~(133)Ba面源效率转移法刻度~(133)Xe气体源HPGe探测效率

提出使用面源进行效率转移法刻度133 Xe气体体源的探测效率。使用133 Ba面源模拟刻度放射性133 Xe气体体源的探测效率,在刻度的过程中利用137 Cs与133Ba的峰效率比值及133 Ba面源远端无符合相加效应,137 Cs能峰远近都无符合相加效应等,结合软件模拟建立远近距离峰效率比值的关系,来解决133 Ba的符合相加效应。最后,使用面源效率转移法进行气体体源效率刻度,转移法刻度结果和面源模拟法刻度计算值的偏差在±1%以内。     

基于虚拟刻度原理对放射性体源效率刻度方法的发展

采用蒙特卡罗模拟计算的方法,标定能够适用于不同HPGe探测器上的体源虚拟点源位置,需要对点源、体源的探测效率进行模拟计算。通过²⁴¹Am、¹³⁷Cs、⁶⁰Co点源和体源研究了HPGe晶体尺寸、类型对它们的虚拟点源位置的影响,模拟结果表明²⁴¹Am虚拟点源位置随着探测器尺寸、类型的不同呈现明显的差异性,说明虚拟刻度原理对于探测器表面的小体积样品测量在低能区间是不可取的,最后采用¹³⁷Cs、⁶⁰Co源得出体源高度与虚拟点源位置的半经验公式。通过实验室两台HPGe探测系统对尺寸Φ70 mm×65 mm标准土壤体源的探测效率进行计算,与模拟探测效率值、虚拟点源效率表征结果对比分析,验证了拟合一组表征虚拟点源位置的半经验公式是可行的。实验结果表明,对γ能量区间在300～2 000 keV进行体源的虚拟刻度时,选择可靠的定标源即可建立体源和虚拟点源位置的关系。这为解决监测工作中样品与探测器之间重复进行探测效率校准的问题提供了新的途径。     

基于CdZnTe像素阵列探测技术的伽玛源成像

根据高能射线针孔成像理论,采用CdZnTe像素阵列探测器建立了直接成像探测模式的伽玛源针孔探测系统。测试分析了CdZnTe像素阵列探测器的能量分辨力及峰值效率,讨论研究了针孔成像探测系统的调制传递函数和附加噪声特性,测试获得直径5mm137Cs源的探测图像,采用Lucy-Richardson迭代算法得到了137Cs源的复原图像。实验结果表明:CdZnTe探测器对662keV137Cs源的能量分辨力为6.25%～7.50%,峰值效率65.0%～72.5%;成像系统探测图像存在一定扩散现象,所采用的Lucy-Richardson迭代复原算法能较好地修正图像扩散,提高探测图像中心区域细节分辨力;估算所得137Cs源尺寸误差约0.5mm,所建立的CdZnTe针孔成像探测系统能有效得到小尺寸伽玛源的辐照强度分布及尺寸信息。     

积分中值定理在放射性氙样品γ谱效率刻度技术中的应用

在实验室HPGe探测器测量分析氙气样品过程中需要用到相对方法进行分析, 然而由于放射性氙同位素半衰期较短, 制作相应的标准气体比较难, 给谱仪系统刻度带来困难.针对该问题, 提出使用混合面源模拟刻度四种氙同位素气体源的效率, 并提出将积分中值定理应用到面源刻度实验过程, 大大简化了中间流程, 提高了测量的准确性.利用氙分离纯化系统在高氡环境下采集样品, 获得了天然放射性¹³³Xe气体γ谱, 通过计算得到其活度浓度. 积分中值位置处的混合面源刻度结果和气体源刻度结果一致, 说明本文提出的面源刻度技术及积分中值定理的思想在HPGe 探测器效率刻度测量分析中是有效可行的. 关键词： 面源 无源效率刻度软件 效率刻度     

基于放射性气体源体积的虚拟源刻度技术

大气中放射性气体氙同位素的活度浓度值是判断核裂变反应的关键数据, 长期以来其准确测量一直是个难题. 针对该问题, 本文提出和定义了虚拟点源的概念, 并使用LabSOCS软件模拟了不同尺寸气体体源和不同高度点源的探测效率, 根据计算的数据建立了气体源体积和虚拟点源位置的函数关系. 理论上证实了气体源体积和虚拟点位置有良好的线性关系, 在理论上为解决虚拟源刻度技术提供了新的途径.     

