Nuclear Engineering and Design最新文献_第6页

Operation risk assessment and management with a front-controlled PRA in an AP1000 nuclear power plant AP1000核电站前置控制PRA运行风险评估与管理

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-24 DOI: 10.1016/j.nucengdes.2025.114703

Guocai Chen , Jiejuan Tong , Jun Zhang , Xiaofei Zhai , Zhenqi Wang

Against the backdrop of a renewed global wave of nuclear energy revival, in which international commercial giants, such as Microsoft, are actively promoting the development of small modular reactors, fast reactors, and even fusion technologies, it is challenging for Generation III (Gen-III) technology, such as the AP1000 and HPR1000, to seize the opportunity in a timely manner while demonstrating its advantages both in economy and in safety. In response to the feedback such as Gen-III's safety advantages not being effectively transformed into the operational benefits, the article proposes a probabilistic risk assessment (PRA) approach that incorporates the concept of checkpoint front-controlled (Front-PRA). A pilot study is being conducted at an AP1000 nuclear power plant in China to demonstrate the methodological feasibility of the proposed front-PRA. During the study, a new risk metric is identified: the triggering frequency of the Automatic Depressurization System (ADS-TF). The feasibility of establishing a four color-coded zone solution with ADS-TF and the traditional risk metric (Core Damage Frequency - CDF) to support operational risk management is thoroughly examined. The study reveals that, for advanced reactors like the AP1000, the current practice of defining the yellow zone as twice the baseline is inappropriate because it fails to translate safety advantages into operational benefits. In addition, the proposed approach also demonstrates its advantage to improve the perceived effectiveness of risk-informed applications.

在全球新一轮核能复兴浪潮的背景下，微软等国际商业巨头积极推动小型模块化反应堆、快堆甚至核聚变技术的发展，AP1000、HPR1000等第三代技术如何在及时抓住机遇的同时，展示其经济和安全优势，是一个挑战。针对“Gen-III”的安全优势未能有效转化为作战效益等反馈意见，本文提出了一种融合检查点前端控制（Front-PRA）概念的概率风险评估（PRA）方法。一项试点研究正在中国的AP1000核电站进行，以证明拟议的前置pra方法的可行性。在研究过程中，确定了一个新的风险度量：自动降压系统（ADS-TF）的触发频率。研究了利用ADS-TF和传统风险度量（核心损伤频率- CDF）建立四色编码区域解决方案以支持操作风险管理的可行性。研究表明，对于AP1000这样的先进反应堆，目前将黄色区域定义为基线的两倍的做法是不合适的，因为它无法将安全优势转化为运营效益。此外，所提出的方法还证明了其在提高风险知情应用的感知有效性方面的优势。

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引用次数: 0

Experimental study on circumferential non-uniform heat transfer characteristics in helical tubes 螺旋管内周向非均匀换热特性的实验研究

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-23 DOI: 10.1016/j.nucengdes.2025.114689

Xiong Zheng, Xin Wang, Xiting Chen, Shuqi Meng, Wei Tang, Tianming Ruan, Biao Li, Yaxin Gao, Yisong Hu, Desheng Jin, Dechang Cai, Yulong Mao

Helical tube once-through steam generator has been widely employed in advanced small modular reactors. However, studies on circumferential non-uniform heat transfer characteristics within helical tubes remain limited, especially their impact on the onset of nucleate boiling (ONB) and dryout (DO) phenomena. This lack of research directly affects the accurate prediction of flow and heat transfer behavior in helical tubes. In this study, the effects of pressure (7–15 MPa), mass flow rate (800–2700 kg/(m²·s)), and heat flux (200–550 kW/m²) on circumferential non-uniform heat transfer characteristics were experimentally investigated in helical tubes with tube inner diameter to coiled diameter ratios ranging from 0.01 to 0.0216. It was found that, in the same cross section, the inner wall exhibits the highest temperature, the outer wall the lowest, while the temperatures at the top and bottom positions are relatively comparable. This circumferential non-uniformity becomes more pronounced with decreasing pressure, increasing mass flow rate, and smaller coil diameter. Furthermore, the ONB and DO typically occur first at the inner wall, while DO might be observed simultaneously at different circumferential locations under low mass flow rate condition. The influences of mass flow rate, pressure, and heat flux on ONB and DO have been systematically clarified. Based on experimental data, improved heat transfers empirical correlations have been developed. These findings provide new experimental data and enhanced modeling capability for accurately predicting heat transfer behaviors in helical tubes operating at high-parameter conditions.

螺旋管一次性蒸汽发生器在先进的小型模块化反应堆中得到了广泛的应用。然而，关于螺旋管内周向非均匀换热特性的研究仍然有限，特别是它们对核沸腾（ONB）和干干（DO）现象发生的影响。这种研究的缺乏直接影响了螺旋管内流动和换热行为的准确预测。实验研究了管内压力（7 ~ 15 MPa）、质量流量（800 ~ 2700 kg/(m2·s)）和热流密度（200 ~ 550 kW/m2）对管内周向非均匀换热特性的影响，管内内径与盘管直径比为0.01 ~ 0.0216。结果表明，在同一截面上，内壁温度最高，外壁温度最低，而顶部和底部位置的温度具有相对可比性。这种环向不均匀性随着压力的降低、质量流量的增加和线圈直径的减小而变得更加明显。此外，ONB和DO通常首先发生在内壁，而在低质量流量条件下，DO可能同时出现在不同的周向位置。系统地阐明了质量流量、压力和热流密度对ONB和DO的影响。基于实验数据，改进了传热经验关系式。这些发现为准确预测高参数条件下螺旋管的传热行为提供了新的实验数据和增强的建模能力。

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引用次数: 0

Long-term cooling characteristics during transients with bypassing the safety protection system of a molten salt fast reactor 熔盐快堆旁路安全保护系统瞬态长期冷却特性

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-22 DOI: 10.1016/j.nucengdes.2025.114706

Hiroyasu Mochizuki

The objective of the present study is to confirm how long the reactor cooling of a molten salt fast reactor system can be maintained even after an abnormal transient with a bypass of the safety protection system under conditions that the decay heat removal system is activated or not activated. The system implements two heat storage tanks and a decay heat removal system using natural convection air cooling. The connection locations of the decay heat removal system are set to two heat storage tanks. The heat capacity of the molten salt stored in these tanks (capacity: 12,566 m³) is primarily intended to enable advanced load-following operation, but it can also be effectively utilized for long-term cooling. System analyses are performed under the conditions where the abnormal transients with the bypass of the safety protection system, and turbine of the energy conversion system is manually tripped at 10 min or 10 h and the air cooler is manually activated simultaneously. It has been confirmed that the heat capacity of the molten salt in the heat storage tanks is properly utilized by natural circulation, and the fuel temperature after turbine tripping becomes less than 900 K. It has been also confirmed through calculations that even if the decay heat removal system cannot be manually activated, if the turbine is tripped within 10 h, the molten fuel temperature can be kept below 1250 K for up to 1000 h (41.7 days) or longer with the heat capacity of molten salt in the heat storage tanks. This term increases as a function of the capacity of the heat storage tank, and it has been deemed that even without the heat storage tank, there is a time margin of more than a couple of hours before the temperature rises to 1250 K. The results of these analyses clearly show that the levels of defense in depth immediately prior to a severe accident are enhanced, and the probability of transition to SA due to internal factors such as unprotected abnormal transients is almost negligible.

本研究的目的是确定在衰变排热系统被激活或未被激活的情况下，即使在安全保护系统旁路的异常瞬态之后，熔盐快堆系统的反应堆冷却可以维持多长时间。该系统采用两个储热罐和一个采用自然对流空气冷却的衰减散热系统。衰变排热系统的连接位置设置为两个储热罐。储存在这些储罐中的熔盐的热容量（容量：12,566 m3）主要用于实现先进的负载跟随操作，但它也可以有效地用于长期冷却。在安全保护系统旁路的异常暂态下，分别在10min或10h手动跳闸换能系统水轮机，同时手动启动空冷器的情况下进行系统分析。结果表明，储热槽内熔盐的热容通过自然循环得到充分利用，涡轮脱扣后燃料温度低于900 K。通过计算也证实，即使不能手动启动衰变排热系统，如果在10 h内跳闸涡轮，利用储热罐中熔盐的热容量，熔燃料温度可以保持在1250 K以下长达1000 h（41.7天）或更长时间。这一项随着储热罐容量的增大而增大，我们认为即使没有储热罐，在温度上升到1250k之前也有几个小时以上的时间余量。这些分析的结果清楚地表明，在严重事故发生前的深度防御水平得到了提高，并且由于内部因素（如未受保护的异常瞬变）过渡到SA的可能性几乎可以忽略不计。

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引用次数: 0

Source term distribution and mobility – WP3 results of EU project SAMOSAFER 源项分布和流动性——欧盟SAMOSAFER项目WP3结果

