Nuclear Engineering and Design最新文献_第4页

Experimental investigation of subcooled flow boiling characteristics of water in vertical helically coiled tubes 水在垂直螺旋盘管中过冷流动沸腾特性的实验研究

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

Nuclear Engineering and Design

Pub Date : 2024-11-17 DOI: 10.1016/j.nucengdes.2024.113716

Yuqing Su, Xiaowei Li, Xinxin Wu

Helically coiled tubes are employed as heat transfer tubes in Once Through Steam Generator (OTSG) of High Temperature Gas-cooled Reactor (HTGR) due to their compact structure, large heat transfer area and excellent thermal expansion adaptability. However, the helical geometry induces centrifugal forces and secondary flows in the tube, resulting in notable differences in flow and heat transfer characteristics compared to that of straight tubes. This study conducted an experimental investigation on the onset of nucleate boiling (ONB) and the subcooled boiling heat transfer coefficient in helically coiled tubes with a large curvature ratio (δ = 0.109). The experimental parameters cover broad ranges. The system pressures are ranging from 3.5 to 7 MPa, mass fluxes are from 300 to 1100 kg/(m²·s) and heat fluxes are from 50 to 600 kW/m2. The experimental results indicate that the ONB can occur even when the average inner wall temperature is below the fluid’s saturation temperature. An increase in heat flux advances the ONB, while increases in mass flux and system pressure delay it. Enhancements in both heat flux and mass flux improve the subcooled boiling heat transfer coefficient. Additionally, higher system pressure also increases the heat transfer coefficient, although this effect diminishes as the quality increases. Based on the experimental data and dimensionless analysis, new correlations were proposed for predicting ONB and calculating the subcooled boiling heat transfer coefficient in helically coiled tubes. Both new correlations exhibit more accurate predictive capabilities, with mean absolute percentage error (MAPE) values of 6.20 % and 8.86 %, respectively.

由于螺旋盘管结构紧凑、传热面积大、热膨胀适应性强，因此被用作高温气冷堆（HTGR）一次蒸汽发生器（OTSG）的传热管。然而，螺旋几何形状会在管内产生离心力和二次流，导致流动和传热特性与直管相比存在显著差异。本研究对大曲率比（δ = 0.109）螺旋盘绕管中的成核沸腾（ONB）起始点和过冷沸腾传热系数进行了实验研究。实验参数范围很广。系统压力从 3.5 到 7 MPa，质量通量从 300 到 1100 kg/（m2-s），热通量从 50 到 600 kW/m2。实验结果表明，即使内壁平均温度低于流体的饱和温度，也会出现 ONB。热通量的增加会使 ONB 提前，而质量通量和系统压力的增加则会使 ONB 推迟。热通量和质量通量的增加都会提高过冷沸腾传热系数。此外，较高的系统压力也会提高传热系数，不过这种影响会随着质量的提高而减弱。根据实验数据和无量纲分析，提出了预测 ONB 和计算螺旋盘管过冷沸腾传热系数的新相关性。两种新的相关性都显示出更精确的预测能力，平均绝对百分比误差 (MAPE) 值分别为 6.20 % 和 8.86 %。

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引用次数: 0

Assessment of stratification and entrainment models in CATHARE 3 code during FONESYS activities 在 FONESYS 活动期间评估 CATHARE 3 代码中的分层和夹带模型

