Nuclear Engineering and Design最新文献

{"title":"Investigating the possibility of using iridium as a burnable absorber in new fuel pellet designs of VVER-1200 for reactivity management","authors":"Sultan J. Alsufyani ,&nbsp;Nassar Alnassar ,&nbsp;Mohammed Sallah ,&nbsp;Mohamed A.E. Abdel-Rahman ,&nbsp;Naima Amrani ,&nbsp;A. Abdelghafar Galahom","doi":"10.1016/j.nucengdes.2025.113989","DOIUrl":"10.1016/j.nucengdes.2025.113989","url":null,"abstract":"<div><div>Searching for the optimal design of the fuel assembly and the material needed to manage the reactivity in the nuclear reactor is still vital. Therefore, new geometry configurations and materials have been investigated in this work to handle the excess reactivity. Two designs of burnable absorber fuel pellets concentric shell model (CSM) and outer shell model (OSM) have been proposed for VVER-1200 assembly. Four BA materials including Gd₂O₃, Eu₂O₃, Er₂O₃ and Ir₂O₃ have been suggested to be investigated in the proposed fuel pellet designs. Various dimensions and concentrations of the suggested BAs have been studied in the OSM and CSM to obtain the optimum model. From both designs, the optimal models from k_inf behavior point of view have been selected and integral studies have been done on them. The neutronic analysis confirms the effectiveness of using erbium and iridium in the suggested models in managing the excess reactivity of VVER-1200.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113989"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610019","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":"Convolutional neural network use in fast neutron computed tomography for void fraction measurements","authors":"Qasim Siddiq,&nbsp;Garik G. Patterson,&nbsp;Basma Foad,&nbsp;David R. Novog","doi":"10.1016/j.nucengdes.2025.113975","DOIUrl":"10.1016/j.nucengdes.2025.113975","url":null,"abstract":"<div><div>Subchannel analysis codes are used for safety analysis and require suitable data to accurately model complex conditions that occur in fuel bundle geometries. To ensure accurate modelling of the complex two-phase flow phenomena in a reactor bundle, experimental data of the void distribution are required. Computed tomography provides a non-invasive method of measurement of the 2D or 3D distribution of void fraction within a bundle geometry. However, thick pressure vessels in full-scale conditions make it difficult to image bundle geometries and maintain contrast of the internal structures such as water/vapour when using photon-based sources. Fast neutron systems provide good penetration capability while maintaining sensitivity to the water-vapour contrast within the bundle but require long scan times (order of hours) because of the relatively poor detection efficiency and low source strength, which may preclude the application in full-scale thermal–hydraulic testing scenarios. Therefore, to effectively use Fast Neutron Computed Tomography (FNCT) in thermal-hydraulics safety experiments, the large scan times must be reduced, and the increased noise and decreased image fidelity that accompanies this reduction must be addressed.</div><div>This challenge is addressed in this paper using convolutional neural networks (CNNs), a branch of machine learning that excels in image processing tasks. A Synthetic nuclear fuel bundle image reconstructions training dataset, including noise and blurring effects, was generated using a custom MATLAB and Python. The CNN model uses the dataset to create a mapping between these noisy reconstructed images and their respective ground truth image. In this work, the Residual U-Net model significantly improves image reconstructions, leading to more accurate measurements of subchannel void fraction compared to the initial noisy images. The model is able to predict the subchannel void fraction with a mean absolute percentage error (MAPE) of 3.2 % ± 2.4 % on a subchannel basis. Here, MAPE refers to the mean of the absolute differences between the predicted and true void fractions, expressed in percentage points. This means that a void fraction prediction of 50 % with a true value of 53 % results in a 3 % error, not a relative percentage of the void fraction itself. The void fractions were predicted within 8.43 % of the true values for 95 % of the subchannels, demonstrating that the majority of predictions are highly accurate.