Nuclear Engineering and Technology最新文献_第10页

Study of circulating liquid fuel in a 1D critical system with thermal feedback 带有热反馈的一维临界系统中的循环液体燃料研究

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-14 DOI: 10.1016/j.net.2024.07.028

Mathis Caprais , Daniele Tomatis , André Bergeron

This research focuses on the description and modeling of a one-dimensional molten salt reactor (MSR), in the presence of thermal feedback. Following the example of previous works, a simple one-dimensional system is proposed, describing a molten salt reactor with a main neutron-multiplying zone called core and a recirculation loop where the salt cools down. Specific attention is paid to the precursors’ drift by modifying the neutron balance equation. Liquid nuclear fuels are characterized by a high volumetric expansion coefficient in comparison to customary solid fuels. Therefore, a strong coupling between neutronics and thermal-hydraulics is expected. As a consequence, a highly negative density coefficient characterizes the thermal feedback on the neutron reactivity. The precursor equation is here inverted analytically and combined with the neutron balance equation to obtain a generalized eigenvalue problem with the neutron flux distribution as the unknown. The balance equations are derived by finite volume integration over a discretized mesh, and the coupling between the two physical models is treated by Picard iterations. The numerical solution is finally extended to time-dependent calculations and compared to an analytical work for a one-dimensional circulating fuel reactor already existing in the literature.

这项研究的重点是在存在热反馈的情况下，对一维熔盐反应堆（MSR）进行描述和建模。借鉴前人的研究成果，我们提出了一个简单的一维系统，描述了一个熔盐反应堆，它有一个被称为堆芯的主中子增殖区和一个盐冷却的再循环回路。通过修改中子平衡方程，对前驱体漂移给予了特别关注。与常见的固体燃料相比，液体核燃料具有体积膨胀系数高的特点。因此，预计中子学和热力学之间会有很强的耦合。因此，高负密度系数是中子反应性热反馈的特征。在这里，前体方程被反演分析，并与中子平衡方程相结合，从而得到一个以中子通量分布为未知数的广义特征值问题。平衡方程是通过离散网格上的有限体积积分得出的，两个物理模型之间的耦合是通过皮卡尔迭代处理的。最后将数值解决方案扩展到随时间变化的计算，并与文献中已有的一维循环燃料反应堆的分析工作进行比较。

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引用次数: 0

Assessing the Phebus FPT-1 experiment: Insights from MELCOR 2.2 and ASYST codes 评估 Phebus FPT-1 试验：从 MELCOR 2.2 和 ASYST 代码中获得的启示

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-14 DOI: 10.1016/j.net.2024.07.021

Peter Mician , Tomas Cerny , Taron Petrosyan , Stepan Foral , Karel Katovsky , Michal Ptacek

Phebus is an experimental facility that represents a scaled-down version of the French 900 MWe pressurized water reactor (PWR) with a ratio of 1/5000. In order to study phenomena occurring during severe accidents in light water reactors, five experiments (FPT-0 to FPT-4) were performed under different fuel and cooling conditions. This paper provides a comprehensive description of the Phebus FPT-1 experiment, MELCOR and ASYST computer codes, discusses the modelling approaches employed and compares the calculated results with the experimental data. To quantitatively assess the results, the Figure of Merits of selected parameters were calculated for both computer codes using ACAP software.

Phebus 是一个实验设施，是法国 900 兆瓦压水反应堆（PWR）的缩小版，比例为 1/5000。为了研究轻水反应堆发生严重事故时的现象，在不同的燃料和冷却条件下进行了五次实验（FPT-0 至 FPT-4）。本文全面介绍了 Phebus FPT-1 实验、MELCOR 和 ASYST 计算机代码，讨论了所采用的建模方法，并将计算结果与实验数据进行了比较。为了对结果进行定量评估，使用 ACAP 软件计算了两种计算机代码选定参数的优劣图。

引用次数: 0

Effect of thermal neutron flux on borate and silicate glasses: Experimental and theoretical investigation for distribution, decay period, and absorbed dose rate of produced isotopes 热中子通量对硼酸盐和硅酸盐玻璃的影响：对所产生同位素的分布、衰变周期和吸收剂量率的实验和理论研究

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-14 DOI: 10.1016/j.net.2024.07.029

