Study of radiation dose reduction of buildings of different sizes and materials

The dependence of radiation dose reduction on the sizes and materials of buildings was studied by numerical analyses using the Monte Carlo simulation code, PHITS. The dose rates inside the buildings were calculated by simulating gamma-ray transport from radioactive cesium deposited at the ground surface. Three building models were developed: the wooden house, the open-space concrete building, and the thin-wall building, to study the effect of building size and construction material on dose reduction inside these structures. Here the floor-area sizes of the building models were varied to clarify the influence of building configuration on dose reduction. The results demonstrated that the dose rates inside the buildings linearly decreased with increasing floor area on a logarithmic scale for all types of buildings considered. The calculated dose distribution inside a building indicated that the distance from the outer walls was a determining factor for the dose rate at each position in the building. The obtained tendency was verified by comparison with data reflecting the dose reduction of typical buildings in Japan.     

聚变驱动次临界堆第一壁材料辐照损伤的初步研究

介绍了中子对材料的辐照损伤原理及化合物原子平均离位(DPA)截面计算方法;使用辐照损伤计算程序SPECTER计算了聚变驱动次临界堆(FDS-I)第一壁材料CLAM钢的辐照损伤参数,并将CLAM钢的辐照损伤计算结果与相同条件下316SS、SiC等聚变堆结构材料的计算结果进行了比较.     

An evaluation of activation and radiation damage effects for the European Spallation Source Target

The details concerning the generation and accumulation of radionuclides produced by spallation reactions within the neutron generation target to be installed at the European Spallation Source are described. The resulting radiation damage for the target tungsten blocks constituents as well as the target encasing is evaluated and explicit results are given for the effects of knock-on atoms as well as those resulting from the accumulation of light and volatile hydrogen and helium isotopes resulting from spallation reactions. Their effects on the mechanical properties are discussed on the light of available experimental data and, on this basis, an estimation of the target lifetime is given.     

Simulation study of personal dose equivalent for external exposure to radioactive cesium distributed in soil

The personal dose equivalent was calculated for the public (newborns; 1-, 5-, 10-, and 15-year-old children; and adults) in an environment contaminated with radioactive cesium (¹³⁴Cs and ¹³⁷Cs) distributed in a soil at specific depths of 0.0, 0.5, 2.5, 5.0, 10.0, and 50.0 g/cm², and the conversion coefficients for converting activity concentration to personal dose equivalent rate were provided for each age group. Monte Carlo calculations were performed using pediatric and adult computational phantoms incorporated into a particle and heavy ion transport code system. Compared with the effective dose and ambient dose equivalent at a height of 100 cm above the ground, the personal dose equivalent was found to provide a good estimate as a measurable quantity for the effective dose and did not exceed the ambient dose equivalent even in the environmental radiation fields, while the personal dose equivalent values increased for younger subjects. The weighted-integral method to obtain the personal dose equivalent for a volumetric source was applied to the analysis of exponential radioactive cesium distributions in the soil observed in Fukushima, and the calculation results successfully reproduced the measured data.     

Progress and prospects of calculation methods for radiation shielding

Progress in calculation methods for radiation shielding are reviewed based on the activities of research committees related to radiation shielding fields established in the Atomic Energy Society of Japan. A technological roadmap for the field of radiation shielding; progress and prospects for specific shielding calculation methods such as the Monte Carlo, discrete ordinate Sn transport, and simplified methods; and shielding experiments used to validate calculation methods are presented in this paper.     

