Pub Date : 2026-03-01Epub Date: 2026-01-06DOI: 10.1016/j.pnucene.2025.106235
Ze Zhu , Xiaojie Guo , Zhiwu Ke , Kelong Zhang , Pengfei Wang
The false water level caused by the so-called “swell and shrink”, and the measurement errors, are the two major challenges in the water level control of U-tube steam generators (UTSGs). This paper proposes a fuzzy adaptive sliding mode controller (FASMC), which effectively mitigates the adverse impacts of false water levels, thus improving the water level control performance. Firstly, an ideal SMC (ISMC) was designed based on a detailed nonlinear UTSG model. Then, to reduce the adverse impacts of measurement errors, a water level estimator was constructed based on the backstepping and Lyapunov's second method, and an adaptive SMC (ASMC) was designed. Finally, the fuzzy algorithm was used to calibrate the hyperparameters in the ASMC online. The simulation results show that the established ISMC, ASMC, and FASMC outperform the PID controller under different operating conditions, with the integrated absolute errors (IAEs) of water level reduced by at least 5.68 %. And the FASMC has the best performance, with the IAE reduced by up to 74.8 %. This demonstrates the effectiveness and superiors of the proposed FASMC, which can provide a valuable reference for mitigating the adverse impacts of false water level in engineering practices and improving the UTSG water level control performance.
{"title":"Fuzzy adaptive sliding mode control of U-tube steam generator water level based on a nonlinear dynamic model","authors":"Ze Zhu , Xiaojie Guo , Zhiwu Ke , Kelong Zhang , Pengfei Wang","doi":"10.1016/j.pnucene.2025.106235","DOIUrl":"10.1016/j.pnucene.2025.106235","url":null,"abstract":"<div><div>The false water level caused by the so-called “swell and shrink”, and the measurement errors, are the two major challenges in the water level control of U-tube steam generators (UTSGs). This paper proposes a fuzzy adaptive sliding mode controller (FASMC), which effectively mitigates the adverse impacts of false water levels, thus improving the water level control performance. Firstly, an ideal SMC (ISMC) was designed based on a detailed nonlinear UTSG model. Then, to reduce the adverse impacts of measurement errors, a water level estimator was constructed based on the backstepping and Lyapunov's second method, and an adaptive SMC (ASMC) was designed. Finally, the fuzzy algorithm was used to calibrate the hyperparameters in the ASMC online. The simulation results show that the established ISMC, ASMC, and FASMC outperform the PID controller under different operating conditions, with the integrated absolute errors (IAEs) of water level reduced by at least 5.68 %. And the FASMC has the best performance, with the IAE reduced by up to 74.8 %. This demonstrates the effectiveness and superiors of the proposed FASMC, which can provide a valuable reference for mitigating the adverse impacts of false water level in engineering practices and improving the UTSG water level control performance.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106235"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927142","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}
Pub Date : 2026-03-01Epub Date: 2026-01-06DOI: 10.1016/j.pnucene.2025.106233
Shilong Shi , Yaoshuang Wan , Guofeng Qu , Runyu Zhang , Jiajian Song , Xuanhao Huang , Jijun Yang , Zhihui Li , Tu Lan , Songdong Ding , Yuanyou Yang , Jiali Liao , Wen Feng , Jing Peng , Ning Liu
Understanding the α-radiolysis of extractants is crucial for their applications in nuclear fuel reprocessing. However, current methodologies face challenges in conducting α-irradiation experiments and elucidating the underlying mechanisms. Herein, a 4–7 MeV 4He2+ beam provided by a CS-30 cyclotron was employed as a fast, convenient, and versatile α-irradiation source to simulate actual radiative scenarios in nuclear fuel reprocessing, which was firstly employed to investigate the α-radiolysis of tri-iso-amyl phosphate (TiAP) in alkane diluent, an alternative extractant in Plutonium Uranium Recovery by Extraction. The dominant radiolysis products of TiAP including hydrogen, methane and di-iso-amyl phosphate (DiAP) were qualitatively and quantitatively determined by GC and HR-MS. Ehrenfest dynamic simulations and DFT calculations revealed the ionization of TiAP induced by electronic stopping is the dominant process in its α-radiolysis process, as confirmed by experimental results. The C‒O bond cleavage in TiAP, leading to the formation of DiAP, is attributed to the decomposition of TiAP+ and TiAP∗, as well as reactions between TiAP and ·OH or secondary electrons. Finally, the radiolysis mechanism of TiAP in alkane diluent was proposed based on the analysis of radiolytic products and multi-time-scale theoretical calculations. This study paves the way for advancing research on the α-radiolysis of extractants for spent fuel reprocessing.
