Nuclear Engineering and Technology最新文献_第6页

IVR-UNED: Interactive virtual environments to understand radiation fields

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.2024.09.002

Mario Belotti, Rafael Juárez

As a result of the evolution of High-Performance Computing (HPC) and new cutting-edge projects such as ITER, nuclear analysis has greatly increased in sophistication. Nowadays, nuclear facilities can be modelled in unprecedented detail for radiation transport calculations. Radiation maps can reach great levels of complexity, including multiple radiation sources in vast geometries. These capabilities must be accompanied by an equal capacity to process the results obtained. Nowadays clients are provided with static views pre-decided by nuclear analysts to understand radiation fields. Since the ability to understand such information depends on the unevenly distributed spatial intelligence, this practice can induce biases and limit the usability of the calculations. But beyond analyst-client communication, analysts themselves often fail to identify cleanly all the aspects of a complex radiation field. To overcome to these limitations, we have expanded the videogame engine Unity to create IVR-UNED. It permits to build 3D videogame-like interactive virtual immersive environments, boosting the visualization and insight of the radiation fields through easy on-demand and real-time radiation field postprocessing and visualization. To demonstrate its features, the application to two relevant examples for fusion-related facilities, ITER and IFMIF-DONES, will be presented.

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引用次数: 0

An innovative and efficient implementation of matrix-free Newton krylov method for neutronics/thermal-hydraulics coupling simulation 用于中子/热-水耦合模拟的无矩阵牛顿克雷洛夫方法的创新和高效实施

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.2024.09.012

Peijun Li , Chen Hao , Ning Xu , Yanling Zhu , Yizhen Wang , Zhigang Zhang

The core physical behavior of reactors is essentially the result of multi-physical fields coupling feedback. High-fidelity neutronics/thermal-hydraulics (N/TH) analysis can simulate and predict nuclear reactor core phenomena realistically, providing advanced and reliable technical means during the design and safety analysis of nuclear reactor. In this work, an efficient and robustness coupling method using power density as the coupling parameter, Matrix-Free Newton Krylov (MFNK) method, is successfully developed and innovatively implemented in HNET for high-fidelity N/TH coupling simulation. To enhance the efficiency and stability, the multi-level generalized equivalence theory-based CMFD (ML-gCMFD) iterative acceleration method and ML-gCMFD coupling acceleration method are proposed. In addition, the nonlinear preconditioning and hybrid perturbation size formula are implemented to further improve the convergence. Finally, to evaluate the numerical accuracy, convergence, efficiency and stability of MFNK method, a series of representative problems, including a three-dimensional (3D) single fuel pin problem, VERA Benchmark Problem 6, and VERA Benchmark Problem 7, are analyzed by comparing with the current N/TH coupling methods. Numerical results indicate that MFNK method can obtain strong stability, high convergence performance, and relatively high computational efficiency while ensuring high accuracy. It demonstrates that MFNK method has significant performance advantages and potential for high-fidelity N/TH coupling simulation.

反应堆堆芯物理行为本质上是多物理场耦合反馈的结果。高保真中子/热工水力（N/TH）分析可以真实地模拟和预测核反应堆堆芯现象，为核反应堆设计和安全分析提供先进可靠的技术手段。本研究成功开发了一种以功率密度为耦合参数的高效稳健耦合方法--无矩阵牛顿克雷洛夫（MFNK）方法，并在 HNET 中创新性地实现了高保真 N/TH 耦合模拟。为了提高效率和稳定性，提出了基于广义等效理论的多层次 CMFD（ML-gCMFD）迭代加速方法和 ML-gCMFD 耦合加速方法。此外，还采用了非线性预处理和混合扰动大小公式来进一步提高收敛性。最后，为了评价 MFNK 方法的数值精度、收敛性、效率和稳定性，分析了一系列具有代表性的问题，包括三维（3D）单燃料针问题、VERA 基准问题 6 和 VERA 基准问题 7，并与当前的 N/TH 耦合方法进行了比较。数值结果表明，MFNK 方法在保证高精度的同时，还能获得较强的稳定性、较高的收敛性能和较高的计算效率。这表明 MFNK 方法在高保真 N/TH 耦合仿真方面具有显著的性能优势和潜力。

