样式: 排序: IF: - GO 导出 标记为已读
-
Multifractal detrended fluctuation analysis of boiling water reactors Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-11 Alberto Quezada Tellez, Francisco A. Godínez, Guillermo Fernández-Anaya, Marco A. Polo-Labarrios, Sergio Quezada García
The aim of this work is to determine a relationship between multifractal parameters and Boiling Water Reactor (BWR) stability; therefore, a Multifractal Detrended Fluctuation Analysis (MF-DFA) of the data series obtained from the Average Power Range Monitor (APRM) and the Local Power Range Monitors (LPRM) during stable operation and during an instability event of a BWR-5, which concludes with the SCRAM
-
Structural optimization of gas-solid separator used in TMSR-SF based on computational fluid dynamics Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-11 Guo-Yan Zhou, Mengdan Wu, Chunfeng Bao, Ning Zhang, Shan-Tung Tu
The fission gas removal system plays a critical role in the Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF). A gas–solid separator is used to separate the waste gas and solid particles in the gas removal system. As a key component of the system, the gas–solid separator directly determines the system energy efficiency. Nowadays, a new combined gas–solid separation equipment combining axial flow blade
-
Progress of sodium-cooled fast reactor developments in Japan taking into account total lifecycle, risk-informed approach, and sustainability Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-11 Hideki Kamide, Tai Asayama, Takashi Wakai, Toshiki Ezure, Akihiro Uchibori, Shigenobu Kubo, Masayuki Takeuchi
A sodium cooled fast reactor (SFR) is one of the most relevant and decarbonized energy supply system with higher sustainability on natural resources, footprint, and waste management. It was planned in a “strategic roadmap” of fast reactor decided by Inter-Ministerial Council for Nuclear Power Japan in 2022 to start a conceptual design of a demonstration reactor from 2024 with a background of accumulated
-
Optimization of airfoil fin PCHE for the power conversion system of lead-based reactor based on reinforcement learning Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-11 Haoqi Wang, Chong Gao, Zhiyi Peng, Hao Wu, Houjian Zhao, Zhangpeng Guo, Ke Zhang, Yang Liu
As the Gen IV nuclear reactors, the lead-based reactor employs the S-CO Brayton cycle. As the intermediate heat exchangers, high or low temperature regenerators and precoolers, the thermal hydraulic characteristics of the PCHEs directly affect the efficiency of S-CO power cycle efficiency. In this work, two reinforcement learning algorithms, namely the Q-learning and DQN are used to optimize the key
-
Prediction of stress corrosion crack growth rate of cold worked stainless steels in BWR conditions by the slip oxidation model Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-11 Masato Koshiishi, Dan Akazawa, Yasufumi Miura, Kenji Kako
Stress corrosion cracking (SCC) is a critical issue for boiling water reactor (BWR) components, and their cold working is considered to be an accelerating factor for SCC crack growth rates (CGRs). However, no method has been proposed to quantitatively evaluate the SCC CGRs of cold worked stainless steels. In this study, the changes in material properties due to cold working were reflected in the input
-
Foreword to “Outcomes and achievements from researches orienting the future in nuclear fission technology: China (+Taiwan)–Gen III” Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-10 Changjun Hu, He Bai, Xuesong Wang, Yun Hu
-
The SARENA Programme: Master's degree studies in nuclear engineering combining different educational and cultural environments Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-10 Ivo Kljenak, Abdesselam Abdelouas, Manon Desmytter, Francisco J. Elorza, Eduardo Gallego, Andreas Hartnack, Christoph Hartnack, Juhani Hyvärinen, Juhani Vihavainen, Iztok Tiselj
Within the SARENA programme, students from different parts of the world have the opportunity to obtain a Master's degree in nuclear engineering by participating in a two-year curriculum combining academic programs in France, Spain, Finland and Slovenia (depending on the selected study track). The concept and organization of the SARENA programme are described. The experience from the first four cohorts
-
Validation of a MELCOR model of HTR-10 in normal operation and accident conditions Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-10 Edward M. Duchnowski, Nicholas R. Brown
We apply the thermal-hydraulics accident progression tool, MELCOR, to thermal–hydraulic benchmark problems on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) listed in the International Atomic Energy Agency (IAEA) Coordinated Research Project on Evaluation of High Temperature Gas-Cooled Reactor (HTGR) Performance (CRP-5). MELCOR results are compared to other solutions reported in
