2021年度核データ研究会

1Development of FRENDY Version 2
Kenichi TADA (JAEA),Akio YAMAMOTO, Tomohiro ENDO (Nagoya Uni.),
Go CHIBA (Hokkaido Uni.),
Michitaka ONO, Masayuki TOJO (GNF-J)2

2021年度核データ研究会

Overview of FRENDY• JAEA started developing a new nuclear dataprocessing system FRENDY in 2013.

• FRom Evaluated Nuclear Data librarY to any application• To process nuclear data library by JAEA’s nuclear application

code users with simple input file.

• FRENDY Version 1 was released in 2019.
• FRENDY Ver. 1 only generates ACE files.
• https://rpg.jaea.go.jp/main/en/program_frendy/• FRENDY Version 2 will be released in Mar. 2022.
• FRENDY Ver. 2 generates multi-group XS file.
Nucleardata library

Nuclear dataprocessingsystem
FRENDYk-eff , flux, …

Nuclearcalculation codes

2021年度核データ研究会

3Features of FRENDY• Utilization of modern programming techniques• C++, BoostTest library, Git• Improvement of quality and reliability• Consideration of maintainability, modularity, andflexibility• Encapsulate all classes• Minimize the function of module• Maintain the independence of each module• Simple input format• Processing mode and nuclear file name is required at minimum.
• NJOY input format is also available.
Ref. K. Tada, et. al., “Development and verification of a new nuclear data processing system
FRENDY,” J. Nucl. Sci. Technol., 54 [7], pp.806-817 (2017).
(http://www.tandfonline.com/doi/abs/10.1080/00223131.2017.1309306)

2021年度核データ研究会

4

Input format of FRENDY• FRENDY treats two types of the input formats.
• FRENDY original input format• NJOY compatible• Simple and easy input data
• The simplest input: Nuclear data file name and processing
mode are only required for the processing.
• FRENDY has recommended parameters in the source code.
• User can also change (override) parameters.

2021年度核データ研究会

Input format of
FRENDY and NJOY
• Input parameters of

FRENDY consist of “inputdata name” and “inputdata”.
• Comment line is similar toC/C++.
• //~ or /* ~ */• Input format of FRENDY

is easy to understand.
• This input format is suitablefor beginners.

5[Sample input of FRENDY]
ace_fast_mode // Processing mode
nucl_file_name U235.dat
ace_file_name U235.ace
temp
296.0

[Sample input of NJOY]
reconr/ command
20 21/ input(tape20), output(tape21)‘PENDF file of U235’/ identifier for PENDF
9228
/ mat
1.00e-03 0.00
/ err, temp0/
broadr/ command
20 21 22
/ endf, pendf(in), pendf(out)9228 1/ mat, temp no1.00e-03 -5.0E+2/ err, thnmax
296.0/ temp0/
gaspr/ command
20 22 23
/ endf, pendf(in), pendf(out)purr/ command
20 23 25
/ endf, pendf(in), pendf(out)9228 1 10 20 100/ mat, temp no, sig no, bin no, lad no296.0/ temp
1E10 1E4 1E3 300 100 30 10 1.0 0.1 1.0E-5 / sig zero
0/acer/ command
20 25 0 30 31/ nendf, npend, ngend, nace, ndir
1 1 1 0.30
/ iopt(fast), iprint(max), itype, suffix‘ACE file of U235’/ descriptive character
9228 296.0/ mat, temp
1
/ newfor(yes)/ acer end
stop/2021年度核データ研究会

Release of FRENDY Ver. 1• FRENDY Ver.1 was released from our website.
• https://rpg.jaea.go.jp/main/en/program_frendy/• Only generates ACE files.
• Generation of multi-group cross section will be implementedsoon.

• Open-source software• 2-Clause BSD license• Presentations of FRENDY training course andexercise are also found in this website.
• Manual of FRENDY Ver. 1• JAEA-Data/Code 2018-014• https://jopss.jaea.go.jp/pdfdata/JAEA-Data-Code-2018-014.pdf

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Development of FRENDY Ver. 2• Many capabilities are prepared after FRENDYVer. 1.
• Neutron induced multi-group cross section generation• Perturbation of ACE file for uncertainty quantification• This function was implemented in FRENDY Ver.1.01.001.
• Modification of evaluated nuclear data file
• Uncertainty quantification for probability tables• Improvement of input checker to reduce input errors• FRENDY version 2 will be released includingthese functions in Mar. 2022.

