Nuclear Engineering and Design最新文献

{"title":"Experimental investigation of subcooled flow boiling characteristics of water in vertical helically coiled tubes","authors":"Yuqing Su,&nbsp;Xiaowei Li,&nbsp;Xinxin Wu","doi":"10.1016/j.nucengdes.2024.113716","DOIUrl":"10.1016/j.nucengdes.2024.113716","url":null,"abstract":"<div><div>Helically coiled tubes are employed as heat transfer tubes in Once Through Steam Generator (OTSG) of High Temperature Gas-cooled Reactor (HTGR) due to their compact structure, large heat transfer area and excellent thermal expansion adaptability. However, the helical geometry induces centrifugal forces and secondary flows in the tube, resulting in notable differences in flow and heat transfer characteristics compared to that of straight tubes. This study conducted an experimental investigation on the onset of nucleate boiling (ONB) and the subcooled boiling heat transfer coefficient in helically coiled tubes with a large curvature ratio (<em>δ</em> = 0.109). The experimental parameters cover broad ranges. The system pressures are ranging from 3.5 to 7 MPa, mass fluxes are from 300 to 1100 kg/(m²·s) and heat fluxes are from 50 to 600 kW/m2. The experimental results indicate that the ONB can occur even when the average inner wall temperature is below the fluid’s saturation temperature. An increase in heat flux advances the ONB, while increases in mass flux and system pressure delay it. Enhancements in both heat flux and mass flux improve the subcooled boiling heat transfer coefficient. Additionally, higher system pressure also increases the heat transfer coefficient, although this effect diminishes as the quality increases. Based on the experimental data and dimensionless analysis, new correlations were proposed for predicting ONB and calculating the subcooled boiling heat transfer coefficient in helically coiled tubes. Both new correlations exhibit more accurate predictive capabilities, with mean absolute percentage error (MAPE) values of 6.20 % and 8.86 %, respectively.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113716"},"PeriodicalIF":1.9,"publicationDate":"2024-11-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660863","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":"Assessment of stratification and entrainment models in CATHARE 3 code during FONESYS activities","authors":"Sofia Carnevali,&nbsp;Philippe Fillion","doi":"10.1016/j.nucengdes.2024.113700","DOIUrl":"10.1016/j.nucengdes.2024.113700","url":null,"abstract":"<div><div>This paper presents the assessment of CATHARE 3 code against tests performed in the horizontal TPTF (Two Phase Tests Facility) and Mantilla facilities. This activity falls within the framework of a benchmark conducted by the Forum &amp; Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). The aim of this benchmark is to highlight the capabilities of system thermal–hydraulic codes to predict the horizontal stratification criteria and the onset of droplet entrainment and entrainment rate. One of the objectives of horizontal TPTF (TPTF-H) experiments, conducted by JAERI (Japan), was to analyse the thermal–hydraulic responses in the horizontal legs of Light Water Reactors. The considered tests were at steady state and saturated steam/water two-phase flow conditions for a pressure varying between 3 MPa and 12 MPa. CATHARE code simulations show a general rather good agreement with TPTF-H experimental results although some improvements like the stratification regime prediction seem necessary. In two horizontal test sections with different diameters, Mantilla carried out air/water experiments at low pressure, from stratified to annular flow conditions, including droplet entrainment. Entrainment fraction obtained both with the two-fluid 6-equation model and the 3-field model of CATHARE are compared against the experimental data. Some improvements of the existing models are proposed for the entrainment and deposition processes to better fit the experiments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113700"},"PeriodicalIF":1.9,"publicationDate":"2024-11-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661297","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":"Velocity and pressure fluctuations downstream analytical spacer grids: Structure and transport","authors":"N. Turankok ,&nbsp;T. Lohez ,&nbsp;V. Biscay ,&nbsp;L. Rossi","doi":"10.1016/j.nucengdes.2024.113682","DOIUrl":"10.1016/j.nucengdes.2024.113682","url":null,"abstract":"<div><div>In Pressurized Water Reactors (PWR), fluid–structure interaction provokes vibrations of the fuel rods leading to grid-to-rod fretting. To explore the origins of the local excitations of the rods by the flow, analytical experiments are performed within a 5 × 5 rod bundle maintained by spacer grids. Experiments are performed using two sets of grids, i.e. configurations with and without mixing vanes, over a wide range of Reynolds number, i.e. from about 10,000 to 120,000. Particle Image Velocity measurements near the central instrumented rod reveal different structures of velocity fields and their fluctuations for the two configurations. Mean fields are in U inverted shape without mixing vanes and in λ shape with mixing vanes. Fields of