Annals of Nuclear Energy最新文献

{"title":"Treatment of molten fluoride salt-exposed graphite: Cleaning made simple","authors":"Y. Huang ,&nbsp;G. Zheng ,&nbsp;D.J. Sprouster ,&nbsp;L.L. Snead","doi":"10.1016/j.anucene.2025.111386","DOIUrl":"10.1016/j.anucene.2025.111386","url":null,"abstract":"<div><div>To ensure chemical stability in storage and transportation containers, graphite waste from Fluoride-cooled High-temperature Reactors (FHRs) requires treatment to remove residual fluorides and beryllium before disposal. This study demonstrates a simple, effective cleaning method for modern fine-grained nuclear graphite exposed to molten FLiBe salt under neutron irradiation. A straightforward soaking in deionized water successfully removed near-surface salt, confirmed through cross-sectional scanning electron microscopy, energy-dispersive spectroscopy mapping, and X-ray diffraction. Additional in situ electrochemical monitoring and ex situ fluorometer assays validated this approach. Post-cleaning analyses confirm the process effectively decontaminates graphite surfaces exposed to FLiBe during irradiation, offering a practical solution for FHR graphite waste management.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111386"},"PeriodicalIF":1.9,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706360","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":"Impact of delayed neutrons on bifurcation and stochastic self-excited oscillations in nuclear reactors with discrete control systems","authors":"Amel Raked ,&nbsp;Abdeslam Seghour ,&nbsp;Azzedine Chafa","doi":"10.1016/j.anucene.2025.111372","DOIUrl":"10.1016/j.anucene.2025.111372","url":null,"abstract":"<div><div>The present study investigates the influence of delayed neutrons on stability boundaries, chaos, and bifurcation phenomena in nuclear reactors with discrete control systems. It focuses on a point kinetics model of a reactor with a regulatory system that includes a single group of delayed neutrons and temperature feedback related to either the moderator or fuel temperature, as well as power feedback. Bifurcation diagrams are used to examine the transition to chaos, an area previously underexplored. Our research includes the development of an analytical model and detailed numerical analysis to identify stability regions and chaotic dynamics in reactor models incorporating delayed neutrons. Furthermore, we introduce innovative one-dimensional piecewise linear maps that exhibit diverse behaviors, including periodic motion, n-band chaotic attractors, exterior crisis phenomena, type I intermittency, and periods of regular motion interrupted by short quasiperiodic motion and chaotic bursts. Comparisons are drawn between models with and without delayed neutrons.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111372"},"PeriodicalIF":1.9,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706475","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":"Modeling of neutron fluctuations with temperature feedback using stochastic differential equations","authors":"B. Dechenaux","doi":"10.1016/j.anucene.2025.111375","DOIUrl":"10.1016/j.anucene.2025.111375","url":null,"abstract":"<div><div>Neutron fluctuations, also coined zero power noise, is concerned with the study of the random fluctuations observed in the neutron population evolving in a fissile medium, due to the fundamentally intrinsic stochastic nature of the neutrons’ interactions in matter.</div><div>The canonical theoretical modeling of such fluctuations rely on the establishment of a microscopic probability balance equation, from which one can extract, with the highest precision, all of the information about the stochastic behavior of the neutron population.</div><div>An approximate yet practical scheme has been gaining in popularity in recent years. It relies on a coarse grained approach of the problem, where the evolution of the neutron population is modeled through (Itô prepoint) Stochastic Differential Equations (SDE). In this simplified scheme, the crux is to find the SDE that best approximates the fluctuations of the underlying microscopic stochastic process.</div><div>In the present work, an alternative approach to the scheme usually found in the neutronics literature is proposed. A family of SDE that best approximate the fine grain behavior of the neutron population is derived from an explicit truncation of the translation operators found in the microscopic master equation of the problem. The approach is developed over a concrete and relatively original application case, namely that of the point reactor model including mono-energetic neutrons, one group of precursors, and a temperature feedback process.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111375"},"PeriodicalIF":1.9,"publicationDate":"2025-03-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143697192","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":"Research on variable speed operation scheme for mobile microreactor coupled with helium gas turbine cycle","authors":"Xuyao Geng,&nbsp;Jie Wang","doi":"10.1016/j.anucene.2025.111389","DOIUrl":"10.1016/j.anucene.2025.111389","url":null,"abstract":"<div><div>The micro high temperature gas-cooled reactor coupled with gas turbine direct cycle has the characteristics of inherent safety, high outlet temperature and no phase change of working medium. It is a competitive mobile microreactor technology scheme. Mobile microreactors mainly operate in remote off-grid areas, and need to cope with frequent load power variations. They mainly operate in variable speed mode. While, there are few studies on the variable speed operation characteristics of mobile microreactors. In this paper, a 10MWth mobile microreactor coupled with helium gas turbine cycle (MMR-GT) is selected as research object, and the differential algebraic equations mathematical model is established. The system analysis program is developed based on Modelica language. Firstly, the