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

Nuclear Engineering and Design

Pub Date : 2024-11-17 DOI: 10.1016/j.nucengdes.2024.113700

Sofia Carnevali, Philippe Fillion

This paper presents the assessment of CATHARE 3 code against tests performed in the horizontal TPTF (Two Phase Tests Facility) and Mantilla facilities. This activity falls within the framework of a benchmark conducted by the Forum & Network of System Thermal-Hydraulics Codes in Nuclear Reactor Thermal-Hydraulics (FONESYS). The aim of this benchmark is to highlight the capabilities of system thermal–hydraulic codes to predict the horizontal stratification criteria and the onset of droplet entrainment and entrainment rate. One of the objectives of horizontal TPTF (TPTF-H) experiments, conducted by JAERI (Japan), was to analyse the thermal–hydraulic responses in the horizontal legs of Light Water Reactors. The considered tests were at steady state and saturated steam/water two-phase flow conditions for a pressure varying between 3 MPa and 12 MPa. CATHARE code simulations show a general rather good agreement with TPTF-H experimental results although some improvements like the stratification regime prediction seem necessary. In two horizontal test sections with different diameters, Mantilla carried out air/water experiments at low pressure, from stratified to annular flow conditions, including droplet entrainment. Entrainment fraction obtained both with the two-fluid 6-equation model and the 3-field model of CATHARE are compared against the experimental data. Some improvements of the existing models are proposed for the entrainment and deposition processes to better fit the experiments.

本文介绍了根据在水平 TPTF（两相试验设施）和 Mantilla 设施中进行的试验对 CATHARE 3 代码进行的评估。这项活动属于核反应堆热工水力论坛（FONESYS）的基准测试框架。该基准的目的是突出系统热工水力代码在预测水平分层标准、液滴夹带开始和夹带率方面的能力。日本 JAERI 公司进行的水平 TPTF（TPTF-H）试验的目标之一是分析轻水反应堆水平支腿的热-水力反应。所考虑的试验是在稳态和饱和蒸汽/水两相流动条件下进行的，压力介于 3 兆帕和 12 兆帕之间。CATHARE 代码的模拟结果与 TPTF-H 的实验结果基本吻合，但在分层机制预测等方面仍有改进的必要。Mantilla 在两个不同直径的水平测试断面上进行了低压空气/水实验，实验条件从分层流到环形流，包括液滴夹带。将双流体 6 方程模型和 CATHARE 的 3 场模型得出的夹带分数与实验数据进行了比较。针对夹带和沉积过程，对现有模型提出了一些改进建议，以更好地适应实验。

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引用次数: 0

Velocity and pressure fluctuations downstream analytical spacer grids: Structure and transport 分析间隔网格下游的速度和压力波动：结构与传输

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

Nuclear Engineering and Design

Pub Date : 2024-11-15 DOI: 10.1016/j.nucengdes.2024.113682

N. Turankok , T. Lohez , V. Biscay , L. Rossi

In Pressurized Water Reactors (PWR), fluid–structure interaction provokes vibrations of the fuel rods leading to grid-to-rod fretting. To explore the origins of the local excitations of the rods by the flow, analytical experiments are performed within a 5 × 5 rod bundle maintained by spacer grids. Experiments are performed using two sets of grids, i.e. configurations with and without mixing vanes, over a wide range of Reynolds number, i.e. from about 10,000 to 120,000. Particle Image Velocity measurements near the central instrumented rod reveal different structures of velocity fields and their fluctuations for the two configurations. Mean fields are in U inverted shape without mixing vanes and in λ shape with mixing vanes. Fields of velocity fluctuations are in butterfly shape without mixing vanes and in Y (low Re) and K (high Re) shapes with mixing vanes. Near the grid,

{u^{'}}_{rms} / U_{flow}

increases by about 30% for

{Re}_{Dh}

varying from 10,000 to 120,000. The presence of eddies is highlighted by visualisations of the fields of the velocity fluctuations. The spacing of these eddies and their sizes are found to be in agreement with the periodic length scales (obtained from frequency peaks) and the integral length-scales measured. Frequencies of the observed frequency peaks are found to be the same for both velocity and pressure fluctuations. Consequently, a new Strouhal versus Reynolds map is built over all existing data for Reynolds numbers between 10,000 and 120,000 and for the two grids configurations. The integral length-scales are found to be about the same for velocity and pressure fluctuations near the grid. It is shown that events of pressure fluctuations are persistent and transported with a speed close to the one of the mean flow. Moreover, this transport is correlated with the displacement of velocity events. The coherence between measured turbulence statistics, the observed eddies and cross-correlations of pressure and velocity fluctuations support the conclusion that the large-scale eddies (about 0.2

D_{h}

) are at the origin of main pressure fluctuations and their transport. To the authors’ knowledge, this is the first experimental evidence of the origin and transport of pressure fluctuations downstream analytical grids with different geometries relevant to PWR fuel assemblies.

