Nuclear Engineering and Design最新文献_第8页

Impact assessment of internal explosives on physical barriers within nuclear facilities through demonstration testing 通过演示试验评估内部爆炸物对核设施内物理屏障的影响

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-28 DOI: 10.1016/j.nucengdes.2024.113653

Taegwan Do, Yun Seon Chung, Hyeseung Kim, Seung Rae Kim, Wooseub Kim, Sun Do Choi

The objective of this study was to assess the impact of internal explosives on the physical barriers (reinforced concrete and fireproof doors) of nuclear facilities by conducting explosives demonstration tests and comparing the results with computer code results. In this study, we conducted internal explosion tests on physical barriers (reinforced concrete and fireproof doors) within a nuclear facility, with the weights of the explosives set at 20 g, 100 g and 150 g (TNT criteria), to measure the pressure changes corresponding to each weight. These tests aimed to analyzed the pressure distribution and displacement effects on the structure. An array of sensors, including LVDTs, and incident and reflect pressure gauges were used to record the blast tests and capture dynamic pressure and displacement responses at critical structural points. The experimental results indicated significant variations in pressure distributions according to the placement and quantity of the explosives. A notable finding in test B-2 revealed that despite the explosives being detonated at the same location as in other tests, the resulting pressures were usually higher at the ceiling rather than at the walls, contrary to the outcomes observed in other experiments. This pattern demonstrates the complexity of internal blast dynamics and suggests interference from reflected waves. In addition, the experiments indicated that as the weight of the explosives increased, the time intervals between successive pressure peaks decreased, suggesting a faster propagation of pressure waves with heavier explosives. To complement the experimental data, computational simulations using AUTODYN were conducted. These closely reflected the experimental results, with a maximum displacement discrepancy of 14.5 %. This research will contribute to the field by providing empirical data and validated models that can be used to enhance the design standards for blast protection for concrete walls and doors in major national facilities, particularly nuclear facilities.

本研究的目的是通过进行爆炸物演示试验并将结果与计算机代码结果进行比较，评估内部爆炸物对核设施物理屏障（钢筋混凝土和防火门）的影响。在这项研究中，我们对核设施内的物理屏障（钢筋混凝土和防火门）进行了内部爆炸试验，将炸药重量设定为 20 克、100 克和 150 克（TNT 标准），以测量与每种重量相对应的压力变化。这些测试旨在分析压力分布和位移对结构的影响。包括 LVDT、入射和反射压力表在内的一系列传感器用于记录爆炸试验，并捕捉关键结构点的动态压力和位移响应。实验结果表明，根据炸药的位置和数量，压力分布存在很大差异。在 B-2 试验中的一个显著发现是，尽管炸药是在与其他试验相同的位置引爆的，但所产生的压力通常是在天花板处而不是在墙壁处，这与其他试验中观察到的结果相反。这种模式表明了内部爆炸动力学的复杂性，并暗示了反射波的干扰。此外，实验表明，随着炸药重量的增加，连续压力峰值之间的时间间隔缩短，这表明炸药越重，压力波的传播速度越快。为了补充实验数据，我们使用 AUTODYN 进行了计算模拟。这些模拟结果与实验结果密切相关，最大位移差异为 14.5%。这项研究将通过提供经验数据和验证模型为该领域做出贡献，这些数据和模型可用于提高主要国家设施（尤其是核设施）混凝土墙和门的防爆设计标准。

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引用次数: 0

Modelica and Arduino-based hardware-in-the-loop simulation of a nuclear-powered engineering ship 基于 Modelica 和 Arduino 的核动力工程船硬件在环仿真

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-28 DOI: 10.1016/j.nucengdes.2024.113650

Ao Zhang , Xun He , Antonio Cammi , Xiang Wang

Hardware-in-the-loop simulation (HILS) can enhance the authenticity and reliability of offline simulation at the design stage, and reduce the risk of equipment performance commissioning for complex systems. This paper focuses on the HILS of the control mechanism of a nuclear-powered engineering ship including a two-loop nuclear power system, an electric power system, and a mechanical system. In the software part, the mathematical model is implemented in Modelica language with OpenModelica. The nuclear power system is demonstrated in steady-state referring to the Japanese nuclear power merchant ship “NS Mutsu” and is debugged under transient conditions with the help of an operation control module. The electric power system and mechanical system were developed to have certain functions for marine engineering. In the hardware part, the system is built based on an Arduino microcontroller, the Modelica open-source library, and a Bluetooth-based communication protocol between the computer and the microcontroller. The study proved that HILS is capable of simulating the multi-physical joint operation on the software level, establishing the real-time action response and data feedback between the software and hardware parts, and completing steady-state as well as various transient simulations.

