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Littérature scientifique sur le sujet « Neutronics and thermal-hydraulics coupling »

Auteur : Grafiati

Publié le 9 mars 2023

Mis à jour le 10 mars 2023

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Articles de revues sur le sujet "Neutronics and thermal-hydraulics coupling"

Królikowski, Igor P., et Jerzy Cetnar. « Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent ». Nukleonika 60, n^o 3 (1 septembre 2015) : 531–36. http://dx.doi.org/10.1515/nuka-2015-0097.

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Abstract Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD) code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

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Blanco, J. A., P. Rubiolo et E. Dumonteil. « NEUTRONIC MODELING STRATEGIES FOR A LIQUID FUEL TRANSIENT CALCULATION ». EPJ Web of Conferences 247 (2021) : 06013. http://dx.doi.org/10.1051/epjconf/202124706013.

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Framework • A detailed and highly flexible numerical tool to study criticality accidents has been developed • The tool implements a Multi-Physics coupling using neutronics, thermal-hydraulics and thermal-mechanics models based on Open FOAM and SERPENT codes • Two neutronics models: Quasi-Static Monte Carlo and SPN Objective: In this work a system composed by a 2D square liquid fuel cavity filled with a fuel molten salt has been used to: • Investigate the performance of the tool’s thermal-hydraulics and neutronics solvers coupling numerical scheme • Evaluate possible strategies for the implementation of the Quasi-Static (QS) method with the Monte Carlo (MC) neutronics code • Compare the QS-MC approach precision and computational cost against the Simplified P3 (SP3) method

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Tollit, Brendan, Alan Charles, William Poole, Andrew Cox, Glynn Hosking, Ben Lindley, Peter Smith, Andy Smethurst et Jean Lavarenne. « WHOLE CORE COUPLING METHODOLOGIES WITHIN WIMS ». EPJ Web of Conferences 247 (2021) : 06006. http://dx.doi.org/10.1051/epjconf/202124706006.

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The ANSWERS® WIMS reactor physics code is being developed for whole core multiphysics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding and pin-by-pin differential depletion. An integrated thermal hydraulic solver permits differential temperature and density variations to feedback to the neutronics calculation. This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver. Two coupling routes are presented and compared using a challenging PWR SMR benchmark. The first route, called GEOM, dynamically calculates the resonance shielding and homogenisation with the whole core flux solution. The second coupling route, called CAMELOT, separates the resonance shielding and pincell homogenisation from the whole core solution via generating tabulated cross sections. Both routes can use the MERLIN homogenised pin-by-pin whole core flux solver and couple to the same integrated thermal hydraulic solver, called ARTHUR. Heterogeneous differences between the neutronics and thermal hydraulics are mapped via thermal identifiers for neutronics materials and thermal regions. The ability for the integrated thermal hydraulic solver to call an external code via a Fortran-C-Python (FCP) interface is also summarised. This flexible external coupling permits one way coupling to an external fuel performance code or two way coupling to an external thermal hydraulic code.

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Wu, Jianhui, Jingen Chen, Xiangzhou Cai, Chunyan Zou, Chenggang Yu, Yong Cui, Ao Zhang et Hongkai Zhao. « A Review of Molten Salt Reactor Multi-Physics Coupling Models and Development Prospects ». Energies 15, n^o 21 (6 novembre 2022) : 8296. http://dx.doi.org/10.3390/en15218296.

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Molten salt reactors (MSRs) are one type of GEN-IV advanced reactors that adopt melt mixtures of heavy metal elements and molten salt as both fuel and coolant. The liquid fuel allows MSRs to perform online refueling, reprocessing, and helium bubbling. The fuel utilization, safety, and economics can be enhanced, while some new physical mechanisms and phenomena emerge simultaneously, which would significantly complicate the numerical simulation of MSRs. The dual roles of molten fuel salt in the core lead to a tighter coupling of physical mechanisms since the released fission energy will be absorbed immediately by the molten salt itself and then transferred to the primary heat exchanger. The modeling of multi-physics coupling is regarded as one important aspect of MSR study, attracting growing attention worldwide. Up to now, great efforts have been made in the development of MSR multi-physics coupling models over the past 60 years, especially after 2000, when MSR was selected for one of the GEN-IV advanced reactors. In this paper, the development status of the MSR multi-physics coupling model is extensively reviewed in the light of coupling models of N-TH (neutronics and thermal hydraulics), N-TH-BN (neutronics, thermal hydraulics, and burnup) and N-TH-BN-G (neutronics, thermal hydraulics, burnup, and graphite deformation). The problems, challenges, and development trends are outlined to provide a basis for the future development of MSR multi-physics coupling models.

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Ta, Duy Long, Huy Hiep Nguyen, Tuan Khai Nguyen, Vinh Thanh Tran et Huu Tiep Nguyen. « Coulped neutronics/thermal-hydraulics calculation of VVER-1000 fuel assembly ». Nuclear Science and Technology 6, n^o 2 (24 septembre 2021) : 31–38. http://dx.doi.org/10.53747/jnst.v6i2.153.

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This paper presents a computational scheme using MCNP5 and COBRA-EN for coupling neutronics/thermal hydraulics calculation of a VVER-1000 fuel assembly. A master program was written using the PERL script language to build the corresponding inputs for the MCNP5 and COBRA-EN calculations and to manage the coupling scheme. The hexagonal coolant channels have been used in the thermal hydraulics model using CORBRA-EN to simplify the coupling scheme. The results of two successive iterations were compared with an assigned convergence criterion and the loop calculation can be broken when the convergence criterion is satisfied. Numerical calculation has been performed based on a UO2fuel assembly of the VVER-1000 reactor.

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Price, Dean, Majdi I. Radaideh, Travis Mui, Mihir Katare et Tomasz Kozlowski. « Multiphysics Modeling and Validation of Spent Fuel Isotopics Using Coupled Neutronics/Thermal-Hydraulics Simulations ». Science and Technology of Nuclear Installations 2020 (26 juillet 2020) : 1–14. http://dx.doi.org/10.1155/2020/2764634.

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Multiphysics coupling of neutronics/thermal-hydraulics models is essential for accurate modeling of nuclear reactor systems with physics feedback. In this work, SCALE/TRACE coupling is used for neutronic analysis and spent fuel validation of BWR assemblies, which have strong coolant feedback. 3D axial power profiles with coolant feedback are captured in these advanced simulations. The methodology is applied to two BWR assemblies (2F2DN23/SF98 and 2F2D1/F6), discharged from the Fukushima Daini-2 unit. Coupling is performed externally, where the SCALE/T5-DEPL module transfers axial power data in all axial nodes to TRACE, which in turn calculates the coolant density and temperature for each of these nodes. Within a burnup step, the data exchange process is repeated until convergence of all coupling parameters (axial power, coolant density, and coolant temperature) is observed. Analysis of axial power, criticality, and coolant properties at the assembly level is used to verify the coupling process. The 2F2D1/F6 benchmark seems to have insignificant void feedback compared to 2F2DN23/SF98 case, which experiences large power changes during operation. Spent fuel isotopic data are used to validate the coupling methodology, which demonstrated good results for uranium isotopes and satisfactory results for other actinides. This work has a major challenge of lack of documented data to build the coupled models (boundary conditions, control rod history, spatial location in the core, etc.), which encourages more advanced methods to approximate such missing data to achieve better modeling and simulation results.

