Characterization and densification studies on ThO2–UO2 pellets derived from ThO2 and U3O8 powders

Kutty, T.R.G.; Hegde, P.V.; Khan, K.B.; Jarvis, Tom; Sengupta, Arijit; Majumdar, S.; Kamath, H.S.

doi:10.1016/j.jnucmat.2004.08.003

Cited by 23 publications

(11 citation statements)

References 45 publications

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“…According to them, U 3 O 8 phase was precipitated in the solid solution phase at higher U 3 O 8 concentrations, but a single phase region of the cubic solid solution was observed to exist at lower U 3 O 8 concentrations. Kutty et al [42,43] have demonstrated that the addition of small quantities of U 3 O 8 to ThO 2 enhances sintering, resulting in formation of high quality ThO 2 -UO 2 pellets without the use of conventional dopants such as CaO and Nb 2 O 5 , which also causes to reduce the impurity level in the pellets. The significance of U 3 O 8 addition for enhancing sintering, especially in UO 2 , has been discussed by many researchers [42][43][44].…”

Section: Discussionmentioning

confidence: 99%

“…Kutty et al [42,43] have demonstrated that the addition of small quantities of U 3 O 8 to ThO 2 enhances sintering, resulting in formation of high quality ThO 2 -UO 2 pellets without the use of conventional dopants such as CaO and Nb 2 O 5 , which also causes to reduce the impurity level in the pellets. The significance of U 3 O 8 addition for enhancing sintering, especially in UO 2 , has been discussed by many researchers [42][43][44]. Chevrel et al [44] indicated that the composition of UO 2.25 appeared to be the most appropriate for the low temperature sintering.…”

Section: Discussionmentioning

confidence: 99%

“…This is predominantly due to the increased concentration of uranium vacancies in UO 2+x . It has been established that U 3 O 8 acts as a sintering aid for ThO 2 [42,43]. The O/U values of U 3 O 8Àz oxide in air at 1000, 1100 and 1200 o C are calculated 2.637, 2.633 and 2.627, respectively [43].…”

Section: Sinteringmentioning

confidence: 99%

“…It has been established that U 3 O 8 acts as a sintering aid for ThO 2 [42,43]. The O/U values of U 3 O 8Àz oxide in air at 1000, 1100 and 1200 o C are calculated 2.637, 2.633 and 2.627, respectively [43]. This means that on increasing the temperature, there may be a considerable amount of deviation from stoichiometry generating more point defects.…”

Section: Sinteringmentioning

confidence: 99%

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Characterization of ThO2–UO2 pellets made by co-precipitation process

Kutty

Khan

Achuthan

et al. 2009

Journal of Nuclear Materials

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Section: Discussionmentioning

confidence: 99%

Section: Discussionmentioning

confidence: 99%

Section: Sinteringmentioning

confidence: 99%

Section: Sinteringmentioning

confidence: 99%

See 2 more Smart Citations

Characterization of ThO2–UO2 pellets made by co-precipitation process

Kutty

Khan

Achuthan

et al. 2009

Journal of Nuclear Materials

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“…The fuel rods are filled with ThO 2 and UO 2 mixed by four different mixing methods to compare powers provided in the blanket and to supply a quasi-constant fission power density. In another study by [17], it was investigated to develop mixed thorium-uranium dioxide (ThO 2 -UO 2 ) fuel by powder metallurgy technique using ThO 2 and U 3 O 8 powders as the starting material.…”

Section: Introductionmentioning

confidence: 99%

Analysis of temperature distribution, burn-up and breeding parameters in nuclear fuel rod in fusion-fission reactor system fueled with mixed ThO2-UO2 fuel

Ipek

2010

Int. J. Energy Res.

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SUMMARYTemperature distribution in nuclear fuel rod, burn-up (BU), possibility of nuclear fuel rejuvenation and breeding parameters are investigated for different coolants under various first wall loads (Pw 5 2, 5 and 7 MW m À2 ) in a deuterium-tritium driven fusion-fission reactor system, fueled with mixed UO 2 -ThO 2 fuel. The fuel mixture is considered to be mixed with various mixture fractions.A (D-T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The plasma chamber dimension, DR, with a line fusion neutron source is 300 cm. In the tritium breeding zone of the blanket Li 2 O is used and blanket is reflected by graphite for neutron economy. The fissile fuel zone is considered to be cooled with three different coolants [gas (He, CO 2 ), Flibe (Li 2 BeF 4 ) and natural lithium (Li)] with the volumetric ratio of coolant-to-fuel [e 5 (V coolant /V fuel ) 5 1:1 (45.515% coolant, 45.515% fuel and 8.971% clad)] for nuclear heat transfer. The behavior of the blankets fueled with mixed fuel mentioned above are observed for 48 months for discrete time intervals of Dt 5 15 days and by a plant factor of 75%. The fissile fuel zone, containing 10 fuel rod rows in the radial direction, covers the cylindrical fusion plasma chamber. The fissile fuel breeding occurs through the neutron capture reaction in the 232 Th (ThO 2 ), in the 238 U (UO 2 ) isotopes. As a result of the calculations, fuel mixtures having the best performance parameters have been obtained as follows for different coolants and under the 7 MW m À2 first wall load during operation period without reaching the melting point:The blanket fueled with 70 wt% UO 2 À30 wt% ThO 2 for Flibe (Li 2 BeF 4 ) coolant The blanket fueled with 80 wt% UO 2 À20 wt% ThO 2 for Gas (He, CO 2 ) coolant The blanket fueled with 90 wt% UO 2 À10 wt% ThO 2 for natural lithium (Li) coolantClad surface temperature T c controlled by coolant velocity is taken between certain operation temperatures (T c 5 200, 300, 400, 500 and 6001C). The best Cumulative Fissile Fuel Enrichment (CFFE) grade of the nuclear fuel calculated as the sum of the isotopic ratios of all fissile materials ( Pu) is obtained in Flibe (Li 2 BeF 4 ) coolant blankets, followed by Gas (He, CO 2 ), whereas natural lithium (Li) coolant shows a poor rejuvenation performance in all fuels. CFFE reaches the maximum value (CFFE 5 10.6200%) in blanket cooled with Li 2 BeF 4 coolant, followed by gas coolant with 9.0319% and natural lithium coolant with 7.9863% without reaching the melting point (T max 5 T m 4 28301C) of the fuel materials. At this point, the maximum temperature in centerline of the fuel rod (T m ) reach to 2727.571C in blanket cooled with Li 2 BeF 4 coolant. However, in the blanket fueled with pure UO 2 fuel and cooled with Flibe, CFFE and Centerline Temperature of the fuel rods (T m ) are reached to 9.7277% and 2811.131C respectively at the end of 39 months. In the blanket fueled with pure UO 2 and cooled with gas, CFFE and Centerline Temperature of the fuel rods (T m ) are reached to 7...

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