Wasteform is the first barrier to prevent radionuclide release from repositories into the biosphere. Since leaching rates of nuclides in wasteform significantly impact on safety assessment of the repository, clarifying the leaching behavior is critical for accurate safety assessment. However, the current waste acceptance criteria (WAC) of the Gyeongju repository only evaluates leachability indexes for Cs, Sr, and Co, which are tracers for nuclear power plant waste streams. Furthermore, ANS 16.1, the current leaching test method used in WAC, applies deionized water (DI) as leachant. However, the interactions between wasteform and groundwater environment in the repository may not be reflected. Therefore, it is necessary to review the current leaching test method and nuclides that may require the extra evaluation of leachability beyond the Cs, Sr, and Co. Tc and I are key nuclides contributing to high radioactive dose in safety assessment due to their high mobility and low retardation factor. The groundwater conditions within the repository, such as pH and Eh significantly affect the chemical form of Tc and I. For example, Tc in H2O system tends to form hydroxide precipitates in neutral pH condition and TcO4 - in strong alkaline environments according to the Pourbaix diagram. In case of I, it generally exists in the form of I-, while it exists as IO3 - as Eh increases. Although the current leaching test at the Gyeongju repository applies DI as a leachant, the actual repository is expected to have a highly alkaline environment with a substantial amount of various ions in the groundwater. Consequently, the leaching behavior in the ANS 16.1 test and the actual disposal condition is different. Thus, it is necessary to analyze the leaching behavior of Tc and I with reflecting the actual disposal environment. In this study, the leaching behavior of Tc and I is investigated by following ANS 16.1 leaching test method. The solidified waste specimens containing 10 mmol of Re and I were manufactured with cement, which is widely used as a solidification material. Re was applied instead of Tc, which has similar chemical behavior to Tc, and NH4ReO4 and NaI were used as surrogates for Re and I. As a leachant, deionized water and cement-saturated groundwater were prepared and the concentration of nuclides in the leachant is analyzed by ICP-OES. As the result of this study, experimental data can be applied to improve the WAC and disposal concentration standards in the future.

Low- and intermediate level waste (LILW) repository in Gyeongju, Korea is in operation and the radioactive waste should satisfy the waste acceptance criteria (WAC) of the repository. Among the WAC of the Gyeongju LILW repository, the leachability index criterion is considered to be the criterion that is directly related to the isolation of the radionuclides from biosphere. Cesium, strontium, and cobalt should satisfy the leachability index larger than six by following the ANS 16.1 leaching test method. Several research were performed for the leachability index of Cs, Sr, and Co by following the ANS 16.1 leaching test method. However, the test condition of the previous research is expected to be different to the condition of the actual waste. Due to the radioactivity of the radionuclide such as Cs, Sr and Co, most of the research applied the surrogate of those radionuclides. The concentration of those nuclides was generally measured by the inductively coupled plasma (ICP) equipment, however, high concentration compared to the disposal limit of those nuclides due to the detection limit of the ICP was applied. From the Freundlich and Langmuir adsorption isotherms, the adsorption of the nuclides differs according to the concentration of the nuclides. As the leachability index of the nuclides is affected by the adsorption of the nuclides on the binding material, the effect of nuclide concentration is expected to be not ignorable. Therefore, the leachability index difference according to the nuclide concentration should be compared to avoid over- or underestimation of the leachability index. In this study, the difference in the leachability index according to the concentration of nuclides is aimed to be checked. Cs, Sr, and Co, which should satisfy the leachability index criterion in the WAC of the Gyeongju repository, were selected as target nuclides. Three concentrations were selected to compare the leachability index: 0.1 mol, 0.001 mol and below the regulatory exemption concentration. The concentration of non-radioactive nuclides in the leachant was measured by ICPOES and ICP-MS while the concentration of radionuclides was measured by HPGe. The result of this study can be applied as background data enhancing the WAC or disposal concentration limit of the radionuclides in Gyeongju LILW repository.

