Annals of Nuclear Energy最新文献_第2页

Calculation of the multiplicity moments in nuclear safeguards with forward transport theory 利用前向传输理论计算核保障措施中的多重性时刻

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

Annals of Nuclear Energy

Pub Date : 2024-11-23 DOI: 10.1016/j.anucene.2024.111046

Liliane Basso Barichello , Imre Pázsit

The theoretical basis of multiplicity counting of nuclear safeguards lies in the calculation of the factorial moments of the number of neutrons emitted from the item. While the traditional method to derive these moments uses the so-called point model in which the spatial transport of neutrons in the item is neglected, the theoretical framework has recently been re-derived in a one-speed transport model, which is inherently of the backward (adjoint) type. The arising integral equations for the moments were solved numerically with a collision number type (iterated kernel or Neumann-series) expansion. In this paper, we show that effective methods of analytical character, originally developed for direct (forward-type) transport problems, can be associated with the solution of the adjoint-type moment equations. The theory is described, and quantitative results are given for selected representative cases. The accuracy and computational speed of the method is investigated and compared favourably with those of the collision number expansion method. The quantitative results also lend some new insight into the properties of statistics of the multiplicative process for the exiting neutrons.

核保障措施的倍率计数的理论基础在于计算物品发射的中子数的阶乘矩。计算这些阶乘矩的传统方法采用所谓的点模型，其中忽略了中子在物品中的空间传输，而最近的理论框架则是通过单速传输模型重新推导出来的，该模型本质上属于后向（邻接）类型。所产生的矩积分方程通过碰撞数型（迭代核或诺依曼数列）展开进行数值求解。在本文中，我们展示了最初为直接（正向）传输问题开发的有效分析方法，可以与邻接型矩方程的求解联系起来。本文对理论进行了描述，并给出了部分代表性案例的定量结果。对该方法的精度和计算速度进行了研究，并与碰撞数扩展法进行了比较。定量结果还使我们对流出中子的乘法过程的统计特性有了一些新的认识。

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引用次数: 0

The effects of different fuel cladding materials on neutronic behavior and fuel depletion performance in the VVER-1200 reactor 不同燃料包层材料对 VVER-1200 反应堆中子行为和燃料耗竭性能的影响

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

Annals of Nuclear Energy

Pub Date : 2024-11-22 DOI: 10.1016/j.anucene.2024.111071

Ayhan Kara, Emil Mammadzada

This study investigates the impact of different fuel cladding materials on the performance and safety of VVER-1200 reactors using the Serpent 2 Monte Carlo code. Cladding materials evaluated include Zircaloy, 304SS, 310SS, FeCrAl, APMT, TiC, ZrC, and SiC. Key parameters assessed are fuel performance, neutronic behavior, infinite multiplication factor (kinf), and radioactive fission product levels. Results indicate that Zircaloy, ZrC, and SiC claddings retain criticality longer (kinf > 1) with favorable neutron flux and fission neutron production. TiC, however, loses criticality early and generates high neutron poisons and fission products. Steel alloys (304SS, 310SS), APMT, and FeCrAl demonstrate moderate performance affecting reactor criticality and neutron flux. Overall, Zircaloy is identified as the most effective cladding, balancing criticality, minimizing plutonium buildup, and reducing radioactive fission products, with ZrC and SiC as close competitors.

本研究使用 Serpent 2 Monte Carlo 代码研究了不同燃料包壳材料对 VVER-1200 反应堆性能和安全性的影响。评估的包层材料包括锆合金、304SS、310SS、铁铬铝、APMT、TiC、ZrC 和 SiC。评估的主要参数包括燃料性能、中子行为、无限倍增因子（kinf）和放射性裂变产物水平。结果表明，锆合金、碳化锆和碳化硅包壳保持临界状态的时间更长（kinf >1），中子通量和裂变中子产生量也更多。然而，TiC 很早就失去临界状态，并产生大量中子毒物和裂变产物。钢合金（304SS、310SS）、APMT 和 FeCrAl 在影响反应堆临界度和中子通量方面的性能适中。总体而言，锆合金被认为是最有效的包层，它能平衡临界状态、最大限度地减少钚的积累并减少放射性裂变产物，而 ZrC 和 SiC 则是最接近的竞争对手。

引用次数: 0

Investigation of nuclide inventory of cladding material irradiated in the Goesgen PWR core 调查戈斯根压水堆堆芯中经过辐照的包壳材料核素清单

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

Annals of Nuclear Energy

Pub Date : 2024-11-22 DOI: 10.1016/j.anucene.2024.111061

R. Dagan , T. König , M. Herm , F. Alvarez , E. Dorval , S. Häkkinen , E. Vlassopoulos , A. Shama , A. Smaizys , P. Schillebeeckx

A characterization of the spent nuclear fuel (SNF) for its radionuclide (RN) inventory is vital for various back-end stages of the nuclear fuel cycle. It concerns both the fuel and the metallic (i.e., cladding and structural material) components of the spent fuel assemblies, where different calculation approaches and methods should be deployed for their characterization. This study concentrates on fuel traces and other impurities within the cladding. During the operating cycles, the Zircaloy cladding is exposed to a considerable amount of irradiation. The impact of the exposure should be checked to assure the integrity of the cladding and thus the safety of the stored spent fuel. Within the work package “Spent Nuclear Fuel Characterization and Evolution until Disposal” (SFC) of the EURAD project, dedicated samples were produced, irradiated and the radionuclide inventory of the cladding was analysed and compared. In parallel a blind test was performed, in which different partners used different codes to simulate the irradiation quantity. The blind test showed good agreement between most of the codes, in particular in view of the small amount of the evolved fuel traces. Furthermore, the presence of actinides, caused by precipitation of uranium on the inner surface of the cladding during manufacturing, was found to be negligible in comparison to precipitation of traces of fuel pellets on the cladding during operation. The good agreement between the simulating codes enables to depict further the initial amount of alloying elements of the cladding material itself in a better manner. In particular specific isotopes of cobalt, nickel and iron, which are directly connected to the unique properties of each cladding material can be better identified based on the accurate measuring techniques used in this study.

乏核燃料（SNF）放射性核素（RN）清单的表征对于核燃料循环的各个后端阶段至关重要。它涉及乏燃料组件的燃料和金属（即包层和结构材料）成分，在表征时应采用不同的计算方法和手段。本研究主要针对燃料痕量和包壳内的其他杂质。在运行周期中，锆合金包壳会受到大量辐照。应检查辐照的影响，以确保包壳的完整性，从而保证乏燃料储存的安全。在 EURAD 项目的 "乏核燃料表征和处置前的演变"（SFC）工作包中，制作了专用样品，对其进行了辐照，并分析和比较了包壳的放射性核素清单。同时还进行了盲测，不同的合作伙伴使用不同的代码来模拟辐照量。盲测结果表明，大多数代码之间的一致性都很好，特别是考虑到燃料痕量较小。此外，与运行过程中燃料芯块在包壳上的沉淀相比，制造过程中铀在包壳内表面沉淀所导致的锕系元素的存在可以忽略不计。模拟代码之间的良好一致性使我们能够以更好的方式进一步描述包壳材料本身合金元素的初始含量。特别是与每种包壳材料的独特性质直接相关的钴、镍和铁的特定同位素，可以根据本研究中使用的精确测量技术更好地识别出来。

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引用次数: 0

Application of risk informed safety margin characterization to the analysis of a pressurized water reactor 在压水反应堆分析中应用风险知情安全裕度表征法

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

Annals of Nuclear Energy

Pub Date : 2024-11-22 DOI: 10.1016/j.anucene.2024.111066

Y.M. Chen , D.W. Wu , T.C. Wang , M. Lee

The sequence Core Damage Frequency of a pressurized water reactor has been quantified in three initiating events that are Medium Break Loss of Coolant Accident, Small Break Loss of Coolant Accident and Steam Generator Tube Rupture, following a realistic methodology called risk informed safety margin characteristic. The surrogate plant analyzed in the study is a typical pressurized water reactor. The plant adopted two Westinghouse Three-Loop Pressurized Water Reactors with rated thermal power of 2,830 MWt. The phenomenon identification and ranking table is applied for uncertainty analysis. The mitigation actions as described in plant specific Probabilistic Risk Assessment include cooldown and depressurization, emergency cooldown and depressurization, high head safety injection, high head safety recirculation, low head safety recirculation and Refueling Water Storage Tank replenishment. These mitigation actions are analyzed by thermal hydraulic system analysis code RELAP5-3D to determine the successfulness of the actions. The uncertainties of input parameters of the plant conditions are included, and the time of mitigation action executed is treated as one of the input uncertainties. The results of realistic methodology show a decrease in Core Damage Frequency for all three analyzed events in comparison with conventional methodology. The differences between three initiating events are also discussed.

