• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.465-479
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• 2020

Two waste forms, namely cement and geopolymer, were investigated and tested in this study to solidify the corrosive sludge generated from the surface and precipitates of the tubes of steam generators in nuclear power plants. The compressive strength of the cement waste form cured for 28 days was inversely proportional to waste loading (24.4 MPa for 0wt% to 2.7 MPa for 60wt%). The corrosive sludge absorbed the free water in the hydration reaction to decrease the cementation reaction. When the corrosive sludge waste loading increased to 60wt%, the cement waste form showed decreased compressive strength (2.7 MPa), which did not satisfy the acceptance criteria of the repository (3.45 MPa). Meanwhile, the compressive strength of the geopolymer waste form cured for 7 days was proportional to waste loading (23.6 MPa for 0wt% to 31.9 MPa for 40wt%). The corrosive sludge absorbed the free water in the geopolymer when the water content decreased, such that a compact geopolymer structure could be obtained. Consequently, the geopolymer waste forms generally showed higher compressive strengths than cement waste forms.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.65-82
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• 2017

The result of a preliminary safety assessment that was completed by applying the radionuclide inventory calculated on the basis of available data from radioactive waste generation agencies suggested that many difficulties are to be expected with regard to disposal safety and operation. Based on the results of the preliminary safety assessment of the entire disposal system, in this paper, a unit package exceeding the safety goal is selected that occupies a large proportion of radionuclides in intermediate-level radioactive waste. We introduce restrictions on the amount of radioactivity in a way that excludes the high surface dose rate of the package. The radioactivity limit for disposal will be used as the baseline data for establishing the acceptance criteria and the disposal criteria for each disposal facility to meet the safety standards. It is necessary to draw up a comprehensive safety development plan for the Gyeongju waste disposal facility that will contribute to the construction of a Safety Case for the safety optimization of radioactive waste disposal facilities.

• Proceedings of the Korean Geotechical Society Conference
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• 2005.03a
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• pp.1434-1441
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• 2005

For anchored system applications, each ground anchor is tested after installation and prior to being put into service to loads that exceed the design. This load testing methodology, combined with specific acceptance criteria, is used to verify that the ground anchor can carry the design load without excessive deformations and that the assumed load transfer mechanisms have been properly developed behind the assumed critical failure surface. After acceptance, the ground anchor is stressed to a specified load and the load is locked-off. The two types of load tests conducted during the research program included performance test and creep test which were carried out in accordance with testing procedures by AASHTO(AASHTO 1990) and FHWA(Weatherby 1998) at Samsung-Dong 00 Site. Form the measurements, ultimate load and creep rate of anchors are proposed for straight shaft pressured grouted anchors in waste landfill. The load distribution on the grout was obtained from the measured strain data at each fraction of the ultimate load during the load tests.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2889-2896
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• 2023

The waste acceptance criteria for heavy metals in mixed waste should be developed by reflecting the leaching behaviors that could highly depend on the repository design and environment surrounding the waste. The current standards widely used to evaluate the leaching characteristics of heavy metals would not be appropriate for the silo-type repository since they are developed for landfills, which are more common than a silo-type repository. This research aimed to explore the leaching behaviors of cementitious waste with Pb, Cd, and Sb metallic and oxide powders in an environment simulating a silo-type radioactive waste repository. The Toxicity Characteristic Leaching Procedure (TCLP) and the ANS 16.1 standard were employed with standard and two modified solutions: concrete-saturated deionized and underground water. The compositions and elemental distribution of leachates and specimens were analyzed using an inductively coupled plasma optical emission spectrometer (ICP-OES) and energy-dispersive X-ray spectroscopy combined with scanning electron microscopy (SEM-EDS). Lead and antimony demonstrated high leaching levels in the modified leaching solutions, while cadmium exhibited minimal leaching behavior and remained mainly within the cement matrix. The results emphasize the significance of understanding heavy metals' leaching behavior in the repository's geochemical environment, which could accelerate or mitigate the reaction.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.271-281
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• 2023

A study was conducted on the vitrification of the rare earth oxide waste generated from the PyroGreen process. The target rare earth waste consisted of eight elements: Nd, Ce, La, Pr, Sm, Y, Gd, and Eu. The waste loading of the rare earth waste in the developed borosilicate glass system was 20wt%. The fabricated glass, processed at 1,200℃, exhibited uniform and homogeneous surface without any crystallization and precipitation. The viscosity and electrical conductivity of the melted glass at 1,200℃ were 7.2 poise and 1.1 S·cm^-1, respectively, that were suitable for the operation of the vitrification facility. The calculated leaching index of Cs, Co, and Sr were 10.4, 10.6, and 9.8, respectively. The evaluated Product Consistency Test (PCT) normalized release of the glass indicated that the glass satisfied the requirements for the disposal acceptance criteria. Furthermore, the pristine, 90 days water immersed, 30 thermal cycled, and 10 MGy gamma ray irradiated glasses exhibited good compressive strength. The results indicated that the fabricated glass containing rare earth waste from the PyroGreen process was acceptable for the disposal in the repository, in terms of chemical durability and mechanical strength.

