1FRENDY:A New Nuclear Data Processing
Code being Developed at JAEA
Japan Atomic Energy Agency (JAEA)
Kenichi Tada

2Outline• Background• Overview of FRENDY• Comparison of processing results between

FRENDY and NJOY
• Conclusions

3

Overview of nuclear data processing
and FRENDY

4Importance of nuclear data processing
Reactor analysis,
Dose evaluation,…
Neutronics calculation codes(MVP,PHITS,MCNP,…)

Cross section
library
Nuclear datalibrary(JENDL,ENDF
JEFF)• Cross section library is thefundamental data for theneutronics calculations• Reliability of the cross section

library has large impact on
the neutronics calculation

NJOY is widely used togenerate cross section libraryin Japan5

Number of engineers in Japan• Neutronics calculation codeusers

• More than 1,000• Nuclear data processing codeusers

• 1~2 in each company• Total : 20~30?

• Expert of nuclear dataprocessing• Less than 10• Technical tradition of nuclear

data processing is important• Deeply understanding of thenuclear data processing is required
to appropriately generate the crosssection library6

Present situation of nuclear data processing
in JAEA• JAEA provides nuclear data library and many

neutronics calculation codes• The nuclear data processing code had not beendeveloped• Imported nuclear data processing code are used
• JAEA cannot release the nuclear data processing code forour neutronics calculation codes• Development of domestic nuclear data processing

code were desired
Nuclear datalibrary

Imported nuclear data
Domestic
nuclearprocessingcodedata processing code

NJOY, PREPRO

Neutronics
calculation code

MVP, MARBLE2,
PHITS

7

Development of nuclear data processing
code FRENDY• JAEA started developing a new nuclear dataprocessing code FRENDY in 2013
• FRom Evaluated Nuclear Data librarY to any application• To process the nuclear data library by JAEA’s nuclear

application codes users with simple input file
• The first goal is processing the nuclear data forcontinuous energy Monte Carlo codes• For MVP, PHITS of JAEA and MCNP of LANL
Nucleardata library

Nuclear dataprocessing
code
FRENDY

Nuclearapplication code
MVP、MARBLE2、
PHITS

k-eff , flux, …

8

Features of FRENDY• Utilization of modern programming techniques• C++, BoostTest library, Git• Improvement of quality and reliability• Consideration of maintainability, modularity, portability

and flexibility• Encapsulate all classes• Minimize the function• Maintain the independence of each module• Processing methods of FRENDY is similar to NJOY99• Reflecting requests of nuclear data processing codeusers

• Development of FRENDY is supported by many organizations

and companies in Japan

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)

9Development system of FRENDY• Development of FRENDY is supported manyorganization concerning to nuclear data processingin Japan• Reflecting request of nuclear data processing code users

・Reactor physics group・Nuclear data group
Discussdevelopmentof FRENDY

Development team

Report thedevelopment
statusRequests(function, userinterface, …)Users group
・JENDL committee
Nuclear data processing WG
Memberuniversity, regulatory agency,
manufacturer

10Structure of FRENDY• Modularity is carefully considered• Modules of FRENDY can be used other calculation code byadding only a few lines
ENDF-6
formatEndf6Parser/Writer
Resonance
Reconstructor
DopplerBroader

Endf6
Converter

Gnds
Converter

NuclearData
ObjectAceDataGenerator

GasProduction
CrossSection
Calculator

AceDataObject

Implemented
moduleAceDataParser/Writer

Not implemented
moduleACE format

GndsParser/Writer

GNDS
formatHeatingCross
SectionGenerator
ThermalScattering
DataProcessor

UnresolvedResonance
DataProcessor

11

GNDS format• Developed by OECD/NEA/NSC/WPEC/SG38• Currently, maintained by WPEC/EGGNDS• Completely different from ENDF-6 format• Utilizing Extensible Markup Language (XML)• It will be used not only for nuclear data file, but also other data
file, e.g., cross section library and nuclear structure data file
• LLNL develops FUDGE code to convert ENDF-6

format to GNDS format• FUDGE code also processes nuclear data file to generatecross section library for LLNL’s neutronics calculation codes
Ref. C. M. Mattoon, et al., “Generalized Nuclear Data: a New Structure (with Supporting
Infrastructure) for Handling Nuclear Data,” Nucl. Data Sheets, 113, pp.3145-3171 (2012).
https://ndclx4.bnl.gov/gf/project/gnd/
https://www.oecd-nea.org/science/wpec/gnds/

12

Example of ENDF-6 format (MF=3)(n,2n) XS of Fe-56 from JENDL-4.0

MF
MAT
MT
2.605600+4 5.545440+100002631 3 16
1 HEAD-1.120270+7-1.120270+7001112631 3 16
2
11
2
0
0
0
02631 3 16
31.140470+7 0.000000+0 1.170000+7 1.622410-2 1.200000+7 4.800450-22631 3 164TAB1
1.300000+7 2.138200-1 1.400000+7 3.891650-1 1.500000+7 5.134000-12631 3 1651.600000+7 5.817500-1 1.700000+7 6.107500-1 1.800000+7 6.118000-12631 3 1661.900000+7 5.977000-1 2.000000+7 5.759000-1
2631 3 16
72631 3 099999 SEND