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113975"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610017","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":"Transient local effects analysis during typical accidents of a heat pipe cooled reactor","authors":"Guanghui Jiao,&nbsp;Genglei Xia,&nbsp;Tao Zhou,&nbsp;Minjun Peng,&nbsp;Jianjun Wang","doi":"10.1016/j.nucengdes.2025.113988","DOIUrl":"10.1016/j.nucengdes.2025.113988","url":null,"abstract":"<div><div>Heat pipe cooled reactors (HPCRs) utilize high-temperature heat pipes to export fission heat to the energy conversion unit without requiring a coolant loop, simplifying the reactor core design while enhancing safety. Owing to its high safety standards and compact, lightweight design, this technology is the preferred solution for miniature nuclear energy applications in space and marine environments. Analyzing the transient local effects of HPCRs during typical accidents is essential for precise reactor control. However, a fast and practical computational model is still lacking. Therefore, a system code was established to assess the transient accidents between neutronics and thermal hydraulics within an HPCR core. The code includes a three-dimensional neutron physics model, a full-core thermal dynamics model for hexagonal fuel elements (FTDMH), and an improved multi-node heat pipe flow and heat transfer model. Validation was conducted using experimental data and COMSOL Multiphysics simulation results. Heat pipe failure and control drum misoperation accidents were simulated using the developed code, and the local effects of both accidents were analyzed. The results indicate that when a single control drum undergoes an unintended minor outward rotation, the reactor successfully returns to equilibrium through inherent negative feedback mechanisms. Fuel elements near the affected control drum show slight increases in temperature and power generation compared to normal operations, exhibiting a maximum temperature variation of 1.3  K and a peak power discrepancy of 1.02 %. In case of heat pipe failure, the resulting temperature rise in adjacent fuel elements remains below 30  K, and there is no initiation of a chain reaction leading to further heat pipe failures. The code provides a fast and efficient tool for the design and safety analysis of HPCRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113988"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610018","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":"Predictive control of a heat pipe-cooled reactor based on a neural network model","authors":"Xi Bai ,&nbsp;Jiajun Huang ,&nbsp;Zheyu Chen,&nbsp;Peiwei Sun,&nbsp;Xinyu Wei","doi":"10.1016/j.nucengdes.2025.113983","DOIUrl":"10.1016/j.nucengdes.2025.113983","url":null,"abstract":"<div><div>The adoption of heat pipes for heat transfer makes the heat pipe-cooled reactor (HPR) with significant time delays, making it difficult for traditional proportional integral derivative control systems to meet the rapid and precise power control demands. Therefore, neural network-based model predictive control is applied for HPR. NUSTER-100 model predictive control system utilizes a feedforward neural network model as the prediction model. By employing numerical optimization algorithms to obtain the optimal control variables, the system effectively overcomes the impact of significant time delays on the control system, achieving rapid and precise regulation. It can effectively suppress the impact of abnormal states on control performance, and ensure that heat pipe-cooled reactor can operate safely and stably under abnormal conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113983"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143620575","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":"Uncertainty in CFD simulation of reactors and approaches to improve the confidence of simulation results","authors":"Xu Lu ,&nbsp;Wengpeng Jia ,&nbsp;Lingyu Dong ,&nbsp;Yang Li ,&nbsp;Dandan Chen ,&nbsp;Genshen Chu","doi":"10.1016/j.nucengdes.2025.113974","DOIUrl":"10.1016/j.nucengdes.2025.113974","url":null,"abstract":"<div><div>Computational Fluid Dynamics (CFD) plays an important role in the design and optimization of reactor thermal-hydraulic systems and serves as an essential component of the ”virtual reactor” projects globally. However, the credibility of CFD simulations is often questioned due to various sources of error and uncertainty, limiting their broader applications in engineering practice. A rigorous process of Verification, Validation, and Uncertainty Quantification (V&amp;V&amp;UQ) is widely acknowledged as essential for assessing the credibility of CFD simulation results. Nevertheless, existing guidelines and standards remain largely theoretical, lacking practical implementation strategies, which prevents a comprehensive understanding of V&amp;V&amp;UQ among researchers. With the enhancement of computational resources, there is potential to improve the credibility of CFD simulation of reactors by reducing the introduction of errors and uncertainties while enhancing V&amp;V&amp;UQ tools and data accessibility. This paper identifies key sources of error and uncertainty across different stages of CFD simulation of reactors from a software developer’s perspective. Furthermore, it explores approaches to improve the credibility of CFD simulation of reactors, emphasizing four critical areas: minimizing model uncertainty, reducing numerical uncertainty, establishing a robust mesh quality evaluation framework, and advancing supportive tools and datasets. By clarifying sources of error and uncertainty and providing guidance for future developments, this paper aims to contribute to the enhanced credibility of CFD simulation of reactors in the future.