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More configuraton options in fissionSource #101

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Nov 21, 2023
Merged

More configuraton options in fissionSource #101

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8262018
Update fissionSource
Mikolaj-A-Kowalski Nov 20, 2023
1fc5c3f
Add manual entry for fissionSource
Mikolaj-A-Kowalski Nov 20, 2023
f2861de
Update docs/User Manual.rst
Mikolaj-A-Kowalski Nov 21, 2023
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65 changes: 50 additions & 15 deletions ParticleObjects/Source/fissionSource_class.f90
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Original file line number Diff line number Diff line change
Expand Up @@ -3,7 +3,8 @@ module fissionSource_class
use numPrecision
use endfConstants
use universalVariables, only : OUTSIDE_MAT, VOID_MAT
use genericProcedures, only : fatalError, rotateVector
use genericProcedures, only : rotateVector, numToChar
use errors_mod, only : fatalError
use dictionary_class, only : dictionary
use RNG_class, only : RNG

Expand All @@ -28,17 +29,17 @@ module fissionSource_class
!! Neutron Source from distributed fission sites
!!
!! Places fission sites uniformly in regions with fissile material.
!! Spectrum of the fission neutron is such as if it fission was caused by incdent
!! neutron with CE energy E or MG with group G.
!! Spectrum of the fission neutron is such as if it fission was caused by
!! incident neutron with CE energy E or MG with group G.
!! Angular distribution is isotropic.
!!
!! Private members:
!! isMG -> is the source multi-group? (default = .false.)
!! bottom -> Bottom corner (x_min, y_min, z_min)
!! top -> Top corner (x_max, y_max, z_max)
!! E -> Fission site energy [MeV] (default = 1.0E-6)
!! G -> Fission site Group (default = 1)
!!
!! isMG -> is the source multi-group? (default = .false.)
!! bottom -> Bottom corner (x_min, y_min, z_min)
!! top -> Top corner (x_max, y_max, z_max)
!! E -> Fission site energy [MeV] (default = 1.0E-6)
!! G -> Fission site Group (default = 1)
!! attempts -> Maximum number of attempts to find a fissile material
!! Interface:
!! source_inter Interface
!!
Expand All @@ -48,6 +49,9 @@ module fissionSource_class
!! #data MG; #
!! #E 15.0; #
!! #G 7; #
!! #attempts 100000; # (default: 10000)
!! #top (1.0 1.0 1.0); #
!! #bottom (0.0 0.0 0.0); #
!! }
!!
type, public,extends(source) :: fissionSource
Expand All @@ -57,6 +61,7 @@ module fissionSource_class
real(defReal), dimension(3) :: top = ZERO
real(defReal) :: E = ZERO
integer(shortInt) :: G = 0
integer(shortInt) :: attempts = 10000
contains
procedure :: init
procedure :: sampleParticle
Expand All @@ -76,6 +81,7 @@ subroutine init(self, dict, geom)
class(geometry), pointer, intent(in) :: geom
character(nameLen) :: type
real(defReal), dimension(6) :: bounds
real(defReal), allocatable, dimension(:) :: temp
character(100), parameter :: Here = 'init (fissionSource_class.f90)'

! Provide geometry info to source
Expand All @@ -98,10 +104,39 @@ subroutine init(self, dict, geom)
call dict % getOrDefault(self % E, 'E', 1.0E-6_defReal)
call dict % getOrDefault(self % G, 'G', 1)

! Load the number of allowed attempts
call dict % getOrDefault(self % attempts, 'attempts', 10000)

if (self % attempts < 1) then
call fatalError(Here, 'Number of attempts must be greater than 0. Is: ' // numToChar(self % attempts))
end if

