Possible use of bisphenol-a polycarbonate in radioactive waste embedding

Frequent use of bisphenol-a polycarbonate in daily life results with a huge amount of polycarbonate waste. The proper utilization of this waste would be an environmental friendly approach. As polymeric materials are candidate materials for radioactive waste embedding, having an aromatic structure within the main chain, the bisphenol-a polycarbonate is a candidate material to be used as an embedding matrix for the confinement of low and intermediate level radioactive waste. This possibility would partly solve the problem of huge amount of waste generated due to the usage of polycarbonate.In this study, the possibility for the usage of bisphenol-a polycarbonate in radioactive waste management was investigated regarding to radiation stability and resistance. For that purpose, changes in the properties of polycarbonate irradiated with different doses were determined. Irradiations, which were carried out for the duration of more than 8 months with an initial dose rate of 993 Gy/h, ended with a total irradiation dose of 4341 kGy to the samples. A detailed characterization of irradiated polycarbonate samples were performed by tensile, DMA, TGA, ATR-FTIR, AFM, EDX and SEM tests. The modifications within the polymer matrix have been reported. The total dose at which bisphenol-a polycarbonate could resist regarding to mechanical properties was determined.     

Use of poly(methyl methacrylate) in radioactive waste management: I. Radiation stability and degradation

In this study, in order to understand the possible use of PMMA in radioactive waste management as a solidifying agent, radiation stability of the PMMA was studied by gamma irradiations at two different dose rates of 1485 and 82.8 Gy/h. The total dose of irradiation was up to 523 kGy. Degradation nature was tested by studying the changes in mechanical and thermal properties with rate and total dose of irradiation. Ultimate tensile strength and toughness first increased and then decreased with total irradiation dose. Half value dose (HVD) for elongation was 148 kGy and it was 178 kGy for tensile strength at the dose rate of 1485 Gy/h. Half value dose was found from the extrapolation of experimental data as 228 kGy for elongation and 205 kGy for tensile strength at the dose rate of 82.8 Gy/h. The FTIR spectral analysis showed depolymerization degradation of polymer with irradiation. It was concluded from experimental results that PMMA can be used for embedding radioactive wastes.     

Monte Carlo simulation of radionuclide transport through fractured media

The presence of fractures within the host matrix may strongly affect the radionuclide mobility in the far-field of the current proposed solutions for underground radioactive waste repositories. In this paper a Monte Carlo simulation approach, based on analogies with neutron transport, has been developed for describing radionuclide transport in fractured media. The Monte Carlo approach here adopted is verified by means of comparison with analytic and numerical solutions for two different case studies. An application to a more complex and realistic case is also reported to show the potential of the proposed approach.     

Optimization of GEANT4 settings for Proton Pencil Beam Scanning simulations using GATE

This study reports the investigation of different GEANT4 settings for proton therapy applications in the context of Treatment Planning System comparisons. The GEANT4.9.2 release was used through the GATE platform. We focused on the Pencil Beam Scanning delivery technique, which allows for intensity modulated proton therapy applications. The most relevant options and parameters (range cut, step size, database binning) for the simulation that influence the dose deposition were investigated, in order to determine a robust, accurate and efficient simulation environment. In this perspective, simulations of depth-dose profiles and transverse profiles at different depths and energies between 100 and 230 MeV have been assessed against reference measurements in water and PMMA. These measurements were performed in Essen, Germany, with the IBA dedicated Pencil Beam Scanning system, using Bragg-peak chambers and radiochromic films. GEANT4 simulations were also compared to the PHITS.2.14 and MCNPX.2.5.0 Monte Carlo codes. Depth-dose simulations reached 0.3 mm range accuracy compared to NIST CSDA ranges, with a dose agreement of about 1% over a set of five different energies. The transverse profiles simulated using the different Monte Carlo codes showed discrepancies, with up to 15% difference in beam widening between GEANT4 and MCNPX in water. A 8% difference between the GEANT4 multiple scattering and single scattering algorithms was observed. The simulations showed the inability of reproducing the measured transverse dose spreading with depth in PMMA, corroborating the fact that GEANT4 underestimates the lateral dose spreading. GATE was found to be a very convenient simulation environment to perform this study. A reference physics-list and an optimized parameters-list have been proposed. Satisfactory agreement against depth-dose profiles measurements was obtained. The simulation of transverse profiles using different Monte Carlo codes showed significant deviations. This point is crucial for Pencil Beam Scanning delivery simulations and suggests that the GEANT4 multiple scattering algorithm should be revised.     

