Processing and benchmarking of evaluated nuclear data file/b-viii.0b4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016

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Processing and benchmarking of evaluated nuclear data file/b-viii.0b4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016. The results obtained with the ENDF/BVIII.0b4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/BVII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.
1 255
Contents lists available at ScienceDirect
Nuclear Engineering and Technology
journal homepage: www.elsevier.com/locate/net
Original Article
Processing and benchmarking of evaluated nuclear data le/b-viii.0b4
cross-section library by analysis of a series of critical experimental
benchmark using the monte carlo code MCNP(X) and NJOY2016
Kabach Ouadie*, Chetaine Abdelouahed, Jalil Abdelhamid, Darif Abdelaziz,
Saidi Abdelmajid
Nuclear Reactor and Nuclear Security Group Energy Center, Nuclear Physics Laboratory, Physics Department, Faculty of Science, Mohammed V University, 4
Avenue Ibn Battouta B.P. 1014 RP, Rabat 10000, Morocco
a r t i c l e
i n f o
a b s t r a c t
Article history:
To validate the new Evaluated Nuclear Data File (ENDF)/B-VIII.0b4 library, 31 different critical cores were
Received 26 May 2017
Received in revised form
8 August 2017
Accepted 24 August 2017
Available online 18 September 2017
selected and used for a benchmark test of the important parameter keff. The four utilized libraries are
processed using Nuclear Data Processing Code (NJOY2016). The results obtained with the ENDF/B-
VIII.0b4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-
VII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff
values with these four libraries were compared with the experimentally measured results, which are
Keywords:
Benchmark
Critical
Evaluated nuclear data File (ENDF)/B-VI.8
available in the International Critically Safety Benchmark Evaluation Project. The obtained results are
discussed and analyzed in this paper.
© 2017 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the
ENDF/B-VII.0
ENDF/B-VII.1
ENDF/B-VIII.0b4
MCNP(X)
NJOY2016
Validation
1. Introduction
company, U.S.A, Product of China) [i5 CPU 3.20 GHz, third genera-
tion (4 cores), 4 GB RAM, and 6 MB Cache Memory] under Win 10
The evaluated nuclear data library Evaluated Nuclear Data File
system and using MPICH2 [4]. Calculations for keff were performed
(ENDF)/B-VIII.0b4 was released on February 28, 2017 [1]. The
using these four libraries with 600 iterations on a nominal source
Neutron General Purpose Librarycontains incident neutron data for
446 isotopes from 1H to 100Fm, and the Scattering Thermal Library
Sða;bÞ covers 26 different moderators.
size of 60,000 particles per cycle, in order to decrease statistical
error estimates. Initial 100 cycles were skipped to ensure homo-
geneous neutron source distribution.
The code system used for processing and generating the ACE
format (a compact ENDF format for the MCNP) is Nuclear Data
2. Methodology
Processing Code (NJOY2016) [2] for all isotopes and Sða;bÞ thermal
scattering, especially the ACER module (the ACER module prepares
2.1. Nuclear data processing converted from evaluated nuclear data
libraries in ACE format), for correct processing of the libraries. A
to ACE format
locally executable version of NJOY2016 was created using 64 bits
under a Linux system.
NJOY is a modular computer code used for converting evaluated
As part of the validation process, comparisons with the ENDF/B-
nuclear data in the ENDF format into different types of libraries
VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries were carried out.
useful for the calculations of criticality and shielding applications.
In this study, Monte Carlo N-Particle (MCNP(X)) v2.6.0 [3] cal-
Since the ENDF format is used worldwide, NJOY gives its users ac-
culations were performed with an HP Pro Intel (Hewlett-Packard
cess to a wide variety of the most up-to-date nuclear data [2]. NJOY
provides comprehensive capabilities for processing evaluated data,
* Corresponding author.
serving applications that include continuous-energy Monte Carlo-
E-mail address: ouadie.kabach10@gmail.com (K. Ouadie).
like MCNP(X) [3].
