• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.391-402
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• 2017

The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility's design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a $2{\times}10$ arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the $2{\times}10$ arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

• Proceedings of the Korean Society of Marine Engineers Conference
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• 2011.06a
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• pp.3-6
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• 2011

The utilization of gas as ship fuel requires a new set of regulations by IMO and society of classification. Maritime Safety Committee(MSC) and the subcommittee Bulk-Liquids and Gases(BLG) in IMO developed "Interim Guidelines on Safety for Natural Gas-fueled Engine Installation in Ships(Res.MSC.285(86))" for the use of natural gas in internal combustion engine. According to the requirement of Res.MSC.285(86) for natural gas-fueled engine installations in ships, several parts of ships should follow safety criteria in terms of Fuel bunkering, Gas safe Machinery spaces, Gas Fuel Storage and etc. In this thesis, details of the IGF code shall be described and development of the IGF code in IMO shall be illustrated.

• Journal of Korea Water Resources Association
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• v.53 no.4
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• pp.285-291
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• 2020

The Noah 3.3 Land Surface Model (LSM) was used to estimate the global soil moisture in this study and these soil moisture datasets were assessed against satellite-based and reanalysis soil moisture products. The Noah 3.3 LSM simulated soil moistures in four soil layers and root-zone soil moistures defined as a depth-weighted average in the first three soil layers (i.e., up to 1.0 m deep). The Noah LSM soil moisture products were then compared with a satellite-based soil moisture dataset (European Space Agency Climate Change Initiatives (ESA CCI) SM v04.4) and reanalysis soil moisture datasets (ERA-interim). In addition, the five major basins (Yangtze, Mekong, Mississippi, Murray-Darling, Amazon) were selected for the assesment with the Gravity Recovery and Climate Experiment (GRACE)-based Total Water Storage Anomaly (TWSA) and TWS Change (TWSC). The results revealed that high anomaly correlations were found in most of the Asia-Pacific regions including East Asia, South Asia, Australia, and Noth and South America. While the anomaly correlations in the Murray-Darling basin were somewhat low, relatively higher anomaly correlations in the other basins were found. It is concluded that this study can be useful for the development of soil moisture based drought indices and subsequently can be helpful to reduce damages from drought by timely providing an efficacious strategy.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.05a
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• pp.190-191
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• 2007

Cleanout of the F Reactor Fuel Storage Basin (FSB) is an element of the FSB decontamination and decommissioning (D&D) and is required to complete interim safe storage (ISS) of the F Reactor. Following reactor shutdown and in preparation for a deactivation layaway action in 1970, the water level in the FReactor FSB was reduced to approximately 0.6 m (2 ft) over t]to floor. Basin components and other miscellaneous items were left or placed in the FSB. The item placement was performed with a sense of finality, and no attempt was made to place the items in an orderly manner. The F Reactor FSB was then filled to grade level with 6(20of local surface material (essentially a fine sand). The reactor FSB backfill cleanout has the potential of having to remove spent nuclear fuel (SNF) that may have been left unintentionally. Based on previous cleanout of six water-filled FSBs with similar designs (i.e., the B, C, D, and DR FSBs in the 1980's), it was estimated that up to five SNF elements could be discovered in the F FSB (I). In reality about 17 full SNF elements were found in the excavation. This paper covers the technical and programmatic challenges of performing this decommissioning effort with some of the controls used for SNF management. The paper also will highlight how many various technologies were married into a complete package to address the issue at hand and show how no one tools could complete the job, but combined, good progress is being made.

• Journal of the Korean Society for Marine Environment & Energy
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• v.17 no.2
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• pp.153-165
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• 2014

Carbon dioxide capture and storage (CCS) technology is recognizing one of method responding the climate change with reduction of carbon dioxide in atmosphere. In Korea, due to its geological characteristics, sub-seabed geological $CO_2$ storage is regarded as more practical approach than on-land storage under the goal of its deployment. However, concerns on potential $CO_2$ leakage and relevant acidification issue in the marine environment can be an important subject in recently increasing sub-seabed geological $CO_2$ storage sites. In the present study effect data from literatures were collected in order to conduct an effect assessment of elevated $CO_2$ levels in marine environments using a species sensitivity distribution (SSD) various marine organisms such as microbe, crustacean, echinoderm, mollusc and fish. Results from literatures using domestic species were compared to those from foreign literatures to evaluate the reliability of the effect levels of each biological group and end-point. Ecological effect guidelines through estimating level of pH variation (${\delta}pH$) to adversely affect 5 and 50% of tested organisms, HC5 and HC50, were determined using SSD of marine organisms exposed to the $CO_2$-induced acidification. Estimated HC5 as ${\delta}pH$ of 0.137 can be used as only interim quality guideline possibly with adequate assessment factor. In the future, the current interim guideline as HC5 of ${\delta}pH$ in this study will look forward to compensate with supplement of ecotoxicological data reflecting various trophic levels and indigenous species.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.163-169
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• 2015

