Pub 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":"2026-01-10","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 : 2026-01-09DOI: 10.1016/j.pnucene.2025.106230
Jack A. Lanza , Lei R. Cao
Self-powered neutron detectors (SPNDs) have been essential instruments for in-core flux monitoring in light-water reactors (LWRs) for over six decades. This paper provides a comprehensive review of SPND principles, materials, and configurations, along with their roles in the third-generation (Gen-III and Gen-III+) reactor designs. The review compares performance characteristics of common emitter materials such as Rh, V, Co, and Ag, and examines emerging candidates, including Cd, Er, and Hf, for prompt response applications. Developments in modeling, compensation algorithms, and signal processing, including analytical, analog, and digital methods such as Kalman filtering, are summarized to highlight improvements in accuracy and response time. Recent advancements have extended to fast-neutron and gamma-sensitive SPNDs, along with opportunities for integration in digital twin and AI-based reactor monitoring frameworks. The paper also identifies research gaps in detector-level integration, materials optimization, and adaptability for small modular and microreactor environments. These findings underscore the continued importance of SPNDs as reliable, radiation-tolerant sensors within evolving nuclear instrumentation and safety systems.
{"title":"Review of self-powered neutron detectors for reactor instrumentation","authors":"Jack A. Lanza , Lei R. Cao","doi":"10.1016/j.pnucene.2025.106230","DOIUrl":"10.1016/j.pnucene.2025.106230","url":null,"abstract":"<div><div>Self-powered neutron detectors (SPNDs) have been essential instruments for in-core flux monitoring in light-water reactors (LWRs) for over six decades. This paper provides a comprehensive review of SPND principles, materials, and configurations, along with their roles in the third-generation (Gen-III and Gen-III+) reactor designs. The review compares performance characteristics of common emitter materials such as Rh, V, Co, and Ag, and examines emerging candidates, including Cd, Er, and Hf, for prompt response applications. Developments in modeling, compensation algorithms, and signal processing, including analytical, analog, and digital methods such as Kalman filtering, are summarized to highlight improvements in accuracy and response time. Recent advancements have extended to fast-neutron and gamma-sensitive SPNDs, along with opportunities for integration in digital twin and AI-based reactor monitoring frameworks. The paper also identifies research gaps in detector-level integration, materials optimization, and adaptability for small modular and microreactor environments. These findings underscore the continued importance of SPNDs as reliable, radiation-tolerant sensors within evolving nuclear instrumentation and safety systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106230"},"PeriodicalIF":3.2,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927224","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 : 2026-01-09DOI: 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":"2026-01-09","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 : 2026-01-09DOI: 10.1016/j.pnucene.2026.106242
Yunho Kim, Jaehyun Cho
Following the Three Mile Island accident, risk has become a fundamental concept in evaluating nuclear power plant (NPP) safety. Probabilistic Safety Assessment (PSA) is widely used to quantify risk by integrating accident frequencies and radiological consequences. However, the comprehensive PSA process, particularly level 3 consequence assessment using codes such as MACCS, demands significant computational resources and is not well-suited for rapid evaluations during design optimization. This challenge is intensified by recent Risk-Informed and Performance-Based (RIPB) regulatory frameworks, which require comprehensive risk quantification explicitly accounting for accident consequences across diverse accident sequences. To address this limitation, this study proposes a computationally efficient methodology based on a correlation model linking source-term releases to resulting radiological consequences. This approach enables the development of intuitive risk profiles that can enhance the practical utilization of risk information. As a case study, the proposed framework was applied to core damage accident sequences identified in the OPR-1000 level 1 PSA model, focusing on two initiating events: 1) loss of feedwater (LOFW), and 2) small break loss of coolant accident (SLOCA). Frequency-Consequence (F-C) curves were developed to compare the relative risks of the two events and evaluate the effects of severe-accident mitigation measures. The results demonstrate the initial applicability of the proposed methodology and indicate its potential to support more efficient risk-informed safety evaluations and decision-making.
