Nuclear Engineering and Design最新文献

{"title":"PIV measurements of natural circulation driven flow inside and around SMR fuel assemblies","authors":"Gergely Imre Orosz,&nbsp;Levente Schaul,&nbsp;Bence Barnabás Mészáros,&nbsp;Dániel Kacz,&nbsp;Attila Aszódi","doi":"10.1016/j.nucengdes.2025.114475","DOIUrl":"10.1016/j.nucengdes.2025.114475","url":null,"abstract":"<div><div>Buoyancy-driven circulation has gained popularity in water-cooled Small Modular Reactor (SMR) designs, serving as a means to facilitate heat removal from the fuel assemblies in normal operation conditions. Natural convection cooling has found application not only in SMRs but also in numerous pool-type research reactors like the Training Reactor in operation at Budapest University of Technology and Economics (BME) in Hungary. Heat transfer efficiency depends on flow conditions, so high-accuracy models are needed to predict coolant behaviour around fuel rods. The nuclear industry uses many Computational Fluid Dynamics (CFD) codes for this. However, as in the case of any numerical simulations, the need for experimental validation remains essential. The Particle Image Velocimetry (PIV) method provides a high-accuracy solution to measure flow characteristics in at least two spatial dimensions without intervening in the flow. To facilitate the investigation of thermal-hydraulics inside and around fuel pin bundles of BME’s Training Reactor (EK-10 type fuel) and also of future Small Modular Reactors, a new equipment has been designed and constructed in the PIV Laboratory at the Institute of Nuclear Techniques at BME. The Transparent mOdel for Water-coolEd Reactors (TOWER) system includes a single full scale electrically heated fuel assembly model, comprising 16 fuel rods arranged in a 4x4 square lattice, with adjustable power outputs. PIV measurements have been performed at four different power levels focusing on the flow behaviour inside subchannels and right at the fuel assembly outlet where the buoyancy-induced free jet of lower density warm water exits the bundle.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114475"},"PeriodicalIF":2.1,"publicationDate":"2025-09-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145119070","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":"Effect of neutron fluence on the safety of the RPV of the VVER 1200","authors":"K M Rakib Al Hasan ,&nbsp;Md.Imtiaj Hossain ,&nbsp;Md.Shafiqul Islam","doi":"10.1016/j.nucengdes.2025.114478","DOIUrl":"10.1016/j.nucengdes.2025.114478","url":null,"abstract":"<div><div>The Reactor Pressure Vessel (RPV) is the most critical safety component in a nuclear power plant (NPP), especially in light-water reactor (LWR) designs. It is subjected to continuous neutron irradiation, extreme coolant pressure, and temperature during normal operation. Neutron irradiation of the RPV degrades its mechanical properties, highlighting the necessity of accurately predicting embrittlement phenomena to prevent brittle failure. Although the RPV of VVER-1200 is manufactured using advanced steel grades that have been developed and optimized through decades of experience with earlier VVER designs, such as the VVER-230 and VVER-440, it has a short operational history. As a result, there is still limited long-term surveillance data. Therefore, continued research and modeling are essential to reliably predict the RPV’s lifetime and ensure its structural integrity under extended service conditions. This work offers a unique investigation by combining the full-core neutronic behavior of the VVER-1200 and its impact on the RPV lifetime using OpenMC simulation. In this study, a VVER-1200 full core model is created to evaluate the fast neutron flux (&gt;0.5 MeV) spectrum . The analysis reveals a maximum fast neutron fluence of 4.11 × 10¹⁹ neutrons/cm² over 60 years of operation, resulting in a ductile to brittle transition temperature (DBTT) shift of about 68 °C for the RPV base metal and 69 °C for the weld metal, impacting RPV safety and longevity. Burnup analysis shows an initial k_eff &gt; 1.2 with fresh fuel, decreasing as fissile isotopes deplete over time. After 60 years operation, the displacement per atom (DPA) reaches 0.021, with the base metal’s fracture toughness (<span><math><mrow><msub><mi>K</mi><mrow><mi>I</mi><mi>C</mi></mrow></msub><mrow><mo>)</mo></mrow></mrow></math></span> decreasing to 85 MPa√m and the weld metal dropping from 136 MPa√m to 87 MPa√m. The DBTT shift is also influenced by impurity concentrations, with Cu at 0.30 %, Ni at 1.3 %, and P at 0.020 % showing the most significant shifts under the same neutron fluence. The results can be used for aging management of the RPV of a VVER-1200.