When damaged nuclear fuel is stripped and re-fabricated into stabilized pellets, it is necessary to analyze the characteristics of the stabilized pellets, such as density, leaching behavior, and compressive strength, for final disposal. In this study, simulated nuclear fuel with UO2 and burn-up of 35 GWd/tU and 55 GWd/tU was used to measure the compressive strength of the stabilization pellet. In order to change the density of the sintered pellet, a sintered pellet was prepared by heat treatment at 1,550°C and 1,700°C for 6 hours in a reducing atmosphere of 4% H2/Ar. In the case of UO2, the density was 10.4 g/cm3 (94.5% of T.D.) and 10.6 g/cm3 (96.6% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). In the case of simulated fuel with a burn-up of 35 GWd/tU, the density was 8.8 g/cm3 (80.9% of T.D.) and 10.2 g/cm3 (93.6% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). In the case of simulated fuel with a burn-up of 55 GWd/tU, the density was 8.3 g/cm3 (77.0% of T.D.) and 10.0 g/cm3 (92.3% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). It was found that the compressive strength of simulated nuclear fuel decreased with increasing burn-up and increased with increasing density. In the case of UO2, the compressive strengths were 717.8 MPa and 897.4 MPa when the densities were 10.4 g/cm3 and 10.6 g.cm3, respectively. In the case of simulated nuclear fuel with a burn-up of 35 GWd/tU, the compressive strengths were 472.1 MPa and 732.3 MPa when the densities were 8.8 g/cm3 and 10.2 g/cm3. In the case of simulated nuclear fuel with a burn-up of 55 GWd/tU, the compressive strengths were 301.4 MPa and 515.5 MPa when the densities were 8.3 g/cm3 and 10.0 g/cm3, respectively.

Korea Atomic Energy Research Institute (KAERI) has investigated Pyroprocessing technology in order to decrease the burden of disposal system and increase availability of useful radionuclides in the spent nuclear fuel (SNF) for future. The treatment and the disposal of SNF, however, are very sensitive issues socially. In addition, under the energy transition policy phasing out nuclear energy gradually there have been demands for alternatives so far. Thus various alternatives should need to be investigated in preparation for unexpected situations. This study has been conducted roughly in effectiveness point of view of alternative pre-managements for SNF, not pyroprocessing technology, in disposal system, consisting of three stages according to the degree of burden in disposal system. Stage I is the case for making safety increase with removing highly-mobile radionuclides from SNF. Stage II is the case for eliminating high-heat radionuclides additionally, alleviating thermal risk in the disposal system. And Stage III is the case for recovering Uranium in addition to Stage II. These options of pre-management are thought to be able to provide an intuitive strategy for effective diversification of the disposal system. Because several types of waste form from pre-management make it possible to develop the effective, newly-composed waste disposal system according to the properties of radionuclides. And the processability of SNF through pre-management might be combination with available core-drilling technology, being able to design various disposal system as well. Even though the whole, detailed unit processes have not designed yet, mass balance and distributions of radionuclides are performed under the appropriate assumption of engineering processes. As a first step the alternative approaches for SNF pre-management for disposal system might be expected to be widely used in implementing SNF management policy in the future.

염화희토류 수화물(RECl3·xH2O) 내 존재하는 수분을 제거하기 위하여 탈수화 장치를 제작하여 8가지 (La, Ce, Nd, Pr, Sm. Eu, Gd, Y)Cl3·xH2O에 대한 탈수화 실험을 수행하였다. 탈수화 과정 중 희토류옥 시염화물의 형성을 억제하기 위하여 TGA 분석을 바탕으로 하여 단계적인 온도 상승(80→150→230℃)구 간을 설정하였으며 증발된 수분의 원활한 이동을 위하여 예열된 Ar 가스와 vacuum pump를 이용하였다. 각 온도구간에서의 탈수화 정도를 살펴본 결과 YCl3·xH2O를 제외한 염화희토류 수화물은 원자번호가 높을수록 높은 온도에서 더 많은 탈수화가 일어남을 알 수 있었다. 탈수화 과정 후 희토류옥시염화물의 형성은 보이지 않았으며 염화 희토류 수화물 내 수분을 10%이하로 감소시킬 수 있었다.

본 연구에서는 산화조건하 LiCl-KCl 공융염내에서 란탄계 염화물의 하나인 PrCl3의 열적거동을 살펴보았다. 먼저 산소를 주입하면서 PrCl3의 열중량분석(TGA; thermogravimetric analysis)을 실시하였고, 이 때 얻어진 결과들을 바탕으로, 산소분산법을 이용하여 온도에 따른 LiCl-KCl 공융염내 PrCl3의 산화실험을 수행하였다. PrCl3의 열중량분석 결과에 따르면, 약 380 ℃까지 PrCl3에서 염소의 해리가 급격하게 발생되었고 약 600 ℃에 서 PrCl3가 PrOCl로 전환되는 반응이 종료되는 것으로 확인되었다. 산소분산법에 의한 LiCl-KCl 공융염내 PrCl3의 열적거동은 산화조건에서 열중량분석시 나타난 PrCl3의 열적거동과 유사하였고, 발생된 PrOCl은 공 융염내에서 불용성 화합물로써 바닥으로 침전하였다. 산소분산법에 의한 공융염내 PrCl3의 PrOCl로의 전환 은 650 ℃ 이상의 온도에서 활발하게 진행되었고, 이 때 발생되는 배기가스내 Cl2의 농도분석을 통해 공융염내 PrCl3의 전환상태를 예측할 수 있을 것으로 판단된다