U–Mo alloy fuel for TRU-burning advanced fast reactors

Kim, Yeon Soo; Hofman, G.L.; Yacout, Abdellatif M.; Kim, T.K.

doi:10.1016/j.jnucmat.2013.01.324

Cited by 13 publications

(6 citation statements)

References 4 publications

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“…Meeting the conditions of high thermal conductivity, low thermal expansion, and stability under irradiation and corrosion resistance, U-Mo compositions also have low potential for interacting with cladding materials. All these points show the precedence of U-Mo over other examined high-dense metal fuels, for example, based on U-Zr [21]-also considered as an important fuel material for fast-neutron reactors [22]. The results of the theoretical comparison between the phase stability data for U-Mo and U-Zr provided in [23] also indicates the preference of γ -U-Mo due to an ability to remain in a bcc structure and corresponding very weak constituent redistribution in the typical fuel operation temperatures (compared with U-Zr, which undergoes several phase transitions).…”

Section: Introductionmentioning

confidence: 57%

A ternary EAM interatomic potential for U–Mo alloys with xenon

Smirnova

Kuksin

Starikov

et al. 2013

Modelling Simul. Mater. Sci. Eng.

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A new interatomic potential for a uranium-molybdenum system with xenon is developed in the framework of an embedded atom model using a forcematching technique and a dataset of ab initio atomic forces. The verification of the potential proves that it is suitable for the investigation of various compounds existing in the system as well as for simulation of pure elements: U, Mo and Xe. Computed lattice constants, thermal expansion coefficients, elastic properties and melting temperatures of U, Mo and Xe are consistent with the experimentally measured values. The energies of the point defect formation in pure U and Mo are proved to be comparable to the density-functional theory calculations. We compare this new U-Mo-Xe potential with the previously developed U and Mo-Xe potentials. A comparative study between the different potential functions is provided. The key purpose of the new model is to study the atomistic processes of defect evolution taking place in the U-Mo nuclear fuel. Here we use the potential to simulate bcc alloys containing 10 wt% of intermetallic Mo and U 2 Mo.

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Section: Introductionmentioning

confidence: 57%

A ternary EAM interatomic potential for U–Mo alloys with xenon

Smirnova

Kuksin

Starikov

et al. 2013

Modelling Simul. Mater. Sci. Eng.

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“…However, since the γ phase is unstable at low temperatures, alloying is necessary to stabilize the BCC structure. Accordingly, there has been increasing interest in the study and use of uranium in the γ phase alloyed with other transition metals, primarily molybdenum [3,[9][10][11][12].…”

Section: Introductionmentioning

confidence: 99%

Thermal neutron scattering cross sections of ²³⁸U and ²³⁵U in the γ phase

Saltos

Peters

Hammond

2018

J. Phys.: Condens. Matter

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The development of metallic, low-enrichment uranium fuels requires accurate prediction of their neutron transport properties and reactivity parameters, which in turn require thermal neutron scattering data. Accurate prediction of thermal neutron scattering data, including thermal cross sections, requires knowledge of the phonon scattering properties of the medium, but such matrix binding effects in next-generation fuels such as U-Mo, U-Zr, and U-Si are typically neglected because these effects are often difficult to measure or calculate. Using molecular dynamics simulations with previously published interatomic potentials, we calculate the phonon dispersion relations and phonon densities of states for U andU in the α and γ phases. The performance of these potentials was evaluated using published ab initio simulation data and inelastic neutron scattering data. The phonon densities of states obtained by each potential were then utilized to calculate the thermal neutron scattering cross sections of U andU at 1113 K using the NJOY program. The resulting thermal neutron scattering cross sections are assessed by comparison to data obtained from available experimental densities of states. The cross sections generated show how the addition of binding effects decreases the cross section by up to a factor of six over the free-atom model. A definite effect on reactivity is also demonstrated by the use of these thermal libraries on a simple core model. As a consequence, the cross sections generated in this work provide a better description of the true cross section than the free-atom data currently available. We also discuss the sensitivity of the thermal scattering cross sections to the phonon density of states.

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“…There were 4 U-10Mo alloy fuels irradiated in these experiments. Early in metal fuel development for fast reactors, U-Mo alloys were considered [5,6] and, subsequently, have been considered to possibly have favorable properties for high burnup and transmutation applications [28]. The most significant use of U-10Mo fuel was in the Mark II fuel pins of the Dounreay Fast Reactor.…”

Section: U-mo Fuelsmentioning

confidence: 99%

Baseline Postirradiation Examination of the AFC-3C, AFC-3D, and AFC-4A Experiments

Harp

Capriotti

Cappia

2018

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The AFC-3C, AFC-3D, and AFC-4A capsule irradiations were irradiated at the Idaho National Laboratory Advanced Test Reactor. These irradiations were planned to test several different fast reactor fuels that could be used to facilitate ultra-high burnup applications in sodium fast reactors. Several different alloys, fuel geometries, bonding materials, and the use of additives were tested in ferritic-martensitic HT-9 cladding. The AFC-3C and 3D experiment took various different U-Mo, U-Zr, U-Pd-Zr, U-Mo-Ti-Zr, and U-Mo-Ti-Zr-Pd alloys to burnups of 2.1% to 4.5% FIMA (fissions per initial heavy metal atoms) and fission densities of 6.13x10 20 to 1.27x10 21 fissions / cm 3 . These alloys were evaluated against the historical fuel performance of previously irradiated fuel from literature. Fuel performance was irradiated through a suite of postirradiation examination techniques including neutron radiography, gamma spectrometry, dimensional inspection, fission gas release measurements, chemical burnup analysis, and optical microscopy. The irradiations were performed at relatively aggressive temperatures that resulted in significant interaction between the cladding and the fuel in many cases. In spite of this, there were no in-pile cladding breaches in these tests. The U-Zr alloys performed better than the U-Mo or U-Mo-Ti-Zr alloys. Some of the fuel-cladding interaction issues seen in helium bonded annular fuel that was not machined to tight tolerances appear to have been resolved by machining the outer diameter of annular fuel in this irradiation. Higher Zr concentrations in additive bearing fuel appears to have improved the performance of Pd additive fuel.Baseline Postirradiation Examination of the AFC-3C, AFC-3D, and AFC-4A Experiments iv

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U–Mo alloy fuel for TRU-burning advanced fast reactors

Cited by 13 publications

References 4 publications

A ternary EAM interatomic potential for U–Mo alloys with xenon

A ternary EAM interatomic potential for U–Mo alloys with xenon

Thermal neutron scattering cross sections of ²³⁸U and ²³⁵U in the γ phase

Baseline Postirradiation Examination of the AFC-3C, AFC-3D, and AFC-4A Experiments

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U–Mo alloy fuel for TRU-burning advanced fast reactors

Cited by 13 publications

References 4 publications

A ternary EAM interatomic potential for U–Mo alloys with xenon

A ternary EAM interatomic potential for U–Mo alloys with xenon

Thermal neutron scattering cross sections of 238U and 235U in the γ phase

Baseline Postirradiation Examination of the AFC-3C, AFC-3D, and AFC-4A Experiments

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Thermal neutron scattering cross sections of ²³⁸U and ²³⁵U in the γ phase