Nuclear Engineering and Design最新文献

{"title":"Thermal effects on gas migration in saturated bentonite under rigid boundary conditions","authors":"Sai Li ,&nbsp;Weimin Ye ,&nbsp;Qian Zhang ,&nbsp;Qiong Wang ,&nbsp;Yonggui Chen","doi":"10.1016/j.nucengdes.2025.114462","DOIUrl":"10.1016/j.nucengdes.2025.114462","url":null,"abstract":"<div><div>The influence of temperature on gas migration behavior constitutes a critical consideration for both the design and operational safety of deep geological repositories for high-level radioactive wastes (HLW). In this study, water injection and subsequent gas injection tests were performed on specimens with dry densities of 1.3, 1.5 and 1.7 Mg/m³ at temperatures 20, 40 and 60 °C. Meanwhile, the specimens that experienced related gas injection tests were subjected to mercury intrusion porosimetry (MIP) tests. The results indicate that the effects of temperature on the effective gas permeability are dependent on both the injection pressure and the initial dry density. Under low gas injection pressures, an increase in temperature leads to a rise in effective gas permeability, while under high gas injection pressures, the temperature impact on the effective gas permeability depends on dry density. For specimens with high dry densities, the effective gas permeability positively correlates with temperature, while for low dry densities, the value at 40 °C exceeds those at 20 and 60 °C. Additionally, the gas breakthrough pressure decreases with increasing temperature. Higher dry density specimens are more likely to experience capillary breakthrough, while interfacial breakthrough more commonly happens in lower dry density specimens. According to the microstructural observations from the MIP tests, increasing temperature reduces the specimen pore space due to the shrinkage of the bentonite matrix. These findings indicate that gas migration in saturated bentonite is governed by a competitive mechanism between pore structure and the state of pore water, while both of which are influenced by temperature.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114462"},"PeriodicalIF":2.1,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020682","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":"Performance of dry electrostatic precipitator for the removal of CsI aerosol particles under simulated severe nuclear accident conditions","authors":"Satish Basnet ,&nbsp;Atte Jäntti ,&nbsp;Pasi Yli-Pirilä ,&nbsp;Miika Kortelainen ,&nbsp;Olli Sippula ,&nbsp;Anna Lähde","doi":"10.1016/j.nucengdes.2025.114440","DOIUrl":"10.1016/j.nucengdes.2025.114440","url":null,"abstract":"<div><div>The performance of electrostatic precipitators (ESPs) in reducing the emission of radioactive aerosols and hydrogen was investigated under varying experimental conditions to improve the safety features of nuclear power plants (NPPs). The effects of gaseous atmosphere, humidity, flow rate, and temperature on particle properties and filtration efficiency were evaluated using caesium iodide as a model aerosol. The results indicate that the ESP achieved a maximum particle mass filtration efficiency of over 90 % for the lab-scale ESP and more than 99.5 % for the industrial ESP. However, the particle number concentration varied with the industrial ESP, highlighting effective removal of larger particles that acted as a condensation sink, allowing higher concentrations of newly formed ultrafine particles to persist in the system. Hydrogen mitigation experiments revealed no measurable impact of ESPs on hydrogen concentrations under typical operating conditions, confirming their safety within current NPP protocols. The study highlights the crucial role of particle properties, carrier gas composition, and ESP design in determining filtration efficiency, emphasising the need for further research on reactor-specific conditions to optimise ESP performance and enhance source term reduction strategies.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114440"},"PeriodicalIF":2.1,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020684","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":"An integral effect test and its code simulation on the complete loss of reactor coolant system flowrate for the SMART100 design","authors":"Jin-Hwa Yang,&nbsp;Byong-Guk Jeon,&nbsp;Hwang Bae,&nbsp;Hyun-Sik Park","doi":"10.1016/j.nucengdes.2025.114431","DOIUrl":"10.1016/j.nucengdes.2025.114431","url":null,"abstract":"<div><div>A complete loss of reactor coolant system flowrate (CLOF) scenario was successfully tested using the integral effect test facility of SMART-ITL. The steady-state conditions were well achieved to satisfy initial test conditions presented in the test requirement; its boundary conditions were accurately simulated, and the CLOF scenario was reproduced properly for the SMART100 design. The CLOF test was performed over a long period of 60,000 s to understand the long-term behavior of the passive residual heat removal system (PRHRS) of the SMART100 design. The test results from the SMART-ITL were analyzed using the MARS-KS code to assess its capability to simulate a CLOF scenario for the SMART100 design. As the passive safety systems will operate for no less than 72 h to fulfill their function, the precise prediction of thermal–hydraulic behaviors in reactor coolant system (RCS) and passive safety systems of the SMART100 in the long-term sense is very crucial for