Nuclear Engineering and Design最新文献

{"title":"Adaptive learning observer and radial basis function neural networks based fixed-time fault-tolerant control of load following for a MHTGR with CRDM faults","authors":"Wangheng Liu ,&nbsp;Hongliang Liu ,&nbsp;Zigen Ouyang ,&nbsp;Wenjie Zeng ,&nbsp;Hua Liu","doi":"10.1016/j.nucengdes.2025.113872","DOIUrl":"10.1016/j.nucengdes.2025.113872","url":null,"abstract":"<div><div>Load following of the Modular High-Temperature Gas-Cooled Reactor (MHTGR) under Control Rod Drive Mechanism (CRDM) faults and disturbances remains a major challenge. This paper focuses on proposing a fixed-time fault-tolerant control method for this issue without considering the sensitivities associated with parameter setting. Firstly, to reconstruct some unmeasurable states of the MHTGR and the values of CRDM faults, an adaptive learning observer is established. Based on the learning characteristic of Radial Basis Function Neural Networks (RBFNN), the lumped uncertainties can be approximated. And then a fixed-time fault-tolerant controller is developed to ensure that the actual load output of the MHTGR actually tracks the expected output power within a fixed time, which can be determined through the system and controller parameters. Finally, simulations under two operational conditions demonstrate the control method is effective and feasible to the MHTGR system under disturbance and CRDM faults.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113872"},"PeriodicalIF":1.9,"publicationDate":"2025-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143349963","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Analysis of separation characteristics of variable plate-hook spacing corrugated plate separator","authors":"Qi Qin ,&nbsp;Rulei Sun ,&nbsp;Shouxu Qiao ,&nbsp;Ruifeng Tian ,&nbsp;Sichao Tan","doi":"10.1016/j.nucengdes.2025.113885","DOIUrl":"10.1016/j.nucengdes.2025.113885","url":null,"abstract":"<div><div>As an important steam-water separation device in nuclear power steam generators, the corrugated plate separator can separate water from steam, improve the dryness of the steam delivered, and ensure the safe operation of steam turbines and other equipment. This paper designs a corrugated plate structure with variable plate-hook spacing, and based on this structure, designs single-hook and double-hook corrugated plates with variable plate-hook spacing. At the same time, using numerical calculation methods, the performance of the corrugated plates with variable plate-hook spacing is compared with that of single-hook and double-hook corrugated plates with fixed plate-hook spacing. The separation efficiency and pressure drop are evaluated from two aspects, and the flow characteristics of internal airflow and the separation characteristics of droplets are discussed. Compared with traditional single-hook and double-hook corrugated plates, the gradual increase in plate-hook spacing promotes continuous acceleration of airflow, enhances inertial collisions between droplets and between droplets and solid walls, and achieves higher separation efficiency at lower inlet airflow speeds.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113885"},"PeriodicalIF":1.9,"publicationDate":"2025-02-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143349964","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Implementation of assembly shuffling for maximizing fuel utilization in the VVER-1200 reactor","authors":"Afroza Shelley ,&nbsp;H. Rainad Khan Rohan ,&nbsp;Md. Abidur Rahman Ishraq","doi":"10.1016/j.nucengdes.2025.113901","DOIUrl":"10.1016/j.nucengdes.2025.113901","url":null,"abstract":"<div><div>This study aims to maximize reactor fuel utilization by employing assembly shuffling instead of traditional refueling. A homogeneous VVER-1200 core containing 163 fuel assemblies (4.95 wt.% enriched) and excluding control rods/absorber banks is taken as the reference and burnt for 950 EFPDs. It is found that the discharge burnup of the central 54 fuel assemblies (Zone-1) is 41.2 MWD/kg, the middle 55 assemblies (Zone-2) is 36.9 MWD/kg, and the outermost 54 assemblies (Zone-3) is 27.5 MWD/kg. Burnup of Zone-3 increased by 27.4% after interchanging fuel assemblies with those of Zone-1 at 500 EFPD, which is considered as the first cycle. Simultaneously, the k_eff jumped from 1.093 to 1.151, maintaining criticality up to 900 EFPD. However, further interchange of assemblies among the zones did not improve k_eff significantly. The fuel and moderator temperature coefficients at BOL were −1.898 pcm/K and −58.106 pcm/K respectively, and these became more negative through shuffling, especially towards the end of core life. Despite an immediate rise in radial power peaking factor, the assembly-wise linear power distribution became more uniform as burnup progressed. The beta-effective peaked immediately after shuffling but declined gradually afterwards. Although 1.16% rise in fissile ²³⁹Pu was observed in the discharged fuel, the total quantity of minor actinides and long-lived fission products decreased by 1.63%.