样式: 排序: IF: - GO 导出 标记为已读
-
Experimental investigation of countercurrent flow limitation (CCFL) in a 1/19 scale of VVER-1000 geometry Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-15 Aref Rahimian, Ebrahim Abedi, Seyed Khalil Mousavian, Masood Amin Mozafari, Seyed Mohammad Mirvakili
Countercurrent flow limitation (CCFL) has been experimentally considered in a VVER-1000 hot-leg. The proposed test facility consists of a hot-leg pipe, a reactor vessel simulator (RVs), and a steam generator simulator (SGs). It was designed and constructed based on power-to-volume scaling approach by preserving the Froud number in the hot leg. The scaling ratio was considered 1/19 of the Bushehr NPP
-
Subdiffusive processes in BWRs Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-15 Gilberto Espinosa-Paredes, Vishwesh A. Vyawahare, Érick-G. Espinosa-Martínez
The objective of this work is the analysis of subdiffusive processes in BWR reactors. Nuclear reactors are highly heterogeneous systems where the phenomena at the reactor scale are not possible to describe with constitutive standard laws (normal diffusion) of energy, momentum, and mass. However, there are methods to achieve it such as the volumetric average to upscale the reactor and incorporate up-scaled
-
Numerical investigation of the migration and retention of corrosion particles in a lead-cooled fast reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-15 Yishou Xu, Xinggang Li, Xiaodong Huang, Xiaoxin Zhang, Chen Hu, Xian Zeng, Qingzhi Yan
The lead–bismuth eutectic (LBE) coolant corrodes structural materials during nuclear reactor operations. The corrosion products accumulate and are deposited after long-term operations, blocking the narrow flow paths inside the reactor loop and reducing reactor performance. This study investigates the effects of the flow field and particle characteristics on the corrosion products. The discrete phase
-
Critical configuration of a SMR based on CAREM 25 using the SERPENT code Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-14 Marco Antonio C. Lima, Edson Henrice, Daniel A.P. Palma, Amir Zacarias Mesquita
In this paper, the simulation of a conceptual reactor based on the Argentinian prototype SMR CAREM 25 is presented. The reference reactor is a project of the National Atomic Energy Commission of Argentina (CNEA). CAREM 25 represents the strides made on the development and construction of SMRs in Latin America. The reason for choosing this particular model is that it fulfills the main requirement for
-
Vibration response analysis of mobile liquid Pb-Bi micro-reactors under transportation with liquid sloshing Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-13 Tiandi Fan, Yong Zhang, Guowei Yang, Yong Song, Jieqiong Jiang, Tao zhou
Mobile micro-reactors, which used the liquid lead–bismuth eutectic (LBE) as coolants, providing energy to remote regions had become a popular topic in nuclear energy due to their desirable characteristics, including inherent safety and modularity. However, the liquid sloshing effect in a reactor may exhibit fluid–structure interaction (FSI) coupling effects, which could have a significant impact on
-
Fuel element microreactor integrating a square UO2 fuel rod with an internal heat pipe Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-12 Horus Ibrahim Orlandi, Pedro Carlos Russo Rossi, Adolfo Aguiar Braid, João Manoel Losada Moreira
In this study we propose the Fuel Element Micro Reactor (FEMR), with 308 square UO fuel rods, each one containing inside a heat pipe. We selected Mercury as the working fluid for the heat pipes and operate the microreactor at average temperature around 900 K or 637 °C. Heat pipes introduce empty regions into the core that increase neutron leakage and severely reduce core reactivity. Square fuel lattices
-
Augmentation of hot cell facility for pyro-process of irradiated metal fuels Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-12 D. Bola Sankar, T. Kalaiyarasu, R. Karunakaran, S. Rajeswari, R. Padmanaban, B. Arul Kumar, M. Masanam, E. Mohanraj, N. Ravi, P. Manoravi, V. Jayaraman
Hot cells with an inert atmosphere are used for handling air-sensitive, hygroscopic, and radioactive materials like mixed carbide fuels and for pyro-processing metallic fuels. Pyro-processing of spent metal fuels was based on the electrochemical recovery of actinides in high-temperature molten salts (LiCl-KCl eutectic with 58.5 mol.% LiCl). To demonstrate the remote operation feasibility of the electro-refining
-
Fractional scaling analysis of cladding temperature in large break loss of coolant accident Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-12 B.H. Yan, B. Long, L.G. Li, J. Xing, D.H. Lu
In integral test facilities, some phenomena relevant in the separate effect test are usually neglected due to scaling limitations and design compromises. The typical heat transfer phenomenon in reactor core in large break loss of coolant accident (LBLOCA) scenario is analyzed with fractional scaling analysis (FSA) method in this work. A conservation form of cladding temperature is derived on the basis
-
