Nuclear Engineering and Design最新文献_第7页

Integrated multiphysics assessment of grid-to-rod fretting mitigation: enhancing nuclear safety through improved fuel integrity 电网到棒微动缓解的综合多物理场评估：通过提高燃料完整性加强核安全

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

Nuclear Engineering and Design

Pub Date : 2026-01-12 DOI: 10.1016/j.nucengdes.2025.114696

Quazi Md. Zobaer Shah , Md Mahabub Hasan Mousum , Debashis Datta , Md. Arefin Kowser , Mohammad Asaduzzaman Chowdhury , Md. Shojib Mia , Shahpara Sheikh Dola

Grid-to-rod fretting (GTRF) represents a persistent challenge to the mechanical integrity and service life of nuclear fuel assemblies across various reactor technologies. As a leading cause of fuel failures in pressurized water reactors and a growing concern in other light water reactor systems, GTRF has drawn significant attention from both experimental and computational communities. Recent investigations have expanded in scope and precision, with advances in multiphysics simulation frameworks, in-reactor diagnostics, and surface-engineered cladding materials. These studies have illuminated complex dependencies on flow turbulence, spacer grid geometry, contact dynamics, and material behavior under irradiation, yet several technical uncertainties continue to limit predictive confidence. This work presents a comprehensive synthesis of developments surrounding GTRF, highlighting current approaches that span high-resolution fluid-structure interaction modeling, fretting-wear characterization, and emerging mitigation strategies. Specific phenomena such as fretting wear response under variable coolant conditions, coating-induced changes in contact fatigue, and the evolving mechanical role of spacer supports are examined through a multidisciplinary lens. While no single mechanism dominates across all contexts, the convergence of insights from structural mechanics, tribology, and reactor operation points toward integrated pathways for addressing the problem. GTRF arises from tightly coupled turbulent forcing, nonlinear rod-support dynamics and evolving material states. Integrated LES→FSI→wear pipelines, validated against autoclave and flow-loop experiments, offer the most promising path toward predictive life-assessment. It was recommended that coordinated high-burnup testing, open benchmark datasets, and development of reduced-order multiphysics tools with uncertainty quantification to enable quantified life-predictions for advanced claddings and higher-burnup operation. Viewing GTRF as a systems-level challenge highlights its implications for fuel reliability and safety and underscores the need for focused research to achieve quantitative, licensing-grade predictions.

网格-棒微动（GTRF）是对各种反应堆技术中核燃料组件的机械完整性和使用寿命的持续挑战。作为压水堆燃料失效的主要原因和其他轻水堆系统日益受到关注，GTRF引起了实验界和计算界的极大关注。随着多物理场模拟框架、反应堆内诊断和表面工程包层材料的进步，最近的研究在范围和精度上都有所扩大。这些研究揭示了流动湍流、间隔网格几何形状、接触动力学和辐照下材料行为的复杂依赖关系，但一些技术上的不确定性继续限制了预测的可信度。这项工作介绍了围绕GTRF的全面综合发展，重点介绍了目前跨越高分辨率流固耦合建模、微动磨损表征和新兴缓解策略的方法。通过多学科的视角，研究了不同冷却剂条件下的微动磨损响应、涂层引起的接触疲劳变化以及垫片支撑的机械作用演变等具体现象。虽然没有单一的机制在所有情况下都占主导地位，但从结构力学、摩擦学和反应堆操作方面的见解的融合指向了解决问题的综合途径。GTRF是由紧密耦合的湍流强迫、非线性杆支撑动力学和不断变化的材料状态引起的。集成LES→FSI→磨损管道，通过高压灭菌器和流动回路实验验证，为预测寿命评估提供了最有希望的途径。建议协调高燃耗测试，开放基准数据集，开发具有不确定性量化的降阶多物理场工具，以实现高级包层和高燃耗操作的量化寿命预测。将GTRF视为系统层面的挑战，凸显了其对燃料可靠性和安全性的影响，并强调了集中研究以实现定量、许可级预测的必要性。

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引用次数: 0

Investigation on diversity of trip parameters in reactor protection system for SALUS under various anticipated operational occurrences SALUS反应堆保护系统在各种预期运行工况下跳闸参数多样性的研究