characteristics of MMR-GT under constant speed and variable speed operation schemes are compared when 100pcm reactivity is inserted and the load power varies the same. In variable speed operation, the performance parameters of turbomachinery changes more obviously than that in constant speed operation. Then, the steady-state operation characteristics of MMR-GT at different rotational speed is analyzed. With the decrease of speed, the power of each component, mass flow rate and system pressure all drop rapidly. The recuperator effectiveness is higher at low flow rate, and it is increased by 15.5% when the rotational speed is reduced to 60% of the rated value. Finally, the response of MMR-GT to load power fluctuations is analyzed. Without inserting external reactivity, the decrease of load power leads to the increase of rotational speed and flow rate, the decrease of reactor outlet temperature, and the rise in reactor power due to negative feedback effect. For load power increase, the positive reactivity needs to be inserted to increase driving torque. The speed control scheme is tested for partial load power loss. It indicates that the shaft speed can be controlled at the set value through bypass valve.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111389"},"PeriodicalIF":1.9,"publicationDate":"2025-03-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143697084","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":"Effect of swirling strength on droplets impact on concave wall: Experimental and numerical investigation","authors":"Shuo OUYANG ,&nbsp;Zhenqin Xiong ,&nbsp;Hua Bei ,&nbsp;Kaicheng Yu ,&nbsp;Lichen Xue ,&nbsp;Jiyun Zhao","doi":"10.1016/j.anucene.2025.111282","DOIUrl":"10.1016/j.anucene.2025.111282","url":null,"abstract":"<div><div>Droplets impact on walls in a rotating gas–liquid flow field is common in industrial applications. To reveal the effect of swirling strength on droplet impact outcome, droplet trajectory and impingements on the concave wall with three kinds of swirlers are investigated by numerical simulation and experiment. Three impact outcomes are observed: downward slide spreading, horizontal slide spreading, and upward slide spreading. It is found that the swirling strength affects the impact outcome and the spreading factors. The transition from the upward slide spreading to other impact outcomes is promoted by increasing the swirling strength. Moreover, a theoretical model is deduced for the maximum transverse spreading factor of the droplets impact on the concave wall in a rotating gas flow field by modifying the surface energy and adding the auxiliary dissipation. The discrepancies between the predicted maximum transverse spreading factors and the experimental results are in the range of + 5 %∼-10 %.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111282"},"PeriodicalIF":1.9,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143685348","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":"Preliminary criticality analysis of a partially damaged reactor core under different scenarios","authors":"Hoang Hai Nguyen","doi":"10.1016/j.anucene.2025.111361","DOIUrl":"10.1016/j.anucene.2025.111361","url":null,"abstract":"<div><div>This study examined the criticality characteristics of a partially damaged reactor model, in which fuels located at the core center melt into fuel debris of varying shapes, while fuels situated at the core edge remain intact. The investigation was conducted using the Serpent code with the JENDL-5 library. The results indicate that when the volume of fuel debris is constant, the shape of the fuel debris does not cause significant alterations in the variation of the k_eff of the system. In contrast, for the scenario of the variable fuel-debris volume, the k_eff variation law can be divided into two groups for the reference case with a system temperature of 300 K and no boron in the water. The first group comprises fuel debris with cuboid and cylindrical shapes, while the second group includes spherical, conical, and truncated conical shapes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111361"},"PeriodicalIF":1.9,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143685356","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":"Study on effect of impact parameter on droplets successively impacting on liquid film by CLSVOF method","authors":"Bowen Chen ,&nbsp;Yandong Hou ,&nbsp;Chuntian Gao ,&nbsp;Weichao Li ,&nbsp;Yuyang Wang ,&nbsp;Ruifeng Tian","doi":"10.1016/j.anucene.2025.111359","DOIUrl":"10.1016/j.anucene.2025.111359","url":null,"abstract":"<div><div>In the corrugated plate dryer, there are a lot of phenomena of multi-droplets impacting on the liquid film. Therefore, the analysis of droplets successively impacting on liquid film has important engineering significance. In this paper, the CLSVOF method is used to simulate the successive impact of two droplets on the liquid film. By analyzing the velocity and pressure distribution of the liquid film and the motion state of the residual film, the influence of the droplets spacing, the initial liquid film thickness and the impact velocity on the impact outcomes is obtained, and the influence mechanism is analyzed. The simulation results show that the inner crown is easy to splash with the increase of the droplets spacing; with the increase of impact velocity, the inner crown is easy to splash；with the increase of the initial liquid film thickness, the inner crown is not easy to splash.