• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.11-16
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• 2016

Fuel assembly's mechanical characterization test facility (FAMeCT) in KAERI was constructed with upgraded functional features such as increased loading capacity, underwater vibration testing and severe earthquake simulation for extended fuel design guideline. This facility is designed and developed to provide out-pile fuel data for accident analysis model and fuel licensing. Functional tests of FAMeCT were performed to confirm functionality, structural integrity, and validity of newly-built fuel assembly mechanical test facility. Test program includes signal check of data acquisition system, load delivering capacity using real-sized fuel assemblies and a standard loading cylindrical rigid specimen. Fuel assembly's lateral bending test was carried out up to 30 mm of pull-out displacement. Limit case axial compression loading test up to 33 kN was performed to check structural integrity of UCPS (Upper Core Plate Simulator) support frame. Test results show that all test equipment and measurement system have acceptable range of alignment, signal to noise ratio, load carrying capacity limit without loss of integrity. This paper introduces newly constructed fuel assembly's mechanical test facility and summarizes results of functional test for the mechanical test equipment and data acquisition system.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.495-500
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• 2019

When a new fuel is developed, various mechanical properties are absolutely necessary for a safety analysis of the fuel for the licensing and prediction of its mechanical behavior during operation and accident conditions. In this paper, a mechanically equivalent surrogate plate of U-Mo dispersion fuel is presented using tungsten, substitute material of U-Mo particle. A surrogate plate, composed of tungsten/aluminum dispersion meat and aluminum alloy cladding, is manufactured with the same fabrication process with that of fuel plate except that a tungsten powder is used instead of U-Mo powder. A modal test showed that the surrogate plate and fuel plate have similar dynamic characteristics, and a tensile test demonstrated the similarity of the material property up to the yield strength range. The conducted tests proved that the surrogate tungsten plate has equivalent mechanical behaviors with that of a fuel plate, which leads to the acceptable use of a surrogate fuel assembly using tungsten/aluminum dispersion meat in various mechanical tests. The surrogate fuel assembly can be utilized for various out-of-pile characteristic tests, which are necessary for the licensing achievement of a research reactor that uses U-Mo dispersion fuel as a driver.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.1026-1031
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• 1998

A licensing approach for new fuel design is propose. It includes the possible licensing options to be taken according to the level of significance of design changes. This approach can be applied on legal grounds of the AEA, Article 44.3 and Article 21, but implies that the related ERAEAs should be revised or other alternative regulatory guidelines should be prepared. However, it is not intended to invoke additional requirements but rather to streamline and formulate the current practice by using one of supplementary provisions of the AEA, Article 104.2 which is recently enacted for approval of the technical report on a special topic.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.147-151
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• 1997

Fuel management simulation have been performed for CANFLEX-0.9％ RU fuel in the CANDU 6 reactor. In this study, the bi-directional 4-bundle shift fuelling scheme was assumed The lattice cell and time-average calculation were carried out. The refuelling simulation calculations were performed for 600 full power days. Time-averaged results show good axial power profile with the CANFLEX-RU fuel. During the simulation period, the maximum channel and bundle power were maintained below the licensing limit of CANDU 6 reactor.

• Nuclear Engineering and Technology
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• 제42권3호
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• pp.249-258
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• 2010

Recently, many utilities have considered interim dry storage of spent nuclear fuel as an option for increasing spent fuel storage capacity. Foreign nuclear regulatory committees have provided some regulatory and licensing requirements for relatively low- and medium-burned spent fuel with respect to the prevention of spent fuel degradation during transportation and interim dry storage. In the present study, the effect of cladding creep and hydride distribution on spent fuel degradation is reviewed and performance tests with high-burned Zircaloy-4 and advanced Zr alloy spent fuel are proposed to investigate the effect of burnup and cladding materials on the current regulatory and licensing requirements. Creep tests were also performed to investigate the effect of temperature and tensile hoop stress on hydride reorientation and subsequently to examine the temperature and stress limits against cladding material failure. It is found that the spent fuel failure is mainly caused by cladding creep rupture combined with mechanical strength degradation and hydride reorientation. Hydride reorientation from the circumferential to radial direction may reduce the critical stress intensity that accelerates radial crack propagation. The results of cladding creep tests at $400^{\circ}C$ and 130MPa hoop stress performed in this study indicate that hydride reorientation may occur between 2.6% to 7.0% strain in tube diameter with a hydrogen content range of 40-120ppm. Therefore, it is concluded that hydride re-orientation behaviour is strongly correlated with the cladding creep-induced strain, which varies as functions of temperature and stress acting on the cladding.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2017년도 추계학술논문요약집
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• pp.115-116
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• 2017

