• Current Optics and Photonics
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• 제1권5호
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• pp.500-504
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• 2017

In this paper, we developed a temperature measurement system based on an infrared temperature imaging module for thermal safety verification of a plasma skin treatment device (PSTD). We tested a pilot product of the low-temperature PSTD using the system, and the temperature increase of each plasma torch was well-monitored in real-time. Additionally, through the approximation of the temperature increase of the plasma torches, a certain limitation of the plasma treatment time on skin was established with the International Electrotechnical Commission (IEC) guideline. We determined an appropriate plasma treatment time ($T_{Safe}$ < 24 minutes) using the configured temperature measurement system. We believe that the temperature measurement system has a potential to be employed for testing thermal safety and suitability of various medical devices and industrial instruments.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.185-194
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• 2006

Transient operations for an integral type reactor, SMART-P, have been experimentally investigated using a thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), in order to verify the system design and performance of the SMART-P, a pilot plant of SMART. The VISTA facility was subjected to various accident conditions such as feedwater increase and decrease, loss of coolant flow, and control rod withdrawal accidents in order to elucidate the thermal-hydraulic responses following such accidents and finally to verify the system design of the SMARTP. Full functional control logics have been implemented in the VISTA facility in order to control the required control action for an accident simulation. As one of the sensitivity tests to verify the PRHRS performance, the effects of the initial water level in the compensation tank are experimentally investigated. When the initial water level is 16%, the water is quickly drained and nitrogen gas is then introduced into the PRHR system, resulting in deterioration of the PRHRS performance. It is thus found that nitrogen ingression should be prevented to ensure stable PRHRS operation.

• Journal of Radiation Protection and Research
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• 제28권4호
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• pp.281-290
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• 2003

The KN-12 spent nuclear fuel transport cask, which is a Type B(U) package designed to comply with the requirements of Korea Atomic Energy Act[1], IAEA Safety Standards Series No.TS-R-1[2] and US 10 CFR Part 71[3], is designed for carrying up to 12 PWR spent fuel assemblies in a basket structure. The cask has been licensed in accordance with Korea Atomic Energy Act and was fabricated in Korea in accordance with the requirements of ASME B&PV Sec.III, Div.3[4]. The cask must maintain thermal integrity in accordance with the related regulations and be evaluated to verify that the thermal performance of the cask complies with the regulatory requirements. The temperatures of the cask and components were determined by using finite elements methods with a numerical tool, safety tests using an 1/8 height slice model of the real cask were conducted to demonstrate verification of the numerical tool and methods, and heat transfer tests for normal transport conditions were performed as a fabrication acceptance test to demonstrate the heat transfer capability of the cask.

• International Journal of Aeronautical and Space Sciences
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• 제17권4호
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• pp.526-534
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• 2016

A MEMS solid propellant thruster array shall be operated within an allowable range of operating temperatures to avoid ignition failure by incomplete combustion due to a time delay in ignition. The structural safety of the MEMS thruster array under severe on-orbit thermal conditions can also be guaranteed by a suitable thermal control. In this study, we propose a thermal control strategy to perform on-orbit verification of a MEMS thruster module, which is expected to be the primary payload of the STEP Cube Lab mission. The strategy involves, the use of micro-igniters as heaters and temperature sensors for active thermal control because an additional heater cannot be implemented in the current design. In addition, we made efforts to reduce the launch loads transmitted to the MEMS thruster module at the system level structural design. The effectiveness of the proposed thermo-mechanical design strategy has been demonstrated by numerical analysis.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1231-1240
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• 2019

Unlike land-based nuclear power plants, a marine or floating reactor is affected by external forces due to ocean conditions. These external forces can cause additional accelerations and affect each system and equipment of the marine reactor. Therefore, in designing a marine reactor and evaluating its performance and stability, a thermal hydraulic safety analysis code is necessary to consider the thermal hydrodynamic effects of ship motion. MARS, which is a reactor system analysis code, includes a dynamic motion model that can simulate the thermal-hydraulic phenomena under three-dimensional motion by calculating the body force term included in the momentum equation. In this study, it was verified that the dynamic motion model can simulate fluid motion with reasonable accuracy using conceptual problems. In addition, two modifications were made to the dynamic motion model; first, a user-supplied table to simulate a realistic ship motion was implemented, and second, the flow regime map determination algorithm was improved by calculating the volume inclination information at every time step if the dynamic motion model was activated. With these modifications, MARS could simulate the thermal-hydraulic phenomena under ocean motion more realistically.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2021년도 가을 학술논문 발표대회
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• pp.192-193
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• 2021

Recently, the need to strengthen the Air Environment Conservation Act and secure alternative heat sources during the winter by carbon neutrality policies has been raised. Accordingly, winter construction, which has safety and quality measures, is emerging as an essential factor. It is believed that eco-friendly tropical type electric hot air heaters will be able to solve most of the problems of winter construction at construction sites, especially prevention of suffocation and fire accidents. In addition, as a result of on-site performance verification, it has secured more than the same performance as the existing curing method, and the curing technology can create an eco-friendly and pleasant working environment while considering safety and construction.

• 국제초고층학회논문집
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• 제2권1호
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• pp.23-30
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• 2013

This paper explains the Japanese present situations relevant to the fire resistance performance. Performance-based fire provisions was introduced in 1998 for the first time when the Building Standard Law was amended. However, performance-based fire resistance design had been used since long before the official introduction of performance-based provisions. A Comprehensive Technology Development Project of Ministry of Construction from 1982 to 1986 established a technical basis for performance-based fire safety engineering in Japan. A system of calculation methods for fire resistance verification was prescribed in the Ministry Notification in 2000 utilizing the results of this project as a background. This method, referred to as the Fire Resistance Verification Method (FRVM), is the standard method to verify the fire resistance performance of principal building parts such as columns, beams, and walls of steel, concrete, or wood structured buildings. For tall buildings, however, more advanced method for performance verification is often necessary because new building materials or structural systems are often used for these buildings. An example project of tall building owned by a major newspaper company is presented in this paper. Advanced thermal deformation analysis is executed to secure the fire resistance of the building.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.2053-2058
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• 2004

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant.