System assessment of an FPGA-based RPS for ABWR nuclear power plant

The instrumentation and control (I&C) systems for the Lungmen nuclear power plant (LMNPP) are fully digitized based on microprocessor and software technology, and extensively utilize multiplexing networks. That is, undetectable software faults and common cause failures due to software errors may occur, and that will defeat the redundancy of a nuclear power plant (NPP). A diverse backup implementation for the digital I&C systems is an important means to defense against undetectable software faults.This paper presents system assessment of a quad-redundant reactor protection system (RPS) design for an Advanced Boiling Water Reactor (ABWR) by utilizing the field programmable gate array (FPGA) technology. The FPGA-based RPS has been assessed by using a full-scope engineering simulator for the LMNPP. Accident scenarios and abnormal conditions are inserted into the engineering simulator in order to activate the function of the FPGA-based RPS. In this study, conceptual design of the proposed quad-redundant FPGA-based RPS, including preliminary hardware architecture, software design and system assessment will be presented. The results demonstrate that the FPGA-based RPS system is a practical approach to implement a diverse backup for the digital I&C system of nuclear power plant applications.Also, the sensitivity study of probabilistic risk assessment (PRA) shows that RPS combined with ARI (Alternative Rod Insertion) contributes significant influence on the core damage frequency (CDF) calculation of LMNPP. The PRA sensitivity study is independent of the RPS technology.     

核电站综合设计和分析平台IDAP的研究

对于核电站的设计和安全分析,各国发展了许多设计和系统分析程序。按处理的问题分,主要有结构设计软件、燃料管理和堆物理设计分析软件、热工水力学设计和分析软件以及严重事故分析软件等等。本文结合 200MW低温核供热堆的设计和分析需要,把以上各个方面的软件统一到一个综合的平台 IDAP(Integrated Design＆ Analysis Platform),各系统实行可视化建模、自动参数转换和传递、耦合计算,大大加快了核电站的设计和分析过程。     

基于CPLD的CCD通用驱动电路设计方法

建立CCD通用测试平台有助于系统研究各类CCD器件的辐射效应及损伤机理。探讨了一种基于CPLD的线阵CCD通用驱动电路设计方法与实现途径。利用MAX-PL USII开发系统,选用MAX7000S系列CPLD芯片,设计实现了核心驱动主控制器,用于读取外部存储器驱动文件,设置相关参数寄存器,并产生符合参数要求的驱动时序脉冲。在此方法的基础上,完成了基本驱动模块电路的设计。基本驱动模块电路输出波形的测试结果表明,这种设计方法是完全可行的。     

Analysis of electromagnetic loads on EU-DEMO inboard and outboard blanket vertical segments

An analysis of the EM loads acting on a DEMO reactor configuration based on Multi Module Segment (MMS) design is presented in this work as part of the ongoing EU DEMO studies. Lorentz's forces and moments, both on the single module as well as on the complete blanket segment, are calculated for both the European HCPB and HCLL concepts. The system is analyzed considering a major central disruption scenario with a linear current quench of 42 ms using the ANSYS finite element software. The results are also compared to linear analyses to underline the effect of the non-linearity of the ferromagnetic materials.     

Design of a Spacer Grid Using Axiomatic Design

Recently, much attention is focused on the design of fuel assemblies in the Pressurized Light Water Reactor (PWR). The spacer grid is one of the main structural components in a fuel assembly. It supports fuel rods, guides cooling water, and maintains geometry from external impact loads. In this research, a new shape of the spacer grid is designed by axiomatic approach. The Independence Axiom is utilized for the design. For the conceptual design, functional requirements (FRs) are defined and corresponding design parameters (DPs) are found to satisfy corresponding FRs in sequence. Overall configuration and shapes are determined in this process. Detailed design is carried out based on the sequence from axiomatic design. For the detailed design, the system performances are evaluated by using linear and nonlinear finite element analyses. The dimensions are determined by optimization. Some commercial codes are utilized for the analysis and design.     

Kinematic and dynamic analysis of a serial-link robot for inspection process in EAST vacuum vessel

The present paper introduces a serial-link robot which is named flexible in-vessel inspection robot (FIVIR) and developed for Experimental Advanced Superconducting Tokamak (EAST). The task of the robot is to carry process tools, such as viewing camera and leakage detector, to inspect the components installed inside of EAST vacuum vessel. The FIVIR can help to understand the physical phenomena which could be happened in the vacuum vessel during plasma operation and could be one part of EAST remote handling system if needed. The FIVIR was designed with the consideration of having easy control and a good mechanics property which drives it resulted in function modular design. The workspace simulation and kinematic analysis are given in this paper. The dynamic behavior of the FIVIR is studied by multi-body system simulation using ADAMS software. The study result shows the FIVIR has ascendant kinematic and dynamic performance and can fulfill the design requirement for inspection process in EAST vacuum vessel.     

减压精馏分离稳定同位素18O的模拟优化研究

研究建立了1座采用蒸馏水减压精馏分离稳定同位素¹⁸O的实验装置,其填料层高20m,塔径为0.1m,塔内装填自主开发的PAC-¹⁸O专用填料。首先利用AspenPlus建立模型并验证其可靠性,然后利用此模型得出1组模拟数据,并利用人工神经网络（ANN）及Statistica软件对这组模拟数据进行优化设计。综合分析了塔顶压力及产品采出量对¹⁸O产品丰度的影响。结果表明,产品¹⁸O的丰度随塔顶压力和产品采出量的降低而升高,并得出最优化的塔压及¹⁸O丰度与产品采出量间的数学关联式。本研究采用的模拟优化方法可应用到¹⁸O的产业化设计及推广至传统精馏过程的优化设计中。     

Transient thermal–hydraulic simulations of direct cycle gas cooled reactors

This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to “active” heat extraction systems.     

Design methodology for insulating and cooling of a small modular reactor head by high temperature structural-thermal-fluid analysis

The design of a sodium-cooled fast reactor (SFR) head can be complicated due to its shape and functions. The head is a component placed in the pressure boundary to shield nuclear radioactive radiation. At the same time, it needs to seal the reactor vessel, support penetrating components, and minimize heat losses. This paper presents a new insulating and cooling design concept of a small SFR head. For a new design, this study shows a comprehensive design approach considering fluid-thermal-structural computations. The interactive design approach refers to dependent simulation steps of three-dimensional (3D) thermal-structural, one-dimensional (1D) heat-transfer, and 3D computational fluid dynamics (CFD) analysis. This multi-domain approach was applied to the head of the large sodium integral effect test facility called sodium test loop for safety simulation and assessment (STELLA-2). And the STELLA-2 head design was proposed as a thick plate with a sandwich type of insulation, cooling the perimeter annulus of the round head-top surface. For the structural design, the ASME design code was utilized, and the head temperature of 346?°C was calculated as its initial design temperature target. In an axial heat-transfer mode from the in-vessel to the head, a 1D finite element model gave 57 and 75 mm insulation thicknesses with a thermal conductivity of 0.07 W/m/K. The cooling effectiveness of the proposed head design was shown through a commercial CFD package.     

Progress in design and verification of the AMPS nuclear-electric plant

While nuclear plants of reduced scale and output specification supported by various degrees of inherent safety are being advanced for land-based applications, the AMPS development seeks to introduce nuclear plants with inherent safety to the marine industry, and in particular to small, highly mobile submersibles for commercial, oceanographic and coast-guard service. The AMPS development is rooted in an inherent safety design approach, as a matter of both practical necessity and opportunity. Verification of the AMPS concepts and design analyses is in progress using a full-scale thermalhydraulic facility and an organic Rankine engine representing the 100-kWe AMPS Prototype plant. Also, the design of an AMPS unit of over 1000 kWe and using a low-pressure steam cycle is subjected to analytical studies of accident scenarios.     

Security by facility design for sabotage protection

Facility design of nuclear power plant (NPP) for a sabotage protection is investigated and an effect of the design change for damage control on reduction of sabotage risk is shown using the vital area identification methodology. In a sabotage incident, it is not straightforward to identify the most credible scenario for NPP. However, the loss of offsite power leading to the station blackout is assumed to be a typical example for further evaluation. In this study, the vulnerability of vital area is considered in terms of the accessibility, the distribution of vital equipment, and the adversary's interference. As seen in the past report, the built-in measures for damage control are important in case of the existence of adversary's interference until neutralization. It is confirmed that not only the physical protection system, but also the facility design on structures, systems, and components play an important role in the effective and efficient sabotage protection. To reduce any vulnerability in the design of NPP, it is very important to introduce a security by design approach in an initial stage of the NPP construction while considering the interface between safety and security.     

Simulation-based design process for the verification of ITER remote handling systems

The work behind this paper takes place in the EFDA's European Goal Oriented Training programme on Remote Handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. One of the projects of this programme focuses on the verification and validation (V&V) of ITER RH system requirements using digital mock-ups (DMU). The purpose of this project is to study and develop efficient approach of using DMUs in the V&V process of ITER RH system design utilizing a System Engineering (SE) framework. Complex engineering systems such as ITER facilities lead to substantial rise of cost while manufacturing the full-scale prototype. In the V&V process for ITER RH equipment, physical tests are a requirement to ensure the compliance of the system according to the required operation. Therefore it is essential to virtually verify the developed system before starting the prototype manufacturing phase. This paper gives an overview of the current trends in using digital mock-up within product design processes. It suggests a simulation-based process design centralized around a simulation lifecycle management system. The purpose of this paper is to describe possible improvements in the formalization of the ITER RH design process and V&V processes, in order to increase their cost efficiency and reliability.     

Trial Visualization of Fast Reactor Design Knowledge

In design problems of large-scale systems like fast breeder reactors, inter-relations among design specificationsare very important where a selected specification option is transferred to other specification selections as a premise to be taken account in engineering judgments. These inter-relations are also important in design case studies with the hypothetical adoption of rejected design options for the evaluation of deviation propagations among design specifications. Some of these rejected options have potential worth for future reconsideration by some circumstance changes (e.g., advanced simulations to exclude needs for mock-up tests, etc.), to contribute to flexibility in system designs. In this study, a computer software is built to visualize a design problem structure by representing engineering knowledge nodes on individual specification selections along with inter-relations of design specifications, to validate the knowledge representation method and to derive open problems.     

