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The behavior of a jet passing through a grid-type obstacle: An experimental investigation Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-13 Satoshi Abe, Yasuteru Sibamoto
During a severe accident in a nuclear power plant, jets released from the primary system exhibit complex thermohydraulic behavior due to buoyancy effects and impingement on internal obstacles such as inner walls and floors. Thus, the obstacle-influenced jets are of interest in recent research activities. This paper describes an experimental investigation of the behavior of jets passing through a grid-type
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Itô-calculus based mathematical models for stochastic nuclear reactor kinetics and dynamics simulations of low neutron source nuclear power plant (NPP) start-up Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-12 T.L. Gordon, M.M.R. Williams, M.D. Eaton, P. Haigh
This paper investigates the effect thermal feedback has on the stochastic nuclear reactor dynamics of low neutron source nuclear power plant (NPP) start-ups. Stochastic mathematical and computational models are required to determine the probability of a stochastic power surge occurring during nuclear reactor start-up that would damage the nuclear fuel. The aim is to design the nuclear reactor, the
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Multiphysics and multi-scale design and analysis of nuclear thermal propulsion pellet bed reactor based on spherical cermet fuel element Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-12 Linyuan Lu, Shiqi Sun, Bin Tang, Guoqiang Wang, Ye Dai, Yajuan Zhong, Jun Lin
Due to the advantages of high thrust and high specific impulse, nuclear thermal propulsion (NTP) is a preferred near-term rocket engine technology for future Mars missions. This article introduces the reactor design based on spherical cermet fuel element, with the aim of studying the feasibility of applying this new fuel in NTP system. First, design requirements are determined based on the unique features
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Assessment of radiological doses of raw building materials and CEN room model using RESRAD-BUILD Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-11 Nguyen Quang Dao, Vu Ngoc Ba, Phan Thi Xuan Mai, Truong Thi Hong Loan
Environmental radioactivity in building materials can pose radiological risks to residents by exposure to the emitted gamma radiation and radon. In this study, we evaluated the radioactivity concentration of 53 raw building material samples (fly ash, cement, sand, and gravel) and estimated the dosimetric quantities in outdoor exposure scenarios for residents. The concentrations ranged from 40.1 to
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Nuclear power plant pipeline detection robot based on a new radiation-proof material Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-10 Tengfei Zheng, Yupu Liang, Zhikang Li, Xudong Wang, Wen Zhou, Zhikai Zhang, Chaohui Wang, Guang Hu
Soft robots are gradually gaining recognition in recent years for their ability to work in extreme environments. However, radiation shielding is a major issue when designing soft robots for use in nuclear environments. In this paper we present an effective soft robot design and develop new radiation shielding materials that enable the robot to perform a variety of tasks in the complex pipelines of
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Evaluation of existing correlations and development of a new correlation for single-phase friction factor in rod bundle assembly Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-07 Vikrant Siddharudh Chalgeri, Xiuzhong Shen, Toshihiro Yamamoto
Single-phase water flow in rod bundles plays a crucial role in removing the heat generated in the nuclear reactor core. The design and safety analysis of light-water nuclear reactors require us to know the pressure drop experienced by the coolant in a fuel assembly. The friction factor, a key parameter to determine the pressure drop, is influenced by various factors, including the geometry of the rod
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Forward and inverse predictive transient models of TREAT using surrogate reactivity models Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-07 Mustafa K. Jaradat, Sebastian Schunert, Frederick N. Gleicher, Vincent M. Labouré, Mark D. DeHart
In this work, we present two novel approaches for predicting the power evolution and control rod height of the Transient Reactor Test Facility (TREAT) to support experiment modeling; specifically, transient analysis of the NASA-sponsored Sirius series of experiments. These approaches utilize steady-state Monte Carlo model, point kinetics model, and surrogate models to predict power evolution and control
