Progress in Nuclear Energy最新文献_第3页

Data assimilation-based momentum source parameter calibration in subchannel code CUNLUN 基于数据同化的子信道动量源参数标定

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-06 DOI: 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

Z = 0.5 D_{h}

), 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.

为了提高燃油组件子通道内三维流动模拟的精度，将集成卡尔曼滤波器（Ensemble Kalman Filter, EnKF）集成到细网格子通道热工代码CUNLUN中，提出了一种数据同化框架（DA-DRM）。在该框架中，DA-DRM作为整体数据同化方案，EnKF作为核心算法迭代更新模型参数和状态变量。该方法在保持物理一致性的前提下，根据仿真-观测残差的协方差动态标定关键动量源参数并更新状态变量。从而提高了流量预测的局部适应性和全局一致性。通过MATiS-H国际热工基准测试，验证了该方法的有效性。设计了多个观测配置，以系统地评估传感器放置对优化性能的影响。结果表明，基于enkf的DA-DRM框架显著提高了代表性截面上轴向和横向速度的空间一致性。在混合间隔网格下游速度梯度较大的区域（Z=0.5Dh区域），速度预测的均方根误差（RMSE）从0.127降低到0.039，降低了69.6%。进一步的收敛性分析表明，设计良好的观测布局不仅可以提高预测精度，而且可以加速和稳定数据同化过程。简化的配置产生更快的收敛，而更复杂的设置提供更好的鲁棒性。总的来说，所提出的框架为校准子通道模型和提高复杂反应堆部件热工仿真的保真度提供了一种有效和通用的策略。

{"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}

引用次数: 0

Experimental study on CHF with low mass flux in a full-scale rod based on the heated perimeter equivalent method 基于热周长等效法的全尺寸棒材低质量通量CHF实验研究

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-06 DOI: 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.

采用加热周长等效法设计的单杆试验段，研究了低质量通量条件下的临界热流密度（CHF）现象及其相关参数变化趋势。实验结果表明，CHF随质量流量和进口过冷度的增加而增加，而其对压力的响应呈先上升后下降的非线性趋势。在零质量通量条件下，CHF主要发生在试验段的中上部区域，低压时壁温波动明显。这些发现为核反应堆安全设计提供了有价值的见解，表明与棒束配置相比，从加热周长等效单棒获得的CHF值产生了更保守的估计。此外，还提出了一种基于回流准则数的识别回流的新方法，该准则数在湍流流型中超过90，在层流和过渡流型中超过400。实验数据的分析证实了加热周长等效方法在准确捕获实际燃料组件中观察到的CHF行为方面的有效性。本研究有助于改进CHF预测模型，增强对核热工系统低质量流量条件下CHF机理的基本认识。

{"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}

引用次数: 0

Dynamic analysis of China's load capacity factor: The roles of energy heterogeneity and economic globalization through ARDL and machine learning methods 基于ARDL和机器学习方法的中国负荷能力因子动态分析：能源异质性与经济全球化的作用

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-06 DOI: 10.1016/j.pnucene.2025.106222

Zhipeng Yang , Yingjie Wang , Jiacheng Yu , Hanrui Qiu , Mingjun Wang , Wenxi Tian , G.H. Su

This study investigates the dynamic impacts of energy heterogeneity, economic globalization, and gross domestic product on China's Load Capacity Factor (LCF) within the context of its dual carbon goals. Using annual data from 1993 to 2021, the study integrates econometric and machine learning approaches to examine dynamic relationships and forecast future energy trends. The Autoregressive Distributed Lag (ARDL) model is used for empirical analysis, while the Gated Recurrent Unit (GRU) model supports forecasting. The robustness of the ARDL results is verified through cointegration tests, stability diagnostics, and Fully Modified Least Squares (FMOLS) estimation. The long-run results indicate that fossil fuel consumption has the strongest negative impact on LCF, with a coefficient of −1.110, while renewable energy consumption also negatively affects LCF with a coefficient of −0.137. In contrast, nuclear energy consumption, economic globalization, and gross domestic product contribute positively, with coefficients of +0.077, +0.564, and +0.322, respectively. In the short run, nuclear energy consumption (+0.116), renewable energy consumption (−0.275), and gross domestic product (+0.524) exhibit significant effects, whereas fossil fuel consumption and economic globalization primarily influence long-term dynamics. GRU forecasts project a substantial transformation of China's energy structure by 2030: the fossil fuel consumption share is expected to fall to 54.95 %, while nuclear and renewable energy shares are projected to rise to 7.96 % and 37.09 %, respectively. LCF is projected to stabilize overall, with a potential turning point in 2027, followed by a gradual upward trend. These findings underscore the imperative of transitioning toward a lower-carbon energy system to enhance LCF and meet sustainability goals.

