Thermal-hydraulic System Code Research Articles

RELAP5/Mod3.3 MHD module development and validation: WCLL-TBM mock-up model

Magnetohydrodynamic (MHD) phenomena significantly impact the design of magnetic-confinement fusion reactors, particularly in the context of breeding blankets (BB) utilizing liquid metals (LMs) as working fluids. These effects arise from the interaction between the electro-conductive flowing metal and the magnetic field used for plasma confinement in the reactor chamber. Induced electrical currents in the liquid generate Lorentz forces, thereby altering flow behaviour in comparison to standard hydrodynamic conditions. For example, MHD effects modify velocity distribution and mass transport within ducts, amplify pressure losses, and influence heat transfer mechanisms. Accurate estimation of these impacts is crucial for the effective design of a liquid metal breeding blanket. While computational tools are essential for fusion-related physical analyses, no comprehensive MHD code currently exists for simulating all relevant phenomena in a liquid metal blanket. In this context, models predicting both distributed and concentrated MHD pressure drops have been integrated into the thermal-hydraulic system code RELAP5/Mod3.3. The Verification and Validation (V&V) process compares code results to direct numerical simulations and experimental data. For validation, RELAP5 recreates experimental results of a Water Cooled Lithium Lead test blanket module at magnetic field intensities ranging from Ha=500−3000, confirming the reliability of the newly implemented MHD subroutines for predicting pressure drops within this parameter range.

Transient simulation and analysis of a supercritical CO2 heat removal system under different abnormal operation conditions

The supercritical carbon dioxide (sCO2) heat removal system, which is based on multiple closed Brayton cycles with sCO2 as the working fluid, is an innovative, self-propelling and modular heat removal system for existing and future nuclear power plants. Previous studies analysed its design, layout, control and operation. In addition, this study considers different sudden failures during the accident progress, e.g. failure of single sCO2 cycles, control systems and valves. These abnormal conditions were investigated with the thermal-hydraulic system code ATHLET for a generic Konvoi pressurized water reactor. In most cases, the failure of a single sCO2 cycle can be compensated. On the one hand, failure of the fans of the gas cooler leads to a pressure increase which may be mitigated by an inventory control system or cycle shutdown. On the other hand, unintended fan speed-up can cause compressor surge without adequate countermeasures. Furthermore, the system can operate under the cyclic blow-off from the steam generator safety valves when the relief valves are not available. Finally, the unintended closure of the valve which controls the steam flow through the compact heat exchanger triggers a fast cycle shutdown but a subsequent restart might be possible.

Comparison of thermal hydraulic performance between horizontal and vertical steam generators in nuclear power plants

Abstract Steam generators (SG) for Nuclear Power Plants (NPPs) are important large-sized metal consuming equipment of nuclear power installations. Efficiency of steam generator operation determines the overall service life of the whole nuclear facility. Vertical steam generators are used normally in North America, Europe and South East Asia, but horizontal steam generators are currently of great importance to reduce building heights and construction costs that govern the overall cost of new nuclear power plants. The objective of the current study is to measure the performance of nuclear steam generators in NPPs through comparative assessments of vertical and horizontal configurations. For this purpose, the vertical U-tube steam generators of a typical 4-loop Pressurized Water Reactor (PWR) was replaced by a VVER-1000 horizontal steam generators to evaluate the horizontal SG performance, at the same plant conditions and characteristics, in comparison with the vertical SG. The thermal hydraulic system code RELAP5 was used in the current work to perform the required analysis, during steady and transient scenarios.

