• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.311-312
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• 2009

가압 경수로에서 53,000 MWd/tU으로 연소된 사용후 핵연료의 PCI 영역에 대해 방사선 차폐형 성분 분석기기( Shielded EPMA)를 사용하여 반경방향에 대한 성분분포를 분석하였다. PCI 영역에서 산화층의 두께는 13um 이었으며, 핵분열생성물의 침투 두께는 시료에서 약 10 um 이내로 나타났다. 이 두께에 침투된 핵종의 총 농도는 1~2 wt%로 관찰되었다. 주요핵종은 Cs 0.5~0.7 wt%, Mo 0.2~0.3 wt%, Pd, Ru, Nd, Ce등이 0.1~0.2 wt% 로 관찰되었다.

• 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.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.335-343
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• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with $3.2\%$ of enrichment had been irradiated up to 35,000 MWd/MTU(reference data). The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, chemical analysis (destructive method) has been used but it mattes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

• Applied Microscopy
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• v.35 no.3
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• pp.113-119
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• 2005

The Methodology of burnup calculation with EPMA test set up in this study. The spent fuel from PWR nuclear power plant was used as specimen. This $UO_2$ fuel with 3.2% of enrichment had been irradiated up to 35,000 MWd/MTU. The burnup is very important factor for nuclear fuel to estimate all fuel behaviors in reactor. To measure amounts of fission products and actinides for the burnup calcualation, destructive method analysis has been used but it makes long experimental time and second radio-wastes. In this study, EPMA test was available to measure amount of fission products. Neodymium is able to be detected and quantified. It can be compared with the results from chemical analysis and ORIGEN-2 code calculation. Concentration of Nd from EPMA test showed good agreement with result of ORIGEN-2 code in the same burnup.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.329-333
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• 2005

To identify the effect of engineered barriers on the leach rate of cesium from spent PWR fuel under a synthetic granitic groundwater, the related leach tests with and without bentonite or metals have been performed up to about 6 years. The leach rates were decreased as a function of leaching time and then became a constant after a certain period. The period in a bare spent fuel was much longer than that with bentonite or metal sheets. The cumulative fraction of cesium released from the spent fuel with bentonite or with copper and stainless steel sheets was steadily increased, but the fraction from bare fuel was rapidly and then sluggishly increased. However, the values deducted its gap inventory from the cumulative fraction of cesium released from the bare fuel was almost very close to the others. These suggest that the initial release of cesium from bare fuel might be dependant on its gap inventory and the effect of engineered barriers on the long-term leach rate of cesium would be insignificant but the rate with engineered barriers could be reduced in the initial transient period due to their retardation effect. And the long-term leach rate of cesium from spent fuel in a repository would be approached to a constant rate of $2\times10^{-2}g/m^2-day$.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.92-100
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• 2015

For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.197-202
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• 1997

ORIGEN2코드의 검증계산을 통해 PWR 사용 후 핵연료 조성핵종의 핵종량에 대한 핵임계측면에서 보수성을 가지는 안전인자를 산출하였고, MCNP코드의 검증계산으로 95/95 신뢰구간에서의 계산오차를 구하였다. 이를 바탕으로 직경이 1.2567cm이고 길이가 380.5cm인 196 개 금속봉을 장전한 캐니스터 ( 금속저장체 )가 x-y 방향으로 무한히 배열된 경우에 대해 캐니스터의 두께, 간격 및 외부의 공기중 수분농도에 따른 핵임계 안전해석을 수행하였다. 그 결과, 캐니스터의 두께가 7mm일 때 공기중 수분농도가 0.30 g/㎤이고 캐니스터간의 간격이 6.0cm인 경우의 최종핵 임계도값은 0.94130로서 최대허용핵임계값 (0.942)보다 적은 값을 보였다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.41-50
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• 2006

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm-container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by $\sim20$ tons.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.270-275
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• 2003

The heat transfer analysis was conducted for the conceptual design of high level waste canisters. The temperature distribution due to the heat generation from four PWR spent fuel bundles which were contained in a canister located in a borehole 500 m below the surface was obtained. NISA computer program based upon FEM was used for the numerical solution. The temperature distribution in the composite system of $\ulcorner$canister ＋ buffer ＋ tunnel ＋ rock$\lrcorner$ due to heat generation from the spent fuel was obtained. In the case of 40m tunnel spacing and 6m borehole spacing the temperature showed the maximum value of $87.5^{\circ}C$around 15-16 years after disposal and decreased.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.237-247
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• 1993

Activity ratio of two radioactive primary fission products which had sufficiently different half-lives was expressed as functions of cooling time and irradiation histories in which average burnup, irradiation time, cycle interval time and the dominant fissile material of the spent fuel were included. The gamma-ray spectra of 36 samples from 6 spent PWR fuel assemblies irradiated in Kori unit-1 reactor were obtained by a spectrometric system equipped with a high purity germanium gamma-ray detector. Activity ratio $^{l44}$Ce $^{l37}$Cs, analyzed from each spectrum, was used for the calculation of cooling time. The results show that the radioactive fission products $^{l44}$Ce and $^{l37}$Cs are considered as useful monitors for cooling time determination because the estimated cooling time by detection of activity ratio $^{l44}$Ce $^{l37}$Cs agreed well with the operator declared cooling time within relative difference of $\pm$5 % despite the low counting rate of the gamma-ray of $^{l44}$Ce (about 10$^{-3}$ count per second). For the samples with several different irradiation histories, the determined cooling time by modeled irradiation history showed good agreement with that by known irradiation history within time difference of $\pm$0.5 year. From this result, it would be expected to be possible to estimate reliably the cooling time of spent nuclear fuel without the exact information about irradiation history. The feasibility study on identification of and/or sorting out spent nuclear fuel by applying the technique for cooling time determination was also performed and the result shows that the detection of activity ratio $^{l44}$Ce $^{l37}$Cs by gamma-ray spectrometry would be usefully applicable to certify spent nuclear fuel for the purpose of safeguards and management in a facility in which the samples dismantled or cut from spent fuel assemblies are treated, such as the post irradiation examination facility.mination facility.