Changes in OpenMC to imitate MCNP's TOTNU option

totnu.diff

diff --git a/src/cross_section.F90 b/src/cross_section.F90
index 2b3b32a9c..4c101f05b 100644
--- a/src/cross_section.F90
+++ b/src/cross_section.F90
@@ -187,7 +187,7 @@ contains
 
         if (nuc % fissionable) then
           micro_xs(i_nuclide) % fission = sigF
-          micro_xs(i_nuclide) % nu_fission = sigF * nuc % nu(E, EMISSION_TOTAL)
+          micro_xs(i_nuclide) % nu_fission = sigF * nuc % nu(E, EMISSION_PROMPT)
         else
           micro_xs(i_nuclide) % fission    = ZERO
           micro_xs(i_nuclide) % nu_fission = ZERO
diff --git a/src/nuclide_header.F90 b/src/nuclide_header.F90
index 38e89b648..ff4dc8872 100644
--- a/src/nuclide_header.F90
+++ b/src/nuclide_header.F90
@@ -691,7 +691,7 @@ contains
       if (this % fissionable) then
         do i = 1, size(this % sum_xs(t) % fission)
           this % sum_xs(t) % nu_fission(i) = this % nu(this % grid(t) % energy(i), &
-               EMISSION_TOTAL) * this % sum_xs(t) % fission(i)
+               EMISSION_PROMPT) * this % sum_xs(t) % fission(i)
         end do
       else
         this % sum_xs(t) % nu_fission(:) = ZERO
diff --git a/src/physics.F90 b/src/physics.F90
index d13313e05..0149c8267 100644
--- a/src/physics.F90
+++ b/src/physics.F90
@@ -1208,8 +1208,8 @@ contains
     site % uvw(3) = sqrt(ONE - mu*mu) * sin(phi)
 
     ! Determine total nu, delayed nu, and delayed neutron fraction
-    nu_t = nuc % nu(E_in, EMISSION_TOTAL)
-    nu_d = nuc % nu(E_in, EMISSION_DELAYED)
+    nu_t = nuc % nu(E_in, EMISSION_PROMPT)
+    nu_d = ZERO
     beta = nu_d / nu_t
 
     if (prn() < beta) then