This study investigated the effectiveness of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, the effects of using MgCl2, NH4Cl, and Cl2 as a single chlorinating agent or applying MgCl2, NH4Cl, and Cl2 sequentially for spent fuel chlorination were evaluated Furthermore, in this study, assuming the actual process operation situation, where only a part of the semi-volatile nuclides is removed during the heat treatment process, and including the process of precipitating the molten salt from the chlorination process with K3PO4 and K2CO3 precipitants, the percentage distribution of 50 nuclides in the light water reactor spent fuel into each process stream was quantitatively calculated using the simulation function of the HSC program and tabulated for intuitive viewing. Compared to a single chlorinator, sequential chlorination more effectively separated the heat and radioactivity of the spent fuel from the uranium-dominated product solids. Specifically, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore that sequential chlorination can be an effective method for spent fuel partitioning, either as a standalone approach or in combination with other partitioning processes such as pyroprocessing.

This study examined the efficacy of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, we assessed the outcomes of employing MgCl2, NH4Cl, and Cl2 as chlorinating agents. A comparison was drawn between using a single agent and a sequential approach involving all three agents (MgCl2, NH4Cl, and Cl2). Following heat treatment, the utilization of MgCl2 as the sole chlorinating agent resulted in a moderate separation. Specifically, this method yielded a solid separation with 96.9% mass retention, 31.7% radioactivity, and 44.2% decay heat, relative to the initial spent fuel. In contrast, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore the potential effectiveness of a sequential chlorination strategy for partitioning spent fuel. This approach holds promise as a standalone technique or as a complementary process alongside other partitioning processes such as pyroprocessing. Overall, our findings contribute to the advancement of spent fuel management strategies.

The stabilization techniques are highly required for damaged nuclear fuel to strengthen safety in terms of transportation, storage, and disposal. This technique includes recovering fuel materials from spent fuel, fabrication of stabilized pellets, and fabrication of fuel rods. Thus, it is important to identify the leaching behavior of the stabilized pellets to verify their stability in humid environments which are similar to storage conditions. In this study, we introduce various leaching experiment methods to evaluate the leaching behavior of the stabilized pellets, and determine the most suitable leaching test methods for the pellets. Also, we establish the leaching test conditions with various factors that can affect the dissolution and leaching behavior of the stabilized pellets. Accordingly, we prepare the simulated high- (55 GWd/tU) and low- (35 GWd/tU) burnup nuclear fuel (SIMFUEL) and pure UO2 pellets sintered at 1,550°C and 1,700°C, respectively. Each pellet is placed in a vessel and filled with DI water and perform the leaching test at three different temperature to verify the leaching mechanism at different temperature range. Based on the standard leaching test method (ASTM C1308-21), the test solution is removed from the pellet after specific time intervals and replaced in the fresh water, and the vessel is placed back into the controlled-temperature ovens. The test solutions are analyzed by using ICP-MS.

The Ag0-containing sorbents synthesized by Na, Al, and Si alkoxides have higher maximum iodine capture capacity and textural properties than zeolite-based Ag0-containing sorbents. However, these sorbents were prepared in the form of granules via a step for cutting cylindrical alcogels. Since asmade sorbents decreased packing density, they must be additionally crushed and then classified into an appropriate size for increasing packing density. The bead formation in the step of sol-gelation could bring about the simplification of sorbent preparation process and an improvement of packing density. In the Na, Al, and Si alkoxides as starting materials, sol solution was hydrophilic and lower density than vegetable oil, which transformed sol droplets to sol-gel beads. Thus, in these precursors, sol droplets, which must be sprayed by single nozzle placed at bottom side of oil column, can rise up through oil column. Acetic acid (HOAc) was used as the catalyst for the hydrolysis of Na alkoxide (TEOS) and gelation of the Na+AlSi-OH alcosol. For obtaining sol-gel beads, experiments were performed by the flowrate change of sol solution and HOAc at different nozzle sizes using soybean oil column of 1 m in length. At a sol/HOAc flowrate ratio of 3.85, some Na+AlSi-OH alcogel beads were obtained. After the Ag/Na ion-exchange, Ag content in Ag+AlSi-OH hydrogel was low due to reaction between Na+ and HOAc during sol-gelation and aging step. The Ag+AlSi-OH hydrogel with high Ag content could be prepared by Na addition. After the solvent exchange and drying at ambient pressure, the bead sorbents had higher Ag0 content and larger pore size than granular sorbents. However, further experiments are needed to increase yield rate in bead sorbent.

Irradiated uranium dioxide in damaged used fuel could oxidize during transportation, interim storage or disposal, resulting that the fuel pellet fragments are reduced to a grain-sized powder that can easily escaped from the damaged rod. It has been reported that oxidized spent fuel (i.e. U4O9+x) that was in contact with water could increase the dissolution rate by making the grain boundaries more accessible to the water. Therefore, the damaged used fuel requires stabilization technology including nuclear material recovery, pellet manufacturing process, and stabilization fuel rod manufacturing that can secure safety in terms of permanent disposal. In this study, we prepared pure UO2 and SIMFUEL pellets that are a mixture of UO2 and surrogated metallic oxides for fission products equivalent to a burn-up of 35 GWd/tU and 55 GWd/tU as the stabilized spent fuel. The UO2 and fission products powders were milled and pressed into pellets at 250 MPa and sintered at 1,550°C and 1,700°C for 6 hours in an atmosphere of 4%H2-Ar. The prepared UO2 and SIMFUEL pellets were placed in PTFE Teflon vessels and filled with deionized water to identify the leaching behavior by a long-term leaching experiment under the similar condition to a repository for the safe disposal.