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114478"},"PeriodicalIF":2.1,"publicationDate":"2025-09-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145119081","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":"Counterpart testing on the SBLOCA scenario using two integral effect test facilities for the integral type reactor SMART","authors":"Byong Guk Jeon,&nbsp;Jin-Hwa Yang,&nbsp;Hwang Bae,&nbsp;Hyun-Sik Park","doi":"10.1016/j.nucengdes.2025.114473","DOIUrl":"10.1016/j.nucengdes.2025.114473","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114473"},"PeriodicalIF":2.1,"publicationDate":"2025-09-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145119071","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 thermal hydraulic analysis of supercritical natural circulation loop","authors":"A.K. Vias ,&nbsp;V.K. Garg ,&nbsp;P.K. Vijayan ,&nbsp;G. Dutta","doi":"10.1016/j.nucengdes.2025.114457","DOIUrl":"10.1016/j.nucengdes.2025.114457","url":null,"abstract":"<div><div>In the present numerical study, a supercritical natural circulation loop (SCNCL) with a closed configuration is analyzed from a thermal-hydraulic (TH) perspective. An in-house model is developed to capture the axial variation of TH field variables considering a single channel in the heated section. This TH model efficiently accounts for local property variations at supercritical pressures and is integrated with models for the pressurizer, cooling heat exchanger (CHX), and wall heat conduction to accurately represent the behavior of a closed SCNCL. To evaluate the predictive capability of the integrated TH model, validation is performed against both experimental data and numerical results available in the literature, covering steady state and transient scenarios. Steady state simulations are then conducted to assess the SCNCL performance and determine the mass flow rate under various operating conditions, with an emphasis on identifying the underlying physical mechanisms. Finally, transient simulations are carried out to investigate the risk of density wave oscillations (DWOs) and to determine the marginal stability boundary (MSB). A comprehensive parametric study is performed to explore the influence of different factors on the MSB.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114457"},"PeriodicalIF":2.1,"publicationDate":"2025-09-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097501","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":"Dynamic response and failure mechanism of steel-UHPC-steel nuclear containment impacted by the large commercial aircraft through a numerical approach","authors":"Yilinke Tan ,&nbsp;Zefang Wang ,&nbsp;Jun Gong ,&nbsp;Zhenyu Cheng ,&nbsp;Feng Fan","doi":"10.1016/j.nucengdes.2025.114477","DOIUrl":"10.1016/j.nucengdes.2025.114477","url":null,"abstract":"<div><div>This study proposed, for the first time, a single-layer nuclear containment structure utilizing a steel–concrete–steel (SCS) sandwich configuration. Ultra-high performance concrete (UHPC) with steel fiber reinforcement is selected as the core material, capitalizing on its superior ductility and high strength. The dynamic response and failure mechanisms of steel-UHPC-steel (SUHPCS) containment under high-velocity aircraft impact are systematically investigated. A refined finite element (FE) model of a Boeing 747-400 airliner and SUHPCS containment is developed through Abaqus/Explicit, employing an efficient coupled simulation approach to accurately capture the interaction between the airliner and the containment. The effects of core concrete, steel plate thickness, and tie rod spacing are examined on the impact resistance of SUHPCS containment. Results demonstrate that the SUHPCS containment effectively mitigates concrete fragmentation and has excellent shielding ability, owing to the membrane effect of steel plates and superior energy absorption of UHPC reinforced with steel fibers. The core concrete thickness significantly enhances impact resistance, altering failure modes from perforation to penetration, while steel plate thickness and tie rod spacing show more limited influence. A theoretically calculated method for critical penetration energy is proposed, providing a basis for designing SUHPCS containment to withstand airliner impacts. This work offers valuable insights and practical recommendations for enhancing the safety and resilience of nuclear containment structures under extreme impact loading conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114477"},"PeriodicalIF":2.1,"publicationDate":"2025-09-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097506","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":"Effect of brazing parameter and Zr-Cu-Ti powder filler metal