• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.349-349
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• 1999

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.305.2-305
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• 1999

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 춘계학술논문요약집
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• pp.251-252
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• 2017

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.156-164
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• 2005

The challenges China is facing in energy security are briefly discussed. Then, the development of nuclear power in China in the first half of 21 st century is envisioned, and it is expected that Generation-3 PWR nuclear power plants (NPPs) would be the leading units of nuclear power in the coming $30\~40$ years. As part of the nuclear power program, the R&D work on nuclear fuel cycle is generally proposed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• 방사성폐기물학회지
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• 제21권4호
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• pp.439-450
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• 2023

In this study, the impact load resulting from collision with the fuel rods of surrogate spent nuclear fuel (SNF) assemblies was measured during a rolling test based on an analysis of the data from surrogate SNF-loaded sea transportation tests. Unfortunately, during the sea transportation tests, excessive rolling motion occurred on the ship during the test, causing the assemblies to slip and collide with the canister. Hence, we designed and conducted a separate test to simulate rolling in sea transportation to determine whether such impact loads can occur under normal conditions of SNF transport, with the test conditions for the fuel assembly to slide within the basket experimentally determined. Rolling tests were conducted while varying the rolling angle and frequency to determine the angles and frequencies at which the assemblies experienced slippage. The test results show that slippage of SNF assemblies can occur at angles of approximately 14° or greater because of rolling motion, which can generate impact loads. However, this result exceeds the conditions under which a vessel can depart for coastal navigation, thus deviating from the normal conditions required for SNF transport. Consequently, it is not necessary to consider such loads when evaluating the integrity of SNFs under normal transportation conditions.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1994년도 가을 학술발표회 논문집
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• pp.183-190
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• 1994

The paper provides a review of the analysis methods currently being used to perform seismic analysis of free standing high density spent fuel storage racks for PWR. On the basis of the analysis techniques obtained by KAERI from the design experience of Yonggwang unit 3&4 and Ulchin unit 3&4, the analysis procedure and modeling methods are discussed. The analysis of free standing fuel racks requires consideration of complex phenomena such as hydrodynamic coupling, impact through gap between fuel assembly and poison box and racks, frictional effect, rigid body sliding and tipping and etc. The present modeling of these factors is reviewed in comparison with the recommendation of regulatory group. Further improvement of analysis method and the current issues for the development are discussed.

• 방사성폐기물학회지
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• 제18권spc호
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• pp.63-74
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• 2020

The operational safety assessment is an important part of a safety case for the deep geological repository of spent fuels. It consists of different stages such as the identification of initiating events, event tree analysis, fault tree analysis, and evaluation of exposure doses to the public and radiation workers. This study develops a probabilistic safety assessment method for the operational safety assessment and establishes an assessment framework. For the event and fault tree analyses, we propose the advanced information management system for probabilistic safety assessment (AIMS-PSA Manager). In addition, we propose the Radiological Safety Analysis Computer (RSAC) program to evaluate exposure doses to the public and radiation workers. Furthermore, we check the applicability of the assessment framework with respect to drop accidents of a spent fuel assembly arising out of crane failure, at the surface facility of the KRS⁺ (KAERI Reference disposal System for SNFs). The methods and tools established through this study can be used for the development of a safety case for the KRS⁺ system as well as for the design modification and the operational safety assessment of the KRS⁺ system.

• 방사성폐기물학회지
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• 제6권3호
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• pp.225-232
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• 2008

제3차 전력수급기본계획에 근거하여 현재 운영중이거나 계획중인 원자력발전소에서 발생할 사용후핵연료의 양과 특성을 추정하였다. 본 연구에서 고려된 대상 특성은 핵연료집합체에 대한 제원, 핵연료봉 배열, 무게, $^{235}U$ 초기 농축도 및 방출연소도이다. 이들은 파이로공정 시설을 설계하는데 필수적인 것이다. 2077년말까지 가압경수로 사용후핵연료의 예상발생량은 약 23,000 tU이 될 것으로 보인다. $^{235}U$ 초기 농축도 4.5 wt.% 이하를 갖는 사용후핵연료의 비율은 전체 발생량의 약 95%를 차지할 것이며, 16$\times$16 배열을 갖는 핵연료집합체는 74%를 차지할 것 같다. 현재 사용후핵연료의 평균연소도는 45 GWd/tU인데 반해, 2010년대 중 후반 이후 발생할 사용후 핵연료의 평균연소도는 55 GWd/tU이 될 것 같다. 이상의 결과에 따라 파이로공정 시설의 설계를 위한 기준 사용후핵연료를 도출하였다. 예상 사용후핵연료는 21.4 cm $\times$ 21.4 cm의 단면적, 453 cm의 길이, 672 kg의 질량, 4.5 wt.%의 $^{235}U$ 초기 농축도 및 55 GWd/tU의 방출연소도를 갖는 16$\times$16 한국표준형연료가 타당할 것으로 판단된다.

• 방사성폐기물학회지
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• 제8권4호
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• pp.347-353
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• 2010

사용후핵연료를 파이로 건식처리하면 사용후핵연료 자체 내에 존재하는 세슘, 스트론튬, 초우라늄 계열 등이 중간저장 되어 영구처분 방사선원항에서 제외되므로 사용후핵연료집합체를 구성하는 구조재, 즉 금속폐기물의 방사선원항이 중요해지게 된다. 따라서 본 연구에서는 $17{\times}17$ KOFA 사용후핵연료 10 톤이 파이로 건식처리 되었을 경우를 가정하여 각 구조재 부품별로 방사선원항 특성을 분석하였다. 우선 구조재 부품별로 질량 및 부피를 상세히 계산하였다. 핵연료 상단 및 하단 고정체에서의 중성자스펙트럼이 노심과 다르므로 각 구조재 부품별로 핵반응단면적라이브러리를 KENO-VI/ORIGEN-S 모듈로 직접 생산하였으며, 이를 적용하여 ORIGEN-S 코드로 방사화 방사선원항을 평가하였다. 평가결과 원자로 방출후 10 년 시점에서의 방사능세기, 붕괴열, 위해지수 값은 각각 $1.40{\times}10^{15}$ Bequerels, 236 Watts, $4.34{\times}10^9m^3$-water 로 나타났으며, 이는 사용후핵연료 자체 값의 0.7 %, 1.1 %, 0.1 %에 해당하는 값이다. 방사능세기, 붕괴열, 위해지수 모든 측면에서는 금속폐기물 전체물량의 1 %만을 차지하는 인코넬 718 그리드판이 가장 중요한 것으로 평가되었으며, 특히 이를 따로 분리하여 관리하면 금속폐기물 전체 방사능세기를 20~45 % 정도, 위해지수를 30~45 % 정도 감소시킬 수 있는 것으로 나타났다. 전체적으로 볼 때, 금속폐기물의 방사능세기 및 위해지수는 처분시스템 설계 시 중요한 인자로 고려되어야 하나, 붕괴열은 그 열량이 작아 중요하지 않은 것으로 나타났다.