本研究探讨了能源异质性、经济全球化和国内生产总值在双重碳目标背景下对中国负荷能力因子的动态影响。该研究使用1993年至2021年的年度数据，将计量经济学和机器学习方法结合起来，研究动态关系并预测未来的能源趋势。实证分析采用自回归分布滞后（ARDL）模型，预测采用门控循环单元（GRU）模型。ARDL结果的稳健性通过协整检验、稳定性诊断和完全修正最小二乘（FMOLS）估计得到验证。长期结果表明，化石燃料消费对LCF的负向影响最大，系数为- 1.110，可再生能源消费也对LCF产生负向影响，系数为- 0.137。核能消费、经济全球化和国内生产总值的贡献为正，系数分别为+0.077、+0.564和+0.322。在短期内，核能消费（+0.116）、可再生能源消费（- 0.275）和国内生产总值（+0.524）表现出显著影响，而化石燃料消费和经济全球化主要影响长期动态。GRU预测，到2030年，中国能源结构将发生重大转变：化石燃料消费份额预计将降至54.95%，而核能和可再生能源份额预计将分别上升至7.96%和37.09%。预计LCF总体将趋于稳定，并可能在2027年出现拐点，随后呈逐步上升趋势。这些发现强调了向低碳能源系统过渡以提高LCF和实现可持续发展目标的必要性。

{"title":"Dynamic analysis of China's load capacity factor: The roles of energy heterogeneity and economic globalization through ARDL and machine learning methods","authors":"Zhipeng Yang , Yingjie Wang , Jiacheng Yu , Hanrui Qiu , Mingjun Wang , Wenxi Tian , G.H. Su","doi":"10.1016/j.pnucene.2025.106222","DOIUrl":"10.1016/j.pnucene.2025.106222","url":null,"abstract":"<div><div>This study investigates the dynamic impacts of energy heterogeneity, economic globalization, and gross domestic product on China's Load Capacity Factor (LCF) within the context of its dual carbon goals. Using annual data from 1993 to 2021, the study integrates econometric and machine learning approaches to examine dynamic relationships and forecast future energy trends. The Autoregressive Distributed Lag (ARDL) model is used for empirical analysis, while the Gated Recurrent Unit (GRU) model supports forecasting. The robustness of the ARDL results is verified through cointegration tests, stability diagnostics, and Fully Modified Least Squares (FMOLS) estimation. The long-run results indicate that fossil fuel consumption has the strongest negative impact on LCF, with a coefficient of −1.110, while renewable energy consumption also negatively affects LCF with a coefficient of −0.137. In contrast, nuclear energy consumption, economic globalization, and gross domestic product contribute positively, with coefficients of +0.077, +0.564, and +0.322, respectively. In the short run, nuclear energy consumption (+0.116), renewable energy consumption (−0.275), and gross domestic product (+0.524) exhibit significant effects, whereas fossil fuel consumption and economic globalization primarily influence long-term dynamics. GRU forecasts project a substantial transformation of China's energy structure by 2030: the fossil fuel consumption share is expected to fall to 54.95 %, while nuclear and renewable energy shares are projected to rise to 7.96 % and 37.09 %, respectively. LCF is projected to stabilize overall, with a potential turning point in 2027, followed by a gradual upward trend. These findings underscore the imperative of transitioning toward a lower-carbon energy system to enhance LCF and meet sustainability goals.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106222"},"PeriodicalIF":3.2,"publicationDate":"2026-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927137","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}

引用次数: 0

Neutronic study of ThF4-UF4-LiF fuel mixture in the molten salt hybrid reactor for 233U denaturing 熔盐混合堆中ThF4-UF4-LiF燃料混合物对233U变性的中子研究

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-06 DOI: 10.1016/j.pnucene.2026.106237

Hacı Mehmet Şahin , Sümer Şahin , Güven Tunç , Hüseyin Şahiner

This study investigates a fusion-fission hybrid reactor concept using a thorium-based molten salt fuel mixture to enhance proliferation resistance and operational sustainability. Neutron transport and reaction rates were modeled using the Monte Carlo N-particle code (MCNP6) with ENDF/B-VIII.0 data. Thorium is mixed homogeneously with 2.25 % depleted uranium (DU) in order to denaturate the ²³³U fuel. The analysis. showed that with 75 % ⁶Li enrichment and a 50 cm coolant layer, the tritium breeding ratio (TBR) remained above 1.05 for a period of four years. The energy multiplication factor (M) increased from 1.88 to 2.2, consistently exceeding the minimum target of 1.5. Under the hard fusion neutron flux, more than 96 % of the plutonium produced was ²³⁹Pu, heavier plutonium isotopes were burnt in situ. The production of the low enriched ²³³U fuel increased to about 12 % after 34 months. These results indicate the technical feasibility of a thorium-based fusion-fission hybrid reactor with improved proliferation resistance, efficient energy multiplication, and sustainable fuel cycle characteristics.

