Progress in Nuclear Energy最新文献

{"title":"Preliminary neutronics optimization on the concept of multi-beam accelerator driven system","authors":"Xunchao Zhang ,&nbsp;Neng Pu ,&nbsp;Huan Jia ,&nbsp;Yuanshuai Qin ,&nbsp;Hanjie Cai ,&nbsp;Yuan He","doi":"10.1016/j.pnucene.2024.105542","DOIUrl":"10.1016/j.pnucene.2024.105542","url":null,"abstract":"<div><div>A new subcritical reactor concept is proposed as one accelerator beam splitting into multiple beams to drive the subcritical reactor in this work, which is so called Multi-Beam Accelerator Driven System (MB-ADS). The spallation target is designed as a target assembly to the fuel assembly. The high current proton beam is divided into multiple parts and injected into different targets scattered among the core to improve the beam efficiency and flatten the spatial power distribution of the core. Based on different MB-ADS schemes, neutronics were conducted on the effects of splitting beam number, target assembly arrangements, fuel zoning, and neutron data libraries. The results show that a reasonable multi-beam scheme can significantly improve the efficiency of the proton beam and flatten the power distribution of the reactor compared to the one target ADS scheme. With the improved beam efficiency, the beam density on the target window is greatly reduced.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105542"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141576","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 research on the thermal-hydraulic characteristics of the large sodium-cooled fast reactor roof slab","authors":"Yuhao Zhang ,&nbsp;Yuanshu Qu ,&nbsp;Jiaqi Feng ,&nbsp;Daogang Lu ,&nbsp;Han Wang ,&nbsp;Zongyu Yu","doi":"10.1016/j.pnucene.2024.105584","DOIUrl":"10.1016/j.pnucene.2024.105584","url":null,"abstract":"<div><div>The roof slab of a large sodium-cooled fast reactor is a structure subjected to bottom heating and external cooling. Preserving the roof slab's integrity is paramount for ensuring the safety of the fast reactor, which is profoundly influenced by the temperature distribution. However, owing to its intricate internal configuration, obtaining a detailed and comprehensive temperature distribution of the roof slab is challenging. To address this, a refined and comprehensive 1:1 prototypical model of the fast reactor roof slab was established and subjected to numerical calculations. These calculations determined the thermal-hydraulic characteristics of the roof slab during operating conditions, as well as the natural circulation characteristics within the annular gaps. In order to assess the reliability of the calculation methods utilized for evaluating the thermal-hydraulic properties of the fast reactor roof slab, this study developed a scaled-down model based on the scaled-down Heat-transfer Experimental Facility for Argon space at the top of the main vessel (HEFA). A comprehensive comparison was conducted between the calculated results and the corresponding experimental data. The resulting discrepancy was found to be less than 3%, indicating the effectiveness of the employed calculation methods and the credibility of the obtained results. Finally, an empirical temperature-<em>Ra</em> equation for the fast reactor roof slab under operating conditions was obtained by fitting the calculation results. This research provides a valuable reference for the design of the fast reactor roof slab and its associated penetration.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105584"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141637","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 on bubble growth interphase heat transfer of rapid decompression","authors":"Jiewen Deng ,&nbsp;Tao Huang ,&nbsp;Longxiang Zhu ,&nbsp;Liangming Pan ,&nbsp;Wan Sun ,&nbsp;Luteng Zhang ,&nbsp;Zaiyong Ma","doi":"10.1016/j.pnucene.2024.105578","DOIUrl":"10.1016/j.pnucene.2024.105578","url":null,"abstract":"<div><div>In a loss of coolant accident in a nuclear reactor, steam bubbles undergo superheated growth, where the interphase heat transfer coefficient significantly influences both the bubble growth rate and the void fraction of the two-phase flow. The bubbles should be grown within a homogeneous temperature field in order to facilitate a more comprehensive investigation into the characteristics of interphase heat transfer. Most existing studies are based on vapor bubbles generated within vaporization holes on walls. However, due to temperature gradients near the wall, these studies have limitations regarding heat transfer conditions and bubble shape. To investigate the realistic growth behavior characteristics of vapor bubbles during reactor loss-of-coolant accidents, the experiment was conducted on bubble growth in superheated liquid phases using a visual pressure vessel capable of creating instantaneous pressure drops. The system pressure ranged from 0.20 to 2.93 MPa. By employing high-speed cameras, pressure