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引用次数: 0

Control strategy for the core power in an accelerator drive sub-critical system 加速器驱动亚临界系统中的核心功率控制策略

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.2024.08.066

Xinxin Li , Yuan He , Wenjing Ma , Wenjuan Cui , Zhiyong He , Detai Zhou , Hai Zheng , Feng Yang , Yuhui Guo , Haihua Niu , Kai Yin , Shiwu Dang

This paper reports the control strategy for the core power in an accelerator drive sub-critical (ADS) system. In an ADS system, the intense external neutron source provided by a proton accelerator coupled to a spallation target is used to drive a sub-critical reactor. The proposed control strategy is to control the reactor power by adjusting the proton beam power, where the beam power is adjusted by changing either the duty factor or the intensity of the proton beam. As an example, the reactor power control of the China initiative Accelerator Driven System (CiADS) facility has been studied by adjusting the beam power. Firstly, the beam power is set roughly by assigning a new duty factor, where the duty factor is set by changing the beam macro-pulse length and the pulse repetition rate of the proton beam. Both the pulse length and the repetition rate are assigned by a timing system. Secondly, the power is adjusted precisely by changing the beam intensity. To change continuously the beam intensity, an adjustable aperture is used to block the outer particles of the beam line in the accelerator. In order to evaluate the proposed control strategy, a CiADS core model is built based on the multi-node point reactor dynamics model. Three cases, the start of the facility, the decrease of core power and the increase of core power, have been simulated with the model. The simulation results indicate that the control strategy for the core power by changing either the duty factor or the intensity of the proton beam works very well during the operation of the facility.

本文报告了加速器驱动亚临界（ADS）系统中堆芯功率的控制策略。在 ADS 系统中，由质子加速器提供的高强度外部中子源耦合到溅射靶上，用于驱动亚临界反应堆。所提出的控制策略是通过调整质子束功率来控制反应堆功率，其中质子束功率是通过改变质子束的占空比或强度来调整的。以中国主动加速器驱动系统（CiADS）设施为例，研究了通过调整质子束功率来控制反应堆功率的问题。首先，通过分配一个新的占空比来粗略地设置质子束功率，其中占空比是通过改变质子束的宏观脉冲长度和脉冲重复率来设置的。脉冲长度和重复率均由定时系统分配。其次，通过改变光束强度来精确调节功率。为了连续改变束流强度，加速器中使用了一个可调光圈来阻挡束流线的外部粒子。为了评估所提出的控制策略，我们在多节点点反应堆动力学模型的基础上建立了 CiADS 核心模型。该模型模拟了设施启动、堆芯功率下降和堆芯功率上升三种情况。模拟结果表明，在设施运行期间，通过改变占空比或质子束强度来控制堆芯功率的策略效果非常好。

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引用次数: 0

Performance simulation study of plastic scintillator arrays for stability monitoring applications

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

Nuclear Engineering and Technology

Pub Date : 2025-01-01 DOI: 10.1016/j.net.2024.08.037

Z. Chen , Z.W. Pan , Z. Wang , Z.Y. He , Z.B. Lin , T.Y. Yang , Y. Wang , Z.Y. Zhang , J.D. Liu , S.B. Liu , B.J. Ye , L.W. Chen , Y.H. Yu

The characteristics of high energy cosmic-ray muons make them strongly penetrative in materials. It has great potential to monitor the stability of buildings by reconstructing the tracks of cosmic-ray muons. In this work, a buildings stability monitoring system based on plastic scintillator strips is proposed. The monitoring system consists of a single-plane upper telescope to record the incoming track of a muon, and a three-plane lower telescope to track its outgoing trajectory. Every detector plane is formed by orthogonally placed scintillator strips. The impacts of the scintillator geometry and spatial placement on the stability monitoring precision are carefully studied using Geant4 simulations. A sub-millimeter position resolution is achievable by selecting a proper data taking time. The triangular scintillator strips always outperform rectangular ones of the same size and placement. Two empirical formulas have been derived to quantitatively estimate the position resolution with respect to the abovementioned influential factors. In addition, the comprehensive effects of multiple parameters are also studied. Accordingly, triangular scintillator strips will be utilized in the construction of a real monitoring system with better performance and less detection channels. The simulation study in this work can provide good guidance for the manufacturing of triangular strips and their experimental configurations.

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引用次数: 0

Conceptual safeguards method proposal for milling facilities based on nuclear isotopic ratios in uranium mill tailings 基于铀矿厂尾矿中核同位素比率的制粉设施概念保障方法提案

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

Nuclear Engineering and Technology

Pub Date : 2025-01-01 DOI: 10.1016/j.net.2024.08.025

Danwoo Ko , Seunguk Cheon , Jiyoung Kim , Seungmin Lee , Seung Min Woo

There is an argument that the Comprehensive Safeguards Agreement (CSA) should be applied to the milling process, which is the first stage of nuclear fuel cycle, to reduce the possibility of nuclear proliferation. Therefore, this study aims to propose new and conceptual safeguards applicable to milling facilities and design Nuclear Material Accountancy for its application. Mill tailings reach a secular equilibrium state. While approaching this equilibrium state, the ratio of each isotope changes. First, the Bateman equation was used to analyze this phenomenon and evaluate the feasibility of utilizing the specific isotope ratio to track undeclared nuclear activities. Second, the gamma spectrum analysis of mill tailings was conducted using the Monte Carlo N-Particle Code to validate undeclared nuclear activities. This study shows that the ratio of U-235/Th-234 and U-235/Pa-234m can estimate the production time within a year. Furthermore, gamma spectrum analysis of mill tailings revealed visible differences in the low-energy region due to the decay of Th-234. Finally, a conceptual Material Balance Area, Key Measurement Points, and Material Balance Period for milling facilities were designed to apply CSA. It is anticipated that applying CSA to milling facilities could enhance not only the nuclear fuel cycles but also nuclear non-proliferation system.

有一种观点认为，《全面保障监督协定》（CSA）应适用于作为核燃料循环第一阶段的制粉过程，以减少核扩散的可能性。因此，本研究旨在提出适用于制粉设施的新概念保障措施，并为其应用设计核材料衡算。制粉尾矿会达到一个世俗平衡状态。在接近这一平衡状态时，每种同位素的比例都会发生变化。首先，利用贝特曼方程分析这一现象，并评估利用特定同位素比值追踪未申报核活动的可行性。其次，利用蒙特卡洛 N 粒子代码对选矿厂尾矿进行伽马能谱分析，以验证未申报的核活动。这项研究表明，铀 235/Th 234 和铀 235/Pa 234m 的比值可以估算出一年内的生产时间。此外，对选矿厂尾矿的伽马能谱分析显示，由于 Th-234 的衰变，低能区存在明显差异。最后，为应用 CSA 设计了制粉设施的概念性物料平衡区、关键测量点和物料平衡期。预计将 CSA 应用于制粉设施不仅能加强核燃料循环，还能加强核不扩散系统。

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引用次数: 0

Unveiling the potential of Nd2O3 in optimizing the radiation shielding performance of B2O3–TiO2–BaO–ZnO-Nd2O3 glasses 揭示 Nd2O3 在优化 B2O3-TiO2-BaO-ZnO-Nd2O3 玻璃辐射屏蔽性能方面的潜力

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

Nuclear Engineering and Technology

Pub Date : 2025-01-01 DOI: 10.1016/j.net.2024.08.004

Maryam Al Huwayz , Aljawhara H. Almuqrin , F.F. Alharbi , M.I. Sayyed , B. Albarzan

This research focuses on the preparation of a new glass system designed specifically for applications in radiation shielding materials. These glasses are based on the general formula (56-x)B₂O₃–10TiO₂–8BaO–27ZnO-(x-1)Nd₂O₃, where x takes the values of 2, 4, 6 and 8 mol%. For the examination of the designed glasses' radiation attenuation performance, Phy-X software was used, which is a useful approach for predicting the linear attenuation coefficient (LAC), the half value layer (HVL), and effective atomic number. The LAC decreases from 1.489 cm⁻¹ to 0.551 cm⁻¹ for the glass with x = 1 mol%, while the glass with 7 mol% Nd₂O₃ saw a decrease in the LAC from 2.483 cm⁻¹ to 0.718 cm⁻¹. Introducing Nd₂O₃ increases the glasses' LAC, suggesting enhanced radiation shielding performance. Also, Nd₂O₃ addition influences the HVL within the glasses, with higher content reducing the HVL. At 0.122 MeV, the HVL and tenth value layer (TVL) are 0.456 and 1.546 cm, respectively. At 0.245 MeV, the TVL is about 3.32 times higher than the HVL. The lowest mean free path (MFP) is found at 0.122 MeV, which varies between 0.672 cm for Nd1 and 0.403 cm for Nd4.

这项研究的重点是制备一种专门用于辐射屏蔽材料的新型玻璃系统。这些玻璃的通式为(56-x)BO-10TiO-8BaO-27ZnO-(x-1)NdO，其中 x 的值为 2、4、6 和 8 摩尔%。为了检测所设计玻璃的辐射衰减性能，使用了 Phy-X 软件，该软件是预测线性衰减系数（LAC）、半值层（HVL）和有效原子序数的有效方法。x = 1 摩尔%的玻璃的线性衰减系数从 1.489 厘米降至 0.551 厘米，而含有 7 摩尔% NdO 的玻璃的线性衰减系数则从 2.483 厘米降至 0.718 厘米。引入氧化钕可增加玻璃的 LAC，表明辐射屏蔽性能增强。此外，NdO 的添加也会影响玻璃内部的 HVL，含量越高，HVL 越低。在 0.122 MeV 时，HVL 和十值层（TVL）分别为 0.456 厘米和 1.546 厘米。在 0.245 MeV 时，TVL 约为 HVL 的 3.32 倍。在 0.122 MeV 时，平均自由路径（MFP）最低，钕 1 为 0.672 厘米，钕 4 为 0.403 厘米。

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引用次数: 0

Corrigendum to “A study on the applicability of simplified few-group GET (Generalized Equivalence Theory) to cylindrical molten salt fast reactor” [Nucl. Eng. Technol. 56 (10) (2024) 4207–4218]

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

Nuclear Engineering and Technology

Pub Date : 2025-01-01 DOI: 10.1016/j.net.2024.11.035

Sungtaek Hong , Taesuk Oh , Yonghee Kim

引用次数: 0

High-power performance studies of an S-band high-gradient accelerating cavity for medical applications 医疗应用 S 波段高梯度加速腔的大功率性能研究

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

Nuclear Engineering and Technology

Pub Date : 2025-01-01 DOI: 10.1016/j.net.2024.08.033

P. Martinez-Reviriego , N. Fuster-Martínez , D. Esperante , M. Boronat , B. Gimeno , C. Blanch , D. González-Iglesias , P. Martín-Luna , E. Martínez , A. Menendez , L. Pedraza , J. Fernández , J. Fuster , A. Grudiev , N. Catalan Lasheras , W. Wuensch

High-Gradient accelerating cavities are one of the main research lines in the development of compact linear accelerators. However, the operation of such accelerating cavities is currently limited by non-linear electromagnetic effects that are intensified at high electric fields, such as RF breakdowns, dark currents and radiation. A novel normal-conducting High Gradient S-band Backward Travelling Wave accelerating cavity for medical application (v = 0.38c) has been designed and constructed at CERN with a design gradient of 50 MV/m. In this paper, the high-power performance studies of this novel design carried out at the IFIC high-power laboratory are presented, as well as the analysis of the conditioning parameters in combination with numerical simulations.

高梯度加速腔是开发紧凑型直线加速器的主要研究方向之一。然而，这种加速腔的运行目前受到非线性电磁效应的限制，这些效应在高电场下会加剧，如射频击穿、暗电流和辐射。欧洲核子研究中心（CERN）设计并建造了一个用于医疗应用的新型常导高梯度 S 波段后向游波加速腔（v = 0.38c），其设计梯度为 50 MV/m。本文介绍了在 IFIC 高功率实验室对这种新型设计进行的高功率性能研究，以及结合数值模拟对调节参数进行的分析。

引用次数: 0

Statistical sampling method to verify the homogeneity of full-scale cement-solidified radioactive waste 验证全尺寸水泥固化放射性废物均匀性的统计取样方法

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

Nuclear Engineering and Technology

Pub Date : 2025-01-01 DOI: 10.1016/j.net.2024.08.020

Hyeongjin Byeon, Ugyu Jeong, Jaeyeong Park

Homogeneity is an important factor for ensuring the structural stability of solidified radioactive waste, and the most effective approach for assessing its homogeneity is by performing compressive strength measurements using the minimum amount of coring specimens. The efficiency of detecting inhomogeneous waste is affected by the coring position and number of coring positions. However, no guidelines exist for coring solidified waste for compressive-strength tests. Therefore, this study compared uniform, random, and quasi-Monte Carlo sampling methods to determine the most effective core position. Further, the effects of different sampling amounts on the detection rate of inhomogeneous solidified waste were observed, and the detection rate of the inhomogeneous waste was obtained by modeling the coring procedure of solidified radioactive waste using MATLAB. Thus, a sampling method and a method for increasing the specimen amount, both of which can efficiently detect inhomogeneous waste during compressive strength tests, were presented in this paper. The results of this study can be applied as background data for developing homogeneity assessment guidelines for solidified radioactive waste.

均匀性是确保固化放射性废物结构稳定性的一个重要因素，评估其均匀性的最有效方法是使用最少的取芯试样进行抗压强度测量。检测不均匀废料的效率受到取芯位置和取芯数量的影响。然而，目前还没有关于在压缩强度测试中对固化废物取芯的指南。因此，本研究比较了均匀、随机和准蒙特卡洛取样方法，以确定最有效的取芯位置。此外，还观察了不同取样量对非均质固化废物检测率的影响，并通过使用 MATLAB 对固化放射性废物取芯过程进行建模，得出了非均质废物的检测率。因此，本文提出了一种取样方法和一种增加试样量的方法，这两种方法都能在抗压强度试验中有效地检测出不均匀废物。本研究的结果可作为制定固化放射性废物均匀性评估准则的背景数据。

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引用次数: 0

Exploring helium retention from technical materials: Development and investigation 探索技术材料中的氦保留：开发与研究

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

Nuclear Engineering and Technology

Pub Date : 2025-01-01 DOI: 10.1016/j.net.2024.08.005

Andrew K. Gillespie, Cuikun Lin, Django Jones, Sandeep Puri, R.V. Duncan

Materials used to study nuclear fusion can retain atmospheric helium unless pretreated before an experiment. Understanding helium outgassing is important for accurate diagnostics in experiments surrounding nuclear fusion. The presence of helium is often cited as the primary evidence that a nuclear reaction has occurred, so it is imperative that known sources of helium are mitigated prior to proceeding with novel nuclear experiments. It is also necessary to ensure hermeticity when transferring gas aliquots from an experiment to a mass spectrometer. In this article, we present studies of helium leak rates in systems used in novel nuclear experiments. We also present studies of helium retention in materials subjected to various heating profiles and atmospheric concentrations. Without pretreatment, 12-inch lengths of both 3/8” diameter tubes and 1/2″ diameter stainless-steel 316 tubing yielded an average areal outgassing amount of 0.64 pmol/cm². If pretreatment is impractical, then the results may be scaled based on the tubing length necessary for constructing custom experimental equipment. It also may reabsorb ⁴He from the atmosphere in time. These studies also demonstrate that it is necessary to pretreat most materials prior to performing experiments where the presence of ⁴He is being used as an indicator for novel nuclear reactions.

用于研究核聚变的材料会保留大气中的氦，除非在实验前进行预处理。了解氦放气对于核聚变实验的准确诊断非常重要。氦气的存在通常被认为是发生核反应的主要证据，因此在进行新的核实验之前，必须减少已知的氦气来源。此外，在将实验中的等分气体转移到质谱仪时，也有必要确保密封性。在这篇文章中，我们介绍了对新型核实验所用系统中氦泄漏率的研究。我们还介绍了在不同加热曲线和大气浓度条件下材料中氦气保留情况的研究。在没有预处理的情况下，12 英寸长的直径为 3/8 英寸的管子和直径为 1/2 英寸的 316 不锈钢管的平均放气量为 0.64 pmol/cm。如果预处理不可行，则可根据定制实验设备所需的管子长度来调整结果。它还可以及时从大气中重新吸收 He。这些研究还表明，在进行以 He 的存在作为新型核反应指标的实验之前，有必要对大多数材料进行预处理。

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引用次数: 0