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Ma, Yugao, Jinkun Min, Jin Li, Shichang Liu, Minyun Liu, Xiaotong Shang, Ganglin Yu, Shanfang Huang, Hongxing Yu et Kan Wang. « Neutronics and thermal-hydraulics coupling analysis in accelerator-driven subcritical system ». Progress in Nuclear Energy 122 (avril 2020) : 103235. http://dx.doi.org/10.1016/j.pnucene.2019.103235.

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Zhang, Dalin, Limin Liu, Minghao Liu, Rongshuan Xu, Cheng Gong et Suizheng Qiu. « Neutronics/Thermal-hydraulics Coupling Analysis for the Liquid-Fuel MOSART Concept ». Energy Procedia 127 (septembre 2017) : 343–51. http://dx.doi.org/10.1016/j.egypro.2017.08.075.

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Pascal, V., Y. Gorsse, N. Alpy, K. Ammar, M. Anderhuber, AM Baudron, G. Campioni et al. « MULTIPHYSICS MODELISATION OF AN UNPROTECTED LOSS OF FLOW TRANSIENT IN A SODIUM COOLED FAST REACTORS USING A NEUTRONIC-THERMAL-HYDRAULIC COUPLING SCHEME ». EPJ Web of Conferences 247 (2021) : 07001. http://dx.doi.org/10.1051/epjconf/202124707001.

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Sodium cooled fast neutron reactors (SFR) are one of the selected reactor concepts in the framework of the Generation IV International Forum. In this concept, unprotected loss of cooling flow transients (ULOF), for which the non-triggering of backup systems is postulated, are regarded as potential initiators of core melting accidents. During an ULOF transient, spatial distributions of fuel, structure and sodium temperatures are affected by the core cooling flow decrease, which will modify the spatial and energy distribution of neutron in the core due to the spatial competition of neutron feedback effects. As no backup systems are triggered, sodium may reach its boiling temperature at some point, leading to local sodium density variations and making the transient fluctuate in a two-phase flow physics where thermal-hydraulics and neutronics may interact with each other. The transient phenomenology involves several physic disciplines at different time and spatial scales, such as core neutronics, coolant thermal-hydraulics and fuel thermo-mechanics. This paper presents the results of thermal-hydraulic/neutronic coupled simulations of an ULOF transient on the SFR project ASTRID. These coupled calculations are based on the supervisor platform SALOME to link the neutron code APOLLO3® to the system thermal-hydraulic code CATHARE3. The physical approach used by the coupling to describe the neutron kinetic is a quasi-static adiabatic one, updating the normalized spatial power distribution periodically by performing static neutron calculations, while a point kinetic model associated to a neutron feedback model calculates the power amplitude variations.

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Yang, Ping, Liangzhi Cao, Hongchun Wu et Changhui Wang. « Core design study on CANDU-SCWR with 3D neutronics/thermal-hydraulics coupling ». Nuclear Engineering and Design 241, n^o 12 (décembre 2011) : 4714–19. http://dx.doi.org/10.1016/j.nucengdes.2011.03.036.

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Thèses sur le sujet "Neutronics and thermal-hydraulics coupling"

Guyot, Maxime. « Neutronics and thermal-hydraulics coupling : some contributions toward an improved methodology to simulate the initiating phase of a severe accident in a sodium fast reactor ». Thesis, Aix-Marseille, 2014. http://www.theses.fr/2014AIXM4345.

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Le sujet de la thèse s'inscrit dans le cadre de la rénovation des outils et des méthodes de calculs appliqués aux accidents graves des Réacteurs à Neutrons Rapides refroidis au Sodium (RNR-Na). En particulier, on s'intéresse aux biais et conservatismes liés à la méthodologie de calculs de la phase primaire d'un accident grave. Pour évaluer les conséquences d'un accident de fusion du coeur d'un RNR-Na, une approche déterministe est généralement réalisée en considérant des hypothèses dites "best-estimate". Cette approche repose sur l'utilisation de codes informatiques pour simuler numériquement le comportement du coeur en conditions accidentelles.La phase primaire de dégradation concerne les évènements se produisant tant que les boîtiers inter-assemblages sont intègres. Les assemblages combustibles conservent alors une indépendance les uns par rapport aux autres. Pour cette raison, la simulation de la phase primaire repose sur une approche multi-canaux. Cette approche consiste à regrouper les assemblages semblables en classes d'assemblages appelés canaux. Le modèle thermo-hydraulique en canaux est couplé à un calcul neutronique pour évaluer le niveau de puissance et de réactivité au cours du transitoire accidentel. La méthodologie de calcul de la phase primaire d'un accident grave repose sur des hypothèses fortes en termes de modélisation neutronique et thermo-hydraulique. Après avoir identifié les principales sources d'erreur, la thèse a consisté à développer un nouvel outil de calcul pour la phase primaire en vue d'évaluer les biais et conservatismes méthodologiques
This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and recriticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios.During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. Inthe multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level.In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling

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Faucher, Margaux. « Coupling between Monte Carlo neutron transport and thermal-hydraulics for the simulation of transients due to reactivity insertions ». Thesis, Université Paris-Saclay (ComUE), 2019. http://www.theses.fr/2019SACLS387/document.

