Annals of Nuclear Energy最新文献_第10页

Analysis of swarm flow and bubble residence time under pool scrubbing conditions 水池洗涤条件下的群流和气泡停留时间分析

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

Annals of Nuclear Energy

Pub Date : 2024-10-07 DOI: 10.1016/j.anucene.2024.110956

Fanli Kong, Xu Cheng

In severe accidents of nuclear power plants, large amounts of fission products existing as radioactive aerosols are released. Pool scrubbing plays an important role in the removal of radioactive aerosols. Bubble residence time is one of the key parameters to determine the efficiency of aerosol removal, especially in the swarm flow region which makes a very important contribution to the total aerosol removal. In this study, the Euler-Euler-Lagrangian approach is built to track the evolution of bubble motion and to determine the bubble residence time in the liquid pool. Specifically, the Euler-Euler two-fluid approach is utilized to resolve the flow field of gas and liquid phases, while the Lagrangian approach is employed to track the discrete bubbles and to obtain the bubble residence time. The results reveal that the present approach is feasible to predict the bubble dynamics and residence time in the liquid pool. Bubble residence time is dependent on the initial position, where bubbles deviating from the central region could remain inside the liquid pool for a longer physical time. The bubble diameter, volume flow rate and submergence height are key parameters affecting the bubble residence time. And comparison between the simulated bubble residence time and the model-predicted results is carried out, indicating the discrepancy of simulated residence time and limitations of the existing model at high volume flow rate and high submergence.

在核电站的严重事故中，大量裂变产物以放射性气溶胶的形式释放出来。水池洗涤在清除放射性气溶胶方面发挥着重要作用。气泡停留时间是决定气溶胶去除效率的关键参数之一，尤其是在群流区域，它对气溶胶的总去除率有非常重要的贡献。本研究采用欧拉-欧拉-拉格朗日方法来跟踪气泡运动的演变，并确定气泡在液池中的停留时间。具体来说，欧拉-欧拉双流体方法用于解析气相和液相流场，而拉格朗日方法用于跟踪离散气泡并获得气泡停留时间。结果表明，本方法可用于预测气泡动力学和在液池中的停留时间。气泡停留时间取决于初始位置，偏离中心区域的气泡在液体池中停留的物理时间更长。气泡直径、体积流量和浸没高度是影响气泡停留时间的关键参数。模拟气泡停留时间与模型预测结果进行了比较，表明在高体积流量和高浸没度条件下，模拟停留时间与现有模型的局限性存在差异。

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引用次数: 0

Alternative core configurations analysis to improve the neutronics performance of modular gas cooled fast reactor 提高模块式气冷快堆中子性能的备选堆芯配置分析

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

Annals of Nuclear Energy

Pub Date : 2024-10-07 DOI: 10.1016/j.anucene.2024.110951

Shohanul Islam, Md Tanvir Ahmed

This study investigates the neutronics characterization of the Allegro-75 MW_th modular reactor by analyzing three alternative heterogeneous core configurations-axial, radial, axial + radial across three fuel candidates-UPuC, UPuN, and UPuO along with reflector materials namely ZrC, SiC, BeO, and Zr₃Sc₂ to improve neutronics performance, identify the most suitable core configuration and optimal axial reflector thickness. The study revealed that axial + radial heterogeneous core configuration exhibited better performance across each fuel type compared to other heterogeneous models. UPuC with axial + radial heterogeneity was identified as the optimal model as it demonstrated cycle length over ten years, satisfactory neutron spectrum, uniform neutron flux distribution, low radial and axial PPF, high beta effective, and negative Doppler Constant. Analyzing reflector materials with the most suitable fuel model revealed that the optimum axial reflector thickness is 60 cm for all reflector models where BeO emerged as the most favorable reflector due to its superior results in other neutronics parameters.