1738-5733/© 2017 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/
K. Ouadie et al. / Nuclear Engineering and Technology 49 (2017) 1610e1616
1611
MODER
GASPR
THERMR
RECONR
HEATR
ACER
(create ACE file)
PURR
BROADR
ACER
(check ACE file)
Fig. 1. Flow diagram of NJOY2016 processing for ACE format library construction. ACE: A Compact ENDF format for the MCNP; ACER: The ACER module prepares libraries in ACE
format; BROADR, Doppler broadens and thins pointwise cross sections; GASPR, generates gas-production cross sections in pointwise format from basic reaction data in an ENDF
evaluation. These results can be converted to multigroup form using GROUPR, passed to ACER, or displayed using PLOTR; HEATR, generates pointwise heat production cross sections
(KERMA coefcients) and radiation-damage cross sections; MODER, converts ENDF tapesback and forth between ASCII format and the special NJOY blocked-binary format; NJOY,
Nuclear Data Processing Code developed at Los Alamos National Laboratory; PURR, generates unresolved-resonance probability tables for use in representing resonance self-
shielding effects in the MCNP Monte Carlo code; RECONR, reconstructs pointwise (energy-dependent) cross sections from ENDF resonance parameters and interpolation schemes;
THERMR, produces cross sections and energy-to-energy matrices for free or bound scatterers in the thermal energy range.
The ENDF/B-VIII.0b4, ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-
(2) RECONR reconstructs pointwise (energy-dependent) cross
VII.1 libraries contain continuous-energy neutron cross sections;
sections from ENDF resonance parameters and interpolation
resonance parameters, angular data, and other data for all the
schemes. The cross-section accuracy in this module is of the
isotopes composing the materials of studied benchmarks havebeen
order of 0.5% (err ¼ 0.005).
processed using NJOY2016 in ACE format (compatible with the
(3) PURR generates unresolved-resonance probability tables for
Monte Carlo code MCNP) at a temperature of 300 K. In addition,
use in representing resonance self-shielding effects in the
probability tables (ptable) have been generated for benchmark
MCNP Monte Carlo code.
materials containing unresolved resonance data. In addition, Sða;bÞ
(4) GASPR generates gas-production cross sections in pointwise
thermal scattering has been processed at 300 K [5] for hydrogen
format from basic reaction data in an ENDF evaluation. These
bound in polyethylene (poly.10t), graphite (grph.10t), H in H2O
results can be converted to multigroup form using GROUPR;
(lwtr.10t), and beryllium (Be) metal (be.10t).
they are then passed to ACER or displayed using PLOTR.
The processing sequence for generating an ACE-formatted li-
(5) HEATR generates pointwise heat production cross sections
brary suitable for the MCNP code is shown in Fig. 1.
(kinetic energy released per unit mass (KERMA) coefcients)
These modules are explained below. The reason for running the
and radiation-damage cross sections.
ACER module twice (Fig. 1) is that it is used to evaluate the con-
(6) BROADR Doppler broadens and thins pointwise cross sec-
sistency of the ACE format, which is checked and for which prob-
tions. The cross-section accuracy in this module is of the
lems are corrected [2].
order of 0.5% (err ¼ 0.005).
(1) MODER converts ENDF tapesback and forth between ASCII
format and the special NJOY blocked-binary format.
Table 2
ICSBEP benchmark systems.
Table 1
ICSBEP benchmark systems
ICSBEP abbreviations used [8].
