• Title/Summary/Keyword: HLW

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Performance Assessment for Radionuclides Transport from HLW Repository (고준위방사성폐기물 처분장으로부터 핵종이동 평가)

  • 김성기;강철형;이연명;황용수
    • Proceedings of the Korean Society of Soil and Groundwater Environment Conference
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    • 2001.09a
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    • pp.41-46
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    • 2001
  • 요오드나 세순 같은 핵종들은 고용해도 핵 종들로서 사용 후 핵연료 내 피복관 이나 연료 결정 경계면에 위치하고 있다가 고준위 방사성폐기물 처분 후 지하수가 용기를 부식시키고 용기 내부로 침투하면 고용해도를 가지고 유출된 후 공학적, 천연 방벽을 통해 최종적으로 유출되게 된다. 본 연구에서는 한국원자력연구소에서 개발한 MASCOT-K글 이용하여 고용해도 핵 종들이 조화 유출과 고용해도 유출할 경우 유출 량을 평가 분석해 보았다. 평가 결과 요오드와 같은 고용해도 핵 종인 경우 전체 핵 종 재고량의 최대 10%만이 고용해도 유출을 하지만 그 영향은 조차 유출에 비해 훨씬 중요한 것으로 판명되었다. 이러한 결과를 바탕으로 현재 국내 고 준위 처분 환경에서 보수적인 시나리오로 주목받고 있는 우물 굴착 시나리오를 대상으로 우물까지의 거리 등 입력 자료의 불확실성을 평가해 보았다. 36,000 톤의 사용 후 핵연료를 처분 대상으로 했을 때 성능 평가 결과는 현재 처분 개념이 안전함을 입증한다.

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ASSESSMENT OF THE COST OF UNDERGROUND FACILITIES OF A HIGH-LEVEL WASTE REPOSITORY IN KOREA

  • Kim, Sung-Ki;Choi, Jong-Won
    • Nuclear Engineering and Technology
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    • v.38 no.6
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    • pp.561-574
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    • 2006
  • This study presents the results of an economic analysis for a comparison of the single layer and double layer alternatives with respect to a HLW-repository. According to a cost analysis undertaken in the Korean case, the single layer option was the most economical alternative. The disposal unit cost was estimated to be 222 EUR/kgU. In order to estimate such a disposal cost, an estimation process was sought after the cost objects, cost drivers and economic indicators were taken into consideration. The disposal cost of spent fuel differs greatly from general product costs in the cost structure. Product costs consist of direct material costs and direct labor and manufacturing overhead costs, whereas the disposal cost is comprised of construction costs, operating costs and closure costs. In addition, the closure cost is required after a certain period of time elapses following the building of a repository.

Mechanical behavior of an underground research facility in Korea Atomic Energy Research Institute

  • Kwon S.K.;Cho W.J.;Hahn P.S.
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2005.11b
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    • pp.245-252
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    • 2005
  • An underground research facility (KURF) is under construction at KAERI for the in situ studies related to the validation of a HLW disposal system. For the safe construction and long-term researches at KURF, mechanical stability of the facility should be evaluated. In this study, 3D mechanical stability analysis using the rock mass properties determined from various in situ as well as laboratory tests was carried out. From the analysis, it was possible to predict the rock deformation, stress concentration, and plastic zone developed before and after the excavation. A test blasting was performed to characterize the site dependent dynamic response, which can be used for the prediction of the blasting impact on the facilities in KAERI.

