Relevant bibliographies by topics / Neutronics and thermal-hydraulics coupling

Contents

  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Neutronics and thermal-hydraulics coupling'

Author: Grafiati
Published: 9 March 2023
Last updated: 31 July 2025

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Journal articles on the topic "Neutronics and thermal-hydraulics coupling"

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Królikowski, Igor P., and Jerzy Cetnar. "Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent." Nukleonika 60, no. 3 (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 n
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Blanco, J. A., P. Rubiolo, and 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 implemen
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Tollit, Brendan, Alan Charles, William Poole, et al. "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
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Wu, Jianhui, Jingen Chen, Xiangzhou Cai, et al. "A Review of Molten Salt Reactor Multi-Physics Coupling Models and Development Prospects." Energies 15, no. 21 (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 absor
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Ta, Duy Long, Huy Hiep Nguyen, Tuan Khai Nguyen, Vinh Thanh Tran, and Huu Tiep Nguyen. "Coulped neutronics/thermal-hydraulics calculation of VVER-1000 fuel assembly." Nuclear Science and Technology 6, no. 2 (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 convergen
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Jiang, Duoyu, Peng Xu, Tianliang Hu, et al. "Coupled Monte Carlo and Thermal-Hydraulics Modeling for the Three-Dimensional Steady-State Analysis of the Xi’an Pulsed Reactor." Energies 16, no. 16 (2023): 6046. http://dx.doi.org/10.3390/en16166046.

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The Xi’an Pulsed Reactor (XAPR) is characterized by its small core size and integrated fuel moderator structure, which results in a non-uniform core power and temperature distribution. Consequently, a complex coupling relationship exists between its core neutronics and thermal hydraulics, necessitating the assurance for the operational safety of the XAPR. To optimize the experimental scheme in the reactor, a refined three-dimensional steady-state nuclear-thermal coupling analysis is imperative. This study focuses on investigating the coupling calculation of a three-dimensional steady-state neu
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Price, Dean, Majdi I. Radaideh, Travis Mui, Mihir Katare, and Tomasz Kozlowski. "Multiphysics Modeling and Validation of Spent Fuel Isotopics Using Coupled Neutronics/Thermal-Hydraulics Simulations." Science and Technology of Nuclear Installations 2020 (July 26, 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
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Pascal, V., Y. Gorsse, N. Alpy, 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 feedb
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Ma, Yugao, Jinkun Min, Jin Li, et al. "Neutronics and thermal-hydraulics coupling analysis in accelerator-driven subcritical system." Progress in Nuclear Energy 122 (April 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, and Suizheng Qiu. "Neutronics/Thermal-hydraulics Coupling Analysis for the Liquid-Fuel MOSART Concept." Energy Procedia 127 (September 2017): 343–51. http://dx.doi.org/10.1016/j.egypro.2017.08.075.

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Dissertations / Theses on the topic "Neutronics and thermal-hydraulics coupling"

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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
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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
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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 o
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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], and 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], and 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). T
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Alzaben, Yousef Ibrahim [Verfasser], and 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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Books on the topic "Neutronics and thermal-hydraulics coupling"

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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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Book chapters on the topic "Neutronics and thermal-hydraulics coupling"

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Yang, Mingrui, Chixu Luo, Dan Wang, Tianxiong Wang, Xiaojing Liu, and Tengfei Zhang. "Development and Preliminary Verification of a Neutronics-Thermal Hydraulics Coupling Code for Research Reactors with Unstructured Meshes." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_58.

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AbstractTo maximize their adaptability and versatility, research reactors are designed to adapt to various operational conditions. These requirements result in more complex configurations and irregular geometries for research reactors. Besides, there is usually a strong coupling of neutronics-thermal hydraulics (N-TH) fields inside the reactor. A three-dimensional N-TH coupling code has been developed named CENTUM (CodE for N-Th coupling with Unstructured Mesh). Steady-state and transient neutronic analyses are performed using a 3D triangular-z nodal transport solver with the stiffness confine
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Han, Wenbin, Zechuan Guan, Shanfang Huang, and Jian Deng. "Multi-physics Coupling Analyses of Nuclear Thermal Propulsion Reactor." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_80.

