New Package for the Transport of Fresh MOX Fuel Assemblies in Europe

2000 ◽  
Vol 11 (1-2) ◽  
pp. 101-108
Author(s):  
P. Purcell
Keyword(s):  
Mox Fuel ◽  
Fuel Assemblies

scholarly journals HEXAGONAL PWR-CORE MODELING AND SIMULATION WITH APPLICATION OF NECP-BAMBOO

2021 ◽  
Vol 247 ◽  
pp. 06017
Author(s):  
Cheng Zhang ◽  
Liangzhi Cao ◽  
Yunzhao Li ◽  
Guowei Hua
Keyword(s):  
Monte Carlo ◽  
Modeling And Simulation ◽  
Monte Carlo Code ◽  
Constructive Solid Geometry ◽  
Modeling And Simulations ◽  
Nodal Method ◽  
Solid Geometry ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
1D Model

In this paper, the modeling and simulation of the PWRs loaded with hexagonal fuel assemblies has been implemented with the NECP-Bamboo code. NECP-Bamboo, consisting of a 2D lattice code named Bamboo-Lattice and a 3D steady-state core code named Bamboo-Core, was primitively designed for the PWRs loaded with the rectangular fuel assemblies. As the capability extension for PWRs with hexagonal fuel assemblies, four aspects of improvement have been implemented in NECP-Bamboo. Firstly, the Constructive Solid Geometry (CSG) has been implemented in Bamboo-Lattice for the lattice modeling. Secondly, the explicit modeling of the reflector assembly has been applied to provide more reliable few-group constants, compared with the conventional 1D model for the reflector assembly. Thirdly, the assembly-homogenization capability has been extended to the hexagonal assembly. Fourthly, the diffusion solver in Bamboo-Core based on the Variational Nodal Method (VNM) has been extended to handle hexagonal geometry. With application of the capability-extended NECP-Bamboo, the modeling and simulations for the VVER-1000 benchmark loaded with MOX fuel has been implemented. It can be observed that the numerical results provided by NECP-Bamboo can agree well with corresponding results by the Monte-Carlo code.

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.

Mixed Reload Design Using MOX and UOX Fuel Assemblies

2002 ◽  
Author(s):  
J. Ramo´n Rami´rez Sa´nchez ◽  
R. T. Perry
Keyword(s):  
Conceptual Design ◽  
Cycle Length ◽  
Multiplication Factor ◽  
The Core ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
Current Cycle ◽  
Full Power ◽  
Control Rods ◽  
Mixed Load

As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10×10 BWR fuel assemblies but different fisil material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fisil plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO2 fresh fuel were tested to verify the shutdown margin, the UO2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper.

Analysis of minor actinide recycling using MOX fuel assemblies in a Boiling Water Reactor

2018 ◽  
Vol 120 ◽  
pp. 8-26 ◽  
Author(s):  
Gustavo Alonso ◽  
Eduardo Martinez ◽  
J. Ramón Ramírez ◽  
Rogelio Castillo ◽  
Alejandro Castillo
Keyword(s):  
Boiling Water ◽  
Minor Actinide ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Mox Fuel ◽  
Fuel Assemblies

Transport of Fresh Mox Fuel Assemblies for the Monju Initial Core

1997 ◽  
Vol 8 (2) ◽  
pp. 123-126
Author(s):  
J. Kurakami ◽  
Y. Ouchi ◽  
M. Usami
Keyword(s):  
Mox Fuel ◽  
Fuel Assemblies

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.

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