• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.596-597
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• 2005

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.379-380
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.387-388
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.329-330
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• 2004

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.561-586
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• 2012

This paper describes a summary of final prediction results by system-scale safety analysis codes during the OECD/NEA/CSNI ISP-50 exercise, targeting a 50% Direct Vessel Injection (DVI) line break integral effect test performed with the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS). This ISP-50 exercise has been performed in two consecutive phases: "blind" and "open" phases. Quantitative comparisons were performed using the Fast Fourier Transform Based Method (FFTBM) to compare the overall accuracy of the collected calculations. Great user effects resulting from the combination of the possible reasons were found in the blind phase, confirming that user effect is still one of the major issues in connection with the system thermal-hydraulic code application. Open calculations showed better prediction accuracy than the blind calculations in terms of average amplitude (AA) value. A total of nineteen organizations from eleven countries participated in this ISP-50 program and eight leading thermal-hydraulic system analysis codes were used: APROS, ATHLET, CATHARE, KORSAR, MARS-KS, RELAP5/MOD3.3, TECH-M-97, and TRACE.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.753-764
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• 2009

JAEA started the LSTF experiments in 1985 for the fourth stage of the ROSA Program (ROSA-IV) for the LWR thermal-hydraulic safety research to identify and investigate the thermal-hydraulic phenomena and to confirm the effectiveness of ECCS during small-break LOCAs and operational transients. The LSTF experiments are underway for the ROSA-V Program and the OECD/NEA ROSA Project that intends to resolve issues in thermal-hydraulic analyses relevant to LWR safety. Six types of the LSTF experiments have been done for both the system integral and separate-effect experiments among international members from 14 countries. Results of four experiments for the ROSA Project are briefly presented with analysis by a best-estimate (BE) code and a computational fluid dynamics (CFD) code to illustrate the capability of the LSTF and codes to simulate the thermal-hydraulic phenomena that may appear during SBLOCAs and transients. The thermal-hydraulic phenomena dealt with are coolant mixing and temperature stratification, water hammer up to high system pressure, natural circulation under high core power condition, and non-condensable gas effect during asymmetric SG depressurization as an AM action.

• 한국안전학회지
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• 제22권4호
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• pp.43-50
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• 2007

Lack of understanding of the process chemistry and thermodynamics are the major reasons that can is lead to thermal runaway reaction in the chemical reaction process. The evaluation of reaction factors and thermal behavior in neutralization process of pigment plant are described in this paper. The experiments were performed in the C 80 calorimeter, and Thermal Screening Unit($TS^{u}$). The aim of the study was to evaluate the results of thermal stability in terms of safety reliability to be practical applications. It suggested that we be proposed safe operating conditions and securities for accident prevention through this study.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.871-894
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• 2013

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This $2^{nd}$ ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

• 한국안전학회지
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• 제26권2호
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• pp.26-31
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• 2011

In this paper, the safety diagnosis using thermal image analysis is described for power equipments. The conventional three-phase comparison method has only provided the results of thermal comparison for the equipments. The proposed method defines the conditions of poor connection by visual checks, and supports the criteria with each thermal rise step. As a result, the thermal difference from $5^{\circ}C$ to $10^{\circ}C$ meant the warning state. In addition, the thermal difference more than $10^{\circ}C$ meant that the connection status was unbalanced. In this case, the countermeasure might be the internal load distribution. If the thermal difference more than $20^{\circ}C$ is observed, it means a hot spot at the poor connection. If the hot spot is observed all over the surface, its cause was the unbalanced load, which made the conductive parts discolored and raised the possibility of oxidization or $Cu_2O$ generation. This diagnostic technology employing thermal image analysis method can be directly applied in the field and ensures the safety of equipments.

• 한국유체기계학회 논문집
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• 제10권5호
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• pp.20-26
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• 2007

Some gate valves are susceptible to pressure locking and thermal binding which prevent the safety function. The safety related gate valves susceptible to pressure locking and thermal binding shall be identified and taken preventive actions to ensure the safety function. The identification of the gate valves susceptible to pressure locking and thermal binding needs the evaluation of system design, valve and piping arrangement, test requirements, and operating conditions. Application of preventive methods should consider the system safety function, applicability, effectiveness, interface with system design, and cost. The selection procedure of valves susceptible to pressure locking and thermal binding can be effectively used in industry including nuclear power plants. In order to prevent the pressure locking, the hole can be drilled through the one disc of upstream side or down stream and the external equalizing line can be installed from bonnet to downstream or upstream. The double disc parallel seat valve type can be used instead of flexible wedge gate valve to prevent the thermal binding. The identification of gate valves susceptible to pressure locking and thermal binding, and preventive actions will meet the regulatory requirements and enhance the availability and safety of plants.