本研究调查了 Allegro-75 MWth 模块化反应堆的中子特性，分析了三种备选的异质堆芯构型--轴向、径向、轴向+径向三种候选燃料--UPuC、UPuN 和 UPuO 以及反射器材料（ZrC、SiC、BeO 和 Zr3Sc2），以提高中子性能，确定最合适的堆芯构型和最佳轴向反射器厚度。研究表明，与其他异质模型相比，轴向+径向异质堆芯配置在每种燃料类型中都表现出更好的性能。具有轴向+径向异质结构的 UPuC 被确定为最佳模型，因为它的循环长度超过十年，中子谱令人满意，中子通量分布均匀，径向和轴向 PPF 低，β 有效率高，多普勒常数为负。利用最合适的燃料模型对反射器材料进行分析后发现，所有反射器模型的最佳轴向反射器厚度均为 60 厘米，其中 BeO 因其在其他中子参数方面的优异结果而成为最有利的反射器。

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引用次数: 0

A study on in-situ characterization technology development for clearance verification of radioactive waste from nuclear decommissioning 核退役放射性废物清除核查原位表征技术开发研究

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

Annals of Nuclear Energy

Pub Date : 2024-10-05 DOI: 10.1016/j.anucene.2024.110945

Kyungmin Kim, Minseung Ko, Sangtae Kim, Yongsoo Kim

Although the clearance level of every radioactive nuclide was published by the IAEA to promote the recycling and reuse of decontaminated radioactive waste worldwide, technical and regulatory issues have always been raised around the application of the criteria. Therefore, several countries are developing in-situ characterization equipment or apparatus for on-site verification to check if the clearance criteria is met.

In this study authors developed a pilot radiation detection and measurement system using in-situ characterization technology to solve the issues, which consists of a 3D scanning camera system and a built-in Monte Carlo simulation program. Measurement results show that MDA (Minimum Detectable Activity) of the current design was indisputably below the clearance level and built-in Monte Carlo simulation package closely predicts the measurements results with the error of less than 5%. This implicates that it can determine with enough margin whether the radioactivity level of decontaminated metallic components meets the clearance criteria at decommissioning site or not.

Practically when we measure the radioactivity from gamma ray source mass attenuation always takes place during the photon transports through the medium. In fact, the reduction depends on the material, shapes, and radioactive sources. In this study the reduction factors were experimentally examined according to the influencing parameters and the results were saved as DCF (Density Correction Factor) in the data base. As expected, it turned out that the factor is somewhat affected by medium material and radioactive sources, but it is basically proportional to the distance of gamma ray passage.

It is expected that upgraded design with more accurate and reliable instruments can make it easier for regulators to accept the application of the in-situ characterization technology on-site.

尽管国际原子能机构（IAEA）公布了每种放射性核素的清除水平，以促进全球范围内去污染放射性废物的回收和再利用，但围绕标准的应用一直存在技术和监管问题。因此，一些国家正在开发原位表征设备或仪器，用于现场验证是否符合清除标准。在这项研究中，作者利用原位表征技术开发了一个试验性辐射探测和测量系统来解决这些问题，该系统由三维扫描摄像系统和内置蒙特卡洛模拟程序组成。测量结果表明，当前设计的 MDA（最小可探测活度）无可争议地低于许可水平，而内置的蒙特卡洛模拟软件包可密切预测测量结果，误差小于 5%。实际上，当我们测量伽马射线源的放射性时，在光子通过介质的过程中总是会发生质量衰减。事实上，衰减取决于材料、形状和放射源。在这项研究中，根据影响参数对衰减因子进行了实验研究，并将结果作为 DCF（密度校正因子）保存在数据库中。正如预期的那样，该系数在一定程度上受介质材料和放射源的影响，但基本上与伽马射线通过的距离成正比。

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引用次数: 0

Radiation shielding properties of sustainable concrete with novel plastering techniques 采用新型抹灰技术的可持续混凝土的辐射屏蔽性能

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

Annals of Nuclear Energy

Pub Date : 2024-10-05 DOI: 10.1016/j.anucene.2024.110958

Mohamed A.E.M. Ali , Mohammed A.Y. Hafez , Nabil M. Nagy , Neveen S. Abed

In concrete applications. Major/critical applications of such concrete are radiation-shielding facilities. Both steel slag and silica fume are examples of common by-product materials that can be used as a replacer of aggregates and cement. Thus, in this research work, steel slag was utilized as heavy aggregate in concrete production besides silica fume to present sustainable concrete mixtures probably with better radiation-shielding properties. Different cementitious plasters were applied on the conducted sustainable concrete mixture using different powdery materials; hematite, magnetite, barite, bentonite, and steel slag powders in addition to nano-titanium dioxide as full replacers for sand. The proposed plasters were presented to determine the optimum plaster technique in terms of static performance and attenuation capability against gamma and neutron radiations. The results exhibited that utilizing steel slag and silica fume in concrete mixtures enhanced compressive strength by up to 9.09 % compared to conventional concrete, while the addition of nano-titanium to conventional plaster led to superior enhancement in the compressive strength by up to 38.65 % relative to traditional plaster. Conversely, fully replacing conventional silica sand with the abovementioned powdery materials generally reduced the compressive strength of cementitious plasters by up to 30.83 %. However, the radiation shielding properties against Cs-137, and Co-60 energies have been enhanced by up to 20 % and 26 %, respectively.

在混凝土应用中。这类混凝土的主要/关键应用是辐射屏蔽设施。钢渣和硅灰都是可用作骨料和水泥替代物的常见副产品材料。因此，在这项研究工作中，除了硅灰之外，钢渣还被用作混凝土生产中的重集料，以提供可能具有更好辐射屏蔽性能的可持续混凝土混合物。在进行的可持续混凝土混合物上使用了不同的水泥基抹灰，除了纳米二氧化钛作为砂的完全替代物外，还使用了不同的粉末材料：赤铁矿、磁铁矿、重晶石、膨润土和钢渣粉。对提出的灰泥进行了介绍，以确定在静态性能和对伽马射线和中子辐射的衰减能力方面的最佳灰泥技术。结果表明，与传统混凝土相比，在混凝土混合物中使用钢渣和硅灰可提高抗压强度达 9.09%，而在传统抹灰中添加纳米钛可提高抗压强度达 38.65%。相反，用上述粉末状材料完全取代传统硅砂后，水泥基抹灰的抗压强度普遍降低了 30.83%。不过，对 Cs-137 和 Co-60 能量的辐射屏蔽性能分别提高了 20% 和 26%。

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引用次数: 0

Framework for the correct treatment of model input parameters for Bayesian updating problems in nuclear engineering 正确处理核工程中贝叶斯更新问题模型输入参数的框架

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

Annals of Nuclear Energy

Pub Date : 2024-10-05 DOI: 10.1016/j.anucene.2024.110930

Michał Jędrzejczyk , Piotr Kopka , Basma Foad

Bayesian updating (BU) tools are increasingly used in nuclear engineering for inverse uncertainty quantification (IUQ) and calibration combined with uncertainty reduction. Their goals are quantifying or reducing the uncertainty of some of the uncertain model input parameters. Since it is often the case that available experimental data only allows for updating the most influential input parameters, researchers often ignore the less important ones during BU. This paper explores the consequences of neglecting the uncertainties of uncalibrated model input parameters (UMIP). It also proposes how to include them properly and which BU algorithms are the best choices for various types of inverse problems. The analysis is based on exploring two toy problems and one in nuclear engineering concerning multiplication factor calculations. The results clearly show that the improper treatment of UMIP during BU often leads to underestimating posterior uncertainties — either of the calibrated input parameters or the simulated integral parameters, depending on how the BU was conducted. The proposed methods of correct UMIP treatment will improve the rigorousness of the BU processes and boost confidence in the resulting posterior distributions.

贝叶斯更新（BU）工具越来越多地用于核工程中的反向不确定性量化（IUQ）和校准以及不确定性降低。其目标是量化或减少某些不确定模型输入参数的不确定性。由于可用的实验数据通常只允许更新影响最大的输入参数，因此研究人员在进行 BU 时往往会忽略不太重要的参数。本文探讨了忽略未校准模型输入参数（UMIP）不确定性的后果。本文还提出了如何正确地包含这些不确定性，以及哪些 BU 算法是各类逆问题的最佳选择。分析基于对两个玩具问题和一个核工程中的乘法因子计算问题的探讨。结果清楚地表明，BU 期间对 UMIP 的不当处理往往会导致低估后验不确定性--无论是校准输入参数还是模拟积分参数的后验不确定性，这取决于 BU 是如何进行的。所提出的正确处理 UMIP 的方法将提高 BU 过程的严谨性，并增强对所得后验分布的信心。

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引用次数: 0

Californium-252 production at the High Flux Isotope Reactor − I: Validation study using campaign data 高通量同位素反应堆的锎-252 生产--I：利用运行数据进行验证研究

