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Yoo, Jae-Hyung;Hong, Kwon-Pyo;Lee, Han-Soo
- Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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- v.6 no.3
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- pp.233-244
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- 2008
A conceptual design study for a pyroprocesing facility, has been carried out in this study, which is available for the recovery of uranium and transuranic elemental group(TRU), that is, reusable as a nuclear fuel especially in a next generation-type reactor. The scale of this facility has been chosen as 20 kg HM/batch, comparatively small engineering size in order to collect scale-up data for the design of a commercial facility as well as to get operational experience. The spent fuel to be handled in this process is as follows : 3.5 % enriched uranium fuel, 35,000MWd/tU and 5-year cooled. The major items considered in the conceptual study are a building lay-out including various hot cells, safety management of the process operation in conjunction with material balance control, radiation safety, inert atmosphere control in shielded hot cells, and criticality control of uranium and TRU products.
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Kang, Seong-Ku;Song, Jong-Soon
- Journal of Radiation Protection and Research
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- v.25 no.2
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- pp.89-96
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- 2000
Nuclear fuel cycle choices and costs are important in considering energy policies, fuel diversity, security of supply and associated social and environmental impacts. Particularly, the nuclear spent fuel is very important in view of high activity and the need of long term management. This study focuses on the comparison of reprocessing and direct disposal of nuclear spent fuel in terms of cost, safety and public acceptability. The results of the study show that the direct disposal is about 7% more economical than the reprocessing. In terms of safety, the results show that the risk of vitrified HLW (high-level radioactive waste) is less than directly disposed spent fuel. For the public acceptability, both of the methods are not well understood and therefore they are not accepted. In conclusion, it is necessary to guarantee the safety of the both spent fuel processing methods through continuous development of associated technology and to have a fuel cycle policy which should consider not only the economics but also social and environmental impacts.
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방경식;이주찬;주준식;서기석;김호동
- Proceedings of the Korean Radioactive Waste Society Conference
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- 2003.11a
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- pp.609-613
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- 2003
The confinement of a storage system to accommodate spent fuel and radioactive material must be designed and estimated so that the storage system is safe during a storage period. The confinement can be estimated by calculation of the allowable leakage rate in compliance with ANSI Nl4.5 or ISO 12807. Accordingly, the allowable leakage rate was estimated in compliance with ISO 12807 in the case of storage of 24 PWR spent fuels and 24-conditioned spent fuels. In the case of the 24 PWR spent fuel assemblies, the allowable leakage rate was estimated as $7.43{\times}10_{-11}m_3/s$, in the case of the 24 conditioned spent fuel assemblies, the allowable leakage rate was estimated as $1.80{\times}10_{-10}m_3/s$. Therefore, the confinement condition in the storage of the conditioned spent fuel is easier than that for the storage of the PWR spent fuel.
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Koo, Dae-Seo;Park, Seong-Won
- Journal of the Korean Society for Nondestructive Testing
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- v.22 no.1
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- pp.9-13
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- 2002
A pool image processing measurement method has been developed to improve the examination efficiency and to minimize the errors of dimensional measurements of spent fuel assemblies in pool. Diameter and length measurements of mock-up fuel rods using the image processing system are $-0.24{\pm}0.03mm,\;0.34{\pm}0.06mm$ on the basis of the true value and their maximum errors are within -0.3 and 0.4mm, respectively, According to the result of dimensional measurement of spent fuels in pool, the upper and lower part diameter and mid part diameter of fuel rods of the J44 fuel assembly irradiated for 2 cycles in the Kori-2 nuclear reactor were decreased by about 2.0 and 3.0% in comparison with design values, respectively. The length of fuel rods was elongated by about 0.4%. The change behavior of diameter and length. of fuel rods of the F02 fuel assembly irradiated for 3 cycles in the Kori-1 nuclear reactor showed a trend similar to the results of J44.
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김명현;윤진규
- Proceedings of the Korean Nuclear Society Conference
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- 1996.05a
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- pp.112-116
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- 1996
신형경수로의 대안으로서 가압경수로의 단점을 보완하고, 가압중수로의 장점을 채택한 중수감속 경수로의 핵적 개념설계를 제안하였다. 냉각재와 감속재가 서로 다른 채넬을 통해 흐르는 기존 가압중수로의 Pressure-Tube 설계의 장점을 채택하여, 냉각재는 경수를 감속재는 중수를 사용하는 중수감속 가압경수로(DPWR, Deuterium-moderated PWR)의 설계 타당성을 검토하였다. 기본적으로 CANDU의 system설계를 Proven Technology로서 가능한 많이 채택하고, CANFLEX 핵연료 설계도 기존 연구 결과로서 최대한 활용하였다. 월성 2,3,4호기 FSAR의 사양을 그대로 사용하여 기존 중수로의 37봉 핵연료 다발을 6$\times$6 직각 배열 등가 핵연료집합체로 재구성한 후, SEU $UO_2$ 핵연료에 대해 HELIOS코드를 사용하여 핵적 특성을 검토하였다. 냉각재 온도계수가 음의 안전성을 갖고 있으며, 기존 중수로보다 연소도가 훨씬 큰 원자로가 설계될 수 있음을 확인하였다. 또한 발전소 이용률의 증대, 사용후 핵연료 발생량의 감소를 기대할 수 있었다.
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