Nuclear Engineering and Technology最新文献_第2页

Assemblies design and modeling analysis of a new fine mesh neutronics/thermal-hydraulics coupling benchmark for plate-type PWR core 板式压水堆堆芯新型精细网格中子/热液压耦合基准的装配设计和建模分析

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.058

Zhigang Li , Wei Lu , Shanfang Huang , Xiafeng Zhou , Yingwei Wu , Bangyang Xia , Junji Chen , Tao He , Guodong Liu , Yangyu Zhong , Zhiying Yue

In order to support the verification of neutronics/thermal-hydraulics coupling calculation method or simulation codes at the fine mesh for plate-type pressurized water reactor (PWR) with high parameters (which the ratio of power to mass flow rate is greater than 235 kW/kg, the core outlet enthalpy exceeds 1500 kJ/kg), a set of coupling calculation of plate-type PWR based on high parameters (COPHP) is design by Nuclear Power Institute of China (NPIC). Multiple industry research teams, including Tsinghua University, Xi'an Jiaotong University, and Huazhong University of Science and Technology, participated in the production of COPHP benchmark. This article provides a detailed explanation of the completed assembly design and modeling calculations, and provides the k_eff, ~~pin-by-pin wise~~ fine mesh relative power distribution, and deviation results calculated for 10 conditions of 6 assemblies using RMC, OpenMC, and KYLIN V2 software. The results show that: 1) when using the same cross-sectional library, the OpenMC calculation results are in good agreement with the RMC results. Taking ENDF/B-VII.1 as an example, the maximum deviation of k_eff in the entire burnup process of all assemblies is −157pcm, the maximum deviation of relative power is −1.13 %, and the maximum power weight error(PWE) is 0.226 %. 2) Compared between KYLIN V2 and the RMC by using ENDF/B-VII.1, the maximum deviation of k_eff is −468pcm, the maximum deviation of relative power is −1.49 %, and the maximum PWE is 0.316 % when calculating the standard assemblies with all control rod out(ARO) condition and two burnable poison assemblies. The maximum deviation of k_eff is −795pcm, the maximum deviation of relative power is 1.81 %, and the maximum PWE is 0.369 % when calculating the standard assemblies with all control rod inserted(ARI) conditions.

为支持高参数（功率与质量流量比大于 235 kW/kg，堆芯出口焓大于 1500 kJ/kg）板式压水堆中子/热工耦合计算方法或仿真代码的精细网格验证，中国核动力研究院设计了一套基于高参数的板式压水堆耦合计算（COPHP）。包括清华大学、西安交通大学和华中科技大学在内的多个行业研究团队参与了 COPHP 基准的制作。本文对已完成的组件设计和建模计算进行了详细说明，并提供了使用 RMC、OpenMC 和 KYLIN V2 软件对 6 个组件的 10 种工况计算出的、细网格相对功率分布和偏差结果。结果显示1) 在使用相同截面库的情况下，OpenMC 计算结果与 RMC 结果一致。以ENDF/B-VII.1为例，所有组件在整个燃烧过程中的最大偏差为-157pcm，相对功率的最大偏差为-1.13 %，最大功率权重误差（PWE）为 0.226 %。2) 使用ENDF/B-VII.1对KYLIN V2和RMC进行比较，在计算全部控制棒退出（ARO）条件下的标准组件和两个可燃毒物组件时，最大偏差为-468pcm，最大相对功率偏差为-1.49 %，最大功率权重误差（PWE）为0.316 %。在计算全部控制棒插入（ARI）条件下的标准组件时，最大偏差为 -795pcm，最大相对功率偏差为 1.81 %，最大 PWE 为 0.369 %。

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引用次数: 0

Optimization of a large-area grid electrode for negative ion source in fusion neutral beam injector 聚变中性束注入器负离子源大面积网格电极的优化

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.051

Jinhong Wang , Yan Wang , Yuwen Yang , Yuming Gu , Yahong Xie , Lizhen Liang , Jianglong Wei

At present, the CRAFT (Comprehensive Research Facility for Fusion Technology) dual-driver negative ion source for neutral beam injection applications has made progress. In order to solve the problem of heat removal function in large-area extraction grid, the multi-physics simulation, manufacturing and testing are carried out. Two simulation conditions before Cs seeding (J_e∼350 A/m², J_H-∼50 A/m²) and after sufficient Cs conditioning (J_e∼150 A/m², J_H-∼250 A/m²) were considered to evaluate the power load on EG, which peak power density were 11 and 6 MW/m² respectively and were concentrated on locate near the two aperture rows. This paper mainly studies three designs of minor cooling channels (double bending channels, double straight channels, and single straight channel). First, the grid temperature, thermal stress and thermal deformation are compared with an inlet water rate of 1.21 kg/s at 35 °C, and the channel forming methods and welding types mechanisms are compared. The simulation results show that the thermos-mechanical property of double bending channels is the best, and that of double straight channels is slightly worse. In terms of processing and manufacturing, the yield of double bending channels is lower than double straight channels, and it is easy to leak during the experiment.

