• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.60-67
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• 2004

The exposure dose form recycling of a large amount of the steel scrap from the KRR-1＆2 decommissioning activities was evaluated, and also the clearance level(draft) was derived. The maximum individual dose and collective dose were evaluated by modifying internal dose conversion factor which was based on the concept of effective dose in ICRP 60, applied to the RESRAD-RECYCLE ver 3.06 computing code, IAEA Safety Series 111-P-1.1 and NUREG-1640 as the assessment tool. The result of assessment for individual dose and collective dose is 23.9 $\mu$Sv per year and 0.11 man$.$Sv per year respectively. The clearance levels were ultimately determined by extracting the most conservative value form the results of the generic assessment and specific assessment methodologies. The result of clearance level for radionuclides( $Co^{60}$ , C $s^{l37}$) is less than 1.14${\times}$10$^{-1}$ Bq/g to comply with the clearance criterion(maximum individual dose : 10 $\mu$Sv per year, collective dose : 1 man$.$Sv per year) provided for Korea Atomic Energy Act and relevant regulations.s.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2008.05a
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• pp.47-48
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• 2008

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.11a
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• pp.59-60
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• 2007

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.220-221
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.645-649
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• 2003

The variety conditions and working situations during decontamination and decommissioning practice may cause the production of metallic wastes, which can differ in character or in quantity from the waste anticipated during D&D planning. The management of such kind of wastes involves the need to evaluate existing waste management system in order to determine how metallic wastes should be well handled and treated. In this study, the investigation of existing management practices which can be also applied to the metallic waste management, was carried. Simultaneously, assistance in selection of appropriate technologies and processes is investigated which can be used when anticipated situations occur during decontamination and decommissioning practice.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.202-207
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• 2012

The radioactivity inventory significantly affects all steps of decommissioning projects including planning, cost estimation, risk assessment, waste management and site remediation. The decommissioning project of the KRR-2 was completed in 2009 and a large amount of activated concrete waste was generated. The bioshield concrete, containing minute amount of impurity elements, was activated by neutron reaction during the operation of the reactor. A variety radionuclides was generated in the concrete, including $^3H$, $^{14}C$, $^{55}Fe$, $^{60}Co$ $^{63}Ni$, $^{134}Cs$, $^{152}Eu$ and $^{154}Eu$. In this paper, the comparison between the calculated results and previous measured results was carried out to estimate the inventory of the bioshield concrete of the KRR-2. The combined computer codes of MCNP5 and ORIGEN 2.1 for calculation of the distribution of neutron flux, cross-section and generation of radionuclides were used. The results were shown that 99.8% of the total radioactivity of $^3H$, $^{55}Fe$, $^{60}Co$ and $^{152}Eu$ in the bioshield concrete 12 years after shutdown. The effects on the variation of inventory were analysed depending on the operation periods and the cooling times in the bioshield concrete.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1275-1276
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.836-837
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• 2005

• Proceedings of the IEEK Conference
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• 2001.10a
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• pp.198-201
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• 2001

Porous structures of aluminum foam have been studied by using return aluminum scrap. The apparent foam shape, foam height, density, pore size and their distributions in various section areas of the experimental samples have been investigated. The sample have been cast into metallic mold, using aluminum foam prepared from a precursor based on pure Al ingot and return aluminum scrap mixed with various amounts of 1-2wt% increasing viscosity and foam agent materials. The process provides for flexibility in design of foam structures via relatively easy control over the amount of hydrogen evolution and the drainage processes which occur during foam formation. This is facilitated by manipulating parameters such as the foaming agent, thermal histories during solidification and mix melt viscosities. A metal for producing the foamed are decomposing a foaming agent in a molten metal such that there is an initial and a subsequent expansion due to foaming agent. It has been found that the Al porous foaming with variation amount of 1∼2wt% foam agent and at 2min holding time, which melting temperature has appeared homogeneous pore size at 650∼700$^{\circ}C$. The compression strength were 10-13 kg/min at 125ppi, and increased by higher pore density. The acoustical performance of the panel made with the foamed aluminum is considerably improved; its absorption coefficient shows NRC 0.6-0.8. It has been found that the Al foam is very preferable for the compactness of the thermal system.

• Journal of the Korean Society of Safety
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• v.37 no.4
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• pp.129-138
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• 2022

A new method for injecting cooling water into the Korean research reactor (KRR) in the event of beam tube rupture is proposed in this paper. Moreover, the research evaluates the risk to the reactor core in terms of core damage frequency (CDF). The proposed method maintains the cooling water in the chimney at a certain level in the tank to prevent nuclear fuel damage solely by gravitational coolant feeding from the emergency water supply system (EWSS). This technique does not require sump recirculation operations described in the current procedure for resolving beam tube accidents. The reduction in the risk to the core in the event of beam tube rupture that can be achieved by the proposed change in the cooling water injection design is quantified as follows. 1) The total CDF of the KRR for the proposed design change is approximately 4.17E-06/yr, which is 8.4% lower than the CDF of the current design (4.55E-06/yr). 2) The CDF for beam tube rupture is 7.10E-08/yr, which represents an 84.1% decrease compared with that of the current design (4.49E-07/yr). In addition to this quantitative reduction in risk, the modified cooling water injection design maintains a supply of pure coolant to the EWSS tank. This means that the reactor does not require decontamination after an accident. Thermal hydraulic analysis proves that the water level in the reactor pool does not cause damage to the nuclear fuel cladding after beam tube rupture. This is because the amount of water in the chimney can be regulated by the EWSS function. The EWSS supplies emergency water to the reactor core to compensate for the evaporation of coolant in the core, thus allowing water to cover the fuel assemblies in the reactor core over a sufficient amount of time.