• Journal of Radiation Protection and Research
• /
• v.36 no.4
• /
• pp.174-182
• /
• 2011

With the reactor operation conditions - 4.3 wt% $^{235}U$ initial enrichment, burn-up 55,000 MWd/MTU, average power 34 MW/MTU for three periods burned time for 539.2 days per period and cooling time for 100 hours after shut down, to set up the condition to determine the minimum height (depth) of spent fuel storage pool to shut off the radiation out of the spent fuel storage pool and to store spent fuels safely, the dose rate on the specific position directed to the surface of spent fuel storage pool induced by the neutron and gamma-ray from spent fuels are evaluated. The length of spent fuel is 381 cm, and as the result of evaluation on each position from the top of spent fuel to the surface of spent fuel storage pool, it is difficult for neutrons from spent fuels to pass through the water layer of maximum 219 cm (600 cm from the floor of spent fuel storage pool) and 419 cm (800 cm from the floor of spent fuel storage pool) for gamma-ray. Therefore, neutron and gamma-ray from spent fuels can pass through below 419 cm (800 cm from the floor) water layer directed to the surface of spent fuel storage pool.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.19 no.2
• /
• pp.215-223
• /
• 2021

The domestic CANDU nuclear power plants have been operated for a long time and various unforeseen spent fuel defects have been discovered. As the spent fuel defects are important factors in the safety of the nuclear power plant, a study on the analysis of the spent fuel defects to prevent their recurrence is necessary. However, in cases where the fuel rods inside the fuel assembly are defected, it is difficult to dismantle the fuel assembly owing to their welded structure and the facility conditions of the plant. Therefore, it is impossible to analyze the spent fuel defect because it is difficult to visually check the shape of the fuel defect. To resolve these problems, an analysis technology that can predict the number of defected fuel rods and defect size was developed. In this study, we developed a methodology for investigating the root cause of spent fuel defects using a database of the earlier fuel defects in the plants. It is anticipated that in the future this analysis technology will be applied when spent fuel defects occur.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2062-2066
• /
• 2022

Nuclear energy sources are indispensable in cost effectively achieving carbon neutral economy, where public opinion is critical to adoption as the consequences of nuclear accident can be catastrophic. In this context, discussion on spent nuclear fuel is a prerequisite to expanding nuclear energy, as it leads to the issue of radioactive waste disposal. Given the dearth of study on spent nuclear fuel public acceptance, we use text mining and big data analysis on the news article and public comments data on Naver news portal to identify the Korean public opinion on spent nuclear fuel. We identify that the Korean public is more interested in the nuclear energy policy than spent nuclear fuel itself and that the alternative energy sources affect the position towards spent nuclear fuel. We recommend relating spent nuclear fuel issue with nuclear energy policy and environmental issues of alternative energy sources to further promote spent nuclear fuel.

• 제어로봇시스템학회:학술대회논문집
• /
• 2000.10a
• /
• pp.238-238
• /
• 2000

In this paper, to increase the utilization of uranium resources contained in the spent fuel, the spent fuel is reused. for this, the spent fuel is dismantled or spent fuel rod is extracted from the spent fuel assembly. Therefore, to achieve the performance of compacting the spent fuel assembly, we proposed the controller consisting of adaptive and fuzzy with teaming algorithm. In order to show the performance of proposed algorithm compares, we compared the controller with conventional controller in plant.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 1997.10a
• /
• pp.630-633
• /
• 1997

Spent fuel decladding device and dry voloxidizer is to separate the spent pellet from spent fuel rod cut by 250mm and to convert the spent pellet into powder form for reuse and/or disposal of the spent fuel. There are two methods in decladding and voloxidation of spent fuel, that is, wet method with chemical material and dry method with mechanical device. In this study, to examine the fuel rod decladding process and the pellet voloxidation process, the devices for the spent fuel decladding and the pellet voloxidation with dry method are developed. The decladding machine is designed to separate pellets from fuel rod by slitting device. And, the voloxidizer is designed to convert the spent pellet which is ceramic form into powder form by oxidation using the multi step mesh, vibrator, and air in the high temperature environment. The result of this study, such as operation condition et., will be utilized in the design of the machine for demonstration.

• Nuclear Engineering and Technology
• /
• v.31 no.6
• /
• pp.23-38
• /
• 1999

This study addresses the reference spent fuel and its characteristics for developing a geological repository concept. As a disposal capacity of the reference repository system to be developed, spent fuel inventories were projected based on the basis of the Nuclear Energy Plan of the Long-term National Power Program. The reference spent fuel encompassing a variability in characteristics of all existing and future spent fuels of interest was defined. Key parameters in the reference fuel screening processes were the nuclear and mechanical design parameters and the burnup histories for existing spent fuels as of 1996 and for future spent fuels with the more extended burnup the initial enrichment and its expected turnup. The selected reference fuel was characterized in terms of initial enrichment, bumup, dimension, gross weight and age. Also the isotopic composition and the radiological properties are quantitatively identified. This information provided in this study could be used as input for repository system development and performance assessment and applied in fuel material balance evaluation for the various types of back-end fuel cycle studies.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2017.10a
• /
• pp.301-302
• /
• 2017

• Nuclear Engineering and Technology
• /
• v.45 no.1
• /
• pp.29-40
• /
• 2013

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.

• Journal of the Korean Society of Systems Engineering
• /
• v.17 no.2
• /
• pp.82-90
• /
• 2021

The spent fuel pool (SFP) of a nuclear power plant functions to store the spent fuel. The spent fuel pool is designed to properly remove the decay heat generated from the spent fuel. If the cooling function is lost and proper operator action is not taken, the spent fuel in the storage pool can be damaged. Probabilistic safety assessment (PSA) is a safety evaluation method that can evaluate the risk of a large and complex system. So far, the probabilistic safety assessment of nuclear power plants has been mainly performed on the reactor. This study defined the requirements and the functional architecture for the probabilistic safety assessment of the spent fuel pool (SFP-PSA) by applying the systems engineering process. And, a systematic and efficient methodology was defined according to the architecture.

• 제어로봇시스템학회:학술대회논문집
• /
• 2005.06a
• /
• pp.2407-2409
• /
• 2005

As the cumulative amount of spent fuel increases, the reliable and effective management of the spent fuel has become a world-wide mission. For this mission, KAERI is developing the Advanced Spent Fuel Conditioning Process (ACP) as a pre-disposal treatment process for spent fuel. Conventional approach to the development of the process and the remote operation technology is to fabricate the process equipment on the same scale as the real environment and demonstrate the remote handling operation using simulated fuel called a mock-up test. But this mock-up test is expensive and time consuming, since the design may need to be modified and the equipment fabricated again to account for the problems found during a testing. To deal with this problem, we developed a digital mockup for the ACP. Also, for an effective utilization of the digital mockup, we developed user interface modules such as the data acquisition and display module and the external input device interface module. The result of this implementation shows that a continuous motion of the manipulator using the external device interface can be represented easily and the information display screens responded well to the simulation situation.