Progress in Nuclear Energy最新文献_第5页

A neutronics-fuel coupling model for the simulation of constituent redistribution in U–Zr fuel 用于模拟 U-Zr 燃料中成分再分布的中子-燃料耦合模型

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

Progress in Nuclear Energy

Pub Date : 2024-10-03 DOI: 10.1016/j.pnucene.2024.105467

Simulating the dynamic redistribution of elements within metallic fuels, such as U–10Zr (wt.%), is a complex challenge in nuclear power research. While existing models can adequately simulate constituent redistribution throughout the fuel pin, they rely heavily on pre-known power history data or power calculated a priori. Furthermore, existing models do not calculate an accurate actinide inventory throughout the fuel pin. This limitation leads to less precise power distribution calculations, which can negatively affect the efficiency and safety of fuel design and performance. This work aims to introduce a new approach to modeling constituent redistribution for U–10Zr (wt.%) fuel by developing a new code called PFPS that accounts for the depletion and diffusion of Zr throughout the fuel pin during burnup through an explicit coupling method. The approach independently calculates the power distribution by coupling with the reactor physics code, Serpent 2, thereby eliminating the need for externally supplied power history data. The main contribution of this study lies in this coupling to obtain a more accurate history of the compositional evolution of U–Zr fuel and its fission products. This procedure can enhance the understanding of constituent redistribution by calculating a more accurate power distribution and lays the foundation for future attempts to simulate the compositional and microstructural evolution of all materials in the fuel.

模拟 U-10Zr（重量百分比）等金属燃料内部元素的动态再分布是核电研究中的一项复杂挑战。虽然现有模型可以充分模拟整个燃料针内的成分再分布，但它们在很大程度上依赖于已知的功率历史数据或先验计算的功率。此外，现有模型无法计算整个燃料销的精确锕系元素库存。这种限制导致功率分布计算不够精确，从而对燃料设计和性能的效率和安全性产生负面影响。这项工作旨在通过开发一种名为 PFPS 的新代码，为 U-10Zr （重量百分比）燃料的成分再分布建模引入一种新方法。该方法通过与反应堆物理代码 Serpent 2 的耦合，独立计算功率分布，从而无需外部提供功率历史数据。本研究的主要贡献在于通过这种耦合获得了更准确的 U-Zr 燃料及其裂变产物成分演变史。这一程序可以通过计算更精确的功率分布来加深对成分再分布的理解，并为今后模拟燃料中所有材料的成分和微结构演变奠定基础。

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引用次数: 0

Main lessons learnt from 40 years of R&D on iodine source term prediction: Identification of the main parameters governing iodine volatility in PHEBUS FP tests 从 40 年的碘源期预测研发中汲取的主要经验教训：确定 PHEBUS FP 试验中影响碘挥发的主要参数

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

Progress in Nuclear Energy

Pub Date : 2024-10-03 DOI: 10.1016/j.pnucene.2024.105473

Iodine chemistry and phenomenology in the containment has been studied for several decades. The main phenomena leading to the formation of volatile iodine have been identified step by step and their kinetics has been modeled and capitalized over the years in ASTEC-SOPHAEROS IRSN Severe Accident (S.A) code. Recently, the uncertainties for each phenomenon have been quantified and uncertainty propagation calculations have been performed on PHEBUS FPT-0/1/2/3 tests within the objective to identify which phenomena govern iodine volatility. The main highlights from PHEBUS studies are that (1) the sump reactions do not contribute to iodine volatility and (2) the gaseous phase chemical reactions are the main contributor to iodine volatility and (3) only a few gaseous reactions govern iodine volatility in PHEBUS containment. Another objective was to narrow the estimated range of %I_{2_RCS} (gaseous iodine fraction coming from the RCS). The results show that, considering 43 uncertain parameters, the iodine volatility plume is compatible with the experimental data whatever 2% < %I_{2_RCS} < 50% that mostly govern iodine volatility in the first days. It also indicates that, as soon as the FP release from the core is stopped and whatever 2% < %I_{2_RCS} < 50%, the influence of %I_{2_RCS} decreases over time so that the main processes leading to iodine volatility are slowly switched from %I_{2_RCS} (short term) to other gaseous phenomena (long term). The influence of %I_{2_RCS} on iodine volatility is thus important in the short term but becomes less and less significant in the long term (after several days). A more complete analysis is necessary for reactor applications to identify if the same conclusions can be drawn.