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2021年度核データ研究会

Multi-group XS generation• FRENDY/MG generates multi-group

XS files from ACE files.

• FRENDY/MG can also generate amulti-group XS library from the existing
ACE library.

…

multi-group XS files.

Linearization

Probability-tablecalculation• NJOY input are also available for

ACE file generation• Input of GROUPR and MATXSRACE lib.

multi-group XS file generation.
modules are available.

*) A. Yamamoto, K. Tada, G. Chiba, T. Endo, “Multi-groupneutron cross section generation capability for FRENDY
nuclear data processing code,” J. Nucl. Sci. Technol., 2021.
https://doi.org/10.1080/00223131.2021.1921631

FRENDY/MG
Multi-group
XS lib.

FRENDY Version 1• FRENDY/MG*) is used for aEvaluatednuclear data lib.

2021年度核データ研究会

9Major capabilities of FRENDY/MG• Focus on neutron cross section generation.
• It can treat fast continuous and thermal scattering law data.
• Output format of multi-group cross sections• GENDF and MATXS• Angular/energy distributions• LAW=3, 4, 7, 9, 11, 44, 61 66 in ACE file
• All nuclides in JENDL-4.0, ENDF/B-VII.1, B-VIII.0, JEFF-3.3,and TENDL-2019 are available.
• Scattering matrices• It can treat all reactions including anisotropic scattering.
•：free gas model, incoherent inelastic, incoherent elastic,coherent elastic

2021年度核データ研究会

10New functions of FRENDY/MG• Back-ground cross-section set can be automatically

set with minimum number of background crosssections*).
• A compound of different isotope can be specified to
explicitly consider the resonance interference effect.
• For example, U-235, U-238, and O-16 in UO2.

• Current (P1 flux) weighted total cross-section can becalculated.
• Any energy grid points can be used for ultra-fine groupslowing down calculation.
*) A. Yamamoto, T. Endo, K. Tada, “Adaptive setting of background cross sections forgeneration of effective multi-group cross sections in FRENDY nuclear data processing code,”
J. Nucl. Sci. Technol., 2021. https://doi.org/10.1080/00223131.2021.1944930

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2021年度核データ研究会

Verification of FRENDY/MG• Comparison of all processing results between

FRENDY/MG and NJOY
• All nuclides in JENDL-4.0, ENDF/B-VII.1, B-VIII.0, andJEFF-3.3• The processing results shows good agreement.
• Comparison of neutronics calculations• PWR pin-cell (5 wt% UO2, 600 K)[Comparison of neutron spectrum]
Neutron flux, UO2 pincell• MOC (GENESIS code), 172 gr.

:free gas

• K-effective• NJOY：1.4089250• FRENDY/MG：1.4089277

Neutron flux (per unit lethargy)•

6.0E+00
5.0E+00
4.0E+00

Pellet(NJOY-GENESIS)
Pellet(MG-GENESIS)

3.0E+00
2.0E+00
1.0E+00
1.0E-07
1.0E-03

1.0E-01

1.0E+01

1.0E+03
Energy [eV]

1.0E+05

1.0E+0712

2021年度核データ研究会

ACE file perturbation tool
• Implementation of a sampling tool to perturb

cross section, number of neutrons per fission,and fission spectrum of ACE file*).
• Reaction type, energy region, and amount ofperturbation are required as input parameters.

• This tool is available from FRENDY Ver. 1.01.001.

XS[Example of XS
perturbation]*) R. Kondo, et al., “Implementation of random
sampling for ACE-format cross sections using
FRENDY and application to uncertainty
reduction,” Proc. M&C2019, Aug. 25-29 (2019).

×1.1

Incident neutron energy

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2021年度核データ研究会

ENDF modification function• This function removes, adds, exchanges specified

MF/MT data.