! Set bounding region
bounds = self % geom % bounds()
self % bottom = bounds(1:3)
self % top = bounds(4:6)
if (dict % isPresent('top') .or. dict % isPresent('bottom')) then
! Get top and bottom from dictionary
call dict % get(temp, 'top')
if (size(temp) /= 3) then
call fatalError(Here, "Top point must have 3 coordinates.")
end if
self % top = temp

call dict % get(temp, 'bottom')
if (size(temp) /= 3) then
call fatalError(Here, "Bottom point must have 3 coordinates.")
end if
self % bottom = temp

! Check that top and bottom are valid
if (any(self % top < self % bottom)) then
call fatalError(Here, "Top point must have all coordinates greater than bottom point.")
end if

else ! Get bounds from geometry
bounds = self % geom % bounds()
self % bottom = bounds(1:3)
self % top = bounds(4:6)
end if


end subroutine init

Expand Down Expand Up @@ -136,9 +171,9 @@ function sampleParticle(self, rand) result(p)
rejection : do
! Protect against infinite loop
i = i +1
if ( i > 200) then
call fatalError(Here, 'Infinite loop in sampling of fission sites. Please check that&
& defined volume contains fissile material.')
if (i > self % attempts) then
call fatalError(Here, "Failed to find a fissile material in: "// numToChar(self % attempts) // " attempts.&
& Increase the number of maximum attempts or verify that fissile materials are present.")
end if

! Sample Position
Expand Down
71 changes: 46 additions & 25 deletions docs/User Manual.rst
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Original file line number Diff line number Diff line change
Expand Up @@ -25,8 +25,8 @@ eigenPhysicsPackage, used for criticality (or eigenvalue) calculations
* XSdata: keyword to the name of the nuclearDataHandle used
* seed (*optional*): initial seed for the pseudo random number generator
* outputFile (*optional*, default = 'output'): name of the output file
* outputFormat (*optional*, default = ``asciiMATLAB``): type of output file.
Choices are ``asciiMATLAB`` and ``asciiJSON``
* outputFormat (*optional*, default = ``asciiMATLAB``): type of output file.
Choices are ``asciiMATLAB`` and ``asciiJSON``
* printSource (*optional*, default = 0): 1 for true; 0 for false; requests
to print the particle source (location, direction, energy of each particle
in the particleDungeon) to a text file
Expand Down Expand Up @@ -71,19 +71,19 @@ fixedSourcePhysicsPackage, used for fixed source calculations
* XSdata: keyword to the name of the nuclearDataHandle used
* seed (*optional*): initial seed for the pseudo random number generator
* outputFile (*optional*, default = 'output'): name of the output file
* outputFormat (*optional*, default = ``asciiMATLAB``): type of output file.
Choices are ``asciiMATLAB`` and ``asciiJSON``
* outputFormat (*optional*, default = ``asciiMATLAB``): type of output file.
Choices are ``asciiMATLAB`` and ``asciiJSON``
* printSource (*optional*, default = 0): 1 for true; 0 for false; requests
to print the particle source (location, direction, energy of each particle
in the particleDungeon) to a text file
* buffer (*optional*, default = 50): size of the particle bank used by each
* buffer (*optional*, default = 50): size of the particle bank used by each
OpenMP thread to store secondary particles
* commonBufferSize (*optional*): if not included, the common buffer is not
used; if included, after each particle history the particles in each
thread-private buffer (or bank, or dungeon) are moved to a buffer
* commonBufferSize (*optional*): if not included, the common buffer is not
used; if included, after each particle history the particles in each
thread-private buffer (or bank, or dungeon) are moved to a buffer
common to all threads to avoid long histories
* bufferShift (*optional*, default = 10): threshold of particles to be
stored in a thread-private buffer, after which particles are shifted to
* bufferShift (*optional*, default = 10): threshold of particles to be
stored in a thread-private buffer, after which particles are shifted to
the common buffer

Example: ::
Expand Down Expand Up @@ -154,7 +154,7 @@ Example: ::
Source
------

For the moment, the only possible external **source** types in SCONE are point source
For the moment, the only possible external **source** types in SCONE are point source
and material source.