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Dans le contexte de la physique des réacteurs, l’analyse du comportement non stationnaire de la population neutronique avec contre-réactions dans le combustible et dans le modérateur se rend indispensable afin de caractériser les transitoires opérationnels et accidentels dans les systèmes nucléaires et d’en améliorer par conséquent la sûreté. Pour ces configurations non stationnaires, le développement de méthodes Monte-Carlo qui prennent en compte la dépendance en temps du système neutronique, mais aussi le couplage avec les autres physiques, comme la thermohydraulique et la thermomécanique, a pour but de servir de référence aux calculs déterministes.Ce travail de thèse a consisté à mettre en place une chaîne de calcul pour la simulation couplée neutronique Monte-Carlo, avec le code TRIPOLI-4, en conditions non stationnaires et avec prise en compte des contre-réactions thermohydrauliques.Nous avons d'abord considéré les méthodes cinétiques dans TRIPOLI-4, c'est-à-dire avec prise en compte du temps mais sans prise en compte des contre-réactions, en incluant une évaluation des méthodes existantes ainsi que le développement de nouvelles méthodes. Ensuite, nous avons développé un schéma de couplage entre TRIPOLI-4 et le code de thermohydraulique sous-canal SUBCHANFLOW. Enfin, nous avons réalisé une analyse préliminaire de la propagation des incertitudes au sein du calcul couplé sur un modèle simplifié. En effet, les fluctuations statistiques sont inhérentes à notre schéma de par la nature stochastique de TRIPOLI-4. De plus, les équations de la thermohydraulique étant non-linéaires, la propagation des incertitudes au long du calcul doit être étudiée afin de caractériser la convergence du résultat
One of the main issues for the study of a reactor behaviour is to model the propagation of the neutrons, described by the Boltzmann transport equation, in the presence of multi-physics phenomena, such as the coupling between neutron transport, thermal-hydraulics and thermomecanics. Thanks to the growing computer power, it is now feasible to apply Monte Carlo methods to the solution of non-stationary transport problems in reactor physics, which play an instrumental role in producing reference numerical solutions for the analysis of transients occurring during normal and accidental behaviour.The goal of this Ph. D. thesis is to develop, verify and test a coupling scheme between the Monte Carlo code TRIPOLI-4 and thermal-hydraulics, so as to provide a reference tool for the simulation of reactivity-induced transients in PWRs.We have first tested the kinetic capabilities of TRIPOLI-4 (i.e., time dependent without thermal-hydraulics feedback), evaluating the different existing methods and implementing new techniques. Then, we have developed a multi-physics interface for TRIPOLI-4, and more specifically a coupling scheme between TRIPOLI-4 and the thermal-hydraulics sub-channel code SUBCHANFLOW. Finally, we have performed a preliminary analysis of the stability of the coupling scheme. Indeed, due to the stochastic nature of the outputs produced by TRIPOLI-4, uncertainties are inherent to our coupling scheme and propagate along the coupling iterations. Moreover, thermal-hydraulics equations are non linear, so the prediction of the propagation of the uncertainties is not straightforward

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CHIESA, DAVIDE. « Development and experimental validation of a Monte Carlo simulation model for the Triga Mark II reactor ». Doctoral thesis, Università degli Studi di Milano-Bicocca, 2014. http://hdl.handle.net/10281/50064.

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In recent years, many computer codes, based on Monte Carlo methods or deterministic calculations, have been developed to separately analyze different aspects regarding nuclear reactors. Nuclear reactors are very complex systems, which require an integrated analysis of all the variables which are intrinsically correlated: neutron fluxes, reaction rates, neutron moderation and absorption, thermal and power distributions, heat generation and transfer, criticality coefficients, fuel burnup, etc. For this reason, one of the main challenges in the analysis of nuclear reactors is the coupling of neutronics and thermal-hydraulics simulation codes, with the purpose of achieving a good modeling and comprehension of the mechanisms which rule the transient phases and the dynamic behavior of the reactor. This is very important to guarantee the control of the chain reaction, for a safe operation of the reactor. In developing simulation tools, benchmark analyses are needed to prove the reliability of the simulations. The experimental measurements conceived to be compared with the results coming out from the simulations are really precious and can provide useful information to improve the description of the physics phenomena in the simulation models. My PhD research activity was held in this framework, as part of the research project Analysis of Reactor COre (ARCO, promoted by INFN) whose task was the development of modern, flexible and integrated tools for the analysis of nuclear reactors, relying on the experimental data collected at the research reactor TRIGA Mark II, installed at the Applied Nuclear Energy Laboratory (LENA) at the University of Pavia. In this way, once the effectiveness and the reliability of these tools for modeling an experimental reactor have been demonstrated, these could be applied to develop new generation systems. In this thesis, I present the complete neutronic characterization of the TRIGA Mark II reactor, which was analyzed in different operating conditions through experimental measurements and the development of a Monte Carlo simulation tool (relied on the MCNP code) able to take into account the ever increasing complexity of the conditions to be simulated. First of all, after giving an overview of some theoretical concepts which are fundamental for the nuclear reactor analysis, a model that reconstructs the first working period of the TRIGA Mark II reactor, in which the “fresh” fuel was not heavily contaminated with fission reaction products, is described. In particular, all the geometries and the materials are described in the MCNP simulation model with good detail, in order to reconstruct the reactor criticality and all the effects on the neutron distributions. The very good results obtained from the simulations of the reactor at low power condition -in which the fuel elements can be considered to be in thermal equilibrium with the water around them- are then used to implement a model for simulating the full power condition (250kW), in which the effects arising from the temperature increase in the fuel-moderator must be taken into account. The MCNP simulation model was exploited to evaluate the reactor power distribution and a dedicated experimental campaign was performed to measure the water temperature within the reactor core. In this way, through a thermal-hydraulic calculation tool, it has been possible to determine the temperature distribution within the fuel elements and to include the description of the thermal effects in the MCNP simulation model. Thereafter, since the neutron flux is a crucial parameter affecting the reaction rates and thus the fuel burnup, its energy and space distributions are analyzed presenting the results of several neutron activation measurements. Particularly, the neutron flux was firstly measured in the reactor's irradiation facilities through the neutron activation of many different isotopes. Hence, in order to analyze the energy flux spectra, I implemented an analysis tool, based on Bayesian statistics, which allows to combine the experimental data from the different activated isotopes and reconstruct a multi-group flux spectrum. Subsequently, the spatial neutron flux distribution within the core was measured by activating several aluminum-cobalt samples in different core positions, thus allowing the determination of the integral and fast flux distributions from the analysis of cobalt and aluminum, respectively. Finally, I present the results of the fuel burnup calculations, that were performed for simulating the current core configuration after a 48 years-long operation. The good accuracy that was reached in the simulation of the neutron fluxes, as confirmed by the experimental measurements, has allowed to evaluate the burnup of each fuel element from the knowledge of the operating hours and the different positions occupied in the core over the years. In this way, it has been possible to exploit the MCNP simulation model to determine a new optimized core configuration which could ensure, at the same time, a higher reactivity and the use of less fuel elements. This configuration was realized in September 2013 and the experimental results confirm the high quality of the work done. The results of this Ph.D. thesis highlight that it is possible to implement analysis tools -ranging from Monte Carlo simulations to the fuel burnup time evolution software, from neutron activation measurements to the Bayesian statistical analysis of flux spectra, and from temperature measurements to thermal-hydraulic models-, which can be appropriately exploited to describe and comprehend the complex mechanisms ruling the operation of a nuclear reactor. Particularly, it was demonstrated the effectiveness and the reliability of these tools in the case of an experimental reactor, where it was possible to collect many precious data to perform benchmark analyses. Therefore, for as these tools have been developed and implemented, they can be used to analyze other reactors and, possibly, to project and develop new generation systems, which will allow to decrease the production of high-level nuclear waste and to exploit the nuclear fuel with improved efficiency.

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Waata, Christine Lylin. « Coupled neutronics, thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly ». Karlsruhe : FZKA, 2006. http://bibliothek.fzk.de/zb/berichte/FZKA7233.pdf.