本文研究了一种采用钍基熔盐燃料混合物的聚变-裂变混合反应堆概念，以提高其抗核扩散能力和运行可持续性。中子输运和反应速率采用蒙特卡罗n粒子代码（MCNP6）和ENDF/B-VIII进行建模。0数据。为了使233U燃料变性，钍与2.25%贫铀（DU）均匀混合。分析。结果表明，在6Li浓度为75%、冷却层厚度为50 cm的条件下，氚增殖比（TBR）在4年内保持在1.05以上。能量倍增系数(M)从1.88增加到2.2，持续超过1.5的最低目标。在硬聚变中子通量下，96%以上的钚是239Pu，较重的钚同位素在原地燃烧。34个月后，低富集233U燃料的产量增加到12%左右。这些结果表明，基于钍的聚变-裂变混合反应堆具有更好的抗扩散能力、高效的能量倍增和可持续的燃料循环特性。

{"title":"Neutronic study of ThF4-UF4-LiF fuel mixture in the molten salt hybrid reactor for 233U denaturing","authors":"Hacı Mehmet Şahin , Sümer Şahin , Güven Tunç , Hüseyin Şahiner","doi":"10.1016/j.pnucene.2026.106237","DOIUrl":"10.1016/j.pnucene.2026.106237","url":null,"abstract":"<div><div>This study investigates a fusion-fission hybrid reactor concept using a thorium-based molten salt fuel mixture to enhance proliferation resistance and operational sustainability. Neutron transport and reaction rates were modeled using the Monte Carlo N-particle code (MCNP6) with ENDF/B-VIII.0 data. Thorium is mixed homogeneously with 2.25 % depleted uranium (DU) in order to denaturate the <sup>233</sup>U fuel. The analysis. showed that with 75 % <sup>6</sup>Li enrichment and a 50 cm coolant layer, the tritium breeding ratio (TBR) remained above 1.05 for a period of four years. The energy multiplication factor (M) increased from 1.88 to 2.2, consistently exceeding the minimum target of 1.5. Under the hard fusion neutron flux, more than 96 % of the plutonium produced was <sup>239</sup>Pu, heavier plutonium isotopes were burnt <em>in situ</em>. The production of the low enriched <sup>233</sup>U fuel increased to about 12 % after 34 months. These results indicate the technical feasibility of a thorium-based fusion-fission hybrid reactor with improved proliferation resistance, efficient energy multiplication, and sustainable fuel cycle characteristics.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106237"},"PeriodicalIF":3.2,"publicationDate":"2026-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927144","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}

引用次数: 0

PuO2 surrogate behaviour in a nuclear waste incineration process 核废料焚烧过程中PuO2的替代行为

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-06 DOI: 10.1016/j.pnucene.2025.106229

A. Quintas , P. Charvin , S. Lemonnier , M. Tribet , A. Russello , K. Poizot , H. Pablo , B. Frasca

In this study, the behaviour and fate of plutonium oxide within a nuclear waste incineration reactor was investigated. A comprehensive literature review on the use of PuO₂ surrogate in the nuclear R&D sector was first undertaken to assess different scientific approaches and related limitations. A thorough analysis of the different phenomenologies involved in the incineration process was also performed which resulted in the selection of a set of four oxide surrogates (ZrO₂, CeO₂, Gd₂O₃ and HfO₂) that were eventually used during seven inactive full-scale mock-up test campaigns. After description of the innovative methodology and the experimental set-up, the results obtained were presented. Surrogate retention factors were calculated for a wide range of operating conditions and systematically stands below 1 %. The use of several surrogate oxide powders with different particle size distributions ranging from 0.05 to 120 μm allowed to put in evidence interesting granulometric effects which were supported by some preliminary numerical simulation studies. It has also been confirmed that the fiberglass bag used to wrap the waste load in the incineration reactor plays a key role in the containment of surrogate powders. The data and observations were finally consistent with the existence of a dynamic deposition process on the reactor cooled walls that combines growth and elimination stages. This mechanism requires to be confirmed by further investigations over longer period of time and could open the way to a steady state process operation with a constant equilibrium mass of actinide trapped within the reactor deposit.

