• Journal of the Korea Organic Resources Recycling Association
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• v.15 no.3
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• pp.106-112
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• 2007

The purpose of this study is to derive the mixing ratio with stable heating value to be used as fuel and secondary fuel by mixing sewage sludge cake, tar, peat moss, and verify the combustion property of produced solid fuel and the applicability of RDF as alternative fuel. Tar shows the highest heating value with 7,000kcal/kg and the heating value of sewage sludge cake and peat moss ranges from 4,000 to 4,500kcal/kg. Also, the solid fuel with length 1.6cm, diameter 1.3cm and weight 2.3g was produced using the heating value of over 6,000kcal/kg and proper mixing ratio (sewage sludge cake: tar: peat moss) from 1 : 4 : 1 to 1 : 7 : 1. Upon the analysis of the RDF applicability of produced solid fuel, the exhaust gas analysis finds that the composition concentration of exhaust gas occurred according to the mixing ratio did not change significantly and the flame lasting time was found to be around 5 minutes, similar to the lasting time of the same mass (2.3g) of general anthracite burned. Therefore, it can be concluded that solid fuel produced in this study can be used as fuel and secondary fuel.

• Economic and Environmental Geology
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• v.28 no.3
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• pp.251-257
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• 1995

Three PEA (Pulverised Fuel Ash) samples, which were fresh, 17 and some 40 years weathered, were collected from two major British power plants. Batch leaching tests with these samples using distilled water and simulated industrial leachate showed higher amounts of element liberation from fresh ash, including Ca, Na, K, S (as $SO^{2-}_4$, $Cr_{total}$, Cu, Li Ni, Mo and CI and this seems to indicate their surface association and easier dissolution when contact with water. On the contrary Mg, Al, Ba, Si, V, As and Se do not show such readily leachable concentrations and these elements might be more associated with glass fraction in PFA particle rather than surface. Although element concentrations in the weathered ash are much lower than those in the initial leachate from the fresh ash, elements are still detected as resonable concentrations, with rather constant levels and this seems to demonstrate the element release from unstable glass phase of PFA particle. Fe, Ca, $Cr_{total}$, Cu, Ni, Zn and Hg were removed from the synthetic leachate by PFA and this is also confirmed by gain in solid PFA. The order of element retention is Meaford weathered ash > Drax weathered ash > Drax fresh ash in decreasing order and this conforms with the degree of weathering. Namely, the more wethered, the more wethered, the more effective in metal retention from the synthetic leachate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.53-64
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• 2007

The dynamic release behavior of Cs from high burn-up spent PWR fuel was experimentally performed under the conditions of a thermal treatment process such as voloxidation and sintering conditions. In voloxidation process, influence of the oxidation and reduction atmosphere on the Cs release characteristic using fragment type of spent fuel heated up to $1,500^{\circ}C$ was compared. In sintering process, temperature history effect on Cs release behavior was evaluated using green pellet under 4% $H_2/Ar$ environment. Temperature range for complete Cs release from spent fuel fragment under voloxidation condition was about $800^{\circ}C{\sim}1,200^{\circ}C$, but that of green pellet under the reduction atmosphere was $1,100^{\circ}C{\sim}1,400^{\circ}C$. Key parameters on Cs release behavior from spent fuel was powder formation as well as the diffusion rate of Cs compound to grain boundary and fuel surface.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.79-84
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• 2007

Inventories of soluble elements in the gap and grain boundaries of domestic used PWR fuel pellets were measured to estimate the quantities of radionuclides that are liable to be rapidly released into the groundwater of a disposal site. The gap inventory of cesium for the pellets in the used fuel with a burn-up range of 45 to 66 GWD/MTU showed 0.85 to 1.7% of its total inventory, which was close to 1/6 to 1/3 of the fission gas release fraction (FGRF). However, the amounts of cesium released from the gaps of the pellets below 40 GWD/MTU of a burn-up and less than 1% FGRF were so erratic that the gap inventory could not be defined by ie FGRF. Strontium inventories in the gap and grain boundaries of the pellets in the same rod were not significantly varied, and the iodine inventory in the gap of the used PWR fuels was estimated to be less than or the same as the FGRF.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.135-149
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• 2017

