Progress in Nuclear Energy最新文献_第3页

Non-linear cascade control with gain-scheduling and startup control strategy study for thermionic space reactor TOPAZ-II 热离子空间反应堆 TOPAZ-II 的增益调度非线性级联控制和启动控制策略研究

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

Progress in Nuclear Energy

Pub Date : 2024-10-26 DOI: 10.1016/j.pnucene.2024.105514

Zongyun Wu, Tiancai Liu, Yufeng Lyu, Chunqiu Guo, Lin Sun

The TOPAZ-II thermionic space reactor system, which was designed by the Soviet Union, is characterized by its high degree of nonlinearity and positive temperature reactivity feedback. The thermionic space reactor exhibits characteristics of high inertia and significant delay in controlling its electrical power and outlet temperature. The simple PID controller is difficult to achieve good performance. To carry out controller design for thermionic space reactor, the simulation platform for thermionic space reactor is developed based on coupling between reactor system thermal-hydraulic code RESYS and control system simulator in this study. After that, based on the cascade control strategy and gain-scheduling, the reactor thermal controller, electric power controller, outlet temperature controller applicable for the full power range is designed with transfer function model and frequency domain analysis method. To validate the nonlinear electric power controller performance, the continuous minor step disturbances, major step disturbances, and ramp variation of electric power setpoint is simulated. The performance of outlet temperature controller is verified with the simulation result of step and ramp variation of outlet temperature setpoint. Thereafter, the start-up process of the thermionic space reactor TOPAZ-II is simulated and analyzed. The simulation result reveals that the controller designed in this paper can overcome the nonlinearity of the thermionic space reactor system and has good performance throughout the entire power range. Compared to traditional simple PID controller, the cascade controller has better performance and can achieve good control performance even in situations where simple PID controllers cannot function properly.

前苏联设计的 TOPAZ-II 热离子空间反应堆系统具有高度非线性和正温度反应反馈的特点。热离子空间反应堆在控制其电功率和出口温度方面具有高惯性和显著延迟的特点。简单的 PID 控制器难以实现良好的性能。为了进行热离子空间堆的控制器设计，本研究基于反应堆系统热-液压代码 RESYS 和控制系统模拟器的耦合，开发了热离子空间堆的模拟平台。然后，基于级联控制策略和增益调度，利用传递函数模型和频域分析方法设计了适用于全功率范围的反应堆热控制器、电功率控制器和出口温度控制器。为了验证非线性电力控制器的性能，模拟了电力设定点的连续小阶跃扰动、大阶跃扰动和斜坡变化。出口温度控制器的性能通过出口温度设定点的阶跃和斜坡变化仿真结果得到验证。随后，对热离子空间反应堆 TOPAZ-II 的启动过程进行了模拟和分析。仿真结果表明，本文设计的控制器能克服热离子空间堆系统的非线性问题，在整个功率范围内性能良好。与传统的简单 PID 控制器相比，级联控制器具有更好的性能，即使在简单 PID 控制器无法正常工作的情况下也能实现良好的控制性能。

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引用次数: 0

Efficacy of advanced concretes for attenuation of ionizing radiations: A comprehensive review and comparison 高级混凝土在衰减电离辐射方面的功效：全面审查和比较

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

Progress in Nuclear Energy

Pub Date : 2024-10-22 DOI: 10.1016/j.pnucene.2024.105502

Rajni Devi , Poonamjot , Mohinder Singh , Amandeep Sharma

Radiation shielding materials are key components to suppress the hazardous effects of ionizing radiation, especially energetic gamma rays and penetrative neutrons. This review includes the most ancient building material, concrete, in its different composition obtained by introducing a variety of additives, aggregates and nanomaterials in its conventional form. The objective of present work is to critically review and compare the variety of concretes, reported through various experimental and computational methods, so that the best composition among diverse concrete categories can be highlighted. For this purpose the essential shielding parameters namely mass attenuation coefficient (MAC) and half value layer (HVL) at useful gamma energies of 662, 1173 and 1332 keV have been compared graphically. The protective shielding concretes against neutrons have also been studied through a plot corresponding to Fast Neutron Removal Cross-sections (FNRCS) data of different concretes. Lastly, shielding competency of granite and pyroclastic rock samples for light and heavy charged particles have been included by taking into consideration the interaction parameters namely mass stopping power and projected range. Apart from this, numerous advanced applications of radiation shielding concretes, proper utilization of different forms of waste in concrete mix and few shortcomings of concrete specimens are also listed in this review paper. From the comparative plots of various concretes it is concluded that marble based concretes are best for gamma ray attenuation and nanomaterials based compositions are top if lesser thickness is to employ for attenuation. On the basis of acquired knowledge from literature, the present work will highlight the future perspectives of concretes as shielding materials and would be quite helpful for the selection of appropriate compositions by the community interacting directly or indirectly with ionizing radiations.