{"title":"Development of a risk profile for accident sequences considering release-to-impact correlation","authors":"Yunho Kim, Jaehyun Cho","doi":"10.1016/j.pnucene.2026.106242","DOIUrl":"10.1016/j.pnucene.2026.106242","url":null,"abstract":"<div><div>Following the Three Mile Island accident, risk has become a fundamental concept in evaluating nuclear power plant (NPP) safety. Probabilistic Safety Assessment (PSA) is widely used to quantify risk by integrating accident frequencies and radiological consequences. However, the comprehensive PSA process, particularly level 3 consequence assessment using codes such as MACCS, demands significant computational resources and is not well-suited for rapid evaluations during design optimization. This challenge is intensified by recent Risk-Informed and Performance-Based (RIPB) regulatory frameworks, which require comprehensive risk quantification explicitly accounting for accident consequences across diverse accident sequences. To address this limitation, this study proposes a computationally efficient methodology based on a correlation model linking source-term releases to resulting radiological consequences. This approach enables the development of intuitive risk profiles that can enhance the practical utilization of risk information. As a case study, the proposed framework was applied to core damage accident sequences identified in the OPR-1000 level 1 PSA model, focusing on two initiating events: 1) loss of feedwater (LOFW), and 2) small break loss of coolant accident (SLOCA). Frequency-Consequence (F-C) curves were developed to compare the relative risks of the two events and evaluate the effects of severe-accident mitigation measures. The results demonstrate the initial applicability of the proposed methodology and indicate its potential to support more efficient risk-informed safety evaluations and decision-making.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106242"},"PeriodicalIF":3.2,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927141","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 : 2026-01-09DOI: 10.1016/j.pnucene.2026.106241
Yanqun Wu , Junfeng Wang , Quan Ma , Mingxing Liu , Meiyuan Chen
The nuclear safety video display unit (SVDU) is a critical human-machine interface in the nuclear reactor protection system, and its cybersecurity is increasingly challenged by the sector's full digitalization. This paper proposes a novel cybersecurity risk assessment method integrating an attack tree model with the STRIDE threat framework to evaluate the risk status of the nuclear SVDU. The approach innovatively incorporates the latest CVSS4.0 standard to define security attributes (attack cost, technical difficulty, discovery difficulty, and impact) for leaf nodes, which are quantified through hierarchical scoring. Crucially, an objective entropy weighting method is employed to calculate the weights of these attributes, effectively eliminating the subjective bias inherent in traditional expert-dependent methods like FAHP. The risk probability is then propagated from leaf to root nodes based on their dependency relationships. Comparative analysis demonstrates that the proposed method offers a more sensitive and discriminating assessment, identifying critical threats that subjective approaches might overlook and providing a robust, data-driven basis for prioritizing security protection strategies.
{"title":"Cybersecurity risk assessment of nuclear SVDU based on attack tree","authors":"Yanqun Wu , Junfeng Wang , Quan Ma , Mingxing Liu , Meiyuan Chen","doi":"10.1016/j.pnucene.2026.106241","DOIUrl":"10.1016/j.pnucene.2026.106241","url":null,"abstract":"<div><div>The nuclear safety video display unit (SVDU) is a critical human-machine interface in the nuclear reactor protection system, and its cybersecurity is increasingly challenged by the sector's full digitalization. This paper proposes a novel cybersecurity risk assessment method integrating an attack tree model with the STRIDE threat framework to evaluate the risk status of the nuclear SVDU. The approach innovatively incorporates the latest CVSS4.0 standard to define security attributes (attack cost, technical difficulty, discovery difficulty, and impact) for leaf nodes, which are quantified through hierarchical scoring. Crucially, an objective entropy weighting method is employed to calculate the weights of these attributes, effectively eliminating the subjective bias inherent in traditional expert-dependent methods like FAHP. The risk probability is then propagated from leaf to root nodes based on their dependency relationships. Comparative analysis demonstrates that the proposed method offers a more sensitive and discriminating assessment, identifying critical threats that subjective approaches might overlook and providing a robust, data-driven basis for prioritizing security protection strategies.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106241"},"PeriodicalIF":3.2,"publicationDate":"2026-01-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927222","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 : 2026-01-07DOI: 10.1016/j.pnucene.2025.106234
D. Timpano , A. Vasiliev , D. Rochman , M. Hursin
The European project Experiments for Validation and Enhancement of the REactor preSsure vessel fluence assessmenT (EVEREST) was launched in 2024 with the goal to employ advanced multiphysics tools to contribute to Long Term Operation (LTO) of nuclear power plants (NPP). The improvement of the reactor pressure vessel (RPV) fluence calculations is one of the overarching aims of this research endeavor, alongside a quantification of the modeling biases and uncertainties related to the standard computational methods. In the framework of the EVEREST project, EPFL has set up a new methodology for vessel fluence calculation based on Polaris/PARCS/Serpent and applied it to the Turkey Point 3 Pressurized Water Reactor (PWR), leveraging on public core design and operating data. This paper aims at presenting the results of the sensitivity analysis and uncertainty quantification performed for this application case. A novel integrated workflow to propagate uncertainties in the whole computational chain of vessel fluence calculations is described. The analysis covers core follow modeling, source preparation and shielding calculations, tackling the uncertainty due to material densities, geometry and nuclear data. Both standard Sandwich Formula (SF) and Total Monte-Carlo (TMC) techniques have been used for this scope. The UQ performed in this study revealed an analytical uncertainty associated with the fast flux at the thermal shield and the reactor pressure vessel location between 8 and 10%. The main contributors are the uncertainties on nuclear data and manufacturing tolerances in shielding calculations. This work contributes to the implementation of the Best Estimate Plus Uncertainty (BEPU) approach for assessing radiation damage in nuclear reactor structural materials.
