Annals of Nuclear Energy最新文献

{"title":"Experimental study on the release of fission gas from defective cladding of PWR based on steady-state escape rate coefficient","authors":"Huaxiang Chen ,&nbsp;Bing Dong ,&nbsp;Junlian Yin ,&nbsp;Dezhong Wang ,&nbsp;Yuchen Song","doi":"10.1016/j.anucene.2025.111758","DOIUrl":"10.1016/j.anucene.2025.111758","url":null,"abstract":"<div><div>This study experimentally investigates the release behavior of fission gas through cladding defects in a pressurized water reactor, utilizing non-radioactive krypton as a surrogate for radioactive fission products. The results demonstrate a transient-to-steady evolution of gas release rates, characterized by an initial rapid surge followed by gradual decay to equilibrium. Minor defects tend to exhibit extended transient phases as a result of limited gas flux, whereas larger defects are characterized by elevated peak release rates along with abbreviated transient phases. Elevated temperature and pressure enhance peak release rates through intensified phase transitions. This process is modeled through a first-order kinetic equation incorporating the steady-state escape coefficient (<span><math><mrow><mi>ε</mi></mrow></math></span>). This coefficient is estimated via the Chapman-Enskog diffusion equation. Both experimental and theoretical results indicate that <span><math><mrow><mi>ε</mi></mrow></math></span> is in the order of 10⁻⁹ to 10⁻⁸ s⁻¹ for micrometer defects.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111758"},"PeriodicalIF":2.3,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738427","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":"Refined numerical analysis of prototype steam generator for sodium-cooled fast reactor","authors":"Bo Wang,&nbsp;Yan Zhang,&nbsp;Yang Yang,&nbsp;Zhengrong Guo,&nbsp;Jun Fan,&nbsp;Mingyang Li","doi":"10.1016/j.anucene.2025.111776","DOIUrl":"10.1016/j.anucene.2025.111776","url":null,"abstract":"<div><div>This paper presents a comprehensive three-dimensional numerical simulation of a prototype steam generator for a sodium-cooled fast reactor. The model features a 1:1 geometric representation of the sodium side with detailed meshing, while the water side utilizes a User-Defined Function for accurate thermal modeling. The simulation integrates the coupled flow and heat transfer processes on both sides of the sodium-water steam generator. Validation against experimental data confirms the model’s reliability, showing good agreement in sodium temperature and heat load distribution under both rated and 28% power conditions. The study reveals detailed flow and heat transfer characteristics, including sodium-side flow fields, pressure fields, and temperature fields, as well as water-side fluid and wall temperatures and heat transfer coefficients. Results indicate a relatively uniform heat load distribution in most areas of the tube bundle. This work provides a robust theoretical foundation and effective technical support for the design optimization, performance improvement, and safety assessment of sodium-cooled fast reactor steam generators.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111776"},"PeriodicalIF":2.3,"publicationDate":"2025-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144724721","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":"Oscillatory heat transfer analysis of assisting and opposing radiating flow of Williamson-nanofluid along heat-exchanger plate in nuclear-power reactors: numerical simulation","authors":"Zia Ullah ,&nbsp;Mohammed Alkinidri","doi":"10.1016/j.anucene.2025.111777","DOIUrl":"10.1016/j.anucene.2025.111777","url":null,"abstract":"&lt;div&gt;&lt;div&gt;Exothermic catalytic chemical reaction and thermal solar energy aspects for steady and fluctuating heat and mass flow rate using Williamson nanofluid is very important in nuclear power reactors. Main purpose of this work is to analyze viscous dissipation effects for fluid temperature field, concentration reactants, stead and oscillatory heat-mass transport, streamlines and isotherms along heat exchanger plate of nuclear power reactor using opposing and assisting flow. Significance of this study is to display oscillations and periodical heat transfer and skin friction using opposing and assisting flow. Mathematical formulation is developed for the computational assessment of laminar stream rate and fluctuating heat-mass transmission by using amplitude and phase change analysis. For programming algorithm and numerical results, the mathematical model is solved by dimensionless variables, Stokes oscillating factors, primitive transformations, implicit finite-difference methodology, and Gaussian elimination computation. For numerical results, the following range of parameters exothermic reaction rate &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mi&gt;r&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;2.0&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Eckert number &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;E&lt;/mi&gt;&lt;mi&gt;c&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;2.