the safety analyses of the SMART100 design. The measured thermal–hydraulic data from the CLOF test using the SMART-ITL were properly simulated using the MARS-KS code, which showed its reasonable simulation capability for the CLOF scenario of the SMART100 design with PRHRS. The thermal–hydraulic phenomena depend more on flow resistance both in RCS and PRHRS loops and heat losses from the RCS, steam generator, PRHRS loop and emergency cooldown tank during the long-term simulation. It is considered that accurate simulation is possible with proper consideration on flow resistance in the loops and heat loss through the heat structure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114431"},"PeriodicalIF":2.1,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020685","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":"Infinite refuelling equilibrium burnup analysis in pebble-bed HTR","authors":"Wang Yizhen ,&nbsp;Guo Jiong ,&nbsp;Zhang Han ,&nbsp;Wu Yingjie ,&nbsp;Hao Chen ,&nbsp;Li Fu","doi":"10.1016/j.nucengdes.2025.114432","DOIUrl":"10.1016/j.nucengdes.2025.114432","url":null,"abstract":"<div><div>Pebble-bed High Temperature Reactor (HTR) adopts multi-pass refuelling fuel management or MEDUL cycle, where fuel pebbles would run through the core multiple times before reaching their burnup value limits and being discharged. This cycle ultimately leads reactor to an equilibrium state whose characteristics are closely related to the allowed refuelling times in the cycle. Typically, increasing refuelling times permits a more homogeneous equilibrium state with lower maximum power density. Also, atomic density uncertainty, e.g. contributed from nuclear data, inside this equilibrium core would be reduced with increased refuelling times. Although it is beneficial for reactor operation safety, increasing refuelling times also burdens fuel handling system and shortens their service life. Analysing equilibrium state under hypothetical infinite refuelling times will reveal the limiting effect of refuelling times on the characteristics of equilibrium state in pebble-bed HTR, and it could be used to justify fuel management design. This work proposes a Lagrangian burnup framework based infinite refuelling burnup model. A burnup model constructed from HTR-PM (High-Temperature gas-cooled Reactor Pebble-bed Module) is used to verify the proposed infinite refuelling model, and results are compared with finite refuelling calculations. It is found that infinite refuelling highlights the limiting effects of refuelling times on equilibrium state in terms of neutron flux, regional power density, actinides and fission products’ batch averaged atomic density. Increasing refuelling times makes equilibrium state approaching infinite refuelling equilibrium state nonlinearly, and equilibrium state obtained from fifteen times refuelling is quite close to that obtained from infinite times refuelling. As a hypothetical model, the infinite refuelling equilibrium burnup model developed in this work could balance out the randomness as well as uncertainty associated with fuel pebbles’ movement inside pebble-bed HTR. It is expected to be a reference for multiple design and analysis of pebble-bed HTR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114432"},"PeriodicalIF":2.1,"publicationDate":"2025-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145005116","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":"Physics-Informed hybrid machine learning for critical heat flux prediction: A comparative analysis of modeling approaches","authors":"Huakang Wu,&nbsp;Minyang Gui,&nbsp;Di Wu","doi":"10.1016/j.nucengdes.2025.114434","DOIUrl":"10.1016/j.nucengdes.2025.114434","url":null,"abstract":"<div><div>Critical Heat Flux (CHF) is a crucial safety parameter in two-phase flow boiling systems, playing a fundamental role in the design and operation of nuclear reactors, heat exchangers, and other thermal systems. Traditional CHF prediction methods, such as the empirical correlations, Look-up Table (LUT) or mechanistic models, offer valuable physical insights but often struggle with accuracy under complex and untested operating conditions. Recently, machine learning (ML) techniques, particularly artificial neural networks (ANNs), have shown promise in capturing the nonlinear dependencies in CHF prediction. However, standalone ML models frequently suffer from limited physical consistency and poor extrapolation capability, restricting their reliability in engineering applications. To address these challenges, this study examines a physics-informed gray-box framework that integrates physical models with Multi-layer Perceptrons (MLPs) to enhance predictive accuracy and generalization. Two hybrid modeling approaches are explored: (1) LUT + MLP, where MLP refines the residuals of LUT-based predictions to improve accuracy, and (2) Mechanistic Model + MLP, which maintains physical consistency while optimizing empirical parameters critical to CHF prediction. The data evaluation results demonstrate that the proposed framework significantly outperforms traditional methods, including the standalone LUT and liquid sublayer dry-out model—achieving superior predictive accuracy and robustness within the evaluated operating envelope, particularly in complex flow conditions represented in the dataset.