Annals of Nuclear Energy最新文献

{"title":"Development of system analysis code and its application in transient simulation of supercritical CO2 Brayton cycle direct cooled reactors","authors":"Haijie Wu ,&nbsp;Minyun Liu ,&nbsp;Shenghui Liu ,&nbsp;Jiatao Meng ,&nbsp;Shuangqing Chen ,&nbsp;Ruohan Zheng ,&nbsp;Yanping Huang ,&nbsp;Houjun Gong ,&nbsp;Xiaoliang Zhu ,&nbsp;TianYang Xing ,&nbsp;Mudi Jiang","doi":"10.1016/j.anucene.2024.111031","DOIUrl":"10.1016/j.anucene.2024.111031","url":null,"abstract":"<div><div>The supercritical carbon dioxide (s-CO₂) Brayton cycle direct cooled reactor system has the characteristics of high cycle efficiency, compact equipment, and simple layout, which has attracted the attention of researchers recently. The use of s-CO₂ Brayton cycle and the elimination of intermediate heat exchangers make the transient characteristics of this reactor different from light water reactors which have rich operating experience. Transient simulation and safety analysis with system code is an essential stage in reactor design and safety verification. Therefore, a system code, named NUSAS (NUclear Safety Analysis and Simulation), was developed in this study for the transient simulation of s-CO₂ Brayton cycle direct cooled reactor system. This code is developed based on the advanced two-fluid two-pressure seven-equation model and has the capability of modelling the turbomachinery in s-CO₂ Brayton cycle direct cooled reactor system. The transient characteristics during core power step and precooler cooling water mass flow rate step were analyzed. The load tracking and operation control strategy of the reactor was designed, and transient simulations of the wide range (100% to 50%) load tracking process were conducted. According to the simulation results of NUSAS, the transient characteristics of s-CO₂ Brayton cycle direct cooled reactor were demonstrated when operating conditions change.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111031"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146039","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":"Stability and bifurcation analysis of a Sodium-cooled Fast Reactor under two-phase flow using 1D multiphysics model","authors":"R. Abbé ,&nbsp;C. Patricot ,&nbsp;M. Médale","doi":"10.1016/j.anucene.2024.111006","DOIUrl":"10.1016/j.anucene.2024.111006","url":null,"abstract":"<div><div>The Generation IV Sodium-cooled Fast Reactor (SFR) R&amp;D program of the CEA (French Alternative Energies and Atomic Energy Commission), embodied in France by the ASTRID program (600 MWe, 2010–2019), must meet more stringent safety requirements than previous programs. For this reason, the behavior of the reactor in accident phenomenology is studied. We can distinguish three main initiator families that can lead to severe accident conditions in an SFR core: reactivity insertions, local subassembly failures and loss of primary flow. They can evolve in different ways depending on the transient and the reactor characteristics, especially the latter (boiling stabilization, primary power excursion). In order to assess the accidental behavior of innovative SFRs, in particular in case of sodium boiling, dedicated tools have been developed at CEA in recent years. In general, these tools are based on “fine” 3D codes that have highlighted the possibility of stabilized boiling. However, it is difficult to analyze stability with such tools in a dynamic scenario and it is also very complicated to validate complex numerical tools on a dynamic nonlinear phenomenology. In this paper, we propose a methodology that allows to study the stability of SFR nuclear cores under boiling conditions resulting from an unprotected loss-of-coolant accident. This methodology relies on a semi-analytical approach directly applied on a discretized model with possibly thousands of unknowns, rather than on a handful of unknowns derived from a reduced-order model as usually done in the literature when it comes to stability analyses. The nuclear reactor is modeled by a 1D drift thermal-hydraulic model, a 1D neutron diffusion model, and a OD heat transfer model. The semi-analytical methodology developed in this article is based on a continuation method, the Numerical Asymptotic Method, which allows us to compute all the stationary solutions of interest (single-phase liquid, two-phase) with good numerical efficiency (accuracy and computational time). We then perform a linear stability analysis of some of these stationary solutions, which amounts to solving generalized eigenvalue problems with a singular mass matrix. This allows