压水堆的顺序堆芯损坏频率是按照一种称为风险知情安全裕度特征的现实方法，在中型破裂失去冷却剂事故、小型破裂失去冷却剂事故和蒸汽发生器管破裂这三种启动事件中进行量化的。研究中分析的替代电厂是一个典型的压水反应堆。该电厂采用了两台西屋三回路压水堆，额定热功率为 2,830 兆瓦。不确定性分析采用了现象识别和排序表。工厂特定概率风险评估中描述的缓解措施包括冷却和减压、紧急冷却和减压、高水头安全注入、高水头安全再循环、低水头安全再循环和补充燃料水储罐。这些缓解措施通过热液压系统分析代码 RELAP5-3D 进行分析，以确定这些措施是否成功。其中包括电厂条件输入参数的不确定性，以及执行缓解措施的时间作为输入不确定性之一。现实方法的结果显示，与传统方法相比，所有三个分析事件的核心损坏频率都有所下降。此外，还讨论了三个启动事件之间的差异。

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引用次数: 0

Condition monitoring and breakage assessment of steam generator heat transfer tubes in nuclear power plants 核电站蒸汽发生器传热管的状态监测和破损评估

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

Annals of Nuclear Energy

Pub Date : 2024-11-21 DOI: 10.1016/j.anucene.2024.111032

Xue-ying Huang, Hong Xia, Wen-zhe Yin, Yong-kuo Liu

The steam generator (SG) is a critical component of the steam power conversion system in nuclear power plants. The heat transfer tubes of steam generators are susceptible to mechanical and chemical damage due to prolonged exposure to high-temperature, high-pressure environments, and high-radiation media. Timely detection of abnormal states and accurate assessment of the breach degree in the heat transfer tubes are crucial for enhancing the economic and operational safety of nuclear power plants. This study focuses on simulating the normal state of steam generator heat transfer tubes and different degrees of abnormal states using a simulator, while collecting characteristic parameters that can be monitored by sensors. In order to improve the fidelity of the simulated signals to real-world engineering signals, in cases where the breach degree is significant, the reactor undergoes an emergency shutdown, resulting in a smaller amount of effective signal data collected for larger breach degrees. To address these issues, this paper employs the Synthetic Minority Oversampling Technique (SMOTE) to expand the capacity of small sample data. Additionally, to mitigate the impact of high-dimensional feature parameters on subsequent condition monitoring and breach degree assessment, a Denoised AutoEncoder (DAE) is employed to reduce the dimensionality of the feature parameters. The One-Class Support Vector Machine (One-Class SVM) is then utilized to monitor the condition of the steam generator heat transfer tubes. When an abnormality is detected in the heat transfer tubes, a Bi-directional Long Short-Term Memory (Bi-LSTM) model is used to evaluate the magnitude of the tube leakage. The experimental results demonstrate that the developed system achieves a high monitoring accuracy and provides a good assessment of the fault degree.

蒸汽发生器（SG）是核电站蒸汽动力转换系统的关键部件。由于长期暴露在高温、高压环境和高辐射介质中，蒸汽发生器的传热管很容易受到机械和化学损伤。及时发现传热管的异常状态并准确评估其破损程度对于提高核电站的经济性和运行安全性至关重要。本研究的重点是利用模拟器模拟蒸汽发生器传热管的正常状态和不同程度的异常状态，同时收集可由传感器监测的特征参数。为了提高模拟信号与实际工程信号的保真度，在破损程度较大的情况下，反应堆会进行紧急停堆，导致在破损程度较大时收集到的有效信号数据较少。为了解决这些问题，本文采用了合成少数过采样技术（SMOTE）来扩大小样本数据的容量。此外，为了减轻高维特征参数对后续状态监测和破损程度评估的影响，本文采用了去噪自动编码器（DAE）来降低特征参数的维度。然后利用单类支持向量机（One-Class SVM）来监测蒸汽发生器传热管的状态。当检测到传热管出现异常时，就会使用双向长短期记忆（Bi-LSTM）模型来评估传热管泄漏的程度。实验结果表明，所开发的系统实现了较高的监测精度，并能很好地评估故障程度。

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引用次数: 0

Research on application of heterogeneous resonance Integral for double heterogeneous system 双异质系统的异质共振积分应用研究

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

Annals of Nuclear Energy

Pub Date : 2024-11-21 DOI: 10.1016/j.anucene.2024.111051

Shuai Qin , Qian Zhang , Kai Wang , Dong Huang , Song Li , Yuechao Liang

The Double Heterogeneous (DH) system, where fuel particles are randomly dispersed in the non-fissile matrix, is challenging for the reactor physics calculation. The Sanchez-Pomraning method accurately handles the DH system, but integrating it into existing reactor physics code requires code development. This study adopts the Sanchez-Pomraning coupled Ultra-Fine-Group (SP-UFG) slowing-down calculation to generate the heterogeneous Resonance Integral (RI) for DH system treatment with simple volume homogenization. Fully Ceramic Micro-encapsulated (FCM) fuel pin-cells and plates with varying configurations are calculated for verification. Effective cross-sections (XSs) and k_eff calculated by the heterogeneous RI are compared with SP-UFG results. Results show that the maximum bias of XSs and k_eff caused by the XS biases are less than 5% and 200 pcm, respectively. The maximum bias of k_eff when compared with Monte Carlo calculated results is −213 pcm, demonstrating that only considering the DH effect in the resonance energy region is acceptable.