HEU-MET-FAST-001
IEU-MET-FAST-004-case-2
Abbreviation
meaning
HEU-MET-FAST-004
HEU-MET-FAST-008
LEU-SOL-THERM-001
LEU-SOL-THERM-002-case-1
Fissile material
HEU-MET-FAST-009-case-1
MIX-MET-FAST-001
HEU
High enriched uranium
HEU-MET-FAST-011
MIX-MET-FAST-003
(235U  60 wt%)
HEU-MET-FAST-015
MIX-COMP-THERM-002-case-pnl30
IEU
Intermediate or mixed enrichment uranium
HEU-MET-FAST-018
PU-MET-FAST-001
(60 wt% > 235U > 10 wt%)
HEU-MET-FAST-019-case-2
PU-MET-FAST-002
LEU
Low enriched, natural, or depleted uranium
HEU-MET-FAST-020-case-2
PU-MET-FAST-005
(235U  10 wt%)
HEU-MET-INTER-006-case-1
PU-MET-FAST-006
PU
Plutonium
HEU-SOL-THERM-013-case-1
PU-MET-FAST-025
MIX
Mixed uranium and plutonium
HEU-SOL-THERM-032
PU-MET-FAST-026
U233
Uranium 233U systems
IEU-MET-FAST-001-case-1
U233-MET-FAST-002-case-1
Physical form of ssile material
IEU-MET-FAST-002
U233-SOL-INTER-001-case-1
MET
Metal
IEU-MET-FAST-003-case-2
U233-SOL-THERM-001-case-1
SOL
Solution
U233-SOL-THERM-008
COMP
Spectrum
FAST
THERM
Compound system, e.g., lattice in water
Fast system (50% of ssions above 100 keV)
Thermal system (50% of ssions below 0.625 eV)
COMP, compound system; FAST, fast system; HEU, high enriched uranium; ICSBEP,
International Critically Safety Benchmark Evaluation Project; IEU, intermediate
enriched uranium; INTER, intermediate; LEU, low enriched uranium; MET, metal;
MIX, mixed uranium and plutonium system; PU, plutonium; SOL, solution; THERM,
ICSBEP, International Critically Safety Benchmark Evaluation Project.
thermal.
1612
K. Ouadie et al. / Nuclear Engineering and Technology 49 (2017) 1610e1616
Table 3
MCNP(X) calculations of keff values with four data libraries and benchmark keff.
Case name
Benchmark keff
ENDF/B-VI.8
ENDF/B-VII.0
ENDF/B-VII.1
ENDF/B-VIII.0b4
HEU-MET-
1.00000 ± 0.00100
0.99656 ± 0.00010
0.99985 ± 0.00010
0.99978 ± 0.00010
0.99994 ± 0.00010
FAST-001
HEU-MET-
1.00200 ± 0.00100
0.99833 ± 0.00010
1.00302 ± 0.00010
1.00305 ± 0.00010
1.00250 ± 0.00010
FAST-004-
case-1
HEU-MET-
0.99890 ± 0.00160
0.99240 ± 0.00010
0.99577 ± 0.00010
0.99573 ± 0.00010
0.99562 ± 0.00010
FAST-008
HEU-MET-
0.99920 ± 0.00150
0.99484 ± 0.00010
0.99507 ± 0.00010
0.99749 ± 0.00010
0.99615 ± 0.00010
FAST-009-
case-1
HEU-MET-
0.99890 ± 0.00150
1.00050 ± 0.00010
0.99881 ± 0.00010
1.00437 ± 0.00010
1.00489 ± 0.00010
FAST-011
HEU-MET-
0.99960 ± 0.00170
0.99146 ± 0.00010
0.99491 ± 0.00010
0.99466 ± 0.00010
0.99478 ± 0.00010
FAST-015
HEU-MET-
1.00000 ± 0.00140
0.99599 ± 0.00010
0.99971 ± 0.00010
0.99959 ± 0.00010
0.99936 ± 0.00010
FAST-018-
case-2
HEU-MET-
1.00000 ± 0.00280
1.00310 ± 0.00010
1.00746 ± 0.00010
1.00713 ± 0.00010
