• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.183-202
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• 2019

For design and performance assessment of a high-level radioactive waste (HLW) disposal system, it is necessary to understand the characteristics of coupled thermo-hydro-mechanical (THM) behavior. However, in previous studies for the Korean Reference HLW Disposal System (KRS), thermal analysis was performed to determine the spacing of disposal tunnels and interval of disposition holes without consideration of the coupled THM behavior. Therefore, in this study, TOUGH2-MP/FLAC3D is used to conduct THM modeling for performance assessment of the Korean Reference HLW Disposal System (KRS). The peak temperature remains below the temperature limit of $100^{\circ}C$ for the whole period. A rapid rise of temperature caused by decay heat occurs in the early years, and then temperature begins to decrease as decay heat from the waste decreases. The peak temperature at the bentonite buffer is around $96.2^{\circ}C$ after about 3 years, and peak temperature at the rockmass is $68.2^{\circ}C$ after about 17 years. Saturation of the bentonite block near the canister decreases in the early stage, because water evaporation occurs owing to temperature increase. Then, saturation of the bentonite buffer and backfill increases because of water intake from the rockmass, and bentonite buffer and backfill are fully saturated after about 266 years. The stress is calculated to investigate the effect of thermal stress and swelling pressure on the mechanical behavior of the rockmass. The calculated stress is compared to a spalling criterion and the Mohr-Coulumb criterion for investigation of potential failure. The stress at the rockmass remains below the spalling strength and Mohr-Coulumb criterion for the whole period. The methodology of using the TOUGH2-MP/FLAC3D simulator can be applied to predict the long-term behavior of the KRS under various conditions; these methods will be useful for the design and performance assessment of alternative concepts such as multi-layer and multi-canister concepts for geological spent fuel repositories.

• Economic and Environmental Geology
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• v.56 no.4
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• pp.421-433
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• 2023

The final disposal of spent nuclear fuel(SNF) from nuclear power plants takes place in a deep geological repository. The metal canister encasing the SNF is made of cast iron and copper, and is engineered to effectively isolate radioactive isotopes for a long period of time. The SNF is further shielded by a multi-barrier disposal system comprising both engineering and natural barriers. The deep disposal environment gradually changes to an anaerobic reducing environment. In this environment, sulfide is one of the most probable substances to induce corrosion of copper canister. Stress-corrosion cracking(SCC) triggered by sulfide can carry substantial implications for the integrity of the copper canister, potentially posing a significant threat to the long-term safety of the deep disposal repository. Sulfate can exist in various forms within the deep disposal environment or be introduced from the geosphere. Sulfate has the potential to be transformed into sulfide by sulfate-reducing bacteria(SRB), and this converted sulfide can contribute to the corrosion of the copper canister. Bentonite, which is considered as a potential material for buffering and backfilling, contains oxidized sulfate minerals such as gypsum(CaSO4). If there is sufficient space for microorganisms to thrive in the deep disposal environment and if electron donors such as organic carbon are adequately supplied, sulfate can be converted to sulfide through microbial activity. However, the majority of the sulfides generated in the deep disposal system or introduced from the geosphere will be intercepted by the buffer, with only a small amount reaching the metal canister. Pyrite, one of the potential sulfide minerals present in the deep disposal environment, can generate sulfates during the dissolution process, thereby contributing to the corrosion of the copper canister. However, the quantity of oxidation byproducts from pyrite is anticipated to be minimal due to its extremely low solubility. Moreover, the migration of these oxidized byproducts to the metal canister will be restricted by the low hydraulic conductivity of saturated bentonite. We have comprehensively analyzed and summarized key research cases related to the presence of sulfates, reduction processes, and the formation and behavior characteristics of sulfides and pyrite in the deep disposal environment. Our objective was to gain an understanding of the impact of sulfates and sulfides on the long-term safety of high-level radioactive waste disposal repository.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.1
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• pp.23-32
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• 2011