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Waata, Christine Lylin. « Coupled neutronics, thermal hydraulics analysis of a high-performance light water reactor fuel assembly ». Karlsruhe FZKA, 2005. http://bibliothek.fzk.de/zb/berichte/FZKA7233.pdf.

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Waata, Christine Lylin [Verfasser], et Eckart [Akademischer Betreuer] Laurien. « Coupled neutronics thermal hydraulics analysis of a high-performance light-water reactor fuel assembly / Christine Lylin Waata. Betreuer : Eckart Laurien ». Stuttgart : Universitätsbibliothek der Universität Stuttgart, 2006. http://d-nb.info/1081642378/34.

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Waata, Christine Lylin [Verfasser]. « Coupled neutronics, thermal hydraulics analysis of a high-performance light water reactor fuel assembly / Kernforschungszentrum Karlsruhe GmbH, Karlsruhe. Christine Lylin Waata ». Karlsruhe : FZKA, 2006. http://d-nb.info/982286341/34.

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Basualdo, Perelló Joaquín Rubén [Verfasser], et R. [Akademischer Betreuer] Stieglitz. « Development of a Coupled Neutronics/Thermal-Hydraulics/Fuel Thermo-Mechanics Multiphysics Tool for Best-Estimate PWR Core Simulations / Joaquín Rubén Basualdo Perelló ; Betreuer : R. Stieglitz ». Karlsruhe : KIT-Bibliothek, 2020. http://d-nb.info/1220359068/34.

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Silva, Rodney Aparecido Busquim e. « Implications of advanced computational methods for reactivity initiated accidents in nuclear reactors ». Universidade de São Paulo, 2015. http://www.teses.usp.br/teses/disponiveis/3/3139/tde-20072016-142605/.

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Advanced computational tools are applied to simulate a nuclear power plant (NPP) control rod assembly ejection (CRE) accident. The impact of these reactivity-initiated accidents (RIAs) on core reactivity behavior, 3D power distribution and stochastic reactivity estimation are evaluated. The three tools used are: the thermal-hydraulic (TH) RELAP5 (R5) code, the neutronic (NK) PARCS (P3D) code, and the coupled version P3D/R5, with specially developed linkage using the environment code MATLAB. This study considers three different-size cores: NPP1 (2772 MWt); NPP2 (530 MWt); and NPP3 (1061 MWt). The three cores have the same general design and control rod assembly (CRA) positions, and the ejected CRA has similar worth and at the same rod ejection pace. The CRE is assessed under both hot zero power (HZP) and hot full power (HFP) conditions. The analyses indicate that RIA modeling and simulation should be carried out through a systematic coding and configuration approaches, otherwise the results will not capture the true transient behavior of the core under analysis. The simulation of one code depends on the appropriate configuration of parameters generated by the other code and on the correct determination of the TH/NK mapping weight factors for the various mesh regions in each of the models. From the design point of view, the standalone codes predict milder magnitude of power and reactivity increase compared to the coupled P3D/R5 simulation. The magnitudes of reduced peak power and reactivity become larger as the core size shrinks. The HFP simulation shows that the three NPPs have the same transient peak value, but the post-transient steady power is lower for a smaller core. The HZP analysis indicates that the transient peak is lower for the smaller core, but the post-transient power occurs at the same level. The three-dimensional (3D) power distributions are different among the HFP and HZP cases, but do not depend on the size of the core. The results indicate: i) HFP: core power increases in the area surrounding the ejected rod/bank assembly, and this increase becomes lower as the NPPs shrinks however, the power is well-distributed after the transient; and ii) HZP: the area surrounding the CRA stays hotter, but the 3D peak assembly factor becomes lower, during and after the transients, as the NPPs shrinks. These features confirm that the smaller cores yield a safer response to a given inserted reactivity compared to larger cores. A stochastic extended Kalman filter (EKF) algorithm is implemented to estimate the reactivity based on the reactor power profile, after the addition of random noise. The inverse point kinetics (IPK) deterministic method is also implemented and the results of the application of EKF and IPK are compared to the P3D/R5 simulation. The following sophisticated strategies made the EKF algorithm robust and accurate: the system is modeled by a set of continuous time nonlinear stochastic differential equations; the code uses a time step directly based on the power measured and applies that to the model for online discretization and linearization; filter tuning goes automatically up from the first time step; and the state noise covariance matrix is updated online at each time step. It was found that the IPK reactivity has higher noise content compared to the EKF reactivity for all cases. Thus, the EKF presents superior and more accurate results. Furthermore, under a small reactivity insertion, the IPK reactivity varies widely from positive to negative values: this variation is not observed within the EKF. A sensitivity analysis for three distinct standard deviation (SD) noise measurements suggests that EKF is superior to IPK method, independent of the noise load magnitude. As the noise content increases, the error between the IPK and P3D/R5 reactivity also increases. A sensitivity analysis for five distinct carry-over effects of different random noise loads indicates that the random addition of different noise loads to the reactor power does not change the overall performance of both algorithms.
Este trabalho aplica métodos computacionais avançados para simular a ejeção de barras de controle (CRE) em uma planta térmica nuclear (NPP). São avaliados o impacto da ocorrência de acidentes iniciados por reatividade (RIAs) na reatividade total, na distribuição da potência em três dimensões (3D) e na determinação da reatividade. As ferramentas utilizadas são: o código termo-hidráulico (TH) RELAP5 (R5), o código neutrônico (NK) PARCS (P3D), a versão acoplada P3D/R5, e o ambiente computacional MATLAB. Este estudo considera três reatores nucleares de diferentes tamanhos: NPP1 (2772 MWT); NPP2 (530 MWt); e NPP3 (1061 MWt). Os três núcleos possuem projeto similar e idêntica posição dos grupos das barras de controle (CRA), além do mesmo valor de reatividade diferencial das CRA ejetadas e idêntica velocidade de ejeção. A ocorrência da CRE é avaliada sob condições de hot zero power (HZP) e de hot full power (HFP). As análises indicam que a modelagem e a simulação de RIAs devem ser realizadas sistematicamente, caso contrário os resultados não irão refletir o comportamento em regime transitório do núcleo. A simulação de um modelo em um código depende da apropriada configuração de parâmetros gerados pelo outro código e da determinação adequada do mapeamento TH/NK para as várias malhas dos modelos. Do ponto de vista de projeto, a utilização de códigos independentes resulta em cálculos de potência e reatividade conservadores em comparação com os resultados utilizando-se P3D/R5. Os picos de potência e de reatividade são menores à medida que o núcleo encolhe. A simulação em condições de HFP resulta em valores de pico de potência similares durante transitório para as três NPPs, mas a potência de pós-transitórios é menor para o menor núcleo. A análise em condições de HZP também indica que o valor máximo durante o transitório é menor para o menor núcleo, mas o pós-transitórios ocorre aos mesmos níveis de potência das demais NPPS. A distribuição de potência em 3D também apresenta resultados distintos para condições de HFP e HZP, mas tais resultados são independentes do tamanho do núcleo: i) HFP: há um aumento da potência do núcleo em torno da CRE, mas tal comportamento diminui para núcleos menores - no entanto, a potência é bem distribuída após o transitório; e ii) HZP: há aumento de potência na área do CRE, mas o pico de potência em 3D é menor durante e depois dos transitórios para núcleos menores. Tais características indicam que os núcleos menores respondem de forma mais segura quando da inserção de reatividade em comparação a reatores de maiores dimensões. O método estocástico de filtragem de Kalman estendido (EKF) foi codificado para estimar a reatividade com base no perfil de potência da NPP, após a adição de ruído aleatório. O método determinístico da cinética pontual inversa (IPK) também foi implementado e os resultados da aplicação dos algoritmos do EKF e IPK foram comparados com os resultados da simulação do P3D/R5. As seguintes estratégias, implementadas neste trabalho, possibilitaram a aplicação robusta e precisa do EKF: o sistema foi modelado por um conjunto de equações diferenciais não-lineares estocásticas de tempo contínuo; o algoritmo obtém o passo de tempo diretamente da potência medida e aplica-o ao modelo para a discretização e linearização online; o ajuste do filtro ocorre automaticamente a partir do primeiro passo de tempo; e a matriz de covariância do ruído no estado é atualizada online. Verificou-se que a reatividade calculada pelo método IPK possui maior nível de ruído quando comparada ao EKF para todos os casos estudados. Portanto, o EKF apresenta resultados superiores e mais precisos. Além disso, sob uma pequena inserção de reatividade, a reatividade calculada pelo método IPK varia consideravelmente de valores positivos para negativos: esta variação não é observada com o EKF. Uma análise de sensibilidade para três desvios padrão (SD) sugere que o algoritmo EKF é superior ao método IPK, independente da magnitude do ruído. Com o aumento da magnitude do ruído, o erro entre as reatividades calculadas pelo IPK e pelo P3D/R5 aumenta. A análise de sensibilidade para cinco ruídos aleatórios indica que a adição de ruído na potência do reator não altera o desempenho global de ambos os algoritmos.