Structural stability of a waste form can be provided by the waste form itself (steel components, etc.), by processing the waste to a stable form (solidification, etc.), or by emplacing the waste in a container or structure that provides stability (HICs or engineered structure, etc.). The waste or container should be resistant to degradation caused by radiation effects. In accordance with the requirements for the domestic waste acceptance criteria, irradiation testing of solidified waste forms containing spent resin should be conducted on specimens exposed to a dose of 1.0E+6 Gy and other material 1.0E+7 Gy. Expected cumulative dose over 300 years is about 1.770E+6 Gy for spent resin and 0.770E+6 Gy for dried concentrated waste generated from NPPs generally. According to NRC Waste Form Technical Position, to ensure that spent resins will not undergo adverse degradation effects from radiation, resins should not be generated having loadings that will produce greater than 1E+6 Gy total accumulated dose. If it necessary to load resins higher than 1E+6 Gy, it should be demonstrated that the resin will not undergo radiation degradation at the proposed higher loading. This is the recommended maximum activity level for organic resins based on evidence that while a measurable amount of damage to the resin will occur at 1E+6 Gy, the amount of damage will have negligible effect on disposal site safety. Cementitious materials are not affected by gamma radiation to in excess of 1E+6 Gy. Therefore, for cement-stabilized waste forms, irradiation qualification testing need not be conducted unless the waste forms contain spent resins or other organic media or the expected cumulative dose on waste forms containing other materials is greater than 1E+7 Gy. Testing should be performed on specimens exposed to IE+6 Gy or the expected maximum dose greater than 1E+6 Gy for waste forms that contain ion exchange resins or other organic media or the expected maximum dose greater than 1E+7 Gy for other waste forms. This is suggestion as a review result that requirement for irradiation testing of solidified waste forms has something to be revise in detail and definitively.

The design and fabrication of suitable waste forms with high thermal and structural stability are essential for the safe management and disposal of radioactive wastes. In particular, the thermal properties and temperature distribution of waste form containing high heat-generating nuclides such as Cs and Sr can be used to evaluate its thermal stability, but also provide useful information for the design of canisters, storage systems, and repositories. In this study, a new program code-based thermal analysis framework has been developed to facilitate the characterization, design, and optimization of the waste form. Matlab was used as a software development platform because it provides powerful mathematical computation and visualization components such as the partial differential equation (PDE) toolbox for solving heat transfer problems using finite element method, the App Designer for graphical user interface (GUI), and the MATLAB Compiler for sharing MATLAB programs as standalone applications and web applications. The thermal analysis results such as temperature distribution, heat flux, maximum/ minimum temperature, and centerline/surface temperature profile are visualized with graphs and tables. To evaluate the effectiveness of the developed program, several design and optimization studies were carried out for the SrTiO3 waste form, selected as a stable form of strontium nuclide.

In domestic nuclear power plants, drums of concentrated radioactive waste solidified with paraffin that do not meet radioactive waste disposal standards are stored temporarily. In this paper, the design of a machine that separates these paraffin drums into paraffin and concentrated waste using heating vaporization and pressure difference is described. The separation process is as follows. First, the paraffin solid is indirectly heated by heating the outside of the drum. The paraffin solid is partially melted to increase the fluidity and is easily detached from the drum. The detached solid is transferred to the melting tank, and further heated in the melting tank. When the temperature is sufficiently high, paraffin is melted and becomes a mixture of liquid paraffin and concentrated waste homogeneously. The mixed solution is transferred to a paraffin recovery vessel and further heated. The vaporization point of paraffin is 370°C under atmospheric pressure, and becomes lower depending on the pressure decreasing in the vessel. The vaporization point of the paraffin is a relatively low value compared to the radioactive elements in the concentrated waste, and therefore only paraffin would be vaporized. A paraffin transfer pipe is installed on the upper part of the paraffin recovery vessel, and is connected to another tank called the paraffin capture vessel. The pressure of the paraffin capture vessel is reduced (i.e. vacuum condition), only gaseous paraffin is transferred to the paraffin capture vessel by the pressure difference. When the paraffin capture vessel is cooled below the vaporization point of the paraffin, the paraffin is liquefied or solidified, and only the paraffin is recovered. Based on the above process, the solidified paraffin could be separated into pure paraffin and concentrated waste. However, if a radioactive element with a lower vaporization point than paraffin exists in the concentrated waste, it may be mixed with paraffin and separated together. Therefore, it is necessary to measure the radioactivity or radiation dose rate for the separated paraffin, and to verify that it is sufficiently low. If necessary, additional separation process may be considered for removing radioisotopes from the paraffin.

