Progress in Nuclear Energy最新文献_第5页

Sensitivity and uncertainty analysis of neutron activation source terms based on statistical sampling method 基于统计采样法的中子活化源项敏感性和不确定性分析

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-08 DOI: 10.1016/j.pnucene.2024.105472

Xiao Yu, Jingyu Zhang, Yixin Xiao, Wangmuhong Ni, Juanjuan Wang

The calculation of neutron activation source terms is an essential task in the radiation of nuclear reactors, and its sensitivity and uncertainty (S&U) analysis can help better identify and control the influencing factors and safety margin of activation source terms. In this paper, we studied the statistical sampling strategies for the S&U analysis of activation source terms, established the calculation process of sensitivity coefficient and uncertainty, and the S&U analysis module based on sampling methods was integrated for the proprietary developed activation computation program ABURN. Two sampling ways, Monte Carlo and Latin hypercube sampling treatments, were employed to figure out the cases of pure decay single chain, pure activation single chain, and complex reaction network. The sensitivity coefficients and uncertainty of nuclide inventory were evaluated by taking the uncertainty of decay constant or reaction cross-section into account. The simulation results indicate that after introducing the LHS sampling method, the activation calculation program ABURN can improve computational efficiency in sensitivity and uncertainty analysis. Subsequent results demonstrated that the sensitivity coefficients and uncertainty solutions based on the sampling methods in the ABURN program were in good agreement with the analytical solutions and the FISPACT program, which verified the accuracy of the approaches and program in this paper.

中子活化源项的计算是核反应堆辐射中的一项重要工作，其灵敏度和不确定性（S&U）分析有助于更好地识别和控制活化源项的影响因素和安全裕度。本文研究了活化源项 S&U 分析的统计采样策略，建立了灵敏度系数和不确定性的计算流程，并在自主开发的活化计算程序 ABURN 中集成了基于采样方法的 S&U 分析模块。采用蒙特卡洛和拉丁超立方取样处理两种取样方法，分别计算了纯衰变单链、纯活化单链和复杂反应网络的情况。考虑衰变常数或反应截面的不确定性，评估了核素清单的敏感系数和不确定性。模拟结果表明，在引入 LHS 采样方法后，活化计算程序 ABURN 可以提高灵敏度和不确定性分析的计算效率。随后的结果表明，ABURN程序中基于采样方法的灵敏度系数和不确定性解与分析解和FISPACT程序的结果吻合良好，验证了本文方法和程序的准确性。

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引用次数: 0

Gamma radiation induced tailoring the structural, optical, surface and mechanical properties of UHMWPE 伽马辐射诱导定制超高分子量聚乙烯的结构、光学、表面和机械性能

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-08 DOI: 10.1016/j.pnucene.2024.105481

Sultan J. Alsufyani , M.F. Zaki , T.S. Soliman , Nadi Mlihan Alresheedi , Tayseer I. Al-Naggar

The present investigation examined how the resultant gamma radiation affected the ultra-high molecular weight polyethylene's (UHMWPE) mechanical, optical, surface, and structural characteristics. Different gamma doses of 75, 150, 250, and 350 kGy were applied to the UHMWPE samples. The study of physical and chemical qualities has involved a variety of spectroscopy techniques, including Fourier Transform Infrared spectroscopy, X-ray diffraction, mechanical property changes, and biocompatibility properties. New bands are formed, as indicated by FT-IR analysis, and this process is linked to the oxidation of irradiation polymer chains and the production of carbon dioxide. The crystallinity of irradiated samples increases as the gamma radiation increases, according to XRD patterns. The analyzed optical results exhibit improvements in the optical characteristics of the irradiated samples. The absorbance spectra of the irradiated samples showed a shift toward the high of wavelength values in the absorption edge as compared to the pristine sample. On the other hand, the optical absorption edge has an increasing tendency as the dose of gamma-ray radiation is increased. As the gamma-ray dose increases, the absorption edge moves toward the longer wavelength. The measurements of the contact angle indicate that the surface free energy rises with increasing gamma irradiation. It was detailed how the mechanical properties of the irradiated UHMWPE samples were measured. The mechanical measurements indicate that the mechanical properties are dose-dependent to some extent. Furthermore, as gamma doses rise, so does the hardness of irradiated UHMWPE.

