Date of modification: 2024/Jan./18

Manual of FRENDY Version 2(Preliminary version)- 1 -Contents
Contents ................................................................................................................................................2
1

Input Instructions ..........................................................................................................................4
1.1

Input Format..........................................................................................................................4

1.2

FRENDY Original Input Format...........................................................................................4

1.2.1

Processing Mode ...........................................................................................................6

1.2.2

Input Parameters ...........................................................................................................7

1.2.3

Examples of FRENDY Input Formatted File ..............................................................64

1.3

NJOY Compatible Format ..................................................................................................66

1.4

Sample Input Data for ACE File Generation ......................................................................68

1.4.1

Simplest Input Data.....................................................................................................68

1.4.2

Input Data to Modify Default Input Parameters..........................................................69

1.4.3

Input Data to Reproduce NJOY99 Input .....................................................................70

1.5

Sample Input Data for Multi-Group Cross-Section Generation ..........................................78

1.5.1

Sample Input Data to Generate Neutron-Induced Multi-Group Cross-Section File ...78

1.5.2

Sample Input Data to Generate Multi-Group Cross-Section File of TSL Data ...........85

1.5.3

Sample Input Data to Generate Multi-Nuclide Data ...................................................87

1.5.4

Sample Input Data for Automatic Setting of Background Cross-Sections..................92

1.5.5

Sample Input Data for Automatic Setting of Energy Group Structure ........................94

1.5.6

Sample Input Data for Resonance Up-Scattering Correction .....................................95

1.5.7

Sample Input Data to calculate IR parameter..............................................................96

1.6

Input Instruction of ACE File Perturbation Tools ...............................................................99

1.6.1

Input Instruction of make_perturbation_factor ...........................................................99

1.6.2

Sample Input of make_perturbation_factor .............................................................. 100

1.6.3

Input Instruction of perturbation_ace_file................................................................. 101

1.7

Input Instruction of ENDF Modification Tool .................................................................. 102

1.8

Sample Input of ENDF Modification Tool ....................................................................... 103

1.9

Sample Input of XS Output Tool ...................................................................................... 104

1.9.1

Sample Input Data to Output Continuous Energy Cross Section Data from an ENDF

Formatted File ........................................................................................................................... 105
1.9.2

Sample Input Data to Output Multi-Group Cross Section Data from an ENDF

Formatted File ........................................................................................................................... 106
1.9.3

Sample Input Data to Output Continuous Energy Cross Section Data from an ACE

Formatted File ........................................................................................................................... 107- 2 -1.9.4

Sample Input Data to Output Multi-Group Cross Section Data from an ACE

Formatted File ........................................................................................................................... 108
1.9.5

Sample Input Data to Output Multi-Group Data from a GENDF Formatted File .... 109

1.9.6

Sample Input Data to Output Comparison Results of Continuous Energy Cross

Section Data from ENDF/ACE Formatted Files ....................................................................... 109
1.9.7

Sample Input Data to Output Comparison Results of Multi-Group Cross Section Data

from ENDF/ACE Formatted Files ............................................................................................ 111
2

Installation of FRENDY ........................................................................................................... 113
2.1

Directory Structure............................................................................................................ 113

2.2

How to Install FRENDY on Linux, UNIX, or macOS Platforms ..................................... 114

2.3

How to Execute FRENDY ................................................................................................ 115

2.4

Test Calculation with Samples .......................................................................................... 115

2.5

Test Programs for Boost Test Library ............................................................................... 115

2.6

How to Install FRENDY on Visual Studio 2019 .............................................................. 116

2.6.1

Installation of Boost Library ..................................................................................... 116

2.6.2

Installation of CLAPACK Library ............................................................................ 117

2.6.3

Installation of FRENDY ........................................................................................... 117

2.6.4

Running FRENDY with bat Files ............................................................................. 118

- 3 -

1

Input Instructions

1.1

Input Format

FRENDY accepts two types of input formats:⚫

FRENDY original input format,⚫

NJOY compatible format.

The original input format requires only the processing mode name and evaluated nuclear data filename at the minimum. FRENDY has default values in the source code for nuclear data processing.
Users can give the parameters in the input file if they want to change the parameters. The original
input format is simple and does not require expert knowledge of nuclear data processing.
FRENDY can also treat the input files for NJOY Many users process the evaluated nuclear data
file with NJOY. FRENDY interprets the input as the NJOY compatible format when the firstparameter is the NJOY module name, e.g., MODER and RECONR. The available module is
MODER, RECONR, BROADR, PURR, UNRESR, THERMR, ACER, GROUPR, and MATXSR.
Note that the UNRESR module is not prepared in FRENDY. FRENDY calculates the effective
self-shielded cross-sections using the probability table method even if the user selects the UNRESR
module. Users can easily use FRENDY without changing the input files for NJOY. They cantherefore replace NJOY modules with FRENDY ones as they need. In addition, the modules of
FRENDY and NJOY can be used in combination. For example, users can generate the multi-groupcross-section data library using the GROUPR module of NJOY with the PENDF file generated by
FRENDY.

1.2

FRENDY Original Input Format

In the conventional processing code including NJOY and PREPRO, users must select the running
modules and prepare the input parameters for these modules to generate the cross-section data library.
FRENDY automatically generates the cross-section data library with the recommended processing
flow as shown in Fig. 1.2.1 when users select an appropriate processing mode at least. They do notneed to worry about what modules are required to generate the cross-section data library when theyuse FRENDY original input format. FRENDY also prepares the skip option to manually select themodules to be executed. Using this option, users can skip some modules in the recommended
processing flow.

- 4 -Nuclear data

Read and convert nuclear data

Resonance reconstruction

Doppler broadening

Dosimetry ACE fileCalculation of gas production
cross-sections

Generation of dosimetry ACE file

Generation of probability table

Processing thermal scattering law
data

Generation of fast ACE file

Generation of thermal ACE file

Fast ACE file

Thermal ACE file

Generation of GENDF, MATXS and
KRAM files
GENDF, MATXS, andKRAM files
Figure 1.2.1 Processing flow to generate ACE, GENDF, MATXS, and KRAM files- 5 -1.2.1 Processing Mode
The order of the input data is free except for the first parameter. Users must set the processing
mode as the first parameter. The available processing modes are listed in Table 1.2.1. The
representation of the ACE file processing mode is identical to NJOY.
The neutron-induced data and the thermal scattering law data are available to generate themulti-group cross-section file. The multi-group cross-section generation mode is not divided by theprocessing type since the FRENDY checks the input parameter and judges which type of data isprocessed.
FRENDY generates the multi-group cross-section file from the ACE file. The input file format is
not only the ENDF-6 formatted file but also the ACE formatted file. The start file format does not set
in the processing mode. The start file format is set in the input parameter “mg_start_file_mode”.
Table 1.2.1 Available processing mode
Processing mode name

Description

ace_file_generation_fast_mode
ace_fast_mode
ace_fast

Generation of the neutron incident ACE file

ace_file_generation_normal_mode
ace_normal_mode
ace_normale
ace_file_generation_thermal_scattering_mode
ace_file_generation_thermal_scatter_mode
ace_therm_mode

Generation of ACE file of the thermal

ace_therm

scattering law data

ace_file_generation_tsl_mode
ace_tsl_mode
ace_tsl
ace_file_generation_dosimetry_mode

Generation of ACE file of the dosimetry

ace_dosi_mode

data

ace_dosi
mg_generation_neutron_mode
mg_neutron_mode

Generation of the neutron-induced

mg_neutron

multi-group XS file.

mg_modemg- 6 -endf_file_modification_mode
endf_file_modify_mode

Modification of ENDF-6 formatted file.

endf_mod_mode

The explanation of the input format is

modify_endf_file_mode

shown in Sec. 1.7.

mod_endf_file_mode
mod_endf_mode
output_xs_mode

Output 1-D table data from ENDF,

plot_mode

GENDF, and ACE formatted file.

comp_mode

1.2.2 Input Parameters1.2.2.1 General information
The input data except the processing mode consists of “parameter name” and “parameter value”.
Users need to set the parameters if they want to modify the default values. FRENDY original input
format accepts comment lines. The C++ style comments are available, i.e., “//” for a single line
comment and “/* … */” for multi-line comments. FRENDY can read four types of data, i.e., integer,
real, string, and text, and the vector data of integer and real. The available types are as follows:
integer : integer number, e.g., -2, -1, 0, 1, 2,
real: real number, e.g., -1.0, 0.0, 1.0, 1.0E-1, 1.0E+1, 1.0D-1, 1.0D+1, 1.0-1, 1.0+1,

string : character data without space, e.g., nuclide_name, calculation-type,text: multiple lines character data.

Users have to set the vector data in a bracket when they want to set the vector data of integer andreal value. If users want to set three values, i.e., 1.0, 2.0, and 3.0, users write as follows:(1.0 2.0 3.0).
Since FRENDY can read data specified in multiple lines, the following input style is also acceptable
(

1.0
2.0
3.0

),

or

- 7 -

(
1.0
2.0
3.0
).
If users want to set the text data, they need to enclose single or double quotation marks. The text data
allows multiple lines as follows:“92-U-238 from JENDL-4.0
Processed with FRENDY
Processed day: 2017/10/13”
or‘92-U-238 from JENDL-4.0
Processed with FRENDY
Processed day: 2017/10/13’
If users want to use double quotation marks in the comment line, single quotation marks are used to
enclose the text data as follows:‘This is the comment line for the PENDF file.
Users can use the “double quotation marks” when the comment line is enclosed by the single
quotation marks.’
If users want to use single quotation marks in the comment line, double quotation marks are used to
enclose the text data as follows:“This is the comment line for the PENDF file.
Users can use the ‘single quotation marks’ when the comment line is enclosed by the double
quotation marks.”- 8 -1.2.2.2 Input parameter name and recommended value for common utilities

nucl_file_name
Data type:stringDefault value:

This parameter must be required for the processing.

Explanation of this parameter: ENDF file name
Sample of this parameter
nucl_file_name ( U235.dat )pendf_label_data
Data type:stringDefault value:

none

Explanation of this parameter: Label for new PENDF tape (Max. 66 words)
Sample of this parameter
pendf_label_data ( “PENDF file of U235 from JENDL-4.0” )comment_data
Data type:list(string)

Default value:

none

Explanation of this parameter: Comment line for the PENDF file.
FRENDY recognizes the comment line that is enclosed by double or single quotation.
If the user wants to use the double/single quotation in the comment line, please use thesingle/double quotation to enclose the comment.
Sample of this parameter
comment_data("This is the comment line for the PENDF file.
User can use the 'single quotation mark' when the comment line is enclosed by the double
quotation.")error
Data type:real

Default value:

1.00×10-3

Explanation of this parameter: Tolerance value for linearization.
Sample of this parameter
error 1.0E-3- 9 -error_max
Data type:real

Default value:error×10.0

Explanation of this parameter: Maximum tolerance value for linearization.
Sample of this parameter
error_max 1.0E-2

error_integral
Data type:real

Default value:error/20000.0

Explanation of this parameter: Maximum integral error for linearization.
Sample of this parameter
error_integral

5.0E-8

add_grid_data
Data type:list(real)

Default value:

none

Explanation of this parameter: Additional energy grid [eV].
Sample of this parameter
add_grid_data ( 0.625 1.000 1.0E+6 )temp or temperature
Data type:real

Default value:

293.6 [K]Explanation of this parameter: Temperature [K].
Sample of this parameter
temp 300.0

max_broading_ene
Data type:real

Default value:

1.00×106 [eV]

Explanation of this parameter: Maximum energy for the Doppler broadening [eV].
If the upper limit of the resolved resonance energy Eh is smaller than max_broading_ene, themaximum energy for the Doppler broadening is modified to Eh.
Sample of this parameter
max_broadning_ene

1.0E+6- 10 -

probability_bin_no
Data type:

integer

Default value:

20

Explanation of this parameter: Number of probability table bins
Sample of this parameterprobability_bin_no 20

ladder_no
Data type:

integer

Default value:

100

Explanation of this parameter: Number of resonance ladders for generating the probability
table.
Sample of this parameter
ladder_no 100

sigma_zero_data
Data type:list(real)

Default value:

1.0×1010, 1.0×106, 1.0×105, 10000.0, 1000.0, 100.0, 35.0, 10.0, 1.0, 0.1 [barns]

Explanation of this parameterσ0 values for the Bondarenko-type self-shielded cross-section in the unresolved resonanceregion [barns]. Maximum σ0 is considered as the σinf value.
The multi-group cross-section generation function can automatically set the backgroundcross-section. The input format for the automatic setting of the background cross-section is asfollows:
sigma_zero_data ( auto

err Nmax

σmin Target INT)

where,
err: Tolerance value.
Nmax: Maximum number of the background cross-section.
σmin: Minimum background cross-section [barns]
Target: Target of interpolation. (factor or rr)
INT: Integerpolation method (cubic or linear)rr: Interpolation error is estimated for reaction rate
linear: linear interpolation
cubic: monotone cubic interpolation.
Note that the background cross-section of PURR uses the recommended value when the user
selects the automatic setting of the background cross-section.
- 11 -

If the upper limit of the resolved resonance energy Eh is smaller than max_broading_ene, themaximum energy for the Doppler broadening is modified to Eh.
Sample of this parametersigma_zero_data(1.0E10

1.0E6 1.0E5 10000.0 1000.0 100.0 35.0 10.0 1.0 0.1)

sigma_zero_data(auto 0.1 50 1.0e-10 rr linear)

ene_grid_no_par_ladder or ene_grid_par_ladder orene _no_par_ladder
Data type:

integer

Default value:

10000

Explanation of this parameter: Sampling energy grid number for each ladder number.
Sample of this parameter
ene_grid_no_par_ladder 10000

random_seed
Data type:

integer

Default value:

12345

Explanation of this parameter: Random number seed for probability table generation.
Sample of this parameter
random_seed 11111

err_p_table or err_p_tab or err_ptab
Data type:real

Default value:

1.0E-2

Explanation of this parameter: Tolerance value for probability table generation.
If this parameter is used, the ladder_no parameter is disabled.
Sample of this parameter
err_p_table 0.01

ace_file_name
Data type:stringDefault value:“nucl_file_name”.ace

Explanation of this parameter: ACE file name.
Sample of this parameter
ace_file_name U235.ace- 12 -

ace_dir_file_name or ace_dir or mcnp_dir_file_name or mcnp_dir
Data type:stringDefault value:"nucl_file_name".ace.dir

Explanation of this parameter: MCNP directory information for the ACE file
When the ace_file_name is set, the default value is "ace_file_name".ace.dir.
Sample of this parameter
ace_dir_file_name U235.xsdir

suffix_id
Data type:real

Default value:.00

Explanation of this parameter: The suffix ID for the ACE file.
Sample of this parameter
suffix_id 0.50

ace_label_data
Data type:stringDefault value:

none

Explanation of this parameter: Label for the ACE file. (Max. 70 words)
Sample of this parameter
ace_label_data“PENDF file of U235 from JENDL-4.0”

iz_aw_data
Data type:list(real)

Default value:

none

Explanation of this parameter: The list of (iz, aw) pairs for the ACE file. (iz=1000.0×Z+A,
aw=mass)
Sample of this parameter
iz_aw_data(92235.0 2.330250E+2)

cumulative_angle_distribution_format
Data type:string or integer

Default value:

string = yes, integer = 1

Explanation of this parameter: Calculation option of whether new cumulative angular
distributions for ACE file generation are used or not.
Available value is use, yes, no, default, 0, and 1 (0=no, 1=use/yes).
- 13 -

whereuse, yes, default, 1: Use the new cumulative angular distribution.
no, 0: Do not use the new cumulative angular distribution.
Sample of this parameter
cumulative_angle_distribution_format

yes- 14 -

1.2.2.3 Input parameter name and recommended value which are used only for thermal
scattering law data

nucl_file_name_tsl
Data type:stringDefault value:

This parameter must be required for the processing.

Explanation of this parameter: ENDF file name for thermal scattering law data.
Sample of this parameter
nucl_file_name_tsl ( HinH2O.txt )equi_probable_angle_no
Data type:

integer

Default value:

10

Explanation of this parameter: Number of equiprobable angles for thermal scattering law data.
Sample of this parameter
equi_probable_angle_no 10

principal_atom_no
Data type:

integer

Default value:

FRENDY automatically set from the ENDF file.

Explanation of this parameter: Number of principal atoms for thermal scattering law data.
Sample of this parameter
principal_atom_no 1

atom_type_no
Data type:

integer

Default value:

FRENDY automatically set from the ENDF file.

Explanation of this parameter: Number of atom types in the mixed moderator.
Sample of this parameter
atom_type_no 1

inelastic_reaction_type_no
Data type:

integer

Default value:

221

Explanation of this parameter: Reaction type (MT) number for the inelastic reaction.
Sample of this parameterinelastic_reaction_type_no1- 15 -

max_thermal_ene
Data type:real

Default value:max(10.0, temp/300.0)

Explanation of this parameter: Maximum energy for thermal treatment.
Sample of this parameter
max_thermal_ene

10.0

thermal_za_id_name
Data type:stringDefault value:

ZA value of the ENDF file.

Explanation of this parameter: ZA ID name for the thermal ACE file (ZA=1000.0×Z+A, Max. 6
words).
Sample of this parameter
thermal_za_id_name lwtr

moderator_za_data
Data type:

integer

Default value:

none

Explanation of this parameter: Moderator component ZA value.
Sample of this parameter
moderator_za_data 1001

- 16 -

1.2.2.4 Input parameter name and recommended value which are used only for
multi-group cross-section generation

mg_tsl_data_type
Data type:

list (string)

Default value:

This parameter must be required for the processing of the TSL data.

