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Analysis of Fission Matrix Databases with Nonuniform Fuel Temperature Profiles for Static and Transient Calculations Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-22 Maximiliano Dalinger, William Walters
Monte Carlo codes are the most accurate way to solve the neutronics in a reactor core but can be computationally expensive, especially for when feedback effects are considered or for transient calc...
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Optimizations of the Accelerated Axial Polynomial Method of Characteristics of the APOLLO3® Deterministic Neutron Transport Code Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-22 Arthur Le Bars, Andrea Gammicchia, Simone Santandrea
For some years now, the TDT (two- and three-dimensional transport) solver of the APOLLO3® deterministic neutron transport code has been able to perform lattice calculations on three-dimensional ext...
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On the Possibility of Achieving Optimal Operating Modes of Hyperspeed Model Gas Centrifuges: 3D Numerical Simulation Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-22 S. V. Bogovalov, I. V. Tronin, A. V. Vasilyev
In this paper, numerical simulation methods are used to study issues related to the optimal operating modes of hyperspeed (rotor velocity 1000 m/s and above) model gas centrifuges (GCs) of various ...
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MaCaw: Domain-Decomposed Monte Carlo Neutral Particle Transport on Unstructured Mesh in MOOSE Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-17 G. Giudicelli, R. Crowder, L. Harbour, D. Gaston
MaCaw is a Multiphysics Object-Oriented Simulation Environment (MOOSE)–based application that enables domain-decomposed neutral particle transport calculations in MOOSE. It leverages MOOSE’s ray-tr...
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A Multilevel Approach for Direct Core Calculation Schemes Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-15 Antonio Galia
A method of dynamic homogenization was recently proposed as an alternative technique for three-dimensional (3D) core calculations using a direct approach. This technique allows for producing homoge...
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Hot Hydrogen Testing of W-Coated UN Kernels in a Mo30W Matrix Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-15 Arne Cröll, Jamelle K. P. Williams, Brian Taylor, Martin P. Volz, Christopher McKinney, Timothy Coons, Jhonathan Rosales
Ceramic uranium mononitride (UN) is being considered as a reactor fuel for nuclear thermal propulsion. To avoid or reduce the dissociation of UN at the high temperatures needed, embedding it in a m...
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Convergence Study of 2D Transport Method for Multigroup k-Eigenvalue Problems with High-Order Scattering Source Using Fourier Analysis Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-12 Boran Kong, Longfei Xu, Baiwen Li
The convergence behavior of a two-dimensional (2D) transport method has been derived by Fourier analysis for single-group problems with isotropic sources. However, in real calculation, to pursue pr...
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Comparing Coupled Multiphysics Simulations of Pump-Driven Transients for a Pressurized Heavy Water Reactor Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-12 Santiago Bazzana, Juan I. Beliera, Dumitru Serghiuta, Alexandre Trottier
Comparison of results for global responses predicted by different multiphysics simulations of benchmark problems may fail to reveal potentially significant local modeling issues. An examination of ...
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Predicting Model Biases of the APOLLO3 Neutronics Code Using Machine Learning: Toward a Multiscale and Multifidelity Approach Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-10 Claire Terrazzoni, Laurent Buiron, Jean-Marc Palau
As part of the verification, validation, and uncertainty quantification process applied to neutronics deterministic codes, there is a requirement to expand the validation domain, especially to acco...
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Demonstrating a Quantitative and Systematic Approach to Reducing Excess Conservativism in Nuclear Criticality Safety Analyses Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-10 Sharbrenai Anise Holyk, Robert B. Hayes
Although reducing conservatism would alleviate unnecessary constraints in processing, storage, transportation, and disposal of nuclear materials, excessively conservative approaches are still utili...
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A Diffusion Scale Approximation for Stochastic Fission Chains with State Feedback Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-04 Chen Dubi, Anil K. Prinja
Modeling and simulation of stochastic fission chains, often referred to as zero power reactor noise or stochastic transport, is a central topic in nuclear science and engineering, with important pr...
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Some Results Concerning the Singular Solution of the Conservative Transport Equation in Spherical Geometry Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-03 R. D. M. Garcia
We examine in this work one of the exact solutions of the conservative transport equation for isotropic scattering in spherical geometry, specifically the solution that is singular at the origin an...
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Implementation and Testing of Generalized Perturbation Theory Capabilities in TRIPOLI-4® Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-03 Alexis Jinaphanh, Giorgio Valocchi, Andrea Zoia
In this paper, we describe the implementation, verification, and numerical validation of Generalized Perturbation Theory (GPT) capabilities in the TRIPOLI4® Monte Carlo code. The GPT is applied to ...
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Prototyping of a Machine Learning–Based Burnup Measurement Capability for Pebble Bed Reactor Fuel Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-02 Nick Rollins, India Allan, Jason Hou
The pebble bed reactor is a unique reactor design due to its capacity for continuous multipass circulation of the fuel elements, without causing interruption to reactor operation, with the assistan...
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Coupled Neutronics-Thermal-Hydraulic Modeling of a Molten Salt Reactor: The Aircraft Reactor Experiment Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-01 François Martin, André Bergeron, Guillaume Campioni, Yannick Gorsse, Nathan Greiner, Elsa Merle
The CEA multiphysics tool combining the deterministic neutronics code APOLLO3® and the computational fluid dynamics (CFD) platform TRUST/TrioCFD is used to model the first-ever-built molten salt nu...
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Reactor Physics Monitoring of a Source-Driven Subcritical System in Stationary State by Deterministic and Probabilistic Deep Neural Networks Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-01 Ronald Daryll E. Gatchalian, Pavel V. Tsvetkov
Reactivity measurement methods, like the Amplified Source Method (ASM), link observable quantities to integral physics parameters characterizing subcritical assemblies (SCAs). These methods were mo...
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Experimental Covariance Determination for Critical Integral Experiments Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-04-01 Eric Aboud, Jesse Norris, Daniel Siefman
Integral benchmarks for criticality safety and nuclear data validation require expensive uncertainty quantification studies. In general, uncertainty quantification techniques ignore correlations be...
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Three-Dimensional Full-Core BEAVRS Using OpenMOC with Transport Equivalence Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-19 G. Giudicelli, B. Forget, K. Smith
Using an optimized implementation of the three-dimensional (3D) method of characteristics for neutron transport, along with a novel equivalence method for transport calculations that was designed t...
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On the Assumptions Behind Statistical Sampling: A 235U Fission Yield Uncertainty Propagation Case Study Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-19 Enrica Belfiore, Federico Grimaldi, Luca Fiorito, Pablo Romojaro, Gašper Žerovnik, Pierre-Etienne Labeau, Sandra Dulla
Monte Carlo sampling is frequently employed for uncertainty quantification in depletion calculations. Several assumptions are needed to perform this analysis. In this work, an assessment of these a...
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Multiscale Modeling of Flow in Rod Bundles: From Direct Numerical Simulation and Subchannel to Coarse-Mesh CFD Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-14 Adam R. Kraus, Elia Merzari
Fast and accurate evaluation of flow and heat transfer phenomena in rod bundles is a problem of long-standing interest in nuclear engineering. Computational fluid dynamics (CFD) can provide accurat...
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Improved Collision Probability Analysis of the Double-Heterogeneity Problem Based on Matrix Chord Length Correction Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-14 Zhaoyu Liang, Ding She, Yutong Wen, Lei Shi, Zuoyi Zhang
Dispersion fuel exhibits excellent safety performance and effectively reduces the risk of radioactive leakage, making it widely applied in high-temperature gas-cooled reactors (HTGRs) and other adv...