本研究考察了伽马辐射如何影响超高分子量聚乙烯（UHMWPE）的机械、光学、表面和结构特性。对超高分子量聚乙烯样品施加了 75、150、250 和 350 kGy 的不同伽马剂量。物理和化学特性的研究涉及多种光谱技术，包括傅立叶变换红外光谱、X 射线衍射、机械特性变化和生物相容性特性。傅立叶变换红外光谱分析显示，辐照聚合物会形成新的波段，这一过程与辐照聚合物链的氧化和二氧化碳的产生有关。根据 XRD 图谱，辐照样品的结晶度随着伽马辐射的增加而增加。光学分析结果表明，辐照样品的光学特性有所改善。与原始样品相比，辐照样品的吸收光谱显示出吸收边缘向高波长值移动。另一方面，随着伽马射线辐射剂量的增加，光吸收边沿呈上升趋势。随着伽马射线剂量的增加，吸收边缘会向长波长方向移动。接触角的测量结果表明，表面自由能随着伽马射线照射剂量的增加而上升。详细介绍了如何测量辐照超高分子量聚乙烯样品的机械性能。机械性能测量结果表明，机械性能在一定程度上与剂量有关。此外，随着伽马剂量的增加，辐照超高分子量聚乙烯的硬度也会增加。

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引用次数: 0

A subchannel analysis code for advanced liquid metal fast reactor cores and study on heat transfer characteristics of core geometry parameters 先进液态金属快堆堆芯子通道分析代码及堆芯几何参数传热特性研究

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-07 DOI: 10.1016/j.pnucene.2024.105477

Bin Han , Yuanyuan Yin , Xiaoliang Zhu , Bao-Wen Yang , Shenghui Liu , Lingping Song , Aiguo Liu , Tianyang Xing

The liquid metal fast reactor represents an important reactor type for the future development of nuclear energy. Accurately modeling and predicting the subchannel flow and heat transfer phenomenon in the rod bundles plays a key role in ensuring core safety. Consequently, a subchannel code suitable for the liquid metal fast reactor is analyzed based on the subchannel analysis code of the Pressurized Water Reactor (PWR). The modified code incorporates various types of liquid metals, such as lead, lead-bismuth eutectic (LBE), and sodium, along with their constitutive models, enhancing the applicability of the liquid metal reactor systems. This code has been verified and validated at several levels, including comparing the code simulation results with other similar subchannel codes results, high-dimensional Computational Fluid Dynamics (CFD) simulations, and experiment data. The sensitivity analysis of core geometric parameters and turbulent mixing coefficients is performed based on the modified version code. The influence of the core geometry, including the rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) and the wire wrap pitch (H) on the sensitivity of outlet temperature has been investigated. The results show that the new subchannel analysis code suitable for liquid metal cooled reactor shows reasonable agreement with the verified and validated results. The geometric parameters, such as P/D, RTW and H have noticeable effects on the outlet temperature difference of the subchannels. Additionally, the outlet temperature distribution in the subchannel is significantly affected by different turbulent mixing coefficients and the distribution of the turbulent mixing coefficient also influenced by the reactor core sizes. Overall, this study could support the rod bundles design and safety analysis in the liquid metal reactors.