目前，用于中性束注入应用的 CRAFT（聚变技术综合研究设施）双驱动负离子源已取得进展。为了解决大面积萃取栅的散热功能问题，进行了多物理场仿真、制造和测试。考虑了铯播种前（J∼350 A/m，J-∼50 A/m）和铯充分调节后（J∼150 A/m，J-∼250 A/m）两种模拟条件，以评估 EG 的功率负荷，其峰值功率密度分别为 11 MW/m和 6 MW/m，并集中在两排孔附近。本文主要研究了三种小冷却通道设计（双弯曲通道、双直线通道和单直线通道）。首先，比较了 35 °C、进水速度为 1.21 kg/s 时的栅格温度、热应力和热变形，并比较了通道形成方法和焊接类型机制。模拟结果表明，双弯曲水道的热机械性能最好，双直水道的热机械性能稍差。在加工制造方面，双弯曲槽钢的成品率低于双直槽钢，且在实验过程中容易发生泄漏。

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引用次数: 0

Multi-step multivariate forecasting of transmission power in NPPs using operational and meteorological data 利用运行和气象数据对核电站输电功率进行多步骤多变量预测

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.038

Jaeseok Yoo , Young-jin Oh , Nam-hyun Kim , Soo-ill Lee , Jaepil Ko

As the proportion of renewable energy has increased in the national power grid of Republic of Korea, various efforts are needed to maintain the stability of total power generation. All kinds of power plants, including nuclear power, must notify the grid operation organization of their expected transmission power. Even in NPPs, the accuracy of transmission power forecasting can increase the plant owner's economic benefits as well as the stability of the power grid. The transmission power of a NPP is affected by various plant conditions and environmental conditions, including the temperature of circulating water (sea water). In this study, we explored how to effectively handle the long-term dependence problem and various data characteristics to increase the forecasting accuracy of transmission power in NPPs by introducing a Seq2Seq model with an encoder-decoder structure and an attention mechanism, beyond traditional time series deep learning models, especially LSTM. This approach will improve the accuracy of transmission power forecasting and contribute to a stable power supply. Additionally, the model is expected to provide a realistic and practical solution for the power demand response of power plants.

随着大韩民国国家电网中可再生能源比例的增加，需要做出各种努力来维持总发电量的稳定。包括核电在内的各类发电厂都必须向电网运行机构通报其预期输电功率。即使是核电站，输电功率预测的准确性也能提高电站业主的经济效益和电网的稳定性。核电站的输电功率受各种电站条件和环境条件的影响，包括循环水（海水）的温度。在本研究中，我们探索了如何通过引入具有编码器-解码器结构和注意力机制的 Seq2Seq 模型，超越传统的时间序列深度学习模型（尤其是 LSTM），有效地处理长期依赖性问题和各种数据特征，从而提高核电站的输电功率预测精度。这种方法将提高输电功率预测的准确性，有助于实现稳定的电力供应。此外，该模型有望为发电厂的电力需求响应提供现实可行的解决方案。

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引用次数: 0

Performances of halide scintillators for the dosimetry based on gamma-ray spectrometry for environmental monitoring systems 卤化物闪烁体在环境监测系统伽马射线分光计基础上的剂量测定性能

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.042

Wanook Ji, Eunjoong Lee, Young-Yong Ji

High pressure ion chambers (HPIC) and NaI(Tl) scintillation detectors are widely used to monitor the ambient dose equivalent rate H∗(10) within and around the Korean nuclear facilities. However, HPIC cannot provide spectrometric information and NaI(Tl) detector is limited in identifying nuclides, such as ¹³¹I, ¹³⁴Cs, and ¹³⁷Cs, released from nuclear facilities owing to its insufficient energy resolution. This study employed four halide scintillators – LaBr₃(Ce), CeBr₃, and SrI₂(Eu) – to measure the ambient dose equivalent rate and detect gamma nuclides from measured energy spectrum. First, the pulse–shaping time in the signal processing unit was optimized for each scintillator. Second, energy resolution and counting efficiency were estimated for ¹³⁷Cs and ⁶⁰Co. Finally, an irradiation test was performed to estimate the dose rate. Based on these results, LaBr₃(Ce) and NaI(Tl) were selected as in situ gamma spectrometry system for measuring environmental radiation, and field experiments were conducted near the Fukushima Daiichi nuclear power plant to measure the dose rate.

高压离子室（HPIC）和 NaI（Tl）闪烁探测器被广泛用于监测韩国核设施内部和周围的环境剂量当量率 H*（10）。然而，HPIC 无法提供光谱信息，而 NaI（Tl）探测器由于能量分辨率不足，在识别核设施释放的 I、Cs 和 Cs 等核素方面受到限制。这项研究采用了四种卤化物闪烁体--LaBr(Ce)、CeBr 和 SrI(Eu)--来测量环境剂量当量率，并从测量的能谱中探测伽马核素。首先，针对每种闪烁体优化了信号处理装置中的脉冲整形时间。其次，对铯和钴的能量分辨率和计数效率进行了估算。最后，进行了辐照试验，以估算剂量率。根据这些结果，我们选择了 LaBr（Ce）和 NaI（Tl）作为测量环境辐射的原位伽马能谱系统，并在福岛第一核电站附近进行了实地实验，以测量剂量率。

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引用次数: 0

Laser cutting study of zirconium alloys for nuclear decommissioning 用于核退役的锆合金激光切割研究

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.053

Jae Sung Shin , Joonsoo Ock , Sungyeol Choi

We conducted laser cutting studies on zirconium alloys, specifically Zircaloy-2 and Zr-2.5%Nb alloy, which are used as constituent materials in the nuclear fuel channel of a pressurized heavy water reactor. The study measured the maximum cutting speed, amount of secondary emissions, and aerosol characteristics for each material using 10 mm thick plate specimens. The cutting performance of the zirconium alloys was similar to each other. At a laser power of 1–5 kW, the maximum cutting speed ranged from 750 to 1900 mm/min, and the amount of secondary emissions per length ranged from 32 to 53 g/m. Compared to 304 L stainless steel, the maximum cutting speed was 1.7–1.9 times higher, and the amount of secondary emissions was about 60–70 % of that of stainless steel. Analyzing the physical properties of aerosols, both zirconium alloys generated larger particles with a count median aerodynamic diameter of 0.25 μm, which is approximately 15–17 % larger than that of 304 L stainless steel.

我们对锆合金（特别是 Zircaloy-2 和 Zr-2.5%Nb 合金）进行了激光切割研究，这两种合金是加压重水反应堆核燃料通道的组成材料。研究使用 10 毫米厚的板材试样测量了每种材料的最大切割速度、二次排放物量和气溶胶特性。锆合金的切割性能彼此相似。在激光功率为 1-5 kW 时，最大切割速度为 750 至 1900 mm/min，单位长度的二次辐射量为 32 至 53 g/m。与 304 L 不锈钢相比，最大切割速度提高了 1.7-1.9 倍，二次排放量约为不锈钢的 60-70%。分析气溶胶的物理性质，两种锆合金产生的颗粒都较大，气动直径的计数中值为 0.25 μm，比 304 L 不锈钢大约 15-17%。

引用次数: 0

Electrokinetic in situ leaching of U from low-permeability uranium ore

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.09.015

Chunguang Li , Wenji Wang , Zhenzhong Liu , Yongmei Li , Qi Liu , Longcheng Liu

In this study, the impacts of electrokinetic in situ leaching of uranium (EK-ISLU) on the pore structure and connectivity of the low-permeability uranium ore and its leaching efficiency were investigated. Our results demonstrated that a direct current field with an intensity of 7.5 V/cm increased the porosity by 7.07 % and enhanced the uranium leaching efficiency by 73.73 % using a sulfuric acid solution. Under electrical stimulation, the leaching agent exhibited uniform penetration ability through combined electromigration and electroosmosis effects, significantly improving the directed migration of the target ions. Various characterization techniques, such as Fourier transform infrared (FTIR) spectroscopy, scanning electron microscopy-energy dispersive X-ray spectrometry (SEM-EDS), three-dimensional computed tomography (CT) pore scanning and Brunauer-Emmett-Teller (BET), showed that the electrokinetic-induced action overcame the preferential flow phenomena, promoted the dissolution of the target minerals and gangue rocks, and notably improved the pore structure to ultimately enhance the uranium leaching efficiency. Our results provide valuable insights for the development of innovative technologies aimed at efficiently recovering U from low-permeability uranium ores.