硬件在环仿真（HILS）可以提高设计阶段离线仿真的真实性和可靠性，降低复杂系统设备性能调试的风险。本文重点研究了一艘核动力工程船的控制机制的 HILS，包括双回路核动力系统、电力系统和机械系统。在软件部分，数学模型是用 Modelica 语言和 OpenModelica 实现的。核动力系统参照日本核动力商船 "NS Mutsu "号进行了稳态演示，并在运行控制模块的帮助下进行了瞬态调试。开发的电力系统和机械系统具有一定的海洋工程功能。在硬件部分，系统基于 Arduino 微控制器、Modelica 开源库以及计算机与微控制器之间基于蓝牙的通信协议构建。研究证明，HILS 能够在软件层面模拟多物理关节的运行，在软件和硬件部分之间建立实时动作响应和数据反馈，并完成稳态和各种瞬态模拟。

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引用次数: 0

Applications of deep reinforcement learning in nuclear energy: A review 深度强化学习在核能中的应用：综述

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-26 DOI: 10.1016/j.nucengdes.2024.113655

Yongchao Liu , Bo Wang , Sichao Tan , Tong Li , Wei Lv , Zhenfeng Niu , Jiangkuan Li , Puzhen Gao , Ruifeng Tian

In recent years, Deep reinforcement learning (DRL), as an important branch of artificial intelligence (AI), has been widely used in physics and engineering domains. It combines the perceptual advantages of deep learning (DL) and the decision-making advantages of reinforcement learning (RL), and is very suitable for solving the “perception-decision” problem with high-dimensional and nonlinear characteristics. In this paper, firstly, the algorithm principle, mainstream framework, characteristics and advantages of DRL are summarized. Secondly, the application research status of DRL in other energy fields is reviewed, which provides reference for the possible impact and future research direction in the field of nuclear energy. Thirdly, the main research directions of DRL in the field of nuclear energy are summarized and commented, and the application architecture and advantages of DRL are illustrated through specific application cases. Finally, the advantages, limitations and future development direction of DRL in the field of nuclear energy are discussed. The goal of this review is to provide an understanding of DRL capabilities along with state-of-the-art applications in nuclear energy to researchers wishing to address new problems with these methods.

近年来，深度强化学习（DRL）作为人工智能（AI）的一个重要分支，在物理和工程领域得到了广泛应用。它结合了深度学习（DL）的感知优势和强化学习（RL）的决策优势，非常适合解决具有高维和非线性特征的 "感知-决策 "问题。本文首先总结了 DRL 的算法原理、主流框架、特点和优势。其次，回顾了 DRL 在其他能源领域的应用研究现状，为其在核能领域可能产生的影响和未来研究方向提供参考。第三，总结并评述了 DRL 在核能领域的主要研究方向，并通过具体应用案例说明了 DRL 的应用架构和优势。最后，讨论了 DRL 在核能领域的优势、局限性和未来发展方向。本综述的目的是让希望使用这些方法解决新问题的研究人员了解 DRL 的能力以及在核能领域的最新应用。

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引用次数: 0

Characterization of long-term evolution of leakage rates of O-ring seals in nuclear power plant under high-temperature and high-pressure conditions 高温高压条件下核电站 O 型圈密封件泄漏率的长期演变特征

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-25 DOI: 10.1016/j.nucengdes.2024.113666