对安全壳中碘的化学性质和现象的研究已有几十年的历史。多年来，ASTEC-SOPHAEROS IRSN 严重事故（S.A）代码逐步确定了导致形成挥发性碘的主要现象，并对其动力学进行了建模和资本化。最近，对每种现象的不确定性进行了量化，并对 PHEBUS FPT-0/1/2/3 试验进行了不确定性传播计算，目的是确定哪些现象会影响碘的挥发性。PHEBUS 研究的主要亮点是：(1) 底盘反应不会导致碘挥发；(2) 气相化学反应是导致碘挥发的主要因素；(3) PHEBUS 容器中只有少数气相反应会导致碘挥发。另一个目标是缩小 %I2_RCS（来自 RCS 的气态碘部分）的估计范围。结果表明，考虑到 43 个不确定参数，碘挥发羽流与实验数据相符，无论 2%<%I2_RCS<50%，这些参数在最初几天主要控制碘挥发。这也表明，一旦堆芯中的FP释放停止，无论2%< %I2_RCS<50%是多少，%I2_RCS的影响都会随着时间的推移而减小，这样导致碘挥发的主要过程就会慢慢地从%I2_RCS（短期）转向其他气体现象（长期）。因此，%I2_RCS 对碘挥发的影响在短期内很重要，但在长期内（几天后）则越来越小。有必要对反应器应用进行更全面的分析，以确定是否可以得出相同的结论。

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引用次数: 0

Engineering multifunctional polypropylene nanocomposites: Tailoring structural, thermal, and gamma-ray shielding properties with Ni0.9Zn0.1Fe2O4 doping 多功能聚丙烯纳米复合材料工程：通过掺杂 Ni0.9Zn0.1Fe2O4 来定制结构、热和伽马射线屏蔽性能

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

Progress in Nuclear Energy

Pub Date : 2024-10-02 DOI: 10.1016/j.pnucene.2024.105478

This study delves into the structural, thermal and shielding processing of polypropylene (PP) modified with Ni_0.9Zn_0.1Fe₂O₄ NPs, synthesized via melt processing. The structural analysis confirms the effective integration of Ni_0.9Zn_0.1Fe₂O₄ NPs into the PP matrix, leading to enhanced thermal stability. Furthermore, the impact of incorporating Ni_0.9Zn_0.1Fe₂O₄ NPs on the radiation shielding properties of the fabricated PP-based composites was assessed using Monte Carlo simulation. The results reveal a significant increase in the linear attenuation coefficient of the composites with higher concentrations of Ni_0.9Zn_0.1Fe₂O₄ NPs, resulting in improved radiation shielding efficiency, particularly at lower gamma-ray energies.

本研究深入探讨了通过熔融加工合成的 Ni0.9Zn0.1Fe2O4 NPs 改性聚丙烯（PP）的结构、热和屏蔽加工。结构分析证实，Ni0.9Zn0.1Fe2O4 NPs 有效地融入了聚丙烯基体，从而提高了热稳定性。此外，还利用蒙特卡洛模拟评估了掺入 Ni0.9Zn0.1Fe2O4 NPs 对所制备的聚丙烯基复合材料辐射屏蔽性能的影响。结果表明，Ni0.9Zn0.1Fe2O4 NPs 的浓度越高，复合材料的线性衰减系数就越大，从而提高了辐射屏蔽效率，尤其是在较低的伽马射线能量下。

引用次数: 0

Use of hybrid resin to reduce silica from spent fuel pool borated water of power plant at pilot scale 使用混合树脂减少发电厂乏燃料池硼酸水二氧化硅含量的试验规模

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

Progress in Nuclear Energy

Pub Date : 2024-10-02 DOI: 10.1016/j.pnucene.2024.105479

Removal of silica from the borated water of primary systems of the nuclear power plant remains a long-term challenge for the industry. The same was tried to solve by membrane technology, however, membrane technology has its issues like generations of additional radioactive wastes. To address this issue a specialized hybrid resin for selective silica removal from borated water was prepared and tested for its efficiency at a pilot scale on an inventory taken from the spent fuel pool of an operational nuclear power plant having 2110 ppm boron and 400 ppb silica. The developed hybrid resin technique offers a simple and easy procedure to overcome several problems encountered in other methods of silica removal such as additional equipment, additional waste handling and maintenance. The validation shows that the application of the developed technique has a silica removal efficiency of 38% without affecting boron concentration.

从核电厂一级系统的硼酸盐水中去除二氧化硅仍然是该行业面临的一项长期挑战。同样的问题也曾试图通过膜技术来解决，但膜技术也有其自身的问题，比如会产生额外的放射性废物。为了解决这个问题，我们制备了一种用于从含硼水中选择性去除二氧化硅的专用混合树脂，并在试验规模上对取自运行中核电厂乏燃料池的库存（含硼 2110 ppm，二氧化硅 400 ppb）进行了效率测试。所开发的混合树脂技术提供了一个简单易行的程序，克服了其他去除二氧化硅方法所遇到的几个问题，如额外的设备、额外的废物处理和维护。验证结果表明，在不影响硼浓度的情况下，所开发技术的二氧化硅去除效率为 38%。

引用次数: 0

Numerical investigation on the formation of focusing effect in the IVR strategy 关于 IVR 策略中聚焦效应形成的数值研究

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

Progress in Nuclear Energy

Pub Date : 2024-10-02 DOI: 10.1016/j.pnucene.2024.105476

The “focusing effect” is the main challenging issue to the success of the IVR strategy, since the heat flux to the RPV wall could be significantly larger in the thin metal layer region than that in the oxide layer region. This paper numerically investigates the formation of focusing effect using validated CFD approach. The influences of the top cooling condition, layer height and material properties on the formation of focusing effect are investigated. Results indicate that, enhancing the top cooling mitigates the focusing effect. For the insufficiently-cooled top radiation situation, reducing the pool height significantly increases the focusing effect. For the sufficiently-cooled top surface (e.g., with top water cooling), the focusing effect is not formed for all the cases regardless of the pool height. It demonstrates/supports the benefit of adding in-vessel flooding to IVR strategy as a supplementary measurement. It means that once the in-vessel flooding could be established in engineering to allow for a sufficient top cooling, the focusing effect would not likely be formed regardless of the pool height. It also confirms enhancing top cooling condition an efficient way to reduce focusing effect. As two main influential material properties, the effects of thermal conductivity and viscosity are also investigated. Either decreasing the thermal conductivity or increasing the viscosity (e.g., by addition of other materials) may reduce the focusing effect. Since IVR is a widely adopted severe accident mitigation strategy, this study could provide some insights in the formation of focusing effect and help inspiring or supporting possible new engineering features for a safety IVR design.

"聚焦效应 "是 IVR 战略取得成功的主要挑战问题，因为到达 RPV 壁的热通量在薄金属层区域可能明显大于氧化层区域。本文采用有效的 CFD 方法对聚焦效应的形成进行了数值研究。研究了顶部冷却条件、层高和材料特性对聚焦效应形成的影响。结果表明，加强顶部冷却可减轻聚焦效应。对于顶部冷却不足的辐射情况，降低水池高度会显著增加聚焦效应。对于充分冷却的顶面（例如，采用顶部水冷却），无论水池高度如何，聚焦效应在所有情况下都不会形成。这证明/支持了在 IVR 策略中加入舱内充水作为辅助测量的好处。这意味着，一旦在工程中确定了容器内充水，使顶部冷却充分，则无论水池高度如何，都不会形成聚焦效应。这也证实了加强顶部冷却是减少聚焦效应的有效方法。热导率和粘度是影响材料性能的两个主要因素，我们也对它们的影响进行了研究。降低导热系数或增加粘度（如通过添加其他材料）都可能降低聚焦效应。由于 IVR 是一种广泛采用的严重事故缓解策略，这项研究可为聚焦效应的形成提供一些启示，并有助于激发或支持安全 IVR 设计的新工程特性。

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引用次数: 0

On the prediction and optimization of the flow boiling heat transfer in mini and micro channel heat sinks 小型和微型通道散热器中流动沸腾传热的预测与优化

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

Progress in Nuclear Energy

Pub Date : 2024-10-01 DOI: 10.1016/j.pnucene.2024.105466

The most common methods for predicting flow boiling heat transfer in mini/micro channels-based heat sinks rely on semi-empirical correlations derived from experimental data. However, these correlations are often limited to specific testing conditions. This study proposes a novel approach using deep learning and genetic algorithms (GA) to predict and optimize refrigerants' flow boiling heat transfer coefficients (FBHTC) in mini/microchannels-based heat sinks. The dataset used in this study includes FBHTC observations from the literature for seven refrigerants (R1234yf, R1234ze, R134A, R513A, R410A, R22, and R32). The optimal input parameters identified include hydraulic diameters ranging from 1 to 7 mm, saturation temperature from 0 to 20 °C, flow qualities from 0.006 to 0.972, heat flux from 3 to 78.8 kW/m², and mass fluxes between 100 and 1200 kg/m²s. Gradient-boost regression trees were employed to develop the deep learning and GA models for accurate estimation and optimization. Correlation analysis and feature engineering selected the most influential parameters to construct a precise and simple model. The results demonstrate that the models could estimate refrigerants' FBHTC with high accuracy, achieving an R² of 0.988 and a mean squared error (MSE) of 0.05%. The GA-based method effectively optimized the FBHTC for each refrigerant by determining the appropriate input parameters, including the saturation temperature, heat and mass fluxes, quality, and hydraulic diameter. Additionally, a parametric analysis using explainable artificial intelligence was conducted to interpret the impact of each input parameter on the FBHTC.