Explanation of this parameter: The S(α,β) type for the MATXS file.
The S(α,β) type used in FRENDY is shown in Table 1.2.2. The S(α,β) type has no impact on
the GENDF file. Please set “free” when the user only generates the GENDF file.
Sample of this parameter
mg_tsl_data_type ( hh2o )mg_tsl_data_type ( free )Table 1.2.2 S(α,β) type name list and corresponding material name

Material name

S(α,β) type

Material name

S(α,β) typeAlal

H in YH2

hyh2

Be

be

H in ZrH

hzrh

Be in BeO

bebeo

Liquid Methane (CH4)

lch4

Benzenebenz

N in UNnun

C in SiC

csic

O in BeO

obeo

C5O2H8c5o2h8O in D2O

od2o

D in D2O

dd2o

O in ICE (H2O)

oice

Ortho-DdorthoO in UO2

ouo2

Para-D

dpara

Polyethylene (CH2)

poly

Fe

fe

Solid Methane (CH4)sch4

Graphitegraph

Si in SIC

sisic

H in H2O

hh2o

U in UNuun

H in Ice (H2O)

hice

U in UO2

uuo2

Ortho-H

horthoY in YH2

yyh2

Para-Hhpara

Zr in ZrH

zrzrh

- 17 -

mg_file_name
Data type:stringDefault value:

input file name

Explanation of this parameter: Multi-group XS file name.
The output file name is as follows:"mg_file_name"_"mg_file_mode"_"ZAID of the ACE file".mg
Sample of this parameter
mg_file_name “U235”

mg_edit_mode or mg_edit_option or mg_edit_xs
Data type:list(string)

Default value:

GENDF

Explanation of this parameter: Output format and output data of the multi-group cross-section
generation.
The available output format and output data are as follows:
MATXS, SimpleMATXS, GENDF, FullMATXS, SimpleGENDF, FullGENDF, KRAMXS,
1DXS, 2DXS, NuChi, MGFlux, MGCurrent, UFG, UFG1DXS, and RUC.
where
MATXS: Microscopic cross-sections in MATXS format, consistent with NJOY2016.
SimpleMATXS: Microscopic cross-sections in MATXS format, consistent with NJOY99.
FullMATXS: Microscopic cross-sections in MATXS format without truncation ofshielded cross-sections. Output file size may be large.
GENDF: Microscopic cross-sections in GENDF format consistent with NJOY2016.
SimpleGENDF: microscopic cross-sections in GENDF format, consistent with NJOY99.
FullGENDF: microscopic cross-sections in GENDF format without truncation ofshielded cross-sections. Output file size may be large.
KRAMXS: microscopic cross-sections in KRAM format.
1DXS: One-dimensional cross-sections such as total, fission, radiative capture.
2DXS: Two-dimensional cross-sections such as elastic scattering, inelastic scattering,(n,2n) reaction.
NuChi: Nu-value and fission spectrum.
MGFlux: Multi-group flux (group integrated values).
MGCurrent: Multi-group current (group integrated values).
UFG: Ultra-fine group spectrum, total cross-sections, the slowing down source, totalsource.
- 18 -

UFG1DXS: One-dimensional ultra-fine group cross-sections such as total, fission,
radiative capture.
RUC: The 0 K scattering cross section data. This data is used for the input file of theresonance upscattering resonance up-scattering correction (reso_upscat).
The user can select the specified MT number when 1DXS, 2DXS, and UFG1DXS options are
selected as follows:"1DXS 1, 2, 4, -50"
The minus value for MT means all MT numbers between the previous MT number.
Sample of this parameter
mg_edit_mode “U235”

mg_start_file_mode
Data type:stringDefault value:

ENDF

Explanation of this parameter: Format of the start file.
FRENDY generates the multi-group cross-section file from the ENDF-6 formatted file and theACE file. The user selects which data is used to generate the multi-group cross-section file.
The available value is ENDF and ACE.
ENDF: A multi-group cross-section file is generated from the ENDF-6 formatted file.
ACE: A multi-group cross-section file is generated from the ACE file.
Sample of this parameter
mg_start_file_mode ENDF

mg_label_data
Data type:stringDefault value:

ENDF

Explanation of this parameter: The label of the multi-group cross-section file.
The label is output on the first line of the GENDF file.
Sample of this parameter
mg_label_data“Multi-group XS file of U235”

legendre_order
Data type:

integer

Default value:

3

Explanation of this parameter: Maximum Legendre order (PL order)- 19 -

Sample of this parameterlegendre_order3

max_thermal_ene_e_out
Data type:real

Default value:

max_thermal_ene

Explanation of this parameter: Maximum energy of the thermal treatment for the outgoingparticle energy [eV].
Sample of this parameter
max_thermal_ene_e_out

10.0

mg_weighting_spectrum_mode
Data type:stringDefault value:

1/E

Explanation of this parameter: Weighting spectrum.
The available output format and output data are 1/E and Fission+1/E+Maxwell.
1/E: 1/E spectrum for the whole energy range.
Fission+1/E+Maxwell: Fission for fast energy range, 1/E for intermediate energy range,
Maxwell for the thermal energy range.
The other parameters are required to use "Fission+1/E+Maxwell" option as follows:
mg_weighting_spectrum( Fission+1/E+Maxwell Eh El Tfis

E1 E2 )

where
Eh: Highest energy [eV],
El: Lowest energy [eV],
Tfis: Fission temperature [eV],E1: Energy boundary between fission and 1/E spectra [eV],E2: Energy boundary between 1/E and Maxwell spectra [eV].
The sample of "Fission+1/E+Maxwell" is as follows:
mg_weighting_spectrum( Fission+1/E+Maxwell 2.0e+7 1.0e-5 1.6e+6 1.0e+6 0.625)
The default value of Eh, El, Tfis, E1, and E2 are 2.0e+7, 1.0e-5, 1.6e+6, 1.0e+6, and 0.0. When E2
is 0.0, E2 is automatically set by this program.
Sample of this parameter
mg_weighting_spectrum_mode“1/E”- 20 -

mg_weighting_spectrum_data or mg_weighting_spectrum
Data type:

list (real)

Default value:

none

Explanation of this parameter: Weighting spectrum. The user can manually set the weightspectrum.
The format of the weight spectrum is similar to the TAB1 record in the ENDF-6 format asfollows:
mg_weighting_spectrum_data(E1 W1

E2 W2 … Ei-1 Wi-1 Ei Wi)

where
Ei is the ith value of energy [eV],
Wi is the ith value of weight.
NJOY's weight spectrum, i.e., the “iwt” number of the GROUPR module and the weighted
spectrum name shown in Table 1.2.3, is also available. The input format to select the NJOY's
weight spectrum is as follows:
mg_weighting_spectrum_data( iwtXX )where XX is the “iwt” number (iwt02 ~ iwt12).
Sample of this parameter
mg_weighting_spectrum_data (1.0E-5 1.0 2.0E7 1.0)
mg_weighting_spectrum_data ( iwt06 )mg_weighting_spectrum_data ( epri-cell-lwr )- 21 -

Table 1.2.3 Weighted spectrum name list and corresponding “iwt” number

iwt

Weighting spectrum nameiwt

constant

Weighting spectrum namevitamine-e

2
const

vit-e
11

3

1/e

ornl-5505

4

fission+1/e+maxwell

vitamine-e-with-t-dependentepri-cell-lwr

vitamine-e-with-t-depend

epri-cellvitamine-e-with-t

5

epri

vit-e-with-t-dependent
12

claw-weight-function

vit-e-with-t-depend

claw-weight

vit-e-with-t

claw-with-t-dependent

vit-e-w-t

9claw-with-t-depend
10claw-with-t
claw-tmg_weighting_spectrum_data_int or mg_weighting_spectrum_int
Data type:

list (integer)

Default value:

2 (linear-linear)

Explanation of this parameter: Interpolation method of the weight spectrum.
The format of the weight spectrum is similar to the TAB1 record in the ENDF-6 format asfollows:
mg_weighting_spectrum_data_int(NBT1 INT1 NBT2 INT2 … NBTi-1 INTi-1

NBTi INTi)

where
NBTi is the value of n separating the ith and (i+1)th interpolation range,
INTi is the interpolation scheme identification number used in the ith range.
The available interpolation scheme is 1 (const), 2 (linear-linear), 3 (linear-log), 4 (log-liner), and 5(log-log).
Sample of this parameter
mg_weighting_spectrum_data_int ( 2 1 ) //NBT=2, INT=1 (const)

mg_structure or mg_structure_neutron
Data type:

list (real)

Default value:

XMAS 172-group structure- 22 -

Explanation of this parameter: Energy group structure.
The user can manually set the energy group structure [eV].
NJOY's energy group structure, i.e., the “ign” number of the GROUPR module and the energy
group structure name shown in Table 1.2.4, is also available. The input format to select the
NJOY's energy group structure is as follows:
mg_structure( ignXX )where XX is the ign number (ign02 ~ ign34, ign101).
The energy group structure can be automatically set by the user input.
The input format for the automatic setting of the energy group structure is as follows:
mg_structure ( auto E1 N1 O1

E2 N 2 O 2... Ei-1 Ni-1 Oi-1

Ei N i O i

Ei+1 )

where
Ei: Energy boundary i [eV],Ni: Number of divisions,
Oi: Option for divisions.
The available option for divisions Oi is as follows:
EL: Divide the energy range by equi-lethargy width,
EE: Divide the energy range by equi-energy width.
Sample of this parameter
mg_structure ( ign18 ) //XMAS 172-group structure
mg_structure ( 1.0E-5 0.625 2.0E+7 ) //2-group structure
mg_structure ( auto
2.0e+07 3200

EL

100.0

EL

1

1.0e-5 )mg_structure_g or mg_structure_gam or mg_structure_gamma
Data type:

list (real)

Default value:

none

Explanation of this parameter: Photon (gamma) production energy group structure.
The user can manually set the energy group structure [eV].
If the user does not set this parameter, the photon production data generation is skipped.
NJOY's energy group structure, i.e., the “igg” number of the GROUPR module and the energy
group structure name shown in Table 1.2.5, is also available. The input format to select the
NJOY's energy group structure is as follows:
mg_structure( iggXX )- 23 -

where XX is the igg number (igg02 ~ igg10).
Table 1.2.4 Energy group structure name list and corresponding “ign” number

ign
2
3
4
5

6

Energy groupstructure namecsewg-239
csewg
lanl-30

8

9

13
14

anl-27anl

Energy groupstructure namelanl-80eurlib-100
eurlibsand-iia-640

15

sand-iia

rrd-50sand-640

rrd

vitamin-e-174

gam-i-68

vitamin-e

gam-i
gam-687ign

16vitamin-174
vit-e-174

gam-ii-100

vit-e

gam-iivit-174

gam-100

vitamin-j-175

laser-thermos-35

vitamin-jlaser-thermos
laser

17

vitamin-175

22
23
24
25
26
27
28

Energy groupstructure namexmas-lwpc-172
xmas-lwpc
vit-j-lwpc-175
vit-j-lwpc
shem-cea-281
shem-cea
shem-epm-295
shem-cea-epm-361
shem-cea-epm
shem-epm-315
rahab-aecl-89
rahab-aecl

28
29

rahab
ccfe-660
ccfe

vit-j-175

30

ukaea-1025

laser-35

vit-j

31

ukaea-1067

epri-cpm-69

vit-175

32

ukaea-1102

epri-cpm

xmas-nea-lanl-172

33

ukaea-142

xmas-nea-lanl

34

lanl-618

epri
epri-69

18

lanl-187

11

lanl-70

19

ecco-33

sand-ii-620

20

ecco-1968

sand-ii
sand-620

xmas-nea

21

vitamin-b6-199

xmas-lanl

10

12

ign

tripoli-315
tripoli- 24 -

101

vitamin-199
vit-b6-199
vit-199

Table 1.2.5 Energy group structure name list and corresponding “igg” number

igg

Energy groupigg

structure name2
3

structure namecsewg-94

vitamin-e-38

csewg

vitamin-e

lanl-129steiner-21

4Energy groupvitamin-38
vit-e-38

steinervit-estraker-22

vit-38straker

vitamin-j-42

6lanl-48

vitamin-j

7lanl-245

10vitamin-c-36

8vitamin-42
vit-j-42

vitamin-c

vit-jvitamin-36

vit-42

vit-c-36
vit-c
vit-36

mg_ufg_structure or mg_ultra_fine_group_structure
Data type:

list (real)

Default value:( 2.0e+07

10000

EL

52475.0

56000

EL9118.812000

EL4307.412000

EL961.128000

EL130.0712000

EL0.3224210000

EL1.0e-5 )Explanation of this parameter: Energy group structure to generate neutron flux in the ultra-finegroup structure.
The format of this parameter is as follows:
mg_ufg_structure( E1 N1 O1 E2 N2 O2... Ei-1 Ni-1 Oi-1 Ei Ni Oi Ei+1 )where
Ei: Energy of ith boundary [eV],Ni: Number of divisions,- 25 -

Oi: Option for divisions
The available option for divisions Oi is as follows:
EL: Divide the energy range by equi-lethargy width.
EE: Divide the energy range by equi-energy width.
Sample of this parameter
mg_ufg_structure( 2.0e+07

10000

EL

52475.0

56000

EL9118.812000

EL4307.412000

EL961.128000

EL130.0712000

EL0.3224210000

EL1.0e-5 )mg_number_density
Data type:

list (real)

Default value:

1.0 (use only one nuclide)

Explanation of this parameter: Number density of each nuclide [1/barn/cm].
If the user processes the one nuclide data, this parameter can be skipped.
Sample of this parameter
mg_number_density( 1.0 )

mg_flux_calc_mode
Data type:stringDefault value:

SLD

Explanation of this parameter: Calculation option at the resonance region.
The available option at the resonance region is as follows:
NR: Use narrow resonance calculation.
SLD: Use ultra-fine group slowing down calculation.
Sample of this parameter
mg_flux_calc_mode

SLD

mg_flux_calc_w_eh_el
Data type:

list (real)

Default value:

W=0.999167, Eh=1.0×104, El=1.0- 26 -

Explanation of this parameter: Calculation condition at the resonance region.
The format of this parameter is as follows:
mg_flux_calc_w_eh_el( W Eh El )where
W: Atomic weight (relative to neutron mass) used for the background moderator nuclide,
Eh: Upper energy to use ultra-fine group slowing down spectrum [eV],
El: Lower energy to use ultra-fine group slowing down spectrum [eV].
Sample of this parameter
mg_flux_calc_w_eh_el( 0.999167 1.0E4 1.0 )mg_thermal_upscatter_treatment or mg_thermal_xs_treatment
Data type:stringDefault value:on

Explanation of this parameter: Option for calculation of the thermal scattering using the freegas model.
Available value is use, on, yes, off, no, default, and def
on, yes, default, def: Calculation of the thermal scattering cross-section
off, no: Skip calculation of the thermal scattering cross-section
Sample of this parameter
mg_thermal_upscatter_treatment

onreso_upscat or reso_up_scat or reso_upscat_data or reso_up_scat_data
Data type:

list (string)

Default value:

none

Explanation of this parameter: The file name list of the 0 K scattering cross section data.
If the user wants to consider the resonance upscattering correction, please use this option. The
user has to set the file name of the 0 K scattering cross section data which is generated by the
RUC option in mg_edit_option.
Sample of this parameter
reso_upscat ( U235_RUC_MT2.mg )reso_upscat ( skip U235_RUC_MT2.mg U238_RUC_MT2.mg )//Skip means that this nuclide is not considered the resonance upscattering correlation

reso_upscat_mode or reso_up_scat_mode
Data type:stringDefault value:ALL- 27 -

Explanation of this parameter: Calculation option for the resonance up-scattering correction.
ALL: Resonance up-scattering correction is considered not only for 1D effective cross sections
but also for the elastic scattering matrix.
XS: Resonance up-scattering correction is considered only for 1D effective cross sections. No
resonance up-scattering is considered for the elastic scattering matrix.
Sample of this parameter
reso_upscat_mode ( ALL )potential_scat_xs
Data type:

list (real)

Default value:

none

Explanation of this parameter: Potential scattering cross-section of each nuclide [barns].
If the user does not set this parameter, FRENDY uses the potential scattering cross-sections
obtained by JENDL-4.0, ENDF/B-VIII.0, or JEFF-3.3.
Sample of this parameter
potential_scat_xs

1.15825E+01 //Potential scattering XS of U-235[barns]

mg_mat_no or mg_mat_list or mg_mat_no_list
Data type:

list (integer)

Default value:

none

Explanation of this parameter: The MAT number of the GENDF file.
This parameter is only required when the ACE file is used as the start format of the multi-groupcross-section generation since the ACE file does not have the MAT number.
Sample of this parameter
mg_mat_no( 825 9228 9237 ) //O16, U235, U238

ace_output_option or ace_edit_option
Data type:stringDefault value:on

Explanation of this parameter: Output ACE file option until the multi-group cross-section
generation.
Available value is on, output, edit, off, no, skip, default, def.
on, output, edit, default, def: Output the ACE file.
off, no: Do not output the ACE file.
Sample of this parameter
ace_output_option output- 28 -

1.2.2.5 Input parameter name and recommended value which are used only for XS
output tool

gendf_ name or gendf_file_name
Data type:stringDefault value:

none

Explanation of this parameter: GENDF formatted file name to output the (x, y) table.
Sample of this parameter
gendf_name ( "./gendf/U235.gendf" )output_name or output_file_name
Data type:stringDefault value:

none

Explanation of this parameter: Output file name for XS output tool.
Sample of this parameter
output_name ( "./output_xs/xs_data_01" )mat_no or mat
Data type:

integer

Default value:

none

Explanation of this parameter: Specified MAT number to output cross section.
This option is available for ENDF and GENDF formatted files.
Sample of this parametermat_no ( 9228 )mf_list or mf_data
Data type:vector<integer>

Default value:

3

Explanation of this parameter: Selection of output MF data.
This option is only available for a GENDF formatted file.
Sample of this parametermf_list ( 3 6 )mt_list or mt_data_list or xs_type_list or reaction_type_list
Data type:vector<integer>

Default value:

none (all)Explanation of this parameter: Selection of output MT data. If the user does not set this option,- 29 -

this program outputs all cross section data.
Sample of this parametermt_list ( 1 2102 )mt_list ( 1 250 - 102 ) //1, 2, 50, 51, 52, 53, …, 99, 100, 101, 102

mt_list ( -102 ) //1, 2, 3, …, 99, 100, 101, 102
mt_list ( “all” ) //To output all MT data

edit_flag or edit_mode
Data type:stringDefault value:

histogram

Explanation of this parameter: Energy grid data option for multi-group cross section data.
Sample of this parameter
edit_flag( "histogram" )If the user selects the non-histogram case, this tool only outputs cross sections at the lowest
energy of each energy group, such as (Eli, XSi), (Eli+1, XSi+1). If the user selects the histogram case,this tool outputs cross sections at the lowest and highest energy of each energy group, such as (Eli,XSi), (Ehi, XSi), (Eli+1, XSi+1), (Ehi+1, XSi+1). Figure 1.2.2 shows an example of the difference
between histogram and non-histogram plots using Excel. The histogram option is useful for plotting
the multi-group cross section distribution by an Excel scatter plot.

1,000.00

Cross section [barn]

100.00histogram
non-histogram

10.00
1.00
0.10
0.01
1,000

Incodent neutron energy [eV]10,000Figure 1.2.2 Difference between histogram and non-histogram plots- 30 -

1.2.2.6 Optional input parameter

print_set_data_resonance_reconstruction or print_set_data_linearize
Data type:stringDefault value:on

Explanation of this parameter: Output input information for the resonance reconstruction.
Available value is on, yes print write output, off, no, skip, default, and def.
on, yes, print, write, output, default, def: Output the input information.
off, no, skip: Do not output the input information.
Sample of this parameter
print_set_data_resonance_reconstruction on

process_resonance_reconstruction or process_linearize
Data type:stringDefault value:on

Explanation of this parameter: Processing option for the resonance reconstruction.
Available value is on, yes process, run, off, no, skip, default, and def.
on, yes, process run, default, def: Process the resonance reconstruction.
off, no, skip: Do not process the resonance reconstruction.
Sample of this parameter
process_resonance_reconstructiononwrite_pendf_resonance_reconstruction or write_pendf_linearize
Data type:stringDefault value:

off

Explanation of this parameter: Output PENDF file option for the resonance reconstruction.
Available value is on, yes, print, write, output, off, no, skip, default, and def.
on, yes, print, write, output: Output the PENDF file of the resonance reconstruction.
on, yes, print, write, output + PENDF file name: Output the PENDF file of the resonancereconstruction with a specific name.
off, no, skip, default, def: Do not output the PENDF file of the resonance reconstruction.
Sample of this parameter
write_pendf_resonance_reconstruction ( on “U235_reso.pendf” )- 31 -

pendf_file_name_resonance_reconstruction orpendf_file_name_linearize
Data type:stringDefault value:

none

Explanation of this parameter: The PENDF file name for the resonance reconstruction.
The write_pendf_linearize parameter is changed from off to on when this parameter is set.
Sample of this parameter
pendf_file_name_resonance_reconstruction ( “U235_reso.pendf” )print_set_data_doppler_boroadening or print_set_data_dop
Data type:stringDefault value:on

Explanation of this parameter: Output input information for the Doppler broadening.
Available value is on, yes print write output, off, no, skip, default, and def.
on, yes, print, write, output, default, def: Output the input information.
off, no, skip: Do not output the input information.
Sample of this parameter
print_set_data_doppler_broadening on

process_doppler_broadening or process_dop
Data type:stringDefault value:on

Explanation of this parameter: Processing option for the Doppler broadening.
Available value is on, yes process, run, off, no, skip, default, and def.
on, yes, process run, default, def: Process the Doppler broadening.
off, no, skip: Do not process the Doppler broadening.
Sample of this parameter
process_doppler_broadeningonwrite_pendf_doppler_broadening or write_pendf_dop
Data type:stringDefault value:

off

Explanation of this parameter: Output PENDF file option for the Doppler broadening.
Available value is on, yes, print, write, output, off, no, skip, default, and def.
on, yes, print, write, output: Output the PENDF file of the Doppler broadening.
on, yes, print, write, output + PENDF file name: Output the PENDF file of the Doppler- 32 -

broadening with a specific name.
off, no, skip, default, def: Do not output the PENDF file of the Doppler broadening.
Sample of this parameter
write_pendf_doppler_broadening( on “U235_dop.pendf” )pendf_file_name_doppler_broadening or pendf_file_name_dop
Data type:stringDefault value:

none

Explanation of this parameter: The PENDF file name for the Doppler broadening.
The write_pendf_dop parameter is changed from off to on when this parameter is set.
Sample of this parameter
pendf_file_name_doppler_broadening ( “U235_dop.pendf” )print_set_data_gas_production_cross_section or print_set_data_gas_xs
Data type:stringDefault value:on

Explanation of this parameter: Output input information for the calculation of the gas
production cross-section.
Available value is on, yes print write output, off, no, skip, default, and def.
on, yes, print, write, output, default, def: Output the input information.
off, no, skip: Do not output the input information.
Sample of this parameter
print_set_data_gas_production_cross_section on

process_gas_production_cross_section or process_gas_xs
Data type:stringDefault value:on

Explanation of this parameter: Processing option for the calculation of the gas production
cross-section.
Available value is on, yes process, run, off, no, skip, default, and def.
on, yes, process run, default, def: Process the calculation of the gas production cross-section.
off, no, skip: Do not process the calculation of the gas production cross-section.
Sample of this parameter
process_gas_xs on- 33 -

write_pendf_gas_production_cross_section or write_pendf_gas_xs
Data type:stringDefault value:

off

Explanation of this parameter: Output PENDF file option for the calculation of the gas
production cross-section.
Available value is on, yes, print, write, output, off, no, skip, default, and def.
on, yes, print, write, output: Output the PENDF file of the calculation of the gas production
cross-section.
on, yes, print, write, output + PENDF file name: Output the PENDF file of the calculation of
the gas production cross-section with a specific name.
off, no, skip, default, def: Do not output the PENDF file of the calculation of the gas production
cross-section.
Sample of this parameter
write_pendf_gas_xs ( on “U235_gas_xs.pendf” )pendf_file_name_gas_production_cross_section orpendf_file_name_gas_xs
Data type:stringDefault value:

none

Explanation of this parameter: The PENDF file name for the calculation of the gas production
cross-section.
The write_pendf_gas_xs is changed from off to on when this parameter is set.
Sample of this parameter
pendf_file_name_gas_xs ( “U235_gas_xs.pendf” )print_set_data_probability_table or print_set_data_prob_table
Data type:stringDefault value:on

Explanation of this parameter: Output input information for the probability table generation.
Available value is on, yes print write output, off, no, skip, default, and def.
on, yes, print, write, output, default, def: Output the input information.
off, no, skip: Do not output the input information.
Sample of this parameter
print_set_data_probability_table on- 34 -

process_probability_table or process_prob_table
Data type:stringDefault value:on

Explanation of this parameter: Processing option for the probability table generation.
Available value is on, yes process, run, off, no, skip, default, and def.
on, yes, process run, default, def: Process the probability table generation.
off, no, skip: Do not process the probability table generation.
Sample of this parameter
process_probability_tableonwrite_probability_table or write_pendf_prob_table
Data type:stringDefault value:

off

Explanation of this parameter: Output PENDF file option for the probability table generation.
Available value is on, yes, print, write, output, off, no, skip, default, and def.
on, yes, print, write, output: Output the PENDF file of the probability table generation.
on, yes, print, write, output + PENDF file name: Output the PENDF file of the probability table
generation with a specific name.
off, no, skip, default, def: Do not output the PENDF file of the probability table generation.
Sample of this parameter
write_pendf_probability_table ( on “U235_ptab.pendf” )pendf_file_name_probability_table or pendf_file_name_prob_table
Data type:stringDefault value:

none

Explanation of this parameter: The PENDF file name for the probability table generation.
The write_pendf_prob_tablle parameter is changed from off to on when this parameter is set.
Sample of this parameter
pendf_file_name_probability_table ( “U235_ptab.pendf” )print_set_data_thermal_scattering_law or print_set_data_tsl
Data type:stringDefault value:on

Explanation of this parameter: Output input information for the calculation of the thermal
scattering law data.
Available value is on, yes print write output, off, no, skip, default, and def.
- 35 -

on, yes, print, write, output, default, def: Output the input information.
off, no, skip: Do not output the input information.
Sample of this parameter
print_set_data_thermal_scattering_law

onprocess_thermal_scattering_law or process_tsl
Data type:stringDefault value:on

Explanation of this parameter: Processing option for the calculation of the thermal scattering
law data.
Available value is on, yes process, run, off, no, skip, default, and def.
on, yes, process run, default, def: Process the calculation of the thermal scattering law data.
off, no, skip: Do not process the calculation of the thermal scattering law data.
Sample of this parameter
process_thermal_scattering_law on

write_pendf_thermal_scattering_law or write_pendf_tsl
Data type:stringDefault value:

off

Explanation of this parameter: Output PENDF file option for the calculation of the thermal
scattering law data.
Available value is on, yes, print, write, output, off, no, skip, default, and def.
on, yes, print, write, output: Output the PENDF file of the calculation of the thermal scattering
law data.
on, yes, print, write, output + PENDF file name: Output the PENDF file of the thermal
scattering law data with a specific name.
off, no, skip, default, def: Do not output the PENDF file of the calculation of the thermal
scattering law data.
Sample of this parameter
write_pendf_thermal_scattering_law( on “U235_tsl.pendf” )pendf_file_name_thermal_scattering_law or pendf_file_name_tsl
Data type:stringDefault value:

none

Explanation of this parameter: The PENDF file name for the calculation of the thermal
scattering law data.
- 36 -

The write_pendf_tsl is changed from off to on when this parameter is set.
Sample of this parameter
pendf_file_name_thermal_scattering_law( “U235_tsl.pendf” )print_set_data_ace_data_generator or print_set_data_ace
Data type:stringDefault value:on

Explanation of this parameter: Output input information for the ACE file generation.
Available value is on, yes, print, write, output, off, no, skip, default, and def.
on, yes, print, write, output, default, def: Output the input information.
off, no, skip: Do not output the input information.
Sample of this parameter
print_set_data_ace_data_generator on

process_ace_data_generator or process_ace
Data type:stringDefault value:on

Explanation of this parameter: Processing option for the ACE file generation.
Available value is on, yes, process, run, off, no, skip, default, and def.
on, yes, process, run, default, def: Process the ACE file generation.
off, no, skip: Do not process the ACE file generation.
Sample of this parameter
process_ace_data_generator on- 37 -

1.2.2.7 Other input parameter

nan_err_mes_opt or err_mes_nan_opt or err_nan_opt or nan_opt
Data type:stringDefault value:

runtime_error

Explanation of this parameter: Policy for dealing with NaN (Not a Number) when data is
output to ENDF, PENDF, and GENDF file.
Available value is runtime, runtime_error, error, caution, warning, default, and def.
runtime, runtime_error, error, default, def: The process is aborted when the “NaN” value isfound.
caution, warning: The process is not aborted and the “NaN” value is converted -99999999.0
when the “NaN” value is found.
Sample of this parameter
nan_err_mes_opt runtime_error

1.2.2.8 Input parameter name and recommended value list
The input parameter name and its recommended value for the FRENDY original input format arelisted in Tables 1.2.6 - 1.2.16. Note the parameters and explanation of them are identical to Sections
1.2.2.2 - 1.2.2.7.
The parameters shaded in Tables 1.2.6 - 1.2.15 always need input. In the default option, FRENDY
does not output the PENDF file and users do not need to select the modules to be executed. Users
have to set the print and process flags as listed in Tables 1.2.8 - 1.2.15 when they want to write the
PENDF file or skip the modules. FRENDY can recognize synonymous words. For example,
FRENDY prepares “ene_grid_no_per_ladder”, “ene_grid_per_ladder”, and “ene_no_per_ladder” as
the parameter name to change the number of ladders to generate the probability table and “on”,“yes”, “print”, “write”, and “output” to write the PENDF file.
Table 1.2.16 shows the optional input parameter. The input parameters in Table 1.2.16 are used
for debug and checking the processing results.

- 38 -

Table 1.2.6 Input parameter name and recommended value for common parameter
Parameter nameType

Recommended value

nucl_file_namestring-pendf_label_data

stringnone

Description
ENDF file nameLabel for new PENDF tape(Max. 66 words)
Descriptive comments for PENDF file (MF=1, MT=451)

FRENDY recognize the comment line that is enclosed bydouble or single quotation.
If user wants to use double/single quotation in the comment- 39 -

comment_data

list(string)none

line, please use single/double quotation to enclose thecomment line.

Ex)"This is the comment line for the PENDF file.
User can use the 'single quotation mark' when the comment
line is enclosed by the double quotation."
error

real

1.00×10-3

Tolerance value for linearization

error_max

realerror×10.0

Maximum tolerance value for linearization

error_integral

real

error/20000.0

Maximum integral error for linearization

add_grid_data

list (real)

none

Additional energy grid [eV]

temp
or

real

293.6

Temperature [K]max_broading_ene

real

1.00×106

Maximum energy for the Doppler broadening [eV]probability_bin_no

integer

20

ladder_no

integer100

integer

10000

Sampling energy grid number for each ladder number

random_seed

integer

12345

Random number seed for probability table generation

err_p_table

real

1.00E-02

ace_file_namestring"nucl_file_name".ace

ace_label_data

stringnone

iz_aw_data

list (real)none

temperature

Number of probability table bins
Number of resonance ladders for generating the probability
table

ene_grid_no_par_ladderor
ene_grid_par_ladder- 40 -

orene_no_par_ladder

Tolerance value for probability table generation.
If this parameter is used, the ladder_no parameter is desabled.
ACE file nameLabel for ACE file(Max. 70 words)
list of (iz, aw) pairiz=1000.0×Z+A, aw=mass

σ0 values for the Bondarenko-type self-shielded cross sectionin the unresolved resonance region [barns]
Maximum σ0 is considered as the σinf value.
The multi-group cross-section generation function can
automatically set the background cross-section.
The input format for the automatic setting of the backgroundcross-section is as follows:
sigma_zero_data ( auto "error" "maximum number of
1.0×1010, 1.0×106,- 41 -

sigma_zero_data

list (real)

1.0×105, 10000.0,
1000.0, 100.0,
35.0, 10.0, 1.0, 0.1

cross-section" "minimum background cross-section [b]""target of interpolation (factor/rr)" "interporation method(cubic/linear)")factor: Interpolation error is estimated for self-shielding factor
rr: Interpolation error is estimated for reaction rate
linear: linear interpolation
cubic: monotone cubic interpolation

The sample of input is as follows:
sigma_zero_data(auto 0.1 50 1.0e-10 rr linear)

Note that the background cross-section of PURR uses therecommended value when the user selects the automatic
setting of background cross-section.

ene_grid_no_par_ladderor
ene_grid_par_ladderinteger

10000

Sampling energy grid number for each ladder number

or
ene_no_par_ladder
ace_dir_file_name"nucl_file_name".ace.dir

oror

ace_diror

stringmcnp_dir_file_name

MCNP directory information for ACE file- 42 -

is changed by user)

mcnp_dirreal
stringcumulative_angle_distribution_format(When theace_file_nameorsuffix_id"ace_file_name".ace.dirorinteger.00

Suffix ID for ACE fileCalculation option of whether new cumulative angularstring = yes,

distributions for ACE file generation are used or not

integer = 1

Available value is use, yes, no, default or 0-1(0=no, 1=use/yes)Table 1.2.7 Input parameter name and recommended value which are used only for thermal scattering law data
Parameter nameType

Recommended value

nucl_file_name_tsl

string-equi_probable_angle_no

integer

10

Description
ENDF file name for thermal scattering law data
Number of equi-probable angles for thermal scattering law
data
Number of principal atoms for thermal scattering law data

principal_atom_no

integer-(If user doesn't set this parameter, FRENDY automatically set
from the ENDF file)- 43 -

inelastic_reaction_type_no

integer

221

max_thermal_ene

realmax(10.0, temp/300.0)

thermal_za_id_namestring

ZA value of ENDF filemoderator_za_data

integernone

Reaction type (MT) number for inelastic reaction.
(MT=221-250 only)
Maximum energy for thermal treatment
ZA ID name for thermal ACE file
(ZA=1000.0×Z+A, Max. 6 words)
Moderator component ZA value
Number of atom types in mixed moderator

atom_type_no

integer-(If user doesn't set this, FRENDY automatically set fromENDF file name and so on)

stringweight_option

or
integerstring = tabulated,
integer = 2

Weighting option for thermal ACE file
Available value is variable, constant, tabulated, default or 0-2(0=variable, 1=constant, 2=tabulated)

Table 1.2.8 Input parameter name and recommended value which are used only for multi-group cross-section generation
Parameter nameType

Recommended value

Description
S(α,β) type for MATXS file. This parameter must be required when user processes TSL

mg_tsl_data_type

list (string)-data. S(α,β) type used in FRENDY is shown in Table 1.2.2. S(α,β) type has no impact on
GENDF file. Please set “free” when the user only generates GENDF file.
Multi-group XS file namemg_file_namestringinput file name

The output file name is as follows:"mg_file_name"_"mg_file_mode"_"ZAID of the ACE file".mg
Format of the start file.
FRENDY can generate the multi-group cross-section file from the ENDF-6 formatted file- 44 -

mg_start_file_mode

stringENDF

and the ACE file.
The available value is ENDF and ACE.
ENDF: Multi-group cross-section file is generated from the ENDF-6 formatted file.
ACE: Multi-group cross-section file is generated from the ACE file.

mg_label_data

string-legendre_orderinteger3

real

max_thermal_ene

max_thermal_ene_e_outLabel of multi-group cross-section file.
The label is output on the first line of the GENDF file.

Maximum Legendre order (PL order)

Maximum energy of thermal treatment for outgoing energy [eV]

Output format and output data of the multi-group cross-section generation.
The available output format and output data are MATXS, SimpleMATXS, GENDF, FullMATXS,
SimpleGENDF, FullGENDF, KRAMXS, 1DXS, 2DXS, NuChi, MGFlux, MGCurrent, UFG, UFG1DXS,
and URC.

MATXS: Microscopic cross sections in MATXS format, consistent with NJOY2016.
SimpleMATXS: Microscopic cross sections in MATXS format, consistent with NJOY99.
FullMATXS: Microscopic cross sections in MATXS format without truncation of shielded cross sections.
Output file size may be large.
mg_edit_modeor- 45 -

mg_edit_option
ormg_edit_xs

GENDF: Microscopic cross sections in GENDF format consistent with NJOY2016.
list (string)

GENDF

SimpleGENDF: microscopic cross sections in GENDF format, consistent with NJOY99.
FullGENDF: microscopic cross sections in GENDF format without truncation of shielded cross sections.
Output file size may be large.

KRAMXS: microscopic cross sections in KRAM format.

1DXS: One-dimensional cross sections such as total, fission, radiative capture.
2DXS: Two-dimensional cross sections such as elastic scattering, inelastic scattering, (n,2n) reaction.
NuChi: Nu-value and fission spectrum.
MGFlux: Multi-group flux (group integrated values).
MGCurrent: Multi-group current (group integrated values).

UFG: Ultra-fine group spectrum, total cross sections, slowing down source, total source.
UFG1DXS: One-dimensional ultra-fine group cross sections such as total, fission, radiative capture.

RUC: The 0 K scattering cross section data. This data is used for the input file of the resonanceupscattering resonance up-scattering correction (reso_upscat).

User can select specified MT number when 1DXS, 2DXS, and UFG1DXS options are selected as follows:"1DXS 1, 2, 4, -50"

The minus value for MT means all MT number between previous MT number.
- 46 -

Weighting spectrum.
The available output format and output data are 1/E and Fission+1/E+Maxwell.
1/E: 1/E spectrum for the whole energy range.
Fission+1/E+Maxwell: Fission for fast energy range, 1/E for intermediate energy range,
Maxwell for the thermal energy range.