MF=1
MF=2, MT=151
MF=3, MT=1
MF=3, MT=2
MF=6remove

add
MF=1

MF=1

MF=2, MT=151

MF=2, MT=151

MF=3, MT=1

MF=3, MT=1

MF=3, MT=2

MF=6

MF=3, MT=102
MF=6The modified evaluatednuclear data file must be
checked carefully since
FRENDY does not checkthe new file.
exchange

MF=1
MF=2, MT=151
MF=3, MT=1
MF=3, MT=2
MF=6

2021年度核データ研究会

Conclusions• About FRENDY• FRENDY Ver. 1 was released in 2019 as an opensource software.
• Utilization of modern programming techniques• Consideration of maintainability, modularity, and flexibility• Simple input format• New capabilities in FRENDY Ver. 2• Neutron induced multi-group cross section generation• Perturbation of ACE file for uncertainty quantification• Modification of evaluated nuclear data file
• FRENDY Ver. 2 will be released in Mar. 2022.

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2021年度核データ研究会

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2021年度核データ研究会

Perturbation of ACE file
• Perturbation tool decreases or increases XS or number ofneutron per fission (ν) or fission spectrum (χ).
• XS, ν or χ is multiplied by perturbation factor

range.

within arbitrary energy• This tools can be adopted to two analyses.
• Sensitivity analysis with direct perturbation method• Uncertainty analysis with random sampling method
100000

Energy range ofperturbation

XS [b]
微視的断面積10000
1000

XS is discontinuity
at boundaries ofenergy region.

100
10
6.0

6.2

6.4 6.6 6.8 7.0 7.2
Incident neutron energy [eV]

7.4

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2021年度核データ研究会

Relations of each reaction type
MT = 2MT = 1MT = 5, 11, 17, 22 - 26, 28 - 37,
41 - 42, 44 - 45, 152 - 154, 156 Fission
181, 183 - 190, 194 - 196,
198 - 200
Elastic

MT = 18MT = 19 - 21, 38
MT = 102, 108 - 117,155, 182, 191 - 193, 197

MT = 103 (n, γ)
MT = 600 - 649

Total

MT = 3MT = 4

MT = 16

MT = 50 - 91

MT = 875 - 891

MT = 27
MT = 101(n, n’)(n, 2n)MT = 104

MT = 106

MT = 650 - 699

MT = 750 - 799

MT = 105

MT = 107

MT = 700 - 749

MT = 800 - 849

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2021年度核データ研究会

Example of fission XS perturbation
Perturbed fission XS (MT=18):
MT18 contains MT=19-21 and 38.
MT=19-21 and 38 are also perturbed.
,,,(Perturabtion factor: )MT = 1MT = 3MT = 27

MT=1, 3, and 27 contain MT=18.
XS of MT=1, 3, and 27 are modified.
,MT = 18
MT=19, 20,21, and 38

2021年度核データ研究会

19Random sampling
Covariance matrix of nuclear data

inp/Nddd_000x
102 2.00E+01 6.43E+00 8.78E-01
102 6.43E+00 4.30E+00 8.67E-01
102 4.30E+00 3.00E+00 8.62E-01．．．

• User must prepare covariance matrix.
• We are now developing converter from GENDF file ofNJOY/ERROR to input of random sampling tool.
• Generation of perturbation factors using random

sampling method• See “/frendy_20yymmdd/tools/make_perturbation_factor/sample”

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2021年度核データ研究会

Uncertainty quantification using random
sampling method [1]Godiva (HMF-001)
GeometrySphere
Radius: 8.7 cm

Composition

U-235: 93.71 wt.%
U-238: 5.27 wt.%
U-234: 1.02 wt.%ୣ୤୤1.000 ± 0.001







MCNP6.2
Number of perturbed ACE file: 100
Covariance data: 56groupcov7.1(from SCALE6.2.3)
MT=2,4,16,18,102,452, and 1018
(MT=452: ν, MT=1018: χ)

Godiva [2][1] R. Kondo, et al., “Implementation of random
sampling for ACE-format cross sections using
FRENDY and application to uncertainty reduction,”
Proc. M&C2019, Aug. 25-29, Portland, USA (2019).
[2] ICSBEP NEA/NSC/DOC(95)03, Organization for
Economic Co-operation and Development-Nuclear
Energy Agency (OECD-NEA) (September 2016).

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Calculation results (k-effective uncertainty)-uncertainty due to all nuclides and reactions[%]
Sensitivity analysis (SA) Random sampling method usingof MCNP6.2
perturbation tool
1.11
1.12 [0.98 – 1.24]
Comparison of-uncertainty due to individual nuclide andreaction[%]