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113901"},"PeriodicalIF":1.9,"publicationDate":"2025-02-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143291727","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Modal seismic analysis of reactor internals considering the initial stress at normal operating conditions: A study on Atucha-I NPP control rods","authors":"Alejandro E. Albanesi ,&nbsp;Santiago M. Rabazzi ,&nbsp;Braian A. Desía ,&nbsp;Juan E. Ramos Nervi ,&nbsp;Javier W. Signorelli","doi":"10.1016/j.nucengdes.2025.113890","DOIUrl":"10.1016/j.nucengdes.2025.113890","url":null,"abstract":"<div><div>Modal seismic analysis evaluates the seismic response of nuclear plants by combining dominant vibration modes under a prescribed pseudo-acceleration spectrum. Such analyses typically consider only the thermal and kinematic boundary conditions for normal operation, but initial stresses in this state significantly affect vibration modes and modal seismic response. Simply considering all boundary conditions is insufficient to account for initial stresses; a prior thermo-mechanical analysis is mandatory to compute the geometric stiffness and redistribution of internal forces under initial stresses. This study investigates the modal seismic response of the guide tubes and control rods of the Atucha-1 Pressurized Heavy Water Reactor, comparing configurations with and without initial stresses, given their oblique arrangement makes them susceptible to developing such initial stresses. Boundary conditions were provided by Nucleoeléctrica Argentina S.A. (NA-SA) and the 3-D finite element model was built and analyzed with the open-source suite Salome<span><math><mo>−</mo></math></span>Meca v2022. Results indicate that initial stresses alter the effective stiffness of the piece by redistributing internal forces, increasing the natural frequencies by about 50%, shifting away from the seismic excitation frequency range. This reduces the modal seismic displacement by one order of magnitude, leading to more realistic responses that may extend operational limits within safety margins.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113890"},"PeriodicalIF":1.9,"publicationDate":"2025-02-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170315","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A comprehensive review of pulsed research reactors: Focus on Russian periodic fast reactors","authors":"A.A. Hassan ,&nbsp;M.V. Bulavin ,&nbsp;D. Myktybekov ,&nbsp;Ayman M. Abdalla ,&nbsp;O. Ashraf","doi":"10.1016/j.nucengdes.2024.113810","DOIUrl":"10.1016/j.nucengdes.2024.113810","url":null,"abstract":"<div><div>Research reactors play a crucial role in advancing scientific understanding, medical diagnostics, and nuclear technology development. This review article provides a comprehensive examination of research reactors, emphasizing classification and highlighting the unique features of periodic pulsed reactors. The classification of research reactors is elucidated, distinguishing between stationary and pulsed reactors. Stationary reactors serve diverse purposes, while pulsed reactors offer distinct advantages in certain applications as time-of-flight research and neutron generation lifetime calculation. Particular attention is given to research fast periodic pulsed reactors in Russia, showcasing their significant contributions to scientific research and technological innovation. Detailed discussions include the IBR-1, IBR-30, IBR-2, and NEPTUNE reactors, elucidating their design, capabilities, and scientific achievements. Furthermore, the article explores the utilization of Neptunium-237 as a promising nuclear fuel for fast pulsed reactors, highlighting advantages such as low neutron generation time, high neutron flux, and the ability to burn up part of dangerous nuclear waste. Overall, this review consolidates essential knowledge on research reactors and underscores the importance of ongoing research and development efforts, particularly in harnessing innovative fuels like Neptunium-237, to meet the evolving demands of science and technology.