液态金属快堆是未来核能发展的重要堆型。准确模拟和预测棒束中的副通道流动和传热现象对确保堆芯安全起着关键作用。因此，在压水堆子通道分析代码的基础上，分析了适合液态金属快堆的子通道代码。修改后的代码纳入了各种类型的液态金属，如铅、铅铋共晶（LBE）和钠，以及它们的构成模型，增强了液态金属反应堆系统的适用性。该代码在多个层面上进行了验证和确认，包括将代码模拟结果与其他类似子通道代码结果、高维计算流体动力学（CFD）模拟和实验数据进行比较。在修改版代码的基础上，对核心几何参数和湍流混合系数进行了敏感性分析。研究了核心几何参数，包括杆距直径比 (P/D)、杆壁间隙 (RTW) 和绕线间距 (H) 对出口温度灵敏度的影响。结果表明，适用于液态金属冷却反应堆的新子通道分析代码与经过验证和确认的结果显示出合理的一致性。P/D、RTW 和 H 等几何参数对子通道出口温差有明显影响。此外，子通道的出口温度分布受不同湍流混合系数的显著影响，而湍流混合系数的分布也受反应器堆芯尺寸的影响。总之，这项研究可为液态金属反应堆的棒束设计和安全分析提供支持。

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引用次数: 0

A review on methods and applications of artificial intelligence on Fault Detection and Diagnosis in nuclear power plants 人工智能在核电站故障检测和诊断中的方法和应用综述

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-05 DOI: 10.1016/j.pnucene.2024.105474

Aicheng Gong , Zhongjian Qiao , Xihui Li , Jiafei Lyu , Xiu Li

Nuclear power plants are critical facilities, which can generate a lot of energy through nuclear power and reduce environmental pollution. At the same time, should an accident occur at nuclear power plants, such as a nuclear leak, the consequences can be more severe than those associated with traditional power generation facilities. Therefore, Fault Detection and Diagnosis (FDD) has been an important technology in nuclear power plants. Traditional FDD methods mostly rely on the precise mathematical system model, which can be sometimes difficult to obtain in reality, and the detection accuracy of existing methods is thus limited. With the development of artificial intelligence (AI) technologies, FDD methods based on AI have been widely used. In this work, we make a systematic review of AI-based FDD methods, in conjunction with the introduction of the traditional FDD methods, and present the corresponding application scenarios of them. We hope that this work will help researchers incorporate more advanced AI models in nuclear power plants FDD and enlighten those interested in this field.

核电站是关键设施，可以通过核能产生大量能源，减少环境污染。同时，核电站一旦发生核泄漏等事故，后果可能比传统发电设施更为严重。因此，故障检测与诊断（FDD）一直是核电站的一项重要技术。传统的故障检测与诊断方法大多依赖于精确的系统数学模型，而这在现实中有时很难获得，因此现有方法的检测精度有限。随着人工智能（AI）技术的发展，基于 AI 的 FDD 方法得到了广泛应用。在这项工作中，我们结合对传统 FDD 方法的介绍，对基于人工智能的 FDD 方法进行了系统综述，并介绍了其相应的应用场景。我们希望这项工作能帮助研究人员将更先进的人工智能模型纳入核电站 FDD，并为对该领域感兴趣的人员提供启迪。

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引用次数: 0

Design of experimental facility for helium gas component testing 设计氦气组件测试实验设施

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-05 DOI: 10.1016/j.pnucene.2024.105471

Silvino A. Balderrama Prieto , Zachary D. Sellers , David Arcilesi , Piyush Sabharwall

The current state-of-the-art helium gas test loops for component testing are constrained by the maximum operating temperatures, pressures, and flow rates of such facilities. Idaho National Laboratory is designing a new experimental facility known as HECTOR (HElium Component Testing Out-of-pile Research), which is a helium gas test loop for evaluating the performance of components to helium gas pressures, temperatures, and flow rates up to 8 MPa, 800 °C, and 0.15 kg/s, respectively. The construction of this facility will enable researchers to perform experiments across a wide range of Reynolds and Nusselt numbers, which could result in the maturation of technology for high-temperature gas-cooled reactor (HTGR) components.

