• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.312-312
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.268-273
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• 1997

사용후핵연료는 장기간 강한 방사선과 붕괴열이 방출된다. 따라서 사용후핵연료를 안전하게 운반하기 위하여 수송용기는 방사선차폐의 건전성, 격납경계의 유지 및 내부 붕괴열의 적절한 제거 등의 설계기준을 만족하도록 설계되어야 한다. 본 연구에서는 28개의 PWR 사용후핵연료집합체를 운반할 수 있는 KSC-28 수송용기의 적절한 열전달 특성을 갖는 copper 냉각핀 및 aluminum 전열판을 설정하였다. 또한, 정상수송조건 및 화재사고조건에 대한 열전달해석을 수행하여 수송용기의 열적 건전성을 평가하였고 여기에서 얻어진 온도를 열하중으로 고려하여 열응력해석을 수행함으로써 수송용기의 온도변화에 따른 구조적 건전성을 평가하였다.

• Analytical Science and Technology
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• v.14 no.4
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• pp.291-299
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• 2001

A method of separation and purification of Sm for quantitation of Sm isotopes from various fission products in PWR spent nuclear fuels has been studied. Simulated solution containing inactive metal ions(Cs, Ba, Gd, Eu, Sm and Nd) in place of radioactive fission products was prepared. Sm was separated with 0.5 M $HNO_3$/80% MeOH after washing with 1 M $HNO_3$/90% MeOH on AG $1{\times}8$, anion exchange resin. Sm was purified on cation exchange resin, AG $50W{\times}8$, pretreated with 0.2 M alpha-hydroxisobutyric acid(pH 4.5-4.6) to remove Ba causing isobaric effect Sm from PWR spent fuel. As a result of mass spectrometric measurement, eluted Sm portion did not include isobars form other elements such as Gd, Eu, Pm, Nd and BaO. The contents of Sm and its isotopes($^{147}Sm$, $^{148}Sm$, $^{149}Sm$, $^{150}Sm$, $^{151}Sm$, $^{152}Sm$ and $^{154}Sm$) in spent fuel were determined by isotope dilution mass spectrometric method spiking $^{154}Sm$.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.11a
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• pp.91-92
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• 2006

• The Journal of Engineering Geology
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• v.13 no.2
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• pp.171-191
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• 2003

The objective of this present study is to understand long term(500 years) thermohydromechanical interaction behavior on joints adjacent to a repository cavern, when high level radioactive wastes are disposed of within discontinuous granitic rock masses, and then, to contribute this understanding to the development of a disposal concept. The model includes a saturated discontinuous granitic rock mass, PWR spent nuclear fuels in a disposal canister surrounded with compacted bentonite inside a deposition hole, and mixed bentonite backfilled in the rest of the space within a repository cavern. It is assumed that two joint sets exist within a model. Joint set 1 includes joints of $56^{\circ}$ dip angle, spaced 20m apart, and joint set 2 is in the perpendicular direction to joint set 1 and includes joints of $34^{\circ}$ dip angle, spaced 20m apart. The two dimensional distinct element code, UDEC is used for the analysis. To understand the joint behavior adjacent to the repository cavern, Barton-Bandis joint model is used. Effect of the decay heat from PWR spent fuels on the repository model has been analyzed, and a steady state flow algorithm is used for the hydraulic analysis.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.406-410
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• 2003

DUPIC Fuel Development Facility(DFDF) is the facility to fabricate CANDU-type fuel from spent PWR fuel material without any separation of fissile elements and fission products. Unattended continuous surveillance systems for safeguards of nuclear facility result in large amounts of image and radiation data, which require much time and effort to inspect. Therefore, it is necessary to develop system that automatically pinpoints and diagnoses the anomalies from data. In this regards, this paper presents a novel concept of the continuous surveillance system that integrates visual image and radiation data by the use of neural networks. This surveillance system is operating for safeguards of the DFDF in KAERI.

• Nuclear industry
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• v.34 no.3
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• pp.39-43
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• 2014

한국원자력연구원에서는 사용후핵연료 관리 방안의 일환으로 파이로 공정 기술을 개발하고 있다. 파이로 기술은 PWR 사용후핵연료를 처리함으로써 사용후핵연료의 부피, 방사성 독성, 열부하 및 처분장 면적 등을 감소시킬 수 있을 뿐 아니라, TRU 핵종들을 함께 회수하여 소듐냉각고속로의 금속 연료로 제공이 가능하므로 핵저항성과 핵연료의 재활용률을 증대시킬 수 있다는 장점을 가지고 있다. 한국원자력연구원에서 개발되고 있는 파이로 공정 기술에 대해 전처리 공정에서부터 마지막 폐기물 처리 공정에 이르기까지의 공정 기술에 대해 설명하고 이와 관련된 연구 시설인 DFDF 시설과 ACPF 시설, 그리고 공학 규모 파이로 일관 공정 시험 시설인 PRIDE 3) 시설의 개발 현황을 설명하고자 한다.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.383.1-383
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.339-343
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• 1995

PWR 사용후핵연료 집합체를 운반할 수 있는 수송용기를 개발하기 위하여 단면이 수송용기의 실제 크기인 slice 모델을 사용하여 법규에서 규정하고 있는 정상조건인 주변온도 38$^{\circ}C$에서 냉각 매체로 nitrogen 과 helium 인 경우에 대하여 열시험을 수행하여 수송용기의 열전달 특성 및 핵연료봉의 건전성을 평가하였다. 열시험결과 내부핵연봉의 최대 은도는 각각 448$^{\circ}C$ 와 416$^{\circ}C$로 측정되었다. 이 값들은 핵연료봉의 건전성 유지에 필요한 허용치 이내 만족하는 것으로 수송 용기의 열전달성능이 우수함을 입증하는 것이다.