Progress in Nuclear Energy最新文献_第3页

Upgrade of FROBA code to derive new insights in fuel performance of Lead-Bismuth Eutectic Cooled Fast Reactor 升级FROBA代码以获得对铅铋共晶冷却快堆燃料性能的新见解

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-16 DOI: 10.1016/j.pnucene.2026.106252

Yangbin Deng , Zijian Huang , Cong Liu , Ruikuan Li , Yuefeng Wu , Yingwei Wu , Guanghui Su

Lead-Bismuth Eutectic Cooled Fast Reactor (LBEFR) uses low melting point and high boiling point lead-bismuth as the coolant, and is one of the most promising Generation IV nuclear systems. However, fuel rods in LBEFR have to be subjected to more severe operation conditions compared with Light Water Reactor (LWR) fuels, such as harder neutron spectrum, higher temperature, and enhanced coolant corrosiveness, resulting in distinct evolution processes of fuel behavior. In this study, with full consideration of geometry and material design characteristics, models on distinct phenomena of LBEFR fuels, such as fuel constituent migration, pellet restructuring, cladding corrosion, have been developed and implanted into the program FROBA initially developed for LWR fuel analysis. The capability of the upgraded FROBA for LBEFR fuel performance analysis was preliminarily verified with several benchmark cases. Subsequently, a fuel behavior simulation was conducted for MOX-T91 fuel in a compact LBEFR under long-term normal operation and a power ramp following the normal operation. Compared with traditional LWR fuels, the LBEFR fuel operated at significantly higher temperatures, triggering substantially more fission gas release. To accommodate the large volume of released fission gases and suppress internal pressure rise, LBEFR fuel requires a large plenum design. Benefiting from the coolant outlet temperature controlled around 500 °C, the corrosion of T91 cladding was limited, resulting in a minimal oxide layer thickness that poses no significant threat to cladding structural integrity. However, at high burnup stages, the T91 cladding exhibited pronounced irradiation swelling specifically within the critical temperature range of 420 ± 40 °C. To ensure cladding deformation remains within acceptable limits, T91 should avoid operation within this swelling-sensitive temperature range. Power ramp sensitivity analysis reveals that elevated fuel temperatures drive fuel restructuring-induced central void formation, significantly influencing the thermo-mechanical response. In specific, the central void formation could reduce fuel center temperatures and alleviate contact pressure during pellet-cladding mechanical interaction. These findings provide valuable insights crucial for the design and optimization of LBEFR fuels.

铅铋共晶冷却快堆（LBEFR）采用低熔点和高沸点的铅铋作为冷却剂，是最有前途的第四代核系统之一。然而，与轻水堆燃料相比，低沸水堆燃料棒必须经受更严酷的运行条件，如更硬的中子谱、更高的温度和更强的冷却剂腐蚀性，从而导致燃料行为的演变过程不同。本研究在充分考虑几何和材料设计特点的情况下，开发了针对低沸水堆燃料不同现象的模型，如燃料成分迁移、颗粒重构、包层腐蚀等，并将其植入到最初为低沸水堆燃料分析开发的程序FROBA中。通过几个基准算例，初步验证了升级后的FROBA对LBEFR燃料性能分析的能力。随后，对MOX-T91燃料在紧凑型LBEFR中长期正常运行和正常运行后功率斜坡下的燃料行为进行了模拟。与传统的轻水堆燃料相比，LBEFR燃料在更高的温度下运行，引发更多的裂变气体释放。为了容纳大量释放的裂变气体和抑制内部压力的上升，LBEFR燃料需要一个大的静压设计。得益于冷却剂出口温度控制在500℃左右，T91包层的腐蚀受到限制，导致氧化层厚度最小，对包层结构完整性没有明显威胁。然而，在高燃耗阶段，T91包层表现出明显的辐照膨胀，特别是在420±40℃的临界温度范围内。为了确保包层变形保持在可接受的范围内，T91应避免在此膨胀敏感温度范围内运行。功率斜坡敏感性分析表明，燃料温度升高会驱动燃料重构引起的中心空洞形成，显著影响热-机械响应。具体而言，中心空洞的形成可以降低燃料中心温度，减轻颗粒-包层机械相互作用过程中的接触压力。这些发现为LBEFR燃料的设计和优化提供了有价值的见解。

