• 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.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2803-2811
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• 2020

With the aging of nuclear reactors, embrittlement of the reactor pressure vessel (RPV) steel, as a consequence of routine operations, is highly probable. To ensure operational integrity and safety, prediction and mitigation of compromising damage, brought on by pressurized thermal shock (PTS) following an emergency procedure, is of utmost importance. Computational fluid dynamics (CFD) codes can be employed to predict these events and have therefore been an acceptable method for such assessments. In this paper, CFD simulations of a density driven ECC state in the ROCOM facility are analyzed. Obtained numerical results are validated with the experimental measurements. Considerable attention is attributed to the boundary conditions and their influence, specifically outlet definitions, in order to determine and adequately replicate the non-active pumps in the facility. Consequent analyses focused on initial conditions as well as on the temporal discretization and inner iterations. Disparities due to different turbulent modelling approaches are investigated for standard RANS models. Based on observed trends for different cases, a definitive simulation setup has been established, results of which have been ultimately compared to the measurements.

• 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.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3567-3579
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• 2022

This paper provides an assessment of fluid transport and mixing processes inside the primary circuit of the test facility ROCOM through the numerical simulation of Test 2.1 with the system code ATHLET. The experiment represents an asymmetric injection of cold and non-borated water into the reactor coolant system (RCS) of a pressurized water reactor (PWR) to restore core cooling, an emergency procedure which may subsequently trigger a core re-criticality. The injection takes place at low velocity under single-phase subcooled conditions and presents a major challenge for the simulation in lumped parameter codes, due to multidimensional effects in horizontal piping and vessel arising from density gradients and gravity forces. Aiming at further validating ATHLET 3-D capabilities against horizontal geometries, the experiment conditions are applied to a ROCOM model, which includes a newly developed horizontal pipe object to enhance code prediction inside coolant loops. The obtained results show code strong simulation capabilities to represent multidimensional flows. Enhanced prediction is observed at the vessel inlet compared to traditional 1-D approach, whereas mixing overprediction from the descending denser plume is observed at the upper-half downcomer region, which leads to eventual deviations at the core inlet.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4359-4372
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• 2022

Turbulent mixing in buoyant flows is an essential mechanism involved in many scenarios related to nuclear safety in nuclear power plants. Comprehensive understanding and accurate predictions of turbulent buoyant flows in the reactor are of crucial importance, due to the function of mitigating the potential detrimental consequences during postulated accidents. The present study uses URANS methodology to investigate the buoyancy-influenced flows in the reactor pressure vessel under the main steam line break accident scenarios. With a particular focus on the influence of turbulent heat flux closure models, various combinations of two turbulence models and three turbulent heat flux models are utilized for the numerical simulations of three ROCOM tests which have different characteristic features in terms of the flow rate and fluid density difference between loops. The simulation results are compared with experimental measurements of the so-called mixing scalar in the downcomer and at the core inlet. The study shows that the anisotropic turbulent heat flux models are able to improve the accuracy of the predictions under conditions of strong buoyancy whilst in the weak buoyancy case, a major role is played by the selected turbulence models with essentially a negligible influence of the turbulent heat flux closure models.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.941-951
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• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• 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.

• Journal of computational fluids engineering
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• v.18 no.4
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• pp.53-60
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• 2013

The boron transport model has been implemented into the CUPID code to simulate the boron transport phenomena of the PWR. The boron concentration conservation was confirmed through a simulation of a conceptual boron transport problem in which water with a constant inlet boron concentration injected into an inlet of the 2-dimensional vertical flow tube. The step wise boron transport problem showed that the numerical diffusion of the boron concentration can be reduced by the second order convection scheme. In order to assess the adaptability of the developed boron transport model to the realistic situation, the ROCOM test was simulated by using the CUPID implemented with the boron transportation.

• 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.