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AN ANALYSIS OF THE THERMAL AND MECHANICAL BEHAVIOR OF ENGINEERED BARRIERS IN A HIGH-LEVEL RADIOACTIVE WASTE REPOSITORY

  • Kwon, S.;Cho, W.J.;Lee, J.O.
    • Nuclear Engineering and Technology
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    • v.45 no.1
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    • pp.41-52
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    • 2013
  • Adequate design of engineered barriers, including canister, buffer and backfill, is important for the safe disposal of high-level radioactive waste. Three-dimensional computer simulations were carried out under different condition to examine the thermal and mechanical behavior of engineered barriers and rock mass. The research looked at five areas of importance, the effect of the swelling pressure, water content of buffer, density of compacted bentonite, emplacement type and the selection of failure criteria. The results highlighted the need to consider tensile stress in the outer shell of a canister due to thermal expansion of the canister and the swelling pressure from the buffer for a more reliable design of an underground repository system. In addition, an adequate failure criterion should be used for the buffer and backfill.

Effect of Bentonite Type on Thermal Conductivity in a HLW Repository

  • Lee, Gi-Jun;Yoon, Seok;Cho, Won-Jin
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.19 no.3
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    • pp.331-338
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    • 2021
  • Extensive studies have been conducted on thermal conductivity of bentonite buffer materials, as it affects the safety performance of barriers engineered to contain high-level radioactive waste. Bentonite is composed of several minerals, and studies have shown that the difference in the thermal conductivity of bentonites is due to the variation in their mineral composition. However, the specific reasons contributing to the difference, especially with regard to the thermal conductivity of bentonites with similar mineral composition, have not been elucidated. Therefore, in this study, bentonites with significantly different thermal conductivities, but of similar mineral compositions, are investigated. Most bentonites contain more than 60% of montmorillonite. Therefore, it is believed that the exchangeable cations of montmorillonite could affect the thermal conductivity of bentonites. The effect of bentonite type was comparatively analyzed and was verified through the effective medium model for thermal conductivity. Our results show that Ca-type bentonites have a higher thermal conductivity than Na-type bentonites.

Measuring thermal conductivity and water suction for variably saturated bentonite

  • Yoon, Seok;Kim, Geon-Young
    • Nuclear Engineering and Technology
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    • v.53 no.3
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    • pp.1041-1048
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    • 2021
  • An engineered barrier system (EBS) for the disposal of high-level radioactive waste (HLW) is composed of a disposal canister with spent fuel, a buffer material, a gap-filling material, and a backfill material. As the buffer is located in the empty space between the disposal canisters and the surrounding rock mass, it prevents the inflow of groundwater and retards the spill of radionuclides from the disposal canister. Due to the fact that the buffer gradually becomes saturated over a long time period, it is especially important to investigate its thermal-hydro-mechanical-chemical (THMC) properties considering variations of saturated condition. Therefore, this paper suggests a new method of measuring thermal conductivity and water suction for single compacted bentonite at various levels of saturation. This paper also highlights a convenient method of saturating compacted bentonite. The proposed method was verified with a previous method by comparing thermal conductivity and water suction with respect to water content. The relative error between the thermal conductivity and water suction values obtained through the proposed method and the previous method was determined as within 5% for compacted bentonite with a given water content.

Important Parameters Related With Fault for Site Investigation of HLW Geological Disposal

  • Jin, Kwangmin;Kihm, You Hong;Seo, Dong-Ik;Kim, Young-Seog
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.19 no.4
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    • pp.533-546
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    • 2021
  • Large earthquakes with (MW > ~ 6) result in ground shaking, surface ruptures, and permanent deformation with displacement. The earthquakes would damage important facilities and infrastructure such as large industrial establishments, nuclear power plants, and waste disposal sites. In particular, earthquake ruptures associated with large earthquakes can affect geological and engineered barriers such as deep geological repositories that are used for storing hazardous radioactive wastes. Earthquake-driven faults and surface ruptures exhibit various fault zone structural characteristics such as direction of earthquake propagation and rupture and asymmetric displacement patterns. Therefore, estimating the respect distances and hazardous areas has been challenging. We propose that considering multiple parameters, such as fault types, distribution, scale, activity, linkage patterns, damage zones, and respect distances, enable accurate identification of the sites for deep geological repositories and important facilities. This information would enable earthquake hazard assessment and lower earthquake-resulted hazards in potential earthquake-prone areas.