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Empirical model of flow-accelerated corrosion at elbow of carbon steel pipeline based on dimensional analysis Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-06 Guozhi Bao, Weiguang Qin, Dailong Pan, Xiaodong Si
Flow-accelerated corrosion (FAC) can result in continuous thinning of the wall thickness of water and steam-water transportation pipelines, thereby significantly compromising the safety of related systems. This study aims to investigate the behavior of FAC at a 90° elbow made of carbon steel using an array electrode technique and computational fluid dynamics (CFD) simulation. The electrochemical measurements
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Sensitivity study of core characteristic parameter of ATF loaded APR1400 core and cycle length compensation by enrichment adjustment Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-06 Kibeom Park, Tongkyu Park, Sung Kyun Zee
Following the Fukushima Daiichi accident, efforts have been made in a variety of disciplines to respond to nuclear accidents. From the perspective of nuclear fuel, various approaches have been attempted, and development has been made as an Accident Tolerant Fuel (ATF). Advanced fuel technology, often known as advanced fuel concepts, is ATF. The ATF is a type of fuel designed to be more resistant to
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Stochastic model updating for analysis of a nuclear containment vessel under internal pressure Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-06 Meng-Yan Song, Yu-Xiao Wu, De-Cheng Feng, Di Jiang, Pei-Yao Zhang
The prestressed concrete containment vessel (PCCV) is a critical safety measure for nuclear power plants. In the analysis of PCCV, the parameters of PCCV are usually determined through experience and a small number of experiments, which are highly accidental, and this paper provides a new idea for clarifying the scope of PCCV material parameters and clarifying some parameters that are important to
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Model order reduction of a once-through steam generator via dynamic mode decomposition Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-06 Yifan Xu, Minjun Peng, Antonio Cammi, Carolina Introini, Genglei Xia
Once-Through Steam Generators (OTSG), as the heat exchanger and the radioactive shield that connecting the primary system and the secondary system, is of vital importance for nuclear safety and economy. Whereas large-scale and coupling simulation models can provide high-fidelity estimations of the flow and heat exchange in OTSGs, there is an extra computational burden when applied to multi-query tasks
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Impact of the extended computational domain at inlet and outlet on natural convection in a vertical rectangular channel Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-06 Yongan Ji, Changwei Li, Ming Ding, Zehua Guo, Zhongning Sun
The natural convection flow in the vertical rectangular channels is frequently encountered in many engineering applications. Currently, most relevant simulation studies neglected the influence of the extended domain on the calculation results, it could cause the heat transfer coefficient to deviate by more than 10% from the actual result. This paper aims to study the influence of the extended domain
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Small rotational fuel-shuffling breed-and-burn sodium-cooled and nitride-fueled fast reactor with LBE reflectors Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-04 Tsendsuren Amarjargal, Jun Nishiyama, Toru Obara
In this work, a Rotational Fuel-shuffling Breed-and-Burn fast reactor with nitride fuel and sodium coolant (RFBB-NS) was analyzed. The purpose of the study was to design an RFBB-NS, including assembly duct and control rod assemblies, and to show the feasibility. Monte Carlo simulation code Serpent was used for neutronic and fuel burnup analyses and COMSOL Multiphysics was used for heat removal analysis
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Development of coarse mesh finite difference acceleration in the three-dimensional discrete-ordinates discontinuous finite element transport code TARS Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-04 Hu Zhang, Guangchun Zhang, Henglin Hu
The coarse mesh finite difference method (CMFD) is implemented to expedite the computation of the 3D S-DFEM code TARS, supporting both cuboid and hexagonal prism coarse cells. To enhance iteration stability, the optimally diffusive CMFD (odCMFD) technique is implemented. Moreover, in order to make the odCMFD method more straightforward to implement, we propose the technique of using only one uniform
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Investigation of the second-order nodal expansion method for the forward and adjoint neutron flux in the hexagonal geometry Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-03 Sajjad Abbasi Fashami, Mahdi Zangian, Abdolhamid Minuchehr, Ahmadreza Zolfaghari
