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Bibliographies thématiques / Neutronics and thermal-hydraulics coupling / Articles de revues

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

Auteur : Grafiati

Publié le 9 mars 2023

Mis à jour le 31 juillet 2025

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1

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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Ouazzani, Ayoub, Yannick Gorsse, and Antoine Gerschenfeld. "Coupled neutronics and thermal-hydraulics using TRUST-NK for high performance computing in molten salt reactor simulation." EPJ Web of Conferences 302 (2024): 03007. http://dx.doi.org/10.1051/epjconf/202430203007.

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Molten Salt Reactors display a strong coupling between neutronics and thermal-hydraulics for which various simulation tools have been developed. In this paper we describe TRUST-NK, a neutronics code based on TRUST, CEA’s High Performance Computing (HPC) platform. TRUST-NK provides a multi-group diffusion solver with transport of delayed neutron precursors, for both steady-state and transient simulations. Being based on TRUST allows TRUST-NK to take advantage of the platform’s HPC capabilities and to easily couple to CFD solvers provided by other derivatives of TRUST. After a brief description

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Yang, Ping, Liangzhi Cao, Hongchun Wu, and Changhui Wang. "Core design study on CANDU-SCWR with 3D neutronics/thermal-hydraulics coupling." Nuclear Engineering and Design 241, no. 12 (2011): 4714–19. http://dx.doi.org/10.1016/j.nucengdes.2011.03.036.

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Yu, Jiankai, Hyunsuk Lee, Matthieu Lemaire, Hanjoo Kim, Peng Zhang, and Deokjung Lee. "MCS based neutronics/thermal-hydraulics/fuel-performance coupling with CTF and FRAPCON." Computer Physics Communications 238 (May 2019): 1–18. http://dx.doi.org/10.1016/j.cpc.2019.01.001.

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Nagaya, Yasunobu. "Review of JAEA’s Monte Carlo codes for nuclear reactor core analysis." EPJ Nuclear Sciences & Technologies 11 (2025): 1. https://doi.org/10.1051/epjn/2024020.

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Japan Atomic Energy Agency (JAEA) has been developing a general-purpose continuous-energy Monte Carlo code MVP for nuclear reactor core analysis. Recently improvements to MVP have been focused on the development of an advanced neutronics/thermal-hydraulics coupling code. JAEA has also developed a new Monte Carlo solver Solomon for criticality safety analysis. Solomon aims to calculate the criticality of fuel debris. This paper provides an overview of the capabilities and reviews recent applications of MVP and Solomon.

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Holler, David, Sandesh Bhaskar, Grigirios Delipei, Maria Avramova, and Kostadin Ivanov. "A Framework for Multi-Physics Modeling, Design Optimization and Uncertainty Quantification of Fast-Spectrum Liquid-Fueled Molten-Salt Reactors." Applied Sciences 14, no. 17 (2024): 7615. http://dx.doi.org/10.3390/app14177615.

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The analysis of liquid-fueled molten-salt reactors (LFMSRs) during steady state, operational transients and accident scenarios requires addressing unique reactor multi-physics challenges with coupling between thermal hydraulics, neutronics, inventory control and species distribution phenomena. This work utilizes the General Nuclear Field Operation and Manipulation (GeN-Foam) code to perform coupled thermal-hydraulics and neutronics calculations of an LFMSR design. A framework is proposed as part of this study to perform modeling, design optimization, and uncertainty quantification. The framewo

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Elhareef, Mohamed, Zeyun Wu, and Massimiliano Fratoni. "A Consistent One-Dimensional Multigroup Diffusion Model for Molten Salt Reactor Neutronics Calculations." Journal of Nuclear Engineering 4, no. 4 (2023): 654–67. http://dx.doi.org/10.3390/jne4040041.

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Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model t

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Li, Kaiwen, Shichang Liu, Juanjuan Guo, Zhen Luo, Shanfang Huang, and Kan Wang. "An internal coupling method between neutronics and thermal-hydraulics with RMC and CTF." Annals of Nuclear Energy 187 (July 2023): 109793. http://dx.doi.org/10.1016/j.anucene.2023.109793.

