• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.186-195
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• 1997

When asymmetric thermal-hydraulic conditions occur between cold legs, the core inlet temperature will be nonuniform if the coolant is not mixed perfectly in the lower plenum. These uneven core inlet conditions may induce the change in core power distribution. Thus realistic prediction of thermal mixing is important in such abnormal conditions. In this study, reactor internals, which are scaled down as to conserve the flow area ratio, are set up in the model of KORI Unit 1 with the scaling factor of 1/710 by volume and coolant temperatures are measured beneath the lower core plate. Based on experimental results, the ability of COMMIX-1B code to simulate the coolant mixing phenomena in the lower plenum is estimated. The results show that complete mixing never occurs in any conditions and the mixing pattern is characterized according to the plant type.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.466-475
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• 2017

Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal-hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.23 no.7
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• pp.831-840
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• 1999

In this work, an aerothermodynamic calculation model for cooled axial flow turbine blades with trailing edge ejection is suggested and a mean line performance analysis of a turbine stage with nozzle cooling is carried out. A unique model regarding the interaction between coolant and main gas is proposed, while existing correlations are adopted to predict viscous loss and blade outflow angle. The interactions considered are the heat transfer from main gas to coolant and the temperature and pressure losses by the mixing of two streams due to the trailing edge coolant ejection. For a stator blade without ejection, trailing edge loss calculated by the trailing edge analysis is compared with that calculated by loss correlation. The effect of heat transfer effectiveness of coolant passage on the mixing loss is analyzed. For a model turbine stage with nozzle cooling, parametric analyses are carried out to investigate the effect of main design variables(coolant mass flow ratio, temperature and ejection area) on the stage performance.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.51-54
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• 2002

When a cold HPSI (High Pressure Safety Injection) fluid associated with an design basis accident, such as LOCA (Loss of Coolant Accident), enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters a reactor pressure vessel downcomer, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. Previous thermal-mixing analyses have assumed that the thermal stratification phenomena generated in stagnated loop of a partially stagnated coolant loop are neutralized in the vessel downcomer by strong flow from unstagnated loop. On the basis of these reasons, this paper presents the thermal-mixing analysis results in order to identify the fact that the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is affected by the strong flow of the unstagnated loop.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.95-103
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• 2019

Vapor explosions can be classified in terms of modes of contact between the hot molten fuel and the coolant, since different contact modes may affect fuel-coolant mixing and subsequent vapor explosion energetics. It is generally accepted that most vapor explosion phenomena fall into three different modes of contact; fuel pouring into coolant, coolant injection into fuel and stratified fuel-coolant layers. In this study, we review previous stratified steam explosion experiments as well as recent experiments performed at the KTH in Sweden. While experiments with prototypic reactor materials are minimal, we do note that generally the energetics is limited for the stratified mode of contact. When the fuel mass involved in a steam explosion in a stratified geometry is compared to a pool geometry based on geometrical aspects, one can conclude that there is a very limited set of conditions (when melt jet diameter is small) under which a steam explosion is more energetic in a stratified geometry. However, under these limited conditions the absolute energetic explosion output would still be small because the total fuel mass involved would be limited.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.327-336
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• 2007

The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields. This paper presents a ROCOM experiment in which water with higher density was injected into a cold leg of the reactor model. Wire-mesh sensors measuring the tracer concentration were installed in the cold leg and upper and lower part of the downcomer. The experiment was run with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water especially for the validation of the Computational Fluid Dynamics (CFD) software ANSYS CFX. A mesh with two million control volumes was used for the calculations. The effects of turbulence on the mean flow were modelled with a Reynolds stress turbulence model. The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

• Proceedings of the KSME Conference
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• 2001.06e
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• pp.482-487
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• 2001

A CFD study was conducted to evaluate the nuclear fuel assembly coolant mixing that is promoted by the flow-mixing vanes on the grid spacer. Four mixing vanes (split vane, swirl vane, twisted vane, hybrid vane) were chosen in this study. A single subchannel of one grid span is modeled using the flow symmetry. The three mixing vanes other than swirl vane generate a large crossflow between the subchannels and a skewed elliptic swirling flow in the subchannel near the grid spacer. The swirl vane induces a circular swirling flow in the subchannel and a negligible crossflow. The split vane and the twisted vane were predicted to result in relatively larger pressure drop across the grid spacer. Since the average turbulent kinetic energy in the subchannel rapidly decreases to a fully developed level downstream of the spacer, turbulent mixing caused by the mixing vanes appears to be not as effective as swirling flow mixing in the subchannel. In summary, the CFD analysis represented the overall characteristics of coolant mixing well in a nuclear fuel assembly with the flow mixing vanes on the grid spacer. The CFD study is therefore quite useful for the development of an advanced flow-mixing vane.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.71-81
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• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3182-3195
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• 2021

The implementation and validation of multi-dimensional (multi-D) features in thermal-hydraulic system codes aims to extend the application of these codes towards multi-scale simulations. The main goal is the simulation of large-scale three-dimensional effects inside large volumes such as piping or vessel. This novel approach becomes especially relevant during the simulation of accidents with strongly asymmetric flow conditions entailing density gradients. Under such conditions, coolant mixing is a key phenomenon on the eventual variation of the coolant temperature and/or boron concentration at the core inlet and on the extent of a local re-criticality based on the reactivity feedback effects. This approach presents several advantages compared to CFD calculations, mainly concerning the model size and computational efforts. However, the range of applicability and accuracy of the newly implemented physical models at this point is still limited and needs to be further extended. This paper aims at contributing to the validation of the multi-D features of the system code ATHLET based on the simulation of the Tests 1.1 and 2.1, conducted at the test facility ROCOM. Overall, the multi-D features of ATHLET predict reasonably well the evolution from both experiments, despite an observed overprediction of coolant mixing at the vessel during both experiments.