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Alzaben, Yousef Ibrahim [Verfasser], et R. [Akademischer Betreuer] Stieglitz. « Neutronics and Thermal-Hydraulics Safety Related Investigations of an Innovative Boron-Free Core Integrated Within a Generic Small Modular Reactor / Yousef Ibrahim Alzaben ; Betreuer : R. Stieglitz ». Karlsruhe : KIT-Bibliothek, 2019. http://d-nb.info/1199459127/34.

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Javadi, M. Neutronics and Thermal Hydraulics Feedback Models of the Harwell Materials Testing Reactors DIDO and PLUTO. AEA Technology Plc, 1986.

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Demazière, Christophe. Modelling of Nuclear Reactor Multiphysics : From Local Balance Equations to Macroscopic Models in Neutronics and Thermal-Hydraulics. Elsevier Science & Technology, 2019.

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Demazière, Christophe. Modelling of Nuclear Reactor Multi-Physics : From Local Balance Equations to Macroscopic Models in Neutronics and Thermal-Hydraulics. Elsevier Science & Technology Books, 2019.

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Chapitres de livres sur le sujet "Neutronics and thermal-hydraulics coupling"

Zhao, Chuanqi, Kunpeng Wang, Liangzhi Cao, Hongchun Wu et Youqi Zheng. « Coupled Neutronics and Thermal–Hydraulics Analysis of Annular Fuel Assembly for SCWR ». Dans Proceedings of The 20th Pacific Basin Nuclear Conference, 93–104. Singapore : Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2314-9_8.

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Rouault, Jacques, P. Chellapandi, Baldev Raj, Philippe Dufour, Christian Latge, Laurent Paret, Pierre Lo Pinto et al. « Sodium Fast Reactor Design : Fuels, Neutronics, Thermal-Hydraulics, Structural Mechanics and Safety ». Dans Handbook of Nuclear Engineering, 2321–710. Boston, MA : Springer US, 2010. http://dx.doi.org/10.1007/978-0-387-98149-9_21.

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Cinotti, Luciano, Craig F. Smith, Carlo Artioli, Giacomo Grasso et Giovanni Corsini. « Lead-Cooled Fast Reactor (LFR) Design : Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design ». Dans Handbook of Nuclear Engineering, 2749–840. Boston, MA : Springer US, 2010. http://dx.doi.org/10.1007/978-0-387-98149-9_23.

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Demazière, Christophe. « Neutronic/thermal-hydraulic coupling ». Dans Modelling of Nuclear Reactor Multi-physics, 311–36. Elsevier, 2020. http://dx.doi.org/10.1016/b978-0-12-815069-6.00006-4.

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H. Khalafi, Farshad Faghihi, et S. M. « A Literature Survey of Neutronics and Thermal-Hydraulics Codes for Investigating Reactor Core Parameters ; Artificial Neural Networks as the VVER-1000 Core Predictor ». Dans Nuclear Power - System Simulations and Operation. InTech, 2011. http://dx.doi.org/10.5772/16521.

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Suikkanen, H., J. Ritvanen, P. Jalali et R. Kyrki-Rajamäki. « Modeling Packing of Spherical Fuel Elements in Pebble Bed Reactors Using DEM ». Dans Discrete Element Modelling of Particulate Media, 175–83. The Royal Society of Chemistry, 2012. http://dx.doi.org/10.1039/bk9781849733601-00175.

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Pebble bed reactor is a nuclear reactor type where the fuel material is in the form of coated particles dispersed inside graphite spheres of approximately 60 mm in diameter. Typical reactor contains about half a million of these pebbles that form a heat-producing column. Pebbles are fed on top and removed from the bottom of the column so that they flow through the core slowly. The column is cooled by helium gas flowing through the pebble bed. Knowledge of the packing characteristics and behaviour of the fuel pebbles is important for resolving the core neutronics, burnup and thermal-hydraulics accurately. DEM can be used to model the mechanical interactions between pebbles accurately. The results can then be used in further analyses, e.g. as detailed pebble positions or packing density profiles. In this work, an in-house DEM code was used to compact 450 000 pebbles inside an annular cylinder representing the core of a pebble bed reactor. Three pebble beds with different average packing densities were obtained by varying input parameters. Pebble-scale packing fraction data was extracted from the packed beds using Voronoi tessellation. The data was used to plot packing density profiles for different parts of the columns. The local packing fractions were also used for analyses with statistical and 3D visualization methods to reveal packing structure details. Local dense regions with crystallization were observed especially when the average packing density was high. The results are useful in further analyses, e.g. to identify hot spots in the reactor core.