velocity fluctuations are in butterfly shape without mixing vanes and in Y (low Re) and K (high Re) shapes with mixing vanes. Near the grid, <span><math><mrow><msub><mrow><msup><mrow><mi>u</mi></mrow><mo>′</mo></msup></mrow><mrow><mi>rms</mi></mrow></msub><mo>/</mo><msub><mi>U</mi><mrow><mi>flow</mi></mrow></msub></mrow></math></span> increases by about 30% for <span><math><mrow><msub><mrow><mi>Re</mi></mrow><mrow><mi>Dh</mi></mrow></msub></mrow></math></span> varying from 10,000 to 120,000. The presence of eddies is highlighted by visualisations of the fields of the velocity fluctuations. The spacing of these eddies and their sizes are found to be in agreement with the periodic length scales (obtained from frequency peaks) and the integral length-scales measured. Frequencies of the observed frequency peaks are found to be the same for both velocity and pressure fluctuations. Consequently, a new Strouhal versus Reynolds map is built over all existing data for Reynolds numbers between 10,000 and 120,000 and for the two grids configurations. The integral length-scales are found to be about the same for velocity and pressure fluctuations near the grid. It is shown that events of pressure fluctuations are persistent and transported with a speed close to the one of the mean flow. Moreover, this transport is correlated with the displacement of velocity events. The coherence between measured turbulence statistics, the observed eddies and cross-correlations of pressure and velocity fluctuations support the conclusion that the large-scale eddies (about 0.2 <span><math><mrow><msub><mi>D</mi><mi>h</mi></msub></mrow></math></span>) are at the origin of main pressure fluctuations and their transport. To the authors’ knowledge, this is the first experimental evidence of the origin and transport of pressure fluctuations downstream analytical grids with different geometries relevant to PWR fuel assemblies.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113682"},"PeriodicalIF":1.9,"publicationDate":"2024-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660859","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":"Numerical study on the impact characteristics of molten lead-bismuth droplets into water","authors":"Yue Lin ,&nbsp;Dalin Zhang ,&nbsp;Yutong Chen ,&nbsp;Xisi Zhang ,&nbsp;Jian Deng ,&nbsp;Peng Du ,&nbsp;Wenxi Tian ,&nbsp;Suizheng Qiu ,&nbsp;Guanghui Su","doi":"10.1016/j.nucengdes.2024.113689","DOIUrl":"10.1016/j.nucengdes.2024.113689","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113689"},"PeriodicalIF":1.9,"publicationDate":"2024-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660857","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":"An investigation of structural integrity of CSB − for normal contact with RV outlet nozzle","authors":"Niniek Ramayani Yasintha,&nbsp;Omondi Christopher Ondwasi,&nbsp;Namgung Ihn","doi":"10.1016/j.nucengdes.2024.113704","DOIUrl":"10.1016/j.nucengdes.2024.113704","url":null,"abstract":"<div><div>This research is about improving the design of Core Support Barrel (CSB) by eliminating the gap between its hot-leg opening and the reactor vessel (RV) outlet nozzle. This gap causes significant bypass flow, impacting reactor performance. A proposed design modification aimed to establish contact between these components, requiring a thorough analysis of the contact interface. The study determined that a contact overlap of 15 mm meets ASME code standards and other relevant criteria. Moreover, this design improvement offers the added benefit of enhancing the seismic response of the reactor internals due to the increased horizontal support provided by the RV outlet nozzle to the CSB. Modal analysis revealed a substantial upward shift in the CSB’s natural frequency from 15.6 Hz-20.7 Hz to 21.9 Hz-38 Hz, representing a more than 40 % increase. The research demonstrates that the proposed design effectively eliminates bypass flow, and significantly improves the seismic response of the reactor internals.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113704"},"PeriodicalIF":1.9,"publicationDate":"2024-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660862","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":"Robust multipole approach for continuous nuclear Data: RKFIT implementation for X2 VVER-1000 reactor benchmark","authors":"Abdolbaset Agh,&nbsp;Mahdi Zangian,&nbsp;Abdolhamid Minuchehr","doi":"10.1016/j.nucengdes.2024.113699","DOIUrl":"10.1016/j.nucengdes.2024.113699","url":null,"abstract":"<div><div>The windowed multipole method stands out as a promising approach for effectively conducting on-the-fly Doppler-broadening for continuous nuclear cross-section data. Its implementation was predominantly relied on the conversion of the resolved resonance parameters and vector-fitting technique into the multipole representation. Recently, progress has been made in deriving multipole representations from continuous cross-section data, with one notable method being RKFIT mothed, that is a robust least square fitting method. The advantages and disadvantages of this method have been thoroughly investigated. Detailed instructions on utilizing this method are provided within the OpenMC Monte Carlo calculation code. The experimental reactor physics results of the X2 reactor (a VVER-1000 type reactor) have been used to benchmark the effectiveness of the continues nuclear data generated by this approach via the OpenMC code simulations. Also, the simulation results are compared with those obtained using continuous cross-section libraries and other multipole libraries.