Leaching of Unirradiated Mox Fuel in Sea Water

Abstract

MOX fuel, in the form of pellets or crushed powder, has been leached in sea water under various conditions to simulate and determine the potential effects of a transport accident leading to a breaching of the fuel cladding and subsequent attack by sea water. The simulation involved the immersion of MOX in sea water over a period of 11 months and measuring the leach rates by analysing the leachate water at intervals for Pu and Am and also, at the end of the tests, for U. The Pu and Am leach rates obtained were equivalent to a surface removal rate of ～0.3 μm per century for leachates adjusted to pH 4 and to less than 0.03 μm per century for samples leached at a more realistic pH 8. The Pu and Am concentrations in the leachates were probably limited by their solubilities and by sorbtion on the surrounding surfaces as would presumably occur in any real case.     

核燃料包装容器桶体缝焊工艺可靠性建模

从缝焊工艺的质量要求出发，详细分析了缝焊工艺可靠性的影响因素，针对焊透率与重叠量建立了工艺可靠性概率模型，给出了工艺可靠性试验方案与工艺可靠度计算步骤。     

Computerised Modelling of Radiolysis and Corrosion Effects in Transport Containers for Fuel Spent Nuclear

Abstract

The radiolysis of water and/or gases within transport containers for spent nuclear fuel may result in the generation of hydrogen and oxygen gases and also the enhanced corrosion of the materials in contact with the water. These effects are important, particularly when the fuel container is also used for storage post-transport prior to reprocessing or disposal. The behaviour of a range of radiolytic systems has been studied. Plant behaviour has been simulated in numerous laboratory experiments: plant and experimental results have been linked by a computerised model describing the radiolysis mechanism and predicting the quantities and production rate of gaseous and corrosive species. This allows prediction of plant performance over a long time scale. The model is based on a well-accepted radiolysis mechanism supplemented with specific measurements made at the Harwell laboratory. Model capabilities include inert atmospheres, materials corrosion, variations in water and gas volumes or aqueous chemistry. The model has been applied to design stage radiolysis assessments of transport containers; information from operating plant has been interpreted to advise on design improvement, e.g. diminution of gas production using easily corroded scavengers to remove oxygen. Radiolysis in gas filled dry storage containers for spent nuclear fuel has been studied; corrosive product production (e.g. nitric acid), which is important for fuel cladding integrity has been assessed. The development and use of this computerised model is described with a current summary.     

核电站钢制安全壳焊后热处理仿真技术研究

利用SYSWELD软件模拟安全壳筒体第1大环和第2大环之间环焊缝的焊后热处理过程,分析局部整圈和局部分段热处理后安全壳的残余应力及变形情况;通过实测残余应力与数值模拟残余应力结果的对比,证明数值模拟结果的正确性和可靠性,同时结合工程实际情况,证明局部分段热处理方法的合理性,并确定了合理且经济的局部热处理分段方案。研究结果为大壁厚、大体积的钢制安全壳热处理提供了一种有效的途径。      

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。     

Radiation Effects on the Long-term Ageing of Spent Fuel Storage Containers

Abstract

The properties of a long-term interim storage low-carbon steel container could be changed by radiation-induced damage. In this paper, the gamma and neutron flux in the container are estimated over 300 years for pressurised water reactor spent fuels. Gamma radiation is the main contributor to the damage flux during the whole storage period (from 3 × 10^?8 dpa y^?1 after 2 years of storage to 4 ×10^?11 dpa y^?1 after 300 years, where dpa is displacements peratom). A cluster dynamics model is used to calculate the point defect concentrations and their effect on the copper precipitation kinetic and point defect clustering for different storage conditions. The shear stress increases in the first years of storage, from 23 to 50 MPa for a Cu content varying from 0.05 to 0.25%. The temperature or the damage rate has an effect on the kinetic only at the beginning of the storage. The hardening is entirely caused by copper precipitation enhanced by vacancy supersaturation.     

后处理中试厂放射性流体输送设备应用总结

核燃料后处理厂处理的对象具有高放射性,输送放射性流体应尽量采用免维修或易维修设备.我国核燃料后处理中试厂采用了多种放射性流体输送设备,例如:空气升液器、蒸汽喷射泵、真空虹吸装置、特殊机械泵、扬液器、戽斗流量计、压空喷射器、罗茨鼓风机等.本文结合这些流体输送设备在中试厂主工艺厂房的应用情况和调试经验,对这些设备的特点、使用要求和实际运行情况进行了技术总结,对各种放射性流体输送设备性能进行了评价,对放射性流体输送设备的选型提出了一些建议,最后提出一些能够在大厂放化厂房中推广使用的一些设备的改进方向,为大型核燃料后处理厂等项目中放化厂房设计及相关科研提供参考.     

An Experimental Study of Spray Cooling in Nuclear Reactor Containers

As a contribution to the safety analyses of condition following loss-of-coolant accidents in light water reactor, experiments were performed to obtain data on container spraying and on accompanying heat transfer characteristics, by means of a one-tenth scale model container. The loss of coolant was simulated injecting high temperature saturated water into the model container in which air had been enclosed at atmospheric and room temperature, and then the steam-air mixture was cooled by sprays of water.

The rate of heat absorption by the spray was obtained from both stationary and transient data, and it was confirmed that the absorption becomes incomplete as the container pressure is lowered.

The evaporation of the water collected at the bottom and of that from sprayed walls of the container was derived from container pressure variation in time, by measuring its rate of change immediately following stoppage of the spray; the rate of evaporation was found more active as the rate of spray cooling increases.

The evaporation from the sprayed walls was obtained from thermal data taken on a copper cylinder placed in the container, and it was found that the evaporation becomes more active as the pressure increases.     