The other parameters are required to use "Fission+1/E+Maxwell" option as follows:
mg_weighting_spectrum( Fission+1/E+Maxwell Eh El Tfis
mg_weighting_spect
rum_modeE1 E2)

Eh: Highest energy [eV].
string

1/E

El: Lowest energy [eV].
Tfis: Fission temperature [eV].

- 47 -

E1: Energy boundary between fission and 1/E spectra [eV].
E2: Energy boundary between 1/E and Maxwell spectra [eV].
The sample of "Fission+1/E+Maxwell" is as follows:
mg_weighting_spectrum( Fission+1/E+Maxwell 2.0e+7 1.0e-5 1.6e+6 1.0e+6
0.625)The default value of Eh, El, Tfis, E1, and E2 are 2.0e+7, 1.0e-5, 1.6e+6, 1.0e+6, and 0.0.
When E2 is 0.0, E2 is automatically set by this program.

Weight spectrum.
The user can manually set the weight spectrum.
The format of weight spectrum is similar to the TAB1 record in the ENDF-6 format as
mg_weighting_spect

follows:
mg_weighting_spectrum_data(E1 W1 E2 W2 … Ei-1 Wi-1

rum_dataorlist (real)

Ei Wi)-

mg_weighting_spect

NJOY's weight spectrum, i.e., iwt of the GROUPR module and the weighted spectrum

rum

name shown in Table 1.2.3, is also available.The input format to select the NJOY's weightspectrum is as follows:
mg_weighting_spectrum_data( iwtXX )where XX is the iwt number (iwt02 ~ iwt12).

- 48 -

mg_weighting_spect
rum_data_int
or
mg_weighting_spect
rum_int

Interpolation method of the weight spectrum.
list

2(integer)(linear-linear)

The format of weight spectrum is similar to the TAB1 record in the ENDF-6 format asfollows:
mg_weighting_spectrum_data_int(NBT1 INT1 NBT2 INT2 … NBTi-1 INTi-1
INTi)

NBTi

Energy group structure.
The user can manually set the energy group structure [eV].

NJOY's energy group structure, i.e., ign of the GROUPR module and energy groupstructure name shown in Table 1.2.4, is also available. The input format to select the
NJOY's energy group structure is as follows:
mg_structure( ignXX )where XX is the ign number (ign02 ~ ign34, ign101).

The energy group structure can be automatically set by the user input.

mg_structure- 49 -

ormg_structure_neutronlist (real)

XMAS 172-groupstructure

The input format for the automatic setting of the energy group structure is as follows:
mg_structure ( auto E1 N1 O1 E2 N2 O2 ... Ei-1 Ni-1 Oi-1
Ei: Energy boundary i [eV]Ni: Number of divisions
Oi: Option for divisions
The available option for divisions is EL and EE.
EL: Divide the energy range by equi-lethargy width.
EE: Divide the energy range by equi-energy width.
The sample of input is as follows:
mg_structure ( auto
2.0e+07

3200 EL100.0

11.0e-5 )EL

Ei Ni Oi Ei+1)

Photon (gamma) production energy group structure.
The user can manually set the energy group structure [eV].

mg_structure_g

If the user does not set this parameter, the photon production data generation is skipped.

ormg_structure_gam
orlist (real)-

NJOY's energy group structure, i.e., igg of the GROUPR module and the energy groupstructure name shown in Table 1.2.5 is also available.

mg_structure_gamm

The input format to select the NJOY's energy group structure is as follows:amg_structure_gam( iggXX )where XX is the igg number (igg02 ~ igg10).

- 50 -( 2.0e+07 10000 EL
52475.0 56000 EL
mg_ufg_structure

9118.8 12000 EL

or

4307.4 12000 EL

mg_ultra_fine_grou

list (real)

p_structure

961.12 8000 EL
130.07 12000 EL
0.32242 10000 EL
1.0e-5 )mg_number_density

list (real)

1.0(use only one nuclide)

Energy group structure to generate neutron flux in the ultra-fine group structure.
The format of this parameter is as follows:
mg_ufg_structure( E1 N1 O1 E2 N2 O2... Ei-1 Ni-1 Oi-1

Ei Ni Oi Ei+1)

Ei: Energy boundary i [eV]Ni: Number of divisions
Oi: Option for divisions
The available option for divisions is EL and EE.
EL: Divide the energy range by equi-lethargy width.
EE: Divide the energy range by equi-energy width.
Number density of each nuclide [1/barn/cm].
If the user processes the one nuclide data, this parameter can be skipped.

Option for calculation of the thermal scattering using the free gas model.

mg_thermal_upscatt
er_treatmentor

stringon

mg_thermal_xs_trea

off/noreso_upscat
reso_upscat_data

on/yes/def/default<Skip calculation of the thermal scattering cross-section>

tment

reso_up_scat<Calculation of the thermal scattering cross-section>

The file name list of the 0 K scattering cross section data.
list (string)-

reso_up_scat_data

If the user wants to consider the resonance upscattering correction, please use this option.
The user has to set the file name of the 0 K scattering cross section data which is generated
by the RUC option in mg_edit_option.
Calculation option for the resonance up-scattering correction.

- 51 -

reso_upscat_mode
orALL: Resonance up-scattering correction is considered not only for 1D effective crossstringALL

reso_up_scat_mode

sections but also for the elastic scattering matrix.
XS: Resonance up-scattering correction is considered only for 1D effective cross sections.
No resonance up-scattering is considered for the elastic scattering matrix.
Potential scattering cross-section of each nuclide [barns].

potential_scat_xs

list (real)-

If the user does not set this parameter, FRENDY uses the potential scatteringcross-sections obtained by JENDL-4.0, ENDF/B-VIII.0, or JEFF-3.3.
Calculation option at the resonance region.

mg_flux_calc_mode

stringSLD

NR: Use narrow resonance calculation.
SLD: Use ultra-fine group slowing down calculation.

Calculation condition at the resonance region.

The format of this parameter is as follows:
mg_flux_calc_w_eh_ellist (real)

0.999167, 1.0×104, 1.0

mg_flux_calc_w_eh_el( W Eh El )W: Atomic weight (relative to neutron mass) used for the background moderator nuclide.
Eh: Upper energy to use ultra-fine group slowing down spectrum [eV].
El: Lower energy to use ultra-fine group slowing down spectrum [eV].
mg_mat_noor- 52 -

mg_mat_list
orlist(integer)

The MAT number of the GENDF file.
-

This parameter is only required when the ACE file is used as the start format of themulti-group cross-section generation since the ACE file does not have the MAT number.

mg_mat_no_list
Output ACE file option until multi-group cross-section generation.
ace_output_option
or
ace_edit_option

stringon<Writing ACE case>on/output/edit/def/default<Skip ACE case>
off/no/skipTable 1.2.9 Input parameter name and recommended value which are used only for XS output toolParameter nameType

Recommended value

Description

stringnone

GENDF formatted file name to output the (x, y) table.

stringnone

Output file name for XS output function.

gendf_nameorgendf_file_nameoutput_nameor
output_file_namemat_noorSpecified MAT number to output cross section.
integernone- 53 -

matfiles.

mf_listorThis option is available for ENDF and GENDF formatted

vector<Integer>3

mf_data

Selection of output MF data.
This option is only available for a GENDF formatted file.

mt_listor

Selection of output MT data.

mt_data_list
or
xs_type_list
orreaction_type_list

vector<Integer>none (all)If the user does not set this option, this program outputs all
cross section data.

Energy grid data option for multi-group cross section data.

If the user selects the non-histogram case, this tool only
outputs cross section at the lowest energy of each energy
groups, such as (Eli, XSi), (Eli+1, XSi+1).
If the user selects the histogram case, this tool outputs crosssections at the lowest and the highest energy of each energy
edit_flagorgroup, such as (Eli, XSi), (Ehi, XSi), (Eli+1, XSi+1), (Ehi+1,
stringhistogram

XSi+1).

edit_mode
The histogram option is useful for plotting the multi-group- 54 -

cross section distribution by an Excel scatter plot.

<histgram case>histogram/hist/two_point/two<Non histgram case>
non_histogram/non_hist/one_point/one

Table 1.2.10 Input parameter name and recommended value for resonance reconstruction
Parameter nameType

Recommended

Description

value

Output input information for resonance reconstruction
print_set_data_resonance_reconstruction
or<Writing input information>
stringon

print_set_data_linearize

on/yes/print/write/output/default/def<Skip information>
off/no/skip
Processing option for resonance reconstruction

process_resonance_reconstruction
or<Running resonance reconstruction case>stringon- 55 -

process_linearize

on/yes/process/run/default/def<Skip resonance reconstruction case>
off/no/skip
Output PENDF option for resonance reconstruction<Writing PENDF case>write_pendf_resonance_reconstructionoron/yes/print/write/output
stringoffwrite_pendf_linearize<Writing PENDF case with specific name>on/yes/print/write/output + PENDF file name
<Skip PENDF case>
off/no/skip/default/def

pendf_file_name_resonance_reconstructionorpendf_file_name_linearize

PENDF file name for resonance reconstruction
stringnone

When this parameter is set, write_pendf_linearize parameter is
automatically changed from off to on.

Table 1.2.11 Input parameter name and recommended value for Doppler broadening
Parameter nameType

Recommended value

Description
Output input information for Doppler broadening

print_set_data_doppler_broadeningor<Writing input case>stringon

print_set_data_dop

on/yes/print/write/output/default/def<Skip case>
off/no/skip
Processing option for Doppler broadening

process_doppler_broadening
or<Running Doppler broadening case>stringon- 56 -

process_dop

on/yes/process/run/default/def<Skip Doppler broadening case>
off/no/skip
Output PENDF option for Doppler broadening<Writing PENDF case>write_pendf_doppler_broadeningoron/yes/print/write/output
stringoffwrite_pendf_dop<Writing PENDF case with specific name>on/yes/print/write/output + PENDF file name
<Skip PENDF case>
off/no/skip/default/def

pendf_file_name_doppler_broadeningorpendf_file_name_dop

PENDF file name for Doppler broadening
stringnone

When this parameter is set, write_pendf_dop parameter is
automatically changed from off to on.

Table 1.2.12 Input parameter name and recommended value for gas production cross-section calculation
Parameter nameType

Recommended

Description

value

print_set_data_gas_production_cross_section,
print_set_data_gas_production_xs,

Output input information for gas production cross section generation

print_set_data_gas_production,<Writing input case>print_set_data_gas_prod_xs,

stringon

print_set_data_gas_xs,

on/yes/print/write/output/default/def<Skip case>oroff/no/skip

print_set_data_gas
process_gas_production_cross_section,- 57 -

process_gas_production_xs,

Processing option for gas production cross section generation

process_gas_production,<Gas production cross section generation case>process_gas_prod_xs,
process_gas_xs,
orprocess_gas

stringon

on/yes/process/run/default/def<Skip gas production cross section generation case>
off/no/skipwrite_pendf_gas_production_cross_section,

Output PENDF option for gas production cross section generation

write_pendf_gas_production_xs,<Writing PENDF case>write_pendf_gas_production,
write_pendf_gas_prod_xs,

on/yes/print/write/output
stringoffwrite_pendf_gas_xs,<Writing PENDF case with specific name>on/yes/print/write/output + PENDF file nameor

<Skip PENDF case>write_pendf_gas

off/no/skip/default/def

pendf_file_name_gas_production_cross_section,
pendf_file_name_gas_production_xs,
pendf_file_name_gas_production,
pendf_file_name_gas_prod_xs,- 58 -

pendf_file_name_gas_xs,
orpendf_file_name_gas

PENDF file name for gas production cross section generation
stringnone

When this parameter is set, write_pendf_gas is automaticallychanged from off to on.

Table 1.2.13 Input parameter name and recommended value for probability table generation
Parameter nameType

Recommended value

print_set_data_probability_table_generator,

Output input information for probability table generation

print_set_data_probability_table,
print_set_data_prob_table_generator,
print_set_data_prob_table,<Writing input case>stringonoff/no/skip

print_set_data_unreso_utils
process_probability_table_generator,

Processing option for probability table generation

process_probability_table,- 59 -

process_prob_table,<Probability table generation case>stringonoff/no/skip

process_unreso_utils

Output PENDF option for probability table generation

write_pendf_probability_table_generator,<Writing PENDF case>write_pendf_probability_table,

write_pendf_prob_table,orwrite_pendf_unreso_utils

on/yes/process/run/default/def<Skip probability table generation case>orwrite_pendf_prob_table_generator,

on/yes/print/write/output/default/def<Skip case>orprocess_prob_table_generator,

Description

on/yes/print/write/output
stringoff<Writing PENDF case with specific name>on/yes/print/write/output + PENDF file name
<Skip PENDF case>
off/no/skip/default/def

pendf_file_name_probability_table_generator,
pendf_file_name_probability_table,
pendf_file_name_prob_table_generator,
pendf_file_name_prob_table,orpendf_file_name_unreso_utils

PENDF file name for probability table generation
stringnone

When this parameter is set, write_pendf_prob_table parameteris automatically changed from off to on.

- 60 -

Table 1.2.14 Input parameter name and recommended value for thermal scattering law data
Parameter nameType

Recommended

Description

value

Output input information for processing thermal scattering law data
print_set_data_thermal_scattering_law,
print_set_data_thermal_scattering,<Writing input case>stringon

print_set_data_tsl

on/yes/print/write/output/default/def<Skip case>
off/no/skip
Processing option for thermal scattering law dataprocess_thermal_scattering_law,- 61 -

process_thermal_scattering,<Processing thermal scattering law data case>stringonprocess_tsl

on/yes/process/run/default/def<Skip thermal scattering law data case>
off/no/skip
Output PENDF option for thermal scattering law data
<Writing PENDF case>write_pendf_thermal_scattering_law,
write_pendf_thermal_scattering,on/yes/print/write/output
stringoffwrite_pendf_tsl<Writing PENDF case with specific name>on/yes/print/write/output + PENDF file name
<Skip PENDF case>
off/no/skip/default/def

pendf_file_name_thermal_scattering_law,
pendf_file_name_thermal_scattering,
pendf_file_name_tsl

PENDF file name for thermal scattering law data
stringonWhen this parameter is set, write_pendf_tsl parameter is
automatically changed from off to on.

Table 1.2.15 Input parameter name and recommended value for ACE file generation
Parameter nameType

Recommended value

Description
Output input information for ACE file generation

print_set_data_ace_data_generatororstringon

print_set_data_ace<Writing input case>on/yes/print/write/output/default/def<Skip case>
off/no/skip
Processing option for ACE file generation

process_ace_data_generator- 62 -

orprocess_ace

stringon<ACE file generation case>on/yes/process/run/default/def<Skip ACE file generation case>
off/no/skipTable 1.2.16 Optional input parameter name and recommended value
Parameter nameType

Recommended value

Description
Policy for dealing with NaN (Not a Number) when data is
output to ENDF, PENDF, and GENDF file.

nan_err_mes_opt,
err_mes_nan_opt,
orerr_nan_opt,
or
nan_opt

stringruntime_error<Process is aborted when the “NaN” value is found>runtime/runtime_error/error/default/def<Process is not aborted and the “NaN” value is converted-99999999.0 when the “NaN” value is found >caution/warning- 63 -

1.2.3 Examples of FRENDY Input Formatted File
The examples of the input files for the FRENDY original input format are shown in Figs. 1.2.3 1.2.7. As shown in Fig. 1.2.3, the “parameter name” and “parameter value” do not need to be placedin the same line. Comment lines can be put among the “parameter name” and the “parameter value”.
As shown in Fig. 1.2.5, the minimum input parameters are the processing mode and the evaluatednuclear data file name. FRENDY prepares the recommended values for the other parameters as listed
in Tables 1.2.5 - 1.2.15.
//Input file of FRENDY for continuous-energy neutron data
ace_fast_mode//processing mode (generate fast ACE file)//Nuclear data file namenucl_file_nameU235.dat/* ace file name */ace_file_nameU235.ace

temp

296.0 // temperature [K]

add_grid_data(0.625 4.0) //Additional energy grid point

sigima_zero_data //sigma-infinity is 1.0E10 [b](1.0E10 1.0E06 1.0E05 1.0E04 1.0E03 1.0E02
3.5E02 1.0E02 1.0E01 1.0E-1)
Figure 1.2.3 Example of input file for continuous-energy neutron data//Input file of FRENDY for thermal scattering law data
ace_therm_mode//process mode (generate thermal ACE file)//Nuclear data file namenucl_file_name
H001.dat//Nuclear data file name for TSL
nucl_file_name_tsl
H_in_H2O.txt/* ace file name */ace_file_name
H_in_H2O.ace
temp /* temperature */ 296.0 //[K]Figure 1.2.4 Example of input file for continuous-energy thermal scattering law data
- 64 -

//Input file of FRENDY for continuous-energy neutron data
ace_dosi_mode //processing mode (generate dosimetry ACE file)//Nuclear data file namenucl_file_nameU235.dat

Figure 1.2.5 Example of input file for continuous-energy dosimetry data//Input file of FRENDY for multi-group neutron-induced data
mg_neutron_mode //processing mode (generate multi-group XS file)//Nuclear data file namenucl_file_nameU235.dat/* multi-group XS file name */mg_file_name

U235

temp

296.0 // temperature [K]/* Output data type list */mg_edit_option ( GENDF MATXS )/* energy group structure */mg_structure ( ign18 ) // XMAS NEA LANL 172 groups/* weight spectrum */mg_weighting_spectrum ( iwt03 ) // 1/e
Figure 1.2.6 Example of input file for multi-group neutron-induced data//Input file of FRENDY for thermal scattering law data
mg_neutron_mode//process mode (generate multi-group XS file)//Nuclear data file namenucl_file_name

H001.dat

nucl_file_name_tsl H_in_H2O.txt/* multi-group XS file name */mg_file_name

H_in_H2O

temp /* temperature */ 296.0 //[K]/* energy group structure */mg_structure ( xmas_nea-lanl_172 ) // ign=18/* weight spectrum */mg_weighting_spectrum ( 1/e ) // iwt=3
Figure 1.2.7 Example of input file for multi-group thermal scattering law data

- 65 -

1.3

NJOY Compatible Format

The input format of NJOY is explained in the source code and manuals of the NJOY code system.
Users can consult them for the details of the format. FRENDY can process the evaluated nuclear
data file with the input file of the following modules:
MODER, RECONR, BROADR, GASPR, PURR, UNRESR, THERMR, ACER, GROUPR,
and MATXSR.
Though FRENDY processes the evaluated nuclear data file using the NJOY compatible format, users
should pay attention to the following points.
⚫

FRENDY reads and writes the PENDF file only in the text format.