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, radiation &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;d&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;6.0&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Prandtl 0.1 &lt;span&gt;&lt;math&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;/math&gt;&lt;/span&gt; Pr &lt;span&gt;&lt;math&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;/math&gt;&lt;/span&gt; 10.0, local-modified buoyancy forces &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;λ&lt;/mi&gt;&lt;mi&gt;T&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;λ&lt;/mi&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;4.0&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Schmidt number &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;c&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;1.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, thermophoresis &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;N&lt;/mi&gt;&lt;mi&gt;T&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;0.8&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Brownian motion &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;N&lt;/mi&gt;&lt;mi&gt;B&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;0.6&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Weissenberg number &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mi&gt;W&lt;/mi&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;2.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, and activation energy &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;E&lt;/mi&gt;&lt;mi&gt;A&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;1.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; are considered. First steady flow rate is calculated and used in fluctuating formula to examine oscillatory flow rate. The percentage rate (28 %) of heat transfer is increased at small choice of Eckert number&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;E&lt;/mi&gt;&lt;mi&gt;c&lt;/mi&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; = 0.1. The maximum percentage (16 %) of mass rate is observed at high Eckert number&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;E&lt;/mi&gt;&lt;mi&gt;c&lt;/mi&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; = 2.0. It is observed that streamlines contour increases as radiation-energy increases and exothermic/catalytic reaction flow decreases. It is deduc","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111777"},"PeriodicalIF":2.3,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144722642","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":"Improved XGPT method for uncertainty quantification of reactor physics analysis with continuous-energy covariance data","authors":"Jinlong Huang,&nbsp;Liangzhi Cao,&nbsp;Qingming He,&nbsp;Mohan Wang,&nbsp;Chenghui Wan,&nbsp;Hongchun Wu","doi":"10.1016/j.anucene.2025.111759","DOIUrl":"10.1016/j.anucene.2025.111759","url":null,"abstract":"<div><div>Continuous-energy (CE) cross section and covariance data were employed to carry out the uncertainty quantification (UQ) of reactor physics analysis, ensuring that the entire UQ process is conducted based on CE nuclear data, free from multi-group approximation. In traditional “Sandwich Rules” for UQ, when tallying group-averaged sensitivity coefficients within fine energy intervals, significant statistical fluctuations often occur. To resolve this issue, the eXtendend Generalized Perturbation Theory (XGPT) method was proposed to transform the tally of sensitivity coefficients from fine energy intervals into projections onto basic vectors and the Singular Value Decomposition (SVD) method was utilized for matrix decomposition of the CE covariance matrix. The sensitivity coefficient projections can be tallied using the Iterated Fission Probability (IFP) method implemented in the Monte Carlo code. Nevertheless, in certain cases, a vast number of sensitivity coefficient projection tallies are required. This substantially increases the memory consumption and computational time for sensitivity calculations based on the IFP method. To overcome this challenge, a novel approach, the Sensitivity coefficient Weighted-XGPT (SW-XGPT) method, was proposed to reduce the number of sensitivity coefficient projection tallies during UQ. The performance of the SW-XGPT