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

Nuclear Engineering and Design

Pub Date : 2026-01-12 DOI: 10.1016/j.nucengdes.2025.114727

Ji-Woong Han, Sun Rock Choi, In Sub Jun, Seungjoon Baik, Huee-Youl Ye, Jewhan Lee

This paper investigates the diversity of trip parameters in the reactor protection system (RPS) of SALUS, a Small, Advanced, Long-cycled, and Ultimate safe Sodium-cooled fast reactor, for various anticipated operational occurrences (AOOs). Three representative AOO events are selected: transient overpower (TOP), loss of heat sink (LOHS), and loss of core flow (LOF). The MARS-LMR code is used to analyze the thermal-hydraulic behavior and the safety of reactor. The primary and secondary trip variables are identified and delayed effects were analyzed for each event. The results show that the reactor can be safely cooled down in both primary and secondary trip scenarios for all events, with the secondary trip parameters providing adequate protection. The cumulative damage fraction (CDF) values for the fuel cladding integrity remain within safety acceptance criteria. The study demonstrates the selected trip parameters in RPS is proper to ensure the safety of SALUS for the representative AOOs.

本文研究了小型、先进、长周期、终极安全钠冷快堆SALUS的反应堆保护系统（RPS）在各种预期运行事故（AOOs）下跳闸参数的多样性。选择了三种具有代表性的AOO事件：瞬态过电（TOP）、散热器损失（LOHS）和堆芯流量损失（LOF）。采用MARS-LMR程序对反应堆的热工性能和安全性进行了分析。确定了主要和次要行程变量，并分析了每个事件的延迟效应。结果表明，在所有事件的一次和二次脱扣情况下，反应堆都可以安全冷却，并且二次脱扣参数提供了足够的保护。燃料包壳完整性的累积损伤分数（CDF）值保持在安全可接受标准范围内。研究表明，对于具有代表性的AOOs， RPS中选取的行程参数是合适的，可以保证SALUS的安全性。

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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 : 2026-01-10 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

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 : 2026-01-10 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

SMR safety through HTTF modeling and benchmark efforts for code validation for gas-cooled reactor applications SMR安全性通过http建模和基准测试工作，为气冷反应堆应用程序进行代码验证

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

Nuclear Engineering and Design

Pub Date : 2026-01-10 DOI: 10.1016/j.nucengdes.2025.114737

Aaron Epiney , Gerhard Strydom , Robert Kile , Jonathan Barthle , Izabela Gutowska , Benjamin Nakhnikian-Weintraub , Thanh Hua , Ling Zou , Jun Fang , Krishna Podila , Xianmin Huang , Qi Chen , Tariq Jafri , Geoffrey Waddington , Peter Pfeiffer

Accurate modeling and simulation tools for thermal-hydraulics calculations are a key element needed to design and license new advanced reactors including Small Modular Reactors (SMR) and Microreactors. Uncertainties in modeling and simulation can have significant safety and economic implications.

The High Temperature Test Facility (HTTF) at Oregon State University (OSU) is a scaled integral effects experiment designed to investigate transient behavior in high-temperature gas-cooled prismatic-block nuclear reactors. High-quality measurement data is available from the HTTF that is suitable for a thermal-hydraulics code validation benchmark for gas-cooled reactor simulations.

This paper summarizes individual HTTF modeling efforts to date for tool validation at Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Oregon State University (OSU) and Canadian Nuclear Laboratories (CNL) using system thermal-hydraulics codes, Computational Fluid Dynamics (CFD) codes and system-CFD code couplings. Also, the paper introduces the ongoing OECD Nuclear Energy Agency (NEA) High Temperature Gas Reactor Thermal-Hydraulics (HTGR T/H) benchmark that allows for better comparisons of results between different international modeling teams. The benchmark provides well defined computational problems that include code-to-code comparisons and comparisons to measured data. These problems provide an avenue for quantifying accuracy and identifying sources of uncertainty in thermal-hydraulics calculations, including in measured thermophysical properties, as part of validation for gas-cooled reactor simulation tools.

