Progress in Nuclear Energy最新文献_第8页

Experimental study of spacer effects on post-dryout heat transfer in a tight-lattice rod bundle

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-10 DOI: 10.1016/j.pnucene.2025.105713

Shuo Chen , Chen Zeng , Yao Xiao , Hanyang Gu

{"title":"Experimental study of spacer effects on post-dryout heat transfer in a tight-lattice rod bundle","authors":"Shuo Chen , Chen Zeng , Yao Xiao , Hanyang Gu","doi":"10.1016/j.pnucene.2025.105713","DOIUrl":"10.1016/j.pnucene.2025.105713","url":null,"abstract":"<div><div>Tight-lattice fuel assembly is an advanced design for a compact water-cooled reactor core, but it presents challenges during loss of flow accidents, potentially suffering from critical heat flux (CHF). Downstream of the CHF point, dispersed flow with steam-entrained droplets becomes critical, significantly deteriorating heat transfer abilities. This paper experimentally studies the heat transfer capabilities downstream of spacer grids under post-dryout dispersed flow conditions. Experiments were conducted based on the high-temperature, high-pressure open-loop system at the Reactor Thermal-Hydraulics Laboratory of Shanghai Jiao Tong University. The working medium is water and the conditions tested included pressures ranging from 6 to 10 MPa, mass fluxes between 65 and 200 kg⸱m<sup>−2</sup>⸱s, heat fluxes from 75 to 200 kW⸱m<sup>−2</sup>, and inlet qualities between 0.543 and 0.887. The heat transfer enhancement ratio is defined to assess the impact of spacer grids. It is shown that the heat transfer enhancement ratio is influenced by quality, mass flux, and system pressure. Specifically, spacer grids exhibit better heat transfer enhancement capabilities under conditions of low qualities, low mass flow rates, and high pressures. The circumferential wall temperature distribution is quite uniform before dryout, while local hot spots appear near the 30°, 90°, and 180° directions after dryout. Spacer grids exhibit optimal heat transfer enhancement under low qualities, low mass flow rates, and high pressure conditions, with the maximum effect observed at the spacer grid outlet, decreasing exponentially thereafter. Available correlations for spacer grid heat transfer characteristics are evaluated, with a novel correlation according to experimental data proposed. This novel correlation closely aligns with experimental results, maintaining an error range within ±15%.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105713"},"PeriodicalIF":3.3,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579287","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}

引用次数: 0

Corrosion mechanisms and differences of Inconel 617 and Incoloy 800H under high-temperature air ingress accident

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-10 DOI: 10.1016/j.pnucene.2025.105725

Haoxiang Li , Dapeng Yan , Guodong Li , Changqing Tian , Jiayi An , Jiaqi Tan , Wei Zheng , Bin Du , Huaqiang Yin

{"title":"Corrosion mechanisms and differences of Inconel 617 and Incoloy 800H under high-temperature air ingress accident","authors":"Haoxiang Li , Dapeng Yan , Guodong Li , Changqing Tian , Jiayi An , Jiaqi Tan , Wei Zheng , Bin Du , Huaqiang Yin","doi":"10.1016/j.pnucene.2025.105725","DOIUrl":"10.1016/j.pnucene.2025.105725","url":null,"abstract":"<div><div>In the high-temperature operation of very high-temperature gas-cooled reactors (VHTRs), \"microclimate reaction\" are one of the key corrosion reactions of concern. This study investigates the corrosion behavior and mechanisms of the two mainstream high-temperature alloys, Incoloy 800H and Inconel 617, under minor air ingress accidents. The main focus is on comparing and analyzing the \"microclimate reaction\" corrosion behavior and mechanisms of the two alloys, as well as exploring changes in their mechanical properties. The results show that during the experiment, both alloys primarily underwent oxidation and \"microclimate reaction\". The double-layer oxide of Cr<sub>2</sub>O<sub>3</sub>-MnCr<sub>2</sub>O<sub>4</sub>/MnFe<sub>2</sub>O<sub>4</sub> and the SiO<sub>2</sub> layer on the surface of Incoloy 800H provide good resistance to \"microclimate reaction,\" whereas the single-layer Cr<sub>2</sub>O<sub>3</sub> oxide on Inconel 617 is not effective in preventing \"microclimate reaction\". The mechanical property test results indicate that the decarburization of the alloy caused by the \"microclimate reaction\" does not significantly affect the overall basic mechanical properties of the alloy on a macroscopic level. Research on the \"microclimate reaction\" should primarily focus on preventing the oxide layer from cracking and reducing the time during which the alloy matrix is unprotected by a stable oxide layer.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105725"},"PeriodicalIF":3.3,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143591962","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}