在这项研究中，研究了钚氧化物在核废料焚烧反应堆中的行为和命运。首先对在核研发部门使用PuO2替代品进行了全面的文献综述，以评估不同的科学方法和相关局限性。对焚烧过程中涉及的不同现象进行了彻底的分析，结果选择了一组四种氧化物替代品（ZrO2, CeO2， Gd2O3和HfO2），最终在七个非活动的全尺寸模型测试活动中使用。在描述了创新的方法和实验设置之后，给出了得到的结果。在广泛的操作条件下计算了替代保留系数，系统地保持在1%以下。使用几种不同粒径分布（从0.05到120 μm）的替代氧化物粉末，可以证明一些初步数值模拟研究支持的有趣的颗粒效应。还证实，焚烧反应堆中用于包裹废物负荷的玻璃纤维袋在遏制替代粉末方面起着关键作用。这些数据和观测结果最终证实了在反应堆冷却壁上存在一个结合生长和消除阶段的动态沉积过程。这一机制需要通过更长时间的进一步研究来证实，并可能为反应堆沉积物中捕获恒定平衡质量的锕系元素的稳态过程操作开辟道路。

{"title":"PuO2 surrogate behaviour in a nuclear waste incineration process","authors":"A. Quintas , P. Charvin , S. Lemonnier , M. Tribet , A. Russello , K. Poizot , H. Pablo , B. Frasca","doi":"10.1016/j.pnucene.2025.106229","DOIUrl":"10.1016/j.pnucene.2025.106229","url":null,"abstract":"<div><div>In this study, the behaviour and fate of plutonium oxide within a nuclear waste incineration reactor was investigated. A comprehensive literature review on the use of PuO<sub>2</sub> surrogate in the nuclear R&D sector was first undertaken to assess different scientific approaches and related limitations. A thorough analysis of the different phenomenologies involved in the incineration process was also performed which resulted in the selection of a set of four oxide surrogates (ZrO<sub>2</sub>, CeO<sub>2</sub>, Gd<sub>2</sub>O<sub>3</sub> and HfO<sub>2</sub>) that were eventually used during seven inactive full-scale mock-up test campaigns. After description of the innovative methodology and the experimental set-up, the results obtained were presented. Surrogate retention factors were calculated for a wide range of operating conditions and systematically stands below 1 %. The use of several surrogate oxide powders with different particle size distributions ranging from 0.05 to 120 μm allowed to put in evidence interesting granulometric effects which were supported by some preliminary numerical simulation studies. It has also been confirmed that the fiberglass bag used to wrap the waste load in the incineration reactor plays a key role in the containment of surrogate powders. The data and observations were finally consistent with the existence of a dynamic deposition process on the reactor cooled walls that combines growth and elimination stages. This mechanism requires to be confirmed by further investigations over longer period of time and could open the way to a steady state process operation with a constant equilibrium mass of actinide trapped within the reactor deposit.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106229"},"PeriodicalIF":3.2,"publicationDate":"2026-01-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927226","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}

引用次数: 0

Effect of Si addition on the oxidation of a 12Cr1Al ODS alloy in liquid LBE 添加Si对12Cr1Al ODS合金在液态LBE中氧化的影响

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-05 DOI: 10.1016/j.pnucene.2025.106226

You Wang , Zhangjian Zhou , Junqiang Lu , Lingzhi Chen , Haodong Jia , Zhenfeng Tong , Carsten Schroer , Yang Liu , Kai Chen , Zhao Shen , Xiaoqin Zeng

The corrosion behavior of Fe-12Cr-1Al oxide dispersion strengthened (ODS) alloys, with and without 2 wt% Si addition, was examined in oxygen-controlled (10⁻⁶ wt%) liquid lead–bismuth eutectic (LBE) at 650 °C for 2000 h. Both alloys developed protective surface scales that suppressed dissolution attack, covering nearly the entire exposed surface. For the Si-containing alloy, the oxide scale was thinner and more compact, accompanied by the formation of a discontinuous SiO₂ layer at the interface between the outer and inner oxides. This interfacial SiO₂ impeded the outward diffusion of Fe and Cr while facilitating Al outward transport, leading to denser (Fe,Al)₃O₄ spinels in the outermost layer and a reduced defect density within the Al₂O₃-rich inner layer. A dissolution–oxidation–redeposition mechanism was proposed to rationalize the evolution of the multi-layered scales. These findings demonstrate that minor Si additions enhance the compactness and protectiveness of oxide scales on ODS alloys, offering guidance for the design of corrosion-resistant structural materials for LBE-cooled systems.