{"title":"The α-radiolysis behavior and mechanism of tri-isoamyl phosphate (TiAP): New insights from 4He2+ beam irradiation experiments and theoretical calculations","authors":"Shilong Shi , Yaoshuang Wan , Guofeng Qu , Runyu Zhang , Jiajian Song , Xuanhao Huang , Jijun Yang , Zhihui Li , Tu Lan , Songdong Ding , Yuanyou Yang , Jiali Liao , Wen Feng , Jing Peng , Ning Liu","doi":"10.1016/j.pnucene.2025.106233","DOIUrl":"10.1016/j.pnucene.2025.106233","url":null,"abstract":"<div><div>Understanding the α-radiolysis of extractants is crucial for their applications in nuclear fuel reprocessing. However, current methodologies face challenges in conducting α-irradiation experiments and elucidating the underlying mechanisms. Herein<strong>,</strong> a 4–7 MeV <sup>4</sup>He<sup>2+</sup> beam provided by a CS-30 cyclotron was employed as a fast, convenient, and versatile α-irradiation source to simulate actual radiative scenarios in nuclear fuel reprocessing, which was firstly employed to investigate the α-radiolysis of tri-iso-amyl phosphate (TiAP) in alkane diluent, an alternative extractant in Plutonium Uranium Recovery by Extraction. The dominant radiolysis products of TiAP including hydrogen, methane and di-iso-amyl phosphate (DiAP) were qualitatively and quantitatively determined by GC and HR-MS. Ehrenfest dynamic simulations and DFT calculations revealed the ionization of TiAP induced by electronic stopping is the dominant process in its α-radiolysis process, as confirmed by experimental results. The C‒O bond cleavage in TiAP, leading to the formation of DiAP, is attributed to the decomposition of TiAP<sup>+</sup> and TiAP∗, as well as reactions between TiAP and ·OH or secondary electrons. Finally, the radiolysis mechanism of TiAP in alkane diluent was proposed based on the analysis of radiolytic products and multi-time-scale theoretical calculations. This study paves the way for advancing research on the α-radiolysis of extractants for spent fuel reprocessing.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106233"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927139","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}
Pub Date : 2026-03-01Epub Date: 2026-01-03DOI: 10.1016/j.pnucene.2025.106232
Zhi-qi Guo , Wen-hua Yang , Jing-yi Han , Shuo Zhang , Chun-hui Zhang , Zhan Li , Ding-jun Zhu , Jian-xiong Shao
As a key instrument for measuring neutron flux within a reactor core, the signal current of a Self-Powered Neutron Detector (SPND) is subject to change due to continuous material burnup. Accurately accounting for the long-term evolution of SPND signals is essential for reliable power monitoring and safe reactor operation. This study focuses on the Rhodium Self-Powered Neutron Detector (Rh-SPND) and, within the context of a typical Pressurized Water Reactor (PWR), develops a predictive model for signal current variation using a Back Propagation Neural Network (BPNN). The model incorporates changes in emitter material composition due to burnup and utilizes a layered approach to calculate the radial distribution of burnup. Key parameters such as beta electron generation rate, escape probability, gamma photon generation, and gamma-induced electron escape rate are all taken into account. Results show that over a 10-year period at a neutron flux of 1 × 1014 n/(cm2·s), the smaller the radius of the rhodium wire, the greater the variation in detector sensitivity. Specifically, a 0.1 mm wire radius results in a 51.3 % sensitivity reduction, while a 1 mm wire radius yields a 25.2 % reduction. The proposed model enables accurate sensitivity compensation for Rh-SPNDs over time, thereby minimizing neutron flux measurement errors and enhancing reactor safety and stability.
