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,当退火制度为500 ℃×2 h时,残余应力处于最低水平;退火后,板的微观结构以小角度晶界为主,且随着退火温度的升高,小角度晶界的密度先增加后趋于稳定。     

后处理料液中氨基羟基脲的分析方法研究

氨基羟基脲（HSC)是一种新型无盐试剂，可有效控制后处理料液中钚的价态，实现铀钚分离和钚的纯化，其含量是后处理工艺中的一项重要参数，需准确测定。本文根据HSC和对二甲氨基苯甲醛在酸性介质中发生显色反应，其产物在456 nm处有明显特征吸收的原理，建立了后处理料液中HSC的分析方法，并优化了分析条件。结果表明，HSC的最佳测量条件为测量波长456 nm、显色温度80～90 ℃（水浴）、显色时间45 min，NO^-₂、Fe³⁺、N₂H₄及其他共存组分对HSC的测定无明显影响。HSC含量在2.5～60.0 μmol/L范围内符合比耳定律，R²为0.999 9，测量精密度优于5.0%。在45 d内工作曲线重复性好，具有良好的长期稳定性。真实样品的测量结果显示，其重加回收率在99%～105%之间。     

The economic competitiveness of promising nuclear energy system: A closer look at the input uncertainties in LCOE analysis

In this study, we aimed to provide important information about the potential economic benefits and risks of nuclear electricity generation associated with existing and prevailing nuclear technologies and to examine the economic effects of nuclear fuel cycle strategies in Korea. An economic analysis model that evaluates the overall life‐cycle costs of nuclear energy systems coupled with multiple fuel cycle options was specially developed by using the levelized cost of electricity (LCOE) as the fundamental methodology. This model is capable of identifying a range of techno‐economic uncertainties underlying each individual nuclear energy system taking into account the state of the art in fuel cycle technologies. It can also quantify and incorporate the resulting impacts into a system‐wide LCOE distribution for each fuel cycle option based on Monte Carlo probabilistic simulation. We analyzed and discussed examples of the economic performance of 13 promising candidates for nuclear energy systems integrated with extensive fuel cycle technologies (including one direct disposal and 12 specific reprocessing and recycling fuel cycle options). We also conducted a sensitivity analysis to investigate the major sensitivity factors of the system component cost in each fuel cycle option and their impacts on individual economic performances. Furthermore, a closer look at the techno‐economic uncertainties of advanced fuel cycle technologies in a break‐even analysis offers evidence of the potential economic feasibility and cost‐reduction opportunities in the reprocessing and recycling options relative to the direct disposal of spent nuclear fuel.     

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.     

Behavior of Ruthenium in Fluoride-Volatility Processes, (IV)

The products of the following three reactions were studied in relation to the reprocessing of oxide fuels: (i) fluorination of Ru by F₂ (ii) fluorination of Ru by a O₂-F₂ mixture, and (iii) secondary process of RuO₂-F₂ reaction. The product of Ru-F₂ reaction was only RuF₅; the mass spectrum of RuF₅ was obtained. Fluorination of Ru by a O₂-F₂ mixture resulted in the production of RuF₅ (85～75%) and RuOF₄ (15～25%); these results are different from those reported by earlier workers. The use of radioactive Ru^*O₂ traced the behavior of RuOF₄ in the apparatus. RuOF₄ decomposes on the wall which was not preliminarily coated with ruthenium; refluorination was effective for removal of the deposit. These results suggest that the fluorination of irradiated oxide fuels volatilizes the ruthenium as a mixture of RuOF₄, RuF₅, and a small amount of RuO₄.     

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.     

Removal of Ruthenium from PUREX Process, (II) Fundamental Research of Electrolytic Oxidation of Ruthenium

A new method was developed for the removal of ruthenium from nitric acid solutions. Ruthenium in 3 M nitric acid solution is oxidized electrolytically to ruthenium tetraoxide, and the tetraoxide is extracted into n-paraffin. The ruthenium tetraoxide extracted in paraffin is readily reduced to ruthenium dioxide by the solvent as black suspension, and it can be easily filtered off through an ordinary cellulose fiber filter. By this method, ruthenium, one of the most troublesome fission elements in PUREX Process, can be removed.     

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.