Progress in Nuclear Energy最新文献

{"title":"Experimental measurement and theoretical calculation of Ar-41 radioactivity in the primary coolant of HTR-10","authors":"Rui Nie,&nbsp;Yu Wang,&nbsp;Feng Xie,&nbsp;Weihua Zhang,&nbsp;Jianzhu Cao,&nbsp;Liqiang Wei,&nbsp;Tao Ma,&nbsp;Xiaowei Li","doi":"10.1016/j.pnucene.2025.105686","DOIUrl":"10.1016/j.pnucene.2025.105686","url":null,"abstract":"<div><div>Ar-41 is a key activation product in the primary circuit of the high-temperature gas-cooled reactor (HTGR) and a concern for reactor radiation safety, impacting doses for nuclear power plant staff and nearby residents. Currently, the source of Ar-40 in pebble-bed HTGRs is unclear, and systematic research is lacking. In this study, the Ar-41 source term of the 10 MW high-temperature gas-cooled experimental reactor (HTR-10) was determined based on the operating status. The Ar-41 source term calculation model for HTGRs was established with three-types Ar-41 source term: pulse, short-time constant, and constant sources. The accuracy of the calculation model for the Ar-41 source term was confirmed by comparing the theoretical and experimental radioactivities of Ar-41 in the primary circuit, both of which varied in 10⁷–10⁸ Bq during normal operation and exhibited consistent trends. Finally, we proposed reducing radioactivity in the primary circuit by controlling the entry of Ar-40 or increasing the HPS's flow rate while maintaining purification efficiency. Current research fills a gap in the Ar-41 source term study in the primary circuit and systematically constructs a framework and methodology for calculating the Ar-41 source term in HTGRs, which provides strong support for the safe operation of HTGRs.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105686"},"PeriodicalIF":3.3,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143519708","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 investigation of separation and flow characteristics of gas-liquid separators for lithium-cooled space nuclear reactor","authors":"Wendong Li ,&nbsp;Ming Jin ,&nbsp;Chao Liu ,&nbsp;Yong Song ,&nbsp;Yang Li ,&nbsp;Danna Zhou ,&nbsp;Tao Zhou ,&nbsp;Jieqiong Jiang","doi":"10.1016/j.pnucene.2025.105682","DOIUrl":"10.1016/j.pnucene.2025.105682","url":null,"abstract":"<div><div>High-power space reactors utilizing liquid lithium as a coolant exhibit significant potential due to the high specific power, extended operational life, and reduced environmental dependence. A significant issue is that liquid lithium generates helium under core irradiation. Such phenomenon provokes challenges on separating and storing helium in a microgravity environment. To solve this problem, the current work proposes a specific design for a passive centrifugal Gas-Liquid Separator (GLS). The performance of the designed GLS was verified and validated by parametric investigations using computational fluid dynamics (CFD) techniques. The performance parameters, i.e., flow, separation, and gas-storage were analyzed in detail. The results confirmed that the stable separation and storage of helium can be achieved successfully by the GLS. The mechanisms of separation and gas-storage were revealed through dynamic gas-liquid interface variations, static pressure distribution, three-dimensional streamlines, axial and tangential velocities, and velocity magnitudes. The characteristics of flow, separation, and gas-storage within the GLS were precisely identified. According to the analysis of the three-dimensional flow field, gas within the gas-storage zone of the GLS exhibited stable and slow internal circulation, which was the critical indicator for maintaining effective gas-storage. Flow rate, helix angle (<em>θ</em>_s) and height to diameter ratio (<em>H</em>/<em>D</em>_s) were critical parameters influencing the separation and gas-storage performance of GLS. Notably, with varying flow rates, separation efficiency and maximum gas storage mass exhibited opposite variation trends. Decreasing <em>θ</em>_s can significantly enhance the separation and gas storage capabilities of the GLS. When the volume of gas-storage zone remains constant, variations in <em>H</em>/<em>D</em>_s lead to non-monotonic changes in tangential velocity and axial velocity, which lead to opposite effects on maximum gas storage mass and separation efficiency.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105682"},"PeriodicalIF":3.3,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143519709","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 filtration performance optimization of venturi scrubber with stirring baffle","authors":"Yangfeng Wu ,&nbsp;Yanmin Zhou ,&nbsp;Zhiming Zhang ,&nbsp;Jingmei Zhu ,&nbsp;Zhongning Sun ,&nbsp;Haifeng Gu ,&nbsp;Qianchao Ma","doi":"10.1016/j.pnucene.2025.105622","DOIUrl":"10.1016/j.pnucene.2025.105622","url":null,"abstract":"<div><div>As the core equipment of wet-type Filtered Containment Venting System (FCVS), the self-priming venturi scrubber has the characteristics of passive operation, reliable work, and superior filtration performance, and is widely used in the design of active nuclear power plant systems. In this paper, the filtration performance of the venturi scrubber is taken as the