预测基于微型/微通道的散热器中流动沸腾传热的最常用方法依赖于从实验数据中得出的半经验相关性。然而，这些相关性往往局限于特定的测试条件。本研究提出了一种使用深度学习和遗传算法（GA）的新方法，用于预测和优化基于微型/微通道的散热器中制冷剂的流动沸腾传热系数（FBHTC）。本研究使用的数据集包括文献中对七种制冷剂（R1234yf、R1234ze、R134A、R513A、R410A、R22 和 R32）的 FBHTC 观察结果。确定的最佳输入参数包括 1 至 7 毫米的液压直径、0 至 20 °C 的饱和温度、0.006 至 0.972 的流量质量、3 至 78.8 kW/m2 的热通量以及 100 至 1200 kg/m2s 的质量通量。采用梯度提升回归树来开发深度学习和 GA 模型，以进行精确估算和优化。相关性分析和特征工程选择了最有影响力的参数，从而构建了一个精确而简单的模型。结果表明，模型可以高精度地估计制冷剂的 FBHTC，R2 为 0.988，均方误差（MSE）为 0.05%。基于 GA 的方法通过确定适当的输入参数（包括饱和温度、热通量和质量通量、质量和液压直径），有效优化了每种制冷剂的 FBHTC。此外，还使用可解释人工智能进行了参数分析，以解释每个输入参数对 FBHTC 的影响。

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引用次数: 0

Development of a liquid metal subchannel code applied to ocean conditions and study on the effect of core geometry parameters and ocean motions on flow and heat transfer characteristics 开发适用于海洋条件的液态金属子通道代码，研究核心几何参数和海洋运动对流动和传热特性的影响

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

Progress in Nuclear Energy

Pub Date : 2024-09-30 DOI: 10.1016/j.pnucene.2024.105469

The Generation IV International Forum (GIF) proposes a type of liquid metal reactor, due to its thermal properties and efficient heat transfer in a relatively small system. This special heat transfer characteristic allows the liquid metal reactor to be modular in design and flexible in installation on a nuclear-powered ship. Accurate prediction of the coolant temperature and flow distribution between subchannels is important to ensure nuclear safety in the complex ocean environment. In this study, an ocean version of subchannel code is developed by adding the motion terms in the momentum equations for the liquid metal reactor core based on the previously land-based subchannel code. Since the research on the characteristics of flow resistance and heat transfer mechanism under ocean conditions is not mature, the original empirical model is still used in the ocean version code. The constitutive models of the liquid metal are analyzed and incorporated into the ocean code. The code has been verified and validated by comparing with Computational Fluid Dynamics (CFD) simulation results, and experimental data. The effect of geometric parameters such as rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) is studied in the liquid metal lead 7 rod bundles. In addition, the effect of heaving start-up and shutdown motions between liquid metal lead and water is discussed in the 5 × 5 bare rod bundles. In general, the modified subchannel analysis code can well complete the calculation of liquid metal reactor under ocean conditions, as well as provide support for the design and safety analysis of a liquid metal cooled fast reactor.

第四代核反应堆国际论坛（GIF）提出了一种液态金属反应堆，因为它具有热特性，可以在一个相对较小的系统中高效传热。这种特殊的传热特性使液态金属反应堆可以模块化设计，灵活地安装在核动力船舶上。准确预测冷却剂温度和子通道之间的流量分布对于确保复杂海洋环境中的核安全非常重要。本研究在之前陆基子通道代码的基础上，通过在液态金属反应堆堆芯动量方程中增加运动项，开发了海洋版子通道代码。由于对海洋条件下流阻特性和传热机理的研究尚不成熟，因此海洋版代码仍采用原有的经验模型。对液态金属的构成模型进行了分析，并将其纳入海洋代码。通过与计算流体动力学（CFD）模拟结果和实验数据的比较，对代码进行了验证和确认。在液态金属铅 7 杆束中，研究了杆间距与直径比 (P/D)、杆与壁间隙 (RTW) 等几何参数的影响。此外，还讨论了 5 × 5 裸棒束中液态金属铅和水之间起伏启动和关闭运动的影响。总之，改进后的子通道分析代码可以很好地完成海洋条件下液态金属反应堆的计算，并为液态金属冷却快堆的设计和安全分析提供支持。