辐射屏蔽材料是抑制电离辐射（尤其是高能伽马射线和穿透性中子）有害影响的关键部件。本综述包括最古老的建筑材料--混凝土，它通过引入各种添加剂、集料和纳米材料，以传统形式获得不同的成分。本研究的目的是对通过各种实验和计算方法报告的各种混凝土进行批判性审查和比较，从而突出不同混凝土类别中的最佳成分。为此，以图表形式比较了在 662、1173 和 1332 千伏有用伽马能量下的基本屏蔽参数，即质量衰减系数（MAC）和半值层（HVL）。此外，还通过与不同混凝土的快速中子去除截面（FNRCS）数据相对应的曲线图，研究了屏蔽混凝土对中子的防护能力。最后，考虑到相互作用参数，即质量停止力和投射范围，研究了花岗岩和火成岩样品对轻重带电粒子的屏蔽能力。除此之外，本文还列举了辐射屏蔽混凝土的许多先进应用、混凝土混合物中不同形式废物的合理利用以及混凝土试样的一些不足之处。从各种混凝土的对比图中可以得出结论，大理石基混凝土最适合用于伽马射线衰减，而纳米材料基混凝土则最适合用于厚度较小的衰减。根据从文献中获得的知识，本研究将突出混凝土作为屏蔽材料的未来前景，并将对直接或间接接触电离辐射的群体选择适当的成分大有帮助。

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引用次数: 0

Calculation acceleration for fuel cycle simulation of molten salt reactor based on multi-group cross section method 基于多组截面法的熔盐反应堆燃料循环模拟计算加速系统

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

Progress in Nuclear Energy

Pub Date : 2024-10-22 DOI: 10.1016/j.pnucene.2024.105505

Zhenghao Xu , Guifeng Zhu , Shuyang Jia , Yafen Liu , Changqing Yu , Yuhan Fan , Yu Zhong , Rui Yan , Yang Zou , Hongjie Xu

The time-consuming issue of transport calculations is prominent in the burnup calculation of nuclear reactor. Multi-Group Cross Section (MGXS) method is an acceleration technique developed based on the characteristics of Monte Carlo simulation, which can significantly reduce the computation time required to solve a single group cross section in transportation calculations. The effectiveness of the method has been verified in the test calculations of water reactor pins. However, liquid molten salt reactors (MSRs) exhibit significant differences from conventional water reactors in terms of neutron energy spectra and fuel cycle mode. The effectiveness of the MGXS method in MSR burnup simulations remains to be validated, and targeted adjustments are required during its application. In this study, OpenMC and ORIGEN2 are coupled to develop an accelerated calculation method for MSR burnup simulations based on the MGXS approach. The reasonable grouping structure of the MGXS method is explored, and the performance of different grouping structures is tested. Results show that the transport calculation can be accelerated by an average factor of 2.4 for a single burnup zone by using MGXS method and the acceleration effect is generally independent of the grouping structure adopted. The nuclide mass bias compared to the traditional direct solution can be reduced to approximately 1% when the fuel burnup is 250MWd/kg for the LEU loading scheme with the 10000 groups structure. For the TRU loading scheme, the mass bias compared to the traditional direct solution of important nuclides (such as U-233, U-235, Pu-239 and so on) can be controlled below 0.5% at a burnup of 230 MW d/kg. The results indicate that the grouping strategy proposed in this study can achieve the adaptation of MGXS to MSRs, and the 10000 groups structure adopted in the study exhibits good accuracy.

