• 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.

• International Journal of Automotive Technology
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• v.8 no.5
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• pp.533-542
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• 2007

The two stroke spark ignition engine is the greatest contributor of the total vehicular pollution in a country like India. It is therefore an item that requires great attention in order to reduce fuel consumption and its concomitant pollution. The use of strong magnetic charge in the fuel line gives a complete and clean burn so that power is increased while operating expenses are reduced. The magnetic flux on the fuel line dramatically reduces harmful exhaust emissions while increasing mileage, thereby saving money and improving engine performance. It increases combustion efficiency and provides higher-octane performance. The experimental results show that the magnetic flux on fuel reduces the carbon monoxide emission up to 13% in a base engine, 23% in a copper-coated engine and 29% in a zirconia-coated engine.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.191-198
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• 2008

The influence of fission products' contents on the DUPIC fuel powder and pellet properties was experimentally evaluated using SIMFUEL as a surrogate for actual spent PWR fuel due to the high radioactivity of spent fuel. Pure $UO_2$ and SIMFUEL pellets with fission products equivalent to a burn-up of 35,000 MWd/tU and 60,000 MWd/tU were used as impurities in this study. The specific surface area of the powder milled after the OREOX treatment increased and resulted in sintered pellets with a theoretical density (TD) higher than 95%, regardless of the impurity contents. However, the grain size of the sintered pellets decreased with the increasing impurity contents. As a result of the dissolved oxides in $UO_2$ from the impurity groups, the specific surface area of the OREOX powder increased with an increase of the impurities. The grain size of the sintered pellets was significantly decreased by the metallic and oxide precipitates.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.143-148
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• 2009

The increasing power and other demands on PWR fuel is leading to a demand for cladding that has low corrosion but that should also be robust with regard to mechanical behavior, impact of the irradiation environment and the coolant chemistry. The Optimized $ZIRLO^{TM}$ cladding is an evolutionary development of $ZIRLO^{TM}$ taking advantage of the long experience of the ZIRLO cladding but has significantly improved corrosion behavior. Recently, operation of Optimized ZIRLO to above 73 kWd/kgU has shown a reduction of the corrosion of almost 50%.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.565-572
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• 2013

The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade $SiC_f/SiC$ composites are briefly reviewed. A CVI-processed $SiC_f/SiC$ composite with a PyC or $(PyC-SiC)_n$ interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.913-914
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.339-340
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• 2005

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.301-307
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• 2005

The morphology of the high burnup $UO_2$ spent fuel, which was oxidized and annealed in a PIA (Post Irradiation Annealing) apparatus, has been observed. The high burnup fuel irradiated in Ulchin Unit 2, average rod burnup 57,000 MWd/tU, was transported to the KAERI's PIEF. The test specimen was used with about 200 mg of the spent $UO_2$ fuel fragment of the local burnup 65,000 MWd/tU. This specimen was annealed at $1400^{\circ}C$ for 4hrs after the oxidation for 3hrs to grain boundary using the PIA apparatus in a hot-cell. In order to oxidize the grain boundary, the oxidation temperature increased up to $500^{\circ}C$ and held for 3hrs in the mixed gas (60 ml He and 100 ml STD-air) atmosphere. The amount of 85Kr during the whole test process was measured to know the fission gas release behavior using the online system of a beta counter and a gamma counter. The detailed micro-structure was observed by a SEM to confirm the change of the fuel morphology after this test. As the annealing temperature increased, the fission products were observed to move to the grain surface and grain boundary of the $UO_2$ matrix. This specimen was re-structured through the reduction process, and the grain sizes were distributed from 5 to $10\;{\mu}m$.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.576-581
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• 2010

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 ${\mu}m$ was higher than that of the other regions. The grain growth occurred from 10 ${\mu}m$ to 12 ${\mu}m$ in the central region of rod 6 during irradiation.