Annals of Nuclear Energy最新文献

Review on development status of comprehensive prevention and control technology for disaster-causing floating bodies at water intake of nuclear power plants 核电站进水口致灾浮体综合防治技术发展现状综述

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

Annals of Nuclear Energy

Pub Date : 2024-12-02 DOI: 10.1016/j.anucene.2024.111087

Yingying Zheng , Rongyong Zhang , Yun Long , Xinshu Jiang , Rongsheng Zhu , Ji Xing

Nuclear energy, classified as a clean energy source, enjoys extensive application in numerous countries worldwide. In recent years, the growth rate of marine organisms has accelerated, resulting in the formation of large aggregations of marine organisms and other large congregations that have the potential to obstruct the cold source water intake system of nuclear power plants. The safety situation about the cold source is of significant concern. This paper presents a summary of water intake safety events at nuclear power plants around the world in recent years. It also analyzes the current status of comprehensive prevention and control technology for disaster-causing floating bodies at nuclear power plants’ water intakes. The analysis is based on the latest developments in global nuclear power technology. This study provides a comprehensive overview of the advancements in fine offshore monitoring and early warning systems, passive interception techniques for open channel dredging, the evolution of active defense measures, the development of composite water intake open channel structures, bubble curtain interception methods, cutting and grinding techniques, and the subsequent transportation of marine organisms. Nevertheless, shortcomings remain in the water intake safety of nuclear power plants. These include an inadequate design basis for water intake open channels and sewage interception networks, as well as deficiencies in marine organism early warning and monitoring equipment and layout. Furthermore, the issue of sea ice represents a novel challenge confronting the nuclear power industry in coastal regions situated at high latitudes. Therefore, it is essential to undertake research and implement improvements to the comprehensive technology used to prevent and control water intake safety in nuclear power plants. It is imperative to provide robust backing for the long-term advancement of nuclear power technology.

核能作为一种清洁能源，在世界许多国家得到了广泛的应用。近年来，海洋生物的生长速度加快，形成了大型海洋生物群落和其他大型群落，有可能阻碍核电站冷源取水系统。冷源的安全状况是一个值得关注的问题。本文对近年来发生在世界各地的核电站取水安全事件进行了总结。分析了核电站进水口致灾浮体综合防治技术的现状。该分析基于全球核电技术的最新发展。本研究全面概述了精细近海监测和预警系统的进展、明渠疏浚的被动拦截技术、主动防御措施的演变、复合进水口明渠结构的发展、气泡幕拦截方法、切割和研磨技术以及随后的海洋生物运输。然而，核电站的进水安全仍存在不足。这些问题包括取水、明渠和污水截流网的设计基础不足，以及海洋生物早期预警和监测设备和布局的不足。此外，海冰问题是高纬度沿海地区核动力工业面临的一个新挑战。因此，有必要对核电站进水安全综合防治技术进行研究和改进。为核电技术的长期发展提供强有力的支持是当务之急。

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引用次数: 0

Validating automated resonance evaluation with synthetic data 用合成数据验证自动共振评价

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

Annals of Nuclear Energy

Pub Date : 2024-12-01 DOI: 10.1016/j.anucene.2024.111081

Oleksii Zivenko , Noah A.W. Walton , William Fritsch , Jacob Forbes , Amanda M. Lewis , Aaron Clark , Jesse M. Brown , Vladimir Sobes

The integrity and precision of nuclear data are crucial for a broad spectrum of applications, from national security and nuclear reactor design to medical diagnostics, where the associated uncertainties can significantly impact outcomes. A substantial portion of uncertainty in nuclear data originates from the subjective biases in the evaluation process, a crucial phase in the nuclear data production pipeline. Recent advancements indicate that automation of certain routines can mitigate these biases, thereby standardizing the evaluation process and enhancing reproducibility. This research aims to provide a methodology, framework, and metrics for the validation of automated nuclear data evaluation software leveraging high-quality synthetic data that closely mimic real experimental observables. An introduced error metric provides a scale and intuitive measure of the evaluation quality by quantifying the estimate’s accuracy and performance across the specified energy range. Synthetic data provides access to experimental observables and underlying resonance parameters, enabling comparison of different evaluations. The methodology is demonstrated using Ta-181 isotope data in the resolved resonance region. The Automated Resonance Identification Subroutine (ARIS), which operates without prior resonance information, was used to test and showcase the framework’s capabilities utilizing the proposed error metrics. The results demonstrate the effectiveness of the proposed approach and framework for optimizing software parameters and testing hypotheses through “what-if” controlled experiments, such as modifying assumptions about experimental conditions or average resonance parameters.