Demonstration programs for spent nuclear fuel dry storage have been carried out to produce important and confirmatory data to support safety of dry storage systems and integrity of spent nuclear fuel stored in dry condition. The US initiated the dry storage of spent nuclear fuel and has strict and explicit regulatory stipulations on the integrity of spent nuclear fuel in dry storage. The US has carried out several notable demonstration programs for the initiation and license extension of dry storage. At the very early stage of dry storage, the demonstration programs were focused on proof of the safety of dry storage systems and a demonstration project called the dry cask storage characterization project was performed for the license extension of low burn-up fuel dry storage. Currently, a demonstration program for the license extension of high burn-up fuel dry storage is under way and is expected to continue for at least 10 years. Korea has not yet begun the dry storage of PWR fuel and the US programs can be a good reference and can provide lessons to safely begin and operate dry storage in Korea. In this paper, past and current demonstration programs of the US are analyzed and several recommendations are provided for demonstration programs for the dry storage of spent nuclear fuel in Korea.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.237-246
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• 2005

A tip-over test of the 1/3 scale model is conducted to verify the tip-oner analysis of a dry concrete storage cask under a hypothetical accident condition. The tip-oner analysis is executed using the velocity at each point which are determined from the initial angular velocity as the initial conditions of the model just before the impact. To confirm the structural integrity of the canister of a dry concrete storage cask, the non-detective testing such as Liquid Penetrants testing and Ultrasonic Testing are conducted. The strains and tile accelerations acquired by the tip-over test are compared with those by the analysis to verify the tip-over analysis. The lid of a storage calk are plastically deformed at the impact point. Liquid

• Journal of Energy Engineering
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• v.19 no.4
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• pp.240-245
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• 2010

In this study, characteristics of wood pellet gasification was studied using a Two Stage Gasifier which is consisted of pyrolysis reactor and ultra high temperature reformer. The average yields of $H_2$, $CH_4$, CO, $CO_2$ were 16.7, 11.3, 37.2, 26.6 L/mim, conversion rate from biomass to gas was 65% in pyrolysis reactor and gas yields in reformer were 55.4, 0.8, 120.8, 56.8 L/mim, respectively. The hydrogen flow rate from reformer is obtained 360.1 L/hr. The most of $CH_4$ was decomposed from 12.3 to 0.3 vol.% while $H_2$ is from 18.2 to 23.7 vol.% in reformer by methane dry reforming, Boudouard reaction, oxidation and/or steam reforming. The amount of $H_2O$ generated by hydration reaction from reformer was 1111.8 g, its accelerated conversion of $CH_4$ to other products. The conversion rate from $CH_4$ to other Compounds was 97.2%. Cold gas efficiency was 53.2%.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.35-36
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• 2009

