Despite its advantages such as safety, unnecessary pretreatment, and decontamination of waste with complex geometry, conventional ultrasonic decontamination technology has been only used to remove loose contaminants, oil and grease, not fixed contaminants due to the limitations in increasing the intensity in the high frequency range. Thus, ultrasound has been used as an auxiliary method to accelerate chemical decontamination of radioactive wastes or chemicals were added to the solution to increase the decontamination efficiency. The recently developed high-intensity focused ultrasound (HIFU) decontamination technology overcomes these limitations by combining multiple frequencies of ultrasonic waves in a specific arrangement, making it possible to remove most fixed contaminants, including radioactive micro particles less than 1 micrometer within half an hour. KEPCO NF and EnesG developed mobile HIFU decontamination equipment and successfully demonstrated the decontamination effect on various radionuclides found in nuclear power plants by treating radioactive metal waste to the level below free release criteria. The mobile HIFU decontamination equipment used in the demonstration can be operated anywhere where water is supplied, including controlled area in nuclear power plants, and is expected to be used widely for decontamination and free release of metal radioactive wastes.

Currently, the most promising fuel candidate for use in sodium fast reactors (SFRs) is metallic fuel, which is produced by a modified casting method in which the metallic fuel material is sequentially melted in an inert atmosphere to prevent volatilization, followed by melting in a graphite crucible, and then injection casting in a quartz (SiO2) mold to produce metallic fuel slugs. In previous studies, U-Zr metallic fuel slugs have been cast using Y2O3 reaction prevent coatings. However, U-Zr alloy-based metallic fuel slugs containing highly reactive rare earth (RE) elements are highly reactive with Y2O3-coated quartz (SiO2) molds and form a significant thickness of surface reaction layer on the surface of the metallic fuel slug. Cast parts that have reacted with nuclear fuel materials become radioactive waste. To decrease amount of radioactive waste, advanced reaction prevent material was developed. Each RE (Nd, Ce, Ln, Pr) element was placed on the reaction prevent material and thermal cycling experiments were carried out. In casting experiments with U-10wt% Zr, it was reported that Y2O3 layer has a high reaction prevent performance. Therefore, the reaction layer properties for RE elements with higher reactivity than uranium elements were evaluated. To investigate the reaction layer between RE and NdYO3, the reaction composition and phase properties as a function of RE content and location were investigated using SEM, EDS, and XRD. The results showed that NdYO3 ceramics had better antireaction performance than Y2O3.

Normally, non-metallic wastes, such as sands, concrete and asbestos are regarded as electrically non-conductive materials. However, when the temperatures are increased up to the melting point, their electrical conductivities can be greatly improved, flowing arc current. Accordingly, these nonmetallic wastes can be efficiently treated by heating them up to the electrically conducting temperatures by using a non-transferred type plasma torch, and then, melting them completely with arc currents in transferred mode of plasma torch. For this purpose, we propose a convertible plasma torch consisting of three cylindrical electrodes (rear electrode, front electrode and exit nozzle). Compared with conventional plasma torch with two cylindrical electrodes (rear electrode and front electrode), the proposed plasma torch can provide more stable plasma jet in high powered and non-transferred mode due to the presence of exit nozzle, resulting in rapid heating of the non-conductive materials.

The treatment of radioactive waste by melting has been mainly discussed with low-level waste (LLW). Considering that a large amount of waste in RV or RVI is intermediate-level waste (ILW), however, it is necessary to examine the possibility of treatment by melting of ILW. Different from LLW, melting of ILW with a high content of long-lived nuclides would lead to no free releasee, but has advantages in volume reduction, homogenization, and delay of release. In this paper, the possibility of melting as an alternative technology for the treatment of ILW in the future is reviewed by analyzing the benefits generated by melting ILW in the following aspects: 1) Similar to melting techniques of LLW, them of ILW are mostly based on well-known techniques, but it is necessary to review the feasibility of performing operations such as removal of solidified melt using remote equipment in abnormal situations such as loss of electricity. 2) It is necessary to specify radiation limits for the melting operation unless the ILW melting operation technique can guarantee that the risk of abnormal occurrence is very low. The main quantified radiation parameter is the ingot dose rate, which of 10 mSv/h is considered more reasonable. 3) Although the treatment of ILW by melting leads to a reduction in volume, the main characteristics of the waste still remain, and no waste can be disposed of for free release. Thus, the main potential benefits are improved long-term safety and reduced waste volume. 4) Reducing the surface-to-volume ratio of the molten material could reduce the amount of corrosive material per unit time and, consequently, increase long-term safety. Its effect on long-term safety is difficult to quantify precisely as it depends on several factors, such as the geometry of the original component or whether radionuclides were distributed on the surface of the original component or the induced radioactivity. 5) The volume reduction of ILW is estimated to be reduced by about 1/4 compared to the generated amount when assuming a disposal volume reduction factor of 3 and considering the dose reduction due to radioactive decay after long-term storage, however, due to the lack of knowledge about non-hazardous facility alternatives, it is difficult to evaluate cost-benefit. This is heavily influenced by both the final volume reduction and the potential to reduce the complexity of the repository’s technical barriers.

