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

{"title":"Development of coupled analysis system for PWR design basis accidents using RETAC and ATHROC","authors":"Jian Wu,&nbsp;Shihao Wu,&nbsp;Jingchi Yang,&nbsp;Yongkang Li,&nbsp;Huifang Zhang","doi":"10.1016/j.jandt.2026.08.004","DOIUrl":"10.1016/j.jandt.2026.08.004","url":null,"abstract":"<div><div>A coupled analysis framework based on RETAC and ATHROC was developed to support the safety evaluation of pressurized water reactors under design-basis accident conditions. The framework considers two coupling interfaces: one between the primary coolant system and the containment, and the other between the safety injection system and the primary system. Through data exchange between the two codes, the transient behavior of the reactor coolant system and the containment can be evaluated in an integrated manner. The proposed RETAC–ATHROC coupling scheme was applied to a CPR1000 small-break loss-of-coolant accident accompanied by auxiliary feedwater failure. The results show that the coupled calculation preserves the main thermal-hydraulic response of the primary coolant system predicted by RETAC, while the containment pressure, gas temperature, refueling water storage tank level, and sump water level follow trends consistent with ATHROC predictions. These results indicate that the coupled program can reasonably reproduce the interaction between the primary system and the containment during accident progression. Therefore, the developed method provides a useful tool for integrated safety analysis of PWR plants and supports transient thermal-hydraulic evaluation under coupled system conditions.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 3","pages":"Pages 403-409"},"PeriodicalIF":0.0,"publicationDate":"2026-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"148935052","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Research on verification benchmarks for reactor thermal-hydraulic system analysis software","authors":"Xinli Gao,&nbsp;Guangyu Zhu,&nbsp;Pingfei Du,&nbsp;Ke Wang,&nbsp;Yiran Chen,&nbsp;Jiaxu Zuo,&nbsp;Hao Zhang,&nbsp;Tao Huang,&nbsp;Kui Zhang,&nbsp;Ronghua Chen,&nbsp;Caihong Xu","doi":"10.1016/j.jandt.2026.08.003","DOIUrl":"10.1016/j.jandt.2026.08.003","url":null,"abstract":"<div><div>Verification serves as the prerequisite for the engineering application of independently developed nuclear safety analysis software. Verification methods for reactor thermal-hydraulic system analysis software are studied, and the framework and workflow for verification were proposed. Based on the development and application of reactor thermal-hydraulic numerical algorithms, nine benchmark problems were identified and selected. Multiple nuclear power research institutions of China were invited to perform comparative calculations using independently developed software. Analysis of the computational results revealed the influence of factors such as boundary conditions, numerical algorithms, interphase friction, and virtual mass force on software verification results. Consequently, the boundary conditions and application requirements for the benchmark problems were standardized, the comparison and acceptance standards for the benchmark problems were unified, and the standards framework have been established for the verification work of China's independent developed reactor thermal-hydraulic system analysis software.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 3","pages":"Pages 380-387"},"PeriodicalIF":0.0,"publicationDate":"2026-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"148854249","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Simulation and analysis of typical operational conditions in a PWR based on Modelica","authors":"Zhibo Gao,&nbsp;Ziao Xiang,&nbsp;Di Wu,&nbsp;Jie Cheng,&nbsp;Jianjun Wang","doi":"10.1016/j.jandt.2026.08.007","DOIUrl":"10.1016/j.jandt.2026.08.007","url":null,"abstract":"<div><div>With advantages such as rapid and simplified modeling and better compatibility with emerging technologies, Modelica is one of the key technical pathways driving the digitalization of future nuclear power systems. Based on the Modelica, the primary loop, secondary loop, and residual heat removal system of a typical pressurized water reactor were modeled in this study. By analyzing simulation results under steady-state and transient conditions, the accuracy of the model and its feasibility for application in the nuclear energy field were evaluated. The results show that under steady-state conditions, with the reactor-following-turbine control strategy, the deviation between the predicted values of the established simulation model and the design values is within ±3%. During transient conditions, the simulation results accurately reflect the operational trends of the system. This study can provide references for the modeling and simulation analysis of nuclear energy systems.