Progress in Nuclear Energy最新文献

{"title":"Post-irradiation examination of silicide fuel (U3Si2-Al) with 4.8 gU/cm3 for RSG-GAS","authors":"Aslina Br Ginting ,&nbsp;Rohmad Sigit Eko Budi Prasetyo ,&nbsp;Boybul ,&nbsp;Sungkono ,&nbsp;Supardjo ,&nbsp;Surian Pinem ,&nbsp;Wahid Luthfi","doi":"10.1016/j.pnucene.2025.105673","DOIUrl":"10.1016/j.pnucene.2025.105673","url":null,"abstract":"<div><div>The GA Siwabessy Multipurpose Reactor (RSG-GAS) has a nominal thermal power of 30 MWth and uses silicide fuel (U₃Si₂-Al) with a 2.96 gU/cm³ uranium density. To extend the reactor operating cycle, the RSG-GAS fuel can be converted to use a higher density fuel such as silicide fuel with a 4.8 gU/cm³ density. Irradiation tests of silicide fuel with 4.8 gU/cm³ density in the RSG-GAS core have been carried out to achieve burnup fractions of 20%, 40%, and 60% burned U-235. As part of a study for the development of new research reactor fuels, the post-irradiation examination (PIE) has been done for mini fuel plates that have been irradiated at the RSG-GAS. Several aspects of the PIE that were considered were physical changes, fuel swelling, and burnup distribution on the fuel plate. An in-depth analysis was conducted to verify the performance of 4.8 gU/cm³ silicide fuel so it can be used in the reactor core. Non-destructive testing in the form of visual inspection showed that the dimensions of the fuel plate did not change and there were no physical anomalies that could affect the performance of the silicide fuel plate for up to 60% burnup fraction. Measurement of the axial burnup distribution on the fuel using destructive tests, either based on the isotopic masses of Cs-137 or U-235 showed a consistency. In general, the physical properties of the silicide fuel plate that has been burned up to 60% burnup fraction are still quite relevant so the RSG-GAS fuel plate with a 4.8 gU/cm³ uranium density can be a candidate to replace the typical 2.56 gU/cm³ silicide fuel in RSG-GAS which currently has an average discharge burnup of 56%.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"183 ","pages":"Article 105673"},"PeriodicalIF":3.3,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143438232","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":"Hydraulic characteristics investigation on a novel multi-plate throttle entry tube of fast reactor control rod subassembly","authors":"Haiqi Qin ,&nbsp;Daogang Lu ,&nbsp;Dawen Zhong ,&nbsp;Qiong Cao ,&nbsp;Xiaowei Li","doi":"10.1016/j.pnucene.2025.105669","DOIUrl":"10.1016/j.pnucene.2025.105669","url":null,"abstract":"<div><div>As the only reactivity control method, the Control Rod Subassembly (CRSA) is employed to control the startup, shutdown and power regulation of Sodium-cooled Fast Reactor (SFR). The entry tube is an essential structure component of CRSA, serving not only as the inlet of liquid sodium but also as the support and guidance for CRSA. To address the shortcomings of traditional straight-through entry tube in terms of throttling effect and experiment cost, a novel Multi-plate Throttle Entry Tube (MTET) is designed and manufactured in this investigation. MTET hydraulic characteristics and geometrical parameter effect are investigated detailed by hydraulic experiments and numerical simulations. Based on the similarity analysis, the alternative hydraulic experiments are conducted on MTETs with different geometrical parameters. Experiment results reveal that MTET can achieve 299.12 kPa at the rated mass flow rate of 1.20 kg/s, with a relative error of −0.29%, which also verifies the rationality of this new MTET. In view of the design requirements of SFR, the optimal geometrical parameters are determined as three throttling holes of 10 mm in diameter and five throttling plates of 41 mm in width. Additionally, numerical results provide an evident that the geometrical parameters are pivotal influence factors to the MTET hydraulic characteristics, such as throttling plate width, throttling hole arrangement, connecting rod length, and throttling plate pitch. Increasing the throttling plate width and extending the throttling plate pitch can both improve the throttling effect of MTET. By comparison, the throttling effect of MTET is more sensitive to the throttling hole arrangement rather than the connecting rod length. It's important to note that the modification of connecting rod length needs to be carefully evaluated. Meanwhile, the non-uniform arrangement in multi-plate throttle is another solution to adjust the MTET hydraulic characteristics, which only needs to modify the width of single throttling plate. Compared to the fifth throttling plate (along the coolant flow direction), the adjustment efficiency of the first throttling plate increases by more than 4%. Therefore, the significant pressure drop should be adjusted by modifying the widths of upstream throttling plates, while the modification of downstream throttling plates is more suitable for the minor pressure drop adjustment. This innovative design demonstrates the improved throttling effect of MTET, as the promising structure optimization for CRSA and other subassembly of SFR.