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Scaling Analysis for DVI Line Break Accident of APR1400 based on ATLAS Experiment
Abstract
The Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) facility has been established by Korea Atomic Energy Research Institute to conduct integral effect tests for Advanced Power Reactor 1400 (APR1400). ATLAS has been scaled by using the three-level scaling methodology suggested by Ishii et al with the scaling ratio of 1/2 and 1/144 in length and flow area, respectively. Thus the transient in ATLAS occurs 1.414 times faster than that in the prototype, APR1400. In order to address the scalability of ATLAS, a DVI (Direct Vessel Injection) line guillotine break accident at APR1400 has been analyzed by using a system code, MARS-KS. Since the main idea of the analysis is to figure out if the phenomena at ATLAS during the accident are reproduced at APR1400, initial and boundary conditions are taken from the relevant experiment at ATLAS. The final analysis result reveals that plant general behavior, important phenomena and parameters during the accident including break flow, loop seal clearing and peak cladding temperature are well reproduced in the analysis, which indicates the scalability of ATLAS to APR1400 for the DVI line guillotine break accident.
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KAERI has been operating an integral effect test facility, ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), for accident simulations of advanced PWRs. Regarding integral effect tests, a database for major design basis accidents has been accumulated and a Domestic Standard Problem (DSP) exercise using the ATLAS has been proposed and successfully performed. The ATLAS DSP aims at the effective utilization of an integral effect database obtained from the ATLAS, the establishment of a cooperative framework in the domestic nuclear industry, better understanding of thermal hydraulic phenomena, and an investigation of the potential limitations of the existing best-estimate safety analysis codes. For the first ATLAS DSP exercise (DSP-01), integral effect test data for a 100% DVI line break accident of the APR1400 was selected by considering its technical importance and by incorporating comments from participants. Twelve domestic organizations joined in this DSP-01 exercise. Finally, ten of these organizations submitted their calculation results. This ATLAS DSP-01 exercise progressed as an open calculation; the integral effect test data was delivered to the participants prior to the code calculations. The MARS-KS was favored by most participants but the RELAP5/MOD3.3 code was also used by a few participants. This paper presents all the information of the DSP-01 exercise as well as the comparison results between the calculations and the test data. Lessons learned from the first DSP-01 are presented and recommendations for code users as well as for developers are suggested.
The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. 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 did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.
A thermal-hydraulic integral-effect test facility [advanced thermal-hydraulic test loop for accident simulation (ATLAS)] is being constructed at the Korea Atomic Energy Research Institute. The ATLAS is a one-half-reduced-height and 1/288-volume-scaled test facility based on the design features of the APR1400, an evolutionary pressurized water reactor developed by the Korean industry. The simulation capability of the ATLAS for major design-basis accidents (DBAs), including a large-break loss-of-coolant accident and direct vessel injection line-break and main-steam-line-break accidents, is evaluated by the best-estimate system code MARS with the same control logics, transient scenarios, and nodalization scheme. The validity of the applied scaling law and the thermal-hydraulic similarity between the ATLAS and the APR1400 for the major DBAs are assessed. It is confirmed that the ATLAS can maintain an overall similarity with the reference plant APR1400 for the major DBAs considered in the study. However, depending on the accident scenarios, there are some inconsistencies in certain thermal-hydraulic parameters, such as cladding temperature, subcooling at the lower plenum of the core, break flow rate, core and downcomer water level, and secondary pressure. The causes of the inconsistencies are carefully investigated by considering the detailed design features of the ATLAS. It is found that the inconsistencies are mainly due to the reduced power effect and the increased stored energy in the structure. The similarity analysis was successful in obtaining a greater insight into the unique design features of the ATLAS and would be used for developing optimized experimental procedures and control logics.
A vortex valve, called fluidic device, is to be installed inside a Safety Injection Tank (SIT) of Advanced Power Reactor 1400MWe (APR1400) that passively controls an Emergency Core Cooling (ECC) water discharge flow rate without any moving part or any action of the plant operator. The fluidic device was designed, and its performance was evaluated by a series of repetitive experiments using VAlve Performance Evaluation Rig (VAPER), a prototypical full-scale test facility. The passive flow controlling SIT satisfied the major performance requirements of the APR1400 plant design, in view of peak discharge flow rate, pressure loss coefficient, and duration time of the ECC water discharge.
Scaling criteria for a natural circulation loop under single phase and two-phase flow conditions have been derived. For a single phase case the continuity, integral momentum, and energy equations in one-dimensional area average forms have been used. From this, the geometrical similarity groups, friction number, Richardson number, characteristic time constant ratio, Biot number, and heat source number are obtained. The Biot number involves the heat transfer coefficient which may cause some difficulties in simulating the turbulent flow regime. For a two-phase flow case, the similarity groups obtained from a perturbation analysis based on the one-dimensional drift-flux model have been used. The physical significance of the phase change number, subcooling number, drift-flux number, friction number are discussed and conditions imposed by these groups are evaluated. In the two-phase flow case, the critical heat flux is one of the most important transients which should be simulated in a scale model. The above results are applied to the LOFT facility in case of a natural circulation simulation. Some preliminary conclusions on the feasibility of the facility have been obtained.
The onset of downwards penetration of a bubbly water pool above a perforated plate was studied with both air and steam upflows. An interpolative scaling length is developed empirically, which, when introduced into the Wallis countercurrent flow equation, fits the air-water data for a variety of perforatedplate geometries, as well as full-length tube bundle data with saturated water and steam. The same equation, when suitably corrected for steam condensation in the immediate neighborhood of the plate, fits the steam-water data also. Implications for the cooling of overheated nuclear reactor cores are discussed.
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