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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 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 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。
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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 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 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
Techno-Economic optimization of sandstone uranium Mining: A Case study of uranium content per square meter 砂岩铀矿开采技术经济优化——以每平方米含铀量为例
IF 2.3 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-20 DOI: 10.1016/j.anucene.2026.112125
Jiabing Li , Chuanfei Zhang , Xiangxue Zhang , Meifang Chen , Mingtao Jia
This study introduces an improved non-dominated sorting genetic algorithm II (INSGA-II) to optimize the boundary delineation of sandstone-type uranium deposits by determining the threshold of uranium content per square meter (UCPSM, kg/m2) for minable units. A multi-objective optimization model was developed to maximize both economic and resource benefits, which was solved using the INSGA-II. Key enhancements include: (1) population initialization via symmetric Latin hypercube design (SLHD); (2) adaptive mutation and crossover parameters. Applied to real data from a Chinese mining area, the model and algorithm demonstrated practical effectiveness. The Pareto solution set derived from the optimization enabled the determination of UCPSM thresholds, supporting a novel mining-area boundary definition method based on aggregating minable units and unlocking the resource and potential economic value of idle uranium deposits. This approach offers a new decision-making tool for sandstone-type uranium mining area design.
通过确定可采单位每平方米铀含量阈值(UCPSM, kg/m2),引入改进的非主导分选遗传算法II (INSGA-II),优化砂岩型铀矿边界圈定。建立了以经济效益和资源效益最大化为目标的多目标优化模型,并利用INSGA-II进行求解。主要改进包括:(1)通过对称拉丁超立方体设计(SLHD)进行种群初始化;(2)自适应变异和交叉参数。将该模型和算法应用于某矿区的实际数据,验证了该模型和算法的实用性。基于优化得到的Pareto解集能够确定UCPSM阈值,支持基于可采单元聚合的矿区边界定义新方法,解锁闲置铀矿床的资源和潜在经济价值。该方法为砂岩型铀矿区设计提供了一种新的决策工具。
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引用次数: 0
Efficient critical safety analysis model for PWR fuel assembly under transport accidents by neutron worth iteration 基于中子值迭代的压水堆燃料组件运输事故临界安全分析模型
IF 2.3 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-20 DOI: 10.1016/j.anucene.2026.112158
Xinling Dai , Dechang Cai , Jin Cai
Criticality safety analysis is essential for fuel assembly transport, as it ensures subcriticality under all potential accident scenarios. Traditional methods are computationally expensive, requiring hundreds of input cases. The Neutron Worth Iteration method was developed to efficiently determine conservative k envelopes for PWR fuel assemblies under transport impacts. By iteratively adjusting the configuration of fuel rods according to the neutron worth distribution within a fuel assembly, the method optimizes rod positioning to maximize k. This method significantly reduces computational effort while maintaining a slight k deviation of less than 2.62. This approach enables rapid yet rigorous criticality safety assessments.
临界安全分析对燃料组件运输至关重要,因为它可以确保在所有潜在的事故情景下都处于亚临界状态。传统的方法在计算上很昂贵,需要数百个输入案例。为了有效地确定压水堆燃料组件在输运冲击下的保守k∞包线,提出了中子价值迭代法。该方法根据燃料组件内的中子值分布,迭代调整燃料棒的配置,优化燃料棒的位置,使k∞最大化。该方法在保持k∞偏差小于2.62‰的情况下,显著减少了计算量。这种方法可以实现快速而严格的临界安全评估。
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引用次数: 0
Evaluation of tantalum–tungsten–oxygen compounds as lead-free radiation shielding materials 钽钨氧化合物作为无铅辐射屏蔽材料的评价
IF 2.3 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-20 DOI: 10.1016/j.anucene.2026.112141
H.C. Manjunatha , B.M. Sankarshan , P.S. Damodara Gupta , L. Seenappa , K.N. Sridhar , R. Munirathnam
This study investigates alternative materials to lead for radiation shielding, addressing the need for safer and more effective options. Traditional materials like lead, although effective due to their high atomic number, are toxic and pose environmental risks. The study explores a set of tantalum–tungsten–oxygen (Ta–W–O) compounds, including TaW2O3, TaW2O4, Ta2W2O5, TaWO3, and others. These compounds offer promising shielding properties due to their high density, atomic number, and stability. Key shielding parameters such as mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), half-value layer (HVL), and effective atomic number (Zeff) were calculated and compared to lead. Among all the studied Ta–W–O compounds, TaW2O3 was identified as the most efficient and thermodynamically stable lead-free shielding material, exhibiting the highest photon attenuation performance across low- and intermediate-energy ranges. Across various energy ranges, these compounds demonstrate superior radiation protection efficiency (RPE) and electron density, essential for shielding in healthcare, nuclear, and aerospace applications. The findings suggest that tantalum–tungsten compounds could serve as viable lead-free shielding materials, offering a safer and more sustainable alternative for radiation protection.
本研究探讨了铅辐射屏蔽的替代材料,解决了更安全、更有效的选择需求。像铅这样的传统材料,虽然由于其高原子序数而有效,但却是有毒的,并构成环境风险。该研究探索了一组钽钨氧(Ta-W-O)化合物,包括TaW2O3, TaW2O4, Ta2W2O5, TaWO3等。这些化合物由于其高密度、原子序数和稳定性,提供了有前途的屏蔽性能。计算了质量衰减系数(MAC)、线性衰减系数(LAC)、半值层(HVL)和有效原子序数(Zeff)等关键屏蔽参数,并与铅进行了比较。在所有被研究的Ta-W-O化合物中,TaW2O3被认为是最有效和热力学稳定的无铅屏蔽材料,在中低能量范围内表现出最高的光子衰减性能。在各种能量范围内,这些化合物显示出卓越的辐射防护效率(RPE)和电子密度,这对于医疗保健、核和航空航天应用中的屏蔽至关重要。研究结果表明,钽钨化合物可以作为可行的无铅屏蔽材料,为辐射防护提供更安全、更可持续的替代方案。
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引用次数: 0
Numerical investigation of spacer grid-induced flow disturbances and impact on fuel rod flow-induced vibrations 间隔栅诱导的流动扰动及其对燃料棒流致振动影响的数值研究
IF 2.3 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-20 DOI: 10.1016/j.anucene.2026.112153
Yilong Li , Yan Guo , Junyong Liu, Hongyu Yang, Guangyun Min, Naibin Jiang
This study aims to investigate the flow disturbance effect of different spacer grids in a Pressurized Water Reactor (PWR) and their influence on the flow-induced vibration characteristics of fuel rods. Based on the model from the Subchannel and Bundle Test (PSBT). Differences between two-way flow-structure interaction and one-way flow-structure interaction analyzed. Subsequently, the one-way flow-structure interaction method was adopted. Under the condition of consistent average inlet mean flow velocity, compared the results of each spacer grid subchannel: transverse flow, pressure distribution, and pressure drop. These effects show significant correlation with grid structural features such as rigid protrusion, spring, and mixing vanes. Thus amplitude are significant in subchannels: the influence of the simple supported grid is negligible, spacer grids with no mixing vanes has a significant influence, and spacer grids with mixing vanes have the most significant effects.
研究了压水堆中不同间隔栅的流动扰动效应及其对燃料棒流激振动特性的影响。基于子通道和捆绑测试(PSBT)的模型。分析了双向流-结构相互作用与单向流-结构相互作用的区别。随后,采用单向流-结构相互作用方法。在平均进口平均流速一致的情况下,比较各间隔栅子通道的横向流动、压力分布和压降结果。这些影响与网格结构特征(如刚性突出、弹簧和混合叶片)有显著的相关性。因此,在子通道中振幅是显著的:简支网格的影响可以忽略不计,没有混合叶片的间隔网格的影响显著,有混合叶片的间隔网格的影响最显著。
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引用次数: 0
A non-negative Lasso Orthogonal Matching Pursuit method for gamma radiation field reconstruction with sparse measurement data 稀疏测量数据下γ辐射场重建的非负Lasso正交匹配追踪方法
IF 2.3 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-20 DOI: 10.1016/j.anucene.2026.112147
Longji Qiu, Nan Chao, Yong-kuo Liu, Zongzhen Shi, Joseph Daniel
This paper proposes a gamma radiation field reconstruction method based on sparse detection data. By integrating compressed sensing (CS) theory with clustered transport theory, a sensing matrix for radiation field reconstruction is constructed. A novel Non-negative Lasso Orthogonal Matching Pursuit (NNLasso-OMP) algorithm is developed, combining the high efficiency and flexibility of OMP with the overfitting resistance of Lasso. To evaluate the proposed method, three simulation scenarios are conducted, using Monte Carlo simulation results as reference benchmarks. The reconstruction performance of NNLasso-OMP is compared with that of OMP, inverse distance weighting (IDW) and 3DCNN algorithms. Results show that the average relative error (ARE) of NNLasso-OMP remains below 10% across all scenarios, achieving a reconstruction success rate (SR) exceeding 95%, while accurately identifying source locations. The proposed NNLasso-OMP method outperforms both OMP and IDW, demonstrating its effectiveness in achieving high-quality gamma radiation field reconstruction from sparse measurements.
提出了一种基于稀疏检测数据的伽马辐射场重建方法。将压缩感知理论与聚类输运理论相结合,构造了辐射场重构的感知矩阵。将非负Lasso正交匹配追踪算法(NNLasso-OMP)的高效率和灵活性与Lasso的抗过拟合性相结合,提出了一种新的非负Lasso正交匹配追踪算法。为了评估所提出的方法,以蒙特卡罗模拟结果作为参考基准,进行了三种模拟场景。将NNLasso-OMP算法与OMP、逆距离加权(IDW)和3DCNN算法的重建性能进行了比较。结果表明,在所有场景下,NNLasso-OMP的平均相对误差(ARE)保持在10%以下,重建成功率(SR)超过95%,同时准确识别出源位置。提出的NNLasso-OMP方法优于OMP和IDW,证明了其在从稀疏测量中获得高质量伽马辐射场重建方面的有效性。
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引用次数: 0
Identification of the containment heating mechanism and temperature distribution by high-temperature gas leakage under severe accident conditions 严重事故条件下高温气体泄漏安全壳加热机理及温度分布的识别
IF 2.3 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Pub Date : 2026-01-20 DOI: 10.1016/j.anucene.2025.112009
Restu Kojo
This study evaluates the thermal-hydraulics in the containment of boiling water reactors caused by superheated gas leakage under severe accident conditions. The research involved characterizing the heat transfer paths, which included identifying superheated gas leakage positions and heat release from the reactor pressure vessel boundaries, classifying the containment volumes, and categorizing potential failure sections of the containment. Based on these insights, the heat transfer paths during severe accidents were clarified, and accident scenarios considering leakage from the safety relief valve and traversing in-core probe tubes were selected as representative scenarios.
A three-dimensional computational fluid dynamics (CFD) model of the Mark I containment was developed to evaluate the thermal-hydraulics of an entire drywell of the containment. Special attention was given to modeling the detailed structures of the containment top head flange, radiation from the reactor pressure vessel upper head, condensation in the suppression pool, and heat release from the containment top head to the reactor well. The CFD analyses focused on two scenarios: safety relief valve leakage and traversing in-core probe tube leakage, which can result in significant temperature distribution in the upper and lower drywell, respectively. This study identified the high temperature location on the containment boundary with higher possibility of failure by high-temperature gas leakage under severe accident conditions based on temperature distribution obtained by the present detailed three-dimensional CFD analysis.
本文研究了严重事故条件下沸水反应堆过热气体泄漏引起的安全壳热工水力学特性。该研究涉及表征传热路径,包括识别过热气体泄漏位置和反应堆压力容器边界的热量释放,对安全壳体积进行分类,并对安全壳的潜在故障部分进行分类。在此基础上,明确了严重事故中的传热路径,并选择了考虑安全阀泄漏和穿过堆芯内探针管的事故场景作为代表场景。建立了Mark 1型安全壳的三维计算流体动力学(CFD)模型,对整个安全壳干井的热工水力学进行了评估。特别注意对安全壳顶部法兰的详细结构、反应堆压力容器顶部的辐射、抑制池中的冷凝以及从安全壳顶部向反应堆井释放的热量进行建模。CFD分析主要集中在两种情况下:安全阀泄漏和岩心内穿越探针管泄漏,这两种情况分别会导致干井上部和下部的温度分布明显。本研究基于详细的三维CFD分析得到的温度分布,确定了在严重事故条件下高温气体泄漏破坏可能性较大的安全壳边界高温位置。
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引用次数: 0
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