The nuclide management process for reducing the environmental burden being developed by the Korea Atomic Energy Research Institute is performed in molten salts, resulting in contaminated salt wastes containing fission products such as Cs, Sr, Ba, and rare-earth nuclides. In addition, the spent fuel of a molten salt reactor (MSR) contains a variety of fission products, and a purification process may be required for the reuse of the salt and the separation and disposal of the fission products in the spent nuclear fuel. The melt-crystallization method is a technique used for the purification and separation of chemicals or metals based on the different melting points of the different substances. In a recent study, our group developed a reactive-crystallization method using Li2CO3 precipitation agent to precipitate metal corrosion from the reactor through a chlorination reaction by HCl and Cl2, which may occur in chloride molten salt, and successfully precipitated the metal precipitate and purified and recovered LiCl salt. In this study, reactive-crystallization method has been established for removing fission products and corrosive materials. Using the reactive crystallization method, white LiCl-KCl salt that was not discolored by metal corrosion was recovered through the crystallization plates, and fission products and metal elements were shown to be suppressed to several ppm in the purified salt. Consequently, high-purity salts were recovered with high nuclide and corrosive separation efficiencies. The reactive crystallization procedure can also be applied to other salt waste systems, such as MSR nuclear fuel treatment and molten salt chemistry for the elimination of corrosive substances.

Present study investigated the waste form integrity of melted products generated from PAM-MSO system, which is proposed and developed to compensate the drawbacks of each system. The disposal suitability of the melting solidification products generated from the plasma arc melting treatment of pulverized cement debris spiked by Pb, Cd and Cs, as indicators of typical hazardous metals and radionuclides existed in the low-level mixed waste in the KHNPPs. The final waste form obtained by the test was evaluated for suitability for disposal. The compressive strength was 261.10 MPa, showing much higher values when compared to other waste form products. The compressive strength of both the sample after irradiation with 107 Gy radiation and that after long-term submersion test (90 days) satisfied the disposal criteria. As a result of the leaching test conducted according to the ANS 16.1 test method, it was confirmed that the leaching index satisfies the disposal criteria.

High-performance carbon materials were prepared via a one-step molten salt carbonization of tobacco waste used as electrode materials for supercapacitors. Carbon material prepared by carbonization for 3 h in molten CaCl2 at 850 °C exhibits hierarchically porous structure and ideal capacitive behavior. In a three-electrode configuration with 1 mol L− 1 H2SO4 aqueous solution, it delivers specific capacitance of 196.5 F g− 1 at 0.2 A g− 1, energy density of 27.2 Wh kg− 1 at 0.2 A g− 1, power density of 983.5 W kg− 1 at 2 A g− 1, and excellent cyclic stability with 94% capacitance retention after 5000 charge–discharge cycles at 1 A g− 1. Moreover, in a symmetrical two-electrode configuration with 6 mol L− 1 KOH aqueous solution, it delivers specific capacitance of 111.1 F g− 1 at 0.2 A g− 1, energy density of 3.8 Wh kg− 1 at 0.2 A g− 1, and power density of 482.0 W kg− 1 at 2 A g− 1. The relationship between hierarchically porous structure and capacitive performance is also discussed.

A simplified flowsheet for pyroprocessing commercial spent fuel is proposed in which the only salt treatment step is actinide drawdown from electrorefiner salt. Actinide drawdown can be performed using a simple galvanic reduction process utilizing the reducing potential of gadolinium metal. Recent results of equilibrium reduction potentials for Gd, Ce, Nd, and La are summarized. A description of a recent experiment to demonstrate galvanic reduction with gadolinium is reviewed. Based on these experimental results and material balances of the flowsheet, this new variant of the pyroprocessing scheme is expected to meet the objectives of minimizing cost, maximizing processing rate, minimizing proliferation risk, and optimizing the utilization of geologic repository space.