Accident classification methodology with don’t know response for PWR nuclear reactors using the cuckoo optimization algorithm and principal component analysis method Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-11 Diego J.S.N. de Souza, Marcelo C. Santos, Andressa S. Nicolau, Roberto Schirru
This study tackles the Nuclear Accident Identification Problem (NAIP) in Nuclear Power Plants (NPPs), focusing on employing the Cuckoo Optimization Algorithm (COA). The methodology involves classifying anomalous events using data from normal operational conditions and three design basis accidents within the simulated plant state dataset of the Brazilian NPP Angra 2. The classification process is enhanced
-
Transient Monte Carlo simulations with OpenMC(TD): A catalyst towards advancing research in next-generation reactors and to improve fission nuclear data Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-11 J. Romero-Barrientos, F. Molina, M. Zambra, F. López-Usquiano
OpenMC(Time-Dependent) or OpenMC(TD), is a modified version of OpenMC which includes the time dependence related to the emission of -delayed neutron emission from individual precursors. Although the code is still in ongoing development and has been tested in simple systems, its results so far have been promising. The aim of this work is to show that OpenMC(TD) can be a valuable Monte Carlo tool and
-
A non-equilibrium phase change model based on the Lee model for flashing flows in converging–diverging nozzles Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-11 Zihua Liu, Shun Wang, Zhikang Lin, Xiong Wang, Di Wang, Yong Ouyang, Dalin Zhang
Flashing flow plays a crucial role in the occurrence of steam generator tube rupture (SGTR) accidents in advanced nuclear reactors. In this study, a simplified non-equilibrium phase change model is developed by modifying the Lee model. The modified model expands the scope of Lee model, allowing it to simulate phase change scenarios in response to abrupt pressure changes. Through adjustments to the
-
A new compartmental fractional neutron point kinetic equations with different fractional orders Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-11 Gilberto Espinosa-Paredes, Carlos-Antonio Cruz-López
-
Numerical investigation on steam condensation and heat transfer in an emergency condenser tube with the thermo-hydraulic system code ATHLET Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-11 Matthias Jobst, Frank Schäfer, Sören Kliem
-
Thin liquid film method for analyzing gas–liquid annular flow in nonstraight pipe components Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-09 Ri Zhang, Mengyan Ding
This study employs the thin liquid film method (TLFM) to analyze annular flow in nonstraight pipe components. The TLFM considers the interaction among the liquid film, gas phase, and droplets and requires a specific numerical method for implementation. Following thorough verification against experimental results from straight pipelines, the TLFM is applied to investigate the evolution of annular flow
-
Human reliability analysis of intelligent mine human–computer interaction based on improved SPAR-H Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-09 Xiaofang Yuan, Xin Jiang, Zepeng Shi, Linhui Sun
In order to evaluate the human reliability of human–computer–interaction (HCI) in intelligent mines and prevent the human performance error of the operators, this study base on the SPAR-H method and take the chute monitoring center as the research object to develop a human reliability analysis method suitable for smart mines. First, the basic error probability of human error in the HCI is determined
-
Numerical study of TRISO particles with random size and location distribution in cuboid and cylindrical matrix: A validation for two-regime heat conduction model Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-08 Chunxi Wu, Guitao Yang, Wei Zhang, Xianglong Guo, Maolong Liu
During reactor operation, the precise size values or specific size distribution of tri-structural isotropic (TRISO) fuel particles randomly dispersed within the fully ceramic microencapsulated (FCM) fuel elements are often unknown. To investigate the impact of the random distribution of positions and sizes of TRISO particles within a cuboid or cylindrical matrix on the temperature distribution of fuel
-
A PSA-based approach for the evaluation of plant level testing strategies Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-08 Douglas Barreto de Oliveira Fonseca, José de Jesús Rivero Oliva
While existing literature comprehensively outlines the pros and cons of simultaneous, sequential, or staggered testing of redundant trains, the combined effect of these strategies on plant level risk has been underexplored. In order to investigate plant level testing strategies, this paper proposes an approach based on Probabilistic Safety Assessment (PSA) methods that utilizes time-dependent reevaluation
-
A mass and energy balance for a specific refuelling of Angra 2 nuclear power plant from a life cycle perspective Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-06 P.J.M. Figueiredo, P.F. Frutuoso e Melo, A.C.M. Alvim, A.S.M. Alves
-
Tri-Loop design and thermoeconomic analysis for the high temperature gas cooled nuclear reactor coupling energy storage system Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-06 Zhuojun Jiang, Wan Sun, Simiao Tang, Longxiang Zhu, Luteng Zhang, Liangming Pan