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引用次数: 0

Review of the development of turbulence models for low Prandtl number fluid flows and heat transfer based on machine learning techniques 基于机器学习技术的低普朗特数流体流动和传热湍流模型的发展综述

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-15 DOI: 10.1016/j.pnucene.2026.106264

Ikram Ul Haq, Mohsin Raza, Waqar Ul Hassan, Li-Qi Guo, Jun-Liang Guo, Hong-Na Zhang, Xiao-Bin Li, Feng-Chen Li

Nuclear energy is a clean and sustainable power source with minimal environmental impact. Its effectiveness relies on advanced cooling systems that use low Prandtl number (

P r

) fluids, such as liquid metals, to manage extreme heat in liquid-metal cooled fast reactors (LMFRs). However, predicting the turbulent heat transfer and fluid flow behaviour of these fluids is challenging for traditional Reynolds-averaged Navier–Stokes (RANS) models due to their non-equilibrium turbulence and complex flow patterns. While large-eddy simulation (LES) and direct numerical simulation (DNS) methods are sufficiently accurate, they remain impractical for engineering applications due to their high computational cost. Machine learning (ML) offers innovative solutions to improve turbulence modelling, enabling more accurate predictions and efficient designs. This review highlights the need to advance ML-assisted turbulence modelling for low-

P r

fluids by reviewing traditional models, ML-augmented approaches and their real-world applications. It also addresses current limitations such as limited data, model reliability and generalization issues, while outlining opportunities to develop smarter, safer and more efficient nuclear thermal hydraulics cooling systems for the nuclear energy sector and other cutting-edge technologies.

核能是一种清洁、可持续的能源，对环境的影响最小。它的有效性依赖于先进的冷却系统，该系统使用低普朗特数（Pr）流体，如液态金属，来管理液态金属冷却快堆（LMFRs）中的极端热量。然而，由于这些流体的非平衡湍流和复杂的流动模式，对于传统的reynolds -average Navier-Stokes （RANS）模型来说，预测这些流体的湍流传热和流体流动行为是具有挑战性的。虽然大涡模拟（LES）和直接数值模拟（DNS）方法足够精确，但由于计算成本高，它们在工程应用中仍然不切实际。机器学习（ML）提供了创新的解决方案，以改善湍流建模，实现更准确的预测和高效的设计。这篇综述强调了通过回顾传统模型、ml增强方法及其实际应用来推进ml辅助低pr流体湍流建模的必要性。它还解决了当前的局限性，如有限的数据，模型可靠性和泛化问题，同时概述了为核能部门和其他尖端技术开发更智能，更安全，更高效的核热工液压冷却系统的机会。

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引用次数: 0

Experimental study and software simulation of air-water periodic flooding in a 3×3 bundle channel 3×3束状通道气-水周期性注水试验研究及软件模拟

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-15 DOI: 10.1016/j.pnucene.2026.106262

Guangyuan Jin , Weilian Li , Jinghu Bai , Rui Wang , Yandong Hou

Investigating the behavior of gas-liquid two-phase flow within rod bundle channels can provide theoretical support for the smooth operation and emergency response of reactor systems. The characteristics of flow in a 3 × 3 rod bundle channel under periodic flooding conditions were investigated using numerical simulation methods. Comparing the simulation results with the experimental data revealed a high degree of agreement in the gas-liquid behavior and the time fractions across different flooding regions. This study examines the evolution of the void fraction along the axial direction and the variation of liquid phase velocity within the subchannels. The study of the local liquid phase velocity field analyzes the mechanisms underlying the onset of the flooding front and the formation of liquid phase accumulation. The relationship between the dimensionless pressure gradient and liquid holdup was analyzed, and an examination of the force conditions on the liquid film led to the conclusion that periodic flooding results from the alternating effects of the gas phase's carrying force and the gravity of the liquid phase in different regions. These research findings further enhance the understanding of thermal-hydraulic characteristics within rod bundle channels.

