• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.34 no.7
• /
• pp.935-941
• /
• 2010

HANARO is an open-tank-in-pool-type Korean research reactor that generates 30MW of thermal power. It differs from power plant reactor in that the heat generated by HANARO is exhausted into the atmosphere through a secondary cooling tower, thus maintaining the core temperature constant. During every monthly inspection of the cooling tower, large vibrations that exceeded the permissible limit were observed at cooling fan gear reducer No. 4 of the cooling tower. The purpose of this study is to identify the origin of the large vibration and to repair it. FFT spectrum analysis is performed to identify the part that caused the large vibration. The results of the frequency analysis showed that the vibration frequency was 354 Hz, which is twice the natural frequency of the pinion gear. A check of the pinion gear revealed that there was a crack on the surface of the pinion gear. After the gear was replaced, the reducer operated normally.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.3085-3094
• /
• 2022

The design concept of a Small High-flux Multipurpose LBE(Lead Bismuth Eutectic) cooled Fast Reactor (SHMLFR) was proposed in the paper. The primary cooling system of the reactor is forced circulation, and the fuel element form is arc-plate loaded high enrichment MOX fuel. The core is cylindrical with a flux trap set in the center of the core, which can be used as an irradiation channel. According to the requirements of the core physical design, a series of physical design criteria and constraints were given, and the steady and transient parameters of the reactor were calculated and analyzed. Regarding the thermal and hydraulic phenomena of the reactor, a simplified model was used to conduct a preliminary analysis of the fuel plates at special positions, and the temperature field distribution of the fuel plate with the highest power density under different coolant flow rates was simulated. The results show that the various parameters of SHMLFR meet the requirements and design criteria of the physical design of the core and the thermal design of the reactor. This implies that the conceptual design of SHMLFR is feasible.

• Proceedings of the KSME Conference
• /
• 2001.11b
• /
• pp.735-740
• /
• 2001

In the nuclear power plant, emergency core coolant system(ECCS) is furnished at reactor coolant system(RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, it occurs thermal stratification phenomena in case that there is the mixing of cooling water and high temperature water due to valve leakage in ECCS. This thermal stratification phenomena raises excessive thermal stresses at pipe wall. Therefore, this phenomena causes the accident that reactor coolant flows in reactor containment in the nuclear power plant due to the deformation of pipe and thermal fatigue crack(TFC) at the pipe wall around the place that it exists. Hence, in order to fundamental identification of this phenomena, it requires the experimental research of modeling test in the pipe flow that occurs thermal stratification phenomena. So, this paper models RCS and ECCS pipe arrangement and analyzes the mechanism of thermal stratification phenomena by measuring of temperature in variance with leakage flow rate in ECCS modeled pipe and Reynold number in RCS modeled pipe. Besides, results of this experiment is compared with computational analysis which is done in advance.

• Nuclear Engineering and Technology
• /
• v.41 no.3
• /
• pp.307-318
• /
• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

• Nuclear Engineering and Technology
• /
• v.50 no.8
• /
• pp.1412-1420
• /
• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.12 no.10
• /
• pp.4288-4295
• /
• 2011

In this paper, variable reactor and capacitor for high-frequency welder are analyzed by optimum design. As the polar panel of high-frequency welder has the role of condenser, the material with the high rate of induced electricity has to be selected in order to manufacture the condenser with the great power cut. And the area of polar panel must be large and the gap between panels must be thin. On the contrary, the resistance is generated and the heat is happened because the large current is flown. To prevent the thermal deformation of this polar panel, the temperature can be lowered by using cooling water and so on. At this point, the speed of cooling water due to deformation and temperature of polar panel can be optimized.