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引用次数: 0

Human-machine cooperation in safety management: An intelligent decision-making framework for cooling water intake reliability 安全管理中的人机合作：冷却水取水口可靠性智能决策框架

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

Progress in Nuclear Energy

Pub Date : 2024-09-28 DOI: 10.1016/j.pnucene.2024.105452

This paper develops an intelligent decision-making framework for urgent response in the water intake path before marine blockage. Using machine learning and decision-making techniques (heuristic approach and systematic approach), human-machine cooperation supports exploring the pattern of the decision-making process concerning multiple knowledge areas in the nuclear power plant (NPP) reliability of the large cooling water intake. The decision-making framework contains: (1) monitoring datasets with IoT-supported detection, (2) a dynamic expert system, and (3) lump-sum treatments for improving the reliability of NPPs' cooling water intake systems. Through dynamic data collection and analysis, the framework improves the decision quality through higher-level cooperation between human-driven and machine-driven data. Interviews were conducted to illustrate the operating value of the proposed framework in practice. The proposed framework responds to operational, tactical, and strategic requirements via human-machine cooperation design. It effectively finds a valid solution before incidents can happen. It also sheds light on the NPP's operation safety epistemologies from an intelligence-based decision-making viewpoint. Theoretically, the framework presents a human-machine cooperated method for projects involving various decision-makers and multiple datasets. It is expected to bring insights into other projects with similar decision-making processes, like NPP water intake issues.

本文开发了一个智能决策框架，用于在海洋堵塞前对进水路径进行紧急响应。利用机器学习和决策技术（启发式方法和系统化方法），人机合作支持探索核电站（NPP）大型冷却水取水口可靠性中涉及多个知识领域的决策过程模式。决策框架包括：（1）支持物联网检测的监测数据集；（2）动态专家系统；（3）提高核电站冷却水取水口系统可靠性的一次性处理方法。该框架通过动态数据收集和分析，在人类驱动数据和机器驱动数据之间进行更高层次的合作，从而提高决策质量。为说明拟议框架在实践中的操作价值，还进行了访谈。所提出的框架通过人机合作设计满足了作战、战术和战略要求。它能有效地在事故发生前找到有效的解决方案。它还从基于情报的决策角度揭示了国家核电厂的运行安全认识论。从理论上讲，该框架为涉及不同决策者和多个数据集的项目提出了一种人机合作方法。它有望为其他具有类似决策过程的项目（如核电站取水问题）带来启示。

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引用次数: 0

Improvements of the dynamics code TMSR3D for molten salt reactor leveraging triangular prism nodal method 利用三角棱柱节点法改进熔盐反应堆动力学代码 TMSR3D

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

Progress in Nuclear Energy

Pub Date : 2024-09-28 DOI: 10.1016/j.pnucene.2024.105454

Dynamic simulation with coupled model of neutronics/thermal-hydraulics (TH) is one of the crucial aspects in safety analysis for molten salt reactor (MSR), which requires a deeper insight into the interactions among hydrodynamics, heat transfer and neutron kinetics because of its special characteristics induced by fuel drift, in comparison to the traditional reactors using solid fuels. To improve the geometric adaptability and calculation precision of the MSR specific dynamics code TMSR3D based on quadrilateral nodal method and TH module including the multi-channel flow and single-channel heat transfer models, the analytic basis function expansion nodal (ABFEN) method using triangular prism nodes is implemented to solve the static neutron diffusion equations, an implicit backward difference method along with exponential transformation is adopted to deal with the neutron kinetics problems, and the concentration equations of delayed neutron precursors (DNP) considering fuel flow are solved by the method of characteristics. The VVER-440 transient benchmark is first selected to validate capacity of the updated code for simulating core with hexagonal assemblies. Subsequently, the Molten Salt Reactor Experiment (MSRE) is simulated for further validation, in which the steady-state characteristics including delayed neutron loss induced by circulating fuel, profiles of reactor temperature and DNP, impact of fuel residence time in external loop on the effective yields of delayed neutrons, two fuel pump driven transients, and the natural convection are analyzed and discussed in detail. The results agree well with that from similar codes and the experimental values, which indicate that the improvements conducted for TMSR3D are effective and correct.