evaluation index, and the venturi scrubber is optimized by using a combination of numerical calculation and experimental based on the structure of the stirring baffle. By controlling the two structural parameters of the axial setting height and the area ratio of the gas phase opening, the effect of structural parameters of the stirring battle on the two-phase flow characteristic and atomization characteristic of the venturi scrubber was obtained. Subsequently, coupling CFD and dust particle capture theory to theoretically calculate the filtration efficiency, combined with experimental verification, the structural parameters of the stirring baffle with the best filtration performance were obtained. The results show that compared with the traditional venturi scrubber, the venturi scrubber with a stirring baffle with an axial height of 120 mm and an area ratio of gas phase opening of 0.8 can achieve a filtration efficiency of more than 99.9% for aerosols with a low flow resistance. Finally, the optimization suggestions for the structure of the stirring baffle were proposed. This paper is of great significance for the upgrading of the venturi scrubber and the optimization of the nuclear industry and systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105622"},"PeriodicalIF":3.3,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143519706","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 study of forced circulation and natural circulation transition characteristics of a small modular reactor equipped with helical-coiled steam generators","authors":"Jianping Zhou ,&nbsp;Maolong Liu ,&nbsp;Ziyi Xu ,&nbsp;Limin Liu ,&nbsp;Jianhua Cao","doi":"10.1016/j.pnucene.2025.105672","DOIUrl":"10.1016/j.pnucene.2025.105672","url":null,"abstract":"<div><div>The transient thermal hydraulic characteristics of the helical-coiled steam generator (H-OTSG) is important for the safety of the nuclear reactor. No research has yet addressed the transition between natural circulation and forced circulation in small modular pressurized water reactors equipped with H-OTSGs. In the present study, a numerical simulation study was carried out to investigate the transient thermal hydraulic characteristics of a H-OTSG during a transient process between forced circulation and natural circulation using the SGTH-1D code. Different transition schemes have significant impact on the thermal-hydraulics characteristics of the H-OTSG. An optimized transition scheme was established to perform a smooth transition between forced circulation and natural circulation without significant perturbation on the shell- and tube-side parameters. The results of the present study can be used to optimize the operation procedure of the small modular reactor equipped with H-OTSGs.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105672"},"PeriodicalIF":3.3,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512578","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":"Fast prediction and inverse modeling for parameterized depressurization from a CO2 vessel using POD-BPNN reduced-order method","authors":"Yanjie Kang ,&nbsp;Gengyuan Tian ,&nbsp;Yanping Huang ,&nbsp;Sulin Qin ,&nbsp;Qingyu Huang ,&nbsp;Yuan Yuan ,&nbsp;Yuan Zhou","doi":"10.1016/j.pnucene.2025.105684","DOIUrl":"10.1016/j.pnucene.2025.105684","url":null,"abstract":"<div><div>Rapid and accurate prediction and inversion of Loss-of-Coolant Accident (LOCA) are crucial for reactor safety. Existing methods struggle with computational efficiency and heavy reliance on detectors, which are difficult to deploy in confined, high-pressure environments and prone to noise and failure. This study introduces a parameterized model using Reduced-Order Modeling (ROM) for predicting depressurization and leak location inversion in pressure vessels. A CO₂ vessel case study utilizes a phase separation model as the Full-Order Model (FOM) to generate pressure and liquid level snapshots under various initial conditions, such as temperature, pressure, and leak height. Proper Orthogonal Decomposition (POD) is employed to extract reduced-order bases and coefficients, while Back Propagation Neural Networks (BPNNs) map initial conditions to these coefficients, enabling rapid depressurization prediction. The model is integrated into an inversion framework based on pressure response, with a grid search algorithm refining leak location estimates. Predictions for 0–1200 s take less than 1 s, with errors under 0.2% compared to the FOM. Inversion errors remain below 4% across noise levels, showcasing robustness. This POD-BPNN ROM method holds promise for efficient LOCA prediction and inversion.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105684"},"PeriodicalIF":3.3,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512599","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":"Characteristics of melt-jet fragmentation behavior in coolant in nuclear reactor based on VTMCI experiments","authors":"Shaojie Tan ,&nbsp;Hui Cheng ,&nbsp;Songbai Cheng","doi":"10.1016/j.pnucene.2025.105683","DOIUrl":"10.1016/j.pnucene.2025.105683","url":null,"abstract":"<div><div>Aimed at systematically investigating the mechanisms underlying the melt-jet fragmentation behavior in liquid metal cooled fast reactors, several series of experiments were skillfully engineered and carried out. Different molten metals were released into a water pool using the VTMCI facility. The molten metals used included the Bi-Sn-In alloy, lead-bismuth eutectic (45 wt% Pb-55 wt% Bi), lead, and lead-bismuth non-eutectic (70 wt% Pb-30 wt% Bi). This study presents a comprehensive synthesis and detailed comparative analysis of experimental data, examining the potential impact of key parameters on melt-jet fragmentation behavior. The results show that the increase in water subcooling results in an increase in particle size by up to 70%, an increase in the debris bed porosity by up to 67%, and a reduction in the debris sphericity by up to 78%. Raising the melt superheat causes particle size to decrease by as much as 33%, debris bed porosity to drop by up to 34%, and debris sphericity to rise by as much as 144%. Higher melt penetration velocity or smaller jet cross section diameter results in smaller fragment sizes by up to 34%. Increasing the water depth leads to a decrease in particle size of up to 9%, a reduction in debris bed porosity by as much as 7%, and an enhancement in debris sphericity of up to 76%. The article offers a wealth of experimental data that can enhance the validation of reactor safety analysis codes in China.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105683"},"PeriodicalIF":3.3,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143511205","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":"Optimization of water chemistry to mitigate corrosion products in nuclear power plants using big data and multiple linear regression in machine learning","authors":"Jinsoo Choi ,&nbsp;Jong-Il Yun","doi":"10.1016/j.pnucene.2025.105685","DOIUrl":"10.1016/j.pnucene.2025.105685","url":null,"abstract":"<div><div>The primary system of a nuclear power plant operates under high temperature and high pressure conditions, making material integrity crucial to prevent accidents and ensure safe operation. Effective coolant chemistry control is essential for mitigating corrosion product releases, preventing material degradation and minimizing radiation exposure within the plant. To accurately predict corrosion product concentrations, a model was developed using data collected under reactor-critical operating conditions. where consistent parameters like pressure and temperature were maintained. Co-58, a nickel activation product, was selected as a representative corrosion product due to its high concentration and presence in primary materials. Key chemical factors influencing corrosion, such as lithium (Li), pH, and dissolved hydrogen (DH), were analyzed, and operational parameters were verified to confirm reactor-critical conditions. Effects of Li and DH on Co-58 concentration were evaluated using Multiple Linear Regression in Machine Learning. Between Li and pH, Li was ultimately selected for its higher regression coefficient (β) and direct role in controlling pH. Then, Li and DH were used to develop a predictive model for corrosion diagnostics. Findings indicate that Li has a stronger impact on Co-58 reduction than DH, with the β value being 4.01 times higher. The model is statistically significant (F = 282.6, p &lt; 0.000) and demonstrates an adjusted R² value of 0.351, with machine learning results achieving an R² value of 0.442 on the test set. These results suggest that maintaining Li and DH conditions at the upper allowable ranges can play a crucial role in reducing corrosion products.</div><div>The significance of this study lies in its reliance on operational, reactor-critical data rather than laboratory-only data, offering a robust and practical approach to optimize corrosion management in both existing nuclear power plants (NPPs) and next-generation reactors, including Small Modular Reactors (SMRs). This model provides a novel method that enhances plant safety and operational efficiency by enabling more accurate prediction and control of Co-58 concentrations than traditional approaches.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105685"},"PeriodicalIF":3.3,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512577","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":"Density functional theory study of radiotoxic PbPo adsorption on Ag surfaces","authors":"Ao Gan ,&nbsp;Hui Du ,&nbsp;Lijie Shi ,&nbsp;Man Jiang ,&nbsp;Muyi Ni","doi":"10.1016/j.pnucene.2025.105680","DOIUrl":"10.1016/j.pnucene.2025.105680","url":null,"abstract":"<div><div>Radiotoxic PbPo molecules are produced in the lead-bismuth eutectic and vaporized and transported with the cover gas in the lead-bismuth fast reactor, the leakage of PbPo causes radioactive safety issues in the reactors. PbPo should be filtered and silver (Ag) is considered a promising candidate for filtration due to its unique properties. This study employs density functional theory to investigate the adsorption behavior of PbPo on Ag surfaces with and without vacancy. It was found that PbPo is adsorbed either in molecular form with the adsorption energies ranging from −0.44 to −1.72 eV, or in dissociated form with the adsorption energies ranging from −1.55 to −2.99 eV. The results indicate that the PbPo is chemisorbed on all low-index Ag surfaces considered and prefers to occupy the vacancy and hollow sites in the dissociated form. The charge density differences analysis shows that the charge of the electrons is mostly transferred from Pb and Po to nearby Ag atoms on the surface. The density of states result demonstrates that strong hybridization of Pb 6p, Po 6p and Ag 4d orbitals contributes to the strong chemisorption.