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

Annals of Nuclear Energy

Pub Date : 2024-10-04 DOI: 10.1016/j.anucene.2024.110960

Donny Hartanto, David Chandler, Hailey Green, Jin Whan Bae, Kevin M. Burg, Yves Robert, Carol Sizemore

This paper presents a series of ²⁵²Cf production validation and code-to-code comparison studies performed based on data from the production campaigns at the High Flux Isotope Reactor (HFIR). These studies support efforts to convert HFIR from using highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. HFIR must maintain its world-class performance and missions following this conversion, and because ²⁵²Cf is a vital neutron-emitting radioisotope used for a variety of high-impact applications (e.g., reactor startup, cancer treatment), the ability to efficiently produce ²⁵²Cf must be preserved. In this work, the HFIRCON, Shift, ORIGEN, and TCOMP codes were deployed, and several sets of data libraries were investigated to better understand the calculation codes and the data biases. As-loaded target composition data, as-run irradiation history data, and post-irradiation measurements from recent multi-cycle irradiation campaigns of the HEU core were used to validate and determine methodology biases. The findings demonstrated a good agreement, with results falling within 3 standard deviations of measurements. This paper lays the ground work for the second paper, which evaluates and compares ²⁵²Cf production and safety metrics with the HEU core and a proposed LEU core.

本文基于高通量同位素反应堆（HFIR）生产活动的数据，介绍了一系列 252Cf 生产验证和代码间比较研究。这些研究为高通量同位素堆从使用高浓铀 (HEU) 燃料转换为低浓铀 (LEU) 燃料提供了支持。HFIR 必须在转换后保持其世界一流的性能和任务，由于 252Cf 是一种重要的中子发射放射性同位素，可用于各种高影响应用（如反应堆启动、癌症治疗），因此必须保持高效生产 252Cf 的能力。在这项工作中，部署了 HFIRCON、Shift、ORIGEN 和 TCOMP 代码，并对几组数据库进行了调查，以更好地了解计算代码和数据偏差。使用加载的目标成分数据、运行时的辐照历史数据以及高浓缩铀内核最近多周期辐照活动的辐照后测量数据来验证和确定方法偏差。研究结果表明，两者的一致性很好，结果与测量值的偏差在 3 个标准偏差以内。这篇论文为第二篇论文奠定了基础，第二篇论文将评估和比较高浓缩铀堆芯和拟议的低浓缩铀堆芯的 252Cf 生产和安全指标。

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引用次数: 0

Advancing source reactor-type discrimination using machine learning techniques and SFCOMPO-2.0 experimental database 利用机器学习技术和 SFCOMPO-2.0 实验数据库推进源反应堆类型辨别工作

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

Annals of Nuclear Energy

Pub Date : 2024-10-04 DOI: 10.1016/j.anucene.2024.110952

Tianxiang Wang, Hao Yang, Shengli Chen, Cenxi Yuan

In recent years, nuclear forensic analysis has become crucial due to the growing global threat of nuclear terrorism and smuggling. Since 2005, extensive research has been conducted on identifying the origin of spent nuclear fuel, focusing on the source reactor-type discrimination, ²³⁵U enrichment of the fresh fuel, and the fuel exposure in the reactor (known as burnup). However, the majority of research relies on computed databases, which may lead to tracing discrepancies compared with actual situations. The present study employs the isotopic measurements from the experimental SFCOMPO-2.0 database to predict nuclear reactor types using Factor Analysis (FA) and various machine learning classification algorithms. The results reveal that FA is an effective method for dimension reduction and visualization. The FA-KNN, Random Forest (RF), and Multilayer Perceptron (MLP) algorithms are applied using a consistent dataset partition to ensure unbiased comparisons. The prediction results based on 10-fold stratified cross-validation are quite promising and the Receiver Operating Characteristic (ROC) curves for multi-class classification confirm the excellent generalization ability of models. Therefore, the application of machine learning techniques is highly effective for reactor-type forensics analysis, especially for RF and MLP.