研究棒束通道内气液两相流动特性可以为反应堆系统的平稳运行和应急响应提供理论支持。采用数值模拟的方法研究了周期性注水条件下3 × 3杆束通道内的流动特性。将模拟结果与实验数据进行比较，发现不同驱油区域的气液行为和时间分数高度一致。本研究考察了孔隙率沿轴向的演化和子通道内液相速度的变化。局部液相速度场的研究分析了洪水锋面发生和液相堆积形成的机理。分析了无量纲压力梯度与液含率之间的关系，并对液膜受力条件进行了考察，得出了气相携载力和液相重力在不同区域交替作用的结论。这些研究结果进一步加深了对杆束通道热水力特性的认识。

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引用次数: 0

NURBS-enhanced control volume finite element spatial discretisation methods for the steady-state multigroup neutron diffusion equation 稳态多群中子扩散方程的nurbs增强控制体积有限元空间离散化方法

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-14 DOI: 10.1016/j.pnucene.2025.106197

J. Trainor , M.D. Eaton , J. Kópházi , S.G. Wilson , C. Latimer , L. Smith , D. Baker , I. Jordan

The paper describes a novel spatial discretisation termed the NURBS-enhanced control volume finite element Method (NECVFEM) and applies it to the solution of the steady-state multigroup neutron diffusion equation. The NECVFEM combines the geometric flexibility of FEM with the local, and global, conservation properties of finite volumes and the exact geometric representation associated with Non-Uniform Rational B-Splines (NURBS). In this paper, the NECVFEM and CVFEM discretisations are used to solve three nuclear reactor physics benchmark verification test cases. The numerical accuracy, and computational efficiency, of the NECVFEM and CVFEM have been compared to standard Lagrangian FEM. The Lagrangian FEM has utilised both volume and surface preserving approaches for curvilinear computational domains. The NECVFEM method has been shown to be highly effective in reducing the geometric errors associated with modelling curvilinear computational domains. It has achieved lower numerical errors than the CVFEM across all spatial refinements and benchmark verification test cases. In addition, the NURBS-enhanced methods have been shown to produce geometric errors that are lower, or of similar magnitude, to that of the volume and surface preserving methods while avoiding the need for complex mesh modification steps. A comparison of computational solution times associated with each discretisation method showed that the NECVFEM was more computationally efficient than the CVFEM and comparable to the FEM, particularly for linear spatial discretisations.

本文提出了一种新的空间离散化方法——nurbs -增强控制体积有限元法（NECVFEM），并将其应用于求解稳态多群中子扩散方程。NECVFEM将有限元的几何灵活性与有限体积的局部和全局守恒特性以及与非均匀有理b样条（NURBS）相关的精确几何表示相结合。本文采用NECVFEM和CVFEM离散方法求解了三个核反应堆物理基准验证试验用例。将NECVFEM和CVFEM的数值精度和计算效率与标准拉格朗日有限元法进行了比较。拉格朗日有限元法对曲线计算域采用了体积保持和曲面保持两种方法。NECVFEM方法在减小与曲线计算域建模相关的几何误差方面已被证明是非常有效的。在所有空间细化和基准验证测试用例中，它比CVFEM实现了更低的数值误差。此外，与体积和表面保持方法相比，nurbs增强方法产生的几何误差更低或相似，同时避免了复杂的网格修改步骤。与每种离散化方法相关的计算求解时间的比较表明，NECVFEM比CVFEM计算效率更高，与FEM相当，特别是对于线性空间离散。

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引用次数: 0

foamForNuclear: A unified OpenFOAM multi-physics platform for nuclear applications foamForNuclear：一个统一的OpenFOAM多物理场核应用平台

