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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 10 mars 2023

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1

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

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Dai, Tao, Liangzhi Cao, Qingming He, Hongchun Wu et Wei Shen. « A Two-Way Neutronics/Thermal-Hydraulics Coupling Analysis Method for Fusion Blankets and Its Application to CFETR ». Energies 13, n^o 16 (6 août 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 results. In this paper, the MCNP/FLUENT code system is developed based on the 3D-1D-2D hybrid coupling method. The one-way and two-way N/TH coupling calculations for two typical blanket concepts, the helium-cooled solid breeder (HCSB) blanket and the water-cooled ceramic breeder (WCCB) blanket, are carried out to study the two-way N/TH coupling effect in CFETR. The numerical results show that, compared with the results from the one-way N/TH coupling calculation, the tritium breeding ration (TBR) calculated with the two-way N/TH calculation decreases by −0.11% and increases by 4.45% for the HCSB and WCCB blankets, respectively. The maximum temperature increases by 1 °C and 29 °C for the HCSB and WCCB blankets, respectively.

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

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

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Guo, Juanjuan, Shichang Liu, Xiaotong Shang, Qicang Shen, Xiaoyu Guo, Shanfang Huang, Kan Wang et Xiaoming Chai. « Versatility and stabilization improvements of full core neutronics/thermal-hydraulics coupling between RMC and CTF ». Nuclear Engineering and Design 332 (juin 2018) : 88–98. http://dx.doi.org/10.1016/j.nucengdes.2018.03.028.

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Delipei, Gregory K., Pascal Rouxelin, Agustin Abarca, Jason Hou, Maria Avramova et Kostadin Ivanov. « CTF-PARCS Core Multi-Physics Computational Framework for Efficient LWR Steady-State, Depletion and Transient Uncertainty Quantification ». Energies 15, n^o 14 (19 juillet 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 modeling (CTF-PARCS). The computational framework is discussed in detail from the Polaris lattice calculations up to the CTF-PARCS coupling approaches. Sampler is used to perturb the multi-group microscopic cross-sections, fission yields and manufacturing parameters, while Dakota is used to sample the CTF input parameters and the boundary conditions. Python scripts were developed to automatize and modularize both pre- and post-processing. The current state of the framework allows the consistent perturbation of inputs across neutronics and thermal-hydraulics modeling. Improvements to the standard thermal-hydraulics modeling for such coupling approaches have been implemented in CTF to allow the usage of 3D burnup distribution, calculation of the radial power and the burnup profile, and the usage of Santamarina effective Doppler temperature. The uncertainty quantification approach allows the treatment of both scalar and functional quantities and can estimate correlation between the multi-physics outputs of interest and up to the originally perturbed microscopic cross-sections and yields. The computational framework is applied to three exercises of the LWR Uncertainty Analysis in Modeling Phase III benchmark. The exercises cover steady-state, depletion and transient calculations. The results show that the maximum fuel centerline temperature across all exercises is 2474K with 1.7% uncertainty and that the most correlated inputs are the 238U inelastic and elastic cross-sections above 1 MeV.

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Mala, P., A. Pautz, H. Ferroukhi et 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 axial and radial power profile with the Monte Carlo code SERPENT for a small minicore and with the state-of-the-art nodal code SIMULATE5 for a quarter core without feedback. Then the results of Tortin+COBRA-TF are compared with SIMULATE5 for one assembly problem with feedback. The axial profiles of power and moderator temperature show good agreement, while the fuel temperature differ by up to 40 K. This is caused mainly by different gap and fuel conductance parameters used in COBRA-TF and in SIMULATE5.

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Chen, Juan, Tao Zhou, Zhou Sen Hou, Wan Xu Cheng et 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 (février 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 chosen its axial power distribution would give an obvious deviation from the cosine function that used for non-coupled model. Although the cladding temperature at most of the axial nodes rises with a shifted power peak, the maximum cladding temperature is finally decreased. For the above coupling condition, the maximum cladding temperature would appear at the external assemblies with lower coolant temperature but not at inner assemblies with higher coolant temperature. As the detailed description for coupling characteristics of supercritical water-cooled reactor is given, a certain theory reference for its system safety could be provided.

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Yang, Dongmei, Xiaojing Liu, Jinbiao Xiong, Xiang Chai et Xu Cheng. « Coupling of neutronics and thermal-hydraulics codes for the simulation of reactivity insertion accident for LFR ». Progress in Nuclear Energy 106 (juillet 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 et Yunlin Xu. « The comparison between nonlinear and linear preconditioning JFNK method for transient neutronics/thermal-hydraulics coupling problem ». Annals of Nuclear Energy 132 (octobre 2019) : 357–68. http://dx.doi.org/10.1016/j.anucene.2019.04.053.

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Tuominen, Riku, Ville Valtavirta, Manuel García, Diego Ferraro et 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 recently implemented energy deposition treatment of Serpent 2. The new treatment offers four energy deposition modes each of which offers a different combination of accuracy and required computational time. As a test case, a 3D PWR fuel assembly is modelled with different energy deposition modes. The resulting effective multiplication factors are within 30 pcm. Differences of up to 100K are observed in the fuel temperatures.

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

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Safavi, A., M. H. Esteki, S. M. Mirvakili et M. Khaki. « Validation of a new neutronics/thermal hydraulics coupling code for steady state analysis of light water reactors ». Kerntechnik 85, n^o 5 (12 octobre 2020) : 351–58. http://dx.doi.org/10.3139/124.190087.

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

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Ye, Linrong, Mingjun Wang, Xin’an Wang, Jian Deng, Yan Xiang, Wenxi Tian, Suizheng Qiu et G. H. Su. « Thermal Hydraulic and Neutronics Coupling Analysis for Plate Type Fuel in Nuclear Reactor Core ». Science and Technology of Nuclear Installations 2020 (28 août 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 neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.

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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 (juillet 2021) : 111191. http://dx.doi.org/10.1016/j.nucengdes.2021.111191.

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Yang, Dongmei, Xiaojing Liu, Tengfei Zhang et 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 (décembre 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 et 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 (avril 2023) : 112164. http://dx.doi.org/10.1016/j.nucengdes.2023.112164.

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Rais, A., D. Siefman, G. Girardin, M. Hursin et 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 modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

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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 (avril 2022) : 111669. http://dx.doi.org/10.1016/j.nucengdes.2022.111669.

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Zhang, Yijun, Liangzhi Cao, Zhouyu Liu et 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 (août 2018) : 220–34. http://dx.doi.org/10.1016/j.anucene.2018.04.016.

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Forestier, M., G. Girault, F. Jacq et 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 current technique usually adopted in France for the safety demonstrations is the so-called ‘conservative' approach, which consists of reducing all the feedback (Doppler and moderator effects) and in modifying some physical quantities in such a way to increase a power peak in an accidental transient. For this reason, these facilities (‘penalties’) have been implemented in ANTARES. In this paper we will give two examples of accidental transients that can be simulated with ANTARES: a REA (Rod Ejection Accident) and an inadvertent boron dilution event.

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Furuya, Masahiro, Takanori Fukahori et 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, n^o 2 (mai 2007) : 191–207. http://dx.doi.org/10.13182/nt07-a3835.

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Xu, Yuchao, Jason Hou et 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 coupling coupling between zeroth- and second-order flux moments was established. Two methods for generating second order discontinuity factors (DFs) has ben developed, one based on the Generalized Equivalence Theory (GET) and one based on Parial Current Equivalence Theory (PCET). DFs were generated with three lattice sizes: single pin, 2 pins and assembly level. These developments were tested using the C5G7 benchmark. The results of the SP3 solver improvement, by using P2 and P3 scattering cross sections, show a 50% decrease in the eigenvalue (keff) prediction error compared to the reference transport solution. The GET DFs are applied in the C5G7 core pin by pin calculation and are compared with PCET DFs. The results show that PCET have a better performance in global results (eigenvalue). Concerning the different lattice sizes studies, the results show that DFs generated in smalll colorsets can improve local solutions. However, in order to reveal strong global trends, DFs should be generated in a larger corloset representative of the whole core. For the core calculations, DFs generated with the three colorsets together with an additional mixed type DFs were tested. For the mixed type, DFs generated from assembly size lattice were used for the internal interfaces and DFs generated from 2 pins size lattice were used for the assemblies boundary interfaces. These mixed DFs outperformed all the other configurations indicating that they manage to accomplish a satisfying compromise between global and local trends.

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Zhang, Qingyang, Tianji Peng, Guangchun Zhang, Jie Liu, Xiaowei Guo, Chunye Gong, Bo Yang et Xukai Fan. « An Efficient Scheme for Coupling OpenMC and FLUENT with Adaptive Load Balancing ». Science and Technology of Nuclear Installations 2021 (24 septembre 2021) : 1–16. http://dx.doi.org/10.1155/2021/5549602.

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This paper develops a multi-physics interface code MC-FLUENT to couple the Monte Carlo code OpenMC with the commercial computational fluid dynamics code ANSYS FLUENT. The implementations and parallel performances of block Gauss–Seidel-type and block Jacobi-type Picard iterative algorithms have been investigated. In addition, this paper introduces two adaptive load-balancing algorithms into the neutronics and thermal-hydraulics coupled simulation to reduce the time cost of computation. Considering that the different scalability of OpenMC and FLUENT limits the performance of block Gauss–Seidel algorithm, an adaptive load-balancing algorithm that can increase the number of nodes dynamically is proposed to improve its efficiency. Moreover, with the natural parallelism of block Jacobi algorithm, another adaptive load-balancing algorithm is proposed to improve its performance. A 3 x 3 PWR fuel pin model and a 1000 MWt ABR metallic benchmark core were used to compare the performances of the two algorithms and verify the effectiveness of the two adaptive load-balancing algorithms. The results show that the adaptive load-balancing algorithms proposed in this paper can greatly improve the computing efficiency of block Jacobi algorithm and improve the performance of block Gauss–Seidel algorithm when the number of nodes is large. In addition, the adaptive load-balancing algorithms are especially effective when a case demands different computational power of OpenMC and FLUENT.

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Mochizuki, Hiroyasu. « Neutronics and thermal-hydraulics coupling analysis using the FLUENT code and the RELAP5-3D code for a molten salt fast reactor ». Nuclear Engineering and Design 368 (novembre 2020) : 110793. http://dx.doi.org/10.1016/j.nucengdes.2020.110793.

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Lian, Qiang, Wenxi Tian, Suizheng Qiu et G. H. Su. « Development of a three-dimensional method for thermal-hydraulics/neutronics coupling analysis and its application on CFETR helium-cooled solid breeder blanket ». International Journal of Advanced Nuclear Reactor Design and Technology 3 (2021) : 154–65. http://dx.doi.org/10.1016/j.jandt.2021.09.002.

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Yuan, Baoxin, Jie Zheng, Jian Wang, Herong Zeng, Wankui Yang, Huan Huang et Songbao Zhang. « Numerical Calculation Scheme of Neutronics-Thermal-Mechanical Coupling in Solid State Reactor Core Based on Galerkin Finite Element Method ». Energies 16, n^o 2 (5 janvier 2023) : 659. http://dx.doi.org/10.3390/en16020659.

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It is of practical significance to study the multi-physical processes of solid state nuclear systems for device design, safety analysis, and operation guidance. This system generally includes three multi-physical processes: neutronics, heat transfer, and thermoelasticity. In order to analyze the multi-physical field behavior of solid state nuclear system, it is necessary to analyze the laws of neutron flux, temperature, stress, and other physical fields in the system. Aiming at this scientific goal, this paper has carried out three aspects of work: (1) Based on Galerkin’s finite element theory, the governing equations of neutronics, heat transfer, and thermoelasticity have been established; (2) a neutronics-thermal-mechanical multi-physical finite element analysis code was developed and verified based on benchmark examples and third-party software for multi-physical processes; (3) for a solid state nuclear system with a typical heat pipe cooled reactor configuration, based on the analysis code developed in this work, the neutronics-thermal-mechanical coupling analysis was carried out, and the physical field laws such as neutron flux, temperature, stress, etc., of the device under the steady-state operating conditions were obtained; and (4) finally, the calculation results are discussed and analyzed, and the focus and direction of the next work are clarified.