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111359"},"PeriodicalIF":1.9,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143685358","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 hybrid optimized Picard method for neutronic/thermal-hydraulic coupling problem","authors":"Xinru Peng,&nbsp;Han Zhang,&nbsp;Lixun Liu,&nbsp;Qinrong Dou,&nbsp;Yingjie Wu,&nbsp;Jiong Guo,&nbsp;Fu Li","doi":"10.1016/j.anucene.2025.111380","DOIUrl":"10.1016/j.anucene.2025.111380","url":null,"abstract":"<div><div>Picard iteration is a widely-used method for Neutronic/Thermal-Hydraulic (N/TH) coupling calculation in nuclear reactor, but it usually suffers from computational efficiency and stability, especially for complicated large-scale coupling issues. Residual balance method (RB) and Anderson acceleration method (AA) are two well-known optimized variants, where the former focusing on reducing the computational cost per outer iteration by adaptive tolerance, and the latter on decreasing the number of outer nonlinear iterations by using multiple iteration history information. However, combining these two methods causes incompatibility because the inaccurate function evaluations from the adaptive tolerance in the inner iterations disrupt the convergence rate of Anderson acceleration. In this work, a new hybrid optimized Picard iteration method (Picard-H) is proposed to combine the advantages of residual balance method and Anderson acceleration method by using an adaptive switching factor, avoiding the incompatibility from the inner iteration when approaching convergence. This new method could further enhance the efficiency of Anderson Acceleration, which could finish calculation much faster than both RB and Picard. A simplified neutronic k-eigenvalue problem in PWR with thermal–hydraulic feedback effect is utilized to evaluate the performance of the new optimized Picard method. The results demonstrate that for the simplified PWR model, the Picard-H method is 10% more computationally efficient than the AA method, more than twice as fast as the RB method, and more than four times faster than the traditional Picard method. In addition, the simple combination of RB and AA methods diverges due to the inexact fixed-point mapping function evaluation from the RB method. Furthermore, the hybrid method is robust against variations in parameters, such as the bounding parameter in the residual balance method and the memory depth in Anderson acceleration. The computational performance of the new method also is not sensitive to the choice of switching factor which is suggested to be adaptive when the residual drops to about 2 to 3 orders of magnitude of the initial residuals. The supplementary numerical results from the simplified HTR-10 model confirm that the Picard-H method exhibits 5% greater computational efficiency than the AA method and is over nine times faster than the traditional Picard method and the RB method.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111380"},"PeriodicalIF":1.9,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143685357","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 validation of neutron transport in chlorine in LR-0 reference field","authors":"Tomáš Peltan ,&nbsp;Tomáš Czakoj ,&nbsp;Vlastimil Juříček ,&nbsp;Evžen Losa ,&nbsp;Jan Šimon ,&nbsp;Michal Košťál","doi":"10.1016/j.anucene.2025.111356","DOIUrl":"10.1016/j.anucene.2025.111356","url":null,"abstract":"<div><div>Nuclear energy is one of the promising ways of reaching carbon neutrality, but this goal requires the development of new reactor types. One such concept is the molten salt reactor system, which uses chlorine-based fuel salts. Developing this system needs reliable neutronic data for all materials used, with particular emphasis on core components. While uranium cross-sections remain well-studied, sufficient experimental data for chlorine is lacking, particularly for validations of current data evaluations. This paper presents new, precise integral experiments to validate various chlorine cross-section evaluations. Notably, compared with experimental results, the currently available data sets exhibit significant discrepancies, suggesting that simulations based on current nuclear data evaluations may be unreliable.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111356"},"PeriodicalIF":1.9,"publicationDate":"2025-03-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143685350","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":"Seaknot: Looking ahead of severe accident research","authors":"L.E. Herranz ,&nbsp;S. Gupta ,&nbsp;S. Paci ,&nbsp;P. Piluso","doi":"10.1016/j.anucene.2025.111390","DOIUrl":"10.1016/j.anucene.2025.111390","url":null,"abstract":"<div><div>Severe Accidents (SAs) dominate the risk associated to the commercial production of nuclear energy. Despite the major achievements made in their research, still existing gaps, upcoming new technologies as Accident Tolerant Fuels (ATFs) and Small Modular Reactors (SMRs), more stringent safety requirements, optimization of SA management, and other factors, point the need for an efficient use of research resources in the years to come. Three major elements should integrate any SA roadmap to be proposed: preservation of knowledge and know-how; identification of key issues which research would result in the best accident management (AM) feasible; and, no less important, strengthening the workforce who will be responsible for such research. The SEAKNOT project (SEvere Accident research and KNOwledge managemenT for LWRs) was born to address this need in all and every aspect. The present article outlines the major pillars of SEAKNOT and synthesizes the progress made since its onset at the end of 2022. The methodologies adopted to develop a SA PIRT (Phenomena Identification and Ranking Table) and to build a Validation Database Directory (VADD) are described along with the ongoing phenomena listing and ranking. Besides, the first steps towards an experimental infrastructure capable of dealing with present and future needs (SAINET) are included. No less relevant the actions already made and the novelties coming on the side of knowledge and know-how transfer are also discussed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111390"},"PeriodicalIF":1.9,"publicationDate":"2025-03-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143685349","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}