NFCSim: A Dynamic Fuel Burnup and Fuel Cycle Simulation Tool

2005 ◽  
Vol 151 (1) ◽  
pp. 35-50 ◽  
Author(s):  
Erich A. Schneider ◽  
Charles G. Bathke ◽  
Michael R. James
Keyword(s):  
Fuel Cycle ◽  
Fuel Burnup ◽  
Simulation Tool ◽  
Cycle Simulation

Neutronic predictors for PWR fuelled with multi-recycled plutonium and applications with the fuel cycle simulation tool CLASS

2017 ◽  
Vol 100 ◽  
pp. 33-47 ◽  
Author(s):  
Fanny Courtin ◽  
Baptiste Leniau ◽  
Nicolas Thiollière ◽  
Baptiste Mouginot ◽  
Xavier Doligez ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Simulation Tool ◽  
Cycle Simulation

Characterizing the United States Nuclear Used Fuel Inventory Using the Cyclus Fuel Cycle Simulator

2016 ◽  
Author(s):  
Cem Bagdatlioglu ◽  
Robert Flanagan ◽  
Erich Schneider
Keyword(s):  
United States ◽  
Reference Data ◽  
Fuel Cycle ◽  
Nuclear Reactors ◽  
The United States ◽  
Energy Information Administration ◽  
Reactor Modeling ◽  
Cycle Simulation ◽  
The Us ◽  
Energy Information

The used fuel inventory of the United States commercial nuclear fleet has been accumulating since the inception of nuclear reactors. In order to understand the mass and composition of the used fuel inventory, a nuclear fuel cycle simulation package (Cyclus) is used with a reactor modeling tool (Bright-lite). The parameters for the simulation are obtained as historical operation and burnup data for every reactor in the US fleet, taken from the U.S. Energy Information Administration. The historical burnup data is used to calculate the fuel enrichment of every reactor at every refueling. Discharged uranium inventories calculated by the software are shown to closely match the reference data. The total mass of three major actinide groups are presented as they build up over time. In addition, the evolution of the plutonium composition in discharged fuel is also presented, illustrating Cyclus’ ability to track the composition of material flowing through a large, evolving reactor fleet over decades.

A Study of Transverse Buckling Effect on the Characteristics of Nuclides Burnup Wave in a Fast Neutron Multiplying Media

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Manish Kumar ◽  
Om Pal Singh
Keyword(s):  
Fast Neutron ◽  
Fuel Cycle ◽  
Fuel Burnup ◽  
Full Width ◽  
Half Maximum ◽  
Transient Time ◽  
Velocity Of Propagation ◽  
Neutron Leakage ◽  
Uranium Plutonium ◽  
Plutonium Fuel

A study of transverse buckling effect on the characteristics of nuclides burnup wave in multiplying media (cylindrical geometry) has been carried out. The burnup wave is characterized in terms of velocity of propagation, transient length (TL), and transient time (TT) in establishing the burnup wave and full width at half maximum (FWHM) in the established region of the wave. The uranium–plutonium fuel cycle is considered. The sensitivity of the results is studied for different radial buckling led leakage of neutrons. It is discovered that the velocity of the wave increases with the increase in the radius of the cylinder (i.e., reduction in the transverse buckling and hence increase in radial neutron leakage). FWHM is relatively insensitive to radial neutron leakage. The transient time and transient length are very large for smaller radius; these decrease with the increase in radius. The study provides insight on the build-up of burnup wave in the neutron multiplying media and brings out the importance of transverse buckling led radial neutron leakage on the characteristics of fuel burnup wave in multiplying media.

Possibility of Different Fuel Cycles Usage in GT-MHR

2002 ◽  
Author(s):  
N. Kodochigov ◽  
Yu. Sukharev ◽  
E. Marova ◽  
N. Ponomarev-Stepnoy ◽  
E. Glushkov ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Block Design ◽  
Average Power ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Fuel Load ◽  
Fuel Cycles ◽  
Partial Cycle ◽  
Control Rods ◽  
Neutronic Characteristics

The GT-MHR reactor core is characterized by flexibility of neutronic characteristics at the given average power density and fixed geometrical dimensions of reactor core. Such flexibility makes it possible to start the reactor operation with one fuel cycle, and then to turn to another type of core fuel load without changes of main reactor elements: fuel block design, core and reflector size, control rod number etc. Preliminary analysis reindicates the commercial viability of the GT-MHR, part of which is due to the ability to accommodate different fuel types and cycles. This paper presents the results of studies of the neutronic characteristics of reactor cores using different fuel (low- and high-enriched uranium, MOX fuel). Comparison of different fuel cycles is carried out for a three-batch refueling option with respect to following characteristics: discharged fuel burnup, reactivity change during one partial cycle of fuel burnup, consumption of fissile isotopes per unit of produced energy, power distribution, reactivity effects, control rods worth. It is shown, that the considered options of fuel loads provide the three-year fuel campaign (with accounting of capacity factor ∼ 0,8) without change of core design, number and design of control rods at transition from the one fuel type to another.

