模拟不可燃放射性废物等离子体熔融处理试验研究

在自制的等离子体熔融试验台架上对玻璃纤维、混凝土、土壤的单体玻璃固化配方及三元混合废物玻璃固化配方开展等离子体熔融处理和同位素示踪实验。四种样品在1 100~1 300 ℃条件下熔融1 h均可得到玻璃固化体,经检测,玻璃固化体的密度、抗浸出性能以及机械性能均满足放射性废物玻璃固化体性能要求。示踪实验结果表明,等离子体熔融系统对示踪元素Co、Cs和Sr有较高截留率,且玻璃固化体对Co和Sr的固化能力较高、对Cs固化能力相对较低。在工程应用中,建议在熔融炉系统前端增设造粒等预处理系统,减少物料直接进入烟气净化系统的比例,以提高物料固化效率。     

核电站低中放固体废物热等离子体处理研究进展

结合核电站中产生的低中放固体废物现有处理技术的不足,介绍了热等离子体处理废物的原理、优点和处理放射性废物的进展,重点介绍了现有典型装置的反应器与系统构成,探讨了等离子体装置处理放射性废物值得注意的问题、技术难点与解决方向,以期为国内开展相关的研究提供参考。     

Design of an Anti-Compton Spectrometer for Low-Level Radioactive Wastes using Monte Carlo Techniques

An anti-Compton spectrometer with semi-2π Compton suppression is designed to identify the photons emitted from low-level radioactive wastes from radioisotope usage and nuclear research laboratory. Since the objective sample is massive and large, the system has a full opening toward the sample position. The characteristics and features of the system concerning Compton suppression and reduction of the background component due to natural radioactive source are estimated by the Monte Carlo simulations. The anti-Compton technique is shown to be quite advantageous for the reduction of the surrounding natural background radiation, as well as the suppression of the background for the higher energy photons.     

Technique for Radioactivity Measurement in Drum Package Waste by Using Scattered Gamma-Rays

An evaluation method for the radionuclide content in low level drum package waste by using scattered γ-rays has been investigated. Gamma-rays are counted by a detector after passing through a collimator having appropriate vertical and horizontal openings both for the segmented scan and for the average geometrical efficiency being almost independent of the source position. The attenuation of unscattered γ-rays is estimated from the intensity ratio of scattered rays to unscattered ones of the nuclide emitting the highest γ-ray energy in the drum. Satisfactory results were obtained in numerical simulations and basic experiments carried out using test equipment with a pure Ge detector and a rectangular collimator, and sample 200l drums, containing sealed ⁶⁰Co and ¹³⁷Cs sources, and having a uniform density of 2.2 g/cm³. The error of the evaluated content, which depended slightly on the activity distribution, was a maximum of about ±30%.     

Three Year Experimental Study of Leaching Phenomena from Low Level Radioactive Cement-Based Waste Forms

Abstract

This study was carried out in order to demonstrate the safety of homogeneous cementbased waste forms (hereinafter called cement forms) for BWR's low level radioactive wastes as engineered barriers. Eighteen full scale simulated cement forms were manufactured with the addition of ¹³⁷Cs, ⁶⁶Co and ⁹⁰Sr.

Leaching tests on these forms were carried out for approximately three years. In order to study the relationship of leachability to environments at disposal sites, this Three Year Leaching Test was conducted for three kinds of environmental conditions, sea water, land water and soil. After the tests, all of these forms were cut to measure the distribution of the radionuclide's density within them. In case of the soil tests, the distribution of radionuclide in the soil was also measured.

The radionuclide leachability results reveal that ⁶⁰Co was almost completely retained in the cement forms and that ¹³⁷Cs leached from cement forms was mostly adsorbed by the soil. On the other hand, ⁹⁰Sr was not trapped in the forms and leaked through the soil around them in retard. This study also showed that simulated cement forms buried in the soil were more physically and chemically stable, and had longer term stable radionuclide containment capability than those which were submerged in sea or land water.