1.00658 ± 0.00010
FAST-019-
case-2
HEU-MET-
1.00000 ± 0.00280
0.99677 ± 0.00010
1.00087 ± 0.00010
1.00078 ± 0.00010
1.00031 ± 0.00010
FAST-020-
case-2
HEU-MET-
0.99770 ± 0.00080
0.98567 ± 0.00010
0.99286 ± 0.00010
0.98734 ± 0.00010
0.98805 ± 0.00010
INTER-
006-case-1
HEU-SOL-
1.00120 ± 0.00260
0.99930 ± 0.00010
0.99872 ± 0.00010
0.99872 ± 0.00010
0.99844 ± 0.00010
THERM-
013-case-1
HEU-SOL-
1.00150 ± 0.00260
0.99880 ± 0.00010
0.99956 ± 0.00010
0.99951 ± 0.00010
0.99850 ± 0.00010
THERM-
032
IEU-MET-
0.99890 ± 0.00100
0.99645 ± 0.00010
1.00087 ± 0.00010
1.00073 ± 0.00010
0.99901 ± 0.00010
FAST-001-
case-1
IEU-MET-
1.00000 ± 0.00300
1.00305 ± 0.00010
0.99920 ± 0.00010
0.99876 ± 0.00010
0.99613 ± 0.00010
FAST-002
IEU-MET-
1.00000 ± 0.00170
0.99902 ± 0.00010
1.00251 ± 0.00010
1.00277 ± 0.00010
0.99968 ± 0.00010
FAST-003-
case-2
IEU-MET-
1.00000 ± 0.00300
1.00362 ± 0.00010
1.00747 ± 0.0001 0
1.00751 ± 0.00010
1.00533 ± 0.00010
FAST-004-
case-2
LEU-SOL-
0.99910 ± 0.00290
1.01004 ± 0.00010
1.01201 ± 0.00010
1.01172 ± 0.00010
1.01164 ± 0.00010
THERM-
001
LEU-SOL-
1.00380 ± 0.00400
0.99838 ± 0.00010
0.99993 ± 0.00010
0.99992 ± 0.00010
0.99605 ± 0.00010
THERM-
002-case-1
IX-MET-
1.00000 ± 0.00160
0.99698 ± 0.00010
0.99947 ± 0.00010
0.99952 ± 0.00010
0.99951 ± 0.00010
FAST-001
MIX-MET-
0.99930 ± 0.00160
0.99833 ± 0.00010
1.00087 ± 0.00010
1.00067 ± 0.00010
1.00073 ± 0.00010
FAST-003
MIX-COMP-
1.00240 ± 0.00600
0.99207 ± 0.00010
1.00117 ± 0.00010
1.00049 ± 0.00010
0.99957 ± 0.00010
THERM-
002-case-
pnl30
PU-MET-
1.00000 ± 0.00200
0.99752 ± 0.00010
0.99985 ± 0.00010
0.99987 ± 0.00010
0.99974 ± 0.00010
FAST-001
PU-MET-
1.00000 ± 0.00200
0.99792 ± 0.00010
1.00004 ± 0.00010
0.99981 ± 0.00010
1.00149 ± 0.00010
FAST-002
PU-MET-
1.00000 ± 0.00130
1.00753 ± 0.00010
1.00943 ± 0.00010
1.00085 ± 0.00010
0.99940 ± 0.00010
FAST-005
PU-MET-
1.00000 ± 0.00300
1.00275 ± 0.00010
1.00123 ± 0.00010
1.00110 ± 0.00010
0.99970 ± 0.00010
FAST-006
PU-MET-
1.00000 ± 0.00200
0.99649 ± 0.00010
0.99868 ± 0.00010
0.99867 ± 0.00010
0.99967 ± 0.00010
FAST-025-
case-2
PU-MET-
1.00000 ± 0.00240
0.99700 ± 0.00010
0.99866 ± 0.00010
0.99844 ± 0.00010
1.00083 ± 0.00010
FAST-026-
case-2
U233-MET-
1.00000 ± 0.00100
0.99530 ± 0.00010
0.99897 ± 0.00010
0.99933 ± 0.00010
1.00003 ± 0.00010
FAST-002-
case-1
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Processing and benchmarking of evaluated nuclear data file/b-viii.0b4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016. The results obtained with the ENDF/BVIII.0b4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/BVII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper..