Ming Li , Lei He , Dengke Zheng , Xing Fang , Xiaoming Huang , Guoliang Xu

Under accident conditions, the sustained high temperature and high pressure (HTHP) environment within the containment structure poses a threat to the penetration seals. This study integrates finite element analysis of the mechanical properties of O-ring, both hyperelastic and viscoelastic characteristics, and an interfacial leakage model to predict the variation in O-ring leakage rates over time in HTHP environments. A series of leakage test experiments are conducted to validate the predictive model, indicating good agreement between experimental and predicted values. The effects of HTHP on non-aged O-rings (short-term service) are analyzed through mechanical simulations and leakage rate calculations. The results reveal that high temperatures positively and reversibly affect the O-ring seals, with hazardous conditions mainly resulting from over-pressurization. However, during the long-term service of aged O-rings, thermal aging caused by high temperatures significantly influences leakage rates. The thermal aging coupled with high pressure can cause material damage (such as rubber being squeezed out) and functional failures (excessive leakage rates). The long-term leakage rates of O-rings at high temperatures in further investigation exhibits a time–temperature equivalence. The master curve is plotted to derive an equation that describes the relationship between leakage rates, temperature, and time under specific pressure conditions. The equation indicates that the dimensionless leakage rate serves as an indicator of seal degradation and enables the quantitative evaluation of the long-term service life of the O-rings using the maximum allowable leakage rate. These findings are applicable within the range of accidental operating conditions for containment structures, including temperatures up to 160 °C and pressures up to 0.75 MPa.

在事故条件下，安全壳结构内的持续高温高压（HTHP）环境对贯穿密封件构成威胁。本研究整合了 O 形环机械性能的有限元分析（包括超弹性和粘弹性特性）和界面泄漏模型，以预测 HTHP 环境中 O 形环泄漏率随时间的变化。为验证预测模型，进行了一系列泄漏测试实验，结果表明实验值与预测值之间具有良好的一致性。通过机械模拟和泄漏率计算，分析了 HTHP 对非老化 O 形环（短期使用）的影响。结果表明，高温会对 O 形圈密封件产生正向和可逆的影响，危险情况主要来自于过压。然而，在老化 O 形环的长期使用过程中，高温引起的热老化会对泄漏率产生重大影响。热老化加上高压会导致材料损坏（如橡胶被挤出）和功能失效（泄漏率过高）。在进一步调查中，O 形圈在高温下的长期泄漏率呈现出时间-温度等效关系。绘制主曲线可以得出一个方程，描述特定压力条件下泄漏率、温度和时间之间的关系。方程式表明，无量纲泄漏率可作为密封退化的指标，并能利用最大允许泄漏率对 O 形圈的长期使用寿命进行定量评估。这些发现适用于安全壳结构的意外运行条件范围，包括温度高达 160 °C 和压力高达 0.75 MPa。

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引用次数: 0

Study and analysis of the effect of dynamic load on the application of LBB technology to the primary pipe of reactor coolant system 研究和分析动载荷对反应堆冷却剂系统一次管道应用 LBB 技术的影响

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-24 DOI: 10.1016/j.nucengdes.2024.113646

Feng He , Furui Xiong, Xiaoming Bai, Kang Yang, Xifeng Lu, Bingjin Li, Xinjun Wang

Reactors may be subjected to dynamic loads, and in order to ensure that Leak-Before-Break (LBB) technology is safely and reliably applied to the primary piping of reactor coolant system, dynamic load impact analysis of LBB technology applied to the primary piping was performed. The dynamic mechanical properties test was carried out on the primary piping, and the crack leakage rate test under dynamic load and the impact load loading test under different loading speeds were carried out on the primary piping test piece with circumferential penetrating crack. Based on experimental data, the crack leakage rate analysis method and crack stability analysis method in quasi-static LBB technology were used to analyze the impact of applying LBB technology to the primary piping under dynamic load. The analysis results indicate that, when LBB technology is applied to the primary piping of reactor coolant system, the analysis results under dynamic load are more conservative than those under quasi-static load, and it is safer and more reliable under dynamic loads in practical engineering applications.