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Cui, Shijie, Dalin Zhang, Jian Ge, Jie Cheng, Wenxi Tian, G. H. Su et Suizheng Qiu. « Development and application of a neutronics/thermal-hydraulics coupling optimization code for the CFETR helium cooled solid breeder blanket with mixed pebble beds ». Fusion Engineering and Design 125 (décembre 2017) : 24–37. http://dx.doi.org/10.1016/j.fusengdes.2017.10.020.

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Ferraro, Diego, Manuel García, Uwe Imke, Ville Valtavirta, Riku Tuominen, Jaakko Leppänen et Víctor Sanchez-Espinoza. « SERPENT/SUBCHANFLOW COUPLED BURNUP CALCULATIONS FOR VVER FUEL ASSEMBLIES ». EPJ Web of Conferences 247 (2021) : 04005. http://dx.doi.org/10.1051/epjconf/202124704005.

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The continuous improvement in nuclear industry safety standards and reactor designers’ and operators’ commercial goals represent a push for the development of highly accurate methodologies in reactor physics. This fact, combined with the availability of vast computational resources, allowed the development of a wide range of coupled state-of-the-art neutronic-thermal-hydraulic calculation tools worldwide during last decade. Under this framework, the McSAFE European Union project is a coordinated effort aimed to develop multiphysics tools based on Monte Carlo neutron transport and subchannel thermal-hydraulics codes, suitable for high-fidelity calculations for PWR and VVER reactors. This work presents the results for a pin-by-pin coupled burnup calculation using the Serpent 2 code (developed by VTT, Finland) and the subchannel thermal-hydraulics code SUBCHANFLOW (SCF, developed by KIT, Germany) for two different VVER-type fuel assembly types. For such purpose, a recently refurbished master-slave coupling scheme is considered, which provides several new features such as burnup and transient calculations capabilities for square and hexagonal geometries. Main aspects of this coupling are presented for this burnup case, showing some of the capabilities now available. On top of that, the obtained global results are compared with available published data from a similar high-fidelity approach for the same FA design, showing a good agreement. Finally, a brief analysis of the main resources requirement and main bottlenecks identification are also included. The results presented here provide valuable insights and pave the way to tackle the final goals of the McSAFE project, which includes full-core pin-by-pin depletion calculation within a fully coupled MC-TH approach.

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Zheng, Lei, Zhiyuan Feng et Kan Wang. « ON-THE-FLY INTERPOLATION OF CONTINUOUS TEMPERATURE-DEPENDENT THERMAL NEUTRON SCATTERING DATA IN RMC CODE ». EPJ Web of Conferences 247 (2021) : 09012. http://dx.doi.org/10.1051/epjconf/202124709012.

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Thermal neutron scattering data have an important influence on the high-fidelity neutronics calculation of thermal reactors. Due to the limited storage capabilities of computers, a discrete ACE representation of the secondary neutron energy and angular distribution has been used for Monte Carlo calculation since the early 1980s. The use of this discrete representation does not produce noticeable effects in the integral calculations such as keff eigenvalues, but can produce noticeable deficiencies for differential calculations. A new continuous representation of the thermal neutron scattering data was created in 2006, but was not widely known. Recently, the continuous representation of the thermal neutron scattering ACE data based on ENDF/B-Ⅷ.0 library was officially released and was available for all users. The new representation shows great difference compared with the discrete one. In order to utilize the more physical and rigorous representation data for high fidelity neutronic-thermohydraulic coupling calculation, the on-the-fly treatment capability was proposed and implemented in RMC code. The two-dimensional linear-linear interpolation method was used to calculate the inelastic scattering cross sections and the secondary neutron energies and angles. The on-the-fly treatment capability was tested by a pressurized water reactor assembly. Results show that the on-the-fly treatment capability has high accuracy, and can be used to consider the temperature feedback in the neutronic-thermohydraulic coupling calculations. However, the efficiency of the on-the-fly treatment still need to be improved in the near future.

