高温气冷堆新燃料元件运输容器临界安全分析

采用基于蒙特卡罗方法的MCNP5程序对高温气冷堆所用的球形燃料元件进行描述;根据包覆燃料颗粒在燃料球内的分布性质构建了8种不同模型,并研究不同模型对有效增殖因子（keff）和计算时间的影响,获得了临界计算问题中最优的燃料球模型;运用MCNP5描述燃料球运输容器,并研究了容器中子吸收板厚度、外容器壁厚、缓冲层材料、反射层材料、容器形状、容器结构缺失和水密度等影响运输容器临界安全的因素。结果表明,所研究的高温气冷堆新燃料元件运输容器在正常运输条件下和事故运输条件下均处于临界安全状态,其临界安全指数（CSI）可定为0。      

高温气冷堆燃料运输容器热工计算分析

采用CFX程序模拟高温气冷堆燃料运输容器内外导热、对流、热辐射等传热方式。计算结果表明:容器各部件温度不会超过限值、热工结构符合安全运输要求。将计算结果与容器火烧试验相比较,证明了计算模型的保守性与合理性。     

AP1000新燃料组件运输货包的临界安全计算

新燃料组件运输过程中最主要的核安全问题是临界安全。在对运输货包进行临界安全分析中必须要同时考虑多货包阵列形式、事故后货包损伤情况、最佳水慢化条件等因素。本文采用MCNP程序针对美国西屋公司XL型运输容器装载AP1000新燃料组件货包的实例进行了临界安全计算。结果表明,在XL型运输容器设计许可书中允许装载货包数N=75的限制条件下,临界安全是有保障的。     

CEFR-MOX新燃料组件运输货包临界安全计算

易裂变材料运输过程中重要的安全问题之一是临界安全。在对运输货包进行临界安全分析中必须要同时考虑多货包阵列形式、事故后货包损伤对临界安全影响、最佳水慢化条件等因素。本文采用MCNP 程序针对CEFR-MOX新燃料组件运输货包进行了临界安全计算。计算结果表明:MCNP程序(采用核截面库为ENDF/B-V库)对本问题的次临界限值为0.924 6;正常运输条件下无限个运输货包的最大k_eff值为0.574 4,运输事故条件下无限个运输货包的最大k_eff值为0.659 7。根据临界安全指数的定义,确定CEFR-MOX新燃料组件运输货包的临界安全指数为0。     

高温气冷堆燃料元件发展前景

第4代核能系统主要候选堆型之一的超高温气冷反应堆(very-high-temperature reactor,简称VHTR)氦气出口温度要求大于1 000℃;从经济性考虑,模块式高温气冷堆的单堆功率愈高愈好,燃料的燃耗深度也愈高愈好,这些对近代低富集度3层包覆颗粒(modern LEU TRISO particles)燃料元件提出了更高燃耗深度和耐更高温的要求.为满足上述要求,本文介绍了ZrC层代替包覆燃料颗粒的SiC层、UCO(UO2 UC2)核芯代替包覆燃料颗粒的UO2核芯和进一步降低现有低富集度3层包覆颗粒SiC层破损率的高温气冷堆燃料元件的研究和发展.     

高温气冷堆新型燃料球传感器的设计

对高温气冷堆中燃料球运行情况的准确监测是保障反应堆安全可靠运行的关键。针对原有探测器的不足,利用穿透式涡流检测原理提出了新型对装式燃料球传感器。运用有限元方法建立了该传感器的电磁场数值计算模型,对传感器结构参数和检测参数进行了分析和优化设计。实验结果表明,该传感器过球信号信噪比高,对连续过球具有很好的分辨率,满足反应堆现场使用要求。     

高温气冷堆燃料元件发展现状和趋势

本文介绍了高温气冷堆燃料元件的发展历史，现状和趋势。经过３０多年的研究和发展，燃料元件的设计，制造工艺和质量鉴证技术已相当成熟，燃料元件可在１２５０℃长期工作。２１２０００个ＴＲＩＳＯ颗粒辐照试验的时没有一个破损，１６００℃下辐照后退火５００ｈ，阻挡裂变产物释放的能力没有下降。     

高温气冷堆包覆燃料颗粒包覆层性能测试方法

高温气冷堆包覆燃料颗粒由ＵＯ２燃料堆饼和在它表面沉积的热解碳和ＳｉＣ层材料构成。这些热解碳和ＳｉＣ层的厚度只有３０－９０μｍ，为测量这些微小颗粒包覆层材料的性能，专门研究了热解碳和ＳｉＣ层的厚度，密度和热解碳层的各项异性能，ＳｉＣ层的弹性模量等的测量方法，并研制了颗粒尺寸分析仪，小试样弹性模量测定仪设备等。     

10MW高温气冷堆球形燃料元件碳化工艺研究

通过对高温气冷堆球形燃料元件压制坯体及酚醛树脂碳化过程的研究，确定了碳化工艺制度的制订原则。在碳化过程中，低温开裂主要是由压制工艺中产生的应变不均匀性造成原，高温开裂则主要受加热速率的影响，采用加压碳化工艺可提高基体材料的机械性能。     

高温气冷实验堆燃料元件双向探测器的研制

介绍了高温气冷实验堆燃料元件双向探测器的基本原理和实现方法。它以两个并联的感应线圈为敏感元件,通过双通道法采集信号,以８９Ｃ５１单片机为处理核心,系统软件采用循环扫描输入端口的方式获取过球信号,经智能分析、判断,实现了燃料元件的双向检测。     

Simulation study on candle burnup applied to block-type high temperature gas cooled reactor

The CANDLE burnup is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top (or from top to bottom) of the core and without any change in their shapes. It can be applied easily to a block-type high temperature gas cooled reactor (HTGR) using an appropriate burnable poison with a high neutron absorption cross section mixed with uranium oxide fuel. In this study, natural gadolinium is used as burnable poison. In the present paper, the simulation of the burnup for the steady state and the startup is performed.