{"title":"Uncertainty quantification for a vessel fluence calculation: A PWR case study","authors":"D. Timpano , A. Vasiliev , D. Rochman , M. Hursin","doi":"10.1016/j.pnucene.2025.106234","DOIUrl":"10.1016/j.pnucene.2025.106234","url":null,"abstract":"<div><div>The European project Experiments for Validation and Enhancement of the REactor preSsure vessel fluence assessmenT (EVEREST) was launched in 2024 with the goal to employ advanced multiphysics tools to contribute to Long Term Operation (LTO) of nuclear power plants (NPP). The improvement of the reactor pressure vessel (RPV) fluence calculations is one of the overarching aims of this research endeavor, alongside a quantification of the modeling biases and uncertainties related to the standard computational methods. In the framework of the EVEREST project, EPFL has set up a new methodology for vessel fluence calculation based on Polaris/PARCS/Serpent and applied it to the Turkey Point 3 Pressurized Water Reactor (PWR), leveraging on public core design and operating data. This paper aims at presenting the results of the sensitivity analysis and uncertainty quantification performed for this application case. A novel integrated workflow to propagate uncertainties in the whole computational chain of vessel fluence calculations is described. The analysis covers core follow modeling, source preparation and shielding calculations, tackling the uncertainty due to material densities, geometry and nuclear data. Both standard Sandwich Formula (SF) and Total Monte-Carlo (TMC) techniques have been used for this scope. The UQ performed in this study revealed an analytical uncertainty associated with the fast flux at the thermal shield and the reactor pressure vessel location between 8 and 10%. The main contributors are the uncertainties on nuclear data and manufacturing tolerances in shielding calculations. This work contributes to the implementation of the Best Estimate Plus Uncertainty (BEPU) approach for assessing radiation damage in nuclear reactor structural materials.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106234"},"PeriodicalIF":3.2,"publicationDate":"2026-01-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927143","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 : 2026-01-06DOI: 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":"2026-01-06","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 : 2026-01-06DOI: 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":"2026-01-06","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 : 2026-01-06DOI: 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":"2026-01-06","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}
Pub Date : 2026-01-06DOI: 10.1016/j.pnucene.2025.106219
Binzhuo Xia , Di Liu , Kui Zhang , Jinquan Yan , Zhan Liu , Fanting Xia , Ronghua Chen , Fujun Gan , Wenxi Tian , Bo Yang , Suizheng Qiu
This study investigates the critical heat flux (CHF) phenomenon and the relevant parametric trends under low mass flux conditions, employing a single rod test section designed using the heated perimeter equivalent approach. Experimental results indicate that CHF increases with increasing mass flux and inlet subcooling, whereas its response to pressure exhibits a non-linear trend, initially rising and subsequently decreasing. Under zero mass flux conditions, CHF predominantly occurs in the middle and upper regions of the test section, accompanied by pronounced wall temperature fluctuations at low pressures. These findings provide valuable insights for nuclear reactor safety design, demonstrating that CHF values obtained from a heated perimeter equivalent single rod yield more conservative estimates compared to those derived from rod bundle configurations. Moreover, a novel methodology for identifying backflow was developed based on the backflow criterion number, which exceeds 90 in the turbulent regime and surpasses 400 in laminar and transitional flow regimes. Analysis of the experimental data confirms the effectiveness of the heated perimeter equivalent approach in accurately capturing the CHF behavior observed in real fuel assemblies. This research contributes to the improvement of CHF prediction models and enhances the fundamental understanding of CHF mechanisms under low mass flux conditions in nuclear thermal-hydraulic systems.
{"title":"Experimental study on CHF with low mass flux in a full-scale rod based on the heated perimeter equivalent method","authors":"Binzhuo Xia , Di Liu , Kui Zhang , Jinquan Yan , Zhan Liu , Fanting Xia , Ronghua Chen , Fujun Gan , Wenxi Tian , Bo Yang , Suizheng Qiu","doi":"10.1016/j.pnucene.2025.106219","DOIUrl":"10.1016/j.pnucene.2025.106219","url":null,"abstract":"<div><div>This study investigates the critical heat flux (CHF) phenomenon and the relevant parametric trends under low mass flux conditions, employing a single rod test section designed using the heated perimeter equivalent approach. Experimental results indicate that CHF increases with increasing mass flux and inlet subcooling, whereas its response to pressure exhibits a non-linear trend, initially rising and subsequently decreasing. Under zero mass flux conditions, CHF predominantly occurs in the middle and upper regions of the test section, accompanied by pronounced wall temperature fluctuations at low pressures. These findings provide valuable insights for nuclear reactor safety design, demonstrating that CHF values obtained from a heated perimeter equivalent single rod yield more conservative estimates compared to those derived from rod bundle configurations. Moreover, a novel methodology for identifying backflow was developed based on the backflow criterion number, which exceeds 90 in the turbulent regime and surpasses 400 in laminar and transitional flow regimes. Analysis of the experimental data confirms the effectiveness of the heated perimeter equivalent approach in accurately capturing the CHF behavior observed in real fuel assemblies. This research contributes to the improvement of CHF prediction models and enhances the fundamental understanding of CHF mechanisms under low mass flux conditions in nuclear thermal-hydraulic systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106219"},"PeriodicalIF":3.2,"publicationDate":"2026-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927225","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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