Assessment of NEPTUNE_CFD Code Capabilities to Simulate Two-Phase Flow in the OECD/NRC PSBT Subchannel Experiments

This paper deals with the validation of the multifield computational fluid dynamics code NEPTUNE_CFD v2.0.1 against experimental data available from the OECD/NRC NUPEC PWR subchannel and bundle tests (PSBT) international benchmark. The present work is performed in the framework of the NURESAFE European collaborative project and focuses on the steady-state single subchannel void fraction tests. From overall 126 PSBT experiments covering wide range of test conditions and 4 different geometrical configurations of PWR subchannel, 42 tests have been selected and simulated using NEPTUNE_CFD. Following the NEA/CSNI (Nuclear Energy Agency / Committee on the Safety of Nuclear Installations) best practice guidelines about computational grid design and grid quality, mesh sensitivity analysis has been performed using axial and radial grid refinement. Both axial and radial mesh sensitivity studies do not exhibit any significant change in the predicted results, which thus result to be grid-converged. Besides, a series of sensitivity calculations have been performed in order to investigate the influence of uncertainties of the experimental boundary conditions on the code predictions. The influence of code physical and closure models on the void fraction prediction has been studied and discussed in detail. Generally, the calculated cross-sectional averaged void fraction at the measurement plane differs from the measured one by maximum of +/- 8%. This discrepancy is comparable to the 2σ experimental uncertainty range on void fraction measurement. The performed investigations have shown the ability of NEPTUNE_CFD to predict reasonably the void fraction in PSBT subchannel using appropriate modelling

Validation of CATHARE TH-SYS Code Against Experimental Reflood Tests

This paper presents results of a code validation activity that has been carried out at the University of Pisa within the EC-funded NURESAFE project, aimed to assess CATHARE2 v2.5_3 Mod3.1 code capabilities to simulate scenarios featuring reflood conditions. For such purpose, experimental data available from FEBA and ACHILLES separate-effect test facilities was used. In order to set-up a reference calculation model, rigorous sensitivity studies have been performed for every of the selected experimental test facilities. Quantitative analysis of the results has been carried out for all of the considered tests, using the Fast Fourier Transform Based Method (FFTBM) for accuracy quantification of code predictions. The calculations of experimental tests of ACHILLES facility have been performed with CATHARE2 v2.5_3 mod 3.1 using both 1-D and 3-D models. The no-regression of the results predicted by such code was successfully checked through qualitative and quantitative comparison with results obtained by the one of previous code versions: CATHARE2 v2.5_2 mod 7.1. An assessment of the capabilities of the new CATHARE3 v1.3.13 code to simulate reflood phenomena using both two- and three-field 1-D models has then been carried out, based on the same ACHILLES tests. Simulations by CATHARE3 (three-field) exhibit faster quenching than CATHARE2, mainly due to the presence of the droplet field enhancing the heat exchange from the fuel rod simulators. The performed qualitative analysis has shown the ability of CATHARE2 code to capture the main features of the reflood phenomena using appropriate modeling. Nonetheless, the quantitative analysis shows a systematic underprediction of the PCT and faster quenching in the majority of tests

FONESYS and SILENCE Networks: Looking to the Future of T-H Code Development and Experimentation

