Nuclear Engineering and Design最新文献

{"title":"Experimental investigation on interfacial area concentration in narrow-channel steam-water flow based on printed circuit sensor","authors":"Yiang Yang,&nbsp;Jinbiao Xiong","doi":"10.1016/j.nucengdes.2025.114091","DOIUrl":"10.1016/j.nucengdes.2025.114091","url":null,"abstract":"<div><div>Due to deficiency of void fraction measurement of steam-water flow in narrow channels, existing interfacial area concentration (IAC) models for narrow channels have mostly been developed based on air–water flow. The applicability of these models to steam-water flow remains uncertain. To address this deficiency, a ceramic-substrate printed circuit (CSPC) sensor is utilized to obtain the instantaneous interface topology of steam-water flow in a narrow rectangular channel. Experiments were conducted under pressures ranging from 0.2 to 0.9 MPa, covering flow patterns from bubbly to annular flow. The average bubble shape, projected interfacial area concentration (PIAC), and projected interfacial length concentration (PILC) for bubbles of different sizes were comprehensively analyzed. The critical diameter separating distorted and cap bubbles, as utilized in existing IAC models, was verified from a statistical averaging perspective. Within the experimental conditions, PIAC and PILC showed no significant variation under different pressures. The case-averaged PIAC and PILC of steam-water flow and air–water flow exhibited good agreement within the same channel geometry, although some divergence was observed for bubbles of different sizes. Among the existing IAC correlations, the Yang-Xiong model demonstrated the best performance in predicting IAC compared to the present experimental data. The relative error of the total IAC was within ± 10 % for most cases and decreased with increasing void fraction.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114091"},"PeriodicalIF":1.9,"publicationDate":"2025-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143864082","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":"Design of electrical feedouts for hard-core components in nuclear power plants","authors":"Jorge Rafael González-Teodoro ,&nbsp;Jordi Mena Bravo ,&nbsp;Francisco de Asís Salguero Rodríguez","doi":"10.1016/j.nucengdes.2025.114088","DOIUrl":"10.1016/j.nucengdes.2025.114088","url":null,"abstract":"<div><div>Hard-Core Components (HCCs) have become a key element in the defense-in-depth strategy for nuclear safety, particularly in scenarios involving extreme events such as seismic activity, fires, or total loss of power. This paper presents the conceptual design and qualification considerations for electrical feedouts—specifically, the penetrations that ensure the structural and functional integrity of power and signal transmission across containment barriers under severe conditions. The study emphasizes innovative design features such as bellows, service loops, and advanced fire barriers, which minimize the mechanical stress transmitted to containment walls and enhance system resilience. Through a combination of mechanical, thermal, and seismic design strategies, the proposed feedouts contribute to maintaining containment integrity and electrical functionality during beyond-design-basis events. This work constitutes a significant step forward in the development of reliable and qualified electrical penetrations as part of the HCC framework in modern nuclear facilities.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114088"},"PeriodicalIF":1.9,"publicationDate":"2025-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143864081","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":"Research on the oscillation characteristic of thermal–hydraulic parameters for natural circulation conditions of small pressurized water reactors under rolling condition","authors":"Ang Li,&nbsp;Yuqing Chen,&nbsp;Yuxian Rao","doi":"10.1016/j.nucengdes.2025.114076","DOIUrl":"10.1016/j.nucengdes.2025.114076","url":null,"abstract":"<div><div>The ocean conditions of stormy waves, especially the rolling motion condition, have a large impact on the natural cycle operation of small integrated pressurized water reactors. The thermal–hydraulic parameters will fluctuate drastically and deviate from the original equilibrium position. For this small integrated pressurized water reactor, its thermal–hydraulic models of the primary and secondary loops are developed. And the calculation module under the rolling motion condition is supplemented. On this basis, the effects of rolling condition on the steady-state operation of the natural circulation and the conversion process from forced circulation to natural circulation are analyzed. “The factor of oscillation characteristic under rolling condition” is proposed to quantitatively evaluate the fluctuation characteristics of the parameters. The results show that an increase in the angular acceleration of the rolling makes the parameter oscillations more violent. Although the angular acceleration is small, the large-angle and large-period rolling has a greater effect on the parameters. The results can provide a basis for the analysis of the operating characteristics and optimization of the control scheme of a small integrated pressurized water reactor under the rolling condition.