法国先进核燃料组件在中国应用前景浅析

介绍了法国先进燃料组件（ＡＦＡ）系列核燃料的特点及其在中国的应用现状，阐述了广东核电集团有限公司核电发展战略和第三代先进燃料组件（ＡＦＡ－３Ｇ）在中国的应用前景，并从物理，热工水力和燃料组件的机械完整性等方面作了初步论证，对当前开展的有关工作进行了讨论。     

核燃料

在法国核然料工业组织中,法杰马公司主要销售燃料组件。法比燃料公司(FBFC)的3个从属工厂都负责燃料组件的制造,该公司每年生产装铀量为1500t 的燃料组件。自1985年以来,法杰马公司又销售先进燃料组件(AFA)。该 AFA 的主要特点是使用了锆合金定位格架和可拆式上、下管嘴。大亚湾核电站要用的燃料组件正是该种与一般组件不同的先进燃料组件。法杰马公司采用钆作可燃毒物,以保证燃料组件的良好特性。近来该公司又推出混合氧化物燃料组件(MOX)。由于法杰马公司在设计和制造的各阶段都严格遵守了质量保证和质量控制制度,所以其产品质量优良、可靠性好。展望未来,法杰马公司将与法国核燃料工业中的其它集团一起,努力为用户提供尽可能好的产品。     

AFA 3G及其它高性能燃料组件

文章着重介绍了国际上大规模入堆的高性能AFA 3G燃料组件的设计特点和制造特点、Performance+组件的设计特点及目前正开发的其它高性能燃料组件.介绍了高性能燃料组件的使用现状,并对我国压水堆高性能燃料的发展提出了一些建议.     

中核北方核燃料元件有限公司企业技术创新工程简介

<正>中核北方核燃料元件有限公司是我国核燃料、核材料生产和科研基地。通过引进、消化、吸收,先后掌握了CANDU-6核电燃料等3种水堆燃料制造技术。实现了具有我国自主知识产权的高温气冷堆球形元件制造技术的工程化应用。公司积极致力于先进燃料研发,突破了CAP1400自主化燃料组件及压水堆环形燃料组件制造技术,ATF燃料等先进燃料研发取得一定进展。     

AP1000燃料组件设计特性

AP1000型燃料组件是西屋公司在40多年的燃料组件没小重行经验的基础上,改进开发的用于AP1000反应堆的高性能燃料组件。本文介绍了西屋压水堆燃料组件的设计发展,重点描述了AP1000型燃料组件的设计特点。     

中国先进研究堆燃料组件设计及试验进展

燃料组件是中国先进研究堆(CARR)的核心部件,燃料组件设计成平板型,使用低浓铀U3Si2-Al弥散体燃料.经多项堆内外验证试验证明在设计要求条件下,燃料组件结构稳定,使用安全.     

秦山二期工程燃料组件LOCA和SSE下的事故分析

介绍了秦山二期工程燃料组件在ＬＯＣＡ和ＳＳＥ同时发生的情况下,燃料组件与组件间、组件与围板间的撞击力计算方法和结果以及燃料组件各部分的应力分析和组件的稳定性分析。     

AFA3G LE燃料组件的结构特点和设计论证

简要介绍AFA3G LE型燃料组件的设计特点,就AFA3G LE型燃料组件与AFA3G型燃料组件的不同点进行比较,并对AFA3G LE型燃料组件的试验论证进行简要介绍。     

CF2系列燃料组件热态冲刷试验研究

针对CF2系列燃料组件,采用全尺寸和截短型(应用于模块化小堆)整组件分别开展了1:1的模拟燃料组件高温高压冲刷试验研究,考验了CF2系列燃料组件结构的可靠性及耐磨蚀等情况,获得了控制棒在全尺寸和截短型组件内的落棒性能数据,同时探讨了冲刷试验的工况参数确定方法。研究结果为国产先进燃料组件设计优化、安全评定及应用提供了重要的试验依据。     

中国先进研究堆标准燃料组件三维流场分析

采用计算流体力学软件CFX4.4和CFX5.5对中国先进研究堆标准燃料组件流场进行了数值模拟。计算得到了额定工况下标准燃料组件内各个冷却剂通道的流量分布和不等间隙通道燃料板两侧压差。根据不同流量下的压降计算结果,给出了标准燃料组件的阻力特性曲线,并与试验结果进行了比较,符合较好。     

Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.     

SMART behavior under over-pressurizing accident conditions

SMART (System-integrated Modular Advanced ReacTor) is an integral reactor of 330 MW capacity with passive safety features under development in Korea. The design is developed by combining the firmly-established commercial reactor technologies with new and advanced technologies such as industry proven KOFA (Korea Optimized Fuel Assembly) based nuclear fuels, self-pressurizing pressurizer, helically coiled once-through steam generators, and new control concepts. The design of SMART focuses on enhancing the safety and reliability of the reactor by employing inherent safety features such as low core power density, elimination of large break loss of coolant accident, etc. In addition, in order to prevent the progression of emergency situations into accidents, the SMART is provided with a number of engineered safety features such as Passive Residual Heat Removal System, Passive Emergency Core Cooling System, Safeguard Vessel, and Passive Containment Over-Pressure Protection System. This paper presents an overview of the SMART design, characteristics of it’s safety systems, and results of over-pressure accident analyses. The results of the accident analyses show that the SMART provides the inherent over-pressure protection capability for design basis accidents without actuation of any protection devices such as safety valves, rupture disks, etc.     

