• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.9-15
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• 2006

For the accurate safety assessment of potential radioactive waste disposal site which is located in the crystalline rock it is important to simulate the mass transportation through engineered and natural barrier system precisely, characterized by porous and fractured media respectively. In this work the methods to construct discrete fracture network for the analysis of flow and mass transport through fractured-porous medium are described. The probability density function is adopted in generating fracture properties for the realistic representation of real fractured rock. In order to investigate the intersection between a porous and a fractured medium described by a 2 dimensional rectangular and a cuboid grid respectively, an additional imaginary fracture is adopted at the face of a porous medium intersected by a fracture. In order to construct large scale flow paths an effective method to find interconnected fractures and algorithms of swift detecting connectivities between fractures or porous medium and fractures are proposed. These methods are expected to contribute to the development of numerical program for the simulation of radioactive nuclide transport through fractured-porous medium from radioactive waste disposal site.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.165-175
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• 2005

Measurement of spent resin activity was initiated in 2004 in order to develop the C-14 removal technology for safe disposal. As part of this program, spent resins were sampled and measured in the in-station resin storage tank 2 at Wolsong Nuclear Power Plant Unit 1. At the time of sampling, the resins had been in storage tank from 3 to 23 years. Total 72 resin samples were sampled, which were collected from both man-hole (68 samples) and test-hole (4 samples) in the in-station resin storage tank 2. They were separated into liquid, activated carbon, zeolite, and spent resin. The spent resins were oxidized with sample oxidizer and analyzed for C-14. Ten of collected mixed resin samples were separated by density into cation and anion resins using a sugar solution. The C-14 concentration in anion exchange resin was approximately 2 times higher than in the mixed resin. The average concentration of C-14 in the cation/anion mixed exchange resin was $460\;GBq/m^3$ from test-hole and $53.1\;GBq/m^3$ from man-hole. We have found that concentration of C-14 in the spent resin is about from 0.4 to $1,321\;GBq/m^3$. So it could be a problem, when dispose of at a repository, since there is a disposal limit of $222\;GBq/m^3$. This means we should develop the C-14 removal technology.

• Journal of Radiation Protection and Research
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• v.45 no.4
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• pp.178-186
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• 2020

Background: Reactor pressure vessel (RV) with internals (RVI) are activated structures by neutron irradiation and volume contaminated wastes. Thus, to develop safe and optimized disposal plan for them at a disposal site, it is important to perform exact activation calculation and evaluate the dose rate on the surface of casks which contain cut-off pieces. Materials and Methods: RV and RVI are subjected to neutron activation calculation via Monte Carlo methodology with MCNP6 and ORIGEN-S program-neutron flux, isotopic specific activity, and gamma spectrum calculation on each component of RV and RVI, and dose rate evaluation with MCNP6. Results and Discussion: Through neutron activation analysis, dose rate is evaluated for the casks containing cut-off pieces produced from decommissioned RV and RVI. For RV cut-off ones, the highest value of dose rate on the surface of cask is 6.97 × 10-1 mSv/hr and 2 m from it is 3.03 × 10-2 mSv/hr. For RVI cut-off ones, on the surface of it is 0.166 × 10-1 mSv/hr and 2 m from it is 1.04 × 10-1 mSv/hr. Dose rates for various RV and RVI cut-off pieces distributed lower than the limit except the one of 2 m from the cask surface of RVI. It needs to adjust contents in cask which carries highly radioactive components in order to decrease thickness of cask. Conclusion: Two types of casks are considered in this paper: box type for very-low-level waste (VLLW) as well as low-level waste (LLW) and cylinder type for intermediate-level waste (ILW). The results will contribute to the development of optimal loading plans for RV and RVI cut-off pieces during the decommissioning of nuclear power plant that can be used to prepare radioactive waste disposal plans for the different types of wastes-ILW, LLW, and VLLW.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.531-541
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• 2023

The Wolsong Nuclear Power Plant (NPP) operates an on-site spent fuel dry storage facility using concrete silo and vertical module systems. This facility must be safely maintained until the spent nuclear fuel (SNF) is transferred to an external interim or final disposal facility, aligning with national policies on spent nuclear fuel management. The concrete silo system, operational since 1992, requires an aging management review for its long-term operation and potential license renewal. This involves comparing aging management programs of different dry storage systems against the U.S. NRC's guidelines for license renewal of spent nuclear fuel dry storage facilities and the U.S. DOE's program for long-term storage. Based on this comparison, a specific aging management program for the silo system was developed. Furthermore, the facility's current practices-periodic checks of surface dose rate, contamination, weld integrity, leakage, surface and groundwater, cumulative dose, and concrete structure-were evaluated for their suitability in managing the silo system's aging. Based on this review, several improvements were proposed.