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114076"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863346","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":"Study of flow-induced vibrations of spiral tube bundles under nonlinear support in lead-bismuth eutectic environment based on large eddy simulation","authors":"Guangchao Yang ,&nbsp;Shanshan Bu ,&nbsp;Ke Zhang ,&nbsp;Xiaofei Yu ,&nbsp;Guo Chen ,&nbsp;Hui He ,&nbsp;Deqi Chen","doi":"10.1016/j.nucengdes.2025.114094","DOIUrl":"10.1016/j.nucengdes.2025.114094","url":null,"abstract":"<div><div>The flow-induced vibration (FIV) phenomenon of multi-row helical tube bundles in lead–bismuth eutectic (LBE) environments may be more severe with the tube-support gap included. This study utilizes the large eddy simulation (LES) turbulence model and a nonlinear dynamics model to investigate the effects of contact types, tube-support gaps, and the number of supports on the FIV characteristics of spiral tube bundles in LBE environment. The reliability of the numerical model is confirmed through validation against existing experimental data. The results show that, in the absence of gaps, linear and nonlinear contact conditions have a negligible impact on vibration characteristics. However, the influence of these contact types becomes significant when gaps are present. The presence of the gap causes a reduction in the critical flow rate at which fluid-elastic instability occurs and significantly alters the vibration characteristics during instability so that it no longer vibrates at the natural frequency of the structure. For flutter vibrations, the frequency domain characteristics of dynamic parameters remain consistent. For non-flutter vibrations, the displacement root mean square (RMS) is proportional to the gap size before support failure. As the gap increases (0 ∼ 0.5 mm), the contact pressure RMS decreases monotonically, while the destructiveness caused by contact vibro-impact initially increases and then decreases. The structural stress at the tube-support exhibits a trend of first decreasing and then increasing as the gap size grows. Additionally, increasing the number of supports (2 ∼ 4 within 90°) significantly reduces the vibration displacement and stress levels of the spiral tube.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114094"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863345","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":"SMR core thermal hydraulic experiments and code validation with the MOTEL test facility","authors":"Heikki Suikkanen ,&nbsp;Joonas Telkkä ,&nbsp;Antti Räsänen ,&nbsp;Eetu Kotro ,&nbsp;Michael Böttcher ,&nbsp;Lucia Rueda-Villegas ,&nbsp;Veronika Sunkova","doi":"10.1016/j.nucengdes.2025.114056","DOIUrl":"10.1016/j.nucengdes.2025.114056","url":null,"abstract":"<div><div>Thermal hydraulic experiments with the modular integral test facility, MOTEL, were performed as a part of the European McSAFER research project. The facility models an integral pressure water small modular reactor (SMR) with a helical steam generator and a core with separate heater rod groups, in which power can be individually controlled. Different asymmetric and ring-shaped radial core power distributions were imposed in the experiments to provoke cross flows in the buoyancy-driven coolant flow. The purpose of the experiments was to produce new SMR-relevant data for the validation of computational fluid dynamic (CFD) and thermal-hydraulic subchannel codes. The experimental measurements revealed cross flow mixing effects, mainly in the top part of the core. Obtaining visible differences in the fluid temperature measurements between different heater regions required significant power gradients between the regions. CFD simulations were performed using ANSYS CFX with a detailed model comprising the whole primary side of the facility, and additional investigations were conducted with a stand-alone model of the heat exchanger. Good agreement with the measurements was obtained with the CFD simulations, which also revealed further details of the core flow characteristics in an asymmetric heating case. Furthermore, simulations with the subchannel codes, CTF and VIPRE-01, were performed. The simulations with CTF highlighted the code’s capability to handle flow rates typical to natural circulation driven SMRs, as the results agreed well with the experiments and were able to predict the correct axial temperature profiles in the different regions of the core. VIPRE-01 solution stability was found to be highly sensitive to the flow rate, the power level, and the axial nodalization. Simulations with VIPRE-01 ended unsuccessfully due to convergence issues, and it was concluded that the conditions of the experiments are beyond the current capabilities of the code.