• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.433-435
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• 2002

The relation between sintered densities and porosities in UO$_2$ pellets is investigated. The open porosity decreases linearly up to about 95% T.D.,(theoretical density) as the sintered density increases whereas, above 96% T.D., sintered UO$_2$ pellets do not have any open pores. The fraction of open porosity to the total porosity also decreases linearly as the sintered density increases, though the slope is lower than that of open porosity and, above 95% T.D., the fraction decreases rapidly to approach a zero.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.493-520
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• 2009

During the last four decades, 16 Pressurized Water Reactors (PWR) and 4 Pressurized Heavy Water Reactors (PHWR) have been constructed and operated in Korea, and nuclear fuel technology has been developed to a self-reliant state. At first, the PWR fuel design and manufacturing technology was acquired through international cooperation with a foreign partner. Then, the PWR fuel R&D by Korea Atomic Energy Research Institute (KAERI) has improved fuel technology to a self-reliant state in terms of fuel elements, which includes a new cladding material, a large-grained $UO_2$ pellet, a high performance spacer grid, a fuel rod performance code, and fuel assembly test facility. The MOX fuel performance analysis code was developed and validated using the in-reactor test data. MOX fuel test rods were fabricated and their irradiation test was completed by an international program. At the same time, the PWR fuel development by Korea Nuclear Fuel (KNF) has produced new fuel assemblies such as PLUS7 and ACE7. During this process, the design and test technology of fuel assemblies was developed to a self-reliant state. The PHWR fuel manufacturing technology was developed and manufacturing facility was set up by KAERI, independently from the foreign technology. Then, the advanced PHWR fuel, CANFLEX(CANDU Flexible Fuelling), was developed, and an irradiation test was completed in a PHWR. The development of the CANFLEX fuel included a new design of fuel rods and bundles.. The nuclear fuel technology in Korea has been steadily developed in many national R&D programs, and this advanced fuel technology is expected to contribute to a worldwide nuclear renaissance that can create solutions to global warming.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.541-548
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• 2017

Uranium metal used for fabrication of fuel for research reactors in India is generally produced by magnesio-thermic reduction of $UF_4$. Performance of magnesio-thermic reaction and recovery and quality of uranium largely depends on properties of $UF_4$. As ammonium diuranate (ADU) is first product in powder form in the process flow-sheet, properties of $UF_4$ depend on properties of ADU. ADU is generally produced from uranyl nitrate solution (UNS) for natural uranium metal production and from uranyl fluoride solution (UFS) for low enriched uranium metal production. In present paper, ADU has been produced via both the routes. Variation of uranium recovery and crystal structure and composition of ADU with progress in precipitation reaction has been studied with special attention on first appearance of the precipitate Further, ADU produced by two routes have been calcined to $UO_3$, then reduced to $UO_2$ and hydroflorinated to $UF_4$. Effect of two different process routes of ADU precipitation on the characteristics of ADU, $UO_3$, $UO_2$ and $UF_4$ were studied here.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2003.05a
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• pp.236-239
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• 2003

The spent fuel slitting device is an equipment developed in order to feed UO$_2$pellet to the dry pulverizing/mixing device. In this study, we have compared and analyzed the handling method of the slitting and that of the pellet and hull, processing time, separating time for 20kgHM, the number of blades, on the existing slitting device using in DUPIC, and spent fuel management technology research and test facility. Also, we have compared and analyzed about an advantage and weak point, designing and producing, processing, establishment, operation, maintenance about the vertical and horizontal slitting device. Based on these results, we have developed the vertical slitting device. By using the results, we have enhanced the slitting processing time(over 40%)in comparison with DUPIC device, and it will is effectively applied to available data for designing and producing of the hot test facility.

• Nuclear Engineering and Technology
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• v.31 no.5
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• pp.455-464
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• 1999

The effect of TiO$_2$ on the sintering behavior of mixed UO$_2$-U$_3$O$_{8}$ Powder compacts has been investigated using the U$_{3}$O$_{8}$ powder made tv oxidation of defective UO$_{2}$ pellets. Without TiO$_2$, UO$_2$ pellet density is inversely proportional to U$_3$O$_{8}$ content and is below 94 %TD in the U$_3$O$_{8}$ range above 15 wt%. Using more than 0.1 wt % TiO$_2$, however, the density decreases slightly with U$_3$O$_{8}$ content and thus is higher than about 94% TD in the whole range of U$_3$O$_{8}$ content. The grain sizes of UO$_2$ pellets with more than 0.1 wt % TiO$_2$are larger than about 30${\mu}{\textrm}{m}$. Therefore, the U$_3$O$_{8}$ Powder can be reused without any restriction on its amount in UO$_2$ pellet fabrication by sintering the mixed UO$_2$-U$_3$O$_{8}$ compact with the aid of TiO$_2$. Mechanisms for densification and grain growth are proposed and discussed, based on a dilatometry study and an examination of microstructure. microstructure.

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

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.344-357
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• 2002

The present paper reports spontaneous steam explosions observed in fuel coolant interaction experiments using prototypic reactor materials. Pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$ are used. A high temperature molten material in the form of a jet is poured into a subcooled water pool located in a pressure vessel. An induction skull melting technique is used for the melting of the reactor material. In both tests using pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$, either a quenching or a spontaneous steam explosion was observed. The morphology of debris and pressure profile clearly indicate the differences between the qunching cases and explosion cases. The dynamic pressure. dynamic impulse, water temperature, melt temperature, and static pressure Inside the containment chamber were measured . As the spontaneous steam explosion for the reactor material is firstly observed in the present experiments, the results of present experiments could be a siginificant step forward the understanding the explosion of the reactor material.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.255-263
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• 2006

Vol-oxidizer is a device to convert $UO_2$ pellets into $U_3O_8$ powder and to feed a homogeneous powder into a Metal Conversion Reactor in the ACP(Advanced Spent Fuel Conditioning Process). In this paper, we propose a design model of the vol-oxidizer, develop the new vol-oxidizer with a capacity of 20 kg HM/batch in $UO_2$ pellets, and conduct a verification for the device. Design considerations include the internal structure, the capacity, the heating position of the device, and the size. The dimensions of the new vol-oxidizer are decided by the design model. We determine a permeability test of the $U_3O_8$ measuring the temperature distribution, and the volume of $UO_2$ and $U_3O_8$. We manufactured the new vol-oxidizer for a 20 kg HM/batch in $UO_2$ pellets, and then analyzed the characteristics of the $U_3O_8$ powder for the verification. The experimental results show that the permeability of the $U_3O_8$ throughout mesh enhance more than old vol-oxidizer, the oxidation time takes only 8 hours when compared with the 13 hours of the old device, and the average distribution of particle size is $40{\mu}m$. The capacities of new vol-oxidizer for a 20 kg HM/batch in $UO_2$ pellets were agree well with the predictions of design model.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.49-55
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• 1991

The microstructure and pore characteristics have been studied on the sintered UO$_2$pellet which was made of the UO$_2$powder manufactured via AUC process. The open porosity decrease with the density and is nearly annihilated above the density of 10.45 g/㎤. The round pore smaller than 3 $\mu$m exist In all densities. The large and elongated pore appears additionally In low density The pore in low density is more elongated than the pore in high density The distribution of the pore area versus the pore size is monomodal and shows its peak on the pore size of 2 to 3 $\mu$m. As the density decreases, the related area of large pore Increases.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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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.