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 3","pages":"Pages 395-402"},"PeriodicalIF":0.0,"publicationDate":"2026-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"148854250","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Analysis of the effectiveness of in-vessel retention strategy under a three-layer corium pool configuration","authors":"Guangyu Zhu,&nbsp;Pingfei Du,&nbsp;Jipu Wang,&nbsp;Xinli Gao,&nbsp;Kunpeng Wang","doi":"10.1016/j.jandt.2026.08.006","DOIUrl":"10.1016/j.jandt.2026.08.006","url":null,"abstract":"<div><div>To cope with the corium cooling issue in severe PWR accidents, Generation III reactors commonly employ the IVR (in-vessel retention) approach, which involves flooding the reactor cavity. The stratification of different melted core internals influences the temperature field and velocity field of the corium pool, thereby affecting the distribution of heat load on the reactor pressure vessel steel shell. The success criterion for the IVR strategy in present PWR safety assessments is that the local heat flux acting on the RPV wall from the molten corium must stay below the lower-head boiling CHF. In this paper, a CFD model is built to simulate the three-layer condition composed of metal and oxide, providing a basis for the further study of thermal-hydraulic phenomena in the corium pool and the integrity of the RPV lower head. Simulation findings indicate that adequate external cooling of the RPV vessel wall ensures the retention of lower head structural integrity under severe accident conditions.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 3","pages":"Pages 388-394"},"PeriodicalIF":0.0,"publicationDate":"2026-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"148854251","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental investigation of the spatial evolution of local flow parameters in vertical LBE–N2 bubbly flow","authors":"Zihao Zhu ,&nbsp;Bin Li ,&nbsp;Bin Hou ,&nbsp;Lina Zhu ,&nbsp;Kefan Geng ,&nbsp;Bo Pang ,&nbsp;Yangbin Deng ,&nbsp;Xi Huang","doi":"10.1016/j.jandt.2026.05.005","DOIUrl":"10.1016/j.jandt.2026.05.005","url":null,"abstract":"<div><div>To support closure-model development for thermal-hydraulic analysis of SGTR/HXTR-related two-phase flow in lead-cooled fast reactors, a separate-effect experimental study was conducted on lead–bismuth eutectic (LBE)–nitrogen bubbly flow in a vertical circular pipe using a loop-type high-temperature liquid-metal two-phase-flow facility. The test section was a vertical pipe with an inner diameter of 50 mm and a length of 2000 mm. A single-sensor conductivity probe and four-sensor conductivity probes were employed for different measurement objectives. The single-sensor probe was used mainly to examine local void-fraction trends in the upper developed section, whereas the four-sensor probes were used to obtain local void fraction, bubble frequency, bubble velocity, equivalent bubble radius, and interfacial area concentration (IAC) at two representative axial sections and multiple radial positions.</div><div>The measurements indicate that the void fraction exhibits clear spatial evolution along both the radial and axial directions. The radial void-fraction profile gradually changes from a relatively flat distribution in the near-inlet region to a center-peaked distribution at the upper measurement section. With increasing superficial gas velocity, both the local void fraction and the three-point area-weighted estimate of the cross-sectional void fraction increase. At relatively high superficial liquid velocity, the estimated cross-sectional void fraction tends to decrease because of the shortened gas residence time. Bubble velocity and equivalent bubble radius generally decrease along the flow direction, while the IAC varies in close correspondence with bubble frequency.</div><div>These results provide a local-parameter database for the validation and improvement of interphase-interaction and interfacial-area closure models in LBE two-phase-flow simulations. The present study focuses on dispersed gas transport and bubbly-flow evolution in a high-density liquid-metal system, and should therefore be interpreted as a separate-effect bubbly-flow study relevant to SGTR/HXTR analysis rather than as a full reproduction of prototypical tube-rupture transients involving high-pressure steam release, condensation, flashing, and thermal shock.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 2","pages":"Pages 252-270"},"PeriodicalIF":0.0,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"148185088","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":"Multiphysics study on phase separation and thermal aggregation effects in the melt pool of the reactor lower head under severe accidents","authors":"Zongyan Bai ,&nbsp;Tao Zhou ,&nbsp;Mubashar Zaman ,&nbsp;Yao Yao ,&nbsp;Peng Liu","doi":"10.1016/j.jandt.2026.05.004","DOIUrl":"10.1016/j.jandt.2026.05.004","url":null,"abstract":"<div><div>In the progression of severe accidents in light water reactors, the retention of core melt in the reactor pressure vessel (RPV) lower head, known as In-Vessel Retention (IVR), constitutes a core strategy for ensuring nuclear power plant safety and preventing large-scale release of radioactive materials. Targeting the complex multi-component liquid-liquid phase separation phenomena within the melt pool and the resulting “thermal focusing effect,” a coupled phase-field-flow-heat transfer multiphysics model based on free energy functionals was established. By simultaneously solving the Cahn-Hilliard equation, incompressible Navier-Stokes equations, and energy conservation equation in a two-dimensional axisymmetric domain, the dynamic evolution of