Mechanical analysis of fuel fretting problem

Fuel fretting is studied by contact mechanics approach. Shear force produced by flow-induced vibration is regarded as the major factor of the fuel fretting. Contact dimension is examined for the Korean PWR Fuel Assembly using finite element method. Axial direction is incorporated with transverse one for the shear force. As for the sequence of the shear, a closed rectangular as well as an oblique path are considered to simulate the actual behaviour due to the vibration. The shear stresses on the contact surface between fuel rod and spacer grid is evaluated numerically. It is supposed that a partial slip regime prevails on the contact at the early stage of fuel life. In case of gross slip, the present method can be applied without modification. The dissipation of friction energy on the contact is evaluated and discussed for a wear model and a spacer grid design.     

Phase characteristics of a number of U-Pu-Am-Np-Zr metallic alloys for use as fast reactor fuels

Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. A series of test designs for the Advanced Fuel Cycle Initiative (AFCI) have been irradiated in the Advanced Test Reactor (ATR), designated as the AFC-1 and AFC-2 designs. Metal fuel compositions in these designs have included varying amounts of U, Pu, Zr, and minor actinides (Am, Np). Investigations into the phase behavior and relationships based on the alloy constituents have been conducted using X-ray diffraction and differential thermal analysis. Results of these investigations, along with proposed relationships between observed behavior and alloy composition, are provided. In general, observed behaviors can be predicted by a ternary U-Pu-Zr phase diagram, with transition temperatures being most dependent on U content. Furthermore, the enthalpy associated with transitions is strongly dependent on the as-cast microstructural characteristics.     

Helios, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark

As part of an effort to test the ability of current transport codes to treat reactor core problems without spatial homogenization, the lattice code HELIOS was employed to perform criticality calculations. The test consists in seven-group calculations of the C5 MOX fuel assembly problem specified by Cavarec et. al. [1]. This problem, known as C5G7 MOX Benchmark, is described in the Benchmark Specification [2] and comprises two cases — two and three-dimensional geometry. There are four fuel assemblies — two with MOX fuel, the other two with UO₂ fuel. Each fuel assembly is made up of a 17×17 lattice of square fuel-pin cells. Fuel pin compositions are specified in the Benchmark Specification, which also provides seven-group transport-corrected isotropic scattering cross-sections for U0₂, the three MOX enrichments, the guide tubes, the fission chamber and the moderator. This paper preset is the methodology employed in solving the C5G7 MOX Fuel Assembly Problem using the transport code HELIOS.     

The first results of deuterium retention on EAST with a full graphite wall via particle balance

This paper introduces the first results of deuterium retention on the Experimental Advanced Superconducting Tokamak (EAST) using particle balance.In the fall 2010 EAST experiments with a full graphite wall,the average deuterium retention fraction was about 19％ (including disruptive shots) and 38％ (not including disruptive shots).Fuel retention for the short-and long-pulse discharge was different.The H-mode discharges had a slightly lower fuel retention than the L-mode discharges.However,it was observed that disruptions introduced outgassing from the wall.Wall conditioning,such as lithium coating,increases retention.     

Increase in pyrolytic carbon optical anisotropy and density during processing of coated particle fuel due to heat treatment

Measurements of pyrolytic carbon optical anisotropy and density have been made on a series of tri-isotropic (TRISO) coated particles prepared for the United States Department of Energy’s Advanced Gas Reactor Fuel Development and Qualification (AGR) program. These measurements show the effect of varying the deposition conditions, especially the deposition temperature, on the density and optical anisotropy of the carbon layers. Additional heat treatment studies of the coated particles at various stages illustrate the strong effect of post-deposition thermal processing on these two pyrolytic carbon properties. Such post-deposition heat treatment occurs during SiC deposition and fuel compact firing, resulting in increased anisotropy and density of the pyrolytic carbon layers.     

A conceptual design study on innovative FBR

An Actinide Recycle Reactor (ARR) with ductless fuel assemblies and mixed nitride fuel is studied in accordance with an Advanced Fuel Recycle System. The core is designed so that yield more economical efficiencies (high breeding ratio and high burnup), safety aspects (high Doppler reactivity coefficient, low void reactivity coefficient and reactor dynamic characteristics) in comparison with mixed oxide or metal fuel on a suitable condition. Preliminary calculations about key parameters of the core design performances had been done to compare with mixed oxide or metal fuel. Results that the mixed nitride fuel with a sodium bond and ZrH has promising capacity.     

在线粘度仪的信号处理和控制

本文介绍了在线粘仪简单可原理和特性，由框图描述了信号处理和控制电路，并采用汇编和Ｃ语言软件进行了电路特性在线测试。     

日本动燃团沥青固化示范工厂着火／爆炸事件分析及教训

本文介绍了日本动燃团沥青固化示范工厂着火／爆炸事件的过程,并讨论了事件发生的原因,提出了值得吸取的教训。