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105680"},"PeriodicalIF":3.3,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143487468","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":"Thermo-mechanical response of SiCf/SiC composite cladding: Effect of loss-of-coolant accident duration","authors":"Shuang Liang ,&nbsp;Chong Wei ,&nbsp;Bowen Qiu ,&nbsp;Cheng Zhang ,&nbsp;Haoyu Liao ,&nbsp;Xingjie Ren ,&nbsp;Zhong Xiao","doi":"10.1016/j.pnucene.2025.105681","DOIUrl":"10.1016/j.pnucene.2025.105681","url":null,"abstract":"<div><div>SiC_f/SiC composite is one of the candidate materials for fuel cladding in Generation IV advanced nuclear reactors. However, there are still few risks under accident conditions, such as large physical property changes and poor thermal conductivity, the performance of the cladding remains to be studied especially in extreme accident conditions (loss-of-coolant accident, LOCA and reactivity-insertion accident, RIA). In this study, the thermo-mechanical responses of SiC_f/SiC composite cladding under different LOCA durations were investigated by a meso-macroscopic coupling finite element simulation method. At mesoscopic scale, we present a mathematical model between the longitudinal tensile ultimate strength and temperature (200K∼2000K), which enables us to define the failure of macroscopic cladding. At macroscopic scale, the established mathematical model was applied to the failure analysis of macroscopic cladding, then the temperature and hoop stress of the inner and outer surfaces of the cladding were studied under different LOCA durations of 200s, 300s and 400s. The results showed that under LOCA condition, the hoop stress difference between the inner and external surfaces of the SiC_f/SiC composite cladding increases about 200% compared with that under normal service conditions. Meanwhile, the hoop stress difference increases slowly with the increase of LOCA duration, and the maximum increment can reach about 2.8%. In addition, the peak temperature increases with the increase of LOCA duration, which the maximum increment can reach about 5.6%, while the peak temperature difference between the inner and external surfaces increases slowly with the increase of LOCA duration, and the maximum increment can reach about 5.1%.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105681"},"PeriodicalIF":3.3,"publicationDate":"2025-02-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143471152","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":"Investigation of heat transfer characteristics of carbon dioxide transcritical flow in a reactor channel","authors":"Mao Luo ,&nbsp;Hui Bao ,&nbsp;Fulong Zhao ,&nbsp;Ruibo Lu ,&nbsp;Sichao Tan ,&nbsp;Ruifeng Tian","doi":"10.1016/j.pnucene.2025.105670","DOIUrl":"10.1016/j.pnucene.2025.105670","url":null,"abstract":"<div><div>The fourth generation of supercritical carbon dioxide (S-CO₂) cooled advanced reactor has the characteristics of compact volume and high efficiency, and is widely used in space reactors and small reactors. For the purpose of the heat transfer characteristics of S-CO₂ under transcritical conditions, the effects of changing inlet velocity, inlet CO₂ fluid temperature, CO₂ fluid operating pressure and wall heat flux on <em>h</em> and <em>Nu</em> in a single cylinder without gravity were analyzed by numerical simulation. The influence of different inlet flow rates on transcritical conditions under horizontal tube gravity is analyzed, and the temperature and velocity distributions of normal transcritical conditions and heat transfer deterioration conditions are compared. The consequence show that the rise of wall heat flux and the decrease of pressure will make the critical position move towards the inlet end. Reducing the inlet rate and increasing the inlet temperature will also shift the critical position towards the inlet end and result the heat transfer to worsen at the inlet end. In the horizontal pipe, when the heat transfer deteriorates, the CO₂ fluid temperature differences between the top and lower parts exceeds 20 K, and the coefficient of heat transfer at underside is 2.5 times that at the top. When the deterioration degree of heat transfer deepens, the difference of heat transfer coefficients between the upper and lower parts of the channel will also increase due to ascent the temperature difference. Finally, the heat transfer deterioration can be attributed to the dual effects of the near-wall CO₂ fluid criticality and the increase in the CO₂ fluid mainstream thermal conductivity.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105670"},"PeriodicalIF":3.3,"publicationDate":"2025-02-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143471153","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}