基于虚拟点探测器的虚拟点源效率刻度技术

主要针对尺寸为?75 mm×25 mm的241Am标准体源HPGe探测器探测效率展开研究,通过LabSOCS软件模拟及蒙卡计算研究发现,使用同一种结构类型的探测器测量特定形状的体源,其虚拟点探测器位置与虚拟点源位置满足线性、二次函数、指数等关系。针对确定的探测器,通过点源在对称轴线上不同高度位置实验就可以标定出虚拟点探测器位置,再结合它们的线性函数进一步计算出虚拟点源位置,最后将点源放置于虚拟点位置刻度即可得到体源的探测效率。因此,该技术方法首次提出通过虚拟点探测器位置找到了虚拟点源的位置,是一种全新的刻度方法,已达到实际应用的程度,可以大面积推广应用该技术成果。     

γ分布与特征峰绝对探测效率关系的Monte-Carlo模拟

用γ谱法评估放射性物质的量时，必须获得相应的探测效率。当测量比较复杂体系中的放射性物质的量时，被测对象的材料密度、材料组分以及源分布状态的差异使得用标准源刻度法是非常困难的。因此，无标定量分析方法具有更适用的价值。该方法的理论基础是光子的输运理论，而解决光子输运问题最好的方法是蒙特卡洛(Monte-Carlo，简称MC)模拟方法。MC法能够逼真地描述真实的物理过程、在解决光子的输运问题时具有无可比拟的优点。为此，用MC方法来模拟γ源分布与特征峰绝对效率的关系具有重要的意义。对假定长方体内的点源、线源、面源和体源，以几种形式稀疏分布(均匀、抛物)，模拟计算了相应条件的特征射线的绝对探测效率，并讨论了模拟计算绝对探测效率与能量、吸收体厚度、源分布和探测几何位置等关系。模拟计算程序结构见图1。     

不同射线能量及线衰减系数条件下体源探测效率的函数模型

基于放射性废物桶内所含介质材料和放射性核素分段均匀分布的假设, 采用蒙特卡罗程序模拟计算多种射线能量、多种样品密度条件下的线衰减系数和体源探测效率, 通过多元非线性拟合得到以线衰减系数和能量为自变量的体源探测效率函数。基于壳源法, 通过多个点源组合测试近似替代均匀体源的实验测量, 对3种能量射线和3种密度的样品进行验证实验, 利用蒙卡模拟与数值拟合结合方法和实验测量两种方式获得各自的体源探测效率, 完成活度估算。两种方式活度估算结果相对误差均小于23%。通过蒙卡模拟与数值拟合结合方法所得结果与实验结果基本吻合, 证明了该方法的有效性。     

基于蒙卡模拟的分段γ扫描无源效率刻度方法

针对200 L核废物桶分段γ扫描(SGS)过程中的效率刻度问题,提出了一种效率刻度函数模型,采用MCNP程序计算不同基质密度和γ射线能量条件下的离散断层效率,经过多元非线性回归获取函数参数,从而建立效率刻度函数,实现核废物桶SGS断层效率刻度。对核废物桶样品进行实验分析,结果表明:对于桶内基质分别为密度0.310 g·cm^-3的硅酸铝、密度0.595 g·cm^-3的木质纤维,桶内核素分别为活度3.110×10^5 Bq的点源137Cs、活度1.371×10^5 Bq的点源60Co,在桶内仅有单个点源存在的核素分布极端不均匀情况下,桶内核素活度重建误差在-37.68%~31.52%范围内。本文的方法能够准确有效实现核废物桶SGS断层效率矩阵计算,并确定核废物桶内放射性核素活度,满足实际检测要求。     

井型高纯锗探测器体源样品无源效率刻度

采用蒙特卡罗程序MCNP建立物理模型,对井型高纯锗探测器的效率进行虚拟刻度。模拟计算在密度为0.4,1.2 gcm-3的6种不同成分环境样品中,探测器对射线的探测效率,当能量高于0.10 MeV时,体源样品的探测效率主要与样品密度、射线能量相关。以土壤、水和植物油样品为代表,结合所选取的函数模型,确定了密度范围0.1~1.6 gcm-3的固体源、密度1.0 gcm-3的水溶液和密度0.92 gcm-3的油溶液体源的效率函数及参数。实验采用标准源,对探测效率模拟结果进行了验证,探测效率的模拟值与实验值符合较好,二者误差均在3%以内。说明MCNP程序可以较为准确地模拟计算井型高纯锗探测器对射线的探测效率,验证了该无源效率刻度方法的准确性和可行性。     