反应堆可能会受到动态载荷的影响，为了确保在反应堆冷却剂系统一次管道上安全可靠地应用 "先漏后破"（LBB）技术，对应用于一次管道的 "先漏后破 "技术进行了动态载荷冲击分析。对一次管道进行了动态力学性能试验，并对带有圆周贯穿裂纹的一次管道试件进行了动载荷下的裂纹泄漏率试验和不同加载速度下的冲击载荷加载试验。根据试验数据，采用准静态 LBB 技术中的裂纹泄漏率分析方法和裂纹稳定性分析方法，分析了在动载荷下应用 LBB 技术对主管道的影响。分析结果表明，将 LBB 技术应用于反应堆冷却剂系统一次管道时，动载荷下的分析结果比准静载荷下的分析结果更为保守，在实际工程应用中，动载荷下的分析结果更为安全可靠。

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引用次数: 0

Investigation on nonlinear dynamics and fatigue damage of a simply supported tube with clearance restriction in tube bundles subjected to two-phase flow 受两相流作用的管束中带有间隙限制的简支管的非线性动力学和疲劳损伤研究

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-24 DOI: 10.1016/j.nucengdes.2024.113644

Shihao Yang , Jiang Lai

In this study, a mathematical model was developed to analyze the vibration of a simply supported tube with clearance restriction in an otherwise rigid rotated triangular tube array subjected to two-phase flow. The study evaluated the fatigue damage and collision behavior of the tube bundles system. The numerical results showed that the reduced velocity significantly affects the dominant frequency, collision frequency, and stable equilibrium position of the system. This suggests that the nonlinear dynamic characteristics of the simply supported tube with clearance restriction play a crucial role in determining fatigue damage.

本研究建立了一个数学模型，用于分析在受到两相流作用的刚性旋转三角管阵列中带有间隙限制的简支撑管的振动。研究评估了管束系统的疲劳损伤和碰撞行为。数值结果表明，速度降低会显著影响系统的主频、碰撞频率和稳定平衡位置。这表明，带有间隙限制的简支管的非线性动态特性在决定疲劳损伤方面起着至关重要的作用。

引用次数: 0

Generic material irradiation database for delayed heating calculations 用于延迟加热计算的通用材料辐照数据库

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-24 DOI: 10.1016/j.nucengdes.2024.113634

Roberto E. Fairhurst-Agosta, Tomasz Kozlowski

Safety analyses in research reactors require the estimation of heat deposited in experiments and reactor structures after shutdown. The accurate assessment of the deposited energy across a reactor geometry better determines the heat removal requirements and ensures effective cooling. However, the development of experiment safety analyses on a case-by-case basis often proves to be effort and time-consuming. This article introduces a method based on the creation of a generic material irradiation database to expedite the process. The irradiation database is created by calculating the delayed heating in experiments of individual chemical elements. Then, the created database enables the quick calculation of the delayed heating of experiments of arbitrary material composition. This article showcases two applications to demonstrate the delayed heating calculation workflow and verify the generic material irradiation database method. These applications include a simple demonstration exercise and an Advanced Test Reactor experiment. The results display an overall good agreement between the generic material irradiation database method and reference values for a wide variety of experiments.

研究反应堆的安全分析需要对停堆后沉积在实验和反应堆结构中的热量进行估算。对反应堆几何结构中沉积能量的准确评估可以更好地确定散热要求，并确保有效冷却。然而，根据具体情况进行实验安全分析往往费时费力。本文介绍了一种基于创建通用材料辐照数据库的方法，以加快这一过程。辐照数据库是通过计算单个化学元素实验中的延迟加热而创建的。然后，创建的数据库可以快速计算任意材料成分实验的延迟加热。本文展示了两个应用程序，以演示延迟加热计算工作流程并验证通用材料辐照数据库方法。这些应用包括一个简单的演示练习和一个高级试验反应堆实验。结果表明，通用材料辐照数据库方法与各种实验的参考值之间总体上具有良好的一致性。

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引用次数: 0

COAL reflooding experiments during a loss of coolant Accident: Effect of the water flow rate, the pressure and the rod power with ballooned rods 冷却剂损失事故期间的 COAL 再充水实验：水流速度、压力和气球棒功率的影响

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-23 DOI: 10.1016/j.nucengdes.2024.113641

G. Repetto , Q. Grando , S. Eymery , R. Van Lochem

During a loss of coolant accident (LOCA) in a pressurized water reactor, the drying of the fuel assemblies leads to an increase in the fuel temperature and a deformation of the fuel rod cladding.