The main purpose of this paper is to infer the second-order nodal equations in hexagonal-z geometries so that the forward and adjoint neutron parameters of the reactor core can be calculated more accurately. To achieve this goal, we developed a multigroup diffusion code (NRCC-Hex) for hexagonal-z geometries which can approximate the three-dimensional forward and adjoint neutron flux, based on the higher-orders
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Numerical study on surface corrosion deposition of fuel elements and its influence on flow heat transfer Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-03 Yandong Hou, Tianbo Chen, Weichao Li, Chuntian Gao, Bowen Chen, Chao Zhang, Yan Xiang
Corrosion of pressurized water reactors (PWR) in nuclear power plants can lead to serious safety hazards. This study aims to analyze the deposition of corrosion products using FLUENT software. Deposition models and thermal resistance models were developed, and the effects of deposits on the reactor’s thermal–hydraulic characteristics were evaluated. Additionally, the impact of various parameters on
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Numerical simulation investigations on the flow induced vibration of a nuclear fuel rod with relaxed clamping structures Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-01 Yongjun Jiao, Yihu Wang, Wei Li, Meiyin Zheng, Jing Zhang, Yingwei Wu, GuangHui Su, Suizheng Qiu
Grid-to-rod fretting (GTRF) wear is an important mechanism of nuclear breakage and fuel failure, which relates to the multi-physics field of fluid, structure, and irradiation. In this work, numerical simulation on the flow induced vibration of a nuclear fuel rod with relaxed clamping structures under the irradiation effect has been investigated by ANSYS FLUENT and ANSYS Workbench. The user’s defined
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Study of microstructure and properties of SiCf/SiC-Zr composite cladding prepared through PIP Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-01 Ying Meng, Yueqing Qian, Haining Feng, Yiming Gao, Yongheng Lu
As one of the new research direction of the fourth generation cladding material, the interface strength of SiC/SiC-Zr and the sustain properties of Zr alloy is the key point of the research. The study utilizes the third generation SiC fiber to form the 2.5D wave structure among the outside layer of Zr alloy tube and prepare the SiC/SiC-Zr composite material cladding tube with PyC interface through
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Spectrum features of the eigenvalue formulations in neutron transport Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-01 Nicolò Abrate, Sandra Dulla, Piero Ravetto, Paolo Saracco
The steady state neutron transport equation can be studied resorting to different eigenvalue formulations, useful to investigate the criticality state of a nuclear system. In this respect, the knowledge of the eigenvalue spectra featuring the main eigenproblems is fundamental, from a practical standpoint, to guide the eigenvalue solvers searching for the dominant eigenpairs. The study of the eigenvalue
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Multi-input and multi-output shielding calculation agent model method based on improved neural network with Swish-NNTA Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-03-01 Long Gui, Yingming Song, Xiaomeng Li, Weiwei Yuan
Researchers have utilized neural networks as agent models to predict shielded data sets for reactor shielding designs. However, previous models faced challenges related to unbalanced multi-scale characteristics in input and output data sets, resulting in significant generalization errors. Additionally, these models did not consider source term parameters like energy spectrum probability density distribution
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Intelligent Radiation: A review of Machine learning applications in nuclear and radiological sciences Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-28 Abbas J. Jinia, Shaun D. Clarke, Jean M. Moran, Sara A. Pozzi
Modern advancements in computing power and the ability of machine learning (ML) to model complex relationships between input and output have opened new prospects for data processing. This ML technology finds applications in nuclear and radiological sciences to extract meaningful information from data and drive intelligent decision-making. The literature review performed in the present manuscript encompasses
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Conceptual study of a new space nuclear propulsion system based on direct propulsion of fission fragments Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-28 Dacai Zhang, Ganglin Yu, Minzhi Xiong, Xirui Zhang, Guanghui Zhong, Kan Wang