在核反应堆的燃耗计算中，输运计算的耗时问题十分突出。多组截面（MGXS）方法是根据蒙特卡罗模拟的特点而开发的一种加速技术，它可以大大减少运算中求解单组截面所需的计算时间。该方法的有效性已在水反应堆销钉的试验计算中得到验证。然而，液态熔盐反应堆（MSR）在中子能谱和燃料循环模式方面与传统的水反应堆存在显著差异。MGXS 方法在 MSR 烧毁模拟中的有效性仍有待验证，在应用过程中需要进行有针对性的调整。本研究将 OpenMC 和 ORIGEN2 结合起来，开发了一种基于 MGXS 方法的 MSR 烧损模拟加速计算方法。探讨了 MGXS 方法的合理分组结构，并测试了不同分组结构的性能。结果表明，采用 MGXS 方法，单个燃烧区的输运计算平均可加速 2.4 倍，而且加速效果与采用的分组结构基本无关。与传统的直接解法相比，当燃料燃烧度为 250MWd/kg 时，采用 10000 组结构的 LEU 装载方案的核素质量偏差可降至约 1%。对于 TRU 装载方案，当燃料燃烧度为 230 MW d/kg 时，重要核素（如铀-233、铀-235、钚-239 等）的质量偏差与传统的直接解法相比可控制在 0.5% 以下。结果表明，本研究提出的分组策略可以实现 MGXS 对 MSR 的适应，研究中采用的 10000 组结构具有良好的精度。

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引用次数: 0

Modification of Portland cement matrix with diethyldithiocarbamate for technetium immobilization 用二乙基二硫代氨基甲酸二乙酯改性硅酸盐水泥基质以固定锝

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

Progress in Nuclear Energy

Pub Date : 2024-10-22 DOI: 10.1016/j.pnucene.2024.105508

Elena Abramova, Mikhail Volkov, Anton Novikov, Mikhail Grigoriev, Konstantin German, Alexey Safonov

The paper considers the effect of sodium diethyldithiocarbamate (SDDC) addition on the immobilization of technetium in a Portland cement matrix. The leaching process was evaluated in a model solution that simulated the conditions of the future radioactive waste (RW) storage site at the Yeniseisky site. The results demonstrated that the addition of 1.0 wt % SDDC to the cement composite increased the incorporation of immobilized technetium into the cement matrix by 35% in comparison to the blank sample. Furthermore, the retention of technetium in the cement matrix was observed to be enhanced, with a retention of up to 90% observed on the 200th day of the experiment. Resulting materials fulfill the necessary technical characteristics for use as a matrix and engineered safety barrier in the concept of deep RW disposal. The addition of SDDC into cement results in microbial respiratory activity decreasing. In order to evaluate the mechanisms of technetium immobilization, the structure of the technetium compound with diethyldithiocarbamate (DDC) [Tc(C₅H₁₀NS₂)₄]TcO₄ is described for the first time. This compound contains a Tc(V) atom which coordinates four diethyldithiocarbamate C₅H₁₀NS₂ moieties through eight sulfur atoms to form a complex cation. The addition of SDDC in cement is presumed to result in a cascade of disproportionation reactions and the formation of stable Tc(IV) compounds, as it was evidenced by XANES spectroscopy.

本文探讨了添加二乙基二硫代氨基甲酸钠（SDDC）对锝在波特兰水泥基质中固定化的影响。在模拟叶尼塞斯基场址未来放射性废物 (RW) 储存地条件的模型解决方案中对沥滤过程进行了评估。结果表明，与空白样品相比，在水泥复合材料中添加 1.0 wt % 的 SDDC 可使固定锝在水泥基质中的结合率提高 35%。此外，还观察到锝在水泥基质中的保留率得到了提高，在实验的第 200 天，锝的保留率高达 90%。所得材料满足了深层 RW 处置概念中用作基质和工程安全屏障的必要技术特征。在水泥中添加 SDDC 后，微生物的呼吸活动减少。为了评估锝的固定机制，首次描述了二乙基二硫代氨基甲酸锝化合物（DDC）[Tc(C5H10NS2)4]TcO4 的结构。该化合物含有一个锝（V）原子，通过八个硫原子与四个二乙基二硫代氨基甲酸盐 C5H10NS2 分子配位，形成一个复合阳离子。正如 XANES 光谱所证明的那样，水泥中加入 SDDC 后，可能会发生一连串的歧化反应，并形成稳定的 Tc(IV) 化合物。

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引用次数: 0

Development of fuel depletion code for molten salt reactor with very deep burnup 为超深燃耗熔盐反应堆开发燃料耗竭代码