研究了Fe-12Cr-1Al氧化物弥散强化（ODS）合金在含和不含2 wt% Si的情况下，在650℃的含氧（10 - 6 wt%）铅铋共晶（LBE）中腐蚀2000小时的行为。两种合金都形成了保护表面的鳞片，抑制了溶解攻击，几乎覆盖了整个暴露表面。含硅合金的氧化层更薄、致密，在内外氧化物界面处形成不连续的SiO2层。这种界面SiO2阻碍了Fe和Cr的向外扩散，同时促进了Al的向外迁移，导致最外层（Fe,Al）3O4尖晶石密度增大，而富al2o3的内层缺陷密度减小。提出了溶蚀-氧化-再沉积机制，为多层鳞片的演化提供了理论依据。这些研究结果表明，少量添加Si可以增强ODS合金氧化层的致密性和防护性，为lbe冷却系统耐腐蚀结构材料的设计提供指导。

{"title":"Effect of Si addition on the oxidation of a 12Cr1Al ODS alloy in liquid LBE","authors":"You Wang , Zhangjian Zhou , Junqiang Lu , Lingzhi Chen , Haodong Jia , Zhenfeng Tong , Carsten Schroer , Yang Liu , Kai Chen , Zhao Shen , Xiaoqin Zeng","doi":"10.1016/j.pnucene.2025.106226","DOIUrl":"10.1016/j.pnucene.2025.106226","url":null,"abstract":"<div><div>The corrosion behavior of Fe-12Cr-1Al oxide dispersion strengthened (ODS) alloys, with and without 2 wt% Si addition, was examined in oxygen-controlled (10<sup>−6</sup> wt%) liquid lead–bismuth eutectic (LBE) at 650 °C for 2000 h. Both alloys developed protective surface scales that suppressed dissolution attack, covering nearly the entire exposed surface. For the Si-containing alloy, the oxide scale was thinner and more compact, accompanied by the formation of a discontinuous SiO<sub>2</sub> layer at the interface between the outer and inner oxides. This interfacial SiO<sub>2</sub> impeded the outward diffusion of Fe and Cr while facilitating Al outward transport, leading to denser (Fe,Al)<sub>3</sub>O<sub>4</sub> spinels in the outermost layer and a reduced defect density within the Al<sub>2</sub>O<sub>3</sub>-rich inner layer. A dissolution–oxidation–redeposition mechanism was proposed to rationalize the evolution of the multi-layered scales. These findings demonstrate that minor Si additions enhance the compactness and protectiveness of oxide scales on ODS alloys, offering guidance for the design of corrosion-resistant structural materials for LBE-cooled systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106226"},"PeriodicalIF":3.2,"publicationDate":"2026-01-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927079","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}

引用次数: 0

Study on the melting and relocation behavior of dispersion helical cruciform fuel under accidents 分散螺旋十字形燃料在事故下的熔化和再安置行为研究

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-05 DOI: 10.1016/j.pnucene.2025.106228

Shihao Wu, Xingyu Wang, Yapei Zhang, Yicong Lan, Wenxi Tian, Suizheng Qiu, G.H. Su

The Helical Cruciform Fuel (HCF) has a large heat transfer area and is self-supporting, showing great potential for improving reactor efficiency and safety. However, there is still a lack of research on the melting and relocation behavior of the dispersion HCF under accident conditions. Based on the Incompressible Smoothed Particle Hydrodynamics (ISPH) method, this paper uses the Discrete Phase Model (DPM) and the Discrete Element Model (DEM) to solve the coupled motion of dispersion fuel particles and molten matrix, and further analyzes the dispersion fuel settling behavior in the melted matrix inside the cladding and the relocation behavior of matrix and dispersion fuel outside the cladding. The analysis results indicate that the settling behavior of fuel particles in the fin zone is influenced by the twisted structure, leading to local overheating of the cladding. After the failure of the cladding, the relocation states of the molten material in the cladding groove include the transition from droplet slip to stream flow, followed by the separation of small droplets, ultimately forming a narrow-wide-narrow blockage structure. This paper provides a theoretical basis and guidance for the safety assessment and accident mitigation strategy formulation of the dispersion HCF.