Opposing Mixed Convection Heat Transfer for Turbulent Single-Phase Flows

Convection, wherein forced and natural convections are prominent, is known as mixed convection. Specifically, when a forced convection flow is downward, this flow is called opposing flow. The objectives of this study are to gain a comprehensive understanding of opposing flow mixed convection heat transfer and to establish the prediction methodology by evaluating existing correlations and models. Several heat transfer correlations have been reported related to single-phase opposing flow; however, these correlations are based on experiments conducted in various channel geometries, working fluids, and thermal flow parameter ranges. Because the definition of nondimensional parameters and their validated range confirmed by experiments differ for each correlation reported in previous studies, establishing a guideline for deciding which correlation should be selected based on its range of applicability and extrapolation performance is important. This study reviewed the existing heat transfer correlations for turbulent opposing flow mixed convection and the single-phase heat transfer correlations implemented in the thermal–hydraulic system codes. Furthermore, the authors evaluated the predictive performance of each correlation by comparing them with the experimental data obtained under various experimental conditions. The Jackson and Fewster, Churchill, and Swanson and Catton correlations can accurately predict all the experimental data. The effect of the difference in the thermal boundary conditions, i.e., uniform heat flux and uniform wall temperature, on the turbulent mixed convection heat transfer coefficient is not substantial. The authors confirmed that heat transfer correlations using the hydraulic-equivalent diameter as a characteristic length can be used for predictions regardless of channel-geometry differences. Furthermore, correlations described based on nondimensional dominant parameters can be used for predictions regardless of the differences in working fluids. The authors investigated the extrapolation performance of the mixed convection heat transfer correlations for a wide range of nondimensional parameters and observed that the Jackson and Fewster, Churchill, and Aicher and Martin correlations exhibit excellent extrapolation performance with respect to natural and forced convection flows, indicating that they can be applied beyond the parameter range validated experimentally.

Thermal-hydraulic study of the EU-DEMO Helium Cooled Pebble Bed Breeding Blanket Primary Heat Transport System

The European DEMO reactor will work under normal operating conditions in accordance with a pulsed duty cycle. However unplanned and planned transients of plasma overpower may occur compromising the integrity of its plasma-facing components structures. Consequently, adopting appropriate tools is essential to accurately and consistently model the thermal-hydraulic response of the involved cooling systems under both normal operating conditions and accidental events. Given this background, the University of Palermo, in collaboration with EUROfusion, started a research work to study the thermal-hydraulic behaviour of the Primary Heat Transport System (PHTS) of the Helium Cooled Pebble Bed Breeding Blanket (HCPB BB) of the DEMO device under steady-state and transient conditions. The activity has been performed adopting a computational approach, employing the thermal-hydraulic system code TRACE version 5.0 patch 6. The key point of the work has been the code-to-code benchmark with the outcomes previously obtained with the RELAP5-3D code, to estimate the impact of the physical models, numerical resolution schemes and modelling techniques adopted on the predictive capabilities of the system codes considered. The models and the analysis results are presented and critically discussed herein.

Preliminary thermal-hydraulic deterministic safety analysis of an in-vessel LOCA for the DTT facility

The DTT (Divertor Tokamak Test) facility is a new experimental tokamak under development by an Italian consortium in cooperation with several high-standard European laboratories. The ENEA division FSN-SICNUC is involved in the project for carrying out deterministic safety analysis of postulated accidents. In the first year of activity, it has been analyzed an In-Vessel LOCA (IVLOCA) scenario. The IVLOCA scenario selected is caused by a break in the divertor’s cassette cooling tubes with the consequent release of coolant inside the Vacuum Vessel (VV). The best estimate thermal-hydraulic system code TRACE, developed by USNRC, was selected to conduct the preliminary thermal-hydraulic analysis of this accident. DTT is currently under design, so not all the data are frozen. Therefore, based on some engineering assumptions and scaling considerations from similar facilities, the nodalization of DTT VV was developed using the three-dimensional TRACE component “VESSEL”. Then, a sensitivity analysis was carried out to simulate the break and the consequent water injection in the VV. This was done to compare the system behavior and test different nodalization approaches. Subsequently, the data of the DTT divertor cooling system were used to run additional simulations. The results allow to compare different nodalization approaches and to have a preliminary estimate of the pressure peak and temperature behavior in the VV for an IVLOCA. Finally, a first uncertainty analysis was carried out using the DAKOTA toolkit, coupled with TRACE code in SNAP. Two uncertain input parameters were selected: the rupture area of a cooling divertor tube and the temperature of the divertor coolant. The uncertainty analysis allows having a wider spectrum of system behavior and a preliminary insight on the dispersion of the VV pressure, selected as figure of merit. This paper aims to show the results of this preliminary analysis, characterizing the phenomena that occurred during the selected transient.