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AbstractNuclear thermal propulsion (NTP) reactors have high-temperature solid-state characteristics and significant thermal expansion, which therefore require multi-physics coupling analyses. In this paper, the framework of Neutronics, Thermal-Hydraulics and Mechanics coupling (N/T-H/M) of nuclear thermal propulsion reactor is developed, and the typical reactor XE-2 is analyzed with this method. The results show that the N/T-H/M coupling will bring -1049 pcm negative reactivity, of which the thermal expansion effect accounts for 22%, indicating that the nuclear thermal propulsion reactor has a
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Li, Xiangyue, Xiaojing Liu, Xiang Chai, and Tengfei Zhang. "Preliminary Multi-physics Coupled Simulation of Small Helium-Xenon Cooled Mobile Nuclear Reactor." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_59.

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AbstractFor the prediction of the internal physical process of SIMONS (Small Innovative helium-xenon cooled MObile Nuclear power Systems), this research created a coupled three-dimensional high-fidelity calculation platform of the neutronics/ thermo-elasticity analysis called FEMAS (FEM based Multi-physics Analysis Software for Nuclear Reactor). This platform allows for the multi-physics coupling calculations of neutron diffusion/ transport, thermal diffusion, and thermal elasticity. It is based on the open-source Monte Carlo code OpenMC and the open-source finite element codes Dealii and Feni
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Zhou, Xingguang, Dalin Zhang, Xinyu Li, et al. "Neutronics and Thermal-Hydraulics Coupling Analysis of Integral Inherently Safe Fluoride-Salt-Cooled High-Temperature Advanced Reactor - Fustar." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-19-8899-8_7.

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Qin, Junwei, and Yunzhao Li. "Three-Dimensional Pin-by-Pin Transient Analysis for PWR-Core." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_67.

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AbstractTo ensure the safety of PWR-core operation, three-dimensional whole-core transient analysis needs to be carried out for the sake of the pin-power distribution. For this purpose, this paper presents “Bamboo-Transient 2.0”, a three-dimensional pin-by-pin transient analysis program. The program adopts a fully-implicit backward method with finite difference for time variable discretization, a method of exponential function expansion nodal (EFEN) SP3 for the neutron transport calculation, and a multi-channel model for the thermal feedback calculation. In addition, Picard iteration is used t
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Zhao, Chuanqi, Kunpeng Wang, Liangzhi Cao, Hongchun Wu, and Youqi Zheng. "Coupled Neutronics and Thermal–Hydraulics Analysis of Annular Fuel Assembly for SCWR." In Proceedings of The 20th Pacific Basin Nuclear Conference. 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, et al. "Sodium Fast Reactor Design: Fuels, Neutronics, Thermal-Hydraulics, Structural Mechanics and Safety." In Handbook of Nuclear Engineering. Springer US, 2010. http://dx.doi.org/10.1007/978-0-387-98149-9_21.

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Kalra, Hemant, Sanjay Singh, Paresh Patra, Y. K. Pandey, and N. Rama Mohan. "Validation of System Thermal Hydraulics Neutronics Computer Code ATMIKA LWR for KKNPP Reactors." In Lecture Notes in Mechanical Engineering. Springer Nature Singapore, 2024. http://dx.doi.org/10.1007/978-981-97-3087-2_63.

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Zhou, Shuyong, Zesong Huang, Junming Tang, et al. "Research on Interface Technology of Coupling Thermal-Hydraulics and Other Codes." In Springer Proceedings in Physics. Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-19-8899-8_77.

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

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Conference papers on the topic "Neutronics and thermal-hydraulics coupling"

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Stewart, R., J. Cavaluzzi, P. Balestra, and G. Strydom. "Multiphysics Pebble-Bed Reactor Run-In Simulation Using a High-Fidelity Neutronics--Thermal Hydraulics Framework." In International Conference on Physics of Reactors (PHYSOR 2024). American Nuclear Society, 2024. http://dx.doi.org/10.13182/physor24-43676.

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Komu, Rebekka, Ville Valtavirta, Riku Tuominen, and Seppo Hillberg. "LDR Lite Benchmark: Coupled 3D Neutronics and Thermal-Hydraulics Analysis of a Control Rod Drop Transient." In International Conference on Physics of Reactors (PHYSOR 2024). American Nuclear Society, 2024. http://dx.doi.org/10.13182/physor24-43668.