The purpose of this paper is to present briefly the projects called FONESYS (Forum & Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics) and SILENCE (Significant Light and Heavy Water Reactor Thermal Hydraulic Experiments Network for the Consistent Exploitation of the Data), their participants, their motivations, their main targets and working modalities. System Thermal-Hydraulics (SYS-TH) codes, also as part of the Best Estimate Plus Uncertainty (BEPU) approaches, are expected to achieve a more-and-more relevant role in nuclear reactor safety and design technology. Namely, the number of code-users is likely to increase in the countries where nuclear technology is exploited. Thus, the idea of establishing a forum and a network among the code developers and with possible extension to code users has started to have major importance and value. In this framework, the FONESYS initiative has been launched in 2010 aiming at creation of a common ground for discussing current limitations and envisaged improvements in various areas of SYS-TH and their application in the licensing process and safety analysis. According to FONESYS statute, there are seven signatory Institutions and two observer Institutions currently participating in the project. Signatory Institutions are AREVA-NP, Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA), San Piero a Grado Nuclear Research Group - University of Pisa (GRNSPG-UNIPI), Gesellschaft für Anlagenund Reaktorsicherheit (GRS), Korea Atomic Energy Research Institute (KAERI), Korea Institute of Nuclear Safety (KINS), and VTT Technical Research Centre of Finland. SILENCE is a network that intends to promote the cooperation among teams of experimentalists managing or involved in significant experimental projects in nuclear reactor thermal-hydraulics, with the aim to contrast the risk of losing expertise and vision in this important area of the nuclear technology. This network was launched in 2012, replicating for the TH experimental domain the role that FONESYS plays in the code-development domain. Currently, the following Organizations are Members of SILENCE: AREVA GmbH, Helmholtz Zentrum Dresden-Rossendorf (HZDR), Korea Atomic Energy Research Institute (KAERI), Hungarian Academy of Sciences Centre for Energy Research (MTA EK),Lappeenranta University of Technology (LUT), and Paul Scherrer Institute (PSI). SILENCE is currently organizing a “Specialists Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal Hydraulics” (SWINTH-2016). The San Piero a Grado Nuclear Research Group - University of Pisa (GRNSPG-UNIPI) is the Host Institution and plays as a Scientific Secretariat for both Networks

Prospective For Nuclear Thermal Hydraulic Created By Ongoing And New Networks

International audienceThis paper introduces the FONESYS, SILENCE and CONUSAF projects run by some of the leading organizations working in the nuclear sector.The FONESYS members are developers of some of the major System Thermal-Hydraulic (SYS-TH) codes adopted worldwide, whereas the SILENCE members own and operate important thermal-hydraulic experimental facilities. The two networks work in a cooperative manner and have at least one meeting per year where top-level experts in the areas of thermal-hydraulic code development and experimentation are gathered.The FONESYS members address various topics such as hyperbolicity and numerics in SYS-TH codes, 3-field modeling, transport of interfacial area, 3D modeling, scaling of thermal-hydraulic phenomena, two-phase critical flow (TPCF), critical heat flux (CHF), and others. As part of the working modalities, some numerical benchmarks were proposed and successfully conducted by the network, addressing some of the most relevant topics selected by the FONESYS members.On the other hand, SILENCE addresses topics such as identification of current measurement needs and main gaps for further SYS-TH and CFD codes development and validation, definition of similar tests and counterpart tests in Integral Tests Facilities (including containment thermal-hydraulics) to be possibly conducted on Members' test facilities, scaling issue, and other subjects. Furthermore, SILENCE organized a Specialists Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics (SWINTH) which was held in Italy on June 2016. A second edition of the Workshop, namely SWINTH-2019, will be held in Italy in 2019 under the umbrella of the OECD/NEA/CSNI/WGAMA.Recently a new initiative is being taken by launching an international consortium of nuclear thermal-hydraulics code users, the CONUSAF. The main idea is to enhance the interactions between the users of computational tools in nuclear TH, noticeably including SYS-TH and CFD codes, the code developers and the experimentalists. The proposed initiative is expected to have a positive impact on the entire ecosystem by pursuing the assessment of the current code limitations and capabilities, analyzing and addressing issues raised by the users and promoting common RandD efforts on topics of high relevance

Modeling of buoyancy driven flow-mixing experiments in Rocom test facility using 3D capabilities of Cathare code

