FRENDY Ver.2

Last Update: 2024/01/18

A nuclear data processing code
➢ Nuclear data processing➢ Development of FRENDY
Japan Atomic Energy Agency (JAEA) developed
the new nuclear data processing code FRENDY1, 2)(FRom Evaluated Nuclear Data librarY to anyapplication) to process evaluated nuclear data files.
Features of FRENDY are as follows:✓ Simple input format✓ Extensible & Modular✓ Open source software under the 2-clause
BSD license
FRENDY will help users who want to generatecross section libraries, process evaluated nuclear
data files for their own code, or modify evaluatednuclear data files or cross section libraries.

➢ How to get FRENDY?

Cross Section (b)Nuclear data processing is an important interface
between an evaluated nuclear data and nuclear
transport calculation codes as shown in Fig. 1. The
transport codes require a cross section library whichis generated by a nuclear data processing code.
The nuclear data processing code is not just aconverter. It performs many processes, e.g.,linearization, reconstruction of the resonance region,
Doppler broadening, etc.

Evaluated Nuclear Data(JENDL, ENDF, JEFF)

105

Nuclear Data Processing

Total XS of 235U (293.6 K)
104
103
102
101
100 10-4 10-2 100 102 104 106
Incident Neutron Energy (eV)

Nuclear Calculation
Cross Section
LibraryTransport Codes

Fig. 1 Calculation flow fromevaluated nuclear data tonuclear calculation

Everyone can download FRENDY from the following website:

https://rpg.jaea.go.jp/main/en/program_frendy/
Some introduction articles are also found in the above website.
Input instructions and overview of FRENDY are written in Ref. 2and the manual in the FRENDY package.

➢ References
1)
2)
3)
4)

K. Tada, et al., “Development and verification of nuclear data processing code FRENDY version 2,” J.
Nucl. Sci. Technol. (2023).
K. Tada, et al., “Nuclear Data Processing Code FRENDY Version 2,” JAEA-Data/Code 2022-009 (2023).
(https://doi.org/10.11484/jaea-data-code-2022-009)
K. Tada, et al., “Development of ACE file perturbation tool using FRENDY,” J. Nucl. Sci. Technol., 60,
pp.624-631 (2023).
A. Yamamoto, et al., "Multi-group neutron cross section generation capability for FRENDY nuclear dataprocessing code,“ J. Nucl. Sci. Technol., 58, pp.1165-1183 (2021).

Features of FRENDY➢ Simple input format
FRENDY does not require the expertise ofthe nuclear data processing as shown in Fig.
2. The other input parameters areautomatically set from the evaluated nuclear
data file. User can also specify theparameters in the input file if they want tochange them.
FRENDY has an alternative capability ofreading ordinary NJOY inputs without anymodifications.

➢ Extensible & ModularFRENDY is developed not only toprocess evaluated nuclear data files but
also to apply the modules to other codes.
FRENDY is written in C++ and all themodules are encapsulated. Users caneasily reuse many modules shown in Fig.
3 into their own codes. For example, aprogram for random sampling from
ACE files is realized by using thesemodules.3) This program is applied tothe uncertainty quantification using acontinuous energy Monte Carlo codeand now included in FRENDY.
There is no restriction to introduce theFRENDY modules into other codes
since this code is an open source
software under the 2-clause BSD license.

➢ Future plan

ace_fast_mode // Processing mode
nucl_file_name U235.dat
ace_file_name U235.ace
temp
296.0 //[K]Fig. 2 Sample input of FRENDY
ENDF-6format

GNDSformat

Endf6Parser/
Writer/Converter

GndsParser/
Writer/Converter

Resonance
Reconstructor

NuclearData
ObjectDoppler
Broader

AceData
Generator

Gas
Production
CrossSection
Calculator

AceData
Generator

Covariance
DataProcessor

Implemented
Future plan

AceDataParser/Writer

ThermalScattering
DataProcessor
Unresolved
Resonance
DataProcessor
HeatProduction
CrossSection
Calculator

ACEformatMGDataGenerator

MGformat

MGDataParser/WriterFig. 3 System structure of FRENDY

FRENDY version 2 generates the ACE files which are used for continuous energy Monte Carlocodes such as PHITS, MCNP, Serpent, and Open MC and the multi-group cross-section files(GENDF and MATXS) for deterministic code. As shown in Fig. 3, the covariance data processing
module (Covariance DataProcessor), the heating production cross-section calculation module(HeatProductionCrossSectionCalculator),andthetreatmentofGNDS
format(GndsParser/Writer/Converter) will be also implemented.

➢ Contact information
Dr. Kenichi Tada (tada.kenichi@jaea.go.jp)
Copyright 2024 Japan Atomic Energy Agency. All Rights Reserved.