核数据的完整性和准确性对于从国家安全和核反应堆设计到医疗诊断等广泛应用至关重要，相关的不确定性可能对结果产生重大影响。核数据的很大一部分不确定性源于评估过程中的主观偏差，这是核数据生产管道中的关键阶段。最近的进展表明，某些例程的自动化可以减轻这些偏差，从而使评估过程标准化并提高可重复性。本研究旨在为自动化核数据评估软件的验证提供一种方法、框架和指标，利用高质量的合成数据来模拟真实的实验观测结果。引入的误差度量通过在指定能量范围内量化估计的准确性和性能，为评估质量提供了一个尺度和直观的度量。合成数据提供了对实验观测值和潜在共振参数的访问，从而可以对不同的评估进行比较。利用Ta-181同位素数据对该方法进行了验证。自动共振识别子程序（ARIS）在没有事先共振信息的情况下运行，利用所提出的误差度量来测试和展示框架的功能。结果表明，所提出的方法和框架在优化软件参数和通过“假设”控制实验（如修改关于实验条件或平均共振参数的假设）测试假设方面是有效的。

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引用次数: 0

Investigation on heat split in a vertical 3 × 3 dual-cooled annular rod bundle 垂直3 × 3双冷环形棒束热分裂研究

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

Annals of Nuclear Energy

Pub Date : 2024-11-30 DOI: 10.1016/j.anucene.2024.111082

Miao Gui , Jingxin Wang , Shan Zhou , Yu Liu , Jianqiang Shan , Yu Liang , Pan Wu

Single-phase heat transfer experiment in a 3 × 3 dual-cooled annular rod bundle was performed to study the heat split behavior and evaluate the applicability of heat transfer correlations to annular fuel. The parametric effects on heat split were investigated. The results showed that the heat split obviously increased with the flow split, and slightly increased with increasing inlet temperature and heating power. The thermal-conduction resistances of gaps played a leading role in heat split mechanism and the convective heat transfer was the secondary factor affecting the heat split. The applicability of heat transfer correlations was assessed by the experimental data. The predictive performance was significantly improved when replacing the hydraulic equivalent diameter by the heated equivalent diameter in the Nusselt number calculation. The DB and El-Genk correlations had the best predictive effect on the heat transfer in the external channel, with the MAE of 0.082 and 0.094 respectively.

在3 × 3双冷环形棒束中进行了单相传热实验，研究了传热行为，并评价了传热关系式对环形燃料的适用性。研究了参数对热分裂的影响。结果表明：随着流量分流的增大，热分流明显增大，随着进口温度和加热功率的增大，热分流略有增大；间隙的热传导阻力在热分裂机制中起主导作用，对流换热是影响热分裂的次要因素。用实验数据对传热关系式的适用性进行了评价。在努塞尔数计算中，用热当量直径代替水力当量直径，预测性能明显提高。DB和El-Genk相关性对外通道换热的预测效果最好，MAE分别为0.082和0.094。

引用次数: 0

Impact of temperature- and phase-dependent zirconium hydride phonons on criticality 温度和相位依赖的氢化锆声子对临界的影响

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

Annals of Nuclear Energy

Pub Date : 2024-11-28 DOI: 10.1016/j.anucene.2024.111034

Amelia Trainer , Benoit Forget , Jesse Holmes , Jonathan Wormald , Michael Zerkle

Zirconium hydride is a solid moderator material that has a high (albeit variable) hydrogen content. Proper modeling of the thermal neutron population in zirconium hydride requires that the thermal neutron scattering law be used, which incorporates the material structure into the cross section. In the incoherent approximation, this scattering law may be calculated using a material’s phonon distribution and other material parameters. Historically, the partial phonon distributions for zirconium hydride (ZrH