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105669"},"PeriodicalIF":3.3,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429035","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 schemes to evaluate HELB-induced blast waves","authors":"Tae-Yong Kim,&nbsp;Yoon-Suk Chang","doi":"10.1016/j.pnucene.2025.105665","DOIUrl":"10.1016/j.pnucene.2025.105665","url":null,"abstract":"<div><div>High energy line breaks (HELBs) have been postulated as a design basis event for protection of major structures, systems and components in nuclear power plants (NPPs). However, it was recently pointed out that plausible phenomena such as blast wave and structural resonance caused by jet impact need to be carefully evaluated. In this context, the present study investigates analysis schemes for evaluating the blast wave effects during design stage of new reactors. At first, three types of comparative analyses were conducted with a focus on the steam generator (SG) in an operating 1400 MWe NPP under a representative HELB scenario, from which the most conservative but reasonable one was identified. As the next step, the adequate scheme was applied for parametric analyses of the SG in a developing 60 MWe NPP considering nine hypothesized conditions. Ultimately, structural behaviors and integrity of the smaller SG according to design acceptance criterion as well as impacts of different pipe diameters and rupture distances on dynamic responses were assessed to discuss capability withstanding the HELB-induced blast waves.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105665"},"PeriodicalIF":3.3,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143420134","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 on flow-induced vibration of LBE-cooled wire-wrapped rod and rod bundle","authors":"Sipeng Wang ,&nbsp;Jingbo Hui ,&nbsp;Zhuoyao Li ,&nbsp;Hang Zhang ,&nbsp;Bao-Wen Yang ,&nbsp;Kun Zhuang","doi":"10.1016/j.pnucene.2025.105664","DOIUrl":"10.1016/j.pnucene.2025.105664","url":null,"abstract":"<div><div>Among the Generation-IV nuclear reactors, LBE-cooled reactors stand out as particularly promising due to their high neutron economy, high thermal conductivity, and high boiling point. However, the high density of lead-bismuth coolants and the mixing effect of helical structures lead to flow-induced vibrations, which can cause fuel rod vibrations leading to the leakage of radioactive materials and posing a significant threat to the safe operation of nuclear reactors. Consequently, this study employs CFD method to conduct an in-depth analysis of the flow-induced vibration characteristics of LBE-cooled wire-wrapped rod and rod bundles under various flow velocities to elucidate the intrinsic link between fluid excitation and structural vibration characteristics. Through multi-scenario analysis, this study not only clarifies the direct relationship between fluid excitation frequencies and fuel rod vibration responses but also integrates micro-scale fluid dynamic characteristics with macro-scale structural vibration characteristics, offering a novel perspective for understanding the physical mechanisms behind flow-induced vibrations. The results indicate that the main low-frequency components in the fluid excitation have induced low-frequency forced vibrations in the fuel rods that correspond to their response frequencies, which is consistent with numerous studies related to vibration. There is a significant difference in vibration behavior between single rod and rod bundle. As for rod bundle, due to the distinct pressure distribution characteristics of center channel and corner channel, the edge rods vibrate in diagonal directions and away from the center with a large amplitude, and the amplitude increase as flow velocity increases. However, flow velocity is not the definitive factor for vibration intensity. The vibration intensity and amplitude of the central rod increase as flow velocity increases, and fuel rods resonate in the studied flow velocities. In this study, not only the relationship between fluid excitation and structural vibration was revealed, but also the vibration characteristics of fuel rods are obtained. This provides a reference for further research and engineering applications in nuclear field.