scholarly journals DISSOLVER FOR NATURAL URANIUM FUEL ELEMENTS

1958 ◽  
Author(s):  
V. P. Caracciolo
Keyword(s):  
Natural Uranium ◽  
Uranium Fuel ◽  
Fuel Elements

scholarly journals Mitigation of End-Flux-Peaking in Fresh CANDU Fuel Bundles Using Neutron Absorbers

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Dylan J. C. Pierce ◽  
Paul K. Chan ◽  
Wei Shen
Keyword(s):  
Three Dimensional ◽  
Computer Code ◽  
Natural Uranium ◽  
Axial Flux ◽  
Fission Rate ◽  
Uranium Fuel ◽  
Fuel Bundle ◽  
Flux Profile ◽  
Thin Disks ◽  
Fuel Elements

Abstract End-flux-peaking (EFP) is a phenomenon where a region of elevated neutron flux occurs between two adjoining fuel bundles, leading to an increase in fission rate and therefore greater heat generation. It is known that the addition of neutron absorbers into fuel bundles can mitigate EFP, yet the implementation in Canada deuterium uranium (CANDU) type reactors using natural uranium fuel has not been pursued. The computer code Monte Carlo N-Particle code (MCNP) 6.1 was used to develop a three-dimensional CANDU bundle–bundle contact model and simulate the addition of neutron absorbers positioned strategically within various locations of the fuel bundle. The burnable absorbers of interest include Gd2O3 and Eu2O3. The locations investigated include within the end pellets of a fuel stack, within the CANDU lubricant (CANLUB) layer, within thin disks located at the ends of the fuel stack, and alloyed in the endplate. Concentrations of the absorbers were varied to gain better insight into their effect on the thermal neutron axial flux profile of the fuel bundle. The results of the study indicated that adding a combination of ∼4 mg/∼12 mg of Eu2O3 into the pellet adjacent to the end pellet and the end pellet, respectively, at each end of all six of the fuel elements in the inner fuel ring, as well as, ∼2 mg/∼6 mg of Eu2O3 into the same respective pellets, at each end of the 18 fuel elements in the outer fuel ring, provides the most effective mitigation of the EFP phenomena in fresh CANDU fuel bundles.

Direct Assay for Molybdenum and Ruthenium in Uranium Alloys by X-Ray Emission Spectrometry

1957 ◽  
Vol 1 ◽  
pp. 387-398 ◽  
Author(s):  
D. S. Flikkema ◽  
R. V. Schablaske
Keyword(s):  
Silicon Carbide ◽  
Analytical Procedure ◽  
Flat Surface ◽  
Emission Lines ◽  
Emission Spectrometry ◽  
Chemical Dissolution ◽  
X Ray ◽  
Direct Assay ◽  
Uranium Fuel ◽  
Fuel Elements

AbstractIt has been found possible to determine quickly the concentrations of molybdenum and ruthenium in non-radioactive alloys representative of high burn-up reactor fuels by the method of X-ray emission spectrometry. Preliminary steps of chemical dissolution and separation are not required. The alloys, essentially ternaries of molybdenum and ruthenium with uranium, are being studied because they are considered to typify the alloys which will result from cycling uranium fuel elements through the sequence of fabrication, use and pyro.metallurgical processing.The analytical procedure involves sampling of the ingot by slicing with a silicon carbide wheel at the plane of interest and reducing the surface to the flatness and finish obtained by a five-minute grinding and polishing operation. In the X-ray spectrograph the flat surface is examined for the intensities of its molybdenum and ruthenium K emission lines, with counting times of one to eight minutes. Calibration plots of intensity versus chemically determined weight per cent are established and used for subsequent sets of analyses.

Achieving High Exposure in Metallic Uranium Fuel Elements

1970 ◽  
Vol 9 (5) ◽  
pp. 673-681 ◽  
Author(s):  
R. D. Leggett ◽  
R. K. Marshall ◽  
C. R. Hann ◽  
C. H. McGilton
Keyword(s):  
High Exposure ◽  
Uranium Fuel ◽  
Fuel Elements ◽  
Metallic Uranium

Neutron-physical and heat-engineering aspects of uranium-graphite reactors loaded with enriched-uranium fuel elements

Atomic Energy ◽  
2006 ◽  
Vol 101 (2) ◽  
pp. 606-610
Author(s):  
Yu. A. Artel’nyi ◽  
P. M. Gavrilov ◽  
A. A. Tsyganov
Keyword(s):  
Heat Engineering ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Fuel Elements

