International Journal of Advanced Nuclear Reactor Design and Technology最新文献

{"title":"Study of cosCONT heat transfer model based on NUPEC M-7-1","authors":"Chaowei Chen ,&nbsp;Yu Yang ,&nbsp;Lin Wang ,&nbsp;Ting Huang ,&nbsp;Shengfei Wang ,&nbsp;Zhaocan Meng","doi":"10.1016/j.jandt.2026.03.001","DOIUrl":"10.1016/j.jandt.2026.03.001","url":null,"abstract":"<div><div>The cosCONT module within the COSINE code package is designed for containment thermal-hydraulic analysis. This study evaluates its heat transfer modeling capabilities by simulating the International Standard Problem ISP35 (NUPEC M-7-1) experiment, which investigated helium mixing and distribution under steam/helium release and dome spray conditions. The analysis focuses on key modeling factors: empirical condensation correlations, fixed heat transfer coefficients, flow resistance, spray initialization, and adiabatic boundary conditions. Results demonstrate that cosCONT effectively reproduces the overall experimental trends. The simulations exhibit a somewhat lower condensation heat removal, primarily stemming from the empirical condensation models. Imposing a fixed heat transfer coefficient leads to significant deviations by further limiting condensation, while increased flow resistance amplifies this underestimation. In contrast, results show minimal sensitivity to spray initialization or adiabatic boundaries. This work verifies the applicability of cosCONT for safety assessment and clarifies the behavior of its heat transfer models.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 1","pages":"Pages 75-88"},"PeriodicalIF":0.0,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147398413","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Numerical study on cladding burst behavior considering irradiation and oxidation with the multiphysics coupling method","authors":"Runhua Gao ,&nbsp;Xin Zhang ,&nbsp;Shangke Li ,&nbsp;Junjie Yu ,&nbsp;Gen Li","doi":"10.1016/j.jandt.2026.01.003","DOIUrl":"10.1016/j.jandt.2026.01.003","url":null,"abstract":"<div><div>This study investigates alterations in the performance of fuel rod cladding during loss of coolant accident through the development of a two-dimensional axisymmetric burst analytical model using COMSOL Multiphysics. The model incorporates irradiation and oxidation effects while accounting for cladding fluid-solid heat transfer, high-temperature anisotropic creep with gap conductivity, and temperature-time dependent oxidation and irradiation processes. This comprehensive approach enables multi-physics field coupling of heat-fluid-solid-irradiation-oxidation interactions. To elucidate the complex oxidation effects, the analysis separates three key components: oxidative oxygen uptake, crack depth, and degree of hydrogenation, examining their individual impacts on cladding behavior. The results indicated that under irradiation conditions with fast neutron flux of 2 × 10¹⁸∼8 × 10¹⁸ n/m²s, the model predicts an acceleration of burst time by approximately 8 s and a 41 % reduction in burst strain, attributed to lattice structure deterioration from irradiation-induced defects. The oxidation analysis shows that when the oxygen uptake increases from 3 % to 20 %, the burst time is delayed. This clearly demonstrates the protective role of the ZrO₂ layer. Crack depth analysis shows that increasing depth from 0.1 mm to 0.5 mm substantially reduces both cladding burst strain and strength. Hydrogen content shows a phase-dependent effect: in the low-temperature <em>α</em>-phase, it decreases cladding plastic toughness, with 600 ppm hydrogen reducing the burst strain to about 10 % at 900 K and 1000 ppm hydrogen reducing the burst strain to 5 %, while enhancing performance in the high-temperature <em>β</em>-phase. These findings provide crucial theoretical support and reference data for nuclear reactor safety design and operation, contributing significantly to the understanding of cladding behavior under complex operating conditions.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 1","pages":"Pages 21-32"},"PeriodicalIF":0.0,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039143","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Hybrid burnable absorber design based on ATF and application in 24-month long cycle fuel management","authors":"Hong Cao,&nbsp;Wenhua Zhang","doi":"10.1016/j.jandt.2026.01.001","DOIUrl":"10.1016/j.jandt.2026.01.001","url":null,"abstract":"<div><div>In order to meet the requirement of reactivity control for 24-month long cycle, an optimization design of burnable absorber based on accident tolerant fuel (ATF) is carried out. The design of hybrid burnable absorber which combines low ¹⁰B coating loading in a higher number of fuel rods with high mass content