• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.775-784
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• 2009

This paper summarizes the tests performed in the ATLAS facility during its first two years of operation (2007${\sim}$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.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.655-676
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• 2009

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.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.9
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• pp.643-657
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• 2009

This paper introduces the current nuclear experimental research activities in KAERI, the unique nuclear research institute in Korea, and the universities in Korea to solve and assess the issues which have been faced in the design of new reactors such as APR1400, SMART, GEN-IV reactors as well as fusion reactor. Also the experimental evaluations of safety for operating nuclear plants have been presented. The nuclear thermalhydraulic experiments performed in such organizations are classified the fundamental test, the separated effect test, and the integral effect test with ATLAS and SNUF. Introduction is deployed according to institutes. Finally, the future works and the direction of research voyage in the nuclear thermal-hydraulic field were suggested.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.25-44
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• 2011

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.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1504-1513
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• 2019

Recently, ocean nuclear reactors have received attention due to enhanced safety features. The movable and transportable characteristics distinguish ocean nuclear reactors from land-based nuclear reactors. Therefore, for safety/design analysis of the ocean reactor, the thermos-hydraulics must be investigated in the moving system. However, there are no studies reporting the general two-fluid equations that can be used for multi-dimensional simulations of two-phase flows in moving systems. This study is to systematically formulate the multi-dimensional two-fluid equations in the non-inertial frame of reference. To demonstrate the applicability of the formulated equations, we perform a total of six different simulations in 2D tanks with translational and/or rotational motions.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.855-870
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• 2012

The paper deals with three main topics: a) the definition of System Thermal-Hydraulics (SYS TH), b) a historical outline for SYS TH and, c) the description of elements for reflection when planning research projects or improvement activities, this last topic being the main reason for the paper. Distinctions between basic thermal-hydraulics and computational Fluid-Dynamics (CFD) on the one side and SYS TH on the other side are considered under the first topic; stakeholders in the technology are identified. The proposal of Interim Acceptance Criteria for Emergency Core Cooling Systems in 1971 by US NRC (AEC at the time) is recognized as the starting date or the triggering event for SYS TH (second topic). The complex codes and the main experimental programs (list provided in the paper) constitute the pillars for SYS TH. Caution or warning statements are introduced in advance when discussing the third topic: a single person (or a researcher) has little to no possibility, or capability, of streamlining the forthcoming investments or to propose a roadmap for future activities. Nevertheless, the ambitious attempt to foresee developments in this area has been pursued without constraints connected with the availability of funds and with industrial benefits or interests. Demonstrating the acceptability of current SYS TH limitations and training in the application of those codes are mentioned as the main challenges for forthcoming research activities.

• Journal of computational fluids engineering
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• v.21 no.3
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• pp.89-97
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• 2016

In this study, direct code coupling, in which two codes share a single flow field, was conducted using 3-dimensional high resolution thermal hydraulics code, CUPID and 1-dimensional system analysis code, MARS. This approach provide the merit to use versatile capability of MARS for nuclear power plants and 3-dimensional T/H analysis capability of CUPID. Numerical Method to directly couple CUPID and MARS was described in this paper. The straight flow and manometer flow oscillation were calculated to verify conservation of coupled CUPID/MARS code in mass, momentum, and energy. This verification calculations indicates that the CUPID/MARS is coupled appropriately in numerical aspect and the coupled code can be applied to nuclear reactor thermal hydraulics after validation against integral transient experiments.

• Proceedings of the Korea Society for Simulation Conference
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• 2004.05a
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• pp.101-108
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• 2004

KEPRI has developed an Nuclear Steam Supply System(NSSS) thermal-hydraulics simulation program (called ARTS-KORI), based on the best-estimate system code, RETRAN, as a part of the development project for the KORI unit 1 Nuclear Power Plant Simulator. A number of code modifications, such as simplifications and removing of discontinuities of the physical correlations, were made in order to change the RETRAN code as an nuclear Steam Supply System thermal-hydraulics engine in the simulator. Some simplified models and a backup system were also developed. This paper briefly presents the results of non-integrated standalone test of ARTS-KORI.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.257-270
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• 2011

A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.481-488
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• 2014

An analysis of the uncertainty quantification related to LBLOCA using the Monte-Carlo calculation has been performed and compared with the tolerance level determined by the Wilks' formula. The uncertainty range and distribution of each input parameter associated with the LOCA phenomena were determined based on previous PIRT results and documentation during the BEMUSE project. Calulations were conducted on 3,500 cases within a 2-week CPU time on a 14-PC cluster system. The Monte-Carlo exercise shows that the 95% upper limit PCT value can be obtained well, with a 95% confidence level using the Wilks' formula, although we have to endure a 5% risk of PCT under-prediction. The results also show that the statistical fluctuation of the limit value using Wilks' first-order is as large as the uncertainty value itself. It is therefore desirable to increase the order of the Wilks' formula to be higher than the second-order to estimate the reliable safety margin of the design features. It is also shown that, with its ever increasing computational capability, the Monte-Carlo method is accessible for a nuclear power plant safety analysis within a realistic time frame.