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

Progress in Nuclear Energy

Pub Date : 2024-10-21 DOI: 10.1016/j.pnucene.2024.105506

Shuning Chen , Shaopeng Xia , Xiangzhou Cai , Chunyan Zou , Jingen Chen

A liquid-fueled molten salt reactor (MSR) can reach a deep burnup based on online reprocessing and continuously refueling, which requires significantly different burnup calculation methods for MSRs compared with those for the traditional reactors. To address the unique burnup features and consider the fidelity of isotopic evolution in an MSR, a fuel depletion code ThorMCB is developed based on the OpenMC coupled with a specific depletion code, MODEC. Furthermore, to lower the computational cost of acquiring the equilibrium state through the time evolution step by step for an MSR, an equilibrium burnup calculation code ThorMCB-eq based on the OpenMC and MODEC is developed, which can obtain the equilibrium burnup efficiently. A single fuel lattice of MSR and an a Molten Salt Fast Reactor (MSFR) benchmark are applied for verifying the correctness of the ThorMCB and ThorMCB-eq codes. Compared with a neutron transport calculation code KENO-VI coupled with MODEC, the maximum deviation of the dominant heavy nuclides (HNs) at equilibrium state by ThorMCB is less than 10%, and that of the total mass of fission products (FPs) is less than 3%. For the MSFR benchmark, the neutronic parameters including temperature reactivity coefficient, the mass evolution of main HNs and FPs and breeding ratio (BR) from ThorMCB agree with the references. The equilibrium behavior can be quickly obtained with ThorMCB-eq, and the relative mass deviations of most nuclides keep around 2% in comparison with the results of step-by-step burnup evolution with ThorMCB. Furthermore, the same fuel contents and micro one-group cross sections at equilibrium are obtained with two different types of start-up fuels and a constant power density and fuel reprocessing scheme. In conclusion, the verified results indicate that ThorMCB and ThorMCB-eq can both provide reliable simulation for depletion evolution and equilibrium burnup for MSR fuel cycle.

液体燃料熔盐反应堆（MSR）可以通过在线后处理和持续加注达到深度燃烧，这就要求 MSR 的燃烧计算方法与传统反应堆的燃烧计算方法大不相同。针对 MSR 独特的燃耗特征，并考虑到 MSR 中同位素演变的保真度，我们在 OpenMC 的基础上开发了燃料耗尽代码 ThorMCB，并结合了特定的耗尽代码 MODEC。此外，为了降低通过时间演化逐步获取 MSR 平衡态的计算成本，还开发了基于 OpenMC 和 MODEC 的平衡燃烧计算代码 ThorMCB-eq，该代码可高效获取平衡燃烧。为了验证 ThorMCB 和 ThorMCB-eq 代码的正确性，应用了 MSR 的单燃料晶格和熔盐快堆（MSFR）基准。与结合 MODEC 的中子输运计算代码 KENO-VI 相比，ThorMCB 计算的平衡态主要重核素（HNs）的最大偏差小于 10%，裂变产物（FPs）总质量的最大偏差小于 3%。对于 MSFR 基准，中子参数包括温度反应系数、主要 HNs 和 FPs 的质量演化以及 ThorMCB 的繁殖比（BR）与参考文献一致。使用 ThorMCB-eq 可以快速获得平衡行为，与 ThorMCB 逐步燃烧演化的结果相比，大多数核素的相对质量偏差保持在 2% 左右。此外，在使用两种不同类型的启动燃料以及恒定功率密度和燃料后处理方案的情况下，可以获得相同的燃料含量和平衡时的微单组截面。总之，验证结果表明 ThorMCB 和 ThorMCB-eq 都能为 MSR 燃料循环的耗竭演化和平衡燃烧提供可靠的模拟。

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引用次数: 0

Research on the radioactive safety of industrial steam supplied by PWR nuclear power plant 压水堆核电站工业蒸汽放射性安全研究

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

Progress in Nuclear Energy

Pub Date : 2024-10-19 DOI: 10.1016/j.pnucene.2024.105489

Lu Li, Junnan Zhang, Dachao Lin, Mi Xu, Qingning Zuo, Jingyun Huang, Xiaoping Bai, Qiming Wei, Da Zhang

The heating source of the nuclear power steam supply installation is PWR secondary steam, but the secondary steam is potentially radioactive, and the radioactive material may migrate to the industrial steam, which goes through the leakage of the heat exchanger. Compared with the conventional industrial steam, the steam supplied by the nuclear power has potential radioactivity. Nuclear power is first introduced to produce the industrial steam, in order to analyze the specific activity and radiation effects of the industrial steam on the staffs at the user side, an analytical model of industrial steam radioactivity is developed based on the technological process of industrial steam under operating conditions, and then the exposure pathways of the staff at the user side from the industrial steam is analyzed. The specific activity of the industrial steam and the exposure dose of the staff is evaluated. According to the evaluation results, it can be seen that the dose to staff caused by the industrial steam is far smaller than the radiation impact on the public caused by the nuclear power plant, and can meet the requirements of relevant regulations and standards. Therefore from the radiation safety point of view, the industrial steam supplied by the nuclear power steam supply installation is safe and acceptable for industrial use.

