• Nuclear Engineering and Technology
• /
• v.46 no.2
• /
• pp.147-158
• /
• 2014

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

• Nuclear Engineering and Technology
• /
• v.38 no.5
• /
• pp.417-422
• /
• 2006

The present paper describes the 45 MW High Flux Reactor (HFR) which is located in Petten, The Netherlands. This paper focuses on selected technical aspects of this reactor and on nuclear fuel irradiation experiments. These fuel experiments are mainly experiments on Light Water Reactor (LWR) and Very/High Temperature Reactor (V/HTR) fuels, but also on Fast Reactor (FR) fuels, transmutation fuels and Material Test Reactor (MTR) fuels.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.19 no.1
• /
• pp.61-74
• /
• 2021

A paradigm shift in the government's energy policy was reflected in its declaration of early closure of old nuclear plants as well as cancellation of plans for the construction of new plants. To this end, unit 1 of Kori Nuclear Power Plant was permanently shut down and is set for decommission. Based on these changes, the off-site transport of spent fuels from nuclear power plants has become a critical issue. The purpose of this study is to develop an optimized method for transportation of spent fuels from Kori Nuclear Power Plant's units 1, 2, 3, and 4 to an assumed interim storage facility by simulating the scenarios using the Flexsim software, which is widely used in logistics and manufacturing applications. The results of the simulation suggest that the optimized transport methods may contribute to the development of delivery schedule of spent fuels in the near future. Furthermore, these methods can be applied to decommissioning plan of nuclear power plants.

• Nuclear Engineering and Technology
• /
• v.52 no.4
• /
• pp.689-699
• /
• 2020

In this work, new multi-recycling options of TRU nuclides using PWR fuel assemblies comprised of MOX and FCM (Fully Ceramic Micro Encapsulated) fuels are suggested and neutronically analyzed. These options do not use a fully recycling of TRU but a partial recycling where TRUs from MOX fuels are recycled while the ones from FCM fuels are not recycled due to their high consumption rate resulted from high burnup. In particular, additional external TRU feed in MOX fuels for each cycle was considered to significantly increase the TRU consumption rate and the finally selected option is to use external TRU and enriched uranium feed as a makeup for the heavy metal consumption in MOX fuels. This hybrid external feeding of TRU and enriched uranium in MOX fuel was shown to be very effective in significantly increasing TRU consumption rate, maintaining long cycle length, and achieving negative void reactivity worth during recycling.

• Nuclear Engineering and Technology
• /
• v.47 no.3
• /
• pp.315-322
• /
• 2015

The research carried out on thorium-based fuels indicates that these fuels can be considered as economic alternatives with improved physical properties and proliferation resistance issues. In the current study, neutronic assessment of $UO_2$ in comparison with two $(Th-^{233}U)O_2$, and $(Th-^{235}U)O_2$ thorium-based fuel loads in a heavy water research reactor has been proposed. The obtained computational data showed both thorium-based fuels caused less negative temperature reactivity coefficients for the modeled research reactor in comparison with $UO_2$ fuel loading. By contrast, $^{235}U$-containing thorium-based fuel and $^{235}U$-containing thorium-based fuel loadings in the thermal core did not drastically reduce the effective delayed neutron fractions and delayed neutron fractions compared to $UO_2$ fuel. A provided higher conversion factor and lower transuranic production in the research core fed by the thorium-based fuels make the fuel favorable in achieving higher cycle length and less dangerous and costly nuclear disposals.

