• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.501-510
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• 2022

The nuclear criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of SNF disposal canister: the KBS-3 canister and small standardized transportation, aging and disposal (STAD) canister. Firstly, the KBS-3 & STAD canister containing four SNFs of the initial enrichment of 4.0wt% ²³⁵U and discharge burnup of 45,000 MWD/MTU were modelled. The k_eff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.79108, 0.78803, and 0.78484 & 0.76149, 0.75683, and 0.75444, respectively. Secondly, the KBS-3 & STAD canister with four SNFs of 4.5wt% and 55,000 MWD/MTU were modelled. The k_eff values for the cooling times of 40, 50, and 60 years were 0.78067, 0.77581, and 0.77335 & 0.75024, 0.74647, and 0.74420, respectively. Therefore, all cases met the performance criterion with respect to the k_eff value, 0.95. The STAD canister had the lower k_eff values than KBS-3. The neutron absorber plates in the STAD canister significantly affected the reduction in k_eff values although the distance among the SNFs in the STAD canister was considerably shorter than that in the KBS-3 canister.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.674-679
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• 1995

The hold-down spring forces of Kori nuclear fuels were measured for seven fuel assemblies having 1 to 4 cycles of irradiation histories in the Kori Unit-1 and -2 reactor. The fuel assemblies examined had burnup from 17 to 38 GWD/MTU and the examination was conducted in KAERI PIEF spent fuel storage pool with the newly developed underwater hold-down suing force measuring device. The measurement was made within the elastic deformation ranges and the trends of hold-down spring force relaxation behavour were examined.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.693-698
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• 1995

원자로 조사 핵연료의 제원거동을 조사하기 위하여 고리원자력1호기 핵연료(평균연소도:17,000-38,000MWD/MTU, 농축도: 2.122-3.199 wt.%) 대한 제원을 측정하였다. 핵연료 연소도에 따른 핵연료봉의 길이신장률과 집합체 길이신장률이 각각 0.4-0.6sc, 0.1-0.2%였다. 조사 핵연료의 길이신장과 핵연료 집합체의 휨은 주로 핵연료 연소도에 의존하였으나 핵연료집합체의 비틀림은 핵연료 연소도와 거의 무관하였다.

• Journal of the Institute of Electronics Engineers of Korea TC
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• v.38 no.7
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• pp.1-5
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• 2001

This study accomplish comparison and analysis multimedia access network speed configuration which is effected by different environmental variation. For this testing. Not only LAN configuration which is symmetric network, But also the ADSL which is asymmetric network is tested by network speed checking software which is designed by our lab in a 10BaseT environment. In the case of TCP receive speed, we find out the network speed is changed according to bands of network to 90% of CPU load. Also, If the TCP receive windows's size became 2 times rate more than MTU, it reduced the relation of speed rate.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.197-201
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• 2012

Annual dose at the boundary of the interim storage facility at normal condition was calculated to estimate the site area of the facility of PWR spent nuclear fuel. In this work, source term was generated by ORIGEN-ARP for 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facilities and radiation shielding evaluations were conducted by MCNP code depending on the storage capacity. In the case of the centralized storage system, the required site area was found to have the radius of more than 700 m.

• Nuclear Engineering and Technology
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• v.24 no.2
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• pp.178-182
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• 1992

Spent PWR fuel rods have been scanned axially and sectionally to measure the relative gamma-ray intensity of Cs-137 and then bumups of the scanned rods determined by measuring Nd-148 which has been chemically separated. From these experimental results, a linear relation(LR) between the gamma-ray intensity of Cs-137 and the bumup in the range of 10∼35 GWD/MTU was obtained. In order to validate the LR, the Cs-137 gamma-ray intensity of unknown sample was nondestructively measured and the bumup obtained by the LR was compared with that of the Nd-148 method. It is revealed that the results from both methods are in good agreement, and thus it seems to be possible to estimate the bumup of spent PWR fuel rod by measuring nondestructively gamma-ray of fission product Cs-137.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.189-193
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• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.27-35
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• 1993

A shielding analysis of the shipping cask for transporting the CANDU spent fuel bundles has been studied. Radiation source term has been calculated on spent fuel with burn-up of 7,800 MWD/MTU and 5 years cooling time by ORIGEN2 code. The shielding calculation for the cask capable of transporting 378 bundles of CANDU spent fuel has been made by use of 1-D ANISN and 2-D DOT 4.2 codes. As a result of analysis, the optimum shield thickness of cask was obtained. And it is proved that the safety in radiation shielding under normal transport and hypothetical accident conditions is confirmed to satisfy the allowable values specified in IAEA Safety Series No. 6 and the Korean Atomic Law.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.248-255
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• 1983

The thermal analysis on the spent fuel shipping cask for a PWR fuel assembly is performed. Under the normal and fire-accident conditions the temperature distribution through a multilayer cask calculated in compliance with 10 CFR Part 71. A KNU 5&6 spent fuel assembly is assumed to be the decay heat source, which has the maximum discharge turnup of 45, 000MWD/MTU and has been stored in the spent fuel storage pool for 300 days. As a result of thermal analysis, the maximum cladding temperature in case of dry cavity under fire-accident conditions is calculated to be 455$^{\circ}C$. This value is much less than the limiting value specified in 10 CFR Part 50.46. It indicates that no fuel rod cladding rupture could occur under fire-accident conditions. It was also found that no melting of lead would take place in the major shield region.

• Nuclear Engineering and Technology
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• v.21 no.1
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• pp.48-55
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• 1989

In this study, simple functional forms which could predict decay heat are referred to and modified in order to analyse more easily long-term behavior of decay heat generated from domestic PWR and CANDU spent fuel. To reduce the difference between the predicted data by functional forms and ORIGEN 2 results and to predict the decay heat under the important parameter(s), sensitivity analysis is performed. By introducing the identified hey parameter, turnup, into the functional forms, the decay heat of spent fuels within a limited rangs of cooling time(3~500 years) becomes predictable for various turnup rates. The predicted decay heat of spent fuels with representative turnup rates such as 33, 37 and 40 GWD/MTU by the functional forms is in so good agreement with ORIGEN 2 results within $\pm$10% difference over the cooling time from 1 to 10$^{5}$ years that the functional forms presented here may be used for engineering purposes such as the thermal design and assessment of the facilities associated with spent fuel management.