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OpenMC-MGXS
LaTeX source for "Multigroup Cross-Section Generation with the OpenMC Monte Carlo Particle Transport Code" published in Nuclear Technology (2019)
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addressed comments on delayed neutron quantities
#10
samuelshaner
closed
6 years ago
1
changed my affiliation to MIT
#9
samuelshaner
closed
6 years ago
0
added funding acknowledgement and changed YE affiliation address
#8
samuelshaner
closed
6 years ago
0
Other Romano comments
#7
paulromano
closed
6 years ago
7
Need units on fission rates for Fig 3
#6
paulromano
closed
6 years ago
3
Eq. 10 and 11
#5
paulromano
closed
6 years ago
1
Explanation re: estimators
#4
paulromano
closed
6 years ago
1
Some minor edits
#3
nelsonag
closed
6 years ago
0
Nelson's comments
#2
nelsonag
closed
6 years ago
3
Sections on computing delayed MGXS
#1
samuelshaner
closed
7 years ago
1