{"title":"A study on the long-term variation of signal current in rhodium self-powered neutron detectors under in-core burnup conditions","authors":"Zhi-qi Guo , Wen-hua Yang , Jing-yi Han , Shuo Zhang , Chun-hui Zhang , Zhan Li , Ding-jun Zhu , Jian-xiong Shao","doi":"10.1016/j.pnucene.2025.106232","DOIUrl":"10.1016/j.pnucene.2025.106232","url":null,"abstract":"<div><div>As a key instrument for measuring neutron flux within a reactor core, the signal current of a Self-Powered Neutron Detector (SPND) is subject to change due to continuous material burnup. Accurately accounting for the long-term evolution of SPND signals is essential for reliable power monitoring and safe reactor operation. This study focuses on the Rhodium Self-Powered Neutron Detector (Rh-SPND) and, within the context of a typical Pressurized Water Reactor (PWR), develops a predictive model for signal current variation using a Back Propagation Neural Network (BPNN). The model incorporates changes in emitter material composition due to burnup and utilizes a layered approach to calculate the radial distribution of burnup. Key parameters such as beta electron generation rate, escape probability, gamma photon generation, and gamma-induced electron escape rate are all taken into account. Results show that over a 10-year period at a neutron flux of 1 × 10<sup>14</sup> n/(cm<sup>2</sup>·s), the smaller the radius of the rhodium wire, the greater the variation in detector sensitivity. Specifically, a 0.1 mm wire radius results in a 51.3 % sensitivity reduction, while a 1 mm wire radius yields a 25.2 % reduction. The proposed model enables accurate sensitivity compensation for Rh-SPNDs over time, thereby minimizing neutron flux measurement errors and enhancing reactor safety and stability.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106232"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927080","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}
Pub Date : 2026-03-01Epub Date: 2026-01-10DOI: 10.1016/j.pnucene.2026.106238
Zhiwen Dai , Donghui Zhang , Yuting Yang , Zhiwei Zhou , Chao Lin , Xiuli Xue , Xintai Yu , Dalin Zhang , Xiyong Chen , Shangang Cao , Songping Wang , Chengwen Xing , Shuiwen Jiang
Pool-type sodium-cooled fast reactors (SFR) have become one of the main selections of Generation-IV reactors due to large thermal inertia and inherent safety, which solve the future shortage of natural uranium and the disposal challenges of spent nuclear fuel (SNF). The decay heat removal system (DHRS) is one of the most important safety systems and must be highly reliable. This study illustrates the design and innovations of the DHRS on the China Fast Reactor. A thermal-hydraulic analysis was conducted using the system program (named ERAC) under station blackout (SBO) conditions, and key parameters of the natural circulation process were evaluated. China's fast reactor design is innovative in many respects, and its novel DHRS design ensures the reactor's safety during emergencies. The analysis results show that the DHRS system operates effectively and that the calculations align with the design goals. Under natural circulation, the peak temperature reached approximately 592 °C at 1000 s. As natural circulation progressed, the core outlet temperature gradually decreased; by 5000 s, the average core fuel outlet temperature was 574 °C. The design of the core throttling component meets the requirements and can provide sufficient natural circulation. This study could provide a valuable reference for the design of SFRs.