• Proceedings of the Korea Concrete Institute Conference
• /
• 1999.10a
• /
• pp.333-336
• /
• 1999

The crack of concrete induced by the heat of hydration is a serious problem, particularly in concrete structures such as mat-slab of nuclear reactor buildings, dams or large footings, foundations of high rise buildings, etc.. As a result of the temperature rise and restriction condition of foundation, the thermal stress which may induce the cracks can occur. Therefore the various techniques of the thermal stress control in massive concrete have been widely used. One of them is prediction of the thermal stress, besides low-heat cement which mitigates the temperature rise, pre-cooling which lowers the initial temperature of fresh concrete with ice flake, pipe cooling which cools the temperature of concrete with flowing water, design change which considers steel bar reinforcement, operation control and so on. The Aim of this paper is to verify the effect of low heat blended cement in reducing thermal stress in slab type's mass concrete such as container harbor structures.

• Nuclear Engineering and Technology
• /
• v.53 no.5
• /
• pp.1429-1435
• /
• 2021

The temperature distribution of an electromagnetic pump was analyzed with a flow rate of 1380 L/min and a pressure of 4 bar designed for the sodium thermo-hydraulic test in the Sodium Test Loop for Safety Simulation and Assessment-Phase 1 (STELLA-1). The electromagnetic pump was used for the circulation of the liquid sodium coolant in the Intermediate Heat Transport System (IHTS) of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) with an electric power of 150 MWe. The temperature distribution of the components of the electromagnetic pump was numerically analyzed to prevent functional degradation in the high temperature environment during pump operation. The heat transfer was numerically calculated using ANSYS Fluent for prediction of the temperature distribution in the excited coils, the electromagnet core, and the liquid sodium flow channel of the electromagnetic pump. The temperature distribution of operating electromagnetic pump was compared with cooling of natural and forced air circulation. The temperature in the coil, the core and the flow gap in the two conditions, natural circulation and forced circulation, were compared. The electromagnetic pump with cooling of forced circulation had better efficiency than natural circulation even considering consumption of the input power for the air blower. Accordingly, this study judged that forced cooling is good for both maintenance and efficiency of the electromagnetic pump.

• Progress in Superconductivity and Cryogenics
• /
• v.5 no.2
• /
• pp.36-40
• /
• 2003

Inductive superconducting fault current limiter(SFCL) with DC reator rated on 6.6k $V_{rms}$/200 $A_{rms}$ has been developed in Yonsei University. The development of the DC reactor is the key technology in this type SFCL. This paper deals with the fabrication and characteristic test of the DC reactor. For the development of this magnet, the winding machine for high-Tc superconducting solenoid was manufactured. Using this machine, a large-scale HTS solenoid using Bi-2223 tape was fabricated successfully. This coil has 5 layers which are connected in series each other. The inductance of the DC reactor coil is 84mB. The cooling system was the sub-cooled nitrogen whose temperature is about 65K. The characteristic test of the coil was performed. The full quench current of this coil is about 490A.90A.

• Proceedings of the KSME Conference
• /
• 2001.06a
• /
• pp.243-249
• /
• 2001

In this study, the progressive inelastic deformation, so called, thermal ratchet phenomenon which can occur in high temperature liquid metal reactor was simulated with thermal ratchet structural test facility and 316L stainless steel test cylinder. The inelastic deformation of the reactor baffle cylinder can occur due to the moving temperature distribution along the axial direction as the hot free surface moves up and down under the cyclic heat-up and cool-down of reactor operations. The ratchet deformations were measured with the laser displacement sensor and LVDTs after cooling the structural specimen which experiences thermal load up to $550^{\circ}$ and the temperature differences of about $500^{\circ}C$. During structural thermal ratchet test, the temperature distribution of the test cylinder along the axial direction was measured from 28 channels of thermocouples and the temperatures were used for the ratchet analysis. The thermal ratchet deformation analysis was performed with the NONSTA code whose constitutive model is nonlinear combined kinematic and isotropic hardening model and the test results were compared with those of the analysis. Thermal ratchet test was carried out with respect to 9 cycles of thermal loading and the maximum residual displacements were measured to be 1.8mm. It was shown that thermal ratchet load can cause a progressive deformation to the reactor structure. The analysis results with the combined hardening model were in reasonable agreement with those of the tests.