Thermal energy storage is a proposed solution that enables nuclear power plants to adjust their output without altering power levels. This technology manages fluctuations in the power generation process by storing generated heat above demand levels until it is required to produce steam. Despite the initial investment needed for infrastructure construction, these costs are easily offset by the ultimate
-
Prognostic model and failure mechanisms of steam generators in Sodium-Cooled fast reactors Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-06 Xinyan Wang, Xingang Zhao, Birdy Phathanapirom, Kyle Warns, Junyung Kim, Hyun Gook Kang, Michael Golay
-
Thermal conductivity of lead and bismuth-lead eutectic melts up to 1300 K Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-06 A.Sh. Agazhanov, S.V. Stankus, I.V. Savchenko, D.A. Samoshkin
Using a laser flash method, the thermal conductivity () of Pb and eutectic alloy Bi-Pb (55.2 wt%Bi) melts is measured in the temperature range from the melting point to 1300 K with 3.2–6.0 % uncertainty. The results are compared with the data of other authors. Based on the measurements, the thermal diffusivity () and the Lorenz number of melts are calculated. Tables of recommended data for and along
-
Filtering of high frequency motion due to the gap effect on bolted anchorage Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-05 Thuong Anh Nguyen, Guillaume Hervé-Secourgeon
In the context of the equipment qualification against induced vibration due to the impact on reinforced concrete structures, the question of the transmission of high frequency and high amplitude motion through bolted anchorages of the equipment is raised. To better comprehend this phenomenon, the 3rd phase of the “mproving obustness assessment of structures mpacted by missile” (IRIS 3) impact tests
-
Status and future prospects of nuclear industry development in Bulgaria Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-05 Pavlin Groudev, Neli Zaharieva, Antoaneta Stefanova
This paper presents a study of nuclear energy development in Bulgaria and its contribution to the country's energy sector. Current status and future prospects of the nuclear power industry have been discussed.
-
Design and dynamic analysis of transport cask for SMR fresh fuel assembly Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-05 Xicheng Wang, Dingqu Wang, Songyang Li, Yueyuan Jiang
Transportation of fresh fuel assemblies requires the utilization of transport casks to ensure the internal radioactive material does not affect the environment. In this work, we introduce an innovative design of a transport cask for the fresh fuel assembly of a Small Module Reactor (SMR). The cask is constituted by two containers, supports and various energy-absorbing structures welded at the outer
-
The AGR-like FHR reactor: Assessing the technical limits of the fuel Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-05 A. de Lara, D. Pizzocri, P. Van Uffelen, E. Shwageraus
This work presents the detailed fuel rod simulation with the new TRANSURANUS code in combination with a high fidelity neutronic analysis of an innovative British advanced gas cooled type reactor with new materials by means of Serpent. The Hastelloy cladding material and molten salt properties have been assessed and implemented in our previous work. In this paper, we evaluate the effect of uncertainties
-
National context of the recent Spanish research on nuclear fission technology Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-05 L.E. Herranz, J. Freixa, S. de Carlos, M. Sánchez-Perea
-
Online practical education and training in nuclear engineering: A methodological framework Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-05 Lenka Frybortova, Lubomir Sklenka, Ondrej Novak, Branislav Vrban, Jakub Luley, Stefan Cerba, Agnieszka Korgul, Aleksandra Fijalkowska, Andrea Salvini, Daniele Alloni, Andrea Gandini
The education of students in the field of nuclear engineering must combine theoretical and practical teaching, this is the only way to produce quality graduates. Practical education generally includes two areas — experimental teaching and the use of advanced computational codes for the simulation of physical phenomena. The availability of practical education is not a matter of course for all students
-
Nuclear education and training activities of the Joint Research Centre of the European Commission: Maintaining and enhancing nuclear skills and competences Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-05 Eirini Michailidou, Manuel Martin Ramos, Jean Galy, Brian Eriksen, Arne Eriksson, Margarida Goulart, Willem Janssens, Veronique Berthou, Rachel Eloirdi, Alfred Morgenstern, Alban Kellerbauer, Paul Van Uffelen, Alice Seibert, Klaus Mayer, Janos Bagi, Philippe Raison, Yetunde Aregbe, Peter Schillebeeckx, Franck Wastin, Andrea Piagentini
Within the Euratom Research and Training Programme, the European Commission’s Joint Research Centre (JRC) implements nuclear education and training initiatives which support EU policy priorities and contribute to maintaining and developing EU’s nuclear competence and expertise.
