A thermal-hydraulic integral effect test facility, Advanced Thermal-Hydraulic Test Loop for Accid... more A thermal-hydraulic integral effect test facility, Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), has been constructed at the Korea Atomic Energy Research Institute. It is a 1/2-reduced-height and 1/288-volume-scaled test facility based on the design features of APR1400, an evolutionary pressurized water reactor developed by the Korean industry. ATLAS was used to perform a set of integral effect tests on the reflood period of a large-break loss-of-coolant accident (LBLOCA) after intensive performance tests had been conducted to verify ATLAS’s operational performance and controllability for major thermal-hydraulic components. The present LB-CL-09 test is one of the integral effect reflood tests for investigating the thermal-hydraulic characteristics during an entire reflood period that can be used to provide reliable data to help validate the LBLOCA analysis methodology for APR1400. The main objective of the present test is to identify the major thermal-hydraulic characteristics such as the direct emergency core coolant (ECC) bypass, downcomer boiling, and core cooling behavior during the reflood phase of an LBLOCA for APR1400 under conditions where the downcomer region interacts with the reactor core region and the heat could be transferred through the steam generator. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results. The decay heat and the ECC flow rate from the safety injection tank were simulated from the start of the reflood period. The ECC flow rate from the safety injection pump was 0.32 kg/s. The system pressure was fixed at ~0.1 MPa, and the initial outer-wall temperature was determined to be 205°C. The experimental results showed the typical thermal-hydraulic trends expected to occur during the reflood phase of the LBLOCA scenario.
18th International Conference on Nuclear Engineering: Volume 4, Parts A and B, 2010
A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS... more A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI (Korea Atomic Energy Research Institute). The reference plant of the ATLAS is a 1400 MWe-class evolutionary pressurized water reactor (PWR), the APR1400 (Advanced Power Reactor 1,400 MWe), which was developed by the Korean industry. The ATLAS has a 1/2 reduced height and a 1/288 volume scaled integral test facility with respect to the APR1400. It has a maximum power capacity of 10% of the scaled nominal core power, and it can simulate full pressure and temperature conditions of the APR1400. The ATLAS could be used to provide experimental data on design-basis accidents including the reflood phase of a large break loss of coolant accident (LBLOCA), small break LOCA (SBLOCA) scenarios including the DVI line and cold leg breaks, a steam generator tube rupture, a main steam line break, a feed line break, etc. An inadvertent opening of POSRV test (SB-POSRV-02) was carried out as one of the SBLOCA spectra. The main objectives of this experimental test were not only to provide physical insight into the system response of the APR1400 reactor during a transient situation but also to present integral effect data for the validation of the SPACE (Safety and Performance Analysis Computer Code), which is now under development by the Korean nuclear industry.
A new instrument, an average BDFT (Birectional Flow Tube), was proposed to measure the flow rate ... more A new instrument, an average BDFT (Birectional Flow Tube), was proposed to measure the flow rate in single and two phase flows. Its working principle is similar to that of the Pilot tube, wherein the dynamic pressure is measured. In an average BDFT, the pressure measured at the front of the flow tube is equal to the total pressure, while that measured at the rear tube is slightly less than the static pressure of the flow field due to the suction effect downstream. The proposed instrument was tested in air/water vertical and horizontal test sections with an inner diameter of 0.08m. The tests were performed primarily in single phase water and air flow conditions to obtain the amplification factor(k) of the flow tube in the vertical and horizontal test sections. Tests were also performed in air/water vertical two phase flow conditions in which the flow regimes were bubbly, slug, and churn turbulent flows. In order to calculate the phasic mass flow rates from the measured differential p...
ABSTRACT The thermal-hydraulic integral effect test (IET) program is being progressed by the Kore... more ABSTRACT The thermal-hydraulic integral effect test (IET) program is being progressed by the Korea Atomic Energy Research Institute. This paper presents an overview of the IET program; the scientific design characteristics of the IET facility; ATLAS, which is under construction; and the experimental and analytical validation works. The ATLAS facility has the following characteristics: (a) a 1/2-height, 1/288-volume, full-pressure simulation of the APR1400, (b) geometrical similarity with the APR1400, including 2 (hot legs) x 4 (cold legs) reactor coolant loops, a direct vessel injection (DVI), an integrated annular downcomer, etc., (c) incorporation of the specific design characteristics of the 1000-MW(electric) class Korean Standard Nuclear Power Plant, such as a cold-leg injection and the low-pressure injection pumps, (d) a maximum 8% of the scaled nominal core power, and (e) simulation capability of broad scenarios, including the reflood phase of the large-break loss-of-coolant accidents (LOCAs), small-break LOCA scenarios including the DVI line breaks, steam generator tube ruptures, main steam line breaks, midloop operation, etc. The scientific design of the ATLAS was accomplished rigorously from the viewpoints of both a global and local scaling based on the three-level scaling methodology of Ishii et al. The validation works showed that the scientific design of the ATLAS test facility is sound.