在压水堆（PWR）中，流体与结构的相互作用会引起燃料棒的振动，从而导致栅与栅之间的摩擦。为了探索燃料棒受流动影响而产生局部振动的原因，我们在一个由间隔网格维持的 5 × 5 燃料棒束内进行了分析实验。实验使用了两套网格，即带混合叶片和不带混合叶片的网格，雷诺数范围很广，从大约 10,000 到 120,000 不等。在中心仪器杆附近进行的粒子图像速度测量显示了两种配置下不同的速度场结构及其波动。不带混合叶片的平均场呈倒 U 形，带混合叶片的平均场呈 λ 形。速度波动场为不带混合叶片的蝶形和带混合叶片的 Y（低 Re）和 K（高 Re）形。在网格附近，当 ReDh 在 10,000 到 120,000 之间变化时，u′rms/Uflow 增加了约 30%。速度波动场的可视化突出显示了涡的存在。这些涡流的间距及其大小与周期性长度尺度（从频率峰值获得）和测量的积分长度尺度一致。观测到的频率峰的频率对速度和压力波动都是相同的。因此，在雷诺数介于 10,000 和 120,000 之间以及两种网格配置的所有现有数据基础上，建立了新的 Strouhal 与雷诺图。结果发现，网格附近的速度和压力波动的积分长度尺度大致相同。结果表明，压力波动事件是持续存在的，其传输速度接近平均流速。此外，这种传输与速度事件的位移相关。测量到的湍流统计数据、观测到的漩涡以及压力和速度波动的交叉相关性之间的一致性支持这样的结论，即大尺度漩涡（约 0.2 Dh）是主要压力波动及其传输的起源。据作者所知，这是有关压水堆燃料组件不同几何形状的分析网格下游压力波动的起源和传输的第一个实验证据。

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引用次数: 0

Numerical study on the impact characteristics of molten lead-bismuth droplets into water 关于熔融铅铋液滴入水冲击特性的数值研究

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

Nuclear Engineering and Design

Pub Date : 2024-11-15 DOI: 10.1016/j.nucengdes.2024.113689

Yue Lin , Dalin Zhang , Yutong Chen , Xisi Zhang , Jian Deng , Peng Du , Wenxi Tian , Suizheng Qiu , Guanghui Su

引用次数: 0

An investigation of structural integrity of CSB − for normal contact with RV outlet nozzle CSB 与 RV 出口喷嘴正常接触的结构完整性调查

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

Nuclear Engineering and Design

Pub Date : 2024-11-15 DOI: 10.1016/j.nucengdes.2024.113704

Niniek Ramayani Yasintha, Omondi Christopher Ondwasi, Namgung Ihn

This research is about improving the design of Core Support Barrel (CSB) by eliminating the gap between its hot-leg opening and the reactor vessel (RV) outlet nozzle. This gap causes significant bypass flow, impacting reactor performance. A proposed design modification aimed to establish contact between these components, requiring a thorough analysis of the contact interface. The study determined that a contact overlap of 15 mm meets ASME code standards and other relevant criteria. Moreover, this design improvement offers the added benefit of enhancing the seismic response of the reactor internals due to the increased horizontal support provided by the RV outlet nozzle to the CSB. Modal analysis revealed a substantial upward shift in the CSB’s natural frequency from 15.6 Hz-20.7 Hz to 21.9 Hz-38 Hz, representing a more than 40 % increase. The research demonstrates that the proposed design effectively eliminates bypass flow, and significantly improves the seismic response of the reactor internals.