Software design of the hybrid robot machine for ITER vacuum vessel assembly and maintenance

A specific software design is elaborated in this paper for the hybrid robot machine used for the ITER vacuum vessel (VV) assembly and maintenance. In order to provide the multi-machining-function as well as the complicated, flexible and customizable GUI designing satisfying the non-standardized VV assembly process in one hand, and in another hand guarantee the stringent machining precision in the real-time motion control of robot machine, a client–server-control software architecture is proposed, which separates the user interaction, data communication and robot control implementation into different software layers. Correspondingly, three particular application protocols upon the TCP/IP are designed to transmit the data, command and status between the client and the server so as to deal with the abundant data streaming in the software. In order not to be affected by the graphic user interface (GUI) modification process in the future experiment in VV assembly working field, the real-time control system is realized as a stand-alone module in the architecture to guarantee the controlling performance of the robot machine. After completing the software development, a milling operation is tested on the robot machine, and the result demonstrates that both the specific GUI operability and the real-time motion control performance could be guaranteed adequately in the software design.     

高温气冷堆示范电站一回路舱室区域模块化设计研究

一回路舱室是高温气冷堆示范电站建造的关键区域和难点之一。为提升该区域的建造质量和效率,针对一回路舱室区域开展模块化设计研究。通过方案比选,确定了支架模块方案的总体技术方案,提出模块化划分与设计的原则,完成该区域的模块化设计方案,利用三维设计软件CATIA构建各模块的三维模型,并对模块支承结构的强度和热应力进行设计校验,最后对设计方案的施工可行性与应用效果进行分析。结果表明,该方案可行,压缩工期效果明显。本研究结果在高温气冷堆示范电站工程上得到实际应用,有效提高了一回路舱室的建造效率,经济效益显著。     

核电主设备接管分析法设计一次应力可靠性分析

分析法设计是核电主设备设计的主要方法之一。该方法将结构设计或评定中各输入参量进行偏于安全的假设,以安全-不安全定性反映主设备设计的结构完整性状态。本研究在确定性分析法设计的基础上,利用可靠性分析方法,综合考虑结构设计或评定中涉及的不确定性因素(如结构几何、材料中的输入不确定性等),建立各种失效模式下的极限状态函数,基于概率统计理论求得结构在给定条件下的失效概率或可靠度,并进行相关参数的敏感性分析。以失效概率的形式定量反映部件的结构完整性状态,研究方法对可靠性理论在ASME核电规范与标准的分析法设计中的应用具有积极意义。     

RANS modeling for flow in nuclear fuel bundle in pressurized water reactors (PWR)

This paper presents use of Reynolds-averaged Navier-Stokes (RANS) based turbulence model for single-phase CFD analysis of flow in pressurized water reactor (PWR) assemblies. An open source code called OpenFoam was used for computational fluid dynamics (CFD) study using computational meshes generated using Shari Harpoon. The PWR assembly design used in this analysis represents a 5 × 5 pin design including structural grid equipped with mixing vanes. The design specifications used in this study were obtained from the experimental setup at Texas A&M University and the results obtained are used to validate the CFD software, algorithm, and the turbulence model used in this analysis.     

ITER过渡馈线辅助支撑结构设计及传热计算

过渡馈线系统是为ITER磁体系统提供能量、制冷剂、进行测量诊断等服务的子系统。本工作对过渡馈线直线段内部辅助支撑和S弯箱内部辅助支撑等进行了设计,并借助ANSYS有限元分析软件[JP2]对过渡馈线直线段内管由辅助支撑引起的传导热进行了分析。结果表明,辅助支撑结构的设计是合理的。     

超临界水冷堆堆芯子通道稳态热工分析

超临界水冷堆(SCWR)作为6种第四代未来堆型中唯一的水冷堆,冷却剂出口温度可达500℃,具有良好的经济性.本文采用改进的COBRA-IV程序对超临界水冷堆方形组件子通道进行稳态热工分析.对计算结果进行分析可知:减小慢化剂通道中给水质量流量份额和加大慢化剂通道与相邻子通道之间的热阻,可以降低热管焓升,后者还可以得到较好的慢化效果.通过热通道的传热恶化分析发现,超临界水冷堆的设计不能避免传热恶化,必须精确计算传热恶化条件下的包壳温度才能确定包壳能否保证其完整性.     

微机型船用核动力工程仿真器系统初步设计

基于船用核动力装置安全管理、开发研究的需求及核电站仿真技术的发展，分析了研制微机型船用核动力工程仿真器系统的重要意义。按软件工程的思想，从仿真器系统的功能设计、总体设计、模块设计三方面论述了该系统的设计思路及实现方法，并对日本核动力舰船“陆奥”号的回路系统进行了初步的模拟设计。该系统的实现将进一步提高船用核动力装置的优化设计、安全运行和科学管理水平。