-
Experimental investigation on the effect of direct contact condensation regime on thermal stratification Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-09 Zeji Wang, Haiqiang Liu, Wentao Guo, Zhangpeng Guo, Shengfei Wang, Xiaoping Ouyang
Direct contact condensation affects the performance of a small modular reactor with pressure suppression system. Since the heat and mass transfer varies with the condensation regimes as well as steam-water interface shapes, the identification and classification of the condensation regimes are important. The experiment of steam direct contact condensation was carried out and the corresponding condensation
-
Numerical simulation of the ocean conditions impact on heat pipe-cooled molten salt reactor core thermal-hydraulic performance Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-08 Ziyin Liu, Limin Liu, Ziang Guo, Hui Guo, Tenglong Cong, Maolong Liu, Hanyang Gu
The Heat-Pipe-cooled Molten Salt Reactor (HP-MSR) has been seen as a promising potential candidate for powering ocean vehicles with long endurance while reducing the carbon emissions in the shipping industry. By combining the operation merits of HPs and MSRs, this type of reactor can characterize no need for fuel fabrication, inherent safety, short waste lifespan, and elimination of in-core thermal
-
Thermodynamic equilibrium state calculations for oxidation and corrosion reactions of B4C and oxide-based neutron absorber compounds in reactor control rods Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-08 Dong-Joo Kim, Jae Ho Yang, Dong Seok Kim, Ji-Hae Yoon, Kwang-Young Lim, Jae-Yong Kim
The challenges associated with conventional boron carbide (BC) control rod materials, including helium gas accumulation and susceptibility to oxidation and corrosion in various environments, have been thoroughly explored. To address these issues, a comprehensive investigation into the potential of oxide-based neutron absorber compounds for control rods has been undertaken. Thermodynamic equilibrium
-
Outcomes and achievements from researches orienting the future in nuclear fission technology: NFT-18: Brazil, Peru and Bolivia: Nuclear Fission Technology in Brazil, Peru and Bolivia Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-08 Antonella Lombardi Costa, Daniel Artur Pinheiro Palma, Frederico Genezine, Giovanni Laranjo de Stefani
-
A hybrid model based on STL with simple exponential smoothing and ARMA for wind forecast in a Brazilian nuclear power plant site Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-08 José L.R.N. Cunha, Cláudio M.N.A. Pereira
In the event of a nuclear accident with release of radioactive effluents into the environment, the Environmental Control System (ECS) at the Central Nuclear Almirante Álvaro Alberto (CNAAA) nuclear complex estimates the movement of the radioactive plume produced and provides forecast for one and two hours ahead. The prediction is based on the current accident and atmospheric data collected from meteorological
-
Assessment of the IEA-R1 nuclear reactor using a nonstandard fuel assembly with six fuel plates of the Brazilian Multipurpose Reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-07 Humberto Vitor Soares, Walmir Maximo Torres, Pedro Ernesto Umbehaun, Antonio Belchior, Delvonei Alves de Andrade
In order to qualify the fuel plates of the Brazilian Multipurpose Reactor (RMB), a nonstandard Instrumented Fuel Assembly (IFA) was designed and is being constructed to be burned in the IEA-R1 nuclear research reactor. IFA has fuel plates of different uranium densities (10 fixed fuel plates of 3.0 gU/cm – IEA-R1 standard; 6 removable fuel plates of 3.7 gU/cm – RMB; and a central aluminum plate). This
-
A new procedure for solving multigroup neutron diffusion eigenvalue problems Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-07 YanTing Cheng, Mei Huang, XiaoPing Ouyang, YanPing Huang, DengGao Chen, Hiroshi Matsuda
The spatial distribution of neutron flux within the core of a nuclear reactor plays a crucial role in ensuring nuclear safety. The eigenvalue problem of neutron diffusion equations can be solved to determine this distribution. Typically, the fluxes are solved for one energy group at a time using the power iteration method, which helps reduce computational storage requirements. However, when dealing
-
Analysis of thorium-transuranic fuel deployment in a LW-SMR: A solution toward sustainable fuel supply for the future plants Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-06 R. Akbari, M. Aghili Nasr, F. D'Auria, A. Cammi, J.R. Maiorino, G.L. de Stefani
