Nuclear Engineering and Design最新文献_第7页

Characterization of braided-wire wicks for bent heat pipe applications: Experiments and modeling 弯曲热管用编织丝芯的特性：实验和建模

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-01-14 DOI: 10.1016/j.nucengdes.2026.114762

Yohan Kim, Hyungdae Kim

The bent heat pipe, which connects a compact reactor core to a relatively large power conversion system, is a critical component in the design of kilowatt-scale heat pipe-cooled reactors for space nuclear applications. Conventional screen wicks (SWs) used in straight heat pipes are prone to structural damage when bent, resulting in a loss of capillary performance. To address this issue, bendable braided-wire wicks (BWWs) have recently been proposed as a promising alternative due to their ability to maintain capillary functionality even under bending conditions. However, comprehensive studies on the capillary performance of BWWs in bent configurations remain limited. This study experimentally investigates the capillary flow characteristics of BWWs in both straight and bent geometries. The porosity was calculated through a geometric analysis of a unit cell in the wick and found to be 0.215. Capillary rise in a vertically oriented straight wick was visualized and quantitatively assessed using infrared thermography. The effective pore radius and permeability were determined by fitting the experimental data of the sample, yielding values were 67.3 μm and

1.8 \times 10^{- 11}

m², respectively. Subsequently, a series of rate-of-rise experiments was conducted on bent BWWs to evaluate the impact of bending on their capillary behavior. The experimental results demonstrated that the BWW maintains its intrinsic capillary properties irrespective of geometric configuration. Finally, theoretical models were proposed to predict the porosity, permeability, and effective pore radius of the BWW structure. The predicted values agreed with the experimental measurements within a margin of approximately 20%.

弯曲热管将紧凑的反应堆堆芯与相对较大的功率转换系统连接起来，是设计用于空间核应用的千瓦级热管冷却反应堆的关键部件。用于直热管的传统筛芯在弯曲时容易发生结构损坏，导致毛细管性能下降。为了解决这个问题，可弯曲编织丝芯（BWWs）最近被提出作为一种有前途的替代方案，因为它们即使在弯曲条件下也能保持毛细管功能。然而，对弯曲构型下的水射流的毛细管性能的全面研究仍然有限。本文通过实验研究了直、弯两种几何形状下涡轮增压发动机的毛细流动特性。孔隙率是通过对灯芯内的单胞进行几何分析计算得出的，结果为0.215。利用红外热像仪对垂直定向直芯中的毛细上升进行了可视化和定量评价。通过拟合试样的实验数据，确定了有效孔隙半径和渗透率，屈服值分别为67.3 μm和1.8×10−11 m2。随后，对弯曲的BWWs进行了一系列速率上升实验，以评估弯曲对其毛细行为的影响。实验结果表明，无论几何形状如何，BWW都保持其固有的毛细特性。最后，提出了预测BWW结构孔隙度、渗透率和有效孔隙半径的理论模型。预测值与实验测量值的误差约为20%。

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引用次数: 0

Leveraging ENDF data in an enhanced ORIGEN2 library for advanced VVER-1000 fuel management 利用增强的ORIGEN2库中的ENDF数据进行先进的VVER-1000燃料管理

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-01-29 DOI: 10.1016/j.nucengdes.2026.114758

Saeedeh Arabzadeh , Seyed Pezhman Shirmardi , Nasser Mansour Shariflou

Accurate burn up calculations are critical for nuclear reactor design, particularly for determining the nuclear concentrations of fuel isotopes and fission products throughout the reactor cycle. An updated cross-sectional library is essential for effective fuel behavior analysis and management. This study aims to develop a tailored cross-sectional library for the VVER-1000 reactor to enhance the accuracy of burn up calculations using the ORIGEN2 code, leveraging the ENDF reference library. The Monte Carlo N-Particle (MCNPX) code was used to generate the required cross-sectionals, which were then integrated into ORIGEN2 for burn up calculations. The results were compared with those obtained using the existing library. The new library demonstrates moderately improved accuracy and computational efficiency for burn up calculations in the VVER-1000 reactor compared to the previous library.

准确的燃烧计算对于核反应堆设计至关重要，特别是对于确定整个反应堆循环中燃料同位素和裂变产物的核浓度。更新的截面库对于有效的燃料行为分析和管理至关重要。本研究旨在利用ENDF参考库，为VVER-1000反应堆开发一个定制的横截面库，以提高使用ORIGEN2代码进行燃烧计算的准确性。蒙特卡罗n粒子（MCNPX）代码用于生成所需的横截面，然后将其集成到ORIGEN2中进行燃烧计算。结果与现有文库的结果进行了比较。与以前的库相比，新库在VVER-1000反应堆的燃烧计算中显示出适度提高的精度和计算效率。

引用次数: 0

The application of the fixed-point iteration acceleration in the neutronics and thermal-hydraulics coupling calculation 不动点迭代加速在中子热工耦合计算中的应用

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2026-01-08 DOI: 10.1016/j.nucengdes.2026.114754

An Ping , Shen Jie , Liu Rui , Liu Dong , He Xiao Qiang , Liu Ting , Hu Ming

Many fixed-point problems are involved in the numerical calculation of reactor high-fidelity simulation. The Gauss-Seidel and the SOR (Successive Over-Relaxation) algorithm are commonly used for iterative acceleration. In this paper, based on Anderson's idea, an acceleration algorithm, known as RCA (Reactor Coupling Calculation Acceleration), is proposed. This algorithm is suitable for multi-disciplinary coupling calculation fixed point iteration for reactors. The RCA algorithm takes the subspace as the iteration object, and the numerical format is determined by minimizing the weighted residual 2-norm, which has good convergence. This paper adopts the RCA algorithm to accelerate the coupled calculation of neutronics and thermal-hydraulics, and calculates the boron critical searching under normal or accident operating conditions. The results show that the RCA algorithm can reduce the number of iterations and improve the calculation efficiency with the same calculation accuracy. Our research provides support for the optimization of existing software and the development of new software.

在反应堆高保真仿真数值计算中涉及到许多不动点问题。高斯-赛德尔算法和SOR（连续过松弛）算法是迭代加速的常用算法。本文基于Anderson的思想，提出了一种加速算法，称为RCA （Reactor Coupling Calculation acceleration）。该算法适用于反应器的多学科耦合计算不动点迭代。RCA算法以子空间为迭代对象，通过最小化加权残差2范数来确定数值格式，具有较好的收敛性。本文采用RCA算法加速中子与热工耦合计算，计算了正常工况和事故工况下的硼临界搜索。结果表明，在保证计算精度的前提下，RCA算法可以减少迭代次数，提高计算效率。我们的研究为现有软件的优化和新软件的开发提供了支持。

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引用次数: 0

Long-term cooling characteristics during transients with bypassing the safety protection system of a molten salt fast reactor 熔盐快堆旁路安全保护系统瞬态长期冷却特性

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-22 DOI: 10.1016/j.nucengdes.2025.114706

Hiroyasu Mochizuki

The objective of the present study is to confirm how long the reactor cooling of a molten salt fast reactor system can be maintained even after an abnormal transient with a bypass of the safety protection system under conditions that the decay heat removal system is activated or not activated. The system implements two heat storage tanks and a decay heat removal system using natural convection air cooling. The connection locations of the decay heat removal system are set to two heat storage tanks. The heat capacity of the molten salt stored in these tanks (capacity: 12,566 m³) is primarily intended to enable advanced load-following operation, but it can also be effectively utilized for long-term cooling. System analyses are performed under the conditions where the abnormal transients with the bypass of the safety protection system, and turbine of the energy conversion system is manually tripped at 10 min or 10 h and the air cooler is manually activated simultaneously. It has been confirmed that the heat capacity of the molten salt in the heat storage tanks is properly utilized by natural circulation, and the fuel temperature after turbine tripping becomes less than 900 K. It has been also confirmed through calculations that even if the decay heat removal system cannot be manually activated, if the turbine is tripped within 10 h, the molten fuel temperature can be kept below 1250 K for up to 1000 h (41.7 days) or longer with the heat capacity of molten salt in the heat storage tanks. This term increases as a function of the capacity of the heat storage tank, and it has been deemed that even without the heat storage tank, there is a time margin of more than a couple of hours before the temperature rises to 1250 K. The results of these analyses clearly show that the levels of defense in depth immediately prior to a severe accident are enhanced, and the probability of transition to SA due to internal factors such as unprotected abnormal transients is almost negligible.

本研究的目的是确定在衰变排热系统被激活或未被激活的情况下，即使在安全保护系统旁路的异常瞬态之后，熔盐快堆系统的反应堆冷却可以维持多长时间。该系统采用两个储热罐和一个采用自然对流空气冷却的衰减散热系统。衰变排热系统的连接位置设置为两个储热罐。储存在这些储罐中的熔盐的热容量（容量：12,566 m3）主要用于实现先进的负载跟随操作，但它也可以有效地用于长期冷却。在安全保护系统旁路的异常暂态下，分别在10min或10h手动跳闸换能系统水轮机，同时手动启动空冷器的情况下进行系统分析。结果表明，储热槽内熔盐的热容通过自然循环得到充分利用，涡轮脱扣后燃料温度低于900 K。通过计算也证实，即使不能手动启动衰变排热系统，如果在10 h内跳闸涡轮，利用储热罐中熔盐的热容量，熔燃料温度可以保持在1250 K以下长达1000 h（41.7天）或更长时间。这一项随着储热罐容量的增大而增大，我们认为即使没有储热罐，在温度上升到1250k之前也有几个小时以上的时间余量。这些分析的结果清楚地表明，在严重事故发生前的深度防御水平得到了提高，并且由于内部因素（如未受保护的异常瞬变）过渡到SA的可能性几乎可以忽略不计。

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引用次数: 0

Source term distribution and mobility – WP3 results of EU project SAMOSAFER 源项分布和流动性——欧盟SAMOSAFER项目WP3结果

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-22 DOI: 10.1016/j.nucengdes.2025.114679

J. Křepel , S. Lorenzi , L. Giot , E.M.A. Frederix , A. Cammi , F. Caruggi , S. Deanesi , F. Scioscioli , S. Delpech , P. Souček , J. Uhlíř , M. Mareček , M. Cihlář , J. Serp , H. Pitois , E. Merle , D. Rodrigues , C. Cannes , J. Kalilainen , S. Nichenko , T. Lind

The aim of the EU Horizon 2020 founded project SAMOSAFER was to develop and demonstrate new safety barriers and a more controlled behavior in severe accidents of the Molten Salt Reactor (MSR). The aim of work package 3 was to develop and validate models for tracing the nuclides carrying the radiotoxicity and decay heat, i.e. the source term and their chemical form and mobility during nominal and accidental conditions. The source term distribution at the beginning of an accident depends on the foregoing nominal operation mode. Hence, the liquid salt reprocessing and gaseous and insoluble Fission Products (FPs) separation was modeled before the severe accident simulation were accomplished. Because the source term assessment is a complex topic, only selected phenomena were addressed. As a reference system the MSFR concept was adopted from the previous EU project SAMOFAR. The study included benchmarking of the burnup tools, thermal-hydraulics simulations to confirm the removal rates to the off-gas system. Chemical experiments and calculations of the methods used in the reprocessing unit, simulation of severe accident conditions and finally calculation of the source term distribution. The major outcomes are the benchmark for the nuclide inventory tracing tools, application of multi-physics tools on He-bubbling for gaseous and solid FPs removal rates simulations, refinement of the reprocessing scheme, and first simulation of simplified severe accident.

欧盟地平线2020成立的SAMOSAFER项目的目的是开发和展示新的安全屏障，并在熔盐堆（MSR）的严重事故中更好地控制行为。工作包3的目的是开发和验证模型，以追踪携带放射性毒性和衰变热的核素，即源项及其在名义和意外条件下的化学形式和迁移率。事故开始时的源项分布取决于上述标称运行模式。因此，在完成严重事故模拟之前，对液态盐后处理和气态和不溶性裂变产物（FPs）分离进行了建模。由于源项评估是一个复杂的主题，因此只讨论了选定的现象。作为一个参考系统，MSFR概念是从以前的欧盟SAMOFAR项目中采用的。该研究包括燃耗工具的基准测试，热工模拟，以确认废气系统的去除率。化学实验和计算方法应用于后处理单元，模拟严重事故条件，最后计算源项分布。主要成果包括核素库存追踪工具的基准，he -泡泡多物理场工具在气体和固体FPs去除速率模拟中的应用，后处理方案的改进以及简化严重事故的首次模拟。

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引用次数: 0

Margin to onset of nucleate boiling and flow instability studies for preliminary MITR design-demonstration element thermal-hydraulics 核沸腾起始余量及MITR初步设计的流动不稳定性研究-论证元件热工水力学

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2026-01-10 DOI: 10.1016/j.nucengdes.2026.114766

Palash K. Bhowmik, Mauricio E. Tano, SuJong Yoon, Changhu Xing, Silvino A.B. Prieto, Alexander L. Swearingen, Ann M. Phillips, Piyush Sabharwall, Jeffrey J. Giglio

This study covers the onset of flow instability (OFI) preliminary results obtained from leveraging correlations, in addition to the preliminary thermal hydraulics results such as pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment (DDE) for the Massachusetts Institute of Technology Reactor (MITR). Current computational fluid dynamics (CFD) models in fluid structure interaction (FSI) have added the capability of assessing margins to onset of nucleate boiling (ONB). This study initiates the capability to model the margin to OFI and ONB presented for the MITR. Such study is supportive of the United States High Performance Research Reactor (USHPRR) program. Previous studies provided preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the MITR DDE under conservative approximations for plate power distribution. This study focuses on providing insights into the OFI future research direction optimizing the transport of thermal energy, mass-flow rates, flow-channel geometries, and boundary conditions.

除了麻省理工学院反应堆（MITR）设计演示实验（DDE）上的压力、流速、温度和氧化层等初步热工水力学结果外，本研究还涵盖了利用相关性获得的流动不稳定性（OFI）初始结果。当前流体结构相互作用（FSI）中的计算流体动力学（CFD）模型增加了评估核沸腾（ONB）开始边缘的能力。这项研究启动了为MITR提供的OFI和ONB边际建模的能力。这样的研究支持了美国高性能研究堆（USHPRR）计划。先前的研究在板功率分布的保守近似下，对MITR DDE的水动力效应进行了初步的热水力和力学分析。本研究的重点是为OFI未来的研究方向提供见解，优化热能输运，质量流率，流道几何形状和边界条件。

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引用次数: 0

Momentum equations for scaling analysis of natural circulation loops: Principles and application 自然循环环标度分析的动量方程：原理与应用

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-29 DOI: 10.1016/j.nucengdes.2025.114712

Montanini Marco , Carnevali Sofia , Bestion Dominique , Cottarel Valentin , Rossi Lionel

Developments are in progress at Commissariat á l’énergie atomique et aux Énergies Alternatives (CEA) to use the CATHARE code for scaling analysis. It has been shown that mature system codes can perform more detailed scaling analyses than of analytical methods alone. When an integral effect test facility (IET) is available to simulate a reactor transient and when the code correctly predicts it, a code assisted a-posteriori scaling analysis is able to analyze the origins of distortions and quantify the impact of distortions between the IET facility and the reactor transient simulation. The analysis is now extended to natural convection (NC). Two equations are analyzed: integrated mixed momentum equation (MME), describing mass flow rate, and integrated crossed momentum equation (CME), controlling slip ratio and void fraction in the circuit. A very simple exercise is performed on a loop at reactor scale. The same analysis is run for hypothetical reduced-scale IETs designed using power-to-volume scaling with full-height (PTVS-FH) and three-level-scaling with reduced-height (3LS-RH) approaches. The paper introduces the method developed in CATHARE code based on discretized momentum equations. Nodalizations and boundary conditions which simulate the natural circulation (NC) for a decreasing mass inventory in the circuit are described. Dominant terms of mixture and crossed momentum equation are identified. Distortions of the scaled loops with respect to the reactor-type loop are analyzed and preliminary conclusions are presented.

在化学和化学替代材料委员会（CEA），使用CATHARE代码进行比例分析的工作正在取得进展。成熟的系统代码比单独的分析方法可以进行更详细的尺度分析。当一个积分效应测试设施（IET）可用来模拟反应堆瞬态，并且当代码正确预测它时，代码辅助的后验标度分析能够分析扭曲的来源，并量化IET设施和反应堆瞬态模拟之间的扭曲影响。该分析现已扩展到自然对流（NC）。分析了描述质量流量的积分混合动量方程（MME）和控制电路滑移率和空隙率的积分交叉动量方程（CME）。一个非常简单的练习是在反应堆规模的环路上进行的。同样的分析也适用于采用全高度功率体积缩放（PTVS-FH）和低高度三级缩放（3LS-RH）方法设计的小型IETs。本文介绍了在CATHARE代码中基于离散动量方程开发的方法。描述了电路中模拟自然循环（NC）的节点化和边界条件。确定了混合动量方程和交叉动量方程的主导项。分析了标度环相对于电抗器型环的畸变，并给出了初步结论。

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引用次数: 0

Implementation, verification and validation of spatial kinetics calculations in CONDOR-CITVAP codes CONDOR-CITVAP代码中空间动力学计算的实现，验证和验证

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-22 DOI: 10.1016/j.nucengdes.2025.114713

Ignacio Ferrari , Diego Ferraro , Daniel Hergenreder

INVAP’s methodology for the neutronic design of research and radio-isotope production reactors is based on a deterministic cell-core scheme, which has been successfully applied in numerous projects. The cell stage is handled by CONDOR code, which solves the multi-group integral neutron transport equations by means of the Heterogeneous Response Method, and allows the generation of few-group macroscopic constants for core calculations. The core stage is handled by CITVAP code, which solves the multi-group diffusion equation. In this work, the implementation developed for the solution of spatial kinetics problems is presented, which extends the analysis capabilities beyond the standard criticality and fixed-source calculations. The selected methodology is based on a direct approach where the time-dependent neutron diffusion equation is solved. Several verification cases for rectangular geometry are evaluated and compared against reference results. Besides, validation cases using experimental data from the commissioning of the OPAL and RA-6 reactors are also analyzed.

INVAP的研究中子设计和放射性同位素生产反应堆的方法是基于确定性的细胞核心方案，该方案已成功地应用于许多项目。单元阶段由CONDOR代码处理，该代码采用非均质响应法求解多群积分中子输运方程，并允许生成用于堆芯计算的少数群宏观常数。核心阶段由CITVAP代码处理，求解多群扩散方程。在这项工作中，提出了为解决空间动力学问题而开发的实现，它将分析能力扩展到标准临界和固定源计算之外。所选择的方法是基于求解随时间变化的中子扩散方程的直接方法。对矩形几何的几个验证案例进行了评价，并与参考结果进行了比较。此外，还分析了OPAL和RA-6反应堆调试实验数据的验证案例。

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引用次数: 0

Evaluation of long-term exposure of 310S and 800H under conditions of SCW-SMR SCW-SMR条件下310S和800H长期暴露的评价

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-24 DOI: 10.1016/j.nucengdes.2025.114691

Daniela Marušáková , Jan Vít , Monika Šípová , Marek Vronka

Austenitic stainless steels 310S and Alloy 800H are considered promising candidates for fuel cladding in Small Modular Reactors cooled by Supercritical Water due to their high corrosion resistance and favourable mechanical properties. To evaluate their long-term behaviour in supercritical water environments, full-length tube samples were exposed to supercritical water at 500 °C, 25 MPa, and 150 ppb dissolved oxygen for up to 10,000 h. Weight gain measurements revealed a decelerating oxidation rate over time, with low cumulative mass increases indicative of excellent corrosion resistance. Conservative extrapolation suggests that the wall penetration depth is projected to remain below 5 μm after 30,000 h for both alloys, supporting the expectation of long-term structural integrity. Surface roughness measurements corroborated these trends: 310S showed a gradual increase from 0.18 μm (as received) to 0.29 μm after 10,000 h, whereas 800H exhibited minimal change, attributed to its initially higher surface roughness. Post-exposure characterization by Scanning Electron Microscope with Energy-Dispersive X-ray Spectroscopy and Transmission Electron Microscope confirmed the formation of compact, chromium-rich oxide layers (Cr₂O₃) on both materials, with underlying Cr-depleted zones. Microstructural analysis revealed that 310S developed thicker oxide layer with larger grains and Cr-Ni-rich phases, whereas 800H exhibited finer oxide grains in a thinner layer, with occasional localized corrosion – nodules. These differences underscore the role of alloy composition in oxidation behaviour under supercritical water conditions. Overall, both 310S and 800H demonstrate excellent oxidation resistance and microstructural stability, reinforcing their applicability as fuel cladding materials in Small Modular Reactors cooled by Supercritical Water designs.