In the present study, the 3-D features of the thermal-hydraulic system code CATHARE2 were thoroughly assessed against the OECD/PKL-2 ROCOM tests T1.1, T2.1 and T2.2 which focus on the evaluation of the thermal mixing in the reactor pressure vessel (RPV) under asymmetric buoyant cooling loop conditions. Furthermore, the calculation of the recent ROCOM test T2.3 (OECD/PKL-3), which aims at detailed investigation of the thermal-hydraulic behavior inside the RPV during the emergency core cooling injection from accumulator to the cold leg, was performed in the framework of OECD/NEA/CSNI international benchmark activity. A successful application of CFD code to support a set-up of best estimate 3-D SYS-TH nodalization of RPV was demonstrated. This was achieved by using ANSYS CFX code to evaluate the pressure losses throughout the vessel with further application of additional loss coefficients for the sieve drum and core support plate in CATHARE reference model in order to match the pressure drops predicted by the CFD model. The effect of nodalization refinement (axial, radial and azimuthal) of the ROCOM pressure vessel as well as the influence of the singular pressure losses in the sieve drum and core support plate on the predicted results was studied. It was observed that the number of azimuthal nodes is the most influential parameter on the mixing scalar. It also important to notice that the calculations conducted with refined grids show better general agreement with the experimental data. This is due to the fact that an increase of the node numbers causes a decrease of the numerical diffusion, which plays a vital role in the simulations. The performed qualitative analysis has shown the ability of CATHARE 3-D models to capture the main features of the mixing phenomena in reactor pressure vessel using appropriate modelling, however from the quantitative point of view, the effectiveness of the thermal mixing is generally overpredicted. The obtained results are qualitatively comparable to the ones obtained with CFD codes, like ANSYS CFX, CODE_SATURNE, STAR-CCM+, however require considerably less computational resources. It also important to notice that applicability of the findings of the present studies to NPP scale has to be further investigated both experimentally and analytically, due to scaling distortions which are inevitably present in any scaled test facility

Validation of the CATHARE 1-D and 3-D reflood models against FEBA and ACHILLES experimental tests

This paper presents results of a code validation activity that has been carried out at the University of Pisa within the EC-funded NURESAFE project, aimed to assess CATHARE2 v2.5_3 Mod3.1 code capabilities to simulate scenarios featuring reflood conditions. For such purpose, experimental data available from FEBA and ACHILLES separate-effect test facilities was used. In order to set-up a reference calculation model, rigorous sensitivity studies have been performed for each of the selected experimental test facilities. Quantitative analysis of the results has been carried out for all of the considered tests, using the Fast Fourier Transform Based Method (FFTBM) for accuracy quantification of code predictions. The calculations of experimental tests of ACHILLES facility have been performed with CATHARE2 v2.5_3 mod 3.1 using both 1-D and 3-D models. The non-regression of the results predicted by such code was successfully checked through qualitative and quantitative comparison with results obtained by the one of previous code versions: CATHARE2 v2.5_2 mod 7.1. An assessment of the capabilities of the new CATHARE3 v1.3.13 code to simulate reflood phenomena using both two- and three-field 1-D models has then been carried out, based on the same ACHILLES tests. Simulations by CATHARE3 (three-field) exhibit faster quenching than CATHARE2, mainly due to the presence of the droplet field enhancing the heat exchange from the fuel rod simulators. The performed qualitative analysis has shown the ability of CATHARE2 code to capture the main features of the reflood phenomena using appropriate modeling. Nonetheless, the quantitative analysis shows a systematic underprediction of the peak cladding temperature and faster quenching in the majority of FEBA and ACHILLES tests

Critical Power Prediction by CATHARE2 of the OECD/NRC BFBT Benchmark

The main purpose of this study is to evaluate the performance of the French best estimate thermal-hydraulic code CATHARE 2 regarding critical power prediction, comparing the analytical results with the experiments, available in the framework of the International OECD/NRC Benchmark, based on NUPEC (Nuclear Power Engineering Corporation) BWR Full-size Fine-mesh Bundle Tests (BFBT). Two-phase flow calculations and prediction of the void fraction distribution and the critical power were carried out both in steady state and transient cases, using 1D and 3D modeling. Comparison with the test results shows the ability of CATHARE 2 code to predict reasonably the critical power, using appropriate modeling