Deep space exploration holds immense significance for the future of space exploration, and it requires propulsion systems with high specific impulse characteristics to accomplish long-duration space missions. This paper focuses on studying a space nuclear propulsion system based on the direct propulsion of fission fragments. Am-242 m is utilized as fuel, and two system models, namely uniform magnetic
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Experimental study of turbulent flow in lower plenum with flow skirt of reactor pressure vessel of pressurized water reactor Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-28 Wenhai Qu, Hao Xie, Hanyu Wang, Jinbiao Xiong
The turbulent flow in lower plenum of reactor pressure vessel (RPV) of pressurized water reactor (PWR) is an important phenomenon for design of flow mixing device. A high-precision matched index of refraction (MIR) technique is requisite for time-resolved particle image velocimetry (TR-PIV) measurements in RPV with complicated internals. In this study, turbulent flow in a scaled-down RPV model was
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Research on the flywheel clearance flow and heat transfer characteristics in a horizontal canned motor reactor coolant pump Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-27 Xiaohang Chen, Danrong Song, Rui Xu, Jinqi Lu, Zemin Gao, Dezhong Wang
Flywheel is an important component for the canned motor reactor coolant pump (RCP). The flywheel of horizontal canned motor reactor coolant pump (HRCP) is different from RCP’s flywheel. Its clearance width or different thermal conductivity structure would affect the flow structure of flywheel clearance flow, and then affect temperature distribution of the clearance, which would further affects the
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Sub-modeling numerical analysis on the radionuclide release from a transport cask submerged in the deep sea Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-26 Guhyeon Jeong, Sanghoon Lee
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Enhancement in the safety and reliability of Pressurized Water reactors using Machine Learning approach Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-26 Muhammad Zubair, Yumna Akram
The increasing interest in artificial intelligence and automation within the nuclear industry stems from the hope of elevating the safety and reliability of nuclear power plants by minimizing the impact of human errors. This paper explores the capabilities of machine-learning based fault detection and diagnosis (FDD) models in accurately classifying transient events in a nuclear power plant using the
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Numerical study of transient temperature thermal stress coupled heat transfer in spent fuel storage cask based on Gaussian random heat source Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-24 Jianjie Cheng, Weihao Ji, Hao Xu, Yawen Wang
Spent fuel storage is a necessary part of the entire lifecycle to achieve intrinsic safety in the application of nuclear energy as a clean energy source, and should be adequately investigated. Based on a Gaussian stochastic heat source, this paper presents a three-dimensional coupled numerical model of the heat-stress multiphysics field for the dry storage cask. Comparison with the literature model
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Bayesian update of fragility curves for equipment failure probability in seismic probabilistic safety assessment in nuclear power plant Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-24 Hong Jiang, Changhong Peng
The assessment of equipment failure probability during seismic probabilistic safety assessment (PSA) is a crucial and fundamental component in nuclear power plant seismic PSA. A Monte Carlo program must be formulated to compute the equipment failure probability and uncertainty analysis in seismic PSA. A numerical calculation method is employed for the integration calculation to obtain the conditional
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A memory-efficient neutron noise algorithm for reactor physics Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-23 Paul Cosgrove, Maximilian Kraus, Valeria Raffuzzi
The neutron noise equation contains a fixed source term where the neutron angular flux is present. Deterministic noise solvers have tended to solve for and store this angular flux term in a preprocessing step. This can be a substantial memory burden as the angular flux is a function of space, energy, and angle. This can limit these solvers to smaller problems and can limit the obtainable angular resolution
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Numerical simulation of typical abnormal operating conditions in the secondary circuit system of a Hua-long Pressurized Reactor nuclear power unit Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-22 Chen Yang, Qiang Zhang, Chuntian Lu, Shanglong Huang, Tao Zhang, Zonglong Zhang
