再加工工艺对N18锆合金板残余应力的影响

采用\"热轧+冷轧+退火\"工艺对N18锆合金板进行再加工,通过X射线法分析板的表面残余应力,采用EBSD技术分析晶界取向差角分布和大小角度晶界.结果表明,热轧后N18锆合金板的表面残余应力呈现无规律分布状态;冷轧后板的表面残余应力均为压应力,其大小随着冷轧变形量的增加而增加;退火后板的表面残余应力值处于较低水平,当退火制...     

熔盐电解后处理工艺中的在线监测技术

干法后处理技术是我国核燃料循环体系的重要发展方向,其中有发展前景的熔盐电解法在工业化过程中需要对元素进行准确地监测以保障安全运行。传统离线检测方法程序复杂、数据分析延迟,而在线监测方法因可以获得实时数据而备受瞩目。本文对近几年国内外熔盐电解法中不同元素的在线监测技术进行了总结,重点调研分析了电化学法中的循环伏安法、安培法、脉冲法以及光谱法中的荧光光谱法和紫外光谱法在各种元素的在线监测中的原理、特点和应用前景,以期对熔盐电解中在线监测技术的研发和实际应用提供理论支撑和技术参考。     

利用二次谱盲估计多速率DS/CDMA伪码周期

针对可变扩频长度(VSL)的多速率DS/CDMA信号伪码周期的盲估计问题进行了研究,将一般单速率直扩信号二次功率谱伪码周期估计的方法扩展到多速率DS/CDMA模型。该方法首先将接收的多速率DS/CDMA信号进行采样,并对其求一次功率谱,再将一次谱作为输入信号作傅里叶变换并取模、平方,从而得到信号的二次功率谱。通过推导证明,多速率DS/CDMA信号的二次功率谱在扩频码周期的整数倍处出现尖锐的谱线,且不同速率用户的二次谱线幅度是不同的,利用这些幅度差异区分不同速率并通过估计谱线之间的距离即可获得不同速率信号的伪码周期。仿真表明该算法在低信噪比下适用,在-15 dB能够利用较少数据同时估计多组速率的伪码周期。     

用于乏核燃料后处理的商用 Zr702 在沸腾硝酸溶液中的应力腐蚀开裂行为

研究了商用Zr702在沸腾硝酸中应用于乏核燃料后处理工业的应力腐蚀开裂（SCC）行为和应力腐蚀敏感性（I_SCC）。使用独立设计的慢应变率拉伸（SSRT）系统在沸腾的硝酸中获得应变率为10^-5 s^-1的Zr702的应力-应变曲线。进行微观表征以及数值分析以量化Zr702的SCC行为。结果表明,随着HNO₃浓度的增加,Zr702的I_SCC由5%明显增加到26.67%。商用Zr702在沸腾HNO₃溶液中力学性能的急剧下降归因于试样表面的解理断裂,并且解理裂纹的深度随着HNO₃浓度的增加而增加。最后,提出了预测商业Zr702在沸腾HNO₃溶液中SCC行为的模型和方程。HNO₃浓度、I_SCC、断裂应力（σ_SCC）和解理裂纹深度（d_cc）之间的定量关系可以使用高阶回归方程来描述和预测。     

A Remark to a Formula in K. Ishimatsu's Paper on “Continuous γ-ray Spectrometry with Scintillation Counters”

Spatial distribution of the Cd-ratio of Au foils were measured for various lattice pitches in a natural uranium-light water system. Comparison of the experimental results with theoretical calculation using the THERMOS-code shows some divergence of the theoretical Cd-ratio toward larger than experimental values in the fuel rod.     

An overview of a nuclear reprocessing plant Human Factors programme

This paper presents a case study of a large Human Factors programme applied in the nuclear fuel reprocessing industry (1987-1991). The paper outlines the key Human Factors issues addressed, as well as the impacts achieved, and gives an indication of the resources utilised (approximately 15 person-years of effort). It also considers the starting point of the programme, in terms of the factors that led to the need for such an extensive programme. Some general lessons learned are given at the end of the paper.     

Long Term Change in Level-Volume Relation of an Accountability Tank for Plutonium Nitrate Solution

A slight change in the level-volume relation for an accountability tank for a large amount of plutonium nitrate solution (PuN) was observed at the Plutonium Conversion Development Facility (PCDF) in the Power Reactor and Nuclear Fuel Development Corp. (PNC), Tokai Works. From the results of annual tank re-calibrations for the plutonium receiving tank from 1985 to 1992 using the incremental feed of nitric acid as the density standard, it became clear that the relation between the level and the volume changed slightly, and the rate of the change was a linear function of operating time. Also it became clear that the change was linear in relation to the level. In the PCDF, the cumulative change in the volume at the nominal level was evaluated to be 0.1% during 8 years' operation. It was also evaluated that the repeatability of the re-calibration is much better than 0.1%. A reasonable frequency of tank re-calibration is once every 5 years.     

Isolation of 137Cs with Copper Ferrocyanide-Anion Exchange Resin

The adsorption and desorption of ¹³⁷Cs on copper ferrocyanide-anion exchange resin, prepared in the manner previously reported in this Journal in a Short Note, are presented in detail.

This resin can also be used for the concentration of the ¹³⁷Cs found in sea water. The nuclide is adsorbed effectively on the resin from water, hydrochloric acid below 4 M, and on nitric acid below 1M. After adsorpsion on the resin, ¹³⁷Cs is eluted easily with either nitric acid (>6M), ammonium water, or silver nitrate solution.

Adsorption on the resin is specific for ¹³⁷Cs, and the action is due entirely to the ferrocyanide moiety of the resin.

This method is more efficient than the co-precipitation method with copper ferrocyanide for the concentration of radiocesium from a large volume of sea water.     

Ion Exchange Technology in Spent Fuel Reprocessing

Solvent extraction is the major unit operation employed in spent nuclear fuel reprocessing. The operation yields three streams; fission product waste, uranium product and plutonium product. Ion exchange is primarily used in reprocessing as a tail-end method to concentrate and isolate the plutonium product stream. This review will describe the details of plutonium recovery and purification by both cation- and anion-exchange processing. A brief overview of miscellaneous uses of ion-exchange employed in reprocessing will also be given.