us to study the effect of neutron coupling on the stability of steady-state solutions in the framework of Dynamical Systems Theory through bifurcation diagrams. We can see that neutron coupling extends the stability zone in the studied situation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111006"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146509","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":"Development of a design method for burnable poisons in small modular fast reactors","authors":"Yiwei Wu,&nbsp;Yuyang Shen,&nbsp;Xin Jin,&nbsp;Qufei Song,&nbsp;Yao Xiao,&nbsp;Hui Guo,&nbsp;Hanyang Gu","doi":"10.1016/j.anucene.2024.111008","DOIUrl":"10.1016/j.anucene.2024.111008","url":null,"abstract":"<div><div>Long-cycle small modular fast reactors have high excess reactivity. The use of burnable poison can compensate for excess reactivity and improve core safety. Burnable poison design of fast reactors is a complicated problem, the empirical design method is time-consuming and does not guarantee optimal compensation. In this study, a design method for burnable poison in fast reactors is developed. Firstly, a fast prediction method of compensation ability is formed to optimize poison assembly, the evaluation of optimal concentration in a specific geometry and compensation in different geometries can be advanced. Secondly, a multi-group Monte Carlo depletion method is formed to accelerate calculation, and the optimization of poison arrangement was carried out based on a genetic algorithm. Optimization reduced core burnup reactivity loss from 10,043 pcm to 3841 pcm in 3.5 years, and the residual penalty is 301 pcm. The design method can provide a reference for the subsequent design of burnable poison in a fast reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111008"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146561","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":"Assessment and improvement of numerical model for steam condensation with non-condensable gases on all-curvature surfaces in NPP containments","authors":"Gonglin Li ,&nbsp;Hui Wang ,&nbsp;Shuhang Zhou ,&nbsp;Haozhi Bian ,&nbsp;Ming Ding","doi":"10.1016/j.anucene.2024.111009","DOIUrl":"10.1016/j.anucene.2024.111009","url":null,"abstract":"<div><div>Steam condensation with non-condensable gases is one of the most important phenomena in containment of nuclear power plant (NPP) after accidents. With the application of passive containment cooling system (PCCS), steam is condensed on surfaces with different curvatures of PCCS and structures. Previous investigations indicated the need for suction corrections in the diffusion boundary layer model of steam condensation in the presence of non-condensable gases. By implanting different suction effect corrections in OpenFOAM, this study found that existing suction effect corrections cannot make good predictions for steam condensation with non-condensable gases on varied curvature surfaces, which resulted from cases lack of diversity. Thus, a new correction was proposed based on Gasthof number and previously corrections. This correction was validated by data sets from TOSQAN, COAST, Dehbi and Su’s facilities. The relative deviations of 90.0 % simulation results based on the new correction from the experimental results were within 20.0 %.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111009"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146563","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":"Estimating parameter uncertainty bounds of human error probability using Monte Carlo simulation","authors":"Yochan Kim,&nbsp;Jaewhan Kim,&nbsp;Dong-San Kim","doi":"10.1016/j.anucene.2024.111024","DOIUrl":"10.1016/j.anucene.2024.111024","url":null,"abstract":"<div><div>Characterizing the uncertainty of reliability in risk assessment has been recognized as a critical element in effective decision-making. Although many kinds of risk assessment guidelines have introduced various uncertainty analysis techniques, estimates of the uncertainty bounds of human reliability have not been investigated on an empirical basis. In this study, we predicted how parameter uncertainties in human error probabilities can be formed by employing uncertainty intervals in the failure probabilities of primitive tasks, the multipliers of performance-shaping factors, and recovery failure probabilities. The above component values were incorporated through Monte Carlo simulation and the EMBRACE (Empirical Data-Based Crew Reliability Assessment and Cognitive Error Analysis) method. As a result, distributions of human