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113699"},"PeriodicalIF":1.9,"publicationDate":"2024-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660860","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 study of spherical tubesheet design for APR1400 and AP1000 steam generators","authors":"Seungmin Kim,&nbsp;Namgung Ihn","doi":"10.1016/j.nucengdes.2024.113701","DOIUrl":"10.1016/j.nucengdes.2024.113701","url":null,"abstract":"<div><div>This research investigates the potential of spherical tube sheets as a more efficient alternative to conventional flat tube sheets in APR1400 and AP1000 steam generators. By leveraging the inherent strength of spherical structures, we aim to reduce material consumption, manufacturing costs, and construction time. Finite Element Method simulations were employed to compare the deformation behavior of flat and spherical tube sheets under high-temperature and high-pressure conditions. The analysis identified optimal spherical tube sheet thicknesses that maintained structural integrity while achieving significant reductions in thickness: approximately 20–30 % for APR1400 and 40–50 % for AP1000. Moreover, the spherical design can enhance thermal efficiency by minimizing variations in heat transfer tube lengths, leading to more uniform heat distribution.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113701"},"PeriodicalIF":1.9,"publicationDate":"2024-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660858","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":"Entropy generation analysis for fine flow states in PWR fuel assembly","authors":"Yunsheng Zhang ,&nbsp;Guangliang Chen ,&nbsp;Hao Qian ,&nbsp;Lixuan Zhang ,&nbsp;Jinchao Li ,&nbsp;Hanqi Zhang ,&nbsp;Dabin Sun ,&nbsp;Hansheng Zhi","doi":"10.1016/j.nucengdes.2024.113708","DOIUrl":"10.1016/j.nucengdes.2024.113708","url":null,"abstract":"<div><div>Although the development of pressurized water reactor (PWR) technology has been relatively well developed so far, the depth of research on the thermo-hydraulic characteristics of the in-reactor coolant, especially the energy dissipation characteristics, still needs to be explored further. As an energy conversion system for PWR, entropy generation analysis plays a vital role in obtaining the irreversible loss of the coolant quantitatively and directionally, which is scarce. It is difficult to access the full coolant energy losses and thermal–hydraulic properties, and to provide direction and schemes for optimizing the flow field in the core. In this paper, the energy dissipation of the flow field is finely analyzed in the 5x5 rod bundle of the PWR core. For the irreversible dissipation of the flow energy of the coolant, the pulsation dissipation entropy generation and wall friction entropy generation account for about 90% and 10%, respectively, and the direct dissipation entropy generation is negligible. For the high dissipation region, the distribution of pulsation dissipation entropy generation in space has a directional role. The irreversibly dissipated energy is calculated using the definition of entropy generation, and the dissipation characteristics of cross-flow kinetic energy in different regions are analyzed accordingly. Lastly, the irreversible dissipation of temperature difference heat transfer is considered and correlated with the above hydraulic viscous dissipation, and the negative correlation characteristics of the two are found. Meanwhile, the parameter constructed by the two together has some advantage in flow field evaluation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113708"},"PeriodicalIF":1.9,"publicationDate":"2024-11-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660861","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":"Computational analysis for predicting fission product release of PeLUIt-40 under normal operating conditions","authors":"Anik Purwaningsih ,&nbsp;Dwi Irwanto ,&nbsp;Julwan H. Purba","doi":"10.1016/j.nucengdes.2024.113677","DOIUrl":"10.1016/j.nucengdes.2024.113677","url":null,"abstract":"&lt;div&gt;&lt;div&gt;PeLUIt (&lt;em&gt;Pembangkit Listrik dan Uap untuk Industri&lt;/em&gt;) is an High Temperature Gas Cooled Reactor (HTGR)-based cogeneration reactor designed by the Indonesian National Nuclear Energy Agency of Indonesia (now the Research Organization for Nuclear Energy, BRIN). HTGR uses tri-structural isotropic (TRISO) coated particle fuel, which is the main barrier to the release of fission products. The ability of the fuel to retain fission products both under normal and accident conditions is very important for design licensing. This study was conducted to predict the release of fission products of PeLUIt-40 (PeLUIt with a power of 40 MWt) under normal operating conditions. In this study, the Source Term Analysis Code System (STACY) was used to predict the release of radiologically