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-13 DOI: 10.1016/j.pnucene.2026.106250

Giovanni Nervi , Thomas Guilbaud , Carlo Fiorina , Mathieu Hursin , Alessandro Scolaro

OpenFOAM has been extensively used to develop dedicated physics solvers for nuclear applications, ranging from thermal-hydraulics to structural mechanics and neutronics. However, existing frameworks still lack a general-purpose structure for coupling arbitrary physics across multiple regions and meshes. This work presents foamForNuclear, a new OpenFOAM-based open-source multiphysics architecture that addresses this limitation. Building on the established GeN-Foam and OFFBEAT codes, the platform introduces a flexible, modular structure that enables users to configure simulations with different combinations of physics, solvers, and coupling strategies without source code modifications. The platform architecture, physics libraries, and verification and validation processes are detailed. Its capabilities and applicability to both fission and fusion systems are demonstrated through selected test cases, including the simulation of the ORNL 19-pin sodium-cooled fuel assembly, the sloshing of liquid metals in pool-type reactors under seismic loading, and the thermal performance of a dual-coolant lithium-lead fusion blanket.

OpenFOAM已广泛用于开发核应用的专用物理求解器，范围从热工水力学到结构力学和中子学。然而，现有的框架仍然缺乏一种通用的结构来跨多个区域和网格耦合任意物理。这项工作提出了foamForNuclear，一个新的基于openfoam的开源多物理场体系结构，解决了这个限制。基于已建立的GeN-Foam和OFFBEAT代码，该平台引入了灵活的模块化结构，使用户能够在不修改源代码的情况下配置具有不同物理，求解器和耦合策略组合的模拟。详细介绍了平台架构、物理库以及验证和验证过程。通过选定的测试案例，包括ORNL 19针钠冷却燃料组件的模拟，地震载荷下池式反应堆中液态金属的晃动，以及双冷却剂锂铅聚变包层的热性能，证明了其在裂变和聚变系统中的能力和适用性。

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引用次数: 0

Impact of Am-241 on proliferation-resistant fuel cycle in boiling water reactors Am-241对沸水堆防扩散燃料循环的影响

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-13 DOI: 10.1016/j.pnucene.2025.106220

Mustafa J. Bolukbasi , Marat Margulis

Nuclear power is a cornerstone for achieving global sustainable energy goals, offering a low-carbon alternative to fossil fuels and promising significant contributions to energy security and climate change mitigation. However, the dual-use nature of nuclear technology presents inherent risks of nuclear proliferation, where materials and technologies intended for civilian energy production could be diverted to develop nuclear weapons. In this study, the impact of adding ²⁴¹Am to the fuel composition on reactor operations for BWRs was analysed using CASMO-4/SIMULATE-3, focusing on achieving a proliferation-resistant fuel cycle. An examination was conducted on the breeding behaviours of plutonium atoms, as well as on power distribution and thermal behaviour. Analyses of neutronics and the fuel cycle have demonstrated that the addition of ²⁴¹Am can facilitate the attainment of the necessary ²³⁸Pu ratio for a proliferation-resistant fuel cycle in BWR operations. Despite the requirement to moderately increase uranium enrichment to sustain cycle duration, it was observed that this adjustment does not significantly alter the power or thermal profiles.

核电是实现全球可持续能源目标的基石，是化石燃料的低碳替代品，有望为能源安全和减缓气候变化作出重大贡献。然而，核技术的双重用途性质带来了核扩散的固有风险，用于民用能源生产的材料和技术可能被转用于发展核武器。在本研究中，使用CASMO-4/ simulation -3分析了在燃料成分中添加241Am对沸水堆反应堆运行的影响，重点是实现抗扩散的燃料循环。对钚原子的增殖行为以及功率分布和热行为进行了研究。中子学和燃料循环分析表明，添加241Am可以促进实现沸水堆运行中抗扩散燃料循环所需的238Pu比。尽管需要适度增加铀浓缩以维持循环时间，但观察到这种调整不会显著改变功率或热分布。

