• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1776-1783
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• 2019

To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.

• 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.56 no.7
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• pp.2524-2533
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• 2024

When analyzing containment thermal-hydraulics, computational fluid dynamics (CFD) is a powerful tool because multi-dimensional and local analysis is required for some accident scenarios. According to the previous study, neglecting steam bulk condensation in the CFD analysis leads to a significant error in boundary layer profiles. Validating the condensation model requires the experimental data near the condensing surface, however, available boundary layer data is quite limited. It is also important to confirm whether the heat and mass transfer analogy (HMTA) is still valid in the presence of bulk condensation. In this study, the boundary layer measurements on the vertical condensing surface in the presence of air were performed with the rectangular channel facility WINCS, which was designed to measure the velocity, temperature, and concentration boundary layers. We set the laminar flow condition and varied the Richardson number (1.0-23) and the steam volume fraction (0.35-0.57). The experimental results were used to validate CFD analysis and HMTA models. For the former, we implemented a bulk condensation model assuming local thermal equilibrium into the CFD code and confirmed its validity. For the latter, we validated the HMTA-based correlations, confirming that the mixed convection correlation reasonably predicted the sum of wall and bulk condensation rates.

• Journal of the Korean Society for Geothermal and Hydrothermal Energy
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• v.1 no.1
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• pp.1-8
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• 2005

Aquifer Thermal Energy Storage (ATES) can be a cost-effective and renewable geothermal energy source, depending on site-specific and thermohydraulic conditions. To design an effective ATES system having the effect of groundwater movement, understanding of thermohydraulic processes is necessary. The heat transfer phenomena for an aquifer heat storage are simulated by using FEFLOW with the scenario of heat pump operation with pumping and waste water reinjection in a two layered confined aquifer model. Temperature distribution of the aquifer model is generated, and hydraulic heads and temperature variations are monitored at the both wells during 365 days. The average groundwater velocities are determined with two hydraulic gradient sets according to boundary conditions, and the effect of groundwater flow are shown at the generated thermal distributions of three different depth slices. The generated temperature contour lines at the hydraulic gradient of 0.001 are shaped circular, and the center is moved less than 5 m to the direction of groundwater flow in 365 days simulation period. However at the hydraulic gradient of 0.01, the contour center of the temperature are moved to the end of east boundary at each slice and the largest movement is at bottom slice. By the analysis of thermal interference data between two wells the efficiency of the heat pump system model is validated, and the variation of heads is monitored at injection, pumping and no operation mode.

• 한국신재생에너지학회:학술대회논문집
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• 2005.06a
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• pp.701-704
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• 2005

Aquifer Thermal Energy Storage (ATES) can be a cost-effective and renewable geothermal energy source, depending on site-specific and thermohydraulic conditions. To design an effective ATES system having influenced by groundwater movement, understanding of thermo hydraulic processes is necessary. The heat transfer phenomena for an aquifer heat storage are simulated using FEFLOW with the scenario of heat pump operation with pumping and waste water reinjection in a two layered confined aquifer model. Temperature distribution of the aquifer model is generated, and hydraulic heads and temperature variations are monitored at the both wells during 365 days. The average groundwater velocities are determined with two hydraulic gradient sets according to boundary conditions, and the effect of groundwater flow are shown at the generated thermal distributions of three different depth slices. The generated temperature contour lines at the hydraulic gradient of 0.00 1 are shaped circular, and the center is moved less than 5m to the groundwater flow direction in 365 days simulation period. However at the hydraulic gradient of 0.01, the contour center of the temperature are moved to the end of boundary at each slice and the largest movement is at bottom slice. By the analysis of thermal interference data between two wells the efficiency of the heat pump system model is validated, and the variation of heads is monitored at injection, pumping and no operation mode.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1603-1610
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• 2020

Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

• 한국신재생에너지학회:학술대회논문집
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• 2007.11a
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• pp.489-492
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• 2007

Understanding the thermohydraulic processes in the aquifer is necessary for a proper design of the aquifer thermal energy utilization system under given conditions. Experimental and numerical test were accomplished to evaluate the relationship between the geothermal heat exchanger operation and hydrogeological conditions in the open-loop geothermal system. Sand tank experiments were designed to investigate the open-loop geothermal system. Water injection and extract ion system as open-loop borehole heat exchanger was applied to observe the temperature changes in time at injection well, extraction well and ambient groundwater. The thermohydraulic transfer for heat storage was simulated using FEFLOW for two cases of extraction and injection phase operation in sand tank model. As one case, the movement of the thermal plume was simulated with variable locations of injection and extraction well. As another case, the simulation was performed with fixed location of injection and extraction well. The simulation and experimental results showed that the temperature distribution depends highly on the injected water temperature and the length of injection time and the groundwater flow and pumping rate sensitively affect the heat transfer.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.765-774
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• 2009

For over 30 years, investigations of the thermohydraulic behavior of pressurized-water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany. The PKL facility models the entire primary side and significant parts of the secondary side of a of pressurized water reactor at a height scale of 1:1. Volumes, power ratings and mass flows are scaled with a ratio of 1:145. The experimental facility consists of four primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermohydraulic phenomena. The PKL tests began in the mid 1970s with the support of the German Research Ministry. Since the mid 1980s, the project has also been significantly supported by the German PWR operators. Since 2001, 25 partner organizations from 15 countries have taken part in the PKL investigations with the support and mediation of the OECD/ NEA (Nuclear Energy Agency). After an overview of PKL history and a short description of the facility, this paper focuses on the investigations carried out since the beginning of the international cooperation, and shows, by means of some examples, what insights can be derived from the tests.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.793-803
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• 2021

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1742-1756
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• 2023

The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a Large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high-temperature flow of approximately 390 ℃ was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high-temperature region. The phenomenological discussion in this paper helps understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.