Nuclear Engineering and Design最新文献_第5页

Porous Flow Modeling of Axial Gas Redistribution in Fragmented LWR Fuel Rods using MOOSE 基于MOOSE的碎片化轻水堆燃料棒轴向气体再分布多孔流动模拟

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

Nuclear Engineering and Design

Pub Date : 2025-12-30 DOI: 10.1016/j.nucengdes.2025.114677

Chiara Genoni , Kyle A. Gamble , Davide Pizzocri , Fabiola Cappia , Tommaso Bergomi , Chase Christen , Seongtae Kwon

Understanding how gas axially redistributes within fragmented fuel pellets is crucial for predicting the behavior of Light Water Reactor (LWR) fuel rods during accidental scenarios. Specifically, the time scale of this phenomenon plays a fundamental role in determining the progression and hazard of a Loss Of Coolant Accident (LOCA), especially when high burn-up fuel in a severe state of fragmentation is involved. This study presents a Computational Fluid Dynamics (CFD) model developed within the Multiphysics Object-Oriented Simulation Environment (MOOSE) to predict the time-scale of plenum depressurization in fragmented Light-Water Reactor (LWR) fuel rods. The model examines the effects of incorporating non-linearities in the friction term by comparing the results with experimental data. These data were collected from an experiment that employed surrogate fuel rods containing pellets subjected to mechanical and/or thermal loadings. The objective of the experiement was to reproduce various severity of fuel cracking and to investigate the influence of fuel fragmentation on the dynamics of axial gas redistribution. The results of this study indicate that under certain flow regime conditions – determined by the value of an equivalent Reynolds number – accounting for the non-linear friction term in Navier–Stokes equations guarantees better predictions for the time-scale of plenum depressurization. Also, the model enabled the simulation of the plenum pressure decay by assigning distinct permeability values to each pellet instead of a single uniform value. Multiple simulations were run across all possible combinations of pellets’ positions, having each pellet assigned with values of permeability extracted from the experimental data. This allowed to quantify the impact of the considering various non-uniform distributions of permeability on the dynamics of axial gas redistribution. The present work findings enhance the understanding of axial gas transport, and provide valuable insights for the integration of a model for predicting the axial gas redistribution during a LOCA scenario into the BISON fuel performance code.

了解气体在破碎燃料球团内如何轴向再分布，对于预测轻水反应堆（LWR）燃料棒在意外情况下的行为至关重要。具体来说，这种现象的时间尺度在确定冷却剂损失事故（LOCA）的进展和危害方面起着至关重要的作用，特别是当涉及到处于严重破碎状态的高燃耗燃料时。本研究提出了在多物理场面向对象仿真环境（MOOSE）中开发的计算流体动力学（CFD）模型，用于预测碎片化轻水反应堆（LWR）燃料棒充气降压的时间尺度。该模型通过将结果与实验数据进行比较来检验在摩擦项中加入非线性的影响。这些数据收集自一项实验，该实验使用含有颗粒的替代燃料棒，承受机械和/或热负荷。实验的目的是再现不同程度的燃料裂解，并研究燃料破碎对轴向气体再分布动力学的影响。本研究的结果表明，在一定的流型条件下-由等效雷诺数的值决定-考虑Navier-Stokes方程中的非线性摩擦项可以更好地预测充气降压的时间尺度。此外，该模型通过为每个颗粒分配不同的渗透率值而不是单一的均匀值，从而能够模拟充气压力衰减。在所有可能的颗粒位置组合中进行多次模拟，并为每个颗粒分配从实验数据中提取的渗透率值。这可以量化考虑各种不均匀渗透率分布对轴向气体再分布动力学的影响。目前的研究结果增强了对轴向气体输送的理解，并为将LOCA情景下预测轴向气体再分配的模型集成到BISON燃料性能代码中提供了有价值的见解。

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引用次数: 0

Correlation for CHF at all inclinations in rectangular channels with one side heated 矩形通道中各倾角的CHF的相关关系