⚫

FRENDY does not calculate the self-shielding factor in the unresolved resonance region
using the deterministic method and the input file of the UNRESR module is automaticallyconverted to that of PURR.

⚫

The “iform” option in the THERMR module and the “ismooth” option in the GROUPR
module in NJOY2016 are ignored to treat both NJOY99 and NJOY2016 formats.

⚫

The fast, thermal, and dosimetry ACE file generation functions are only implemented andother functions in ACER are not currently implemented.

⚫

FRENDY cannot generate the MATXS file from the GENDF file. Users must combine the
GROUPR and the MATXSR modules to generate MATXS file.

⚫

FRENDY generates multiple MATXS files when the GENDF file or the PENDF file contains
the multi-temperature data.

NJOY reads and writes PENDF files in binary format when the input tape number is negative. The
PENDF file in the binary format is used to efficiently access the evaluated nuclear data file. It isdifficult to perfectly treat the binary file produced by FORTRAN since FRENDY is written in C++.
Furthermore, the PENDF file in the binary format is mainly used as a temporary file. FRENDY
changes the negative tape number (binary file) to a positive one (text file) when users set a negative
tape number.
FRENDY calculates probability tables to treat the self-shielding effect in the unresolved resonanceregion with the ladder method. NJOY implements the deterministic method to calculate it with the
UNRESR module. According to the manual of NJOY, it is not recommended to use the deterministic
method. The self-shielding factor generated by the deterministic method sometimes shows
inappropriate values, e.g., negative or larger than 1.0. We believe that the calculation time of theladder method is acceptable on the current computational platforms. Therefore, we do not implement- 66 -

the deterministic method and FRENDY calculates the self-shielding effect in the unresolved
resonance region using the ladder method even if users select the UNRESR module. The input
parameters for the ladder method, e.g., the numbers of bins and ladders, are obtained from therecommended value of the FRENDY original input as shown in Table 1.2.2.
The input formats of the THERMR module and the ACER module of NJOY2016 are slightly
different from those of NJOY99. Though FRENDY can treat both input formats, the new input
parameter prepared in NJOY2016, i.e., the “iform” option in the THERMR module and the “ismooth”
option in the GROUPR module are ignored.
The ACE file generation for the photo-atomic and the photo-nuclear data is not currently
implemented. FRENDY terminates with an error message if users select these calculation modes.
FRENDY generates the MATXS file and the GENDF file from the ACE file. FRENDY cannot
generate the MATXS file from the GENDF file. FRENDY shows the warning message andterminates the processing when the input of the MATXSR module is only found in the input data.
NJOY outputs one MATXS file even if the multi-temperature data is found in the GENDF file.
The multi-temperature MATXS file contains the difference data from the base temperature data. The
multi-temperature data has some problems, e.g., the number of significant digits is different in eachdata block. We believe that the multi-temperature MATXS file should be separated in eachtemperature.

- 67 -

1.4

Sample Input Data for ACE File Generation

1.4.1 Simplest Input Data
1.4.1.1 Fast ACE file generation
The simplest input format to generate the neutron incident ACE file requires only the processing
mode name and evaluated nuclear data file name.
ace_file_generation_fast_mode
nucl_file_name// Processing mode name
U235.fast.dat// Nuclear data file name

The above input data generates the ACE file at a temperature of 293.6 K named “U235.fast.dat.ace”
and the directory information file named “U235.fast.dat.ace.dir”. For the other calculation conditions,the default input parameters listed in Tables 1.2.6 - 1.2.14 are used.

1.4.1.2 Thermal ACE file generation
The simplest input format to generate the ACE file of the thermal scattering law data requires only
the processing mode name, evaluated nuclear data file name, and thermal scattering law data file
name.
ace_file_generation_thermal_scatter_mode// Processing mode namenucl_file_name

H001.dat// Nuclear data file namenucl_file_name_tsl

HinH2O.dat// Thermal scattering law data// file name

The above input data generates the ACE file at a temperature of 293.6 K named “HinH2O.dat.ace”,
and the directory information file named “HinH2O.dat.ace.dir”. For the other calculation conditions,the default input parameters listed in Tables 1.2.6 - 1.2.14 are used.

1.4.1.3 Dosimetry ACE file generation
The simplest input format to generate the ACE file for the dosimetry data requires only theprocessing mode name and evaluated nuclear data file name.
ace_file_generation_dosimetry_mode// Processing mode namenucl_file_name// Nuclear data file name

U235.dosi.dat- 68 -

The above input data generates the ACE file at a temperature of 293.6 K named “U235.dosi.dat.ace”,
and the directory information file named “U235.dosi.dat.ace.dir”. For the other calculation
conditions, the default input parameters listed in Tables 1.2.6 - 1.2.14 are used.

1.4.2 Input Data to Modify Default Input Parameters
As shown in Tables 1.2.6 - 1.2.14 many input parameters are prepared in the FRENDY originalformat. However, in conventional applications users usually change only 5 parameters, i.e., thetemperature, the number of equiprobable angle bins for thermal scattering law data, ACE file name,directory information for the ACE file, and a suffix ID for the ACE file. This section illustrates
examples of the input files to change these input parameters.

1.4.2.1 Fast ACE file generation
ace_file_generation_fast_mode// Processing mode namenucl_file_nameU235.fast.dat// Nuclear data file name

temp

600.0

// Temperature [K]suffix_id.50// Suffix ID for ACE file

ace_file_nameU235.fast.ace// ACE file nameace_dir_file_nameU235.fast.ace.dir // MCNP directory information

The above input data generates the ACE file at a temperature of 600.0 K named “U235.fast.ace”, andthe directory information file named “U235.fast.ace.dir”. In this input file, the suffix ID of the ACE
file is changed from 0.00 to 0.50.

1.4.2.2

Thermal ACE file generation

ace_file_generation_thermal_scatter_mode// Processing mode namenucl_file_name

H001.dat// Nuclear data file namenucl_file_name_tsl

HinH2O.dat// TSL data name

temp

600.0

// Temperature [K]equi_probable_angle_no30// Number of equi-probable angles

suffix_id.50// Suffix ID for ACE file

ace_file_name

HinH2O.tsl.ace// ACE file nameace_dir_file_name

HinH2O.tsl.ace.dir// MCNP directory information

The above input data generates the ACE file at a temperature of 600.0 K named “HinH2O.tsl.ace”,
and the directory information file named “HinH2O.tsl.ace.dir”. In this input file, the number of- 69 -

equiprobable angles is changed from 10 to 30 and the suffix ID of the ACE file is changed from 0.00
to 0.50.

1.4.2.3

Dosimetry ACE file generation

ace_file_generation_dosimetry_mode// Processing mode namenucl_file_nameU235.dosi.dat// Nuclear data file name

temp

600.0

// Temperature [K]suffix_id.50// Suffix ID for ACE file

ace_file_nameU235.dosi.ace// ACE file nameace_dir_file_nameU235.dosi.ace.dir // MCNP directory information

The above input data generates the ACE file at a temperature of 600.0 K named “U235.dosi.ace”,
and the directory information file named “U235.dosi.ace.dir”. In this input file, the suffix ID of the
ACE file is changed from 0.00 to 0.50.

1.4.3 Input Data to Reproduce NJOY99 Input
This section illustrates how to make FRENDY input files which is identical to NJOY input files.
All the input parameters in the NJOY input files are explicitly set in the FRENDY input files. The
processing conditions, PENDF file name, ACE file name, and so on of the FRENDY input files are
identical to those of the NJOY input files.

1.4.3.1 Fast ACE file generation
The following input files are typical for the fast ACE file generation. These input files process theevaluated nuclear data file of 238U at 300.0 K.

< NJOY99 compatible format >reconr/ command

20 21/ input(tape20), output(tape21)'pendf tape for JENDL-4.0 U-238'/ identifier for PENDF

9237 3 3/ mat, ncards, ngrid

1.00e-03 0.00

/ err, temp

'92-U-238 from JENDL-4.0' / cards (1)'Processed with FRENDY' / cards (2)'Processed day: 2017/10/13' / cards (3)0.625 4.0 100.0/ enode0/- 70 -

broadr/ command

20 21 22/ input(tape20), pendf(tape21), output(tape22)9237 1 0 1 0/ mat, temp_no, restart_opt, bootstrap, temp_initial

1.00E-03 100000/ err, max_energy

300.0/ temp

0

/

gaspr

/ command

20 22

23

/ input(tape20), output(tape23)

0

/

purr

/ command

20 23 25

/ input(tape20), pendf(tape23), output(tape25)

9237 1 7 20 100 1/ mat, temp_no, sig0_no, bin_no, ladder_no, print_opt

300.0/ temp1.0E10 1.0E4 1.0E3 300.0 100.0 30.0 10.0 / sig0
0

/

acer

/ command

20 25

0 30 31/ nendf, npend, ngend, nace, ndir

1

1 0.30/ iopt(fast), iprint(max), itype, suffix

1'ACE file for JENDL-4 U238'/ descriptive character

9237 300.0/ mat, temp

1 1

/ newfor(yes), iopp(yes)1 1 1/ thin(1), thin(2), thin(3)

stop

/

< FRENDY original input format >ace_file_generation_fast_mode
nucl_file_namependf_label_data// Processing mode nametape20"pendf tape for JENDL-4.0 U238"

error

1.0E-3

temp

300.0

add_grid_data(0.625 4.0 100.0)

max_broadening_ene

100000 // 1 MeV

sigma_zero_data
probability_bin_no(1.0E10 1.0E4 1.0E3 300.0 100.0 30.0 10.0)
20- 71 -

ladder_no100

ace_file_nametape30mcnp_dir_file_nametape31

ace_label_data"ACE file for JENDL-4 U238"

suffix_id

0.30

comment_data“92-U-238 from JENDL-4.0
Processed with FRENDY
Processed day: 2017/10/13”// Write PENDF file option for RECONR
write_pendf_resonance_reconstruction tape21// Write PENDF file option for BROADR
write_pendf_doppler_broadening

tape22// Write PENDF file option for GASPR
write_pendf_gas_prod_xs

tape23// Write PENDF file option for PURR
write_pendf_prob_table_generator

tape25// Skip or running RECONR option
process_resonance_reconstructionon// Skip or running BROADR option
process_doppler_broadeningon// Skip or running GASPR option
process_gas_production_cross_sectionon// Skip or running PURR option
process_probability_table_generatoron// Skip or running ACER option
process_ace_data_generator- 72 -

on1.4.3.2 Thermal ACE file generation
The following input files are typical for the thermal ACE file generation. These input files process
the evaluated nuclear data file of 9Be in BeO at 300.0 K.

< NJOY99 compatible format >reconr/ command

20 21/ input(tape20), output(tape21)'pendf tape for JENDL-4.0 Be-009'/ identifier for PENDF

425

/ mat

1.00e-03 0.00

/ err, temp0/

broadr/ command

20 21 22/ input(tape20), pendf(tape21), output(tape22)

425 1 0 1 0/ mat, temp_no, restart_opt, bootstrap, temp_initial

1.00E-03 100000.0/ err, max_energy

400.0/ temp

0

/

gaspr

/ command

20 22 23

/ input(tape20), pendf(tape22), output(tape23)

0/

thermr/ command

60 23 25

/ input(tape60), pendf(tape23), output(tape25)

27 425 10 1 4 1 1 221/ natde, matdp, nbin, ntemp, iinc, icof

400.0

/ temp1.00E-2 4.0/ tolerance, max energy

purr/ command

20 25 26

/ input(tape20), pendf(tape25), output(tape26)

425 1 7 20 200 1/ mat, temp_no, sig0_no, bin_no, lad_no, print_opt

400.0/ temp1.0E10 1.0E4 1.0E3 300.0 100.0 30.0 10.0 / sig0
0

/

acer

/ command

20 26

0 30 31/ nendf, npend, ngend, nace, ndir

2

1 0.30/ iopt(fast), iprint(max), itype, suffix

1'ACE file for JENDL-4 BeinBeO'/ descriptive character

425 400.0 'bebeo'/ mat, temp
- 73 -

4009 8016

/ iza01, iza02

221 10 222 0 2 4.0 0/ mti, nbint, mte, ielas, nmix, emax, iwt

stop

/

< FRENDY original input format >ace_file_generation_thermal_scatter_mode
nucl_file_nametape20nucl_file_name_tsl

tape60pendf_label_data// Processing mode name"pendf tape for JENDL-4.0 Be-009"

error

1.0E-3

temp

400.0

max_broadening_ene

100000.0 // 1 MeV

sigma_zero_data(1.0E10 1.0E4 1.0E3 300.0 100.0 30.0 10.0)

probability_bin_no20

ladder_no200

ace_file_nametape30_frendy

mcnp_dir_file_nametape31_frendy

ace_label_data"ACE file for JENDL-4 BeinBeO"

thermal_za_id_name"bebeo"suffix_id

0.30//Thermal scattering law data only
equi_probable_angle_no10

principal_atom_no

1 //Beinelastic_reaction_type_no

221

max_thermal_ene

4.0

thermal_za_id_name“bebeo”

moderator_za_data

4009

atom_type_no

2 //Be and O

weight_option

0 //0 = variable// Write PENDF file option for RECONR
write_pendf_resonance_reconstruction// Write PENDF file option for BROADR- 74 -

tape21

write_pendf_doppler_broadening

tape22// Write PENDF file option for GASPR
write_pendf_gas_prod_xs

tape23// Write PENDF file option for THERMR
write_pendf_thermal_scattering

tape25// Write PENDF file option for PURR
write_pendf_prob_table_generator

tape26// Skip or running RECONR option
process_resonance_reconstructionon// Skip or running BROADR option
process_doppler_broadeningon// Skip or running GASPR option
process_gas_production_cross_sectionon// Skip or running THERMR option
process_thermal_scattering_law

on// Skip or running PURR option
process_probability_table_generatoron// Skip or running ACER option
process_ace_data_generator

on

1.4.3.3 Dosimetry ACE file generation
The following input files are typical for the dosimetry ACE file generation. These input files
process the evaluated nuclear data file of 238U at 300.0 K.

< NJOY99 compatible format >reconr/ command

20 21/ input(tape20), output(tape21)'pendf tape for JENDL-4 U238'/ identifier for PENDF

9237

/ mat

1.00e-03 0.0/ err, temp0/

broadr/ command

20 21 22/ input(tape20), pendf(tape21), output(tape22)9237 1 0 1 0/ mat, temp_no, restart_opt, bootstrap, temp_initial

1.00E-03 100000/ err, max_energy- 75 -

300.0/ temp

0

/

gaspr

/ command

20 22 23

/ input(tape20), pendf(tape22), output(tape23)

acer/ command

20 23

0 30 31/ nendf, npend, ngend, nace, ndir3

1 0.30/ iopt(fast), iprint(max), itype, suffix

1'ACE file for JENDL-4 U238'/ descriptive character

9237 300.0/ mat, temp

stop

< FRENDY original input format >ace_file_generation_dosimetry_mode // Processing mode namenucl_file_namependf_label_data

tape20"pendf tape for JENDL-4.0 U-238"

error

1.0E-3

temp

300.0

max_broadening_ene

100000 //1 MeV

ace_file_nametape30mcnp_dir_file_nametape31

ace_label_data"ACE file for JENDL-4 U238"

suffix_id

0.30// Write PENDF file option for RECONR
write_pendf_resonance_reconstruction

tape21// Write PENDF file option for BROADR
write_pendf_doppler_broadening

tape22// Write PENDF file option for GASPR
write_pendf_gas_prod_xs

tape23// Skip or running RECONR option
process_resonance_reconstructionon// Skip or running BROADR option
process_doppler_broadening- 76 -

on// Skip or running GASPR option
process_gas_production_cross_sectionon// Skip or running ACER option
process_ace_data_generator- 77 -

on1.5 Sample Input Data for Multi-Group Cross-Section Generation
1.5.1 Sample Input Data to Generate Neutron-Induced Multi-Group Cross-Section
FileThis section shows the sample input file to generate the neutron-induced multi-group cross-section
file from the evaluated nuclear data and the ACE files. The input files in this section process 1H, and
235

U and 238U at 300.0 K. The sample input files to process 238U show the modification of the atomic

weight of the moderator in the “mg_flux_calc_w_eh_el” to generate multi-group cross section files
for graphite and hydrogen moderators.