One of the main challenges faced by nuclear power plants is the scarcity of U-235, the fissile isotope used in conventional nuclear reactors. While there is enough U-235 in the world to fuel current reactors, the demand for energy is increasing rapidly, especially in emerging economies like China and India, and this could lead to shortages in the future. One potential solution to these challenges is
-
Innovative burnable absorbers: Assessing PaO2 and NpO2 coatings for improved safety in (Th-233U-235U)O2 fuel assemblies Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-06 Ouadie Kabach, El Mahjoub Chakir, Hamid Amsil
Various advanced fuels and burnable absorbers (BAs) have been proposed to improve the cycle burnup of a pressurized water reactor (PWR) fuel assembly (FA). This study looks into the impact of an innovative fuel known as (Th-U-U)O with a novel neutron absorber with double coatings of PaO and NpO. The study focuses on adding the proposed two-layer absorbers in some (Th-U-U)O fuel in an AP1000 assembly
-
An open-source porous media modelling approach to investigate thermohydraulic features of compact printed circuit heat exchangers Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-06 M. McDermott, S. He
An experimental air-foil printed circuit heat exchanger (PCHE) with as a working fluid is numerically modelled and developed within the OpenFOAM environment as a freely available package for more general PCHE designs. The conjugate heat transfer solver (chtMultiRegionFoam) is adapted to include both the hot and cold fluid streams of the PCHE, along with the solid recuperator body, within three uniquely
-
Core reduction for increasing neutron flux and radioisotope production in the IEA-R1 research reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-06 Felipe Viggiano, Giovanni Laranjo de Stefani, Frederico A. Genezini, João M.L. Moreira, Caio J.C.M. Cunha
This work aims to present and analyze a new reactor core configuration for the IEA-R1 reactor core with the objective of enhancing its neutron irradiation capabilities. The reactor was modeled using the KENO-VI module from the SCALE Code System, provided by ORNL. Results have shown that implementing a near-cylindrical configuration in the IEA-R1 core could substantially increase neutron flux, particularlly
-
Development of the nuclear competences based on global trends in the nuclear industry Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-06 Kateryna Piliuhina, Sergiy Bushuyev, Roberta Cirillo, Gabriel Lazaro Pavel, Marco Ricotti, Willem Janssens
The nuclear industry worldwide is experiencing significant transformations driven by various trends. They include rising energy demands, technological advancements, climate crisis challenges, international cooperation agreements, growing public concerns and evolving threats. These trends require robust competences development in the nuclear area to ensure the safe and secure use of nuclear energy.
-
Enhancing higher education through hybrid and flipped learning: Experiences from the GRE@T-PIONEeR project Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-06 C. Demazière, C. Stöhr, Y. Zhang, O. Cabellos, S. Dulla, N. Garcia-Herranz, R. Miró, R. Macian, M. Szieberth, C. Lange, M. Hursin, S. Strola
GRE@T-PIONEeR is a Horizon 2020 project coordinated by Chalmers University of Technology, running over the period 2020–2024. 18 university teachers from 8 different universities located in 6 different countries gathered forces to develop and offer advanced courses in computational and experimental nuclear reactor physics and safety. All courses are flipped hybrid courses, i.e., students work on online
-
Review of the research on the scrubbing of fission products in liquid metal pool Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-05 Gen Jiang, Mou Wang, Kai Wang, Songbai Cheng
In recent times, the risk assessment and design optimization of advanced fourth-generation nuclear reactors have garnered significant global scholarly attention. Specifically, researchers worldwide have been focusing on the comprehensive analysis of the scrubbing process of radioactive fission product aerosols within the reactor containment pool under severe accident scenarios. This study delves into
-
Online parameter adaptation of Pressurized Water reactor during Load-Following operation with bounded axial power distribution via Lyapunov approach Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-05 M. Zaidabadi nejad, G.R. Ansarifar, H. Zayermohammadi Rishehri