In Korea, it is expected that the decommissioning of nuclear reactors will increase due to the license termination of reactors constructed in the 1960s to the 80s. According to the investigation of KORAD, VLLW accounts for 67.10% of decommissioning wastes and amounts to about 413,336 drums. Due to their huge amount, it is necessary to create an appropriate decommissioning waste management plan even though VLLW is disposed at the second-phase disposal facility of the Gyeongju repository. For efficient reduction in decommissioning wastes, it is required to actively use a clearance of metallic and concrete radioactive wastes. Regulations of nuclear safety and security commission notice that the radioactive waste can be reused or recycled if it meets the clearance criterion, 10 μSv·y−1 for individual dose. Therefore, it is important to develop a computational code which calculate individual doses for each scenario, and determine whether the clearance criterion is satisfied. However, in the case of metallic waste, RESRAD-RECYCLE used in dose assessment for the clearance has no longer been maintained or updated since 2005 and there is no code for recycling of concrete waste. For this reason, a dose assessment code RUCAS (Recycle-Underlying Computational dose Assessment System) has been developed by Ulsan National Institute of Science and Technology (UNIST). A point kernel method is adopted into external dose assessment model to calculate more realistic options, which are various geometries of source, and shielding effect. In the case of internal radiation, equations of internal dose from IAEA are used. This research conducts a verification of dose assessment model for recycling of metallic radioactive waste. RESRAD-RECYCLE is the comparison object and results from RESRAD-RECYCLE validation report are referenced. Targets are 14 recycling scenarios composed up to the smelting metal step of four steps, which are arising scrap metal, smelting scrap metal, and fabrication of metal product, and reusing/recycling of product. Seven isotopes, which are Ac-227, Am-241, Co-60, Cs-137, Pu-239, Sr- 90, and Zn-65, are selected for calculation. Validation results for external dose vary by isotopes, but show acceptable differences. It seems to be caused by difference in the calculation method. In the case of internal dose using same calculation formula, results are exactly matched to RESRAD-RECYCLE for all isotopes. Consequently, RUCAS can conduct functions supported by RESRAD-RECYCLE well and future work will be conducted related to domestic recycling scenarios considering public acceptance, and verification with radiation shielding codes for various geometries of source.

방사성금속폐기물의 관리 옵션들을 안전성, 경제성, 작업자 피폭, 부피 감용 등의 선별 기준을 적용하여 비교 평가하였다. 원전 해체로부터 발생하는 금속폐기물의 관리 옵션에는 무구속 방출, 제한적 재사용, 그리고 직접 처분이 있다. 고려된 각 각의 옵션들은 금속폐기물의 절단과 용융에 의한 부피감용을 수반한다. AHP기법을 적용하여 각 옵션들의 순위를 부여하였 다. 방사성금속폐기물을 용융하여 금속 잉곳을 제조한 후 제한적 재이용 또는 무구속 방출하는 방안이 가장 효율적인 옵션 으로 도출되었다.

KAERI에서는 파이로프로세싱에서 발생하는 금속폐기물의 부피 및 무게 감량을 위해 고방사성 장반감기 핵종을 포함하는 anode sludge내 NM의 고화매질로써 폐피복관과 첨가금속을 재활용하는 연구를 진행하고 있다. 본 연구에서는 Cr 함량을 조절한 Zr-17Cr-8NM, Zr-22Cr-8NM, Zr-27Cr-8NM 합금을 유도용융을 통해 제조하였고, 전기화학적 부식시험을 실시하여 부 식특성을 평가하였다. 모든 조성에서 기존 연구 중인 Zr계 합금고화체 조성보다 우수한 부식특성을 나타냈다. 또한 Zr-22Cr- 8NM 시편의 부식시험 후 침출용액 조성 분석 결과, 500 mV 전압 조건 이하에서는 NM 침출이 없었고 이를 통해 우수한 화 학적 안정성을 갖는 합금고화체 조성을 확보하였다.

한국의 가장 오래된 상업 원전인 고리 1호기가 2017년에 해체가 이루어질 예정이다. 원전 해체 폐기물의 적절한 처리는 효율적인 원전해체에 있어 중요한 역할을 할 것이다. 특히, 저준위 또는 오염되지 않은 금속폐기물의 재활용은 폐기물 발생 저감은 물론 처분장의 공간을 절약하는데 기여할 것이다. 본 논문은 재활용 시스템의 개념설계와 정의된 업무 흐름에서 발생 하는 피폭 선량을 평가하는데 그 목적이 있다. 작업의 흐름과 운전 개념을 정립하기 위해 다양한 형태의 다이어그램을 설계 하였다. 선량평가에 필요한 시나리오는 개념설계를 기반으로 선정되었으며, RESRAD-RECYCLE을 이용하여 선량을 평가하였다. 이를 통하여, 결정적 시나리오 선별, 핵종 특성 및 핵종 분배가 선량에 미치는 영향을 분석하였다. 더 나아가, 선량분석은 피폭 시나리오에 대한 대체 방안 수립, 필요한 제염 및 방사선방어 프로세스 그리고 허용 방사능 검토의 정보를 제공 하는데 사용 될 수 있을 것이다.