现场校准用便携式X射线照射装置的优化设计及辐射特性研究

中国原子能科学研究院计量测试部研制了一款用于校准现场固定式X、γ辐射剂量仪的便携式X射线照射装置。首先利用蒙特卡罗软件建立模型,对出射口准直光阑结构进行优化设计,随后,对所建参考辐射场射束范围、均匀性及散射辐射进行模拟计算,并利用TW32005电离室进行了实验验证。在本研究所选辐射质、管电流及参考点-焦斑距离条件下,所建立的辐射场能量范围为60～164 keV,空气比释动能率在0.08～565 mGy/h,周围剂量当量率在0.13～892 mSv/h,为后续利用便携式X射线照射装置开展现场校准技术研究奠定了基础。结果表明,经优化设计后的准直光阑在满足准直限束需求的同时有效减轻了自身重量,便携式X射线参考辐射场特性满足GB/T 12162.1—2000要求,验证了所建模型的正确性及蒙特卡罗方法用于便携式X射线参考辐射场特性研究的有效性。     

For a better estimation of gamma heating in nuclear material-testing reactors and associated devices: status and work plan from calculation methods to nuclear data

The determination of gamma heating levels in material-testing reactors (MTRs) is of crucial importance as gamma heating affects both safety and performance parameters of MTRs [1 Amharrak H, Di Salvo J, Lyoussi A, Carette M, Reynard-Carette C. State of the art on nuclear heating in a mixed (n/gamma) field in research reactors. Nucl Instrum Methods A. 2014;749:57–67.[Crossref], [Web of Science ®] , [Google Scholar],2 Brun J, Reynard-Carette C, Lyoussi A, Merroun O, Carette M, Janulyte A, Zerega Y, Andre J, Bignan G, Chauvin J-P, Fourmentel D, Gonnier C, Guimbal P, Malo J-Y, Villard J-F. Numerical and experimental calibration of calorimetric sample cell dedicated to nuclear heating measurements. IEEE Trans Nucl Sci. 2012;59:3173–3179.[Crossref], [Web of Science ®] , [Google Scholar]]. The required accuracy (5% at one standard deviation) makes it necessary to calibrate bias and uncertainty associated with MTR gamma-heating calculations. Main steps of bias determination for gamma-heating calculations include, first, the development of a calculation methodology with the controlled use of physical approximations; second, the interpretation of gamma-heating measurements with reference calculations so as to determine bias supposed to be mainly due to nuclear data.     

中子比释动能系数和辐射损伤截面计算程序KDC的研发

为弥补国内在由评价核数据出发计算中子比释动能(KERMA)系数和辐射损伤截面的程序方面的空白,建立了中子KERMA系数和辐射损伤截面计算方法,并基于FORTRAN-90程序语言开发了具有自主知识产权的中子KERMA系数和辐射损伤截面计算程序KDC。另外,针对能量平衡检查过程中发现的能量不平衡问题,提出了一种对不合理KERMA系数进行直接修订的方法,即用运动学上限替代不合理KERMA系数,并在KDC程序中实现了这一修订功能。通过将KDC程序与国际上广泛应用的核数据处理程序系统NJOY中的HEATR模块的计算结果进行比对,验证了KDC程序在计算结果和功能上的可靠性。     

Development of a stochastic biokinetic method and its application to internal dose estimation for insoluble cesium-bearing particles

ABSTRACT

Spherical cesium-bearing particles were found around the Fukushima Daiichi Nuclear Power Plant after its accident. These particles are insoluble with high specific activities. If such a particle is incorporated into the body, it is expected that the radioactive cesium contained in the particle will move as a single particulate material without being distributed throughout the body. Consequently, internal doses vary depending on the pathway of the particle in the body. Therefore, general deterministic methods for dose estimation are not applicable to the intake of small number of cesium-bearing particles. In this study, a Monte Carlo method was developed to evaluate a probability density function of internal doses for inhalation of a particulate material. In addition, a biokinetic model was constructed to consider the insolubility of the particles. Probability density functions of committed equivalent and effective doses were estimated using the constructed biokinetic model with the developed computation method. As a result, the uncertainty of the doses was very large for single particle inhalation. However, it was observed that this uncertainty decreased with increasing number of simultaneously inhaled particles. On the other hand, the consideration about the insolubility made the doses somewhat higher than those based on the general cesium model.