pointSource
Expand Down Expand Up @@ -185,21 +185,42 @@ It is a type of volumetric source. For the moment it is constrained to neutrons.
The properties of a material source are:

* mat: the name of the material from which to sample (must be defined in materials).
* data (*optional*, default = continuous energy): data type for source particles. Can be ``ce``
* data (*optional*, default = continuous energy): data type for source particles. Can be ``ce``
or ``mg``.
* E (*optional*, default = 1E-6): energy of the particles emitted, for continuous energy
* E (*optional*, default = 1E-6): energy of the particles emitted, for continuous energy
calculations. [MeV]
* G (*optional*, default = 1): energy group of the particles emitted, for multi-group
* G (*optional*, default = 1): energy group of the particles emitted, for multi-group
calculations.
* boundingBox (*optional*, default is the geometry bounding box):
(x_min y_min z_min x_max y_max z_max) vector describing a bounding box to improve sampling
* boundingBox (*optional*, default is the geometry bounding box):
(x_min y_min z_min x_max y_max z_max) vector describing a bounding box to improve sampling
efficiency or to localise material sampling to a particular region.

Hence, an input would look like: ::

source { type materialSource; mat myMat; data ce; E 2.0;
source { type materialSource; mat myMat; data ce; E 2.0;
boundingBox (-5.0 -3.0 2.0 5.0 4.0 3.0); }

fissionSource
#############

A source intended to initialise eigenvalue calculations. If it is not defined in the input file, it is
used with the default settings. It is a type of volumetric source, which uniformly distributes fission
sites in the geometry. The energy spectrum of the fission neutrons is based on a fixed incident
energy provided by the user. The properties of a fission source are:

* data (*optional*, default='ce'): data type for source particles. Can be ``ce``
or ``mg``.
* E (*optional*, default=1E-6): energy of the incident neutron causing fission [MeV]. Makes
sense for continuous energy source only.
* G (*optional*, default=1): energy group of the incident neutron causing fission. Makes
sense for multi-group source only.
* attempts (*optional*, default=10000): number of attempts to sample a fission site in a cell
before throwing an error.
* bottom (*optional*): Lower point determining axis-aligned bounding box where to sample points. If
provided ``top`` must also be provided.
* top (*optional*): Upper point determining axis-aligned bounding box where to sample points. If
provided ``bottom`` must also be provided.

Transport Operator
------------------

Expand Down Expand Up @@ -634,8 +655,8 @@ vtk
* corner: (x y z) array with the corner of the geometry [cm]
* width: (x y z) array with the width of the mesh in each direction [cm]
* vox: (x y z) array with the number of voxels requested in each direction
* what (*optional*, default = material): defines what is highlighted in the
plot; options are ``material`` and ``uniqueID``, where ``uniqueID``
* what (*optional*, default = material): defines what is highlighted in the
plot; options are ``material`` and ``uniqueID``, where ``uniqueID``
highlights unique cell IDs

Example: ::
Expand All @@ -651,17 +672,17 @@ bmp
with the width of the geometry plotted in each direction [cm]
* res: (y z), (x z) or (x y) array with the resolution of the mesh in each direction
* output: name of the output file, with extension ``.bmp``
* what (*optional*, default = material): defines what is highlighted in the
plot; options are ``material`` and ``uniqueID``, where ``uniqueID``
* what (*optional*, default = material): defines what is highlighted in the
plot; options are ``material`` and ``uniqueID``, where ``uniqueID``
highlights unique cell IDs

Example: ::

plotBMP { type bmp; axis z; centre (0.0 0.0 0.0); width (50 10); res (1000 200); output geomZ; what material; }
.. note::
SCONE can be run to visualise geometry without actually doing transport, by
including ``--plot`` when running the application. In this case the visualiser

.. note::
SCONE can be run to visualise geometry without actually doing transport, by
including ``--plot`` when running the application. In this case the visualiser
has to be included in the file.

Nuclear Data
Expand Down