目前用于组件测试的最先进的氦气测试回路受到此类设施的最高工作温度、压力和流速的限制。爱达荷国家实验室正在设计一个名为 HECTOR（桩外氦气元件测试研究）的新实验设施，这是一个氦气测试环路，用于评估元件在氦气压力、温度和流速分别高达 8 兆帕、800 ℃ 和 0.15 千克/秒时的性能。该设施的建设将使研究人员能够在广泛的雷诺数和努塞尔特数范围内进行实验，从而使高温气冷堆（HTGR）组件的技术更加成熟。

引用次数: 0

The sampling and removal of tritiated water vapor in Taiwan research reactor 台湾研究反应堆中三价水蒸气的取样和去除

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-05 DOI: 10.1016/j.pnucene.2024.105482

Jhih-Jhong Huang, Peng-Yu Chen

The Taiwan Research Reactor (TRR) is a research facility in Taiwan. TRR is a heavy water reactor designed based on the National Research Experiment (NRX) with output thermal power 40 MW. On January 3, 1973, it reach criticality. TRR had been operated for 15 years and was permanently shut down in early 1988. TRR uses heavy water moderation and light water cooling. Hence, some tritium was created when deuterium captures a neutron. The inside of the Calandria has been washed with clean water after shutdown. After draining the water in the interior, it is inevitable that some residual water would remain at the bottom of the Calandria. Air sampling and analysis techniques were employed to detect and quantify tritiated water vapor. The investigation of tritiated water vapor in the TRR Calandria proved potential tritium contamination risks. Adopting the vapor condensation drying technique proved effective in removing tritiated water vapor from the TRR Calandria. After the removal of tritiated water vapor, the air in the Calandria no longer contains rich concentration of tritiated water vapor. That will help to ensure a safe environment for decommissioning operations.

台湾研究堆（TRR）是台湾的一个研究设施。台湾研究堆是根据国家研究实验（NRX）设计的重水反应堆，输出热功率为 40 兆瓦。1973 年 1 月 3 日，该反应堆达到临界状态。德黑兰研究堆运行了 15 年，于 1988 年初永久关闭。德黑兰研究堆使用重水调节和轻水冷却。因此，当氘俘获一个中子时会产生一些氚。关闭后，卡兰迪亚内部已被清水冲洗干净。在排干内部的水后，卡兰迪亚底部不可避免地会残留一些水。为检测和量化三价水蒸汽，采用了空气采样和分析技术。对德黑兰研究堆卡兰迪亚中氚化水蒸气的调查证明了潜在的氚污染风险。事实证明，采用蒸汽冷凝干燥技术可有效去除德黑兰研究堆卡兰迪瑞亚号中的氚水蒸汽。去除氚水蒸气后，卡兰迪亚的空气中不再含有高浓度的氚水蒸气。这将有助于确保退役作业的安全环境。

{"title":"The sampling and removal of tritiated water vapor in Taiwan research reactor","authors":"Jhih-Jhong Huang, Peng-Yu Chen","doi":"10.1016/j.pnucene.2024.105482","DOIUrl":"10.1016/j.pnucene.2024.105482","url":null,"abstract":"<div><div>The Taiwan Research Reactor (TRR) is a research facility in Taiwan. TRR is a heavy water reactor designed based on the National Research Experiment (NRX) with output thermal power 40 MW. On January 3, 1973, it reach criticality. TRR had been operated for 15 years and was permanently shut down in early 1988. TRR uses heavy water moderation and light water cooling. Hence, some tritium was created when deuterium captures a neutron. The inside of the Calandria has been washed with clean water after shutdown. After draining the water in the interior, it is inevitable that some residual water would remain at the bottom of the Calandria. Air sampling and analysis techniques were employed to detect and quantify tritiated water vapor. The investigation of tritiated water vapor in the TRR Calandria proved potential tritium contamination risks. Adopting the vapor condensation drying technique proved effective in removing tritiated water vapor from the TRR Calandria. After the removal of tritiated water vapor, the air in the Calandria no longer contains rich concentration of tritiated water vapor. That will help to ensure a safe environment for decommissioning operations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"177 ","pages":"Article 105482"},"PeriodicalIF":3.3,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142423744","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Aloe barbadensis - Assisted MgBiO/MgCuBiO/MgBaBiO nanocomposites as effective gamma shielding novel materials 芦荟--辅助镁硼氧化物/镁铜硼氧化物/镁钡硼氧化物纳米复合材料作为有效的伽马屏蔽新型材料

