12. Output Control

The RMC output control module is used to customize the output content. Especially in large-scale burnup calculations, which may generate a large amount of output information, the output control module can effectively reduce the size of output files.

12.1. Output Control Module Input Options

The input options for the output control module is:

PRINT
Mat <flag>
Keff <flag>
Source <flag>
CellTally <flag>
SURFACETALLY <flag>
POINTTALLY <flag>
MeshTally <flag>
CsTally <flag>
CYCTALLY <flag>
Inpfile <flag>
FissionSource <flag>
Material2HDF5 <flag>
BurnupCorrector <flag>

Among them,

• PRINT is the keyword for the output control module.

• Input options such as Mat ,Keff etc., specify whether to output related content(refer to Table 12.1). flag = 0 indicates not to output the specified content, input options including Keff, Source, CycTally, Inpfile flag = 1 indicates to output the specified content, input options including Mat, CellTally, SurfTally, PointTally, MeshTally, CsTally flag = n indicates output to the nth decimal place

• FissionSource is used to control the output of the converged (final generation) fission source information, including the positions, directions, and energy information of all sources, to the file inp.Source.h5.

Table 12.1 Input options for the output file control module

Input options	Output Content	Default Option
Mat	Nuclide density list for all materials	flag = 5, output to 5th decimal places
Keff	Keff for each generation	flag = 1, output
Source	Fission source information for each generation	flag = 0, do not output
CellTally	Results of cell tallies	flag = 4, output to 4th decimal places
SurfTally	Results of surface tallies	flag = 4, output to 4th decimal places
PointTally	Results of point tallies	flag = 4, output to 4th decimal places
MeshTally	Results of mesh tallies	flag = 4, output to 4th decimal places
CsTally	Results of cross-section tallies, including one-group cross-section during burnup calculations.	flag = 4, output to 4th decimal places
CycTally	Tallies for each generation	flag = 0, do not output
Inpfile	Continuation file input option	flag = 0, do not output
Material2HDF5	Outputs material information to an HDF5 file	flag = 0, do not output
FissionSource	Converged fission source information	flag = 0, do not output
BurnupCorrector	Calculation results for burnup correction steps	flag = 1, output

Note

It is important to note that Material2HDF5 is only effective when the user chooses to enable Inpfile. In this case, the program will output material information (including nuclear density, nuclide composition, etc.) to the material.h5 file instead of a text file. Compared to previous text file outputs, the materials output to the HDF5 file are easier for the program to read, which can effectively reduce the input card reading time in large-scale nuclear thermal coupling calculations. However, the readability of its content is poorer. Notably, to maintain the readability of materials when outputting to an HDF5 file, the RMC Python module includes the Materials.from_hdf5 function. Users can use this function to read the corresponding HDF5 file and convert it into a text file, as shown below:

from RMC.model.input.Material import Materials
materials = Materials.from_hdf5('material.h5')
with open('material.txt', 'w') as f:
    f.write(str(materials))

Caution

Material2HDF5 option is recommended to be enabled only during large-scale nuclear thermal coupling calculations to reduce the time taken by the program to read material files and accelerate initialization.

Caution

BurnupCorrector option is used to control the output of calculation results for correction steps when the estimated correction method is enabled in burnup calculations. This includes results such as effective multiplication factors and counting statistics for correction steps. It is important to note that in nuclear thermal coupling calculations, due to the existing framework where burnup calculations are split into single-step calculations, the counting results for correction steps will overwrite those for estimated steps (primarily in the power counting file MeshTally1.h5) when the estimated correction method is enabled. However, what is actually needed in coupled calculations are the counting results from the estimated steps. Therefore, it is recommended to set BurnupCorrector 0 in the PRINT options during nuclear thermal coupling calculations.

12.2. Output Control Module Input Example

For burnup calculations with a large number of burnup zones, it is recommended to use the following input cards to suppress the output of material and one-group cross-section information to avoid generating large data files.

PRINT
Mat 0
CsTally 0