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Sanchez-Espinoza, V. H., L. Mercatali, J. Leppänen, E. Hoogenboom, R. Vocka et J. Dufek. « THE McSAFE PROJECT - HIGH-PERFORMANCE MONTE CARLO BASED METHODS FOR SAFETY DEMONSTRATION : FROM PROOF OF CONCEPT TO INDUSTRY APPLICATIONS ». EPJ Web of Conferences 247 (2021) : 06004. http://dx.doi.org/10.1051/epjconf/202124706004.

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The increasing use of Monte Carlo methods for core analysis is fostered by the huge and cheap computer power available nowadays e.g. in large HPC. Apart from the classical criticality calculations, the application of Monte Carlo methods for depletion analysis and cross section generation for diffusion and transport core simulators is also expanding. In addition, the development of multi-physics codes by coupling Monte Carlo solvers with thermal hydraulic codes (CFD, subchannel and system thermal hydraulics) to perform full core static core analysis at fuel assembly or pin level has progressed in the last decades. Finally, the extensions of the Monte Carlo codes to describe the behavior of prompt and delay neutrons, control rod movements, etc. has been started some years ago. Recent coupling of dynamic versions of Monte Carlo codes with subchannel codes make possible the analysis of transient e.g. rod ejection accidents and it paves the way for the simulation of any kind of design basis accidents as an alternative option to the use of diffusion and transport based deterministic solvers. The H2020 McSAFE Project is focused on the improvement of methods for depletion considering thermal hydraulic feedbacks, extension of the coupled neutronic/thermal hydraulic codes by the incorporation of a fuel performance solver, the development of dynamic Monte Carlo codes and the development of methods to handle large depletion problems and to reduce the statistical uncertainty. The validation of the multi-physics tools developed within McSAFE will be performed using plant data and unique tests e.g. the SPERT III E REA test. This paper will describe the main developments, solution approaches, and selected results.

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Li, Shuzhou, Jingchao Feng, Liankai Cao, Muhammad Shehzad Khan et Hongli Chen. « A thermal neutronics coupling analysis method for lead based reactor core ». Annals of Nuclear Energy 107 (septembre 2017) : 82–88. http://dx.doi.org/10.1016/j.anucene.2017.04.021.

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Khan, Salah Ud-Din, Minjun Peng et Shahab Ud-Din Khan. « Neutronics and thermal hydraulic coupling analysis of integrated pressurized water reactor ». International Journal of Energy Research 37, n^o 13 (20 novembre 2012) : 1709–17. http://dx.doi.org/10.1002/er.2981.

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Freile, Ramiro, et Mark Kimber. « Influence of molten salt-(FLiNaK) thermophysical properties on a heated tube using CFD RANS turbulence modeling of an experimental testbed ». EPJ Nuclear Sciences & ; Technologies 5 (2019) : 16. http://dx.doi.org/10.1051/epjn/2019027.