Low-and intermediate level waste (LILW) should be solidified and satisfy the waste acceptance criteria (WAC) to be disposed of in the LILW repository. The LILW should be uniformly solidified and should maintain its structural stability under the expected condition according to the WAC. Compressive strength of cement solidified waste should satisfy at least 3.44 MPa to be disposed of in the repository. In addition, its compressive strength should satisfy at least 3.44 MPa after the irradiation, immersion and leaching test. The compressive strength test and dimension of test specimen differ according to countries. However, measured compressive strength of solidified waste is affected by geometry of specimen and test condition. Diameter, ratio between diameter and height, and porosity are one of factors that affect to the compressive strength of cement solidified waste. Generally, specimen with larger diameter shows higher value of measured compressive strength. The ratio of height and diameter shows similar tendency to the diameter while larger porosity generally lowers the compressive strength. In other hands, higher compressive strength is expected when the loading rate is higher during the compressive strength test. U.S. is applying loading rate from ASTM C39 (0.25±0.05 MPa) for the compressive strength test while Korea is applying loading rate from KS F 2405 (0.6 MPa·s−1). France applies loading rate following FT-02-010 (0.5 MPa·s−1) for cement solidified waste. As the measured compressive strength increases when the loading rate increases, the effect of loading rate to the compressive strength of cement solidified waste should be assessed by quantification and consider its effect on the sight of regulation. In this study, the effect of geometric parameters of specimen and test condition to the compressive strength are checked by manufacturing specimen by solidifying mock sludge waste with cement. To prevent increasing amount of secondary waste, effects of ratio of height and diameter and porosity to the compressive strength are checked while diameter value is fixed. For loading rate, loading rate from ASTM C39 and KS F 2405 were compared. Existence of significant variance of measured compressive strengths of cement solidified waste are check by performing statistical analysis. Finally, by analyzing the relationship between test condition and measured compressive strength, the test method that measures the compressive strength conservatively is aimed to be derived.

The decommissioning of a nuclear power plant generates large amounts of radioactive waste, which is of several types. Radioactive concrete powder is classified as low-level waste, which can be disposed of in a landfill. However, its safe disposal in a landfill requires that it be immobilized by solidification using cement. Herein, a safety assessment on the disposal of solidified radioactive concrete powder waste in a conceptual landfill site is performed using RESRAD. Furthermore, sensitivity analyses of certain selected input parameters are conducted to investigate their impact on exposure doses. The exposure doses are estimated, and the relative impact of each pathway on them during the disposal of this waste is assessed. The results of this study can be used to obtain information for designing a landfill site for the safe disposal of low-level radioactive waste generated from the decommissioning of a nuclear power plant.

우라늄 토양 및 콘크리트 폐기물의 동전기 제염 후 방사성폐기물의 시멘트 고화특성을 분석하기 위하여, 시멘트 고화 유동성 시험을 수행하고 시멘트 고화 시료를 제작하였다. 시멘트 고화시료에 대하여 압축강도, pH, 전기전도도, 방사선조사 효과 및 부피증가를 분석하였다. 방사성폐기물의 시멘트 고화의 작업 적정도는 175~190% 정도였다. 시멘트 고화시료의 방사선 조사 후 압축강도는 방사선 조사 전 압축강도 보다 약 15% 감소하였으나, 한국원자력환경공단 인수기준 (34 kgf·cm-2)을 만족하였다. 동전기 제염 후 방사성폐기물의 시멘트 고화 시료에 대한 SEM-EDS 분석결과, 알루미늄상은 시멘트와 잘 결합 한 형상을 나타낸 반면, 칼슘상은 시멘트와 분리된 형상을 나타내었다. 방사성폐기물의 시멘트 고화 부피는 시멘트에 대한 폐기물의 배합과 수분량에 따라 다르게 나타났다. 방사성폐기물의 시멘트 고화 부피(C-2.0-60)는 약 30% 증가였으며 동전기 제염 후 생성된 방사성폐기물의 영구처분은 적절하다고 판단되었다.

본 연구에서는 암석폐재를 대상으로 직경이 60mm인 고화체를 얻을 수 있는 수열 hot press 장치를 개발하였고, 고형화를 위한 최적조건을 찾아내는 내용을 다루었다. 이어 고화체의 기계적 성질을 평가하였고, 고화기구와 미시적 파괴거동을 규명하기위하여 SEM관찰 및 음향방출실험을 실시하였다. 고형화를 위한 최적조건은 NaOH용액이 10wt%, 수열온도가 300˚C이고, 유지시간이 1시간이었다. 또한 수열반응동안에 다양한 제 2화합물들이 생성되었으며, 이들은 고화체의 기계적 성질에 큰 영향을 미쳤다. 아울러 원석의 경우에는 AE Counts가 초기화중에서부터 나타났으나, 고화체는 초기하중에서 전혀 AE Counts가 검출되지 않았다. 이와같은 사실로부터 수열 hot press법에 의해 얻어진 고화체는 원석과는 다르게 암석 입자간의 결합이 보다 치밀하게 이루어지고 있음을 유추할 수 있다.