• Journal of the Korean Recycled Construction Resources Institute
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• v.2 no.1
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• pp.103-112
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• 2006

Along with recent improvement of recycling technique, the quality of the recycled concrete aggregate have become very competitive to the natural concrete aggregate. Therefore, a practical use of the recycled concrete aggregate may be possible for structural members. Majority studies about the recycled concrete aggregate was emphasized a limitation of fundamental study concerned with a strength characteristics and durability of the recycled aggregate concrete, there is use for the structural members. Therefore, for the extension of application of recycled concrete aggregate, this investigation verifies the strength characteristics recycled concrete aggregate of the spun-concrete products with various coarse and fine recycled aggregate replacement ratio(coarse recycled aggregate: 0%, 20%, 40%, 60%, 100%; fine recycled aggregate: 0%, 30%, 60%, 100%) and with addition of cellulose fibers(0%, 0.01%, 0.03%, 0.05%, 0.08%). From the test results, The strength of spun concrete used with recycled aggregate [NR specimen], was measured as 72MPa, was found to be very approximately to the strength of spun concrete used with the natural aggregate(NN specimen), was measured as 74MPa, when only fine aggregate was replaced with the recycled. Therefore, the fine recycled concrete aggregate can be successfully used in the spun high strength concrete product. The compressive strength of all specimens used the specialty cellulose fiber were measured as about 70M Pa, however, the increasement of the specialty cellulose fiber content is showed to decrease compressive strength of spun concrete. Therefore, it is anticipated that the specialty cellulose fiber can be applied to the various spun concrete products.

• Korean Journal of Mineralogy and Petrology
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• v.35 no.1
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• pp.25-39
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• 2022

This study was conducted to evaluate the manufacturing process of non-sintered cement for the safe containment of radioactive waste using low level or ultra-low level radioactive waste soil generated from nuclear-decommissioning facilities, clay minerals, and blast furnace slag (BFS) as an industrial by-product recycling and to characterize the products using mineralogical and morphological analyses. A stepwise approach was used: (1) measuring properties of source materials (reactants), such as waste soil, clay minerals, and BFS, (2) manufacturing the non-sintered cement for the containment of radioactive waste using source materials and deducing the optimal mixing ratio of solidifying and adjusting agents, and (3) conducting mineralogical and morphological analyses of products from the hydration reactions of manufactured non-sintered cement solidifier (NSCS) containing waste concrete generated from nuclear-decommissioning facilities. The analytical results of NSCS using waste soil and clay minerals confirmed none of the hydration products, but calcium silicate (CSH) and ettringite were examined as hydration products in the case of using BFS. The compressive strength of NSCS manufactured with the optimum mixing ratio and using waste soil and clay minerals was 3 MPa after the 28-day curing period, and it was not satisfied with the acceptance criteria (3.44 MPa) for being brought in disposal sites. However, the compressive strength of NSCS using BFS was estimated to be satisfied with the acceptance criteria, despite manufacturing conditions, and it was maximized to 27 MPa at the optimal mixing ratio. The results indicate that the most relevant NSCS for the safe containment of radioactive waste can be manufactured using BFS as solidifying agent and using waste soil and clay minerals as adsorbents for radioactive nuclides.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1526-1531
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• 2024

¹³¹I is a fission product produced in a nuclear reactor by irradiating tellurium dioxide, with a half-life of 8.02 day. The most important and widely used method for making ¹³¹I is irradiation using a nuclear reactor and post-irradiation followed by dry distillation. The advantage of the dry distillation process is that the process and the equipment are relatively simple, namely TeO₂ (m.p. 750 ℃), which can withstand heating during reactor irradiation. Based on TeO₂ irradiation by neutron following the technique of dry distillation was explained for production of ¹³¹I on a large scale. A dry distillation followed the radioisotope production operation using the 30 MW GA Siwabessy nuclear reactor to meet national demand. TeO₂ targets are 25 and 50 g irradiated for 87-100 h. The resulting ¹³¹I activity is 20.29339-368.50335GBq. According to the requirements imposed on the radionuclide purity of the preparation, the contribution of ¹³¹I training in the resulting preparation was not less than 99.9 %