ISOCS在铀同位素丰度比测量中的应用

本文介绍了应用就地计数系统(ISOCS)测量235U和238U丰度比的方法。该方法使用基于蒙特卡罗原理的无源效率刻度软件对235U和238U子体234mPa进行效率刻度,采用平衡铀测定方法计算出235U和238U的原子比。使用该方法对某一铀样品进行了分析,其测量结果与质谱法测量结果的相对偏差为2.2%。     

SIMS测量铀氧化物中氧同位素比的方法研究

在核法证领域,氧同位素比有可能作为铀矿石、铀氧化物等初级核材料地域来源的一个辅助判据。本文研究了用SIMS测量铀氧化物中氧同位素比时的质量刻度、一次离子束强度、质量分辨率等仪器参数对二次离子发射、氧同位素比测量等的影响因素,优化了仪器测量参数,建立了SIMS测量铀氧化物中氧同位素比的测量方法。在不同时间对同一样品进行了分析,测量结果基本一致,测量相对标准偏差小于1.6%。建立的方法将来可应用于核法证对铀氧化物中氧同位素的分析需要,为核材料/放射源的溯源提供辅助判据。     

Utilization Factor of Nuclear Material in LMFBR

The effects of γ-irradiation on a simulated nuclear waste glass were studied by electron spin resonance spectroscopy (ESR), and were compared with the results on silica glass and Pyrex glass. Three kinds of glasses were γ-irradiated up to the dose of 1.22 MGy and the ESR spectra were obtained. The intensity of ESR spectra were obtained as a function of irradiation dose and annealing temperature.

The spectrum of the waste glass was characteristic of two typical peaks, Peak 1 was the strong resonance at g=4.3 showing the existence of four coordinated Fe³⁺ and Peak 2 was the weak and broad resonance at g= 2.0 showing the existence of six coordinated Fe³⁺. The ESR spectra of the waste glass before and after γ-irradiation were almost overlapped and a little difference only in the intensity was observed. While in silaca glass and Pyrex glass, the peaks from E'γ center and boron-oxygen hole center (BOHC) were observed to arise after irradiation. The absolute intensity of. Peaks 1 and 2 described above changed in complicated way depending on the dose. The result suggests oxidation or reduction of iron takes place in the waste glass depending on the dose. The isochronal annealing of irradiated glasses shows most of γ-ray- induced damages in the waste glass are restored even at room temperature, although most of the damages in silica glass and Pyrex glass are disappeared at the temperature from 550 to 600 K. The results show that the waste glass with a few weight percent of iron is resistent to radiation than other commercial glasses.     

浅谈除核电外核燃料循环中物料泄漏事故的隐患排查

本文首先归纳国内外除核电外核燃料循环中典型的物料泄漏事故,为达到"全面深入、细致彻底、不留死角、不留盲区"的物料泄漏事故隐患排查目标,采用概率安全分析中始发事件分析、失效模式和影响分析、人因分析思想对除核电外核燃料循环各个主要环节进行了事故隐患分析,得到了各环节较为全面的事故隐患清单,可供相关环节的安全检查和隐患排查工...     

乏燃料后处理设施核材料近实时衡算系统概念与设计

为指导后处理设施设计阶段统一部署核材料衡算、在设施运行阶段实施近实时衡算、及时反馈工艺的运行状态和趋势并探知异常情况,保障核设施核材料安全,本文在开展乏燃料后处理设施核材料衡算评价及关键技术研究的基础上,深入调研分析了核材料近实时衡算技术现状,梳理了Purex流程中核材料平衡区内过程监控的重要设备和先进仪器,以及一体化数据信息系统结构及其运行维护需求,提出我国开展乏燃料后处理近实时衡算技术研究的必要性和技术配置建议：结合传统的平衡区划分及关键测量点设置方式,以核材料重要量为目标,增补适宜的在线监控点和战略观察点,采用模型开发验证、分析方法优化评估、信息系统整合技术,在后处理设施全寿命周期内统筹管理控制Purex工艺中设备、管道、阀门、储槽中的核材料,达到近实时衡算目标。     