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A Consequence-Informed Licensing Path Selection for the Design of Physical Protection Systems at Commercial Nuclear Power Facilities Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-12 Thomas Freyman, Karen Vierow Kirkland
The survivability of the domestic nuclear power industry depends on the cost-competitiveness of safe and secure nuclear power generation. Advanced reactor design concepts aim to have increased safe...
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Analytic Error Analysis of the Partial Derivatives Cross-Section Model—I: Derivation Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-05 Thomas Folk, Siddhartha Srivastava, Dean Price, Krishna Garikipati, Brendan Kochunas
Accurate assessment of uncertainties in cross-section data is crucial for reliable nuclear reactor simulations and safety analyses. In this study, we focus on the interpolation procedure of the par...
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User Impressions and Gait Analysis of Exoskeleton Device Usage in Generalized Tank Farm Activities Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-05 Janelle Bottom, David Wood, Tamzidul Mina, Savannah Bradley, Michal Rittikaidachar, Alexandria Miera, Jason Wheeler
Tank farm workers involved in nuclear cleanup activities perform physically demanding tasks, typically while wearing heavy personal protective equipment (PPE). Exoskeleton devices have the potentia...
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ORCA: A Tool for Radiological Consequences for Accidental Releases Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-04 P. Maka, E. Van Heerden, M. Rezaee
Evaluating atmospheric dispersion and radiological doses in the vicinity of buildings is required for small modular reactors (SMRs) because of the reduced size of their exclusion area boundary. The...
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New RMC Energy Deposition Treatment and its Application in Multi-Physics Simulation Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-01 Hao Luo, Kaiwen Li, Nan An, Shanfang Huang, Kan Wang
Accurate estimation of energy deposition is important in core physics and severe accident analyses for design optimizations. In this study, a new energy deposition treatment is implemented in the R...
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On the Use of the Jacobian-Free Newton Krylov Method to Generate One-Group Discontinuity and Super Homogenization Factors for Full-Core Neutron Diffusion Simulations Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-03-01 Bailey Painter, Dan Kotlyar
Coupled Monte Carlo (MC) and thermal-hydraulic analysis is valuable as a design or reference tool but can be slow, especially when implemented in a Picard iteration. Previous work has developed a n...
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Analytic Error Analysis of the Partial Derivatives Cross-Section Model—II: Numerical Results Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-29 Thomas Folk, Siddhartha Srivastava, Dean Price, Krishna Garikipati, Brendan Kochunas
Accurately predicting errors incurred in a cross-section model for two-step reactor analysis enables the development of optimal case matrices and more efficient cross-section models. In a companion...
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A Review of Neutron–Gamma-Ray Discrimination Methods Using Organic Scintillators Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-29 Imane Ahnouz, Hanan Arahmane, Rajaa Sebihi
Neutron detection is increasingly vital in various fields such as homeland security, medical sciences, and high-energy physics. However, interference from accompanying gamma rays poses a critical c...
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Examining the Contributions of Various Source Layers and Transport Pathways to Dose Rates Exterior to a Low- and Intermediate-Level Radioactive Waste Storage Facility Using ADVANTG/MCNP Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-29 Bo-Lun Lai, Szu-Li Chang, Rong-Jiun Sheu
An in-depth shielding analysis of a large indoor facility designed for storing low- and intermediate-level radioactive waste generated from nuclear power plant decommissioning was performed. The fa...
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Pre-Experimental Assessment of Uranyl Nitrate Solution Irradiation in Kartini Reactor Facilities Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-28 Feryantama Putra, Syarip, Sihana
Medical radioisotope production using neutron irradiation via fission reaction requires a sufficient neutron source. The Kartini reactor has been proposed and studied to become a neutron source for...
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The Effect of Branchless Collisions and Population Control on Correlations in Monte Carlo Power Iteration Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-28 Theophile Bonnet, Hunter Belanger, Davide Mancusi, Andrea Zoia
The investigation of correlations in Monte Carlo power iteration has long been dominated by the question of generational correlations and their effects on the estimation of statistical uncertaintie...