error probabilities in various contextual situations were derived. The statistical model most suitable for the distributions was then selected, and the bounds of 90 % uncertainty intervals were estimated according to the selected statistical model. Implications and limitations of this study are also discussed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111024"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146585","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":"Evaluating uncertainties: Heat transfer parameter effects on stratified melt pool simulation","authors":"Pengya Guo ,&nbsp;Fengyang Quan ,&nbsp;Peng Yu ,&nbsp;Yidan Yuan ,&nbsp;Jiyang Yu ,&nbsp;Weimin Ma","doi":"10.1016/j.anucene.2024.110970","DOIUrl":"10.1016/j.anucene.2024.110970","url":null,"abstract":"<div><div>Following Fukushima Daiichi, nuclear safety is paramount in advanced pressurized water reactors. In-Vessel Retention (IVR), notably External Reactor Vessel Cooling (ERVC), offers simplicity and cost-effectiveness. However, uncertainties in the corium thermal load and IVR processes mandate conservative design and safety margins. This study simulates the SAMPO experiment with nitrate salts and thermal oil to investigate the thermal hydraulics of a stratified melt pool. Analysis of power levels, heat transfer coefficients, and radiation heat transfer reveals key insights. Increasing input power raises the temperature and sidewall heat flux in the upper layer, leading to interlayer crust dissolution and enhanced upward heat transfer near the lower layer’s melting point. Higher convection coefficients double the heat flux at the metal layer’s top while reducing sidewall heat flux. Adjusting radiation emissivity of the front and back plates to 0.7 synchronously decreases heat flux from both the top and curved sidewalls, achieving an effect similar to a 50% power reduction without radiation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110970"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146587","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":"Radioactive study of oxidation-corrosion materials activated by neutron in lead-bismuth eutectic reactors","authors":"Jiajie Chen ,&nbsp;Shiwei Wang ,&nbsp;Xiaojing Liu ,&nbsp;Tengfei Zhang ,&nbsp;Xiang Chai ,&nbsp;Hui He","doi":"10.1016/j.anucene.2024.111037","DOIUrl":"10.1016/j.anucene.2024.111037","url":null,"abstract":"<div><div>Lead-bismuth-cooled fast reactors, one of the promising Generation IV reactors, hold significant advantages in providing clean energy. However, the use of liquid lead-bismuth eutectic (LBE) as a coolant poses compatibility issues with structural materials like T91 cladding and heat exchange tubes, leading to the introduction of corrosion products into the LBE. Oxygen control techniques alleviate these issues by forming protective oxide layers on the cladding surfaces. During reactor operation, corrosion products in the LBE become activated under neutron irradiation, producing radionuclides. Evaluating corrosion products in the LBE requires considering the oxide layer status on cladding materials. Limited research has assessed the sources of corrosion products and their impact on LBE radioactivity. This study combines the oxide growth kinetic model of T91 alloy with an oxide layer corrosion model to evaluate the sources of corrosion products. Neutron activation theory is used to analyze the activation of corrosion products. The research findings indicate a close correlation between additional radioactivity and the degree of oxide layer corrosion. Corrosion of the spinel layer may introduce radioactivity levels of up to 10⁹Bq/cm³, primarily from ⁵¹Cr elements. Compared to scenarios with magnetite layer corrosion, spinel layer corrosion increases radioactivity in the LBE by 3–14 %. This study provides insights into the additional levels of radioactivity that T91 alloy may introduce during the operation of lead-bismuth reactors. It also offers recommendations for the oxygen control conditions in such reactor systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111037"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146675","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 realistic TRACE PWR LOCA model with application to high-burnup analysis","authors":"Aaron Wysocki,&nbsp;Nathan Capps","doi":"10.1016/j.anucene.2024.111014","DOIUrl":"10.1016/j.anucene.2024.111014","url":null,"abstract":"<div><div>This paper presents a TRACE model of a postulated large-break loss-of-coolant accident (LBLOCA) event in a four-loop pressurized water reactor (PWR). The