• Journal of Radiation Protection and Research
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• v.16 no.1
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• pp.43-52
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• 1991

A comparison study is performed on dose assessment programs including DOSE, LIMCAL, and PABLM, DOSE is a program developed for preliminary safety assessments of the low- and intermediate- level radioactive waste disposal, and the others are existing programs applicable to similar calculations. The results show acceptable agreement within an order of magnitude(mrem/yr) except for C-14 and Pu-239. PABLM results higher dose for C-14 while lower value for Pu-239 in comparison with those from DOSE or LIMCAL. It is found that the discrepancy in C-14 is due to difference in transport model introduced and that in Pu-239 is from the different value of dose conversion factor to each program.

• Explosives and Blasting
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• v.25 no.1
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• pp.15-29
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• 2007

In situ and laboratory tests were carried out for investigating the Excavation Disturbed Zone(EDZ) generated from blasting at the KAERI Underground Research Tunnel(KURT), which is for the researches related to High-level radioactive waste disposal program. It was found that the EDZ was generated more than In from the laboratory rock tests and in situ experiments. RQD of the rock mass within $0{\sim}2(m)$, where the blasting impact was significant, was 17% lower than in the deeper zones without a serious blasting impact. It was also estimated that the deformability of rock mass was reduced about 40% after the blasting.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.4
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• pp.263-269
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• 2004

During the last two years, Site Information and Total Environmental database management System (SITES) ver. 1.0 has been developed for the systematic SITES Database Module (SDM), which includes site information, facility information and environmental information. The SITES includes the module for site environmental monitoring system and safety assessment (M&A) system for the nuclear facility. SITES is expected to be an effective system for the radioactive waste disposal management facility. Currently, SITES ver.2.0 is under development after the SITES ver.1.0 that is focused on the M&A system. The main purpose of this paper is to introduce and try to account for the major development in the concept of SEMS sub-module of the M&A module. The SEMS is purposed of development of the program for real time environmental monitoring, prediction, and automatic alarm system using SITES Database and related information.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1994.10a
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• pp.201-208
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• 1994

To secure long-term structural safety of underground openings for radioactive waste disposal, the proper structural safety analyses are required. Especially, the structural analysis for underground openings should consider the effects of groundwater pressure. The objective of this study is to develop the structural analysis method for saturated rock masses. In this study, the interaction between groundwater distribution and structural behavior of rock masses are carried out to develop the structural analysis method of saturated rock masses. Then, a 3-Dimensional Multi-Phase Dynamic Analysis Program (MPDAP-3D) has been developed by modifying the existing MPDAP which is based on the concept of 2-dimensional two-phase media.

• Computational Structural Engineering
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• v.4 no.4
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• pp.117-124
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• 1991

In order to predict properly the effects of ground motion associated with earthquakes on underground radioactive waste disposal facilities, an understanding of the structural behavior of an underground opening in discontinuous rock masses subjected to dynamic loadings is essential. This paper includes literature review on computational models for discontinuous rock masses and on mathematical models for the structural analysis of underground opening. Then, structural analyses of underground openings using the distinct element computer program written for the static and dynamic analysis of discontinuous rock masses have been performed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.63-74
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• 2020

The operational safety assessment is an important part of a safety case for the deep geological repository of spent fuels. It consists of different stages such as the identification of initiating events, event tree analysis, fault tree analysis, and evaluation of exposure doses to the public and radiation workers. This study develops a probabilistic safety assessment method for the operational safety assessment and establishes an assessment framework. For the event and fault tree analyses, we propose the advanced information management system for probabilistic safety assessment (AIMS-PSA Manager). In addition, we propose the Radiological Safety Analysis Computer (RSAC) program to evaluate exposure doses to the public and radiation workers. Furthermore, we check the applicability of the assessment framework with respect to drop accidents of a spent fuel assembly arising out of crane failure, at the surface facility of the KRS⁺ (KAERI Reference disposal System for SNFs). The methods and tools established through this study can be used for the development of a safety case for the KRS⁺ system as well as for the design modification and the operational safety assessment of the KRS⁺ system.