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Nuclear Engineering and Technology 49 (2017) 1610e1616 Contents lists available at ScienceDirect Nuclear Engineering and Technology journal homepage: www.elsevier.com/locate/net Original Article Processing and benchmarking of evaluated nuclear data file/b-viii.0b4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016 Kabach Ouadie*, Chetaine Abdelouahed, Jalil Abdelhamid, Darif Abdelaziz, Saidi Abdelmajid Nuclear Reactor and Nuclear Security Group Energy Center, Nuclear Physics Laboratory, Physics Department, Faculty of Science, Mohammed V University, 4 Avenue Ibn Battouta B.P. 1014 RP, Rabat 10000, Morocco a r t i c l e i n f o a b s t r a c t Article history: Received 26 May 2017 Received in revised form 8 August 2017 Accepted 24 August 2017 Available online 18 September 2017 Keywords: Benchmark Critical Evaluated nuclear data File (ENDF)/B-VI.8 ENDF/B-VII.0 ENDF/B-VII.1 ENDF/B-VIII.0b4 MCNP(X) NJOY2016 Validation 1. Introduction To validate the new Evaluated Nuclear Data File (ENDF)/B-VIII.0b4 library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016). The results obtained with the ENDF/B-VIII.0b4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X)) code. All the MCNP(X) calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper. © 2017 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/). company, U.S.A, Product of China) [i5 CPU 3.20 GHz, third genera- tion (4 cores), 4 GB RAM, and 6 MB Cache Memory] under Win 10 The evaluated nuclear data library Evaluated Nuclear Data File (ENDF)/B-VIII.0b4 was released on February 28, 2017 [1]. The Neutron General Purpose Librarycontains incident neutron data for 446 isotopes from 1H to 100Fm, and the Scattering Thermal Library Sða;bÞ covers 26 different moderators. The code system used for processing and generating the ACE format (a compact ENDF format for the MCNP) is Nuclear Data Processing Code (NJOY2016) [2] for all isotopes and Sða;bÞ thermal scattering, especially the ACER module (the ACER module prepares libraries in ACE format), for correct processing of the libraries. A locally executable version of NJOY2016 was created using 64 bits under a Linux system. As part of the validation process, comparisons with the ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries were carried out. In this study, Monte Carlo N-Particle (MCNP(X)) v2.6.0 [3] cal-culations were performed with an HP Pro Intel (Hewlett-Packard * Corresponding author. E-mail address: ouadie.kabach10@gmail.com (K. Ouadie). system and using MPICH2 [4]. Calculations for keff were performed using these four libraries with 600 iterations on a nominal source size of 60,000 particles per cycle, in order to decrease statistical error estimates. Initial 100 cycles were skipped to ensure homo-geneous neutron source distribution. 