Although nuclear accidents have occurred in the past, advancements in technology and safety measures have made nuclear energy a viable and competitive option for energy generation. However, within nuclear reactors, the issue of spatial oscillations in neutron flux distribution caused by reactivity feedback of xenon must be addressed. Uncontrolled oscillations in power distribution may lead to exceeding
-
Comparative study of CUDA-based parallel programming in C and Python for GPU acceleration of the 4th order Runge-Kutta method Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-05 Davi F. Fernandes, Marcelo C. Santos, Adilson C. Silva, Alan M.M. Lima
In this paper, a comparative study is presented on the application of General-purpose Computing on Graphics Processing Units for solving the point reactor kinetics equations through the utilization of the 4th Order Runge-Kutta (RK4) method using the programming languages C and Python. Sequential and parallel algorithms of the RK4 method were developed in C/C++ and Python, with parallel algorithms specifically
-
Modeling of the influence of local heat sources on a light gas stratification formation and erosion in a large-scale experimental facility using eddy resolving numerical approach Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-04 A.A. Kanaev
The main objective of this work is to evaluate the results of application of Implicit Large Eddy Simulation (ILES) turbulence modeling approach to simulate the complex hydrogen safety problem. The paper presents the results of numerical simulation of the OECD/NEA HYMERES (HYdrogen Mitigation Experiments for Reactor Safety) HP2 series tests in the PANDA facility, aimed at studying the thermal effect
-
Flame wrinkling factor in quiescent hydrogen-air mixtures Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-04 Veikko Taivassalo
When modelling the combustion of hydrogen in such a large domain as a nuclear reactor containment building, flame wrinkling needs be quantified in the changing conditions of the combustion process. In addition to turbulence, intrinsic instabilities and buoyancy is known to contribute to the wrinkling of the flame surface and thus increase the total area of the flame front leading to flame acceleration
-
Nuclear material container drop testing using finite element analysis with verification using digital Image correlation Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-02 Jude M. Oka, Raj U. Vaidya, John T. Davis, Kirk W. Webber, Renita K. Walzel
Los Alamos National Laboratory (LANL) uses a vast array of containment vessels to store, handle and transport special nuclear material. One of the most frequent uses of containers is the specific transferring and storing of nuclear material within a glovebox or glovebox line. A glovebox is an engineered barrier that protects the worker and public from special nuclear material processing. Containers
-
Coupled neutronic-thermal–mechanical analysis of a medium temperature heat pipe cooled reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-02 S.N. Li, J.C. Huang, B.H. Yan, J.Y. Zhao
Medium temperature heat pipe cooled reactor (MTR) can better address problems faced in tradition heat pipe cooled reactor such as challenges in high-temperature corrosion-resistant material, possible startup failure accidents and high steel monolithic thermal stress. The unique advantages of MTR include excellent startup performance, high transient negative temperature coefficient, low monolithic thermal
-
A systematic approach for the adequacy analysis of a set of experimental databases: Application in the framework of the ATRIUM activity Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-02 J. Baccou, T. Glantz, A. Ghione, L. Sargentini, P. Fillion, G. Damblin, R. Sueur, B. Iooss, J. Fang, J. Liu, C. Yang, Y. Zheng, A. Ui, M. Saito, R. Mendizábal Sanz, A. Bersano, F. Mascari, T. Skorek, L. Tiborcz, Y. Hirose, T. Takeda, H. Nakamura, C. Choi, J. Heo, A. Petruzzi, K. Zeng, Z. Xie, X. Wu, H. Eguchi, F. Pangukir, P. Breijder, S. Franssen, G. Perret, I.D. Clifford, T.M. Coscia, F. Di Maio
In the Best-Estimate Plus Uncertainty (BEPU) framework, the use of best-estimate code requires to go through a Verification, Validation and Uncertainty Quantification process (VVUQ). The relevance of the experimental data in relation to the physical phenomena of interest in the VVUQ process is crucial. Adequacy analysis of selected experimental databases addresses this problem. The outcomes of the
-
A neutronic evaluation of a thorium-based molten salt breeder reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-01 Clarysson A.M. Silva, Isabella R. Magalhães, María Lorduy-Alós, Sergio Gallardo, Claubia Pereira, Antonella L. Costa, Gumersindo Verdú