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114434"},"PeriodicalIF":2.1,"publicationDate":"2025-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145005045","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":"Three-Field subchannel modeling of CANDU fuel thermal-hydraulics with MEFISTO-T code","authors":"Jean-Marie Le Corre","doi":"10.1016/j.nucengdes.2025.114405","DOIUrl":"10.1016/j.nucengdes.2025.114405","url":null,"abstract":"<div><div>To support the development and validation of core thermal–hydraulic simulation codes, OECD/NEA organized a benchmark focused on thermal–hydraulic behavior of CANDU fuel assemblies. The benchmark was based on full-scale heat transfer experiments conducted in a 28-element horizontal test bundle, covering conditions representative of CANDU core operations. A non-blind phase was included, providing a limited set of axial pressure drop (single-phase and two-phase) and critical power data, followed by a blind phase for predictive assessment. The tested geometries simulated CANDU pressure tubes under both uncrept and crept (i.e., aging) conditions, as well as two different bearing pad heights. The crept geometry introduces a significant bypass on the upper side of the rod bundle, associated with complex crossflows that are notoriously challenging to capture by simulation codes. The Westinghouse MEFISTO-T subchannel analysis code, employing a mechanistic three-field model of annular two-phase flow and originally developed for BWR fuel applications, was applied to simulate all benchmark cases. The code calibration parameters related to form loss and drop deposition enhancement due to structural components were adjusted based on the provided dataset. For the non-blind phase, the code accurately predicts axial pressure drop and critical power, and correctly identifies the dryout rod and orientation, although with a downstream bias in axial location. Further results from the blind phase demonstrate generally good agreement, particularly in capturing the effect of pressure tube creep on axial pressure drop and the influence of system parameters (pressure, mass flow rate and inlet temperature) on critical power. However, the significant reduction in critical power observed in the tests due to pressure tube creep and lower bearing pad height was underpredicted, indicating areas for further model refinement. These results highlight the strong potential of applying the MEFISTO-T three-field model to CANDU fuel geometry, while also revealing opportunities for further improvements.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114405"},"PeriodicalIF":2.1,"publicationDate":"2025-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145005115","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":"Safety margin predictions of reactor pressure vessel integrity assessment procedures for pressurized thermal shocks","authors":"Klaus Heckmann,&nbsp;Jens Arndt,&nbsp;Jürgen Sievers,&nbsp;Ulrike Läuferts,&nbsp;Uwe Jendrich,&nbsp;Sara Beck,&nbsp;Frank Michel","doi":"10.1016/j.nucengdes.2025.114435","DOIUrl":"10.1016/j.nucengdes.2025.114435","url":null,"abstract":"<div><div>The integrity assessment of a reactor pressure vessel in a loss of coolant scenario with pressurized thermal shock (PTS) loading of the vessel wall is a crucial part of the safety review for the long-term operation, since the embrittlement of the vessel steel due to neutron irradiation continues during the ongoing operation. The procedures for the safety demonstration in different countries select different options for this assessment, namely concerning crack postulates, weld residual stresses, fracture toughness model, and warm pre-stress model. In order to understand the impact of these differences, the deterministic procedures of eight European countries are applied to two different example cases, which are based on the Konvoi and the VVER-440 plant type but are partly fictitious and not reflecting any particular unit. The safety margin in terms of ductile–brittle transition temperature is computed for each assessment, allowing to draw conclusions regarding the impact of different options within an assessment. In addition, it is possible to identify general trends of the procedures concerning their safety margin prediction.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114435"},"PeriodicalIF":2.1,"publicationDate":"2025-09-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144988731","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":"Pre-conceptual design of prestressed concrete containment for a GFR nuclear reactor","authors":"Petr Bílý,&nbsp;Jakub Holan,&nbsp;Roman Kubát","doi":"10.1016/j.nucengdes.2025.114425","DOIUrl":"10.1016/j.nucengdes.2025.114425","url":null,"abstract":"<div><div>The paper presents the process of a pre-conceptual design, analysis and optimization of a prestressed concrete containment vessel (PCCV) for a new type of gas-cooled fast reactor (GFR). The feasibility of the PCCV concept has been confirmed by a preliminary study presented in a previous paper. This paper builds on the prior results and presents an advanced analysis of the PCCV conducted using a global 3D non-linear coupled thermomechanical finite element model. The model is presented together with the results, revealing problems regarding unacceptable cracks or excessive compressive stresses in atypical areas of the structure (the gusset, the peripheral part of the dome, the ring beam, and the equipment hatch area). In the main part of the paper, various