• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.831-846
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• 2012

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

• Nuclear Engineering and Technology
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• v.41 no.6
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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.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2262-2275
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• 2022

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.209-224
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• 2001

The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3320-3325
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• 2023

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re = 2,409 and Re = 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

• Journal of Energy Engineering
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• v.24 no.2
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• pp.91-102
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• 2015

A thermal-hydraulic design and performance analysis computer code for a once-through steam generator using straight tubes is developed. To benchmark the developed physical models and computer code, an once-through steam generator developed by other designer is simulated and the calculated results are compared with the design data. Also, the same steam generator is analyzed with the best-estimate thermal-hydraulic system code, MARS, for the code-to-code validation. The overall characteristics of heat transfer area, pressure and temperature distributions calculated by the developed code show general agreements with the published design data as well as the analysis results of MARS. It is demonstrated that the developed code can be utilized for diverse purposes, such as, sensitivity analyses and optimum thermal design of a once-through steam generator.

• Proceedings of the Korea Society for Simulation Conference
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• 2004.05a
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• pp.94-100
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• 2004

KEPRI has developed an Nuclear Steam Supply System(NSSS) thermal-hydraulic simulation program (called ARTS-KORIl) based on the best-estimate system code, RETRAN, as a part of the development project for the KORI unit 1 nuclear power plant simulator. To develop the RETRAN code as an NSSS T/H engine for the simulator, a number of code modifications, such as simplifications and removing of discontinuities of the physical correlations, were made to satisfy the simulator requirements of robustness and real time calculation capability Some simplified models and a backup system were also developed to simulate some transients that cannot be efficiently calculated by the RETRAN part of ARTS-KORIl.

• Nuclear Engineering and Technology
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• v.32 no.2
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• pp.169-179
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• 2000

A detailed numerical analysis of the evolution of thermal stratification in a curved piping system in a nuclear power plant is performed. A finite volume based thermal-hydraulic computer code has been developed employing a body-fitted, non-orthogonal curvilinear coordinate for this purpose. The cell-centered, non-staggered grid arrangement is adopted and the resulting checkerboard pressure oscillation is prevented by the application of momentum interpolation method. The SIMPLE algorithm is employed for the pressure and velocity coupling, and the convection terms are approximated by a higher-order bounded scheme. The thermal-hydraulic computer code developed in the present study has been applied to the analysis of thermal stratification in a curved duct and some of the predicted results are compared with the available experimental data. It is shown that the predicted results agree fairly well with the experimental measurements and the transient formation of thermal stratification in a curved duct is also well predicted.

• Nuclear Engineering and Technology
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• v.33 no.6
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• pp.638-651
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• 2001

The present study is to assess the applicability of the best-estimate thermal-hydraulic code, MARS 1.4, for the analysis of thermal-hydraulic behavior in PWRs during natural circulation conditions. The code simulates a natural circulation test, BETHSY test 4. la, which was conducted on the integral test facility of BETHSY. The test represented the cooling states of the primary cooling system under single-phase natural circulation, two-phase natural circulation and the reflux condensation mode with conditions corresponding to the residual power, 2% of the rated core power value and 6.8 MPa at the secondary system. Based on MARS 1.4 calculations, the major thermal-hydraulic behaviors during natural circulation are evaluated and the differences between the experimental data and calculated results are identified. The calculated results show generally good behavior with regard to the experimental results; the region of single-phase natural circulation is 100-92% of the initial mass inventory, two-phase natural circulation is 84-63 %, and the reflux condensation mode occurred below 58 %. U-tubes empty and the core uncovery are obtained at 39 % and 34 % of the initial mass inventory, respectively.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2332-2342
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• 2024

Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.