In many nuclear power plant sites in Korea, high density storage racks were installed in the spent fuel pool to expand the spent fuel storage capacity. Nevertheless, the capability of the Hanbit nuclear site will be saturated by 2024. Also, 10 NPPs will reach their design life expiration date by 2029. In the case of the US, SFPI (Spent Fuel Pool Island) operated temporarily as a spent fuel storage option before spent nuclear fuels were transported to an interim storage facility or a final disposal facility. As a spent fuel storage option after shutdown during decommissioning, the SFPI concept can be expected to have the following effects: reduced occupational exposure, lower cost of operation, strengthened safety, and so on. This paper presents a case study associated with the regulations, operating experiences, and systems of SFPI in the US. In conclusion, the following steps are recommended for applying SFPI during decommissioning in Korea: confirmation of design change scope of SFPI and expected final cost, the submission of a decommissioning plan which is reflected in SFPI improvement plans, safety assessment using PSR, application of an operating license change for design change, regulatory body review and approval, design change, inspection by the regulatory body, education and commissioning for SFPI, SFPI operation and periodic inspection, and dismantling of SFPI.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1563-1570
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• 2017

In order to perform dry interim storage and transportation of the spent-fuel assemblies of the Bushehr Nuclear Power Plant, dual-purpose casks can be utilized. The effectiveness of different neutron-shield materials for the dual-purpose cask was analyzed through a set of calculations carried out using the Monte Carlo N-Particle (MCNP) code. The dose rate for the dual-purpose cask utilizing the recently developed materials of $epoxy/clay/B_4C$ and $epoxy/clay/B_4C/carbon$ fiber was less than the allowable radiation level of 2 mSv/h at any point and 0.1 mSv/h at 2 m from the external surface of the cask. By utilization of $epoxy/clay/B_4C$ instead of an ethylene glycol/water mixture, the dose rates on the side surface of the cask due to neutron sources and consequent secondary gamma rays will be reduced by 17.5% and 10%, respectively. The overall dose rate in this case will be reduced by 11%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.113-118
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• 2020

An options study was performed for the treatment of residual elemental sodium in driver plenums following the chopping operation during the pyroprocessing of used nuclear fuel. Given the pending availability of a multi-function furnace for distillation and consolidation operations in the Fuel Conditioning Facility, the furnace was considered for the processing of driver plenums. Although two options (oxidation and distillation) could be performed in the multi-function furnace, neither option has been developed sufficiently to date to warrant the use of the furnace for treatment operations. Thus, it was decided to defer the treatment of elemental sodium from driver plenums in the multi-function furnace until more developed technologies and/or furnaces become available. In the interim, storage of the plenums and characterization efforts are recommended.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.1-19
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• 2020

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.681-688
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• 2014

Hydride reorientation behaviors of PWR cladding tubes under typical interim dry storage conditions were investigated with the use of as-received 250 and 485ppm hydrogen-charged Zr-Nb alloy cladding tubes. In order to evaluate the effect of typical cool-down processes on the radial hydride precipitation, two terminal heat-up temperatures of 300 and $400^{\circ}C$, as well as two terminal cool-down temperatures of 200 and $300^{\circ}C$, were considered. In addition, two cooling rates of 2.5 and $8.0^{\circ}C/min$ during the cool-down processes were taken into account along with zero stress or a tensile hoop stress of 150MPa. It was found that the 250ppm hydrogen-charged specimen experiencing the higher terminal heat-up temperature and the lower terminal cool-down temperature generated the highest number of radial hydrides during the cool-down process under 150MPa hoop tensile stress, which may be explained by terminal solid hydrogen solubilities for precipitation, and dissolution and remaining circumferential hydrides at the terminal heat-up temperatures. In addition, the slower cool-down rate generates the larger number of radial hydrides due to a cooling rate-dependent, longer residence time at a relatively high temperature that can accelerate the radial hydride nucleation and growth.