• Explosives and Blasting
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• v.22 no.3
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• pp.79-84
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• 2004

It is well accepted that modem emulsion explosives are intrinsically much less sensitive than traditional products such as dynamites or black powder. However, they have still been involved in a significant number of accidental explosions. In October 1975, Canadian Research, Limited's, Energetic Research Laboratory in Quebec exploded. Although explanations for the incident varied, one logical explanation was that the pump used in transporting the emulsion dead headed, thereby turning mechanical work in to frictional heating under a zero flow rate. There is a minimum pressure required for combustion(MBP) to propagate in emulsion explosives. A stable deflagration may lead to a deflagration-to-detonation transition(DDT) in emulsion explosives. Tests were also performed on sensitized sampled consisting of 6 to 21% waters as well as 1 to 11% aluminium powder. It was founded the emulsion explosives consisting of 6% waters had the lowest minimum homing pressure(MBP) of 3 bar, and the 21% waters were unable to achieve sustained homing at pressures as high as 100 bar. The aluminium contained explosives tested here displayed a MBP higher than that of without emulsion. It appears that this test may offer a firm ground for the classification of emulsion explosives in view of the regulating the hazards associated with the various process used for their manufacturing and transport.

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 2002.05a
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• pp.433-438
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• 2002

화재 및 폭발 특성치로 인화점, 최소발화온도, 폭발한계, 최소발화에너지, 연소열 등을 들 수 있다 이 가운데 폭발한계(explosive limits)는 가연성물질(가스 및 증기)을 다루는 공정 설계 시 고려해야 할 중요한 변수로써, 발화원이 존재할 때 가연성가스와 공기가 혼합하여 일정 농도범위 내에서만 연소가 이루어지는 혼합범위를 말한다/sup 1)/. 특히 폭발범위는 온도, 압력, 불활성가스의 농도, 화임전과 방향, 용기의 크기, 무리리적 상태 등에 의해 변한다/sup 2)/.(중략)

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 1997.11a
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• pp.14-14
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• 1997

추진제의 노화 못지 않게 점화제의 노화도 추진 기관 성능에 큰 영향을 미칠 것으로 예측된다. 따라서 10년 이상 경과된 활성 모터에서 점화기를 분해하여 노화에 의한 점화제의 성능 변화를 알아보았다. 분석에 사용한 점화제는 II-D Bi-Convex형상의 B-KNO$_3$ 펠렛으로 열량, 자동 점화온도, 기계적 물성의 변화를 관찰하였고, 밀폐 용기(Closed bomb)에서 연소시험을 통하여 노화에 따른 점화알약의 압력변화를 측정하고 이론 값과 비교하였다. 또한 비활성 모타를 이용한 연소시험으로 점화기의 점화지연시간, 최대 압력, 최대 압력 도달시간 등을 측정하고 이론식과 비교하여 노화에 의한 변화를 관찰하였다.

• Transactions of the Korean Society of Automotive Engineers
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• v.4 no.1
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• pp.155-164
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• 1996

Dual-fuel engines are being researched with emphasis on the possible types of natural gas supply systems. Hence, a three-dimensional combustion model by using finite volume method was developed to provide a fundamental understanding of the auto-ignition of pilot distillate and subsequent burning of natural gas, when the natural gas as well as the distillate was directly injected into a quiescent diesel engine like combustion bomb tests and the numerical results were investigated for the mixed combustion phenomena. With high-pressure natural gas injection, it was found that the gaseous fuel injection characteristics had to be well harmonised with that of the pilot distillate. For better combustion efficiency, however, further researches are required for the optimisation of injection system in the existence of air motion.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.73-80
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• 2011

The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source ($^{60}Co$ radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.411-422
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• 2016

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2006.05a
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• pp.61-65
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• 2006

A thery was developed to design a high pressure vessel with combined structure of the metal and the composite to withstand the pressure of several tens of thousands psias to reduce the weight of the impulse motor which produces high level of thrust within several tens of seconds. The elastic-plastic stress analyses were carried out to prove the validity of the design theory A combustion chamber of the impulse motor was designed by the design theory, fabricated, and tested by the hydraulic pressure and the static firings. The bursting pressures from the tests were compared to those predicted by tile design theory and the stress analyses and found to be almost the same. It will be possible to design the high pressure vessel with combined structure of the metal and the composite very easily by the proposed design theory.

• Journal of the Korean Society of Propulsion Engineers
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• v.10 no.2
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• pp.23-30
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• 2006

A thery was developed to design a high pressure vessel with combined structure of the metal and the composite to withstand the pressure of several tens of thousands psias to reduce the weight of the impulse motor which produces high level of thrust within several tens of seconds. The elastic-plastic stress analyses were carried out to prove the validity of the design theory. A combustion chamber of the impulse motor was designed by the design theory, fabricated, and tested by the hydraulic pressure and the static firings. The bursting pressures from the tests were compared to those predicted by the design theory and the stress analyses and found to be almost the same. It will be possible to design the high pressure vessel with combined structure of the metal and the composite very easily by the proposed design theory.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.191-198
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• 2011

In general, conventional criticality analyses for spent fuel transport/dry storage systems have been performed based on assumption of fresh fuel concerning the potential uncertainties from number density calculation of Transuranic and Fission Products in spent fuel. However, because of economic loss due to the excessive criticality margin, recently the design of transport/dry storage systems with Burnup Credit(BUC) application has been actively developed. The uncertainties in criticality analyses on transport/storage systems with BUC technique show strong dependance upon initial enrichment and burnup rate, whereas those in the conventional criticality evaluation based on fresh fuel assumption do not show such a dependance. In this study, regulatory-required uncertainties of the criticality analyses for BK 26 Cask, which is conceptually designed spent fuel transport cask with BUC corresponding to the limiting circumstances on nuclear power plants in Korea, are evaluated as a function of initial enrichment and burnup rate. Results of this study will be used as basic data for spent fuel loading curve of BK 26 Cask.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.