sensors, and thermocouples, the study investigated the impact of varying degrees of superheat, bubble Reynolds numbers, and Prandtl numbers on bubble growth rate. In this paper, a novel approach is proposed to investigate the interphase heat transfer coefficient under pressure drop, and its results are compared with previous methods that neglected the influence of vapor expansion. The correlation function incorporates Reynolds number, Jacob number, and Prandtl number, indicating the equation reveals the quantitative correlation between the interphase heat transfer coefficient and convection as well as heat conduction. Average heat transfer coefficients are determined for various operating conditions, leading to the development of an empirical correlation equation for the interphase heat transfer coefficient with an average relative error of 16.6%. Comparisons between the new correlation and other experimental data demonstrate its effectiveness in predicting the growth radius of the bubble.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105578"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141640","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":"The neutronic analysis of heat pipe-cooled traveling wave reactors utilizing natural uranium","authors":"Po Hu,&nbsp;Kunfeng Ma","doi":"10.1016/j.pnucene.2024.105572","DOIUrl":"10.1016/j.pnucene.2024.105572","url":null,"abstract":"<div><div>The traveling wave reactor utilizing natural uranium fuel can achieve enhanced economic performance. This paper presents two novel designs for a heat pipe cooled traveling wave reactor (HPTWR) that incorporates natural uranium. In the initial HPTWR design, the reactor features an ignitor region containing high-assay low-enrichment uranium (HALEU) (20% and 10%) and a breeder region utilizing natural uranium fuel at 0.72% enrichment. The core is configured as a hexagonal prism and consists of 1242 individual UN fuel elements. Each fuel element is a smaller hexagonal prism with a heat pipe channel at its center. The current design employs a lithium and molybdenum/rhenium (Mo/Re) heat pipe. Simulation results indicate that at the beginning of the cycle (BOC), the core reactivity is substantial, with a K_eff value reaching 1.36. The power profile is asymmetric, peaking in the ignition region. After 175 years of operation, when the traveling wave propagates into the breeding region, the power profile assumes a relatively symmetric form. An improved HPTWR design is then proposed, in which the fuel concentration and enrichment of ²³⁵U are adjusted axially in the extended ignition region. This adjustment results in a more symmetric power profile and a significantly lower K_eff of 1.055 at BOC. The initial HPTWR design can achieve a refueling cycle of 350 years with a thermal power output of 100 MW, while the improved HPTWR design can attain a refueling cycle of 335 years with the same breeder region and power level.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105572"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141648","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":"HINT-Sec: Hardware-in-the-loop nuclear power plant testbed for cyber security","authors":"Kihaeng Nam ,&nbsp;Kookheui Kwon ,&nbsp;Aram Kim","doi":"10.1016/j.pnucene.2024.105600","DOIUrl":"10.1016/j.pnucene.2024.105600","url":null,"abstract":"<div><div>The increasing integration of digital technologies within nuclear power plants has heightened their susceptibility to cyber threats, underscoring the need for advanced research frameworks dedicated to cybersecurity. This research introduces HINT-Sec, a high-fidelity hardware-in-the-loop testbed meticulously designed for cybersecurity research in nuclear power plants. HINT-Sec integrates a verified NPP simulator with both safety-grade and non-safety-grade programmable logic controllers, as well as distributed control systems, facilitating the realistic simulation of cyber-attacks and their consequent impacts on nuclear operations. Through comprehensive performance testing and scenario validation, the capability of HINT-Sec to accurately replicate the behavior of the APR1400 reactor under both normal and abnormal conditions is demonstrated. This capability provides a robust platform for the evaluation and development of cybersecurity measures within NPP environments. The design and implementation approach of the HINT-Sec testbed represents a significant contribution to the field of nuclear cybersecurity, ensuring that future research will be equipped to address the intricate challenges posed by cyber threats in critical infrastructure.