66 letters (11 data)

4 2 3

5 letters[MAT, 3, MT/ ZA, AWR, 0, 0, 0, 0] HEAD
[MAT, 3, MT/ QM, QI, 0, LR, NR, NP/ Eint/ σ(E)] TAB1
[MAT, 3, 0/ 0.0, 0.0, 0, 0, 0, 0] SENDZA, AWR : 1000.0×Z＋A, mass quantities for materials
QM：Mass-difference Q value (eV)QI : Reaction Q value
LR : Complex or “breakup” reaction flag

13

Example of GNDS format(n,2n) cross section for Fe-56 from JENDL-4.0(n,2n) reaction

Reaction<reaction label="29" outputChannel="n[multiplicity:'2']
type+ Fe55 + gamma" date="1987-03-01" ENDF_MT="16">
Cross <crossSection nativeData="linear">
Section<linear xData="XYs" length="11" accuracy="0.001"><axes><axis index=“0” label=“energy_in” unit=“eV”
interpolation="linear,linear" frame="lab"/>Interpolation<axis index=“1” label=“crossSection” unit=“b”
frame="lab"/></axes><data> 1.14e7 0.00000 1.17e7 0.0162241 1.20e7 0.0480045
1.30e7 0.21382 1.40e7 0.3891650 1.50e7 0.5134000
Cross section data 1.60e7 0.58175 1.70e7 0.6107500 1.80e7 0.6118000
1.90e7 0.59770 2.00e7 0.5759000 </data></linear></crossSection><outputChannel genre="NBody" Q="-11202700 eV"><product name="n" label="n" multiplicity="2"Secondary
ENDFconversionFlag="MF6">energy and<distributions nativeData="Legendre">angular<Legendre nativeData="LegendrePointwise">
distribution

14

Advantage for using the FRENDY’s
original nuclear data format• FRENDY uses independent internal nuclear data format• NuclearDataObject class• Minimizing the impact by the change of nuclear data

format• Developer and users are not necessary to consider the nuclear

data format• Consideration of a new data format GNDS

• GNDS format can be addressed if another set of parser, writer andconverter classes are implemented

ENDF-6
formatEndf6Parser/Writer
Resonance
Reconstructor
DopplerBroader

Endf6
Converter

Gnds
Converter

NuclearData
ObjectGndsParser/Writer

GNDS
format

ThermalScattering
DataProcessor

15

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

Input format of
FRENDY and NJOY
• Input parameters of

FRENDY consist of “input
data name” and “inputdata”• Comment line is similar to

C/C++• //~ or /* ~ */• Input parameters of

NJOY are hard tounderstand• This input format is sodifficult for beginners

16【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 tape for JENDL-4 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 7 20 500/ mat, temp no, sig no, bin no, lad no296.0/ temp
1E10 1E4 1E3 300 100 30 10 / sig zero0/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 for JENDL-4 U235' / descriptive character
9228 296.0/ mat, temp
1 1
/ newfor(yes), iopp(yes)1 1 1/ thin(1), thin(2), thin(3)
stop
/

17

Development schedule of FRENDY• FRENDY ver.1 will be released in the next spring• Generation of ACE file
• Generation of multi-group cross-section librarywill be implemented in the near future• Processing covariance data and calculation of KERMA

factor will also be implemented

18

Present status of nuclear data processing
code development• Development of nuclear data processing code isstarted in many institute• To process their own nuclear data library• To handle new nuclear data format GNDS
【Nuclear data processing codes development in the world】
CALENDF(CEA)GRUCON(Kurchatov)

Ruller(CIAE)FUDGE(LLNL)AMPX(ORNL)PREPRO (IAEA)GALILEE (CEA),GAIA (IRSN)NJOY(LANL)
NECP-Atlas(Xi’an Jiaotong Uni.)

FRENDY(JAEA)

Existing code
New code

Ref. D. Brown, “The New Evaluated Nuclear Data File Processing Capabilities,” INDC(NDS)-0695.

Close relationship withusers and evaluators
Special focus ondomestic utilization
including nuclear
regulators
Using latest
programming
technique
Treatment of new
nuclear data format
Ease in use forbeginners
NJOY compatible I/OContinuing update
and maintenance

Comparisonof nuclear
data
processing
code

△
○
○
×
×
×
○
△ 1.5

PREPRO

△
○
△
×
×
×
×
△

NJOY21

○
○
△
○
○
×
○
○ 2.5

FRENDY

○
○
○
○
○
○
○
○ 1.5

Human resources

V&V

Existing code
NJOY2016

New code

19

1

20

Comparison of processing results
between FRENDY and NJOY

21Comparison of processing results• Processing results of FRENDY are compared tothose of NJOY99.393 for verification• All nuclei in JENDL-3.3 and JENDL-4.0 are compared• We found several programming errors in NJOY

