伤口摄入241Am内照射回顾性调查及剂量评价

2002年某单位一名工作人员操作241Am固体放射源时,右手食指被螺丝刀穿刺扎伤,伤口摄入颗粒态241Am。2016年对人员伤口摄入241Am进行回顾性调查及剂量评价。对人员头骨中241Am的定量测量约为166 Bq,头骨干重占全身骨干重约18.86%,人员全身骨内241Am约为880 Bq。结合伤口摄入241Am人体骨骼滞留函数,计算人员伤口241Am摄入量约为2 630 Bq,待积有效剂量约为0.85 Sv。对人员伤口内现有241Am滞留量定量测量约为527 Bq,结合伤口污染滞留函数计算伤口241Am摄入量约为5560 Bq,待积有效剂量约为1.81 Sv。综合两种剂量评价方法的结果,该人员伤口摄入241Am待积有效剂量应小于2 Sv。对伤口现有241Am进行干预评估,伤口残留污染物手术切除可节省待积有效剂量约为170 mSv。     

基于尿样测量的~(241)Am内照射常规监测与评价技术研究

~(241)Am属于极毒组α核素,吸入人体将产生内照射危害。尿样核素分析监测技术是工作人员内照射常规监测与评价的重要方法之一。本文通过研究~(241)Am大体积尿样的前处理、分离纯化、电沉积等因素,建立了大体积尿样中~(241)Am的分析监测技术,同时考察了该分析监测技术在~(241)Am内照射监测与评价中的应用。研究结果表明:建立的大体积尿样中~(241)Am分析监测技术化学回收率为86.2%、精密度为3.88%、相对偏差小于10%;对Pu的去污系数大于1.0×10~3;探测限为31.8μBq/L。对于1.6 L 24 h尿样,常规监测周期为360 d时,最小可测待积有效剂量为0.13 mSv,低于常规监测所规定的2 mSv调查水平,表明该监测技术适用于工作人员~(241)Am内照射的常规监测以及剂量评价。     

退役作业中一次异常事件及处理所致241Am内照射剂量的估算

根据某涉及镅的退役作业中一次异常事件及处理相关的4名工作人员尿镅的测量结果,采用多次摄入模式,对由241Am吸入所致的内照射剂量进行了估算.结果表明:最大个人待积有效剂量为9.4 mSv (241Am摄入量:349 Bq),低于年剂量限值(20 mSv)或年摄入量限值(740 Bq).     

用液体闪烁计数法测定^147Pm

用液体闪烁计数法测量含α-羟基异丁酸、HNO_3和~(241)Am等杂质的~(147)Pm。研究了闪烁液用量、α-羟基异丁酸和HNO_3存在量对~(147)Pm测量的影响及有~(241)Am存在时的干扰情况;对~(147)Pm的仪器探测效率进行了刻度。在所选定的测量~(147)Pm条件下,被同时记录的~(241)Am不超过~(241)Am总计数的1%。方法还实测了核燃料后处理高放废液中经化学分离后的~(147)Pm的绝对含量,相对标准偏差<1%。     

~(241)Am俘获反应中~(242)Cm α粒子测量方法研究

241Am发生中子俘获反应,主要生成激发态的242Amg(83.7%),242Amg通过β衰变生成242Cm,242Cmα粒子的最低能量比241Amα粒子的最高能量高约500 keV。用SRIM程序模拟了两种能量的α粒子在铝箔中的穿透情况,26μm的Al箔可以将241Am的α粒子完全吸收掉,242Cm的α粒子对26μm的Al箔穿透率达99.7%。在探测器前加26μm吸收铝箔,可以排除241Amα粒子对242Cm的α粒子测量的影响,实现242Cmα粒子的测量。     

~(243)Am制备中r 核物理测定方法

本文介绍了用几种核物理探测方法,测定用堆照钚靶制备的~(243)Am产品及其杂质核素的含量,也对进堆前的钚料液和极低含量的杂质,特别是~(241)Am进行了测定,测出了存在于~(238)Pu中α放射性约为10~(-6)的~(241)Am的相对含量。文中描述了探测方法及定性定量结果。     