与使用固体燃料的传统反应堆相比，熔盐反应堆（MSR）因其由燃料漂移引起的特殊性，需要更深入地了解流体力学、传热学和中子动力学之间的相互作用。为了提高 MSR 特定动力学代码 TMSR3D 的几何适应性和计算精度，该代码基于四边形节点法和 TH 模块（包括多通道流动和单通道传热模型），采用了使用三角形棱柱节点的解析基函数展开节点法（ABFEN）来求解静态中子扩散方程、采用隐式后向差分法和指数变换来处理中子动力学问题，并通过特性法求解考虑燃料流的延迟中子前体（DNP）浓度方程。首先选择了 VVER-440 瞬态基准，以验证更新后的代码模拟六边形组件堆芯的能力。随后，对熔盐反应堆实验（MSRE）进行了模拟以进一步验证，其中详细分析和讨论了稳态特性，包括循环燃料引起的延迟中子损失、反应堆温度和 DNP 曲线、燃料在外循环中的停留时间对延迟中子有效产率的影响、两个燃料泵驱动的瞬态以及自然对流。结果与类似代码和实验值完全一致，表明对 TMSR3D 进行的改进是有效和正确的。

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引用次数: 0

Thermo-economic and environmental analyses of supercritical carbon dioxide Brayton cycle for high temperature gas-cooled reactor 用于高温气冷堆的超临界二氧化碳布雷顿循环的热经济和环境分析

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

Progress in Nuclear Energy

Pub Date : 2024-09-27 DOI: 10.1016/j.pnucene.2024.105461

The supercritical carbon dioxide (sCO₂) Brayton cycle demonstrates special advantages for the high temperature gas-cooled reactor (HTGR) delivered into commercial operation recently in China, with its high efficiency, compactness, flexibility, and safety compared to the conventional steam Rankine cycle. However, the large temperature rise of 500 °C for the HTGR brings new challenges for the design of sCO₂ cycle. Here, we present the first study on the thermodynamic, economic, and environmental performance of the HTGR-sCO₂ system using the energy, exergy, economic, and environmental (4E) evaluation method. The cascaded sCO₂ cycle made up of two independent sCO₂ cycles is proposed, which are arranged in series on the cold side of reactor heat exchanger. We show that the cascaded sCO₂ cycle can utilize the heat absorption from HTGR effectively by optimizing the cycle configurations of top and bottom sub-cycles. The improved cascaded sCO₂ cycle minimizes the exergy loss, and increases the thermal efficiency to 43.2% when compared to the steam Rankine cycle of HTGR demonstration power plant and the single recompression cycle. By balancing the fixed-capital investment cost with the net power, the levelized cost of electricity can be reduced to 0.0283$/kWh. The life-cycle GHG emission intensity of HTGR-sCO₂ systems is about 6.5gCO_2,eq/kWh, which is much smaller than that of coal-fired power plants, suggesting a great potential for decarbonization of the HTGR-sCO₂ system. Our study may find implications for the advancement of the sCO₂ Brayton cycle in next-generation nuclear power plant.

与传统的蒸汽朗肯循环相比，超临界二氧化碳（sCO2）布雷顿循环具有高效、紧凑、灵活和安全等优点，适用于最近在中国投入商业运行的高温气冷堆（HTGR）。然而，高温气冷堆 500 ℃ 的大幅升温给 sCO2 循环的设计带来了新的挑战。在此，我们首次采用能量、放能、经济和环境（4E）评价方法对高温热电联产-sCO2 系统的热力学、经济和环境性能进行了研究。我们提出了由两个独立的 sCO2 循环组成的级联 sCO2 循环，它们串联布置在反应堆热交换器的冷侧。我们的研究表明，通过优化顶部和底部子循环的循环配置，级联 sCO2 循环可有效利用高温热电联产反应堆的吸热。与 HTGR 示范电站的蒸汽朗肯循环和单一再压缩循环相比，改进后的级联 sCO2 循环最大程度地减少了放能损失，并将热效率提高到 43.2%。通过平衡固定资本投资成本和净电量，平准化电力成本可降至 0.0283 美元/千瓦时。高温热电联供系统的生命周期温室气体排放强度约为 6.5gCO2,eq/kWh，远低于燃煤电厂，表明高温热电联供系统的脱碳潜力巨大。我们的研究可能会对下一代核电厂中二氧化碳布赖顿循环的发展产生影响。

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引用次数: 0