a U–Zr–Fe–O quaternary system from an initially random uniform mixture to a macroscopic stratified state within the RPV lower head was simulated. Results demonstrate that the melt pool, driven by decay heat and interfacial energy, undergoes a “labyrinthine” bicontinuous structure evolution initiated by spinodal decomposition, ultimately forming a stable bilayer structure with a light metal layer above and a heavy oxide layer below under gravitational effects. The flow field evolves from initial small-scale chaotic convection (maximum velocity 0.0541 m/s) to large-scale circulation driven by thermal buoyancy (maximum velocity 0.00192 m/s), forming a stable configuration with an active upper convection zone and a stagnant lower dead zone. Due to the significant thermal conductivity contrast between the two stratified layers, wall heat flux exhibits a notable increase at the metal-oxide interface region, with peak heat flux reaching approximately 0.9 × 10⁶ W/m² during mid-evolution, attaining 60% of the upper CHF limit from ULPU experiments, posing a potential threat to IVR safety margins. Quantitative analysis of the phase separation time scale and its impact on wall thermal loading was conducted, providing numerical evidence and theoretical support for IVR safety margin assessment.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 2","pages":"Pages 207-217"},"PeriodicalIF":0.0,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"148185090","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":"Heat transfer characteristics of two-layer corium pool with additional metallic layer heat source","authors":"Kui Ge ,&nbsp;Jiaxin Wu ,&nbsp;Xiaowei Jiang ,&nbsp;Yapei Zhang ,&nbsp;Wenxi Tian ,&nbsp;Shihao Wu ,&nbsp;Jun Xu ,&nbsp;Jiaxin Zhang ,&nbsp;Dahuan Zhu ,&nbsp;Hongxing Yu","doi":"10.1016/j.jandt.2026.05.006","DOIUrl":"10.1016/j.jandt.2026.05.006","url":null,"abstract":"<div><div>Due to material oxidation, activation, and the presence of fission products and uranium, the metallic layer in a two-layer corium pool contains internal heat sources. This study investigated the influence of these additional heat sources on heat transfer characteristics using 3D CFD simulations, analyzing temperature distributions and wall heat flux patterns. The results reveal that the thickest solidified layer typically forms in the upper-middle section of the oxide layer rather than the cooler lower region. Beyond a certain height, the solidified layer rapidly diminishes, leading to significant wall erosion. While the Nu relationship established for conditions without metallic layer heat sources remains applicable when such sources are present, parameter c substantially impacts the metallic layer's heat transfer behavior. However, heat transfer correlations like the Nu relationship developed for heat-source-free conditions may not be valid when heat sources exist. Even experimental data from studies focusing solely on metallic layer with internal heat sources cannot be directly applied to analyze heat transfer in two-layer corium pools. The coefficient f was proposed to quantify the effect of metallic layer heat sources on heat transfer characteristics.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 2","pages":"Pages 230-237"},"PeriodicalIF":0.0,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"148185086","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":"Advanced UZrH research reactor design, part I: thermal-hydraulic review","authors":"Qifan Yu,&nbsp;Ersheng You,&nbsp;Yiyi Li,&nbsp;Dianchuan Xing,&nbsp;Tianzhou Xie,&nbsp;Liangzi Wang,&nbsp;Lanjiang Zheng,&nbsp;Xianjun Zhang,&nbsp;Zhike Lan","doi":"10.1016/j.jandt.2026.05.002","DOIUrl":"10.1016/j.jandt.2026.05.002","url":null,"abstract":"<div><div>Uranium–Zirconium Hydride (UZrH) research reactor is a type of research reactor that uses UZrH fuel as its core fuel. This reactor type typically employs a pool-type coolant system design, offering advantages such as high inherent safety, versatile applications, simple structure, and flexible operational modes. Internationally, there are numerous design schemes for UZrH research reactors with pool-type configurations, exhibiting significant variations in power levels. This paper reviews the existing coolant systems and accident mitigation measures of pool-type research reactors worldwide. Additionally, it introduces established research reactor designs that incorporate central irradiation channels with dedicated test loops. This paper recommended that the 18 MW UZrH research reactor should adopt a pool-type forced circulation coolant system, and complemented by a passive natural circulation accident mitigation measure. Furthermore, the design should include a central irradiation channel with an independent test loop for neutron flux focusing.