Dynamic simulations of nuclear power plant behaviors are critical in plants’ lifecycle. However, accurate modeling of nuclear plants is quite a challenge due to the intrinsic complexity involved in the highly coupled processes. In this study, a dynamic model is established to investigate the dynamic characteristics of the secondary circuit system of HPR1000. Specifically, the major components are modeled
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An improved parallel Random Sequential Addition algorithm in RMC code for dispersion fuel analysis Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-21 Zhe Chuan Tan, Zhi Yuan Feng, Kan Wang
With growing interest in Accident Tolerant Fuels (ATF) such as Fully Ceramic Micro-Encapsulated (FCM) fuel, explicit modeling processes play an increasingly important role in the precise simulation of particle transport in stochastic media. Current explicit modeling methods in RMC utilize Random Sequential Addition (RSA), where particles are randomly and sequentially packed into fuel. However, at high
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Numerical study on flow instability of parallel helical tubes under convective heating condition Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-21 Maolong Liu, Cong Shen, Ziyi Xu, Chen Zeng, Xiaowen Wang, Limin Liu, Hanyang Gu
The present study numerically investigated the flow instability in parallel helical tubes under various conditions using a one-dimension code SGTH-1D. The effects of non-uniform heat flux, curvature ratio and inlet throttling configuration are investigated. The results show that the convective heating condition can significantly improve the system stability compared with uniform heating. The single-phase
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Development of friction factor and heat transfer correlation of liquid metal flow in helical tube bundles Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-20 Cong Shen, Maolong Liu, Limin Liu, Ziyi Xu, Chen Zeng, Li Liu, Hanyang Gu
There are currently few friction factor and heat transfer correlations for liquid metal flow in the shell-side of helical-coiled once-through steam generators (H-OTSG). The flow and heat transfer characteristics in H-OTSG are affected by various configuration parameters. Therefore, it is crucial to develop friction factor and Nusselt number correlations for the preliminary design of liquid metal H-OTSG
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Numerical simulation prediction of critical heat flux for 5 × 5 rod bundle with multiple grid spacers and different cladding materials Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-20 Kexin Liu, Kejia Li, Yulong Mao, Rui Zhang, Ming Ding, Xiaxin Cao
The critical heat flux (CHF) is one of the safety standards for core thermal design and is crucial for the normal operation of the reactor. The CHF problem of rod bundles with grid spacers is a hot topic in related research. The successful design of grid spacers can improve channel CHF by affecting the flow of downstream coolant, and this detection analysis can be achieved through numerical simulation
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An experimental study on the effect of coolant salinity on steam explosion Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-20 Yucheng Deng, Qiang Guo, Yan Xiang, Di Fang, Andrei Komlev, Sevostian Bechta, Weimin Ma
The steam explosion plays an essential role in the safety analysis of light water reactors (LWRs). Some studies have demonstrated that the occurrence of steam explosions is dependent on many factors such as melt and coolant temperatures, melt and coolant properties, non-condensable gases, etc. After the Fukushima accident, seawater as an emergency coolant and its impact on fuel coolant interactions
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Novel polynomial Abet data augmentation algorithm with GRU paradigm for nuclear power prediction Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-20 Saifullah Khan, Saeed Ehsan Awan, Yasir Muhammad, Ihtesham Jadoon, Muhammad Asif Zahoor Raja
In current study innovative approach is presented to improve the accuracy and efficiency of machine learning models on the dataset for each of the 31 US states, solely comprising a single power output column in Terra watt-hours (TWh) spanning from January 2013 to November 2020 by exploiting the knacks of Long Short-Term Memory (LSTM), Gated Recurrent Unit (GRU), hybrid GRU-LSTM, Polynomial Feature
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Steady and transient solutions of neutron diffusion equations via computational methods based on Hartley series and higher order finite difference schemes Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-20 Yasser Mohamed Hamada
