• Journal of Radiation Protection and Research
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• v.31 no.2
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• pp.105-113
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• 2006

The oxidizing species are generated from the radiolysis of groundwater in the pore of buffer material around the canister used for the disposal of spent fuels. A mathematical model was introduced to calculate the cathodic current density induced by the oxidant around the canister, which determined the corrosion of carbon steel. An analytical solution was derived to get the cathodic current density in the cylindrical coordinate. The cathodic current densities from both the rectangular coordinate and cylindrical coordinate were compared with each other. The source terms and absorbed dose rate for the calculation of the radiolysis were calculated using the ORIGEN2 and MCNP computer code, respectively. The radius of the canister was determined with the new model in order to prevent the local corrosion. The results showed that the new solution made the cathodic current density around 25 % lower than the Marsh model.

• Journal of Radiation Protection and Research
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• v.13 no.2
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• pp.78-87
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• 1988

This article suggests the radionuclides which should be considered more important to the offsite consequence assesment during a nuclear power plant accident. For this purpose, the relative importance to the early health effects of released radionuclides on the major organs during the accident is estimated under the assumption of the same release fraction. The inventories of the 25 elements, 54 nuclides selected in the Reactor Safety Study are calculated by ORIGEN 2 code. The organs of interest in the estimation are G. I. track, bone marrow, thyroid and lung. The result shows the relative potential importance of radionuclides as follows: For G.I. track, Np, Ce, Ru, Y, and Zr are of importance in sequence, Np, I, La, Sr, Ba for bone marrow, I and Te for thyroid, Cm, Ce, Ru, Pu, Zr for lung. In addition to iodine and noble gases, therefore, the potential contribution of those nuclides listed above to the offsite consequences should not be overlooked for some accidents of particular sequence.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.65-70
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• 1988

A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.545-550
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• 1998

PWR 사용후핵연료 내에 존재하는 $^{134}$ Cs/$^{137}$Cs 및 $^{154}$ Eu/$^{137}$Cs의 감마선 핵종비를 써서 각각 연소도를 결정하고, 그들의 차이가 최소가 되는 시간을 찾는 방법으로 사용후핵연료의 냉각시간을 결정하였다. $^{134}$ Cs/$^{137}$ Cs 및 $^{154}$ Eu/$^{137}$Cs의 핵종비로부터 연소도를 구하는 방법은 이들 핵종비에 대한 ORIGEN-5 코드 계산과 감마스캐닝 실험 결과를 비교하는 것이었다$^{[1]}$ . 사용후핵연료의 냉각시간을 임의의 시간으로 가정하고 핵종비 $^{134}$ Cs/$^{137}$ Cs을 써서 구한 연소도와 $^{154}$ Eu/$^{137}$Cs를 써서 구한 연소도의 차이를 계산했으며, 이 차이는 실제 측정대상 핵연료의 냉각시간에서 최소가 될 것을 기대하였다. 감마선 방출 핵분열생성물인 $^{134}$ Cs와 $^{154}$ Eu는 비교적 긴 반감기를 갖고 있으면서도 또 이들의 반감기 차이가 약 6.4년이나 되므로 기존의 방법$^{[2]}$ 에 비해 넓은 범위의 냉각시간을 정확하게 측정할 수 있었다.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.58-63
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• 1998

