Computational methods Neutron transport

1 computational methods

1.1 discretization in deterministic methods
1.2 computer codes used in neutron transport

1.2.1 probabilistic codes
1.2.2 deterministic codes

computational methods

both fixed-source , criticality calculations can solved using deterministic methods or stochastic methods. in deterministic methods transport equation (or approximation of it, such diffusion theory) solved differential equation. in stochastic methods such monte carlo discrete particle histories tracked , averaged in random walk directed measured interaction probabilities. deterministic methods involve multi-group approaches while monte carlo can work multi-group , continuous energy cross-section libraries. multi-group calculations iterative, because group constants calculated using flux-energy profiles, determined result of neutron transport calculation.

discretization in deterministic methods

to numerically solve transport equation using algebraic equations on computer, spatial, angular, energy, , time variables must discretized.

spatial variables typically discretized breaking geometry many small regions on mesh. balance can solved @ each mesh point using finite difference or nodal methods.
angular variables can discretized discrete ordinates , weighting quadrature sets (giving rise sn methods), or functional expansion methods spherical harmonics (leading pn methods).
energy variables typically discretized multi-group method, each energy group represents 1 constant energy. few 2 groups can sufficient thermal reactor problems, fast reactor calculations may require many more.
the time variable broken discrete time steps, time derivatives replaced difference formulas.

computer codes used in neutron transport
probabilistic codes

mcnp - lanl developed monte carlo code general radiation transport
openmc - mit developed open source monte carlo code
shift/keno - ornl developed monte carlo codes general radiation transport , criticality analysis
cog - llnl developed monte carlo code criticality safety analysis , general radiation transport (http://cog.llnl.gov)
rmc - tsinghua university -department of engineering physics developed monte carlo code general radiation transport
mcbend - answers software service developed monte carlo code general radiation transport
serpent - finnish developed monte carlo neutron transport code
tripoli - 3d general purpose continuous energy monte carlo transport code developed @ cea, france

deterministic codes

attila - commercial transport code
dragon - open-source lattice physics code
phoenix/anc - proprietary lattice-physics , global diffusion code suite westinghouse electric
partisn - lanl developed transport code based on discrete ordinates method
newt - ornl developed 2-d sn code
dif3d/variant - argonne national laboratory developed 3-d code developed fast reactors
denovo - massively parallel transport code under development ornl
jaguar - parallel 3-d slice balance approach transport code arbitrary polytope grids developed @ nnl
dantsys
rama - proprietary 3d method of characteristics code arbitrary geometry modeling, developed epri transware enterprises inc.
raptor-m3g - proprietary parallel radiation transport code developed westinghouse electric company
openmoc - mit developed open source parallel method of characteristics code
mpact - parallel 3d method of characteristics code under development university of michigan
dort - discrete ordinates transport
apollo - lattice physics code used cea, edf , areva
milonga - free nuclear reactor core analysis code




^ openmc . 
^ psg2 serpent . 
^ tripoli-4 . 
^ rama . 
^ openmoc . 
^ apollo3 (pdf). 
^ milonga .