of gadolinium oxide (Gd₂O₃) in a lower number of fuel rods in the fuel assembly, is proposed. The study show that the design can effectively balance the performance parameters of the fuel rod, prolong the reactivity hold-down and achieve the 24-month long cycle. For the 24-month long cycle designed with hybrid burnable absorber, the key parameters of the reactor core meet the requirements of safety-related acceptance criteria, and the maximum power peaking factor during the cycle remains low, and the power distribution is well flattened. Furthermore, the fuel economy is equivalent to the 18-month cycle, significantly enhancing the operational benefits of the nuclear power plant. Fuel rod performance assessment—including rod internal pressure, fuel centerline temperature, and cladding oxide-layer thickness—indicates that all key parameters meet current design criteria.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 1","pages":"Pages 1-6"},"PeriodicalIF":0.0,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145957651","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Back propagation neural network-based prediction of critical heat flux in vertical and inclined tubes","authors":"Ya Li,&nbsp;Huanjun Kong,&nbsp;Yujiao Peng,&nbsp;Jianqiang Shan,&nbsp;Miao Gui","doi":"10.1016/j.jandt.2026.01.005","DOIUrl":"10.1016/j.jandt.2026.01.005","url":null,"abstract":"<div><div>Due to their unique advantages, offshore floating nuclear power plants have broad application prospects. Unlike land-based reactors, it is essential to conduct in-depth research on the characteristics of Critical Heat Flux (CHF) under marine conditions, such as inclination, heaving, and rolling. Artificial neural networks renowned for strong fitting, prediction, classification capabilities, and robustness, are widely used in fields such as economics and engineering. This paper employs a back propagation neural network to predict the CHF of vertical and inclined tubes, defining an inclination factor—CHFR to quantify the impact of inclination on CHF. It also reveals the effects of pressure, mass flow rate, thermal equilibrium quality, and inclination angle on CHFR. The BP (Back Propagation)neural network-based method for predicting inclined CHF demonstrates high accuracy. The results of this study offer valuable insights for ensuring the safe operation of reactors under marine conditions.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 1","pages":"Pages 56-64"},"PeriodicalIF":0.0,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147398411","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Reactivity safety analysis of an air-cooled nuclear thermal propulsion reactor","authors":"Zimian Duan ,&nbsp;Lihong Xu ,&nbsp;Jing Zhang ,&nbsp;Yingwei Wu ,&nbsp;Mingjun Wang ,&nbsp;Yanan He","doi":"10.1016/j.jandt.2026.03.002","DOIUrl":"10.1016/j.jandt.2026.03.002","url":null,"abstract":"<div><div>As the most promising alternative to chemical propulsion, nuclear thermal propulsion technology has been extensively studied in recent years. However, most existing research lacks comprehensive safety assessments of reactor designs. This study replicates an air-cooled nuclear thermal propulsion reactor. And using a cross-dimensional coupled solution approach, the reactor design is evaluated from both normal operation and transient responses. For the normal operation, the diversity and redundancy of the reactor's shutdown means are examined, leading to the conclusion that the reactor possesses an adequate shutdown margin and a negative fuel coefficient. For transient conditions, by simulating the reactor response under Proportion-Integration-Differentiation (PID) control, it is demonstrated that the reactor is capable of withstanding multiple types of reactivity insertion accidents (RIAs). This paper can provide a methodological foundation for subsequent safety evaluation in the design of nuclear thermal propulsion reactors.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 1","pages":"Pages 65-74"},"PeriodicalIF":0.0,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147398412","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Study on thermal hydraulic characters of corrugated plate dryer based on DPM coupled ELFM method","authors":"Shunhao Xu ,&nbsp;Bo Wang ,&nbsp;Mengyao Wang ,&nbsp;Ru Li ,&nbsp;Ruifeng Tian ,&nbsp;Sichao Tan","doi":"10.1016/j.jandt.2026.01.002","DOIUrl":"10.1016/j.jandt.2026.01.002","url":null,"abstract":"<div><div>To improve the structures and characteristics of the steam generator corrugated plate dryer and increase the steam-liquid separation efficiency, the discrete phase model (DPM) coupled with the Eulerian liquid film model (ELFM) is employed to calculate the flow field of a corrugated plate dryer, and the flow field characteristics of the double-hooked corrugated plate are analyzed. The motion of liquid droplets and the separation of liquid film in the corrugated plate are numerically simulated and investigated by the DPM coupled with the ELFM and combined with three critical values. The applicability of the model is verified by comparing the numerically calculated and experimental values in the cold-state. And the separation performance of other different types of corrugated plates is compared, which lays an important technical foundation for the selection and structural optimization of corrugated plates in engineering practice.