Measurement of Neutron and Gamma-Ray Absorbed Doses Under Criticality Accident Conditions at TRACY Using Tissue-Equivalent Dosimeters
Measurement of Neutron and Gamma-Ray Absorbed Doses Under Criticality Accident Conditions at TRACY Using Tissue-Equivalent Dosimeters
复制标题
使用组织当量剂量计测量 TRACY 临界事故条件下的中子和伽马射线吸收剂量
DOI：
10.13182/nse01-a2232
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发表时间：
2001
影响因子：
1.2
通讯作者：
Noboru Soramasu
中科院分区：
文献类型：
--
作者：
Hiroki Sono;H. Yanagisawa;Akio Ohno;T. Kojima;Noboru Soramasu
为了评价临界事故中人体对中子和γ射线的吸收剂量，在日本原子能综合研究所瞬态实验临界装置（TRACY）上，用高分子丙氨酸剂量计和7 Li_（211）B_4 O_7热释光剂量计对10%浓缩硝酸铀酰溶液进行了剂量学实验。对于实验，进行了五个瞬态操作，通过改变反应性添加的条件来模拟临界事故。使用聚合物-丙氨酸剂量计成功地测量了1.5至1600戈伊范围内的中子和γ射线的高剂量。通过使用7 Li 211 B4 O 7 TLD，能够在1至900戈伊的范围内确定γ射线剂量。此外，它被证实，剂量成正比的综合功率在瞬态操作过程中，虽然反应性添加的条件是不同的。由于7 Li 211 B4 O 7对γ射线的灵敏度几乎与丙氨酸的灵敏度相同，因此通过从聚合物-丙氨酸剂量计的中子和γ射线剂量的总和中减去7 Li 211 B4 O 7 TLD获得的γ射线剂量，可以容易地评估中子剂量，而无需任何复杂的校正。MCNP 4 B程序计算分析的结果还发现，计算的剂量与测量的剂量在95%的置信区间内一致，MCNP 4 B适用于瞬态期间中子和γ射线吸收剂量的评估。
Abstract To evaluate neutron and gamma-ray absorbed doses in human bodies at criticality accidents, two kinds of tissue-equivalent dosimeters, a polymer-alanine dosimeter and a thermoluminescent dosimeter (TLD) made of 7Li211B4O7, were applied to dosimetry experiments with ˜10% enriched uranyl nitrate solution at the Transient Experiment Critical Facility (TRACY) in the Japan Atomic Energy Research Institute. For the experiments, five transient operations were conducted to simulate criticality accidents by varying the conditions of reactivity addition. Very high doses from both neutrons and gamma rays were successfully measured in the range of 1.5 to 1600 Gy by using polymer-alanine dosimeters. The gamma-ray doses were able to be determined in the range of 1 to 900 Gy by using 7Li211B4O7 TLDs. In addition, it is confirmed that the doses are proportional to integrated power during transient operations although the conditions of reactivity addition are different. Since the sensitivity of 7Li211B4O7 to gamma rays is almost the same as that of alanine, the neutron doses are easily evaluated without any complicated correction by subtracting the gamma-ray doses obtained by the 7Li211B4O7 TLDs from the sum of neutron and gamma-ray doses by the polymer-alanine dosimeters. As a result of computational analyses by the MCNP4B code, it is also found that calculated doses agree with measured ones within 95% confidence intervals and the MCNP4B is applicable to the evaluation of neutron and gamma-ray absorbed doses during the transient.