• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.37-50
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• 2020

In this paper, an overview of the scoping calculation results is provided with respect to criticality and radiation shielding of two KBS-3V type PWR SNF disposal systems and one NWMO-type CANDU SNF disposal system of the improved KAERI reference disposal system for SNFs (KRS⁺). The results confirmed that the calculated effective multiplication factors (k_eff) of each disposal system comply with the design criteria (< 0.95). Based on a sensitivity study, the bounding conditions for criticality assumed a flooded container, actinide-only fuel composition, and a decay time of tens of thousands of years. The necessity of mixed loading for some PWR SNFs with high enrichment and low discharge burnup was identified from the evaluated preliminary possible loading area. Furthermore, the absorbed dose rate in the bentonite region was confirmed to be considerably lower than the design criterion (< 1 Gy·hr^-1). Entire PWR SNFs with various enrichment and discharge burnup can be deposited in the KRS⁺ system without any shielding issues. The container thickness applied to the current KRS⁺ design was clarified as sufficient considering the minimum thickness of the container to satisfy the shielding criterion. In conclusion, the current KRS⁺ design is suitable in terms of nuclear criticality and radiation shielding.

• Journal of the Korean Institute of Gas
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• v.18 no.5
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• pp.6-11
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• 2014

To optimize the design of fuel cylinder for LNG vehicles, we introduced the design parameters of the inner and the outer tank of the vessel support structure by analyzing the structural characteristics of conventional design. We selected the inner and outer diameter of the hollow support bars and a dimension of the inner structure of the vessel among the design parameters for design optimization. In this study the temperature distribution and thermal stress of the support structure were evaluated by using the utility program as MSC/MARC. The evaluation criteria are first mode natural frequency, total transferred energy through support structure and thermal stress. The developed design satisfied the design criteria and it was made of prototype. The prototype was verified through three-dimensional vibration testing and thermal performance test.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.299-312
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• 2014

An uncertainty propagation methodology based on the Monte Carlo method is applied to PWR nuclear design analysis to assess the impact of nuclear data uncertainties. The importance of the nuclear data uncertainties for $^{235,238}U$, $^{239}Pu$, and the thermal scattering library for hydrogen in water is analyzed. This uncertainty analysis is compared with the design and acceptance criteria to assure the adequacy of bounding estimates in safety margins.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1869-1877
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• 2021

This paper presents the neutron dose rate analysis of the new CONSTOR® RBMK-1500/M2 storage cask intended for the spent nuclear fuel storage at Ignalina Nuclear Power Plant in Lithuania. These casks are designed to be stored in a new "closed" type interim storage facility, with the capacity to store up to 202 CONSTOR® RBMK-1500/M2 casks. In 2016 y, the "hot trials" of this new facility were conducted and 10 CONSTOR® RBMK-1500/M2 casks loaded with the spent nuclear fuel were transported to the dedicated storage places in this facility. During "hot trials", the dose rate measurements of the CONSTOR® RBMK-1500/M2 casks were performed as the dose rate is one of the critical parameter to control and it must be below design (and safety) criteria. Therefore, having the actual data of the spent nuclear fuel characteristics, the neutron dose rate modeling of the CONSTOR® RBMK-1500/M2 cask loaded with this particular fuel was also performed. Neutron dose rate modeling was performed using MCNP 5 computer code with very detailed geometrical representation of the cask and the fuel. The obtained modeling results were compared with the measurement results and it was revealed, that modeling results are generally in good agreement with the measurements.