< Input file to process 1H from evaluated nuclear data file >mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF ) //Output format

nucl_file_name( ../lib/H001.dat )mg_file_name

H001

temperature

300.0 //[K]//ENDF file name//Output file namemg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( 1/e )// Identical to iwt=3 in GROUPR/NJOY< Input file to process 235U from evaluated nuclear data file >mg_neutron_mode //Process mode
nucl_file_name ( ../lib/U235.dat ) //ENDF file namemg_file_name U235//Output file namemg_edit_mode ( SimpleGENDF SimpleMATXS GENDF MATXS"MGFlux" )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016
mg_structure ( ign18 )// Identical to ign=18 in GROUPR/NJOY

mg_structure_gam ( vitamin-j-42 ) // Identical to igg=10 in GROUPR/NJOY
mg_weighting_spectrum ( iwt04 ) // Identical to iwt=4 in GROUPR/NJOY
temp

300.0 //[K]legendre_order3- 78 -

< Input file to process 235U from evaluated nuclear data file using specified energy
group structure and weighting spectrum>
mg_neutron_mode//Process mode

nucl_file_name ( ../lib/U235.dat ) //ENDF file namemg_file_name U235//Output file namemg_edit_option ( SimpleGENDF SimpleMATXS GENDF MATXS
MGFlux )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016
temp

300.0 //[K]mg_structure ( 1.0E-5 0.625 5.0E5 1.0E6 2.0E+7 ) //4-groups
mg_weighting_spectrum ( 1.0E-5 1.0 2.0E+7 1.0 )legendre_order//constant3< Input file to process 235U from evaluated nuclear data file for fast reactor>
mg_neutron_mode//Process mode

nucl_file_name ( ../lib/U235.dat ) //ENDF file namemg_file_name U235//Output file namemg_edit_option ( SimpleGENDF SimpleMATXS GENDF MATXS
MGFlux )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016
temp

300.0 //[K]mg_structure ( sand-iia-640 )// Identical to ign=15 in GROUPR/NJOY

mg_weighting_spectrum ( iwt08 ) // Identical to iwt=8 in GROUPR/NJOY
legendre_order3

mg_thermal_upscatter_treatment off //Thermal scattering XS treatment//Thermal scattering cross-section is not required in a fast reactor- 79 -

< Input file to process 238U from evaluated nuclear data file with hydrogenmoderator>
mg_neutron_mode//Process mode

nucl_file_name ( ../lib/U238.dat ) //ENDF file namemg_file_name U238//Output file namemg_edit_option ( SimpleGENDF SimpleMATXS GENDF MATXS
MGFlux )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016
temp

300.0 //[K]mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
legendre_order3

sigma_zero_data ( 1.0e+10 3.0e+4 1.0e+4 3.0e+3 1.0e+3 3.0e+2 1.0e+2 3.0e+1 1.0e+1 3.0 )// Typical background cross section for U238
mg_flux_calc_w_eh_el ( 0.999167, 1.0e+4, 1.0 )// Atomic weight of moderator (hydrogen, default value)// Upper energy [eV] of ultra-fine group slowing down calculation// Lower energy [eV] of ultra-fine group slowing down calculation< Input file to process 238U from evaluated nuclear data file with graphite
moderator>
mg_neutron_mode//Process mode

nucl_file_name ( ../lib/U238.dat ) //ENDF file namemg_file_name U238//Output file namemg_edit_option ( SimpleGENDF SimpleMATXS GENDF MATXS
MGFlux )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99- 80 -

//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016
temp

300.0 //[K]mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
legendre_order3

sigma_zero_data ( 1.0e+10 3.0e+4 1.0e+4 3.0e+3 1.0e+3 3.0e+2 1.0e+2 3.0e+1 1.0e+1 3.0 )// Typical background cross section for U238
mg_flux_calc_w_eh_el ( 12.0, 1.0e+4, 1.0 )// Atomic weight of moderator, (graphite)// Upper energy [eV] of ultra-fine group slowing down calculation// Lower energy [eV] of ultra-fine group slowing down calculation< Input file to process 1H from ACE file >mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF ) //Output format

ace_file_name( ../ace/H001.ace )//ACE file namemg_file_name

H001//Output file nametemperature

300.0 //[K]mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY< Input file to process 235U from ACE file >mg_neutron_mode //Process mode
ace_file_name ( ../ace/U235.ace ) //ACE file namemg_file_name U235//Output file namemg_edit_mode ( SimpleGENDF SimpleMATXS GENDF MATXS"MGFlux" )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016- 81 -

//MATXS: MATXS format consistent with NJOY2016
mg_structure ( ign18 )// Identical to ign=18 in GROUPR/NJOY

mg_structure_gam ( igg10 )// Identical to igg=10 in GROUPR/NJOY

mg_weighting_spectrum ( iwt04 ) // Identical to iwt=4 in GROUPR/NJOY
temp

300.0 //[K]legendre_order3< Input file to process 235U from ACE file using specified energy group structure
and weighting spectrum >mg_neutron_mode//Process mode

ace_file_name ( ../ace/U235.ace ) //ACE file namemg_file_name U235//Output file namemg_edit_option ( SimpleGENDF SimpleMATXS GENDF MATXS )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016
mg_structure ( 1.0E-5 0.625 5.0E5 1.0E6 2.0E+7 ) //4-groups
mg_weighting_spectrum ( 1.0E-5 1.0 2.0E+7 1.0 )legendre_order//constant3< Input file to process 235U from ACE file for fast reactor>
mg_neutron_mode//Process mode

ace_file_name ( ../ace/U235.ace ) //ACE file namemg_file_name U235//Output file namemg_edit_option ( SimpleGENDF SimpleMATXS GENDF MATXS
MGFlux )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016
mg_structure ( ign15 )// Identical to ign=15 in GROUPR/NJOY- 82 -

mg_weighting_spectrum ( iwt08 ) // Identical to iwt=8 in GROUPR/NJOY
legendre_order3

mg_thermal_upscatter_treatment off //Thermal scattering XS treatment//Thermal scattering cross-section is not required in a fast reactor< Input file to process 238U from ACE file with hydrogen moderator>
mg_neutron_mode//Process mode

nucl_file_name ( ../ace/U238.ace ) //ACE file namemg_file_name U238//Output file namemg_edit_option ( SimpleGENDF SimpleMATXS GENDF MATXS
MGFlux )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016
temp

300.0 //[K]mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
legendre_order3

sigma_zero_data ( 1.0e+10 3.0e+4 1.0e+4 3.0e+3 1.0e+3 3.0e+2 1.0e+2 3.0e+1 1.0e+1 3.0 )// Typical background cross section for U238
mg_flux_calc_w_eh_el ( 0.999167, 1.0e+4, 1.0 )// Atomic weight of moderator (hydrogen, default value)// Upper energy [eV] of ultra-fine group slowing down calculation// Lower energy [eV] of ultra-fine group slowing down calculation< Input file to process 238U from ACE file with graphite moderator>
mg_neutron_mode//Process mode

nucl_file_name ( ../ace/U238.ace ) //ENDF file namemg_file_name U238//Output file namemg_edit_option ( SimpleGENDF SimpleMATXS GENDF MATXS- 83 -

MGFlux )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016
temp

300.0 //[K]mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
legendre_order3

sigma_zero_data ( 1.0e+10 3.0e+4 1.0e+4 3.0e+3 1.0e+3 3.0e+2 1.0e+2 3.0e+1 1.0e+1 3.0 )// Typical background cross section for U238
mg_flux_calc_w_eh_el ( 12.0, 1.0e+4, 1.0 )// Atomic weight of moderator, (graphite)// Upper energy [eV] of ultra-fine group slowing down calculation// Lower energy [eV] of ultra-fine group slowing down calculation- 84 -

1.5.2 Sample Input Data to Generate Multi-Group Cross-Section File of TSL Data
This section shows the sample input file to generate the multi-group cross-section file of thethermal scattering law data from the evaluated nuclear data and the ACE files. The input files in thissection process H in H2O at 293.6 K and graphite at 293.6 K.

< Input file to process H in H2O from evaluated nuclear data file >mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF )//Output format

nucl_file_name( ../lib/H001.dat )//ENDF file namenucl_file_name_tsl ( ../lib/01_h_in_h2o.txt ) //TSL file namemg_file_name

H_in_H2O

temperature

293.6 //[K]mg_structure ( xmas_nea-lanl_172 )//Output file name// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
mg_tsl_data_type(hh2o)//This data type is used for MATXS file generation< Input file to process C in Graphite from evaluated nuclear data file >mg_neutron_mode
nucl_file_name//Process mode
( ../lib/C012.dat )//ENDF file namenucl_file_name_tsl ( ../lib/31_graphite.txt ) //TSL file namemg_edit_option( MATXS GENDF )//Output format

mg_file_name

C_in_Graphite//Output file nametemperature

293.6 //[K]mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( 1/e )// Identical to iwt=3 in GROUPR/NJOY

mg_tsl_data_type//This data type is used for MATXS file generation(graph)- 85 -

< Input file to process H in H2O from ACE file >mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF )ace_file_name( ../ace/H001.ace )//Output format//ENDF file nameace_file_name_tsl ( ../ace/01_h_in_h2o.ace ) //TSL file namemg_file_name

H_in_H2O

temperature

293.6 //[K]mg_structure ( xmas_nea-lanl_172 )//Output file name// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
mg_tsl_data_type(hh2o)//This data type is used for MATXS file generation< Input file to process C in Graphite from ACE file >mg_neutron_mode
ace_file_name//Process mode
( ../ace/C012.ace )//ACE file nameace_file_name_tsl ( ../ace/31_graphite.ace ) //ACE file name (TSL data)
mg_edit_option( MATXS GENDF )//Output format

mg_file_name

C_in_Graphite//Output file nametemperature

293.6 //[K]mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( 1/e )// Identical to iwt=3 in GROUPR/NJOY

mg_tsl_data_type//This data type is used for MATXS file generation(graph)- 86 -

1.5.3 Sample Input Data to Generate Multi-Nuclide Data
FRENDY can process multi-nuclide data, e.g., 235U, 238U, and 16O. This section shows the sample
input file to process such multi-nuclide data. The input files in this section process 235U, 238U, and16

O at 600.0 K and H in H2O, D in D2O, and O in H2O at 293.6 K.

< Input file to process 235U, 238U, and 16O from evaluated nuclear data file >mg_neutron_mode //Process mode
nucl_file_name (../lib/O016.dat../lib/U235.dat../lib/U238.dat ) //ENDF file namemg_file_name mix_UO2//Output file namemg_edit_mode (SimpleGENDF
SimpleMATXS
GENDF
MATXS
"1DXS 1, 2, 4, -50"
"2DXS 1, 2, 4, -50"
"MGFlux" )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016//"1DXS 1, 2, 4, -50": 1D cross-section data (MT=1, 2, 4 – 50)//"2DXS 1, 2, 4, -50": 2D cross-section data (MT=1, 2, 4 – 50)//MGFlux: Multi group flux data
mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_structure_gam ( vitamin-j-42 )// Identical to igg=10 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
mg_label_data "This data is UO2 data" //This label is output in the GENDF file
temp

600.0 //[K]- 87 -

legendre_order

3//8016.50c

mg_number_density(4.58e-2

92235.50c 92238.50c
7.18e-4

2.22e-2) //[1/barn/cm]< Input file to process H in H2O, D in D2O, and O in H2O from evaluated nuclear data
file >
mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF ) //Output format

nucl_file_name (../lib/H001.dat../lib/H002.dat../lib/O016.dat ) //ENDF file namenucl_file_name_tsl (../lib/01_h_in_h2o.txt../lib/11_d_in_d2o.txt../lib/02_o_in_h2o.txt )mg_file_name

mix_H2O

temperature

293.6 //[K]mg_structure ( xmas_nea-lanl_172 )//Output file name// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY//1001

1002

8016

mg_number_density(4.68e-2 7.01e-6 2.34e-2 ) //[1/barn/cm]

mg_tsl_data_type( hh2o

dd2o

oh2o

)

< Input file to process H in H2O, D in D2O, and O-16 from evaluated nuclear data file>mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF ) //Output format

nucl_file_name ( ../lib/H001.dat ../lib/H002.dat ../lib/O016.dat ) //ENDF file name- 88 -

nucl_file_name_tsl ( ../lib/01_h_in_h2o.txt ../lib/11_d_in_d2o.txt skip )mg_file_name

mix_H2O

temperature

293.6 //[K]//Output file namemg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY//1001

1002

8016

mg_number_density(4.68e-2 7.01e-6 2.34e-2 ) //[1/barn/cm]

mg_tsl_data_type( hh2o

dd2o

free

)

< Input file to process 235U, 238U, and 16O from ACE file >mg_neutron_mode //Process mode
ace_file_name (../ace/O016.ace../ace/U235.ace../ace/U238.ace ) //ACE file namemg_file_name mix_UO2//Output file namemg_edit_mode (SimpleGENDF
SimpleMATXS
GENDF
MATXS
"1DXS 1, 2, 4, -50"
"2DXS 1, 2, 4, -50"
"MGFlux" )//SimpleGENDF: MATXS format consistent with NJOY99//SimpleMATXS: GENDF format consistent with NJOY99//GENDF: GENDF format consistent with NJOY2016//MATXS: MATXS format consistent with NJOY2016//"1DXS 1, 2, 4, -50": 1D cross-section data (MT=1, 2, 4 – 50)//"2DXS 1, 2, 4, -50": 2D cross-section data (MT=1, 2, 4 – 50)//MGFlux: Multi group flux data
- 89 -

mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_structure_gam ( igg10 )// Identical to igg=10 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
mg_label_data "This data is UO2 data" //This label is output in the GENDF file
temp

600.0 //[K]legendre_order

3//8016.50c

mg_number_density(4.58e-2

92235.50c 92238.50c
7.18e-4

2.22e-2) //[1/barn/cm]< Input file to process H in H2O, D in D2O, and O in H2O from ACE file >mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF ) //Output format

ace_file_name (../ace/H001.ace../ace/H002.ace../ace/O016.ace ) //ACE file nameace_file_name_tsl (../ace/01_h_in_h2o.ace../ace/11_d_in_d2o.ace../ace/02_o_in_h2o.ace )mg_file_name

mix_H2O

temperature

293.6 //[K]mg_structure ( xmas_nea-lanl_172 )//Output file name// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY//1001

1002

8016

mg_number_density(4.68e-2 7.01e-6 2.34e-2 ) //[1/barn/cm]

mg_tsl_data_type( hh2odd2ooh2o- 90 -)

< Input file to process H in H2O, D in D2O, and O-16 from evaluated nuclear data file>mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF ) //Output format

ace_file_name ( ../ace/H001.ace ../ace/H002.ace../ace/O016.ace ) //ACE file nameace_file_name_tsl ( ../ace/01_h_in_h2o.ace ../ace/11_d_in_d2o.ace skip )mg_file_name

mix_H2O

temperature

293.6 //[K]mg_structure ( xmas_nea-lanl_172 )//Output file name// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY//1001

1002

8016

mg_number_density(4.68e-2 7.01e-6 2.34e-2 ) //[1/barn/cm]

mg_tsl_data_type( hh2odd2ofree- 91 -)

1.5.4 Sample Input Data for Automatic Setting of Background Cross-Sections
This section shows the sample input file to generate the neutron-induced multi-group cross-section
file from the evaluated nuclear data and the ACE files. This sample input file automatically set thebackground cross-sections of multi-group cross-section file. The input files in this section process235

U at 600.0 K.

< Input file to process 235U from evaluated nuclear data file >mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF ) //Output format

nucl_file_name ( ../lib/U235.dat ) //ENDF file namemg_file_nameU-235

temperature

600.0 //[K]mg_structure ( xmas_nea-lanl_172 )//Output file name// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
sigma_zero_data(auto 0.1 50 1.0e-10 rr linear)//Tolerance = 0.1 (10%)//Maximum number of background XS = 50//Minimum background XS = 1.0E-10 (barn)//Target of interpolation (factor / rr) = reaction rate)//Interpolation method (cubic / linear) = linear interpolation< Input file to process 235U from ACE file >mg_neutron_mode //Process mode
mg_edit_option( MATXS GENDF ) //Output format

ace_file_name ( ../ace/U235.ace ) //ENDF file namemg_file_nameU-235

temperature

600.0 //[K]mg_structure ( xmas_nea-lanl_172 )//Output file name// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY- 92 -

sigma_zero_data(auto 0.1 50 1.0e-10 rr linear)//Tolerance = 0.1 (10%)//Maximum number of background XS = 50//Minimum background XS = 1.0E-10 (barn)//Target of interpolation (factor / rr) = reaction rate)//Interpolation method (cubic / linear) = linear interpolation- 93 -

1.5.5 Sample Input Data for Automatic Setting of Energy Group StructureThis section shows the sample input file to generate the neutron-induced multi-group cross-section
file from the evaluated nuclear data and the ACE files. This sample input file automatically set theenergy group structure of multi-group cross-section file. The input files in this section process 235U
at 300.0 K.

< Input file to process 235U from evaluated nuclear data file >mg_neutron_mode//Process mode

nucl_file_name ( ../lib/U235.dat ) //ENDF file namemg_file_name U235//Output file namemg_edit_option ( GENDF MATXS )temp

300.0 //[K]mg_structure ( auto 2.0E+7 100 EL
1.0E+2 100 EL
1.0E-5 )// 2.0E+7 eV – 1.0E+2 eV: Divide 100 energy groups by equi-lethagy width// 1.0E+2 eV – 1.0E-5 eV: Divide 100 energy groups by equi-lethagy width//EL: Divide by equi-lethagy width, EE: Divide by equi-energy width
mg_weighting_spectrum ( iwt03 ) // Identical to iwt=3 in GROUPR/NJOY< Input file to process 235U from ACE file >mg_neutron_mode//Process mode

ace_file_name ( ../ace/U235.ace ) //ACE file namemg_file_name U235//Output file namemg_edit_option ( GENDF MATXS )temp

300.0 //[K]mg_structure ( auto 2.0E+7 100 EL
1.0E+2 100 EL
1.0E-5 )// 2.0E+7 eV – 1.0E+2 eV: Divide 100 energy groups by equi-lethagy width// 1.0E+2 eV – 1.0E-5 eV: Divide 100 energy groups by equi-lethagy width//EL: Divide by equi-lethagy width, EE: Divide by equi-energy width
mg_weighting_spectrum ( iwt03 ) // Identical to iwt=3 in GROUPR/NJOY- 94 -

1.5.6 Sample Input Data for Resonance Up-Scattering Correction
FRENDY can consider the resonance up-scattering correction. This section shows the sample
input file to use this function. The input files in this section process 16O, 235U, and 238U at 600.0 K.
The resonance up-scattering correction of 16O is not considered. The scattering cross section data at
0 K is required to consider the resonance up-scattering correction. This section shows two types of
input data i.e., generation of the scattering cross section data at 0 K and generation of themulti-group cross section data considering the resonance up-scattering correction.

< Input file to generate scattering cross section data at 0 K (processing 235U from
evaluated nuclear data file) >mg_neutron_mode //Process mode
mg_edit_option( GENDF RUC ) //Output format

nucl_file_name ( ../lib/U235.dat )//ENDF file namemg_file_nameU235_RUC//Output file nametemperature

0.0 //[K]mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY< Input file to generate scattering cross section data at 0 K (processing 238U from
evaluated nuclear data file) >mg_neutron_mode //Process mode
mg_edit_option( GENDF RUC ) //Output format

nucl_file_name ( ../lib/U238.dat )//ENDF file namemg_file_name

U238_RUC//Output file nametemperature

0.0 //[K]mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY- 95 -

< Input file to generate multi-group cross section considering resonance
up-scattering correction (processing 16O, 235U, and 238U from evaluated nuclear datafile) >
mg_neutron_mode //Process mode
mg_edit_option( GENDF MATXS ) //Output format

nucl_file_name ( ../lib/O016.dat ../lib/U235.dat ../lib/U238.dat ) //ENDF file namemg_file_name

mix_UO2_with_RUC

temperature

600.0 //[K]mg_structure ( xmas_nea-lanl_172 )//Output file name// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY//8016.50c
mg_number_density (4.58e-2

92235.50c 92238.50c
7.18e-4

2.22e-2) //[1/barn/cm]

reso_upscat ( skip U235_RUC_92235.00c_MT2.mg U238_RUC_92238.00c_MT2.mg )1.5.7

Sample Input Data to calculate IR parameter

The IR parameter, which is used to capture the impact of the different atomic mass of moderator
nuclides in a slowing down calculation, is calculated by Eq. (1.1).
𝜎1 − 𝜎0𝑓𝐼𝑅 = 1 + 𝜎 − 𝜎 .
20𝜎𝑏1 − 𝜎𝑏0
𝜎𝑏0

(1.1)

Here, we consider calculating the IR parameter of 16O for 238U. 𝜎0 is an effective cross-section of
238U using slowing down calculation using the typical background cross-section of 238U (𝜎𝑏0 ) anddefault mass for the background nuclide (𝑎0 =0.999167 for light water). 𝜎1 is an effective
cross-section of 238U using slowing down calculation using 𝜎𝑏0 and mass of 16O (𝑎1 =15.8575). 𝜎2
is an effective cross-section of 238U using slowing down calculation using 𝜎𝑏1 (e.g., 𝜎𝑏1 =1.05 × 𝜎𝑏0) and default mass for the background nuclide (𝑎0 =0.999167 for light water).
This subsection shows sample input files to calculate 𝜎0 , 𝜎1 , and 𝜎2 . The IR parameters of
MT=102 are calculated using these sample files. For example, when we use these input files andevaluated nuclear data files from JENDL-5, 𝜎0 , 𝜎1 , and 𝜎2 of 88th group (7.52-6.16 eV) are 𝜎0 =45.45043, 𝜎1 = 42.49696, 𝜎2 = 46.30637. Using Eq. (1.1), the IR parameter 88th group is 0.827.