引用次数: 0

Numerical modeling and parameters analysis of marine radionuclide dispersion under the Fukushima Daiichi nuclear accident

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-09 DOI: 10.1016/j.pnucene.2025.105716

Hongyu Li, Deyi Chen, Baojie Nie, Dezhong Wang

{"title":"Numerical modeling and parameters analysis of marine radionuclide dispersion under the Fukushima Daiichi nuclear accident","authors":"Hongyu Li, Deyi Chen, Baojie Nie, Dezhong Wang","doi":"10.1016/j.pnucene.2025.105716","DOIUrl":"10.1016/j.pnucene.2025.105716","url":null,"abstract":"<div><div>Radioactive contaminated water was released into the Pacific Ocean under the Fukushima Daiichi nuclear accident in 2011. Knowledge of the marine radionuclide dispersion behaviors is significant to accurately evaluate radiation effects due to ocean discharge of radioactive contaminated water. The marine radionuclide dispersion is affected by wind field, tide, seafloor topography, temperature and salinity, etc. Significant uncertainty exists in marine radionuclide dispersion simulation with the consideration of complex environmental factors. MIKE 3 FM computational platform was used to establish a Fukushima marine radionuclide dispersion model driven by ocean current and tidal coupling boundary under consideration of various environmental factors. Further, the model accuracy validation with measurement data and inter-models’ comparison for bias mechanism analysis were performed based on the benchmark under the frame of IAEA MODARIA Programme. The effects of vertical distribution of layers mesh, grid resolution, temperature and salinity, horizontal diffusion parameters, wind characteristics as well as tidal forcing on the dispersion behavior were discussed. The results show that the established model can accurately describe the oceanic currents in the coastal waters of Fukushima and predict the dispersion trajectory, as well as concentration level and distribution. More vertical layers, wind-driven forces data and refined mesh for the area of concern should be considered to improve the predictive accuracy of model. The modeling and parameters analysis can help to decrease uncertainty in predicting the levels of radioactivity in the adjutant waters east of the Fukushima site.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105716"},"PeriodicalIF":3.3,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579385","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}

引用次数: 0

Mechanisms of bubble effects on acoustic propagation characteristics in high-temperature and high-pressure bubble flows

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-09 DOI: 10.1016/j.pnucene.2025.105717

Yu Jiang , Genshan Jiang , Jianhao Sun , Yu Zhou , Hao Li

{"title":"Mechanisms of bubble effects on acoustic propagation characteristics in high-temperature and high-pressure bubble flows","authors":"Yu Jiang , Genshan Jiang , Jianhao Sun , Yu Zhou , Hao Li","doi":"10.1016/j.pnucene.2025.105717","DOIUrl":"10.1016/j.pnucene.2025.105717","url":null,"abstract":"<div><div>In the complex gas-liquid two-phase flow environment of the nucleate boiling region outside the heat-exchange pipes of steam generators, acoustic propagation is influenced by multiple factors. Previous research has largely considered how the bubble distribution and acoustic frequency affect the acoustic velocity and attenuation, whereas the role of boundary-layer thickness in acoustic propagation has not been explored systematically. Using bubble damping theory and the linear model of acoustic propagation in water containing monodisperse bubbles, derived here is the relationship among acoustic velocity, acoustic attenuation, and boundary-layer thickness in bubble flow. Also established is a numerical model for acoustic propagation in bubble flow under high temperature and pressure, and how the bubble distribution and acoustic frequency affect the sound field is analyzed. This study reveals the factors influencing bubble damping and how it regulates acoustic attenuation, providing insight into how acoustic waves propagate in gas-liquid two-phase flows.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105717"},"PeriodicalIF":3.3,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579386","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}

引用次数: 0

Development and steady-state validation of the fine sub-channel thermal-hydraulic code CUNLUN

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-09 DOI: 10.1016/j.pnucene.2025.105712

Hongwei Jiang , Guangliang Chen , Zhenjie Li , Zhaofei Tian , Hao Qian , Xinli Yin