method, in aspects of accuracy, computational time and memory cost, was evaluated through the Jezebel and ZPPR9 benchmark problems. The numerical results demonstrate that the SW-XGPT method can effectively reduce the computational time and memory cost of sensitivity calculations while maintaining a high level of precision in the UQ results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111759"},"PeriodicalIF":1.9,"publicationDate":"2025-07-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144714357","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":"Neutron total and capture cross-section measurement and resolved resonance analysis of Er","authors":"Gerard Rovira ,&nbsp;Atsushi Kimura ,&nbsp;Shoji Nakamura ,&nbsp;Shunsuke Endo ,&nbsp;Osamu Iwamoto ,&nbsp;Nobuyuki Iwamoto ,&nbsp;Tatsuya Katabuchi","doi":"10.1016/j.anucene.2025.111688","DOIUrl":"10.1016/j.anucene.2025.111688","url":null,"abstract":"<div><div>The neutron total and capture cross sections of <span><math><msup><mrow></mrow><mrow><mi>n</mi><mi>a</mi><mi>t</mi></mrow></msup></math></span>Er were measured at the Accurate Neutron-Nucleus Reaction Measurement Instrument beamline in the Materials and Life Science Experimental Facility of the Japan Proton Accelerator Research Complex. For the neutron total cross-section measurements, Li-glass detectors were employed to derive the neutron transmission ratio of three <span><math><msup><mrow></mrow><mrow><mi>n</mi><mi>a</mi><mi>t</mi></mrow></msup></math></span>Er transmission samples with thicknesses of 1 mm, 0.175 mm and 0.05 mm. The neutron capture cross-section was determined from measurements of 0.025-mm-thick <span><math><msup><mrow></mrow><mrow><mi>n</mi><mi>a</mi><mi>t</mi></mrow></msup></math></span>Er samples with NaI(Tl) and Ge spectrometers in separate experiments, since they offer complementary capabilities. Moreover, the capture cross-section of ¹⁶⁷Er was derived below 1 eV, since it accounted for over 95% of the capture events in <span><math><msup><mrow></mrow><mrow><mi>n</mi><mi>a</mi><mi>t</mi></mrow></msup></math></span>Er. For the neutron total cross-section, at the thermal energy, the present results align with the JENDL-5, ENDF/B-VIII.0 and JEFF-3.3 evaluated nuclear data, the past experimental results of Danon <em>et al.</em> and the recommended values of Mughabghab. For the capture cross-section, similar consistencies were found for <span><math><msup><mrow></mrow><mrow><mi>n</mi><mi>a</mi><mi>t</mi></mrow></msup></math></span>Er and ¹⁶⁷Er, with the present results agreeing within uncertainties with the JENDL-5, ENDF/B-VIII.0 and JEFF-3.3 recommended values, the work of Danon <em>et al.</em> and the evaluation of Mughabghab. In addition, the resolved resonance region up to 65 eV was analyzed using the REFIT code. The resonance parameters of 20 ¹⁶⁷Er resonances, together with one resonance for each ¹⁶⁴Er and ¹⁶⁶Er were determined in a least-square fit using a combination of both capture data and the transmission data from the thicker 1 mm and 0.175 mm samples. The fitting results were found to be highly consistent in-between the four measured data sets, providing convincing experimental data to improve the current knowledge of the resonance region for the Er isotopes in nuclear data libraries, which diverged from the present experimental results in terms of both resonance energy and resonance kernels for most of the resonances above 20 eV.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111688"},"PeriodicalIF":1.9,"publicationDate":"2025-07-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144713854","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":"Parameter study of the running-in process for the Generic Pebble Bed Reactor (GPBR-200)","authors":"Ethan Fowler ,&nbsp;Ryan Stewart ,&nbsp;Paolo Balestra ,&nbsp;Benjamin Lindley ,&nbsp;Gerhard Strydom","doi":"10.1016/j.anucene.2025.111719","DOIUrl":"10.1016/j.anucene.2025.111719","url":null,"abstract":"<div><div>The run-in period of a pebble-bed reactor is complex and difficult to model given significant heterogeneity in core composition, power, and temperature. While the initial core composition should eventually result in the same equilibrium core composition, the approach to equilibrium can vary significantly depending on factors such as startup fuel enrichment and power ramp rate. To explore this, we used a high-fidelity model of the General Pebble Bed Reactor to vary power ramp schemes and startup core compositions. Both the initial core composition and power ramp rate had a significant impact on the flow rate of pebbles during early time steps, with higher ramp rates and low enrichment resulting in non-physical flow rates. Power ramp rate alone was found to dictate the maximum pebble power peaking observed during the run-in process, with higher ramp rates resulting in greater peak pebble powers. Startup fuel enrichment and power ramp were found to both impact total fuel consumption, although the impact of startup fuel enrichment was generally secondary to ramp rate.