CellTally

Description

The CellTally class wraps an OpenMC tally with a distributed cell filter to enable data transfer from CSG cells to a MooseMesh mirror of the OpenMC geometry. The cell to element mapping established by the OpenMCCellAverageProblem is used to facilitate this data transfer to the mesh mirror. When using a CellTally, block specifies which blocks in the [Mesh] should be tallied. Then, any OpenMC cells that map to those blocks are added to a cell tally, with one bin for each unique cell ID/instance combination. check_equal_mapped_tally_volumes is used to confirm whether each cell in the CellTally has the same mapped volume on the MooseMesh (to a tolerance specified in equal_tally_volume_abs_tol).

Cardinal supports several tally scores which can be specified in the score parameter when adding a tally. Options include:

• heating: total nuclear heating

• heating_local: same as the heating score, except that energy from secondary photons is deposited locally

• kappa_fission: recoverable energy from fission, including prompt sources (fission fragments, prompt neutrons, prompt gammas) and delayed sources (delayed neutrons, delayed gammas, delayed betas). Neutrino energy is neglected. The energy from photons is assumed to deposit locally.

• fission_q_prompt: the prompt components of the kappa_fission score, except that the energy released is a function of the incident energy by linking to optional fission energy release data.

• fission_q_recoverable: same as the kappa_fission score, except that the score depends on the incident energy by linking to optional fission energy release data

• damage_energy: damage energy production

• flux: particle scalar flux

• H3_production: tritium production reaction rate

• total: the total reaction rate

• absorption: the absorption reaction rate

• scatter: the scattering reaction rate

• nu_scatter: the scattering reaction rate combined with the total (n, xn) reaction rate

• fission: the fission reaction rate

• nu_fission: the fission neutron production rate

• prompt_nu_fission: the prompt fission neutron production rate

• delayed_nu_fission: the delayed fission neutron production rate

• decay_rate: the neutron precursor decay constant weighted by the delayed fission neutron production rate

• inverse_velocity: one over the particle velocity weighted by the particle flux

For more information on the specific meanings of these various scores, please consult the OpenMC tally documentation. Table 1 compares the units from OpenMC and the units of the AuxVariables created for all tally scores supported by Cardinal. Note that for all area or volume units in Table 1, that those units match whatever unit is used in the [Mesh]. The names of the tally auxvariables added by the OpenMCCellAverageProblem can be modified by specifying a name for each score in name.

This tally is capable of controlling OpenMC execution through the use of tally triggers. Tallies currently support active batch termination through a relative error indicator, which can be enabled by setting trigger to rel_err for each score in the tally. Scores which do not need to control active batch termination can have a value of none set. If enabled, an appropriate relative error must be set for each score in trigger_threshold.

Certain aspects of tally results can be output as auxiliary variables to the mesh:

• unrelaxed_tally: unrelaxed tally; this will append _raw to the tally name and output to the mesh mirror

• unrelaxed_tally_rel_error: unrelaxed tally relative error; this will append _rel_error to the tally name and output to the mesh mirror

• unrelaxed_tally_std_dev: unrelaxed tally standard deviation; this will append _std_dev to the tally and output to the mesh mirror

Example Input File Syntax

As an example, this CellTally scores kappa_fission (the default tally score) on block 0 and stores the result in a variable named heat_source. This corresponds to tallying the heating for three different fuel pebbles.

[Problem<<<{"href": "../../syntax/Problem/index.html"}>>>]
  type = OpenMCCellAverageProblem
  verbose = true
  power = 1500.0
  temperature_blocks = '0'
  cell_level = 1
  scaling = 100.0

  [Tallies<<<{"href": "../../syntax/Problem/Tallies/index.html"}>>>]
    [heat_source]
      type = CellTally<<<{"description": "A class which implements distributed cell tallies.", "href": "CellTally.html"}>>>
      block<<<{"description": "Subdomains for which to add tallies in OpenMC. If not provided, tallies will be applied over the entire domain corresponding to the [Mesh] block."}>>> = '0'
      name<<<{"description": "Auxiliary variable name(s) to use for OpenMC tallies. If not specified, defaults to the names of the scores"}>>> = heat_source
    []
  []
[]

Input Parameters