• The Journal of Engineering Geology
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• v.11 no.3
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• pp.255-267
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• 2001

A comparison study is performed to understand the coupling behavior of the thermal, hydraulic, and mechanical interactions in the vicinity of an underground radwaste repository, assumed to be located at a depth of 500 m, within a granitic rock mass with a 58$^{\circ}$ dipping fault passing through the roof-wall intersection of the repository cavern. The two dimensional universal distinct element code, UDEC is used for the analysis. The model includes a granitic rock meas, a canister with PWR spent fuels surrounded by the compacted bentonite inside a deposition hole, and the mixed bentonite backfilled in the rest of the space within a repository cavern. The coupling behavior of hydromechanical, thermomechanical, and thermohydromechanical interaction has been studied and compared. The effect of the time-dependent decaying heat, from the radioactive materials in PWR spent fuels, on the repository and its surroundings has been studied. A steady state flow algorithm is used for the hydraulic analysis.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.339-346
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• 2019

Based on the $8^{th}$ Basic Plan for Electric Power Demand and Supply, an estimation has been made for inventories and characteristics of spent fuel (SF) to be generated from existing and planned nuclear power plants. The characteristics under consideration in this study are dimensions, fuel array, $^{235}U$ enrichment, discharge burnup, and cooling time for each fuel assembly. These are essentially needed for designing a disposal facility for SFs. It appears that the anticipated quantity by the end of 2082 is about 62,500 assemblies for PWR SFs. The inventories of Westinghouse-type and Korean-type SFs were revealed to be 60% and 40%, respectively as of the end of 2018. The proportion of SFs with initial $^{235}U$ enrichment below 4.5 weight percent (wt%) was shown to be approximately 90% in total as of the end of 2018. As of 2077, more than 97% of SFs generated from Westinghouse-type nuclear reactors were shown to have cooling time of over 50 years. As of 2125, more than 98% of SFs generated from Korean-type nuclear reactors were shown to have cooling time of over 45 years. Based on these results, for the efficient design of a disposal system, it is reasonable to adopt two types of reference spent fuel. SF of KSFA with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 50 years was determined as reference fuel for Westinghouse-type SFs; SF of PLUS7 with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 45 years was determined as reference fuel for Korean-type SFs.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2012.10a
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• pp.83-84
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• 2012

SF 집합체 해체장치 개념설계요건 설정을 위해서 PWR $16{\times}16$ SF집합체 제원을 분석하였다. 또한 집합체 해체장치 주요요건을 도출하여 핵심메커니즘을 도출하였다. 주요요건은 다음과 같다. 집합체의 최대 clamping 힘은 각 grid의 경우: 240 kg, 하부노즐의 경우: 900 kg이다. 3 축 방향에서 절단을 위한 정확한 위치공자는 ${\pm}0.25mm$이다. 또한 처분을 위해 cuttings, fines 및 다양한 hardware를 수거하는 기능을 제공해야 한다. SF 집합체 해체를 위하여 드릴링 방식을 채택하였다. PWR SF의 종류에 따라 드릴링 위치가 다르기 때문에 위치제어와 해제장치 하단과 중간에 있는 X, Y, Z 제어를 할 수 있는 구조로 고안 하였다. SF 집합체 해체장치는 국내에서 가동되는 모든 PWR SF 집합체를 해체할 수 있는 구조로서 범용성을 가지고 있다. 원격 유지보수성을 향상하기 위하여 Solid Works 프로그램 툴(tool)을 이용하여 8개의 주요 모듈을 구성하였고, SF 집합체 해제장치 개념을 3D로 설계하였다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.87-93
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• 2006

Inventories, and characteristics such as dimension, fuel rod array, weight, $^{235}U$ enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial $^{235}U$ enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed $\sim36$ GWD/MTU and $\sim40$ GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be $\sim45$ GWD/MTU at the end of 2000's. From the comprehensive study, it was concluded that the imaginary SNF with $16\times16$ Korean Standard Fuel Assembly, cross section of $21.4cm\times21.4cm$, length of 453cm, mass of 672 kg, initial $^{235}U$ enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.143-149
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• 1989