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Recovering From Structural Instability in Nuclear Power Design Optimization Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-27 Dominic J. Brennan, Geoffrey T. Parks
The Scientific Method—parsing a problem into isolated subproblems—is often necessarily employed in optimization efforts to reduce large and complex problems into more tractable parcels. However, ad...
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Resonance Calculation of FCM Fuel Based on the Global-Local Method Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-27 Siyu Yi, Zhouyu Liu, Kunpeng Wang, Wei Shen, Tiejun Zu, Liangzhi Cao, Hongchun Wu
Fully ceramic microencapsulated fuels is a promising potential fuel for pressurized water reactors because of its inherent safety, but double heterogeneity (DH) brings challenges to the neutronic c...
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Verification of Griffin-Pronghorn-Coupled Multiphysics Code System Against CNRS Molten Salt Reactor Benchmark Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-23 Mustafa K. Jaradat, Namjae Choi, Abdalla Abou-Jaoude
The molten salt reactor (MSR) flowing-fuel simulation capability of the Griffin-Pronghorn-coupled multiphysics code system developed by Idaho National Laboratory (INL) was verified against the Cent...
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Optimizing the Fixed Number Detector Placement for the Nuclear Reactor Core Using Reinforcement Learning Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-23 Kai Tan, Fan Zhang
Monitoring three-dimensional flux distribution in a nuclear reactor core is essential for improving safety and economics, which requires strategically placed in-core detectors. However, the deploym...
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Post-Neutron Mass Yield Distribution in the Epi-Cadmium Neutron-Induced Fission of 239Pu Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-22 H. Naik, R. J. Singh, S. P. Dange, W. Jang
In the epi-cadmium neutron-induced fission of 239Pu, cumulative and independent yields of various fission products within the mass ranges of 83 to 117 and 123 to 156 have been measured by using an ...
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Thermal-Hydraulic Analysis of Pebble Bed High Temperature Gas-Cooled Reactor After Shutdown Based on FLUENT Software Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-22 Junbing Zhu, Tianyun Liu, Zhiyuan Ren
In order to provide a reliable tool for thermal-hydraulic simulation of pebble bed high temperature gas-cooled reactors (HTGRs), a two-dimensional model was developed based on the porous media mode...
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A Variant of the Second-Moment Method for k-Eigenvalue Calculations Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-21 Connor Woodsford, James Tutt, Jim E. Morel
The second-moment (SM) method is a linear variant of the quasi-diffusion (QD) method for accelerating the iterative convergence of Sn source calculations. It has several significant advantages rela...
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A Precision Benchmark Suite for Nuclear Reactor Point Kinetics Equations via Converged Accelerated Taylor Series (CATS) Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-21 B. D. Ganapol
Extreme benchmarks of 10 or more places for the point kinetics equations for time-dependent nuclear reactor power transients are rare. Therefore, to establish an extreme benchmark, we employ a Tayl...
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Research on Nuclear Power Plant Operational Task Complexity Assessment Method Based on Group Decision Making and Deep Learning of EEG Signals Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-20 Dongliang Zhang, Jie Wu, Kunpeng Wu, Hanming Tao
This study aims to explore the correlation between the operational task complexity of nuclear power plant (NPP) operators and electroencephalogram (EEG) features. Initially, we segmented EEG signal...
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Study on Unstructured Mesh–Based Monte Carlo/Deterministic Coupled Particle Transport Calculation Method Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-09 Hanlin Shu, Liangzhi Cao, Qingming He, Qi Zheng, Tao Dai
The unstructured mesh (UM)–based Monte Carlo (MC) method can utilize modern computer-aided-design/computer-aided-engineering platforms to obtain geometric models with reduced human effort and is ca...
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Effectiveness of Radiation Transport Variance Reduction Methods for Wide-Area Environmental Contamination Assay Applications Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-07 E. Asano, S. Dewji
This study compares the accuracy, efficiency, and reliability of variance reduction (VR) methods for Monte Carlo radiation transport simulations involving wide-area ground plane (i.e., “surface”) a...
