Avtandilashvili Maia, Puncher Matthew, McComish Stacey L, Tolmachev Sergei Y
US Transuranium and Uranium Registries, Washington State University, 1845 Terminal Drive, Suite 201, Richland, WA 99354-4959, USA.
J Radiol Prot. 2015 Mar;35(1):129-51. doi: 10.1088/0952-4746/35/1/129. Epub 2015 Jan 12.
The United States Transuranium and Uranium Registries' (USTUR) whole-body donor (Case 1031) was exposed to an acute inhalation of uranium hexafluoride (UF6) produced from an explosion at a uranium processing plant 65 years prior to his death. The USTUR measurements of tissue samples collected at the autopsy indicated long-term retention of inhaled slightly enriched uranium material (0.85% (235)U) in the deep lungs and thoracic lymph nodes. In the present study, the authors combined the tissue measurement results with historical bioassay data, and analysed them with International Commission on Radiological Protection (ICRP) respiratory tract models and the ICRP Publication 69 systemic model for uranium using maximum likelihood and Bayesian statistical methods. The purpose of the analysis was to estimate intakes and model parameter values that best describe the data, and evaluate their effect on dose assessment. The maximum likelihood analysis, which used the ICRP Publication 66 human respiratory tract model, resulted in a point estimate of 79 mg of uranium for the occupational intake composed of 86% soluble, type F material and 14% insoluble, type S material. For the Bayesian approach, the authors applied the Markov Chain Monte Carlo method, but this time used the revised human respiratory tract model, which is currently being used by ICRP to calculate new dose coefficients for workers. The Bayesian analysis estimated that the mean uranium intake was 160 mg, and calculated the case-specific lung dissolution parameters with their associated uncertainties. The parameters were consistent with the inhaled uranium material being predominantly soluble with a small but significant insoluble component. The 95% posterior range of the rapid dissolution fraction (the fraction of deposited material that is absorbed to blood rapidly) was 0.12 to 0.91 with a median of 0.37. The remaining fraction was absorbed slowly, with a 95% range of 0.000 22 d(-1) to 0.000 36 d(-1) and a median of 0.000 31 d(-1). The effective dose per unit intake calculated using the dissolution parameters derived from the maximum likelihood and the Bayesian analyses was higher than the current ICRP dose coefficient for type F uranium by a factor of 2 or 7, respectively; the higher value of the latter was due to use of the revised respiratory tract model. The dissolution parameter values obtained here may be more appropriate to use for radiation protection purposes when individuals are exposed to a UF6 mixture that contains an insoluble uranium component.
美国超铀元素与铀登记处（USTUR）的全身供体（病例1031）在其去世前65年，因铀加工厂爆炸产生的六氟化铀（UF6）急性吸入而受到照射。USTUR在尸检时对采集的组织样本进行的测量表明，吸入的轻度富集铀物质（0.85%（235）U）长期滞留在深部肺部和胸部淋巴结中。在本研究中，作者将组织测量结果与历史生物测定数据相结合，并使用最大似然法和贝叶斯统计方法，采用国际放射防护委员会（ICRP）呼吸道模型和ICRP第69号出版物中关于铀的全身模型对其进行分析。分析的目的是估计最能描述数据的摄入量和模型参数值，并评估它们对剂量评估的影响。使用ICRP第66号出版物中的人类呼吸道模型进行的最大似然分析得出，职业摄入量的点估计值为79毫克铀，其中86%为可溶性F型物质，14%为不溶性S型物质。对于贝叶斯方法，作者应用了马尔可夫链蒙特卡罗方法，但这次使用的是经修订的人类呼吸道模型，ICRP目前正在使用该模型来计算工人的新剂量系数。贝叶斯分析估计铀的平均摄入量为160毫克，并计算了特定病例的肺部溶解参数及其相关不确定性。这些参数与吸入的铀物质主要为可溶性且含有少量但显著的不溶性成分一致。快速溶解分数（沉积物质中迅速吸收到血液中的部分）的95%后验范围为0.12至0.91，中位数为0.37。其余部分吸收缓慢，95%范围为0.000 22 d⁻¹至0.000 36 d⁻¹，中位数为0.000 31 d⁻¹。使用从最大似然分析和贝叶斯分析得出的溶解参数计算的每单位摄入量的有效剂量分别比ICRP目前针对F型铀的剂量系数高2倍或7倍；后者较高的值是由于使用了经修订的呼吸道模型。当个体暴露于含有不溶性铀成分的UF6混合物时，此处获得的溶解参数值可能更适合用于辐射防护目的。