The ECC penetration duct of the direct vessel injection (DVI) system for the emergency core cooli... more The ECC penetration duct of the direct vessel injection (DVI) system for the emergency core cooling (ECC) system has been studied to prevent steam-water interaction in the downcomer increasing the ECC bypass fraction during the late reflood phase of a large break loss-of-coolant accident (LBLOCA. The ECC penetration duct is installed from the DVI nozzle at the upper downcomer to the lower downcomer. The performance of the extension DVI duct was evaluated for both LBLOCA and DVI line break of the APR1400 using the MARS code, which is a multi-dimensional best-estimate thermal hydraulic code being developed by KAERI. The shape of an advanced extension DVI duct, which has a two exits at both the top and bottom of the duct, was considered to prevent an inlet-to-outlet switching of the extension duct during the DVI line break. A vent hole is considered at the top of the ECC penetration duct to have the role of the break suction point at the upper downcomer. The DVI break flow is initiated...
To understand boiling phenomena and the effect of gap condition on CHF, experiments have been acc... more To understand boiling phenomena and the effect of gap condition on CHF, experiments have been accomplished with various gap sizes and inclination angle at atmospheric pressure using annulus test section. Inclination angle is changed from horizontal to vertical. The annular gap size is varied from 0.5 to 3.5 mm and the channel length is 200 mm. From experimental results, CHF increases with gap size and inclination angle. One-dimensional two-phase model was derived and friction factor correlation is suggested. Correlation for inclined narrow channel is compared with other gap CHF data. 1.Introduction Sometimes the Critical Heat Flux(CHF) means the maximum heat transfer rate without any deformation or degradation of heat transfer materials. For the research of the severe accident in nuclear power plants, it is important to predict the magnitude of CHF. In the TMI-2 accident, a significant part of the RPV lower head was overheated (one part of the RPV wall is estimated to have reached a...
Abstract The direct vessel injection (DVI)-adopted power plant APR1400 considers a DVI line break... more Abstract The direct vessel injection (DVI)-adopted power plant APR1400 considers a DVI line break among the analyzed small-break loss-of-coolant accidents in safety analysis. The first-ever integral effects test database for various DVI line break sizes from 5% to 100% was established with the Korea Atomic Energy Research Institute’s Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) test facility. This database enhances our physical understanding of the major thermal-hydraulic behaviors of the APR1400 during DVI line break accidents, and it can also be used to examine the prediction capabilities and identify any deficiencies in the existing best-estimate safety analysis codes. Effects of the break size were experimentally investigated, and the best-estimated MARS code was assessed against the experimental database. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than the data just before a loop seal clearing occurs, and it also produced a more rapid decrease in the downcomer water level than the data. These disagreements are the expected consequence of uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effects test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology on the DVI line break accidents of the APR1400.
In the present study a new measurement technique has been developed, which uses an ultrasonic tra... more In the present study a new measurement technique has been developed, which uses an ultrasonic transmission signal in order to identify the vertical two phase flow pattern. The ultrasonic measurement system developed in the present study not only provides the information required for the identification of vertical two phase flow patterns but also makes real time identification possible. Various vertical two phase flow patterns such as bubbly, slug, churn, annular flow etc. have been accurately identified with the present ultrasonic measurement system under atmospheric condition. In addition, the present test apparatus can practically simulate the ultrasonic propagation characteristics under high temperature and high pressure systems. Therefore, it is expected that the present ultrasonic flow pattern identification technique could be applicable to the vertical two phase flow systems under high temperature and high pressure conditions.
An experimental study of heat transfer characteristics near the critical pressure has been perfor... more An experimental study of heat transfer characteristics near the critical pressure has been performed with an internally-heated vertical annular channel cooled by R-134a fluid. Two series of tests have been completed: (a) steady-state critical heat flux (CHF) tests, and (b) heat transfer tests for pressure reduction transients through the critical pressure. In the present experimental range, the steady-state CHF decreases with increase of the system pressure for fixed inlet mass flux and subcooling. The CHF falls sharply at about 3.8 MPa and shows a trend towards converging to zero as the pressure approaches the critical point of 4.059 MPa. The CHF phenomenon near the critical pressure does not lead to an abrupt temperature rise of the heated wall, because the CHF occurs at remarkably low power levels. In the pressure reduction transients, as soon as the pressure passes below the critical pressure from the supercritical pressure, the wall temperatures rise rapidly up to very high values due to the departure from nucleate boiling. The wall temperature reaches a maximum at the saturation point of the outlet temperature, and then tends to decrease gradually.
A thermal-hydraulic integral effect test facility, Advanced Thermal-Hydraulic Test Loop for Accid... more A thermal-hydraulic integral effect test facility, Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), has been constructed at the Korea Atomic Energy Research Institute. It is a 1/2-reduced-height and 1/288-volume-scaled test facility based on the design features of APR1400, an evolutionary pressurized water reactor developed by the Korean industry. ATLAS was used to perform a set of integral effect tests on the reflood period of a large-break loss-of-coolant accident (LBLOCA) after intensive performance tests had been conducted to verify ATLAS’s operational performance and controllability for major thermal-hydraulic components. The present LB-CL-09 test is one of the integral effect reflood tests for investigating the thermal-hydraulic characteristics during an entire reflood period that can be used to provide reliable data to help validate the LBLOCA analysis methodology for APR1400. The main objective of the present test is to identify the major thermal-hydraulic characteristics such as the direct emergency core coolant (ECC) bypass, downcomer boiling, and core cooling behavior during the reflood phase of an LBLOCA for APR1400 under conditions where the downcomer region interacts with the reactor core region and the heat could be transferred through the steam generator. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results. The decay heat and the ECC flow rate from the safety injection tank were simulated from the start of the reflood period. The ECC flow rate from the safety injection pump was 0.32 kg/s. The system pressure was fixed at ~0.1 MPa, and the initial outer-wall temperature was determined to be 205°C. The experimental results showed the typical thermal-hydraulic trends expected to occur during the reflood phase of the LBLOCA scenario.
18th International Conference on Nuclear Engineering: Volume 4, Parts A and B, 2010
A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS... more A thermal-hydraulic integral effect test facility for advanced pressurized reactors (PWRs), ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI (Korea Atomic Energy Research Institute). The reference plant of the ATLAS is a 1400 MWe-class evolutionary pressurized water reactor (PWR), the APR1400 (Advanced Power Reactor 1,400 MWe), which was developed by the Korean industry. The ATLAS has a 1/2 reduced height and a 1/288 volume scaled integral test facility with respect to the APR1400. It has a maximum power capacity of 10% of the scaled nominal core power, and it can simulate full pressure and temperature conditions of the APR1400. The ATLAS could be used to provide experimental data on design-basis accidents including the reflood phase of a large break loss of coolant accident (LBLOCA), small break LOCA (SBLOCA) scenarios including the DVI line and cold leg breaks, a steam generator tube rupture, a main steam line break, a feed line break, etc. An inadvertent opening of POSRV test (SB-POSRV-02) was carried out as one of the SBLOCA spectra. The main objectives of this experimental test were not only to provide physical insight into the system response of the APR1400 reactor during a transient situation but also to present integral effect data for the validation of the SPACE (Safety and Performance Analysis Computer Code), which is now under development by the Korean nuclear industry.
A new instrument, an average BDFT (Birectional Flow Tube), was proposed to measure the flow rate ... more A new instrument, an average BDFT (Birectional Flow Tube), was proposed to measure the flow rate in single and two phase flows. Its working principle is similar to that of the Pilot tube, wherein the dynamic pressure is measured. In an average BDFT, the pressure measured at the front of the flow tube is equal to the total pressure, while that measured at the rear tube is slightly less than the static pressure of the flow field due to the suction effect downstream. The proposed instrument was tested in air/water vertical and horizontal test sections with an inner diameter of 0.08m. The tests were performed primarily in single phase water and air flow conditions to obtain the amplification factor(k) of the flow tube in the vertical and horizontal test sections. Tests were also performed in air/water vertical two phase flow conditions in which the flow regimes were bubbly, slug, and churn turbulent flows. In order to calculate the phasic mass flow rates from the measured differential p...
ABSTRACT The thermal-hydraulic integral effect test (IET) program is being progressed by the Kore... more ABSTRACT The thermal-hydraulic integral effect test (IET) program is being progressed by the Korea Atomic Energy Research Institute. This paper presents an overview of the IET program; the scientific design characteristics of the IET facility; ATLAS, which is under construction; and the experimental and analytical validation works. The ATLAS facility has the following characteristics: (a) a 1/2-height, 1/288-volume, full-pressure simulation of the APR1400, (b) geometrical similarity with the APR1400, including 2 (hot legs) x 4 (cold legs) reactor coolant loops, a direct vessel injection (DVI), an integrated annular downcomer, etc., (c) incorporation of the specific design characteristics of the 1000-MW(electric) class Korean Standard Nuclear Power Plant, such as a cold-leg injection and the low-pressure injection pumps, (d) a maximum 8% of the scaled nominal core power, and (e) simulation capability of broad scenarios, including the reflood phase of the large-break loss-of-coolant accidents (LOCAs), small-break LOCA scenarios including the DVI line breaks, steam generator tube ruptures, main steam line breaks, midloop operation, etc. The scientific design of the ATLAS was accomplished rigorously from the viewpoints of both a global and local scaling based on the three-level scaling methodology of Ishii et al. The validation works showed that the scientific design of the ATLAS test facility is sound.