核电蒸汽供应装置的加热源是压水堆二次蒸汽，但二次蒸汽具有潜在的放射性，放射性物质可能通过热交换器的泄漏迁移到工业蒸汽中。与传统的工业蒸汽相比，核电供应的蒸汽具有潜在的放射性。首先引入核电来生产工业蒸汽，为了分析工业蒸汽的比活度和对用户端工作人员的辐射影响，根据工业蒸汽在运行条件下的工艺流程，建立了工业蒸汽放射性分析模型，然后分析了用户端工作人员从工业蒸汽中暴露的途径。评估了工业蒸汽的比活度和工作人员的暴露剂量。根据评估结果可以看出，工业蒸汽对工作人员造成的剂量远小于核电站对公众造成的辐射影响，能够满足相关法规和标准的要求。因此，从辐射安全的角度来看，核电蒸汽供应装置供应的工业蒸汽是安全的，可用于工业用途。

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引用次数: 0

Neutron and gamma radiation effects on thermal storage properties of polyethylene wax 中子和伽马辐射对聚乙烯蜡蓄热性能的影响

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

Progress in Nuclear Energy

Pub Date : 2024-10-18 DOI: 10.1016/j.pnucene.2024.105491

Ryan Steere, Joshua Schlegel, Sean Drewry, Jack Fletcher, Camden Henke, Joshua Rittenhouse, Galen Selligman, Joseph Graham

Several nuclear reactors use ice condensers to condense steam in the case of a loss of coolant accident. These ice condensers have many problems that could be alleviated by using another material. The effects of low-dose neutron and gamma radiation on the thermal properties of polyethylene wax (PEW) were investigated for this purpose. PEW was irradiated in the Missouri University of Science and Technology Research Reactor (MSTR), the University of Missouri Cyclotron (MUC) and the University of Missouri Research Reactor (MURR) up to doses equivalent to 10 months in a nuclear power reactor's containment structure. The melting temperature and latent heat of fusion were determined using differential scanning calorimetry (DSC). Changes in the molecular bonds was determined using Raman spectroscopy. It was found that there was not a significant change in the thermal properties nor bonding over the investigated doses. This suggests that organic PCMs could be reliable alternatives to ice in nuclear reactor containment applications. The measured melting peak was found to be significantly wider expected by the suppliers' description. The ramifications of wide melting peaks are discussed in the context of reactor accident analysis and further experiments are suggested.

一些核反应堆使用冰冷凝器在发生冷却剂损失事故时冷凝蒸汽。这些冰冷凝器存在许多问题，使用其他材料可以缓解这些问题。为此，我们研究了低剂量中子和伽马辐射对聚乙烯蜡（PEW）热性能的影响。在密苏里科技大学研究反应堆（MSTR）、密苏里大学回旋加速器（MUC）和密苏里大学研究反应堆（MURR）中对聚乙烯蜡进行了辐照，辐照剂量相当于核电反应堆安全壳结构中 10 个月的剂量。使用差示扫描量热法（DSC）测定了熔化温度和聚变潜热。使用拉曼光谱测定了分子键的变化。结果发现，在所调查的剂量范围内，热性能和键合情况都没有发生显著变化。这表明，在核反应堆安全壳应用中，有机 PCM 可以成为冰的可靠替代品。根据供应商的描述，测得的熔化峰明显比预期的要宽。在反应堆事故分析的背景下讨论了宽熔峰的影响，并提出了进一步实验的建议。

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引用次数: 0

Generating optimal reloading patterns for CNPGS Unit-3 core using multi-objective elitist teaching-learning-based optimizer 利用基于教学-学习的多目标精英优化器生成 CNPGS Unit-3 核心的最佳重装模式

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

Progress in Nuclear Energy

Pub Date : 2024-10-18 DOI: 10.1016/j.pnucene.2024.105507

Nadeem Shaukat , Amjad Ali , Ammar Ahmad , Ouadie Kabach , Khaled Al-Athel , Afaque Shams

Pressurized Water Reactors (PWRs) management presents the core reloading pattern optimization as a significant problem that contributes to the improvement of reactor productivity and optimal fuel utilization with considering safety conditions. In present study, Multi-Objective Elitist Teaching-Learning-Based Optimization (MO-ETLBO) technique is suggested to cope up the multi-objective reloading optimization problem of the Chashma Nuclear Power Generating Station (CNPGS) unit-3 core. A multivariable objective function is designed to evaluate the quality of each loading pattern while maximizing critical boron concentration (CBC), minimizing the power peaking factor (PPF), and optimally enhancing the cycle length while ensuring adequate safety margins and design limits. It has been found that the equilibrium cycle can be extended to 16.07 extended full power days (EFPDs) while maintaining the PPF and CBC within design limits. To validate the effectiveness of TLBO, the optimized loading pattern of the equilibrium core was evaluated using the deterministic computer code DONJON5, for neutronic parameter analysis. The results show that the algorithm proposed in this study is a promising approach for reloading pattern optimization in CNPGS unit-3, offering potential improvements in reactor cycle length while ensuring safety and enhancing overall performance.

压水堆（PWR）管理中，堆芯重装模式优化是一个重要问题，它有助于提高反应堆的生产率，并在考虑安全条件的情况下优化燃料利用。本研究建议采用多目标精益教学法优化（MO-ETLBO）技术来解决恰希玛核电站（CNPGS）3 号机组堆芯的多目标重装优化问题。设计了一个多变量目标函数，以评估每种装料模式的质量，同时最大化临界硼浓度（CBC）、最小化功率峰值因数（PPF），并在确保足够的安全裕度和设计限值的情况下优化循环长度。研究发现，平衡周期可延长至 16.07 个延长全功率日 (EFPD)，同时将 PPF 和 CBC 保持在设计限值内。为了验证 TLBO 的有效性，使用确定性计算机代码 DONJON5 对平衡堆芯的优化加载模式进行了评估，以进行中子参数分析。结果表明，本研究提出的算法是 CNPGS 3 号机组重新装料模式优化的一种有前途的方法，在确保安全和提高整体性能的同时，还有可能改善反应堆的周期长度。

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引用次数: 0

Advancements to generalized equivalence theory for preserving cross-sections of whole core simulations 用于保存整个岩心模拟截面的广义等价理论的进展

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

Progress in Nuclear Energy

Pub Date : 2024-10-17 DOI: 10.1016/j.pnucene.2024.105504

Oscar Lastres, Yunlin Xu

The conventional two-step approach, without the use of an equivalence method, can introduce significant error for the simulation of non-LWRs (Light Water Reactors), especially fast reactors, due to the large leakage and high anisotropic neutron distribution. To improve upon the accuracy of the two-step approach, a 3D whole core model is simulated with a transport method to generate region-wise homogenized cross-sections (XS). These XS can then be used in a diffusion whole core solver during the second step to extend the application to cycle and transient analysis. Discontinuity factors (DFs) are then introduced to improve the accuracy during the simulation of the second step. With properly generated DFs from Generalized Equivalence Theory (GET), the region-averaged solutions from the first transport step can then be reproduced by the second step diffusion solver. The Monte Carlo method was selected to perform the whole core transport simulation to generate region-wise XS. However, simulating whole core problems with Monte Carlo may result in poor statistics near the peripheral region especially for partial current tallies. This paper introduces an advancement to GET to reproduce region-wise solutions for select regions when the reaction rates and surface currents have good statistics in these regions but poor statistics in other regions. A reference high-fidelity model was constructed using the Serpent 2 Monte Carlo code based on the EBR-II benchmark evaluation and verification was carried out using the TriPEN-4 method in PARCS. The results show that it is possible to reproduce the exact eigenvalue and power distributions of whole core problems in a feasible manner.