TRPO萃取-电沉积法同时测定食品中的~(239)pu、~(241)Am和~(237)Np

本文叙述了用 TRPO 萃取、TTHA-HLac 分别反萃,电沉积制源、低本底α计数器测定的食品中~(239)Pu、~(241)Am 和~(237)Np 的同时测定方法。18种食品的分析结果表明,该方法对~(239)Pu、~(241)Am 和~(237)Np 的全程回收率分别为:(67.2±3.9)%、(67.0±3.9)%和(71.2±2.3)%;测最时间为48h 时,8g 灰样三种核素的方法灵敏度的均值为35μBq/g(灰);具有良好的去污效果;能满足测定食品中~(239)Pu、~(241)Am 和~(237)Np 的要求。     

DGA树脂萃取色层法分析土壤样品中的~(241)Am

为了对核设施及周边环境的放射性污染水平进行有效监测,需要对大量土壤样品中的~(241)Am进行分析。本工作首先考察了接触时间、介质酸度、温度对~(241)Am在DGA树脂柱上的吸附影响,确定了最佳吸附条件,为动态柱实验提供了输入条件;然后进行了动态柱实验的验证:以萃取色层法为分离纯化手段,采用高酸上柱,EDTA络合解吸,以低本底液闪谱仪为放射性测量手段,建立了一个快速分析土壤中~(241)Am的方法。当土壤样品量为10g,使用8mol/L HNO_3浸取灰化后的模拟污染土壤样品,该方法对~(241)Am的化学回收率约为95%,最低可探测比活度为3.1Bq/kg,全流程分析时间小于3h。对其它核素~(137)Cs、~(90)Sr-~(90)Y、~(239)Pu、~(99)Tc有很好地去污能力。因此该方法可广泛应用于环境土壤样品的~(241)Am分析。     

原子能院钚污染废水处理实验室研究

提出了对原子能院钚污染废水的处理方案。该废液是含有~(239,240)Pu、~(241)Am等α核素的一种去污废液,其α放射性水平约1×10~5Bq/L,不能直接送弱放废水处理车间处理,并已经大罐贮存17年。通过实验室研究对方案中的处理方法加以验证。实验结果表明:在合适条件下,采用化学絮凝沉淀,P301型无机材料吸附或者D001-CC大孔阳离子交换树脂离子交换任何一种方法,经过一级处理,都能使钚污染废水的α放射性水平降低到3.7×10~3Bq/L以下,从而可以满足送往弱放废水处理车间的要求。     

锂玻璃微珠固体闪烁流线监测器

文章描述了一种新型的流线监测器,可直接用于监测高压离子交换流出液中的α和低能β放射性。采用铈激活锂玻璃微珠为固体闪烁体。监测器对~(241)Am,~(242)Cm α射线的绝对探测效率分别达到85.8%,92.8%。对~(14)C,~(90)Sr-~(90)Y的β射线则分别为62.1%和88.6%。在~(241)Am总加入最为7.4Bq,淋洗峰半宽度为0.22ml时,仍能给出清晰的淋洗曲线。     

人体内DTPA的排出及其对~(241)Am排泄量的影响

本文报道给人肌肉注射或雾化吸入 DTPA 后尿中 DTPA 的排出情况。观察结果表明,两种给药方式尿中 DTPA 的排出分数均可用两个指数项之和来描述。肌肉注射给药时,快排部分约占注入量的97%,半减期为88min;慢排部分约占3%,半减期为292min。雾化吸入给药后,快排部分约占81%,半减期为92min;慢排部分约占19%,半减期为624min。DTPA 对~(241)Am 的促排效果是明显的,本实验条件下平均促排倍数大于18。     

