When the recycling technology of spent nuclear fuels (SNF) for future nuclear reactor systems and the treatment technology of SNF for disposing of in a disposal site use a molten salt such as LiCl-KCl eutectic as a processing medium one of the essential unit processes is a distillation process that remove the salt component mixed with fission products recovered. Especially, in case of Pyro-SFR recycling system the recovered nuclear fuel materials such as U, TRU and some of rare earths come from main three processes (electro-refining, electro-winning, and drawdown processes) for recycling of SNF. These recovered fuel materials contain large portion of molten salt or liquid cadmium which requires removal of them by distillation. In spent nuclear fuels discharged from PWR the portion of composing element is as follows. Uranium is about 95%, other actinides such as transuranic elements (TRU; Np, Pu, Am, Cm) is about 1%, the rare earths (lanthanides) is about 1%, and the other elements is about 3%. For example, americium (Am) in the recovered fuel materials has a problem that the reported loss of Am inevitably occurs during the vacuum salt distillation operation. A new segregation method of AMM (actinide metal mixture)–salt system is based on the difference in melting point of the actinide elements. It is possible to apply this segregation method to recovering other actinides from AMM with accompanied salt because of relatively large amount and lower melting point of a specific element in other actinides avoiding vacuum salt distillation. This new segregation method successfully tested using a surrogate element such as aluminum due to its similar melting point with a specific element. The segregation principle is solid-liquid separation, thus the solidified actinides mixture ingot can take out of a molten salt medium.

As temporary storage facilities for spent nuclear fuel (SNF) are becoming saturated, there is a growing interest in finding solutions for treating SNF, which is recognized as an urgent task. Although direct disposal is a common method for handling SNF, it results in the entire fuel assembly being classified as high-level waste, which increases the burden of disposal. Therefore, it is necessary to develop SNF treatment technologies that can minimize the disposal burden while improving long-term storage safety, and this requires continuous efforts from a national policy perspective. In this context, this study focused on reducing the volume of high-level waste from light water reactor fuel by separating uranium, which represents the majority of SNF. We confirmed the chlorination characteristics of uranium (U), rare earth (RE), and strontium (Sr) oxides with ammonium chloride (NH4Cl) in previous study. Therefore, we prepared U-RE-SrOx simulated fuel by pelletizing each elements which was sintered at high temperature. The sintered fuel was again powdered by heating under air environment. The powdered fuel was reacted with NH4Cl to selectively chlorinate the RE and Sr elements for the separation. We will share and discuss the detailed results of our study.

This study proposes a method of separating uranium (U) and minor actinides from rare earth (RE) elements in the LiCl-KCl salt system. Several RE metals were used to reduce UCl3 and MgCl2 from the eutectic LiCl-KCl salt systems. Five experiments were performed on drawdown U and plutonium (Pu) surrogate elements from RECl3-enriched LiCl-KCl salt systems at 773 K. Via the introduction of RE metals into the salt system, it was observed that the UCl3 concentration can be lowered below 100 ppm. In addition, UCl3 was reduced into a powdery form that easily settled at the bottom and was successfully collected by a salt distillation operation. When the RE metals come into contact with a metallic structure, a galvanic interaction occurs dominantly, seemingly accelerating the U recovery reaction. These results elucidate the development of an effective and simple process that selectively removes actinides from electrorefining salt, thus contributing to the minimization of the influx of actinides into the nuclear fuel waste stream.

본 연구는 nPr-BTP/nitrobenzene 추출 계에 의한 악티나이드(III)의 선택적 분리로, 우선 자연친화적 CHN 형 의 nPr-BTP (2.6-Bis-(5.6-n-propyl-1.2.4-triazin-3-yl)-pyridine)를 합성하고, 이의 희석제에 대한 용해성 및 질산에 대한 안정성 등을 평가하였다. 악티나이드(III)의 대표원소로는 Am을 선정하였으며, 0.1M nPr-BTP/nitrobenzene-1M , O/A=2의 조건에서 Am은 약 85%, RE 원소는 Eu가 8%, 기타 Nd, Ce, Y 등은 3% 이하가 추출되어 (이때 Am/Eu의 상호분리 계수 약 60정도) 악티나이드(III)의 선택적 추출에는 별 문제가 없을 것으로 판단되었다. 그러나 Am의 역추출의 경우 0.05M 질산용액으로 O/A=1 에서 약43%가 역추출 되었으며, O/A=0.3에서도 65% 정도만이 역추출 되어 질산 이외의 다른 역추출제의 개발이 요구되고 있다.

본 연구는 실제 HLW 수준의 다성분 계 모의 용액으로부터 금속함유 추출 계에 의한 Am-Cm/RE 원소의 공분리 및 이의 상호분리 연구를 수행하였다. 우선 금속함유 추출제인 Zr-DEHPA를 자체 제조하고, 제 3상 방지 조건 결정과 질산 농도, DEHPA 농도, Zr 함유량 등이 공추출에 미치는 영향을 평가하여 최적 조건으로 (15g/L Zr-1M DEHPA)/NDD-1M 추출 계를 설정하였다. 이때 추출률은 Am (81%), Cm (85%), RE 원소 (80% 이상), Mo (98%), Fe (85%), U (98%), Np (73%), 기타 원소 (5% 이하) 등으로 Am-Cm/RE의 공분리 적용성은 양호하나, U, Np, Mo, Fe의 선제거가 필요하고 특히 제 3상 형성 유발 물질인 Zr이 거의 함유되지 않아야 한다. 그리고 공추출된 Am-Cm/RE를 Am-Cm (역추출제 : 0.05M DTPA-1M Lactic acid-pH 3.6) (역추출제 : 5M ) 순으로 상호 분리하여 각각의 분리계수를 평가하였으며 이때 Am은 65.4%, Cm은 63.9% RE 원소(Y 제외)는 85% 이상이 역추출 되었다.

New pyrochlore-type phases(A2B2O7) were synthesized in the systems: CaO-CeO2-TiO2, CaO-UO2(ThO2)-ZrO2, CaO-UO2(ThO2)-Gd2O3-TiO2-ZrO2, 및 CaO-ThO2-SnO2. The starting materials were pressed with the pressure of 200~400 MPa and sintered at 1500~ 1550℃ for 4~8 hours in air and at 1300~ 1350℃ for 5 ~50 hours under oxygen atmosphere. The products were characterized using XRD, SEM/EDS and TEM. In the bulk compositions of CaCeTi2O7, CaThZr2O7,(Ca0.5 GdTh0.5)(ZrTi)O7) (Ca0.5GdTh0.5)(ZrTi)O7, (Ca0.5GdU0.5)(ZrTi)O7 and CaThSn2O7 , pyrochlore was the major phase, together with other oxide phase of2O7 fluorite structure. In the samples with target compositions CaUZr2O2및 Ca0.5 GdU0.5)Zr2TiO7 pyrochlore was not identified, but a fluorite-structured phase was detected. The formation factor as the stable phase depended on crystal chemical characteristics of the actinide and lanthanide elements of the system concerned.