Nuclear Engineering and Design最新文献

Validation of the ISAA-Na two-phase model for sodium-cooled fast reactors: An assessment against CABRI LOF experiments 钠冷快堆ISAA-Na两相模型的验证：对CABRI LOF实验的评估

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-01-13 DOI: 10.1016/j.nucengdes.2026.114756

Rui Hou , Shaowei Tang , Jingliang Zhang , Yi Lei , Bin Zhang , Jianqiang Shan

{"title":"Validation of the ISAA-Na two-phase model for sodium-cooled fast reactors: An assessment against CABRI LOF experiments","authors":"Rui Hou , Shaowei Tang , Jingliang Zhang , Yi Lei , Bin Zhang , Jianqiang Shan","doi":"10.1016/j.nucengdes.2026.114756","DOIUrl":"10.1016/j.nucengdes.2026.114756","url":null,"abstract":"<div><div>Under Loss-of-Flow (LOF) accident conditions, the boiling behavior of the coolant has a decisive impact on core integrity. A multi-bubble slug model has been implemented and integrated within the integrated severe accident analysis code ISAA-Na to dynamically simulate the two-phase flow and heat transfer behavior in sodium-cooled fast reactors (SFRs) under LOF conditions. To validate the model's effectiveness, experiments BI1, E8, and EFM1 from the French CABRI facility were selected as benchmarks. The predictive capability of the model was assessed through a systematic comparison of the calculation results from ISAA-Na with experimental data and published results from other mainstream codes. The assessment indicates that the model accurately captures key physical phenomena, at boiling inception, the saturation temperature is predicted with a relative error of 0.12%. For the subsequent two-phase conditions, the model captures boiling initiation times with an absolute error of less than 0.3 s and axial locations within a relative error of 6%, while also accurately reproducing interface propagation and flow oscillations. This confirms the reliability of the model's implementation in ISAA-Na for SFR safety analysis, providing a robust basis for predicting subsequent accident progression.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114756"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981788","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

Numerical implementation of the MOOSE subchannel module (SCM) algorithm MOOSE子信道模块（SCM）算法的数值实现

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-02-11 DOI: 10.1016/j.nucengdes.2026.114802

Vasileios Kyriakopoulos, Mauricio Tano

引用次数: 0

Sensitivity analysis of main neutronic parameters of standard PWR and WWER-1000 pin cells to resonance self-shielding treatment methods and cross section libraries 标准PWR和WWER-1000引脚电池主要中子参数对共振自屏蔽处理方法和截面库的敏感性分析

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-01-31 DOI: 10.1016/j.nucengdes.2026.114801

Farrokh Khoshahval

{"title":"Sensitivity analysis of main neutronic parameters of standard PWR and WWER-1000 pin cells to resonance self-shielding treatment methods and cross section libraries","authors":"Farrokh Khoshahval","doi":"10.1016/j.nucengdes.2026.114801","DOIUrl":"10.1016/j.nucengdes.2026.114801","url":null,"abstract":"<div><div>To evaluate the neutronic behavior of a fuel pellet versus burnup, it is necessary to select a proper lattice code. The deterministic nuclear codes are superior to probabilistic codes in terms of computational speed. The deterministic codes such as WIMS-D5 and DRAGON codes are dependent on the method used for treatment of resonance self-shielding cross sections and the library of cross sections. This paper focus on two standard PWR and WWER pin cells and evaluate both equivalence in dilution, and multiband methods for resonance self-shielding calculations. In addition, three different neutron library cross sections (WLUP-69, WLUP-172, DRAGLIB-172) are assessed. It is revealed that the deterministic lattice DRAGON code can accurately treat the self-shielding behavior in the PWR and WWER pin cell and one can trust on the generated main neutronic parameters and multi-group homogenized cross-sections of PWR and WWER fuel assembly and/or whole-core calculations. In addition, it is proved that generalized Stamm'ler (SHI: self-shielding module of the Dragon) depends on the type of geometry. For square geometries, for all three different libraries, the best values of multiplication factor are obtained using GSM-NOLJ and the worst results are attributed to the GSM-LJ. Furthermore, to have a better evaluation of the applied self-shielding methods and their accuracy, the main isotopic concentration variation and fuel temperature coefficient versus burnup are also computed, considering different libraries and self-shielding treatments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114801"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081261","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 and application of a simplified neutron-kinetics model for severe accident recriticality assessment 严重事故临界性评估简化中子动力学模型的建立与应用

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-01-12 DOI: 10.1016/j.nucengdes.2026.114769