{"title":"Development and steady-state validation of the fine sub-channel thermal-hydraulic code CUNLUN","authors":"Hongwei Jiang , Guangliang Chen , Zhenjie Li , Zhaofei Tian , Hao Qian , Xinli Yin","doi":"10.1016/j.pnucene.2025.105712","DOIUrl":"10.1016/j.pnucene.2025.105712","url":null,"abstract":"<div><div>This paper, based on the flow characteristics of coolant between channels in a pressurized water reactor (PWR), proposes a reactor core sub-channel calculation and analysis method that utilizes fine sub-channel control volumes. A three-dimensional momentum source resistance model for mixing vanes was established based on the distributed resistance method. Using the C++ programming language, a PWR reactor core thermal-hydraulic calculation and analysis program called CUNLUN was independently developed based on this method. Numerical simulations were conducted using the Westinghouse 5 × 5 rod bundle experiment and the OECD/NEA MATiS-H international benchmark experiment. The maximum relative error for the Westinghouse 5 × 5 rod bundle experiment was 6.73%, and the simulation results for the MATiS-H benchmark experiment also agreed well with experimental data, demonstrating that this program can improve the accuracy of predicting flow characteristics within the sub-channels of a PWR reactor core. Additionally, Computational Fluid Dynamics (CFD) simulation data were used as benchmark data, and in 112 comparison points, the average calculation accuracy of 102 comparison points improved by 74.21%, verifying the applicability and validity of the three-dimensional momentum source resistance model for mixing vanes based on the distributed resistance method for the CUNLUN code. This code provides an effective design and research tool for thermal-hydraulic analysis of PWR fuel assemblies and offers valuable insights for the development of reactor core thermal-hydraulic analysis programs.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105712"},"PeriodicalIF":3.3,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579403","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}

引用次数: 0

Effect of circumferential and radial position on rewetting time under boil-off condition of 54 pins AHWR

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-09 DOI: 10.1016/j.pnucene.2025.105728

Subodh Kumar Yadav , Mithilesh Kumar , Ravi Kumar , P. Majumdar , D. Mukhopadhyay

引用次数: 0

Debris bed coolability in hypothetical core disruptive accidents: Theory, experiment, and numerical simulation review

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-09 DOI: 10.1016/j.pnucene.2025.105700

Yonglong Li , Shaojie Tan , Yubin Lin , Kai Wang , Songbai Cheng

{"title":"Debris bed coolability in hypothetical core disruptive accidents: Theory, experiment, and numerical simulation review","authors":"Yonglong Li , Shaojie Tan , Yubin Lin , Kai Wang , Songbai Cheng","doi":"10.1016/j.pnucene.2025.105700","DOIUrl":"10.1016/j.pnucene.2025.105700","url":null,"abstract":"<div><div>The coolability of debris beds in severe accidents of sodium-cooled fast reactors (SFRs) is critical for preventing radionuclide release, yet existing research has predominantly focused on light water reactors (LWRs), leaving a systematic analysis for SFRs lacking. This paper reviews debris bed coolability through theoretical models, experimental studies, and numerical simulations, while contrasting the distinctions between LWRs and SFRs. Theoretical models reveal that the inclusion or exclusion of interfacial drag marks a key divergence—models like Reed's, which neglect interfacial resistance, are suitable for high-flow bottom-flooding, whereas formulations such as Schulenberg-Müller's better align with top-flooding scenarios. Experimental findings demonstrate that particle size distribution, geometry, and porosity significantly influence dryout heat flux (DHF). Multidimensional cooling strategies (e.g., bottom-flooding or downcomers) disrupt countercurrent flooding limits (CCFL), enhancing DHF by 1.5–2.7 × compared to conventional top-flooding. Numerical simulations confirm the predictive capability of codes like MEWA and MELCOR for homogeneous debris beds, yet limitations persist in addressing SFR-specific challenges, such as high-temperature liquid metal coolant behavior and complex flow regimes. Future work must prioritize SFR-focused experiments, extend models to liquid metal conditions, and strengthen experimental-simulation synergies to improve risk assessment reliability. This paper provides theoretical and technical foundations for optimizing SFR severe accident management strategies.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105700"},"PeriodicalIF":3.3,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579381","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}

引用次数: 0

Uncertainty analysis of the inverse LASSO estimation scheme on radioactive source distributions inside reactor building rooms from air does rate measurements

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-09 DOI: 10.1016/j.pnucene.2025.105710