O.L. Tashlykov , K.A. Mahmoud , D.O. Kaskov , T.P. Volozheninov

The gamma-ray shielding properties for silicate and borate glass samples were enhanced by heavy metal oxides such as Bi₂O₃, CdO, and Y₂O₃. The current work aims to study experimentally performance of these borate and silicate glasses experimentally when exposed to a flux of thermal neutrons. A nuclear research reactor was utilized to irradiate the fabricated glasses with various doses of thermal neutrons varied between 1.73 and 12.10 MGy. Additionally, the fluency of the thermal neutrons within the dry channel of the IVV-2M reactor containing the fabricated samples varied between 2.27E+17 and 15.86E+17 neutron/cm² at various irradiation times between 1 and 7 days. After that, a gamma-ray spectrometer was utilized to detect the activity concentrations from the irradiated glasses as well as identify the new isotopes created within the fabricated glasses. Additionally, the Monte Carlo simulation code was utilized to estimate the absorbed dose from the irradiated glass samples over a time period between 1 and 120 days after the exposure. The study shows almost all of the activities of the irradiated samples decomposed over 80 days after irradiation. The decomposition of the dose rate and activity concentration for the irradiated samples are attributed to the short lifetime of Ba-131 isotope which represents the major radioactive isotope created within the glass sample.

Bi2O3、CdO 和 Y2O3 等重金属氧化物增强了硅酸盐和硼酸盐玻璃样品的伽马射线屏蔽性能。目前的工作旨在通过实验研究这些硼酸盐玻璃和硅酸盐玻璃在暴露于热中子流时的性能。利用一个核研究反应堆，用 1.73 和 12.10 MGy 之间不同剂量的热中子辐照所制造的玻璃。此外，在 1 至 7 天的不同辐照时间内，IVV-2M 反应堆干通道内的热中子流在 2.27E+17 和 15.86E+17 中子/平方厘米之间变化。然后，利用伽马射线光谱仪检测辐照玻璃的放射性浓度，并确定在制造的玻璃中产生的新同位素。此外，还利用蒙特卡罗模拟代码估算了辐照后 1 至 120 天内辐照玻璃样品的吸收剂量。研究表明，辐照后 80 天内，辐照样品的几乎所有活性都会分解。辐照样品的剂量率和放射性活度浓度的分解是由于玻璃样品中产生的主要放射性同位素 Ba-131 的寿命较短。

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引用次数: 0

Al–B4C-(Gd, Gd2O3) composite materials: Synthesis and characterization for neutron shielding applications Al-B4C-(Gd, Gd2O3) 复合材料：中子屏蔽应用的合成与表征

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-11 DOI: 10.1016/j.net.2024.07.027

Yasin Gaylan , Barış Avar

In this study, Al–20%B₄C-x%Gd and Al–20%B₄C-x%Gd₂O₃ (x = 1, 3, 5) composite powders were prepared using a high-energy planetary ball milling method to enhance the physical properties of Al–B₄C neutron shielding composites. The prepared powders were subjected to uniaxial cold compaction at 500 MPa, resulting in cylindrical specimens. Subsequently, the specimens were sintered in a tube furnace at 600 °C for 1 h under an Ar atmosphere to prevent oxidation. The microstructure of the resulting composites was characterized using X-ray diffraction (XRD) and scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM/EDX). Archimedes density, hardness, and corrosion tests were performed on the compacted samples. Moreover, the composite's thermal and fast neutron absorption rates were calculated using the MCNP6.2 simulation code. The neutron equivalent dose rate was experimentally determined using the Am–Be neutron source. The simulation results demonstrated that the composite materials containing Gd exhibited the highest thermal neutron absorption rate, while those containing Gd₂O₃ demonstrated the highest fast neutron absorption rate. This research contributes valuable insights into the design and utilization of neutron-absorbing materials with suitable mechanical properties.

本研究采用高能行星球磨法制备了 Al-20%B4C-x%Gd 和 Al-20%B4C-x%Gd2O3 (x = 1, 3, 5) 复合粉末，以提高 Al-B4C 中子屏蔽复合材料的物理性能。将制备好的粉末在 500 兆帕的压力下进行单轴冷压实，得到圆柱形试样。随后，在管式炉中以 600 °C 的温度烧结试样 1 小时，并在氩气环境下防止氧化。使用 X 射线衍射（XRD）和扫描电子显微镜与能量色散 X 射线光谱（SEM/EDX）对所得复合材料的微观结构进行了表征。对压实样品进行了阿基米德密度、硬度和腐蚀测试。此外，还使用 MCNP6.2 模拟代码计算了复合材料的热中子吸收率和快中子吸收率。中子当量剂量率是利用 Am-Be 中子源通过实验测定的。模拟结果表明，含 Gd 的复合材料的热中子吸收率最高，而含 Gd2O3 的复合材料的快中子吸收率最高。这项研究为设计和利用具有合适机械性能的中子吸收材料提供了有价值的见解。