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114056"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143863344","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":"Analytical studies for hydrogen distribution & management for 700 MWe IPHWR","authors":"Sanjeev Kumar Sharma ,&nbsp;D.K. Chandraker ,&nbsp;Manvendra Singh ,&nbsp;Vibha Hari ,&nbsp;Sameer Hajela","doi":"10.1016/j.nucengdes.2025.114053","DOIUrl":"10.1016/j.nucengdes.2025.114053","url":null,"abstract":"<div><div>Hydrogen generation and local accumulation of hydrogen in the containment atmosphere during a postulated accident scenario involving multiple failures could pose a threat to the integrity of the containment as the hydrogen can form a flammable mixture with air in the containment. One such unlikely postulated accident sequence is Large Break Loss–Of-Coolant Accident (LBLOCA) along with failure of Emergency Core Cooling System (ECCS). Moderator cooling system works as the heat sink for such accident sequences and hydrogen generation will be limited. Further as a part of Defense-In-Depth (DID) approach, the containment response has been evaluated under more severe conditions where failure of moderator cooling system is also postulated. During such a severe accident large amounts of hydrogen are expected to get generated.</div><div>A system thermal hydraulic computer code, Post Accident Containment System Response (<span><span>Singh et al., 2023</span></span>, <span><span>Sharma et al., 2024</span></span>) has been developed for the containment response calculation during normal operation as well as accident conditions including, severe accident.. The calculations for hydrogen management by using a combination of Passive Catalytic Recombiner Devices (PCRD) and provision of Passive Opening or Forced Mixing are performed for 700 MWe Indian Pressurized Heavy Water Reactors (IPHWRs) by using in house computer code PACSR- SI2.0. This paper presents the studies performed to assess the hydrogen behaviour along with optimised mitigating features during severe accident conditions. From the analysis, it is found that, with optimized number of PCRDs; hydrogen can be effectively managed during severe accident conditions. Lumped parameter approach is considered to be a good technique for the finalisation of the optimised number of PCRD,which can be further verified using 3-D CFD approach.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114053"},"PeriodicalIF":1.9,"publicationDate":"2025-04-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143851881","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":"Analysis of fission products release under DLOFC scenario at different power levels of Indonesian PeLUIt reactor design","authors":"Fitria Miftasani,&nbsp;Anni Nuril Hidayati,&nbsp;Steven Wijaya,&nbsp;Nina Widiawati,&nbsp;Dany Mulyana,&nbsp;Topan Setiadipura","doi":"10.1016/j.nucengdes.2025.114082","DOIUrl":"10.1016/j.nucengdes.2025.114082","url":null,"abstract":"<div><div>Understanding fission product release is essential for ensuring reactor safety during normal operation and accident conditions. This computational study focuses on the fission product release behavior in PeLUIt reactor design under normal and Depressurized Loss of Forced Cooling (DLOFC) scenarios at different power levels, from 10 MWt to 40 MWt. It deployed STACY, a fission product release code, with input fed by TRIAC-BATAN, which evaluates the failure fraction of TRISO-coated fuel particles. Neutronic calculations were conducted using PEBBED and OpenMC while fuel irradiation temperatures and accident conditions were analyzed through a combination of PEBBED and THERMIX-KONVEX. The results indicate that the failure fraction of TRISO particles remains negligible up to approximately 2,700 h of irradiation but increases significantly at higher power due to prolonged exposure with elevated temperatures. The fractional release of some fission products were studied. I-131 exhibits the highest release fluctuations with a rapid initial release followed by stabilization. Cs-137 maintains a stable release profile which increase gradually. Ag-110 m shows the highest cumulative release among metal fission products. Sr-90 is the most retentable one in the fuel, with minimal release observed even at high power. The study confirms that a higher reactor power level effect fission product release positively due to thermal degradation of the TRISO coated fuel particle. At 10 MWt to 30 MWt, the defect fraction increase is minor, but at 40 MWt, it becomes significant due to higher temperatures and fission gas pressure; however, at 30 MWt, fission product release remains within safe limits.