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

Nuclear Engineering and Design

Pub Date : 2025-12-30 DOI: 10.1016/j.nucengdes.2025.114734

Mirza M. Shah

Calculation of CHF (Critical Heat Flux) in rectangular channels with one side heated is needed in applications including nuclear and fusion reactors and cooling of computer chips by boiling liquids. While a well-verified correlation for CHF during horizontal and vertical upflow in such channels is available, there is no well-verified correlation for CHF in other orientations. The present research was done to develop a correlation applicable to all orientations. A new correlation is presented which is in reasonable agreement with data from seven sources. The data include inclinations to the horizontal from 0–316 degree, four fluids (water, R-113, FC-72, and PF-5060), hydraulic equivalent diameter 3.2–23.0 mm, heated equivalent diameter 7.3–281 mm, length to diameter ratio 0.14–83, reduced pressure 0.0045–0.071, mass flux 33–6676 kg/m²s, and inlet quality −0.39 to −0.03. The 463 data points from seven experimental studies are predicted with a MAD (Mean Absolute Deviation) of 23.1 %. The same data were also compared to fourteen other correlations. MAD of those correlations ranged from 58 % - 6523 %. The results of this research are presented and discussed. Suggestions are made for further research, including at very low flow rates for which data are not available.

在核聚变反应堆和沸水冷却计算机芯片等应用中，需要计算一侧受热矩形通道的临界热流密度。虽然在这些通道的水平和垂直向上流动期间，CHF的相关性得到了很好的验证，但在其他方向上，CHF的相关性没有得到很好的验证。本研究的目的是建立一种适用于所有方向的相关性。提出了一种新的相关性，与七个来源的数据基本一致。数据包括水平倾角0-316度，四种流体（水，R-113， FC-72和PF-5060），水力等效直径3.2-23.0 mm，加热等效直径7.3-281 mm，长径比0.14-83，减压0.0045-0.071，质量通量33-6676 kg/m2s，进口质量- 0.39至- 0.03。来自7项实验研究的463个数据点的预测MAD（平均绝对偏差）为23.1%。同样的数据还与其他14种相关性进行了比较。这些相关性的MAD范围为58% - 6523%。本文对研究结果进行了介绍和讨论。提出了进一步研究的建议，包括在没有数据的非常低的流速下。

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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 : 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

Probabilistic response analysis of skewed distribution of nuclear power plant structural parameters 核电厂结构参数偏态分布的概率响应分析

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

Nuclear Engineering and Design

Pub Date : 2025-12-29 DOI: 10.1016/j.nucengdes.2025.114697

Wenfu He , Yuxiang Zhou , Ziduan Shang , Tao Wang , Jiawei Ji

Isolation is widely recognized as one of the most effective technologies for protecting nuclear power structure (NPS) from seismic events. As with NPS, the seismic performance of nuclear plant isolated structure (NPIS) is affected not only by the inherent variability of seismic ground motions but also by significant uncertainties in structural characteristics. The addition of an isolation layer using isolation bearings introduces further uncertainty and raises concerns about the reliability of isolated structures. A probability density evolution method (PDEM) is proposed to evaluate the reliability of NPIS under various operational conditions. First, the NPIS is simplified into a nonlinear model representing the superstructure and the isolation layer, and a probability density evolution analysis procedure (PDEAP) is developed. The analysis considers the uncertainty of the parameters of the superstructure and the isolation layer and uses a Weibull distribution to model the skewed uncertainty. A shaking table test of the NPIS is then conducted. Finally, probability density analyses of the displacement are performed for the base earthquake (OBE), the safe shutdown earthquake (SSE) and the over-design base earthquake (2SSE and 3SSE). The shaking table test results indicate that, compared with the NPS, the NPIS reduces peak displacement by 38.71 %, 50.12 %, and 53.21 % under ground motion amplitudes of 0.3 g, 0.6 g, and 0.9 g, respectively. The probability density evolution analysis further reveals that as the peak ground acceleration (PGA) increases from the OBE to 3SSE, the displacement probability density function (PDF) transitions from a narrow, high peak to a broader, flatter distribution, indicating a significant increase in the variability of structural response.