~(241)Am标准溶液研制及比对测量

本文系统地研究了~(241)Am的分离、纯化、溶液组成及其稳定性、制源技术和放射性活度绝对测量等问题。由此建立了生产~(241)Am放射性标准溶液的方法,并于1975年起向计量、科研、生产、环保和学校等有关部门提供产品,使用情况良好。1983年又提供国防科工委系统和计量分院进行了比对测量,我们的测量结果在0.2%误差范围内同各单位测量结果平均值相一致。这些都表明本文提供的方法是正确的,由此得到的~(241)Am标准溶液是准确可靠和稳定的。完全符合放射性计量标准的要求。     

Si(Li)探测器低能区的效率刻度

通过测量19keV电子束轰击厚碳靶产生的韧致辐射能谱，在低能区（>1keV）对Si(Li)探测器进行了探测效率刻度。厚碳靶的理论韧致辐射能谱由Monte-Carlo程序PENELOPE计算，并用²⁴¹Am标准放射源确定出效率刻度曲线的绝对值。采用本工作的刻度方法确定的效率刻度曲线误差主要来源于用标准放射源绝对化的误差，约为5%。将所得初步实验结果与理论计算值进行了比较，并采用最小二乘法对探测器各厚度参数进行了拟合，除Au接触层厚度外，拟合的各厚度值与探测器生产商提供的数据较为吻合。     

Solvent Extraction of Americium(VI) by Tri-n-Butyl Phosphate

Solvent extraction of Am(VI) by tri-n-butyl phosphate (TBP) from nitric acid solutions was investigated to develop a novel method for partitioning americium from high level liquid waste generated through spent nuclear fuel reprocessing. Am(VI) was prepared using ammonium peroxodisulfate and silver nitrate. The distribution coefficients of Am(VI) were determined for extraction systems of various concentrations of nitric acid and TBP. Sufficiently stable Am(VI) could be extracted and the extraction reaction of Am(VI) was found to be the same as for other hexavalent actinides. The apparent equilibrium constant varied with the concentration of peroxodisulfate used for the valence control, which was ascribed to the competitive reaction of the extraction of Am(VI) and the complex formation of Am(VI) with sulfate ion produced by the decomposition of peroxodisulfate. A distribution coefficient of Am(VI) above 1 was obtained with undiluted TBP and the separation factor between Am(VI) and Nd(III) was 87±9. TBP extraction of Am(VI), after implementing valence control, was proved to be an effective method for the partitioning of americium from fission products such as rare earth elements.     

An experimental investigation of accumulation and transmutation behavior of americium in the MOX fuel irradiated in a fast reactor

Americium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Americium was isolated from the irradiated MOX fuel by a combined method of anion-exchange chromatography and oxidation of Am. The isotopic ratios of americium and its content were determined by thermal ionization mass spectroscopy and α-spectrometry, respectively. The americium isotopic ratio was similar for all the specimens, but was significantly different from that of PWR-MOX. On the basis of present analytical results, the accumulation and transmutation behavior of americium nuclides in a fast reactor is discussed from the viewpoints of neutron spectrum dependence and the isomeric ratio of the ²⁴¹Am capture reaction. The estimated isomeric ratio is about 87%, which is close to the latest evaluated value. A rapid estimation method of Am content by using the ²⁴⁰Pu to ²³⁹Pu ratio was adopted and proved to be valid for the spent fuel irradiated in the fast reactor.     

Behaviors of Americium in Aqueous Solutions Containing Iron

The effect of iron on the colloidal behaviors of americium in near-neutral solutions was studied by applying the ultrafiltration method to americium (²⁴¹Am) solutions, iron (⁵⁹Fe) solutions and americium/iron mixed solutions. The concentration of americium was typically about 10^?9M and that of iron 10^?7M. Some properties of colloidal particles were deduced from the particle-size distributions. It was revealed that the colloidal americium particles may represent pseudo colloids, which behave together with iron true colloids. The proposed simple model, in which the iron true colloid is assumed to be a chain-like polymer composed of hydroxide, is able to describe the particle-size distribution of iron colloids observed in this work. It was also found that the adsorption of americium and iron on the vessel walls depends on the particle-size distributions.     