For the steady state simulation with direct solutions of steady state nuclide densities as inputs, the difference between the results of the steady state analysis and the simulation analysis is very small. It confirms that the steady state analysis is correct. When the initial core is constructed from easily available nuclides, the simulation result gives a reactivity change of 1.7% at a burnup time of 0.7 years.     

Control rod drive for high temperature gas cooled reactor

This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.     

Requirements for high temperature reactor fuel particle design assessment

In the framework of a large Research and Development programme devoted to High Temperature Reactors (HTR) and set up in the CEA from 2000 on, we will address ourselves to the issue of coated fuel performance and design. Although HTR fuel main features have been established for a long time, we need today to reassess the fuel design to make sure that it meets the requirements linked to the most recent projects of High Temperature Reactors. Thus, in collaboration with Framatome and in connection with the Gas Turbine - Modular Helium Reactor (GT-MHR) international project, we are planning to perform parametric thermal and mechanical studies, regarding different particle design options (kernel diameter, layers composition and thickness) and seeking optima concerning particle leak tightness and fission product retention. But to initiate such studies, we have first of all to define the design bases and the requirements for HTR fuel, in terms of kernel composition (fissile element, oxide stoechiometry, enrichment), particle and compact geometry (dimensions, particle volume fraction in the graphite matrix), power density, cooling gas temperature and irradiation conditions (burnup, fast fluence).     

Analysis and design of spent fuel transport cask impact limiters

Abstract

Three Latin American countries which operate research reactors, Argentina, Brazil and Chile, have joined efforts to improve the capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half scale model for materials test reactor fuel was constructed in Argentina and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions.

In this paper both the numerical modelling and mechanical tests to select adequate shock absorbers materials are presented. Results of these tasks are used to improve the cask design.     

Evaluation of high temperature gas reactor for demanding cogeneration load follow

Modular nuclear reactor systems are being developed around the world for new missions among which is cogeneration for industries and remote areas. Like existing fossil energy counterpart in these markets, a nuclear plant would need to demonstrate the feasibility of load follow including (1) the reliability to generate power and heat simultaneously and alone and (2) the flexibility to vary cogeneration rates concurrent to demand changes. This article reports the results of JAEA's evaluation on the high temperature gas reactor (HTGR) to perform these duties. The evaluation results in a plant design based on the materials and design codes developed with JAEA's operating test reactor and from additional equipment validation programs. The 600 MWt-HTGR plant generates electricity efficiently by gas turbine and 900°C heat by a topping heater. The heater couples via a heat transport loop to industrial facility that consumes the high temperature heat to yield heat product such as hydrogen fuel, steel, or chemical. Original control methods are proposed to automate transition between the load duties. Equipment challenges are addressed for severe operation conditions. Performance limits of cogeneration load following are quantified from the plant system simulation to a range of bounding events including a loss of either load and a rapid peaking of electricity.     

Development of a thermal-hydraulic analysis code TAC-DS for spent fuel dry-storage system for high-temperature pebble bed reactor

We report the development of a thermal-hydraulic analysis code (called TAC-DS: Thermal-hydraulic Analysis Code for Dry-storage System). The spent fuel dry-storage system of High-Temperature Reactor Pebble-bed Modules in China is simulated using the TAC-DS to confirm the design basis and to analyze the transient behavior following an accident involving blower failure. The TAC-DS includes mathematical models for the air-coolant system, heat conduction within spent fuel canisters, and thermal radiation between heat structures. The time-dependent hydrodynamic model of the TAC-DS is formulated using one-dimensional mass, momentum and energy equations, and solved using semi-implicit finite-difference scheme. The complicated heat transfer models of heat structure are incorporated into the hydrodynamic system implicitly with enclosure correlations. Code is written in Fortran 90. A validation calculation is performed by solving a simplified model. Thermal performance of the buffer storage region in the system under forced ventilation scenario is studied with TAC-DS to validate the design requirement, as well as to provide the initial condition for the transient analysis. Blower failure accident is studied to assess the performance of the safety features during the transient accident. Since the code is modular, TAC-DS can be easily modified and applied to other spent fuel dry-storage system in the future.     

Development of high temperature gas-cooled reactor (HTGR) fuel in Japan

This paper describes experiences and present status of research and development works for the high temperature gas-cooled reactor (HTGR) fuel in Japan. Recently, Very High Temperature Reactor (VHTR) is evaluated highly worldwide, and is a principal candidate for the Generation IV reactor systems. In Japan, HTGR fuel fabrication technologies have been developed through the High Temperature Engineering Test Reactor (HTTR) project in Japan Atomic Energy Agency since 1960’s. In total about 2 tons of uranium of the HTTR fuel has been fabricated successfully and its excellent quality has been confirmed through the long-term high temperature operation. Based on the HTTR fuel technologies, SiC TRISO fuel has been newly developed for burnup extension targeted VHTR. For ZrC-TRISO coated fuel as an advanced fuel designs, R&Ds for fabrication and inspection have been carried out in JAEA. The irradiation with the Japanese uniform stoichiometric ZrC coating has been completed in the cooperation with Oak Ridge National Laboratory of the United States.     

Analysis of thermal problems in development of spent fuel transport cask for research reactors

A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper, both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled.