Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Study on volume reduction of radioactive perlite thermal insulation waste by heat treatment with potassium carbonate

  • Chou, Yi-Sin (Chemical Engineering Division, Institute of Nuclear Energy Research) ;
  • Singh, Bhupendra (Department of Mechanical Engineering and Advanced Institute of Manufacturing with High-tech Innovations, National Chung Cheng University) ;
  • Chen, Yong-Song (Department of Mechanical Engineering and Advanced Institute of Manufacturing with High-tech Innovations, National Chung Cheng University) ;
  • Yen, Shi-Chern (Department of Chemical Engineering, National Taiwan University)
  • Received : 2021.03.02
  • Accepted : 2021.07.16
  • Published : 2022.01.25
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Abstract

Perlite is one of the major constituents of the radioactive thermal insulation waste (RTIW) originating from nuclear power plants and, for proper waste management, a significant reduction in its volume is required prior to disposal. In this work, the volume reduction of perlite is studied by high-temperature treatment method with using K2CO3 as a flux. The perlite is ground with 0-30 wt% K2CO3, and differential thermal analysis/thermogravimetric analysis is used to monitor the glass transition temperature (Tg) and weight loss. The Tg varied between ~772.2 and 837.1 ℃ with the minima at ~643.5 ℃ with the addition of ~10 wt% K2CO3. It is observed that compared to the pure perlite the volume reduction ratio (VRR) increases with the addition of K2CO3. The VRR of 11.20 is observed with 5 wt% K2CO3 at 700 ℃, as compared to VRR of 5.56 without K2CO3 at 700 ℃. The X-ray photoelectron spectroscopy and scanning electron microscopy are used to characterize perlite samples heat-treated without/with 5 wt% K2CO3 at 700 ℃. Moreover, the atomic absorption spectroscopy indicates that the proposed heat-treatment procedure is able to completely retain the radionuclides present in the perlite RTIW.

Keywords

Acknowledgement

We gratefully thank the financial supports from Fuel Cycle and Materials Administration (FCMA), Atomic Energy Council (AEC) and Institute of Nuclear Energy Research (INER) of Taiwan for the performance of this work under project number: 109FCMA007.

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