diff --git a/src/cross_section.F90 b/src/cross_section.F90
index 86d15ef1d1..1ab4e41d51 100644
--- a/src/cross_section.F90
+++ b/src/cross_section.F90
@@ -12,7 +12,7 @@ module cross_section
   use fission,              only: nu_total
   use global
   use material_header,      only: Material
-  use output,               only: write_message
+  use output,               only: write_message, print_nuclide, header
   use string,               only: split_string, str_to_int, str_to_real, &
                                   lower_case, int_to_str
 
@@ -137,6 +137,9 @@ contains
     ! ==========================================================================
     ! READ ALL ACE CROSS SECTION TABLES
 
+    ! display header in summary.out
+    if (master) call header("CROSS SECTION TABLES", unit=UNIT_SUMMARY)
+
     call dict_create(already_read)
 
     ! Loop over all files
@@ -154,6 +157,10 @@ contains
 
              call read_ace_table(index_nuclides, index_list)
 
+             ! Print out information on table to summary.out file
+             nuc => nuclides(index_nuclides)
+             if (master) call print_nuclide(nuc, unit=UNIT_SUMMARY)
+
              call dict_add_key(already_read, name, 0)
              call dict_add_key(already_read, alias, 0)
           end if
@@ -634,6 +641,8 @@ contains
     rxn % Q_value = ZERO
     rxn % TY      = 1
     rxn % IE      = 1
+    rxn % has_angle_dist = .false.
+    rxn % has_energy_dist = .false.
     allocate(rxn % sigma(nuc % n_grid))
     rxn % sigma = nuc % elastic
     
@@ -650,6 +659,10 @@ contains
     do i = 1, NMT
        rxn => nuc % reactions(i+1)
 
+       ! set defaults
+       rxn % has_angle_dist  = .false.
+       rxn % has_energy_dist = .false.
+
        ! read MT number, Q-value, and neutrons produced
        rxn % MT      = int(XSS(LMT + i - 1))
        rxn % Q_value = XSS(JXS4 + i - 1)
@@ -701,10 +714,6 @@ contains
           nuc % index_fission(i_fission) = i + 1
           nuc % n_fission = nuc % n_fission + 1
        end if
-
-       ! set defaults
-       rxn % has_angle_dist  = .false.
-       rxn % has_energy_dist = .false.
     end do
 
   end subroutine read_reactions
diff --git a/src/output.F90 b/src/output.F90
index 68ddee6bc5..e5e0c5facb 100644
--- a/src/output.F90
+++ b/src/output.F90
@@ -3,7 +3,7 @@ module output
   use ISO_FORTRAN_ENV
 
   use constants
-  use cross_section_header, only: Reaction
+  use cross_section_header, only: Nuclide, Reaction
   use datatypes,            only: dict_get_key
   use endf,                 only: reaction_name
   use geometry_header,      only: Cell, Universe, Surface
@@ -711,6 +711,83 @@ contains
 
   end subroutine print_geometry
 
+!===============================================================================
+! PRINT_NUCLIDE displays information about a continuous-energy neutron
+! cross_section table and its reactions and secondary angle/energy distributions
+!===============================================================================
+
+  subroutine print_nuclide(nuc, unit)
+
+    type(Nuclide), pointer :: nuc
+    integer,      optional :: unit
+
+    integer :: i
+    integer :: unit_
+    integer :: size_total
+    integer :: size_xs
+    integer :: size_angle
+    integer :: size_energy
+    type(Reaction), pointer :: rxn => null()
+
+    ! set default unit for writing information
+    if (present(unit)) then
+       unit_ = unit
+    else
+       unit_ = OUTPUT_UNIT
+    end if
+
+    ! Determine size of cross-sections
+    size_xs = (5 + nuc % n_reaction) * nuc % n_grid * 8
+    size_total = size_xs
+
+    ! Basic nuclide information
+    write(unit_,*) 'Nuclide ' // trim(nuc % name)
+    write(unit_,*) '  zaid = ' // trim(int_to_str(nuc % zaid))
+    write(unit_,*) '  awr = ' // trim(real_to_str(nuc % awr))
+    write(unit_,*) '  kT = ' // trim(real_to_str(nuc % kT))
+    write(unit_,*) '  # of grid points = ' // trim(int_to_str(nuc % n_grid))
+    write(unit_,*) '  Fissionable = ', nuc % fissionable
+    write(unit_,*) '  # of fission reactions = ' // trim(int_to_str(nuc % n_fission))
+    write(unit_,*) '  # of reactions = ' // trim(int_to_str(nuc % n_reaction))
+    write(unit_,*) '  Size of cross sections = ' // trim(int_to_str(&
+         size_xs)) // ' bytes'
+
+    write(unit_,*) '  Reaction    Q-value   Mult    IE    size(angle) size(energy)'
+    do i = 1, nuc % n_reaction
+       ! Information on each reaction
+       rxn => nuc % reactions(i)
+
+       ! Determine size of angle distribution
+       if (rxn % has_angle_dist) then
+          size_angle = rxn % adist % n_energy * 16 + size(rxn % adist % data) * 8
+       else
+          size_angle = 0
+       end if
+
+       ! Determine size of energy distribution
+       if (rxn % has_energy_dist) then
+          size_energy = size(rxn % edist % data) * 8
+       else
+          size_energy = 0
+       end if
+
+       write(unit_,'(3X,A11,1X,F8.3,2X,I3,3X,I6,1X,I11,1X,I11)') &
+            reaction_name(rxn % MT), rxn % Q_value, rxn % TY, rxn % IE, &
+            size_angle, size_energy
+
+       ! Accumulate data size
+       size_total = size_total + size_angle + size_energy
+    end do
+
+    ! Write total memory used
+    write(unit_,*) '  Total memory used = ' // trim(int_to_str(size_total)) &
+         // ' bytes'
+
+    ! Blank line at end of nuclide
+    write(unit_,*)
+
+  end subroutine print_nuclide
+
 !===============================================================================
 ! PRINT_SUMMARY displays summary information about the problem about to be run
 ! after reading all input files