• 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 Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.91-101
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• 2007

In this study, domestic regulatory requirement was investigated for self-disposal of concrete waste from nuclear fuel processing facility. And after self-disposal as landfill or recycling/reuse, the exposure dose was evaluated by RESRAD Ver. 6.3 and RESRAD BUILD Ver.3.3 computing code for radiological assessments of the general public. Derived clearance level by the result of assessments for the exposure dose of the general public is 0.1071Bq/g (3.5% enriched uranium) for landfill and $0.05515Bq/cm^2$ (5% enriched uranium) for recycling/reuse respectively. Also, residual radioactivity of concrete waste after decontamination was investigated in this study. The result of surface activity is $0.01Bq/cm^2\;for\;{\alpha}-emitter$ and the result of radionuclide analysis for taken concrete samples from surface of concrete waste is 0.0297Bq/g for concentration of $^{238}U$, below 2w/o for enrichment of $^{235}U$ and 0.0089Bq/g for artificial contamination of $^{238}U$ respectively. Therefore, radiological hazard of concrete waste by self-disposal as landfill and recycling/reuse is below clearance level to comply with clearance criterion provided for Notice No.2001-30 of the MOST and Korea Atomic Energy Act.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.243-253
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• 2009

In this study, research status on radionuclide and colloid migration in underground research facilities including KURT (KAERI Underground Research Tunnel) was investigated. Some foreign underground research facilities constructed in crystalline rock formations such as granite were briefly introduced and compared. International joint researches concerned with the radionuclide and colloid migration were investigated particularly for the Grimsel Test Site (GTS) and $\ddot{A}$sp$\ddot{o}$ Hard Rock Laboratory by analyzing major research items, on-going research projects, and future plans.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.3
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• pp.217-233
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• 2014

Securing the radiological safety is a prerequisite for the safe management of the naturally occurring radioactive materials (NORM) which cannot be reused. This becomes a crucial focus of our R&D efforts upon the implementation of the Act on Protective Action Guidelines against Radiation in the Natural Environment. To secure the safety, the establishment of technical bases and procedures for securing radiological safety related to the disposal of NORM is required. Thus, it is necessary to analyze the characteristics, to collect the data, to have the radiological safety assessment methodologies and tools, to investigate disposal methods and facilities, and to study the effects of the input data on the safety for the NORM wastes. Here, we assess the environmental impact of the NORM waste disposal with respect to the major domestic and foreign NORM characteristics. The data associated with major industries are collected/analyzed and the status of disposal facilities and methodologies relevant to the NORM wastes is investigated. We also suggest the conceptual design concept of a landfill disposal facility and the management plan with respect to the major NORM wastes characteristics. The radionuclide pathways are identified for the atmospheric transport and leachate release and the environmental impact assessment methodology for the NORM waste disposal is established using a relevant code. The assessment and analysis on the exposure doses and excessive cancer risks for the NORM waste disposal are performed using the characteristics of the representative domestic NORM wastes including flying ash, phosphor gypsum, and redmud. The results show that the exposure dose and the excessive cancer risks are very low to consider any radiation effects. This study will contribute to development in the areas of the regulatory technology for securing radiological safety relevant to NORM waste disposal and to the implementation technology for the Act.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.183-193
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• 2018

Currently, the Korea Atomic Energy Research Institute (KAERI) is planning to build the Ki-Jang Research Reactor (KJRR) in Ki-Jang, Busan. It is important to safely dispose of low-level radioactive waste from the operation of the reactor. The most efficient way to treat radioactive waste is cement solidification. For a radioactive waste disposal facility, cement solidification is performed based on specific waste acceptance criteria such as compressive strength, free-standing water, immersion and leaching tests. Above all, the leaching test is important to final disposal. The leakage of radioactive waste such as $^{137}Cs$ causes not only regional problems but also serious global ones. The cement solidification method is simple, and cheaper than other solidification methods, but has a lower leaching resistance. Thus, this study was focused on the development of cement solidification for an enhancement of cesium leaching resistance. We used Zeolite and Loess to improve the cesium leaching resistance of KJRR cement solidification containing simulated KJRR liquid waste. Based on an SEM-EDS spectrum analysis, we confirmed that Zeolite and Loess successfully isolated KJRR cement solidification. A leaching test was carried out according to the ANS 16.1 test method. The ANS 16.1 test is performed to analyze cesium ion concentration in leachate of KJRR cement for 90 days. Thus, a leaching test was carried out using simulated KJRR liquid waste containing $3000mg{\cdot}L^{-1}$ of cesium for 90 days. KJRR cement solidification with Zeolite and Loess led to cesium leaching resistance values that were 27.90% and 21.08% higher than the control values. In addition, in several tests such as free-standing water, compressive strength, immersion, and leaching tests, all KJRR cement solidification met the waste acceptance or satisfied the waste acceptance criteria for final disposal.

