熔盐堆堆芯流体力学计算的GPU并行方法研究

使用计算流体力学(Computational Fluid Dynamics,CFD)数值方法对熔盐堆堆芯的流动和热传导等相关物理问题进行模拟求解,需要大量的计算时间。利用图形处理器(Graphics Processing Unit,GPU)加速技术对开源CFD软件Code_Saturne进行二次开发,研究求解熔盐堆堆芯流场的GPU并行算法。采用OpenACC语言在GPU上实现了向量运算、矩阵向量相乘等基本线性代数运算,从而实现预处理共轭梯度法(Preconditioned Conjugate Gradients,PCG)的GPU并行算法,并使用该算法求解压力状态方程。模拟了方腔驱动流模型及带下降段的熔盐堆堆芯模型的流场分布。结果表明,GPU加速后的软件与原版软件的结果一致,但计算时间更少,证明了GPU算法的正确性及有效的加速性。     

Convective heat transfer correlations for molten core debris pools growing in concrete

A volumetrically-heated pool with gas injection at the boundaries is used to simulate the heat transfer processes taking place in molten core debris—concrete systems. Measurements of the upward, downward, and sideward heat transfer rates at the pool boundaries have been made over wide ranges of power density, superficial gas velocity, and pool aspect ratio. Pools with either a free upper surface or an isothermal solid upper boundary to simulate an overlaying solid crust of oxides have been examined. Detailed measurements of the temperature distribution within the pool have also been made. A total of nearly 400 experiments have been conducted.Based on the results of these experiments, generalized correlations for the downward and sideward Nusselt numbers have been developed. The results indicate that the Nusselt numbers are nearly independent of the internal Rayleigh number and depend only on the nondimensional quantity (V³/μg) where V is the superficial gas velocity, is the pool density, μ is the viscosity, and g is the gravitational acceleration. The results also indicate that the downward and sideward Nusselt numbers in pools with gas release at the boundaries are comparable. This result is in contrast to the natural convection case where the sideward Nusselt numbers are significantly higher than those in the downward direction.The correlations developed in this investigation cover a wide range of the relevant nondimensional quantity (5 × 10⁻¹⁰ ≤ (V³/μg) ≤ 5 × 10⁻³). They can be used to determine the heat transfer rates at the pool boundaries, and hence, the pool growth rate into the concrete. Such information is necessary since it constitutes the last line of defence against violation of the containment.     

美国放射源监督管理模式

“许可证制度、协议州管理和全面监督检查”构成了美国放射源监管的主要模式框架。通过法规制定、许可证审查、监督检查、运行经验评估和实施监管支持活动,美国核管理委员会对放射性源在医学、工业、科研中的应用实施了有效的监管。了解美国放射源的监管体系,对于加强我国的放射源安全与保安的监管能力是有益的。本文简要介绍了美国的许可证制度、协议州管理和全面监督检查方面的情况。     

A study on concrete degradation during molten core/concrete interactions

A series of MCCI tests was performed in COTELS project at NUPEC to examine concrete degradation characteristics during MCCI with and without water addition onto the debris. Molten stainless steel or a mixture composed of UO₂, ZrO₂, Zr and stainless steel was slumped into a two-dimensional concrete trap, where volumetric decay heat generation was simulated by an induction heating technique. The results of dry MCCI tests implied that concrete ablation was dominated by melting of aggregates when the debris was crusted and cement was thermally weaker than aggregates. Without presence of stable crust, unmolten aggregates were possible to relocate upward due to the density difference from the debris. Concrete responses under a wet condition showed a tendency that water migrated into thermally degraded concrete. A preliminary water migration model was incorporated into COCO code for transient heat conduction. The prediction by COCO code agreed with the tendency of concrete thermal responses observed in the dry and wet MCCI tests.     

我国核安全法律体系研究

在归纳和总结国际核安全领域立法情况的基础上,从纵向和横向两个角度对我国现行核安全法律体系进行了剖析,进而分析中国现有核安全法律体系存在的主要问题:基本法迟迟未能制定,核安全法律法规体系缺失高层法律依据;配套性法律法规不完整,核安全法律法规体系内容尚待完善;技术性文件体系不完善,技术体系与核安全法规体系脱节。并从完善体系设计、优化体系结构、协调体系内容三个层面提出了完善中国核安全法律体系的对策。     

Structural and mechanical engineering research of the U.S. Nuclear Regulatory Commission

Research conducted by NRC Office of Nuclear Regulatory Research in Structural Mechanics for Reactor Technology is reported in this paper. The reported research primarily concerns applications to Civil Structures, Mechanical Systems and Components related to Light Water Reactor Nuclear Power Plants. It concerns the period between the Fourth International Conference in Structural Mechanics for Reactor Technology, held in San Francisco in 1977, and the 1979 Fifth Conference in Berlin. It includes research completed to date; status of research now under way, and future research programs.     

Risk Assessment Review Group Report to the U. S. Nuclear Regulatory Commission

The Risk Assessment Review Group was organized by the U.S. Nuclear Regulatory Commission on July 1, 1977, with four elements to its charter: (1) Clarify the achievements and limitations of WASH-1400, the "Rasmussen Report." (2) Assess the peer comments thereon, and responses to those comments. (3) Study the present state of such risk assessment methodology. (4) Recommend to the Commission how (and whether) such methodology can be used in the regulatory and licensing process.     

A metal fuel fast reactor core for the Self-Consistent Nuclear Energy System

Feasibility of transmutation of the major long-lived FPs (I, Pd, Tc, Sn, Se, Zr, Cs) while maintaining fuel breeding capability for the Self-Consistent Nuclear Energy System is evaluated based on the actinide recycle metal fuel core of a fast reactor. It is shown that I, Pd, Tc, Sn, Se, and Zr can be transmuted simultaneously by an aid of the isotope separation of Pd-107, Zr-93, Sn-126 and Se-79. Cs, which is difficult to transmute with the other FPs, is planned to be utilized as an in-reactor shielding material to confine in the system The overall assessment based on those results indicates that the developed system has the great potential toward the Self-Consistent Nuclear Energy System.     

Numerical study of the dynamic characteristics of a single-layer graphite core in a thorium molten salt reactor

A reactor core in a thorium molten salt reactor uses graphite as a moderator and reflector. The graphite core is a multi-layered arrangement of graphite bricks that are loosely connected to each other using a system of keys and dowels. Consequently, the graphite core is a type of discrete stack structure with highly nonlinear dynamic behavior. Hence, it is important to investigate the dynamic characteristics of the graphite core. In this study, a threedimensional single-layer graphite core model, which is a part of the thorium molten salt reactor side reflector structure, was analyzed using the explicit method in ABAQUS 2016 to study the core dynamic behavior when subjected to different excitations. The design parameters,such as the diameter of the dowel, the gap between key and keyway and the bypass flow gap between two adjacent bricks, were also considered in this model. To reduce excessive demands on available computational resources considering the effect of molten salt, the spring–dashpot model was applied to model the interaction forces between the molten salt and graphite bricks. Numerical simulation results show that the effect of molten salt is a reduction inthe peak maximal principal stress, and a larger gap between two bricks is beneficial to maintain the integrity of the graphite core under earthquake loading. The results obtained by the simulation can be used as a reference for future designs of a molten salt graphite core.     

核安全监管有效性评价指标研究

在我国核电发展的新形势下,持续加强监管的有效性十分必要。本文介绍了核安全监管有效性的基本概念,研究了IAEA和OECD所建立的核安全监管有效性评价指标体系,并进行了对比分析,提出了有助于加强我国核安全监管有效性的建议:继续完善核安全监管的法律法规和标准体系;加强监管能力建设;建立并实施有效的内部管理体系和内部质量保证体系;推动监管对象维持良好的安全业绩;力Ⅱ强与公众的沟通和互动。     

Status of structural and mechanical engineering research at the U.S. Nuclear Regulatory Commission

Research conducted by the NRC Office of Nuclear Regulatory Research related to structural mechanics for reactor technology is reported in this paper. This research concerns applications to mechanical and structural systems in light water reactors and includes materials engineering. Progress between the Sixth International Conference on Structural Mechanics in Reactor Technology, held in Paris in 1981, and the 1983 Seventh Conference in Chicago is discussed. A listing of research projects recently completed and the status of research presently under way are included.     

Development of the models for advection-diffusion of eroded concrete into debris and concrete volume reduction in molten core-concrete interactions

Models for the three-dimensional (3D) advection, diffusion, and volume reduction of eroded concrete into molten core are being developed. As part of the assessment of the reactor interior at TEPCO's Fukushima Daiichi Nuclear Power Plant, analytical models of molten core–concrete interaction (MCCI) to predict locations and condition of molten core (debris) have been improved in the debris spreading analysis (DSA) module of the severe accident analysis code SAMPSON. In addition to the primary model for 3D natural convection with simultaneous spreading, melting, and solidification in an open space, the analysis model to treat phenomena in a closed space, such as debris eroding laterally under concrete floors at the bottom of the sump pits, had been improved. This modeling with practical applicability is referred to as the full-3D MCCI model. This paper presents modeling of the advection and diffusion of eroded concrete into debris melt and calculation processes that were installed for simulating volume reduction when concrete decomposed. They were developed and incorporated into the full-3D MCCI model. The advanced DSA module with the models noted above was validated using MCCI test data. The calculated erosion rates agreed with the test data within a margin of about 16%.     

核电站核级承压设备设计分析专用程序系统研究

本程序系统研究是以ANSYS软件为开发主体,建立了用于核级承压设备力学分析的材料数据库、瞬态热工参数库、参数化模型库,编制了应力分析评定程序,实现了程序系统与CAD软件的数据接口,以确保输入数据的正确性和缩短计算分析的周期同时研究和开发了反应堆压力容器密封分析专用程序和特殊构件传热分析程序,并与 ANSYS软件连接,以弥补专用分析程序在前后处理方面的不足     

Dissolution behavior of core structure materials by molten corium in boiling water reactor plants during severe accidents

For the decommissioning of the Fukushima Daiichi Nuclear Power Plant, it is necessary to consider the access route to the fuel debris for its removal, which can be determined by knowing the corruption situation of the core support structure. To predict the damage condition of reactor vessel, dissolution behavior of the core structure material should be understood. In this study, the dissolution behavior of core structure materials (stainless steel) by molten metallic corium (stainless steel + B₄C) originated from control rod and its cladding was investigated. As a result of immersion experiment, it was found that there were two types of dissolution mode in this system: (1) chemical dissolution by eutectic reaction between Fe and B and (2) physical dissolution caused by the grains falling off from solid steel due to infiltration of molten metal. Moreover, on the basis of kinetic analysis, it was considered that the chemical dissolution in this system was slow. Therefore, the dissolution is considered to mainly occur through the mechanism that physical dissolution precedes chemical dissolution.