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105600"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141784","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Optimization and simulation of startup control for space nuclear power systems with closed brayton cycle based on NuHeXSys","authors":"Chengyuan Li ,&nbsp;Leran Guo ,&nbsp;Shanfang Huang ,&nbsp;Jian Deng","doi":"10.1016/j.pnucene.2025.105610","DOIUrl":"10.1016/j.pnucene.2025.105610","url":null,"abstract":"<div><div>This study presents the optimization of startup control strategies for space nuclear power systems utilizing a Closed Brayton Cycle (CBC) with helium-xenon as the working fluid, addressing the critical challenge of battery limitations in space missions. A novel simulation tool, NuHeXSys, was developed to model the system, incorporating non-ideal gas behavior, multi-channel reactor core dynamics, and turbomachinery performance. Leveraging this model, the Non-dominated Sorting Genetic Algorithm II (NSGA-II) was employed to optimize key control parameters, including reactivity insertion, turbine shaft speed, and cooler temperature. The optimization results demonstrate a significant reduction in startup time by 1260 s, with the turbine inlet temperature reaching its nominal value 1980 s earlier. Additionally, external energy requirements decreased by 17.2%, and peak power demand was reduced by 27.9%, thereby improving overall system efficiency and reducing reliance on onboard batteries. The integration of NuHeXSys with NSGA-II offers a comprehensive framework for enhancing startup performance, providing crucial improvements for long-duration space exploration missions.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105610"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143140985","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":"Scaling analysis of single-phase natural circulation in deep-pool-type nuclear heating reactors","authors":"Gongqing Wang ,&nbsp;Dabin Sun ,&nbsp;Jincheng Wang ,&nbsp;Wei Wang ,&nbsp;Ruifeng Tian ,&nbsp;Ru Li","doi":"10.1016/j.pnucene.2024.105564","DOIUrl":"10.1016/j.pnucene.2024.105564","url":null,"abstract":"<div><div>In the deep-pool-type nuclear heating reactor (DHNR), natural circulation within the pool serves as the primary way of removing core residual heat during the late stages of a loss-of-flow accident (LOFA), and its heat transfer performance directly affects the core safety. In this paper, a scaling analysis is conducted for the single-phase natural circulation occurred in DHNR pool, and the similarity criteria are developed for natural circulation loop. Subsequently, three scaled models with different scaling ratios are designed according to similarity criteria. Finally, these three scaled models are preliminarily evaluated by RELAP5 code. The distortion of the local heat transfer in core has little effect on the response characteristics of natural circulation flow between scaled model and prototype. However, this small distortion led to significant discrepancies in the transient response characteristics of the fuel rod cladding temperature of scaled models. The similarity of the axial power distribution characteristics of the heat source compared to the prototype is the key factor in reproducing the axial distribution and response characteristics of the fuel rod cladding temperature in scaled models.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105564"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141582","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":"Modeling of the decomposition of iodine oxides aerosols (IOx) in the containment - Consequences on the understanding of volatile iodine behaviour in the containment","authors":"L. Bosland ,&nbsp;O. Leroy","doi":"10.1016/j.pnucene.2024.105576","DOIUrl":"10.1016/j.pnucene.2024.105576","url":null,"abstract":"<div><div>Iodine oxides aerosols (IOx) are fine aerosol particles that are formed in a reactor containment in case of severe accident. Their stability was studied under irradiation in representative conditions of the containment. Under the influence of the temperature, humidity and irradiation they are found to be unstable. Their decomposition kinetics and stability are studied in EPICUR facility thanks to the STEM, STEM2 and MIRE projects. An IOx decomposition model is set up in ASTEC-SOPHAEROS code (version 3.1). It considers the influence of the humidity, the temperature, and dose rate. Its application to the PHEBUS FPT-0/1/2/3 tests lead to (1) an improved modeling and understanding of the inorganic iodine volatility in the FPT-3 containment all along the transient and to (2) an improved modeling after the washing when the gaseous temperature increases (FPT-0 and FPT-1) and (3) before the washing when the humidity decreases (FPT-2) as long as a relevant molar gaseous iodine fraction coming (x_{I2_RCS}) from the reactor coolant system (RCS) is set up for FPT-0/1/2. A sensitivity analysis indicates that values of x_{I2_RCS} up to ≈30% would be needed to improve the modeling of iodine volatility before the washing for FPT-0/2. A fraction of 10% would be more appropriate for FPT-1. A complementary approach is on-going to quantify the influence of the uncertainties of each chemical model developed in ASTEC-SOPHAEROS code within the objective to identify which chemical phenomena are the most influent on iodine volatility in the PHEBUS containment.