The global energy demand is continuously rising, requiring the search for attractive and low-emission energy generation options. Molten Salt Reactors (MSRs) have emerged as a promising solution in the Generation IV roadmap due to their inherent safety features and fuel flexibility. Operating at high temperatures, MSRs efficiently convert heat into electricity and offer potential applications beyond
-
Experimental investigation on steam sparger and its effect on steam contact condensation in makeup tank Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-29 D.C. Sun, Y. Li, Z. Xi, Y.F. Zan, X. Yan
The Hualong pressurized water reactor (HPR1000) utilizes a secondary passive residual heat removal system (PRS) to maintain the core in a safe state within 72 h after a plant blackout incident (SBO). Two emergency make-up tanks (EMT) are incorporated into the PRS to provide safe injection to the secondary side of the steam generator (SG). However, during the early stages of PRS operation, direct contact
-
A closer-look on W and Pb alloys: In-depth evaluation in elastic modulus, gamma-ray, and neutron attenuation for critical applications Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-29 Ghada ALMisned, Gulfem Susoy, Duygu Sen Baykal, Hessa Alkarrani, Ömer Güler, H.O. Tekin
This investigation assesses the gamma-ray and neutron attenuation properties of various alloys, including Pb90Cu10, A5, Manganin-R, Cu0.2Ag0.8, SA4, and W-based, to uncover efficient and cost-effective radiation shielding materials. Our study centers on alloys featuring elements such as lead, molybdenum, silver, and tungsten, selected for their unique protective qualities against radiation. Employing
-
Performance analysis and helium behaviour of Am-bearing fuel pins for irradiation in the MYRRHA reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-29 L. Luzzi, A. Magni, S. Billiet, M. Di Gennaro, G. Leinders, L.G. Mariano, D. Pizzocri, M. Zanetti, G. Zullo
Minor actinides are the main contributors to medium- and long-term radiotoxicity and heat production in spent nuclear fuels. Research efforts are currently ongoing to explore different options to dispose of such radionuclides, e.g., their burning in fast reactors within mixed-oxide fuels. The MYRRHA sub-critical reactor is one of the future facilities with envisaged burning and transmutation capabilities
-
Single heated channel analysis of the AP-Th 1000 concept Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-29 Caio Július César Miranda Rodrigues da Cunha, Daniel González Rodríguez, Giovanni Laranjo de Stefani, Fernando Roberto de Andrade Lima, Carlos Alberto Brayner de Oliveira Lira
This paper proposes a three-dimensional and single-phase model using computational fluid dynamics to study the thermal–hydraulic design limits of the AP-Th 1000 concept. This thermal–hydraulic study provides the engineering design limits of the hottest subchannel of the AP-Th 1000 concept during the first fuel cycle. The reactor successfully upheld a maximum temperature that consistently remained near
-
Simulation and analysis of core barrel vibrational modes associated with neutron noise phenomena in WWER-type reactors Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-28 A. Kamkar, O. Safarzadeh, M. Abbasi
This survey focuses on an approach to address various vibrational modes of the core barrel in a hexagonal-structured reactor core. The core barrel vibration is considered one of the main mechanical vibrations in a high-power nuclear reactor, initiating core component malfunctions and safety issues. Here, a well-reasoned method is proposed to handle the triple vibrational modes of such core barrels
-
Self-improving few-shot fault diagnosis for nuclear power plant based on man-machine collaboration Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-28 Guolong Li, Yanjun Li, Site Li, Shengdi Sun, Haotong Wang, Jiarui Zhao, Baozhi Sun, Jianxin Shi
Traditional fault diagnosis methods for nuclear power plants rarely consider the update and improvement of the fault diagnosis model during the system operation. Therefore, we propose a self-improving few-shot fault diagnosis method (SNN-MSU) based on man–machine collaboration. The proposed method uses confidence threshold discriminator to judge the reliability of few-shot fault diagnosis results,
-
Operating characteristics analysis and optimization of loop heat pipe radiation cooling system in space reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-27 Genglei Xia, Tao Zhou, Yuandong Zhang, Shoubao Dai, Guanghui Jiao
Loop heat pipe radiation cooling system has good application prospects in space reactor heat dissipation due to its high heat transfer capacity, long-distance heat transfer without pump dependency, and effective isolation of mechanical vibrations. The code for analyzing the steady-state operational characteristics of loop heat pipe radiation cooling system was developed in this paper by the heat balance