-
Investigation of VVER-1000 fuel assembly bowing effect on power distribution during cycle using neutron noise adiabatic approximation Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-03 Javad Vosoughi, Naser Vosoughi, Ali Akbar Salehi
Thermomechanical and radiation loads in the reactor core cause lateral changes in the shape of Fuel Assemblies (FAs). This phenomenon, known as FA bowing, has significant implications for reactor safety and operation. Bowing results in alterations to the spacing between FAs, which in turn leads to neutron flux perturbations affecting power distribution. These perturbations are time-dependent due to
-
Neutronic evaluation of diverse fuel configurations for the supercritical water reactor (SCWR) core Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-03 Md. Abidur Rahman Ishraq, Fahim Mahmud Razu, H. Rainad Khan Rohan
This study aims to investigate the feasibility of five different fuel configurations for the supercritical water reactor (SCWR) core. Thorium, MOX and transuranic (TRU) fuels were considered alongside conventional UO, and burnup simulations were conducted using the Monte Carlo code SERPENT for an extended period of 1800 EFPDs. All of the fuels, except thorium, successfully demonstrate cycle lengths
-
Potentials in “nonproliferating” nuclear fuel: Design and implications on a PWR fuel cycle Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-03 Mustafa J. Bolukbasi, Marat Margulis
Nuclear energy is considered a critical component for achieving low carbon emissions in light of the urgent need for decarbonisation to meet global climate goals. However, concerns about the potential for nuclear weapon proliferation persist due to the uncontrolled spread of nuclear materials. This study focused on assessing the impact of proliferation-resistant fuel, doped with Am, on operating a
-
Quantifying the impact of risk mitigation measures using SPAR-H and RCM Approaches: Case study based on VVER-1000 systems Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-02 S. Kordalivand, R. Akbari, M. Abbasi
Probabilistic Safety Assessment (PSA) provides a systematic and quantitative approach to assess the risk associated with different components and systems within a nuclear power plant. By quantifying the impact of risk mitigation measures, decision-makers can prioritize and implement solutions that have the most significant impact on reducing risk. Fault Tree Analysis (FTA) is employed as a systematic
-
An improved lump mass stick model of a nuclear power plant based on the Kriging surrogate model Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-02 Dayang Wang, Wanruo Chen, Yong Zhu, Yongshan Zhang, Yaochu Fang
This paper focuses on an improved lump mass stick model of the AP1000 nuclear power plant (NPP). The conventional method is used to calculate the parameters of the lump mass stick (LMS) model of the AP1000 NPP, which has great differences between the LMS model and the finite element (FEM) model in terms of frequency and vibration mode. Therefore, with optimizing the mass condensation size of a shear
-
Experimental study of friction coefficient of graphite for high-temperature gas-cooled reactors Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-02 Zeliang Chen, Nan Gui, Yanfei Sun, Xingtuan Yang, Jiyuan Tu, Shengyao Jiang
-
High temperature nanoindentation of (U,Ce)O2 compounds Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-02 D. Frazer, T.A. Saleh, T. Matsumoto, S. Hirooka, M. Kato, K. McClellan, J.T. White
Continuing to refine our knowledge of the evolving mechanical properties of nuclear fuel over the entire fuel service cycle is necessary to understand the pellet-clad mechanical interaction that occurs in the fuel rods during the operation. A challenge with measuring the mechanical properties of irradiated fuels is their high levels of radioactivity that usually require the use of hot cells making
-
Optimized modular nuclear reactor project utilizing artificial intelligence: Seed-blanket concept Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-01 Marcelo V. Silva, Roberto Schirru, Giovanni Laranjo de Stefani, Andressa S. Nicolau, Alan Miranda M. de Lima, Claudio M.N.A. Pereira, João Victor S.A. Guimaraes, Diego Manoel E. Gonçalves
This article proposes a Light Water Small Modular Reactor (LW-SMR) based on the seed-blanket concept with thorium fuel, utilizing Artificial Intelligence (AI) for optimization. The objectives include defining SMR parameters and developing parallel multi-objective AI models using Genetic Algorithm (GA) and Particle Swarm Optimization PSO. A GA has been developed and applied to a simplified fuel assembly