Björn Engström, Weimin Ma

{"title":"Development and application of a simplified neutron-kinetics model for severe accident recriticality assessment","authors":"Björn Engström, Weimin Ma","doi":"10.1016/j.nucengdes.2026.114769","DOIUrl":"10.1016/j.nucengdes.2026.114769","url":null,"abstract":"<div><div>A simplified neutron-kinetics model for recriticality calculations has been developed and implemented within the severe accident code MELCOR. The model partitions the core into active and inactive regions and determines neutron flux using factorization with adiabatic and prompt-jump approximations. Homogenization of pre-calculated core cell multiplication factors, together with Doppler and xenon reactivity defects calculated by perturbation theory, avoids explicit treatment of cross-sections. This approach allows recriticality analysis to be performed seamlessly within MELCOR at standard time-steps and computational cost. Comparisons with SARA project simulations show reasonable agreement. Integrated with MELCOR, the model extends its capabilities to predict reactivity, recriticality timing, fission power, fuel temperatures, and xenon transients. The model was applied to station blackout scenarios in a Nordic BWR with varying low-pressure safety injection timing and two decay-heat curves. Simulations suggest that the time window for recriticality may be substantial if the core periphery remains intact and coolant flow through a breached vessel is limited. In cases where recriticality occurred, fission power evolution was irregular, and even relatively low fission power accelerated containment heat-up and pressurization. These results demonstrate that the simplified model provides an efficient tool for investigating recriticality phenomena, their impact on severe accident progression, and the effectiveness of mitigation strategies under uncertainty and sensitivity analyses.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114769"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145950224","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

Performance and structural optimization of milliwatt-level radioisotope thermoelectric generators with end-mounted thermoelectric modules 端装热电模块毫瓦级放射性同位素热电发生器的性能与结构优化

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-01-27 DOI: 10.1016/j.nucengdes.2026.114791

Hang Jing , Jing Li , Xiaoxi Chen , Qingpei Xiang , Rende Ze , Heng Yan , Liqun Shi , Shuming Peng

{"title":"Performance and structural optimization of milliwatt-level radioisotope thermoelectric generators with end-mounted thermoelectric modules","authors":"Hang Jing , Jing Li , Xiaoxi Chen , Qingpei Xiang , Rende Ze , Heng Yan , Liqun Shi , Shuming Peng","doi":"10.1016/j.nucengdes.2026.114791","DOIUrl":"10.1016/j.nucengdes.2026.114791","url":null,"abstract":"<div><div>Performance and structural optimization of milliwatt-level radioisotope thermoelectric generators (RTGs) with end-mounted thermoelectric modules (TEMs) are investigated. A one-dimensional heat transfer model was developed to analyze temperature distribution and maximum output power (<em>P</em><sub><em>max</em></sub>) of the RTG. The sensitivity of <em>P</em><sub><em>max</em></sub> to TEM length <em>(L)</em> and cross-sectional area (<em>A</em>) was evaluated for RTGs using five thermoelectric materials. Results show that longer <em>L</em> and optimized <em>A</em> enhance the temperature difference (<em>ΔT</em>) and <em>P</em><sub><em>max</em></sub>. For a Bi<sub>2</sub>Te<sub>3</sub>-based RTG with single end TEM (RTG-1), optimal <em>P</em><sub><em>max</em></sub> reached 160.19 mW on Earth at <em>L</em> = 28 mm and <em>A</em> = 292.41 mm<sup>2</sup>, and 273.17 mW on Titan at <em>L</em> = 28 mm and <em>A</em> = 161.29 mm<sup>2</sup>. Dual-end TEM configurations (RTG-2) yielded identical power outputs. COMSOL simulations validated the model with >90% accuracy. Thermal contact resistance (<em>R</em><sub><em>C</em></sub>) analysis revealed higher <em>R</em><sub><em>C</em></sub> necessitates larger <em>L</em>/<em>A</em> ratios for optimal performance. The model provides a versatile tool for designing RTGs with diverse thermoelectric materials.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114791"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078931","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

Integrated digital control room with human-in-the-loop for effective SMR operation 集成数字控制室与人在环有效的SMR操作

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-02-11 DOI: 10.1016/j.nucengdes.2026.114799

Hossam A. Gabbar, Nathanael Selvaraj, Yasir Mahamat

{"title":"Integrated digital control room with human-in-the-loop for effective SMR operation","authors":"Hossam A. Gabbar, Nathanael Selvaraj, Yasir Mahamat","doi":"10.1016/j.nucengdes.2026.114799","DOIUrl":"10.1016/j.nucengdes.2026.114799","url":null,"abstract":"<div><div>This study introduces an integrated Digital Control Room (DCR) for improved operation performance of Small Modular Reactor (SMR). The Integrated DCR framework encompasses human performance monitoring, semantic reasoning, and adaptive controls into a single system with the considerations of human factors within the context of intelligent automation. The proposed design includes Human Performance Management System (HPMS), Human Performance Semantic Network (HPSN), Human Monitoring System (HMS), Human Simulation (HSIM), Control Room Controller (CRC), and External Environment Interface (EEI). The interactions among these modules provide an effective closed loop for effective and optimum SMR operation. The integrated monitoring system enables real-time assessment of operator performance metrics including stress, workload, and situational awareness, and link these with plant and environmental conditions. Operational scenarios are modeled using a co-simulation environment and integrated with SMR simulator for Integral Pressurized Water Reactor (iPWR) from IAEA, to assess operator states and their correlation with environmental changes and required adaptive control room configurations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114799"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190251","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 of heterogeneous nucleation dynamics and transport of bubbles for void fraction assessment in a molten salt reactor 熔盐反应器中非均相成核动力学及气泡输运的研究

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-02-09 DOI: 10.1016/j.nucengdes.2026.114812

N. Guéritat , S. Maillard , P. Rubiolo , J. Léchelle

{"title":"Study of heterogeneous nucleation dynamics and transport of bubbles for void fraction assessment in a molten salt reactor","authors":"N. Guéritat , S. Maillard , P. Rubiolo , J. Léchelle","doi":"10.1016/j.nucengdes.2026.114812","DOIUrl":"10.1016/j.nucengdes.2026.114812","url":null,"abstract":"<div><div>In Molten Salt Reactors (MSR), the nuclear fuel is liquid, so that one of the primary reactivity feedbacks is the fuel thermal expansion. This feedback is controlled by the speed of sound, which is very sensitive to the void fraction in the liquid salt. A high void fraction could delay the thermal expansion feedback, leading to safety issues during fast reactivity transients. This concern is more pronounced in chloride salts reactors where the Doppler effect is weak. This work presents a generic and efficient method to provide a first-order estimate of the void fraction in any MSR design, considering dissolved fission gases and the plenum cover gas as the only sources of gas in the reactor. Two models are developed and combined to estimate the void fraction: a model of heterogeneous diffusion-driven nucleation, that describes bubble formation from pre-existing gas nuclei trapped in wall cavities, and a model of bubble growth and transport within the reactor. In this study, the method is applied to a specific pre-conceptual design, the Advanced Reactor for Actinide Management in Salt (ARAMIS). The analysis highlights the leading role of the cover gas in the nucleation process. Additionally, this method facilitates parametric studies and physical interpretation of the void fraction behavior. The influence of several parameters on the void fraction (including the solubility, the reactor pressurization, and the fission gases extraction time) is evaluated to identify conditions that either amplify or mitigate it. However, parameter values minimizing void fraction may negatively affect the overall reactor physics, which emphasizes the need for a comprehensive multiphysics approach in MSR design.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114812"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190260","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

Pre-conceptual neutronic design and feasibility analysis of a thorium-based lead‑bismuth cooled small modular reactor 钍基铅铋冷却小型模块化反应堆的概念前中子设计和可行性分析

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-02-06 DOI: 10.1016/j.nucengdes.2026.114815

Xianbao Yuan , Changxiao Guan , Binhang Zhang , Yonghong Zhang

{"title":"Pre-conceptual neutronic design and feasibility analysis of a thorium-based lead‑bismuth cooled small modular reactor","authors":"Xianbao Yuan , Changxiao Guan , Binhang Zhang , Yonghong Zhang","doi":"10.1016/j.nucengdes.2026.114815","DOIUrl":"10.1016/j.nucengdes.2026.114815","url":null,"abstract":"<div><div>Small Modular Reactors (SMRs) are regarded as an important direction for future nuclear energy development due to their enhanced safety, flexible deployment capability and lower investment costs. In this study, neutronics simulations were carried out using OpenMC to design and analyze a long-life and inherently safe thorium-based lead‑bismuth cooled small modular reactor (TL-SMR). The core employs a thorium‑uranium fuel cycle and is cooled by liquid lead‑bismuth eutectic (LBE), enabling low-pressure operation and passive heat dissipation. The core can operate for more than 20 effective full-power years without refueling, achieving deep burnup. The key core parameters were analyzed, including control rod bank worths, radial and axial power factors, effective delayed neutron fractions and reactivity coefficients. The results indicate that the reactor exhibits negative reactivity coefficients, which ensure self-regulation under transient conditions. Throughout the operational cycle, the power distribution remains uniform and reasonable with sufficient shutdown margin. In summary, the TL-SMR demonstrates both long operational lifetime and favorable inherent safety characteristics. This study provides strong evidence for the technical feasibility of such reactors and offers valuable support for the further development of advanced reactor systems with inherent safety features.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114815"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190731","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

