A monte carlo pin cell spectral code for nuclear engineering applications.
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- c -
clearTallies() :
TallyBank
clone() :
Isotope
,
Material
,
Tally
collideNeutron() :
BoundedRegion
,
EquivalenceRegion
,
InfiniteMediumRegion
,
Isotope
,
Material
,
Region
computeBatchStatistics() :
Tally
,
TallyBank
computeFuelFuelCollsionProb() :
EquivalenceRegion
computeGroupXS() :
pinspec.process.GroupXS
computeModeratorFuelCollisionProb() :
EquivalenceRegion
computeParametrizedDistance() :
BoundedRegion
,
Surface
,
XPlane
,
YPlane
,
ZCylinder
computeRIs() :
pinspec.process.RIEff
,
pinspec.process.RITrue
computeScaledBatchStatistics() :
Tally
,
TallyBank
contains() :
BoundedRegion
,
Geometry
containsIsotope() :
Material
,
Region
createTally() :
TallyFactory
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