• Nuclear Engineering and Technology
• /
• v.45 no.3
• /
• pp.377-384
• /
• 2013

During the last three decades, South Korean nuclear power plants have discharged about 5,950 tons of spent fuel and the maximum burn-up reached 55 GWd/MTU in 2002. This study was performed to support the development of Korean dry spent fuel storage alternatives. First, we chose V5H-$17{\times}17$ and KSFA-$16{\times}16$ as representative domestic spent fuels, considering current accumulation and the future generation of the spent fuels. Examination reveals that their average burn-ups have already increased from 33 to 51 GWd/MTU and from 34.8 to 48.5 GWd/MTU, respectively. Evaluation of the fuel characteristics shows that at the average burn-up of 42 GWd/MTU, the oxide thickness, hydrogen content, and hoop stress ranged from $30{\sim}60{\mu}m$, 250 ~ 500 ppm, and 50 ~ 75 MPa, respectively. But when burn-up exceeds 55 GWd/MTU, those characteristics can increase up to 100 ${\mu}m$, 800 ppm, and 120 MPa, respectively, depending on the power history. These results demonstrate that most Korean spent nuclear fuels are expected to remain within safe bounds during long-term dry storage, however, the excessive hoop stress and hydrogen concentration may trigger the degradation of the spent fuel integrity early during the long-term dry storage in the case of high burn-up spent fuels exceeding 45 GWd/MTU.

• Nuclear Engineering and Technology
• /
• v.50 no.5
• /
• pp.794-800
• /
• 2018

The proliferation resistance (PR) of Th/U and Th/Pu fuels used in CANada Deuterium Uranium for the deep geological repository was assessed by combining the multiattribute utility analysis proposed by Chirayath et al., 2015 with the transport model of radionuclides in the repository and comparing with that of the used natural U fuel case. It was found that there was no significant advantage for Th/U and Th/Pu fuels from the viewpoint of the PR in the repository. It was also found that the PR values for used nuclear fuels in the repository of Th/U, Th/Pu, and natural U was comparable with those for enrichment and reprocessing facilities in the pressurized water reactor (PWR) nuclear fuel cycle. On the other hand, the PR values considering the transport of radionuclides in the repository were found to be slightly smaller than those without their transport after the used nuclear fuels started dissolving after 1,000 years.

• Nuclear Engineering and Technology
• /
• v.52 no.6
• /
• pp.1208-1212
• /
• 2020

Knowledge on fuel enthalpy and its temperature derivative, the heat capacity, are important quantities in determination of fuel behavior in normal reactor operation and reactor transients. The aim of this study is to compare the heat capacity of oxide and nitrite fuels by using Einstein-Debye approximation. A simple analytical expression was performed to calculate the heat capacity of fuels. To test the validity and reliability, the calculated formulas were compared to published results for various nuclear fuels including UO₂, ThO₂, PuO₂ and UN. Calculated formulas yielded results in consistent with literature.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.02a
• /
• pp.191-204
• /
• 2004

The research and development of effective management technologies of the spent fuels discharged from power reactors are an important and essential task of KAERI. In resent several years KAERI has focused on a project named "development and demonstration of the Advanced spent fuel Conditioning Process (ACP) in a laboratory scale." The Facility for ACP demonstration consists of two Hot Cells and auxiliary facilities. It is now in the final design stage and will be constructed in 2004. After construction of the facility the ACP equipments will be installed in Hot Cells. The ACP will be demonstrated by some simulated spent fuels first and then by spent fuels.

• Nuclear Engineering and Technology
• /
• v.41 no.4
• /
• pp.413-426
• /
• 2009

Nuclear energy plays an important role in world energy production by supplying 6% of the world's current total electricity production. However, 86% of the energy consumed worldwide to produce industrial process heat, to generate electricity and to power the transportation sector still originates in fossil fuels. To cope with dwindling fossil fuels and climate change, it is clear that a clean alternative energy that can replace fossil fuels in these sectors is urgently required. Clean hydrogen energy is one such alternative. Clean hydrogen can play an important role not only in synthetic fuel production but also through powering fuel cells in the anticipated hydrogen economy. With the introduction of the high temperature gas-cooled reactor (HTGR) that can produce nuclear heat up to $950^{\circ}C$ without greenhouse gas emissions, nuclear power is poised to broaden its mission beyond electricity generation to the provision of nuclear process heat and the massive production of hydrogen. In this paper, the features and potential of the HTGR as the energy source of the future are addressed. Perspectives on nuclear heat and hydrogen applications using the HTGR are discussed.