• Calculation conditions• Temperature: 296.0 K• Tolerance (error) : 0.01%

22Comparison of processing time
• The processing time to generate ACE files iscompared• Processing time of FRENDY is similar to that of NJOY
< Processing time [s] >• Adoption of the fixed
FRENDY NJOY
F/N
energy grid affects thecalculation time of the
TLS data• Cause of difference• Calculation method• Programming language• Adopting dynamic array

1H

0.1

0.2

0.5

16O

3.1

0.8

3.9

56Fe

18.7

9.1

2.1

235U

821.7

841.0

1.0

238U

507.5

709.1

0.7

239Pu

348.7

534.9

0.7

1H in H O
2

213.8

14.8

14.4

1H in ZrH

101.7

58.6

1.7

Graphite

116.9

9.5

12.3

*Intel Xeon CPU E7-8857 v2 (3.00GHz, turbo 3.60GHz)

23

Comparison of Doppler broadening• The processing results of FRENDY are similar to

those of NJOY99• The elastic scattering cross section shows thecharacteristics difference at the low energy region (lessthan 1.0×10-3 eV)• The calculation of the cross section at 0.0 eV is different• Other nuclei also show similar difference

XS [barn]

1E+0

+1%

1E+4
1E+2

1E-4

0%

1E+0

1E-8
1E-12

<238U, elastic scattering, 300 K>

1E-4 1E-2 1E+0 1E+2 1E+4 1E+6
Incident neutron energy [eV]

1E-2

1E-4 1E-2 1E+0 1E+21E+4 1E+6
Incident neutron energy [eV]-1%

(FRENDY-NJOY99)/NJOY99<238U, fission, 300 K>

24

Calculation of cross section at 0.0 eV• The cross section at 0.0 eV is required to calculate theDoppler broadened cross section at low energy region• NJOY approximates that the cross section follows the
1/v law• Since the elastic scattering cross section at the low energy

region is constant, this approximation is not appropriate• FRENDY uses linear extrapolation to calculate it

XS [barn]

appropriate for otherreaction types whichobey the 1/v law

1E+4<238U, radiation, 300 K>

+1%

1E+2
1E+0

0%

1E-2
1E-4

1E-4 1E-2 1E+0 1E+2 1E+4 1E+6
Incident neutron energy [eV]-1%

(FRENDY-NJOY99)/NJOY99• Linear extrapolation is

25

Difference of incoherent inelastic- Utilization of fixed energy grid -• NJOY only calculates the incoherent inelastic XS on 117

energy grids• Other energy grids are interpolated using the 5th order Lagrange

interpolation• The fixed energy grid is not appropriate for a material ofwhich the cross section is oscillated• This difference may have impact on the TRIGA reactor<Incoherent inelastic scattering XS (H in ZrH, 400 K)>
100

XS [barn]

1011.0E-5

1.0E-4

1.0E-3 1.0E-2 1.0E-1 1.0E+0
Incident neutron energy [eV]+1%
(FRENDY-NJOY99)/NJOY99

Difference at low
energy region isobserved in many
materials

0%

-1%

26

Verification of ACE file generating function• Comparison of keff values of ICSBEP benchmark• MCNP sample input files in ICSBEP handbook• 79 benchmark experiments, 752 critical configurations• Calculation results are not compared to the experimental results• Many of sample input files were not intended to be used for the strict
validation• All processes to generate the ACE file are processed by

FRENDY and NJOY99.393• The processing methods of FRENDY are similar to those of

NJOY• The programming errors in NJOY is also

implemented in FRENDY for the verification• Processing condition• Nuclear data library• Temperature• Tolerance (error)• Ladder number： JENDL-4.0： 296.0 K： 0.1 %: 100

27

Comparison for integral experiments• keff values of FRENDY are similar to those of NJOY99• Differences are not so varied with the neutron spectra andthe major fissile materials

FRENDY / NJOY99-1• FRENDY properly generates ACE files
0.04%
0.02%
0.00%
‐0.02%‐0.04%1σ

28Conclusions• Overview of nuclear data processing• Nuclear data processing code is not just a converter• It performs many processes to generate cross section
library• Overview of FRENDY• Utilization of modern programming techniques• Simple input format• Reflecting requests of nuclear data processing codeusers
• Comparison of the processing results• Processing results of FRENDY are compatible to thoseof NJOY99.393/2012.08

29

Release of FRENDY ver. 1• FRENDY Ver.1 will is released• From our web site or NEA Data-Bank• https://rpg.jaea.go.jp/main/en/• FRENDY Ver.1 is only generates ACE files• Generation of multi-group cross section library will be
implemented in the near future• FRENDY Ver.1 is open source software• 2-Clause BSD license• Manual of FRENDY Ver. 1 is published from JAEA
• JAEA-Data/Code 2018-014• https://jopss.jaea.go.jp/pdfdata/JAEA-Data-Code-2018-014.pdf• The input instructions and the details of processing method

are described