Changes in OpenMC to imitate MCNP's TOTNU option

totnu_develop.diff

diff --git a/src/nuclide.cpp b/src/nuclide.cpp
index ae3085aff2..13816f59a5 100644
--- a/src/nuclide.cpp
+++ b/src/nuclide.cpp
@@ -409,7 +409,7 @@ void Nuclide::create_derived(
       for (int i = 0; i < n; ++i) {
         double E = grid_[t].energy[i];
         xs_[t](i, XS_NU_FISSION) =
-          nu(E, EmissionMode::total) * xs_[t](i, XS_FISSION);
+          nu(E, EmissionMode::prompt) * xs_[t](i, XS_FISSION);
       }
     }
   }
@@ -602,7 +602,7 @@ void Nuclide::calculate_xs(
     micro.absorption = sig_a;
     micro.fission = sig_f;
     micro.nu_fission =
-      fissionable_ ? sig_f * this->nu(p.E(), EmissionMode::total) : 0.0;
+      fissionable_ ? sig_f * this->nu(p.E(), EmissionMode::prompt) : 0.0;
 
     if (simulation::need_depletion_rx) {
       // Only non-zero reaction is (n,gamma)
@@ -943,7 +943,7 @@ void Nuclide::calculate_urr_xs(int i_temp, Particle& p) const
 
   // Determine nu-fission cross-section
   if (fissionable_) {
-    micro.nu_fission = nu(p.E(), EmissionMode::total) * micro.fission;
+    micro.nu_fission = nu(p.E(), EmissionMode::prompt) * micro.fission;
   }
 }
 
diff --git a/src/physics.cpp b/src/physics.cpp
index 36161f3ea1..7e03c05502 100644
--- a/src/physics.cpp
+++ b/src/physics.cpp
@@ -1040,8 +1040,8 @@ void sample_fission_neutron(int i_nuclide, const Reaction& rx, double E_in,
 {
   // Determine total nu, delayed nu, and delayed neutron fraction
   const auto& nuc {data::nuclides[i_nuclide]};
-  double nu_t = nuc->nu(E_in, Nuclide::EmissionMode::total);
-  double nu_d = nuc->nu(E_in, Nuclide::EmissionMode::delayed);
+  double nu_t = nuc->nu(E_in, Nuclide::EmissionMode::prompt);
+  double nu_d = 0.0;
   double beta = nu_d / nu_t;
 
   if (prn(seed) < beta) {