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引用次数: 0

Applicability analysis of induction bending process to P91 piping of PGSFR by high-temperature fatigue test 通过高温疲劳试验分析感应弯曲工艺对 PGSFR P91 管道的适用性

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-11 DOI: 10.1016/j.net.2024.07.026

Tae-Won Na , Nak-Hyun Kim , Chang-Gyu Park , Junehyung Kim , Jong-Bum Kim , Il-Kwon Oh

The application of the induction bending process in pipe fabrication is expanding across industries, significantly reducing leakage by minimizing welded sections in curved pipes. In this study, the applicability of the induction bending process to P91 bent pipes of PGSFR was analyzed, focusing on both material fatigue tests and structural fatigue tests for induction bent pipe at high temperatures. First, both high-cycle and low-cycle fatigue tests on specimens from the bent pipe were carried out at 550 °C to confirm that the fatigue properties meet the ASME Code's fatigue requirements. Second, material constants for a Chaboche combined hardening model were identified by using the material test results and an inelastic finite element analysis of the P91 bent pipe were performed to determine the fatigue test load for structural test effectively. Lastly, a high-temperature fatigue test on the bent pipe structure was performed to assess its structural integrity and post-test non-destructive examination confirmed that no fatigue cracks developed, and thereby affirming the applicability of the P91 bent pipe.

感应弯曲工艺在管道制造中的应用正在各行各业不断扩大，它通过最大限度地减少弯曲管道中的焊接部分来显著降低泄漏。本研究分析了感应弯曲工艺对 PGSFR P91 弯管的适用性，重点关注高温下感应弯管的材料疲劳试验和结构疲劳试验。首先，在 550 °C 下对弯管试样进行了高循环和低循环疲劳试验，以确认其疲劳性能符合 ASME 规范的疲劳要求。其次，利用材料试验结果确定了 Chaboche 组合硬化模型的材料常数，并对 P91 弯管进行了非弹性有限元分析，以有效确定结构试验的疲劳试验载荷。最后，对弯管结构进行了高温疲劳试验，以评估其结构完整性，试验后的非破坏性检查证实未出现疲劳裂纹，从而肯定了 P91 弯管的适用性。

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引用次数: 0

Impact of choice of fragility approaches on seismic risk quantification of nuclear power plants 选择脆性方法对核电站地震风险量化的影响

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-10 DOI: 10.1016/j.net.2024.07.023

Baris Kasapoglu, Halil Sezen, Tunc Aldemir, Richard Denning

Development and evaluation of seismic fragility of structures and components is crucial in seismic probabilistic risk assessment of nuclear power plants. Simulation-based fragility approaches are a prevailing trend in the literature, while industry-recommended methodologies rely heavily on engineering judgment and deterministic analysis. In this paper, four critical components are selected, modeled and analyzed as a case study. Their fragilities are evaluated using both state-of-the-art fragility methods and code-recommended methods with approximate models. The impact of the choice of fragility approaches on the fragilities of the components and conditional core damage probability of the plant are assessed. The findings reveal that the recommended approaches employed with approximate models have limitations in estimating the median capacity of complex equipment. While there is a notable variance in the treatment of uncertainty among fragility approaches, its influence on core damage probability remains limited unless the component is the primary contributor to core damage.

开发和评估结构和部件的地震脆性对于核电站的地震概率风险评估至关重要。基于模拟的脆性方法是文献中的主流趋势，而行业推荐的方法则在很大程度上依赖于工程判断和确定性分析。本文选择了四个关键部件作为案例进行建模和分析。采用最先进的脆性方法和规范推荐的近似模型方法对它们的脆性进行了评估。评估了脆性方法的选择对组件脆性和工厂条件核心损坏概率的影响。研究结果表明，采用近似模型的推荐方法在估算复杂设备的中值容量时存在局限性。虽然各种脆性方法在处理不确定性方面存在显著差异，但其对核心损坏概率的影响仍然有限，除非组件是造成核心损坏的主要因素。