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Actes de conférences sur le sujet "Neutronics and thermal-hydraulics coupling"

Ge, Jian, Dalin Zhang, Wenxi Tian, Suizheng Qiu et G. H. Su. « Coupled Analysis of Thermal Hydraulics and Neutronics for a Molten Salt Reactor ». Dans 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67042.

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As one of the six selected optional innovative nuclear reactor in the generation IV International Forum (GIF), the Molten Salt Reactor (MSR) adopts liquid salt as nuclear fuel and coolant, which makes the characteristics of thermal hydraulics and neutronics strongly intertwined. Coupling analysis of neutronics and thermal hydraulics has received considerable attention in recent years. In this paper, a new coupling method is introduced based on the Finite Volume Method (FVM), which is widely used in the Computational Fluid Dynamics (CFD) methodology. Neutron diffusion equations and delayed neutron precursors balance equations are discretized and solved by the commercial CFD package FLUENT, along with continuity, momentum and energy equations simultaneously. A Temporal And Spatial Neutronics Analysis Model (TASNAM) is developed using the User Defined Functions (UDF) and User Defined Scalar (UDS) in FLUENT. A neutronics benchmark is adopted to demonstrate the solution capability for neutronics problems using the method above. Furthermore, a steady state coupled analysis of neutronics and thermal hydraulics for the Molten Salt Advanced Reactor Transmuter (MOSART) is performed. Two groups of neutrons and six groups of delayed neutron precursors are adopted. Distributions of the liquid salt velocity, temperature, neutron flux and delayed neutron precursors in the core are obtained and analyzed. This work can provide some valuable information for the design and research of MSRs.

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Akbas, Sabahattin, Victor Martinez-Quiroga, Fatih Aydogan, Abderrafi M. Ougouag et Chris Allison. « Survey of Coupling Schemes in Traditional Coupled Neutronics and Thermal-Hydraulics Codes ». Dans ASME 2015 International Mechanical Engineering Congress and Exposition. American Society of Mechanical Engineers, 2015. http://dx.doi.org/10.1115/imece2015-52990.

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The design and the analysis of nuclear power plants (NPPs) require computational codes to predict the behavior of the NPP nuclear components and other systems (i.e., reactor core, primary coolant system, emergency core cooling system, etc.). Coupled calculations are essential to the conduct of deterministic safety assessments. Inasmuch as the physical phenomena that govern the performance of a nuclear reactor are always present simultaneously, ideally computational modeling of a nuclear reactor should include coupled codes that represent all of the active physical phenomena. Such multi-physics codes are under development at several institutions and are expected to become operational in the future. However, in the interim, integrated codes that incorporate modeling capabilities for two to three physical phenomena will remain useful. For example, in the conduct of safety analyses, of paramount importance are codes that couple neutronics and thermal-hydraulics, especially transient codes. Other code systems of importance to safety analyses are those that couple primary system thermal-hydraulics to fission product chemistry, neutronics to fuel performance, containment behavior and structural mechanics to thermal-hydraulics, etc. This paper surveys the methods used traditionally in the coupling of neutronic and thermal-hydraulics codes. The neutron kinetics codes are used for computing the space-time evolution of the neutron flux and, hence, of the power distribution. The thermal-hydraulics codes, which compute mass, momentum and energy transfers, model the coolant flow and the temperature distribution. These codes can be used to compute the neutronic behavior and the thermal-hydraulic states separately. However, the need to account with fidelity for the dynamic feedback between the two sets of properties (via temperature and density effects on the cross section inputs into the neutronics codes) and the requirement to model realistically the transient response of nuclear power plants and to assess the associated emergency systems and procedures imply the necessity of modeling the neutronic and thermal-hydraulics simultaneously within a coupled code system. The focus of this paper is a comparison of the methods by which the coupling between neutron kinetics and thermal-hydraulics treatments has been traditionally achieved in various code systems. As discussed in the last section, the modern approaches to multi-physics code development are beyond the scope of this paper. From the field of the most commonly used coupled neutron kinetic-thermal-hydraulics codes, this study selected for comparison the coupled codes RELAP5-3D (NESTLE), TRACE/PARCS, RELAP5/PARCS, ATHLET/DYN3D, RELAP5/SCDAPSIM/MOD4.0/NESTLE. The choice was inspired by how widespread the use of the codes is, but was limited by time availability. Thus, the selection of codes is not to be construed as exhaustive, nor is there any implication of priority about the methods used by the various codes. These codes were developed by a variety of institutions (universities, research centers, and laboratories) geographically located away from each other. Each of the research group that developed these coupled code systems used its own combination of initial codes as well as different methods and assumptions in the coupling process. For instance, all these neutron kinetics codes solve the few-groups neutron diffusion equations. However, the data they use may be based on different lattice physics codes. The neutronics solvers may use different methods, ranging from point kinetics method (in some versions of RELAP5) to nodal expansion methods (NEM), to semi-analytic nodal methods, to the analytic nodal method (ANM). Similarly, the thermal-hydraulics codes use several different approaches: different number of coolant fields, homogenous equilibrium model, separate flow model, different numbers of conservation equations, etc. Therefore, not only the physical models but also the assumptions of the coupled codes and coupling techniques vary significantly. This paper compares coupled codes qualitatively and quantitatively. The results of this study are being used both to guide the selection of appropriate coupled codes and to identify further developments into coupled codes.

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Zheng, Yong, Min-jun Peng, Geng-lei Xia et Ren Li. « Investigation on Coupling Behaviors of Thermal-Hydraulics/Neutronics Under Asymmetrical Inlet Conditions ». Dans 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30522.

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The reactor core is a complex system involving the reactor physics, thermal hydraulics and many other aspects. That means the distribution of the core power largely determines the profile of the thermal parameters, meanwhile the local thermal-hydraulics condition will in turn affect the neutronics calculation by moderator temperature effect and Doppler effects. Issues coupling the thermal-hydraulics with neutronics of nuclear plants still challenge the design, safety and the operation of LWR few years ago. Fortunately, the recent availability of powerful computer and computational techniques has enlarged the capabilities of making more realistic simulations of complex phenomena in NPPs. The current study deals with the development of an integrated thermal-hydraulics/neutronics model for Qinshan phase II NPP project reactor for the analysis of specific plant transients in which the neutronic response of the core is important, application of RELAP5-HD making use of the Helios code to derive the macroscopic cross-sections. Based on the coupled model, the steady state calculation and the transient simulation, involving the abnormal operation mode with asymmetrical coolant flux and temperature on the inlet of reactor, have been performed. The results show that the values obtained from coupled code RELAP5-HD calculation are in good agreement with the available experimental data, and the calculated accident parameters curves can predict all major trends of the transient. Steady state and transient condition calculation results are in accordance with the theoretical analysis from the aspect of coupled thermal-hydraulics/neutronics, this demonstrated a successful best estimate coupled RELAP5-HD model of Qinshan phase II NPP reactor has been developed, and the established model will provide a good foundation for the further analysis of the primary loop. It also can be concluded that the more accurate CFD method coupling three dimensional neutron kinetics code based on neutron diffusion method are necessary for steady-state calculation and analysis of transient/accident conditions when asymmetrical processes take place in the core. It is worth mentioning that RELAP5-HD code has already programmed the human-machine interface and the interface for coupling with other code, hence RELAP5-HD code has a broad application prospect in PWRs safety analysis.