用于热工计算的精确建模和仿真工具是设计和许可新型先进反应堆（包括小型模块化反应堆（SMR）和微反应堆）所需的关键要素。建模和仿真中的不确定性会对安全和经济产生重大影响。俄勒冈州立大学（OSU）的高温试验设施（HTTF）是一个规模积分效应实验，旨在研究高温气冷棱镜堆的瞬态行为。HTTF提供了高质量的测量数据，适用于气冷反应堆模拟的热工水力学代码验证基准。本文总结了迄今为止在爱达荷国家实验室（INL）、阿贡国家实验室（ANL）、俄勒冈州立大学（OSU）和加拿大核实验室（CNL）使用系统热工-水力学代码、计算流体动力学（CFD）代码和系统-CFD代码耦合进行工具验证的各个HTTF建模工作。此外，本文还介绍了正在进行的经合组织核能机构（NEA）高温气体反应堆热工水力学（HTGR T/H）基准，该基准可以更好地比较不同国际建模团队之间的结果。基准测试提供了定义良好的计算问题，包括代码到代码的比较和与测量数据的比较。这些问题为量化精度和确定热工水力计算中的不确定性来源提供了途径，包括在测量的热物理性质中，作为气冷堆模拟工具验证的一部分。

{"title":"SMR safety through HTTF modeling and benchmark efforts for code validation for gas-cooled reactor applications","authors":"Aaron Epiney , Gerhard Strydom , Robert Kile , Jonathan Barthle , Izabela Gutowska , Benjamin Nakhnikian-Weintraub , Thanh Hua , Ling Zou , Jun Fang , Krishna Podila , Xianmin Huang , Qi Chen , Tariq Jafri , Geoffrey Waddington , Peter Pfeiffer","doi":"10.1016/j.nucengdes.2025.114737","DOIUrl":"10.1016/j.nucengdes.2025.114737","url":null,"abstract":"<div><div>Accurate modeling and simulation tools for thermal-hydraulics calculations are a key element needed to design and license new advanced reactors including Small Modular Reactors (SMR) and Microreactors. Uncertainties in modeling and simulation can have significant safety and economic implications.</div><div>The High Temperature Test Facility (HTTF) at Oregon State University (OSU) is a scaled integral effects experiment designed to investigate transient behavior in high-temperature gas-cooled prismatic-block nuclear reactors. High-quality measurement data is available from the HTTF that is suitable for a thermal-hydraulics code validation benchmark for gas-cooled reactor simulations.</div><div>This paper summarizes individual HTTF modeling efforts to date for tool validation at Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Oregon State University (OSU) and Canadian Nuclear Laboratories (CNL) using system thermal-hydraulics codes, Computational Fluid Dynamics (CFD) codes and system-CFD code couplings. Also, the paper introduces the ongoing OECD Nuclear Energy Agency (NEA) High Temperature Gas Reactor Thermal-Hydraulics (HTGR T/H) benchmark that allows for better comparisons of results between different international modeling teams. The benchmark provides well defined computational problems that include code-to-code comparisons and comparisons to measured data. These problems provide an avenue for quantifying accuracy and identifying sources of uncertainty in thermal-hydraulics calculations, including in measured thermophysical properties, as part of validation for gas-cooled reactor simulation tools.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114737"},"PeriodicalIF":2.1,"publicationDate":"2026-01-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939574","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

Integrating best estimate plus uncertainty analysis into model-based systems engineering 将最佳估计和不确定性分析集成到基于模型的系统工程中

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

Nuclear Engineering and Design

Pub Date : 2026-01-09 DOI: 10.1016/j.nucengdes.2026.114752

Alan Matias Avelar , Jian Su , Claudia Giovedi , Fabio de Camargo , Joseph T. Klamo , Oleg Yakimenko