隔震是公认的保护核电结构免受地震影响的最有效技术之一。与核电厂隔离结构一样，核电厂隔离结构的抗震性能不仅受到地震地震动的固有变异性的影响，而且受到结构特征的显著不确定性的影响。使用隔震轴承增加的隔震层引入了进一步的不确定性，并引起了对隔震结构可靠性的担忧。提出了一种概率密度演化法（PDEM）来评估NPIS在各种运行条件下的可靠性。首先，将NPIS简化为表示上部结构和隔震层的非线性模型，并开发了概率密度演化分析程序（PDEAP）。该分析考虑了上部结构和隔振层参数的不确定性，并采用威布尔分布来模拟偏不确定性。然后对NPIS进行了振动台试验。最后，对基础地震（OBE）、安全停堆地震（SSE）和超设计基础地震（2SSE和3SSE）进行了位移概率密度分析。振动台试验结果表明，与NPS相比，NPIS在0.3 g、0.6 g和0.9 g地震动幅值下，峰值位移分别减少了38.71%、50.12%和53.21%。概率密度演化分析进一步表明，随着峰值地加速度（PGA）从OBE到3SSE的增加，位移概率密度函数（PDF）从窄、高的峰值转变为更宽、更平坦的分布，表明结构响应的变异性显著增加。

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引用次数: 0

Verification of PWR-Core power distribution based on precisely calculated SPND response currents 基于精确计算SPND响应电流的压水堆堆芯功率分配验证

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

Nuclear Engineering and Design

Pub Date : 2025-12-27 DOI: 10.1016/j.nucengdes.2025.114726

Sipeng Du, Yunzhao Li, Hangqi Zhang, Ruizhi Shao, Liangzhi Cao

The power distribution of nuclear reactor core is a critical indicator of its operation state. For most reactors, the measured power distribution is generally obtained through the calculated power distribution, the measured and calculated in-core detector response currents. Therefore, the calculation accuracy of the in-core detector response current directly impacts the reliability of the measured power distribution. NECP-Bamboo is a PWR-core physics analysis software developed by the Nuclear Engineering Computational Physics (NECP) Laboratory at Xi'an Jiaotong University in China. Its precise calculation of the Self-Powered Neutron Detector (SPND) response current has been validated based on experimental data. Based on NECP-Bamboo, the paper conducts a quantitative analysis of the impact of the precise calculation of the response current on the verification of the reactor power distribution. The verification results based on the actual measurement data of the AP1000 indicate that the accuracy of NECP-Bamboo in calculating SPND response currents is significantly higher than that of the in-service original dedicated software for AP1000. Using accurately calculated SPND response currents can effectively reduce the error in power distribution, and NECP-Bamboo exhibits a smaller error compared to the in-service original dedicated software.

核反应堆堆芯功率分布是反映反应堆运行状态的重要指标。对于大多数电抗器，测量得到的功率分布一般是通过计算得到的功率分布、实测和计算得到的堆芯内探测器响应电流得到的。因此，芯内检测器响应电流的计算精度直接影响到被测功率分布的可靠性。NECP- bamboo是由中国西安交通大学核工程计算物理（NECP）实验室开发的一款压水堆堆芯物理分析软件。该方法对自供电中子探测器（SPND）响应电流的精确计算得到了实验数据的验证。本文以NECP-Bamboo为基础，定量分析了响应电流的精确计算对电抗器功率分布验证的影响。基于AP1000实际测量数据的验证结果表明，NECP-Bamboo软件计算SPND响应电流的精度明显高于AP1000原有的在用专用软件。利用精确计算的SPND响应电流可以有效地减小功率分配误差，NECP-Bamboo与在用的原有专用软件相比误差更小。