In this work, new accurate solutions based on Hartley series expansion for temporal calculations and higher order finite difference schemes for space derivative calculations are presented for the neutron diffusion systems. Throughout a novel use of a linear combination of Hartley series, a new temporal approximate solution is obtained. Determination of the series coefficients is a basic requirement
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The impact of gamma-irradiation on physical properties and membrane distillation performance of PTFE and PVDF hollow fiber membrane Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-19 Xuxiang Jia, Lijun Lan, Tianping Wang, Yu Wang, Chunsong Ye
The impact of gamma irradiation on physical properties and membrane distillation (MD) performance of polytetrafluoroethylene (PTFE) and poly vinylidene fluoride (PVDF) hollow fiber membranes were investigated. As the gamma irradiation dose increased to 150 kGy using a Co source, the mean pore diameter of PTFE and PVDF decreased by 26.7 % and 13.9 %, membrane surface roughness decreased by 59.3 % and
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Numerical study of mixing vane spacer grid effect on subchannel mixing in a 5 × 5 rod bundle Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-17 Wenhai Qu, Yi Li
Turbulent flow in rod bundle with spacer grid without mixing vanes and with mixing vanes of inclination angles from 10° to 30° was investigated by Reynolds Stress Transportation (RST) model under Reynolds numbers (Re) from 3.96 × 10 to 3.96 × 10. The predicted results were validated by experimental data of lateral velocity field in 5 × 5 rod bundle. The cross-flow velocity was relatively well predicted
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Evaluation of probabilistic safety margin of nuclear power plant based on optimized adaptive sampling method Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-17 Haoyin Chen, He Wang, Longcong Wang, Qiang Zhao
The risk-informed safety margin characterization (RISMC) method utilizes a probabilistic safety margin (PSM) comparison between a load and capacity distribution, rather than a deterministic comparison between two values. The RISMC method can be effective in exploiting the potential safety margins and improving the economics of nuclear power plants (NPPs). However, the high computational cost of the
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Traditional machine learning methods in predicting the physics of subcritical systems in source-equilibrium Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-16 Ronald Daryll E. Gatchalian, Pavel V. Tsvetkov
Reactivity measurement methods, that are formulas relating integral physics quantity to system observable, were mostly derived from Point Reactor Kinetics (PRK). PRK presupposes fundamental mode that is driven solely by fissions; however, in Subcritical Assemblies (SCAs), an independent source is present to maintain flux. Deviation from PRK assumptions causes non-ideal response of standard techniques
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Heaving motion effect on the transient thermodynamic characteristics of helical coil once-through steam generators in mixed ocean conditions Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-16 Chao Zhang, Jiangping Li, Haoyu Sun, Weichao Li, Yandong Hou, Yan Xiang
Based on the finite difference method, the fixed boundary method, and a staggered grid method, this paper develops a one-dimensional thermal–hydraulic analysis code to simulate the effects of heave motion and more complex oceanic motions, namely heaving coupled with inclination and heaving coupled with swaying, on the thermal–hydraulic parameters of the International Reactor Innovative and Secure (IRIS)
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Method of characteristics with a new Macro-track transport acceleration for complex general geometry Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-16 Guo Jian, Zhu Guifeng, Yan Rui, Zou Yang
The method of characteristics is a flexible method in heterogeneous geometry and has been widespread use for reactor physics lattice calculation and whole core simulation. However, many acceleration methods commonly used in traditional lattice code and nodal method are not applicable to or inefficient to heterogeneous-geometry, high-dominance-ratio problems. This paper presents a new macro-track transport
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A study on the roll compaction of dry concentrated waste liquid for disposal Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-15 Sang-Hyun Lim, Sang-Heon Lee, Jong-Soon Song, Ki-Hong Kim
The roll compaction process can be used to reduce the physical volume of the powdered wastes. And polymer solidification technology can be used to manufacture polymer-solidified body incorporated with high-density pellets. Powdered wastes (soil, crushed concrete, dry concentrated waste liquid) were converted to rigid pellets of certain shape and size (H 6.5 W 9.4 mm) by a lab-scale roll compactor.