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-22 DOI: 10.1016/j.nucengdes.2025.114679

J. Křepel , S. Lorenzi , L. Giot , E.M.A. Frederix , A. Cammi , F. Caruggi , S. Deanesi , F. Scioscioli , S. Delpech , P. Souček , J. Uhlíř , M. Mareček , M. Cihlář , J. Serp , H. Pitois , E. Merle , D. Rodrigues , C. Cannes , J. Kalilainen , S. Nichenko , T. Lind

The aim of the EU Horizon 2020 founded project SAMOSAFER was to develop and demonstrate new safety barriers and a more controlled behavior in severe accidents of the Molten Salt Reactor (MSR). The aim of work package 3 was to develop and validate models for tracing the nuclides carrying the radiotoxicity and decay heat, i.e. the source term and their chemical form and mobility during nominal and accidental conditions. The source term distribution at the beginning of an accident depends on the foregoing nominal operation mode. Hence, the liquid salt reprocessing and gaseous and insoluble Fission Products (FPs) separation was modeled before the severe accident simulation were accomplished. Because the source term assessment is a complex topic, only selected phenomena were addressed. As a reference system the MSFR concept was adopted from the previous EU project SAMOFAR. The study included benchmarking of the burnup tools, thermal-hydraulics simulations to confirm the removal rates to the off-gas system. Chemical experiments and calculations of the methods used in the reprocessing unit, simulation of severe accident conditions and finally calculation of the source term distribution. The major outcomes are the benchmark for the nuclide inventory tracing tools, application of multi-physics tools on He-bubbling for gaseous and solid FPs removal rates simulations, refinement of the reprocessing scheme, and first simulation of simplified severe accident.

欧盟地平线2020成立的SAMOSAFER项目的目的是开发和展示新的安全屏障，并在熔盐堆（MSR）的严重事故中更好地控制行为。工作包3的目的是开发和验证模型，以追踪携带放射性毒性和衰变热的核素，即源项及其在名义和意外条件下的化学形式和迁移率。事故开始时的源项分布取决于上述标称运行模式。因此，在完成严重事故模拟之前，对液态盐后处理和气态和不溶性裂变产物（FPs）分离进行了建模。由于源项评估是一个复杂的主题，因此只讨论了选定的现象。作为一个参考系统，MSFR概念是从以前的欧盟SAMOFAR项目中采用的。该研究包括燃耗工具的基准测试，热工模拟，以确认废气系统的去除率。化学实验和计算方法应用于后处理单元，模拟严重事故条件，最后计算源项分布。主要成果包括核素库存追踪工具的基准，he -泡泡多物理场工具在气体和固体FPs去除速率模拟中的应用，后处理方案的改进以及简化严重事故的首次模拟。

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引用次数: 0

Implementation, verification and validation of spatial kinetics calculations in CONDOR-CITVAP codes CONDOR-CITVAP代码中空间动力学计算的实现，验证和验证

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-22 DOI: 10.1016/j.nucengdes.2025.114713

Ignacio Ferrari , Diego Ferraro , Daniel Hergenreder

INVAP’s methodology for the neutronic design of research and radio-isotope production reactors is based on a deterministic cell-core scheme, which has been successfully applied in numerous projects. The cell stage is handled by CONDOR code, which solves the multi-group integral neutron transport equations by means of the Heterogeneous Response Method, and allows the generation of few-group macroscopic constants for core calculations. The core stage is handled by CITVAP code, which solves the multi-group diffusion equation. In this work, the implementation developed for the solution of spatial kinetics problems is presented, which extends the analysis capabilities beyond the standard criticality and fixed-source calculations. The selected methodology is based on a direct approach where the time-dependent neutron diffusion equation is solved. Several verification cases for rectangular geometry are evaluated and compared against reference results. Besides, validation cases using experimental data from the commissioning of the OPAL and RA-6 reactors are also analyzed.