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-04 DOI: 10.1016/j.pnucene.2024.105470

K. Kruthika , S.M. Rumana Farheen , H.C. Manjunatha , Y.S. Vidya , S. Manjunatha , R. Munirathnam , S. Krishnaveni , K.N. Sridhar

For the first of its kind, MgO/Bi₂O₃ (1:1) (MBO), MgO/CuO/Bi₂O₃ (1:1:1) (MCBO) and MgO/BaO/Bi₂O₃ (1:1:1) (MBBO) binary/ternary nanocomposites (NCs) were synthesized by Aloe barbadensis (Aloe vera) mediated solution combustion method. The as-formed NCs are subjected to calcination for 3 h at 600^oC.The Mg-based NCs acquired are subjected to various analytical characterization methodologies, including PXRD (Powder X-ray Diffraction), SEM (Scanning Electron Microscopy), EDS (Energy Dispersive X-ray Spectroscopy), FTIR (Fourier Transform Infrared Spectroscopy), and Ultraviolet–Visible Spectroscopy (UV–Vis). The Bragg reflections confirm the formation of monoclinic phase BaO, Bi₂O₃, cubic MgO and tetragonal CuO nanoparticles. The surface morphology of MBO, MCBO and MBBO NCs consists of irregular sized, leafy and burrow like structured agglomerated NPs with hollows and voids/pores. The Eg (direct energy band gap) determined from W-H (Wood and Tauc's) plot was found to be 3.09, 3.19 and 3.12 eV for MBO, MCBO and MBBO NCs respectively. The evaluation of gamma radiation shielding properties was conducted using a NaI (Tl) detector connected to a multi-channel analyzer. Shielding parameters of MBO, MCBO and MBBO, such as μ/ρ (mass attenuation coefficient), λ (mean free path), HVL (half-value layer), TVL (tenth-value layer) are measured. Among the as-synthesized Mg-based NCs, MBO exhibits better shielding properties compared to MCBO and MBBO. Thus, the procured Mg-based NCs are suitable for applications where weight reduction is critical, such as aerospace or portable radiation shielding systems.

首次采用芦荟（Aloe barbadensis）介导的溶液燃烧法合成了 MgO/Bi2O3 (1:1) (MBO)、MgO/CuO/Bi2O3 (1:1:1) (MCBO) 和 MgO/BaO/Bi2O3 (1:1:1) (MBBO) 二元/三元纳米复合材料（NCs）。获得的镁基纳米复合材料采用了多种分析表征方法，包括粉末 X 射线衍射（PXRD）、扫描电子显微镜（SEM）、能量色散 X 射线光谱（EDS）、傅立叶变换红外光谱（FTIR）和紫外可见光谱（UV-Vis）。布拉格反射证实了单斜相 BaO、Bi2O3、立方 MgO 和四方 CuO 纳米粒子的形成。MBO、MCBO 和 MBBO NCs 的表面形态由不规则大小、叶状和毛刺状结构的团聚 NPs 组成，并带有空洞和空隙/孔隙。根据 W-H（Wood and Tauc's）图确定的 MBO、MCBO 和 MBBO NCs 的 Eg（直接能带隙）分别为 3.09、3.19 和 3.12 eV。伽马辐射屏蔽性能的评估是使用与多通道分析仪相连的 NaI（Tl）探测器进行的。测量了 MBO、MCBO 和 MBBO 的屏蔽参数，如 μ/ρ（质量衰减系数）、λ（平均自由路径）、HVL（半值层）、TVL（十值层）。在新合成的镁基 NCs 中，与 MCBO 和 MBBO 相比，MBO 表现出更好的屏蔽性能。因此，所制备的镁基 NC 适用于对减重要求较高的应用领域，如航空航天或便携式辐射屏蔽系统。