{"title":"Innovative design and safety evaluation of the decay heat removal system for the China fast reactor","authors":"Zhiwen Dai , Donghui Zhang , Yuting Yang , Zhiwei Zhou , Chao Lin , Xiuli Xue , Xintai Yu , Dalin Zhang , Xiyong Chen , Shangang Cao , Songping Wang , Chengwen Xing , Shuiwen Jiang","doi":"10.1016/j.pnucene.2026.106238","DOIUrl":"10.1016/j.pnucene.2026.106238","url":null,"abstract":"<div><div>Pool-type sodium-cooled fast reactors (SFR) have become one of the main selections of Generation-IV reactors due to large thermal inertia and inherent safety, which solve the future shortage of natural uranium and the disposal challenges of spent nuclear fuel (SNF). The decay heat removal system (DHRS) is one of the most important safety systems and must be highly reliable. This study illustrates the design and innovations of the DHRS on the China Fast Reactor. A thermal-hydraulic analysis was conducted using the system program (named ERAC) under station blackout (SBO) conditions, and key parameters of the natural circulation process were evaluated. China's fast reactor design is innovative in many respects, and its novel DHRS design ensures the reactor's safety during emergencies. The analysis results show that the DHRS system operates effectively and that the calculations align with the design goals. Under natural circulation, the peak temperature reached approximately 592 °C at 1000 s. As natural circulation progressed, the core outlet temperature gradually decreased; by 5000 s, the average core fuel outlet temperature was 574 °C. The design of the core throttling component meets the requirements and can provide sufficient natural circulation. This study could provide a valuable reference for the design of SFRs.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106238"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927223","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}
Pub Date : 2026-03-01Epub Date: 2025-12-30DOI: 10.1016/j.pnucene.2025.106205
Jieming Hou , Bo Kuang , Shirui Li , Wenjun Hu
This study experimentally investigated the mixed convective heat transfer characteristics of liquid lead-bismuth eutectic (LBE) under buoyancy-aided flow condition. Over 30 steady-state tests were conducted on the Multi-Application LBE Thermal-Hydraulic test facility (MATH). By independently varying heating power, tube diameter and loop resistance, the combined influence of forced and natural convection was systematically explored. Dimensionless parameters (Gr, Re, Nu) reveal that buoyancy effects remain negligible when < 0.01, with Nu close to the existing forced convection correlation. As buoyancy increases, heat transfer decreases, and the flow enters the mixed-convection regime. An empirical correlation for the mixed-convective heat transfer of lead-bismuth eutectic with buoyancy correction was proposed based on theoretical derivation and experimental data. Deviations between predicted and experimental data remain within ±10 %. The results improve the accuracy of thermal-hydraulic simulations for LBE-cooled fast reactors under natural-circulation accident scenarios and provide a validated benchmark for further studies at higher heat fluxes and different flow orientations.
实验研究了浮力辅助流动条件下液态铅铋共晶（LBE）混合对流换热特性。在多用途LBE热液试验装置（MATH）上进行了30多次稳态试验。通过单独改变加热功率、管径和回路阻力，系统探索了强迫对流和自然对流的综合影响。无因次参数（Gr, Re, Nu）表明，当GrRe2 <； 0.01时，浮力的影响可以忽略不计，Nu接近现有的强迫对流相关。随着浮力的增加，换热减少，气流进入混合对流状态。基于理论推导和实验数据，提出了铅铋共晶混合对流换热的浮力修正经验关系式。预测数据与实验数据之间的偏差保持在±10%以内。研究结果提高了lbe冷却快堆在自然循环事故情景下的热工模拟精度，并为进一步研究高热流通量和不同流动方向下的热工模拟提供了验证基准。
{"title":"Study on mixed convection heat transfer characteristics of liquid lead-bismuth eutectic","authors":"Jieming Hou , Bo Kuang , Shirui Li , Wenjun Hu","doi":"10.1016/j.pnucene.2025.106205","DOIUrl":"10.1016/j.pnucene.2025.106205","url":null,"abstract":"<div><div>This study experimentally investigated the mixed convective heat transfer characteristics of liquid lead-bismuth eutectic (LBE) under buoyancy-aided flow condition. Over 30 steady-state tests were conducted on the Multi-Application LBE Thermal-Hydraulic test facility (MATH). By independently varying heating power, tube diameter and loop resistance, the combined influence of forced and natural convection was systematically explored. Dimensionless parameters (<em>Gr</em>, Re, <em>Nu</em>) reveal that buoyancy effects remain negligible when <span><math><mrow><mfrac><mtext>Gr</mtext><msup><mtext>Re</mtext><mn>2</mn></msup></mfrac></mrow></math></span> < 0.01, with <em>Nu</em> close to the existing forced convection correlation. As buoyancy increases, heat transfer decreases, and the flow enters the mixed-convection regime. An empirical correlation for the mixed-convective heat transfer of lead-bismuth eutectic with buoyancy correction was proposed based on theoretical derivation and experimental data. Deviations between predicted and experimental data remain within ±10 %. The results improve the accuracy of thermal-hydraulic simulations for LBE-cooled fast reactors under natural-circulation accident scenarios and provide a validated benchmark for further studies at higher heat fluxes and different flow orientations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106205"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884671","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}
Pub Date : 2026-03-01Epub Date: 2026-01-01DOI: 10.1016/j.pnucene.2025.106223
Jianwen Huo , Mingrun Ling , Junling Wang , Ying Zhou
In this study, we develop a search technique involving heterogeneous robots to achieve a rapid and accurate search for a lost radioactive source by effectively utilizing radiation data. The system includes robots with different levels of radiation resistance. Two fixed robots with lower radiation resistance are positioned at the edge of the environment to collect low-value radiation data and apply a maximum likelihood estimation algorithm based on the natural gradient (NG-MLE) for localization. Concurrently, a mobile robot with high radiation resistance enters the warning area to collect high-value radiation data, and the localization of the radioactive source is performed based on particle filtering and prior information obtained from the fixed robots. The movement of the mobile robot is directed by a decision-making algorithm based on Fisher information distance (FID). Simulation results show that the proposed algorithm achieves higher localization accuracy and requires fewer radiation detections.