-
Recent developments in coupled experiments and simulation to understand the fluid–structure dynamics of a Pressurized Water Reactor fuel assembly Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-27 Vincent Faucher, Hervé Palancher, Guillaume Ricciardi, Maria-Adela Puscas, Emmanuel Lo Pinto, Lionel Rossi
The proposed paper is dedicated to the description of an extensive research program devoted to understanding fluid–structure interaction mechanisms at all scales of a fuel assembly for a Pressurized Water Reactor (PWR). This work carried out over the last decade, closely combines modeling and experiments at the highest level, leading to a production of knowledge in the fields of instrumentation, fluid
-
Optimal path planning in a real-world radioactive environment: A comparative study of A-star and Dijkstra algorithms Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-26 Miyombo Ernest Miyombo, Yong-kuo Liu, Chishinga Milton Mulenga, Anthony Siamulonga, Martin Chihango Kabanda, Phillimon Shaba, Chunli Xi, Abiodun Ayodeji
Navigating complex radioactive environments while minimizing radiation exposure to workers is a critical challenge faced by the nuclear industry. Although various shortest-path algorithms and radiation dose calculation techniques have been employed for optimal path finding, most existing models are based on simulations that do not accurately represent real-world environments. To address this limitation
-
Research on the removal of charged colloidal filter materials from coolants in nuclear power plants Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-26 Tao Liu, Haixia Kong, Yongguo Li, Jia Wang, Haifeng Yu, Yuan Li
In this paper, nano-grade aluminum powder was used as raw material to prepare glass fiber filter material with positive Zeta potential by hydrothermal method. Since colloids of activated corrosion products in primary circuit of nuclear power plant are mainly electronegative, the filter material prepared in this experiment was used to remove colloidal particles in coolant by combining traditional mechanical
-
Active fault tolerant control of a heat pipe-cooled reactor based on state feedback method Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-26 Jiajun Huang, Songmao Pu, Peiwei Sun, Xinyu Wei
Heat pipe-cooled reactor (HPR) holds great potential for applications in deep-space exploration, deep-sea diving, and land-based power supply. However, many of these scenarios involve unmanned operation, demanding the HPR control system to exhibit exceptional reliability and fault tolerant capability in the face of faults. To address this need, an active fault-tolerant control (FTC) system specifically
-
Design optimization of noise reduction for labyrinth control valve in secondary circuit flow regulation system Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-26 Runlin Gan, Baoren Li, Jingrui Chu, Chang Yuan, Zhixin Zhao, Gang Yang
The labyrinth control valve has been widely used in the secondary circuit flow regulation system of nuclear power field for good pressure reduction and cavitation suppression characteristics. The design of the labyrinth channel is a key to realizing pressure regulation and cavitation suppression for the labyrinth control valve. However, an unreasonable labyrinth flow channel design will be challenging
-
Theoretical modeling of the boiling annular-mist flow in vertical channels Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-26 Wen He, Chenru Zhao, Jinyu Han, Hanliang Bo
Accurate prediction of the boiling annular-mist flow as well as the occurrence of the dryout is of particular importance for the safe operation of heat systems. Therefore, this paper focuses on the analysis of the boiling annular-mist flow and aims to propose a mechanistic model to describe this boiling process. Firstly, the physical model is established which divides the flow field into the bubble
-
Numerical study on the characteristics of sodium heat pipes for space application at operation limits Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-24 Jian-Shu Liu, Yu-Peng Mou, Xiao-Bin Li, Feng-Chen Li, Hong-Na Zhang, Ze-Ming Wang, Ye Han, Bao-Hua Chai
Sodium heat pipes are important thermal conductive components in space nuclear power sources. They efficiently transfer heat within the range of 500–1000 °C by utilizing the latent heat of phase change. However, their maximum heat transfer capacity is limited by various factors. In this study, numerical simulations were conducted to investigate the operating conditions of sodium heat pipes when encountering
-
