基于133Ba标准源刻度131I取样滤盒γ总探测效率的方法研究

介绍了一种基于133Ba标准源代替131I标准源刻度NaI(Tl)闪烁体探测器对131I取样滤盒的γ总探测效率的原理和方法.并从实验结果上比较了用133Ba标准源、131I标准源做的该总效率的刻度值,两者在6%的范围内一致.     

福岛核事故后上海气溶胶中~(131)I、~(134)Cs和~(137)Cs的来源途径分析

利用对气溶胶中典型放射性核素(~(131)I和134,~(137)Cs)的分析,可以评估福岛核事故产生的放射性物质对上海及全球的大气放射性本底水平造成的影响。本工作结合核事故释放过程、核素的天然衰变以及气象条件等因素,获得核事故期间上海的气溶胶中~(131)I和134,~(137)Cs活度浓度及其比值的分布特征:~(131)I被检出的时间(2011-03-27)早于~(134)Cs(2011-04-06)和~(137)Cs(2011-04-08),~(131)I的活度浓度(0.01~1.20 mBq/m3)比~(134)Cs(0.01~0.58mBq/m3)和~(137)Cs(0.01~0.65mBq/m3)大2~10倍,而且在不同的时间段出现相应的多峰值现象;~(131)I/~(137)Cs活度浓度比值(1.3~10.6)在2011年4月5日之后呈递减趋势,但是~(134)Cs/~(137)Cs活度浓度比值(0.8~2.9)则一直在1.1左右波动。利用HYSPLIT模型模拟放射性气团运移轨迹的分析方法,表明在核事故期间输入到上海的放射性气溶胶的途径有东北和西北两条主要迁移路径。同时通过结合国内相关城市核事故期间大气放射性监测数据,证实了东北路径在中国境内的控制地位。另外,通过总结和分析北半球大气监测数据中~(131)I/~(137)Cs和~(134)Cs/~(137)Cs活度浓度比值最大值的分布特征,验证了日本核事故产生的放射性气溶胶在北半球的传输过程。     

碘[131I]化钠诊断胶囊活度筛选仪的研制

碘[~(131)I]化钠诊断胶囊活度筛选仪为解决微居级碘[~(131)I]化钠诊断胶囊活度测量均匀性问题而研制的自动化设备。碘[~(131)I]化钠诊断胶囊活度筛选仪采用了核探测技术及自动化技术相结合的方式,提高了自动化测量的程度,缩短了活度测量时间,大幅降低人工活度测量的强度,减少了工作人员的操作剂量。     

一种确定活性碳取样滤盒中131I活度的新方法

代替传统的通过效率刻度测量131I活度的方法,借助于MCNP5模拟和求解病态矩阵方程得出131I在取样滤盒中的深度分布和总活度.该方法稳定性良好,无需制备标准样品和效率刻度,有利于快速准确地监测131I的排放；可以给出深度分布,有助于了解131I在取样盒中的物理进程.     

福清核电站1、2号机组首循环燃料组件破口当量判断

福清核电站1、2号机组首循环均出现燃料包壳破损,通过在线啜吸装置和离线啜吸装置,查找出了破损燃料组件,并给出了破损燃料组件的破口当量。针对破损燃料组件的破口当量计算,福清核电站摸索出了一套标准化的判断方法,该方法以法国原子能委员会卡达拉什中心(CEA Cadarache)编制的简易解释指南(下文简称S.I.G.,简易解释指南是法国原子能委员会卡达拉什中心实际试验的总结,给出了一系列133 Xe释放动力学曲线,该曲线适用于微米级别破口当量的判断)。为基础开发,适用于M310和"华龙一号"等采用AFA-3G及其改进型燃料组件的压水堆核电厂。本文结合福清核电站1、2号机组首循环破损燃料破口当量判断的经验,介绍破损燃料组件破口当量的判断方法。     

133Xe和131mXe混合气体刻度氙符合系统探测效率

作为全面禁止核试验条约(CTBT)国际监测系统(IMS)16个放射性核素实验室之一,北京放射性核素实验室对惰性气体氙-符合系统测量方法进行了研究,用于氙同位素(131mXe、133mXe、133Xe和135Xe)放射性活度测量。准确标定系统的探测效率是放射性氙活度准确测量的重点和难点。本文在剖析氙-符合系统测量原理的基础上,研究建立了采用未知活度的133Xe和131mXe混合气体进行-符合系统效率刻度的方法,对北京放射性核素实验室-符合系统测量符合能谱中部分感兴趣区的探测效率进行了刻度。     