核仪表在钢铁工业的应用与发展

核仪表(核子控制系统)具有非接触测量、精度高、反应快、稳定性好,采用计算机技术实现数字化、智能化、网络化的特点,因此已越来越广泛地应用在生产过程复杂、自动化程度高、现场环境比较恶劣的钢铁制造流程在线检测,并在实时控制系统中作为不可缺少的基础仪表提供各种重要检测数据。本文主要介绍核仪表在国内外钢厂应用现状、发展、技术与经济效益。     

电位滴定法测定基准U_3O_8中铀的不确定度评定

分析了电位滴定法测定基准U3O8中铀总量的检测过程,建立了测量过程分量的数学模型,讨论和比较了检测过程中各种不确定度因素,发现各不确定度因素中称样量的不确定度贡献最大,其次是测量重复性和消耗重铬酸钾质量的不确定度,K2Cr2O7相对分子质量和U的相对原子质量对不确定度结果的影响可忽略。结果表明,采用电位滴定法测定基准U3O8中铀总量,当样品中铀的总量为84.795%时,扩展不确定度为0.022%(k=2)。     

XMA Analysis of Undissolved Materials Observed in NSRR Short Irradiated Uranium Dioxide Fuel

After a short irradiation of UO₂ fuels with and without SiO₂+ Al₂O₃ additive in NSRR, a comparative study of the fission products (FP) dissolved in nitric acid and undissolved materials was performed. Results of the study were: (1) The additive fuel behaved similarly to non-additive fuel. (2) In both types of fuels, the undissolved materials in dried out solution were not FP but metals existing from the fuel fabrication stage. (3) The NSRR short irradiation had great influence on the shape of the undissolved components without changing the as-fabri-cated compositions.     

离子液体体系的萃取行为及其在乏燃料后处理中的应用前景

离子液体由于其特有的性质,在乏燃料后处理中的应用已受到广泛关注。本文综述了不同种类离子液体中多种萃取剂对乏燃料所含若干锕系元素及放射性裂片元素的萃取,重点分析了不同萃取体系的萃取效率、萃取选择性、萃取机理和反萃等关键问题。综合目前的研究成果,可发现:离子液体-萃取剂体系由于其独特的萃取机理,通常比传统萃取体系具有更高的萃取效率;一些萃取体系具有高选择性使其在乏燃料后处理中有很好的应用前景。在简要介绍阳离子交换机理、阴离子交换机理和中性复合物机理三种离子液体体系萃取机理的同时,重点总结了萃取中三相问题和协同萃取效应。本文还总结了液-液反萃、超临界CO2反萃和电化学反萃三种常见的反萃方法,讨论了各自的优缺点。本文最后对离子液体在乏燃料中的应用进行了总结与前景展望。     

Overview oil Development of High Reliablility Nuclear Power Plant in Japan with Particular Emphasis on Fuel and Materials

To begin with the concept of nuclear safety in Japan, efforts to achieve power reactor operation with higher reliability and safety have been discribed. More strict regulatory criteria, standards as well as the better quality control with excellent workmanships have led to this world's best performance today.

Successful development work of fuel and materials, in fundamental work as well as in practical work through the co-operation among research organizations, university people, reactor, fuel and cladding tube manufacturers together with electric utilities has contributed a great deal to achieve this.

Improvement and standardization work followed by the development of ALWR and then next-generation LWR have been going on by reflecting our experiences during 18 years of operation as well as the results of our technological development in fuel and materials.     

Computation on Fuel Particle Size Capable of Being Regarded as Homogeneous in Nuclear Criticality Safety Analysis

For powdered fuel processed in the nuclear fuel facilities, flooding is often thought to be the severest condition regarding the nuclear criticality safety evaluation. The reactivity of such a heterogeneous system as powdered fuel in water should be almost equal to that of the homogeneous one, when the fuel particle size is very small. The neutron multiplication factor was calculated for an infinite cubic array of slightly enriched UO₂ sphere particles immersed in water with various enrichments, water to fuel ratios and fuel particle sizes. The calculations were performed with a computer code module based on the collision probability method to solve the ultra-fine energy group equations of neutrons. The change in the neutron multiplication factor from the homogeneous system is dominated first by the change in the resonance escape probability and second by the change in the thermal utilization factor; these changes and therefore their sum, depend almost completely on the mean uranium concentration (or water to fuel volume ratio) and rarely on uranium enrichment up to 10 wt% for a fuel particle size of 1mm. The dependence determines the fuel particle size regarded as homogeneous in proportion to the negligible relative error of the neutron multiplication factors.