System for high-level radioactive waste using microchannel chip — extraction behavior of metal ions from aqueous phase to organic phase in microchannel

In order to clarify the extraction behavior of U(VI) from aqueous phase to organic one in microchannel, we have carried out extraction experiments of U(VI) from HNO₃ aqueous solution of 3 M (M = mol/dm³) to 30% or 100% TBP phase in microchannel. From the results of extraction experiments, it was found that the extraction of U(VI) in microchannel could be performed in a short time for approximately 1 s with a good extractability in both organic phases of 30% and 100% TBP, and suggested that the other nuclides could be extracted with high extraction efficiency in microchannel. Furthermore, it is expected that the innovative and sophisticated nuclide separation systems can be developed by using microchannel extraction with selective extractants for specific nuclide.     

Phase identification,microstructural characterization,phase microanalyses and leaching performance evaluation of SYNROC-FA crystalline ceramic waste form

SYNROC-FA, a crystalline ceramic waste form designed to contain 50 wt% Amine process, uranium-rich, high-level, radioactive waste for ultimate deep-geologic disposal, has been characterized using X-ray diffractometry (XRD), scanning electron microscopy (SEM) and transmission electron microscopy (TEM). Phase identification was carried out using XRD, electron diffraction in TEM, and backscattered electron imaging in SEM. Phase microanalyses were carried out using energy-dispersive X-ray analyzers (EDX) during SEM and TEM examinations. X-ray diffraction and grain microanalyses using EDX revealed the existence of a pyrochlore-structured phase CaU(Ti³⁺, Ti⁴⁺)₂O₇, perovskite (Ca, U)(Ti³⁺, Ti⁴⁺)O₃ and uraninite (U, Ca, Ti)O₂, while Ba-hollandite Ba(Al³⁺, Ti³⁺)₂Ti₅O₁₄ was identified using only XRD.The leaching resistance of SYNROC-FA was determined by carrying out a modified MCC-1 leach test in a simulated Canadian Shield groundwater at 90°C for 120 days. The normalized leach rate of Ba was $\text{6 × 10}{}^{\mathrm{-3}}\text{g · m}{}^{\mathrm{-2}}\text{· d}{}^{\mathrm{-1}}$ while the concentrations of U and other simulated fission products in the leachants were below the detection limits of inductively coupled plasma spectrometry and atomic absorption techniques. The leach rates of U and Ti were estimated to be less than $\text{6 × 10}{}^{\mathrm{-5}}$ and $\text{3 × 10}{}^{\mathrm{-5}}\text{g · m}{}^{\mathrm{-2}}$, respectively.     

Use of poly(methyl methacrylate) in radioactive waste management: I. Radiation stability and degradation

In this study, in order to understand the possible use of PMMA in radioactive waste management as a solidifying agent, radiation stability of the PMMA was studied by gamma irradiations at two different dose rates of 1485 and 82.8 Gy/h. The total dose of irradiation was up to 523 kGy. Degradation nature was tested by studying the changes in mechanical and thermal properties with rate and total dose of irradiation. Ultimate tensile strength and toughness first increased and then decreased with total irradiation dose. Half value dose (HVD) for elongation was 148 kGy and it was 178 kGy for tensile strength at the dose rate of 1485 Gy/h. Half value dose was found from the extrapolation of experimental data as 228 kGy for elongation and 205 kGy for tensile strength at the dose rate of 82.8 Gy/h. The FTIR spectral analysis showed depolymerization degradation of polymer with irradiation. It was concluded from experimental results that PMMA can be used for embedding radioactive wastes.     

可用于远程辐射探测系统的数据采集与监控平台的开发

针对目前企业、工厂的分布式布局以及远程监控的需要,同时基于产权的纠纷考虑,开发了充分利用现有可用的设备和技术,现场采用80x86硬件,软件系统MSDOS,应用软件自行开发,远程采用windows平台商用微机,利用oracle数据库读取配置,通过电话灵活监控远程现场的数据采集监控系统.该系统通过少许修改即可应用于多种场合,如机场监控、调度监控等.     

Flashing-induced density wave oscillations in a natural circulation BWR—mechanism of instability and stability map

The SIRIUS-N facility was designed and constructed for highly accurate simulation of core-wide and regional instabilities of a natural circulation BWR. A real-time simulation was performed in the digital controller for modal point kinetics of reactor neutronics and fuel-rod conduction on the basis of measured void fractions in reactor core sections of the thermal-hydraulic loop. Stability experiments were conducted for a wide range of thermal-hydraulic conditions, power distributions, and fuel rod time constants, including the nominal operating conditions of a typical natural circulation BWR. The results show that there is a sufficiently wide stability margin under nominal operating conditions, even when void-reactivity feedback is taken into account. The stability experiments were extended to include a hypothetical parameter range (double-void reactivity coefficient and inlet core subcooling increased by a factor of 3.6) in order to identify instability phenomena. The regional instability was clearly demonstrated with the SIRIUS-N facility, when the fuel rod time constant matches the oscillation period of density wave oscillations.     

Forced convection heat transfer in rectangular ducts—general case of wall resistances and peripheral conduction for ventilation cooling of nuclear waste repositories

A numerical solution for laminar flow heat transfer between a flowing gas and its containing rectangular duct has been obtained for many different boundary conditions which may arise in nuclear waste repository ventilation corridors. The problem has been solved for the cases of insulation on no walls, one wall, two walls, and three walls with various finite resistances on the remaining walls. Simplifications are made to decouple the convective heat transfer problem from the far field conduction problem, but peripheral conduction is retained. Results have been obtained for several duct aspect ratios in the thermal entrance and in the fully developed regions, including the constant temperature cases. When one wall is insulated and the other three are at constant temperature, the maximum temperature occurs in the fluid rather than on the insulated wall. This maximum moves toward the insulated wall with increasing axial distance. Nusselt numbers for the same constant flux on all four walls with peripheral conduction lie in a narrow band bounded by zero and infinite peripheral conduction cases. A dimensionless wall conduction group of four can be considered infinite for the purpose of estimating fully developed Nusselt numbers to within an accuracy of 3%. A decrease in wall and bulk temperatures by finite wall conduction has been demonstrated for the case of a black body radiation boundary condition. Nusselt numbers for the case of constant temperature on the top and bottom walls and constant heat flux on the side walls exhibited unexpected behavior.     

船用堆辐射屏蔽优化设计平台开发与验证

船用堆对核反应堆屏蔽设计提出了更高的要求,传统辐射屏蔽设计方法及设计软件已不能满足要求。为了得到更加精确的辐射屏蔽设计,本文基于开源的SALOME框架建立了一套集“几何建模-材料建模-屏蔽优化-结果可视化”功能为一体的船用堆辐射屏蔽多目标优化平台——MOSRT。MOSRT平台可实现屏蔽结构三维CAD实体建模、基于遗传算法的辐射屏蔽多目标优化以及屏蔽计算结果剂量场三维可视化。基于Savannah和MRX船用堆模型对MOSRT平台进行了辐射屏蔽优化验证,优化方案与初始方案相比,在剂量、质量、体积方面均得到了良好的优化效果,证明了MOSRT平台初步具备辐射屏蔽优化设计功能,可为船用堆工程及概念屏蔽设计提供辅助设计手段。      

Static, stability, and dynamic analysis of shells of revolution by numerical integration — A comparison

The use of numerical integration for the analysis of practical shell-of-revolution structures was documented almost simultaneously in the United States by three independent groups of researchers (Cohen, Kalnins, Mason et al.). These early efforts have been refined, reformulated, and increased in scope and applicability to become major program systems (SRA, Kalnins, STARS). While all three programs utilize basically the same mathematical formulation for integrating the shell differential equations, the matrix solution procedures from this point are basically different.The purpose of this paper is twofold, as follows: (1) to present the differences in solution procedures of the largest system (the Grumman — NASA STARS) from the other two, and point out the inherent advantages of this approach; and (2) compare the numerical integration procedure, as utilized in the STARS, with finite difference and finite element procedures, noting the relative advantages of each in the analysis of shells of revolution for static, buckling, and dynamic loadings.To fulfill the above purpose, a brief review of the numerical integration procedure for the analysis of shells of revolution is presented, and the matrix solution procedures of the SRA, Kalnins, and STARS programs are contrasted. The limitations imposed by the relative procedures are discussed. The unique formulation utilized by STARS for the solution of stability and vibration problems, and its advantages, are discussed in detail. The STARS program's analytical capabilities, capacity, and user options are compared with those of other major systems utilizing either finite differences or finite elements for the analysis of shells of revolution. Comparisons are made in terms of program size, program accuracy, number of degrees of freedom required for analysis, ease of idealization and user inputs, limitations imposed on analysis capability or output, running time, and so forth.All advantages and differences are demonstrated by use of solutions for realistic shell problems in the areas of statics, stability (including dead and live load distributions), vibrations, and dynamic response of shells subjected to time-dependent loadings.     

