment is $0.3^{\circ}C$, so the thermal expansion is negligible in these experirrents. The feedback or temperature control function should be added for the precise position control actuator.

• Journal of the Korea Society of Computer and Information
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• v.10 no.4 s.36
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• pp.87-92
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• 2005

This study showed the possibility of the medical treatment by thermal feedback as the laser medical treatment had given by design of the digital I/O interfaces of the electronic shutter to control the laser beam and the temperature controlled algorithm. The electronic shutter is economical and that is designed to be automatically controlled within the range of an extent temperature by such development of its driving interfaces and the controlled algorithm of the electronic shutter. The possibility of local therapy for the patients by the treatment of the laser beam within an extent temperature controlled, is proposed by improvement of the problems on the current treatment methods such as radiotherapy, high frequency treatment or medical therapy of drug stuffs which even kill the normal cells.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.13 no.4
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• pp.47-55
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• 1999

In this paper, we construct the controller for the heat exchanger system using iterative method. For awlying the linear quadratic control theory to the heat exchanger system which is represented by the bilinear system, we fomrulate the bilinear system to execute iteration We also propose Extended Kalman Filter to estimate bilinear system state for the purpose of state feedback controller design. We also awly the iterative controller to the thennal power plant superheater system temperature control, and computer simulation show that the estimated value follows the superheater steam temperature under the variation of the external inputs, and that the output steam temperature is properly maintained.tained.

• Proceedings of the Korea Institute of Fire Science and Engineering Conference
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• 1997.11a
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• pp.138-145
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• 1997

Models of upward flame spread have been attempted in the past, but in the current work an emphasis has been placed on developing a practical model that will be useful across a broad range of materials. Some of the important aspects of the model we: the addition of external radiation to simulate a wall that is a part of an enclosure fire and has flaming walls radiating to it, the use of a correlation for flame heat feedback distribution to the sample surface based on data available in the literature, and the use of an experimentally measured mass loss rate for the sample material, In this paper, the development of the numerical model is presented along with predictions of flame spread for three materials: hardboard, a relatively homogeneous wood-based material; plywood, which is made of laminated wood bonded by adhesives; and a composite material made of fiberglass matrix embedded in epoxy. Predictions are compared with measured data at several levels of external radiation for each material. For the materials tested, the model correctly predicts trends and does a reasonable job predicting flame heights. The need for thermal property data for practical materials, which would be appropriate for flame spread models, is indicated by this work.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1195-1208
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• 2019

This paper introduces a new two-step procedure for PWR depletion analyses. This procedure adopts the albedo-corrected parameterized equivalence constants (APEC) method to correct the lattice-based raw cross sections (XSs) and discontinuity factors (DFs) by accounting for neutron leakage. The intrinsic limitations of the conventional two-step methods are discussed by analyzing a 2-dimensional SMR with the commercial DeCART2D/MASTER code system. For a full-scope development of the APEC correction, the MASTER nodal code was modified so that the group constants can be corrected in the middle of a microscopic core depletion. The basic APEC methodology is described and color-set problems are defined to determine the APEC functions for burnup-dependent XS and DF corrections. Then the new two-step method was applied to depletion analyses of the SMR without thermal feedback, and its validity was evaluated in terms of being able to predict accurately the reactor eigenvalue and nodal power profile. In addition, four variants of the original SMR core were also analyzed for a further evaluation of the APEC-assisted depletion. In this work, several combinations of the burnup-dependent and -independent XS and DF corrections were also considered. The results show that the APEC method could enhance the nodal equivalence significantly with inexpensive additional costs.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2670-2689
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• 2022

STREAM - a lattice transport calculation code with method of characteristics for the purpose of light water reactor analysis - has been developed by the Computational Reactor Physics and Experiment laboratory (CORE) of the Ulsan National Institute of Science and Technology (UNIST). Recently, efforts have been taken to develop a photon module in STREAM to assess photon heating and the influence of gamma photon transport on power distributions, as only neutron transport was considered in previous STREAM versions. A multi-group photon library is produced for STREAM based on the ENDF/B-VII.1 library with the use of the library-processing code NJOY. The developed photon solver for the computation of 2D and 3D distributions of photon flux and energy deposition is based on the method of characteristics like the neutron solver. The photon library and photon module produced and implemented for STREAM are verified on VERA pin and assembly problems by comparison with the Monte Carlo code MCS - also developed at UNIST. A short analysis of the impact of photon transport during depletion and thermal hydraulics feedback is presented for a 2D core also from the VERA benchmark.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.922-932
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• 2024

This paper presents the design features and preliminary design analysis results of the Wielenga Innovation Static Salt Reactor (WISSR). The WISSR incorporates features that make it both flexible and inherently safe. It is based on innovative technology that controls a nuclear reactor by moving molten salt fuel into or out of the core. The reactor is a low-pressure, fast spectrum transuranic (TRU) burner reactor. Inherent shutdown is achieved by a large negative reactivity feedback of the liquid fuel and by the expansion of fuel out of the core. The core is made of concentric, thin annular fuel chambers containing molten fuel salt. A molten salt coolant passes between the concentric fuel chambers to cool the core. The core has both fixed and variable volume fuel chambers. Pressure, applied by helium gas to fuel reservoirs below the core, pushes fuel out of a reservoir and up into a set of variable volume chambers. A control system monitors the density and temperature of the fuel throughout the core. Using NaCl-(TRU,U)Cl₃ fuel and NaCl-KCl-MgCl₂ coolant, a road-transportable compact WISSR core design was developed at a power level of 1250 MWt. Preliminary neutronics and thermal-hydraulics analyses demonstrate the technical feasibility of WISSR.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1116-1123
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• 2024

-In recent years, unmanned underwater vehicles (UUV) have been vigorously developed, and with the continuous deepening of marine exploration, traditional energy can no longer meet the energy supply. Nuclear energy can achieve a huge and sustainable energy supply. The heat pipe reactor has no flow system and related auxiliary systems, and the supporting mechanical moving parts are greatly reduced, the noise is relatively small, and the system is simpler and more reliable. It is more favorable for the control of unmanned systems. The use of heat pipe reactors in unmanned underwater vehicles can meet the needs for highly compact, long-life, unmanned, highly reliable, ultra-quiet power supplies. In this paper, a heat pipe reactor scheme named UPR-S that can be applied to unmanned underwater vehicles is designed. The reactor core can provide 1 MW of thermal power, and it can operate at full power for 5 years. UPR-S has negative reactive feedback, it has inherent safety. The temperature and stress of the reactor are within the limits of the material, and the core safety can still be guaranteed when the two heat pipes are failed.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.