双异质（DH）系统，即燃料颗粒随机分散在非易裂变基质中，对反应堆物理计算具有挑战性。Sanchez-Pomraning 方法能准确处理 DH 系统，但将其集成到现有反应堆物理代码中需要代码开发。本研究采用桑切斯-波姆兰宁耦合超细群（SP-UFG）减速计算，生成异质共振积分（RI），用于 DH 系统的简单体积均质化处理。对不同配置的全陶瓷微胶囊（FCM）燃料针形电池和板进行了计算验证。将异质 RI 计算出的有效截面 (XS) 和 keff 与 SP-UFG 结果进行了比较。结果表明，XSs 和 keff 的最大偏差分别小于 5%和 200 pcm。与蒙特卡洛计算结果相比，keff 的最大偏差为 -213 pcm，这表明只考虑共振能量区域的 DH 效应是可以接受的。

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引用次数: 0

Machine learning-assisted correlations for prediction of fission gas fractions and hydrogen concentration in VVER-1000 fuel 机器学习辅助关联预测 VVER-1000 燃料中的裂变气体分数和氢浓度

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

Annals of Nuclear Energy

Pub Date : 2024-11-21 DOI: 10.1016/j.anucene.2024.111073

Yalcin Ilteris Kaan , Khashayar Sadeghi , Seyed Hadi Ghazaie , Ekaterina Sokolova , Victor Modestov , Vitaly Sergeev , Puzhen Gao

This study aims to develop correlations for predicting the fission gas fractions and hydrogen gas concentration during a fuel cycle, using gene expression programming as an evolutionary machine learning approach. The well-known FRAPCON code is used for generating a straightforward dataset under steady-state conditions. The two-step sensitivity analysis is carried out to identify the most influential parameters for correlation development. Wilks’ statistical method is used to generate 59 scenarios to distribute input parameter uncertainties evenly, which leads to a confidence level of 95 %. The mean squared error for xenon, krypton, and helium is 0, while hydrogen exhibited a value of 59.36 since fraction values are between 0 and 1 and concentration ranged from 5 PPM to 200 PPM.

R^{2}

values exceeded 0.97, indicating strong correlation accuracy. The high accuracy achieved from the correlations demonstrates that selecting a 59-sample dataset based on Wilk’s method is sufficient to obtain accuracy exceeding 95 %.

本研究旨在利用基因表达编程作为进化机器学习方法，开发用于预测燃料循环过程中裂变气体分数和氢气浓度的相关性。使用著名的 FRAPCON 代码生成稳态条件下的直接数据集。通过两步敏感性分析，确定对相关性发展最有影响的参数。使用 Wilks 统计方法生成 59 种情景，均匀分布输入参数的不确定性，从而得出 95 % 的置信度。氙、氪和氦的均方误差为 0，而氢的均方误差为 59.36，因为分数值在 0 到 1 之间，浓度范围在 5 PPM 到 200 PPM 之间。R2 值超过 0.97，表明相关精度很高。相关性达到的高精确度表明，根据 Wilk 方法选择 59 个样本数据集足以获得超过 95 % 的精确度。

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引用次数: 0

Designing a combined guide and investigating the role of collimator in neutron intensity uniformity 设计组合导轨并研究准直器在中子强度均匀性中的作用

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

Annals of Nuclear Energy

Pub Date : 2024-11-21 DOI: 10.1016/j.anucene.2024.111064

E. Tayebfard, M. Shayesteh, R. Razavi, M. Eshghi

The present study investigates the effect of some geometrical parameters on the performance of the neutron guide system, including the length of the collimator and the angle between successive guides. This study investigates the use of optical components to remove fast neutrons in straight and angled guides, and their role in focusing the neutron beam and increasing the intensity uniformity on the sample. In this research, the McStas code, which was developed with the Monte Carlo method, was used. The effect of the length of the collimator on the creation of a neutron beam with a certain energy, and the uniformity of the intensity of the beam at the sample location have been investigated. For angled guides, the intensity of the neutron beam and its uniformity at the sample location, in terms of the angle of the guides with each other, were calculated using McStas and Vitess codes and the results were compared with each other. Calculations have been done using an americium-beryllium source with an intensity of about 5 Curie. The results show that the combination of two convergent and divergent collimators have greater ability to uniform the neutron intensity than the linear collimator.

本研究调查了一些几何参数对中子导向系统性能的影响，包括准直器的长度和连续导向器之间的角度。本研究调查了在直导和斜导中使用光学元件去除快中子的情况，以及它们在聚焦中子束和提高样品上的强度均匀性方面的作用。在这项研究中，使用了用蒙特卡罗方法开发的 McStas 代码。研究了准直器的长度对产生一定能量的中子束的影响，以及样品位置上中子束强度的均匀性。对于有角度的准直器，使用 McStas 和 Vitess 代码计算了中子束的强度及其在样品位置的均匀性，并将结果进行了比较。计算是使用强度约为 5 居里的镅铍源进行的。结果表明，与线性准直器相比，两个会聚和发散准直器的组合具有更强的均匀中子强度的能力。

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引用次数: 0

Development of a decommissioning waste assessment module for heavy water reactors 开发重水反应堆退役废物评估模块

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

Annals of Nuclear Energy

Pub Date : 2024-11-21 DOI: 10.1016/j.anucene.2024.111001

Hyun Young Shin, Kyu Tae Park, Chan Hee Park, Sung Ryul Kim, Jung Min Oh, Ba Ro Lee, Ji Ung Kim, Hye Jin Kim, So Yun Jeong, Wook Jae Yoo

In this study, we developed a Decommissioning Waste Assessment Module (DWAM) for Heavy Water Reactors (HWRs). To evaluate the waste amount of radioactive structures that make up a significant portion of decommissioning waste, we establish a comprehensive DB containing detailed information on the volume, weight, material type, and radioactivity of each decommissioning waste segment. Decommissioning waste assessment algorithms were developed and applied in the proposed module to optimize the calculated results, and their performance was evaluated to ensure accuracy and reliability. The DWAM currently being developed for the Wolsong Unit 1 commercial pressurized heavy water reactor is a technology that has the potential to be utilized not only for domestic reactors but also for reactors worldwide. In addition, the technology is expected to contribute to the development of technology related to the decommissioning of both radioactive and non-radioactive structures.

在本研究中，我们开发了重水反应堆退役废物评估模块（DWAM）。为了评估占退役废物重要部分的放射性结构的废物量，我们建立了一个综合数据库，其中包含每个退役废物部分的体积、重量、材料类型和放射性的详细信息。我们开发了退役废物评估算法，并将其应用于拟议模块中，以优化计算结果，并对其性能进行了评估，以确保准确性和可靠性。目前正在为卧龙一号机组商用压水重水反应堆开发的 DWAM 是一项不仅有可能用于国内反应堆，而且有可能用于全球反应堆的技术。此外，预计该技术将有助于开发与放射性和非放射性结构退役有关的技术。

引用次数: 0

Optimization study of PSF weighting affecting operators’ organizational performance in digital NPPs 影响数字核电站运营商组织绩效的 PSF 权重优化研究

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

Annals of Nuclear Energy

Pub Date : 2024-11-20 DOI: 10.1016/j.anucene.2024.111033

Qian Hu, Yongle Zheng, Yanqi Liu

The organizational performance of main control room (MCR) operators is crucial in digital nuclear power plants (NPPs). A comprehensive study determined the weights of performance-shaping factors (PSFs) influencing operators’ organizational performance. Firstly, a questionnaire survey was conducted, and an interval judgment matrix was innovatively established based on the expectation and standard deviation of statistical data acquired. Next, interval hierarchical analysis (IHA) was employed to derive reasonable intervals for each PSF weight. Finally, a hierarchical interval weighting optimization mathematical model was established for high-precision weighting calculation. This method offers a significant reference for optimizing the PSF weighting process. The approach presented in this paper is highly adaptable and provides new technical and theoretical support for the study of weighting theory.

在数字化核电站（NPP）中，主控室（MCR）操作员的组织绩效至关重要。一项综合研究确定了影响操作员组织绩效的绩效塑造因素（PSF）的权重。首先，进行了问卷调查，并根据所获统计数据的期望值和标准差，创新性地建立了区间判断矩阵。其次，采用区间层次分析法（IHA）得出了每个 PSF 权重的合理区间。最后，建立了分层区间权重优化数学模型，用于高精度权重计算。这种方法为优化 PSF 加权过程提供了重要参考。本文提出的方法具有很强的适应性，为权重理论研究提供了新的技术和理论支持。

引用次数: 0