_{x}

) have often been assumed to be independent of hydrogen content, crystalline phase, and temperature. The present work aims to relax this approximation by generating phonon distributions using molecular dynamics while varying (1) the hydrogen content, (2) crystalline phase, and (3) material temperature. Through this work, hydrogen content and crystalline phase have shown to have mild impacts on the vibrational structure, while temperature appears to hold strong influence over the vibrational properties of the materials considered. Temperature-dependent H(ZrH

_{x}

) phonon distributions for

δ

-ZrH_1.67,

ϵ

-ZrH_1.82, and

ϵ

-ZrH₂ were tested on models of a TRIGA reactor and a SNAP reactor, showing how material-specific vibrational models can noticeably influence

k_{eff}

, flux distributions, and reactivity coefficients.

氢化锆是一种固体慢化剂材料，氢含量高（尽管是可变的）。对氢化锆中热中子居数的正确建模需要使用热中子散射定律，该定律将材料结构纳入截面。在非相干近似中，这种散射定律可以用材料的声子分布和其他材料参数来计算。历史上，氢化锆（ZrHx）的部分声子分布通常被认为与氢含量、晶相和温度无关。目前的工作旨在通过在改变(1)氢含量，(2)晶相和(3)材料温度的情况下使用分子动力学产生声子分布来放松这种近似。通过这项工作，氢含量和晶相对振动结构有轻微的影响，而温度似乎对所考虑的材料的振动特性有很强的影响。δ-ZrH1.67、ϵ-ZrH1.82和ϵ-ZrH2的温度依赖H（ZrHx）声子分布在TRIGA反应堆和SNAP反应堆模型上进行了测试，显示了特定材料的振动模型如何显著影响keff、通量分布和反应性系数。

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引用次数: 0

Application of data partitioned Kriging algorithm with GPU acceleration in real-time and refined reconstruction of three-dimensional radiation fields 在实时和精细重建三维辐射场中应用带 GPU 加速的数据分区克里金算法

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

Annals of Nuclear Energy

Pub Date : 2024-11-27 DOI: 10.1016/j.anucene.2024.111047

Ningbiao Xiao, Jinsen Guo, Zijia Kuang, Wei Wang

Accurately assessing personal radiation doses in real radiation environments like nuclear power plants requires precise and real-time reconstruction of three-dimensional radiation fields. The Kriging algorithm, known for its accuracy in spatial interpolation, provides a promising approach for this task. However, its computational demands can be significant, especially in real-time scenarios. To address this, we enhance the Kriging algorithm with GPU acceleration and data partitioning strategies, enabling efficient and accurate reconstruction of three-dimensional nuclear radiation fields. Using Fluka software for Monte Carlo simulations, we generated a virtual radiation field of dimensions 5 m

\times

5 m

\times

5 m for a single-source, unshielded scenario, and a field of dimensions 20 m

\times

6 m

\times

8 m for a multi-source, shielded scenario. Using the simulated data, we compared the prediction accuracy of the improved algorithm with the conventional Kriging algorithm and further explored factors influencing the acceleration ratio of the improved algorithm. The results indicate that the GPU-accelerated and data-partitioned Kriging algorithm achieves nearly identical accuracy compared to the traditional method. In the single-source, unshielded scenario, with more than 343 known (measurement) points and predicting

95 \times 95 \times 95 =

857,375 points, the prediction accuracy remains above 92.25%. In the multi-source, shielded scenario, with more than 8000 known (measurement) points and predicting

95 \times 95 \times 95 =

857,375 points, the prediction accuracy remains above 91.17%. The acceleration performance of the improved algorithm is consistent across both scenarios, with the acceleration ratio increasing as the number of known and predicted points grows, reaching approximately 20 for smaller datasets and up to 93 for larger datasets. Additionally, the acceleration effect of the improved algorithm varies with data partition size, initially increasing and then decreasing as the partition size increases. When the number of known points is 512 and the number of predicted points is 884,736, the optimal partition size lies between 80,000 and 90,000, resulting in a prediction time of only 0.24 s.