• International Journal of Fluid Machinery and Systems
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• v.3 no.4
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• pp.285-291
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• 2010

To provide required flow rates of cooling or circulating water properly, approach flow conditions of vertical pump systems should be in compliance with state of the art acceptance criteria. The direct inflow should be vortex free, with low pre-rotation and symmetric velocity distribution. Physical model investigations are common practice and the best tool of prediction to evaluate, to optimize and to document flow conditions inside intake structures for vertical pumping systems. Optimization steps should be accomplished with respect to installation costs and complexity on site. The report shows evaluation of various approach flow conditions inside a compact waste water pumping station. The focus is on the occurrence of free surface vortices and the evaluation of air entrainment for various water level and flow rates. The presentation of the results includes the description of the investigated intake structure, occurring flow problems and final recommendations.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.60-70
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• 2003

The stability of various waste forms to meet waste acceptance criteria was evaluated by using standard test methods of U.S.A and France. Compressive strength of waste forms were above 176.03 kgf/$\textrm{cm}^2$(cement), 15 kgf/$\textrm{cm}^2$(paraffin). In the thermal cycling test, there were no any change in their feature and volume, the loss of weight was 6.15% on the average. In the immersion test for 120 days, the loss of weight of paraffin waste form was 8.85-5.14% pH=3.83. The G-Value of $H_2$ and $CH_4$ in paraffin wax at $10^8rads$ rads of exposure dose were 2.65, 0.016.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.411-417
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• 2023

During the operation of a nuclear power plant (NPP), the generation of radioactive waste, including dry active waste (DAW), concentrates, spent resin, and filters, mandates the implementation of appropriate disposal methods to adhere to Korea's waste acceptance criteria (WAC). In this context, this study investigates the potential use of polymer concrete (PC) as a high-integrity container (HIC) material for solidifying and packaging these waste materials. PC is a versatile composite material comprising binding polymers, aggregates, and additives, known for its exceptional strength and chemical stability. A comprehensive analysis of PC's long-term integrity was conducted in this study. First, its compressive strength, which is crucial for ensuring the structural stability of HICs over extended periods, was evaluated. Subsequently, the resilience of PC was tested under various stress conditions, including biological, radiological, thermal, and chemical stressors. The findings of this study indicate that PC exhibits remarkable long-term properties, demonstrating exceptional stability even when subjected to diverse stressors. The results therefore underscore the potential viability of PC as a reliable material for constructing high-integrity containers, thus contributing to the safe and sustainable management of radioactive waste in NPPs.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.149-160
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• 2007

The IAEA launched the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) and developed the INPRO Methodology to provide guidelines and to assess the characteristics of a future innovative nuclear energy system in areas such as safety, economics, waste management, and proliferation resistance. The proliferation resistance area of the INPRO Methodology is reviewed here, and modifications for further improvements are proposed. The evaluation metrics including the evaluation parameters, evaluation scales and acceptance limits are developed for a practical application of the methodology to assess the proliferation resistance. The proliferation resistant characteristics of the DUPIC fuel cycle are assessed by applying the modified INPRO Methodology based on the developed evaluation metrics and acceptance criteria. The evaluation procedure and the metrics can be utilized as a reference for an evaluation of the proliferation resistance of a future innovative nuclear energy system.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.280-289
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• 2005

A pyrochemical process has been introduced and utilized so that the transmutation of spent PWR fuel in PEACER can produce mainly low and intermediate level waste for near surface disposal. Major radioactive nuclides from PEACER pyroprocessing are composed of TRU and LLFP. In this study, the requirement for the final waste from PEACER is evaluated based on the methodology for establishment of waste acceptance criteria. Also, sensitivity analysis for several input parameters is conducted in order to determine acceptable decontamination factor (DF) and LLFP removal efficiency and to find out input parameter that extremely have an effect on DE As a result of the study, LLFP removal efficiency, especially Sr-90 and Tc-99, is proved to be a major nuclide which contributes to annual dose by human intrusion scenario rather than TRU DF. More than $98.5\%$ of LLFP have to be removed to meet below dose constraint within the DF more than 5.0E+03. Besides, because of the relative short half-life of Sr-90, the increasing of the institutional control period is recommended for most important input parameter to determine DF.