An engineering scale test, which is called KENTEX, was carried out to understand and to analyze the coupled thermal, hydrological and mechanical phenomena in the engineered barrier system(EBS) of Korean reference disposal system. Using the experimental data obtained from KENTEX, the water saturation processes in bentonite could be analyzed. From the comparison between the model calculation using ABAQUS and the experimental results, the difference of the water content between them in the unsaturating part was large because the drying phenomena due to moisture redistribution by the temperature gradient could not be included in the model. In the saturating part, the difference of the water content between them was decreased gradually and showed to be small in the full saturation. And the time of about 95% saturation could be estimated about 500 days from the model calculation and experimental results. Also it could be known that the moisture redistribution in the unsaturated part could not be affected on the saturation time of bentonite in the repository. Therefore, it is considered that this model could be used to quantitatively predict the water saturation time in bentonite as EBS for the disposal system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.11-21
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• 2013

The Geyoungju Ca-bentonite with dry density of 1.6 g/$cm^3$ has been considered as a standard buffer material for the disposal of high level waste in KAERI disposal system design. But it had relatively lower thermal conductivity compared with other surrounding materials, that was one of key parameters to limit the increase of the disposal density in the disposal system. In this study, various additives were selected and mixed with the Ca-bentonite in different mixing methods in order to increase the thermal conductivity from 0.8 W/mK to 1.0 W/mK. As an additive, CNT (Cabon Nano Tube), graphite, alumina, CuO, and $Fe_2O_3$ were selected, which are chemically stable and have good thermal conductivity. As mixing methods, dry hand-mixer mixing, wet milling and dry ball mill mixing were applied for the mixing. Above all, the ball mill mixing was proved to be most effective since the produced mixture was most homogeneous and showed higher increase in the thermal conductivity. From this study, it was confirmed that the thermal conductivity for the Geyoungju Ca-bentonite could be improved by adding small amount of highly thermal conductive material to 1.0 W/mk. In conclusion, it was believed that the experimental results will be valuable in the disposal system design if the additive effects on the swelling and permeability on the compact bentonite are also approved in further studies.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.101-109
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• 2008

For the Kyungju bentonite which is considered as a candidate material for the buffer and backfill in the high-level waste repository, the thermal conductivities of compacted bentonite and a bentonite-sand mixture were measured. The thermal conductivities of the compacted bentonites with a dry density of 1.2 to $1.8\;Mg/m^3$ and the bentonite-sand mixture with a dry density of 1.6 and $1.8\;Mg/m^3$ were measured within the gravimetric water content range of 10wt% to 20wt% and the sand fraction range of 10 to 30wt%. The thermal conductivity of compacted bentonite and a bentonite-sand mixture increases with increasing dry density and sand weight fraction in the case of constant water weight fraction, and increases with increasing water weight fraction and sand weight fraction in the case of constant dry density. The empirical correlations to describe the thermal conductivity of compacted bentonite and a bentonite-sand mixture as a function of water fraction at each dry density were suggested. These correlations can predict the thermal conductivities of bentonite and a bentonite-sand mixture with a difference below 10%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.329-338
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• 2019

Suction of unsaturated bentonite buffers is a very important input parameter for hydro-mechanical performance assessment and design of an engineered barrier system. This study analyzed suction measurement techniques and constitutive models of unsaturated porous media reported in the literature, and suggested suction measurement techniques and constitutive models suitable for bentonite buffer in an HLW repository. The literature review showed the suction of bentonite buffer to be much higher than that of soil, as measured by total suction including matric suction and osmotic suction. The measurement methods (RH-Cell, RH-Cell/Sensor) using a relative humidity sensor were suitable for suction measurement of the bentonite buffer; the RH-Cell /Sensor method was more preferred in consideration of the temperature change due to radioactive decay heat and measurement time. Various water retention models of bentonite buffers have been proposed through experiments, but the van Genuchten model is mainly used as a constitutive model of hydro-mechanical performance assessment of unsaturated buffers. The water characteristic curve of bentonite buffers showed different tendencies according to bentonite type, dry density, temperature, salinity, sample state and hysteresis. Selection of water retention models and determination of model input parameters should consider the effects of these controlling factors so as to improve overall reliability.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.453-462
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• 1995