Optimum condition of efficiency functions for HPGe γ-ray detectors in the 121-1408 keV energy range

The optimum condition of three commonly used functions in the Genie 2000 γ spectra analysis software have been studied in the 121-1408 keV energy range.The three functions are applied for fitting the full-energy peak effciency of the HPGe gammaray detector.A detailed procedure to obtain the optimum condition is described.The HPGe detector is calibrated at 11 cm by three radioactive sources of point form(152Eu,137Cs,60Co) providing 11 energy peaks.After data processing,results shows that the three functions used in the Genie 2000 gamma spectra analysis software fit best at orders 3-5.Lastly the standard radioactive source 133Ba is chosen to validate the results.Differences between the standard activity of 133Ba and the result obtained from the fitting functions are below 1.5%.Therefore the optimum orders of the three functions used in the Genie 2000 γ spectra analysis software are 3-5 with the 11 energy peaks.     

132，134，136，137Ba（n，p）、132，134Ba（n，2n）反应截面测量

本文叙述１４．２±０．１ＭｅＶ中子能量的１３７Ｂａ（ｎ,ｐ）１３７Ｃｓ、１３６Ｂａ（ｎ,ｐ）１３６Ｃｓ、１３４Ｂａ（ｎ,ｐ）１３４Ｃｓ、１３２Ｂａ（ｎ,ｐ）１３２Ｃｓ、134Ｂａ（ｎ,２ｎ）１３３Ｂａ和１３２Ｂａ（ｎ,２ｎ）１３１Ｂａ等反应截面的测量,测量方法采用以５４Ｆｅ（ｎ,ｐ）５４Ｍｎ和５４Ｆｅ（ｎ,ａ）５１Ｃｒ反应为标准截面的相对活化法,并用现有的数据对本结果进行了比较． The cross sections of the 137Ba (n, p) 137Cs, 136Ba (n, p)136Cs, 134Ba (n, p ) 134Cs, 132Ba (n, p) 132Cs, 134Ba(n, 2n) 133Ba and 132Ba(n, 2n ) 131Ba reactions have been measured by using the activation method relative the cross sections of 54Fe(n, p ) 54Mn and 54Fe(n,a)51Cr reactions at the neutron energy of 14. 2 ±0. 1MeV. The results obtained are compared with the data on hand.     

Nuclear model calculations on the production of <Superscript>125,123</Superscript>Xe and <Superscript>133,131,129,128</Superscript>Ba radioisotopes

In this study, production rates of ^125,123Xe and ^{133,131,129,128}Ba medical isotopes produced by ¹²⁷I(p, 3n)¹²⁵Xe, ¹²⁷I(p, 5n)¹²³Xe, ¹³³Cs(p, n)^133mg Ba, ¹³³Cs(p, 3n)^131mg Ba, ¹³³Cs(p, 5n)¹²⁹Ba, and ¹³³Cs(p, 6n)¹²⁸Ba reactions have been investigated up to 100 MeV incident proton energy. The preequilibrium calculations involve the hybrid model, the geometry-dependent hybrid model and the cascade exciton model. The calculated results are compared with the experimental data taken from the literature.     

Applicability of an X-ray and gamma-ray portable spectrometer for activity measurement of environmental soil samples