폐-광석으로부터 금속구리분말을 회수하기 위하여 더미 미생물용출, Fe 제거와 전기분해실험을 수행하였다. Cu가 0.034% 함유된 폐-광석시료에 대하여 더미 용출실험을 수행한 결과, Cu 용출률은 박테리아 용출-용액에서 61%, 황산 용출-용액에서 62%로 나타났다. Fe를 효과적으로 제거하기 위하여 더미 용출-용액에 NaOH, H2O2 및 Ca(OH)2를 각각 적용한 결과 H2O2가 가장 효과적인 Fe 제거제로 선정되었다. 전해질 용액을 준비하기 위하여 H2O2를 더미 용출-용액에 처리한 결과 박테리아 용출-용액에서 Fe가 99%, 황산 용출-용액에서 60%로 제거된 반면에 Cu 제거율은 각각 5%와 7%로 나타났다. 이 용액에 대하여 전기분해 실험을 수행한 결과 Cu 회수율이 박테리아 용출-용액에서 98%, 황산 용출-용액에서 76%로 나타났다. 모수석 형태의 금속구리분말이 양쪽 용출-용액에서 회수되었다.

In Korea, two decommissioning projects have been carried out due to retire of nuclear research facilities such as Korean research reactors (KRR-1 & KRR-2) and a uranium conversion plant (UCP). The decommissioning of the KRR-2 and a uranium conversion plant (UCP) at KAERI were finished completely by 2011, whereas the decommissioning of KRR-1 is currently underway. The large quantity of radioactive waste was generated during the decommissioning the KRR and UCF such as concrete waste, soil, combustible and non combustible waste. The volume reduction of the combustible wastes through the incineration technologies has merits from the view point of decrease in the amount of waste to be disposed of resulting in a reduction of the disposal cost. Incineration is generally accepted as a method of reducing the volume of radioactive waste. The incineration technology is effective treatment method that contains hazardous chemical as well as radioactive contamination. Incinerator burns waste at high temperatures. Incineration of a mixture of chemically hazardous and radioactive materials, known as“mixed waste,”has two principal goals: to reduce the volume and total chemical toxicity of the waste. Incineration itself does not destroy the metals or reduce the radioactivity of the waste. A proven melting technology is currently used for low-level waste (LLW) at several facilities worldwide. These facilities use melting as a means of processing LLW for unrestricted release of the metal or for recycling within the nuclear sector. Fig. 1 shows the schematic diagram of the oxygen-enriched incineration (OEI) and melting facility. The oxygen-enriched incinerator located at the KAERI. The system consists of a waste preparation system, incineration system, off-gas cooling system, and off-gas treatment system. Demonstration incineration facility took over the responsibilities of KHNP for decommissioned combustible waste. After taking over the demonstration incineration facility from KHNP, the facility was modified, and work toward the licensing procedure, and an extension of the object waste including alpha-bearing waste and increase incineration capacity, began in June 2011. The melt decontamination technology is the most effective treatment method for decommissioned metal waste. Melting for size reduction would require no prior surface decontamination and very little sorting of the waste material. Also, the recycling or volume reduction of the metallic wastes through the melt decontamination technologies has merits from the view point of an increase in resource recycling as well as a decrease in the amount of waste to be disposed of resulting in a reduction of the disposal cost and an enhancement of the disposal safety. Melt facility consist of four system such as preparation system, melting system, ingot treatment, and off-gas treatment system. The decommissioned combustible waste has been incineration by incinerator from last year. In case of metal waste, metal waste will be melt for self-disposal and volume reduction by induction furnace. Combustible wastes were treated by incinerator and ash dispose permanently site. In case of metal wastes is treated by induction furnace and slag dispose permanently site and ingot will be reuse.

The aim of this study was to investigate the accumulation of metallic elements and the control effect of marine pollution caused by ocean dumping in the sediments at a waste disposal area in the Yellow Sea. In July 2009, concentrations of organic matter and metallic elements (Al, Fe, As, Cd, Cr, Co, Hg, Ni, Mn, Pb, and Zn) were measured in surface sediments at the site. The ignition loss (IL) in the surface sediments showed a mean value of 15.4%, about 1.5 times higher than the mean value of the sediments in the coastal areas of Korea. The chemical oxygen demand (COD) at some disposal sites exceeded 20 ㎎ O2/g·dry, which signifies the initial concentration of marine sediment pollutants in Japan. The disposal sites contain higher concentrations of Cr, Cu and Zn than the sediments of bays and estuaries that might be contaminated. The magnitude of both metal enrichment factors (EF) and adverse biological effects suggest that pollution with Cr and Ni occurred due to the dumping of waste in the study area. In addition, the geoaccumulation index (Igeo) showed that the surface sediments were moderately contaminated. By the mid-2000s, when the amount of waste dumped at this site was the highest, the concentration of metallic elements was higher than ever recorded. On the other hand, in 2008-09, the need for environmental management was relatively low compare with the peak. As a result, the quality of marine sediment has been enhanced, considering the effect of waste reduction and natural dilution in the disposal area.