Around 40 years ago, in the mid-1980s, Swedish government approved the KBS-3 method for the direct disposal of spent nuclear fuels (SNF) in Sweden. Since then, this method has become a reference for many countries including Korea, Republic of. The main ideas of the KBS-3 method are to locate SNF at 500 m below the ground surface using a copper disposal canister and a bentonite buffer. In 2016, our government announced the National Plan (NP 2016) regarding the final management of high-level waste (HLW) in Korea. In 2019, new committee were organized to review the NP 2016, and they submitted the final recommendations to the government in 2021. Finally, the government announced the 2nd National Plan in December, 2021. So far, KAERI has developed the technologies related to the final management of SNF in two directions. One follows ‘direct disposal’ based on the KBS-3 concept, and the other ‘recycling’ based on ‘pyroprocessing-and-SFR’ (PYRO-SFR). Even though Posiva and SKB obtained the construction permits with the KBS-3 method in Finland and Sweden, respectively, there are still several technical obstacles to applying directly to our situations. Some examples are as follows: high burnup, huge amounts of SNF, and high geothermal gradient in Korean peninsula. In this work, we try to illustrate some limits of the KBS-3 method. Within our country, currently, the most probable disposal option is the KBS-3 type geological disposal, but no one knows what the best option will be in 20 or 30 years if those kinds of drawbacks are considered. That is, we compare the effects of the drawbacks using our geological data and characteristics of spent fuels. Last year, we reviewed alternative disposal concepts focusing on the direct disposal of SNF and compared the pros and cons of them in order to enhance the disposal efficiency. We selected four candidate concepts. They were multi-level disposal, deep borehole disposal, sub-seabed disposal and mined deep borehole matrix. As mentioned before, KAERI has developed a pyroprocessing technology based on the SFR to reuse fissile radionuclides in SNF. Even though we can consume some fissile nuclides such as 239Pu and 241Pu using PYRO-SFR cycle, there still remain many long-lived radionuclides such as 129I and 135Cs waiting for the final disposal. The authors review and propose several concepts for the future final management of the long-lived radionuclides.

The Na, Al, and Si akoxides-based sorbents for iodine capture have higher maximum iodine capture capacity and pore properties than zeolite-based sorbents. However, these sorbents were prepared in the form of granules via a step for cutting cylindrical alcogels. Since as-made sorbents decreased packing density, they must be additionally crushed and then classified into an appropriate size for increasing packing density. The bead formation in the step of sol-gelation could bring about the simplification of sorbent fabrication process and an improvement of packing density. For the formation of gel bead, characteristics such as hydrophilic or hydrophobic property and density of sol solution were investigated to design sol-gelation equipment. The sol-gel bead preparation equipment in the reflection of sol solution characteristics was fabricated through selection of oil for formation of sol bead, solvent for collection of gel bead, and nozzle for spray of sol droplet formation. The continuous or discontinuous formation of sol beads from NaAlSi-OH sol solution were observed according to flow rates of 6 to 8 mL·min−1 and nozzle diameters of 0.4 to 0.8 mm. In the sphericity of sol bead, the best sol beads were obtained from 0.5 mm nozzle without clogging by sol solution in the flow rate range of 6–8·min−1.

사용후핵연료의 효율적 관리를 위하여 한국원자력연구원에서 수행 중인 파이로 공정으로부터 발생되는 폐기물 처리기술에 대한 최근 연구동향을 종합적으로 고찰하였다. 파이로 폐기물 처리기술은 처분 대상 폐기물의 감용 및 포장, 저장과 최종 처분에 적합한 고화체 제조를 목표로 하고 있다. 한국원자력연구원에서 수행 중인 파이로 폐기물 처리 기술개발 접근 방향은 공정 흐름으로부터 발생한 폐기물내 주요 핵종들을 분리하고 회수한 물질 등을 재사용함으로서 폐기물 발생량을 최소로 하며 동시에 분리한 핵종을 별도로 고화처리하는 것이다. 폐기물 처리 주요 기술 특성은 먼저 전해환원용 원료물질 제조를 위하여 전처리 고온 열처리 공정을 사용하며, LiCl 과 LiCl-KCl 염으로부터 핵종을 분리하고 회수염의 재사용 및 핵종 함유량을 증대시킨 최종 고화체 제조 기술을 개발하는 것이다. 따라서 실험실 규모 실험 결과를 토대로 최근에는 공정 용량 증대를 위한 자료 확보를 목적으로 공학규모 시험을 수행 중에 있다.

공기 분위기하 UO2의 독특한 산화거동을 모사하기 위해 기존 Crackling Core Model (CCM)을 개선하였다. UO2가 U3O8으로 전환될 때 시간-전환율 곡선에서 나타나는 실험적 sigmoid 거동을 근사하게 재현할 수 있도록 모델 개선에 파편화 효과로 인한 반응 표면적 증대 및 결정립 가변 전환시간 개념을 고려하였다. UO2는 U3O7을 거쳐 U3O8으로 전환되며 최종 결정립 산화 소요 시간은 초기 결정립 산화 소요 시간의 10배에 해당한다는 가정을 도입했을 때, 개선된 모델은 599 - 674 K에서의 UO2 구 형 입자의 실험적 산화거동과 근사한 계산결과를 나타내었으며 핵종성장모델(Nucleation and Growth Model) 및 자촉매반응모델(AutoCatalytic Reaction Model)과 비교할 때 가장 작은 오차를 보여주었다. 개선된 모델을 통해 U3O8으로의 100% 전환시 계산된 활성화에너지값은 57.6 kJ·mol-1로 자촉매반응모델로 계산된 값인 48.6 kJ·mol-1보다 크며, 외삽에 의해 결정된 실험값에 더 근사함이 밝혀졌다.