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引用次数: 0

Experimental study on practical application of optical fiber sensor (OFS) for high-temperature system 高温系统光纤传感器（OFS）实际应用实验研究

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-09 DOI: 10.1016/j.net.2024.07.025

Byeongyeon Kim , Youngwoong Kim , YunSook Lee , Ki-Ean Nam , Jung Yoon , Yong-Hoon Shin , Hyeonil Kim , Jewhan Lee , BongWan Lee

This study explores the application of Raman scattering-based optical fiber sensors (OFSs) in extreme environments, specifically focusing on a loop heater vessel with temperatures ranging from 200 °C to 680 °C. This condition generally covers the advanced reactor designs, such as Sodium-cooled Fast Reactor and High Temperature Reactor. Various optical fiber combinations were employed for temperature measurements, taking into consideration the operating temperature of the target equipment. Two types of OFSs, gold-coated and polyimide-coated, were utilized. Protective tubes made of stainless steel (STS) and carbon were introduced to ensure reliable temperature data collection in high-temperature settings. Results indicate that the STS tube with a gold-coated OFS exhibited the highest consistency and agreement with thermocouple measurements, making it suitable for extreme environments. The study emphasizes the applicability of this system in high-temperature environments, such as liquid metal reactors, high-temperature thermal energy storage system, and hydrogen production system, for environmental monitoring.

本研究探讨了基于拉曼散射的光纤传感器（OFS）在极端环境中的应用，尤其侧重于温度范围为 200 °C 至 680 °C 的循环加热器容器。这种条件一般涵盖先进的反应堆设计，如钠冷快堆和高温反应堆。考虑到目标设备的工作温度，采用了各种光纤组合进行温度测量。使用了两种类型的 OFS：金涂层和聚酰亚胺涂层。为了确保在高温环境下可靠地收集温度数据，还采用了不锈钢（STS）和碳纤维制成的保护管。结果表明，带有金涂层 OFS 的 STS 管与热电偶测量结果的一致性和一致性最高，因此适用于极端环境。这项研究强调了该系统在高温环境中的适用性，如液态金属反应堆、高温热能存储系统和制氢系统的环境监测。

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引用次数: 0

Thermal dynamics aspect identification of loop heat pipe with capillary tube wick using nonlinear autoregressive exogenous neural network 利用非线性自回归外源神经网络识别带毛细管芯的环形热管的热动力学特性

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-09 DOI: 10.1016/j.net.2024.07.022

Yoyok Dwi Setyo Pambudi , Giarno , Sumantri Hatmoko , Anhar Riza Antariksawan , Mukhsinun Hadi Kusuma

The loop heat pipe (LHP) has the potential to be used as a passive cooling system in small modular reactors. The research objective is to study the thermal dynamics of LHP with capillary tube wick using a non-linear autoregressive exogenous (NARX) based on a neural network. The neural network identification of LHP with capillary tube wick was carried out on the MATLAB platform. The experiment data obtained is used to identify the neural network of LHP with capillary tube wick. The temperature of the water as an evaporator heat source was varied at 60, 70, 80, and 90 °C. The LHP was charged with demineralized water with a filling ratio of 100 %. The air as a coolant in condenser section was blown at velocity of 2.5 m/s. The LHP was vacuumed with an initial pressure of 2690 Pa. The result confirmed that NARX based on the neural network model can predict the temperature of the condenser section with a given input set under the steady-state and transient conditions. The coefficient of determination is higher than 0.998 and Mean Square Error (MSE) is below 0.0082. The result obtained shows that the NARX neural network model can predict thermal dynamics phenomena in LHP quickly and precisely.

环形热管（LHP）具有在小型模块化反应堆中用作被动冷却系统的潜力。研究目的是利用基于神经网络的非线性自回归外生（NARX）来研究带有毛细管芯的 LHP 的热动力学。在 MATLAB 平台上对带毛细管芯的 LHP 进行了神经网络识别。获得的实验数据用于识别带毛细管芯的 LHP 神经网络。作为蒸发器热源的水的温度分别为 60、70、80 和 90 °C。LHP 中装有填充率为 100 % 的去矿物质水。冷凝器部分作为冷却剂的空气以 2.5 米/秒的速度吹出。LHP 被抽成真空，初始压力为 2690 Pa。结果证实，基于神经网络模型的 NARX 可以在稳态和瞬态条件下预测给定输入集的冷凝器部分温度。确定系数大于 0.998，平均平方误差 (MSE) 小于 0.0082。结果表明，NARX 神经网络模型可以快速、准确地预测 LHP 的热动力学现象。

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引用次数: 0

Methodology for estimating the contribution of forest fires in loss of offsite power events 估算森林火灾在异地电力损失事件中所占比例的方法

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-09 DOI: 10.1016/j.net.2024.07.018

Joonseok Lim, Seungsu Han, Hyungdae Kim, Gyunyoung Heo

Forest fires are disasters caused by natural or human factors and have environmental as well as economic impacts such as loss of biodiversity, agricultural damage, and property damage. According to statistics, the number of forest fires and the area damaged are gradually increasing in Korea. In addition, there are cases where the units 1, 2, and 6 of the Hanul Nuclear Power Plant (NPP) have already been affected by forest fires. Korea has recently initiated research on forest fires under multi-unit probabilistic safety assessment (PSA) project. This paper deals with the impact on the loss of offsite power (LOOP), which is one of the NPP incidents and accidents affected by forest fires. To estimate the frequency of LOOP induced by forest fires, the PSA methodology for earthquakes, a representative external event, was referred.

森林火灾是由自然或人为因素引起的灾害，会对环境和经济造成影响，如生物多样性丧失、农业损失和财产损失。据统计，韩国的森林火灾次数和受害面积都在逐渐增加。此外，哈努尔核电站（NPP）的 1 号、2 号和 6 号机组已经受到森林火灾的影响。韩国最近在多机组概率安全评估（PSA）项目下启动了森林火灾研究。本文讨论的是受森林火灾影响的核电站事件和事故之一--场外功率损失（LOOP）的影响。为了估算森林火灾诱发 LOOP 的频率，参考了具有代表性的外部事件--地震的 PSA 方法。

引用次数: 0

Improvement of macroscopic turbulence model for subchannel analysis in the rod bundle array 改进用于杆束阵列子通道分析的宏观湍流模型

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2024-07-09 DOI: 10.1016/j.net.2024.07.020

Seok Kim, Jee Min Yoo, Sang-Ki Moon

The PRIUS program was established to generate an experimental database for the 6 × 4 and 12 × 6 rod bundle geometry. The database will be used to address the subchannel and CFD code analysis required for modeling and validation. This is necessary because Small Break Loss of Coolant Accident (SBLOCA) and Intermediate Break Loss of Coolant Accident (IBLOCA) present three-dimensional phenomena in the core due to the radial power profile, crossflow, and diffusion-dispersion. Therefore, specific experimental programs are required, especially during core reflooding, to investigate the large-scale three-dimensional effects. However, validating each sensitive model of the code separately in the presence of 3D effects is not possible due to the inability to implement instrumentation at high pressure and temperature steam-water flow conditions. The PRIUS test program uses a single-phase flow test to simulate a non-homogeneous velocity distribution and provide information on crossflow with radial mixing effects between subchannels. The CUPID code, which uses a macroscopic turbulence model, has been validated using the PRIUS-II experimental database. Existing macroscopic turbulence models were also validated for their prediction capabilities with different inlet flow conditions. However, the validation revealed significant errors in the shear region between subchannels. An improved macroscopic turbulence model showed promising results in predicting turbulence kinetic energy in porous media analysis.

PRIUS 程序的建立是为了生成 6 × 4 和 12 × 6 杆束几何形状的实验数据库。该数据库将用于进行建模和验证所需的子通道和 CFD 代码分析。这一点很有必要，因为小断裂损失冷却剂事故（SBLOCA）和中断裂损失冷却剂事故（IBLOCA）由于径向功率曲线、横流和扩散弥散，会在堆芯中产生三维现象。因此，特别是在堆芯再充水期间，需要特定的实验计划来研究大规模的三维效应。然而，由于无法在高压和高温蒸汽-水流动条件下安装仪器，因此无法在存在三维效应的情况下分别验证代码中的每个敏感模型。PRIUS 测试程序使用单相流测试来模拟非均质速度分布，并提供子通道之间具有径向混合效应的横流信息。使用宏观湍流模型的 CUPID 代码已通过 PRIUS-II 试验数据库进行了验证。此外，还验证了现有宏观湍流模型在不同入口流动条件下的预测能力。然而，验证结果表明，子通道之间的剪切区域存在明显误差。改进后的宏观湍流模型在预测多孔介质分析中的湍流动能方面显示出良好的效果。

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引用次数: 0