-
Research on sensor data optimization technology for thermal hydraulic experiment of nuclear reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-01 Liu Yongchao, Li Tong, Xiao Kai, Chen Jie, Tan Xin, Cheng Jiahao, Tan Sichao, Wang Bo, He Zhengxi, Shen Jihong, Gao Puzhen, Tian Ruifeng
-
ARTISANS—Artificial Intelligence for Simulation of Advanced Nuclear Systems for Nuclear Fission Technology Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-04-01 Alexandra Akins, Aidan Furlong, Lauren Kohler, Jason Clifford, Christopher Brady, Farah Alsafadi, Xu Wu
The objective of this Technical Opinion Paper (TOP) is to provide an overview of the research topics in the ARTISANS (Artificial Intelligence for Simulation of Advanced Nuclear Systems) research group at the North Carolina State University. We will discuss the connections between our research with the key items outlined in the Virtual Special Issues (VSI) Nuclear Fission Technology (NFT) series. NFT
-
A mechanism for spontaneous thermal fragmentation with coolant entrainment during the molten fuel–sodium interaction Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-30 Michael Johnson, Yuki Emura, Remi Clavier, Ken-ichi Matsuba, Kenji Kamiyama, Claude Brayer, Christophe Journeau
Experimental investigation of two interactions between metallic corium jets and sodium, pertaining to severe accidents in a sodium-cooled fast reactor, have been undertaken at the MELT facility. X-ray imaging and debris analysis reveal rapid formation of a crust at the melt coolant-interface, instigating thermal fragmentation events. Heat transfer calculations at the jet-coolant interface, supported
-
Transient analysis of temperature field and fission gas about the arc plate fuel element Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-30 Linjuan Zhu, Qin Zeng, Yangbin Xiong, Jinggang Xu
Obtaining the temperature field and fission gas release of fuel elements are crucial to the study on fuel behaviors and ensure the safety of elements. In order to enhance fuel efficiency, providing a design idea to make the fast reactor with high flux and multi-function, we have designed a core of the lead-based fast reactor with arc plate fuel elements. Based on above, a transient performance analysis
-
Experimental investigation of the effect of geometrical and operating variables on condensation induced water hammer in horizontally oriented pipes Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-29 Anu Dutta, I. Thangamani, P. Goyal, V. Verma, J. Chattopadhyay, Seik Mansoor Ali, L. Raj
Condensation-Induced Water Hammer (CIWH) phenomenon involves dynamic pressure changes caused by rapid condensation of steam in subcooled water. The magnitude of pressure spike is often large and can cause damage to the pipes and components in steam-water systems. Understanding the phenomena is vital for obtaining a safe and reliable design. This paper describes experimental investigations carried out
-
A knowledge gap analysis for transient CHF prediction within RELAP5-3D Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-29 Nicholas A. Meehan, Charles P. Folsom, Nicholas R. Brown
We compare several experimental pool and flow boiling critical heat flux experiments with corresponding computational models developed with RELAP5-3D. Encompassed within this paper are several sensitivity analyses for the boiling heat transfer correlations, critical heat flux (CHF) prediction within RELAP5-3D and the thermal properties of the cladding for experiments in the Transient Reactor Test Loop
-
Multiphysics and multiscale simulation for an experimental sodium-cooled fast reactor Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-29 Roberto Lopez-Solis, Gilberto Espinosa-Paredes, Alejandría D. Perez-Valseca, Carlos-Antonio Cruz-López
-
Study on surface erosion threshold and erosion mass of bentonite applied in nuclear waste repository based on fractal structure characteristics Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-28 Xiaoyue Li, Xinjiang Zheng, Yongfu Xu
Bentonite exhibits good self-sealing performance and low permeability after hydration, and therefore is selected as a buffer material for high-level radioactive waste repositories, isolating the canisters containing nuclear waste from the external environment. The bentonite cushion directly contacts the surrounding rock and generates gels at the contact interface with fracture water. Surface erosion
-
Study on neutronic behavior of VVER-1000 fuel assembly with duplex fuel rod design Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-28 Zuhair, Wahid Luthfi, Hery Adrial, Azizul Khakim, Suwoto
Various studies suggested that duplex fuel rods have better neutronic characteristics than ordinary single-oxide fuel pellets. A duplex fuel rod consists of an inner uranium dioxide (UO) layer and an outer thorium dioxide (ThO) layer. This paper aims to study the neutronic behavior of the VVER-1000 fuel assembly with duplex fuel rod design. The duplex fuel rods were loaded into the fuel assembly in
-
Mock-up conception for experimentally investigate the creep phenomenon caused by the fluid–structure interaction in a rod bundle Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-28 Guilherme Vidal, Guillaume Ricciardi, Emmanuel Lo Pinto, Vincent Faucher, Nicolas Lamorte, Julien Pacull
Pressurized water reactors (PWRs) are employed worldwide and continue to expand in capacity. They require thorough investigation, particularly concerning their safety and operational efficiency. To ensure these factors, comprehensive research into their various components and an understanding of the diverse phenomena they experience within a PWR are imperative. These phenomena can be of mechanical
-
A comprehensive analysis of thermal–hydraulic signatures in neutron noise of WWER-type reactors Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-28 A. Kamkar, M. Abbasi, O. Safarzadeh, A. Sheikhi
The neutron noise phenomenon, occurring in all types of nuclear reactors, provides valuable insights into the dynamic behavior of these reactors. In this work, a well-justified approach for modeling neutron noise induced by thermal–hydraulic sources in a hexagonal reactor core is presented in detail. These sources encompass fluctuations in key features of the inlet coolant, including flow rate, temperature
-
The research on flow-induced vibration of compact curved tube bundles subjected to liquid cross flow Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-28 He Zhu, Guoxun Zhang, Guangdong Liu, Shujian Tang, Li Zhang, Kun He
This paper presents a series of dynamic tests and analyses on single and compactly bundled straight and curved tubes subjected to liquid cross flow. The purpose of this research is to study the flow-induced vibration behavior of compact curved tube bundles, with particular interest in their differences to those of the straight tubes. Both modal and dynamic response tests are performed with tube structures
-
Improving the reactor safety aspects by the implementation of (Th-U233-Pu) fuel in a PWR assembly Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-28 Sayed Saeed Mustafa
-
Mechanistic model of critical heat flux in rod bundles based on a high-precision subchannel code Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-27 Junliang Guo, Jianqiang Shan, Li Jiang, Yujiao Peng, Miao Gui
A mechanistic model for predicting critical heat flux (CHF) in rod bundles is developed based on the high-precision subchannel code ATHAS. To account for the non-uniform distribution of quality in subchannels caused by the presence of mixing vane grids (MVGs) and guide tubes (GTs), further subdivision of conventional subchannels and the incorporation of MVG crossflow models are implemented in ATHAS
-
Development and validation of a new one-dimensional parallel transient system solver with coordinate transformation method developed in OpenFOAM Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-27 Gonglin Li, Hui Wang, Haozhi Bian, Ming Ding
Passive containment cooling system is an important component of both 3rd and 4th generation nuclear power plants. To simulate two-phase flow in loop-type passive containment cooling system, a new one-dimensional parallel transient system solver, ComspaFOAM-SYS1D, was developed based on OpenFOAM. Compared to the previously developed one-dimensional solver, the new solver can be accelerated using multithreading
-
A cyclic-track decomposition method for 3D MOC neutron transport simulation Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-27 An Wang, Junying Wang, Zhezhao Ding, Xiaoxu Geng, Haodong Shan, Yun Hu, Dandan Chen
The method of characteristics (MOC) is a promising numerical technique for solving the neutron transport equation. This method has been parallelized to accommodate its high computing power and memory requirements. Angle decomposition is widely used by parallel MOC codes to distribute MOC tracks over CPU cores, but it may lead to load imbalance. Based on modular ray tracing, we present a cyclic-track
-