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 1","pages":"Pages 33-48"},"PeriodicalIF":0.0,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039142","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental study on gravity settling of aerosols inside the containment vessel of nuclear power plants","authors":"Likai Fei ,&nbsp;Haowei Yuan ,&nbsp;Feng Shen ,&nbsp;Pucheng Fan ,&nbsp;Qiliang Mei ,&nbsp;Kun Zhang ,&nbsp;Yaru Fu ,&nbsp;Minyang Peng","doi":"10.1016/j.jandt.2025.12.002","DOIUrl":"10.1016/j.jandt.2025.12.002","url":null,"abstract":"<div><div>Gravity setting is an important factor in the natural removal of aerosols in the containment. In this text, it is carried out to the experimental study of aerosol gravity setting in the containment based on the CERT (CAP1400 Passive Containment Cooling System Comprehensive Performance Test Bench). By studying the influence of aerosol particle size and initial concentration of aerosol in the containment on the experiment of aerosol gravity setting mechanism in the containment, it is found that when the aerosol concentration is high, due to the polymerization, the attenuation degree of small particle size is higher.While the concentration of large particle size increases, the attenuation degree is lower, and the removal efficiency of aerosol is higher. The ratio of the test bench to the prototype is 1 : 8, and the volume is about 180 m³ in the containment. This large space aerosol experimental study has certain engineering verification value.The test data provide technical support for software development and design safety evaluation.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 1","pages":"Pages 7-12"},"PeriodicalIF":0.0,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981528","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Study on the radioactive sludge retrieval technology","authors":"Guangkai Wang ,&nbsp;Jing Ma,&nbsp;Xin Li,&nbsp;Fang Bao,&nbsp;Tongtong Sun","doi":"10.1016/j.jandt.2025.12.001","DOIUrl":"10.1016/j.jandt.2025.12.001","url":null,"abstract":"<div><div>In this paper, through the large radioactive liquid waste tank sludge retrieval engineering requirements, tank bottom sludge back technology status, combined with the large radioactive liquid waste tank retired technical difficulties, through analysis, determine the process route, and the key technology problems calculation and analysis. The paper verifies the sludge retrieval operation process parameters, gives the best process operation parameters, solves the key problems and technical difficulties of the sludge retrieval operation, and puts forward corresponding improvement suggestions for the system, which is of guiding significance for the sludge retrieval project of large radioactive liquid waste storage tank in the future. At the same time, the possible waste minimization measures in the operation process are also proposed. The test and verification results show that the performance of the sludge retrieval robot equipment developed in this subject meets the requirements, the process parameters are reasonable, and the sludge retrieval can be continuously and stably conveyed with a small amount of sludge retrieval. In this study, for the first time, the crawler chassis equipped with two robotic arms (retrieval robot) is applied to the sludge retrieval operation. The high pressure water jet mixing slurry method is adopted to achieve stable sludge transportation, which adapts to the needs of actual engineering conditions. The technology has been appraised and reached the international advanced level.