在核电厂等真实辐射环境中准确评估个人辐射剂量需要精确、实时地重建三维辐射场。克里金算法以其空间插值的精确性而著称，为这项任务提供了一种很有前途的方法。然而，它的计算要求可能很高，尤其是在实时场景中。为了解决这个问题，我们利用 GPU 加速和数据分区策略增强了克里金算法，从而实现了高效、精确的三维核辐射场重建。我们使用 Fluka 软件进行蒙特卡罗模拟，为单源、无屏蔽场景生成了尺寸为 5 m × 5 m × 5 m 的虚拟辐射场，为多源、有屏蔽场景生成了尺寸为 20 m × 6 m × 8 m 的虚拟辐射场。利用模拟数据，我们比较了改进算法与传统克里金算法的预测精度，并进一步探讨了影响改进算法加速比的因素。结果表明，与传统方法相比，经过 GPU 加速和数据分区的克里金算法达到了几乎相同的精度。在单源、无屏蔽情况下，已知（测量）点超过 343 个，预测 95×95×95= 857,375 个点，预测精度保持在 92.25% 以上。在多源、屏蔽情况下，已知（测量）点超过 8000 个，预测 95×95×95= 857375 个点，预测准确率保持在 91.17% 以上。改进算法在两种场景下的加速性能是一致的，加速比随着已知点和预测点数量的增加而增加，较小数据集的加速比约为 20，较大数据集的加速比可达 93。此外，改进算法的加速效果随数据分区大小的变化而变化，最初随着分区大小的增加而增加，然后随着分区大小的增加而减少。当已知点数为 512 个，预测点数为 884 736 个时，最佳分区大小介于 80 000 和 90 000 之间，预测时间仅为 0.24 秒。

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引用次数: 0

Evaluation of uranium-233 neutron capture cross section in keV region 评估铀-233 在 keV 区域的中子俘获截面

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

Annals of Nuclear Energy

Pub Date : 2024-11-27 DOI: 10.1016/j.anucene.2024.110977

Naohiko Otuka , Kenichi Tada , Oscar Cabellos , Osamu Iwamoto

The uranium-233 neutron capture cross section between 3 keV and 1 MeV was evaluated with the alpha value recently measured at the Los Alamos National Laboratory LANCE facility and compiled in the EXFOR library. The obtained capture cross section is systematically lower than those in the latest versions of the major general purpose nuclear data libraries, and the reduction from the JENDL-5 library is close to 50% around 20 keV. The newly evaluated cross section was benchmarked against 166 criticality experiments chosen from the ICSBEP handbook by performing Monte Carlo neutron transport calculation with the JENDL-5 library, and slight reduction of the cumulative chi-square value was achieved by adoption of the newly evaluated capture cross section.

利用最近在洛斯阿拉莫斯国家实验室 LANCE 设施测量并编入 EXFOR 数据库的阿尔法值，对 3 keV 至 1 MeV 之间的铀-233 中子俘获截面进行了评估。所获得的俘获截面系统性地低于主要通用核资料库的最新版本，而在 20 keV 附近，JENDL-5 资料库的降低幅度接近 50%。通过使用 JENDL-5 库进行蒙特卡洛中子输运计算，以从 ICSBEP 手册中选取的 166 个临界实验为基准，对新评估的截面进行了比对。

引用次数: 0

Boiling critical characteristics in narrow rectangular channel under local heat flux concentration conditions 局部热流集中条件下窄矩形水道的沸腾临界特性

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

Annals of Nuclear Energy

Pub Date : 2024-11-26 DOI: 10.1016/j.anucene.2024.111077

Kepiao Li , Kui Zhang , Tianyi Qi , Ronghua Chen , Wenxi Tian , Suizheng Qiu

For the establishment of predicting method of local concentrated critical heat flux (CHF) in narrow rectangular channels, and also technical support for the preparation of long-life and high-performance dispersive plate fuel, this work simulates Dry-out type CHF under local concentration of heat flux condition by using Eulerian multiphase model and Enhanced near-wall treatment. The distribution of near-wall temperature and void fraction under different heat flux was calculated by Computational Fluid Dynamics (CFD) method. Dry-out CHF was predicted based on highly and quickly wall temperature rise. The influence of system pressure, mass flow rate and inlet subcooling degree were discussed on Dry-out CHF. Compared with uniform heating, the predicted average heat flux on CHF is lower under local heat flux concentration, and location of CHF is related to the position of heat flux concentration area. This work’s research method in Ansys Fluent could be applied to forecast Dry-out CHF in narrow rectangular channel under local concentration of heat flux condition.