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113810"},"PeriodicalIF":1.9,"publicationDate":"2025-02-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143291728","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Fabrication of surrogate oxide spent fuel with various cracking patterns and design of an axial gas transport apparatus","authors":"Seongtae Kwon,&nbsp;Kyle A. Gamble,&nbsp;Fabiola Cappia,&nbsp;Chase E. Christen,&nbsp;Kaustubh Bawane","doi":"10.1016/j.nucengdes.2025.113883","DOIUrl":"10.1016/j.nucengdes.2025.113883","url":null,"abstract":"<div><div>Understanding gas transport behavior in nuclear fuel rods is important for the design, performance, and safety of nuclear fuels. Surrogate materials help enable efficient research by reducing both the costs and the amount of time required. A parametric study using Darcy’s law is completed that demonstrates the feasibility of observing pressure decay over short 13-to-15-cm specimens to enable full characterization of the fabricated specimens using x-ray computed tomography. This paper demonstrates that thermally shocked and mechanically compressed alumina pellets produce surrogate samples whose various cracking patterns are representative of the severity of cracking observed as a function of burnup in irradiated nuclear fuels. Furthermore, image analyses of the cracking patterns—in conjunction with gas transport testing using surrogate samples—affords a valuable accelerated basis for developing gas transport simulations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113883"},"PeriodicalIF":1.9,"publicationDate":"2025-02-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169295","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Dynamic study of a steam Rankine cycle using CATHARE-3 system thermal-hydraulic code: Application to Superphenix fast reactor","authors":"Vincent Audoly ,&nbsp;Laura Matteo ,&nbsp;Claire Vaglio-Gaudard ,&nbsp;Gédéon Mauger ,&nbsp;Nicolas Tauveron","doi":"10.1016/j.nucengdes.2025.113898","DOIUrl":"10.1016/j.nucengdes.2025.113898","url":null,"abstract":"<div><div>There is a growing interest for innovative power conversion systems for nuclear power plants in the context of the decarbonization of the energy mix. This motivates the need of dynamic models of Rankine cycle to be integrated in conception and safety analysis. In this paper, a simplified version of the steam Rankine cycle of the French sodium-cooled fast reactor Superphenix is modeled using the thermal–hydraulic code CATHARE-3. This completes previous works on the modeling of the primary and secondary circuits of this reactor with CATHARE and enables the simulation of the entire power plant. In the first part, parametric studies of the Rankine cycle enable the computation of the gross efficiency as a function of cooling water temperature and as a function of the power load. In the second part, transient simulations are performed. The regulations of the Rankine cycle are implemented in order to simulate a load following scenario. Then, the Rankine cycle is coupled with the CATHARE-3 models of the primary and the secondary sodium circuits (including the neutron point kinetics model for the core). Free dynamics scenarios are simulated: a step variation of the turbine admission valve and a step variation of the cooling water flow rate.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113898"},"PeriodicalIF":1.9,"publicationDate":"2025-02-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170317","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"NEAMS IRP challenge problem 3: Mixing in large enclosures and thermal stratification","authors":"A. Manera ,&nbsp;A.S. Iskhakov ,&nbsp;V.C. Leite ,&nbsp;Jiaxin Mao ,&nbsp;C. Tai ,&nbsp;Vf. Vishwakarma ,&nbsp;R. Wiser ,&nbsp;T. Nguyen ,&nbsp;C.J. Cummings ,&nbsp;E. Baglietto ,&nbsp;I.A. Bolotnov ,&nbsp;N.T. Dinh ,&nbsp;Y. Hassan ,&nbsp;V. Petrov ,&nbsp;E. Merzari","doi":"10.1016/j.nucengdes.2025.113847","DOIUrl":"10.1016/j.nucengdes.2025.113847","url":null,"abstract":"<div><div>Mixing in large enclosures and thermal stratification play critical roles in advanced reactor designs, including liquid metal-cooled and high-temperature gas reactors. Lessons from a recent international benchmark (<span><span>IAEA, 2017</span></span>), using system-level codes for Sodium-Cooled Fast Reactors (SFRs) highlight the need for improved models to accurately capture mixing and thermal stratification in the reactor hot pool upper plenum. These improvements are essential for predicting the propagation of stratification fronts and the effects on natural circulation and heat transfer between primary and intermediate loops. Current computational dynamics (CFD) codes, particularly those relying on Reynolds-averaged Navier-Stokes (RANS)-based turbulence models and the Simple Gradient-Diffusion Hypothesis (SGDH), underperform in simulating buoyancy-driven flows, leading to inaccurate predictions of stratified fronts. The NEAMS IRP Challenge Problem 3 (CP3) aims to develop multi-fidelity, multi-scale models for mixing and stratification in large enclosures. This includes models ranging from high-fidelity Large Eddy Simulations / Direct Numerical Simulations (DNS/LES) to system-level code models. High-resolution experiments and LES/DNS inform the development of these models, providing accurate and computationally affordable predictions. This paper provides an overview of ongoing experimental and modeling activities within CP3, showcasing advancements in understanding and predicting mixing and stratification in large enclosures