{"title":"Heterogeneity: A multi-robot radioactive source search technique","authors":"Jianwen Huo , Mingrun Ling , Junling Wang , Ying Zhou","doi":"10.1016/j.pnucene.2025.106223","DOIUrl":"10.1016/j.pnucene.2025.106223","url":null,"abstract":"<div><div>In this study, we develop a search technique involving heterogeneous robots to achieve a rapid and accurate search for a lost radioactive source by effectively utilizing radiation data. The system includes robots with different levels of radiation resistance. Two fixed robots with lower radiation resistance are positioned at the edge of the environment to collect low-value radiation data and apply a maximum likelihood estimation algorithm based on the natural gradient (NG-MLE) for localization. Concurrently, a mobile robot with high radiation resistance enters the warning area to collect high-value radiation data, and the localization of the radioactive source is performed based on particle filtering and prior information obtained from the fixed robots. The movement of the mobile robot is directed by a decision-making algorithm based on Fisher information distance (FID). Simulation results show that the proposed algorithm achieves higher localization accuracy and requires fewer radiation detections.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106223"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884609","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}
Pub Date : 2026-03-01Epub Date: 2026-01-09DOI: 10.1016/j.pnucene.2025.106214
Ayodeji A. Ala , Lorlornyo Abusah , Shouxu Qiao , Bin Ye , Sichao Tan
Precise prediction of the flow pattern and proper estimation of the void fraction distribution are needed to design and optimize multiphase reactors and bubble columns. This research presents the cross-sectional and averaged void fraction distributions of adiabatic air-water mixtures across square-to-square channel cross-section expansion and contraction, with gas superficial velocity (jg) ranging from 0.003 m/s to 0.0054 m/s and water superficial velocity (jl) from 0.08 m/s to 0.36 m/s. A high-speed imaging system captures the bubble information, while key liquid phase flow features are extracted using the particle image velocimetry technique. Changing jg and jl have different effects on the bubble size distributions. In addition to the slowing flow after the expansion plane, the higher void fraction near the walls before the expansion influences the void fraction distribution in the expansion area. Void fraction distribution in channels with contraction displays a wall and core peaking profile. Conversely, a reduction in the contraction ratio in the channel increases the wall-peaking void fraction compared to the core. The decreasing averaged void fraction across the contraction transitions from linear to polynomial as the contraction in the channel increases. A code was developed to estimate the void fraction distribution before the contraction plane in a square channel, and the output was compared with available data.