这项研究旨在改进堆芯支撑筒（CSB）的设计，消除其热脚开口与反应堆容器（RV）出口喷嘴之间的间隙。该间隙会导致大量旁路流，影响反应堆性能。拟议的设计修改旨在建立这些组件之间的接触，需要对接触界面进行全面分析。研究确定，15 毫米的接触重叠符合 ASME 规范标准和其他相关标准。此外，由于 RV 出口喷嘴增加了对 CSB 的水平支撑，这一设计改进还能增强反应堆内部的地震响应。模态分析显示，CSB 的固有频率从 15.6 Hz-20.7 Hz 大幅上移至 21.9 Hz-38Hz，增幅超过 40%。研究表明，拟议的设计有效地消除了旁路流，并显著改善了反应堆内部的地震响应。

引用次数: 0

Robust multipole approach for continuous nuclear Data: RKFIT implementation for X2 VVER-1000 reactor benchmark 连续核数据的稳健多极方法：X2 VVER-1000 反应堆基准的 RKFIT 实施

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

Nuclear Engineering and Design

Pub Date : 2024-11-15 DOI: 10.1016/j.nucengdes.2024.113699

Abdolbaset Agh, Mahdi Zangian, Abdolhamid Minuchehr

The windowed multipole method stands out as a promising approach for effectively conducting on-the-fly Doppler-broadening for continuous nuclear cross-section data. Its implementation was predominantly relied on the conversion of the resolved resonance parameters and vector-fitting technique into the multipole representation. Recently, progress has been made in deriving multipole representations from continuous cross-section data, with one notable method being RKFIT mothed, that is a robust least square fitting method. The advantages and disadvantages of this method have been thoroughly investigated. Detailed instructions on utilizing this method are provided within the OpenMC Monte Carlo calculation code. The experimental reactor physics results of the X2 reactor (a VVER-1000 type reactor) have been used to benchmark the effectiveness of the continues nuclear data generated by this approach via the OpenMC code simulations. Also, the simulation results are compared with those obtained using continuous cross-section libraries and other multipole libraries.

带窗多极子方法是一种很有前途的方法，可以有效地对连续核截面数据进行即时多普勒扩频。其实施主要依赖于将解析的共振参数和矢量拟合技术转换为多极表示。最近，从连续横截面数据推导多极表示法取得了进展，其中一种值得注意的方法是 RKFIT mothed，这是一种稳健的最小平方拟合方法。对这种方法的优缺点进行了深入研究。OpenMC 蒙特卡罗计算代码中提供了使用这种方法的详细说明。X2 反应堆（VVER-1000 型反应堆）的反应堆物理实验结果被用来衡量这种方法通过 OpenMC 代码模拟生成的持续核数据的有效性。此外，还将模拟结果与使用连续截面库和其他多极库获得的结果进行了比较。

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引用次数: 0

A study of spherical tubesheet design for APR1400 and AP1000 steam generators APR1400 和 AP1000 蒸汽发生器球形管板设计研究

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

Nuclear Engineering and Design

Pub Date : 2024-11-15 DOI: 10.1016/j.nucengdes.2024.113701

Seungmin Kim, Namgung Ihn

This research investigates the potential of spherical tube sheets as a more efficient alternative to conventional flat tube sheets in APR1400 and AP1000 steam generators. By leveraging the inherent strength of spherical structures, we aim to reduce material consumption, manufacturing costs, and construction time. Finite Element Method simulations were employed to compare the deformation behavior of flat and spherical tube sheets under high-temperature and high-pressure conditions. The analysis identified optimal spherical tube sheet thicknesses that maintained structural integrity while achieving significant reductions in thickness: approximately 20–30 % for APR1400 and 40–50 % for AP1000. Moreover, the spherical design can enhance thermal efficiency by minimizing variations in heat transfer tube lengths, leading to more uniform heat distribution.