The COAL experiments focused on the coolability issue of a partially deformed fuel assembly during water injection with the safety systems using a 7x7 bundle of electrically heated rods. The relocation of the fragmented fuel in the balloons is taken into account by a local increase in power by a factor of 1.5, and the effect of the flow area restriction is provided with various flow blockage (intact geometry up to moderate and long ballooning (100 and 300 mm) with different blockage ratios (80 and 90 %)).

These experiments, in the frame of the PERFROI project, were launched by the “Institut de Radioprotection et de Sureté Nucléaire” (IRSN).

This paper presents the thermal hydraulics parameters and the main results of some experiments carried out in a facility of the STERN Laboratories. We studied the effect of the inlet water flow rate which is the consequence of the amount of water entering the reactor core after the break of the primary circuit, the effect of the pressure and the effect of the rod power as a function of the moment of availability of the safety pumps after the reactor scram. We provide experiments data on the coolability limits for different rod powers, which is given by the minimum of water flow to consider that the reflooding may be not impaired (PCT below the LOCA criterium of 1204 °C). The needed flow is ranging from 7 7 kg/s/m² (with intact rods geometry) at low power up 35 kg/s/m² (with at the high power that remaining in the core 1 min after the reactor scram) with a strong effect of the presence a partially local area due to rod ballooning during the large break LOCA accident. We outlined also the effect of the system pressure with a strong effect on the reflooding process above 10 bar up to 30 (for medium break LOCA).

These results are used to improve and validate the heat exchange models of thermal hydraulics codes dealing with the complex reflooding processes in such a configuration.

在压水反应堆发生冷却剂损失事故（LOCA）期间，燃料组件的干燥会导致燃料温度升高和燃料棒包壳变形。COAL 实验的重点是在使用 7x7 电加热棒束的安全系统注水期间，部分变形燃料组件的可冷却性问题。通过局部功率增加 1.5 倍来考虑气球中碎裂燃料的重新定位，并通过各种流动阻塞（从完整几何形状到中等和较长的气球（100 和 300 毫米），以及不同的阻塞率（80% 和 90%））来提供流动区域限制的影响。这些实验是在 PERFROI 项目框架内进行的，由 "辐射防护与核安全研究所"（IRSN）发起。本文介绍了热水力学参数以及在 STERN 实验室设施中进行的一些实验的主要结果。我们研究了进水流量的影响（进水流量是一回路断开后进入堆芯的水量的结果）、压力的影响和棒功率的影响（棒功率是反应堆熄火后安全泵可用时刻的函数）。我们提供了不同棒功率下冷却极限的实验数据，即考虑到不影响再充水的最小水流量（PCT 低于 LOCA 标准 1204 °C）。所需的水流量从低功率时的 7.7 千克/平方米（棒材几何形状完好）到高功率时的 35 千克/平方米（反应堆熄火 1 分钟后留在堆芯中的水流量）不等，在大断裂 LOCA 事故中，棒材气球对局部区域的影响很大。我们还概述了系统压力的影响，该压力对高于 10 巴至 30 巴（中等断裂 LOCA）的再充水过程有很大影响。这些结果可用于改进和验证热工水力学代码的热交换模型，以处理这种配置下的复杂再充水过程。

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引用次数: 0

Development of the ENIGMA fuel performance code for LWR applications with chromium-coated cladding 为使用铬涂层包壳的轻水反应堆应用开发 ENIGMA 燃料性能代码

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-22 DOI: 10.1016/j.nucengdes.2024.113656

Glyn Rossiter , Kerr Fitzgerald , Aiden Peakman

Zirconium-alloy cladding with a chromium coating is the most advanced of the near-term concepts amongst Accident Tolerant Fuel (ATF) materials for Light Water Reactor (LWR) applications. The ENIGMA fuel performance code has been updated to model the thermo-mechanical behaviour of such cladding in both normal and off-normal operating conditions. The focus was on accurately simulating the behaviour in Loss Of Coolant Accident (LOCA) conditions to evaluate the increase in coping time during design-basis accidents. New low-temperature and high-temperature models were incorporated for cladding oxidation and hydriding and cladding creep which take into account the impact of the chromium coating on the overall cladding behaviour. Furthermore, the consumption of the chromium coating due to high-temperature diffusion of chromium into the cladding base alloy’s β-Zr phase is simulated. The new models have been validated using measurements on chromium-coated cladding from irradiated rods, high-temperature annealing experiments and semi-integral LOCA tests. The validation showed good agreement between ENIGMA’s predictions and the experimental data; thereby demonstrating the applicability of the new models for simulating the performance of LWR fuel rods with chromium-coated cladding in both normal operation and accident conditions.