{"title":"Void fraction migration in developing bubbly flow through sudden cross-section contraction and expansion in square channels","authors":"Ayodeji A. Ala , Lorlornyo Abusah , Shouxu Qiao , Bin Ye , Sichao Tan","doi":"10.1016/j.pnucene.2025.106214","DOIUrl":"10.1016/j.pnucene.2025.106214","url":null,"abstract":"<div><div>Precise prediction of the flow pattern and proper estimation of the void fraction distribution are needed to design and optimize multiphase reactors and bubble columns. This research presents the cross-sectional and averaged void fraction distributions of adiabatic air-water mixtures across square-to-square channel cross-section expansion and contraction, with gas superficial velocity (<em>j</em><sub><em>g</em></sub>) ranging from 0.003 m/s to 0.0054 m/s and water superficial velocity (<em>j</em><sub><em>l</em></sub>) from 0.08 m/s to 0.36 m/s. A high-speed imaging system captures the bubble information, while key liquid phase flow features are extracted using the particle image velocimetry technique. Changing <em>j</em><sub><em>g</em></sub> and <em>j</em><sub><em>l</em></sub> have different effects on the bubble size distributions. In addition to the slowing flow after the expansion plane, the higher void fraction near the walls before the expansion influences the void fraction distribution in the expansion area. Void fraction distribution in channels with contraction displays a wall and core peaking profile. Conversely, a reduction in the contraction ratio in the channel increases the wall-peaking void fraction compared to the core. The decreasing averaged void fraction across the contraction transitions from linear to polynomial as the contraction in the channel increases. A code was developed to estimate the void fraction distribution before the contraction plane in a square channel, and the output was compared with available data.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106214"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927140","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}
Pub Date : 2026-03-01Epub Date: 2026-01-06DOI: 10.1016/j.pnucene.2026.106236
Hongwei Jiang , Xian Zhang , Guangliang Chen , Zhaofei Tian , Jinchao Li , Hao Qian , Xinli Yin , Hang Wang
To enhance the accuracy of 3D flow simulations in fuel assembly subchannels, a data assimilation framework (DA-DRM) is proposed by integrating the Ensemble Kalman Filter (EnKF) into the fine-mesh subchannel thermal-hydraulic code CUNLUN. In this framework, DA-DRM serves as the overall data assimilation scheme, while the EnKF functions as the core algorithm to iteratively update model parameters and state variables. This approach dynamically calibrates key momentum source parameters and updates the state variables based on the covariance of simulation–observation residuals, while maintaining physical consistency.As a result, both local adaptability and global consistency of flow predictions are improved.The method is validated against the MATiS-H international thermal-hydraulic benchmark. Multiple observation configurations are designed to systematically assess the impact of sensor placement on optimization performance. Results show that the EnKF-based DA-DRM framework significantly improves the spatial agreement of both axial and lateral velocities across representative cross-sections. In regions with steep velocity gradients downstream of the mixing spacer grid (region ), the root mean square error (RMSE) of velocity predictions is reduced from 0.127 to 0.039, corresponding to a 69.6 % reduction.Further convergence analysis reveals that well-designed observation layouts not only enhance prediction accuracy but also accelerate and stabilize the data assimilation process. Simplified configurations yield faster convergence, while more complex setups offer improved robustness. Overall, the proposed framework provides an effective and generalizable strategy for calibrating subchannel models and improving the fidelity of thermal-hydraulic simulations in complex reactor components.
