• Journal of Powder Materials
• /
• v.15 no.2
• /
• pp.87-94
• /
• 2008

In order to investigate a nitriding process of spent oxide fuel and the subsequent change in thermal properties after nitriding, simulated spent fuel powder was converted into a nitride pellet with simulated fission product elements through a carbothermic reduction process. Nitriding rate of simulated spent fuel was decreased with increasing of the amount of fission products. Contents of Ba and Sr in simulated spent fuel were decreased after the carbothermic reduction process. The thermal conductivity of the nitride pellet was decreased by an addition of fission product element but was higher than that of the oxide fuel containing fission product elements.

• Nuclear Engineering and Technology
• /
• v.35 no.4
• /
• pp.309-317
• /
• 2003

The mass balance of the unit processes of the Advanced spent fuel Conditioning Process was calculated to obtain basic information. Based on this mass balance, the changes in decay heat and radioactivity of the spent fuel due to the metallization in the high temperature molten salt system were estimated. The decay heat and the radioactivity were calculated by using the ORIGEN2 computer code, and the result showed that the decay heat and the radioactivity of the metallized spent fuel ingot were $24.27\%\;and\;24.24\%$, respectively, compared to those of oxide spent fuel.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.1
• /
• pp.77-84
• /
• 2010

Reduction of oxides has been investigated for the volume reduction and recycling of the spent oxide fuel from commercial nuclear power plants. Various oxide reduction methods were proposed and KAERI (Korea Atomic Energy Research Institute) is currently developing an electrochemical reduction process using a LiCl-$Li_2O$ molten salt as a reaction medium. The electrochemical reduction process, the front end of the pyroprocessing, can connect the PWR (Pressurized Water Reactor) oxide fuel cycle to a metal fuel cycle of the sodium cooled fast reactor. This paper summarizes KAERI efforts on the development, improvement, and scale-up of the oxide reduction process.

• Nuclear Engineering and Technology
• /
• v.43 no.4
• /
• pp.317-328
• /
• 2011

Pyroprocessing technology was developed in the beginning for metal fuel treatment in the US in the 1960s. The conventional aqueous process, such as PUREX, is not appropriate for treating metal fuel. Pyroprocessing technology has advantages over the aqueous process: less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, compact equipment, etc. The addition of an oxide reduction process to the pyroprocessing metal fuel treatment enables handling of oxide spent fuel, which draws a potential option for the management of spent fuel from the PWR. In this context, KAERI has been developing pyroprocessing technology to handle the oxide spent fuel since the 1990s. This paper describes the current status of pyroprocessing technology development at KAERI from the head-end process to the waste treatment. A unit process with various scales has been tested to produce the design data associated with the scale up. A performance test of unit processes integration will be conducted at the PRIDE facility, which will be constructed by early 2012. The PRIDE facility incorporates the unit processes all together in a cell with an Ar environment. The purpose of PRIDE is to test the processes for unit process performance, operability by remote equipment, the integrity of the unit processes, process monitoring, Ar environment system operation, and safeguards related activities. The test of PRIDE will be promising for further pyroprocessing technology development.

• Nuclear Engineering and Technology
• /
• v.17 no.3
• /
• pp.185-192
• /
• 1985

The fissile content of PWR spent fuel is higher than that of natural uranium which is normal fuel for CANDU type reactor. Investigated are the concepts of PWR spent fuel utilization in CANDU type reactor to diversify uranium resource and partially to solve storage problems of PWR spent fuel being gradually accumulated. Nuclear characteristics of uranium-plutonium mixed oxide fuel loaded in CANDU type reactor are analysed using the WIMS/D computer code. In this study, analyses are solely carried out upon the current CANDU type reactor design without changingany reactivity control devices.

• Journal of Powder Materials
• /
• v.7 no.3
• /
• pp.123-130
• /
• 2000

The simulated DUPIC fuel provides a convenient way to investigate fuel properties and behaviours such as thermal conductivity, thermal expansion, fission gas release, leaching and so on. Several pellets simulating the composition and microstructure of the DUPIC fuel were fabricated from resintering powder through the OREOX process of the simulated spent fuel pellets, which were prepared from the mixture of stable forms of constituent nuclides. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent fuel was in agreement with the previous studies. The densities and the grain size of simulated DUPIC fuel was pellets are higher than those of simulated spent fuel pellets. Small metallic precipitates and oxide precipitates were observed on matrix grain boundaries.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.11b
• /
• pp.86-91
• /
• 2005

A metal product obtained from an electrolytic reduction process, possesses less volume and radioactivity than those of the unprocessed spent oxide fuels. The chemical composition of the metal product varies according to the process condition. In this work, a basic study was performed to evaluate the chemical forms of the spent oxide fuel components in an electrolytic reduction process with the operation conditions. One of the most important operation conditions is the cell potential applied for the reduction cell. It is expected that $PU_{2}O_3$ is difficult to reduce even though the cell potential is negative enough to reduce the lithium oxide when the activity of $Li_{2}O$ exceeds 0.003. The reduction of actinide oxides via the reduction of $Li_{2}O$ is assumed to have a greater reduction yield than a direct reduction of the actinide oxides.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.19 no.2
• /
• pp.255-266
• /
• 2021

Solid-state mechanochemical reduction combined with subsequent melting consolidation was suggested as a technical option for the oxide reduction in pyroprocessing. Ni ingot was produced from NiO as a starting material through this technique while Li metal was used as a reducing agent. To determine the technical feasibility of this approach for pyroprocessing, which handles spent nuclear fuels, thermodynamic calculations of the phase stabilities of various metal oxides of U and other fission elements were made when several alkaline and alkali-earth metals were used as reducing agents. This technique is expected to be beneficial, not only for oxide reduction but also for other unit processes involved in pyroprocessing.

• Journal of Energy Engineering
• /
• v.13 no.4
• /
• pp.259-266
• /
• 2004

Specific heat mechanism of oxide fuel is contributed by lattice vibration, dilatation, conduction electron and defect and excess specific heat. Model of oxide fuel for specific heat consists of specific heat at constant pressure term, dilatation specific heat term and defect specific heat term. In this study experimental and published data on the specific heats of oxide nuclear fuels have been reviewed and analyzed to recommend the best fitting model. The oxide fuels considered in this paper were UO$_2$, mixed (U, Pu) oxides and spent fuel. The specific heat data of spent fuel has been replaced by that of simulated fuel.