BEPU analysis of an Upgraded ICE facility test by TRACE/DAKOTA coupling

The Ingress of Coolant Event (ICE) in the plasma chamber is one of the safety issues in fusion nuclear plants. The best estimate thermal-hydraulic system codes adopted to perform deterministic safety analysis should be validated against the phenomena typical of accidental transients in fusion installations. TRACE (TRAC/RELAP Advanced Computational Engine), best estimate thermal-hydraulic system code developed by USNRC, has been adopted to simulate an ICE. The calculated results have been compared to the experimental data obtained in one test performed in the upgraded Integrated ICE facility at JAERI. In this updated configuration the pressure suppression system is connected to the top of the plasma chamber instead of the bottom of the vacuum vessel. The facility nodalization has been developed in the SNAP environment/architecture. To qualify the code and the nodalization, an accuracy evaluation has been performed both from a qualitative and quantitative point of view. Then, considering the presence of some uncertainties in the input-deck development, an uncertainty analysis has been carried out. The probabilistic method to propagate the input uncertainties has been selected and the analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in SNAP. In the uncertainty analysis, some relevant statistical parameters have been considered to characterize the dispersion of the results and the correlation between the uncertain input parameters selected and the PC pressure chosen as figure of merit.

Analysis of main steam line break accident on a BWR test facility using TRACE

PUMA-E is an integral test facility, which is used for extensive experimental investigations to study the behavior of Boiling Water Reactor (BWR) plants under accident conditions. The TRACE code is the best estimate thermal hydraulic system code developed by the U.S. Nuclear Regulatory Commission for performing nuclear power plant safety analysis. The aim of the present work is to conduct a short validation study of the TRACE code applied to the PUMA-E facility. Firstly, a TRACE model of the PUMA-E facility was developed based on the existing RELAP5 input deck. Then, an additional facility TRACE model with reactor pressure vessel modeled by three dimensional “VESSEL” component was also established. After that, a Main Steam Line Break accident test was simulated, and the calculation results were compared with experimental data and RELAP5 simulation results. The TRACE simulation capability to predict the accident phenomenon of BWR test facility was evaluated using both qualitative and quantitative methods. On the basis of the above research, the base case accompanying with one Gravity Driven Cooling System tank failure accident was carried out based on the three facility models. The validity and reliability of the simulation results of the different facility numerical models were mutually verified through a qualitative comparison of the simulation results.

Development of a Fully-Implicit Second-Order system Thermal-Hydraulics analysis code for reactor modeling and simulations

There has been continuous advancement in numerical techniques and software engineering in the field of system thermal-hydraulics over the past decade. The current state-of-the-art involves the use of fully-implicit high-resolution schemes in conjunction with the Jacobian-Free Newton-Krylov (JFNK) method for both single-phase and two-phase flows. However, the application of these advanced schemes in system thermal-hydraulics codes is not a common practice. Given the opportunity and the need to develop a new modern system thermal–hydraulic code from scratch, there is no doubt that these recent advancements should be leveraged. In this work, we introduce a new modern object-oriented system thermal-hydraulics code named RETA. RETA is characterized by several key features, including: second-order temporal and spatial discretization schemes for both incompressible and compressible flows, as well as single-phase and two-phase flows; fully-implicit solution schemes utilizing both the Newton and PJFNK methods; and an object-oriented design of major fluid components and physical models, which allows for easy extension. A series of verification and demonstration studies are conducted to verify the implementation, demonstrate the accuracy improvement, and evaluate the performance of the new code. It is concluded that results from the second-order scheme are generally more reliable. Without the concern of a code stability issue, the second-order scheme should be the preferred and default choice for a code user or an analyst.