{"title":"Data assimilation-based momentum source parameter calibration in subchannel code CUNLUN","authors":"Hongwei Jiang , Xian Zhang , Guangliang Chen , Zhaofei Tian , Jinchao Li , Hao Qian , Xinli Yin , Hang Wang","doi":"10.1016/j.pnucene.2026.106236","DOIUrl":"10.1016/j.pnucene.2026.106236","url":null,"abstract":"<div><div>To enhance the accuracy of 3D flow simulations in fuel assembly subchannels, a data assimilation framework (DA-DRM) is proposed by integrating the Ensemble Kalman Filter (EnKF) into the fine-mesh subchannel thermal-hydraulic code CUNLUN. In this framework, DA-DRM serves as the overall data assimilation scheme, while the EnKF functions as the core algorithm to iteratively update model parameters and state variables. This approach dynamically calibrates key momentum source parameters and updates the state variables based on the covariance of simulation–observation residuals, while maintaining physical consistency.As a result, both local adaptability and global consistency of flow predictions are improved.The method is validated against the MATiS-H international thermal-hydraulic benchmark. Multiple observation configurations are designed to systematically assess the impact of sensor placement on optimization performance. Results show that the EnKF-based DA-DRM framework significantly improves the spatial agreement of both axial and lateral velocities across representative cross-sections. In regions with steep velocity gradients downstream of the mixing spacer grid (region <span><math><mrow><mi>Z</mi><mo>=</mo><mn>0.5</mn><msub><mi>D</mi><mi>h</mi></msub></mrow></math></span>), the root mean square error (RMSE) of velocity predictions is reduced from 0.127 to 0.039, corresponding to a 69.6 % reduction.Further convergence analysis reveals that well-designed observation layouts not only enhance prediction accuracy but also accelerate and stabilize the data assimilation process. Simplified configurations yield faster convergence, while more complex setups offer improved robustness. Overall, the proposed framework provides an effective and generalizable strategy for calibrating subchannel models and improving the fidelity of thermal-hydraulic simulations in complex reactor components.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106236"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927138","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 investigation of turbulent flow in a rod bundle with flow blockage","authors":"Denggao Chen, Jingliang Bi, Yanping Huang, Dewen Yuan, Yuanfeng Zan, Jianjun Xu","doi":"10.1016/j.pnucene.2025.106215","DOIUrl":"10.1016/j.pnucene.2025.106215","url":null,"abstract":"","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106215"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884668","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}
Pub Date : 2026-03-01Epub Date: 2025-12-22DOI: 10.1016/j.pnucene.2025.106196
Ji Hwan Lim
This review critically examines advanced methodologies for drying spent nuclear fuel (SNF), emphasizing the optimization of vacuum drying strategies and the role of forced helium drying (FHD). SNF drying is vital for safe storage and prevents issues like moisture-induced corrosion and hydrogen radiolysis. Traditional vacuum drying remains a staple, yet challenges like ice formation and ineffective moisture removal in damaged fuel necessitate innovations like FHD, which ensures thorough moisture elimination through active gas circulation. Empirical studies and enhanced simulations have refined understanding, showing automation and real-time moisture sensors could streamline drying, improve safety, and reduce time. Also, developments in hydride reorientation in zirconium cladding due to high temperatures during drying highlight the need for precise thermal management to prevent cladding embrittlement. Recent research emphasizes tailored drying criteria for diverse fuel types, focusing on damage-prone fuel requiring extended, heat-assisted drying strategies or supportive agents like hydrogen getters to mitigate gas build-up risks. Monitoring technologies are advancing to include dew point sensors and acoustic devices for real-time drying assessment. Thus, while vacuum methods are robust, an international collaborative effort is essential to standardize best practices for complex scenarios involving high-burnup and damaged SNF, ensuring fuel integrity and regulatory compliance across the nuclear industry.
{"title":"A critical review of spent nuclear fuel drying","authors":"Ji Hwan Lim","doi":"10.1016/j.pnucene.2025.106196","DOIUrl":"10.1016/j.pnucene.2025.106196","url":null,"abstract":"<div><div>This review critically examines advanced methodologies for drying spent nuclear fuel (SNF), emphasizing the optimization of vacuum drying strategies and the role of forced helium drying (FHD). SNF drying is vital for safe storage and prevents issues like moisture-induced corrosion and hydrogen radiolysis. Traditional vacuum drying remains a staple, yet challenges like ice formation and ineffective moisture removal in damaged fuel necessitate innovations like FHD, which ensures thorough moisture elimination through active gas circulation. Empirical studies and enhanced simulations have refined understanding, showing automation and real-time moisture sensors could streamline drying, improve safety, and reduce time. Also, developments in hydride reorientation in zirconium cladding due to high temperatures during drying highlight the need for precise thermal management to prevent cladding embrittlement. Recent research emphasizes tailored drying criteria for diverse fuel types, focusing on damage-prone fuel requiring extended, heat-assisted drying strategies or supportive agents like hydrogen getters to mitigate gas build-up risks. Monitoring technologies are advancing to include dew point sensors and acoustic devices for real-time drying assessment. Thus, while vacuum methods are robust, an international collaborative effort is essential to standardize best practices for complex scenarios involving high-burnup and damaged SNF, ensuring fuel integrity and regulatory compliance across the nuclear industry.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106196"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841096","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}
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