In order to investigate the in-reactor performance of the 14$\times$14 PWR fuel burner: for 3 cycles in power reactor, non-destructive examination was carried out in KAERI Hot Facility. The results obtained are as follows. 1) The surface of middle and bottom parts of the fuel rod was dark and the upper part was gray. 2) Severe defect such as through-hole was not found. 3) The diameter of rod was shrinked by about 0.65%, while the length was increased by about 0.55% Compared with the design values. 4) The burnup was decreased by about 2% at the inconel grid region compared to other parts.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.283-295
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• 2007

The technical feasibility of a pyroprocessing of PWR spent fuels to recover nuclear fuel materials, uranium and transuranic elements group(TRU), was examined in this study. Also its applicability as a new fuel cycle technology in terms of non-proliferation was investigated. First, various unit processes were combined to a pyroprocess. Then the flow aspects of such materials of issue as uranium, transuraniums, rare earth, noble metals and heat generating elements were examined on the flowsheet, which was obtained by the assumptions on the basis of various experimental results in this work or separation data collected from literatures. Consequently, the calculated results of the material balance for the whole process showed that uranium and TRU could be recovered as products by 98.0 % and 97.0 %, respectively, from a PWR spent fuel while removing the other elemental groups into radioactive wastes. On the one hand, the TRU product was found to emit a considerable amount of ${\gamma}$-ray as well as neutrons favorably contributing to the strategy of proliferation resistance.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.05a
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• pp.301-302
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• 2010

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.10a
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• pp.107-108
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• 2011

• Tunnel and Underground Space
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• v.11 no.3
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• pp.257-269
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• 2001

Numerical simulation is performed to understand the structural behavior of an underground radwaste repository, assumed to be located at the depth of 500 m, in a granitic rock mats, in which a fault intersects the roof of the repository cavern. Two dimensional universal distinct element code, UDEC is used in the analysis. The numerical model includes a granitic rock mass, a canister with PWR spent fuels surrounded by the compacted bentonite inside the deposition hole, and the mixed bentonite backfilled in the rest of the space within the repository cavern. The structural behavior of three different cases, each case with a fault of an angle of $33^{\circ},\;45^{\circ},\;and\;58^{\circ}$ passing through the cavern roof-wall intersection, has been compared. And then fro the case with the $45^{\circ}$ fault, the hydro-mechanical, thermo-mechanical, and thermo-hydro-mechanical interaction behavior have been studied. The effect of the time-dependent decaying heat, from the radioactive materials in PWR spent fuels, on the repository and its surroundings has been studied. The groundwater table is assumed to be located 10m below the ground surface, and a steady state flow algorithm is used.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.135-149
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• 2017

Demonstration programs for spent nuclear fuel dry storage have been carried out to produce important and confirmatory data to support safety of dry storage systems and integrity of spent nuclear fuel stored in dry condition. The US initiated the dry storage of spent nuclear fuel and has strict and explicit regulatory stipulations on the integrity of spent nuclear fuel in dry storage. The US has carried out several notable demonstration programs for the initiation and license extension of dry storage. At the very early stage of dry storage, the demonstration programs were focused on proof of the safety of dry storage systems and a demonstration project called the dry cask storage characterization project was performed for the license extension of low burn-up fuel dry storage. Currently, a demonstration program for the license extension of high burn-up fuel dry storage is under way and is expected to continue for at least 10 years. Korea has not yet begun the dry storage of PWR fuel and the US programs can be a good reference and can provide lessons to safely begin and operate dry storage in Korea. In this paper, past and current demonstration programs of the US are analyzed and several recommendations are provided for demonstration programs for the dry storage of spent nuclear fuel in Korea.