SiRENE: A new generation of engineering simulator for real-time simulators at EDF

For Safety Assisted Engineering works, real-time simulators have emerged as a mandatory tool among all the key actors involved in the nuclear industry (utilities, designers and safety authorities). EDF, Electricité de France, as the leading worldwide nuclear power plant operator, has a crucial need for efficient and updated simulation tools for training, operating and safety analysis support.This paper will present the work performed at EDF/DT to develop a new generation of engineering simulator to fulfil these tasks. The project is called SiRENE, which is the acronym of Re-hosted Engineering Simulator in French. The project has been economically challenging. Therefore, to benefit from existing tools and experience, the SiRENE project combines:- A part of the process issued from the operating fleet training full-scope simulator.- An improvement of the simulator prediction reliability with the integration of High-Fidelity models, used in Safety Analysis. These High-Fidelity models address Nuclear Steam Supply System code, with CATHARE thermal-hydraulics system code and neutronics, with COCCINELLE code.- And taking advantage of the last generation and improvements of instructor station.The intensive and challenging uses of the new SiRENE engineering simulator are also discussed. The SiRENE simulator has to address different topics such as verification and validation of operating procedures, identification of safety paths, tests of I&C developments or modifications, tests on hydraulics system components (pump, valve etc.), support studies for Probabilistic Safety Analysis (PSA). etc. It also emerges that SiRENE simulator is a valuable tool for self-training of the newcomers in EDF nuclear engineering centers.As a modifiable tool and thanks to a skillful team managing the SiRENE project, specific and adapted modifications can be taken into account very quickly, in order to provide the best answers for our users' specific issues.Finally, the SiRENE simulator, and the associated configurations, has been distributed among the different engineering centers at EDF (DT in Lyon, DIPDE in Marseille and CNEPE in Tours). This distribution highlights a strong synergy and complementarity of the different engineering institutes at EDF, working together for a safer and a more profitable operating fleet.

Validation of modeling scheme implementing MULTID component of MARS-KS code to predict shell-side pressure drop in shell-and-tube type heat exchangers

A modeling scheme to predict the pressure drop on the shell side of the shell-and-tube type heat exchanger is developed, which is using the system thermal hydraulic code, MARS-KS. In the present scheme, the shell side of the heat exchanger is modeled by the MULTID component in rectangular shape in which, the part blocked by the segmental baffle plate and the opening part by the baffle cut are modeled by separate nodes. The form loss coefficients and its distribution for the junctions of the shell side are estimated by the assumption of the same pressure drop in each cross pass of the shell and then finalized by iterative calculation. To prove the validity of the modeling scheme, calculations are made to simulate the experiments conducted at the test heat exchanger having a size used in the industries. It is discussed whether each effect of the size of inlet/outlet nozzle and the height of the baffle cut on the overall pressure drop of the heat exchanger is properly predicted and consistent with the experimental observations. Also, the effectiveness of consideration of leakage at the clearance between the shell and the segmental baffle plates is discussed in resolving the problem of over-predicted pressure drops in the high flow region. For the simplification, the use of one-dimensional modeling to this problem is attempted. The applicability of the present method and results to the different heat exchangers is also discussed.

Analysis of thermal hydraulic system code LOCUST V1.2 with ECC thermal mixing test facility

The safety injection (SI) system might be put into use, and the coolant at high temperature will mix with the supcooled water under Loss Of Coolant Accident (LOCA). The thermo-hydraulic phenomena associated with the thermal mixing of coolant and supcooled water will directly affect the judgment on core reflooding. China General Nuclear Power Group (CGN) independently developed and designed a thermal hydraulic system code named LOCUST, and the verification of thermal mixing models in LOCUST is necessary. This paper will introduce the thermal mixing tests at the T-junction based on Emergency Core Cooling System in Xi'an Jiaotong University (ECCS-XJTU) experimental facility, which were mainly conducted for the mixing between subcooled water injected from the SI pipe with a range of 25–125 kg/h in mass flow and the pure steam in the primary pipe with a range of 100–500 kg/h in mass flow. Besides, the fluid temperature and the dynamic vapor quality after mixing in tests are analyzed, and the simulations of 25 thermal mixing tests using the LOCUST 1.2 code are performed. The results show that the maximum relative error of LOCUST in mass flow of liquid is within 13.8 %, and the maximum relative error of LOCUST in temperature is within 8 %, which validates the reliability and accuracy of simulations of LOCUST for two-phase thermal mixing in LOCA.

Benchmark analysis of the FFTF LOSWOS test #13 with OpenMC and THACS

In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi’an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie’ outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR.