Investigation of the various properties of several candidate additives as buffer materials

  • Gi-Jun Lee;Seok Yoon;Taehyun Kim;Seeun Chang
    • Nuclear Engineering and Technology
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    • v.55 no.3
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    • pp.1191-1198
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    • 2023
  • Bentonite buffer material is a critical component in an engineered barrier system (EBS) for disposing high-level radioactive waste (HLW). The bentonite buffer material protects the disposal canister from groundwater penetration and releases decay heat to the surrounding rock mass; thus, it should possess high thermal conductivity, low hydraulic conductivity, and moderate swelling pressure to safely dispose the HLWs. Bentonite clay is a suitable buffer material because it satisfies the safety criteria. Several additives have been suggested as mixtures with bentonite to increase the thermal-hydraulic-mechanical-chemical (THMC) properties of bentonite buffer materials. Therefore, this study investigated the geotechnical, mineralogical, and THMC properties of several candidate additives such as sand, graphite, granite, and SiC powders. Datasets obtained in this study can be used to select adequate additives to improve the THMC properties of the buffer material.

Glass Dissolution Rates From MCC-1 and Flow-Through Tests

  • Jeong, Seung-Young
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2004.06a
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    • pp.257-258
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    • 2004
  • The dose from radionuclides released from high-level radioactive waste (HLW) glasses as they corrode must be taken into account when assessing the performance of a disposal system. In the performance assessment (PA) calculations conducted for the proposed Yucca Mountain, Nevada, disposal system, the release of radionuclides is conservatively assumed to occur at the same rate the glass matrix dissolves. A simple model was developed to calculate the glass dissolution rate of HLW glasses in these PA calculations [1]. For the PA calculations that were conducted for Site Recommendation, it was necessary to identify ranges of parameter values that bounded the dissolution rates of the wide range of HLW glass compositions that will be disposed. The values and ranges of the model parameters for the pH and temperature dependencies were extracted from the results of SPFT, static leach tests, and Soxhlet tests available in the literature. Static leach tests were conducted with a range of glass compositions to measure values for the glass composition parameter. The glass dissolution rate depends on temperature, pH, and the compositions of the glass and solution, The dissolution rate is calculated using Eq. 1: $rate{\;}={\;}k_{o}10^{(ph){\eta})}{\cdot}e^{(-Ea/RT)}{\cdot}(1-Q/K){\;}+{\;}k_{long}$ where $k_{0},\;{\eta}$ and Eaare the parameters for glass composition, pH, $\eta$ and temperature dependence, respectively, and R is the gas constant. The term (1-Q/K) is the affinity term, where Q is the ion activity product of the solution and K is the pseudo-equilibrium constant for the glass. Values of the parameters $k_{0},\;{\eta}\;and\;E_{a}$ are the parameters for glass composition, pH, and temperature dependence, respectively, and R is the gas constant. The term (1-Q/C) is the affinity term, where Q is the ion activity product of the solution and K is the pseudo-equilibrium constant for the glass. Values of the parameters $k_0$, and Ea are determined under test conditions where the value of Q is maintained near zero, so that the value of the affinity term remains near 1. The dissolution rate under conditions in which the value of the affinity term is near 1 is referred to as the forward rate. This is the highest dissolution rate that can occur at a particular pH and temperature. The value of the parameter K is determined from experiments in which the value of the ion activity product approaches the value of K. This results in a decrease in the value of the affinity term and the dissolution rate. The highly dilute solutions required to measure the forward rate and extract values for $k_0$, $\eta$, and Ea can be maintained by conducting dynamic tests in which the test solution is removed from the reaction cell and replaced with fresh solution. In the single-pass flow-through (PFT) test method, this is done by continuously pumping the test solution through the reaction cell. Alternatively, static tests can be conducted with sufficient solution volume that the solution concentrations of dissolved glass components do not increase significantly during the test. Both the SPFT and static tests can ve conducted for a wide range of pH values and temperatures. Both static and SPFt tests have short-comings. the SPFT test requires analysis of several solutions (typically 6-10) at each of several flow rates to determine the glass dissolution rate at each pH and temperature. As will be shown, the rate measured in an SPFt test depends on the solution flow rate. The solutions in static tests will eventually become concentrated enough to affect the dissolution rate. In both the SPFt and static test methods. a compromise is required between the need to minimize the effects of dissolved components on the dissolution rate and the need to attain solution concentrations that are high enough to analyze. In the paper, we compare the results of static leach tests and SPFT tests conducted with simple 5-component glass to confirm the equivalence of SPFT tests and static tests conducted with pH buffer solutions. Tests were conducted over the range pH values that are most relevant for waste glass disssolution in a disposal system. The glass and temperature used in the tests were selected to allow direct comparison with SPFT tests conducted previously. The ability to measure parameter values with more than one test method and an understanding of how the rate measured in each test is affected by various test parameters provides added confidence to the measured values. The dissolution rate of a simple 5-component glass was measured at pH values of 6.2, 8.3, and 9.6 and $70^{\circ}C$ using static tests and single-pass flow-through (SPFT) tests. Similar rates were measured with the two methods. However, the measured rates are about 10X higher than the rates measured previously for a glass having the same composition using an SPFT test method. Differences are attributed to effects of the solution flow rate on the glass dissolution reate and how the specific surface area of crushed glass is estimated. This comparison indicates the need to standardize the SPFT test procedure.