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引用次数: 0

Cross working conditions fault diagnosis of marine nuclear power system secondary loops based on multi-source domains adaptation method 基于多源域自适应方法的船用核电系统二次回路交叉工况故障诊断

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-12 DOI: 10.1016/j.pnucene.2026.106257

Haotong Wang , Yanjun Li , Haonan Sha , Guolong Li , Shengdi Sun , Xin Zhou

Marine nuclear power systems are difficult to obtain comprehensive maintenance during navigation, especially for the secondary loops with numerous devices and pipelines. Therefore effective fault diagnosis methods are needed to seize maintenance opportunities. However, marine nuclear power systems have variable working conditions and random faults, making it difficult to collect all fault types’ data under the same working condition. This leads to the need to train models with other multiple working conditions data to diagnose faults under specific working condition, which is the multi-source domains data condition. In response to the cross working conditions fault diagnosis problem of marine nuclear power system secondary loops, a simulation model was firstly built and validated to obtain multiple working conditions’ faults data. Subsequently, a novel CNN feature engineering module and a novel joint loss function were proposed to improve the MLDAN domain adaptation classification algorithm. Finally, the novel method was compared with other methods under different cross working conditions domain adaptation fault diagnosis situations. Due to the novel method’s abilities to capture global and detailed features, effectively utilize samples’ feature and labels information and combine different loss functions’ advantages, it achieved the highest accuracy in various situations even compared to advanced competitors.

船舶核动力系统在航行过程中难以获得全面的维护，特别是对于设备和管道众多的二次回路。因此，需要有效的故障诊断方法来抓住维护机会。然而，船舶核电系统的工作状态是可变的，故障是随机的，很难收集到相同工作状态下的所有故障类型的数据。这就导致需要用其他多工况数据训练模型来诊断特定工况下的故障，这就是多源域数据条件。针对船舶核电系统二次回路的跨工况故障诊断问题，首先建立仿真模型并进行验证，获得多工况故障数据。随后，提出了一种新的CNN特征工程模块和一种新的联合损失函数来改进MLDAN域自适应分类算法。最后，对不同工况域自适应故障诊断情况下的其他方法进行了比较。由于新方法能够捕获全局和详细的特征，有效地利用样本的特征和标签信息，结合不同损失函数的优势，即使与先进的竞争对手相比，它在各种情况下也取得了最高的准确性。

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引用次数: 0

Machine learning-enhanced dynamic probabilistic safety assessment for station blackout in Bushehr nuclear power plant 基于机器学习的布什尔核电站停电动态概率安全评估

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-12 DOI: 10.1016/j.pnucene.2026.106240

Mehdi Yarizadeh-Bene , Mahdi Zangian , Abdolhamid Minuchehr , Hamed Kargaran

Dynamic Probabilistic Safety Assessment (D-PSA) faces significant computational challenges when simulating complex accident scenarios with high-fidelity thermal-hydraulic codes, necessitating innovative approaches to balance accuracy and efficiency. This study introduces a machine learning-enhanced Dynamic Event Tree (DET) module that improves risk analysis quantification by integrating deep learning and Support Vector Machine (SVM). This module utilizes a FORTRAN-based submodule for automated generation and parallel execution of thermal hydraulic model input files, enabling concurrent simulation of several of dynamic scenarios while a random sampling strategy ensures comprehensive coverage of failure sequences with minimized training data requirements. In this research a Station blackout (SBO), as initiating event at Bushehr nuclear power plant (BNPP) is considered for thermal-hydraulic model benchmark. Consequently, dynamic effect of diesel generators recovery in different time steps are evaluated. As a result, SBO turns into the loss of offsite power (LOOP) accident. The proposed module investigates LOOP accident for comprehensive risk analysis at BNPP. The results show that the machine learning architecture developed achieves good predictive performance, surpassing at least 97 % accuracy in classifying core damage states and reducing scenario evaluation time. High-resolution dynamic modeling combined with computational feasibility in this module represents fast and effective method in nuclear safety analysis.

动态概率安全评估（D-PSA）在使用高保真的热液代码模拟复杂事故场景时面临着巨大的计算挑战，需要创新的方法来平衡准确性和效率。本研究引入了一个机器学习增强的动态事件树（DET）模块，通过整合深度学习和支持向量机（SVM）来改进风险分析的量化。该模块利用基于fortran的子模块自动生成和并行执行热水力模型输入文件，实现多个动态场景的并发模拟，而随机抽样策略确保以最小的训练数据需求全面覆盖故障序列。本研究以布什尔核电站的一次电站停电为起始事件，作为热工水力模型基准。在此基础上，对柴油发电机组在不同时间步长的动态回收效果进行了评价。因此，SBO演变为场外失电（LOOP）事故。提出的模块调查LOOP事故，以进行BNPP的综合风险分析。结果表明，所开发的机器学习架构取得了良好的预测性能，在分类堆芯损伤状态和减少场景评估时间方面准确率至少超过97%。该模块的高分辨率动态建模与计算可行性相结合，代表了核安全分析快速有效的方法。

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引用次数: 0

Exploration of grid types for radiation field representing and path-finding 辐射场表示和寻路网格类型的探索

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-12 DOI: 10.1016/j.pnucene.2026.106251

Jiamei Tang , Xiaodan Li , Mengkun Li , Li Liu , Mengxiao Wang

引用次数: 0

Innovative design and safety evaluation of the decay heat removal system for the China fast reactor 中国快堆消热系统的创新设计与安全性评价

IF 3.2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2026-01-10 DOI: 10.1016/j.pnucene.2026.106238

Zhiwen Dai , Donghui Zhang , Yuting Yang , Zhiwei Zhou , Chao Lin , Xiuli Xue , Xintai Yu , Dalin Zhang , Xiyong Chen , Shangang Cao , Songping Wang , Chengwen Xing , Shuiwen Jiang

Pool-type sodium-cooled fast reactors (SFR) have become one of the main selections of Generation-IV reactors due to large thermal inertia and inherent safety, which solve the future shortage of natural uranium and the disposal challenges of spent nuclear fuel (SNF). The decay heat removal system (DHRS) is one of the most important safety systems and must be highly reliable. This study illustrates the design and innovations of the DHRS on the China Fast Reactor. A thermal-hydraulic analysis was conducted using the system program (named ERAC) under station blackout (SBO) conditions, and key parameters of the natural circulation process were evaluated. China's fast reactor design is innovative in many respects, and its novel DHRS design ensures the reactor's safety during emergencies. The analysis results show that the DHRS system operates effectively and that the calculations align with the design goals. Under natural circulation, the peak temperature reached approximately 592 °C at 1000 s. As natural circulation progressed, the core outlet temperature gradually decreased; by 5000 s, the average core fuel outlet temperature was 574 °C. The design of the core throttling component meets the requirements and can provide sufficient natural circulation. This study could provide a valuable reference for the design of SFRs.

池式钠冷快堆（SFR）由于热惯量大、固有安全性好，已成为第四代反应堆的主要选择之一，解决了未来天然铀短缺和乏核燃料（SNF）处理的难题。衰变排热系统（DHRS）是最重要的安全系统之一，必须具有高可靠性。本研究说明了中国快堆的DHRS的设计和创新。利用系统程序（ERAC）进行了电站停电条件下的热液分析，并对自然循环过程的关键参数进行了评价。中国的快堆设计在许多方面具有创新性，其新颖的DHRS设计确保了反应堆在紧急情况下的安全性。分析结果表明，该系统运行有效，计算结果符合设计目标。在自然循环条件下，1000 s温度峰值约为592℃。随着自然循环的进行，岩心出口温度逐渐降低；到5000s时，堆芯燃料出口平均温度为574℃。核心节流元件的设计满足要求，并能提供充分的自然循环。本研究可为SFRs的设计提供有价值的参考。

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引用次数: 0