Investigating geometry adjustments for enhanced performance in a PeLUIt-10 MWt pebble bed HTGR with OTTO refueling scheme Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-26 Fitria Miftasani, R Andika Putra Dwijayanto, Ghulam Abrar, Nina Widiawati, Nuri Trianti, Topan Setiadipura, Dwi Irwanto, Cici Wulandari, Zaki Suud
-
CFD simulations of helium layer erosion in one PANDA vessel following cooler activation Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-26 A. Dehbi, R. Kapulla
We performed CFD simulations of the two H2P6 cooler tests conducted in the PSI PANDA vessel as part of the OECD/NEA HYMERES-2 project. Initially, the vessel atmosphere consists of a 60/40 % vol. steam- air mixture. Then, a helium stratification (15 % vol.) is introduced in the upper half of the vessel. The two experiments H2P6_1 and H2P6_2 are characterized by the activation respectively of 3 and 1
-
Neutronic analysis of high burnup thorium-HALEU fuels in PHWR Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-26 N. Read, V. Raffuzzi
Recent developments have been made in a novel fuel for pressurised heavy water reactors: high burnup oxide fuels composed of a mixture of thorium and high assay low enriched uranium. Several claims are made regarding the performance of such fuels in terms of economics, waste and non-proliferation, among others. This article analyses this class of fuel using an infinite lattice study in SERPENT. First
-
Experimental results for RVK-500 recombiner tested in conditions typical for pressurized water NPP severe accidents Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-26 E.V. Bezgodov, M.V. Nikiforov, S.D. Pasyukov, A.A. Tarakanov, D.L. Moshkin, I.A. Popov, Yu.F. Davletchin, A.A. Ryakin, A.V. Koshcheev
Severe accidents at NPPs with light water coolant can lead to situations when large quantities of hydrogen are released due to zirconium-steam reaction. To avoid explosion consequences, they started to install hydrogen passive autocatalytic recombiners (PARs) at NPPs. Calculating simulation based on PAR models is used to predict and justify the needed quantity of PARs and the chosen place of their
-
Safety evaluation of Multiple Steam Generator Tube rupture accident using the best estimate plus uncertainty approach Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-26 Sebastian Dzień, Aya Diab
Following the disaster in Fukushima Nuclear Power Plant (NPP) in 2011, the awareness of securing the safety of NPP under extreme events has been raised. For this purpose, the concept of Design Extension Conditions (DECs) was introduced, to enhance the plant’s capability to withstand events that are more severe than Design Basis Accidents (DBA), such as Multiple Steam Generator Tube Rupture (MSGTR)
-
Flow fields prediction for data-driven model of 5 × 5 fuel rod bundles based on POD-RBFNN surrogate model Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-26 Guangyun Min, Yu Ma, Yahui Wang, Naibin Jiang
The fuel rod bundles are a crucial component of nuclear reactors. Research the flow characteristics of the fuel rod bundles can be computationally expensive, despite some progress made in the past few decades. In this paper, the fast prediction of flow fields for 5 × 5 fuel rod bundles is implemented based on a data-driven algorithm. First, a refined model of the 5 × 5 fuel rod bundles with spacer
-
Unfolding of the neutron spectrum in a nuclear reactor using genetic algorithms method Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-25 Mohamed Drissi El-Bouzaidi, T. El Bardouni, E. Chham, O. El Hajjaji, A. Nouayti, Abdelghani Idrissi, K. El-Bakkari
The objective of this paper is to present the results of the neutron spectrum unfolding using a high-performance code that we have developed, which is based on genetic algorithm. This paper presents the methodology used for neutron spectrum unfolding and describes the general architecture of our developed code. It also includes results of neutron spectrum unfolding in the case of the ARC-700 reactor
-
Scaling analysis of the Loop Seal Plugging and Clearing phenomena in a SB-LOCA transient on the BETHSY facility Nucl. Eng. Des. (IF 1.7) Pub Date : 2024-03-25 Antoine Ciechocki, Sofia Carnevali, Dominique Bestion, Lionel Rossi
The objective of this work is to investigate what controls the duration between the Loop Seal Plugging (LSP) and Clearing (LSC) phenomena occurring in a 6% cold leg break LOCA transient carried out on the French BETHSY facility. This influences the duration of a possible core uncovery and the height of the associated Peak Clad Temperature (PCT). This work presents these phenomena, which are complex