2. Methodology 2.1. Nuclear data processing converted from evaluated nuclear data to ACE format NJOY is a modular computer code used for converting evaluated nuclear data in the ENDF format into different types of libraries useful for the calculations of criticality and shielding applications. Since the ENDF format is used worldwide, NJOY gives its users ac-cess to a wide variety of the most up-to-date nuclear data [2]. NJOY provides comprehensive capabilities for processing evaluated data, serving applications that include continuous-energy Monte Carlo- like MCNP(X) [3]. https://doi.org/10.1016/j.net.2017.08.017 1738-5733/© 2017 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/ licenses/by-nc-nd/4.0/). K. Ouadie et al. / Nuclear Engineering and Technology 49 (2017) 1610e1616 1611 MODER GASPR RECONR HEATR PURR BROADR THERMR ACER (create ACE file) ACER (check ACE file) Fig. 1. Flow diagram of NJOY2016 processing for ACE format library construction. ACE: A Compact ENDF format for the MCNP; ACER: The ACER module prepares libraries in ACE format; BROADR, Doppler broadens and thins pointwise cross sections; GASPR, generates gas-production cross sections in pointwise format from basic reaction data in an ENDF evaluation. These results can be converted to multigroup form using GROUPR, passed to ACER, or displayed using PLOTR; HEATR, generates pointwise heat production cross sections (KERMA coefficients) and radiation-damage cross sections; MODER, converts ENDF “tapes” back and forth between ASCII format and the special NJOY blocked-binary format; NJOY, Nuclear Data Processing Code developed at Los Alamos National Laboratory; PURR, generates unresolved-resonance probability tables for use in representing resonance self-shielding effects in the MCNP Monte Carlo code; RECONR, reconstructs pointwise (energy-dependent) cross sections from ENDF resonance parameters and interpolation schemes; THERMR, produces cross sections and energy-to-energy matrices for free or bound scatterers in the thermal energy range. The ENDF/B-VIII.0b4, ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries contain continuous-energy neutron cross sections; resonance parameters, angular data, and other data for all the isotopes composing the materials of studied benchmarks havebeen processed using NJOY2016 in ACE format (compatible with the Monte Carlo code MCNP) at a temperature of 300 K. In addition, probability tables (ptable) have been generated for benchmark materials containing unresolved resonance data. In addition, Sða;bÞ thermal scattering has been processed at 300 K [5] for hydrogen bound in polyethylene (poly.10t), graphite (grph.10t), H in H2O (lwtr.10t), and beryllium (Be) metal (be.10t). The processing sequence for generating an ACE-formatted li-brary suitable for the MCNP code is shown in Fig. 1. These modules are explained below. The reason for running the ACER module twice (Fig. 1) is that it is used to evaluate the con-sistency of the ACE format, which is checked and for which prob- lems are corrected [2]. (2) RECONR reconstructs pointwise (energy-dependent) cross sections from ENDF resonance parameters and interpolation schemes. The cross-section accuracy in this module is of the order of 0.5% (err ¼ 0.005). (3) PURR generates unresolved-resonance probability tables for use in representing resonance self-shielding effects in the MCNP Monte Carlo code. (4) GASPR generates gas-production cross sections in pointwise format from basic reaction data in an ENDF evaluation. These results can be converted to multigroup form using GROUPR; they are then passed to ACER or displayed using PLOTR. (5) HEATR generates pointwise heat production cross sections (kinetic energy released per unit mass (KERMA) coefficients) and radiation-damage cross sections. (6) BROADR Doppler broadens and thins pointwise cross