The ECC penetration duct of the direct vessel injection (DVI) system for the emergency core cooli... more The ECC penetration duct of the direct vessel injection (DVI) system for the emergency core cooling (ECC) system has been studied to prevent steam-water interaction in the downcomer increasing the ECC bypass fraction during the late reflood phase of a large break loss-of-coolant accident (LBLOCA. The ECC penetration duct is installed from the DVI nozzle at the upper downcomer to the lower downcomer. The performance of the extension DVI duct was evaluated for both LBLOCA and DVI line break of the APR1400 using the MARS code, which is a multi-dimensional best-estimate thermal hydraulic code being developed by KAERI. The shape of an advanced extension DVI duct, which has a two exits at both the top and bottom of the duct, was considered to prevent an inlet-to-outlet switching of the extension duct during the DVI line break. A vent hole is considered at the top of the ECC penetration duct to have the role of the break suction point at the upper downcomer. The DVI break flow is initiated...
To understand boiling phenomena and the effect of gap condition on CHF, experiments have been acc... more To understand boiling phenomena and the effect of gap condition on CHF, experiments have been accomplished with various gap sizes and inclination angle at atmospheric pressure using annulus test section. Inclination angle is changed from horizontal to vertical. The annular gap size is varied from 0.5 to 3.5 mm and the channel length is 200 mm. From experimental results, CHF increases with gap size and inclination angle. One-dimensional two-phase model was derived and friction factor correlation is suggested. Correlation for inclined narrow channel is compared with other gap CHF data. 1.Introduction Sometimes the Critical Heat Flux(CHF) means the maximum heat transfer rate without any deformation or degradation of heat transfer materials. For the research of the severe accident in nuclear power plants, it is important to predict the magnitude of CHF. In the TMI-2 accident, a significant part of the RPV lower head was overheated (one part of the RPV wall is estimated to have reached a...
Abstract The direct vessel injection (DVI)-adopted power plant APR1400 considers a DVI line break... more Abstract The direct vessel injection (DVI)-adopted power plant APR1400 considers a DVI line break among the analyzed small-break loss-of-coolant accidents in safety analysis. The first-ever integral effects test database for various DVI line break sizes from 5% to 100% was established with the Korea Atomic Energy Research Institute’s Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) test facility. This database enhances our physical understanding of the major thermal-hydraulic behaviors of the APR1400 during DVI line break accidents, and it can also be used to examine the prediction capabilities and identify any deficiencies in the existing best-estimate safety analysis codes. Effects of the break size were experimentally investigated, and the best-estimated MARS code was assessed against the experimental database. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than the data just before a loop seal clearing occurs, and it also produced a more rapid decrease in the downcomer water level than the data. These disagreements are the expected consequence of uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effects test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology on the DVI line break accidents of the APR1400.
In the present study a new measurement technique has been developed, which uses an ultrasonic tra... more In the present study a new measurement technique has been developed, which uses an ultrasonic transmission signal in order to identify the vertical two phase flow pattern. The ultrasonic measurement system developed in the present study not only provides the information required for the identification of vertical two phase flow patterns but also makes real time identification possible. Various vertical two phase flow patterns such as bubbly, slug, churn, annular flow etc. have been accurately identified with the present ultrasonic measurement system under atmospheric condition. In addition, the present test apparatus can practically simulate the ultrasonic propagation characteristics under high temperature and high pressure systems. Therefore, it is expected that the present ultrasonic flow pattern identification technique could be applicable to the vertical two phase flow systems under high temperature and high pressure conditions.
An experimental study of heat transfer characteristics near the critical pressure has been perfor... more An experimental study of heat transfer characteristics near the critical pressure has been performed with an internally-heated vertical annular channel cooled by R-134a fluid. Two series of tests have been completed: (a) steady-state critical heat flux (CHF) tests, and (b) heat transfer tests for pressure reduction transients through the critical pressure. In the present experimental range, the steady-state CHF decreases with increase of the system pressure for fixed inlet mass flux and subcooling. The CHF falls sharply at about 3.8 MPa and shows a trend towards converging to zero as the pressure approaches the critical point of 4.059 MPa. The CHF phenomenon near the critical pressure does not lead to an abrupt temperature rise of the heated wall, because the CHF occurs at remarkably low power levels. In the pressure reduction transients, as soon as the pressure passes below the critical pressure from the supercritical pressure, the wall temperatures rise rapidly up to very high values due to the departure from nucleate boiling. The wall temperature reaches a maximum at the saturation point of the outlet temperature, and then tends to decrease gradually.
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