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Coupled T-H-M Processes Calculations in KENTEX Facility Used for Validation Test of a HLW Disposal System (고준위 방사성 폐기물 처분 시스템 실증 실험용 KENTEX 장치에서의 열-수리-역학 연동현상 해석)

  • Park Jeong-Hwa;Lee Jae-Owan;Kwon Sang-Ki;Cho Won-Jin
    • Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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    • v.4 no.2
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    • pp.117-131
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    • 2006
  • A coupled T-H-M(Thermo-Hydro-Mechanical) analysis was carried out for KENTEX (KAERI Engineering-scale T-H-M Experiment for Engineered Barrier System), which is a facility for validating the coupled T-H-M behavior in the engineered barrier system of the Korean reference HLW(high-level waste) disposal system. The changes of temperature, water saturation, and stress were estimated based on the coupled T-H-M analysis, and the influence of the types of mechanical constitutive material laws was investigated by using elastic model, poroelastic model, and poroelastic-plastic model. The analysis was done using ABAQUS, which is a commercial finite element code for general purposes. From the analysis, it was observed that the temperature in the bentonite increased sharply for a couple of days after heating the heater and then slowly increased to a constant value. The temperatures at all locations were nearly at a steady state after about 37.5 days. In the steady state, the temperature was maintained at $90^{\circ}C$ at the interface between the heater and the bentonite and at about $70^{\circ}C$ at the interface between the bentonite and the confining cylinder. The variation of the water saturation with time in bentonite was almost same independent of the material laws used in the coupled T-H-M processes. By comparing the saturation change of T-H-M and that of H-M(Hydro-Mechanical) processes using elastic and poroelastic material mod31 respectively, it was found that the degree of saturation near the heater from T-H-M calculation was higher than that from the coupled H-M calculation mainly because of the thermal flux, which seemed to speed up the saturation. The stresses in three cases with different material laws were increased with time. By comparing the stress change in H-M calculation using poroelasetic and poroelasetic-plastic model, it was possible to conclude that the influence of saturation on the stress change is higher than the influence of temperature. It is, therefore, recommended to use a material law, which can model the elastic-plastic behavior of buffer, since the coupled T-H-M processes in buffer is affected by the variation of void ratio, thermal expansion, as well as swelling pressure.

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