螺旋十字形燃料（HCF）传热面积大，具有自支撑性，在提高反应堆效率和安全性方面具有很大的潜力。然而，对于事故条件下弥散型HCF的熔化和再安置行为的研究仍然缺乏。基于不可压缩光滑颗粒流体力学（ISPH）方法，采用离散相模型（DPM）和离散元模型（DEM）求解了弥散燃料颗粒与熔融基体的耦合运动，并进一步分析了弥散燃料在熔壳内部的沉降行为和熔壳外基质与弥散燃料的重定位行为。分析结果表明，燃料颗粒在翅片区的沉降行为受到扭曲结构的影响，导致包壳局部过热。包层失效后，熔液在包层槽内的重新定位状态包括由液滴滑移过渡到流流，随后小液滴分离，最终形成窄-宽-窄的堵塞结构。本文为弥散式高速公路的安全评价和事故缓解策略的制定提供了理论依据和指导。

{"title":"Study on the melting and relocation behavior of dispersion helical cruciform fuel under accidents","authors":"Shihao Wu, Xingyu Wang, Yapei Zhang, Yicong Lan, Wenxi Tian, Suizheng Qiu, G.H. Su","doi":"10.1016/j.pnucene.2025.106228","DOIUrl":"10.1016/j.pnucene.2025.106228","url":null,"abstract":"<div><div>The Helical Cruciform Fuel (HCF) has a large heat transfer area and is self-supporting, showing great potential for improving reactor efficiency and safety. However, there is still a lack of research on the melting and relocation behavior of the dispersion HCF under accident conditions. Based on the Incompressible Smoothed Particle Hydrodynamics (ISPH) method, this paper uses the Discrete Phase Model (DPM) and the Discrete Element Model (DEM) to solve the coupled motion of dispersion fuel particles and molten matrix, and further analyzes the dispersion fuel settling behavior in the melted matrix inside the cladding and the relocation behavior of matrix and dispersion fuel outside the cladding. The analysis results indicate that the settling behavior of fuel particles in the fin zone is influenced by the twisted structure, leading to local overheating of the cladding. After the failure of the cladding, the relocation states of the molten material in the cladding groove include the transition from droplet slip to stream flow, followed by the separation of small droplets, ultimately forming a narrow-wide-narrow blockage structure. This paper provides a theoretical basis and guidance for the safety assessment and accident mitigation strategy formulation of the dispersion HCF.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106228"},"PeriodicalIF":3.2,"publicationDate":"2026-01-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927081","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}

引用次数: 0

Development of a transient numerical model for contact melting of molten material on a narrow plate component with varying contact thermal resistance 具有不同接触热阻的窄板构件上熔融材料接触熔化瞬态数值模型的建立

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-05 DOI: 10.1016/j.pnucene.2025.106227

Junjie Ma, Wenzhen Chen

During a core meltdown, the contact melting of the crust of a molten pool against narrow plate components within the reactor exhibits characteristics of a complex convective boundary condition. This complexity arises because the early stage is governed by the coupling of contact thermal resistance and heat conduction mechanisms, while the melting phase is influenced by the overall system heat transfer behavior. Thus, accurately modeling this process poses significant challenges. To address this issue, this paper establishes a transient numerical model to simulate the contact melting behavior between a heat source and a narrow plate of phase change material (PCM), based on contact melting theory. The governing equations are coupled with the heat conduction differential equations of both the heat source and the PCM, with full consideration of the role of contact thermal resistance during the conduction stage. This coupled approach avoids traditional strong numerical coupling computations, significantly reducing computational costs. The results demonstrate that the model effectively tracks the dynamic evolution of the melting rate. Furthermore, the study reveals that modifying the contact thermal resistance by introducing a eutectic reaction substantially improves the prediction accuracy of the melting initiation time, highlighting the importance of considering the influence of the eutectic reaction on contact thermal resistance in the model, which plays a key role in enhancing model accuracy.

在堆芯熔毁过程中，熔池外壳与反应堆内窄板组件的接触熔化表现出复杂对流边界条件的特征。这种复杂性的产生是因为早期阶段受接触热阻和热传导机制的耦合控制，而熔化阶段受整个系统传热行为的影响。因此，准确地对这一过程进行建模提出了重大挑战。为了解决这一问题，本文基于接触熔化理论，建立了相变材料窄板与热源接触熔化过程的瞬态数值模型。控制方程与热源和PCM的导热微分方程耦合，充分考虑了接触热阻在传导阶段的作用。这种耦合方法避免了传统的强数值耦合计算，大大降低了计算成本。结果表明，该模型能有效地跟踪熔化速率的动态变化。此外，研究表明，通过引入共晶反应来修正接触热阻，大大提高了熔化起始时间的预测精度，突出了在模型中考虑共晶反应对接触热阻的影响的重要性，这对提高模型精度起着关键作用。