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引用次数: 0

High temperature mechanical properties of diffusion welded alloy 800H 扩散焊接合金 800H 的高温力学性能

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.046

Jong-Bae Hwang , Injin Sa , Eung-Seon Kim , Dong-Hyun Lee

High temperature mechanical properties of diffusion welded Alloy 800H was investigated to fabricate a printed circuit heat exchanger (PCHE) for high temperature reactor (HTR) systems. Surface treatment was performed on Alloy 800H to transform the solubility product (K_sp) to exceed the reaction quotient (Q). The surface treatment facilitated the grain boundary migration across the interface. The yield strengths exceeded the values described in ASME Section III Division 5 Table HBB-I-14.5 up to 760 °C, while the tensile strengths were comparable to Table HBB-3225-1 up to 700 °C. At 760 °C, the tensile strength was ∼30 MPa lower than the code. The stress-to-rupture values exceeded the expected minimum stress-to-rupture values of Alloy 800H described in Table HBB-I-14.6C. The ductility of the diffusion weldment acquired from the tensile test was comparable to the as-received Alloy 800H. Meanwhile, the formation of the secondary precipitates on the interface during the stress-to-rupture test deteriorated the ductility of the diffusion weldment, which induced intergranular fracture.

研究了扩散焊接合金 800H 的高温机械性能，以制造用于高温反应器（HTR）系统的印刷电路热交换器（PCHE）。对合金 800H 进行了表面处理，使溶度积（）超过反应商（）。表面处理促进了晶界在界面上的迁移。在 760 °C以下，屈服强度超过了 ASME 第 III 章第 5 分部表 HBB-I-14.5 中描述的值，而在 700 °C以下，抗拉强度与表 HBB-3225-1 中描述的值相当。在 760 °C 时，抗拉强度比规范值低 30 MPa。应力-破裂值超过了表 HBB-I-14.6C 中描述的合金 800H 的预期最小应力-破裂值。通过拉伸试验获得的扩散焊接件的延展性与初始合金 800H 相当。同时，在应力-断裂测试期间，界面上形成的二次析出物降低了扩散焊接件的延展性，从而导致晶间断裂。

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引用次数: 0

Searching for an optimal fuel for a supercritical water reactor (SCWR) to dispose of minor actinides resulting from the nuclear waste 为超临界水反应堆（SCWR）寻找最佳燃料，以处理核废料产生的小锕系元素

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.063

Nassar Alnassar , Maha Algarawi , Sitah Alanazi , Muneerah Al-Aqeel , Ahmed Salah Khaliil , A. Abdelghafar Galahom

This work investigates the optimal treatment of Minor Actinides (MAs) produced in the spent fuel of nuclear reactors operating around the world. This process is very important in addressing the challenges associated with nuclear waste and reducing environmental impact. A three-dimensional model of a supercritical water reactor (SCWR) has been designed using MCNPX to make a comprehensive neutronic analysis. MAs with a concentration of 1 % have been added to UO₂, (Th, U)O₂, (Th, ²³³U)O₂ and (Th, rgPu)O₂. These fuels have been uploaded in the SCWR assembly and burned in a separate fuel cycle. The infinity multiplication factor (k_inf) of the suggested has been investigated with and without adding the minor actinides to analyze the effect of MAs on the reactor reactivity. The fuel constituents, plutonium concentration, MAs concentration and transmutation rate have been tracked with fuel burnup. The reactivity temperature coefficients have been calculated for the suggested cases to ensure the validity of the suggested fuels. The power peaking factor (PPF) and radial power distribution have been calculated for the suggested fuels. The neutronic analysis confirms the suitability of the suggested fuel in burning a significant amount of MAs.

这项工作研究的是如何优化处理世界各地运行的核反应堆乏燃料中产生的小锕系元素（MAs）。这一过程对于应对与核废料相关的挑战和减少对环境的影响非常重要。利用 MCNPX 设计了一个超临界水反应堆（SCWR）的三维模型，以进行全面的中子分析。在 UO、(Th, U)O、(Th, U)O 和 (Th, rgPu)O 中添加了浓度为 1 % 的 MA。这些燃料已装入重水反应堆组件，并在单独的燃料循环中燃烧。在加入和不加入次锕系元素的情况下，对所建议的无限倍增因子（k）进行了研究，以分析次锕系元素对反应堆反应性的影响。燃料成分、钚浓度、MAs 浓度和嬗变率随燃料燃烧而变化。计算了建议情况下的反应温度系数，以确保建议燃料的有效性。还计算了建议燃料的功率峰值系数（PPF）和径向功率分布。中子分析证实，建议的燃料适合燃烧大量的 MA。