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 2","pages":"Pages 218-229"},"PeriodicalIF":0.0,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"148185085","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 response characteristics of internal compartments in the containment building under LOCA conditions","authors":"Shilei Dun ,&nbsp;Tao Zhou ,&nbsp;Haoyv Li ,&nbsp;Lanyv Zhou ,&nbsp;Yao Yao ,&nbsp;Changhong Peng ,&nbsp;Shuyv He","doi":"10.1016/j.jandt.2026.04.003","DOIUrl":"10.1016/j.jandt.2026.04.003","url":null,"abstract":"&lt;div&gt;&lt;div&gt;Investigating the thermal-hydraulic response characteristics of internal compartments within the containment structure during LOCA accidents holds significant scientific, engineering, and safety value for elucidating accident evolution mechanisms, evaluating compartment structural loads, and optimizing containment design. During LOCA events, high-energy fluids released through breaches create highly non-uniform velocity, Distribution of Water Vapor, temperature, and pressure distributions within the complex multi-compartment space of the containment structure. This generates intricate dynamic structural loads, making the investigation of these non-uniform response characteristics crucial for assessing and ensuring containment integrity. Taking the Hualong One containment as the research subject, a three-dimensional computational model incorporating the main compartments and connecting passages was established. Transient numerical simulations were conducted for four accident scenarios: passive cooling system with cold-end breach closed, passive cooling system with hot-end breach closed, passive cooling system with cold-end breach open, and passive cooling system with hot-end breach open. The analysis examines the evolution of velocity fields, temperature distributions, temperatures in the breach area, pressures in the breach area, and pressure differentials between the breached compartment and adjacent compartments during the initial high-velocity jet and backflow formation. Regarding velocity, high-velocity zones primarily concentrate near the breach location, forming high-flow regions within the pressure vessel compartment and the top of the containment structure. After activating the passive cooling system, the velocity of moist air within the large upper containment space decreases, while the velocity of moist air within the pressure vessel compartment increases.Regarding water vapor, the PCCS effectively reduces the vapor content within the containment structure through wall condensation, thereby suppressing the rate of temperature rise and pressure increase. For hot-end breaches, where steam locally accumulates, the PCCS directly weakens the pressure differential driving force through immediate condensation near the breach, yielding a slightly higher suppression effect than for cold-end breaches.Regarding temperature, the side of the main pump compartment adjacent to the breach exhibits higher temperatures. Due to insufficient coolant, the pressure vessel experiences a rapid temperature rise, resulting in the highest temperatures in the central region. Temperatures decrease progressively from top to bottom within the large space. The temperature trends in the breach area are similar across the four accident scenarios: a rapid initial increase followed by relative stability. The final temperature in the hot-end breach area is higher than that in the cold-end breach. After activating the passive cooling system, temperature conditions","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 2","pages":"Pages 238-251"},"PeriodicalIF":0.0,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"148185089","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":"Method for measuring the adsorbed radioactivity source term of fission nuclide 131I in lead–bismuth reactors","authors":"Guobing Yu ,&nbsp;Feng Xu ,&nbsp;Xianbao Gu ,&nbsp;Deyun Wen ,&nbsp;Jinnong Wang ,&nbsp;Hao Wang","doi":"10.1016/j.jandt.2026.01.004","DOIUrl":"10.1016/j.jandt.2026.01.004","url":null,"abstract":"<div><div>¹³¹I is a typical fission product in nuclear reactors. It has a strong affinity for the human thyroid gland and is an important nuclide of concern in radiation protection under both normal and accident conditions of reactors. Accurate measurement of its radioactivity is of great significance. The main measurement method involves collecting ¹³¹I in aerosols using an activated carbon adsorption device and then analyzing it with a γ spectrometer. This study proposes a bidirectional ratio method to obtain the γ detection efficiency through integrating the depth distribution coefficient of ¹³¹I in activated carbon. Compared with the traditional methods of preparing standard uniform volume sources or using destructive analysis, this experimental method is more reasonable and reliable. In addition, to verify the reliability of this method, experimental verification of the proportional coefficient of bidirectional measurement was conducted, and experimental work on the correlation analysis between sampling flow rate and distribution coefficient was provided. The experiments show that the detection efficiency of the iodine cartridge source with a real depth distribution coefficient obtained by the volume source integration method incorporating a probability function can quickly and conveniently obtain the γ detection efficiency of non-uniformly distributed iodine cartridge sources, providing a feasible approach for accurately measuring the radioactive activity of the fission product ¹³¹I in the environment.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"8 1","pages":"Pages 49-55"},"PeriodicalIF":0.0,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146090402","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}