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 1","pages":"Pages 13-20"},"PeriodicalIF":0.0,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981529","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental research on siphon break characteristics in pool-type reactor","authors":"Xiaohui Gao ,&nbsp;Xinzhuang Wu ,&nbsp;Jingtian Yuan ,&nbsp;Lijun Zhang ,&nbsp;Zhicong Yuan ,&nbsp;Kai Zhao ,&nbsp;Daotong Chong","doi":"10.1016/j.jandt.2025.09.001","DOIUrl":"10.1016/j.jandt.2025.09.001","url":null,"abstract":"<div><div>In order to ensure that the core of a pool-type reactor is submerged after a loss of coolant accident, a siphon break pipeline needs to be designed to prevent the continuous loss of pool water due to siphon effect. This article conducted experimental research on the siphon break phenomenon of a certain pool reactor using engineering prototype dimensions, the siphon-breaking process was verified, and obtained the pressures at the connection of the siphon break pipeline and the main water tank liquid levels under different flow rates were obtained, and also the siphon break characteristics under different siphon break valve openings were obtained. The experimental results show that the larger the Froude number (<em>Fr</em>), the lower the pressure at the connection of the siphon break pipeline, and the lower the stable liquid level of the main water tank after breaking the siphon. The resistance of the siphon break pipeline is an important factor affecting the characteristics of siphon breaking. The greater the resistance, the less air intake, and the greater the difficulty of bubble aggregation, making it more difficult to break the siphon. Moreover, for the fully open condition of the siphon break valve, experimental correlations were established between the dimensionless minimum pressure at the connection and <em>Fr</em>, as well as between the dimensionless stable liquid level and <em>Fr</em>. The results show that the experimental correlations could predict the minimum pressure and stable liquid level well, with deviations within ±0.5 %.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 4","pages":"Pages 382-387"},"PeriodicalIF":0.0,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145159692","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Implementation of Leak Before Break (LBB) criteria in fast spectrum reactor","authors":"Sitangshu Sekhar Biswas ,&nbsp;Soham Banerjee ,&nbsp;S. Kudiyarasan ,&nbsp;Sudip Ghosh ,&nbsp;Prabhat Kumar","doi":"10.1016/j.jandt.2025.10.004","DOIUrl":"10.1016/j.jandt.2025.10.004","url":null,"abstract":"<div><div>Sodium is employed as a coolant in the primary and secondary loops of fast reactors to transfer heat from the reactor and deliver it to the expander. However, handling sodium poses challenges and can be hazardous, as it reacts exothermically when exposed to air. Again, though there are radioactive barriers, there will be some radiation in the liquid coolant which makes it even more hazardous. Generally, P91 (9 % Cr and 1 % Mo) material is used in sodium-cooled nuclear reactors for its high-temperature strength, excellent creep resistance, and good corrosion and oxidation resistance, making it ideal for the demanding environment within the reactor and coolant line (Kudiyarasan and Biswas, 2024). Although, fast reactors operate at a lower pressure (1̴2 MPa) compared to the pressurized water reactors (5̴10 MPa), the operating temperature in the coolant line for fast reactors being very high (500 ֯ C̴800 ֯ C) unlike PWRs (̴300 ֯ C), there is a risk of brittle failure, due to factors such as material embrittlement from radiation exposure or stress concentrations, due to environmental interactions or a combination of different hydraulic transients (i.e. water hammering, thermal hammering or creep etc.) that can lead to sudden catastrophic failure under certain conditions. After thermal hammering, the main reason of hydraulic transients in fast spectrum reactor is found to be “Water Hammering”, which is nothing but sudden rise in pressure and propagation of pressure waves (shockwaves) resulting in high stress generation in the pipes. There are two primary reasons for “Water Hammering” i.e. i) Sudden closure or opening valves and ii) Due to sudden starting and stopping of pumps. In our paper, we have done numerical simulation on these hydraulic transient cases and found out the necessity to implement Griffith's criteria on designing the coolant line. Based on Griffith's LBB criteria and considering hydraulic transient scenario, we designed pipes for two different cases to find out the applicability of LBB criteria in transient case, experimentally and found out the criteria to be quite suitable to tackle this kind of problems in practical scenarios. Therefore, it seems obvious to always design the coolant line based on Griffith's criteria, so that leakage would occur before the complete failure takes place and this would give us ample time to successfully repair the coolant line before any catastrophic failure happens.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 4","pages":"Pages 423-435"},"PeriodicalIF":0.0,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145363232","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}