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111719"},"PeriodicalIF":1.9,"publicationDate":"2025-07-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144714356","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":"Experimental study on fluctuating pressure characteristics of ellipsoidal flow distribution device","authors":"Tiancai Tan,&nbsp;Hongwei Qiao,&nbsp;Jianzhong Ma,&nbsp;Litao Liu,&nbsp;Xin Long","doi":"10.1016/j.anucene.2025.111766","DOIUrl":"10.1016/j.anucene.2025.111766","url":null,"abstract":"<div><div>One of the key points in the fatigue vibration analysis of thin-walled structures with perforated plate is to obtain the dynamic load distribution on the surface of the perforated plate, which is generated by the fluid flow. In this paper, based on a scaled reactor test model, the characteristics of fluctuating pressure on the surface of ellipsoid flow distribution device under different Reynolds numbers were studied experimentally. The fluctuating pressure on the upstream face and the downstream face of the flow distribution device was measured. The load characteristics at different Reynolds numbers were analyzed from two parameters of the root mean square value and the power spectral density of the fluctuating pressure. The research shows that the fluctuating pressure on the surface of the flow distribution device is proportional to the Reynolds number of the perforated plate. At the same position, the root mean square value and power spectral density on the downstream face are larger than those on the upstream face. At the middle of flow distribution device, the power spectral density on the downstream face is one order of magnitude larger than that on the upstream face. Our previous work, that the flow field inside the reactor has been simulated by CFD, shows that the turbulent kinetic energy in the downstream region is obviously larger than that in the upstream region, which explains the reason of these complex flow phenomena in this paper.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111766"},"PeriodicalIF":1.9,"publicationDate":"2025-07-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144704482","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":"Machine learning applications in nuclear power plant piping inspection: A review of methods, data, and future trends","authors":"Afzal Ahmed Soomro ,&nbsp;Osman K. Siddiqui ,&nbsp;Afaque Shams ,&nbsp;Belal Almomani","doi":"10.1016/j.anucene.2025.111760","DOIUrl":"10.1016/j.anucene.2025.111760","url":null,"abstract":"<div><div>Piping safety is an important concern in nuclear power plants, where an inaccurate prediction of pipe failure will lead to financial loss and fatal incidents. Traditional inspection methods are too laborious and costly, and require expert knowledge. Machine learning (ML) models can learn from the available data and be used for nuclear piping safety. A comprehensive literature search reveals that no existing review has focused exclusively on piping safety using ML in nuclear power plants (NPPs). Keeping this in view, efforts have been made to provide a comprehensive review of this topic. The review reveals the main research gaps by investigating the current studies conducted for the safety of NPP piping systems subjected to corrosion using ML. Various ML models and datasets have been reviewed in this review article. It was observed that convolutional neural network, support vector machine, and artificial neural network are the most widely developed models. Vibration-based datasets have been extensively utilized in ML applications for analyzing pipe degradation due to their effectiveness in capturing structural health changes and predicting failures, pipe degradation severity, pipe wall thinning rate, flow accelerated corrosion rate, and pipe elbow thinning rate. Despite the successful implementation of the ML models on available data, major limitations include data scarcity (due to cost and labor required to run the inspections or simulations) and limited scenarios (restricted operating conditions and geometries of piping). In addition, future recommendations such as applying synthetic minority oversampling technique, generative adversarial neural network, transfer learning, and physics-informed ML are discussed to improve the ML application for NPP piping systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111760"},"PeriodicalIF":1.9,"publicationDate":"2025-07-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144704483","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":"Superiority of clay composite materials of bentonite intercalated with the bimetallic MOFs-Pb/Cu, and nano magnetite to enhance the gamma and neutron radiation shielding","authors":"Islam M. Nabil ,&nbsp;Sameh A. Elhameed ,&nbsp;Elhassan A. Allam ,&nbsp;K.A. Mahmoud ,&nbsp;Shaaban M. Shaaban ,&nbsp;R.A. Elsad ,&nbsp;Islam N. Fathy","doi":"10.1016/j.anucene.2025.111762","DOIUrl":"10.1016/j.anucene.2025.111762","url":null,"abstract":"<div><div>In this study, a series of ternary nanocomposite materials were developed by integrating bentonite clay with nano magnetite and bimetallic metal–organic frameworks (MOFs) based on Pb/Cu and tartaric acid as organic linker. These composites were synthesized in varying compositions and systematically evaluated through both experimental techniques and theoretical modeling for their gamma radiation shielding performance. The composites were formulated with fixed proportions of 40.0 % bentonite as a base material with varying ratios of MOFs and nano magnetite specifically: (60.0–X) % bimetallic Pb/Cu-MOFs and X% nano magnetite, where X = 45.0 %, 30.0 %, and 15.0 %. These formulations are denoted as [B₄₀MPC_X_MG_(60-_X₎]. Comprehensive characterization of the base material of bentonite, and the Pb/Cu-MOFs was conducted using FT-IR, Raman, XRD, XRF, and HR-TEM tests. The radiation shielding properties of the composites were then assessed using Monte Carlo simulation code (MC) and Phy-X software across a broad photon energy range (0.015–15.0 MeV). Key shielding parameters were calculated to evaluate the shielding effectiveness. The results revealed that the MOF-Pb/Cu content controls either the γ-radiation and neutron attenuation performance in the prepared B₄₀MPC_XMG_(60-X) nano-composites. The B₄₀MPC₄₅MG₁₅ sample exhibited the highest linear attenuation coefficient for gamma radiation among the prepared composites, along with the lowest values of H/TV_L. Additionally,the B₄₀MPC₃₀MG₃₀ sample demonstrated the highest fast neutron removal cross-section among the prepared composites. The B₄₀MPC₃₀MG₃₀ sample is recommended as the optimal composition for superior shielding performance against both gamma rays and neutrons.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111762"},"PeriodicalIF":1.9,"publicationDate":"2025-07-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144712891","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":"Flow stability and linear disturbance analysis of single-phase natural circulation loop under different thermal boundaries","authors":"Yiwa Geng,&nbsp;Jiatai Liu,&nbsp;Yuntao Zheng,&nbsp;Yanyu Sun,&nbsp;Jun Fang,&nbsp;Shuliang Huang,&nbsp;Hui Wang,&nbsp;Yuxiang Wu","doi":"10.1016/j.anucene.2025.111752","DOIUrl":"10.1016/j.anucene.2025.111752","url":null,"abstract":"<div><div>Natural circulation is a common flow state that occurs in the pressurized water reactor’s normal operation conditions and accident scenarios. Essentially, the driving force between a lower-level heat source and a higher-level cold source will cause the flow. During the pressurized water reactor design process, the natural circulation concept is applied intelligently in the passive system. Compared to active driving, passive driving is equipped with the characteristics of a low driving force and is easily affected by other factors, like flow resistance, and impurities in the working medium, increasing the system’s uncertainty. Flow instability is an important phenomenon that affects the passive system function. Flow oscillations will cause the components mechanical fatigue and departure from nucleate boiling, which further triggers power oscillations due to thermal–hydraulic and neutron coupling. Nyquist Criterion and Root-Search methods have been proposed by previous researchers for solving rectangular loop stability boundaries. On this basis, this paper proposes an Identifiable-Parameter Stability Criterion (IPSC). A one-dimensional mathematical analytical model of the simplified rectangular loop under different thermal boundary conditions was developed based on the linear perturbation analysis method. This paper removes the restriction of symmetry conditions in previous work, expanding the applicability of thermal boundary conditions. Based on the mesh discrete-independence calculations, the effect of size proportions in non-uniformly arranged loops on stability is analyzed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111752"},"PeriodicalIF":1.9,"publicationDate":"2025-07-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144696498","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}