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113974"},"PeriodicalIF":1.9,"publicationDate":"2025-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143601087","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":"Investigation of neutronic performance of some accident-tolerant fuels for small modular advanced high-temperature reactors","authors":"Shlash A. Luhaib ,&nbsp;Nassar Alnassar ,&nbsp;Sultan J. Alsufyani ,&nbsp;Ahmed Salah Khalil ,&nbsp;Mohammed Sallah ,&nbsp;A. Abdelghafar Galahom","doi":"10.1016/j.nucengdes.2025.113997","DOIUrl":"10.1016/j.nucengdes.2025.113997","url":null,"abstract":"<div><div>The purpose of this study is to examine some accident-tolerant nuclear fuels for small modular advanced high-temperature reactors (SmAHTR) instead of traditional fuel. Various fuels including UCO, UAl, U₃Si₂, (Th, U)CO, (Th, U)Al and (Th, U₃)Si₂ have been investigated as fuels in the TRISO particle of the SmAHTR to investigate the potential advantages of using these fuel types. Neutronic parameters including infinity multiplication factor, fissile concentration, fertile concentration, reactor grade plutonium concentration, fission products concentration, flux per lethargy and radial power distribution have been examined using the cross-section library ENDF/V.II.0. Some safety parameters have been studied to verify the viability of the suggested fuel. The radioactivity of actinides and non-actinides has been tracked with fuel burnup to distinguish their level. This gives a good indication of the precautions required for treating the spent fuel. The neutronic analysis showed the effectiveness of the suggested fuels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113997"},"PeriodicalIF":1.9,"publicationDate":"2025-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143601083","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":"Main outcomes of OECD/NEA THAI-2 project on hydrogen risk and source term investigations: Data application for code validation and containment safety assessment","authors":"S. Gupta ,&nbsp;M. Freitag ,&nbsp;Z. Liang ,&nbsp;F. Funke ,&nbsp;G. Langrock ,&nbsp;S. Beck ,&nbsp;H. Nowack ,&nbsp;A. Bentaib ,&nbsp;L. Cantrel ,&nbsp;J. Ishikawa ,&nbsp;S.W. Hong ,&nbsp;P. Kostka ,&nbsp;J. Glover ,&nbsp;C. Linde ,&nbsp;M. Kotouč ,&nbsp;V. Taivassalo","doi":"10.1016/j.nucengdes.2025.113970","DOIUrl":"10.1016/j.nucengdes.2025.113970","url":null,"abstract":"<div><div>During a core-melt accident, apart from hydrogen, radioactive gases and aerosols are released into the containment, the behaviour of which is of significant importance for determining the radiological source term. The investigation of in-containment combustible gases and fission product behaviour was the<!--> <!-->subject of the OECD/NEA THAI-2 project conducted during 2011–2014. The project focused on experiments on hydrogen behaviour i.e., deflagration in the presence of spray and passive autocatalytic recombiners (PARs) performance in O₂-lean atmosphere, on the interaction of molecular iodine with reactive (silver) and non-reactive (tin oxide) aerosol particles to assess the effect on a potential source term, and on the quantification of the release of gaseous iodine from a flashing jet, representing a PWR steam generator tube rupture scenario during reactor shutdown. The project was supported by 11 countries involving safety organizations, regulatory bodies, research laboratories, universities and industries.</div><div>The experimental programme of the OECD/NEA THAI-2 project strongly contributed to the validation and further development of advanced lumped parameter and computational fluid dynamic codes used for reactor applications by e. g. providing experimental data for code benchmark exercises. The present paper summarizes<!--> <!-->the<!--> <!-->key findings of the project and highlights the importance of project results for mitigation of hydrogen risk and source term related issues. Furthermore, the use of project results by the project partners for code validation and reactor analyses towards management and mitigation of a severe accident in light water reactors is discussed with selected examples.