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Xie, Qiuxia, Xiang Chai, and Xiaojing Liu. "Multi-physical coupling study of neutronics/thermal-hydraulics/material corrosion based on the unified coupling framework." Nuclear Engineering and Design 425 (August 2024): 113339. http://dx.doi.org/10.1016/j.nucengdes.2024.113339.

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Dai, Tao, Liangzhi Cao, Qingming He, Hongchun Wu, and Wei Shen. "A Two-Way Neutronics/Thermal-Hydraulics Coupling Analysis Method for Fusion Blankets and Its Application to CFETR." Energies 13, no. 16 (2020): 4070. http://dx.doi.org/10.3390/en13164070.

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The China Fusion Engineering Test Reactor (CFETR) is a tokamak device to validate and demonstrate fusion engineering technology. In CFETR, the breeding blanket is a vital important component that is closely related to the performance and safety of the fusion reactor. Neutronics/thermal-hydraulics (N/TH) coupling effect is significant in the numerical analysis of the fission reactor. However, in the numerical analysis of the fusion reactor, the existing coupling code system mostly adopts the one-way coupling method. The ignorance of the two-way N/TH coupling effect would lead to inaccurate resu

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Yang, Qing, Qingquan Pan, Hui He, Tengfei Zhang, and Xiaojing Liu. "Improved design of LBE cooled solid reactor using 3D neutronics thermal-hydraulics coupling method." Annals of Nuclear Energy 179 (December 2022): 109441. http://dx.doi.org/10.1016/j.anucene.2022.109441.

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Niu, Yuchen, Dabin Sun, Yuandong Zhang, Lei Chen, Minjun Peng, and Genglei Xia. "Operation characteristics analysis of supercritical CO2 reactor based on neutronics and thermal-hydraulics coupling." Nuclear Engineering and Design 436 (May 2025): 113947. https://doi.org/10.1016/j.nucengdes.2025.113947.

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Delipei, Gregory K., Pascal Rouxelin, Agustin Abarca, Jason Hou, Maria Avramova, and Kostadin Ivanov. "CTF-PARCS Core Multi-Physics Computational Framework for Efficient LWR Steady-State, Depletion and Transient Uncertainty Quantification." Energies 15, no. 14 (2022): 5226. http://dx.doi.org/10.3390/en15145226.

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Best Estimate Plus Uncertainty (BEPU) approaches for nuclear reactor applications have been extensively developed in recent years. The challenge for BEPU approaches is to achieve multi-physics modeling with an acceptable computational cost while preserving a reasonable fidelity of the physics modeled. In this work, we present the core multi-physics computational framework developed for the efficient computation of uncertainties in Light Water Reactor (LWR) simulations. The subchannel thermal-hydraulic code CTF and the nodal expansion neutronic code PARCS are coupled for the multi-physics model

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García-Herranz, N., D. Cuervo, A. Sabater, G. Rucabado, S. Sánchez-Cervera, and E. Castro. "Multiscale neutronics/thermal-hydraulics coupling with COBAYA4 code for pin-by-pin PWR transient analysis." Nuclear Engineering and Design 321 (September 2017): 38–47. http://dx.doi.org/10.1016/j.nucengdes.2017.03.017.

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Guo, Juanjuan, Shichang Liu, Xiaotong Shang, et al. "Versatility and stabilization improvements of full core neutronics/thermal-hydraulics coupling between RMC and CTF." Nuclear Engineering and Design 332 (June 2018): 88–98. http://dx.doi.org/10.1016/j.nucengdes.2018.03.028.

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Mala, P., A. Pautz, H. Ferroukhi, and A. Vasiliev. "DEVELOPMENT OF 3D PIN-BY-PIN CORE SOLVER TORTIN AND COUPLING WITH THERMAL-HYDRAULICS." EPJ Web of Conferences 247 (2021): 02034. http://dx.doi.org/10.1051/epjconf/202124702034.