• 한국전기화학회:학술대회논문집
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• 2004.06a
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• pp.37-42
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• 2004

To maintain proper operating conditions is important to get optimal output power of a PEMFC stack. The air cooled fuel cell stack is widely used in sub kW PEMFC systems. A 500W air cooled PEMFC stack was experimentally investigated to evaluate the design performance and to get optimal operating conditions for the portable application. The relationship between the operating conditions and the performance was analyzed. The results can be used as design criteria for portable PEMFC under various conditions.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.67 no.2
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• pp.337-342
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• 2018

Spent fuel is moved from the reactor into the spent fuel pool when nuclear power plant permanently shutdown. The sole function of a permanently defueled facility is to store spent fuel in a quiescent state. The function of electric system and loads are reduced. It is necessary to establish an operating strategy of electric system in the permanent shutdown nuclear plant. This paper reviews required loads and design criteria considering transition to permanent shutdown. An operating strategy of onsite electric system is proposed considering decommissioning strategy and stage of defueled condition.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2006.05a
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• pp.250-253
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• 2006

The in-pile Section (IPS) is subjected to flow-induced vibration(FIV) due to the flow of the primary coolant and then the structural integrity. The in-pile Section (IPS) of 3-pin Fuel Test Loop(FTL) shall be installed in the vortical hole call IR1 of HANARO reactor core. In order to verify the velocity and displacement both the inside region of IPS at the annular region of IPS, the vibration was measured by varing the flow rate on both regions. The displacements of fuel assembly in the in-pile Section (IPS) were found to be lower than the values of allowable design criteria.

• Corrosion Science and Technology
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• v.11 no.6
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• pp.213-217
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• 2012

The amount of spent fuel from nuclear power plants has been increasing. An effective management plan of the spent fuel becomes a critical issue, because the storage capacity of each plant will reach its storage limit in a few years. The volume of high toxic spent fuel can be reduced through a fuel processing. Advanced Fuel Cycle (AFC) system is considered to be one of the options to reduce the toxicity and volume of the spent fuel. It is necessary to set up a test facility to demonstrate the feasibility of the process at the engineering scale. The objective of the work is a development of the safety evaluation technology for the AFC system. The evaluation technology of the AFC structural integrity and processes were surveyed and reviewed. Key evaluation parameters for the main processes such as electrolytic reduction, electrorefining, and electrowinning were obtained. The survey results may be used for the establishment of the AFC regulatory licensing procedure. The establishment of the licensing criteria minimizes the trials and errors of the AFC facility design. Issues taken from the survey on the regulatory procedure and design safety features for the AFC facility provide a chance to resolve potential issues in advance.

• Proceedings of the KSR Conference
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• 2002.10b
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• pp.876-883
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• 2002

Electric vehicle body has to be subjected to uniform load and requires auxiliary equipment such as air pipe and electric wire pipe. Especially, the cross beam supports the weight of passenger and electrical equipments. a lightweight vehicle body is salutary to save operating costs and fuel consumption. Therefore this study is to perform the size and the shape optimization of crossbeam fur electric vehicle using the method of topology optimization to introduce the concept of homogenization based on optimality criteria method which is efficient for the problem having the number of design variables and a few boundary condition. this provides the method to determine the optimum position and shape of circular hole in the cross beam and then can achieve the optimal design to reduce weight.

• Journal of the Society of Naval Architects of Korea
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• v.49 no.6
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• pp.447-460
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• 2012

In this study, the shutdown system of the fuel gas supply system is designed based on the Safety Integrity Level of IEC 61508 and IEC 61511. First of all, the individual risk($10^{-4}$/year) and the risk matrix which are the risk acceptance criteria are set up for the qualitative risk assessment such as the HAZOP study. The natural gas leakage at the gas supply pipe is identified as the highest risk among the hazards identified through the HAZOP study and as a safety instrumented function the shutdown function for leakage was defined. SIL 2 and PFD($2.5{\cdot}10^{-3}$) for the shutdown function are determined by the layer of protection analysis(LOPA). The shutdown system(SIS) carrying out the shutdown function(SIF) is verified and designed according to qualitative and quantitative requirements of IEC 61508 and IEC 61511. As a result of SIL verification and SIS conceptual design, the shutdown system is composed of two gas detectors voted 1oo2, one programmable logic solver, and two shutdown valve voted 1oo2.