Neutronic Analysis and Fuel Cycle Simulation of the MIT Reactor Using MCODE-FM and Experimental Validation

2013 ◽  
Author(s):  
Kaichao Sun ◽  
Michael Ames ◽  
Thomas Newton ◽  
Lin-wen Hu
Keyword(s):  
Fuel Cycle ◽  
Search Algorithm ◽  
Reactor Core ◽  
Axial Position ◽  
Monte Carlo Code ◽  
Massachusetts Institute Of Technology ◽  
Fuel Management ◽  
Cycle Simulation ◽  
Start Up ◽  
Institute Of Technology

A neutronic analysis of the Massachusetts Institute of Technology Research Reactor (MITR) is performed using state-of-the-art computational tools: the continuous-energy Monte Carlo code MCNP5 and the point-depletion code ORIGEN2.2. These codes are externally coupled by the in-house code package, MCODE (MCNP-ORIGEN Coupled Depletion Program), more recently, it being extended to MCODE-FM (Fuel Management). The latter features automated input file generation, data manipulation, and post-processing of the output data for the fuel cycle analysis, so that it is used to simulate the fuel management of the MITR. MCODE-FM also has an optional criticality search algorithm to simulate control blade movement. The code validation is carried out by comparing the calculated results to experimental data. Two sets of the comparisons are made in the present paper: 1) the Xe-135 reactivity effect during the reactor start-up and shutdown and 2) the thermal and fast neutron flux in an irradiation capsule in the reactor core. Good agreements have been found. The validated MCODE-FM is therefore useful for neutronic analysis and the fuel cycle simulation of the MITR. The time dependent variation of the key parameters, viz. the control blades’ axial position (maintaining criticality) and the fissile inventory in the fuel, is presented.

Physical model of the nuclear fuel cycle simulation code SITON

2017 ◽  
Vol 99 ◽  
pp. 471-483 ◽  
Author(s):  
Á. Brolly ◽  
M. Halász ◽  
M. Szieberth ◽  
L. Nagy ◽  
S. Fehér
Keyword(s):  
Physical Model ◽  
Nuclear Fuel ◽  
Fuel Cycle ◽  
Nuclear Fuel Cycle ◽  
Simulation Code ◽  
Cycle Simulation

VEGAS: A Fuel Cycle Simulation and Preconditioner Tool with Restricted Material Balances

2016 ◽  
Vol 193 (3) ◽  
pp. 416-429 ◽  
Author(s):  
E. A. Schneider ◽  
U. B. Phathanapirom
Keyword(s):  
Fuel Cycle ◽  
Cycle Simulation

The Potential of Different Concepts of Fast Breeder Reactor for the French Fleet Renewal

2006 ◽  
Author(s):  
Simone Massara ◽  
Philippe Tetart ◽  
David Lecarpentier ◽  
Claude Garzenne ◽  
Alexandre Mourogov
Keyword(s):  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Natural Uranium ◽  
Fast Breeder Reactor ◽  
Simulation Code ◽  
Cycle Simulation ◽  
Waste Production ◽  
Highly Sensitive ◽  
Xxist Century ◽  
Transition Scenario

The performances of different concepts of Fast Breeder Reactor (Na-cooled, He-cooled and Pb-cooled FBR) for the current French fleet renewal are analyzed in the framework of a transition scenario to a 100% FBR fleet at the end of the XXIst century. Firstly, the modeling of these three FBR types by means of a semi-analytical approach in TIRELIRE - STRATEGIE, the EDF fuel cycle simulation code, is presented, together with some validation elements against ERANOS, the French reference code system for neutronic FBR analysis (CEA). Afterwards, performances comparisons are made in terms of maximum deployable power, natural uranium consumption and waste production. The results show that the FBR maximum deployable capacity, independently from the FBR technology, is highly sensitive to the fuel cycle options, like the spent nuclear fuel cooling time or the Minor Actinides management strategy. Thus, some of the key parameters defining the dynamic of FBR deployment are highlighted, to inform the orientation of R&D in the development and optimization of these systems.

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