sec-tions. The cross-section accuracy in this module is of the order of 0.5% (err ¼ 0.005). (1) MODER converts ENDF “tapes” back and forth between ASCII format and the special NJOY blocked-binary format. Table 1 ICSBEP abbreviations used [8]. Table 2 ICSBEP benchmark systems. ICSBEP benchmark systems HEU-MET-FAST-001 IEU-MET-FAST-004-case-2 Abbreviation Fissile material HEU IEU LEU PU MIX U233 meaning High enriched uranium (235U 60 wt%) Intermediate or mixed enrichment uranium (60 wt% > 235U > 10 wt%) Low enriched, natural, or depleted uranium (235U 10 wt%) Plutonium Mixed uranium and plutonium Uranium 233U systems HEU-MET-FAST-004 HEU-MET-FAST-008 HEU-MET-FAST-009-case-1 HEU-MET-FAST-011 HEU-MET-FAST-015 HEU-MET-FAST-018 HEU-MET-FAST-019-case-2 HEU-MET-FAST-020-case-2 HEU-MET-INTER-006-case-1 HEU-SOL-THERM-013-case-1 HEU-SOL-THERM-032 IEU-MET-FAST-001-case-1 LEU-SOL-THERM-001 LEU-SOL-THERM-002-case-1 MIX-MET-FAST-001 MIX-MET-FAST-003 MIX-COMP-THERM-002-case-pnl30 PU-MET-FAST-001 PU-MET-FAST-002 PU-MET-FAST-005 PU-MET-FAST-006 PU-MET-FAST-025 PU-MET-FAST-026 U233-MET-FAST-002-case-1 Physical form of fissile material IEU-MET-FAST-002 U233-SOL-INTER-001-case-1 MET Metal SOL Solution IEU-MET-FAST-003-case-2 U233-SOL-THERM-001-case-1 U233-SOL-THERM-008 COMP Spectrum FAST THERM Compound system, e.g., lattice in water Fast system (50% of fissions above 100 keV) Thermal system (50% of fissions below 0.625 eV) COMP, compound system; FAST, fast system; HEU, high enriched uranium; ICSBEP, International Critically Safety Benchmark Evaluation Project; IEU, intermediate enriched uranium; INTER, intermediate; LEU, low enriched uranium; MET, metal; MIX, mixed uranium and plutonium system; PU, plutonium; SOL, solution; THERM, ICSBEP, International Critically Safety Benchmark Evaluation Project. thermal. 1612 K. Ouadie et al. / Nuclear Engineering and Technology 49 (2017) 1610e1616 Table 3 MCNP(X) calculations of keff values with four data libraries and benchmark keff. Case name HEU-MET-FAST-001 HEU-MET-FAST-004-case-1 HEU-MET-FAST-008 HEU-MET-FAST-009-case-1 HEU-MET-FAST-011 HEU-MET-FAST-015 HEU-MET-FAST-018-case-2 HEU-MET-FAST-019-case-2 HEU-MET-FAST-020-case-2 HEU-MET-INTER- 006-case-1 HEU-SOL- THERM-013-case-1 HEU-SOL-THERM-032 IEU-MET-FAST-001-case-1 IEU-MET-FAST-002 IEU-MET-FAST-003-case-2 IEU-MET-FAST-004-case-2 LEU-SOL-THERM-001 LEU-SOL-THERM-002-case-1 IX-MET-FAST-001 MIX-MET-FAST-003 MIX-COMP-THERM-002-case-pnl30 PU-MET-FAST-001 PU-MET-FAST-002 PU-MET-FAST-005 PU-MET-FAST-006 PU-MET-FAST-025-case-2 PU-MET-FAST-026-case-2 U233-MET-FAST-002- case-1 Benchmark keff 1.00000 ± 0.00100 1.00200 ± 0.00100 0.99890 ± 0.00160 0.99920 ± 0.00150 0.99890 ± 0.00150 0.99960 ± 0.00170 1.00000 ± 0.00140 1.00000 ± 0.00280 1.00000 ± 0.00280 0.99770 ± 0.00080 1.00120 ± 0.00260 1.00150 ± 0.00260 0.99890 ± 0.00100 1.00000 ± 0.00300 1.00000 ± 0.00170 1.00000 ± 0.00300 0.99910 ± 0.00290 1.00380 ± 0.00400 1.00000 ± 0.00160 0.99930 ± 0.00160 1.00240 ± 0.00600 1.00000 ± 0.00200 1.00000 ± 0.00200 1.00000 ± 0.00130 1.00000 ± 0.00300 1.00000 ± 0.00200 1.00000 ± 0.00240 1.00000 ± 0.00100 ENDF/B-VI.8 0.99656 ± 0.00010 0.99833 ± 0.00010 0.99240 ± 0.00010 0.99484 ± 0.00010 1.00050 ± 0.00010 0.99146 ± 0.00010 0.99599 ± 0.00010 1.00310 ± 0.00010 0.99677 ± 0.00010 0.98567 ± 0.00010

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