significant fission products Ag110m, Cs137, I131, and Sr90. OpenMC was used to calculate neutronic parameters such as burnup (Fissions per initial heavy metal atom—FIMA), fast neutron fluence, and fission product inventory. A single pebble model with different irradiation times and temperature variations was used to simulate the fission product release in PeLUIt-40. The time and temperature variations were used to investigate the sensitivity of the fission product release fraction in PeLUIt-40 fuel to these parameters and to estimate the maximum safe fuel temperature during operation. The simulation results showed that the largest release fraction was Ag110m release compared to other radionuclide releases. At the normal operating temperature of 977 °C, the fission product release fractions during one-through-one-out (OTTO) and 5-pass cycles were two orders of magnitude lower than the failure fraction for the high-temperature reactor (HTR)-Module &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;1.6&lt;/mn&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;msup&gt;&lt;mrow&gt;&lt;mn&gt;10&lt;/mn&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msup&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; and there was no defective particle during operation. In the OTTO cycle, the maximum fuel temperature that did not cause defective particle was about 1250 °C, but the Ag110m release fraction exceeded &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;1.6&lt;/mn&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;msup&gt;&lt;mrow&gt;&lt;mn&gt;10&lt;/mn&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msup&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;. The release fraction of all fission products in the OTTO cycle is below &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;1.6&lt;/mn&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;msup&gt;&lt;mrow&gt;&lt;mn&gt;10&lt;/mn&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msup&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; when the maximum fuel temperature is 1025 °C. While in the 5-pass cycle, the maximum fuel temperature of about 1200 °C does not cause defective particle, but the release fraction of Ag110m exceeds &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;1.6&lt;/mn&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;msup&gt;&lt;mrow&gt;&lt;mn&gt;10&lt;/mn&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msup&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;. The release fraction of all fission products in the 5-pass cycle is below &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;1.6&lt;/mn&gt;&lt;mi&gt;x&lt;/mi&gt;&lt;msup&gt;&lt;mrow&gt;&lt;mn&gt;10&lt;/mn&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mn&gt;4&lt;/mn&gt;&lt;/mrow&gt;&lt;/msup&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; when the maximum fuel temperature is 1020°","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113677"},"PeriodicalIF":1.9,"publicationDate":"2024-11-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142660850","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":"Fast prediction of key parameters in FEBA using the COSINE subchannel code and artificial neural network","authors":"Yingran Guo,&nbsp;Hao Zhang,&nbsp;Lin Chen,&nbsp;Meng Zhao,&nbsp;Yanhua Yang","doi":"10.1016/j.nucengdes.2024.113709","DOIUrl":"10.1016/j.nucengdes.2024.113709","url":null,"abstract":"<div><div>Numerical techniques have emerged as an essential tool for operators and designers to preemptively acquire key parameters in accidents analysis. However, due to insufficient experience, it is difficult for them to obtain satisfactory numerical results. Moreover, the uncertainty analysis and quantification necessitate the simulation of a substantial number of samples, which requires a significant amount of computational time. Therefore, the development of a fast prediction model becomes imperative. In this work, a prediction model based on the in-house COSINE subchannel code and Multi-Head Perceptron (MHP) is developed. The COSINE subchannel code is employed to provide data sets for training neural networks. Firstly, the numerical results of COSINE subchannel code are compared with experimental data to ensure the accuracy of data sets. Secondly, input features for neural networks are selected by evaluating the impact of input parameters on numerical results, and a series of simulations is carried out to generate data sets. Then, a comparative analysis was conducted between the Multi-Layer Perceptron (MLP) and Support Vector Regression (SVR) models, and the MLP model performs better. Subsequently, the MLP was compared with the MHP, demonstrating the advantage of MHP model. Based on this, the predictions are conducted using the MHP model and the distribution of key parameters is compared with that obtained by COSINE subchannel code. The results illustrate that developed MHP model is an efficient tool for predicting key parameters during the reflooding phase.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"430 ","pages":"Article 113709"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142661298","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}