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引用次数: 0

TRACE assessment of density wave instability onset with void reactivity feedback under natural circulation 利用自然环流下的空隙反应反馈对密度波不稳定性的 TRACE 评估

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.064

Paul Hurley , Yang Liu , Tomasz Kozlowski , Juliana P. Duarte

Density wave oscillation (DWO) is an important safety concern for boiling water reactors (BWR) due to their high void fraction in the core. Power extensions to existing reactors such as the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) lead to increase susceptibility of DWO-type instability following an anticipated transient without scram (ATWS). Experiments performed at the Karlstein Thermal Hydraulic Test Facility (KATHY) have reproduced the reactivity feedback mechanism in BWRs under ATWS conditions. Using a neutronics module simulator, the KATHY facility was able to provide data on the effect of different neutronic parameters on DWO onset. This paper serves to assess the capability of the thermal-hydraulics code TRACE V5P7 for simulating DWO onset and development under natural circulation with neutronic feedback. A model of the KATHY natural circulation facility is created in TRACE and a reactivity feedback mechanism is implemented using a manual control scheme to simulate the parametric effects provided by the tests. This comparison allows for an assessment of the TRACE code as well as a better understanding of the instability mechanisms and behavior under the given conditions.

密度波振荡（DWO）是沸水反应堆（BWR）的一个重要安全问题，因为其堆芯中的空隙率很高。现有反应堆的功率扩展，如最大扩展负荷线极限分析增强版（MELLLA+），导致在预期无扰动瞬态（ATWS）之后更容易出现 DWO 型不稳定性。在卡尔斯坦热工水力试验设施（KATHY）进行的实验再现了在 ATWS 条件下生物质反应堆的反应反馈机制。利用中子模块模拟器，KATHY 设备能够提供不同中子参数对 DWO 发生影响的数据。本文旨在评估 TRACE V5P7 热工水力代码在自然循环和中子反馈条件下模拟 DWO 发生和发展的能力。在 TRACE 中创建了 KATHY 自然循环设施模型，并使用手动控制方案实施了反应反馈机制，以模拟试验提供的参数效应。通过比较，可以对 TRACE 代码进行评估，并更好地了解给定条件下的不稳定机制和行为。

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引用次数: 0

Identifying a probe to visualize the variability of operating teams for supporting the human reliability analysis of nuclear power plants: An explanatory study 确定可视化操作团队可变性的探针，以支持核电厂的人员可靠性分析：解释性研究

IF 2.6 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Technology

Pub Date : 2025-02-01 DOI: 10.1016/j.net.2024.08.062

Jinkyun Park , Wasin Vechgama , Seung Ki Shin

Operating teams consisting of several team members still play a critical role in coping with off-normal conditions in socio-technical systems. Thus, various kinds of human reliability analysis methods have been suggested based on the consideration of diverse performance shaping factors that can affect the performance of team members. Unfortunately, since multiple performance shaping factors can vary across operating teams (i.e., crew-to-crew variability), it is crucial to figure out how to visualize this variability in a systematic way. In this regard, comparing the cultural characteristics of operating teams with their performance would be a good starting point. This study investigates how cultural characteristics can be correlated with the occurrence of unsafe acts based on empirical data collected from operating teams working in the main control room of Korean domestic nuclear power plants. The cultural characteristics of the operating teams were visualized using five Hofstede's cultural indices and compared with the number of unsafe acts observed from simulated off-normal conditions. As a result, a statistically significant correlation is found between the occurrence of unsafe acts and one of the Hofstede's indices. From this finding, it is expected that a relevant probe to scrutinize crew-to-crew variability could be soundly determined in future works.

在社会技术系统中，由多名团队成员组成的操作团队在应对非正常情况时仍然发挥着至关重要的作用。因此，考虑到可能影响团队成员绩效的各种绩效影响因素，人们提出了各种人类可靠性分析方法。遗憾的是，由于多种影响绩效的因素在不同的操作团队中可能存在差异（即机组间的可变性），因此如何系统地将这种可变性可视化至关重要。在这方面，将运营团队的文化特征与其绩效进行比较将是一个很好的起点。本研究基于从韩国国内核电站主控室操作团队收集到的经验数据，探讨了文化特征如何与不安全行为的发生相关联。使用霍夫斯泰德的五个文化指数对操作团队的文化特征进行了可视化，并将其与模拟非正常条件下观察到的不安全行为数量进行了比较。结果发现，不安全行为的发生与其中一个霍夫斯泰德指数之间存在统计学意义上的显著相关性。根据这一发现，预计在今后的工作中可以合理地确定一个相关的探究方法，以仔细研究机组人员之间的差异性。

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引用次数: 0