^241Am（NO3）3在大鼠伤口局部组织中的扩散

本文介绍~(241)Am(NO_3)_3污染大鼠伤口后在局部组织中扩散的实验观测结果。实验大鼠每3只为一组,共10组,分别于污染伤口后的3小时至90天期间活杀,对伤口局部组织作冰冻切片,测量切片上的γ放射性,观察~(241)Am(NO_3)_3污染伤口后沿水平方向向伤口周围组织及沿垂直方向向伤口深部扩散的范围和速度。实验结果提示伤口的早期处理一般应在污染后6小时之前进行,手术切除范围最少为距伤口边缘8mm、深6mm;中期处理一般应在污染后的5天之前,手术切除范围应在距伤口边缘10mm、深7mm。     

Determination of americium in plutonium based nuclear fuel materials: An assessment of different methodologies

Determination of americium (Am) is one of the requirements of chemical quality assurance of plutonium (Pu) bearing fuel materials. Though many methods are published for determining Am at picogram to femtogram levels in environmental and biological matrices, yet a few of them are used routinely for Pu based nuclear fuel samples. This paper gives a brief summary of the different analytical methods available and presents results of our experiments on the determination of Am in Pu bearing fuels using gamma spectroscopy. The methods utilizing gamma emissions from ²⁴¹Am and Pu isotopes are fast as they do not involve chemical separation of Pu and Am, do not require an accurate knowledge of the efficiency values of the detector systems and are not dependent on the availability of a radiometric standard for ²⁴¹Am. In addition, for aged Pu samples containing large amounts of ²⁴¹Am, there is no need for dilution and this reduces the volume of analytical radioactive waste solution. Future requirements of reference materials to validate different methodologies for determining Am isotopes are also highlighted.     

6名工作人员尿,粪中^106Ru含量的测定及初始体负荷量的估算

本文报道6名工作人员摄入~(106)Ru 后尿、粪中~(106)Ru 含量的测定结果。用硫化铜载体沉淀法测定摄入后9—100天工作人员尿样及部分粪样中~(106)Ru 的含量,设尿、粪比为1,拟合了排泄方程和滞留方程,还估算了初始体负荷量。     

Magnetic Properties of Uranium Carbides after Neutron Irradiation

Fundamental investigations on valence control and solvent extraction of americium were carried out to develop a method for americium separation from reprocessing solution. In order to adjust americium valency from III to IV and VI, (NH₄)₁₀P₂W ₁₇O₆₁ synthesized was used as complexant stabilizing Am(IV). Oxidation behavior of americium was investigated as a function of (NH₄)₁₀P₂W ₁₇O₆₁ americium ratio. Using 0.1M (NH₄)₂S₂O₈ and 0.01M AgNO₃ as oxidation reagent, Am(IV) was obtained quantitatively at the ratio of 15. On decreasing the ratio to 0.6, 92% of americium was adjusted to Am(VI). The concentration of (NH₄)₂S₂O₈ could be reduced to 1/15 compared to the previously reported method in which no complexant was used. Americium(IV) was also prepared by reacting O₃ and AgNO₃ but no Am(VI) was obtained even at low (NH₄)₁₀P₂W ₁₇O₆₁ to americium ratio.

Americium(VI) could be extracted by tri-n-butyl phosphate stably without influence of (NH₄)₁₀P₂W ₁₇O₆₁. The distribution coefficient of Am(VI) was 4 between 100% tri-n-butyl phosphate and 1 M nitric acid, and separation factor from Nd(III) was 50.

With regard to the americium separation method which implemented valence control followed by extraction, adding (NH₄)₁₀P₂W ₁₇O₆₁ led to minimization of waste volume and improvement of extraction efficiency.