带铬涂层的锆合金包壳是轻水反应堆（LWR）近期应用的事故耐受燃料（ATF）材料中最先进的概念。ENIGMA 燃料性能代码已经更新，以模拟这种包层在正常和非正常运行条件下的热机械性能。重点是精确模拟失去冷却剂事故（LOCA）条件下的行为，以评估在设计基准事故期间应对时间的增加。针对堆芯氧化和水化以及堆芯蠕变，加入了新的低温和高温模型，其中考虑到了铬涂层对堆芯整体行为的影响。此外，还模拟了由于铬向包层基合金的 β-Zr 相高温扩散而导致的铬涂层消耗。通过对辐照棒铬涂层包层的测量、高温退火实验和半整合 LOCA 试验，对新模型进行了验证。验证结果表明，ENIGMA 的预测结果与实验数据非常吻合，从而证明了新模型适用于模拟带铬涂层的低温反应堆燃料棒在正常运行和事故条件下的性能。

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引用次数: 0

Preliminary analysis of the in-orbit operation characteristic of the inherent safety space reactor power system 固有安全空间反应堆动力系统在轨运行特性初步分析

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2024-10-22 DOI: 10.1016/j.nucengdes.2024.113652

Li Huaqi, Tian Xiaoyan, Wei Mingyan, Zhu Lei, Shi Leitai, Chen Sen, Luo Xiaofei, Li Da, Chen Lixin, Jiang Xinbiao

The inherent safety space reactor power system with a coupled thermoelectric conversion in a liquid metal lithium-cooled reactor represents a highly reliable space power. Compared with the ground reactor, the SNRPS has its own characteristics in safety considerations, mainly manifested in the SNRPS before launch, during launch and during the ascent into orbit will be affected by the launch vehicle. Which can be analyzed by the common methodology of probabilistic risk management. To investigate the response characteristics during in-orbit operation accidents, a transient analysis model of the liquid metal-cooled space reactor power system is established. The system response characteristics of the inherent safety space reactor power system conceptual designs are preliminarily analyzed under four potential typical in-orbit operating conditions, including (1) rated operating condition, (2) control drum misoperation events, (3) partial loss of coolant flow accident, and (4) partial failure of the radiator area accident. The results show that the power system has inherent safety in-orbit operation characteristics due to the system design operating parameters, which the coolant temperature below 1200 K at the rated operating condition. Even under typical operating accidents, the system coolant remains highly supercooled (more than 200 K), preventing boiling from occurring. The maximum temperature of the core fuel pin and cladding materials remains lower than their safety limits, ensuring that no core melting phenomenon occurs.

固有安全空间反应堆动力系统在液态金属锂冷却反应堆中进行耦合热电转换，是一种高度可靠的空间动力。与地面反应堆相比，空间反应堆电源系统在安全考虑方面有其自身的特点，主要表现在空间反应堆电源系统在发射前、发射过程中和上升入轨过程中都会受到运载火箭的影响。这可以用概率风险管理的常用方法进行分析。为了研究在轨运行事故的响应特性，建立了液态金属冷却空间反应堆动力系统的瞬态分析模型。初步分析了固有安全空间堆动力系统概念设计在四种潜在的典型在轨运行条件下的系统响应特性，包括：（1）额定运行条件；（2）控制鼓误操作事件；（3）冷却剂流部分损失事故；以及（4）散热器区域部分失效事故。结果表明，由于系统设计运行参数的原因，动力系统具有固有的安全在轨运行特性，即在额定运行状态下冷却剂温度低于 1200 K。即使在典型的运行事故情况下，系统冷却剂仍保持高度过冷（超过 200 K），防止发生沸腾。堆芯燃料销和包层材料的最高温度仍低于其安全极限，确保不会发生堆芯熔化现象。

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引用次数: 0