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

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

{"title":"Characterization of braided-wire wicks for bent heat pipe applications: Experiments and modeling","authors":"Yohan Kim, Hyungdae Kim","doi":"10.1016/j.nucengdes.2026.114762","DOIUrl":"10.1016/j.nucengdes.2026.114762","url":null,"abstract":"<div><div>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 <span><math><mn>1.8</mn><mo>×</mo><msup><mn>10</mn><mrow><mo>−</mo><mn>11</mn></mrow></msup></math></span> m<sup>2</sup>, 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%.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114762"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981789","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

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

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

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

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

{"title":"The application of the fixed-point iteration acceleration in the neutronics and thermal-hydraulics coupling calculation","authors":"An Ping , Shen Jie , Liu Rui , Liu Dong , He Xiao Qiang , Liu Ting , Hu Ming","doi":"10.1016/j.nucengdes.2026.114754","DOIUrl":"10.1016/j.nucengdes.2026.114754","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114754"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939637","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

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

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

Hiroyasu Mochizuki

{"title":"Long-term cooling characteristics during transients with bypassing the safety protection system of a molten salt fast reactor","authors":"Hiroyasu Mochizuki","doi":"10.1016/j.nucengdes.2025.114706","DOIUrl":"10.1016/j.nucengdes.2025.114706","url":null,"abstract":"<div><div>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 m<sup>3</sup>) 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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114706"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841524","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

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

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

{"title":"Source term distribution and mobility – WP3 results of EU project SAMOSAFER","authors":"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","doi":"10.1016/j.nucengdes.2025.114679","DOIUrl":"10.1016/j.nucengdes.2025.114679","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114679"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841585","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

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

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

{"title":"Margin to onset of nucleate boiling and flow instability studies for preliminary MITR design-demonstration element thermal-hydraulics","authors":"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","doi":"10.1016/j.nucengdes.2026.114766","DOIUrl":"10.1016/j.nucengdes.2026.114766","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114766"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145978183","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

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

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

{"title":"Momentum equations for scaling analysis of natural circulation loops: Principles and application","authors":"Montanini Marco , Carnevali Sofia , Bestion Dominique , Cottarel Valentin , Rossi Lionel","doi":"10.1016/j.nucengdes.2025.114712","DOIUrl":"10.1016/j.nucengdes.2025.114712","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114712"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884681","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

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

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

{"title":"Implementation, verification and validation of spatial kinetics calculations in CONDOR-CITVAP codes","authors":"Ignacio Ferrari , Diego Ferraro , Daniel Hergenreder","doi":"10.1016/j.nucengdes.2025.114713","DOIUrl":"10.1016/j.nucengdes.2025.114713","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114713"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841526","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

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

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

{"title":"Evaluation of long-term exposure of 310S and 800H under conditions of SCW-SMR","authors":"Daniela Marušáková , Jan Vít , Monika Šípová , Marek Vronka","doi":"10.1016/j.nucengdes.2025.114691","DOIUrl":"10.1016/j.nucengdes.2025.114691","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114691"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841582","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

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

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

{"title":"Beyond-design-basis screening by a three-bound critical excitation envelope for base-isolated nuclear power plants","authors":"Ali Ahmadi , Naser Khaji , Hamid Sadegh-Azar","doi":"10.1016/j.nucengdes.2025.114700","DOIUrl":"10.1016/j.nucengdes.2025.114700","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114700"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939545","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