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Currently, safety analyses mostly rely on codes which solve both the neutronics and the thermal-hydraulics with assembly-wise nodes resolution as multiphysics heterogeneous transport solvers are still too time and memory expensive. The pin-by-pin homogenized codes can be seen as a bridge between the heterogeneous codes and the traditional nodal assembly-wise calculations. In this work, the pin-by-pin simplified transport solver Tortin has been coupled with a sub-channel code COBRA-TF. The verification of the 3D solver of Tortin is presented at first, showing very good agreement in terms of axi

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Chen, Juan, Tao Zhou, Zhou Sen Hou, Wan Xu Cheng, and Can Hui Sun. "Influence Analysis of Coupled Neutronics and Thermo-Hydraulics on Steady-State Characteristics of Supercritical Water-Cooled Reactor." Advanced Materials Research 472-475 (February 2012): 278–83. http://dx.doi.org/10.4028/www.scientific.net/amr.472-475.278.

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the coulped neutronics and thermo-hydraulics model for supercritical water-cooled reactor (SCWR) is developed by internal coupling method. It is based on the two group neutron diffusion equations and the one-dimensional junction thermal analysis mode, in which the cross sections used for SCWR are generated by Dragon tool. Compared with the calculation results based on the non-coupling calculation model, the steady state characteristics under coupling calculation condition are detailed analyzed by considering parameters feedback at each axial node. The results show that, as coupled model is cho

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Zhang, Binhang, Zenghao Liu, Xianbao Yuan, et al. "Validation of DDC-3D code system for neutronics and thermal-hydraulics coupling analysis using BEAVRS benchmark." Nuclear Engineering and Design 429 (December 2024): 113583. http://dx.doi.org/10.1016/j.nucengdes.2024.113583.

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Yang, Dongmei, Xiaojing Liu, Jinbiao Xiong, Xiang Chai, and Xu Cheng. "Coupling of neutronics and thermal-hydraulics codes for the simulation of reactivity insertion accident for LFR." Progress in Nuclear Energy 106 (July 2018): 20–26. http://dx.doi.org/10.1016/j.pnucene.2018.02.023.

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Zhang, Han, Jiong Guo, Jianan Lu, Jinlin Niu, Fu Li, and Yunlin Xu. "The comparison between nonlinear and linear preconditioning JFNK method for transient neutronics/thermal-hydraulics coupling problem." Annals of Nuclear Energy 132 (October 2019): 357–68. http://dx.doi.org/10.1016/j.anucene.2019.04.053.

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Luo, Hao, Kaiwen Li, Jie Li, et al. "Enhancing accuracy and efficiency of RMC/SUBCHAN neutronics and thermal-hydraulics coupling system for BEAVRS simulation." Progress in Nuclear Energy 183 (May 2025): 105666. https://doi.org/10.1016/j.pnucene.2025.105666.

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Tuominen, Riku, Ville Valtavirta, Manuel García, Diego Ferraro, and Jaakko Leppänen. "EFFECT OF ENERGY DEPOSITION MODELLING IN COUPLED STEADY STATE MONTE CARLO NEUTRONICS/THERMAL HYDRAULICS CALCULATIONS." EPJ Web of Conferences 247 (2021): 06001. http://dx.doi.org/10.1051/epjconf/202124706001.

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In coupled calculations with Monte Carlo neutronics and thermal hydraulics the Monte Carlo code is used to produce a power distribution which in practice means tallying the energy deposition. Usually the energy deposition is estimated by making a simple approximation that energy is deposited only in fission reactions. The goal of this work is to study how the accuracy of energy deposition modelling affects the results of steady state coupled calculations. For this task an internal coupling between Monte Carlo transport code Serpent 2 and subchannel code SUBCHANFLOW is used along with a recentl

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Liu, Hanxing, and Han Zhang. "A Reduced Order Model Based on ANN-POD Algorithm for Steady-State Neutronics and Thermal-Hydraulics Coupling Problem." Science and Technology of Nuclear Installations 2023 (July 10, 2023): 1–15. http://dx.doi.org/10.1155/2023/9385756.