Detail CFD simulation of the integrated reactor pressure vessel of SMART

This paper presents a detailed analysis of the full Reactor Pressure Vessel (RPV) of SMART with ANSYS CFX 20.2. SMART is a water-cooled SMR designed by KAERI for an operation with forced convection. The thermal power of SMART amounts to 330 MWth. In the integrated SMR-concept, the main primary loop components such as the helical steam generators, pressurizer and canned pumps are located inside the RPV. The CFD model uses a concept of full geometrical detail resolution with partial boundary inflation of the mesh and porous media modelling with sources for momentum and energy for components like the core, pumps and steam generators. The detailed results obtained with a steady-state RANS simulation using ANSYS CFX 20.2 are very promising as the comparison of selected core parameters of the CFD-simulation with the ones of the system thermal hydraulic code TRACE have shown.

Development of a one-dimensional system code for the analysis of downward air-water two-phase flow in large vertical pipes

In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphon-breaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as ⟨jg⟩/⟨j⟩ increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.

Time-series forecasting of a typical PWR system response under Control Element Assembly withdrawal at full power

To expedite the decision-making process under Nuclear Power Plant (NPP) accident conditions, at a reduced computational cost, a Machine Learning (ML) time-series meta-model is proposed to map the relationships between the NPP's real-time parameters to its transient response and hence forecast its future response. To this end, three variations of Recurrent Neural Networks (RNNs), namely: the Long-Short-Term-Memory (LSTM), Gated Recurrent Unit (GRU) and a series combination of a convolutional neural network (CNN) with LSTM were developed. The NPP response was simulated using the thermal hydraulics best estimate code, MARS-KS, and cross-validated against the Design Control Document (DCD) of APR1400. The time-series database required to train, test, and validate the time-series Machine Learning (ML) expert algorithm was acquired by coupling DAKOTA software with the best estimate thermal hydraulics system code, MARS-KS to conduct a Best Estimate Plus Uncertainty (BEPU) analysis. The Monte-Carlo random propagation approach and the Latin Hypercube Sampling technique were used to sample and propagate key uncertain parameters until a statistically significant sample of the NPP transient responses was achieved. All three RNN models were developed and optimized with the help of the Talos tool, and were found to exhibit reasonable performance in forecasting the key most probable NPP transient responses, Furthermore, the uncertainty of the ML predictions was assessed by integrating a Bayesian Neural Network (BNN) with the LSTM model. The prediction accuracy ranged from 86% to 96%, when evaluated on an unseen test dataset. Based on a voting system with five performance metrics (MAE, MSE, RMSE, R2 and Accuracy) the LSTM model and its derivative CNN+LSTM outperformed the GRU model; while the CNN +LSTM is found to be the most computationally efficient.

Fast reactor multi-scale and multi-physics modelling at NRG

With the increase in computational power and capacity, and the advancements being made in numerical modeling, it has become possible to model the various physical phenomena that take place in nuclear reactors with more and more detail and accuracy. This includes phenomena related to structural mechanics, fluid dynamics and reactor physics amongst others. Additionally, there has been an increased interest to simulate these combined, interacting phenomena simultaneously by coupling various numerical tools. This coupling of different codes is currently a topic high on the research and development agenda of the international nuclear community. It is also a focus point within the research done at NRG in the national PIONEER research program funded by the Dutch ministry of economic affairs.This focus has resulted in two branches of research at NRG: multi-scale modeling of the complete primary system of a nuclear reactor by coupling a 3D Computational Fluid Dynamics (CFD) code with a 1D System Thermal Hydraulics (STH) code, and multi-physics modelling through the coupling of dedicated and advanced thermal-dynamics, structural mechanics and reactor physics solvers. The paper presents simulation results of both branches applied to fast reactors. As both fields of research require the coupling of codes, it has led to the creation of an independent, external, Fortran-based coupling tool named myMUSCLE (MultiphYsics MUltiscale Simulation CoupLing Environment) that arranges the efficient and robust coupling of the different codes. The paper presents the proof-of-principle and first validation of the myMUSCLE tool under development. Additionally, results obtained with MUSCLE-Foam will be presented, which is a new multi-physics code being developed at NRG that includes several reactor physics as well as structural mechanics solvers.

