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LBLOCA AND DVI LINE BREAK TESTS WITH THE ATLAS INTEGRAL FACILITY
백원필,김연식,최기용 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.6
This paper summarizes the tests performed in the ATLAS facility during its first two years of operation (2007~2008). Two categories of tests have been performed successfully: (a) the reflood phase of the large-break loss-of-coolant accidents in a cold leg, and (b) the breaks in one of four direct vessel injection lines. Those tests contributed to understanding the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing an evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Several important and interesting phenomena have been observed during the tests. In most cases, the ATLAS shows reasonable accident characteristics and conservative results compared with those predicted by one-dimensional safety analysis codes. A wide variety of small-break LOCA tests will be performed in 2009.
SAFETY ASSESSMENT OF KOREAN NUCLEAR FACILITIES: CURRENT STATUS AND FUTURE
백원필,양준언,하재주 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.4
This paper introduces the development of safety assessment technology in Korea, focusing on the activities of the Korea Atomic Energy Research Institute in the areas of system thermal hydraulics, severe accidents and probabilistic safety assessment. In the 1970s and 1980s, safety analysis codes and methodologies were introduced from the United States, France, Canada and other developed countries along with technology related to the construction and operation of nuclear power plants. The main focus was on understanding and utilizing computer codes that were sourced from abroad up to the early 1990s, when efforts to develop domestic safety analysis codes and methodologies became active. Remarkable achievements have been made over the last 15 years in the development and application of safety analysis technologies. In addition, significant experimental work has been performed to verify the safety characteristics of reactors and fuels as well as to support the development and validation of analysis methods.
A SUMMARY OF 50th OECD/NEA/CSNI INTERNATIONAL STANDARD PROBLEM EXERCISE (ISP-50)
최기용,백원필,강경호,박현식,조석,김연식 한국원자력학회 2012 Nuclear Engineering and Technology Vol.44 No.6
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.
AN EXPERIMENTAL STUDY ON POST-CHF HEATTRANSFER FOR LOW FLOW OF WATER IN A 3x3 RODBUNDLE
문상기,백원필,조석,SE-YOUNG CHUN,SE-YUN KIM 한국원자력학회 2005 Nuclear Engineering and Technology Vol.37 No.5
An experimental study on post-CHF heat transfer has been performed with a 3x3 rod bundle using a vertical steam-water two-phase flow at low flow conditions. The effects of various parameters on the post-CHF heat transfer are investigated and the reasons for the parametric effects are discussed. As the heat transfer regime changes from CHF to post-CHF, the radial wall temperature distribution is changed depending on the pressure and the mass flux conditions. The superheat of the fluid increases considerably with an increase of the wall temperature (or heat flux) and with a decrease of the mass flux. This implies, indirectly, a strong thermal non-equilibrium at high wall temperature and low mass flux conditions. In order to improve the prediction accuracy of the existing post-CHF correlations, it is necessary to perform more experiments, particularly direct measurement of the vapor superheat, and to modify the correlation by considering a strong thermal non-equilibrium at low flow and low pressure conditions.