INVAP的研究中子设计和放射性同位素生产反应堆的方法是基于确定性的细胞核心方案，该方案已成功地应用于许多项目。单元阶段由CONDOR代码处理，该代码采用非均质响应法求解多群积分中子输运方程，并允许生成用于堆芯计算的少数群宏观常数。核心阶段由CITVAP代码处理，求解多群扩散方程。在这项工作中，提出了为解决空间动力学问题而开发的实现，它将分析能力扩展到标准临界和固定源计算之外。所选择的方法是基于求解随时间变化的中子扩散方程的直接方法。对矩形几何的几个验证案例进行了评价，并与参考结果进行了比较。此外，还分析了OPAL和RA-6反应堆调试实验数据的验证案例。

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引用次数: 0

Robust control of small integral pressurized water reactor using μ-synthesis 基于μ-合成的小型整体压水堆鲁棒控制

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-22 DOI: 10.1016/j.nucengdes.2025.114708

Ahmad Salehi , Omid Safarzadeh , Ramazan Havangi

Ensuring robust control of small modular reactors (SMRs) is essential for safety and optimizing the performance of nuclear energy systems. Furthermore, accelerating integration of renewable energy sources like solar and wind into modern electricity grids has introduced certain challenges arising from inherent operational uncertainties and external disturbances. This variability has heightened the importance of using reliable power source such as SMRs with control systems capable of adjusting their power output dynamically in response to fluctuations in electricity demand. The SMART reactor is a small integral pressurized water reactor presented for enhancing the reliability and functionality of next-generation reactors. Achieving effective load-following in the SMART reactor necessitates advanced control systems that can ensure stability, safety, and performance under dynamic and uncertain conditions. The proposed robust control framework is compared with a conventional PID controller. The H_∞ robust controller is also designed to study its performance in managing system uncertainty and external disturbances. This study aims to address these critical challenges by developing a robust control framework. The μ-synthesis method is employed to achieve stable and efficient controllers for the SMART reactor. The control systems are designed to handle complex dynamics of reactor, steam generator, and pressurizer to maintain power, pressure, and water level of the modular reactor. The findings indicate that the designed controllers are highly effective in managing the aforementioned feature, while maintaining stability across the entire operational conditions.

确保小型模块化反应堆（SMRs）的鲁棒控制对于核能系统的安全和性能优化至关重要。此外，加速太阳能和风能等可再生能源与现代电网的整合，也带来了一些固有的运行不确定性和外部干扰带来的挑战。这种可变性提高了使用可靠电源的重要性，例如具有能够根据电力需求波动动态调整其功率输出的控制系统的小型反应堆。SMART反应堆是一种小型整体式压水反应堆，旨在提高下一代反应堆的可靠性和功能性。在SMART反应堆中实现有效的负载跟踪需要先进的控制系统，以确保在动态和不确定条件下的稳定性、安全性和性能。将所提出的鲁棒控制框架与传统的PID控制器进行了比较。设计了H∞鲁棒控制器，研究其在管理系统不确定性和外部干扰方面的性能。本研究旨在通过开发一个强大的控制框架来解决这些关键挑战。采用μ合成方法实现了SMART反应器稳定高效的控制器。控制系统设计用于处理反应堆、蒸汽发生器和稳压器的复杂动态，以维持模块化反应堆的功率、压力和水位。研究结果表明，所设计的控制器在管理上述特征方面非常有效，同时在整个操作条件下保持稳定性。

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引用次数: 0

Enhancing creep life prediction under large-shear deformation based on a modified Kachanov–Rabotnov model 基于改进Kachanov-Rabotnov模型的大剪切变形下蠕变寿命预测

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-22 DOI: 10.1016/j.nucengdes.2025.114694

Jun Hong , Tao Wang , Baoyin Zhu , Dongpeng Li , Haitao Dong , Dungui Zuo , Junlan Huang , Zheng Man , Gongye Zhang

Structural components subjected to simple shear at elevated temperatures are particularly vulnerable to creep deformation and failure, motivating reliable constitutive models and accurate life-prediction tools. This study extends the classical Kachanov-Rabotnov (K-R) model—originally developed for uniaxial tension or small shear deformations—into the large-shear regime, where the traditional formulation becomes less accurate. By introducing equivalent stress and strain measures tailored to finite shear, a modified K-R model is developed that accurately captures creep behavior under large-shear deformation. To demonstrate the applicability of the model, lap-jointed components fabricated with low-melting-point filler metals were selected as case studies, which is used to maintain the safety of the reactor containment vessel. Tensile and creep tests were conducted to fit the model parameters, which were subsequently incorporated into finite element simulations for comparative analysis. Validation against experimental and numerical results shows that the current modified model better replicates creep strain data, achieving closer agreement than the classical K-R model. The proposed model offers a practical and robust tool for creep-life assessment of large-shear structures, with significant implications for applications in nuclear passive-safety systems and brazed assemblies in thermal power and fire-protection equipment.

在高温下经受简单剪切的结构部件特别容易发生蠕变变形和破坏，从而产生可靠的本构模型和准确的寿命预测工具。这项研究将经典的Kachanov-Rabotnov （K-R）模型——最初是为单轴拉伸或小剪切变形而开发的——扩展到大剪切状态，在大剪切状态下，传统的公式变得不那么准确。通过引入为有限剪切量身定制的等效应力和应变测量，开发了一个改进的K-R模型，该模型准确地捕捉了大剪切变形下的蠕变行为。为验证该模型的适用性，选取了用于维护反应堆安全壳安全的低熔点填充金属搭接构件作为案例研究。进行了拉伸和蠕变试验以拟合模型参数，随后将其纳入有限元模拟进行对比分析。实验和数值结果验证表明，修正后的模型能较好地复制蠕变应变数据，与经典K-R模型的一致性更强。所提出的模型为大剪切结构的蠕变寿命评估提供了一个实用而强大的工具，对核被动安全系统和火电和消防设备的钎焊组件的应用具有重要意义。

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引用次数: 0

A CO2-based real options model for assessing the impact of external events on nuclear power plants 基于co2的核电厂外部事件影响评估实物期权模型

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-21 DOI: 10.1016/j.nucengdes.2025.114692

Alessandro Paravano, Giacomo Galeotti, Alessandra Neri, Enrico Cagno, Giorgio Locatelli

Nuclear power plants (NPPs) are essential for decarbonization. However, external events such as prolonged droughts or earthquakes have historically compromised NPPs' availability, undermining their decarbonization potential in national energy mixes. Life Cycle Assessment (LCA) is widely used to evaluate the CO₂e savings of NPPs, yet it typically assumes ideal operating conditions and fails to account for external events. Conversely, models that account for external events mainly adopt Real Options Analysis (ROA). Historically, ROA focuses solely on economic implications and considers monetary indicators (e.g., the price of electricity) while overlooking the impact on CO₂e savings. No ROA study has examined the impact of external events on NPPs from a CO₂e perspective. Yet, such external events are crucial, as empirically demonstrated by the reduced capacity factor of French reactors due to prolonged droughts. In response, we develop a novel model to assess the impact of external events on NPPs' electricity production from a CO₂e perspective. We developed and tested the model through a pseudo-real case study focused on drought events and their effects on NPPs' cooling systems. The model has three key novelties: 1) it consists of an enhanced LCA that incorporates the impact of external events on the NPP LCA; 2) it introduces a ROA approach based on CO₂e emissions as the evaluation metric instead of money; 3) it incorporates uncertainty by considering multiple CO₂e emissions scenarios simultaneously in the case of droughts impacting NPPs.

核电厂（NPPs）对脱碳至关重要。然而，长期干旱或地震等外部事件在历史上损害了核电站的可用性，削弱了它们在国家能源结构中的脱碳潜力。生命周期评估（LCA）被广泛用于评估核电站的CO₂e节约，但它通常假设理想的运行条件，无法考虑外部事件。相反，考虑外部事件的模型主要采用实物期权分析（Real Options Analysis， ROA）。从历史上看，ROA只关注经济影响，并考虑货币指标（例如，电价），而忽略了对CO₂节约的影响。没有一项ROA研究从CO₂e的角度考察了外部事件对核电站的影响。然而，这些外部事件是至关重要的，正如法国反应堆因长期干旱而导致的容量系数下降所证明的那样。为此，我们开发了一个新的模型，从二氧化碳排放的角度来评估外部事件对核电站发电的影响。我们通过一个拟真实的案例研究开发并测试了该模型，该研究主要关注干旱事件及其对核电站冷却系统的影响。该模型有三个关键的新颖之处：1)它包含了一个增强的LCA，该LCA包含了外部事件对NPP LCA的影响；2)引入了以CO₂排放量代替资金作为评价指标的ROA方法；3)在干旱影响核电站的情况下，通过同时考虑多种二氧化碳排放情景，纳入了不确定性。

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引用次数: 0

Effect of simultaneous proton irradiation on the corrosion of 11Cr ferritic/martensitic steel in lead‑bismuth eutectic at 550 °C 质子同步辐照对11Cr铁素体/马氏体钢550℃铅铋共晶腐蚀的影响

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-21 DOI: 10.1016/j.nucengdes.2025.114710