近年来，由于全球核恐怖主义和核走私的威胁日益严重，核鉴识分析变得至关重要。自 2005 年以来，针对乏核燃料来源的鉴定开展了广泛的研究，重点关注源反应堆类型鉴别、新燃料的 235U 丰度以及燃料在反应堆中的暴露情况（称为燃耗）。然而，大多数研究都依赖于计算数据库，这可能会导致追踪结果与实际情况不符。本研究利用来自 SFCOMPO-2.0 实验数据库的同位素测量数据，使用因子分析（FA）和各种机器学习分类算法预测核反应堆类型。结果表明，因子分析是一种有效的降维和可视化方法。FA-KNN 算法、随机森林（RF）算法和多层感知器（MLP）算法都采用了一致的数据集分区，以确保比较无偏。基于 10 倍分层交叉验证的预测结果相当不错，多类分类的接收方操作特征曲线（ROC）证实了模型出色的泛化能力。因此，机器学习技术的应用对于反应堆类型的取证分析非常有效，特别是对于 RF 和 MLP。

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引用次数: 0

A comparative analysis of detailed and reduced CFD approaches to model wire-wrapped fuel bundles for LMFBRs applications 详细和简化 CFD 方法的比较分析，以模拟低密度纤维束燃料元件应用中的线包燃料束

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

Annals of Nuclear Energy

Pub Date : 2024-10-03 DOI: 10.1016/j.anucene.2024.110937

O. Halim , F. Galleni , N. Forgione , I. Di Piazza , A. Pucciarelli

The paper investigates the capabilities of different CFD modelling approaches in reproducing operating conditions relevant for Liquid Metal Fast Breeder Reactors technologies. The selected benchmark is the NACIE-UP facility wire-wrapped fuel bundle using Lead-Bismuth Eutectic (LBE) as coolant: the predictions are compared to the experimental data collected for several operating conditions considered in the frame of two distinct experimental campaigns. Four different modelling approaches have been adopted in this work to model the NACIE-UP Fuel Pin Simulator: Bare, Detailed, Solid-Wire and the Porous-Wire Rod Bundle model. A model-to-model comparison is performed to understand the benefits, limitations, and accuracy of using different modelling approaches for representing wrapped wires fuel bundles. Furthermore, integrating NACIE-UP benchmark experimental data into the comparative analysis reinforce the validation process of the adopted modelling approaches.

本文研究了不同 CFD 建模方法在再现液态金属快中子增殖反应堆技术相关运行条件方面的能力。选定的基准是使用铅铋共晶（LBE）作为冷却剂的 NACIE-UP 设施线包燃料束：将预测结果与在两个不同实验活动框架内考虑的几种运行条件下收集的实验数据进行了比较。这项工作采用了四种不同的建模方法对 NACIE-UP 燃料针模拟器进行建模：裸模型、详细模型、实心线模型和多孔线棒束模型。通过模型与模型之间的比较，可以了解使用不同建模方法表示包线燃料束的优点、局限性和准确性。此外，将 NACIE-UP 基准实验数据纳入比较分析，加强了所采用建模方法的验证过程。

引用次数: 0

Intergranular fracture induced by helium bubbles segregation in NiFe single-phase concentrated solid solution alloys 镍铁单相浓固溶合金中氦气泡偏析诱发的晶间断裂

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

Annals of Nuclear Energy

Pub Date : 2024-10-03 DOI: 10.1016/j.anucene.2024.110959

Jie Li , Xinhua Yang , Peng Wang , Qin Qian

Single-phase concentrated solid solution alloys (SP-CSAs) are potential nuclear structural materials with excellent irradiation resistance. Under irradiation environment, helium (He) atoms are easily segregated and stored at grain boundaries (GBs), thus causing He embrittlement of materials. The intergranular fracture behavior of NiFe SP-CSAs caused by He bubbles segregation is studied with molecular dynamics (MD) method. The effects of Fe atom concentration, GB type and He atom number in He bubbles on the intergranular fracture are analyzed. Since the disordered atoms impede crack initiation and propagation, NiFe SP-CSAs exhibit stronger fracture resistance than pure Ni. With the increase of strain, the number of disordered atoms at the GB increases further, so that the fracture strain of NiFe SP-CSAs with symmetric tilt GB decreases. With the increase of Fe atom concentration, Fe atoms inhibit the intergranular crack initiation and propagation in NiFe SP-CSAs more powerfully. In addition, the increase of the He atom number in the He bubble promotes the fracture. The formation of SP-CSAs is also beneficial to enhancing intergranular fracture resistance of alloys with twist GBs and asymmetrically tilt GBs, and the impact on the latter is more pronounced.