海水中~(131)I的快速检测技术

日本福岛核事故发生以后,对海洋环境中关键放射性核素的快速检测技术提出了更高的要求。人工放射性核素~(131)I在核反应裂变产物中活度相对较高且半衰期短,是用来快速评价核污染的关键性核素之一。本工作从亚铁氰化钾、硝酸铜及硝酸银为原料出发,制备出亚铁氰化铜和亚铁氰化银混合(CuFC/AgFC)吸附材料并分散于聚丙烯纤维富集柱上,来实现水环境中~(131)I现场快速富集。通过室内模拟实验发现:在流速为6.25L/min,~(131)I在海水中I~-初始浓度为24μmol/L时,单次吸附效率即可达到50%以上;而当海水中I~-初始浓度为4μmol/L时,一定体积的海水在连续循环8次后(CuFC/AgFC)聚丙烯纤维富集柱吸附效率达到100%。本方法制源时间约40min,测样时间约12~24h,故最快可在30h内完成海水中~(131)I的分析。本方法的检测限与分析周期均低于目前GB/T 13272-1991水中~(131)I的分析标准,极大地提高了~(131)I的分析时间。除此之外,该法可以同时分析海水中的~(137) Cs(~(134) Cs),有望作为淡水和近岸环境中常规监测和应急时对关键核素~(131)I和~(137) Cs(~(134) Cs)进行快速测定的备选方法之一。     

~(131)I治疗青少年甲亢的疗效观察

采用131I治疗了117例青少年甲亢患者,并进行了一定时间的追踪随访,以探讨131I治疗青少年甲亢的可行性。结果显示,在117例患者中,一次治愈76例(占65.0%),好转28例(23.9%),甲减13例(占11.0%),总有效率为88.9%。14岁以下年龄段的患者使用的活度明显低于14岁以上患者组(15～18岁),但两者的治愈率及甲减发生率均无区别;分次治疗可以提高甲亢的治愈率但对甲减发生率无明显影响。因此,131I可以用于治疗青少年甲亢,但在用药剂量上应视年龄给药。     

131I治疗青少年甲亢的疗效观察

采用^131I治疗了117例青少年甲亢患者，并进行了一定时间的追踪随访，以探讨^131I治疗青少年甲亢的可行性。结果显示，在117例患者中，一次治愈76例(占65．0％)，好转28例(23．9％)，甲减13例(占11．0％)，总有效率为88．9％。14岁以下年龄段的患者使用的活度明显低于14岁以上患者组(15～18岁)，但两者的治愈率及甲减发生率均无区别；分次治疗可以提高甲亢的治愈率但对甲减发生率无明显影响。因此，^131I可以用于治疗青少年甲亢，但在用药剂量上应视年龄给药。     

离线啜吸装置在燃料检测中的应用

辐照后的燃料包壳出现破损时,裂变产物(可溶性固体、气体)释放将增加包壳外部环境介质的放射性活度,通过检测环境介质的放射性活度变化趋势可实现对燃料组件破损程度的定量检测。论文介绍了利用离线啜吸法测量133Xe的释放动力学曲线,根据此动力学曲线定量判断破损燃料组件破口的当量直径,并对燃料组件包壳完整性进行评价。     

Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients

During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the ¹³¹I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent ¹³⁵I, ¹³⁵Xe and ⁸⁹Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.     

Fundamental aspects of defective nuclear fuel behaviour and fission product release

An in-reactor research program with individual, purposely defected, nuclear fuel elements has provided a fundamental understanding of the physical processes of fission product release from defective fuel. On the basis of these experiments, an analytical model has been developed to describe the release of radioactive iodine and noble gas from defective fuel into the primary coolant. An analytic treatment has also been used to model the low-temperature release of fission products from small particles of uranium-bearing compounds (uranium contamination) deposited on in-core surfaces. As a result of this study, a methodology is established whereby release from surface uranium contamination can be distinguished from that resulting from fuel pin failure. Application of this work to power reactor operation is discussed.     

Escape of fission products from defective fuel rods of light water reactors

The escape behaviour of various fission product isotopes from defective fuel rods in PWRs and BWRs is analyzed.Diffusion in the UO₂ is the rate controlling step for the release of noble gases from defective fuel rods. The escape of fission iodine from defective fuel rods is controlled by a mechanism which includes migration and additional delay steps, probably in the nature of a chemical reaction.The inferred effective diffusion constants for fission gases are noticeably higher for defective fuel rods than for intact fuel rods. The difference is about two orders of magnitude. The enhancement of diffusion in defective fuel rods is believed to be due to the increase in the -ratio of the UO₂ in the defective fuel rods.     

Modelling the activity of I in the primary coolant of a CANDU reactor

A mathematical treatment has been developed to describe the activity levels of ¹²⁹I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10⁻⁹ has been determined for the coolant activity ratio of ¹²⁹I/¹³¹I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ∼2×10⁻⁸ for ¹²⁹I/¹³⁷Cs. This ratio can be used to estimate the levels of ¹²⁹I in reactor waste based on a measurement of the activity of ¹³⁷Cs.     

