In KAERI’s previous phosphate precipitation tests, the dispersed powder of lithium phosphate (Li3PO4) as a precipitation agent reacted with various metal chlorides in a simulated LiCl-KCl molten salt. The reaction of metal chlorides composed of actinides such as uranium and three rare earths (Nd, Ce and La) with lithium phosphate is a solid-liquid reaction. A phosphorylation reaction rate is very fast and the metal phosphates as a reaction product precipitated on the bottom of the molten salt crucible. One of the recovery methods of the metal phosphate precipitates is segregation the lower part (precipitates) of the salt ingot using the various cutting tools. Recently, a new phosphorylation experiment using lithium phosphate ingots carried out in order to collect the metal phosphate precipitates into a small recovering vessel, and the test result of this new method was feasible. However, the reaction rate of test using lithium phosphate ingot is extremely slower than that of test using lithium phosphate powder. In this study, the precipitation reactor design (a tapered crucible with polished inner surface) used for phosphorylation reaction showed that the salt ingot with metal phosphate precipitates could be detached from a tapered stainless steel crucible. We propose that the recovery of precipitates from a salt ingot is possible by introducing a dividing plate structure into a molten salt and by positioning it at the interface between salt and precipitated metal phosphate.

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.

A phosphorylation (phosphate precipitation) technology of metal chlorides is considering as a proper treatment method for recovering the fission products in a spent molten salt. In KAERI’s previous precipitation tests, the powder of lithium phosphate (Li3PO4) as a precipitation agent reacted with metal chlorides in a simulated LiCl-KCl molten salt. The reaction of metal chlorides containing actinides such as uranium and rare earths with lithium phosphate in a molten salt was known as solidliquid reaction. In order to increase the precipitation reaction rate the powder of lithium phosphate dispersed by stirring thoroughly in a molten salt. As one of the recovery methods of the metal phosphates precipitated on the bottom of the molten salt vessel cutting method at the lower part of the salt ingot is considered. On the other hand, a vacuum distillation method of all the molten salt containing the metal phosphates precipitates was proposed as another recovering method. In recent study, a new method for collecting the phosphorylation reaction products into a small recovering vessel was investigated resulting in some test data by using the lithium phosphate ingot in a molten salt containing uranium and three rare earth elements (Nd, Ce, and La). The phosphorylation experiments using lithium phosphate ingots carried out to collect the metal phosphate precipitates and the test result of this new method was feasible. However, the reaction rate of test using lithium phosphate ingot is very slower than that of test using lithium phosphate powder. In this presentation, the precipitation reactor design used for phosphorylation reaction shows that the amount of molten salt transferred to the distillation unit will reduce by collecting all of the metal phosphates that will be generated using lithium phosphate powder into a small recovering vessel.

Based on the results of a review for various precipitation methods phosphorylation (phosphate precipitation) of metal chlorides considered as a proper treatment method for recovering of the fission products in a molten salt. In previous precipitation tests, the powder of lithium phosphate (Li3PO4) added into LiCl-KCl molten salt containing metal chlorides as a precipitation agent. The reaction of metal chlorides containing actinides and rare earths to recover with lithium phosphate in a molten salt known as solid-liquid reaction. The powder of lithium phosphate disperse in a molten salt by stirring thoroughly in order to enhance the precipitation reaction. As a result, metal phosphates as the reaction products precipitate on the bottom of the vessel and cutting at the lower part of the salt ingot considered as one of the recovery method of the precipitates. Recently, the vacuum distillation of upper part of the salt proposed as another recovering method. Cutting method of precipitate at the lower part of the salt ingot would be difficult to handle the increased size of the salt ingot produced from the practical scale equipment. In this presentation, a new method for collecting the precipitates of phosphorylation reaction into a small vessel is introduced with test results in a molten salt containing uranium and rare earths such as Nd, Ce, and La. As the first step of a series of test lithium phosphate ingot was prepared by melting the powder at a temperature 1,300°C, and the ingot put into LiCl-KCl molten salt at 500°C for more than three hours to examine the shape of ingot to be deformed or not. The phosphorylation experiments using lithium phosphate ingots carried out to collect the metal phosphate precipitates and the test result of this new method was feasible.

