Heui-Joo Choi

Concept of a Korean Reference Disposal System for Spent Fuels

A deep geologic disposal system for the spent fuels from nuclear power plants has been developed since this program was launched in 1997 in Korea. In this paper, the concept of a Korean reference high-level waste (HLW) vertical disposal system (KRS-V1) is described. Though no site for the underground repository has yet been specified in Korea, a generic site with a granitic rock is considered for a reference spent fuel repository design. The depth of the repository is assumed to be 500 m. The repository consists of a disposal area, a controlled area, and an uncontrolled area. The disposal area consists of disposal tunnels, panel tunnels, and a central tunnel. In the controlled area and the uncontrolled area, there are technical rooms and tunnels and/or shafts to connect them to the ground level, respectively. The repository will be excavated, operated, and backfilled in several phases including an underground research laboratory (URL) phase. The result of this concept development will be used for an evaluation of its feasibility, analyses of its long-term safety, information for public communication, and a cost estimation, among others.

Sorption of cesium and iodide ions onto KENTEX-bentonite

The sorption of cesium and iodide ions onto KENTEX-bentonite was investigated using batch test and in-diffusion test methods. The cesium ions were highly sorbed on the bentonite, and the experimental data fit the Freundlich isotherm well. The distribution coefficient, Kd, of the cesium ions was variably affected by the chemical conditions of the solution (initial ion concentration, pH, salinity) and temperature. An increasing pH of solution increased the Kd. However, there were different Kd values that decrease with an increase in the initial ion concentration, salinity, and temperature. The iodide ions, on the contrary, were negligibly sorptive. The Kd values obtained from the in-diffusion tests were quite lower than those from the batch tests, which could be explained by changes in the pore water chemistry and surface area available for sorption.

DEVELOPMENT OF GEOLOGICAL DISPOSAL SYSTEMS FOR SPENT FUELS AND HIGH-LEVEL RADIOACTIVE WASTES IN KOREA

Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

WASTE CLASSIFICATION OF 17 x 17 KOFA SPENT FUEL ASSEMBLY HARDWARE

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17 x 17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGENS module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and 90Sr, respectively. Finally, it was found that 88.7% of the metal waste from the 17 x 17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

RADIOLOGICAL CHARACTERISTICS OF DECOMMISSIONING WASTE FROM A CANDU REACTOR

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be 1.04 10 16 Bq, 2.09 10³ W, 5.31 10 14 ㎥-water, 4.69 10? ㎏, and 7.38 10¹㎥, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

PYROPROCESS WASTE DISPOSAL SYSTEM DESIGN AND DOSE CALCULATION

PWR spent fuels produced in the Republic of Korea are expected to be recycled by pyroprocess in the long term future. Even though pyroprocess waste amounts can be smaller than that of PWR spent fuel assembly in case of direct disposal, this process essentially will produce various and unique radioactive wastes. The goals of this article are to characterize these wastes, calculate the amount of wastes, design disposal systems for each waste and evaluate the radiation safety of each system by dose assessment. The absorbed dose results of the metal and ceramic waste for the engineering barrier system (EBS) showed 2.21 x 10 -2 Gy/h and 1.15 x 10 -2 Gy/h, which are lower than the recommended value of 1 Gy/h. These results confirmed that the newly proposed disposal systems have a safety margin for the radiation produced from each waste.

Thickness Measurement by Using Cepstrum Ultrasonic Signal Processing

Ultrasonic thickness measurement is a non-destructive method to measure the local thickness of a solid element, based on the time taken for an ultrasound wave to return to the surface. When an element is very thin, it is difficult to measure thickness with the conventional ultrasonic thickness method. This is because the method measures the time delay by using the peak of a pulse, and the pulses overlap. To solve this problem, we propose a method for measuring thickness by using the power cepstrum and the minimum variance cepstrum. Because the cepstrums processing can divides the ultrasound into an impulse train and transfer function, where the period of the impulse train is the traversal time, the thickness can be measured exactly. To verify the proposed method, we performed experiments with steel and, acrylic plates of variable thickness. The conventional method is not able to estimate the thickness, because of the overlapping pulses. However, the cepstrum ultrasonic signal processing that divides a pulse into an impulse and a transfer function can measure the thickness exactly.

Implementation of a virtual environment for development of a spent fuel disposal process

Since the program for the development of a deep geological disposal system for the spent fuel from nuclear power plants was launched in 1997 as a national program in KOREA, a pre-conceptual design of a disposal system for spent fuel in a deep geological host rock formation was carried out and the Korean reference spent fuel vertical disposal system (KRS-VI) is concurrently being developed. In this study, a virtual environment system for developing a spent fuel disposal system and reviewing and analyzing the process for an optimization was implemented. This system consisted of a 3-D graphic simulation module and an analysis module, and was developed in three stages. To develop this system, the design bases like the fuel canisters, the production rate and the functions of the facility, and the detailed processes were reviewed. With the graphic simulation module, the disposal processes were simulated in a virtual work cell according to a process scenario. Also, with the analyses module, the spent fuel disposal processes were preliminarily analyzed. It is necessary to verify this system more specifically and analyze the disposal process in detail. This virtual system can be effectively used for developing a process and the process equipment, as well as optimizing the process for a spent fuel disposal in a deep geological host rock.

R&D Review on the Gap Fill of an Engineered Barrier for an HLW Repository

Abstract I n a high-level waste repository, the gap fill of the engineered barrier is an important component that influences the performance of the buffer and backfill. This paper reviewed the overseas status of RD blowing through the use of shotcrete technology and auger placement and compaction techniques have been used in the gap of horizontal deposition hole and tunnel. However, these emplacement techniques are still technically at the beginning stage, and thus additional research and development are expected to be needed.

A Study on the Conceptual Development for a Deep Geological Disposal of the Radioactive Waste from Pyro-processing

A long-term R&D program for HLW disposal technology development was launched in 1997 in Korea and Korea Reference disposal System(KRS) for spent fuels had been developed. After then, a recycling process for PWR spent fuels to get the reusable material such as uranium or TRU and to reduce the volume of radioactive waste, called Pyro-process, is being developed. This Pyro-process produces several kinds of wastes including metal waste and ceramic waste. In this study, the characteristics of the waste from Pyro-process and the concepts of a disposal container for the wastes were described. Based on these concepts, thermal analyses were carried out to determine a layout of the disposal area of the ceramic wastes which was classified as a high level waste and to develop the disposal system called A-KRS. The location of the final repository for A-KRS is not determined yet, thus to review the potential repository domains, the possible layout in the geological characteristics of KURT facility site was proposed. These results will be used in developing a repository system design and in performing the safety assessment.