为建立窄矩形通道局部集中临界热通量（CHF）的预测方法，同时为制备长寿命、高性能的分散板状燃料提供技术支持，本研究采用欧拉多相模型和增强近壁处理，模拟了局部集中热通量条件下的干馏型CHF。通过计算流体动力学（CFD）方法计算了不同热通量下的近壁温度和空隙率分布。根据高度和快速的壁面温升预测了干化 CHF。讨论了系统压力、质量流量和入口过冷度对干化 CHF 的影响。与均匀加热相比，在局部热通量集中的情况下，CHF 的预测平均热通量较低，且 CHF 的位置与热通量集中区域的位置有关。这项工作在 Ansys Fluent 中的研究方法可用于预测局部热通量集中条件下窄矩形通道中的干化 CHF。

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引用次数: 0

SCALE inventory and reactivity analysis as part of the Hermes 2021 PSAR review 作为赫尔墨斯 2021 PSAR 审查一部分的 SCALE 清单和反应性分析

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

Annals of Nuclear Energy

Pub Date : 2024-11-25 DOI: 10.1016/j.anucene.2024.111063

Friederike Bostelmann , Benjamin R. Betzler , Donny Hartanto , William A. Wieselquist

The readiness of SCALE for comprehensive studies of pebble-bed reactors has been demonstrated through detailed analysis of a fluoride salt–cooled, high-temperature pebble-bed reactor (PB-FHR). The methods developed for pebble-bed reactor modeling in SCALE, particularly for inventory generation, have proven effective in gaining insights into the reactor physics of this advanced reactor. Excellent agreement with another code package has been observed, further highlighting SCALE’s strong performance.

The SCALE results supported the US Nuclear Regulatory Commission’s construction permit application review of the Hermes low-power PB-FHR demonstration reactor. A SCALE model of the Hermes reactor was developed at Oak Ridge National Laboratory using information from the Preliminary Safety Analysis Report (PSAR) and supplemented with publicly available data. SCALE reactivity coefficient simulations reproduced PSAR results within 1

σ

statistical uncertainties. Sensitivity studies emphasized the importance of graphite specifications for accurate k_eff predictions.

通过对氟化盐冷却高温鹅卵石床反应器（PB-FHR）的详细分析，证明了 SCALE 可用于鹅卵石床反应器的综合研究。事实证明，在 SCALE 中开发的鹅卵石床反应器建模方法，特别是生成清单的方法，可以有效地深入了解这种先进反应器的反应器物理特性。SCALE 的结果为美国核管理委员会审查赫尔墨斯低功率 PB-FHR 示范反应堆的施工许可申请提供了支持。橡树岭国家实验室利用初步安全分析报告（PSAR）中的信息，并辅以公开数据，开发出了赫尔墨斯反应堆的 SCALE 模型。SCALE 反应系数模拟再现了 PSAR 的结果，统计不确定性在 1σ 以内。敏感性研究强调了石墨规格对准确预测 keff 的重要性。

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引用次数: 0

Numerical simulation of impurity particles migration and deposition within the 5 × 5 fuel assembly under nucleate boiling conditions 核沸腾条件下 5×5 燃料组件内杂质颗粒迁移和沉积的数值模拟

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

Annals of Nuclear Energy

Pub Date : 2024-11-25 DOI: 10.1016/j.anucene.2024.111076

Hongkang Tian , Tenglong Cong , Maolong Liu , Mengke Cai , Zijian Huang , Hanyang Gu

Deposition of corrosion products on the cladding surface can cause many problems in PWRs, such as CRUD-induced power shift, localized corrosion, and increased radioactive risk. To gain a deeper understanding of fouling behavior within the fuel assembly, the migration and deposition of impurity particles were investigated with CFD method, where the Eulerian two-fluid model, species transport model, and fouling deposition model are considered. The relationship between fouling deposition rate and thermal–hydraulic parameters was investigated, and the distribution of fouling thickness along both the axial and circumferential directions was analyzed. The results show that the maximum fouling thickness reaches 44 µm in a refueling cycle. Compared with the inner rods, the corner rods have a higher fouling thickness in single-phase region and a lower fouling thickness in nucleate boiling region. The results obtained in this study could provide a reference for the investigation of deposition behavior within the reactor core.