on the microstructure and mechanical properties of SiC ceramic joints","authors":"Bofang Zhou ,&nbsp;Zixuan Leng ,&nbsp;Wuman Wang ,&nbsp;Zhaojie Zhang ,&nbsp;Hongxia Zhang","doi":"10.1016/j.nucengdes.2025.114456","DOIUrl":"10.1016/j.nucengdes.2025.114456","url":null,"abstract":"<div><div>Brazing parameters (temperature, holding time) and Ti content in Zr-Cu-Ti filler metal for brazing SiC ceramic are investigated in this paper. Successful bonding was achieved at 950∼1150 °C for 20 min using 5 wt% Ti, forming ZrC and Zr₂Si phases in the reaction layer, and the shear strength peaked at 63 MPa (1050 °C) before declining, fracturing at the interface between the reaction layer and the filler metal. At 1050 °C, reaction layer thickness increased with holding time (5∼45 min), reaching optimal strength (1.4 μm thickness at 20 min). The Ti₅Si₃ phase emerged in the reaction layer on 0∼15 wt% Ti, reducing the shear strength of the joint due to thermal expansion mismatch with Zr₂Si and ZrC. The shear strength first increased then decreased with Ti content, peaking at 5 wt% Ti.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114456"},"PeriodicalIF":2.1,"publicationDate":"2025-09-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097495","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":"Monte Carlo track-chemistry simulations of fast neutron radiolysis in supercritical water at 400–600 °C and 25 MPa","authors":"Md Shakhawat Hossen Bhuiyan,&nbsp;Jintana Meesungnoen,&nbsp;Abida Sultana,&nbsp;Jean-Paul Jay-Gerin","doi":"10.1016/j.nucengdes.2025.114479","DOIUrl":"10.1016/j.nucengdes.2025.114479","url":null,"abstract":"<div><div>Understanding the radiation chemistry and behavior of transient species in supercritical water-cooled reactors (SCWRs), including small modular variants (SCW-SMRs), is essential for evaluating corrosion risks. Operating beyond water’s critical point, these reactors encounter unique challenges as intense radiation alters coolant chemistry and threatens material integrity. Here, we employ Monte Carlo track-chemistry simulations to quantify the radiolytic yields (<em>G</em> values) of e⁻_aq, <strong>^•</strong>OH, H<strong>^•</strong>, H₂, H₂O₂, H₃O⁺, and OH⁻ in SCWR and SCW-SMR coolants exposed to 2-MeV fast neutrons at 400–600 °C and 25 MPa. Calculations track the first three elastically scattered recoil protons, with initial energies of 1.264, 0.465, and 0.171 MeV, from ∼1 ps to 100 μs. The temporal profiles of our calculated yields resemble those reported for low-linear-energy-transfer (LET) radiation, such as ⁶⁰Co γ rays, fast electrons, or 300-MeV protons. However, under fast neutron irradiation, charge recombination between e⁻_aq and H₃O⁺ within spurs or tracks is markedly enhanced, reflecting the high-LET nature of neutrons. In the homogeneous chemical stage of radiolysis, simulations reveal a pronounced rise in <em>G</em>(<strong>^•</strong>OH) and <em>G</em>(H₂) alongside a reduction in <em>G</em>(H<strong>^•</strong>). This behavior is driven by H<strong>^•</strong> + H₂O → <strong>^•</strong>OH + H₂, a key pathway for H₂ formation that may help mitigate net water radiolysis and reduce corrosion. Under supercritical conditions, the very low <em>G</em>(H₂O₂) indicates that H₂O₂ likely contributes little to material oxidation. Furthermore, in situ H₃O⁺ formation by recoil-proton irradiation transiently acidifies native track regions, suggesting that this localized acidity could promote corrosion. Overall, these results provide critical insights into the radiolytic processes in SCWRs and SCW-SMRs, informing strategies for optimized water-chemistry control and enhanced material protection.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114479"},"PeriodicalIF":2.1,"publicationDate":"2025-09-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097505","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":"Size distributions and dynamic characteristics of droplets in gas–liquid annular flow","authors":"Ri Zhang ,&nbsp;Yuntao Hu ,&nbsp;Zhen Yan","doi":"10.1016/j.nucengdes.2025.114482","DOIUrl":"10.1016/j.nucengdes.2025.114482","url":null,"abstract":"<div><div>Droplet breakup and coalescence models are integrated into the previously developed thin liquid film method (TLFM) to investigate in detail the evolution of the droplet population and its dynamic behavior. Validation against experimental data demonstrates that the improved TLFM accurately predicts macroscopic parameters and effectively captures the microscopic dynamics of gas–liquid annular flow. During the evolution of droplet population, atomization and deposition dominate droplet generation and extinction, respectively. Breakup and coalescence primarily influence