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The neutronics and thermal-hydraulics (N/TH) coupling behavior analysis is a key issue for nuclear power plant design and safety analysis. Due to the high-dimensional partial differential equations (PDEs) derived from the N/TH system, it is usually time consuming to solve such a large-scale nonlinear equation by the traditional numerical solution method of PDEs. To solve this problem, this work develops a reduced order model based on the proper orthogonal decomposition (POD) and artificial neural networks (ANNs) to simulate the N/TH coupling system. In detail, the POD method is used to extract

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Akbas, Sabahattin, Victor Martinez-Quiroga, Fatih Aydogan, Chris Allison, and Abderrafi M. Ougouag. "Thermal-hydraulics and neutronic code coupling for RELAP/SCDAPSIM/MOD4.0." Nuclear Engineering and Design 344 (April 2019): 174–82. http://dx.doi.org/10.1016/j.nucengdes.2019.01.009.

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Ye, Linrong, Mingjun Wang, Xin’an Wang, et al. "Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core." Science and Technology of Nuclear Installations 2020 (August 28, 2020): 1–12. http://dx.doi.org/10.1155/2020/2562747.

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The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and n

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Safavi, A., M. H. Esteki, S. M. Mirvakili, and M. Khaki. "Validation of a new neutronics/thermal hydraulics coupling code for steady state analysis of light water reactors." Kerntechnik 85, no. 5 (2020): 351–58. http://dx.doi.org/10.3139/124.190087.

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Xu, Xiaobei, Zhouyu Liu, Hongchun Wu, and Liangzhi Cao. "Neutronics/thermal-hydraulics/fuel-performance coupling for light water reactors and its application to accident tolerant fuel." Annals of Nuclear Energy 166 (February 2022): 108809. http://dx.doi.org/10.1016/j.anucene.2021.108809.

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Yang, Bowen, Jianqiang Shan, and Li Ge. "Development and application of a neutronics/thermal-hydraulics coupling code based on system code and Anderson acceleration." Annals of Nuclear Energy 219 (September 2025): 111490. https://doi.org/10.1016/j.anucene.2025.111490.

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Giudicelli, Guillaume, Fande Kong, Roy Stogner, et al. "Data transfers for full core heterogeneous reactor high-fidelity multiphysics studies." EPJ Web of Conferences 302 (2024): 05006. http://dx.doi.org/10.1051/epjconf/202430205006.

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Multiphysics simulations for nuclear reactor analysis are usually performed by resorting to operator splitting and fixed point iterations between single-physics solvers. This enables the separate solution of each physics, such as neutronics, fuel performance, and thermal hydraulics, on meshes tailored to the requirements of the respective numerical discretizations of the equations. As the equations are coupled, several fields must be transferred between singlephysics solves. Projecting fields between meshes while preserving order of accuracy, conservation properties, and mapping non-overlappin

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Mochizuki, Hiroyasu. "Verification of neutronics and thermal-hydraulics coupling method for FLUENT code using the MSRE pump startup, trip data." Nuclear Engineering and Design 378 (July 2021): 111191. http://dx.doi.org/10.1016/j.nucengdes.2021.111191.

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Yang, Dongmei, Xiaojing Liu, Tengfei Zhang, and Xu Cheng. "A comparison of three algorithms applied in thermal-hydraulics and neutronics codes coupling for lbe-cooled fast reactor." Annals of Nuclear Energy 149 (December 2020): 107789. http://dx.doi.org/10.1016/j.anucene.2020.107789.

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Xie, Qiuxia, Wei Li, Chaoran Guan, Qizheng Sun, Xiang Chai, and Xiaojing Liu. "Development of 3D transient neutronics and thermal-hydraulics coupling procedure and its application to a fuel pin analysis." Nuclear Engineering and Design 404 (April 2023): 112164. http://dx.doi.org/10.1016/j.nucengdes.2023.112164.

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Rais, A., D. Siefman, G. Girardin, M. Hursin, and A. Pautz. "Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL." Science and Technology of Nuclear Installations 2015 (2015): 1–9. http://dx.doi.org/10.1155/2015/237646.

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In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core mo

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Mochizuki, Hiroyasu. "Validation of neutronics and thermal-hydraulics coupling model of the RELAP5-3D code using the MSRE reactivity insertion tests." Nuclear Engineering and Design 389 (April 2022): 111669. http://dx.doi.org/10.1016/j.nucengdes.2022.111669.