单相浓缩固溶体合金（SP-CSAs）是一种潜在的核结构材料，具有优异的抗辐照性能。在辐照环境下，氦（He）原子很容易偏析并储存在晶界（GBs），从而导致材料的氦脆性。本文采用分子动力学（MD）方法研究了氦气泡偏析导致的镍铁 SP-CSA 晶间断裂行为。分析了氦气泡中的铁原子浓度、GB 类型和氦原子数对晶间断裂的影响。由于无序原子阻碍了裂纹的萌发和扩展，镍铁合金 SP-CSA 比纯镍表现出更强的抗断裂性。随着应变的增加，GB 处无序原子的数量进一步增加，因此具有对称倾斜 GB 的镍铁合金 SP-CSA 的断裂应变减小。随着铁原子浓度的增加，铁原子对 NiFe SP-CSA 晶间裂纹萌发和扩展的抑制作用更强。此外，He气泡中He原子数的增加会促进断裂。SP-CSA 的形成还有利于提高具有扭曲 GB 和不对称倾斜 GB 的合金的晶间抗断裂性，而且对后者的影响更为明显。

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引用次数: 0

Fuel performance modelling of Cr-coated Zircaloy cladding under DBA/LOCA conditions DBA/LOCA 条件下铬涂层锆合金包层的燃料性能建模

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

Annals of Nuclear Energy

Pub Date : 2024-10-03 DOI: 10.1016/j.anucene.2024.110950

Pau Aragón , Francisco Feria , Luis E. Herranz , Arndt Schubert , Paul Van Uffelen

Chromium coatings are being developed for advanced technology fuel (ATF) claddings, offering negligible corrosion during normal operation, improved resistance to high-temperature steam oxidation, and superior high-temperature strength, the latter two being of utmost relevance during design basis accidents (DBAs). Demonstrating the improved response of Cr-coated Zircaloy requires the development or extension of fuel performance codes to coating simulations.

In this work, material models and correlations for Cr-coated Zircaloy cladding have been derived or obtained from the literature and implemented into TRANSURANUS and the FRAPTRAN-TUmech suite. These extended tools have been used to simulate two complementary LOCA tests: QUENCH-L1 rod 4 (out-of-pile bundle test on fresh Zircaloy cladding) and IFA-650.10 (in-pile single rod test on high-burnup Zircaloy-UO₂ fuel), enabling a gradual cross-verification of results between codes and a comparative performance analysis between coated and uncoated cladding.

The results indicate negligible impact of coating properties other than creep on the burst time. While the superior high-temperature creep resistance of coated cladding slightly delays the burst time, additional burst data would be necessary to draw sound conclusions on the balloon size. Regarding the modelling approach, treating the coated cladding as a composite material through the definition of effective properties might result in worse performance relative to uncoated cladding, contradicting experimental observations. Therefore, the separate modelling of the coating and the cladding is recommended.

目前正在开发用于先进技术燃料（ATF）包层的铬涂层，该涂层在正常运行期间的腐蚀性可以忽略不计，具有更强的抗高温蒸汽氧化能力和卓越的高温强度，后两者在设计基础事故（DBA）期间具有极大的相关性。要证明铬涂层锆合金的响应性能得到改善，需要开发或扩展燃料性能代码，以进行涂层模拟。在这项工作中，从文献中推导或获得了铬涂层锆合金包层的材料模型和相关性，并将其应用到 TRANSURANUS 和 FRAPTRAN-TUmech 套件中。这些扩展工具已用于模拟两个互补的 LOCA 试验：QUENCH-L1杆4（对新鲜锆合金包层进行的桩外束试验）和IFA-650.10（对高燃耗锆合金-UO2燃料进行的桩内单杆试验），从而能够逐步交叉验证不同代码之间的结果，并对有涂层和无涂层包层的性能进行比较分析。虽然涂层覆层的高温抗蠕变性能优越，会稍微延迟爆破时间，但要就气球大小得出合理结论，还需要更多的爆破数据。关于建模方法，通过定义有效特性将涂层覆层视为复合材料可能会导致其性能不如无涂层覆层，这与实验观察结果相矛盾。因此，建议将涂层和包层分开建模。

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引用次数: 0