Advanced MOX and IMF Fuel Assembly Designs for Multi-Recycling Transuranics

2010 ◽  
Author(s):  
Yunhuang Zhang ◽  
Jean C. Ragusa
Keyword(s):  
Fuel Assembly ◽  
Spent Nuclear Fuel ◽  
Fuel Pellet ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Mixed Oxide Fuel ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
Enriched Uranium

Several new fuel assembly designs for multi-recycling Transuranics from spent nuclear fuel are proposed and investigated. Among these are (1) Mixed Oxide Fuel with Enriched Uranium (MOX-EU), in which Plutonium oxide and U-235 enriched Uranium oxide are mixed (2) MOX fuel with Americium coating, in which a thin layer of Americium is applied to the outer surface of the MOX fuel pellet, and (3) an heterogeneous fuel assembly consists of Inert-Matrix Fuel (IMF) pins at the periphery and UOX pins in the inner zone. All these designs are compatible with standard PWR utilizing 17×17 fuel assemblies. In-reactor fuel depletion simulation and long-term isotopic decay calculation are carried out using DRAGON[1] and ORIGEN[2], separately. Transuranics mass balance and long-term radiotoxicity analyses are implemented and the results are normalized to per 1TWh-electricity produced.

scholarly journals Conceptual Design of a MEDE Treatment System for Sodium Bonded Fuel

2008 ◽  
Author(s):  
Carl E. Baily ◽  
Karen A. Moore ◽  
Collin J. Knight ◽  
Peter B. Wells ◽  
Paul J. Petersen ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Department Of Energy ◽  
Test Facility ◽  
Evaporation Chamber ◽  
Oak Ridge ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Metal Fuel ◽  
Enriched Uranium

Unirradiated sodium bonded metal fuel and casting scrap material containing highly enriched uranium (HEU) is stored at the Materials and Fuels Complex (MFC) on the Idaho National Laboratory (INL). This material, which includes intact fuel assemblies and elements from the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor-II (EBR-II) reactors, as well as scrap material from the casting of these fuels, has no current use under the terminated reactor programs for both facilities. The Department of Energy (DOE), under the Sodium-Bonded Spent Nuclear Fuel Treatment Record of Decision (ROD), has determined that this material could be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for commercial nuclear reactors. A plan is being developed to prepare, package, and transfer this material to the DOE HEU Disposition Program Office (HDPO), located at the Y-12 National Security Complex in Oak Ridge, Tennessee. Disposition of the sodium bonded material will require separating the elemental sodium from the metallic uranium fuel. A sodium distillation process known as MEDE (Melt-Drain-Evaporate), will be used for the separation process. The casting scrap material needs to be sorted to remove any foreign material or fines that are not acceptable to the HDPO program. Once all elements have been cut and loaded into baskets, they are then loaded into an evaporation chamber as the first step in the MEDE process. The chamber will be sealed and the pressure reduced to approximately 200 mtorr. The chamber will then be heated as high as 650 °C, causing the sodium to melt and then vaporize. The vapor phase sodium will be driven into an outlet line where it is condensed and drained into a receiver vessel. Once the evaporation operation is complete, the system is de-energized and returned to atmospheric pressure. This paper describes the MEDE process as well as a general overview of the furnace systems, as necessary, to complete the MEDE process.

scholarly journals Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters

Materials ◽  
2019 ◽  
Vol 12 (3) ◽  
pp. 494
Author(s):  
Alexander Vasiliev ◽  
Jose Herrero ◽  
Marco Pecchia ◽  
Dimitri Rochman ◽  
Hakim Ferroukhi ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Design Parameters ◽  
Fuel Loading ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
National Cooperative ◽  
Criticality Safety

This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, keff, of the canister would comply with the imposed criticality safety criterion.

Long-Term In-Package Spent Fuel Criticality Calculations

2003 ◽  
Author(s):  
Olivier Wantz ◽  
Olivier Smidts ◽  
Alain Dubus ◽  
R. Beauwens
Keyword(s):  
Multiplication Factor ◽  
Spent Fuel ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
Neutron Multiplication

This paper presents MCNP criticality calculations for both UOX and MOX disrupted fuel assemblies canisters systems in the reference Belgian disposal concept and one of its variant. We examine the influence of different parameters (water moderation and geometry alteration) on the neutron multiplication factor, keff. In all the studied cases, the reference concept does not present criticality risks. The variant concept sometimes presents criticality risks. The present results only concern fresh UOX and MOX fuel assemblies. Further developments of this work will include irradiated (UOX and MOX) fuels.

ICONE15-10105 INERT MATRIX FUEL ASSEMBLY AS AN OPTION FOR THE LAGUNA VERDE NPP FUEL RELOADS

2007 ◽  
Vol 2007.15 (0) ◽  
pp. _ICONE1510-_ICONE1510
Author(s):  
Hector Hernandez-Lopez ◽  
Marco A. Lucatero ◽  
Javier Ortiz-Villafuerte
Keyword(s):  
Fuel Assembly ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Laguna Verde

scholarly journals Study on Transmutation of Minor Actinides as Burnable Poison in VVER-1000 Fuel Assembly

2019 ◽  
Vol 2019 ◽  
pp. 1-12
Author(s):  
Vinh Thanh Tran ◽  
Hoai-Nam Tran ◽  
Huu Tiep Nguyen ◽  
Van-Khanh Hoang ◽  
Pham Nhu Viet Ha
Keyword(s):  
Fuel Assembly ◽  
Spent Nuclear Fuel ◽  
Calculation Model ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Reference Case ◽  
Burnable Poison ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Reactivity Insertion

Thermal reactors have been considered as interim solution for transmutation of minor actinides recycled from spent nuclear fuel. Various studies have been performed in recent decades to realize this possibility. This paper presents the neutronic feasibility study on transmutation of minor actinides as burnable poison in the VVER-1000 LEU (low enriched uranium) fuel assembly. The VVER-1000 LEU fuel assembly was modeled using the SRAC code system, and the SRAC calculation model was verified against the MCNP6 calculations and the available published benchmark data. Two models of minor actinide loading in the LEU fuel assembly have been investigated: homogeneous mixing in the UGD (Uranium-Gadolinium) pins and coating a thin layer to the UGD pins. The consequent negative reactivity insertion by minor actinides was compensated by reducing the gadolinium content and boron concentration. The reactivity of the LEU assembly versus burnup and the transmutation of minor actinide nuclides were examined in comparison with the reference case. The results demonstrate that transmutation of minor actinides as burnable poison in the VVER-1000 reactor is feasible as minor actinides could partially replace the functions of gadolinium and boric acid for excess reactivity control.

The irradiation test of inert-matrix fuel in comparison to uranium plutonium mixed oxide fuel at the halden reactor

2001 ◽  
Vol 38 (3-4) ◽  
pp. 309-312 ◽  
Author(s):  
U. Kasemeyer ◽  
Ch. Hellwig ◽  
Y.-W. Lee ◽  
G. Ledergerber ◽  
D.S. Sohn ◽  
...  
Keyword(s):  
Mixed Oxide ◽  
Oxide Fuel ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Mixed Oxide Fuel ◽  
Irradiation Test ◽  
Uranium Plutonium

scholarly journals Study on transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions

2021 ◽  
Vol 9 (4) ◽  
pp. 16-26
Author(s):  
Vinh Thanh Tran ◽  
Thanh Mai Vu ◽  
Van Khanh Hoang ◽  
Viet Ha Pham Nhu
Keyword(s):  
Comparative Analysis ◽  
Fuel Assembly ◽  
Spent Nuclear Fuel ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Burnable Poison ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Actinide Transmutation

The feasibility of transmutation of minor actinides recycled from the spent nuclear fuel in the VVER-1000 LEU (low enriched uranium) fuel assembly as burnable poison was examined in our previous study. However, only the minor actinide vector of the VVER-440 spent fuel was considered. In this paper, various vectors of minor actinides recycled from the spent fuel of VVER-440, PWR-1000, and VVER-1000 reactors were therefore employed in the analysis in order to investigate the minor actinide transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions. The comparative analysis was conducted for the two models of minor actinide loading in the LEU fuel assembly: homogeneous mixing in the UGD (Uranium-Gadolinium) pins and coating a thin layer to the UGD pins. The parameters to be analysed and compared include the reactivity of the LEU fuel assembly versus burnup and the transmutation of minor actinide nuclides when loading different minor actinide vectors into the LEU fuel assembly.

Sign in / Sign up

Export Citation Format

Share Document