{"title":"Development of a transient numerical model for contact melting of molten material on a narrow plate component with varying contact thermal resistance","authors":"Junjie Ma, Wenzhen Chen","doi":"10.1016/j.pnucene.2025.106227","DOIUrl":"10.1016/j.pnucene.2025.106227","url":null,"abstract":"<div><div>During a core meltdown, the contact melting of the crust of a molten pool against narrow plate components within the reactor exhibits characteristics of a complex convective boundary condition. This complexity arises because the early stage is governed by the coupling of contact thermal resistance and heat conduction mechanisms, while the melting phase is influenced by the overall system heat transfer behavior. Thus, accurately modeling this process poses significant challenges. To address this issue, this paper establishes a transient numerical model to simulate the contact melting behavior between a heat source and a narrow plate of phase change material (PCM), based on contact melting theory. The governing equations are coupled with the heat conduction differential equations of both the heat source and the PCM, with full consideration of the role of contact thermal resistance during the conduction stage. This coupled approach avoids traditional strong numerical coupling computations, significantly reducing computational costs. The results demonstrate that the model effectively tracks the dynamic evolution of the melting rate. Furthermore, the study reveals that modifying the contact thermal resistance by introducing a eutectic reaction substantially improves the prediction accuracy of the melting initiation time, highlighting the importance of considering the influence of the eutectic reaction on contact thermal resistance in the model, which plays a key role in enhancing model accuracy.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106227"},"PeriodicalIF":3.2,"publicationDate":"2026-01-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145927082","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}

引用次数: 0

How much residual water remains in a canister after vacuum drying for spent nuclear fuel storage? 在真空干燥乏核燃料储存罐后，剩余的水有多少？

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-03 DOI: 10.1016/j.pnucene.2025.106231

Ji Hwan Lim , Seung-Hwan Yu , Gyung-sun Chae , Kyung-Wook Shin , Nam-Hee Lee

This study presents a rigorous experimental investigation of the direct quantification of residual water remaining inside spent nuclear fuel (SNF) canisters following vacuum drying—a pivotal operation for limiting long-term corrosion, radiolytic gas generation, and structural degradation during dry storage. Using a purpose-built laboratory facility that replicates realistic canister geometries, we examined initial moisture inventories of 10–77.9 g. Drying employed a carefully modulated pressure-reduction protocol, achieving standard-compliant conditions (<3 torr for 30 min), after which the canisters were backfilled to 2 bar with inert gas. Direct moisture measurements used an AquaVolt 2 electrolytic trace analyzer under controlled helium pressurization, returning highly resolved residual-moisture concentrations in ppmv. Across all scenarios, the final internal moisture converged reproducibly to ∼500–600 ppmv (≈0.15 g), indicating a predictable endpoint under standard-compliant drying, with instrument-traceable repeatability across trials. A bounding comparison placed these values 3–4 × below a conservative worst-case assumption (treating the entire residual head as H2O vapor; ∼0.569 g) yet 15–20 × above estimates based on ambient humidity entrainment (∼0.0089 g). Beyond filling a longstanding empirical gap, these measurements have direct engineering and regulatory significance: they supply defendable inputs for hydrogen source-term, corrosion, and pressure-evolution models; enable risk-informed procedure design and QA/QC traceability; and complement current compliance frameworks (e.g., NUREG-2215; ASTM C1553-21) that rely on pressure-based endpoints. Finally, the integrated methodology—controlled vacuum drying, real-time dew-point tracking, and ultra-trace quantitative moisture analysis—provides a practical blueprint for selective, confirmatory direct-moisture verification in SNF dry-storage operations and is transferable to other high-stakes vacuum-dependent industries.

