Thermal-hydraulic Simulation Research Articles

Bayesian inference and non-linear extensions of the CIRCE method for quantifying the uncertainty of closure relationships integrated into thermal-hydraulic system codes

Uncertainty Quantification of closure relationships integrated into thermal-hydraulic system codes is a critical prerequisite in applying the Best-Estimate Plus Uncertainty (BEPU) methodology for nuclear safety and licensing processes. This issue has been subject to several international initiatives, such as BEMUSE and PREMIUM projects, as well as some statistical developments of which the “CIRCE” method. This method has been designed at the end of the twentieth century, then extensively used for quantifying the uncertainty of closure relationships integrated into the CATHARE thermal-hydraulic system code.The purpose of the CIRCE method is to estimate the (log)-Gaussian probability distribution of a multiplicative factor applied to a reference closure relationship in order to assess its uncertainty. Even though this method has been implemented with success in numerous physical scenarios, it can still suffer from substantial limitations such as the linearity assumption and the difficulty of properly taken into account the inherent statistical uncertainty. In the paper, we will extend the CIRCE method in two aspects. On the one hand, we adopt the Bayesian setting putting prior probability distributions on the parameters of the (log)-Gaussian distribution. The posterior distribution of the parameters is then computed with respect to an experimental database by means of Markov Chain Monte Carlo (MCMC) algorithms. On the other hand, we tackle the more general setting where the simulations do not move linearly against the multiplicative factor(s). MCMC algorithms then become time-prohibitive when the thermal-hydraulic simulations exceed a few minutes. This handicap is overcome by using Gaussian process (GP) emulators which can yield both reliable and fast predictions of the simulations.The GP-based MCMC algorithms will be applied to quantify the uncertainty of two condensation closure relationships at a safety injection with respect to a database of experimental tests. The thermal-hydraulic simulations will be run with the CATHARE 2 computer code.

SESAME project: advancements in liquid metal thermal hydraulics experiments and simulations

Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possibility to use natural resources efficiently and to reduce the volume and lifetime of nuclear waste. Sodium and Liquid lead (-alloys) are considered the short and long term solution respectively, as coolant in GEN-IV reactor. Thermal-hydraulics of liquid metals plays a key role in the design and safety assessments of these reactors. Therefore, this is the main topic of a large European collaborative program (the Horizon 2020 SESAME) sponsored by the European Commission. This paper will present the progress in the project with respect to liquid metal cooled reactor thermal-hydraulics (liquid metal heat transfer, fuel assembly thermal-hydraulics, pool thermal-hydraulics, and system thermal-hydraulics). New reference data, both experimental and high-fidelity numerical data is being generated. And finally, when considering the system scale, the purpose is to validate and improve system thermal-hydraulics models and codes, but also to further develop and validate multi-scale approaches under development.

Coupled Neutronics–Thermal-Hydraulic Simulation of BEAVRS Cycle 1 Depletion by the MCS/CTF Code System

The coupled neutronics–thermal-hydraulic simulation of the Benchmark for Evaluation and Validation of Reactor Simulations (BEAVRS) Cycle 1 depletion has been performed by the Monte Carlo–based multiphysics coupling code system MCS/CTF. MCS/CTF is a cyclewise pi-card iteration-based inner-coupling code system that couples the subchannel thermal-hydraulic code CTF as a thermal-hydraulic solver in the Monte Carlo neutron transport code MCS. MCS has been developed by the Computational Reactor Physics and Experiment Lab group at the Ulsan National Institute of Science and Technology for the full-core analysis of large-scale commercial light water reactors with high fidelity at the engineering level. With the high-fidelity performance of MCS, the quarter-core pinwise depletion simulation for the BEAVRS Cycle 1 benchmark has been conducted with thermal-hydraulic feedback including fuel temperature, coolant temperature, and coolant density. Moreover, the MCS internal one-dimensional thermal-hydraulic solver TH1D (MCS/TH1D) has been utilized as the reference. On one hand, the simulated results of the criticality boron concentration and axially integrated assemblywise detector signals were compared with measured data. On the other hand, the comparisons of power, fuel temperature, coolant temperature, and density are also presented in this paper.