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Intelligent multi-severity nuclear accident identification under transferable operation conditions Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-15 Song Xu, Yuantao Yao, Nuo Yong, Dongqin Xia, Daochuan Ge, Jie Yu
Nuclear power plants (NPPs) have witnessed significant advancements in intelligent accident identification in recent years. However, comprehensive research on fine-grained analysis of accident severity levels has been lacking, thus limiting its practical application in real operating environments. This study proposes a novel intelligent nuclear accident identification method based on the designed dual-branch
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Equivalent mechanical model for liquid sloshing in irregular annular cylindrical containers of liquid metal reactors Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-15 Yuxuan Zhu, Daogang Lu, Yu Liu, Donghao Li
The dynamic response of the liquid sloshing in the reactor vessel is an important issue for seismic designs of liquid metal reactors (LMR). In engineering design, equivalent mechanical models are generally used to evaluate the response of liquid containers. However, several equipment support cylinders are arranged in the reactor vessel forming an annular cylindrical container with evenly distributed
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Next event estimation for neutron transport Monte Carlo simulations in moving objects Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-15 Huayang Zhang, Bin Zhong, Longfei Xu, Huayun Shen, Jinhong Li
The next event estimation is one of the important algorithms employed in Monte Carlo simulations for particle transport. It finds wide applications in point flux, image flux, and variance reduction techniques. With the advancement of transport simulations for moving objects, the next event estimation for moving target nuclei has yet to be realized. The next event estimation requires the calculation
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Study on flow regime prediction model for water-cooled reactor core based on machine learning algorithms Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-15 Yichao Ma, Dexiang Kong, Jing Zhang, Mingjun Wang, Wenxi Tian, Yingwei Wu, G.H. Su, Suizheng Qiu
The nuclear reactor core is the pivotal component in a nuclear power plant to generate and transfer heat, so accurate prediction of reactor core flow and heat transfer characteristics is a crucial problem for the reactor system design. Flow regime is closely related to the thermal–hydraulic characteristics of two-phase fluid. Still, flow regime models used in thermal–hydraulic calculation codes rely
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Study of the DVI-LOCA in the AP1000-like reactor with MELCOR code Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-14 Mateusz Włostowski, Piotr Darnowski
The paper presents the outcomes of the deterministic safety studies for a Pressurized Water Reactor based on the AP1000 design. The model was developed with MELCOR computer code using public references and qualified using data for nominal steady-state and double-ended Direct Vessel Injection line break accident. Emphasis was placed on thermal–hydraulic phenomena and modeling of the passive safety features
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Thermal consolidation modeling of unsaturated soil with semi-permeable boundary: Semi-analytical and numerical investigation Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-14 Hongping Meng, Aifang Qin, Lianghua Jiang, Haisheng Liu
This paper focuses on investigating the one-dimensional thermal consolidation of unsaturated soil. Through the utilization of the Laplace transformation, decoupling method, and inverse Laplace transformation, the semi-analytical solutions for excess pore pressures and ground settlement are deduced, particularly in the context of a single-sided semi-permeable boundary. To further investigate the thermal
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Research on protection control strategy of pressurized water reactor nuclear power units for typical power system faults Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-14 Haomin Wang, Muping Li, Peiwei Sun, Xinyu Wei
With the development of nuclear power industry in China, the proportion of nuclear power generation in the total power generation is constantly rising. As the power system is always faced with the threat of faults of different degrees, these faults will not only affect the power quality, but also are potential risks for nuclear power units with high safety requirements. Therefore, it is necessary to
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Autonomous control for Heat-Pipe microreactor using Data-Driven model predictive control Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-13 Linyu Lin, Joseph Oncken, Vivek Agarwal
To enable a self-regulating capability for heat pipe (HP) microreactors, an anticipatory control strategy achieved via model predictive control (MPC) could proactively respond to potential disturbances and deviations in operating setpoints. This paper demonstrates data-driven methods for predicting the distribution and transient of temperatures and heat fluxes at selected components and regions in