In connection with the safe storage of high level nuclear waste, effect of $H_2O$$_2$ on the corrosion behavior of 304L stainless steel was examined. Open circuit potentials and polarization curves were measured with and without $H_2O$$_2$. The experimental results show that $H_2O$$_2$ increased corrosion potential and decreased pitting potential. The passive range, therefore, decreased as $H_2O$$_2$ concentration increased, indicating that pitting resistance was decreased by the existence of $H_2O$$_2$ in the electrolyte. These effect of $H_2O$$_2$ on corrosion of 304L stainless steel are considered to be similar to those of ${\gamma}$-irradiation. To compare the effects of $H_2O$$_2$ with those of $O_2$, cathodic and anodic polarization curves ore made in three types of electrolyte such as aerated, deaerated, and stirred electrolyte. The experimental results show that the effects of $H_2O$$_2$ on the corrosion behavior were tory similar to those of $O_2$ such as increase of corrosion potential, decrease of pitting resistance, and increase of repassivation potential. In acid and alkaline media, the corrosion potential shifts by $H_2O$$_2$ were restricted by the large current density of proton reduction and by the le Chatelier's principle respectively.y.

• Tunnel and Underground Space
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• v.34 no.1
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• pp.15-27
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• 2024

In this case study, detailed survey of the Excavation Damaged Zone (EDZ) evaluation for the deep geological repository for high level nuclear waste was conducted. Oversea and Domestic case studies were compiled and investigated. EDZ is considered a crucial factor in the performance assessment of spent fuel disposal, leading to numerous studies worldwide aiming to understand the characteristics of the EDZ and quantitatively assessment of its extent through field and laboratory tests at Underground Research Laboratory (URL) sites. To enhance the understanding of EDZ, this study begins with defining and exploring the history of EDZ, compiling factors influencing EDZ, and summarizing the impacts caused by EDZ. Subsequently, an analysis of EDZ and rock properties is performed, followed by presenting generalized outcomes, limitations drawn from previous research, and proposing future research directions.

• The Korean Journal of Nuclear Medicine Technology
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• v.12 no.1
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• pp.19-26
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• 2008

Purpose: In general, We discharged radioactive wastewater and sewages less than $8.1{\times}10^{-13}$ Ci/ml in a exclusive water-purifier tank. Our hospital operating three exclusive water-purifier tank for radioactive wastewater and sewages of 60 tons capacity respectively. In order to meet the criteria it need a enough decay more than 125 days per each exclusive tank. However, recently we fell into the serious situation that decay period was decreased remarkably, owing to the wastewater amount increased rapidly by enlarge the therapy ward. For that reason, in this article, I'd like to say the way that reducing of radioactive wastewater and sewages rationally. Materials and Methods: From January, 2006 to October, four hundred and two cases were analyzed. They were all hospitalized during 3 days and 2 nights. We calculated the average amount of water used (include toilet water used, shower water used, washstand water used, $\cdots$), each exclusive water-purifier tank's decay period, as well as try to search the increased factors about water-purifier tank inflow flux by re-analysis of the procedure of radioisotope therapy step by step. Results: We could increase each exclusive water-purifier tank's decay period from 84 days to 130 days through the improvement about following cause: (1) Improvement of conventional toilet stool for excessive water waste $\rightarrow$ Replacement of water saving style toilet stool (2) Prevention of unnecessary shower and wash (3) Stop the diuretics taking during hospitalization (4) Analysis of relationship between water intakes and residual dose of body (5) Education about outside toilet utilization before the administration (6) Changed each water-purifier tank's maximum level from85% to 90% Conclusion: The originality of our efforts are not only software but hardware performance improvements. Incidentally the side of software's are change of therapy procedures and protocols, the side of hardware's are replacement of water saving style toilet stool and change of each water-purifier tank's maximum level. Thus even if a long lapse of time, problem such as return to the former conditions may not happen. Besides, We expect that our trials become a new reasonable model in similar situation.

• Applied Chemistry for Engineering
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• v.28 no.2
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• pp.133-140
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• 2017

Increasing concerns on radioactive wastes have drawn much attention on the development of remediation technologies. Massive amounts of radioactive wastes generated from hospital and nuclear power plants were exposed to our environment. Although physicochemical removal methods were developed, an eco-friendly remediation method has not yet been demonstrated. Recently, an extremophilic bacterium has received much attention due to their extraordinary characteristics. Among them, Deinococcus radiodurans (D. radiodurans) strain was regarded as the best host organism for the removal of radioactive heavy metals and radionuclides, because of their superb characteristics like tolerance against the high level of radioactivity. In this article, we briefly introduced the extraordinary nature of D. radiodurans and also discussed the potential use of D. radiodurans strain for the removal of radioactive wastes.