{"title":"Verification of PWR-Core power distribution based on precisely calculated SPND response currents","authors":"Sipeng Du, Yunzhao Li, Hangqi Zhang, Ruizhi Shao, Liangzhi Cao","doi":"10.1016/j.nucengdes.2025.114726","DOIUrl":"10.1016/j.nucengdes.2025.114726","url":null,"abstract":"<div><div>The power distribution of nuclear reactor core is a critical indicator of its operation state. For most reactors, the measured power distribution is generally obtained through the calculated power distribution, the measured and calculated in-core detector response currents. Therefore, the calculation accuracy of the in-core detector response current directly impacts the reliability of the measured power distribution. NECP-Bamboo is a PWR-core physics analysis software developed by the Nuclear Engineering Computational Physics (NECP) Laboratory at Xi'an Jiaotong University in China. Its precise calculation of the Self-Powered Neutron Detector (SPND) response current has been validated based on experimental data. Based on NECP-Bamboo, the paper conducts a quantitative analysis of the impact of the precise calculation of the response current on the verification of the reactor power distribution. The verification results based on the actual measurement data of the AP1000 indicate that the accuracy of NECP-Bamboo in calculating SPND response currents is significantly higher than that of the in-service original dedicated software for AP1000. Using accurately calculated SPND response currents can effectively reduce the error in power distribution, and NECP-Bamboo exhibits a smaller error compared to the in-service original dedicated software.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114726"},"PeriodicalIF":2.1,"publicationDate":"2025-12-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145841527","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

Forced flow transient safety analysis of irradiation device with adjustable orifice for research reactor fuel assemblies 研究堆燃料组件可调孔辐照装置强制流动瞬态安全性分析

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

Nuclear Engineering and Design

Pub Date : 2025-12-27 DOI: 10.1016/j.nucengdes.2025.114711

B. Rossaert , P. Jacquet , K. Stappers , G. Housley , H. Lin , J. Licht , H. Hartman , L. Capriotti , J. Wight

The Belgium Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) has several irradiation devices or rigs that are dedicated to the fuel performance and qualification demonstration testing of research reactor fuels. In support of the U.S. High Performance Research Reactor (USHPRR) LEU conversion project, a new flexible irradiation apparatus, MUSTANG-R, has been constructed. SCK CEN has completed the design and safety study, in cooperation with Idaho National Laboratory (INL) and Argonne National Laboratory (ANL), to allow for the irradiation testing of a full-size fuel assembly in a 200 mm diameter channel in the BR2 reactor. The moveable valve is a key design feature of the device and acts like an adjustable orifice enhancing or restricting the flow through a coolant channel inlet located in the BR2 upper plenum. This moveable valve allows the flow through the device to be adjusted prior to each BR2 cycle to obtain the necessary conditions for the fuel qualification test. This ensures accurate and representative thermal-hydraulic conditions of the fuel design are achieved. The device was designed and qualified as passively safe, implying verification by a combination of mechanical and thermal-hydraulic analysis and testing. This includes characterization of the safety margin required for a scenario where the moveable valve is assumed to be erroneously closed during irradiation. A simplified and conservative method is proposed for analyzing the corresponding forced flow transient using a critical heat flux criterion. This allows the required minimum valve opening to be determined for the experiments' design and safety studies.

比利时核研究中心（SCK CEN）的比利时2号反应堆（BR2）有几个辐照装置或钻机，专门用于研究堆燃料的燃料性能和资格论证测试。为了支持美国高性能研究堆（USHPRR）低浓铀转换项目，已经建造了一种新的柔性辐照装置MUSTANG-R。SCK CEN已经与爱达荷国家实验室（INL）和阿贡国家实验室（ANL）合作完成了设计和安全研究，允许在BR2反应堆直径200毫米的通道中对全尺寸燃料组件进行辐照测试。可移动阀是该装置的一个关键设计特征，它的作用就像一个可调节的孔口，增强或限制通过位于BR2上部静压室的冷却剂通道入口的流量。该可移动阀允许在每个BR2循环之前调整通过该装置的流量，以获得燃料资格测试的必要条件。这确保了燃料设计的准确和代表性的热工条件。该装置的设计和合格是被动安全的，这意味着通过机械和热工分析和测试相结合的验证。这包括假定在辐照过程中活动阀被错误关闭的情况下所需的安全裕度的特征。采用临界热流密度准则，提出了一种简化的、保守的强制流动瞬态分析方法。这样就可以为实验设计和安全研究确定所需的最小阀门开度。

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引用次数: 0

Critical heat flux prediction for internally heated annuli with uniform power 均匀功率内加热环空临界热流密度预测