New Approach to the Nuclear Fuel Reprocessing in Non-acidic Aqueous Solutions

A new nuclear fuel reprocessing method based on the anodic dissolution of spent fuels in aqueous alkaline solutions (Na₂CO₃-NaHCO₃) has been proposed. Experiments of the anodic dissolution were performed by using a simulated spent fuel in a Na₂CO₃-NaHCO₃ solution. Uranyl ions produced anodically were present in the solution as stable carbonato complexes, and at the same time, most of the simulated fission products (FP) were precipitated as hydroxo or carbonate compounds. Under this condition, Cs of an alkali metal group was dissolved in the solution and precipitated by adding sodium tetraphenylborate. Uranyl ion was recovered as hydroxo compounds by adding NaOH to the solution after removing precipitates of the simulated FP. In view of waste disposal, ⁹⁹Tc having a long half-life should be removed. Precipitation behavior of Tc(VII) was examined by using Re(VII) as a simulant of Tc(VII). It was found that Re(VII) species are completely removed as a precipitate by adding tetraphenylphosphonium chloride. A large amount of Na used in the present method was recovered as NaHCO₃ by blowing CO₂ into alkaline solutions. As a result, it was clarified that the proposed method is fundamentally possible as a new reprocessing method.     

新型无源法核燃料棒~(235)U富集度检测系统研制

介绍了新型无源法核燃料棒~(235)U富集度检测系统的研制,该系统采用4通道多探测器设计,能在核燃料棒匀速通过检测体时有效探测核燃料芯块中~(235)U核素自发衰变产生的γ射线,实时生成~(235)U富集度分布图谱,进而判断各通道核燃料棒质量是否合格并实现好坏料自动分选。该系统经实验验证在单通道检测速度为1 m/min时,能够准确测量核燃料棒~(235)U富集度值并判断棒中是否混有异常芯块。该系统已应用于核燃料元件生产线上。     

六边形套管型燃料堆芯核设计专用程序试验验证

采用六边形套管型燃料堆芯（HCTFR）7个零功率物理试验方案的试验数据对核设计程序（CELL+CPLEV2）的计算精度进行工程验证。验证结果表明,7个临界试验方案的临界棒位有效增殖因子（k_eff）计算偏差均在±0.8%以内,与试验结果符合较好,控制棒价值和停堆深度计算偏差也都在可接受范围内,表明CELL+CPLEV2程序具有较高的计算精度和可靠性,可用于HCTFR的核设计。      

铀同位素标准物质在核燃料系统实验室间比对中的应用

在核燃料系统分析检测实验室间比对活动中,首次开展铀同位素检测项目的设计、策划和实施,选取国家一级标准物质“八氧化三铀中铀同位素标准物质”作为分割样品对,分别采用Z比分数、标准物质测量结果的误差允许限(E)作为能力统计量,对实验室间铀同位素丰度的比对结果进行统计和分析,对测量结果有问题或不满意的情况进行分析和建议.通过首...     

燃料破损在线探测系统在核电厂的应用与分析

国内两台核电机组分别安装了1套由中国原子能科学研究院研制的燃料破损在线探测系统（FDDS）。通过FDDS对两台核电机组发生燃料破损时的连续监测和分析,表明FDDS在核电厂一回路放射性核素活度浓度在线测量及燃料破损监测中发挥了良好的作用,弥补了化学取样分析方法的不足。     

Experience in the European Transport of Spent Nuclear Fuel

Abstract

Nuclear Transport Limited has been responsible for most of the European transport of spent nuclear fuel which has taken place to date, and therefore has unique experience in the field. The services and experience of Nuclear Transport Limited cover a whole range of flask types, large quantities of fuel transported, the design and provision of handling equipment, the logistics of operation, the arrangements necessary to maintain high standards of safety, and the need to alleviate public concern. Transport routes and communications systems have been developed, crossing national boundaries in Europe using road, rail and sea, currently employing the freight rail ferry for the short haul crossing of the English Channel, in contrast to the special purpose ships which are operated by PNTL for the long haul crossing of the ocean from Japan to the United Kingdom. Whilst there has never been a serious accident involving a spent fuel flask, procedures have been established to ensure an effective response to any accident or incident. This approach is consistent with the standards of safety that apply throughout the nuclear industry.     

核燃料循环中的技术与工业演变

本文介绍了从天然铀开采到燃料生产直到后处理的整个燃料循环中的技术和工业演变。核燃料循环的焦点是燃料组件的设计和制造。燃料制造业最重要的发展是以轻水堆 MOX 燃料形式出现的钚再循环。高杰马集团在燃料循环工业中处于领先地位。     

Radiometric Assessment of Nuclear Fuel Burnup in the Romashka Reactor

Atomic Energy - Methods of determining the mass of the burned 235U from the thermal energy released by the reactor and from indirect measurements of fission product activity are considered. A...     

Structural and Fuel Materials for the Fuel Elements of Thermionic Nuclear Power Systems

The characteristic properties of hardened single-crystal alloys (Mo-Nb, W-Nb, W-Ta) and UO_2-x with open thermally stabilized porosity are presented. The combination of the indicated materials makes thermionic fuel elements very stable with a long service life. It is shown that the use of porous uranium dioxide under temperature conditions resulting in restructuring makes it possible to decrease the size of the columnar grains and increase the creep rate of the dioxide with optimal porosity and pore size. A model oxide fuel, which has been developed, with fission-product simulators for rapid determination under extra-reactor conditions of the properties of uranium with deep burnup and for performing accelerated reactor tests is described. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 365–371, November, 2005.