Annals of Nuclear Energy最新文献_第3页

Aerosol generation characteristics during laser cutting of carbon and stainless steel surfaces for nuclear power plant decommissioning 核电厂退役用碳钢和不锈钢表面激光切割时气溶胶产生特性

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-22 DOI: 10.1016/j.anucene.2026.112156

Avadhesh Kumar Sharma , Ruicong Xu , Zeeshan Ahmed , Ravinder Kumar , Shuichiro Miwa , Shunichi Suzuki

The decommissioning of the Fukushima Daiichi Nuclear Power Plant presents significant challenges due to high radiation levels and the need for safe and efficient fuel debris retrieval. Laser cutting is a promising technique for decontamination and dismantling, but it generates submicron-sized radioactive aerosols, necessitating precise aerosol management strategies. This study investigates aerosol generation during laser cutting of carbon steel (CS) and stainless steel (SS) surfaces under varying power levels and surface coatings at the Mitsubishi Heavy Industries Ltd. (MHI) Research & Innovation Center utilizing class-4 laser. Experimental results indicate that increasing laser power leads to higher aerosol concentrations, particularly for larger aerosols, while smaller aerosol concentrations decline. This effect is more pronounced in CS surfaces than in SS. Coated surfaces, especially with zirconium dioxide (ZrO₂), exhibit higher aerosol generation at elevated power levels, suggesting an intensified laser-material interaction. The experimental results highlight the role of coating composition in aerosol generation and importance of dispersion control methods during decommissioning. The analysis of aerosol dispersion results can give insight to enhance radiation worker safety, protect sensitive electronics, and improve the effectiveness of remote laser-based decontamination in high-dose environments.

由于高辐射水平和需要安全和有效地回收燃料碎片，福岛第一核电站的退役提出了重大挑战。激光切割是一种很有前途的去污和拆除技术，但它会产生亚微米大小的放射性气溶胶，需要精确的气溶胶管理策略。本研究在三菱重工（MHI）研究与创新中心使用4级激光对碳钢（CS）和不锈钢（SS）表面进行激光切割时，在不同功率水平和表面涂层下产生的气溶胶。实验结果表明，激光功率的增加导致气溶胶浓度的增加，特别是对于较大的气溶胶，而较小的气溶胶浓度下降。这种效应在CS表面比SS表面更为明显。涂层表面，特别是氧化锆（ZrO2）涂层表面，在高功率水平下表现出更高的气溶胶产生，表明激光与材料的相互作用加剧。实验结果强调了涂层成分在气溶胶产生中的作用以及退役过程中分散控制方法的重要性。对气溶胶分散结果的分析可以为提高辐射工作人员的安全、保护敏感电子设备以及提高高剂量环境中远程激光去污的有效性提供见解。

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引用次数: 0

SaraGR-K: a deterministic transport-based transient analysis code for prismatic gas-cooled micro-reactors SaraGR-K：基于确定性输运的柱形气冷微堆瞬态分析代码

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-22 DOI: 10.1016/j.anucene.2026.112130

Qiming Yang , Yuan Yuan , Dong Huang , Youqi Zheng , Guoming Liu

This paper presents SaraGR-K, a three-dimensional code developed for coupled neutronics/thermal-hydraulics (N/T-H) transient analysis, specifically designed for prismatic gas-cooled microreactors. The code solves the 3D neutron transport equation in the time–space domain using a predictor–corrector quasi-static (PCQS) method. To further enhance computational efficiency in 3D transport simulations, SaraGR-K incorporates a fixed-source scaling factor technique to accelerate the convergence of transient fixed-source equations and utilizes a pipelined parallelization strategy to expedite transport calculations. Temperature distributions within reactor materials are subsequently calculated using a 1D fluid dynamics model coupled with a 1D solid heat conduction model, ensuring consistent integration with the neutronics solution.

With these computational and thermal–hydraulic models in place, SaraGR-K was verified using numerical benchmarks problems and by comparing its results with those from established reference codes. Following the verification, it was applied to the transient analysis of a prismatic gas-cooled microreactor, demonstrating its feasibility for full-core transient simulations in such systems.

本文介绍了SaraGR-K，一个用于耦合中子/热工-水力学（N/T-H）瞬态分析的三维代码，专门为柱形气冷微堆设计。该程序采用预测校正准静态（PCQS）方法求解三维中子输运方程。为了进一步提高3D传输模拟的计算效率，SaraGR-K采用了固定源缩放因子技术来加速瞬态固定源方程的收敛，并利用流水线并行化策略来加快传输计算。随后，使用一维流体动力学模型与一维固体热传导模型耦合计算反应堆材料内部的温度分布，确保与中子溶液的一致集成。有了这些计算模型和热水力模型，SaraGR-K就可以通过数值基准问题进行验证，并将其结果与已有的参考代码进行比较。在验证之后，将其应用于柱形气冷微堆的瞬态分析，证明了其在此类系统中进行全堆芯瞬态模拟的可行性。

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引用次数: 0

Research and application of intelligent loading pattern search for pressurized water reactors 压水堆智能加载模式搜索的研究与应用

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-22 DOI: 10.1016/j.anucene.2026.112117

Peng Sitao, Wei Jinfeng, Yao Jianfan, Yang Shuoyan, Li Wenhuai, Yang Zhengyu, Huang Jie, Mei Gaohui, Huang Yunten, Zhao Changyoug, Yu Chao, Wang Ting, Li Jinggang

Commercial pressurized water reactor (PWR) nuclear power plants require loading pattern (LP) design every 12–24 months. To address the time-consuming nature of routine loading pattern searches and their strong reliance on engineer expertise, this study proposes an AI-based automated loading pattern search method and develops the corresponding intelligent loading pattern design software KAPOK. This software automates tasks previously requiring experienced engineers through automation and intelligent technologies. KAPOK employs semi-empirical methods for automatic selection of spent fuel assemblies, heuristic algorithms for pattern search, and integrates neural network surrogate models for rapid assessment of key refueling parameters with the comprehensive evaluation capabilities of nuclear design software. Typically, KAPOK can identify highly optimized solutions within 30 min, achieving exceptional efficiency through automated processing. Validated extensively against historical cycle patterns, KAPOK has been successfully deployed in routine refueling designs at nuclear power plants.

商用压水堆（PWR）核电站每12-24 个月需要进行一次负荷模式（LP）设计。为了解决常规加载模式搜索耗时和对工程师专业知识依赖程度高的问题，本研究提出了一种基于人工智能的加载模式自动搜索方法，并开发了相应的智能加载模式设计软件KAPOK。该软件通过自动化和智能技术自动化了以前需要经验丰富的工程师的任务。KAPOK采用半经验方法对乏燃料组件进行自动选择，采用启发式算法进行模式搜索，并将神经网络代理模型与核设计软件的综合评估能力相结合，实现关键换料参数的快速评估。通常，KAPOK可以在30 min内确定高度优化的解决方案，通过自动化处理实现卓越的效率。根据历史循环模式进行了广泛的验证，木棉已经成功地部署在核电站的常规换料设计中。

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引用次数: 0

Investigation on the bypass flow characteristics of pebble bed core Gap structure in HTGR based on the fluid network method 基于流体网络方法的高温堆球层岩心间隙结构旁通流动特性研究

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-22 DOI: 10.1016/j.anucene.2026.112123

Wei Liu , Jinsong Guo , Qiong Wu , Yuhang Liu , Lixin Yang , Chaojun Wang

Typical gap structures in the core of a pebble-bed high-temperature gas-cooled reactor (HTGR), including narrow core gaps and the Carbon Brick-Metal Reactor Gap, can induce helium bypass flow, which leads to local overheating that may affect the thermal efficiency and even safety. To investigate the effect of gap structures on bypass flow, the fluid network approach is used and a systematic core network model is established accordingly to represent the spatial features of the HTGR. With the core network model, the effect of gap structures, including the narrow gap width, keyway structure, and Carbon Brick-Metal Reactor Gap, on the bypass flow is analyzed. Results indicate that the keyway structure is the most effective factor in controlling the bypass flow and optimizing its flow resistance can significantly reduce the bypass flow ratio; the effect of narrow gap width on the bypass flow is limited while it is practically challenging to achieve a Carbon Brick-Metal Reactor Gap width range that can effectively restrict bypass flow. In addition, a temperature-based bypass flow severity index for characterizing the degree of impact of bypass flow on the temperature within the core is also proposed.

典型的球床高温气冷堆（HTGR）堆芯间隙结构，包括狭窄的堆芯间隙和碳砖-金属堆间隙，会诱发氦旁路流动，导致局部过热，影响热效率甚至安全。为了研究间隙结构对旁通流的影响，采用流体网络方法，建立了系统的核心网络模型来表征高温高压堆的空间特征。利用核心网络模型，分析了窄间隙宽度、键槽结构、碳砖-金属反应器间隙等间隙结构对旁通流量的影响。结果表明：键槽结构是控制旁通流量的最有效因素，优化键槽流动阻力可显著降低旁通流量比；窄间隙宽度对旁通流量的影响有限，而实现能够有效限制旁通流量的碳砖-金属反应器间隙宽度范围具有实际挑战性。此外，还提出了一种基于温度的旁路流动严重程度指标，用于表征旁路流动对堆芯内温度的影响程度。

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引用次数: 0

Uncertainty quantification with MCMC algorithm for subchannel program based on steady and transient void fraction experiments 基于稳态和瞬态空隙率实验的子信道程序的MCMC算法不确定性量化

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-22 DOI: 10.1016/j.anucene.2026.112154

Hanyu Luo , Jiayu Long , Meiqi Song , Xiaojing Liu

Subchannel codes can provide higher-resolution results in reactor analysis than system codes, yet their uncertainty quantification (UQ) methodologies are less mature compared to those in thermal–hydraulic system codes. Current approaches predominantly focus on steady-state conditions with limited coverage of transient scenarios, and typically employ forward propagation based on parameter distributions derived from expert knowledge. This study develops a UQ framework based on COBRA-IV, combing inverse uncertainty quantification (IUQ) and forward propagation to analyse void fraction distributions for both steady-state and transient conditions of the Nuclear Power Engineering Corporation (NUPEC) Pressurized water reactor Subchannel and Bundle Tests (PSBT) benchmark. The framework also evaluates the effects of initial distributions of key parameters (slip ratio, turbulent mixing coefficient), and surrogate models including back-propagation neural network (BPNN), radial basis function (RBF) network, and Gaussian process regression (GPR). For steady-state conditions, all surrogate models successfully envelope the experimental data within consistent accuracy ranges, with calibrated root mean square error (RMSE) values ranging from 4.0% to 4.3%. In transient analysis, the uncertainty bands of void fraction remain within 10% variation relative to experimental values, indication low sensitivity to initial parameter distributions. All models capture over 88% of void fraction data points. After calibration, the BPNN model achieves an RMSE of 4.64%, while RBF and GPR reach 4.57% and 5.21%, respectively- a 40–45% improvement over the baseline RMSE of 8.15%. These results confirm the framework’s reliability across different implementation choices and its ability to enhance computational accuracy.

子通道码在反应堆分析中可以提供比系统码更高的分辨率结果，但其不确定性量化（UQ）方法与热工系统码相比尚不成熟。目前的方法主要集中在稳态条件下，对瞬态场景的覆盖有限，并且通常采用基于专家知识导出的参数分布的前向传播。本研究基于COBRA-IV开发了一个UQ框架，结合逆不确定性量化（IUQ）和正向传播，分析了核动力工程公司（NUPEC）压水堆子通道和束试验（PSBT）基准稳态和瞬态条件下的空洞分数分布。该框架还评估了关键参数（滑移比、湍流混合系数）初始分布的影响，以及包括反向传播神经网络（BPNN）、径向基函数（RBF）网络和高斯过程回归（GPR）在内的代理模型。对于稳态条件，所有替代模型都成功地将实验数据封装在一致的精度范围内，校准的均方根误差（RMSE）值在4.0%至4.3%之间。在瞬态分析中，相对于实验值，孔隙率的不确定度范围保持在10%以内，表明对初始参数分布的敏感性较低。所有模型都捕获了超过88%的空隙率数据点。校正后，BPNN模型的RMSE为4.64%，RBF和GPR分别达到4.57%和5.21%，比基线RMSE 8.15%提高了40-45%。这些结果证实了该框架在不同实现选择中的可靠性及其提高计算精度的能力。

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引用次数: 0

A spatiotemporal EIM for long-term xenon dynamics prediction in commercial reactor cores 用于商业反应堆堆芯长期氙动力学预测的时空EIM

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-22 DOI: 10.1016/j.anucene.2026.112167

Ziyi Zheng , Lichen Yuan , Xianping Zhong , Lianjie Wang , Jiangyu Wang , Deng Pan , Jun Luo , Helin Gong

For the safe and stable operation of nuclear reactors, accurate prediction of xenon-induced power oscillations is crucial. However, existing prediction methods are constrained by limited sensor coverage, high computational complexity, and substantial algorithmic cost. This study proposes a space–time prediction framework based on the Empirical Interpolation Method (EIM). A large set of parameterized space–time simulations is first generated; unlike typical reduced-basis approaches that only process spatial snapshots, each full space–time solution is treated as one snapshot in a greedy EIM procedure. This yields interpolation bases and a set of space–time magic points, whose temporal locations are restricted to past instants with available measurements. During operation, observations at these magic points are assimilated through a low-dimensional linear system to reconstruct the space–time evolution of the physical fields and to predict future states. The proposed approach is assessed under various power conditions of the HPR1000 reactor core, where it demonstrates superior accuracy and efficiency compared with traditional methods, as well as robustness to measurement noise. Its stability in long-term predictions indicates strong potential as a practical tool for real-time monitoring and prediction of reactor core behavior.

为了保证核反应堆的安全稳定运行，氙致功率振荡的准确预测至关重要。然而，现有的预测方法受到传感器覆盖范围有限、计算复杂度高和算法成本高的限制。本文提出了一种基于经验插值法（EIM）的时空预测框架。首先生成了大量的参数化时空模拟；与只处理空间快照的典型减基方法不同，每个完整的时空解在贪婪EIM过程中被视为一个快照。这产生了插值基础和一组时空魔法点，它们的时间位置被限制在具有可用测量的过去瞬间。在运行过程中，通过低维线性系统吸收这些神奇点的观测结果，重建物理场的时空演化并预测未来状态。在HPR1000堆芯的各种功率条件下对该方法进行了评估，与传统方法相比，该方法具有更高的精度和效率，并且对测量噪声具有鲁棒性。它在长期预测中的稳定性表明了作为实时监测和预测反应堆堆芯行为的实用工具的强大潜力。

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引用次数: 0

A tuning-free POD-FV-ROM with automatic boundary enforcement for practical thermal-hydraulic applications 一种无调谐pod - fv - from，具有自动边界执行，适用于实际热工应用

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-21 DOI: 10.1016/j.anucene.2026.112163

Yang Di , Zhang Chunyu , Lin Jiming , Ding Peng

Reduced order models (ROMs) have been widely adopted to accelerate high-fidelity simulations while retaining essential predictive accuracy. However, two gaps persist for POD-FV-ROMs in the context of thermal-hydraulic problems. First, when Dirichlet and Neumann boundaries coexist, a robust and effective strategy to enforce boundary constraints in the ROM is still missing. Second, the ROM performance has not been convincingly demonstrated on practical three-dimensional cases with complex geometries. To close these gaps, a ROM combining POD-Galerkin projection with supremizer stabilization and POD-RBF interpolation was investigated, together with a penalty formulation for boundary treatment. Two novel a priori estimators are proposed to determine penalty factors (PFs) without manual tuning, one based on residuals relevant to projected snapshots and one derived from optimization to a domain-wise error. The entire ROM framework was assessed on an 2 × 2 helical cruciform fuel assembly discretized into around 24 million cells with parameterized boundary conditions. The results demonstrated that, under appropriate PFs, the ROM delivered satisfying accuracy while achieving a speed up of five orders of magnitude.

降阶模型（ROMs）已被广泛应用于加速高保真仿真，同时保持基本的预测精度。然而，在热液压问题的背景下，pod - fv - rom仍然存在两个空白。首先，当Dirichlet和Neumann边界共存时，仍然缺乏一种鲁棒且有效的策略来强制ROM中的边界约束。其次，ROM的性能还没有在具有复杂几何形状的实际三维情况下得到令人信服的证明。为了缩小这些差距，研究了结合POD-Galerkin投影与上位稳定器和POD-RBF插值的ROM，以及边界处理的惩罚公式。提出了两个新的先验估计器来确定惩罚因子（PFs），而无需手动调优，一个基于与投影快照相关的残差，另一个基于优化到域智能误差。整个ROM框架在一个2 × 2螺旋十字形燃料组件上进行了评估，该组件在参数化边界条件下离散为大约2400万个单元。结果表明，在适当的pf下，ROM提供了令人满意的精度，同时实现了五个数量级的速度提升。

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引用次数: 0

Methodology for significance determination across multiple risk metrics using novel importance measures 使用新颖的重要性度量跨多个风险度量的显著性确定方法

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-21 DOI: 10.1016/j.anucene.2026.112132

Restu Kojo

In risk-informed regulation, the significance of systems, structures, and components (SSCs) is assessed using multiple risk metrics, requiring a systematic method to determine whether SSC degradation has a greater impact on Level 1 or Level 2 probabilistic risk assessments (PRA). A key issue is that risk significance in Level 2 often exceeds that of Level 1 PRA due to the order-of-magnitude difference between target values for core damage frequency (CDF) and that of containment failure frequency (CFF). To address this, a new methodology was developed, including a novel measure—risk difference achievement worth (RDAW)—which enables transparent comparisons across different PRAs. This methodology was applied to large-scale PRA models and confirmed the consistency of the significance comparison results. In summary, a mathematically formulated methodology for comparing significance across multiple PRAs, which is applicable to large-scale practical models, has been established.

在风险知情监管中，系统、结构和组件（SSC）的重要性使用多种风险指标进行评估，需要一种系统的方法来确定SSC退化对1级或2级概率风险评估（PRA）的影响更大。一个关键问题是，由于堆芯损坏频率（CDF）的目标值与安全壳失效频率（CFF）的目标值之间的数量级差异，二级PRA的风险重要性往往超过一级PRA。为了解决这个问题，我们开发了一种新的方法，包括一种新的测量方法——风险差异成就值（RDAW）——它可以在不同的pra之间进行透明的比较。将该方法应用于大规模PRA模型，验证了显著性比较结果的一致性。综上所述，已经建立了一种适用于大规模实际模型的数学公式方法，用于比较多个pra之间的显著性。

引用次数: 0

Performance assessment of spent fuel storage with duplex SiC composite cladding in Pressurized water reactors 压水堆双相SiC复合材料包层乏燃料贮存性能评价

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-21 DOI: 10.1016/j.anucene.2026.112157

Ruixiao Zhang , Yuhang Niu , Yanan He , Zhiwei Lu , Yingwei Wu , Jing Zhang , G.H. Su

Silicon carbide (SiC) composite cladding is a promising candidate for accident-tolerant fuel applications, exhibiting distinct thermomechanical and failure behaviors compared to metallic cladding such as Zircaloy under spent fuel storage conditions. Existing safety criteria for spent fuel storage may therefore be inapplicable to SiC-based assemblies. In this study, the updated FROBA code was used to simulate the performance of high-burnup SiC cladding spent fuel after reactor operation, considering long-term wet and dry storage as well as short-term off-normal dry storage. Results show that SiC cladding performs well during wet storage. During dry storage, the cladding stress slightly exceeds the 90 MPa reference limit for Zr cladding. Due to the probabilistic failure characteristics of monolithic SiC, this corresponds to an estimated failure probability of approximately 0.3%. Elevated internal rod pressure is the main contributor to this risk. The cladding peak temperature of 400℃ at the rod top indicates the highest failure risk. Lower storage temperature limits and optimized pressure balance can effectively mitigate failure.