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In a liquid fuel molten salt reactor (MSR) a key factor to consider upon its design is the strong coupling between different physics present such as neutronics, thermo-mechanics and thermal-hydraulics. Focusing in the thermal-hydraulics aspect, it is required that the heat transfer is well characterized. For this purpose, turbulent models used for FLiNaK flow must be valid, and its thermophysical properties must be accurately described. In the literature, there are several expressions for each material property, with differences that can be significant. The goal of this study is to demonstrate and quantify the impact that the uncertainty in thermophysical properties has on key metrics of thermal hydraulic importance for MSRs, in particular on the heat transfer coefficient. In order to achieve this, computational fluid dynamics (CFD) simulations using the RANS k-ω SST model were compared to published experiment data on molten salt. Various correlations for FLiNaK’s material properties were used. It was observed that the uncertainty in FLiNaK’s thermophysical properties lead to a significant variance in the heat coefficient. Motivated by this, additional CFD simulations were done to obtain sensitivity coefficients for each thermophysical property. With this information, the effect of the variation of each one of the material properties on the heat transfer coefficient was quantified performing a one factor at a time approach (OAT). The results of this sensitivity analysis showed that the most critical thermophysical properties of FLiNaK towards the determination of the heat transfer coefficient are the viscosity and the thermal conductivity. More specifically the dimensionless sensitivity coefficient, which is defined as the percent variation of the heat transfer with respect to the percent variation of the respective property, was −0.51 and 0.67 respectively. According to the different correlations, the maximum percent variations for these properties is 18% and 26% respectively, which yields a variation in the predicted heat transfer coefficient as high as 9% and 17% for the viscosity and thermal conductivity, respectively. It was also demonstrated that the Nusselt number trends found from the simulations were captured much better using the Sieder Tate correlation than the Dittus Boelter correlation. Future work accommodating additional turbulence models and higher fidelity physics will help to determine whether the Sieder Tate expression truly captures the physics of interest or whether the agreement seen in the current work is simply reflective of the single turbulence model employed.

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Hou, Jason, Maria Avramova et Kostadin Ivanov. « Best-Estimate Plus Uncertainty Framework for Multiscale, Multiphysics Light Water Reactor Core Analysis ». Science and Technology of Nuclear Installations 2020 (31 juillet 2020) : 1–18. http://dx.doi.org/10.1155/2020/7526864.

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Tremendous work has been done in the Light Water Reactor (LWR) Modelling and Simulation (M&S) uncertainty quantification (UQ) within the framework of the Organization for Economic Cooperation and Development (OECD)/Nuclear Energy Agency (NEA) LWR Uncertainty Analysis in Modelling (UAM) benchmark, which aims to investigate the uncertainty propagation in all M&S stages of the LWRs and to guide uncertainty and sensitivity analysis methodology development. The Best-Estimate Plus Uncertainty (BEPU) methodologies have been developed and implemented within the framework of the LWR UAM benchmark to provide a realistic predictive simulation capability without compromising the safety margins. This paper describes the current status of the methodological development, assessment, and integration of the BEPU methodology to facilitate the multiscale, multiphysics LWR core analysis. The comparative analysis of the results in the stand-alone multiscale neutronics phase (Phase I) is first reported for understanding the general trend of the uncertainty of core parameters due to the nuclear data uncertainty. It was found that the predicted uncertainty of the system eigenvalue is highly dependent on the choice of the covariance libraries used in the UQ process and is less sensitive to the solution method, nuclear data library, and UQ method. High-to-Low (Hi2Lo) model information approaches for multiscale M&S are introduced for the core single physics phase (Phase II). In this phase, the other physics (fuel and moderator), providing feedback to neutronics M&S in a LWR core, and time-dependent phenomena are considered. Phase II is focused on uncertainty propagation in single physics models which are components of the LWR core coupled multiphysics calculations. The paper discusses the link and interactions between Phase II to the multiphysics core and system phase (Phase III), that is, the link between uncertainty propagation in single physics on local scale and multiphysics uncertainty propagation on the core scale. Particularly, the consistency in uncertainty assessment between higher-fidelity models implemented in fuel performance codes and the rather simplified models implemented in thermal-hydraulics codes, to be used for coupling with neutronics in Phase III is presented. Similarly, the uncertainty quantification on thermal-hydraulic models is established on a relatively small scale, while these results will be used in Phase III at the core scale, sometimes with different codes or models. Lastly, the up-to-date UQ method for the coupled multiphysics core calculation in Phase III is presented, focusing on the core equilibrium cycle depletion calculation with associated uncertainties.