• Journal of the Korean Society of Groundwater Environment
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• v.6 no.3
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• pp.111-119
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• 1999

The purpose of this study is the simulation of discontinuous rockmass and identification of characteristics of discontinuity network as a branch of the study on characteristics of groundwater system in discontinuous rockmass for evaluation of safety on disposal site of radioactive waste. In this study the site for LPG underground storage was selected for the similarities of the conditions which were required for disposal site of radioactive waste. Through the identification of hydraulic properties. characteristics of discontinuities and selection of discontinuity model around LPG underground storage facility. the applications of discrete fracture network model were evaluated for the analysis of pathway. The orientation and spatial density of discontinuities are primarily important elements for the simulation of groundwater and solute transportation in discrete fracture network model. In this study three fracture sets identified and the spatial intensity (P$_{32}$) of discontinuities is revealed as 0.85 $m^2$/㎥. The conductive fracture intensity (P$_{32c}$) estimated for the simulation area around propane cavern (200${\times}$200${\times}$200) is 0.536 $m^2$/㎥. Truncated conductive fracture intensity (T-P$_{32c}$) is calculated as 0.26 $m^2$/㎥ by eliminating the fracture with the iowest transmissivity and based on this value the pathway from the water curtain to PC 2. PC 3 analyzed.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1368-1375
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• 2016

A radiological safety assessment study was performed for the transportation of low level radioactive wastes which are temporarily stored in Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea. We considered two kinds of wastes: (1) operation wastes generated from the routine operation of facilities; and (2) decommissioning wastes generated from the decommissioning of a research reactor in KAERI. The important part of the radiological safety assessment is related to the exposure dose assessment for the incidentfree (normal) transportation of wastes, i.e., the radiation exposure of transport personnel, radiation workers for loading and unloading of radioactive waste drums, and the general public. The effective doses were estimated based on the detailed information on the transportation plan and on the radiological characteristics of waste packages. We also estimated radiological risks and the effective doses for the general public resulting from accidents such as an impact and a fire caused by the impact during the transportation. According to the results, the effective doses for transport personnel, radiation workers, and the general public are far below the regulatory limits. Therefore, we can secure safety from the viewpoint of radiological safety for all situations during the transportation of radioactive wastes which have been stored temporarily in KAERI.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.189-203
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• 2008

The physical and chemical properties of radioactive waste drums, which have been temporarily stored on site, should be characterized before their shipment to a disposal facility in order to prove that the properties meet the acceptance guideline. The investigation of NDT(Nondestructive Test) method was figured out that the contents in drum, the quantitative analysis of free standing water and void fraction can be examined with X-ray NDT techniques. This paper describes the characteristics of X-ray NDT such as its principles, the considerations for selection of X-ray system, etc. And then, the waste drum characteristics such as drum type and dimension, contents in drum, etc. were examined, which are necessary to estimate the optimal X-ray energy for NDT of a drum. The estimation results were that: $(R)\acute{A}$ the proper X-ray energy is under 3 MeV to test the drums of 320 ${\beta}\S$ and less; $(R)\ddot{E}$ both X-ray systems of 450 keV and/or 3 MeV might be needed considering the economical efficiency and the realization. The number of drums that can be tested with 450 keV and 3 MeV X-ray system was figured out as 42,327 and 18,105 drums (based on storage of 2006. 12), respectively. Four testing scenarios were derived considering equipment procurement method, outsourcing or not, etc. The economical and feasibility assessment for the scenarios was resulted in that an optimal scenario is dependent on the acceptance guide line, the waste generator's policy on the waste treatment and the delivery to a disposal facility, etc. For example, it might be desirable that a waste generator purchases two 450 keV mobile system to examine the drums containing low density waste, and that outsourcing examination for the high density drums, if all NDT items such as quantitative analysis for 'free standing water' and 'void fraction', and confirmation of contents in drum have to be characterized. However, one 450 keV mobile system seems to be required to test only the contents in 13,000 drums per year.

• The Journal of Engineering Geology
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• v.27 no.4
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• pp.463-474
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• 2017

Heterogeneous background fractures of granite and sedimentary rocks in Gyeongju LILW (low-intermediate level radioactive waste) facility area have been characterized quantitatively by analyzing fracture parameters (orientation, intensity, and size). Surface geological survey, electrical resistivity survey, and acoustic televiewer log data were used to characterize the heterogeneity of background fractures. Bootstrap method was applied to represent spatial anisotropy of variably oriented background fractures in the study area. As a result, the fracture intensity was correlated to the inverse distance from the faults weighted by nearest fault size and the mean value of electrical resistivity and the average volumetric fracture intensity ($P_{32}$) was estimated as $3.1m^2/m^3$. Size (or equivalent radius) of the background fractures ranged from 1.5 m to 86 m and followed to power-law distribution based on the fractal property of fracture size, using fractures measured on underground silos and identified surface faults.