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114082"},"PeriodicalIF":1.9,"publicationDate":"2025-04-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143848020","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":"Corrigendum to “Steam generator model design parameter sensitivity study for small modular reactor system” [Nucl. Eng. Des. 435 (2025) 1–15, 113973]","authors":"Palash K. Bhowmik ,&nbsp;Congjian Wang ,&nbsp;Nicholas Hernandez ,&nbsp;Tejas Kedlaya ,&nbsp;Piyush Sabharwall","doi":"10.1016/j.nucengdes.2025.114093","DOIUrl":"10.1016/j.nucengdes.2025.114093","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114093"},"PeriodicalIF":1.9,"publicationDate":"2025-04-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143877394","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":"Design and analysis of external steam supply process schemes for HPR1000 nuclear power units","authors":"Rongshun Xie ,&nbsp;Shengguo Wu ,&nbsp;Qiang Zeng ,&nbsp;Jiajun Tang ,&nbsp;Hui’er Sha ,&nbsp;Gonghao Lu ,&nbsp;Gang Hong ,&nbsp;Yaoli Zhang","doi":"10.1016/j.nucengdes.2025.114084","DOIUrl":"10.1016/j.nucengdes.2025.114084","url":null,"abstract":"<div><div>With the global increase in demand for low-carbon energy, the application of nuclear power in the industrial heating field has received significant attention. This paper focuses on the Hua-long Pressurized Reactor (HPR1000), addressing the steam demand of the Gulei Petrochemical Industrial Park in Zhangzhou through the design and optimization of the industrial steam supply system for large PWR cogeneration nuclear power units. The study proposes four steam supply schemes and establishes corresponding thermal models. The results indicate that Scheme B (main steam/condenser) outperforms others regarding fuel utilization coefficient and heat supply power consumption, achieving 48.87% and 29.75%, respectively. Additionally, it has a minimal impact on electricity generation, reducing it by only 12.02%. This study’s evaluation indicator system verifies the feasibility of the HPR1000 nuclear power unit in supplying industrial steam. It provides theoretical support for the comprehensive utilization of the unit’s energy and the optimization of economic benefits.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114084"},"PeriodicalIF":1.9,"publicationDate":"2025-04-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143850772","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 modeling and simulation of the GAMA-AHR","authors":"Andang Widi Harto,&nbsp;Kusnanto,&nbsp;Alexander Agung,&nbsp;Diva Jati Kanaya,&nbsp;M. Yayan Adi Putra","doi":"10.1016/j.nucengdes.2025.114085","DOIUrl":"10.1016/j.nucengdes.2025.114085","url":null,"abstract":"<div><div>The GAMA-AHR is an aqueous fueled nuclear reactor designed to produce 2000 six-day Ci per week of ⁹⁹Mo at thermal power of 200 kW. In this study, the dynamic behavior of the GAMA-AHR was demonstrated. A point reactor dynamics model consisting of point kinetics, heat balance, and reactivity feedback equations was developed to represent the primary system of the GAMA-AHR This reactor dynamics model was implemented in C++, and solved numerically with adaptive Runge Kutta method. Numerical simulation to show the reactor’s steady state and transient behavior is reported in this paper. The transient conditions were simulated for the detection of perturbations caused by a reactivity contribution from the reactivity control system and a reduction in the secondary cooling system’s performance. The model was validated by steady-state simulation at 100 % power, and the results agreed well with previous research and therefore demonstrated its suitability for predicting the reactor’s dynamic behavior. The simulation results showed that reactor power decreased with increasing temperature and decreasing heat transfer capability. It was able to safely operate (without fuel boiling) at 90 % reduced heat transfer capability and power lower than 100 kWh. The simulation results also showed that reactor power can be controlled by adjusting the control rod insertion or the fuel level and the reactor can be independently shut down by the full insertion of the control rod or by full draining of fuel into the fuel drain tank. Thus, the temperature reactivity feedback characteristic and reactivity control systems ensures the inherent safety of the GAMA-AHR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114085"},"PeriodicalIF":1.9,"publicationDate":"2025-04-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143848018","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}