本研究提出了一项严格的实验研究，对真空干燥后乏核燃料（SNF）罐内剩余水的直接量化进行了研究。真空干燥是限制长期腐蚀、放射性溶解气体产生和干燥储存期间结构降解的关键操作。使用一个专门建造的实验室设施，复制真实的罐几何形状，我们检查了10-77.9 g的初始水分库存。干燥采用精心调整的减压方案，达到符合标准的条件（<； 3torr 30分钟），之后用惰性气体回填到2bar。直接水分测量使用AquaVolt 2电解痕量分析仪在受控氦气加压下，返回高度分解的残余水分浓度，单位为ppmv。在所有情况下，最终的内部水分可重复性地收敛到~ 500-600 ppmv(≈0.15 g)，表明在符合标准的干燥条件下可预测的终点，在试验期间具有仪器可追溯的重复性。边界比较使这些值比保守的最坏情况假设（将整个剩余水头视为H2O蒸汽；~ 0.569 g）低3-4倍，但比基于环境湿度夹带（~ 0.0089 g）的估计高15-20倍。除了填补长期的经验空白之外，这些测量还具有直接的工程和监管意义：它们为氢源期、腐蚀和压力演化模型提供了可靠的输入；确保风险知情的程序设计和QA/QC可追溯性；并补充当前依赖压力端点的合规框架（例如，nurg -2215; ASTM C1553-21）。最后，集成的方法-控制真空干燥，实时露点跟踪和超痕量定量水分分析-为SNF干储存操作中的选择性，验证性直接水分验证提供了实用的蓝图，并可转移到其他高风险的真空依赖行业。

{"title":"How much residual water remains in a canister after vacuum drying for spent nuclear fuel storage?","authors":"Ji Hwan Lim , Seung-Hwan Yu , Gyung-sun Chae , Kyung-Wook Shin , Nam-Hee Lee","doi":"10.1016/j.pnucene.2025.106231","DOIUrl":"10.1016/j.pnucene.2025.106231","url":null,"abstract":"<div><div>This study presents a rigorous experimental investigation of the direct quantification of residual water remaining inside spent nuclear fuel (SNF) canisters following vacuum drying—a pivotal operation for limiting long-term corrosion, radiolytic gas generation, and structural degradation during dry storage. Using a purpose-built laboratory facility that replicates realistic canister geometries, we examined initial moisture inventories of 10–77.9 g. Drying employed a carefully modulated pressure-reduction protocol, achieving standard-compliant conditions (<3 torr for 30 min), after which the canisters were backfilled to 2 bar with inert gas. Direct moisture measurements used an AquaVolt 2 electrolytic trace analyzer under controlled helium pressurization, returning highly resolved residual-moisture concentrations in ppmv. Across all scenarios, the final internal moisture converged reproducibly to ∼500–600 ppmv (≈0.15 g), indicating a predictable endpoint under standard-compliant drying, with instrument-traceable repeatability across trials. A bounding comparison placed these values 3–4 × below a conservative worst-case assumption (treating the entire residual head as H2O vapor; ∼0.569 g) yet 15–20 × above estimates based on ambient humidity entrainment (∼0.0089 g). Beyond filling a longstanding empirical gap, these measurements have direct engineering and regulatory significance: they supply defendable inputs for hydrogen source-term, corrosion, and pressure-evolution models; enable risk-informed procedure design and QA/QC traceability; and complement current compliance frameworks (e.g., NUREG-2215; ASTM C1553-21) that rely on pressure-based endpoints. Finally, the integrated methodology—controlled vacuum drying, real-time dew-point tracking, and ultra-trace quantitative moisture analysis—provides a practical blueprint for selective, confirmatory direct-moisture verification in SNF dry-storage operations and is transferable to other high-stakes vacuum-dependent industries.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106231"},"PeriodicalIF":3.2,"publicationDate":"2026-01-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884612","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}

引用次数: 0

A study on the long-term variation of signal current in rhodium self-powered neutron detectors under in-core burnup conditions 在堆芯内燃耗条件下铑自供电中子探测器信号电流长期变化的研究

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-03 DOI: 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.

作为测量堆芯内中子通量的关键仪器，自供电中子探测器（SPND）的信号电流由于材料的持续燃烧而发生变化。准确计算SPND信号的长期演变对可靠的电力监测和反应堆的安全运行至关重要。本研究的重点是铑自供电中子探测器（Rh-SPND），并在典型的压水堆（PWR）的背景下，使用反向传播神经网络（BPNN）开发了信号电流变化的预测模型。该模型包含了由于燃耗引起的发射器材料组成的变化，并利用分层方法计算燃耗的径向分布。关键参数，如β电子产生率，逃逸概率，γ光子产生，和γ诱导电子逃逸率都考虑在内。结果表明，在中子通量为1 × 1014 n/（cm2·s）的10年周期内，铑丝半径越小，探测器灵敏度变化越大。具体来说，0.1 mm的线材半径会导致51.3%的灵敏度降低，而1 mm的线材半径会导致25.2%的灵敏度降低。所提出的模型能够对rh - spnd进行精确的灵敏度补偿，从而最大限度地减少中子通量测量误差，提高反应堆的安全性和稳定性。

{"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":"2026-01-03","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}

引用次数: 0