奥氏体不锈钢310S和合金800H由于其高耐腐蚀性和良好的机械性能被认为是超临界水冷却的小型模块化反应堆燃料包壳的有希望的候选者。为了评估其在超临界水环境中的长期行为，将全长管样品暴露在500°C、25 MPa和150 ppb溶解氧的超临界水中长达10,000小时。重量增加测量显示，随着时间的推移，氧化速率减慢，累积质量增加低表明具有优异的耐腐蚀性。保守推断表明，在30,000 h后，两种合金的壁穿深预计将保持在5 μm以下，从而支持长期结构完整性的期望。表面粗糙度测量证实了这些趋势：310S在10,000 h后从0.18 μm（接收到的）逐渐增加到0.29 μm，而800H的变化很小，这归因于其最初的表面粗糙度较高。利用扫描电子显微镜、能量色散x射线能谱和透射电子显微镜对暴露后的特征进行了表征，证实了两种材料上都形成了致密的富铬氧化物层（Cr₂O₃），下面有Cr-贫区。显微组织分析表明，310S合金的氧化层较厚，晶粒较大，具有富cr - ni相，而800H合金的氧化层较薄，氧化层较细，偶有局部腐蚀结核。这些差异强调了合金成分在超临界水条件下氧化行为中的作用。总体而言，310S和800H都表现出优异的抗氧化性和微观结构稳定性，增强了它们作为超临界水冷却小型模块化反应堆燃料包壳材料的适用性。

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引用次数: 0

Beyond-design-basis screening by a three-bound critical excitation envelope for base-isolated nuclear power plants 基础隔离核电站三界临界激励包络的超设计基础筛选

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2026-01-10 DOI: 10.1016/j.nucengdes.2025.114700

Ali Ahmadi , Naser Khaji , Hamid Sadegh-Azar

In this paper, a resonance-focused Critical Excitation (CE) overlay is developed as a secondary check for base-isolated nuclear power plants. The goal is to capture worst-case, yet physically plausible, motions that align with the isolation period and key equipment periods, while remaining compatible with ASCE 4–16 and ASCE 43–19 practices. The method generates three CE bounds (lower, mean, upper) under explicit Arias intensity, PGA, and PGV constraints. At the same time, routine suite-mean results are read against these bounds to flag under-targeting, proximity, or design-extension concerns. Well-established isolated-reactor models, including two-degree-of-freedom with low-damping rubber, lead rubber bearing, friction pendulum system, and six-degree-of-freedom configurations, are used. As a case study, ground motions are selected for the Diablo Canyon site and matched to the SDC-5 Design Response Spectrum. A code-consistent suite of ground motion records forms the design baseline for averaging. A screening study compares acceleration- and displacement-targeted CE objectives. The displacement objective produces higher peaks in isolator displacement, isolation-plane base shear, and floor acceleration for six of seven seed ground motions; therefore, it is adopted for design-level evaluation. With the displacement objective, suite-mean responses are placed within the CE three-bound for key metrics, indicating conservative and stable estimates without missing resonance. The overlay provides clear decision triggers: (a) below the CE lower bound, supplementation or retuning is indicated; (b) near the CE mean, demand capture is adequate and remaining margins are checked; and (c) trending toward the CE upper bound signals a Beyond-Design-Basis Earthquake (BDBE) condition and prompts targeted checks. The CE overlay thus serves as a transparent, code-compatible safety gate and supports BDBE reasoning without any arbitrary multipliers.

本文提出了一种共振聚焦临界激励（CE）覆盖层，作为基础隔离核电站的二次校核。目标是捕捉与隔离期和关键设备期一致的最坏情况，但物理上合理的运动，同时保持与ASCE 4-16和ASCE 43-19实践的兼容性。该方法在明确的Arias强度、PGA和PGV约束下生成三个CE边界（lower, mean, upper）。同时，根据这些界限读取常规的套件平均值结果，以标记目标不足、接近或设计扩展问题。采用了成熟的隔离反应器模型，包括两自由度低阻尼橡胶、铅橡胶轴承、摩擦摆系统和六自由度配置。作为一个案例研究，选择了Diablo峡谷场地的地面运动，并与SDC-5设计响应谱相匹配。一套与代码一致的地面运动记录形成了平均的设计基线。一项筛选研究比较了以加速度和位移为目标的CE目标。位移目标在隔震器位移、隔震面基底剪切和底板加速度中产生较高的峰值；因此，采用该方法进行设计级评价。对于位移目标，套件平均响应被放置在关键指标的CE三界内，表明保守和稳定的估计而不会丢失共振。叠加层提供了明确的决策触发器：(a)低于CE下限，表示补充或返回；(b)在接近行政长官平均数的情况下，需求已足够，而余下的差额则会受到检查；(c)趋向于CE上限表明超出设计基础地震（BDBE）条件并提示有针对性的检查。因此，CE覆盖层作为一个透明的、代码兼容的安全门，支持BDBE推理，而不需要任何任意乘数。

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引用次数: 0