Hao Liu , Yue Liu , Rongshuo Wang , Jiuguo Deng , Feifei Zhang , Jijun Yang

A coupled 5 MeV proton irradiation and static lead‑bismuth eutectic (LBE) corrosion experiment was conducted on 11Cr ferritic/martensitic (F/M) steel at 550 °C. The influence of irradiation on the corrosion behavior of 11Cr F/M steel was investigated. The oxide layer in the irradiated region exhibited structural differences compared to unirradiated region, manifesting in three principal aspects. Firstly, the formation of a columnar crystal structure was observed on the surface of the Fe₃O₄ layer. Furthermore, the distribution of pores within the Fe-Cr spinel layer became more extensive. In addition, an internal oxidation zone (IOZ) was formed between Fe-Cr spinel layer and the substrate. These discrepancies were attributed to irradiation-enhanced diffusion of Fe and O, thereby accelerating LBE corrosion. The oxide layer thickness in irradiated region was found to be 2.2 times that of unirradiated region.

对11Cr铁素体/马氏体（F/M）钢进行了5mev质子辐照和静态铅铋共晶（LBE）腐蚀实验。研究了辐照对11Cr F/M钢腐蚀行为的影响。与未辐照区相比，辐照区氧化层表现出结构上的差异，主要表现在三个方面。首先，观察到Fe3O4层表面形成柱状晶体结构。此外，Fe-Cr尖晶石层内的孔隙分布更加广泛。Fe-Cr尖晶石层与基体之间形成了内氧化区（IOZ）。这些差异归因于辐照增强了Fe和O的扩散，从而加速了LBE腐蚀。辐照区氧化层厚度是未辐照区氧化层厚度的2.2倍。

引用次数: 0

The ARAMIS-P chloride molten salt concept for actinide conversion: A review of main design results 锕系元素转化用ARAMIS-P氯化物熔盐概念：主要设计结果回顾

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2025-12-19 DOI: 10.1016/j.nucengdes.2025.114681

Vincent Pascal , Christophe Venard , Johann Martinet , Elena Martin-Lopez , Martin Mascaron , Laura Matteo , Barbara Forno , Bertrand Morel , Jérome Serp , Marie-Sophie Chenaud

Within the framework of studies on the fuel cycle and transuranic actinide management (Pu, Am, Cm), the CEA and ORANO launched an R&D program on fast molten salt reactors (MSR) in 2020. The aim was to assess their opportunities with respect to fuel cycle management and to offer insights into their technical feasibility. ARAMIS-P is an abbreviation of ‘Advanced Reactor for Actinides Management in Salt’ with P for plutonium; this project focused on the preliminary design of a small unit for plutonium conversion integrated into a spent fuel reprocessing facility. The goal was to avoid nuclear cycle issues like spent MOX fuel reprocessing, by investigating the possibility of providing a flexible plutonium conversion service (from high-grade plutonium to ex-MOX quality). The reactor power was fixed at 300 MWth, which is in the power range of advanced modular reactors (AMR). From the perspective of a reprocessing unit, the reactor footprint, the fuel salt hold-up, and the need to develop new chemical processes should be limited. A chloride-based fuel salt was selected due to its compliance with known spent fuel processes. A specific design with a high core power density and a 6-month batch feed-up strategy was chosen to limit the overall amount of fuel salt. The design process was also driven by the desire to make maximum use of proven technologies when available, as well as to implement a maintenance-by-design approach. This report presents the design methodology developed to produce a preliminary reactor sketch, to illustrate its burn-up performance, and finally to give insights into component design.

在燃料循环和超铀锕系元素管理（Pu, Am, Cm）研究的框架内，CEA和ORANO于2020年启动了快速熔盐堆（MSR）的研发计划。目的是评估它们在燃料循环管理方面的机会，并对其技术可行性提出见解。amis -P是“盐中锕系元素管理先进反应堆”的缩写，P代表钚；这个项目的重点是初步设计一个小型钚转化装置，该装置与一个乏燃料后处理设施结合在一起。目标是通过调查提供灵活的钚转化服务（从高品位钚到前MOX质量）的可能性，避免核循环问题，如废MOX燃料后处理。反应堆功率固定为300兆瓦，在先进模块化反应堆（AMR）的功率范围内。从后处理装置的角度来看，反应堆占地面积、燃料盐占用率以及开发新化学工艺的需求应该是有限的。选择氯基燃料盐是因为它符合已知的乏燃料过程。采用高堆芯功率密度和6个月间歇补给策略的特定设计来限制燃料盐的总量。设计过程还受到最大限度地利用成熟技术的愿望的推动，并实施按设计维护的方法。本报告介绍了用于生成初步反应堆草图的设计方法，以说明其燃耗性能，并最终提供对组件设计的见解。

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引用次数: 0