The licensing of new nuclear reactors after the Fukushima accident presents significant challenges due to the complexity of nuclear systems and the stringent regulatory requirements involved. Model-based systems engineering (MBSE) has emerged as a useful approach for managing the development of such complex systems, while Best Estimate Plus Uncertainty (BEPU) methodologies have proven valuable within regulatory frameworks for safety evaluation. However, digital models and databases that are needed to provide evidence that the system meets the specified requirements are usually isolated in discipline-specific data repositories. To address this challenge, this article proposes a model breakdown structure (MBS) methodology, using a set of interconnectable models to seamlessly integrate MBSE, computer-aided engineering (CAE) models, and BEPU analysis. The Brazilian Multipurpose Reactor (RMB) served as the system of interest to exemplify the effectiveness of the proposed methodology. A requirement specification was linked to a finite element analysis (FEA) that estimates the peak cladding temperature in a slow loss of flow accident scenario. Additionally, key design factors are identified using design of experiments (DOE) and analysis of variance (ANOVA). Lastly, Wilks' theorem and Monte Carlo simulations are applied for uncertainty quantification. The results indicate that the 95/95 upper tolerance limit of the peak cladding temperature remains below the onset of nucleate boiling. Furthermore, the utilization of Wilks' theorem can reduce computational cost for uncertainty quantification, and the effect of sampling methods is negligible in Monte Carlo simulations with large sample sizes. This approach can enhance the verification and validation (V&V) of regulatory requirements in the licensing process of new reactors.

由于核系统的复杂性和所涉及的严格监管要求，福岛核事故后新核反应堆的许可面临着重大挑战。基于模型的系统工程（MBSE）已成为管理此类复杂系统开发的有用方法，而最佳估计加不确定性（BEPU）方法已被证明在安全评估的监管框架中具有价值。然而，提供系统满足指定需求的证据所需的数字模型和数据库通常被隔离在特定学科的数据存储库中。为了应对这一挑战，本文提出了一种模型分解结构（MBS）方法，使用一组可互连的模型来无缝集成MBSE、计算机辅助工程（CAE）模型和BEPU分析。巴西多用途反应堆（RMB）作为感兴趣的系统，证明了所提出方法的有效性。需求规范与有限元分析（FEA）相关联，该分析用于估算慢流损失事故场景下的峰值包层温度。此外，使用实验设计（DOE）和方差分析（ANOVA）确定关键设计因素。最后，利用Wilks定理和蒙特卡罗模拟对不确定性进行了量化。结果表明，熔覆温度峰值上限为95/95时，熔覆温度仍保持在核沸腾的起始温度以下。此外，利用Wilks定理可以减少不确定性量化的计算成本，采样方法的影响在大样本量的蒙特卡罗模拟中可以忽略不计。这种方法可以加强对新反应堆许可过程中监管要求的核查和确认（V&；V）。

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引用次数: 0

Comparison of URANS and LES predictions for the open phase of the OECD NEA CSNI fluid structure interaction CFD benchmark OECD NEA CSNI流体结构相互作用CFD基准开放阶段的URANS和LES预测比较

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

Nuclear Engineering and Design

Pub Date : 2026-01-09 DOI: 10.1016/j.nucengdes.2025.114748

Graham Macpherson , David Munn , Alan Johnson , Landon Brockmeyer , Elia Merzari , Jerome Solberg

The OECD NEA CSNI WGAMA CFD Task Group ran a benchmark in 2020 and 2021 to assess the predictive capabilities of coupled fluid structure interaction (FSI) CFD analysis methods. This paper presents the predictions made for the open phase of the benchmark using URANS and LES turbulence modelling approaches, and a comparison of the results to the experimental data.

The benchmark comprised a channel containing two inline cylinders in cross-flow. The cylinders were fixed at one end, free at the other, and had measured resonant frequencies and damping properties. The URANS modelling used ANSYS Fluent 2-way coupled to ANSYS Mechanical. The LES modelling used Nek5000, 1-way coupled to Diablo. Comparisons with cross-channel velocity profiles are presented, both for the mean flow and its RMS. Comparisons are also made to the frequency spectra for point measurements of fluid velocity and pressure, and for the accelerations of the free end of each cylinder.

URANS predicts the average velocity profiles relatively well, and is able to predict the velocity and acceleration spectra at the shedding frequency. However, the frequency content at the 4th harmonic of the shedding frequency is low in the URANS flow fields, and so does not excite accelerations at the resonant frequency of the cylinders. LES makes better predictions of the average profiles, and the velocity spectra agree well at both the shedding frequency and at higher frequencies. The 1-way coupled LES results show good agreement for acceleration spectra.