Thermal hydraulic analysis of parallel-channel supercritical natural circulation loop: A numerical approach 平行通道超临界自然循环回路的热水力分析：数值方法

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-02-09 DOI: 10.1016/j.nucengdes.2026.114816

A.K. Vias , P.K. Vijayan , G. Dutta

{"title":"Thermal hydraulic analysis of parallel-channel supercritical natural circulation loop: A numerical approach","authors":"A.K. Vias , P.K. Vijayan , G. Dutta","doi":"10.1016/j.nucengdes.2026.114816","DOIUrl":"10.1016/j.nucengdes.2026.114816","url":null,"abstract":"<div><div>In this numerical study, a closed supercritical natural circulation loop (SCNCL) with a parallel-channel configuration is analyzed from a thermal-hydraulic (TH) perspective. An in-house numerical model is developed to capture the axial variation of TH field variables in each heated channel. The model incorporates local property variations at supercritical pressures and integrates sub-models for the pressurizer, cooling heat exchanger (CHX), and wall heat conduction to accurately represent the behavior of the loop. Validation is carried out using benchmarks from the literature, covering steady state and transient conditions. Steady state simulations are performed to evaluate the flow distribution and overall loop performance under varying operating conditions. Emphasis is placed on identifying the effects of property variations and geometric parameters on flow across channels. Transient simulations are carried out to investigate the possibility of onset of core-wide (in-phase) loop and regional (out-of-phase) modes of the instabilities. While validating against the available experimental data, inlet mass flow rates at the two heated channels are found to be undergoing regional mode of oscillations. For the SCNCL with parallel heated channels considered in this analysis, and under the selected operating conditions, any instability observed in the core occurs without phase differences between the corresponding field variables in the parallel heated channels. This indicates that regional (out-of-phase) oscillations do not occur. Consequently, the marginal stability boundary (MSB) for the core-wide oscillation mode is determined, and a comprehensive parametric study is performed to examine the effects of key parameters, such as heater inlet temperature, loop pressure, loop height, and the number of channels, on the stability thresholds of the entire SCNCL system.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114816"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190728","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

Research on fluid-structure interaction characteristics of transient pressure waves in reactor coolant pump under shaft seizure accident 轴扣事故下反应堆冷却剂泵内瞬态压力波流固耦合特性研究

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2026-04-01 Epub Date: 2026-01-30 DOI: 10.1016/j.nucengdes.2026.114795

Teng Niu , Yi Bin Li , Hai Long Yuan , Xue Zhao , Kong Sheng Liu

{"title":"Research on fluid-structure interaction characteristics of transient pressure waves in reactor coolant pump under shaft seizure accident","authors":"Teng Niu , Yi Bin Li , Hai Long Yuan , Xue Zhao , Kong Sheng Liu","doi":"10.1016/j.nucengdes.2026.114795","DOIUrl":"10.1016/j.nucengdes.2026.114795","url":null,"abstract":"<div><div>This study investigates the fluid-structure interaction (FSI) characteristics of transient pressure waves during a reactor coolant pump (RCP) shaft seizure accident (SSA) through bidirectional FSI numerical simulation of the coolant pipeline. Based on a model of the HPR1000 reactor single-loop system with matched resistance characteristics, the analysis focuses on the RCP flow field pressure, internal pressure fluctuations, and pipeline dynamic response. The results demonstrate that coolant pipeline fluid-structure interaction (CPFSI) significantly alters pressure distributions in RCP flow components. During shaft seizure, CPFSI causes a notable expansion of the low-pressure zone at the impeller inlet and an increase in volute pressure. Immediately after shaft seizure, it induces a widespread pressure decrease at the inlet nozzle, a significant enlargement of the low-pressure region within the guide vane flow passage, and a slight expansion of the low-pressure area at the volute outlet. After shaft seizure ends, CPFSI leads to substantial pressure reductions at the inlet nozzle, impeller inlet, volute annular cavity, and volute outlet, alongside a marked expansion of low-pressure zones at the impeller inlet and IPS and a contraction of the high-pressure zone at the GPS inlet. Throughout the shaft seizure transition process, the transition pipe experiences the most pronounced deformation and fluctuation, followed by the hot leg pipe, with the cold leg pipe showing minimal variation. These structural vibrations intensify pressure fluctuations at RCP monitoring points, leading to either attenuation or amplification of transient pressure wave amplitudes. This research reveals the coupling mechanism between transient pressure waves and pipeline dynamics during an SSA, providing a theoretical basis for accurately assessing RCP operational safety.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114795"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081263","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