[en] Analytic and numerical studies are made to predict space-time-dependent normalized concentrations of radionuclides migrating through the water-saturated fractured porous rock with limited diffusion. In this study, two extended migration models in a discrete fracture surrounded by porous rock matrix with limited diffusion are developed. Applicability and limitation of the models are investigated. First model is the radionuclide migration system through a single, planar fracture with limited thickness, a, of surrounding porous rock matrix. We impose the impermeable boundary condition at the distance from the fracture surface. For smaller values of a, the distance from the fracture and rock interface to the boundary of the impermeable zone, the normalized concentrations deviate from those by the solution without the impermeable zone. Second model, in general sense, includes the surface layer close to the fracture having different characteristics of matrix 1 from the intact rock matrix 2 based on the evident field studies. Although the full analytic solutions are developed for second model in zero dispersion case, numerical inversion of the Laplace Transform is employed to avoid the difficulties associated with the calculation of analytic solution. Higher physical space, which is explained by large porosity, for radionuclide to reside shows lower concentration in a fracture in water. The more radionuclide diffuses into the surrounding porous rock matrix 1, the less radionuclides remain in a fracture and move according to the assumed governing phenomena. As the retardation factors of surrounding porous rock matrix increase, the concentration front moves toward inlet boundary. The existence of difference characteristics of porous rock matrix accelerates the radionuclide migration depending on the thickness of porous rock matrix 1. In spite of the intrinsic problem of numerical inversion, we used the Talbot method to obtain the normalized concentration in a fracture in water. It is because that full analytic solutions is in very complex mathematical form and, thus, difficult to get the concentration numerically. When numerical difficulty such as floating point overflow in the inversion of Laplace solutions occurs, it take much time so that in this case it is more convenient to adopt the another method for computation. Therefore, for actual implementation, two approaches are to be selectively used depending on the combination of parameter value. A traditional full analytic approach for this two models is developed for the following reasons: (1) Even though the numerical inversion techniques are relatively easy, they sometimes create numerical problems in specific time domains and (2) The full analytic solution can be used as a bench-marking tool to verify the numerical inversion program. In the main text, the full analytic approaches are pursued. If the real field data accommodates the existence of impermeable zone in first model or two-layered porous zones, the overall safety assessment for the radioactive repository should reflect these models

[en] Low-level radioactive waste management require the knowledge of the natures and quantities of radionuclides in the immobilized or packaged waste. U. S. NRC rules require programs that measure the concentrations of all relevant nuclides either directly or indirectly by relating difficult-to-measure radionuclides to other easy-to-measure radionuclides with application of scaling factors. Scaling factors previously developed through statistical approach can give only generic ones and have many difficult problem about sampling procedures. Generic scaling factors can not take into account for plant operation history. In this study, a method to predict plant-specific and operational history dependent scaling factors is developed. Realistic and detailed approach are taken to find scaling factors at reactor coolant. This approach begin with fission product release mechanisms and fundamental release properties of fuel-source nuclide such as fission product and transuranic nuclide. Scaling factors at various waste streams are derived from the predicted reactor coolant scaling factors with the aid of radionuclide retention and build up model. This model make use of radioactive material balance within the radioactive waste processing systems. Scaling factors at reactor coolant and waste streams which can include the effects of plant operation history have been developed according to input parameters of plant operation history

[en] Enhancement of a computer code SAGE for evaluation of the Korean concept for a LILW waste disposal facility is discussed. Several features of source term analysis are embedded into SAGE to analyze: (1) effects of degradation mode of an engineered barrier, (2) effects of dispersion phenomena in the unsaturated zone and (3) effects of time dependent sorption coefficient in the unsaturated zone. IAEA's Vault Safety Case (VSC) approach is used to demonstrate the ability of this assessment code. Results of MASCOT are used for comparison purposes. These enhancements of the safety assessment code, SAGE, can contribute to realistic evaluation of the Korean concept of the LILW disposal project in the near future

[en] Engineered barrier test facility is specially designed to demonstrate the performance of engineered barrier system for the near-surface disposal facility under the domestic environmental conditions. Comprehensive measurement systems for the water content, temperature, matricpotential are installed within each test cell. In this study, short-term monitoring of the behavior of multi-layered cover system is implemented with artificial rainfall system. The periodic measurement data are collected and analyzed by a dedicated database management system, and provide a basis for performance verification of the disposal cover design