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Martinez-Quiroga, Victor, Sabahattin Akbas, Fatih Aydogan, Abderrafi M. Ougouag et Chris Allison. « Coupling of RELAP5-SCDAP MOD4.0 and Neutronic Codes ». Dans ASME 2015 International Mechanical Engineering Congress and Exposition. American Society of Mechanical Engineers, 2015. http://dx.doi.org/10.1115/imece2015-52991.

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High-fidelity and accurate nuclear system codes play a key role in the design and analysis of complex nuclear power plants, which consist of multiple subsystems, such as the reactor core (and its fuel, burnable poisons, control elements, etc.), the reactor internal structures, the vessel, and the energy conversion subsystem and beyond to grid demand. Most commonly the interplay between these various subsystems is modeled using coupled codes, each of which represents one of the subsystems. And the most common direct coupling is that of thermal-hydraulics and neutronics codes. The subject of this paper is the coupling of codes that model not only thermal-hydraulics and neutronics, but also structural components damage. Furthermore, the neutronic component is not limited to the sole core solver. The coupled code system encompasses thermal-hydraulics, material performance of the fuel, neutronic solver, and neutronic data preparation. Thus, this paper presents a framework for coupling RELAP5/SCDAPSIM/MOD4.0 with a suite of neutron kinetics codes that includes NESTLE, DRAGON and a version of the ENDF library. The version of the RELAP5/SCDAPSIM/MOD4.0 code used in this work is one developed by Innovate System Software (ISS) as part of the international SCDAP Development and Training Program (SDTP) for best-estimate analysis to model reactor transients including severe accident phenomena. This RELAP5/SCDAPSIM/MOD4.0 code version is also capable of predicting nuclear fuel performance. It uses nodal power distributions to calculate mechanical and thermal parameters such as heat-up, oxidation and meltdown of fuel rods and control rods, the ballooning and rupture of fuel rod cladding, the release of fission products from fuel rods, and the disintegration of fuel rods into porous debris and molten material. On the neutronics side, this work uses the NESTLE and DRAGON codes. NESTLE is a multi-dimensional static and kinetic neutronic code developed at North Carolina State University. It solves up to four energy groups neutron diffusion equations utilizing the Nodal Expansion Method (NEM) in Cartesian or hexagonal geometry. The DRAGON code, developed at Ecole Polytechnique de Montreal, performs lattice physics calculations based on the neutron transport equation and is capable of using very fine energy group structures. In this work, we have developed a coupling approach to exchange data among the various modules. In the coupling process, the generated nuclear data (in fine multigroup energy structure) are collapsed down into two- or four-group energy structures for use in NESTLE. The neutron kinetics and thermal-hydraulics modules are coupled at each time step by using the cross-section data. The power distribution results of the neutronic calculations are transmitted to the thermal-hydraulics code. The spatial distribution of coolant density and the fuel-moderator temperature, which result from the thermal-hydraulic calculations, are transmitted back to the neutron kinetics codes and then the loop is closed using new neutronics results. Details of the actual data transfers will be described in the full length paper.

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Chen, Jun, Liangzhi Cao, Zhouyu Liu, Hongchun Wu et Yijun Zhang. « Preliminary Verification of the High-Fidelity Neutronics and Thermal-Hydraulics Coupling System NECP-X/SUBSC ». Dans 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66511.

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PWR core phenomena can be simulated and predicted more precisely and in more details with high-fidelity neutronics and thermal-hydraulics coupling calculations. An internal coupling between a newly developed high-fidelity neutronics code NECP-X and the sub-channel code SUBSC has been realized. In order to verify the NECP-X/SUBSC coupling system, another high-fidelity neutronics and thermal-hydraulics coupling system OpenMC/SUBSC was developed through external coupling method. Both coupling systems were applied to a simplified PWR 3×3 pin cluster case. The numerical result shows good agreement in both eigenvalue and normalized axial power distribution for a selected pin, demonstrating the success of the internal coupling of NECP-X and SUBSC.

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Guo, Zhangpeng, Yao Xiao, Jianjun Zhou, Dalin Zhang, Khurrum Saleem Chaudri et Suizheng Qiu. « Coupled Neutronics/Thermal-Hydraulics for Analysis of Molten Salt Reactor ». Dans 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15012.

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The Generation IV international Forum (GIF) selected molten salt reactor (MSR) among six advanced reactor types. It is characterized by a liquid circulating fuel that also serves as coolant. In this study, a multiple-channel analysis code (MAC) is developed and it is coupled with MCNP4c to analyze the neutronics/thermal-hydraulics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, such as temperature distribution, flow distribution and pressure drop. MCNP4c performs the analysis of effective multiplication factor, neutron flux and power distribution. A linkage code is developed to exchange data between MAC and MCNP to implement coupling iteration process until the power convergence is achieved. The coupling calculation can achieve converged solution after a few iterations. The results are in reasonable agreement with the analytic solutions from the ORNL. This work was helpful for further design analysis and operation of MSR.

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Xie, Qiuxia, Xiaojing Liu et Xiang Chai. « Three-Dimensional Fine-Mesh Coupled Neutronics and Thermal-Hydraulics Calculation for PWR Fuel Pins ». Dans 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-93140.

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Abstract With the great improvement of computer performance, multi-physics and high-fidelity reactor numerical simulation has attracted widespread attention. The development of a three-dimensional fine-mesh coupled neutronics and thermal-hydraulics procedure mainly using the free open source C++ library OpenFOAM is presented in this research. The coupling codes are allowed to solve the coupled neutronic and thermal-hydraulic problem within the same one procedure for both steady-state and transient conditions, avoiding the data transmission between different programs. The coupling strategy among two physical fields is implemented in the same fully three-dimensional and fine mesh system, which eliminates numerical errors caused by the mapping of the grid. For the thermal-hydraulics, the built-in solid-fluid coupling conjugate heat transfer solver is applied for the calculation of the fluid mass, momentum, and energy equations, together with the solid energy equation to obtain the temperature distribution. The neutron diffusion equation is solved iteratively via the developed three-dimensional multi-group multi-region neutron diffusion solver to get the power distribution. The macroscopic cross-sections are pre-generated by the Monte Carlo code OpenMC and fitted as functions of temperature added in the neutron diffusion solver. The temperature distribution obtained by thermal-hydraulics calculation will change the macroscopic cross-sections and then impact the neutron diffusion calculation, while the power distribution gained from the neutronic calculation is transferred to the thermal-hydraulics calculation and is used as the heat source term. This coupling methodology are tested on a 3 × 3 PWR fuel pins model. The results show that all physical fields conform correct distribution regularity, illustrating the feasibility of the coupling methodology.