Waste treatment technology for the separation and solidification of radioactive nuclides generated from the pyrochemical process has been intensively studied to achieve the reduction of radioactive waste volume. The present study reports the separation efficiency of group II fission products in LiCl waste salt generated from a electrolytic reduction process through a layer- melt crystallization method using Sr and Ba nuclides as a surrogate material of group II fission products. The concentrated group II nuclides are converted into stable oxide form in consideration of solidification by a conversion/distillation process, where selective oxidation of group II nuclides proceeds by Li2O oxidant and residual salts are removed by a vacuum distillation process. Finally, to immobilize separated group II nuclides, a preliminary solidification study was conducted using SiO2-B2O3-Al2O3 matrix, and high density glass-based waste form was fabricated under 50wt% waste loading of strontium oxide surrogate material. Through the verification of the crystallization, conversion/distillation, and solidification processes, the treatment flow for the separation and solidification of group II fission products in LiCl waste salt has been established.

사용후 핵연료내 우라늄 및 초우란원소를 회수하는 파이로프로세싱 공정에서 배출되는 금속염화물계 방사성 폐기물은 높은 휘발특성과 붕규산계 유리와의 낮은 상용성으로 인해 고화처리가 쉽지 않은 폐기 물이다. 이를 위해, 본 연구에서는 고화처리의 한 방법으로 탈염화 반응을 통한 고화체제조 개념을 채택 하였다. 솔젤법을 이용하여 탈염화물질, SiO2-Al2O3-P2O5 (SAP)을 합성하였으며 이를 이용하여 탈염화 반 응거동 반응생성물의 고형화 특성을 조사하였다. LiCl계 폐기물과 달리, LiCl-KCl폐기물의 반응은 두 개 의 온도범위에서 반응이 진행되며, 400℃의 경우에는 LiCl이, 약 700℃에서는 KCl이 주로 반응하는 것으 로 확인되었다. 여러 가지 반응실험을 통하여 LiCl-KCl의 탈염화 반응에 가장 적합한 물질은 SAP 1071 (Si/Al/P=1/0.75/1 in molar)인 것으로 확인되었다. 4가지 종류의 고형화 실험을 통하여 고화체의 bulk shape과 densification은 SAP/Salt의 비에 영향 받는 것을 확인하였다. 제조된 고형화 시료는 Product Consistency Test-A법을 이용하여 기본적인 내구성을 평가하였다. 본 연구는 SiO2, Al2O3, P2O5 로 이루 어진 탈염화 물질을 이용하여 반응특성과 고형화 특성에 대한 기본적인 정보를 제공하였으며, 이와 같은 실험을 통하여, 본 연구에서 제안된 탈염화 고화처리방법이 휘발특성이 높고 기존 유리매질과 상용성이 낮은 금속염화물계 폐기물에 적용이 가능함을 확인하였다.

금속염화물계 방사성 폐기물은 전해공정으로 이루어진 파이로프로세싱공정의 주요한 방사성 폐기물이 다. 이와 같은 폐기물은 탄산염이나 질산염과 달리 고온에서 분해되지 않고 바로 휘발되며, 기존의 규산 계 유리와 상용성이 낮아 처리가 쉽지 않다. 본 연구팀은 금속염화물계 폐기물을 고화처리하는 방법으로 탈염화처리법을 채택하였다. 본 연구에서는 그 후속적인 연구로서, 탈염화물질로 제안된 SAP (SiO2- Al2O3-P2O5)의 조성을 변화시켜 LiCl-KCl과의 반응성을 향상시키고 고화공정을 단순화시키고자 하였다. 기본물질계에 Fe2O3를 첨가할 경우 무게반응비 SAP/Salt를 3에서 2.25로 낮출수 있으며, Fe가 Al을 치환 하는 몰분율이 0.1이상이 될 경우에는 오히려 반응성이 점진적으로 감소하는 것으로 확인되었다. 또한 M-SAP에 B2O3를 첨가할 경우에는 유리매질을 사용하지 않고 monolithic form을 제조할 수 있었다. 침출 시험결과 U-SAP 1071이 가장 높은 내구성을 보여주었으며, 1 g의 금속폐기물을 처리시 약 3∼4 g의 고 화체가 발생되며, 이는 기존의 고화처리법보다 약 ⅓∼¼배정도 최종처분부피가 감소되는 효과를 얻을 수 있다. 이상의 실험결과로부터, 기존의 유리고화공정으로 처리가 어려운 휘발성 금속염화물계 폐기물 을 단 하나의 물질을 이용하여 처리할 수 있음을 확인하였으며, 이러한 처리방법은 고화처리시 발생되는 부피를 최소화활 수 있는 대안적인 고화처리방법이 될 것으로 판단된다.