approaches to the modification of these regions are explored (e.g. modifications of geometry, prestress or steel reinforcement of the structure). The considerations that lead to the selection of the most beneficial modifications are explained in detail. The final version of the pre-conceptual design meets all of the defined criteria for the behaviour of the PCCV in the operation stage and during loss of coolant accident (LOCA). The main contributions of the paper are that it: (a) presents the process of modelling, analysis, and optimisation of a PCCV, (b) explores possible approaches to the modification of the atypical parts of a PCCV, (c) provides solutions for various design problems, and (d) presents and verifies a suitable design of a PCCV for a new type of GFR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114425"},"PeriodicalIF":2.1,"publicationDate":"2025-09-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144988733","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":"Optimization of district heating systems integrated with energy storage for enhanced economic performance","authors":"Ying Deng ,&nbsp;XiaoLong Li ,&nbsp;Shafa Guliyeva ,&nbsp;Umriya Kenjayeva","doi":"10.1016/j.nucengdes.2025.114393","DOIUrl":"10.1016/j.nucengdes.2025.114393","url":null,"abstract":"<div><div>A possible carbon–neutral substitute for district heating systems (DHSs) that rely on fossil fuels is the idea of a heat-only small modular reactor (SMR). However, high capital expenditures and decreased capacity factors—mainly from variations in demand in independent operations—hinder its economic sustainability. To overcome these obstacles, this research suggests an integrated strategy that combines district heating systems (DHSs), heat-only small modular reactors (SMRs), and Organic Rankine Cycle (ORC) power generation. To improve the system’s energy management, auxiliary parts, including heat storage, gas boilers, and electricity storage, are also evaluated. The optimization framework, which focuses on factors including equipment design capacity and hourly operating tactics, is described as a mixed-integer nonlinear programming (MINLP) problem. The main goal criterion is the net present value (NPV). A heat-only SMR (Teplator-150 MWt), a 20 MWe ORC power plant, a 10,000 MWth thermal energy storage unit, a 3.4 MWt gas boiler, and electricity storage with a capacity of 20 MWe/120 MWeh are all part of the ideal design for a typical DHS in Czechia with a peak demand of 41 MWt. Compared to a traditional heat-only supply system, this integrated system improved energy management, increasing the SMR’s capacity from 10% to 83%. This resulted in a 357 million euro profit instead of a 50-million-euro loss. Sensitivity analysis revealed important economic viability variables, such as the ORC system’s size and technology, interest rates, and power market prices. These results demonstrate the importance of energy management and strategic system integration in reaching financial and environmental goals.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114393"},"PeriodicalIF":2.1,"publicationDate":"2025-09-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144988732","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 human factors engineering-based verification of standing consoles design for SMR control rooms","authors":"Tulis Jojok Suryono ,&nbsp;Dian Fitri Atmoko ,&nbsp;Adhika Enggar Pamungkas ,&nbsp;Sudarno ,&nbsp;Kiswanta ,&nbsp;Sigit Santoso ,&nbsp;Restu Maerani ,&nbsp;Anik Purwaningsih ,&nbsp;Muksin Aji Setiawan ,&nbsp;Dhanu Dwiardhika","doi":"10.1016/j.nucengdes.2025.114436","DOIUrl":"10.1016/j.nucengdes.2025.114436","url":null,"abstract":"<div><div>In the context of emerging nuclear power programs, small modular reactor (SMR) control rooms must address both international safety standards and local ergonomic requirements. This study presents a verification-based evaluation of two standing operator console designs for the PeLUIt-40 (Indonesian SMR) main control room, adapted to the anthropometric data of Indonesian workers. The verification process was guided by ergonomic principles and key criteria outlined in NUREG-0700 and relevant national regulations. Using task-based usability tests and post-task questionnaires involving licensed Indonesian operators with substantial practical experience and an iPWR simulator, the consoles were assessed in terms of reachability, comfort, fatigue potential, and display visibility. Results demonstrate that a locally adapted console design improves operator posture, reduces wrist fatigue, and enhances accessibility without compromising operational task performance. The findings contribute to the early-stage verification of control room design and support subsequent validation phases as prescribed in NUREG-0711. This study provides practical insights for nuclear developers in emerging countries aiming to integrate human factors and accommodate operator diversity in the design of SMR control rooms. By embedding ergonomic principles into the conceptual design process, nuclear programs can enhance operator performance, minimize post-deployment modifications, and strengthen safety in complex system environments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114436"},"PeriodicalIF":2.1,"publicationDate":"2025-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144932059","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}