- 96 -

< Input file to calculate 𝝈𝟎 for IR parameter >mg_neutron_mode //Process mode
nucl_file_name ( ../lib/n_092-U-238.dat ) //ENDF file namemg_file_name U238//Output file namemg_edit_mode ( SimpleGENDF SimpleMATXS"1DXS 102"
"MGFlux" )mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
temp

300.0 //[K]legendre_order3

sigma_zero_data ( 1.0e+10 50.0 )// 50.0 barn corresponds to a typical background cross section for U238 in an LWR latticemg_flux_calc_w_eh_el ( 0.999167, 1.0e+4, 1.0 )// Atomic weight of moderator (hydrogen)// Upper energy [eV] of ultra-fine group slowing down calculation// Lower energy [eV] of ultra-fine group slowing down calculation< Input file to calculate 𝝈𝟏 for IR parameter >mg_neutron_mode //Process mode
nucl_file_name ( ../lib/n_092-U-238.dat ) //ENDF file namemg_file_name U238//Output file namemg_edit_mode ( SimpleGENDF SimpleMATXS"1DXS 102"
"MGFlux" )mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
temp

300.0 //[K]legendre_order3- 97 -

sigma_zero_data ( 1.0e+10 50.0 )// 50.0 barn corresponds to a typical background cross section for U238 in an LWR latticemg_flux_calc_w_eh_el ( 15.8575, 1.0e+4, 1.0 )// Atomic weight of moderator (oxygen in UO2)// Upper energy [eV] of ultra-fine group slowing down calculation// Lower energy [eV] of ultra-fine group slowing down calculation< Input file to calculate 𝝈𝟐 for IR parameter >mg_neutron_mode //Process mode
nucl_file_name ( ../lib/n_092-U-238.dat ) //ENDF file namemg_file_name U238//Output file namemg_edit_mode ( SimpleGENDF SimpleMATXS"1DXS 102"
"MGFlux" )mg_structure ( xmas_nea-lanl_172 )// Identical to ign=18 in GROUPR/NJOY

mg_weighting_spectrum ( fission+1/e+maxwell ) // Identical to iwt=4 in GROUPR/NJOY
temp

300.0 //[K]legendre_order3

sigma_zero_data ( 1.0e+10 52.5 ) // sigma_0*1.05 = 50.0*1.05=52.5
mg_flux_calc_w_eh_el ( 0.999167, 1.0e+4, 1.0 )// Atomic weight of moderator (hydrogen)// Upper energy [eV] of ultra-fine group slowing down calculation// Lower energy [eV] of ultra-fine group slowing down calculation- 98 -

1.6 Input Instruction of ACE File Perturbation Tools
The ACE file perturbation tool contains two tools as follows:
tools/make_perturbation_factor/make_perturation_factor.exe
tools/perturation_ace_file/perturbation_ace_file.exe
The input format of both tools is explained in this section. The installation of these tools is found in
the readme file of these tools (tools/README_tools).

1.6.1 Input Instruction of make_perturbation_factor
This tool is the uncertainty analysis tool using the Random Sampling method for the ACE file.
This tool generates perturbation factor from the relative covariance matrix of cross-section andoutputs the input data of the “perturbation_ace_file.exe”. Note that the license of this tool is not the
2-clause BSD license. The make_perturation_factor.exe uses the Eigen library. The license of thistool is the MP2L license since the Eigen library is open-source software under the MP2L license.
The tool requires the following input data. In the input file, these terms are required and the user
has to enclose by "<>" to set the data of each term.
(1) Sample size (int): <sample_size> sample_no
The first parameter is the number of random samplings. The number of random samplings is
identical to the number of output files.
(2) Random seed (int): <seed>

seed_val

The second parameter is the initial random seed.
(3) File name of covariance matrix (string): <relative_covariance> cov_file_name
The third parameter is the name of the covaiance matrix file.
(4) Energy grid of covariance matrix [MeV] (real): <energy_grid > (E1 E2 E3 … Eg-1

Eg)

The fourth parameter is the energy group structure. The number of energy grids of the input file
must be identical to that of the covariance matrix. The energy group structure is enclosed by thebracket “()”.
(5) Nuclide name of the covariance matrix (string): <nuclide > (Nucl1 Nucl2… Nucli-1

Nucli)

The fifth parameter is the energy group structure. The nuclide name is used as the directory
name which stores the output files. If the covariance matrix uses the multiple nuclides, thenuclide name is enclosed by the bracket “()”.

- 99 -

(6) Reaction type (MT) of covariance matrix (int): <reaction > (MT1MT2

MT3 … MTi-1

MTi)

The sixth parameter is the reaction type number. If the covariance matrix uses the multiple
reaction type, the reaction type is enclosed by the bracket “()”.

1.6.2 Sample Input of make_perturbation_factor
The sample input file of this tool is as follows:< Generation of 100 sampling data from U-235 using cov_matrix.csv ><sample_size>

100<seed>

20190504<relative_covariance>

cov_matrix.csv<energy_grid>(1.0E-11<nuclide>(U235)<reaction>(2 18 102)

6.8E-72.0E1)In this case, the sample number is 100, the initial random seed is 20190504, the covariance file nameis “cov_matrix.csv”, energy grid number is 2, the energy grid boundary is 1.0E-11 MeV, 6.8E-7 MeV,
and 20MeV, the nuclide name is U235, and reaction type (MT number) is MT=2, 18, and 102,
respectively.

< Generation of 200 sampling data from U-235 and U-238 using cov_matrix.csv ><sample_size>

200<seed>

1<relative_covariance>

cov_matrix.csv<energy_grid>(1.0E-11<nuclide>(U235 U238)<reaction>(2 18 102) (4 16)

6.8E-72.0E1)In this case, the sample number is 200, the initial random seed is1, the covariance file name is“cov_matrix.csv”, energy grid number is 2, the energy grid boundary is 1.0E-11 MeV, 6.8E-7 MeV,
and 20MeV, the nuclide name is U235 and U238, and reaction type (MT number) is MT=2, 18, and
102 for U235 and MT=4 and 16 for U238, respectively.

- 100 -

1.6.3 Input Instruction of perturbation_ace_file
This tool generates the perturbed ACE file when the user sets the ACE file name, reaction type,energy region, and amount of perturbation. This tool can use the output file of the"make_perturbation_factor" tool.
The perturbation tool perturbs the cross-section, the number of neutrons per fission (νtotal, νdelay,and νprompt), and fission spectrum line by line. The input format of this tool is as follows:“MT number” Emax

Emin for“MT number” Emin Emaxfwhere Emax is the maximum energy of perturbation [MeV], Emin is the minimum energy of
perturbation [MeV], and f is the amount of perturbation, respectively. Note that the above data must
be set in one line. The number of the perturbed data files is identical to the number of lines. The
available reaction type (MT number) is all reaction cross-section type, νtotal (MT=452), νdelay(MT=455), νprompt (MT=456), and fission spectrum (MT=1018).
For example, if the user wants to modify elastic scattering cross-section (MT=2) from 1.0E-11
MeV to 1.0E-10 MeV and fission cross-section (MT=18) from 1.0E-6 eV to 1.0E-5 eV, theperturbation data file is as follows:
2 1.0E-10 1.0E-11 1.1
18 1.0E-5

1.0E-60.9

or
2 1.0E-11 1.0E-10 1.1
18 1.0E-6

1.0E-50.9

The execution command of the perturbation tools is as follows:./perturbation_ace_file.exe “ACE file name” “input file list”
The perturbation tools developed for the uncertainty analysis tool using the Random Sampling
method. The input file list is required to run this tool and the input file name is written in the input
file list to perturb a lot of ACE files simultaneously.
This tool is developed to use the output file of "make_perturbation_factor" as the perturbation data
file. The input file name must be "AAA_nnnn", where "AAA" is the file name and "nnnn" is 4
numbers. For example, if there are three perturbation data files and the target nuclide is U-235, the- 101 -

input file list is as follows:
inp/U235_0001
inp/U235_0020
inp/U235_0300
The perturbation tool reads "U235_0001", "U235_0020", and "U235_0300" files in "inp" directoryand output three perturbed ACE files. The perturbed ACE file name is automatically set using 4
numbers in the perturbation data file name (the "nnnn" part in the input file name). In this case, theperturbed ACE file name is ace_file.ace_0001, ace_file.ace_0020, and ace_file.ace_0300 when theoriginal ACE file name is "ace_file.ace_0001".

1.7 Input Instruction of ENDF Modification Tool
This tool is involved in the executable of FRENDY. The execution method is identical to theprocessing of the nuclear data file.
The user must set the processing mode as the first parameter. The processing mode to modify the
ENDF-6 formatted file is as follows:“endf_file_modification_mode”, “endf_file_modify_mode”, “endf_mod_mode”, or“mod_endf_mode”
The modification tool performs the modifications line by line. If the user wants to remove specified
MF/MT data, the user sets the following command:
remove MF “MF number” MT “MT number” “Original file name” “Modified file name”
remove MF “MF number”
remove“Original file name” “Modified file name”
MT “MT number” “Original file name” “Modified file name”

If the user selects both MF and MT numbers, specified MF/MT data is removed from the original
file. If the user only selects the MF number, specified MF data is removed. If the user only selects
the MT number, specified MT data in all MF data are removed.
If the user wants to replace the specified MF/MT data, the user sets the following command:
replace MF “MF number” MT “MT number” “Original file name” “Replaced file name”“Modified file name”
replace MF “MF number” “Original file name” “Replaced file name” “Modified file name”- 102 -

replace MT “MT number” “Original file name” “Replaced file name” “Modified file name”
The modification tool removes the specified MF/MT data in the original file and copies the specified
MF/MT data in the replaced file. The specified MF/MT data in the replaced file is copied even if theoriginal file does not contain the specified MF/MT data
If the user wants to add the specified MF/MT data, the user sets the following command:
add

MF “MF number” MT “MT number” “Original file name” “Replaced file name”“Modified file name”

add

MF “MF number” “Original file name” “Replaced file name” “Modified file name”

add

MT “MT number” “Original file name” “Replaced file name” “Modified file name”

The modification tool copies the specified MF/MT data in the replaced file. The specified MF/MT
data is replaced if the original file contains the specified MF/MT data.
If the user wants to linearize the specified MF/MT data, the user sets the following command:
linearize MF “MF number” MT “MT number” “Original file name” “Modified file name”
linearize MF “MF number” “Original file name” “Modified file name”
The linearization of the data is only available for the specified MF or MF/MT data.

1.8 Sample Input of ENDF Modification Tool
When the user wants to linearize MF=4, 5, and 6 data of “./j40/Pu239.dat”, the sample input of
modification tools is as follows.
endf_file_modification_mode //processing modelinearize MF 4 ./j40/Pu239.dat./Pu239_mod01.dat

linearize MF 5 ./Pu239_mod01.dat ./Pu239_mod02.dat
linearize MF 6 ./Pu239_mod02.dat ./Pu239_mod03.dat
In this sample input, the modified file name is “./Pu239_mod03.dat”.
This tool can linearize some evaluated nuclear data files in one input file. When the user wants tolinearize MF=4 and MF=5/MT=18 of “./j40/U235.dat”, “./j40/U238.dat”, and “./j40/Pu239.dat”, thesample input of modification tools is as follows.

- 103 -

endf_file_modification_mode //processing modelinearize MF 4./j40/U235.dat./U235_mod01.dat

linearize MF 5 MT 18 ./U235_mod01.dat./U235_mod02.dat

linearize MF 4./U238_mod01.dat./j40/U238.dat

linearize MF 5 MT 18 ./U238_mod01.dat./U238_mod02.dat

linearize MF 4./Pu239_mod01.dat./j40/Pu239.dat

linearize MF 5 MT 18 ./Pu239_mod01.dat ./Pu239_mod02.dat
In this sample input, the modified file name is “./U235_mod02.dat”, “./U238_mod02.dat”, and“./Pu239_mod02.dat”, respectively.
The user can also set the tolerance of the linearization. The default tolerance value is 1.0E-3, i.e.,
0.1%. When the user wants to linearize a small tolerance value, the sample input is as follows:
endf_file_modification_mode //processing mode
tolerance 1.0E-4
linearize MF 4./j40/U235.dat./U235_mod01.dat

linearize MF 5 MT 18./U235_mod01.dat./U235_mod02.dat

In this sample input, the tolerance value is changed from 1.0E-3 to 1.0E-4.

1.9 Sample Input of XS Output Tool
This tool outputs (x, y) tables from ENDF (PENDF), ACE, and GENDF formatted files. FRENDY
does not prepare the plot function. This tool only outputs the (x, y) table and the user has to plot thegraph using gnuplot, Matplotlib, Excel, and so on. The unit of energy is eV. Though the ACE fileuses MeV for energy, this tool changes from MeV to eV. The calculation conditions, e.g., nuclear
data file name, temperature, and MT number, are written in the head of the output file.
The available input parameter of this tool is as follows:
For output the (x, y) table:“nucl_file_name”, “ace_file_name”, “gendf_file_name”, “error”, “error_max”, “temp”,“output_name”, “edit_flag”, “mf_list”, “mt_list”, and “mat_no”
For multi-group cross section generation from the continuous cross section data in the ENDF and theACE formatted files:“mg_structure”, “mg_weighting_spectrum_data”, “mg_weighting_spectrum_data_int”, and“mg_weighting_spectrum_mode”- 104 -

This tool outputs the (x, y) table of a nuclear data file and comparison results (cross sections andrelative differences) for comparing differences between the nuclear data libraries.
This tool processes resonance reconstruction, the Doppler broadening, and the multi-group crosssection generation. There are some differences between nuclear data processing and this tool. The
major differences are as follows:⚫

The default torelance value of the linearization for the resonance reconstruction and the
Doppler broadening is changed from 0.001 (0.1%) to 0.01 (1%) to reduce the processing
time and data size.

⚫

The maximum Doppler broadening energy is fixed at 1 MeV. In nuclear data processing, thedefault Doppler broadening energy is the maximum energy of the resolved resonance region.

⚫

The “Fission+1/E+Maxwell” option which is the weighting spectrum for the multi-groupcross section generation is not available in this tool.

⚫

This tool cannot process the thermal scattering law data. The user has to prepare the PENDF
file or the ACE file to output the cross section of thermal scattering law data.

1.9.1 Sample Input Data to Output Continuous Energy Cross Section Data from an
ENDF Formatted File
The sample input to output MT=1 (total), 2 (elastic), 18 (fission), and 102 (radiative capture) crosssection from the ENDF formatted file is as follows:
plot_mode //processing mode
nucl_file_name ( "../lib/U235.dat " )output_name ( "./output_xs/U235" )mt_list ( 1 2 18 102 )In this sample input, the output data file name is “./output_xs/U235_MT.dat”.
The sample input to change the temperature is follows:
plot_mode //processing mode
nucl_file_name ( "../lib/U235.dat " )output_name ( "./output_xs/U235" )temp ( 300.0 ) //Temperature [K]mt_list ( 1 2 18 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102- 105 -

This tool checks the nuclear data temperature. If the nuclear data temperature is smaller than the
input temperature, it calculates Doppler broadening cross sections. This tool outputs an error
message if the nuclear data temperature is larger than the input temperature,
If the user wants to output consecutive MT numbers, please use “-“, e.g., 18-102, 18 - 102, or 18-102. The sample input file outputs MT=1, 2, 18, 19, 20, …, 99, 100, 101, and 102.
The default torelance value for the linearization (error option) process, i.e., resonancereconstruction and Doppler broadening, is 0.01. This value is 10 times larger than the ACE file
generation to reduce the processing time. The sample input to obtain a more detailed distribution.,
i.e., to change the torelance value, is as follows:
plot_mode //processing mode
nucl_file_name ( "../lib/U235.dat " )output_name ( "./output_xs/U235" )temp 300.0
error 1.0E-3 //1 % -> 0.1 %mt_list ( 1 2 18 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
If the user wants to output all reaction type data, please remove the “mt_list” option as follows:
plot_mode //processing mode
nucl_file_name ( "../lib/U235.dat " )output_name ( "./output_xs/U235" )temp 300.0
error 1.0E-3 //1 % -> 0.1 %1.9.2 Sample Input Data to Output Multi-Group Cross Section Data from an ENDF
Formatted File
The sample input file to output multi-group cross section data from the ENDF formatted file is asfollows:
plot_mode //processing mode
endf_file_name ( "../lib/U235.dat " )output_name ( "./output_xs/U235" )
temp

300.0

error

1.0E-3 //1 % -> 0.1 %- 106 -

mt_list ( 1 2

18 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102

mg_structure ( xmas_nea-lanl_172 ) //Identical to ign=18 in GROUPR/NJOY
mg_weighting_spectrum ( 1/e )//Identical to iwt=3 in GROUPR/NJOY

edit_flag( "histogram" )The available multi-group cross section generation option is “mg_structure”,“mg_weighting_spectrum_data”, “mg_weighting_spectrum_data_int”, and“mg_weighting_spectrum_mode”. Note that “Fission+1/E+Maxwell” option cannot be used in thisfunction.
The sample input file to output the non-histogram cross section data, i.e., to change the “edit_flag”
option from “histogram” to “non-histogram”, is as follows:
plot_mode //processing mode
endf_file_name ( "../lib/U235.dat " )output_name ( "./output_xs/U235" )
temp

300.0

error

1.0E-3 //1 % -> 0.1 %mt_list ( 1 2

18 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102

mg_structure ( xmas_nea-lanl_172 ) //Identical to ign=18 in GROUPR/NJOY
mg_weighting_spectrum ( 1/e )//Identical to iwt=3 in GROUPR/NJOY

edit_flag( "non-histogram" )The difference between histogram and non-histogram is shown in Fig. 1.2.2.