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Effects of boric acid on volatile tellurium in severe accident conditions Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-10 Fredrik Börjesson Sandén, Anna-Elina Pasi, Teemu Kärkelä, Tuula Kajolinna, Christian Ekberg
Boric acid is used in light-water nuclear reactors to control the reactor and is expected to be present as part of the chemistry of a severe accident. Therefore, its influence on other prominent species expected in an accident must be investigated. One such species is tellurium. In the present study, tellurium is volatized, and boric acid is dissolved and injected into the system as a means of studying
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Fuel performance analysis of Cr-coated Zircaloy-4 cladding during a prototypical LOCA event using BISON Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-10 Cole Dunbar, WooHyun Jung, Robert Armstrong, Kumar Sridharan, Michael Corradini, Hwasung Yeom
Deformation and failure of chromium (Cr) coated Zircaloy-4 (Zry-4) were studied in loss-of-coolant accident (LOCA) conditions using the BISON fuel performance code. The BISON validation model simulating Halden research reactor experiments was extended to include Cr coatings and higher rod internal pressures to simulate high-burnup fuel. The transient simulations show Cr coatings help relieve stress
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Prediction of critical heat flux in rod bowing geometries by coupled subchannel analysis and mechanistic model Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-09 Junliang Guo, Jianqiang Shan, Li Jiang, Yujiao Peng, Shan Zhou, Miao Gui
The penalty of fuel rod bowing on the departure from the nucleate boiling ratio limits must be considered in the thermal–hydraulic design. Different from the traditional penalty factor method, in this study, a relatively strong versatile method is proposed to predict the critical heat flux in rod bowing geometries by coupled subchannel analysis and mechanistic model. A subdivisional subchannel method
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Multilayer radiation shields for nuclear and radiological centers using free-lead materials and nanoparticles Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-09 Aljawhara H. Almuqrin, M.I. Sayyed, Dalal A. Alorain, Mohamed. Elsafi
We prepared two types of lead-free shielding materials consist of double and triple layered with different thicknesses and reported the radiation shielding performance of the developed concrete samples. The linear attenuation coefficient (LAC) of microcomposites was calculated theoretically by XCOM-software and compared with experimental results and less than 4 % difference was observed at all discussed
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Proposal of weighting options for influencing factors and method of giving preference to alternatives for selecting the preliminary decommissioning strategy of Kori-1 Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-09 Hye-Jin Jung, Hyung-Woo Seo, Jin-Won Son, Cheon-Woo Kim
For the safe decommissioning of nuclear power plants, establishing a decommissioning strategy is one of the important issues covering the entire decommissioning process. The selection of a decommissioning strategy can affect the overall safety and efficiency of decommissioning projects. This selection process can be complex, and a number of factors that are not generally considered during normal operation
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Machine learning at the edge to improve in-field safeguards inspections Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-06 Nathan Shoman, Kyle Williams, Burzin Balsara, Adithya Ramakrishnan, Zahi Kakish, Jamie Coram, Philip Honnold, Tania Rivas, Heidi Smartt
Artificial intelligence (AI) and machine learning (ML) are near-ubiquitous in day-to-day life; from cars with automated driver-assistance, recommender systems, generative content platforms, and large language chatbots. Implementing AI as a tool for international safeguards could significantly decrease the burden on safeguards inspectors and nuclear facility operators. The use of AI would allow inspectors
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A control rod worth prediction using Adaptive Neuro-Fuzzy Inference System for Pre-Calibration Method at TRIGA PUSPATI Reactor Ann. Nucl. Energy (IF 1.9) Pub Date : 2024-02-06 Teh Zhi Hui, Nur Syazwani Mohd Ali, Mohd Sabri Minhat, Jasman Zainal, Muhammad Arif Sazali, Muhammad Syahir Sarkawi, Khairulnadzmi Jamaluddin, Nor Afifah Basri, Mohsin Mohd Sies, Nahrul Khair Alang Md Rashid
One of the control rod calibration methods in research reactors is the doubling time. However, this method reduces the operation time and limits the number of research activities. A pre-calibration method is proposed in this study by utilizing an ANFIS method. Two data inputs based on the annual rod worth, and the worth drop of the Shim and Transient rods were collected to predict the Safety and Regulating