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Combining Similarity Measures and Left-Right Hidden Markov Models for Prognostics of Items Subjected to Perfect and Imperfect Maintenance Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-02-07 Mattia Zanotelli, J. Wesley Hines, Jamie B. Coble
In the nuclear industry, high system reliability requirements are essential since in-service failure can result in undesirable consequences in terms of costs or safety. However, the current approac...
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Neutron Total Cross Sections of 232Th, 237Np, 235,238U, and 239Pu from 3 to 230 MeV Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-01-25 P. W. Lisowski, M. S. Moore, G. L. Morgan
Neutron total cross-section data for 232Th, 237Np, 235,238U, and 239Pu were measured over the energy range of a few MeV up to 230 MeV at the Weapons Neutron Research Facility of the Los Alamos Neut...
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Parallel Processing of Image Databases for Accelerated Morphological Analysis Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-01-25 Edward Goodell, Glenn Sjoden, Reid Porter, Luther McDonald IV, Kari Sentz
Nuclear forensics relies on different signatures to identify the source of nuclear material. Such signatures include crystalline structure, chemical composition, and particle morphology. One way to...
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Quantifying Safety Significance: An In-Depth Analysis of Importance Measures in Level 1 PSA for the VVR 10-MW Water-Water Research Reactor Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-01-25 Rex Gyeabour Abrefah, Felix Ameyaw
The effectiveness of contemporary strategies for conducting fault tree/event tree (FTET) analyses within the realm of probabilistic risk assessment has recently come under rigorous examination. In ...
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Flux Flattening Large Heavy Water Power Reactors with Accelerator-Driven Photoneutron Source Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-01-25 Arief Rahman Hakim, Douglas A. Fynan
Flux flattening and power uprating of large heavy water power reactors (HWRs) are demonstrated as an application of an accelerator-driven photoneutron source (ADS) in the ADS-CANDU concept where an...
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Evaluation of Kinetic Parameters RSG-GAS Reactor Equilibrium Silicide Core Using Continuous-Energy Monte Carlo Serpent 2 Code Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-01-22 Surian Pinem, Liem Peng Hong, Wahid Luthfi, Tukiran Surbakti, Donny Hartanto
The purpose of this study is to determine the kinetic parameters of the RSG-GAS equilibrium core. The calculated kinetic parameters are the effective delayed neutron fraction βeff, the neutron gene...
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Numerical Evaluation of Sodium Droplet Swarm Combustion and Its Modeling Optimization Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-01-17 Cheng Peng, Jian Deng, Jiang Wu
Because of its superior thermal-hydraulic qualities, liquid sodium has been applied to a variety of industries, including energy storage, solar energy, sodium-cooled fast reactors, and aerospace. H...
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An Innovational Jacobian-Split Newton-Krylov Method Combining the Advantages of the Jacobian-Free Newton-Krylov Method and the Finite Difference Jacobian-Based Newton-Krylov Method Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-01-10 Lixun Liu, Han Zhang, Xinru Peng, Qinrong Dou, Yingjie Wu, Jiong Guo, Fu Li
The Jacobian-free Newton-Krylov (JFNK) method is a widely used and flexible numerical method for solving the neutronic/thermal-hydraulic coupling system. The main property of JFNK is that the Jacob...
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Initial Verification and Validation of a New CASMO5 JENDL-5 Nuclear Data Library for Typical LWR Applications Nucl. Sci. Eng. (IF 1.2) Pub Date : 2024-01-05 Tomoaki Watanabe, Kenya Suyama, Kenichi Tada, Rodolfo M. Ferrer, Joshua Hykes, Charles A. Wemple
A new nuclear data library for the advanced lattice physics code CASMO5 has been prepared based on JENDL-5. In JENDL-5, the range of data, such as the number of nuclides, has been dramatically expa...