这项研究调查了球形管板作为 APR1400 和 AP1000 蒸汽发生器中传统扁平管板的更有效替代品的潜力。通过利用球形结构的固有强度，我们旨在减少材料消耗、制造成本和施工时间。我们采用有限元法模拟来比较扁平管板和球形管板在高温高压条件下的变形行为。分析确定了最佳的球形管板厚度，在保持结构完整性的同时显著减少了厚度：APR1400 和 AP1000 的厚度分别减少了约 20-30% 和 40-50%。此外，球形设计还可以通过最大限度地减少传热管长度的变化来提高热效率，从而使热量分布更加均匀。

引用次数: 0

Entropy generation analysis for fine flow states in PWR fuel assembly 压水堆燃料组件中细微流动状态的熵生成分析

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

Nuclear Engineering and Design

Pub Date : 2024-11-15 DOI: 10.1016/j.nucengdes.2024.113708

Yunsheng Zhang , Guangliang Chen , Hao Qian , Lixuan Zhang , Jinchao Li , Hanqi Zhang , Dabin Sun , Hansheng Zhi

Although the development of pressurized water reactor (PWR) technology has been relatively well developed so far, the depth of research on the thermo-hydraulic characteristics of the in-reactor coolant, especially the energy dissipation characteristics, still needs to be explored further. As an energy conversion system for PWR, entropy generation analysis plays a vital role in obtaining the irreversible loss of the coolant quantitatively and directionally, which is scarce. It is difficult to access the full coolant energy losses and thermal–hydraulic properties, and to provide direction and schemes for optimizing the flow field in the core. In this paper, the energy dissipation of the flow field is finely analyzed in the 5x5 rod bundle of the PWR core. For the irreversible dissipation of the flow energy of the coolant, the pulsation dissipation entropy generation and wall friction entropy generation account for about 90% and 10%, respectively, and the direct dissipation entropy generation is negligible. For the high dissipation region, the distribution of pulsation dissipation entropy generation in space has a directional role. The irreversibly dissipated energy is calculated using the definition of entropy generation, and the dissipation characteristics of cross-flow kinetic energy in different regions are analyzed accordingly. Lastly, the irreversible dissipation of temperature difference heat transfer is considered and correlated with the above hydraulic viscous dissipation, and the negative correlation characteristics of the two are found. Meanwhile, the parameter constructed by the two together has some advantage in flow field evaluation.

尽管压水堆（PWR）技术发展至今已相对成熟，但对堆内冷却剂热工水力特性，尤其是能量耗散特性的研究深度仍有待进一步探索。作为压水堆的能量转换系统，熵生成分析对定量、定向地获取冷却剂的不可逆损耗起着至关重要的作用，而这方面的研究却十分匮乏。要全面了解冷却剂的能量损失和热工水力特性，并为优化堆芯流场提供方向和方案十分困难。本文对压水堆堆芯 5x5 棒束流场的能量耗散进行了精细分析。对于冷却剂流动能量的不可逆耗散，脉动耗散熵生成和壁面摩擦熵生成分别约占 90% 和 10%，直接耗散熵生成可忽略不计。对于高耗散区域，脉动耗散熵的空间分布具有方向性。利用熵产生的定义计算不可逆耗散能量，并据此分析不同区域的横流动能耗散特征。最后，考虑了温差传热的不可逆耗散，并将其与上述水力粘性耗散相关联，发现了两者的负相关特性。同时，两者共同构建的参数在流场评价中具有一定的优势。

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引用次数: 0

Computational analysis for predicting fission product release of PeLUIt-40 under normal operating conditions 预测 PeLUIt-40 在正常运行条件下裂变产物释放的计算分析

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

Nuclear Engineering and Design

Pub Date : 2024-11-14 DOI: 10.1016/j.nucengdes.2024.113677

Anik Purwaningsih , Dwi Irwanto , Julwan H. Purba

<div><div>PeLUIt (<em>Pembangkit Listrik dan Uap untuk Industri</em>) is an High Temperature Gas Cooled Reactor (HTGR)-based cogeneration reactor designed by the Indonesian National Nuclear Energy Agency of Indonesia (now the Research Organization for Nuclear Energy, BRIN). HTGR uses tri-structural isotropic (TRISO) coated particle fuel, which is the main barrier to the release of fission products. The ability of the fuel to retain fission products both under normal and accident conditions is very important for design licensing. This study was conducted to predict the release of fission products of PeLUIt-40 (PeLUIt with a power of 40 MWt) under normal operating conditions. In this study, the Source Term Analysis Code System (STACY) was used to predict the release of radiologically significant fission products Ag110m, Cs137, I131, and Sr90. OpenMC was used to calculate neutronic parameters such as burnup (Fissions per initial heavy metal atom—FIMA), fast neutron fluence, and fission product inventory. A single pebble model with different irradiation times and temperature variations was used to simulate the fission product release in PeLUIt-40. The time and temperature variations were used to investigate the sensitivity of the fission product release fraction in PeLUIt-40 fuel to these parameters and to estimate the maximum safe fuel temperature during operation. The simulation results showed that the largest release fraction was Ag110m release compared to other radionuclide releases. At the normal operating temperature of 977 °C, the fission product release fractions during one-through-one-out (OTTO) and 5-pass cycles were two orders of magnitude lower than the failure fraction for the high-temperature reactor (HTR)-Module <span><math><mrow><mn>1.6</mn><mi>x</mi><msup><mrow><mn>10</mn></mrow><mrow><mo>-</mo><mn>4</mn></mrow></msup></mrow></math></span> and there was no defective particle during operation. In the OTTO cycle, the maximum fuel temperature that did not cause defective particle was about 1250 °C, but the Ag110m release fraction exceeded <span><math><mrow><mn>1.6</mn><mi>x</mi><msup><mrow><mn>10</mn></mrow><mrow><mo>-</mo><mn>4</mn></mrow></msup></mrow></math></span>. The release fraction of all fission products in the OTTO cycle is below <span><math><mrow><mn>1.6</mn><mi>x</mi><msup><mrow><mn>10</mn></mrow><mrow><mo>-</mo><mn>4</mn></mrow></msup></mrow></math></span> when the maximum fuel temperature is 1025 °C. While in the 5-pass cycle, the maximum fuel temperature of about 1200 °C does not cause defective particle, but the release fraction of Ag110m exceeds <span><math><mrow><mn>1.6</mn><mi>x</mi><msup><mrow><mn>10</mn></mrow><mrow><mo>-</mo><mn>4</mn></mrow></msup></mrow></math></span>. The release fraction of all fission products in the 5-pass cycle is below <span><math><mrow><mn>1.6</mn><mi>x</mi><msup><mrow><mn>10</mn></mrow><mrow><mo>-</mo><mn>4</mn></mrow></msup></mrow></math></span> when the maximum fuel temperature is 1020°

PeLUIt（Pembangkit Listrik dan Uap untuk Industri）是一种基于高温气冷堆（HTGR）的热电联产反应堆，由印度尼西亚国家核能机构（现为核能研究组织，BRIN）设计。高温气冷堆使用三结构各向同性（TRISO）涂层颗粒燃料，这是裂变产物释放的主要屏障。燃料在正常和事故条件下保留裂变产物的能力对于设计许可非常重要。本研究旨在预测 PeLUIt-40（功率为 40 兆瓦的 PeLUIt）在正常运行条件下的裂变产物释放情况。在这项研究中，使用源项分析代码系统 (STACY) 预测了具有放射性意义的裂变产物 Ag110m、Cs137、I131 和 Sr90 的释放。OpenMC 用于计算中子参数，如燃烧度（每个初始重金属原子的裂变量-FIMA）、快中子通量和裂变产物存量。使用不同辐照时间和温度变化的单一鹅卵石模型模拟 PeLUIt-40 中的裂变产物释放。利用时间和温度变化来研究 PeLUIt-40 燃料中裂变产物释放分数对这些参数的敏感性，并估算运行期间燃料的最高安全温度。模拟结果表明，与其他放射性核素释放量相比，Ag110m 的释放量最大。在 977 ℃ 的正常运行温度下，一穿一出（OTTO）循环和五通循环的裂变产物释放分数比高温堆（HTR）-模块的失效分数低两个数量级，为 1.6x10-4，运行期间没有出现缺陷粒子。在 OTTO 循环中，不产生缺陷粒子的最高燃料温度约为 1250 ℃，但 Ag110m 的释放分数超过了 1.6x10-4。当最高燃料温度为 1025 ℃ 时，OTTO 循环中所有裂变产物的释放分数都低于 1.6x10-4。而在 5 次循环中，最高燃料温度约为 1200 ℃ 时不会产生缺陷粒子，但 Ag110m 的释放分数超过了 1.6x10-4。当最高燃料温度为 1020 ℃ 时，5 次循环中所有裂变产物的释放分数都低于 1.6x10-4。模拟结果证实，在正常运行期间，PeLUIt-40 燃料中裂变产物的释放量低于 1.6x10-4 的安全要求限值，并且在正常运行期间不会出现缺陷粒子。这表明，TRISO 涂层 PeLUIt-40 颗粒能够防止在正常运行温度下释放裂变产物。

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引用次数: 0

Fast prediction of key parameters in FEBA using the COSINE subchannel code and artificial neural network 利用 COSINE 子信道编码和人工神经网络快速预测 FEBA 中的关键参数

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

Nuclear Engineering and Design

Pub Date : 2024-11-13 DOI: 10.1016/j.nucengdes.2024.113709

Yingran Guo, Hao Zhang, Lin Chen, Meng Zhao, Yanhua Yang

Numerical techniques have emerged as an essential tool for operators and designers to preemptively acquire key parameters in accidents analysis. However, due to insufficient experience, it is difficult for them to obtain satisfactory numerical results. Moreover, the uncertainty analysis and quantification necessitate the simulation of a substantial number of samples, which requires a significant amount of computational time. Therefore, the development of a fast prediction model becomes imperative. In this work, a prediction model based on the in-house COSINE subchannel code and Multi-Head Perceptron (MHP) is developed. The COSINE subchannel code is employed to provide data sets for training neural networks. Firstly, the numerical results of COSINE subchannel code are compared with experimental data to ensure the accuracy of data sets. Secondly, input features for neural networks are selected by evaluating the impact of input parameters on numerical results, and a series of simulations is carried out to generate data sets. Then, a comparative analysis was conducted between the Multi-Layer Perceptron (MLP) and Support Vector Regression (SVR) models, and the MLP model performs better. Subsequently, the MLP was compared with the MHP, demonstrating the advantage of MHP model. Based on this, the predictions are conducted using the MHP model and the distribution of key parameters is compared with that obtained by COSINE subchannel code. The results illustrate that developed MHP model is an efficient tool for predicting key parameters during the reflooding phase.

数值技术已成为操作人员和设计人员在事故分析中预先获取关键参数的重要工具。然而，由于经验不足，他们很难获得令人满意的数值结果。此外，不确定性分析和量化必须对大量样本进行模拟，这需要大量的计算时间。因此，开发快速预测模型势在必行。在这项工作中，开发了一种基于内部 COSINE 子信道编码和多头感知器（MHP）的预测模型。COSINE 子信道编码为训练神经网络提供了数据集。首先，将 COSINE 子信道编码的数值结果与实验数据进行比较，以确保数据集的准确性。其次，通过评估输入参数对数值结果的影响来选择神经网络的输入特征，并进行一系列模拟来生成数据集。然后，对多层感知器（MLP）和支持向量回归（SVR）模型进行了对比分析，结果发现 MLP 模型的性能更好。随后，将 MLP 与 MHP 进行了比较，证明了 MHP 模型的优势。在此基础上，使用 MHP 模型进行了预测，并将关键参数的分布与 COSINE 子信道编码获得的参数进行了比较。结果表明，所开发的 MHP 模型是预测再淹没阶段关键参数的有效工具。

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引用次数: 0