与锆合金等金属包层相比，碳化硅（SiC）复合包层在乏燃料储存条件下表现出不同的热力学和失效行为，是一种很有希望用于耐事故燃料应用的候选材料。因此，现有的乏燃料储存安全标准可能不适用于硅基组件。在本研究中，采用更新后的FROBA代码对高燃耗SiC包层乏燃料在反应堆运行后的性能进行了模拟，同时考虑了长期干湿储存和短期非正常干储存。结果表明，SiC包层具有良好的湿储存性能。在干贮存过程中，包层应力略高于Zr包层的90mpa参考极限。由于单片碳化硅的概率失效特性，这相当于估计的失效概率约为0.3%。杆内压力升高是造成这种风险的主要原因。棒顶包层峰值温度为400℃，失效风险最高。较低的储存温度限制和优化的压力平衡可以有效地减轻故障。

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引用次数: 0

Assessment of self-interrogation safeguards Signatures for pebble bed reactor fuel 球床反应堆燃料自我询问保障标志的评估

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2026-01-21 DOI: 10.1016/j.anucene.2026.112155

Austen Ocanas , Farheen Naqvi , Sudeep Mitra , Jeremy Osborn

Many next generation reactors propose the use of nontraditional nuclear fuel in the form of pebbles filled with thousands of fuel particles. While these next generation pebble bed reactors have been in development for decades, recent support for nuclear energy has bolstered the process, with many reactor designs proposed for deployment in the coming decade. An issue facing pebble bed reactors is safeguarding the fuel itself. Research on the burnup measurement systems is still evolving for developing a nondestructive assay method to quantify the amount of fissile material present in a used fuel pebble, creating a challenge for international safeguards design. The study presented here investigates the potential of neutron self-interrogation of spent fuel pebbles as an innovative method to implement materials accountability in these advanced reactors. Through reactor physics and fuel burnup simulations, spent fuel pebble material compositions are found and a method is developed to equate the delayed gamma-ray emissions resulting from the fissions induced by the pebble neutrons to the mass of key fissile actinides. The feasibility of this self-interrogation method is assessed, leading to the conclusion that the method is suitable for use in a passive counting mode employing a 4π detection geometry. As an example, the mass of ²³⁵U, ²³⁸U, ²³⁹Pu and ²⁴¹Pu can be predicted at a precision of 4.1%, 0.86%, 13% and 13%, respectively, when measuring 100 end-of-life spent fuel pebbles over approximately 12 days.

许多新一代反应堆建议使用非传统的核燃料，即充满数千个燃料颗粒的鹅卵石。虽然这些下一代球床反应堆已经开发了几十年，但最近对核能的支持推动了这一进程，许多反应堆的设计都被提议在未来十年部署。球床反应堆面临的一个问题是如何保护燃料本身。燃耗测量系统的研究仍在不断发展，以开发一种无损分析方法来量化乏燃料卵石中存在的裂变物质的数量，这对国际保障设计提出了挑战。本文提出的研究探讨了乏燃料卵石中子自探询的潜力，作为在这些先进反应堆中实施材料问责制的创新方法。通过反应堆物理和燃料燃耗模拟，得到了乏燃料球团材料的组成，并提出了一种将球团中子诱导裂变产生的延迟伽马射线辐射与关键可裂变锕系元素质量等同起来的方法。评估了这种自我询问方法的可行性，得出结论，该方法适用于采用4π检测几何形状的被动计数模式。例如，在大约12天的时间内测量100个寿命终止的乏燃料鹅卵石，可以分别以4.1%、0.86%、13%和13%的精度预测235U、238U、239Pu和241Pu的质量。

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引用次数: 0