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

Nuclear Engineering and Design

Pub Date : 2025-12-26 DOI: 10.1016/j.nucengdes.2025.114685

Cole Dunbar, Donkoan Hwang, Juliana Pacheco Duarte

Critical heat flux (CHF) prediction methods are tested against a large database of CHF data for internally heated annuli with uniform power. While some of the existing prediction methods perform well within their applicable parameter range, overall performance on the full dataset is poor. Slice analysis of the CHF data shows that the limited quality region (LQR) occurs at lower quality in annuli than tubes for otherwise identical conditions (i.e., pressure, mass flux). Thus, CHF predictions that rely on the 2006 LUT can significantly overpredict CHF near and above the LQR. The Barnett (1966) correlation provides adequate CHF prediction within its applicable range but covers such a limited range that it cannot be broadly recommended. To this end, two modifications of existing prediction methods are presented for better prediction of CHF in internally heated annuli. A quality correction factor, calculated using a function of pressure and mass flux, is proposed for the 2006 LUT, and the Barnett (1966) correlation is transformed to use local conditions with a broader applicable range. These modified prediction methods show greatly improved CHF predictions across the entire dataset and are recommended by the authors for implementation in relevant thermal hydraulic codes.

针对均匀功率内加热环空的临界热流密度预测方法进行了试验研究。虽然现有的一些预测方法在其适用的参数范围内表现良好，但在完整数据集上的整体性能较差。对CHF数据的切片分析表明，在相同条件下（即压力、质量通量），环空中的有限质量区域（LQR）比管内的质量低。因此，依赖2006年LUT的瑞郎预测可能会明显高估LQR附近和以上的瑞郎。Barnett（1966）相关性在其适用范围内提供了充分的CHF预测，但其覆盖范围有限，不能被广泛推荐。为了更好地预测内加热环空CHF，对现有预测方法进行了两种改进。对2006年LUT提出了一个质量校正因子，使用压力和质量通量的函数计算，并将Barnett（1966）相关性转换为使用适用范围更广的当地条件。这些改进的预测方法对整个数据集的CHF预测有了很大的改进，作者建议在相关的热工规范中实施。

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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 : 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

High-fidelity simulation of molten salt natural circulation loops using the spectral element method 用谱元法高保真模拟熔盐自然循环回路

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

Nuclear Engineering and Design

Pub Date : 2025-12-24 DOI: 10.1016/j.nucengdes.2025.114687

Tri Nguyen , John Barton , Haomin Yuan , Casey Emler , Elia Merzari

This study presents high-fidelity Direct Numerical Simulations (DNS) of natural circulation flow of molten salt in a benchmark loop geometry using the GPU-accelerated spectral element code NekRS. The simulations focus on Test 5 from the University of Wisconsin-Madison FLiBe Natural Circulation Loop (UW-FNCL), using a computational model that matches the experimental setup in geometry, boundary conditions, and operational parameters. A low-Mach number formulation is employed to capture the strong temperature-dependent property variations inherent to FLiBe, a high-Prandtl-number molten salt. Validation against experimental data shows good agreement in temperature profiles and Nusselt numbers across a range of Reynolds numbers, demonstrating NekRS's capability to accurately and efficiently simulate buoyancy-driven flows with thermally varying fluid properties. Additionally, the DNS results provide novel insights into the three-dimensional flow and heat transfer characteristics that are challenging to obtain experimentally. Detailed flow analysis reveals pronounced buoyancy-induced velocity asymmetries in the bottom-heated leg, jet-driven shear instabilities in the reservoir, and localized unsteady phenomena near sharp bends. Proper Orthogonal Decomposition (POD) analysis identifies dominant energetic modes, highlighting a four-vortex Dean-like structure at the 90° elbow that deviates from classical two-vortex predictions, attributed to buoyancy-driven thermal stratification and pre-conditioned velocity profiles.