Wei Shi , Masahiko Machida , Susumu Yamada , Koji Okamoto

{"title":"Uncertainty analysis of the inverse LASSO estimation scheme on radioactive source distributions inside reactor building rooms from air does rate measurements","authors":"Wei Shi , Masahiko Machida , Susumu Yamada , Koji Okamoto","doi":"10.1016/j.pnucene.2025.105710","DOIUrl":"10.1016/j.pnucene.2025.105710","url":null,"abstract":"<div><div>Very recently, Least Absolute Shrinkage and Selection Operator (LASSO) has been proposed as a scheme capable to inversely estimate radioactive source distributions inside reactor building rooms from air dose rate measurements together with the predicted lower bound of the measurement numbers for successful reconstructions. However, no one has ever analyzed how the uncertainty of input data including the measurement errors influences the accuracy of the inverse estimation results. In this paper, we therefore perform uncertainty analysis of the LASSO scheme and suggest an uncertainty estimation function derived based on the theory of Candes. We actually demonstrate in two types of numerical tests with different input uncertainties obtained by using Monte Carlo code, Particle and Heavy Ion Transport code System (PHITS) that the calculated errors obey the proposed uncertainty estimation function. Thus, the LASSO scheme allows to successfully estimate radioactive distributions within the predicted uncertainty.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105710"},"PeriodicalIF":3.3,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579402","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}

引用次数: 0

Development of a semiempirical model for the aerosol scrubbing in a sodium pool 开发钠池气溶胶洗涤半经验模型

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-08 DOI: 10.1016/j.pnucene.2025.105699

Gen Jiang , Mou Wang , Jingyu Guo , Kai Wang , Songbai Cheng

{"title":"Development of a semiempirical model for the aerosol scrubbing in a sodium pool","authors":"Gen Jiang , Mou Wang , Jingyu Guo , Kai Wang , Songbai Cheng","doi":"10.1016/j.pnucene.2025.105699","DOIUrl":"10.1016/j.pnucene.2025.105699","url":null,"abstract":"<div><div>In sodium cooled fast reactors, failure of fuel rods may result in the release of radioactive fission products into the coolant pool. Radionuclides may be transported to the cover gas area through bubble entrainment, thereby affecting the normal operation of the reactor system. Therefore, the calculation of the scrubbing mechanism of radioactive fission product aerosols in the rising bubbles of the sodium pool is crucial to ensure the safe operation of fast reactor. In this study, we intend to propose a new sodium pool scrubbing model to evaluate the retention capacity of sodium pools for fission products. The model includes two modules: bubble dynamics module to calculate bubble size, eccentricity, and rising speed; aerosol pool scrubbing module considering four aerosol removal mechanisms: Brownian diffusion, gravitational sedimentation, inertial deposition, and vapor condensation. The model was firstly validated by two transient scenarios, PLOF+ and UTOP+. Moreover, SRT (Simplified Radionuclide Transport) experiments were utilized to validate the model. To further improve the accuracy of the model, we modified the model by introducing a correlation factor to correct the inertial deposition effect. In addition, the influence of various key parameters on the efficiency of pool scrubbing was discussed. The results show that larger bubble diameters and deeper depths have a promoting effect on the pool scrubbing efficiency; The influence of particle diameter is nonmonotonic; The pool temperature has no significant impact.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105699"},"PeriodicalIF":3.3,"publicationDate":"2025-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579384","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}

引用次数: 0

A simulator and controller for the real-time autonomous operation of pressurized water reactors based on expert systems

IF 3.3 3区工程技术

Progress in Nuclear Energy Pub Date : 2025-03-05 DOI: 10.1016/j.pnucene.2025.105675

Marcos A.G.S. Filho , Igor B. Baptista , Alan M.M. Lima , Victor H.C. Pinheiro , Roberto Schirru

{"title":"A simulator and controller for the real-time autonomous operation of pressurized water reactors based on expert systems","authors":"Marcos A.G.S. Filho , Igor B. Baptista , Alan M.M. Lima , Victor H.C. Pinheiro , Roberto Schirru","doi":"10.1016/j.pnucene.2025.105675","DOIUrl":"10.1016/j.pnucene.2025.105675","url":null,"abstract":"<div><div>Pressurized Water Reactor (PWR) power control is carried out by continuously monitoring relevant parameters that depend on operating conditions. It is up to the plant operator to identify when one of these parameters move out of its expected values and perform all the necessary procedures to guarantee that these parameters return to its pre-defined values in a safe way. However, decision-making of which procedures should be done by reactor operators can be subjective, based both on official guidelines and operator expertise. The objective of this work is to implement a real-time autonomous operation system (AOS) that incorporates the knowledge of a human operator for the core power control of a PWR. This AOS is divided in two modules: an expert system controller and a reactivity simulator module. The expert system controller must be able to reproduce all decision processes that would be done by a human operator during their daily activities, while the reactivity simulator reproduces the online behavior of the nuclear plant. The implementation of this AOS is verified and validated against real data of a full operational cycle of a PWR and the presented results show an excellent agreement between AOS simulation and reactor records.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105675"},"PeriodicalIF":3.3,"publicationDate":"2025-03-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548621","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}

引用次数: 0