[en] In a previous paper, the authors have presented the effect of the limiting zone for the matrix diffusion of I-129 from an open fracture into a surrounding porous medium to reflect recent field findings from field studies. In addition to this matrix diffusion into finite porous rock matrix, this study analyzes the back diffusion of Np-237 from finite rock matrix into a fracture with band release condition. When the finite thickness of rock matrix is considered with band release, Np-237 are transported further along the fracture and the radionuclides stored during leaching time in surrounding rock matrix diffuse back into the fracture rapidly making peak diffusive flux

[en] This study presents a method to predict plant-specific and operational history dependent scaling factors. Realistic and detailed approaches are taken to find scaling factors at reactor coolant. This approach begins with fission product release mechanisms and fundamental release properties of fuel source nuclide such as fission product and transuranic nuclide. Scaling factors at various waste streams are derived from the predicted reactor coolant scaling factors with the aid of radionuclide retention and build up model. This model makes use of radioactive material balance within the radioactive waste processing systems. According to input parameters of plant operation history, scaling factors predicted at reactor coolant and waste streams are well brought out the effects of plant operation history

[en] To evaluate radiological impact from the operation of a low- and intermediate-level radioactive waste disposal facility, a logical presentation and explanation of expected accidental scenarios is essential to the stakeholders of the disposal facility. The logical assessment platform and procedure, including analysis of the safety function of disposal components, operational hazard analysis, operational risk analysis, and preparedness of remedial measures for operational safety, are improved in this study. In the operational risk analysis, both design measures and management measures are suggested to make it possible to connect among design, operation, and safety assessment within the same assessment platform. For the preparedness of logical assessment procedure, classifcation logic of an operational accident is suggested based on the probability of occurrence and consequences of assessment results. The improved assessment platform and procedure are applied to an operational accident analysis of the Korean low- and intermediate-level radioactive waste disposal facility and partly presented in this paper

[en] Matrix diffusion into a rock matrix has been regarded to retard radionuclide migration in a fracture. Recent field findings on a fractured system indicate that only a small portion of the rock in a fractured porous medium contributes to holding a radionuclide by matrix diffusion. To understand this effect, radionuclide migration in a fracture and diffusion from a finite rock matrix to a fracture are discussed with limited matrix diffusion under solubility-limited boundary conditions of a target radionuclide for the band-type release. Numerical inversion of the Laplace transform method is applied to estimate concentrations in a fracture and a finite rock matrix and fluxes at the fracture surface. Matrix diffusion into a finite rock matrix shows enhanced radionuclide migration and a higher concentration profile in a fracture. Diffusive flux from a finite rock matrix into a fracture after the end of leaching time shows higher peak values than flux from an infinite rock matrix because of (a) higher saturation of a radionuclide in a finite rock matrix and (b) increase of a radionuclide concentration in a fracture. Therefore, it is more realistic and conservative to apply the finite matrix diffusion for the overall assessment in a potential repository embedded in a fractured porous medium

[en] The suitable disposal plan for disused radioactive sealed sources should be required for their safe management. For assuring the safety of the long half lived spent sealed radioactive wastes on a borehole disposal facility, preliminary safety assessment was performed by SAGE(Safety Assessment of Groundwater Evaluation) code. Spent sealed radioactive sources such as Am-241, Ra-226 and C-14 are considered in safety assessment. Well water drinking scenario is used to calculate annual dose. Ra-226 results in higher annual dose than the other spent sealed sources in the far field. The total annual dose from the suggested borehole disposal system satisfied the regulated dose criteria

[en] To validate the previous conceptual design of cover system, construction of the engineered barrier test facility is completed and the performance tests of the disposal cover system are conducted. The disposal test facility is composed of the multi-purpose working space, the six test cells and the disposal information space for the PR center. The dedicated detection system measures the water content, the temperature, the matric potential of each cover layer and the accumulated water volume of lateral drainage. Short-term experiments on the disposal cover layer using the artificial rainfall system are implemented. The sand drainage layer shows the satisfactory performance as intended in the design stage. The artificial rainfall does not affect the temperature of cover layers. It is investigated that high water infiltration of the artificial rainfall changes the matric potential in each cover layer. This facility is expected to increase the public information about the national radioactive waste disposal program and the effort for the safety of the planned disposal facility.