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引用次数: 0

Study on the effect of Lorentz force on the nuclear reactor core melt stratification in electromagnetic cold crucible 洛伦兹力对电磁冷坩埚中核反应堆堆芯熔体分层的影响研究

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-03 DOI: 10.1016/j.pnucene.2024.105475

Kerong Guo , Yang Li , Houjun Gong , Yuanfeng Zan , Zumao Yang , Yanping Huang

To understand the effect of the Lorentz force on core melt stratification during experiments, a multi-physical field coupling model of electromagnetic induction, non-isothermal flow, and two-phase flow is established in this paper. The evolution of the layered core melt in the electromagnetic cold crucible is studied for different relative densities and phase volume fraction ratios between metal and oxide. The calculation results show that the coupling model can accurately describe the structural evolution of core melt. The Lorentz force cannot change the relative positions of the metal and oxide, but it can significantly affect the morphology of the core melt. If the density of the metal is less than that of the oxide, the Lorentz force causes the light metal with a small phase volume fraction to form a hemispherical shape in the top center of the crucible. As the phase volume fraction increases, the Lorentz force become weaker than those of gravity and buoyancy. The light metal is then flat. When the metal density is greater than the oxide density, the Lorentz force causes the heavy metal with a small phase volume fraction to appear as a hemisphere, but causes the heavy metal with a large phase volume fraction to appear as an ellipsoid in the lower part of the crucible.

为了解洛伦兹力对实验过程中炉芯熔体分层的影响，本文建立了电磁感应、非等温流动和两相流动的多物理场耦合模型。研究了在不同相对密度和金属与氧化物相体积分数比的情况下，电磁冷坩埚中芯熔体分层的演化过程。计算结果表明，耦合模型能准确描述芯熔体的结构演变。洛伦兹力不能改变金属和氧化物的相对位置，但却能显著影响核心熔体的形态。如果金属的密度小于氧化物的密度，洛伦兹力会使相体积分数较小的轻金属在坩埚顶部中心形成半球形。随着相体积分数的增加，洛伦兹力变得比重力和浮力弱。此时轻金属呈扁平状。当金属密度大于氧化物密度时，洛伦兹力会使相体积分数较小的重金属呈半球形，但会使相体积分数较大的重金属在坩埚下部呈椭圆形。

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引用次数: 0

Improvement and validation of discrete-sectional method based code for fast reactor aerosol dynamics 基于离散截面法的快堆气溶胶动力学代码的改进与验证

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-03 DOI: 10.1016/j.pnucene.2024.105457

Shikha Sivakumar, A Jasmin Sudha, V. Subramanian, B. Venkatraman

The study of the dynamics of aerosols produced during a severe accident is vital to the safety analysis of a Sodium-cooled Fast Reactor (SFR). Sodium leakage into the containment following a severe accident may result in the production of sodium aerosols along with fission product aerosols. The in-containment radioactive source term depends upon the settling behaviour of fission product aerosols, which co-agglomerate with sodium fire aerosols. Hence, an in-depth understanding of the dynamics and deposition of aerosols within the containment is essential for the safety assessment of an SFR. The current work uses an open-source code based on the Discrete-Sectional (DS) method to solve the simplified form of General Dynamics Equation (GDE) for aerosols relevant to fast reactors. Aerosols of SrO₂, CeO₂ and sodium are considered in the present work. The DS code has been modified and further improved by including gravitational coagulation, turbulent coagulation, Brownian deposition, gravitational deposition and thermophoretic deposition so that the code can handle the different processes leading to the deposition of aerosols following a sodium fire. The code is also enhanced to account for the effect of relative humidity through the modified Cooper's equation. The improved code is then validated with different experiments conducted in the Aerosol Test Facility (ATF), India; Mini Sodium Fire Facility (MINA), India; and AHMED (Aerosol and Heat Transfer Measurement Device), Finland. Validation from different facilities confirms the applicability of the code to various scenarios. It is found that the modified DS code could predict the decay of suspended mass concentration of aerosols in different enclosures with reasonable accuracy.