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Zhang, Yijun, Liangzhi Cao, Zhouyu Liu, and Hongchun Wu. "Newton-Krylov method with nodal coupling coefficient to solve the coupled neutronics/thermal-hydraulics equations in PWR transient analysis." Annals of Nuclear Energy 118 (August 2018): 220–34. http://dx.doi.org/10.1016/j.anucene.2018.04.016.

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Li, Shuaizheng, Zhouyu Liu, Junji Chen, Minwan Zhang, Liangzhi Cao, and Hongchun Wu. "Development of high-fidelity neutronics/thermal-hydraulics coupling system for the hexagonal reactor cores based on NECP-X/CTF." Annals of Nuclear Energy 188 (August 2023): 109822. http://dx.doi.org/10.1016/j.anucene.2023.109822.

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Yang, Qing, Qingquan Pan, and Xiaojing Liu. "Analysis of lead–bismuth eutectic-cooled solid reactor under flow blockage accident by 3D neutronics thermal-hydraulics coupling method." Nuclear Engineering and Design 407 (June 2023): 112313. http://dx.doi.org/10.1016/j.nucengdes.2023.112313.

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Forestier, M., G. Girault, F. Jacq, and A. Sargeni. "ANTARES: COUPLING PARCS WITH CATHARE-3." EPJ Web of Conferences 247 (2021): 07005. http://dx.doi.org/10.1051/epjconf/202124707005.

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In recent years, the IRSN has launched a new project to couple the first 3D version of the thermal hydraulic code CATHARE-3 (system) with the 3D, neutronic nodal code PARCS (core): ANTARES (Advanced Neutronics and Thermal-hydraulic for the Analysis of the Reactor Safety). The purpose of this project is to increase the IRSN capability to couple different codes, to calculate the core power distribution in CATHARE-3 and to improve the thermal hydraulic boundaries conditions in PARCS. In this way, the IRSN diversifies its available tools to perform safety analysis with improved accuracy. The curre

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Xu, Yuchao, Jason Hou, and Kostadin N. Ivanov. "IMPROVEMENT TO NEM SP3 MODELLING AND SIMULATION." EPJ Web of Conferences 247 (2021): 03008. http://dx.doi.org/10.1051/epjconf/202124703008.

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Accurate reactor core steady state safety analysis requires coupling between thermal-hydraulics and three dimensional multigroup pin by pin neutronics. Concerning the neutronics modeling, the Nodal Expansion Method (NEM) code is developed at North Carolina State University in the framework of high fidelity multiphysics coupling with CTF. NEM includes a simplified third-order Spherical Harmonic (SP3) solver. In this work, the solver has been improved by incorporating higher order scattering matrix library. The boundary conditions were corrected with one dimensional P3 theory and a consistent co

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Furuya, Masahiro, Takanori Fukahori, and Shinya Mizokami. "Development of BWR Regional Stability Experimental Facility SIRIUS-F, Which Simulates Thermal Hydraulics-Neutronics Coupling, and Stability Evaluation of ABWRs." Nuclear Technology 158, no. 2 (2007): 191–207. http://dx.doi.org/10.13182/nt07-a3835.

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Eric, Cervi, Lorenzi Stefano, Luzzi Lelio, and Cammi Antonia. "Multiphysics analysis of the MSFR helium bubbling system: A comparison between neutron diffusion, SP3 neutron transport and Monte Carlo approaches." Annals of Nuclear Energy 132 (June 27, 2019): 227–35. https://doi.org/10.1016/j.anucene.2019.04.029.

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The Molten Salt Fast Reactor is a fast-spectrum molten salt reactor under development in the framework of the European H2020 SAMOFAR Project (http://samofar.eu/). Among the design peculiarities, this circulating fuel reactor features a helium bubbling system aimed at removing on-line gaseous fission products, and metallic particles as well. From a modelling point of view, the presence of helium bubbles in the core needs to be assessed both from a neutronics and thermal-hydraulics point of view. In this paper, the attention is paid to the first aspect, analysing the void reactivity effect induc

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