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Yang, Ping, Liangzhi Cao, Hongchun Wu et Changhui Wang. « Conceptual Design of CANDU-SCWR With Thermal-Hydraulics Coupling ». Dans 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29373.

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A CANDU-SCWR core is designed by using a 3D neutronics/thermal-hydraulic coupling method. In the fuel channel design, a typical 43-element fuel bundle is used, the coolant and the moderator are supercritical water and heavy water respectively. The thickness of the moderator is optimized to ensure the negative coolant coefficient during operation. With 1220 MW electric power, the reactor core is designed with a diameter of 4.8m and length of 4.95m, and there are totally 300 fuel channels, each of which consists of 10 fuel bundles. The inlet coolant temperature is set to be 350 °C °C and the operation pressure is 25 MPa. In order to flatten the radial power distribution, the loading pattern of the equilibrium cycle is optimized, and an optimized fuel management scheme is used with three batches refueling, burnable poison Dy2O3 is used to flatten the power peaking. The numerical results show that the average power density is 42.75 W/cm3, while the maximum linear element rate (LER) is 575W/cm. The average discharged burnup of the equilibrium is 48.3GWD/tU, and a high average outlet coolant temperature of 625 °C is achieved with a maximum cladding surface temperature less than 850 °C. Besides, the coolant temperature coefficient is negative throughout the cycle.

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Zhang, Dalin, Changliang Liu, Libo Qian, Guanghui Su et Suizheng Qiu. « Numerical Research on Steady Coupling of Neutronics and Thermal-Hydraulics for a Molten Salt Reactor ». Dans 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48096.

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The Molten Salt Reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for production of electricity, actinide burning, production of hydrogen, and production of fissile fuels. In this paper, a single-liquid-fueled MSR was selected for conceptual research. For this MSR, a ternary system of 15%LiF-58%NaF-27%BeF2 was proposed as the reactor fuel solvent, coolant and also moderator with ca. 1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. The fuel salt flow makes the MSR different from the conventional reactors using solid fissile materials, and makes the neutronics and thermal-hydraulic coupled strongly, which plays the important role in the research of reactor safety analysis. Therefore, it’s necessary to study the coupling of neutronics and thermal-hydraulic. The theoretical models of neutronics and thermal-hydraulics under steady condition were conducted and calculated by numerical method in this paper. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering flow effect. The thermal-hydraulic model was founded on the base of the fundamental conservation laws: the mass, momentum and energy conservation equations. These two models were coupled through the temperature and heat source. The spatial discretization of the above models is based on the finite volume method (FVM), and the thermal-hydraulic equations are computed by SIMPLER algorithm with domain extension method on the staggered grid system. The distribution of neutron fluxes, the distribution of the temperature and velocity and the distribution of the delayed neutron precursors in the core were obtained. The numerical calculated results show that, the fuel salt flow has little effect to the distribution of fast and thermal neutron fluxes and effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition, and the flow could remove the heat generated by the neutron reactions easily to ensure the reactor safe. The obtained results serve some valuable information for the research and design of this new generation reactor.

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Zhang, Dalin, Zhi-Gang Zhai, Andrei Rineiski, Zhangpeng Guo, Chenglong Wang, Yao Xiao et Suizheng Qiu. « COUPLE, A Time-Dependent Coupled Neutronics and Thermal-Hydraulics Code, and its Application to MSFR ». Dans 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30609.

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Molten salt reactor (MSR) using liquid fuel is one of the Generation-IV candidate reactors. Its liquid fuel characteristics are fundamentally different from those of the conventional solid-fuel reactors, especially the much stronger neutronics and thermal hydraulics coupling is drawing significant attention. In this study, the fundamental thermal hydraulic model, neutronic model, and some auxiliary models were established for the liquid-fuel reactors, and a time-dependent coupled neutronics and thermal hydraulics code named COUPLE was developed to solve the mathematic models by the numerical method. After the code was verified, it was applied to the molten salt fast reactor (MSFR) to perform the steady state calculation. The distributions of the neutron fluxes, delayed neutron precursors, velocity, and temperature were obtained and presented. The results show that the liquid fuel flow affects the delayed neutron precursors significantly, while slightly influences the neutron fluxes. The flow in the MSFR core generates a vortex near the fertile tank, which leads to the maximal temperature about 1100 K at the centre of the vortex. The results can provide some useful information for the reactor optimization.

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Rapports d'organisations sur le sujet "Neutronics and thermal-hydraulics coupling"

Todd S. Palmer et Qiao Wu. Improvements in Neutronics/Thermal-Hydraulics Coupling in Two-Phase Flow Systems Using Stochastic-Mixture Transport Models. Office of Scientific and Technical Information (OSTI), septembre 2003. http://dx.doi.org/10.2172/815998.

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Davidson, Gregory, Mathew Swinney, Seth Johnson, Santosh Bhatt et Kaushik Banerjee. Initial Neutronics and Thermal-Hydraulic Coupling for Spent Nuclear Fuel Canister. Office of Scientific and Technical Information (OSTI), septembre 2019. http://dx.doi.org/10.2172/1659634.

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Mark Anderson, M.L. Corradini, K. Sridharan, P. WIlson, D. Cho, T.K. Kim et S. Lomperski. Supercritical Water Nuclear Steam Supply System : Innovations In Materials, Neutronics & ; Thermal-Hydraulics. Office of Scientific and Technical Information (OSTI), septembre 2004. http://dx.doi.org/10.2172/829883.

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Travis, Adam. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling. Office of Scientific and Technical Information (OSTI), mai 2014. http://dx.doi.org/10.2172/1147719.

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Dugan, Kevin J., Shane W. D. Hart et Bradley T. Rearden. Warthog : Coupling Nek5000 Thermal Hydraulics to BISON Fuel Performance through the Giraffe Interface. Office of Scientific and Technical Information (OSTI), octobre 2018. http://dx.doi.org/10.2172/1479731.

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Barber, D. A., R. M. Miller, H. G. Joo, T. J. Downar, W. Wang, V. A. Mousseau et D. D. Ebert. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes. Office of Scientific and Technical Information (OSTI), mars 1999. http://dx.doi.org/10.2172/329553.

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Forsberg, Charles W., Per F. Peterson, Kumar Sridharan, Lin-wen Hu, Massimiliano Fratoni et Anil Kant Prinja. Integrated FHR technology development : Tritium management, materials testing, salt chemistry control, thermal hydraulics and neutronics, associated benchmarking and commercial basis. Office of Scientific and Technical Information (OSTI), octobre 2018. http://dx.doi.org/10.2172/1485415.

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G. S. Chang, M. A. Lillo et R. G. Ambrosek. Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report. Office of Scientific and Technical Information (OSTI), juin 2008. http://dx.doi.org/10.2172/936617.

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