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Ferraro, Diego, Manuel García, Ville Valtavirta, et al. "Direct Calculation of Safety-Related Parameters for Coupled Transients Using Monte Carlo Neutronics Plus Subchannel Thermal-Hydraulics." In Mathematics and Computation 2021. American Nuclear Society, 2021. https://doi.org/10.13182/xyz-33801.

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Ge, Jian, Dalin Zhang, Wenxi Tian, Suizheng Qiu, and G. H. Su. "Coupled Analysis of Thermal Hydraulics and Neutronics for a Molten Salt Reactor." In 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 neut
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Akbas, Sabahattin, Victor Martinez-Quiroga, Fatih Aydogan, Abderrafi M. Ougouag, and Chris Allison. "Survey of Coupling Schemes in Traditional Coupled Neutronics and Thermal-Hydraulics Codes." In 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
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Zheng, Yong, Min-jun Peng, Geng-lei Xia, and Ren Li. "Investigation on Coupling Behaviors of Thermal-Hydraulics/Neutronics Under Asymmetrical Inlet Conditions." In 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 h
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Kamiya, Tomohiro, Taku Nagatake, Ayako Ono, et al. "Neutronics/Thermal-Hydraulics Coupling Simulation Using JAMPAN in a Single BWR Fuel Assembly." In 2024 31st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2024. http://dx.doi.org/10.1115/icone31-135974.

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Abstract We have aimed to realize high-fidelity neutronics/thermal-hydraulics coupling simulation to provide simulation results that can be used as validation data for reactor analysis codes. We have developed a multi-physics platform, JAMPAN, to conduct neutronics/thermal-hydraulics coupling simulation by connecting independent codes. It is required to reduce empirical correlations as far as possible to perform high-fidelity neutronics/thermal-hydraulics coupling simulation. Hence, a continuous energy Monte Carlo code MVP is adopted as a neutronic analysis code, and a detailed and phenomenolo
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Martinez-Quiroga, Victor, Sabahattin Akbas, Fatih Aydogan, Abderrafi M. Ougouag, and Chris Allison. "Coupling of RELAP5-SCDAP MOD4.0 and Neutronic Codes." In 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 thi
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Chen, Jun, Liangzhi Cao, Zhouyu Liu, Hongchun Wu, and Yijun Zhang. "Preliminary Verification of the High-Fidelity Neutronics and Thermal-Hydraulics Coupling System NECP-X/SUBSC." In 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
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Li, Zhigang, Juejie Pan, Bangyang Xia, Wei Lu, Wenbo Zhao, and Shenglong Qiang. "A Single Channel Thermal-Hydraulic Calculation Module for PWR Pin-by-Pin Wise Coupled Calculation System." In 2024 31st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2024. http://dx.doi.org/10.1115/icone31-124037.

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Abstract Due to the strong feedback effect between neutronics and thermal-hydraulics in the core of pressurized water reactors (PWR), neutronics and thermal-hydraulics coupling calculations are often used in the design and safety evaluation of PWR to provide more accurate results. The coupling calculation of neutronics and thermal-hydraulics at the full core pin-by-pin wise can provide more precise and precise coupling parameters, which is a hot research topic in the industry. How to advance the efficiency and accuracy of neutronics and thermal-hydraulics coupling calculations at the entire co
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Reports on the topic "Neutronics and thermal-hydraulics coupling"

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Todd S. Palmer and 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), 2003. http://dx.doi.org/10.2172/815998.

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

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Spencer, Benjamin, William Hoffman, and Albert Dahal. Coupling Thermal-Hydraulics with Reactor Pressure Vessel Fracture Models. Office of Scientific and Technical Information (OSTI), 2023. http://dx.doi.org/10.2172/2204859.

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Mark Anderson, M.L. Corradini, K. Sridharan, et al. Supercritical Water Nuclear Steam Supply System: Innovations In Materials, Neutronics & Thermal-Hydraulics. Office of Scientific and Technical Information (OSTI), 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), 2014. http://dx.doi.org/10.2172/1147719.

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

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Barber, D. A., R. M. Miller, H. G. Joo, et al. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes. Office of Scientific and Technical Information (OSTI), 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, and 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), 2018. http://dx.doi.org/10.2172/1485415.

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G. S. Chang, M. A. Lillo, and 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), 2008. http://dx.doi.org/10.2172/936617.

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