Under the circumstance of energy transition policy of the previous government in which nuclear energy portion will be gradually reduced, some R&D study looking for alternatives other than Pyro- SFR recycling could be very valuable and timely suitable. New alternative study started to evaluate the possibility of it if there are some advantages in terms of waste burden in case that the spent fuel are appropriately treated and disposed of in a disposal site, instead of recycling of spent nuclear fuels (SNF). The alternative study separate the fission products (minor actinides and rare earths) from SNF in a molten salt medium. The molten salt coming from the alternative study is radioactive and heat generating because it contains the fission products chlorides. It is necessary to collect the fission products from the waste molten salt for minimization of the high-level waste volume and to generate a final waste form containing the fission products compatible to the disposal site. Based on the results of a review for various precipitation methods, phosphorylation (phosphate precipitation) of metal chlorides selected as a proper treatment method for recovering of the fission products in a molten salt. Phosphate precipitation has the potential for removing most of fission product elements from a molten salt arising from the treatment of spent nuclear fuel. The performance of phosphate precipitation method evaluated using a salt mixture with the actinide and rare earth chlorides. The molten salt containing uranium as surrogate of the actinides and three rare earths (Nd, Ce, La) chloride was used for testing a phosphate precipitation method at experimental condition (temperature 500°C, salt stirring 200~300 rpm, and 1~1.2 eq. of phosphorylation agent). A cyclic voltammetry (CV) method monitored in-situ phosphate precipitation progress for determining the precipitation rate and conversion ratio evaluated. The phosphorylation reaction increased greatly at a salt stirring 300 rpm.

액체카드뮴음극(LCC, Liquid Cadmium Cathode)을 사용하여 우라늄과 TRU (TRans Uranium) 원소를 동 시에 회수하는 전해제련공정에서 LCC 표면에서 성장하는 수지상(dendrite) 우라늄의 생성 및 성장을 억제하 기 위한 LCC 구조는 개발은 전해제련공정의 핵심이다. 금속 수지상의 생성과 성장 현상을 관찰하기 위해 상 온에서 실험이 가능하며 육안관찰이 가능한 Zn-Ga 계의 모의실험장치를 제작하였으며 갈륨 계면에서의 수 지상 아연의 성장 현상과 기존의 교반기형과 파운더형 LCC 구조의 성능을 관찰하였다. 이러한 금속 수지상 은 전해용액 내에서 그 기계적 강도가 약한 것으로 보여 여러 가지 음극 구조에 의해 쉽게 파쇄 되지만 액체 금속으로 쉽게 가라앉지는 않았다. 모의 실험결과를 바탕으로, LCC 구조개발에 활용할 수 있는 실험실 규모 의 액체음극 전해제련 실험 장치를 제작하였으며, 수지상 우라늄의 성장 억제를 위한 여러 가지 형태의 LCC 구조의 성능 시험을 수행하였다. 교반기형 LCC 구조의 실험결과 LCC 도가니 내벽에서 성장하는 수지상 우 라늄을 효과적으로 파쇄하지 못하였으며, 일자형과 harrow형 LCC 구조의 성능은 유사하였다. 이에 따라 LCC 표면과 도가니 내벽에서 성장하는 수지상 우라늄을 LCC 도가니 바닥으로 침전시키기 위하여 mesh형 LCC 구조를 개발하였다. 이의 성능실험결과 수지상 우라늄의 성장 없이 약 5 wt%까지의 우라늄을 회수할 수 있었다. 실험 종료 후 LCC 바닥 침전물을 화학 분석한 결과 금속간화합물(UCd11)이 형성되었음을 확인할 수 있었다.

본 연구에서는 방사성폐기물의 화학처리공정에 자주 사용되는 유동관식 장치 중 튜브형 반응기, 다단식 용매추출 장치, 흡착탑 등 물질전달이 수반되는 장치에 있어 각종 매개변수들이 반응수율이나 물질전달수율에 미치는 영향과 민감도를 살펴보았다. 먼저 각 장치에 대한 거동을 묘사하기 위하여 수학적 모델링을 수행하였고 전산모사를 통하여 해당 장치의 거동을 예측하였다. 그리고 그 결과로부터 해당 공정의 고유한 매개변수들이 반응수율 또는 물질전달수율에 미치는 영향과 민감도를 분석하였다. 튜브형 반응기에서는 확산계수, 반응속도상수 등이 반응수율에 미치는 영향을, 다단식 용매추출 장치에서는 연속상과 분산상의 분배계수, 연속상 흐름의 역혼합 등이 추출수율 및 장치 내 농도 분포에 미치는 영향을 고찰하였다. 또 흡착탑에 있어서는 흡착평형상수 및 유체-흡착재간 물질전달계수 등이 흡착 속도에 미치는 영향을 조사하였다.

본 연구에서는 사용후 TRISO 연료 처리를 위한 보다 효과적인 공정개발을 위하여 기존 전처리 기술에 대한 검토를 수행하였다. TRISO 연료 처리에 있어서 가장 중요한 사항은 연료입자에 포함되어 있는 탄소와 SiC성분을 효과적으로 분리하는데 있다. 공정개발 초기에 고려되었던 분쇄 후 배소공정의 경우 처리공정에서 발생되는 2차 폐기물로 인하여 분쇄 후 침출공정으로 대체 되었으나 여전히 해결해야 될 근본적인 문제점이 존재하고 있다. 따라서 본 논문에서는 TRISO 입자의 피복층 제거를 위한 새로운 개념의 열적 파쇄와 용융염 전해반응에 의한 피복층 제거 공정을 제안하였으며 각 공정에 대한 원리를 자세하게 기술하였다.