Pyrochlore-structured titanate ceramics for immobilisation of actinides: Hot isostatic pressing (HIPing) and stainless steel/waste form interactions

A pyrochlore-structured titanate ceramic has been studied in respect of its overall feasibility for immobilisation of impure actinide-rich radioactive wastes through the hot isostatic pressing (HIPing) technique. The resultant waste form contains mainly pyrochlore (∼70%), rutile (∼14%) as well as perovskite (∼12%), hollandite (∼2%) and brannerite (∼1%). Optical spectroscopy confirms that uranium (used to simulate Pu) exists mainly in the stable pyrochlore-structured phase as tetravalent ions as designed. The stainless steel/waste form interactions under HIPing conditions (1280 °C/100 MPa/3 h) do not seem to change the actinide-bearing phases and therefore should have no detrimental effect on the waste form.     

Development of a simple reprocessing process using selective precipitant for uranylions — Engineering studies on systems for precipitating and separating

The simple reprocessing method based on precipitation using N-cyclohexyl-2- pyrrolidone (NCP) as a selective precipitant for U and Pu ions from dissolved solutions of spent FBR fuels has been developed. On the basis of fundamental studies on precipitation behaviors, we designed and built up the scaled-up laboratory equipments (1/20-scale of future plant capacity of 200 tHM/y) to evaluate technological and economical feasibility. This system, which mainly consists of a precipitator and a precipitate separator, should be operated continuously from economical reasons. From the experimental results, it was confirmed that the precipitator is capable of producing UO₂²⁺-NCP precipitate stably with stipulated residence time (approximately 30 min), and the precipitate separator has the highly efficient separation of precipitate from the slurry. Furthermore, the parametric experiments indicated that the rinsing operation increased the efficiency in decontamination of FP elements.     

辐射监测设备维修信息管理系统的设计与开发

介绍了一款辐射监测设备维修信息管理系统的开发过程及采用的方法.包括设计目标、设计构想和数据库的构建,软件的结构和主要界面.该系统减轻了操作人员的工作量,同时有效地提高了积累数据的利用效率.     

对CANDU燃料组件结构的一维瞬态辐射程序开发与验证

在气冷CANDU式燃料组件之中,辐射换热也是不容忽视的一部分。特别是在出现了系统失压/失流事故时,辐射换热将会成为保证燃料安全的主要冷却手段。本文中针对CANDU式压力管编制了针对压力管几何条件下的一维辐射换热瞬态程序。程序中采用将燃料元件棒转化为同心圆环的方式简化辐射角的计算,并加入了隔层辐射模型,使模型更加贴近实际。采用分别将程序中的几个模块的计算结果与CFX计算结果对比的方式来达到程序验证的目的,验证结果显示程序RHTPB具有良好的表现,能够满足于反应堆安全计算的需要。     

辐射加工级电子束吸收剂量量热计的研制

叙述了辐射加工级电子束吸收剂量量热计的设计原理,给出了用自行研制的石墨和盒式水量热计测定电子束吸收剂量的方法和结果。用不同吸收体材料和几何参数是热计测量参考材料的吸收剂量,所得结果在０．８％以内符合。并对现有的吸收剂量深度分布测量模体的关键结构作了改进。由测量结果刻度ＧＡＦ－ＤＭ－１２６０和ＦＷＴ－６０辐射显色薄膜得到的电子束吸收剂量响应曲线与＾６０Ｃｏγ射线响应一致。研制的量热计及其测试技术可作     

A comparative study of engineering options for the disposal of high-level radioactive waste with regard to thermal effects and chemical degradation

