Random Ray Eigenvalue Flux Normalization Change

src/random_ray/flat_source_domain.cpp

@@ -124,7 +124,9 @@ void FlatSourceDomain::update_single_neutron_source(SourceRegionHandle& srh)
         double chi = chi_[material * negroups_ + g_out];
 
         scatter_source += sigma_s * scalar_flux;
-        fission_source += nu_sigma_f * scalar_flux * chi;
+        if (settings::create_fission_neutrons) {
+          fission_source += nu_sigma_f * scalar_flux * chi;
+        }
       }
       srh.source(g_out) =
         (scatter_source + fission_source * inverse_k_eff) / sigma_t;
@@ -351,6 +353,27 @@ void FlatSourceDomain::compute_k_eff()
 
   double k_eff_new = k_eff_ * (fission_rate_new / fission_rate_old);
 
+  // Normalize fluxes by total number of fission neutrons produced. This ensures
+  // consistent scaling of the eigenvector such that its magnitude is
+  // comparable to the eigenvector produced by the Monte Carlo solver.
+  // Multiplying by the eigenvalue is unintuitive, but it is necessary.
+  // If the eigenvalue is 1.2, per starting source neutron, you will
+  // generate 1.2 neutrons. Thus if we normalize to generating only ONE neutron
+  // in total for the whole domain, then we don't actually have enough flux to
+  // generate the required 1.2 neutrons. We only know the flux required to
+  // generate 1 neutron (which would have required less then one starting
+  // neutron). Thus, you have to scale the flux up by the eigenvalue such
+  // that 1.2 neutrons are generated, so as to be consistent with the
+  // bookkeeping in MC which is all done per starting source neutron (not per
+  // neutron produced).
+  double total_fission_neutrons = fission_rate_new * simulation_volume_;
+  double norm_factor = k_eff_new / total_fission_neutrons;
+
+#pragma omp parallel for
+  for (int64_t se = 0; se < n_source_elements(); se++) {
+    source_regions_.scalar_flux_new(se) *= norm_factor;
+  }
+
   double H = 0.0;
   // defining an inverse sum for better performance
   double inverse_sum = 1 / fission_rate_new;

src/random_ray/linear_source_domain.cpp

@@ -68,9 +68,11 @@ void LinearSourceDomain::update_single_neutron_source(SourceRegionHandle& srh)
 
         // Compute source terms for flat and linear components of the flux
         scatter_flat += sigma_s * flux_flat;
-        fission_flat += nu_sigma_f * flux_flat * chi;
         scatter_linear += sigma_s * flux_linear;
-        fission_linear += nu_sigma_f * flux_linear * chi;
+        if (settings::create_fission_neutrons) {
+          fission_flat += nu_sigma_f * flux_flat * chi;
+          fission_linear += nu_sigma_f * flux_linear * chi;
+        }
       }
 
       // Compute the flat source term

tests/regression_tests/random_ray_adjoint_k_eff/results_true.dat

@@ -1,171 +1,171 @@
 k-combined:
 1.006640E+00 1.812969E-03
 tally 1:
-6.684129E+00
-8.939821E+00
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tests/regression_tests/random_ray_auto_convert/material_wise/results_true.dat

@@ -1,2 +1,2 @@
 k-combined:
-7.356667E-01 6.637270E-03
+7.366910E-01 6.872852E-03

tests/regression_tests/random_ray_auto_convert/stochastic_slab/results_true.dat

@@ -1,2 +1,2 @@
 k-combined:
-7.551716E-01 8.117378E-03
+7.551716E-01 8.117379E-03

tests/regression_tests/random_ray_diagonal_stabilization/results_true.dat

@@ -1,2 +1,2 @@
 k-combined:
-7.201808E-01 1.506596E-02
+7.202605E-01 1.464839E-02