Effects of Thickness of Zirconium Liner on Stress-Strain Characteristics of U10Mo Monolithic Plates

2013 ◽  
Author(s):  
Hakan Ozaltun ◽  
Robert M. Allen ◽  
You Sung Han
Keyword(s):  
Diffusion Barrier ◽  
Reactor Core ◽  
Stress Strain ◽  
Strain Behavior ◽  
Uranium Fuel ◽  
Plate Design ◽  
Enriched Uranium ◽  
Test Reactor ◽  
Irradiation Performance ◽  
Fuel Elements

The effects of the thickness of Zirconium liner on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation processes were studied. Monolithic plate-type fuel elements is a new fuel form being developed for research and test reactors to achieve higher uranium densities which allows the use of low-enriched uranium fuel in reactor core. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. Early RERTR experiments indicated that the presence of an interaction layer between the fuel and cladding materials causes mechanical problems. To minimize the fuel/cladding interaction, employing a diffusion barrier between the cladding and the fuel materials was proposed. Current monolithic plate design employs a 0.025 mm thick, 99.8% pure annealed Zirconium diffusion barrier between the fuel foil (U10Mo) and the cladding materials (AL6061-O). To benchmark the irradiation performance, a number of plates were irradiated in the Advanced Test Reactor (ATR) with promising irradiation performance. To understand the effects of the thickness of the Zirconium diffusion barrier on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate from RERTR-12 experiments (Plate L1P7A0) was selected and simulated. Both fabrication and irradiation stages were considered. Simulations were repeated for various Zirconium thicknesses to understand the effects of the thickness of the diffusion barrier. Results of fabrication simulations indicated that Zirconium thickness has noticeable effects on foil’s stresses. Irradiation simulations revealed that the fabrication stresses of the foil would be relieved rapidly in the reactor. Results also showed that Zirconium thickness has little or no effects on irradiation and shutdown stresses.

Burn up Study of an Innovative Natural Uranium-Thorium Fueled Reprocessing Free Fusion-Fission Hybrid Reactor Blanket with Closed Thorium-Uranium Fuel Cycle

2015 ◽  
Vol 68 (3) ◽  
pp. 566-572 ◽  
Author(s):  
S. C. Xiao ◽  
Jing Zhao ◽  
X. Heng ◽  
X. Y. Sheng ◽  
Z. Zhou ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Natural Uranium ◽  
Hybrid Reactor ◽  
Uranium Fuel

scholarly journals Effects of the Foil Flatness on the Stress-Strain Characteristics of U10Mo Alloy Based Monolithic Mini-Plates

2014 ◽  
Author(s):  
Hakan Ozaltun ◽  
Pavel Medvedev
Keyword(s):  
Mechanical Performance ◽  
Peak Stress ◽  
Stress Strain ◽  
Strain Behavior ◽  
Uranium Fuel ◽  
Considerable Impact ◽  
Enriched Uranium ◽  
Fuel Elements ◽  
Cladding Material ◽  
Plate Type

The effects of the foil flatness on stress-strain behavior of monolithic fuel mini-plates during fabrication and irradiation were studied. Monolithic plate-type fuels are a new fuel form being developed for research and test reactors to achieve higher uranium densities. This concept facilitates the use of low-enriched uranium fuel in the reactor. These fuel elements are comprised of a high density, low enrichment, U–Mo alloy based fuel foil encapsulated in a cladding material made of Aluminum. To evaluate the effects of the foil flatness on the stress-strain behavior of the plates during fabrication, irradiation and shutdown stages, a representative plate from RERTR-12 experiments (Plate L1P756) was considered. Both fabrication and irradiation processes of the plate were simulated by using actual irradiation parameters. The simulations were repeated for various foil curvatures to observe the effects of the foil flatness on the peak stress and strain magnitudes of the fuel elements. Results of fabrication simulations revealed that the flatness of the foil does not have a considerable impact on the post fabrication stress-strain fields. Furthermore, the irradiation simulations indicated that any post-fabrication stresses in the foil would be relieved relatively fast in the reactor. While, the perfectly flat foil provided the slightly better mechanical performance, overall difference between the flat-foil case and curved-foil case was not significant. Even though the peak stresses are less affected, the foil curvature has several implications on the strain magnitudes in the cladding. It was observed that with an increasing foil curvature, there is a slight increase in the cladding strains.

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