Concrete ablation depth analysis of OPR1000 NPP during top flooding ex-vessel cooling Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-24 Hanbee Na, Namchul Cho
This study assesses the integrity of OPR1000 nuclear power plant (NPP) containment during top flooding ex-vessel cooling in the event of a postulated severe accident. Maintaining containment integrity during ex-vessel cooling is critical for preventing substantial release of radioactive materials to the environment. To analyze the feasibility of top flooding ex-vessel cooling of the OPR1000 NPP, the
-
Techno-economic feasibility study of coupling low-temperature evaporation desalination plant with advanced pressurized water reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-23 Ammar Alkhalidi, Belal Almomani, A.G. Olabi, Hussam Jouhara
The increasing demand for freshwater necessitates sustainable desalination solutions, and nuclear power plants offer a promising avenue by utilizing their low-grade waste heat. This study assesses a techno-economic feasibility of coupling a 5 MW low-temperature evaporation plant with a UAE-based Advanced Pressurized Water Reactor (APR1400). The system addresses freshwater demand, aligning with sustainability
-
Preface for special issue NFT-17: MX & CU & VE nuclear fission technology in Mexico, Cuba and Venezuela (+Central America) Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-23 G. Espinosa-Paredes, J.-L. François, G. Alonso, A. Castillo
-
Effect of sub-cooling on fuel channel behaviour during LOCA for Indian PHWR Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-23 Madhuri Bhadauria, Ravi Kumar, Arup K. Das, Deb Mukhopadhyay, Prasanna Majumdar
During LOCA with ECCS failure, the removal of decay heat from the channel to a moderator plays a significant role in determining the fuel bundle temperatures. An experimental investigation has been carried out to study the effect of change in moderator temperature (65 ℃, 75 ℃, and 85 ℃) on the fuel channel at a steady state of 900 ℃ center pin temperature. The results show that at different moderator
-
Stochastic modeling of a neutron imaging center at the Brazilian Multipurpose Reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-23 L.P. de Oliveira, A.P.S. Souza, F.A. Genezini, A. dos Santos
Neutron imaging is a non-destructive technique for analyzing a wide class of samples, such as archaeological or industrial material structures. In recent decades, technological advances have had a great impact on the neutron imaging technique, which has meant an evolution from simple radiographs using films (2D) to modern tomography systems with digital processing (3D). The 5 MW research nuclear reactor
-
Deep learning for predicting the residual concentration of sodium hypochlorite in the cooling water OF nuclear power plants Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-23 L.R. Gonçalves, C.H.S. Grecco, C.M.N.A. Pereira
Sea water is used as cooling water in several operating nuclear power plants, as well as in other industries such as petrochemicals. Biofouling is a common problem in systems that use sea water, causing corrosion in pipes and equipment, blockages, and loss of efficiency in heat exchangers. The use of sodium hypochlorite has proven to be effective in minimizing the damages caused by biofouling, provided
-
Preliminary study of a small high-temperature gas-cooled reactor (HTGR) concept with MgO–BeO moderators Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-23 Irwan L. Simanullang, Nozomu Fujimoto
Graphite plays a crucial role as a moderator and reflector in high-temperature gas-cooled reactors (HTGRs). However, under high-irradiation conditions, graphite exhibits microcracking within the operational period. Studies have been investigated the potential of composite-based materials for replacing graphite in HTGRs. This study focused on a magnesium oxide (MgO)-based composite material as the host
-
Analysis of a test facility for transients in a supercritical water-cooled reactor using dynamic system scaling methodology Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-23 Thiago D. Roberto, Mário A.B. da Silva, Celso M.F. Lapa
Depressurization in Supercritical Water-Cooled Reactors (SCWRs) presents challenges due to their unique operational conditions. Scaled-down facilities are preferred for analyzing depressurization phenomena due to the high costs associated with test facilities operating at 25 MPa. This study proposes the use of Dynamical System Scaling (DSS) methodology to analyze supercritical fluid depressurization
-
Utilising 241Am as burnable poison in proliferation resistant PWR Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-22 Mustafa J. Bolukbasi, Marat Margulis