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105576"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141590","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A novel Ansys CFX – RELAP5 coupling tool for the transient thermal-hydraulic analysis of liquid metal systems","authors":"Pietro Cioli Puviani ,&nbsp;Tommaso Del Moro ,&nbsp;Bruno Gonfiotti ,&nbsp;Daniele Martelli ,&nbsp;Fabio Giannetti ,&nbsp;Roberto Zanino ,&nbsp;Mariano Tarantino","doi":"10.1016/j.pnucene.2024.105590","DOIUrl":"10.1016/j.pnucene.2024.105590","url":null,"abstract":"<div><div>Thermal-hydraulic analysis of three-dimensional systems and phenomena is essential for the design and safety assessment of new Generation IV reactor concepts. While System Thermal-Hydraulic (STH) codes serve as established tools for studying Light Water Reactors, their intrinsic limitations cannot allow to accurately model pool-type systems and fluids with high thermal conductivity, where three-dimensional phenomena cannot be accurately replicated. On the other hand, Computational Fluid Dynamics (CFD) codes have demonstrated significant potential in simulating components at a detailed scale and in predicting their thermal-hydraulic behaviour. To exploit the benefits of both code families, coupled tools are gaining attention as a viable solution to enhance the representation of system thermal-hydraulics, while maintaining reasonable computational times. This paper presents a novel coupling tool between Ansys CFX and RELAP5/Mod3.3 codes, developed for different numerical time advancing schemes and domain discretization approaches, and achieved through the development of in-house scripts for memory management, code synchronization, and information exchange. The tool is then applied to three different test cases with increasing modelling complexity, examining the performance and consistency of the coupling method with respect to the RELAP5/Mod3.3 standalone code. As expected, the adoption of different numerical time advancing schemes and domain discretization approaches lead to consistent results, highlighting the efficacy and limitation of each proposed solution. Furthermore, in cases where the approximations of the STH codes are valid (test cases 1 and 2), the coupled tool and the RELAP5/Mod3.3 standalone simulations give similar results. Instead, an application of the tool where RELAP5 is expected to be not reliable is proposed in test case 3, i.e., 3D flow mixing in a large volume, highlighting the potentiality of the tool and the conditions where a TH analysis can benefit from a STH/CFD coupled calculation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105590"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143141593","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":"Experimental study on startup performance of sodium heat pipe with high ratio of length-to-diameter under different coolant flow rates","authors":"Jun Xu ,&nbsp;Hongxing Yu ,&nbsp;Jian Deng ,&nbsp;Yu Liu ,&nbsp;Shuhua Ding ,&nbsp;Muhao Zhang ,&nbsp;Xiaohui Xia","doi":"10.1016/j.pnucene.2025.105623","DOIUrl":"10.1016/j.pnucene.2025.105623","url":null,"abstract":"<div><div>Sodium heat pipes, characterized by a high length-to-diameter ratio, are commonly employed in various heat pipe-cooled nuclear reactors. In these reactors, sodium serves as the working medium, remaining solid at room temperature. The startup process of such heat pipes is complex, involving phenomena such as the melting of the solid sodium and the establishment of a continuous vapor flow. The effectiveness of these heat pipes relies heavily on their startup behavior from a frozen state. To assess the startup performance of sodium heat pipes, an experimental study was conducted on a 2.5-m-long arterial sodium heat pipe, using high-temperature compressed air as the coolant. The investigation examined the impact of different air flow rates on startup performance. The findings revealed that high-temperature compressed air significantly improved the startup performance of the 2.5-m arterial sodium heat pipe under all three airflow conditions. Notably, during the startup process, the heat absorbed in the evaporation section is relatively minimal, making it beneficial to reduce the cooling capacity of the coolant to improve the heat pipe's startup performance. Additionally, when the air flow rate is lowered, the heat pipe requires less startup power and achieves a quicker startup time. By analyzing the data, it is revealed that the evaporator exit temperature increases with the reduction of air flow rate, which can avoid the sonic limit.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"180 ","pages":"Article 105623"},"PeriodicalIF":3.3,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143140940","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}