This paper presents a study on the applicability of a commercial X-ray and gamma-ray portable spectrometer (Rover) for in situ gamma-ray spectrometry. This portable spectrometer consists of a 3 × 3 × 1 mm (CdTe) Cadmium Telluride detector and a 30 × 30 mm [NaI (Tl)] Sodium Iodide detector. The radioactive sources used were ²⁴¹Am, ¹³³Ba, ¹⁵²Eu and ¹³⁷Cs tablet type, sealed with aluminum and polyethylene, as well as soil samples contaminated with ¹³⁷Cs. With the aid of these radioactive sources, this study determined the efficiency curve for both detectors. In addition, measurements were carried out to identify the minimum detectable activity (MDA) for both detectors, using time acquisition change ranging from 900 s to 138 h, depending on which detector was utilized and the sample distance. The results for the tablet-type sources were as follows: The MDA for the CdTe detector, while positioned 4.15 cm from the ¹³⁷Cs source, was 15 kBq and 6 kBq for energy lines 32 keV and 661.65 keV, respectively. However, when the distance between the source and detector was 100 cm, the 661.65 keV line presented a MDA of 68 kBq. Results for the soil samples were as follows: Using the CdTe detector, positioned 4.15 cm from the source, the MDA was 73 Bq for the ¹³⁷Cs 32 keV line. In relation to the ⁷Be soil samples, the MDA was 301 Bq when the detector was 4.15 cm from the source. Using the NaI (Tl) detector to analyze the ¹³⁷Cs 661.65 keV line, the MDA was 8 kBq when the detector was 100 cm from the tablet-type source. For the soil sample containing ¹³⁷Cs, the MDA was 7.4 Bq when the source was 2.8 cm from the NaI (Tl) detector. For those samples containing ⁷Be, and measured at the same distance (2.8 cm), the MDA was of 9.6 Bq. Based on the minimum detected activities obtained for both detectors, it is concluded that the in situ gamma-ray spectrometric system used to quantify soil sample activities with ¹³⁷Cs and ⁷Be is only appropriate when those activities are around one or more orders of magnitude larger than the usual radioactivity levels found in the environment. Even though it is a commercialized machine, the Rover system’s manufacturer does not provide the information cited in this abstract.     

Magneto-optical trap and mass-separator system for the ultra-sensitive detection of <Superscript>135</Superscript>Cs and <Superscript>137</Superscript>Cs

We report the first magneto-optical trapping of radioactive ¹³⁵Cs and ¹³⁷Cs and a promising means for detecting these isotopes at ultra-sensitive levels by coupling a magneto-optical trap (MOT) to a mass separator. A sample containing both isotopes was placed in the source of a mass separator, ionized, mass-separated, and implanted in a Zr foil within the trapping cell. After implantation, atoms were released from the foil by inductive heating and then captured in a MOT that used large-diameter beams and a dry-film-coated cell to achieve high trapping efficiency. Trapped-atom numbers in the case of either isotope ranged from 10⁴ to 10⁷, as determined from the MOT fluorescence signal. Over this trapped-atom range, the MOT fluorescence signal was found to increase linearly with the number of atoms implanted in the foil and without isotopic bias to within 4%. In principle, this method can then provide a measurement of the ¹³⁷Cs/¹³⁵Cs ratio accurate to within 4% through the direct ratio of MOT fluorescence signals. The fluorescence signal from stable ¹³³Cs, when implanted and released from the foil, was suppressed relative to MOT signals by more than seven orders of magnitude when the system was tuned to trap ¹³⁵Cs or ¹³⁷Cs. When combined with the isotopic selectivity of ≥10⁵ for the mass separator, the overall suppression of ¹³³Cs is expected to exceed 10¹². At present our system delivers atoms from sample to MOT with an efficiency of 0.5%, has a trapped-atom detection limit of 4000 atoms, and achieves a sample-detection sensitivity of one million atoms. Received: 23 August 2002 / Published online: 8 January 2003 RID="*" ID="*"Corresponding author. Fax: +1-505/667-0440, E-mail: mdd@lanl.gov     

面向个人剂量计校准的小尺度伽玛参考辐射模拟研究

针对伽玛射线个人剂量计基于标准伽玛参考辐射进行校准时检定效率低、校准工作复杂和需要远程送检的关键技术问题,建构了1 Ci 137Cs放射源小尺度参考辐射场物理模型,采用蒙特卡罗方法,研究了小尺度参考辐射场内的剂量分布、装置结构和待检剂量计变化导致散射射线对剂量场的影响,获得了待检剂量计形状、数量、类型和装置结构产生的散射伽玛射线对小尺度参考辐射量值定度的影响结果。研究结果表明,1 Ci 137Cs可以为小尺度参考辐射辐射场检验点提供1.5 mSv/h的伽玛遂行剂量率,辐照个人剂量计载台直径30 cm束斑上的剂量率相对标准偏差约为0.48%。当载台厚度为20 mm时,散射射线对小尺度参考辐射检验点处剂量率值的影响率为3.27%,高于剂量计尺寸（1.62%）和剂量计数量（0.56%）的影响。