ОБ ОЦЕНКЕ ПОГРЕШНОСТИ РАСЧЕТА МОЩНОСТИ МЕТОДОМ АНАЛИЗА НЕОПРЕДЕЛЕННОСТИ И ЧУВСТВИТЕЛЬНОСТИ ПРИ ВЕРИФИКАЦИИ СОПРЯЖЕННЫХ НЕЙТРОННО-ТЕПЛОГИДРАВЛИЧЕСКИХ КОДОВ

This paper describes the capability of sensitivity and uncertainty analysis (SUA) for estimation of power calculation error during the computer code verification compared to the data received in experiments at operating power plants. Estimation of errors in code verification calculation against measurements at operating plants and assessment of conservatism in safety calculations are considered as two interrelated problems. A defining moment in this approach is that one or more test cases are selected; selected test cases are included in matrix of code verification assuming that these test cases involve the same physical phenomena as in simulated accident scenarios. The criterion of test case “representativeness” is similarity between correlation coefficients of model predictions in accident scenario and test case. The example demonstrates how the SUA approach is used in calculations with the VVER-1000 coupled neutronic and thermal-hydraulic simulation model based on the KORSAR/GP and SAPFIR_95&RC_VVER codes.

Sensitivity analysis based on Morris method of passive system performance under ocean conditions

Passive safety systems, which do not require external input to operate, are widely used to enhance the inherent safety in new generation nuclear power plants. A challenging issue still to be resolved is the quantification of the functional failure probability of passive safety systems. The evaluation of functional failure relies on repeated thermal-hydraulics simulations which is usually time-consuming in practice. Response surface method has been proposed to tackle this issue. However, the number of input parameter limits its efficiency to a great extent. Therefore, it is necessary to identify the uncertain input parameters that have an important impact on the system performance, which helps reduce the dimension of the input parameters. Especially in marine nuclear power plants, the complex ocean motions will introduce more uncertainties in the passive system behavior. It’s of great significance to screen key parameters for passive safety systems under ocean conditions. In this paper, the Morris method is used to perform an efficient sensitivity analysis to identify key parameters for a passive residual heat removal system in IPWR200. Twenty-four parameters related to passive system behavior are screened and ranked for their sensitivity. The results demonstrated that four of the twenty-four parameters have a more significant influence on the passive system performance.

Realistic estimation of human error probability through Monte Carlo thermal-hydraulic simulation

Human reliability analysis (HRA) plays an important role in probabilistic safety assessment because human error probability (HEP) strongly influences not only core damage frequency in the nuclear field, for example, but also risk insight across many industries. Conventional HRA methods though use thermal-hydraulic analysis with best-estimate or conservative conditions to estimate the available time for a given operator action, thus significantly overestimating HEP. In response, this paper suggests a new method to more precisely quantify HEP using realistic thermal-hydraulic analyses with exhaustive conditions. The developed method was applied to the rapid cooldown operation in a small loss-of-coolant accident in the APR1400 nuclear power plant. As a result, it was found that conventional HRA methods currently overestimate the risk-significance of this rapid cooldown operation by a factor of six. It is expected that the new HRA method can be applied to not only the nuclear field but also other complex systems to accurately estimate system failure probabilities for specific scenarios by means of more realistic HEP.

MYRRHA primary heat exchanger experimental simulations on CIRCE-HERO

In the frame of the MYRRHA ADS technological development, an intense research effort has been undertaken by European Commission, in particular for the Heavy Liquid Metal technology assessment. In this scenario, EURATOM HORIZON2020 funded the SESAME & MYRTE projects, coordinating a series of thermal hydraulics experiments and simulations for the safety assessment of liquid metal fast reactors. In particular, in the MYRTE project, a dedicated activity has been defined, consisting of a low pressure experimental campaign realized at the ENEA Brasimone Research Centre and concerning a double wall Steam Generator Bayonet Tube (SGBT), aiming at supporting the development of the Primary Heat Exchanger (PHX) of MYRRHA.The CIRCE (CIRColazione Eutettico) pool facility hosts a devoted Test Section (TS) named HERO (Heavy liquid mEtal pRessurized water cOoled tubes). HERO consists of seven double wall bayonet tubes, with an active length of 6 m, arranged in a hexagonal geometry. This solution aims at improving the plant safety reducing the possibility of water-lead/leadalloy interaction thanks to a double physical separation between them and allowing an easier control of potential leakages of the coolant by pressurizing the separation region with inert gas.The experimental campaign in CIRCE-HERO has been supported by a pre-test analysis using system thermal-hydraulic codes (i.e. RELAP5). A sensitivity analysis has been performed considering the mass flow rate and HERO inlet temperature for both the water and LBE, fixing the operating pressure at 16 bar.A test matrix, as produced as output of the sensitivity analysis and aimed at reproducing as faithfully as possible the thermal-hydraulic behaviour of the MYRRHA PHX under different operating conditions, is here presented. The experimental results collected from the low pressure tests on HERO test section are reported and discussed.

Thermal-hydraulic simulation and evaluation of a natural circulation thermosyphon loop for a reactor cavity cooling system of a high-temperature reactor

Abstract The investigation into a full-scale 27 m high, by 6 m wide, thermosyphon loop. The simulation model is based on a one-dimensional axially-symmetrical control volume approach, where the loop is divided into a series of discreet control volumes. The three conservation equations, namely, mass, momentum and energy, were applied to these control volumes and solved with an explicit numerical method. The flow is assumed to be quasi-static, implying that the mass-flow rate changes over time. However, at any instant in time the mass-flow rate is constant around the loop. The boussinesq approximation was invoked, and a reasonable correlation between the experimental and theoretical results was obtained. Experimental results are presented and the flow regimes of the working fluid inside the loop identified. The results indicate that a series of such thermosyphon loops can be used as a cavity cooling system and that the one-dimensional theoretical model can predict the internal temperature and mass-flow rate of the thermosyphon loop.

Development and verification of numerical study to calculate focal points of temperature in helium cooled ceramic blanket (HCCB) system in CFETR

In this paper, a novel theoretical model has been developed to obtain focal points of temperatures in helium cooled ceramic breeder (HCCB) blanket system of China Fusion Engineering Test Reactor (CFETR). The main objective of the paper is to develop a theoretical-computational model for safety of blanket system, structure stress analysis and feasibility design. The theoretical-computational model was developed and verified by computational techniques using C++/MATLAB program and later with thermo-hydraulic system codes, i.e. RELAP5 and THEATRE. Mathematical model was developed for the thermal conductivity in HCCB blanket system and heat conduction model by using numerical technique. This model was then transferred to C++/MATLAB program and developed a program for calculating the thermal contact temperature of HCCB region. The calculated temperature values were then compared with RELAP5 and THEATRE codes for the accuracy of theoretical model. The preliminary thermal-hydraulic simulation results indicate that the maximum temperature on the first wall (FW) satisfies the design allowable limit and that the cooling channels were designed accurately. The results obtained from these analyses are very close to the experimentally defined values depicting the accuracy of theoretical and simulation results. The calculated results can be applicable to other tokamak devices like CFETR, ITER and JET as well.

Thermal-hydraulic model for axial piston pump with leakage and friction and its application

PurposeThis paper aims to carry out a thermal-hydraulic simulation model for pump and hydraulic system to predict the temperature increasing and pump performance. Based on the model, how to alleviate the temperature is introduced. Besides, the optimization of piston is carried out.Design/methodology/approachThis paper analyzes the heat generation in lubricating interfaces of the pump with energy conversion theory. The heat transfer inside the pump is analyzed with the control volume method. The simulation model is constructed in AMESim because of its operating friendly nature. The experiment is carried out to prove the validity and accuracy of the simulation model.FindingsTemperature has less effect on the mechanical loss of pump. However, it has a great impact on volumetric efficiency. To reduce the temperature on the piston surface, the size of the piston should be optimized.Originality/valueThis paper fulfills a novel thermal-hydraulic model to evaluate the temperature of the pump. Based on the model, the performance of the pump is determined and optimization is carried out.

Thermal-hydraulic and thermo-mechanical simulations of Water-Heavy Liquid Metal interactions towards the DEMO WCLL breeding blanket design

The Water-Cooled Lithium Lead breeding blanket concept foresees the eutectic lithium-lead (Pb-15.7Li) alloy being cooled by pressurized sub-cooled water (temperature 295–328 °C; pressure 15.5 MPa) flowing in double wall tubes. Therefore, the interaction between the Pb-15.7Li and water (e.g. tube rupture) represents one of the main safety concerns for the design and safety analysis. Available LIFUS5/Mod2 experimental data are employed to assess the performances of thermal-hydraulic and thermo-mechanic codes. Thermal-hydraulic simulations, by SIMMER-III code, are focused on the prediction of the thermodynamic interaction among the fluids. ABAQUS Finite Element code, used for the design activities, is adotped to perform the thermo-mechanic simulations, calculating the stress and strain fields of LIFUS5/Mod2 main vessel during the experiments. Code results are compared with the experimental data and the outcomes from the analyses are discussed, in order to derive conclusions on the code assessment.

Thermal and hydraulic evaluation of a linear Fresnel solar collector loop operated with molten salt and liquid metal

In this paper, results of a thermal-hydraulic analysis of a linear Fresnel solar collector loop using molten salt or liquid metals as heat transfer fluid are presented. The purpose of this study is to compare the benefits and challenges of using liquid metals (e.g. sodium) or molten salts (e.g. solar salt) as heat transfer fluid into line focusing Solar Thermal Electric plants. Similar studies have been conducted for point focussing Solar Thermal Electric plants but line focussing plants have not been thoroughly investigated yet. After reviewing and comparing the main thermo-physical properties of sodium and solar salt, results from thermal-hydraulic simulations, using the best-estimate system code TRACE from the US Nuclear Regulatory Commission, are presented for various plant operation scenarios. The results show that sodium offers several advantages over solar salt when used as heat transfer fluid, among which: wider operation temperature range, faster start-up procedures, quicker response of the control system and ultimately potentially higher energy yield, mostly thanks to its hundredfold higher thermal conductivity. These benefits may result in increased interest into this technological concept and might lead to further developments of Solar Thermal Electric plant design with reduced levelised cost of electricity generation.

A data-driven framework for error estimation and mesh-model optimization in system-level thermal-hydraulic simulation

Over the past decades, several computer codes have been developed for simulation and analysis of thermal-hydraulics and system response in nuclear reactors under operating, abnormal transient, and accident conditions. However, simulation errors and uncertainties still inevitably exist even while these codes have been extensively assessed and used. In this work, a data-driven framework (Optimal Mesh/Model Information System, OMIS) is formulated and demonstrated to estimate simulation error and suggest optimal selection of computational mesh size (i.e., nodalization) and constitutive correlations (e.g., wall functions and turbulence models) for low-fidelity, coarse-mesh thermal-hydraulic simulation, in order to achieve accuracy comparable to that of high-fidelity simulation. Using results from high-fidelity simulations and experimental data with many fast-running low-fidelity simulations, an error database is built and used to train a machine learning model that can determine the relationship between local simulation error and local physical features. This machine learning model is then used to generate insight and help correct low-fidelity simulations for similar physical conditions. The OMIS framework is designed as a modularized six-step procedure and accomplished with state-of-the-art methods and algorithms. A mixed-convection case study was performed to illustrate the entire framework.

Implementation and validation of a one-step coupled solution method for the two-fluid model

Most of the system codes for reactor Thermal hydraulics simulation are based on the two-fluid model. The accurate and efficient simulations of the two-fluid model are therefore important. This paper introduces a numerically convenient set of equations and the associated one-step coupled solution method for the two-phase two-fluid model. These equations are devised from the RELAP5 difference equations, but they are cast into a new form having great numerical effectiveness in forming a coupled matrix equation. A one-step solution of the coupled matrix equation is performed to obtain the whole unknown variables of the flow system. Proper numerical treatments, closure relations and coupling of wall conduction are accounted for by modeling boiling phase change convective heat transfer behavior. Finally, the numerical solutions of the new two-fluid model equations were successfully implemented and validated using the water faucet problem and a subcooled flow boiling experiment, and good agreements were obtained.

Dynamic thermal-hydraulic simulations of the JT-60SA cryogenic system for preparing plasma operation

Recent measurements on the JT-60SA toroidal field coils showed a slightly higher pressure drop than expected, resulting in a lower supercritical helium mass flow rate during normal plasma operation. This paper presents some simulations of the cooling loop considering these updated thermal-hydraulic parameters and the expected pulsed heat loads. The model includes a cold circulator, heat exchangers, helium thermal buffer, the distribution system and the superconducting coils. The article presents the differences between the initial scenario which was validated during commissioning of the cryogenic system and the updated scenario to prepare plasma operation with the cryo-magnetic system. The simulation also provides the transient heat load profiles of the cooling loop, highlighting a first smoothing of the thermal loads at the interface with the cryogenic system. The Simcryogenics code, developed at CEA Grenoble, is used to simulate the dynamic loads.

Uncertainty Quantification and Propagation of Multiphysics Simulation of the Pressurized Water Reactor Core

In recent years, the demand to provide best-estimate predictions with confident bounds is increasing for the nuclear reactor performance and safety analysis. The Organisation for Economic Co-operation and Development Nuclear Energy Agency has been developing an international benchmark of the light water reactor (LWR) uncertainty analysis in modeling (UAM) for the examination of uncertainty quantification and propagation methodologies with various modeling and simulation code systems. The objective of the present work is to develop an uncertainty propagation mechanism based on the stochastic sampling method by taking into account the uncertainties of both basic nuclear data and fuel modeling parameters in the simulation of pressurized water reactors (PWRs) that can be incorporated in the conventional LWR simulation approach. More specifically, the Three Mile Island Unit 1–related exercises from the LWR-UAM benchmark have been modeled using the coupled TRACE/PARCS code system in the three-dimensional core representation. The input uncertainties of the neutronics simulation include few-group cross sections and kinetics parameters generated using the Sampler/Polaris sequence of SCALE 6.2.1. Several heat transfer–related variables for the fuel modeling were considered as sources of input uncertainty of the thermal-hydraulics simulations, including the thermal conductivity of fuel and cladding, fuel heat capacity, and the gap conductance. Dakota was used to sample input parameters of the coupled code system and to perform the uncertainty analysis. Two types of simulations were conducted: steady-state calculation at hot full-power condition and transient scenario initiated by the spatially asymmetric rod ejection accident. Quantities of interest for the steady-state calculation, including core multiplication factor and power peaking factors, were calculated with associated uncertainties. For transient calculations, best-estimate plus uncertainty results of the time evolution of core reactivity, core power, and peak fuel temperature were generated and analyzed. The Wilks’ formula was used to determine the necessary sample size to achieve a 95% confidence of the 95% limit of output quantities of interest. Although the uncertainty propagation and quantification method presented in this paper was developed for PWRs, it could be in general applicable to the multiphysics uncertainty quantification of other types of LWR cores.

SLOWPOKE-2 alternative core loading configurations analysis for highly improved reactor performance

A thorough analysis was carried out and is presented here exploring the possibilities for improving the performance of the LEU-fueled SLOWPOKE-2 reactor design, mainly with regard to the fuel pins distribution throughout the core. The chosen performance metric to maximize was the thermal neutron flux at the inner irradiation sites, while the essential constraint was to maintain the same Onset of Nucleate Boiling Ratio (ONBR) at nominal power, as safety criterion. MCNP5 and PLTEMP/ANL codes were used for the neutronics and thermal hydraulics simulations respectively, to analyze the reactor behavior under almost 300 different core configurations or fuel loading patterns. The main proposed configuration, comprising 270 fuel pins and minor modifications to the fuel cage, is predicted to allow a performance improvement of about 75% while maintaining the same safety margins. Additional or alternative, more complex modifications of the reactor for further performance increases, involving changes to the fuel, the annular reflector and the control rod, are also suggested and analyzed in some detail. The basis to reduce or recover safety margin to reach higher power and flux levels, with the potential to at least double the current performance when combined with the suggested modifications, is also briefly examined. Possible effects of the higher reactor power on the reactor operation are considered as well, providing potential modifications to extend the maximum continuous operation time and the period between Be shim adjustments. Overall, this work provides a detailed basis for substantial potential performance improvements for SLOWPOKE-2 reactors that are pending conversion to LEU and those that may need refueling in the future. The work also points out potential performance improvements on MNSR reactors.

Reliability assessment of passive residual heat removal system of IPWR using Kriging regression model

Passive safety systems have been incorporated into new design nuclear power plants to enhance the power plants’ inherent safety. However, due to the lack of the experimental data and operational experience, evaluating the reliability of passive systems is a challenging task. Many uncertainties are involved in the thermal hydraulic process during startup and operation. This may make the system unable to accomplish its expected function even though the hardware is available, known as functional failure. The evaluation of system functional failure probabilities based on the direct Monte Carlo method may be computationally impractical. In order to reduce the computational load, the Kriging regression model was constructed to avoid a large number of thermal hydraulics simulations. Compared with other metamodels, the Kriging regression model obtained better accuracy. The proposed method was applied to evaluate the reliability of a passive residual heat removal system of IPWR200 during a station blackout accident. The Kriging regression model not only ensures the accuracy of the reliability assessment, but also greatly reduces the runs of T-H codes. Furthermore, global sensitivity analysis aimed at determining the contributions of input parameters was carried out. Results indicate that the probability of passive residual heat removal system failing functionally is estimated to be 1.94e-4 and that insufficient decay heat removal is the most likely failure mode. Sensitivity analysis results identified five key uncertain parameters, which is significant for system performance. The sensitivity information can provide guidance for enhancing passive safety systems design and operation, which is of great importance to the safety and reliability of passive safety systems.

Thermal hydraulic simulations of a PWR nuclear power plant

Thermal hydraulic simulations of a PWR nuclear power plant

Thermal hydraulic simulations of a PWR nuclear power plant

This research presents simulation analyses of steady state, station blackout, and loss of coolant accident thermal hydraulic conditions. The plant is a PWR type with output of about 1000 MWe. The s...