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113970"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592560","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":"OpenFOAM-informed TRANSURANUS simulations for assessing the fuel pin performance in the MYRRHA in-pile test section","authors":"M. Di Gennaro,&nbsp;A. Magni,&nbsp;A. Trombetta,&nbsp;D. Pizzocri,&nbsp;S. Lorenzi,&nbsp;L. Luzzi","doi":"10.1016/j.nucengdes.2025.113944","DOIUrl":"10.1016/j.nucengdes.2025.113944","url":null,"abstract":"<div><div>In continuation of previous examinations of the MYRRHA fuel pin performance conducted by <span><span>Magni et al. (2023)</span></span> and <span><span>Luzzi et al. (2024)</span></span>, the present work aims to investigate the impact of the beam power jump (BPJ) transient scenario on the thermo-mechanical response of homogeneous Am-bearing fuel pin (0.49 wt.<span><math><mtext>%</mtext></math></span> - 5 wt.<span><math><mtext>%</mtext></math></span>) irradiated in the in-pile test section (IPS) of the MYRRHA research reactor (core design revision 1.8). The approach pursued in the current work consists of carrying out the pin performance analysis employing the TRANSURANUS fuel performance code informed by more specific and reliable thermo-hydraulic boundary conditions evaluated via the fluid dynamics code OpenFOAM. After illustrating the differences between the conventional approach with TRANSURANUS stand-alone and the approach adopted in this paper, compliance with safety-related limits is discussed with a focus on the transient pin behaviour. The results confirm that Am-MOX fuel options for MYRRHA are suitable and safe, even during the BPJ transient, with wide margins. The adopted approach (namely, complementing the fuel performance code thermo-mechanical analysis with high-fidelity thermo-hydraulic tools) is herein applied to a specific reactor design and irradiation scenario, but it is suitable for the assessment and safety analysis of other/future reactor concepts.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113944"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592559","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":"BORA4-PTS: Experimental reproduction of a Pressurized Thermal Shock, and building of numerical simulation with the CATHARE Code","authors":"Alexis Berny ,&nbsp;Cécile Demathieu ,&nbsp;Joël Nibas ,&nbsp;Alain Ungar","doi":"10.1016/j.nucengdes.2025.113957","DOIUrl":"10.1016/j.nucengdes.2025.113957","url":null,"abstract":"<div><div>When a break occurs in a nuclear reactor, a fast cooldown has to be down to prevent the melting of the core. This is done by the injection of cold water at 7 °C, in a pressurized vessel at 295 °C. This is a Pressurized Thermal Shock. To improve the safety of the nuclear reactor, an experimental facility was built to analyze the mix of hot and cold water in the downcomer of the vessel. This is the BORA4-PTS experiment. Salt water at 45 °C is injected into stagnant pure water at 20 °C, to represent the injection of cold water in hot water. Through this experiment, a scaling was establish to compare it with the reactor case. We then made a numerical simulation of this experimental facility with the CATHARE code. With this simulation come another scaling, in order to properly compare the numerical and the experimental results. The numerical simulations give results that are very similar to the experimental ones. With this experiment, we show the excellent capacity of CATHARE to simulate and model the complex thermohydraulic inside a downcomer.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113957"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143580755","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":"Inverse uncertainty Quantification in the Severe accident Domain: Application to Fission Product release","authors":"G. Tinfena ,&nbsp;M. Angelucci ,&nbsp;L. Sargentini ,&nbsp;S. Paci ,&nbsp;L.E. Herranz","doi":"10.1016/j.nucengdes.2025.113954","DOIUrl":"10.1016/j.nucengdes.2025.113954","url":null,"abstract":"<div><div>This paper presents a pioneering application of Inverse Uncertainty Quantification (IUQ) methodology, proposed by CSNI/WGAMA in the SAPIUM activity, within the Severe Accident (SA) domain. Framed under the SEAKNOT-EU project, such a systematic approach has been used for uncertainty quantification of the Fission Product Release (FPR), particularly the cesium fractional one, as estimated by the MELCOR code. More than 60 experiments have been analyzed and some of them selected based on their completeness and representativeness, as recommended by the adequacy analysis. By using the IUQ CIRCE method, the so called Revised CORSOR-Booth model uncertainty range has been estimated and proved consistent with the database. Finally, the uncertainty characterization has been validated against on-line experimental data recorded in PHEBUS-FPT1.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113954"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143580754","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}