The increased need for energy, as well as the necessity for energy-intensive solutions to tackle climate change, has increased interest in nuclear power generating as a low-carbon energy source. While nuclear energy offers substantial benefits in reducing greenhouse gas emissions, it also raises concerns regarding nuclear proliferation. In this study, the viability of utilising nuclear proliferation-resistant
-
Multi-objective loading pattern optimization of a soluble boron free core using social spider algorithm Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-22 M. Hosseinllu, M. Abbasi, O. Safarzadeh, F. Dehghani
Shortly due to their favorable economic and safety characteristics, Small Modular Reactors (SMR) systems are expected to be in high demand. To ensure effective operation and the proper safety margins of SMRs, it is crucial to find the optimum Loading Pattern (LP). In this paper, we proposed a 300 MW soluble boron free UO fueled core with 20 MWd/KgU discharge burnup. The Fuel Assemblies (FAs) and Burnable
-
Helium-air mixing in simulated reactor cavities of high temperature gas reactors Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-22 Abdullah Abubakar, Mathieu Davis, Zayed Ahmed, Dinesh Kalaga, Masahiro Kawaji
The spatial and temporal variations in air and helium concentration and temperature fields in simulated reactor cavities of High Temperature Gas Reactor (HTGR) were investigated experimentally. This accident scenario follows a hypothetical small pipe break in the Reactor Pressure Vessel and discharge of high temperature helium into the surrounding cavity. A scaled multi-compartment experimental facility
-
High-fidelity PIV measurements of turbulent flow in reactor pressure vessel assisted by high-precision matched index of refraction technique Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-22 Wenhai Qu, Jinbiao Xiong
Understanding and control of turbulent flow is desired for the sake of mixing enhancement or flow distribution in vessels of energy system, such as reactor pressure vessel (RPV) in pressurized water reactor (PWR). However, optical measurement of turbulent flow, e.g., particle image velocimetry (PIV), suffers from refraction at solid–liquid interfaces, especially when the vessels contain complicated
-
Leakage and loss of contact/adjustment effects on slip-joins of a jet-pump assembly of BWR nuclear reactors by dynamic fluid-structure interaction modelling Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-21 Enrique Flores-Cuamatzi, Rubén Cuamatzi-Meléndez, Etel Maya-Flores, Fernando Juárez-López
This work presents a methodology for the evaluation of self-excited vibrations of jet-pump assemblies that make up the water recirculation system of boiling water nuclear reactors. Therefore, for the analysis three different type of models were developed: finite element, computational fluid dynamics and fluid-structure interaction models of a jet-pump assembly. The models were subjected to pressurized
-
The ENEN’s role in shaping the European nuclear education Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-21 Gabriel Lazaro Pavel, Csilla Pesznyak, Francisco Javier Elorza Tenreiro, Joerg Starflinger, Leon Cizelj, Walter Ambrosini
The European Nuclear Education Network (ENEN) just celebrated in 2023 the first twenty years of existence. During this period, the ENEN network grew, reaching today more than ninety Members, Partners and Supporters. The mission of ENEN is the preservation and the further development of expertise in the nuclear fields by higher Education and Training. We target to reach this objective through the co-operation
-
Sustainable online initiatives for the dissemination of nuclear energy culture Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-21 Walter Ambrosini, Roberta Cirillo
The paper reports about three initiatives that partly preceded and then followed the groundbreaking experience of the Covid-19 pandemic, by offering a university level course available online, a series of webinars and an online career event, all including educational content that pertains to the category of nuclear energy culture. These initiatives had different characteristics and varied content,
-
Vocational training related ENEN activities and their impact on nuclear competence development in Europe Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-02-20 Christian Schönfelder
This paper examines the impact of vocational training on the development of nuclear competences and thus on the development of the workforce in the European nuclear industry, which is driven by a surge in demand for nuclear competences in the 2020s. Various actions initiated by the European Commission in this context are described, like Cedefop, EUROPASS, EQAVET, ECVET, plus recently launched initiatives