1.9.3 Sample Input Data to Output Continuous Energy Cross Section Data from an
ACE Formatted File
The sample input file to output continuous energy cross section data from the ACE formatted fileis as follows:
plot_mode //processing mode
ace_file_name ( "../ace/U235.ace" )output_name ( "./output_xs/U235" )temp ( 300.0 )mt_list ( 1 2 18 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
This tool outputs the specified temperature data using the Doppler broadening calculation.
- 107 -

This tool treats the thermal scattering law data. The sample input files to output the thermal
scattering law data are as follows:
plot_mode //processing mode
ace_file_name ( "../ace/HinH2O.ace " )output_name ( "./output_xs/HinH2O" )plot_mode //processing mode
ace_file_name_tsl ( "../ace/HinH2O.ace " )output_name ( "./output_xs/HinH2O" )Note that the Doppler broadening calculation is not carried out when the user selects the thermal
scattering law data. As described in Section 1.9, this tool cannot process the thermal scattering law
data. The user has to prepare the PENDF file or the ACE file to output the cross section of thermal
scattering law data.

1.9.4 Sample Input Data to Output Multi-Group Cross Section Data from an ACE
Formatted File
The sample input file to output multi-group cross section file from the ACE formatted file is asfollows:
plot_mode //processing mode
endf_file_name ( "../lib/U235.dat " )output_name ( "./output_xs/U235" )
temp

300.0

error

1.0E-3 //1 % -> 0.1 %mt_list (1 2 18 - 102) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
mg_structure ( xmas_nea-lanl_172 ) //Identical to ign=18 in GROUPR/NJOY
mg_weighting_spectrum ( 1/e )//Identical to iwt=3 in GROUPR/NJOY

edit_flag( "histogram" )The sample input file to output the multi-group cross section of thermal scattering law data is asfollows:
plot_mode //processing mode
ace_file_name ( "../ace/HinH2O.ace " )- 108 -

output_name ( "./output_xs/HinH2O" )mg_structure ( xmas_nea-lanl_172 ) //Identical to ign=18 in GROUPR/NJOY
mg_weighting_spectrum ( 1/e )//Identical to iwt=3 in GROUPR/NJOY

edit_flag( "histogram" )1.9.5 Sample Input Data to Output Multi-Group Data from a GENDF Formatted File
The sample input file to output multi-group cross section file from the GENDF formatted file is asfollows:
plot_mode //processing mode
gendf_file_name ( "../gendf/U235.gendf" )output_name ( "./output_xs/U235" )mf_list ( 3 6 )mt_list ( 1 2 18 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
edit_flag( histogram )1.9.6 Sample Input Data to Output Comparison Results of Continuous Energy Cross
Section Data from ENDF/ACE Formatted Files
The sample input file to compare the ENDF formatted files is as follows:
plot_mode //processing mode
endf_file_name ( "../lib/U235.dat""../pendf/U235.pendf" )//Reference: ../lib/U235.dat//Comparison : ../pendf/U235.pendf//Relative difference : (comp – ref) / refoutput_name ( "./comp/U235_comp" )temp ( 300.0 )mt_list ( 1 2 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
The sample input file to compare the ENDF and the ACE formatted files is as follows:
plot_mode //processing mode
endf_file_name ( "../lib/U235.dat" ) //Reference
ace_file_name ( "../ace/U235.ace" ) //Comparison//Relative difference : (comp – ref) / ref- 109 -

output_name ( "./comp/U235_comp" )temp ( 300.0 )mt_list ( 1 2 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
The sample input file to compare the ACE formatted files is as follows:
plot_mode //processing mode
ace_file_name ( "../ace/U235_J40.ace""../ace/U235_J50.ace" )//Reference: ../ace/U235_J40.ace//Comparison : ../ace/U235_J50.ace//Relative difference : (comp – ref) / refoutput_name ( "./comp/U235_comp" )temp ( 300.0 )mt_list ( 1 2 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
The sample input files to compare the thermal scattering law data are as follows:
plot_mode //processing mode
ace_file_name ( "../ace/HinH2O_J40.ace" “../ace_HinH2O_J50.ace” )//Reference: ../ace/HinH2O_J40.ace//Comparison : ../ace/HinH2O_J50.ace//Relative difference : (comp – ref) / refoutput_name ( "./comp/HinH2O_comp" )plot_mode //processing mode
nucl_file_name ( "../pendf/HinH2O_J40.pendf")
ace_file_name ( “../ace_HinH2O_J50.ace” )//Reference: ../pendf/HinH2O_J40.pendf//Comparison : ../ace/HinH2O_J50.ace//Relative difference : (comp – ref) / refoutput_name ( "./comp/HinH2O_comp" )- 110 -

1.9.7 Sample Input Data to Output Comparison Results of Multi-Group Cross Section
Data from ENDF/ACE Formatted Files
The sample input file to compare the ENDF formatted files is as follows:
plot_mode //processing mode
endf_file_name ( "../lib/U235.dat""../pendf/U235.pendf" )//Reference: ../lib/U235.dat//Comparison : ../pendf/U235.pendf//Relative difference : (comp – ref) / refoutput_name ( "./comp/U235_comp" )temp ( 300.0 )mt_list ( 1 2 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
mg_structure ( xmas_nea-lanl_172 ) //Identical to ign=18 in GROUPR/NJOY
mg_weighting_spectrum ( 1/e )//Identical to iwt=3 in GROUPR/NJOY

edit_flag( "histogram" )The sample input file to compare the ENDF and the ACE formatted files is as follows:
plot_mode //processing mode
endf_file_name ( "../lib/U235.dat" ) //Reference
ace_file_name ( "../ace/U235.ace" ) //Comparison//Relative difference : (comp – ref) / refoutput_name ( "./comp/U235_comp" )temp ( 300.0 )mt_list ( 1 2 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
mg_structure ( xmas_nea-lanl_172 ) //Identical to ign=18 in GROUPR/NJOY
mg_weighting_spectrum ( 1/e )//Identical to iwt=3 in GROUPR/NJOY

edit_flag( "histogram" )The sample input file to compare the ACE formatted files is as follows:
plot_mode //processing mode
ace_file_name ( "../ace/U235_J40.ace""../ace/U235_J50.ace" )//Reference: ../ace/U235_J40.ace- 111 -

//Comparison : ../ace/U235_J50.ace//Relative difference : (comp – ref) / refoutput_name ( "./comp/U235_comp" )temp ( 300.0 )mt_list ( 1 2 - 102 ) //MT=1, 2, 18, 19, 20, ..., 99, 100, 101, and 102
mg_structure ( xmas_nea-lanl_172 ) //Identical to ign=18 in GROUPR/NJOY
mg_weighting_spectrum ( 1/e )//Identical to iwt=3 in GROUPR/NJOY

edit_flag( "histogram" )The sample input files to compare the thermal scattering law data are as follows:
plot_mode //processing mode
ace_file_name ( "../ace/HinH2O_J40.ace" “../ace_HinH2O_J50.ace” )//Reference: ../ace/HinH2O_J40.ace//Comparison : ../ace/HinH2O_J50.ace//Relative difference : (comp – ref) / refoutput_name ( "./comp/HinH2O_comp" )mg_structure ( xmas_nea-lanl_172 ) //Identical to ign=18 in GROUPR/NJOY
mg_weighting_spectrum ( 1/e )//Identical to iwt=3 in GROUPR/NJOY

edit_flag( "histogram" )plot_mode //processing mode
nucl_file_name ( "../pendf/HinH2O_J40.pendf")
ace_file_name ( “../ace_HinH2O_J50.ace” )//Reference: ../pendf/HinH2O_J40.pendf//Comparison : ../ace/HinH2O_J50.ace//Relative difference : (comp – ref) / refoutput_name ( "./comp/HinH2O_comp" )mg_structure ( xmas_nea-lanl_172 ) //Identical to ign=18 in GROUPR/NJOY
mg_weighting_spectrum ( 1/e )//Identical to iwt=3 in GROUPR/NJOY

edit_flag( "histogram" )- 112 -

Installation of FRENDY

2This chapter describes how to install FRENDY on Linux, UNIX, or macOS platforms.
A disk space of about 1.2 Gbytes is required to make the executables from the program sources1.

and run all test programs and sample files. A disk space of about 210 Mbytes is required if users
make only the executables from the program sources.
A C++ compiler, the Boost library, and the LAPACK library are required to compile source2.

programs. The compilation has been confirmed for an Intel compiler (ICC) version 13.1.3 and a
GNU compiler (GCC) version 4.4.7 with the Boost library version 1.60.0 and the LAPACK
library version 3.8.0.
Since FRENDY stores all nuclear data on memory, a large memory size is required. More than3.

1 Gbytes is recommended to run FRENDY.

2.1

Directory Structure

The directory structure is shown in Fig. 2.1.1. The “frendy” directory contains the source files.
The “sample” directory contains the input files for the test calculations to generate the fast andthermal ACE files. The “test” directory contains source files to run the Boost.Test library.

frendycompile_all.csh
main

frendy.exe

samplerunrun_frendy.csh

tests

run_frendy.csh

README
Figure 2.1 Directory structure of FRENDY- 113 -

2.2

How to Install FRENDY on Linux, UNIX, or macOS Platforms

A shell script of “compile_all.csh” is included in the “frendy” directory to compile FRENDY
source codes. The users run only this shell script or issue the “make” command in “frendy/main”
directory. If the compiler successfully builds the executable file of FRENDY, the following message
is written on the display.
### Completed to make FRENDY ###The executable file named “frendy.exe” is generated in “frendy/main” directory.
FRENDY has the NJOY mode which processes the evaluated nuclear data file with the NJOY
calculation method. The differences between the NJOY mode and the original mode are as follows:

1.

The NJOY mode calculates the unresolved resonance cross-section using cross-section
formulae on the fixed energy grid points and other energy grid points are calculated by theinterpolation method. The fixed energy grid points are 1.00 × 10𝑛 , 1.25 × 10𝑛 , 1.50 ×
10𝑛 , 1.70 × 10𝑛 , 2.00 × 10𝑛 , 2.50 × 10𝑛 , 3.00 × 10𝑛 , 3.50 × 10𝑛 , 4.00 × 10𝑛 , 5.00 ×
10𝑛 , 6.00 × 10𝑛 , 7.20 × 10𝑛 , and 8.50 × 10𝑛 . This difference has an impact on resonancereconstruction.

2.

The calculation method of the complex error function which is defined in Eq. (2.3.41) isdifferent. The complex error function is used to calculate the Doppler broadened cross-sections
with the Single-Level Breit-Wigner representation. This difference has an impact on resonancereconstruction and probability table generation.

3.

The calculation method of the cross-sections at 0 eV is different. The cross-section at 0 eV isrequired to calculate the Doppler broadened cross-sections at the low energy region. The NJOY
mode assumes that the cross-section obeys the 1/v law. This approximation is appropriate formany reactions. However, the elastic scattering cross-sections do not obey the 1/v law since theelastic scattering cross-section is constant due to the potential scattering cross-section at thelow-energy region. In such a case, the linear-linear interpolation is appropriate. The NJOY
mode uses the 1/v law to reproduce the NJOY results. This difference affects the Doppler
broadened elastic scattering cross-sections at the low-energy region.

4.

The NJOY mode uses the fixed incident neutron energy grid to calculate the incoherent inelastic
scattering cross-section. The number of energy grid points is 117 from 1.0 × 10−5 to 10 eV tocalculate the incoherent inelastic scattering cross-section and secondary energy and angular
distributions. The incoherent inelastic scattering cross-section at the other energy grid points isinterpolated using the fifth-order Lagrange interpolation and secondary energy and angular- 114 -

distributions are not calculated. This difference affects some materials, e.g., H in ZrH, for which
the incoherent inelastic scattering cross-section oscillates.
The NJOY mode uses discrete random numbers to calculate the chi-squared random numbers

5.

with 𝑘 degrees of freedom 𝑅𝜒2 (𝑘). This difference has an impact on the probability table
generation.
Users run “compile_all.csh” or “make” command in “frendy/main_njoy_mode” directory if theywant to compile with this calculation mode. The executable file named “frendy_njoy_mode.exe” isgenerated in “frendy/main_njoy_mode” directory.

2.3 How to Execute FRENDY
The execution command is
frendy.exe “input file name”
The input file name is set as a command-line argument.

2.4

Test Calculation with Samples

The sample input files are located in the “sample/input” directory. The sample input files which
are listed in this manual in Sec. 10 are used as the sample calculations. All nuclides and thermal
scattering law data are copied from JENDL-4.0. The “run_all.csh” shell script in the “sample/run”
directory automatically runs all the sample input files.

2.5

Test Programs for Boost Test Library

For quality assurance, the test programs are included to verify the capabilities. The Boost test
library is used for the test programs. Using the Boost test library, comparison of the calculation
results and confirmation of the run-time errors are easily done.
The source files for Boost test library are contained in the “test” directory. Users must run the“run_all_class.csh” shell script in the “test” directory if they want to compile and run all the test
programs. The source file name and the directory structure correspond to those of FRENDY sourcesin the “frendy” directory. Users must recompile and run the test programs when they modify thesource files.
The test programs start with the following message
Running X test cases...

- 115 -

where X is the number of test cases. If the test programs are successfully finished, the following
message is written on the display*** No errors detected
If the test programs find errors, the following message is written on the display*** Y failures detected in test suite "Master Test Suite"where Y is the number of errors.

2.6

How to Install FRENDY on Visual Studio 2019

This section shows the installation of FRENDY using Visual Studio 2019. The author strongly
recommends using the Windows Subsystem for Linux to compile FRENDY since the installation of
FRENDY using Visual Studio 2019 is difficult. The installation of FRENDY using Windows
Subsystem for Linux is explained in “FRENDY installation” on the web page of the FRENDY
training course: https://rpg.jaea.go.jp/main/en/program_frendy/index.html.

2.6.1 Installation of Boost Library
Installation of the Boost library is required before the compilation of FRENDY. The installation of
the Boost library is as follows:
1. Go to the web page of the Boost library: https://www.boost.org/doc/
2. Click “Getting Started”.
3. Click “Next: Getting Started on Microsoft Windows”.
4. Download the latest version of the Boost library.
5. Unzip the downloaded file.
6. Open the x64 native tools command prompt for VS2019. (Windows menu -> Visual Studio 2019-> x64 Native Tools Command Prompt for VS2019)
7. Move to the unzipped Boost directory.
8. Run “bootstrap”.
9. Run “.¥b2”- 116 -

2.6.2 Installation of CLAPACK Library
Installation of the CLAPACK library is required before the compilation of FRENDY. The
CLAPACK library is only used to treat the resonance parameter written in the R-matrix limited
formula. The user can skip this process if the user doesn’t need to treat this resonance parameter.
The installation of the CLAPACK library is as follows:
1. Go to the web page of the CLAPACK library: http://icl.cs.utk.edu/lapack-for-windows/clapack/
2. Download the CLAPACK library from the following web page:
http://icl.cs.utk.edu/lapack-for-windows/clapack/clapack-3.2.1-CMAKE.tgz
3. Decompress the CLAPACK library.
4. Install the CLAPACK library.
5. Run installed CMake and select the source code directory of the CLAPACK library and the new
directory to build.
6. Select Visual Studio 2019.
7. Click the “Finish” button.
8. Select the compiled directory and click the “Configure” button.
9. Click the “Generate” button.
10. Click the “Open Project” button and open the window of Visual Studio 2019.
11. Select “ALL_BUILD” in the solution explorer and run debug and release.
12. Select “INSTALL” in the solution explorer and run debug and release.

2.6.3 Installation of FRENDY
The installation of FRENDY is as follows:
1. Download the “msdirent.h” file instead of the “dirent.h” file from the following web page:
http://svn.apache.org/repos/asf/avro/trunk/lang/c/tests/msdirent.h
2. Copy the “msdirent.h” file to the “frendy¥VisualStudio” directory.
3. Generate a project of FRENDY with “Continue without code”.
4. Click File -> New -> Project From Existing Code.
5. Input source directory of FRENDY.
6. Select “Console Application Project”.
7. Click the “Finish” button.
8. Remove main.cpp, main_frendy_njooy_mode.cpp, and main_frrendy.cpp.
9. Open property of this project. (Project -> Property).
10. Add include directory of the Boost library, the CLAPACK library and FRENDY.
- 117 -

11. Add additional dependency file as follows:
For release : lapack.lib;blas.lib;libf2c.lib;
For debug : lapackd.lib;blasd.lib;libf2cd.lib;
12. Compile FRENDY using release mode.

2.6.4 Running FRENDY with bat Files
The users can run FRENDY using the following two bat files:(1) run.bat
set INPNAME=inp_JENDL-4_U235
call run_frendy.bat(2) run_frendy.bat
set EXE=frendy.exe
rem set EXE= frendy.exe
set INP_DIR=..¥input
set LIB_DIR=..¥lib
set OUT_DIR=out
set PEN_DIR=pendf
set ACE_DIR =ace
mkdir %OUT_DIR%
mkdir %PEN_DIR%
mkdir %ACE_DIR_FAST%
mkdir %ACE_DIR_THERM%
set DIR_DATA=xsdir_list
set LOG=log
rem del %DIR_DATA%
rem del %LOG%
set INP=%INP_DIR%¥%INPNAME%.dat
set LIB=%LIB_DIR%¥%LIBNAME%.dat- 118 -

set LIBK=%LIB_DIR%¥%TSKNAME%.txt
copy %LIB% tape20
if exist %LIBK% copy %LIBK% tape23
echo %LIBNAME%
echo %LIB%
echo %LIBK%
echo %LIBNAME% >> %LOG%
echo "">> %LOG%%EXE% %INP%>> %LOG%echo "">> %LOG%echo "=======================================================" >> %LOG%
echo "">> %LOG%copy output %OUT_DIR%¥frendy_output_%LIBNAME%%TSKNAME%.dat
copy tape25 %PEN_DIR%¥frendy_result_%LIBNAME%%TSKNAME%.dat
copy tape30 %ACE_DIR %¥frendy_acer_%LIBNAME%%TSKNAME%.dat
type tape31 >> %DIR_DATA%
del output
del tape20
del tape21
del tape22
del tape23
del tape25
del tape30
del tape31
rem rmdir /s /q %OUT_DIR%
rem rmdir /s /q %PEN_DIR%
rem rmdir /s /q %ACE_DIR_FAST%
rem rmdir /s /q %ACE_DIR_THERM%- 119 -