Determination of the uncertainties associated to the use of different nuclear data libraries in the analysis of extended-range rem-meters

2021 ◽  
pp. 110012
Author(s):  
Roberto García-Baonza ◽  
Gonzalo F. García-Fernández ◽  
Eduardo Gallego ◽  
Lenin E. Cevallos-Robalino ◽  
Óscar Cabellos
Keyword(s):  
Nuclear Data ◽  
Extended Range ◽  
Data Libraries ◽  
Nuclear Data Libraries

scholarly journals A study on the use of the reactor basic experiments in the U-D2O lattices of the RB critical assembly for validation of modern nuclear data libraries

2016 ◽  
Vol 31 (1) ◽  
pp. 1-15
Author(s):  
Milan Pesic
Keyword(s):  
Critical Level ◽  
Nuclear Data ◽  
Natural Uranium ◽  
Critical Assembly ◽  
Metal Fuel ◽  
Computer Codes ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Basic Experiments

Demand on the availability of well-defined reactor experiments for validation of computer codes for use in nuclear industry and nuclear technology is everlasting. Users must be confident of the results obtained by the proven computer codes and nuclear data libraries chosen in the models. The well-defined (mostly historical) and evaluated reactor experiments (about 5000 in 2015) were collected continuously as the benchmarks within the frame of the OECD/NEA international projects ICSBEP (since 1995) and IRPhEP (since 2003). The Handbooks of the Projects are published in electronic forms (at the NEA web site of the OECD and at a DVD media) every year. This study is aimed to (a) examine and evaluate reactor basic experiments, carried out in the lattice of the natural uranium metal fuel in the heavy water of the RB critical assembly first core in 1958, and (b) demonstrate their possibility for validation of modern nuclear data libraries. These RB reactor basic experiments include: (1) approach to criticality, (2) determination of the reactivity gradient at the D2O critical level, (3) measurement of the dependence of the D2O critical level on the D2O temperature, i. e. dependence of the reactivity with change in the D2O temperature; (4) the critical reactor geometrical parameter (buckling) measurements, (5) the migration length measurements, (6) determination of the neutron multiplication factor in the infinite lattice, and (7) the safety rods reactivity measurements. Results of the experiments are compared to the results obtained using modern nuclear data libraries of the ACE type by applying the MCNP6.1, a well-known and proven computer code based on the Monte Carlo method. A short overview of these experiments (done at the RB assembly) is shown. A brief description of the neutron ACE type nuclear data libraries (created in the LANL, based on the ENDF/B-VII.0 and ENDF/B-VII.1 files, or created in the OECD/NEA, based on the JEFF-3.2 evaluated nuclear data files), used in this validation study, is given. The benchmark models used for this validation study are described and the obtained results were analyzed. It is concluded that most of these reactor basic experiments, carried out in the lattice of the natural uranium metal fuel rods and the heavy water of the RB critical assembly, can be used as the benchmarks for validation of new nuclear data libraries. It may be done after further evaluations of influence of missing data, information and uncertainties in the material composition and geometry dimensions have been prepared according to the IRPhEP criteria and standards.

Criticality Calculations and Basic Sensitivity/Uncertainty Investigation of LR-0 Benchmark Core

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Tomáš Czakoj ◽  
Evžen Losa
Keyword(s):  
Lithium Fluoride ◽  
Three Dimensional ◽  
Reactor Core ◽  
Nuclear Data ◽  
Monte Carlo Code ◽  
Multiplication Factor ◽  
Reference Case ◽  
Energy Version ◽  
Data Libraries ◽  
Nuclear Data Libraries

Three-dimensional Monte Carlo code KENO-VI of SCALE-6.2.2 code system was applied for criticality calculation of the LR-0 reactor core. A central module placed in the center of the core was filled by graphite, lithium fluoride-beryllium fluoride (FLIBE), and lithium fluoride-sodium fluoride (FLINA) compounds. The multiplication factor was obtained for all cases using both ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. Obtained results were compared with benchmark calculations in the MCNP6 using ENDF/B-VII.0 library. The results of KENO-VI calculations are found to be in good agreement with results obtained by the MCNP6. The discrepancies are typically within tens of pcm excluding the case with the FLINA filling. Sensitivities and uncertainties of the reference case with no filling were determined by a continuos-energy version of the TSUNAMI sequence of SCALE-6.2.2. The obtained uncertainty in multiplication factor due to the uncertainties in nuclear data is about 650 pcm with ENDF/B-VII.1.

Sensitivity to Nuclear Data Libraries in the Physics Core Characteristics of Conceptual Thorium Breeders

2011 ◽  
Vol 59 (2(3)) ◽  
pp. 1361-1364
Author(s):  
V. Jagannathan ◽  
U. Pal ◽  
R. Karthikeyan ◽  
A. Srivastava ◽  
S. A. Khan
Keyword(s):  
Nuclear Data ◽  
Data Libraries ◽  
Core Characteristics ◽  
Nuclear Data Libraries

scholarly journals Generation of the 1H in H2O neutron thermal scattering law covariance matrix of the CAB model

2018 ◽  
Vol 4 ◽  
pp. 32
Author(s):  
Juan Pablo Scotta ◽  
Gilles Noguère ◽  
Jose Ignacio Marquez Damian
Keyword(s):  
Covariance Matrix ◽  
Nuclear Data ◽  
Water Model ◽  
Dynamic Simulations ◽  
Light Water ◽  
Thermal Scattering ◽  
Data Libraries ◽  
Hydrogen Bound ◽  
Nuclear Data Libraries ◽  
Evaluated Nuclear Data

The thermal scattering law (TSL) of 1H in H2O describes the interaction of the neutron with the hydrogen bound to light water. No recommended procedure exists for computing covariances of TSLs available in the international evaluated nuclear data libraries. This work presents an analytic methodology to produce such a covariance matrix-associated to the water model developed at the Atomic Center of Bariloche (Centro Atomico Bariloche, CAB, Argentina). This model is called as CAB model, it calculates the TSL of hydrogen bound to light water from molecular dynamic simulations. The performance of the obtained covariance matrix has been quantified on integral calculations at “cold” reactor conditions between 20 and 80∘ C. For UOX fuel, the uncertainty on the calculated reactivity ranges from ±71 to ±155 pcm. For MOX fuel, it ranges from ±110 to ±203 pcm.

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