What is the difference between burnup and depletion calculations?

AI Thread Summary
Burnup calculations measure the energy produced per unit mass of nuclear fuel, typically expressed in various units such as MWd/kgU or GWd/tU, while depletion calculations focus on the reduction of fuel enrichment during irradiation. Burnup indicates how much energy has been extracted from the fuel, whereas depletion tracks the changes in isotopic composition, particularly the competition between U-235 and the buildup of plutonium isotopes. In MOX fuel cores, burnup can be expressed in terms of heavy metal content. Core depletion calculations, often performed using codes like SIMULATE, simulate the fission process over an operational cycle. Understanding these differences is crucial for effective nuclear fuel management and reactor operation.
catseye747
Messages
5
Reaction score
0
What is the difference between burnup calculations and depletion calculations?
 
Engineering news on Phys.org
catseye747 said:
What is the difference between burnup calculations and depletion calculations?
They are the same. Burnup refers to the energy produced per unit mass of fuel, usually in MWd/kgU or GWd/tU, although the Canadians like to use MWh/kgU, the Belgians and some others used to use MWd/kgUO2[/sup], and for a long time GE (GNF) used GWd/stU (st = short ton). In MOX cores, the burnup may expressed in GWd/tHM (HM=Heavy Metal, U+Pu). Finally some folks used FIMA, or fissions per initial metal atom, with a rough equivalence of 1% FIMA = 9.75 GWd/tU.

The term depletion refers to the reduction or depletion of enrichment of the fuel. When fuel is irradiated, most of the fission event occur in U-235 until sufficient Pu-239/Pu-240/Pu-241 build up to compete with the U-235 for neutrons.

Using a code like SIMULATE, one does core depletion calculations which basic simulate the fission process in the core during a cycle of operation.
 

Thread 'MCNP6 BURN Card Issue: Zero Burnup for Material in Fuel Assembly'

Hello everyone, I am currently working on a burnup calculation for a fuel assembly with repeated geometric structures using MCNP6. I have defined two materials (Material 1 and Material 2) which are actually the same material but located in different positions. However, after running the calculation with the BURN card, I am encountering an issue where all burnup information（power fraction(Initial input is 1,but output file is 0), burnup, mass, etc.） for Material 2 is zero, while Material 1...
View full post »

Similar threads

Why Are Burnup Numbers Identical for Different Fuels in MCNP?
Replies
3
Views
2K
Beginner Seeking Help: MCNP6 Burnup Example (OECD-NEA Benchmark)
Replies
6
Views
165
Why is My MCNP Burnup Calculation Failing?
Replies
1
Views
2K
Xenon poisoning (negative xenon load varies with core burnup)
Replies
1
Views
1K
[MCNP] Lost too much Keff with Burnup card
Replies
5
Views
3K
TUBRNP (Transuranus/FRAPCON) model
Replies
3
Views
2K
Anyone knows how to execute burnup process in MCNP5?
Replies
2
Views
3K
Rotational Symmetry in MCNPX core design
Replies
1
Views
2K
Burnup Conversion: Convert MWd/kgU or GWd/tU to %
Replies
1
Views
6K
MCNP6 BURN -- Testing about burnable poison depletion
Replies
1
Views
2K

Hot Threads

Recent Insights

Back
Top