OECD NEA CSNI WGAMA CFD任务组在2020年和2021年运行了一个基准，以评估耦合流固耦合（FSI） CFD分析方法的预测能力。本文介绍了使用URANS和LES湍流模拟方法对基准的开放阶段进行的预测，并将结果与实验数据进行了比较。基准测试包括一个包含两个横流直列圆柱的通道。圆柱体一端固定，另一端自由，并测量了谐振频率和阻尼特性。URANS建模采用ANSYS Fluent双向耦合ANSYS Mechanical。LES建模采用Nek5000，与Diablo单向耦合。与跨通道速度剖面进行了比较，包括平均流量及其均方根。还比较了流体速度和压力点测量的频谱，以及每个圆柱体自由端加速度的频谱。URANS能较好地预测平均速度分布，并能预测脱落频率下的速度和加速度谱。然而，在URANS流场中，脱落频率的四次谐波频率含量很低，因此不会激发圆柱谐振频率处的加速度。LES对平均剖面的预测效果较好，在脱落频率和更高频率处的速度谱吻合较好。单向耦合LES结果与加速度谱吻合较好。

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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 : 2026-01-08 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

Corrigendum to “Effects of nanosilica and aggregate type on the mechanical, fracture and shielding features of heavyweight concrete” [Nucl. Eng. Des. 431 (2025) 113713] “纳米二氧化硅和骨料类型对重型混凝土力学、断裂和屏蔽特性的影响”的勘误[核]。Eng。第431(2025)113713条]

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

Nuclear Engineering and Design

Pub Date : 2026-01-08 DOI: 10.1016/j.nucengdes.2026.114751

Mohsen Ghorbani , Morteza Biklaryan , Morteza Hosseinali Beygi , Omid Lotfi-Omran

引用次数: 0

Thermal treatment and high performance plasma treatment applied to spent ion exchange resins: study of solid product embedding with ordinary Portland cement 废离子交换树脂的热处理及高性能等离子体处理：固体产物与普通硅酸盐水泥包埋的研究

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

Nuclear Engineering and Design

Pub Date : 2026-01-08 DOI: 10.1016/j.nucengdes.2025.114714

Hernán Ariel Castro , Raul Ariel Rodriguez , Hugo Luis Bianchi

Treatment and conditioning of spent ion exchange resins (IERs) from nuclear facilities is a complex process. The direct immobilization of these materials with a hydraulic binder is usually a first option. However, even the operational procedure of immobilization with cement is not complicated, the volume of final solidified waste form increased significantly and its long-term integrity presents certain limitations.

A strategy internationally considered is to apply a prior treatment step to the spent IERs, mainly thermal treatments, in order to reduce the waste volume and stabilize the product.

In the last few years, our research group has developed a novel technology based on low-temperature thermal treatment of the IERs with steam followed by a High Performance Plasma Treatment (HPPT) of the generated off-gas. The process is capable of achieving high volume reduction factors and a non-reactive solid product.

In the present work, the quality of the solid product obtained in a test bench scale of the process is studied, emphasizing the product compatibility with cement. The solid product embedding with ordinary Portland cement (OPC), without any chemical additives or supplementary materials, was examined. The waste incorporation rate was up to roughly 90% (in volume). The waste form obtained was homogenous and presented compressive strength values around 18 MPa. No evidence of deterioration was observed after 90 days of water immersion.

核设施废离子交换树脂（IERs）的处理和调理是一个复杂的过程。用液压粘合剂直接固定这些材料通常是第一选择。然而，即使水泥固定的操作程序并不复杂，但最终固化废物的体积明显增加，其长期完整性也存在一定的局限性。国际上考虑的一种策略是对废弃的IERs采用预先处理步骤，主要是热处理，以减少废物量并稳定产品。在过去的几年里，我们的研究小组开发了一种基于蒸汽低温热处理的新型技术，然后对产生的废气进行高性能等离子体处理（HPPT）。该工艺能够实现高体积缩小系数和非反应性固体产物。在本工作中，研究了该工艺在试验台规模下获得的固体产品的质量，重点研究了产品与水泥的相容性。采用普通硅酸盐水泥（OPC）包埋固体产品，不添加任何化学添加剂和辅助材料。废物掺入率高达约90%（体积）。所得废料形态均匀，抗压强度在18 MPa左右。浸泡90天后，没有观察到恶化的迹象。

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引用次数: 0