低、中放固体废物处置设施的屏蔽计算

天然屏障和工程屏障相结合的多重屏障设计是近地表低、中放固体废物处置场设计的基本原则。本文从辐射防护角度出发,对某低中放固体废物处置场处置单元的设计进行论证。分别采用点核积分法程序QAD和蒙特卡罗法程序MCNP,对处置单元格不同厚度墙体外的剂量率水平进行计算,确定出合理的处置单元构筑物墙体厚度,并得出两种屏蔽计算方法的适用范围。     

透射式CT与自射线式CT检测核废物桶蒙特卡罗模拟

用蒙特止罗方法对透射式CT（TCT）和自射线式CT（ECT）检测核废物进行模拟实验，用TCT的模拟结果对ECT的模拟结果进行材料吸收系数校正，可以较为准确地确定核废物桶中放射性活度的绝对分布，并可确定放射性核素的种类。     

车载放射性废物体在线监测结果的不确定度评定

基于BECK公式建立了车载放射性废物体在线监测模型,其复杂的测量公式使得公式的求导、各影响因素之间的相关系数难以采用解析计算方法实现,导致利用不确定度传播定律进行不确定度评定存在困难。应用蒙特卡罗不确定度合成方法评价了车载放射性废物体在线监测结果的不确定度,解决了就地γ能谱测量中的不确定度评定问题。其评定流程是首先采用A、B类不确定度评定方法评价各测量参数的不确定度,然后使用蒙特卡罗不确定度合成方法对不确定度分项进行合成。结果表明,241Am测量结果的不确定度<40%(1σ),137Cs、60Co和152Eu测量结果不确定度<34%(1σ)。     

Positron implantation profiles in layered samples

The paper presents theoretical and experimental studies of the positron implantation profiles in layered samples. The Monte Carlo simulations performed using the GEANT4 toolkit reveal accumulation of positrons in the denser layer embedded between two less dense layers. That effect is significant not only for low energy positrons at slow monoenergetic positron beams but also for high energy positrons which are emitted from radioactive nuclei in conventional experiments. Measurements of the positron implantation profile into the samples which consist of silver and aluminium foils of different thickness show profile features which correspond well with those simulated by the GEANT4 toolkit for examined cases. We propose to call this phenomenon the accumulation effect of energetic positrons.     

Effects of CT based Voxel Phantoms on Dose Distribution Calculated with Monte Carlo Method

A few CT-based voxel phantoms were produced to investigate the sensitivity of Monte Carlo simulations of X-ray beam and electron beam to the proportions of elements and the mass densities of the materials used to express the patient‘s anatomical structure. The human body can be well outlined by air, lung, adipose, muscle, soft bone and hard bone to calculate the dose distribution with Monte Carlo method. The effects of the calibration curves established by using various CT scanners are not clinically significant based on our investigation. The deviation from the values of cumulative dose volume histogram derived from CT-based voxel phantoms is less than 1% for the given target.     

Evaluation of performance of a new gamma technique for assay of radioactive waste

In order to increase the accuracy of measurement of radioactive waste drum assay by Segmented Gamma Scanning technique a new measurement technique was proposed. Results of the study with noise data of measuring factors are presented. The results confirm that the technique can be successfully developed for assay of radioactive waste drums.     

放射性高盐废液干燥成盐技术研究

放射性废物处理中，放射性废液的体积和所含放射性总量在“三废”中占比较大，为使废物最小化，本研究围绕放射性高盐废液干燥成盐技术开展技术路线论证、工艺设计，研制放射性高盐废液微波干燥成盐工程样机，并完成样机加工制造、安装调试及性能验证。工程样机验证实验结果表明，装置运行过程稳定，干燥速率约6～8 L/h,产物不含游离水，桶内及桶壁温度最高约100℃,桶内压力在1～2 kPa之间。本研究结果可为放射性高盐废液干燥成盐技术及专用工装设计提供参考，并为后续工程应用奠定基础。     

MONTE CARLO CALCULATION OF ENERGY DEPOSITION BY DELTA RAYS AROUND ION TRACKS

The radial distribution of dose around the path of a heavy ion has been studied by a Monte Carlo transport analysis of the delta rays produced along the track of a heavy ion based on classical binary collision dynamics and a single scattering model for the electron transport process.Result comparisons among this work and semi-empirical expression based delta ray theory of track structure,as well as other Monte Carlo calculations are made for 1,3MeV protons and several heavy ions.The results of the Monte Carlo simulations for energetic heavy ions are in agreement with experimental data and with results of different methods.The characteristic of this Monte Carlo calculation is a simulation of the delta rays theory of track structure.     

Improved analysis for matrix effect correction in LLW neutronic assay

The matrix effect correction for the differential die-away (DDA) measurement is an improvement in the fissile material content determination. In low-level radioactive waste (LLW) packages examination, the most widely used methods are based on neutron flux monitoring with ³He tubes, associated to a “matrix interrogation source” (MIS) originally developed for passive neutron measurement and which determine an experimental detection efficiency. This paper describes two new approaches developed with the goal of increasing the accuracy of the matrix effect correction and reducing the measurement time, which is a major objective in the non destructive assay (NDA) of large number of waste packages. The first method is based on an active prompt neutron coincidence measurement using a new generation list mode data card, which is an alternative to the MIS. Monte Carlo simulations have been performed to determine the correction function parameters. An experimental agreement within 20% is obtained with a fissile sample localized at the centre of different matrices provided that the positioning effect remains negligible. Homogeneous distributions of the fissile material have also been simulated and lead to a deviation less than 15% for most of the cases. The second method exploits the effect of matrices on the total active signal. A simulated annealing algorithm, using a reference data base of multi-channel scaling (MCS) spectra, is performed to fit the raw signal. The construction of the MCS library involves a learning phase to define and acquire the DDA signals as representative as possible of the real measurement conditions. Most of the cases are within a 4% agreement interval with the expected experimental value.     

Transformation of a system consisting of plane isotropic source and unit sphere detector into a system consisting of point isotropic source and plane detector in Monte Carlo radiation transport calculation

In a nuclear power plant accident, radioactive nuclides may be released which are distributed uniformly on the ground. If estimation of dose rate from such a source by a Monte Carlo calculation is attempted, some difficulty is encountered because the calculation efficiency is very low. To solve this low efficiency problem, we show that a plane isotropic source can be transformed into a point isotropic source by changing the detector shape from a unit sphere to a plane. We verified the validity of this transformation by the numerical comparison of unscattered photon fluence. As an example of this transformation, the ambient dose rate D _i was calculated from the uniform radioactive nuclide distribution on the ground using the EGS5 Monte Carlo code. We also measured the radioactivity and ambient dose rate (M) on the KEK campus within a month after the releases from the Fukushima No. 1 Nuclear Power Plant accident. Using radioactivity data and D _i, we calculated the ambient dose rate (C). The calculated and measured ambient dose rates agreed reasonably well; their ratio (C/M) was 0.62 to 1.28.     

An improved segmented gamma scanning for radioactive waste drums

In this paper, the equivalent radius of radioactive sources in each segment is determined by analyzing the different responses of the two identical detectors, and an improved segmented gamma scanning is used to assay waste drums containing mainly organic materials, and proved by an established simulation model. The simulated radioactivity distributions in homogenous waste drum and an experimental heterogeneous waste drum were compared with those of traditional segmented gamma scanning. The results show that our method is good in performance and can be used for analyzing the waste drums.     

Leaching properties of bitumen-salt mixtures produced by sedimentation, in water

After the selection of highly radiation resistant types of bitumen for application in radioactive waste management, NaCl grains were introduced into heated bitumen by a new sedimentation technique. Prepared samples were leached with distilled water and the influence of the amount of NaCl, sample geometry, temparature and preirradiation were studied. These experiments showed that a large amount of salt can be embedded in bitumen by the proposed sedimentation method. Leaching rates depended on the salt content and decreased with time and probably with the grain size and were influenced by previous radiation treatment of the bitumen. A dynamic leaching procedure yielded higher leaching rates than static methods; leaching rates were, however, higher than those calculated by diffusion models. Discrepancies can be explained by the existence of microcracks and microholes in the bitumen matrix. Although the determined leaching rates agreed well with published values obtained by other bituminized samples, better retention of embedded salts may be expected by using softer kinds of bitumen which reduce the number of defects.     

西安脉冲堆放射性实验物品暂存库剂量场计算

利用γ射线屏蔽计算公式和蒙特卡罗程序分别对暂存库屏蔽厚度和库内剂量情况进行了计算,结合Matlab软件对计算结果进行了分析和处理,得到了放射性实验物暂存库内剂量场分布和剂量等势线,为废物暂存库的使用提供了参考。     

放射性废物水泥固定过程中抗漂浮装置的初步设计

针对某核电站放射性废物水泥固定过程中出现的废物桶压饼漂浮问题,设计了一种抗漂浮装置。建立数学模型分析压饼长度跟再包装桶装载率之间的关系,结合某核电厂固废处理工艺流程数据进行计算,为操作人员选择适合当前生产工艺的不同挡片长度的320 L再包装桶提供依据。在此基础上,计算出了320 L再包装桶安全装载区间,制作了《压饼优选安全装载区间表》,为放射性压饼装载操作优化提供理论依据。     

城市库废放射源治理工作中的剂量控制

城市放射性废物库废放射源治理过程中影响工作人员受照剂量的因素较多。为实现辐射防护的最优化,本文以全国城市库治理项目统计结果及现场实施经验为依据,分析了城市库废源的特点、辐射风险及剂量控制方法。并对城市库废源治理项目的剂量数据进行统计分析,发现剂量分布规律,为辐射防护工作提供指导和参考。     

高能评价核数据处理方法

为满足加速器驱动次临界系统ADS等高能核装置的蒙特卡罗输运计算需求，通过对相关核数据处理模块的改进，建立了一套基于我国自主的群常数制作软件Ruler与国际公认的核数据处理系统NJOY耦合的中高能评价核数据处理方法及程序系统。通过该方法基于日本高能评价核数据库JENDL-HE-2007制作了ACE格式连续点截面库，并通过一系列绘图及简单问题的蒙特卡罗输运计算，验证了该库是完整的、合理的，可用于蒙特卡罗输运计算，证明了Ruler与NJOY功能模块耦合的方法可用于高能评价核数据的处理。