Kinetic simulation of fission product activity in primary coolant of typical PWRs under power perturbations

Kinetic simulations of fission product activity in primary circuits of a typical PWR under power transients, has been performed. A detailed two-stage model-based methodology has been developed and implemented in a computer coder FPCART which uses LEOPARD and ODMUG codes as subroutines. For normal constant power operation, results for over 39 fission products show that the activity due to fission products in fuel region of PWRs is dominated by ¹³⁴I which is followed by ¹³⁴Te and ¹³³I. The value of the total fission product activity in fuel region predicted by FPCART code has been found to agree with-in 0.36% range with the corresponding values found by using the ORIGEN-2.0 code. The predictions of FPCART code have also been found in good agreement with the corresponding values found in ANS-18.1 Standard as well as with some available power-plant operation data with 2.4% deviation in the value of specific activity of the dominating fission product ¹³⁴I. The saturation value of the fission product activity in coolant depends strongly on the fuel-clad gap escape rate coefficient () and approaches a maximum value with increasing value of . During power transients, the FPCART predictions have been found in good agreement with the corresponding experimental measurements of ¹³¹I specific activity for Beznau and Surry PWRs.     

反应堆燃料棒~(131)I泄漏危害与防护

反应堆如发生燃料破损,~(131)I等裂变气体会通过破损包壳释放到厂房中增加人员内照射风险。以CPR1000机组为例分析表明:即使1根燃料棒破损也会对工作人员带来内照射风险,破损达运行限值0.25%时,即使投运净化系统,也需对人员采取防护措施。本文结合实际核电厂运行经验探讨了放射性碘危害的控制和防护措施。     

Determination of I activity in the RBMK-1500 main circulation circuit

The purpose of this work was to evaluate the content of difficult to measure isotope ¹²⁹I in the RBMK-1500 reactor fuel-to-clad gap and reactor main circulation circuit (MCC) coolant. To determine fission product (FP) release from the defective fuel, the methodology proposed by Lewis and Husain for the CANDU reactor primary coolant activity prediction was applied. The determined effective diffusion coefficient D′ = 1.2E−09 s⁻¹ of iodine in the RBMK-1500 fuel is higher than the one evaluated for the CANDU fuel 6.8E−10 s⁻¹. Results show that the method developed by Lewis and Husain can be applied for the RBMK-1500 fuel gap and reactor main circulation circuit coolant activity prediction.     

基于点源~(133)Ba刻度~(131)I探测效率的面积加权法

介绍了利用~(133)Ba点源的γ能峰(与~(131)I的γ能量相近)通过面积加权法对NaI(Tl)探测器模拟刻度碘盒中的~(131)I探测效率的原理和方法.其模拟刻度结果与标准刻度方法效率(~(133)Ba和~(131)I碘盒标准源刻度探测效率)进行了比较,其相对误差分别为7.15%和1.34%.研究表明,~(133)Ba点源面积加权法刻度NaI(Tl)探测器对碘盒中的~(131)I探测效率是一种简便而可靠的方法.     

Modeling the activity and scaling factor of I in the primary coolant and CVCS resin of operating PWR

To estimate the activity of ¹²⁹I at the primary coolant and chemical and volume control system (CVCS) resin in Korean pressurized water reactor (PWR) plants, a theoretical methodology was developed on the basis of an existing model of primary coolant activity and new model of CVCS resin activity. In order to reflect the difference between ¹²⁹I and ¹³⁷Cs, the different power-related diffusivities in the defective fuel were derived, and the variable removal efficiency of the CVCS resin for ¹³⁷Cs was applied as a function of the coolant activity ratio of ¹³¹I/¹³⁷Cs. The current computational method was validated by using the measured coolant activities of ¹³⁷Cs, and the results show better agreement than a previously suggested parameter correlation method between ¹²⁹I and ¹³⁷Cs. There was also reasonable agreement in a comparison of the results of the test resin columns of the coolant from the PWR plants of other countries. It was shown that the ratio of the effective removal efficiency of ¹²⁹I and ¹³⁷Cs in the CVCS resin linearly influences the activity ratio of ¹²⁹I/¹³⁷Cs in the coolant, but on the other hand, its influence on the activity ratio in the CVCS resin is relatively less sensitive compared with that in the coolant.     

CPR1000型压水堆14C产生量研究

本文阐述了压水堆中¹⁴C的主要产生机理,利用蒙特卡罗程序MCNP5建立了精确的三维堆芯模型,计算了堆芯各辐照区的47群中子注量率,计算得到一回路冷却剂、燃料芯块和包壳及堆芯上下反射层的¹⁴C产生率和年产生量。结果表明,计算模型、参数及计算假设具有一定的代表性,计算结果适用于CPR1000型压水堆核电机组。