由于存在大量泄漏和各向异性的中子分布，不使用等效方法的传统两步法会给非轻水反应堆（尤其是快堆）的模拟带来很大误差。为了提高两步法的精确度，采用传输方法对三维全堆芯模型进行模拟，以生成区域均匀化截面（XS）。在第二步中，这些 XS 可用于扩散整个岩心求解器，从而将应用扩展到循环和瞬态分析。然后引入不连续因子 (DF)，以提高第二步模拟的精度。有了根据广义等效理论（GET）正确生成的 DF，第二步扩散求解器就可以再现第一步传输过程中的区域平均解。我们选择蒙特卡洛方法来进行整个岩心的传输模拟，以生成区域 XS。然而，用蒙特卡罗法模拟整个岩心问题可能会导致外围区域附近的统计数据较差，尤其是部分电流统计。本文介绍了 GET 的一种改进方法，即当反应速率和表面电流在选定区域的统计量较好，而在其他区域的统计量较差时，在这些区域重现区域解。在 EBR-II 基准评估的基础上，使用 Serpent 2 蒙地卡罗代码构建了一个参考高保真模型，并使用 PARCS 中的 TriPEN-4 方法进行了验证。结果表明，有可能以可行的方式再现整个核心问题的精确特征值和功率分布。

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引用次数: 0

Research on sensor condition monitoring and signal reconstruction based on self-correcting anomaly diagnosis model 基于自校正异常诊断模型的传感器状态监测和信号重建研究

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

Progress in Nuclear Energy

Pub Date : 2024-10-17 DOI: 10.1016/j.pnucene.2024.105501

Yudi Zhu , Xinzhi Zhou , Chengping Zhao , Junhui Yu , Jialiang Zhu , Tao Xu , Zhengxi He

Condition monitoring is essential in industrial processes to ensure safe and efficient operations. Sensor signals, which accurately reflect the state of industrial systems, play a central role in this monitoring. However, the harsh conditions in many industrial environments, especially in nuclear power plants, increase the likelihood of sensor failures. Condition monitoring systems detect anomalies by reconstructing input data, with high reconstruction errors indicating the presence of anomalies. The Multivariate State Estimation Technique (MSET) is a widely used nonlinear, non-parametric model for condition monitoring. Traditional nonlinear models assume that training and test data come from the same distribution. This assumption can lead to significant errors when the model encounters anomalies, making it challenging to detect and reconstruct sensor states. To address these challenges, this paper introduces a self-correcting anomaly diagnosis model. Unlike traditional methods, this model establishes a dedicated data structure to store normal sensor patterns and generates a dynamic memory matrix that adapts to changes in industrial processes; The proposed method combines penalized offset projection with multi-scale estimation to mitigate the impact of anomalies on estimation results. Additionally, a variable correlation analysis method is developed to optimize input feature selection for the model. The new approach self-corrects anomalous data in a transformed signal space, achieving accurate reconstruction of sensor states. The model's performance is validated using real sensor data from a nuclear power plant system. Results demonstrate that the proposed model significantly enhances signal reconstruction and anomaly detection capabilities, even under more severe simulated conditions. Compared to traditional nonlinear models, the new method improves the metric for reducing anomaly interference by an order of magnitude. However, we did not change the calculation method of the higher-order kernel in the original method, which still faces the problem of matrix inversion.

为确保安全高效的运行，状态监测在工业流程中至关重要。能够准确反映工业系统状态的传感器信号在这种监测中发挥着核心作用。然而，许多工业环境条件恶劣，尤其是核电站，增加了传感器发生故障的可能性。状态监测系统通过重构输入数据来检测异常情况，重构误差大则表明存在异常情况。多变量状态估计技术（MSET）是一种广泛应用于状态监测的非线性、非参数模型。传统的非线性模型假设训练数据和测试数据来自相同的分布。当模型遇到异常情况时，这一假设可能会导致重大误差，使检测和重建传感器状态的工作面临挑战。为了应对这些挑战，本文介绍了一种自校正异常诊断模型。与传统方法不同，该模型建立了一个专门的数据结构来存储正常的传感器模式，并生成一个动态的记忆矩阵，以适应工业流程的变化；所提出的方法将惩罚偏移投影与多尺度估计相结合，以减轻异常对估计结果的影响。此外，还开发了一种变量相关性分析方法，用于优化模型的输入特征选择。新方法可在转换后的信号空间中对异常数据进行自我修正，从而实现传感器状态的精确重建。利用核电站系统的真实传感器数据对模型的性能进行了验证。结果表明，即使在更恶劣的模拟条件下，所提出的模型也能显著增强信号重建和异常检测能力。与传统的非线性模型相比，新方法将减少异常干扰的指标提高了一个数量级。然而，我们并没有改变原有方法中高阶核的计算方法，它仍然面临着矩阵反演的问题。

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