for advanced reactor applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113847"},"PeriodicalIF":1.9,"publicationDate":"2025-02-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143291726","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"The effect of incorporating Cs, Sr and Eu nitrates on the matrix development of Fe-rich polymers","authors":"Evangelia D. Mooren ,&nbsp;Walter Bonani ,&nbsp;Antonio Bulgheroni ,&nbsp;Jorn Van De Sande ,&nbsp;Glenn Beersaerts ,&nbsp;Sonja Schreurs ,&nbsp;Rudy J.M. Konings ,&nbsp;Wouter Schroeyers","doi":"10.1016/j.nucengdes.2025.113895","DOIUrl":"10.1016/j.nucengdes.2025.113895","url":null,"abstract":"<div><div>Radionuclides like ¹³⁷Cs, ⁹⁰Sr and ^152+154Eu need to be immobilised from liquid radioactive waste to a suitable final encapsulation matrix. Alkali-activated Materials (AAMs) have the potential to be more effective in immobilising Cs⁺ and Sr²⁺ than Portland cement because they can produce stable phases and incorporate them into their structure. Less explored in AAMs is their capacity to immobilise Eu-ions. Nanoparticles are investigated for extracting radionuclides from liquid radioactive waste. CeO₂ nanoparticles have exhibited great potential in their ability to sorb Eu³⁺ but after several adsorption/desorption cycles also these need to be immobilised into a final encapsulation matrix. In this work, AAMs were prepared from synthetic Fe-rich slag. Two sodium silicate ratios were examined and the AAMs were doped with various mixtures of CsNO₃, Sr(NO₃)₂, Eu(NO₃)₃, and CeO₂ nanoparticles. Contaminants added to the AAM matrices can change the properties of the encapsulation system. To understand the impact on the AAM structure and to determine whether the effects are derived from the simulated radioactive Cs⁺, Sr^2+, and Eu³⁺, the presence of the CeO₂ nanoparticles or the presence of the nitrate ions, samples were examined during their matrix development using isothermal calorimetry, and investigations were made on their microstructural and physicochemical characteristics. The introduction of Cs⁺ to the matrix showed no notable impact on the activation kinetics, but Eu³⁺ seems to form Eu(OH)₃ similarly to Sr²⁺ which forms Sr(OH)₂ reducing the available hydroxides during the activation and ultimately hindering the polymerisation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113895"},"PeriodicalIF":1.9,"publicationDate":"2025-02-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169294","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Preparation and performance research of TRISO particle with niobium layer","authors":"Guoqiang Wang ,&nbsp;Lihua Guo ,&nbsp;Linyuan Lu ,&nbsp;Feng Zhang ,&nbsp;Xu Yang ,&nbsp;Jun Lin","doi":"10.1016/j.nucengdes.2025.113896","DOIUrl":"10.1016/j.nucengdes.2025.113896","url":null,"abstract":"<div><div>In this study, a niobium (Nb) layer was prepared by spouted-bed chemical vapor deposition (CVD) to replace the outer pyrolytic carbon (OPyC) layer in TRISO particles as a novel configuration for enhancing fuel performance. Such a configuration in TRISO particles is favorable for the following reasons: Nb has good mechanical properties, which can relieve the tensile stress on the silicon carbide (SiC) layer better with applied compressive stress compared with the OPyC layer. Furthermore, owing to its better thermal conductivity and good contact with the SiC layer, the Nb layer in the TRISO particle can transport the heat generated from the kernel to the coolant in a more efficient manner. In addition, the Nb layer is believed to protect SiC from mechanical damage better with its ductility behavior during the fuel element fabrication, while the OPyC layer is a brittle material. The deposited Nb layer has a thickness of 10 μm, indicating the designed characteristics such as high density and a pure phase. The interface between the SiC and Nb layers exhibits tighten bonding without any visible debonding or gaps. The nanoindentation results suggest that the average hardness of the Nb layer is approximately 5 GPa and the average Young’s modulus is about 110 GPa, which are remarkably higher than the OPyC layer. Simultaneously, the stress and failure fractions for both designs are calculated. The new configuration confirms that the Nb layer exerts a better compressive stress effect on the SiC and thus the failure fraction of fuel particles can be lowered to some extent. In addition, finite element calculations indicate that the application of the Nb layer in TRISO particle is conducive to reducing the peak temperature by about 50 K. Based on the above advantages, the performance of the reactor can be efficiently improved.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"433 ","pages":"Article 113896"},"PeriodicalIF":1.9,"publicationDate":"2025-02-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143169296","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}