AMBIDEXTER(Advanced Molten-Salt Break even Inherently-Safe Dual-Mission EXperiment & TEst Reactor)는 토륨-우라늄 연료주기의 핵적자활성 요건을 설계하는 방법으로써 핵분열중간 생성물인 $^{233}$ Pa의 시간격리, 노내 방사성물질 농도저감, 잉여반응도 및 증식률향상을 위해 핵분열 생성물질의 온라인 정화.처리.재생 개념을 채택하고 있다. 본 연구에서는 AMBIDEXTER 로심의 핵분열성물질의 연소와 온라인 정화.처리에 따른 핵연료내 원소분포 변화를 기술하기 위해 핵분열생성물질의 평형포화농도에 대응하는 등가연소도(Equivalent Burnup)를 정의하고 이를 노심의 핵적자활성 요건에 대해 최적화하는 핵연료 정화공정의 시간상수 특성을 시뮬레이션 하였다. 핵분열생성물질농도의 동특성은 ORIGEN2 코드에 내장된 연속재처리 모델을 이용하여 해석하였으며 실용화가 입증된 후보정화공정들을 고려하여 모든 핵종을 5종의 핵종군으로 분류하여 평가하였다. 시뮬레이션 결과 유효정화주기를 0.1 (노심장전량/일)로 연속재처리 할 때 노심내 포화등 가연소도는 약 650 (MWD/TeH.E.)로 대응되며 이때 동일한 핵연료량으로부터 생성된 노내 핵분 열생성물질 평형농도는 최대연소도 33000MWD/TeU의 PWR 평형노심 BOC시의 대비해 약 1/10 에 해당하는 양이 잔유하는 것으로 나타났다.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.90-96
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• 1984

Estimated dose rates in spent fuel pool storage with the extended fuel cycle core management were reviewed and compared with design limit after calculation with the aid of DLC-23/CASK(22 n, 18 g) nuclear data and ANISN code. Radioactivity and gamma spectrum within spent fuel assemblies were calculated with ORIGEN code by extended fuel cycle model. In the calculation of dose rate, the fuel pool geometry was assumed to be infinite slab. Also, composition materials and radiation source within assemblies which are being stored in pool storage were assumed to be uniformly distributed throughout all the assemblies. As a result of culculation of dose rate from stored assemblies and waterborne radionuclides in pool water, the calculated dose rates appear to be lower than design basis limit under normal condition as well as abnormal condition.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.165-169
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• 1988

As a part of tandem fuel cycle feasibility study, the residual U and Pu nuclide contents of PWR spent fuels are computed using ORICEN2 code for each Korea Nuclear Unit and batch to investigate the potential of utilizing them as CANDU fuels. The annual and accumulated discharged amounts of U and Pu nuclides are computed for the PWRs from KNU 1 through KNU 10. The results of computation show that the spent fuels having 0.7-0.8 w/o U-235 are dominant and considerable amounts of fissile Pu are produced. The enrichment of U-235 is less than the expected 0.8-0.9 w/o U-235 since the burnups offered by KEPCO are higher than those of other PWRs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.237-245
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• 2011

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.39-52
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• 2007

Quantitative analysis on release behavior of the $^{85}Kr\;and\;^{14}C$ fission gases from the spent fuel material during the voloxidation and OREOX process has been performed. This thermal treatment step in a remote fabrication process to fabricate the dry-processed fuel from spent fuel has been used to obtain a fine powder The fractional release percent of fission gases from spent fuel materials with burn-up ranges from 27,000 MWd/tU to 65,000 MWd/tU have been evaluated by comparing the measured data with these initial inventories calculated by ORIGEN code. The release characteristics of $^{85}Kr\;and\;^{14}C$ fission gases during the voloxidation process at $500^{\circ}C$ seem to be closely linked to the degree of conversion efficiency of $UO_2\;to\;U_3O_8$ powder, and it is thus interpreted that the release from grain-boundary would be dominated during this step. The high release fraction of the fission gas from an oxidized powder during the OREOX process would be due to increase both in the gas diffusion at a temperature of $500^{\circ}C$ in a reduction step and in U atom mobility by the reduction. Therefore, it is believed that the fission gases release inventories in the OREOX step come from the inter-grain and inter-grain on $UO_2$ matrix. It is shown that the release fraction of $^{85}Kr\;and\;^{14}C$ fission gases during the voloxidation step would be increased as fuel burn-up increases, ranging from 6 to 12%, and a residual fission gas would completely be removed during the OREOX step. It seems that more effective treatment conditions for a removal of volatile fission gas are of powder formation by the oxidation in advance than the reduction of spent fuel at the higher temperature.