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Seungsu, Yuk, Tak Nam-il et Chang Jo Keun. « DEVELOPMENT OF PIN-LEVEL NEUTRONICS/THERMAL-FLUID ANALYSIS COUPLED CODE SYSTEM FOR A BLOCK-TYPE HTGR CORE ». EPJ Web of Conferences 247 (2021) : 02041. http://dx.doi.org/10.1051/epjconf/202124702041.

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Recently, the coupling between computer codes that simulate different physical phenomena has attracted for more accurate analysis. In the case of high-temperature gas-cooled reactor (HTGR), the coupling between neutronics and thermal-fluid analysis is necessary because of large change of temperature in the reactor core. Korea Atomic Energy Research Institute (KAERI) has developed the coupled code system between a reactor physics analysis code CAPP and a thermal-fluid system safety analysis code GAMMA+ for a block-type HTGR. The CAPP/GAMMA+ coupled code system provides more accurate block-wise distribution data than CAPP or GAMMA+ stand-alone analysis. However, the block-wise distribution data has the limitation in order to predict safety parameters such as the maximum temperature of the nuclear fuel. It is necessary to calculate refined distribution, for example, pin-level (fuel compact level) distribution. In this study, we tried to solve this problem by coupling CAPP and a high-fidelity thermal-fluid analysis code CORONA. CORONA can perform a high-fidelity thermal-fluid analysis of Computational Fluid Dynamics (CFD) level by dividing a block-type HTGR core into small lattices. On the other hand, CAPP can provide a pin power distribution. It is expected that the refined, more accurate distribution data for a block-type HTGR can be obtained by coupling these two codes. This paper presents the development of coupled code system between CAPP and CORONA, and then it is tested on a simple HTGR column problem with encouraging results.

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Al Zain, Jamal, O. El Hajjaji, T. El Bardouni et Y. Boulaich. « Coupling of Neutronics and Thermal-Hydraulic Codes for Simulation of the MNSR Reactor ». Nuclear Science and Engineering 193, n^o 11 (17 juin 2019) : 1276–89. http://dx.doi.org/10.1080/00295639.2019.1622927.

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Duan, Zimian, Jing Zhang, Yingwei Wu, Binqian Li, Mingjun Wang, Yanan He, Wenxi Tian, Siuzheng Qiu et G. H. Su. « Multi-physics coupling analysis on neutronics, thermal hydraulic and mechanics characteristics of a nuclear thermal propulsion reactor ». Nuclear Engineering and Design 399 (décembre 2022) : 112042. http://dx.doi.org/10.1016/j.nucengdes.2022.112042.

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Romano, Paul K., Steven P. Hamilton, Ronald O. Rahaman, April Novak, Elia Merzari, Sterling M. Harper et Patrick C. Shriwise. « DESIGN OF A CODE-AGNOSTIC DRIVER APPLICATION FOR HIGH-FIDELITY COUPLED NEUTRONICS AND THERMAL-HYDRAULIC SIMULATIONS ». EPJ Web of Conferences 247 (2021) : 06053. http://dx.doi.org/10.1051/epjconf/202124706053.

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While the literature has numerous examples of Monte Carlo and computational fluid dynamics (CFD) coupling, most are hard-wired codes intended primarily for research rather than as standalone, general-purpose codes. In this work, we describe an open source application, ENRICO, that allows coupled neutronic and thermal-hydraulic simulations between multiple codes that can be chosen at runtime (as opposed to a coupling between two specific codes). In particular, we outline the class hierarchy in ENRICO and show how it enables a clean separation between the logic and data required for a coupled simulation (which is agnostic to the individual solvers used) from the logic/data required for individual physics solvers. ENRICO also allows coupling between high-order (and generally computationally expensive) solvers to low-order “surrogate” solvers; for example, Nek5000 can be swapped out with a subchannel solver. ENRICO has been designed for use on distributed-memory computing environments. The transfer of solution fields between solvers is performed in memory rather than through file I/O.We describe the process topology among the different solvers and how it is leveraged to carry out solution transfers. We present results for a coupled simulation of a single light-water reactor fuel assembly using Monte Carlo neutron transport and CFD.

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