Fuel development for gas-cooled fast reactors

Meyer, M. K.; Fielding, Randall; Gan, Jian

doi:10.1016/j.jnucmat.2007.05.013

Cited by 103 publications

(43 citation statements)

References 11 publications

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“…[19,20] As either the hydrogen or carbon contents increase the equilibrium shift results in increased formation of methane and therefore higher removal of carbon. Table 2 shows the mole fractions of gaseous species predicted using FactSage, from this it can be seen that HCN is the third highest mole fraction of gas after the H 2 and N 2 atmosphere with the much less stable tautomer of HCN (HNC) and monoatomic hydrogen predicted to form before CH 4 . This higher rate of nitridation seen in samples with higher carbon content would arise from increased HCN formation.…”

Section: Powder Synthesis 311 Chemical Compositionsmentioning

confidence: 99%

“…Fuel materials used in the Gas Cooled Fast Reactor (GFR) nuclear power plant will need to withstand higher temperatures (in the region of 1273-1873K) [4], harsher radiation fields due to the hardened neutron spectrum and possess increased thermophysical properties to improve efficiency of heat transfer in the reactor. Non-oxide fuels contain fewer lighter elements per heavy metal, i.e.…”

Section: Introductionmentioning

confidence: 99%

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On the fabrication of ZrC x N y from ZrO 2 via two-step carbothermic reduction–nitridation

Harrison

Rapaud

Pradeilles

et al. 2015

Journal of the European Ceramic Society

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This study examined the processing parameters of a two-step carbothermal reduction-nitridation of ZrO 2 as a UO 2 /PuO 2 simulant to fabricate ZrN. Variables studied were the reaction time, reaction temperature and the starting powder composition Higher temperatures, longer reaction times and higher initial amounts of carbon in the starting material lead to increased nitridation similar to previous work with uranium oxide but that reaction rate decreases with time. SEM and TEM of the reacted powders reveal regions of ZrC with ledges indicative of crystal defects with small particles of ZrN growing on the ZrC bulk material suggesting a reaction that occurs via Zr transport to the reaction site either through mass diffusion or evaporation-condensation of Zr (g) and with N 2(g) forming the nitride. Annealing experiments on these mixed phases shows formation of a solid solution is rapid compared to the nitridation, where there is no formation of a mixed phase.

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Section: Powder Synthesis 311 Chemical Compositionsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

On the fabrication of ZrC x N y from ZrO 2 via two-step carbothermic reduction–nitridation

Harrison

Rapaud

Pradeilles

et al. 2015

Journal of the European Ceramic Society

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“…Two kinds of fuel system [18] are developed for gas-cooled fast reactor (GFR) in order to increase fuel loading and improve radiation resistivity, which could be applied to PB-FHRs. One is pin-type GFR fuel with refractory cladding material (Fig.…”

Section: Introductionmentioning

confidence: 99%

Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

Zhu

Zou

2016

EPJ Nuclear Sci. Technol.

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Abstract. Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF 2 ) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern) and two kinds of reflector materials (SiC and graphite). This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9 Be(n,2n) reaction and low neutron absorption of 6 Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

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“…Under nuclear reactor operation using uranium-free inert matrix fuel, plutonium will be not newly produced. Also, added fissile plutonium will be almost incinerated [1][2][3][4][5].…”

Section: Introductionmentioning

confidence: 99%

Thermal Conductivity Measurement of Zr-ZrO<sub>2</sub> Simulated Inert Matrix Nuclear Fuel Pellet

Kim

Rhee²,

Oh³

et al. 2013

WJNST

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For an evaluation of a thermal conductivity of Zr + 30 vol% ZrO 2 simulated inert matrix nuclear fuel pellet, a simulated fuel pellet was fabricated using a hot-pressing method at 800˚C in a vacuum and at a 20 MPa load. And several thermophysical properties of the simulated inert matrix fuel pellet were measured and calculated. The thermal diffusivity and linear thermal expansion as a function of temperature of the simulated fuel pellet were measured using a laser flash method and a dilatometry, respectively. Finally, based on the experimental data, the thermal conductivity of the simulated inert matrix fuel pellet was calculated and evaluated.

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Fuel development for gas-cooled fast reactors

Cited by 103 publications

References 11 publications

On the fabrication of ZrC x N y from ZrO 2 via two-step carbothermic reduction–nitridation

On the fabrication of ZrC x N y from ZrO 2 via two-step carbothermic reduction–nitridation

Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

Thermal Conductivity Measurement of Zr-ZrO<sub>2</sub> Simulated Inert Matrix Nuclear Fuel Pellet

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Fuel development for gas-cooled fast reactors

Cited by 103 publications

References 11 publications

On the fabrication of ZrC x N y from ZrO 2 via two-step carbothermic reduction–nitridation

On the fabrication of ZrC x N y from ZrO 2 via two-step carbothermic reduction–nitridation

Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

Thermal Conductivity Measurement of Zr-ZrO&lt;sub&gt;2&lt;/sub&gt; Simulated Inert Matrix Nuclear Fuel Pellet

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Thermal Conductivity Measurement of Zr-ZrO<sub>2</sub> Simulated Inert Matrix Nuclear Fuel Pellet