腐蚀产物沉积在包壳表面会给压水堆带来许多问题，例如 CRUD 引起的功率偏移、局部腐蚀和放射性风险增加。为了深入了解燃料组件内的结垢行为，采用 CFD 方法研究了杂质颗粒的迁移和沉积，其中考虑了欧拉双流体模型、物种迁移模型和结垢沉积模型。研究了污垢沉积速率与热液压参数之间的关系，并分析了污垢厚度沿轴向和圆周方向的分布。结果表明，在一个加油周期内，最大污垢厚度达到 44 µm。与内杆相比，角杆在单相区的污垢厚度较高，而在成核沸腾区的污垢厚度较低。本研究获得的结果可为研究反应堆堆芯内的沉积行为提供参考。

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引用次数: 0

Investigation on thermal response of high temperature heat pipe under thermal mismatch conditions 热失配条件下高温热管的热响应研究

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

Annals of Nuclear Energy

Pub Date : 2024-11-25 DOI: 10.1016/j.anucene.2024.111075

Zhang Jiarui , Guo Kailun , Liu Lining , Xue Meng , Wang Chenglong , Tian Wenxi , Qiu Suizheng , Su Guanghui , Chen Chong

As an efficient and reliable passive heat transfer equipment, high temperature heat pipes (HTHPs) plays an important role in HTHP solid state reactor system. HTHPs often encounter thermal mismatch in reactor applications, which will lead to capillary action, entrainment, and other characteristics, resulting in HTHP failure due to heat transfer limits. In this paper, the thermal response test of liquid metal HTHP under thermal mismatch conditions is studied, and the simulation test platform of HTHP mismatch is designed and built to measure the important physical parameters. The test results show that the heat transfer performance of the HTHP is the best under the condition of horizontal inclination of 60°, and the condenser power is 5.66 kW. This condition is selected as the reference condition for the transient condition test. When the HTHP experienced a transient power increase, the middle of the evaporator section, as the part with the largest heat flux of heating power, appeared a severe overheating phenomenon. The wall temperature of the evaporator section soared from 750 ℃ to 1050 ℃ in a short time, and the heating rate reached 18.9 ℃/min, which affected the smooth operation of the HTHP. When the HTHP encounters an inclination angle transient increase condition, the entrainment phenomenon is more severe, the temperature fluctuation of the condenser section is intensified, and the heat transfer of the HTHP is in an unstable state. When the HTHP encounters an inclination angle transient decrease condition, the temperature rise of the evaporator section is less than 10 ℃, the temperature at the end of the condenser section is increased, and the start-up performance of the HTHP is improved. In summary, this paper obtains the thermal response rule of HTHPs under thermal mismatch conditions through experimental research, which provides support for the safe application of HTHPs for special equipment.

高温热管（HTHP）作为一种高效可靠的无源传热设备，在高温固态反应器系统中发挥着重要作用。高温热管在反应器应用中经常会遇到热失配，从而导致毛细作用、夹带等特性，导致高温热管因传热受限而失效。本文研究了热失配条件下液态金属 HTHP 的热响应测试，设计并搭建了 HTHP 失配模拟测试平台，测量了重要的物理参数。试验结果表明，在水平倾角为 60° 的条件下，HTHP 的传热性能最好，冷凝器功率为 5.66 kW。瞬态工况试验选择该工况作为参考工况。当 HTHP 出现瞬态功率增加时，蒸发器中部作为加热功率热通量最大的部分，出现了严重的过热现象。蒸发器部分的壁温在短时间内从 750 ℃飙升至 1050 ℃，加热速率达到 18.9 ℃/min，影响了 HTHP 的平稳运行。当 HTHP 遇到倾角瞬时增大工况时，夹带现象更加严重，冷凝器段的温度波动加剧，HTHP 的传热处于不稳定状态。当 HTHP 遇到倾角瞬时减小工况时，蒸发器段温升小于 10 ℃，冷凝器段末端温度升高，HTHP 的启动性能得到改善。综上所述，本文通过实验研究得到了热失配条件下 HTHPs 的热响应规律，为 HTHPs 在特种设备中的安全应用提供了支持。

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引用次数: 0