crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-28 14:15:42 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

Move openmc_tally_set_nuclides to C++

Browse source
This commit is contained in:
Sterling Harper 2019-02-05 17:58:45 -05:00
parent 19e7923d08
commit 53a0dc8d97
3 changed files with 49 additions and 124 deletions
Download patch file Download diff file Expand all files Collapse all files

2
include/openmc/tallies/tally.h
View file

@ -61,7 +61,7 @@ public:
std::vector<int> scores_; //!< Filter integrands (e.g. flux, fission)
//! Index of each nuclide to be tallied. -1 indicates total material.
std::vector<int> nuclides_;
std::vector<int> nuclides_ {-1};
//! True if this tally has a bin for every nuclide in the problem
bool all_nuclides_ {false};