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105664"},"PeriodicalIF":3.3,"publicationDate":"2025-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143420133","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 of the dominant frequency of pressure oscillations and condensing pattern in vertically upward steam bubbles jets under rolling conditions","authors":"Pengbo Wei,&nbsp;Zhenghang Luo,&nbsp;Weixiong Chen,&nbsp;Daotong Chong,&nbsp;Junjie Yan","doi":"10.1016/j.pnucene.2025.105655","DOIUrl":"10.1016/j.pnucene.2025.105655","url":null,"abstract":"<div><div>Steam bubbles direct jet condensing technology, known for their exceptional performance in both the transfers of heat or mass, is popular in static and marine industrial energy systems. Nevertheless, it causes intense pressure oscillations, which can be exacerbated by rolling motions in the marine environments. This can severely impact the safe operation of steam bubbles jet equipment. Therefore, the experimental study is carried out to study the effect of rolling motions on condensing pattern and pressure oscillation frequency of vertically upward bubbles jets. The results reveal that, firstly, the condensing pattern of the vapor bubbles changed, and the rolling motions force the vapor bubbles patterns to roll in the same direction as the rolling motions as compared to the static environment. This rolling behavior of the bubbles causes instability during necking and detachment of bubbles and enhances the heat transfers between the bubbles and the subcooled water, which expedites the condensation rate of the bubbles. This phenomenon explains the second finding: the condensing pressure oscillation dominant frequency of vapor bubbles under rolling conditions is larger than that under static conditions. Meanwhile, with a decrease in these rolling periods and an increasing of the maximum angles of rolling, this dominant frequency of bubbles condensing pressure oscillations increases gradually. Finally, a correlation equation is developed to predict dominant frequency of bubbles condensing pressure oscillations for rolling motions with a bias of ±25%.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105655"},"PeriodicalIF":3.3,"publicationDate":"2025-02-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143403462","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":"Thermal-hydraulic performance of natural circulation system with narrow rectangular channel under heaving condition based on 1D/3D coupling analysis","authors":"Qiang Lian ,&nbsp;Luteng Zhang ,&nbsp;Simiao Tang ,&nbsp;Longxiang Zhu ,&nbsp;Bin Liu ,&nbsp;Zaiyong Ma ,&nbsp;Wan Sun ,&nbsp;Liangming Pan","doi":"10.1016/j.pnucene.2025.105663","DOIUrl":"10.1016/j.pnucene.2025.105663","url":null,"abstract":"<div><div>As the demand for ocean exploration grows, nuclear reactors utilizing natural circulation for power generation have become a significant propulsion force for long-term operations. However, the coolant within the reactor system is influenced by ocean conditions, characterized by passive flow fluctuations. In this study, a one-dimensional (1D)/three-dimensional (3D) coupling analysis method is developed and applied to investigate the impact of heaving conditions on the flow and heat transfer performance of natural circulation with a narrow rectangular channel. The flow and heat transfer characteristics within the narrow rectangular channel, which represents the reactor core, are analyzed in detail using a three-dimensional approach, while the overall natural circulation performance of the system is assessed using a one-dimensional method. Validation is conducted based on pulsating flow experiments and natural circulation experiments. Subsequently, detailed thermal-hydraulic parameters for both the rectangular channel and the natural circulation system are obtained. The results indicate that flow fluctuations caused by the heaving conditions significantly influence the transient performance of both the rectangular channel and the entire natural circulation system. However, the time-averaged flow resistance and heat transfer capability remain unaffected by the heaving conditions. An increase in the heaving period results in a decrease in flow fluctuations within the natural circulation system, and the phase delay between heaving displacement and mass flow flux also diminishes. Conversely, as the heaving amplitude increases, flow fluctuations rise, but this has no effect on the phase delay. Due to the ring effect introduced by the heaving motion, the difference between the main flow zone and the wall zone dominates the transient thermal-hydraulic performance in the narrow rectangular channel. Additionally, the heaving motion does not significantly influence the cycle-averaged flow resistance or the cycle-averaged Nusselt number.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"182 ","pages":"Article 105663"},"PeriodicalIF":3.3,"publicationDate":"2025-02-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143403659","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":"Thermal-hydraulic analysis of sequential disposal scenarios in a spent nuclear fuel repository","authors":"Samuel Park ,&nbsp;Pilhyeon Ju ,&nbsp;Taehoon Park ,&nbsp;Richard I. Foster ,&nbsp;GyoSoon Kim ,&nbsp;Sungyeol Choi","doi":"10.1016/j.pnucene.2025.105662","DOIUrl":"10.1016/j.pnucene.2025.105662","url":null,"abstract":"<div><div>In order to address issues related to Spent Nuclear Fuel (SNF), nations using nuclear energy have considered disposal of SNF in a Deep Geological Repository (DGR) located 500m below ground as most promising option. The Korean government has considered the multi-barrier disposal concept developed by Finland and Sweden and has developed a Korean Reference disposal System (KRS). The main consideration of KRS is how effectively it can isolate SNF from ecosystem, including considerations for thermal, mechanical and material degradations such as illitization of bentonite and canister corrosion. Among these criteria, temperature plays a significant role in maintaining repository integrity, influencing both mechanical and material degradation behavior. Consequently, various trials, including analytical and numerical solutions, have been employed to estimate temperature evolutions. However, a significant temperature difference between the analytical and numerical solutions arises due to the simulation region. Specifically, the analytical solution considers a multi-canister geometry, while current numerical solutions only account for a single-canister due to computational cost. As a result, the single canister model does not accurately reflect heat flux from the adjacent canisters. In this study, a quantitative comparison between analytical and numerical solutions using multi-canister geometries was conducted, and the results showed that the analytical and numerical solutions exhibit similar results. Additionally, four different SNF disposal methods were investigated to identify the optimized disposal method in terms of peak temperature. According to the results, it is suggested that the plop-plop method is the optimized disposal method.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"181 ","pages":"Article 105662"},"PeriodicalIF":3.3,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143388086","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":"An overview of the prediction methods for the heat transfer of supercritical fluids","authors":"Abdullah Alasif ,&nbsp;Andrea Pucciarelli ,&nbsp;Osman Siddiqui ,&nbsp;Afaque Shams","doi":"10.1016/j.pnucene.2025.105654","DOIUrl":"10.1016/j.pnucene.2025.105654","url":null,"abstract":"<div><div>Enhancing coolants or working fluid properties has been of interest for decades. One of the most convenient ways is to use fluids operating at conditions higher than the thermodynamically defined critical point. Such fluids are called supercritical fluids. Supercritical fluids can be used in multiple engineering applications, such as jet propulsion systems, power plants, and nuclear reactors. One of the Generation IV reactors is the Supercritical Water Reactor, which utilizes supercritical water as a coolant. However, the inclusion of supercritical fluids in such industries is constrained by the vague nature of heat transfer when used in thermal systems. Predicting heat transfer is vital, especially in thermally sensitive systems like nuclear reactors. Until now, the scientific community has lacked a generalized and accurate method to predict the heat transfer of these fluids. This paper aims to provide an overview of the recent attempts to understand and predict the heat transfer of supercritical fluids by different methods. These methods include experiments, Computational Fluid Dynamics, and machine learning, which are used to generate models for predicting heat transfer.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"181 ","pages":"Article 105654"},"PeriodicalIF":3.3,"publicationDate":"2025-02-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143388085","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":"Development and validation of a subchannel analysis code for PWRs based on the two-phase and three-field model","authors":"Xinyang Zhu,&nbsp;Jinshun Wang,&nbsp;Chenfeng Bao,&nbsp;Ronghua Chen,&nbsp;Wenxi Tian,&nbsp;Suizheng Qiu","doi":"10.1016/j.pnucene.2025.105653","DOIUrl":"10.1016/j.pnucene.2025.105653","url":null,"abstract":"<div><div>The pressurized water reactor (PWR) continues to be the most prevalently operated reactor type worldwide, with core thermal-hydraulic analysis constituting a crucial aspect of overall reactor design. To achieve a refined distribution of key physical fields within the core and enhance both reactor safety and economic efficiency, the SACOS (Subchannel Analysis Code Of Safety) V3.0 (hereinafter referred to as SACOS) code was developed based on a two-fluid, three-field model at the pin-by-pin level for PWR cores. This code incorporates closure relations such as wall friction, heat transfer, turbulent mixing, and interfacial interaction. It utilizes the finite difference method and the SIMPLE algorithm for numerical discretization and solution, facilitating fine-scale simulations of the entire reactor core. The physical models of the code were validated against experiments such as the GE-3✕3 rod bundle mixing test, the CE-5✕5 heat transfer test, the PNL-7✕7 flow blockage test, the CE-15✕15 inlet stream test and the THTF-8✕8 dispersed flow test, demonstrating good agreement between the computed and experimental results, thereby verifying the accuracy of the models. Finally, the code was applied to steady-state full-core calculations of a typical PWR and transient calculations under SBLOCA conditions for a single assembly, with comparative analysis on the impact of the droplet phase on cladding temperature under high void fraction conditions. The results demonstrate that SACOS is capable of both steady-state and transient core simulations. It can be concluded that the SACOS code provides an effective tool for PWR safety analysis and design.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"181 ","pages":"Article 105653"},"PeriodicalIF":3.3,"publicationDate":"2025-02-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143378266","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":"Performance evaluation of flow boiling CHF correlations in short-heated channel","authors":"Guillaume Gesmier ,&nbsp;Shixian Wang ,&nbsp;Shuichiro Miwa ,&nbsp;Koji Okamoto","doi":"10.1016/j.pnucene.2025.105629","DOIUrl":"10.1016/j.pnucene.2025.105629","url":null,"abstract":"<div><div>This study investigates critical heat flux (CHF) in a horizontal saturated flow boiling water loop with a downward-facing heated surface in a macro-channel (40 mm × 10 mm cross-section) at atmospheric pressure. CHF models and correlations were evaluated against a new dataset spanning mass fluxes from <span><math><mrow><mn>120</mn><mo>–</mo><mn>640</mn><mspace></mspace><mi>k</mi><mi>g</mi><mo>/</mo><mi>m</mi><mo>²</mo><mo>·</mo><mi>s</mi></mrow></math></span> and materials such as non-oxidized and oxidized copper and carbon steel. While several correlations accurately predicted CHF, many overlooked the effects of surface conditions and orientation, exposing gaps in existing methods. The study also examines radiation-induced surface activation (RISA) under electron beam and γ-ray irradiation. Results show γ-ray irradiation enhances CHF by up to 47% across metals, while electron beam irradiation has material-dependent effects: aluminum maintains stable CHF, but copper and silver exhibit an initial decrease followed by an increase at higher doses. This study provides insights of CHF behavior on downward-facing surfaces and for irradiated materials, offering guidance for selecting CHF models and informing nuclear safety applications, particularly in-vessel retention (IVR) strategies.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"181 ","pages":"Article 105629"},"PeriodicalIF":3.3,"publicationDate":"2025-02-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143378267","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}