本研究使用gpu加速谱元代码NekRS在基准环几何结构中对熔盐自然循环流动进行了高保真直接数值模拟（DNS）。模拟的重点是来自威斯康星大学麦迪逊分校的flbe自然循环回路（UW-FNCL）的测试5，使用的计算模型在几何形状、边界条件和操作参数方面与实验设置相匹配。采用低马赫数公式来捕获高普朗特数熔盐FLiBe固有的强温度依赖性质变化。对实验数据的验证表明，在一系列雷诺数范围内，温度分布和努塞尔数具有良好的一致性，证明了NekRS能够准确有效地模拟具有热变化流体性质的浮力驱动流动。此外，DNS结果提供了对三维流动和传热特性的新见解，这些特性很难通过实验获得。详细的流动分析表明，在底部加热段存在明显的浮力引起的速度不对称，在储层中存在射流驱动的剪切不稳定，以及在急弯附近存在局部不稳定现象。适当的正交分解（POD）分析确定了主要的能量模式，突出显示了90°弯头处的四涡迪恩式结构，这与经典的双涡预测不同，归因于浮力驱动的热分层和预先调节的速度剖面。

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引用次数: 0

Analyses of Ibscth tests using thermal-hydraulics and fluid dynamic codes 热工水力学和流体力学规范对Ibscth试验的分析

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

Nuclear Engineering and Design

Pub Date : 2025-12-24 DOI: 10.1016/j.nucengdes.2025.114709

Guido Mazzini , Alis Ruscak Musa , Jakub Spacek , Andrea Pucciarelli , Sara Kassem , Armando Nava Dominguez

Numerous computational tools have been developed worldwide to design and evaluate advanced reactor concepts. For Supercritical Water Reactors (SCWRs), various heat transfer correlations have been developed and integrated into codes like AC2 2021. Concerning Computational Fluid Dynamic (CFD), the available models seem suitable for normal heat transfer conditions while several limitations were observed for Deteriorated Heat Transfer (DHT) conditions. Finally, regarding the subchannel codes, as for an example ASSERT-PV, the developed correlation was tested with refined nodalization.

Within the collaborative efforts of China, Canada, and the European Union, under the Joint European Canadian Chinese Development of Small Modular Reactor Technology (ECC-SMART) project and Generation IV Forum (GIF) activities, multiple system and CFD codes are employed to analyse heat transfer simulations in international benchmark on SCWR thermal-hydraulics characteristics (IBSCTH) tests. These tests encompass three configurations: Bare Rod, Local Wire Rod, and Full-Length Wired Rod experiments. This paper presents the work conducted by the University of Pisa using STAR-CCM+, by CNL with ASSERT-PV and the Research Centre Řež (CVR) employing AC2 2021 to simulate the three configurations. The primary objective is to demonstrate a practical combination of computer codes applicable to further SCW technology design. The simulations showed similar prediction of global heat-transfer behaviour across the three configurations, while notable discrepancies emerged in DHT, typically resulting in conservative overprediction of cladding temperatures. STAR-CCM+ generally yields a certain deviation in DHT regions, while ASSERT-PV demonstrates similar consistency with AC2 2021, particularly under conditions closest to SCWR operation.

世界各地已经开发了许多计算工具来设计和评估先进的反应堆概念。对于超临界水反应堆（SCWRs），各种传热相关性已经开发并集成到AC2 2021等代码中。在计算流体动力学（CFD）方面，现有的模型似乎适用于正常传热条件，而在恶化传热（DHT）条件下则存在一些局限性。最后，对于子信道码，以ASSERT-PV为例，对开发的相关性进行了改进的节点化测试。在中国、加拿大和欧盟的共同努力下，在欧洲、加拿大和中国联合开发小型模块化反应堆技术（ec - smart）项目和第四代论坛（GIF）活动下，采用多种系统和CFD代码来分析SCWR热工水力特性（IBSCTH）国际基准测试中的传热模拟。这些测试包括三种配置：裸杆、局部线杆和全长线杆实验。本文介绍了比萨大学使用STAR-CCM+， CNL使用ASSERT-PV和研究中心Řež （CVR）使用AC2 2021模拟三种配置所进行的工作。主要目的是演示适用于进一步SCW技术设计的计算机代码的实际组合。模拟结果显示，在三种结构中，全球传热行为的预测相似，而在DHT中出现了显著差异，这通常导致对包层温度的保守高估。STAR-CCM+通常在DHT区域产生一定的偏差，而ASSERT-PV与AC2 2021具有相似的一致性，特别是在最接近SCWR运行的条件下。

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