Thermal Hydraulic System Codes Performance in Simulating Bouyancy Flow Mixing Experiment in ROCOM Test Facility

The MSLB (Main Steam Line Break) accident scenario is one of the severe abnormal transients that might occur in a NPP. The main concerns of the MSLB are the potential return to power condition and the occurrence of PTS (Pressurized Thermal Shock) as a consequence of both rapid depressurization of the secondary circuit and the entrainment of cold water into the core region. Assessment of these issues is the main objective of integrated experimental tests carried out in the PKL-III and ROCOM facilities. The first test rig is aimed to simulate thermal-hydraulic phenomenology at the system level whereas supporting ROCOM test facility is focused on the coolant mixing phenomenon took place in the Reactor Pressure Vessel (RPV). Combination of these two typologies of experiments (integral effect test (IET) and separate effect test (SET)) provides appropriate experimental data for CFD and TH-SYS (Thermal Hydraulic-SYStem) codes validation against the relevant thermal hydraulic phenomena that occur during the MSLB. The main purpose of this study is to evaluate the capability of two TH-SYS codes TRACE V5 and CATHARE2 V2.5 to predict reasonably buoyancy driven mixing phenomena that affects the IVF (In-Vessel Flow) and the distribution of coolant temperature at the core inlet using 3-D porous media approach. Test 1.1 that had been carried out in ROCOM facility was selected to investigate the coolant mixing inside the RPV under flow conditions typical for a MSLB scenario. Averaging analysis of integral behaviour of the experimental and calculated temperature distributions inside the RPV has been performed

Validation of NEPTUNE CFD two-phase flow models against OECD/NRC PSBT subchannel experiments

This paper deals with the validation of the multifield computational fluid dynamics code NEPTUNE_CFD v2.0.1 against experimental data available from the OECD/NRC NUPEC PWR subchannel and bundle tests (PSBT) international benchmark. The present work is performed in the framework of the NURESAFE European collaborative project and focuses on the steady-state single subchannel void fraction tests. From overall 126 PSBT experiments covering wide range of test conditions and 4 different geometrical configurations of PWR subchannel, 42 tests have been selected and simulated using NEPTUNE_CFD. Following the NEA/CSNI (Nuclear Energy Agency/Committee on the Safety of Nuclear Installations) best practice guidelines about computational grid design and grid quality, mesh sensitivity analysis has been performed using axial and radial grid refinement. Both axial and radial mesh sensitivity studies do not exhibit any significant change in the predicted results, which thus result to be grid-converged. Besides, a series of sensitivity calculations have been performed in order to investigate the influence of uncertainties of the experimental boundary conditions on the code predictions. The influence of code physical and closure models on the void fraction prediction has been studied and discussed in detail. Generally, the calculated cross-sectional averaged void fraction at the measurement plane differs from the measured one by maximum of ±8%. This discrepancy is comparable to the 2Ï\u83 experimental uncertainty range on void fraction measurement. The performed investigations have shown the ability of NEPTUNE_CFD to predict reasonably the void fraction in PSBT subchannel using appropriate modeling

Critical power prediction by CATHARE2 of the OECD/NRC BFBT benchmark

International audienceThis paper presents an application of the French best estimate thermal-hydraulic code CATHARE 2 to calculate the critical power and departure from nucleate boiling (DNB) exercises of the International OECD/NRC BWR Fuel Bundle Test (BFBT) benchmark. The assessment activity is performed comparing the code calculation results with available in the framework of the benchmark experimental data from Japanese Nuclear Power Engineering Corporation (NUPEC). Two-phase flow calculations on prediction of the critical power have been carried out both in steady state and transient cases, using one-dimensional and three-dimensional modelling. Results of the steady-state critical power tests calculation have shown the ability of CATHARE code to predict reasonably the critical power and its location, using appropriate modelling. © 2014 Elsevier B.V