Thermal-hydraulic study of the Primary Heat Transport System of the DEMO Divertor Cassette Body

The EU-DEMO reactor is supposed to undergo a pulsed duty cycle under normal conditions, which might challenge the qualified lifetime of the main equipment inducing undue thermal and mechanical cycling loads. Moreover, during power operation it is not possible to exclude the occurrence plasma overpower transients that might jeopardize the structural integrity of the plasma-facing components. In this context, the University of Palermo in cooperation with ENEA has launched a research campaign within the framework of the EUROfusion action to investigate the thermal-hydraulic behaviour of the Primary Heat Transport System (PHTS) of the EU-DEMO Divertor (DIV) Cassette Body (CB) during the typical DEMO duty cycle. The study was primarily intended to start the definition of the strategies to be followed to minimize thermal fluctuations due to the pulsed operation. In particular, the attention has been focussed on the assessment of the DIV CB PHTS temperature behaviour during the pulse-dwell transition so to check the potential effectiveness of a direct coolant temperature control by properly tuning the feedwater mass flow rate. The activity has been led following a computational approach based on the adoption of the TRACE thermal-hydraulic system code. Models, assumptions, and outcomes of the analyses are herewith reported and critically discussed.

Design and Assessment of the New CIRCE-THETIS Facility for the Development of Heavy Liquid Metal Cooled Fast Reactors

Within the roadmap for the technological development of Generation IV reactors, the HORIZON2020 European Union–funded Partitioning And Transmuter Research Initiative in a Collaborative Innovation Action (PATRICIA) project was launched to support innovative solutions for the development of the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA) accelerator-driven system concept and lead fast reactor technologies in general. The ENEA contributes to the project by involving the experimental infrastructures of the Brasimone Research Center (Italy). In particular, a large-scale pool-type facility named CIRColazione Eutettico (CIRCE), using lead-bismuth eutectic as the primary coolant and pressurized water as the secondary fluid, is under refurbishment with the implementation of a novel test section (TS) named Thermal-hydraulic HElical Tubes Innovative System (THETIS) to be installed in the CIRCE main vessel. The new TS will include a vertical mechanical pump for primary coolant circulation and a new prototypical helical coil steam generator (HCSG). This steam generator concept turns out to be very promising for nuclear power plants since the helical geometry is very compact and it assures high power removed, taking up a minimum amount of space. Accordingly, with the aims of the project, the experimental tests in CIRCE-THETIS will focus on (1) investigating the thermal-hydraulic behavior of the system in steady-state operation (forced circulation regime) during operational and accidental transients (postulated scenarios) and in a natural circulation regime considering as heat sink the HCSG (acting as a decay heat removal system) and the reactor vessel auxiliary cooling system in stand-alone or coupled operation and (2) characterizing the performance of the HCSG. The present work presents the layout of the CIRCE-THETIS facility at the end of the final design phase, describing in detail the main components of the TS, along with the instrumentation installed. Focus will be given to the HCSG mock-up, for which pretest analyses using the system thermal-hydraulic code RELAP5/Mod3.3 and a computational fluid dynamics code have been carried out to support the design of the component and to evaluate its thermal-hydraulic performance under the operative conditions foreseen during the experiments.

A hybrid domain overlapping method for coupling System Thermal Hydraulics and CFD codes

A hybrid multiscale coupling methodology based on a domain overlapping approach has been developed for coupling System Thermal Hydraulics (STH) and Computational Fluid Dynamics (CFD) codes. The method has been implemented between the modern STH code SAM and the CFD code NekRS, using the coupling tool Cardinal. The coupling aims to extend the STH code’s applicability to scenarios where local momentum and energy transfers are important yet difficult for STH codes to capture, such as three-dimensional mixing.Two coupling strategies are implemented and compared: a hybrid domain overlapping method and the conventional domain decomposition method. The strategies are applied to two closed-loop applications, and the present method shows superior stability behavior when compared to the domain decomposition method. Then, the present coupling method is validated against experimental data from a double T-junction experiment. The present STH/CFD coupling shows improved agreement with experimental data when compared to STH standalone simulations.