input files have been made publicly available and serve as a starting point for researchers to improve upon or apply to their purposes. The model, based on a legacy PWR model, contains numerous improvements and modifications consistent with US Nuclear Regulatory Commission LBLOCA analysis guidelines. The model is applied to analyze high-burnup (&gt;62 GWD/MTU) fuel behavior during LBLOCA, improving on previous work to provide realistic thermal hydraulic predictions for future fuel performance analyses and experiments related to fission fragment, relocation, and dispersal (FFRD). Sensitivity studies are presented to quantify the impact of the vessel-modeling approach, core radial discretization, and other phenomena. The model is intended for research purposes and contains sufficient plant geometric, operational, and safety system details to represent the main physical phenomena pertaining to LBLOCA.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111014"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146041","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":"Hermite methods for multi-energy group neutron kinetics model via quasi-static approach and piecewise functions","authors":"Ahmed E. Aboanber ,&nbsp;Abdallah A. Nahla ,&nbsp;Hager M. Atalla","doi":"10.1016/j.anucene.2024.110995","DOIUrl":"10.1016/j.anucene.2024.110995","url":null,"abstract":"<div><div>The prominence of nuclear energy has grown due to its benefits over conventional power. As a result, using an inventive mathematical model is critical to ensuring nuclear reactor security. One of the most significant reactor kinetics models proposed is the neutron kinetics equations (NKEs) with a multi-group of delayed neutrons (MGDN) that use multi-energy group (MEG). The primary purpose of this research is to establish a mathematical methodology for solving the coupled stiff model of MEG NKEs with MGDN, which produces more realistic and meaningful simulations of multidimensional homogeneous reactors while reducing CPU time. The suggested methods utilize piecewise linear, cubic, and quintic polynomials together with a quasi-static approach and the Hermite method. These modifications significantly reduce computational time and accelerate the proposed techniques beyond the established benchmarks. For two and three-dimensional space homogeneous reactors, the results of Hermite methods are tested and compared with conventional and benchmark methodologies for various forms of reactivity, such as step, ramp, and nonlinear reactivity. The results produced by these Hermite methods are comparable to benchmark techniques, and they handle stiffness in the easiest way to implement.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110995"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146562","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":"Power distribution bias from equilibrium xenon effects when insufficient neutron histories per cycle are used in Monte Carlo simulation of CANDU6","authors":"Yeseul Seo,&nbsp;Arief Rahman Hakim,&nbsp;Douglas A. Fynan","doi":"10.1016/j.anucene.2024.111027","DOIUrl":"10.1016/j.anucene.2024.111027","url":null,"abstract":"<div><div>This study examines the impact of insufficient neutron histories per cycle (<em>N</em>) on power distribution accuracy in Monte Carlo (MC) simulations of CANDU reactors when using equilibrium-xenon algorithms. We demonstrate that using an insufficient <em>N</em> with equilibrium xenon introduces significant bias into the power distribution. The observed bias is systematic with a strong flattening affect where power is suppressed in high-power bundles and increased in low-power bundles by up to several percent beyond the real physical flattening effect of xenon on the three-dimensional power distribution. At least <span><math><mrow><mn>5</mn><mo>×</mo><msup><mrow><mn>10</mn></mrow><mn>6</mn></msup></mrow></math></span> to <span><math><msup><mn>10</mn><mn>7</mn></msup></math></span> <em>N</em> must be used for criticality simulation with the 1/8th symmetric CANDU6 model studied to suppress the spatial power tilts and eliminate the bias. This recommended <em>N</em> is orders of magnitude larger than <em>N</em> used in previous full-core Monte Carlo simulations of CANDU reactors. A preliminary analysis of the CANDU bundle power tally variances shows there is an underestimation of the real variance by approximately a factor of two.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111027"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143146589","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}