For the safe transportation of SNF and licensing, the integrity of SNF should be evaluated carefully. Researches to obtain the data for SNF cladding properties, i.e. impact toughness, DBTT (hydride behavior) when evaluating transportation of SNF, shall be precisely implemented by simulating the condition of real SNF to the hilt, accordingly.

• Corrosion Science and Technology
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• 제11권6호
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• pp.213-217
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• 2012

1) 선진 핵연료주기시설 관련 규제기술과 관련하여 인허가 안전심사의 경험이 없으며, 선진 핵연료주기시설 인허가를 위한 규제체계 및 안전성 평가방법 등의 개발이 필요한 단계이며 관련기기와 제반 공정에서 재료의 내식성을 평가하는 기준마련이 시급하다. 2) 선진 핵주기시설 관련 국내 기술수준을 분석하였고 그 핵심 공정인 전해환원, 전해정련, 전해제련공정의 실험변수를 조사하고 평가 필요항목을 정리하였다. 3) 전해환원과 전해정련공정의 경우 Hot-cell 내에 수분 및 산소가 일정 수준 이하로 유지되는 경우, 재료의 부식은 고려하지 않아도 되나 우라늄 잉곳 제조 공정에서 수냉 코일을 사용하게 되는 경우 물에 의한 부식을 고려해야 한다. 4) 전해 제련공정의 경우 LCC, RAR, Cd 증류공정에서 플랜지의 O-ring을 보호하기 위해 수냉 코일을 사용하게 되는 경우 물에 의한 부식을 고려해야 한다.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.421-426
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• 2013

The need for extended storage of used nuclear fuel is increasing globally as disposition schedules for used fuel are pushed further into the future. This is creating a situation where dry storage of used fuel may need to be extended beyond normal regulatory licensing periods. While it is generally accepted that used fuel in dry storage will remain in a safe condition, there is little data that demonstrate used fuel performance in dry storage environments for long periods of time. This is especially true for high burnup used fuel. This paper discusses a technical approach that defines a process that develops the technical basis for demonstrating the safety of used fuel over extended periods of time.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.306-312
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• 2018

There is an increasing need in the United States and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations or interim storage sites. Under normal conditions, the Nuclear Regulatory Commission limits cladding temperature to $400^{\circ}C$ for high-burnup (>45 GWd/mtU) fuel, with higher temperatures allowed for low-burnup fuel. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at $400^{\circ}C$. Results were representative of the majority of US light water reactor fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

• 방사성폐기물학회지
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• 제19권4호
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• pp.435-446
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• 2021

This article examines the consequences of a significant spent fuel management decision or event in the United States, South Korea and Taiwan. For the United States, it is the financial impact of the Department of Energy's inability to take possession of spent fuel from commercial nuclear power companies beginning in 1998 as directed by Congress. For South Korea, it is the potential financial and socioeconomic impact of the successful construction, licensing and operation of a low and intermediate level waste disposal facility on the siting of a spent fuel/high level waste repository. For Taiwan, it is the operational impact of the Kuosheng 1 reactor running out of space in its spent fuel pool. From these, it draws six broad lessons other countries new to, or preparing for, nuclear energy production might take from these experiences. These include conservative planning, treating the back-end of the fuel cycle holistically and building trust through a step-by-step approach to waste disposal.