研究严重事故期间产生的气溶胶的动态对于钠冷快堆的安全分析至关重要。严重事故发生后，钠泄漏到安全壳中可能导致钠气溶胶和裂变产物气溶胶的产生。安全壳内的放射源项取决于裂变产物气溶胶的沉降行为，这些气溶胶会与钠火气溶胶共同聚集。因此，深入了解安全壳内气溶胶的动态和沉降对 SFR 的安全评估至关重要。目前的工作使用基于离散-截面（DS）方法的开源代码来求解与快堆相关的气溶胶的通用动力学方程（GDE）简化形式。本研究考虑了二氧化锰、二氧化 CeO2 和钠的气溶胶。对 DS 代码进行了修改和进一步改进，加入了重力凝结、湍流凝结、布朗沉积、重力沉积和热泳沉积，从而使代码能够处理钠着火后导致气溶胶沉积的不同过程。此外，还通过修改库珀方程增强了代码的功能，以考虑相对湿度的影响。改进后的代码通过在印度气溶胶测试设施（ATF）、印度小型钠火灾设施（MINA）和芬兰 AHMED（气溶胶和传热测量装置）进行的不同实验进行了验证。来自不同设施的验证证实了代码在各种情况下的适用性。结果发现，修改后的 DS 代码能够以合理的精度预测不同围护结构中悬浮气溶胶质量浓度的衰减。

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引用次数: 0

A neutronics-fuel coupling model for the simulation of constituent redistribution in U–Zr fuel 用于模拟 U-Zr 燃料中成分再分布的中子-燃料耦合模型

IF 3.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Progress in Nuclear Energy

Pub Date : 2024-10-03 DOI: 10.1016/j.pnucene.2024.105467

Oscar Lastres, Yunlin Xu, Yi Xie

Simulating the dynamic redistribution of elements within metallic fuels, such as U–10Zr (wt.%), is a complex challenge in nuclear power research. While existing models can adequately simulate constituent redistribution throughout the fuel pin, they rely heavily on pre-known power history data or power calculated a priori. Furthermore, existing models do not calculate an accurate actinide inventory throughout the fuel pin. This limitation leads to less precise power distribution calculations, which can negatively affect the efficiency and safety of fuel design and performance. This work aims to introduce a new approach to modeling constituent redistribution for U–10Zr (wt.%) fuel by developing a new code called PFPS that accounts for the depletion and diffusion of Zr throughout the fuel pin during burnup through an explicit coupling method. The approach independently calculates the power distribution by coupling with the reactor physics code, Serpent 2, thereby eliminating the need for externally supplied power history data. The main contribution of this study lies in this coupling to obtain a more accurate history of the compositional evolution of U–Zr fuel and its fission products. This procedure can enhance the understanding of constituent redistribution by calculating a more accurate power distribution and lays the foundation for future attempts to simulate the compositional and microstructural evolution of all materials in the fuel.

模拟 U-10Zr（重量百分比）等金属燃料内部元素的动态再分布是核电研究中的一项复杂挑战。虽然现有模型可以充分模拟整个燃料针内的成分再分布，但它们在很大程度上依赖于已知的功率历史数据或先验计算的功率。此外，现有模型无法计算整个燃料销的精确锕系元素库存。这种限制导致功率分布计算不够精确，从而对燃料设计和性能的效率和安全性产生负面影响。这项工作旨在通过开发一种名为 PFPS 的新代码，为 U-10Zr （重量百分比）燃料的成分再分布建模引入一种新方法。该方法通过与反应堆物理代码 Serpent 2 的耦合，独立计算功率分布，从而无需外部提供功率历史数据。本研究的主要贡献在于通过这种耦合获得了更准确的 U-Zr 燃料及其裂变产物成分演变史。这一程序可以通过计算更精确的功率分布来加深对成分再分布的理解，并为今后模拟燃料中所有材料的成分和微结构演变奠定基础。

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引用次数: 0