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Comparison of the Baseline USL Calculation Methods for Loosely Coupled and Novel Neutronic Systems Nucl. Sci. Eng. (IF 1.2) Pub Date : 2023-12-22 Bobbi Riedel, Christopher M. Perfetti, Forrest B. Brown
The goal of this study is to evaluate the accuracy of different upper subcritical limit (USL) calculational methods for loosely coupled and novel neutronic systems. This study varied the separation...
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Numerical Simulation of Turbulent Flow and Friction Characteristics Through a Loop of Narrow Rectangular Channel Under Rolling Motions Nucl. Sci. Eng. (IF 1.2) Pub Date : 2023-12-21 Husnain Murtaza, Muhammad Abdul Basit, Romana Basit, Wenxi Tian
Interaction of prevailing ocean waves and wind with the platforms containing the small modular reactors (SMRs) employed in marine environments may significantly alter the flow and friction characte...
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Feasibility of Using Mixed-Oxide Fuel for a Pressurized Water Reactor Instead of Traditional UO2 Fuel Material Nucl. Sci. Eng. (IF 1.2) Pub Date : 2023-12-21 Mohy Sabry, Neveen S. Abed, Ahmed Omar, Moamen G. El-Samrah, Mohamed Y. M. Mohsen
This study examines the feasibility of utilizing mixed-oxide fuel [(U0.9, rgPu0.1) O2] instead of traditional UO2 in nuclear reactors. The utilization of (U0.9, rgPu0.1) O2 is particularly signific...
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Calculation of Thermalized Component of Neutron Spectra in a D-D Neutron Generator Nucl. Sci. Eng. (IF 1.2) Pub Date : 2023-12-19 Nnaemeka Nnamani
The results of the thermalized flux calculation that incorporate radiative capture reactions in the presence and absence of polyethylene blocks that form an enclosure for a deuteron-deuteron (D-D) ...
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Development of HELIOS/BIPR/PARCS/MCNP6 Computation Route for WWER RPV Neutron Fluence Analysis and Validation Against Ex-Vessel Detector Measurement Data Nucl. Sci. Eng. (IF 1.2) Pub Date : 2023-12-19 S. Bznuni, A. Ugujyan, A. Amirjanyan, P. Kohut
A computational route was developed for precise calculation of fast neutron fluence on a WWER-type reactor pressure vessel (RPV). The method is based on the transfer of neutronics data from HELIOS-...
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CFD Analysis of Effects of Grid Spacer Vane on Thermal-Hydraulic Performance of Fluid in a 5×5 Fuel Channel Nucl. Sci. Eng. (IF 1.2) Pub Date : 2023-12-14 Satish Kumar Dhurandhar, S. L. Sinha, Shashi Kant Verma
A grid spacer with a vane is an influential segment in reactor fuel channels. A vane produces significant effects on flow mixing and augmentation of heat transfer in subchannnels. The purpose of th...
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Convergence Analysis for the CMFD Accelerated Linear Axial Expansion Transport Method Based on Fourier Analysis Nucl. Sci. Eng. (IF 1.2) Pub Date : 2023-12-13 Xinyu Zhou, Kun Liu, Haitao Ju, Chen Zhao, Hongbo Zhang, Bo Wang, Wenbo Zhao, Zhang Chen
The linear axial expansion transport method avoids the negative source problem caused by transverse leakage in the traditional two-dimensional/one-dimensional (2D/1D) transport method and has bette...
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Analysis of the Use of Different Backing Layer Materials in the Design of Alpha-Decay Sails for Interplanetary and Interstellar Travel Nucl. Sci. Eng. (IF 1.2) Pub Date : 2023-12-13 Rio Quinn
This paper analyzes the design of an alpha-decay sail. While previous studies have been conducted on the feasibility of alpha-decay sails, none have investigated the effects of different backing ma...