the variation in the number of small droplets. Due to a sorting and screening mechanism within the gas core, large droplets move at lower velocities near its periphery, while small droplets travel faster near the center. The droplet fluctuation velocities in the axial and transverse directions follow Gaussian distributions with a mean of zero, with the axial fluctuation velocity exhibiting significantly greater variance than those in the transverse directions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114482"},"PeriodicalIF":2.1,"publicationDate":"2025-09-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097507","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":"Conditions for bubble recirculation in a molten salt natural circulation loop and heater failure effect on loop conditions and argon bubble dynamics","authors":"Thomas Carson ,&nbsp;Hansol Kim ,&nbsp;Joseph Seo ,&nbsp;Yassin Hassan","doi":"10.1016/j.nucengdes.2025.114449","DOIUrl":"10.1016/j.nucengdes.2025.114449","url":null,"abstract":"<div><div>Off-gas systems employing bubble injection have been proposed as an effective method for removing gaseous fission products in molten salt reactors (MSRs). In a semi-closed loop, bubbles may remain entrained in the fluid even after the gas has passed the point where the bubble should escape the fluid, and then recirculates through the loop. Bubble recirculation poses risks to reactivity control and flow stability in the operation of the off-gas systems of MSRs. The longer the bubble is in the reactor the higher the concentration of fission products in the bubble. This can cause changes in local reactivity and power profiles. This study experimentally investigates the mechanisms behind bubble recirculation in a natural circulation molten salt loop under degraded thermal conditions. Advanced flow diagnostics such as Particle image velocimetry (PIV), particle tracking velocimetry (PTV), and proper orthogonal decomposition (POD) were employed to characterize bubble trajectories, slip velocities, and associated flow structures. A downcomer heater failure created asymmetric heating, leading to partial salt freezing and flow stagnation, while the blocked expansion tank prevented bubble venting. Over time, recirculating bubbles coalesced, increasing in size and altering local flow behavior. Measured slip velocities diverge significantly from drift-flux model predictions, underscoring the limitations under off-normal conditions. POD analysis revealed dominant wake structures and vortex interactions, and temporal cross-correlations revealed a sequence of flow phenomena induced by bubble motion. These findings underscore the need to maintain thermal uniformity and suggest refinements to computational models of off-gas systems in MSRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114449"},"PeriodicalIF":2.1,"publicationDate":"2025-09-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097504","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":"Reliability assessment and maintenance planning of air-operated control valves used in a nuclear power plant","authors":"Guru Prakash ,&nbsp;Mahesh D. Pandey","doi":"10.1016/j.nucengdes.2025.114472","DOIUrl":"10.1016/j.nucengdes.2025.114472","url":null,"abstract":"<div><div>Air-operated valves (AOVs) are used to control and isolate different process systems in a nuclear power plant. The performance of an AOV is adversely affected by degradation in the pneumatic actuator and other control components of the valve. Therefore, valve overhauls (OH) are performed at a fixed interval that is mainly based on the vendor’s recommendations and the experience of the plant personnel. This paper presents a probabilistic approach to assess valve reliability and uses it as a basis to plan maintenance. The paper presents a Bayesian approach to model the valve lifetime distribution and update its parameters based on the in-service data available from the plant. Using this model, the valve OH interval is determined to achieve a target reliability level over the interval. A practical case study is presented that utilizes maintenance data from a fleet of 32-level control valves connected to steam generators at a Canadian nuclear power station. The proposed approach demonstrates that the satisfactory functioning of an in-service valve provides valuable information for extending its OH interval. Thus, the proposed approach can significantly improve the efficiency of the valve maintenance program.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114472"},"PeriodicalIF":2.1,"publicationDate":"2025-09-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097497","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}