지속가능 에너지원인 원자력은 또한 글로벌 에너지원의 특성을 갖추고 있다 핵연료는 원자로에 장전되는 신규 핵연료를 구성하고 있는 우라늄 채광 단계에서부터 연소 후에 발생하는 사용후 핵연료 (SNF 또는 SF) 관리 단계에 이르는 전과정에 걸쳐 핵연료 사이클로 파악되고 있다. 이러한 전과정 관점에서 볼 때, 핵연료 사이클에 관여하는 이해관계자는 자국뿐만 아니라 다국가 (multination)를 포함하고 있다. 특히, 후행 핵연료 사이클인 사용후 핵연료의 저장, 처리, 처분 단계에서는 다국가 시나리오를 배제하지 않는 사용후 핵연료 관리 전략의 도입이 고려될 수 있다. 여기서 다국가는 접경 국가, 인접 국가, 핵연료 공급 국가, 재처리 제공 국가, 재처리 위탁 국가, SNF 통과 허용 국가, SNF 저장 부지 제공 국가, SNF 향후 이용 국가 등이 될 수 있다 [김성호 2006]. 현재 우리나라에서는 여러 국가로부터 수입되고 있는 신규 핵연료 물질을 연소시켜 나온 사용후 핵연료를 부지내에 임시 저장하고 있다. 사용후 핵연료 발생량의 추산에 따르면, 2016년쯤에 현재 임시저장 용량이 포화될 예정이다 이러한 상황에서 다국가 시나리오를 포함한 관리 전략은 다국가 시나리오를 배제한 관리 전략과 다각적인 측면 에서 비교 검토될 필요성이 있다. 사용후 핵연료의 영구 처분장 부지 확보를 해결하기 위한 선행 단계로 공론화 단계가 지금 준비되고 있다. 예컨대, 단기 공론화 관리 방안의 하나로 비록 소극적인 입장이지만 타국 위탁 재처리 방식이 고려되고 있다 [KRS 2009] 이 연구에서는 단기적인 사용후 핵연료 관리 전략으로 여러 가지 다국가 수준의 저장, 처리, 처분 방식으로 바탕으로 다국가 시나리오들을 제안하려고 한다. 이들 다국가 시나리오를 포함한 관리 전략은 현재 다국가 시나리오를 배제한 국내 사용후 핵연료 처분장 부지 선정이 정치적/사회적 수용성 문제로 어려운 상황에 처할 경우에 해결책을 찾는 데에 기여하리라 본다. 또한, 부지 선정 단계에서 바라지 않는 난항이 나타나는 경우에 국가 차원의 한 대비책으로 다음을 제안한다: 한편으로는 자국 저장 시설이 추진되면서, 다른 편으로는 타국 저장 부지를 확보하는 전략이 검토되어야 한다. 이러한 이중 노선 (dual track) 전략은 여러 유럽 국가에서 이미 고려되고 있는 방안이다 [Greenpeace 2005] 다양한 다국성 정도 (a degree of multinationality) 의 저장, 처리, 처분 방식을 연결하는 가능한 다국가 시나리오 구조가 Fig.1에 제시되어 있다. 다국가 시나리오를 구성하는 기본 요소는 다음과 같다’ 1) 자국 임시 저장; 2) 자국 재처리; 3) 자국 중간 저장; 4) 자국 영구 처분; 5) 다국가 중간 저장; 6) 다국가 재처리; 7 ) 다국가 영구 처분. 이들 기본 요소들을 다국성 정도에 따라 결합하면 다양한 다국가 시나리오들이 얻어진다. 이들을 포함한 SNF 관리 전략은 크게는 1) 다국가 재처리 전략， 2) 다국가 저장 및 재처리 전략， 또는 3) 다국가 처분 전략 등으로 분류될 수 있다.

• 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.

• Journal of the Korean Recycled Construction Resources Institute
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• v.8 no.3
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• pp.286-293
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• 2020

In this study, we tried to verify the correlation of the amount of combustible industrial by-products, household waste used as fuels on cement sintering process and the amount of NOx, and CO, harmful components in the exhaust gas. The analysis uses coal as natural fuel, soft plastics (plastics with properties that tend to be scattered by wind, such as vinyls), hard plastics (plastics with properties that are not scattered by wind, such as PETs, wate rubbers), and reclaimed oils as alternative fuels. Utilizing the response surface analysis (RSM) technique using the process data of 2019, such as the fuel input and combustion temperature of a domestic A cement manufacturer's sintering facilities as independent variables, and the NOx, and CO emissions to the stack as dependent variables. Correlation was analyzed. As a result, it was confirmed that the impact on the emission material differs for each waste. In particular, it was analyzed that the hard plastics increase the CO emission but have an excellent effect of reducing NOx.