The disposal of high-level radioactive waste in deep geological repositories requires stable and foreseeable physical conditions over very long time scales. During this period, the chemical stability of both the natural and the engineered barriers is governed by thermally activated processes. These in turn are driven by the heat pulse generated by the nuclear decay of waste products. The technical concept to cap the temperature peak in the repository is thus an important aspect for the proof of safety of disposal facilities. It is shown that densely stocked repositories, as currently foreseen in several countries, do not necessarily represent the optimal choice with regard to temperature effects, long-term reaction kinetics and chemical degradation of components. It is suggested that the optimization of the temperature peak rather than the fulfillment of cut-off conditions for peak temperature should be a cardinal issue in engineering concepts.     

从系统设计分析核电站辐射监测仪表的特点和发展趋势

基于核电站辐射监测系统设计,从辐射监测仪表的类型范围、功能、工程应用以及质量、价格和售后服务角度,对工作中接触较多的国内外主要辐射监测仪表供货商提供的连续辐射监测仪表的现状进行了比较.结合工程设计经验,分析了辐射监测仪表的特点和设计选型原则,探讨了辐射监测仪表的发展趋势和辐射监测仪表国产化应对策略,为新建核电工程辐射监测系统设计选用辐射监测仪表,实现辐射监测仪表制造国产化提供建议和参考.     

Ultimate heat sink for PHWRs—evaluation of existing methods and alternative proposals

Two methods are proposed for removal of decay heat after a reactor trip under loss of class IV power. One is based on natural circulation (thermosyphon) cooling while the other depends on the direct introduction of the standby cooling system (SCS) heat exchanger after the main pump coast-down. The present analysis shows that under bottled condition thermosyphon cooling is adequate to remove up to 10% full power without boiling and up to 12% power with boiling in the primary coolant channels. However, the direct introduction of the SCS obviates the uncertainties of thermosyphon and ensures positive flow driven by a pump available on class III power supply. This flow is so chosen that there is no boiling on the secondary side of SCS within the operable pressure range. The analysis shows that such an operation does not induce undue stresses in the equipment.     

Thermal and stability considerations for a supercritical water-cooled fast reactor with downward-flow channels during power-raising phase of plant startup

This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal-hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal-hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied.     

Development of methodology of probabilistic safety assessment for radioactive waste disposal in consideration of epistemic uncertainty and aleatory uncertainty

Probabilistic safety assessment has not been performed for radioactive waste disposal owing to the difficulty of dealing with the probability distributions of the parameters included in the long-term safety assessment of radioactive waste disposal. In this study, we develop a methodology of probabilistic safety assessment in consideration of both epistemic uncertainty and aleatory uncertainty, in which the probability density function (PDF) and cumulative distribution function (CDF) of the maximum annual dose can be calculated. We also propose an approach to demonstrating dose assessment results in compliance with the stepwise target annual doses of likely and less-likely scenarios according to the occurrence probability of the scenario without classifying the probabilities of parameters involved prior to the safety assessment. For the likely scenario, we can employ the larger of the modal value of the PDF and the 50th percentile of the CDF to meet the target annual dose (10 µSv y^-1). For the less-likely scenario, we can adopt the 95th percentile of the CDF as the assessment result for comparison with the target annual dose (300 µSv y^-1).     

光致荧光剂量计β辐照器的研制及性能研究

本文针对空间辐射、核动力远洋、核应急等特殊环境下个人及环境剂量监测量值无法实时校准的问题,研制了一种便携式的β辐照器,可携带至现场对光致荧光剂量计（OSLD）进行校准,为特殊环境下剂量监测的准确性提供计量保障。β辐照器采用可替换的⁹⁰Sr-⁹⁰Y平面电镀源作为校准源,并可溯源至标准β辐射场,通过电磁阀控制辐照动作,其总质量小于3 kg。研究结果表明：β辐照器周围剂量当量率处于环境水平;参考点平均剂量率为0.060~0.083 mGy/s,其相对标准不确定度为6.9%;辐照剂量重复性为3.9%（n=10）。     

Rules for design of nuclear graphite core components — Some considerations and approaches

The use of graphite as a structural element presents unusual problems both for the designer and stress analyst. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission.