Influence of nuclear data uncertainties on reactor core calculations

Kerntechnik ◽  
2011 ◽  
Vol 76 (3) ◽  
pp. 174-178 ◽  
Author(s):  
M. Klein ◽  
L. Gallner ◽  
B. Krzykacz-Hausmann ◽  
A. Pautz ◽  
W. Zwermann
Keyword(s):  
Reactor Core ◽  
Nuclear Data

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.

scholarly journals Impact of delayed neutron constants on reactivity effects measured in a fast reactor

2019 ◽  
Vol 211 ◽  
pp. 03004
Author(s):  
Antonín Krása ◽  
Anatoly Kochetkov ◽  
Nadia Messaoudi ◽  
Alexey Stankovskiy ◽  
Guido Vittiglio ◽  
...  
Keyword(s):  
Experimental Data ◽  
Reactor Core ◽  
Nuclear Data ◽  
Delayed Neutron ◽  
Neutron Data ◽  
Monte Carlo Calculations ◽  
Data Libraries ◽  
Experimental Data Analysis ◽  
The Impact ◽  
Nuclear Data Libraries

Delayed neutron parameters of fast VENUS-F reactor core configurations are determined with Monte Carlo calculations using various nuclear data libraries. Differences in the calculated effective delayed neutron fraction and the impact of the delayed neutron data (6- or 8-group precursors) that are applied in the experimental data analysis on the measured reactivity effects are studied. Considerable differences are found due to application of 235U and 238U delayed neutron data from JEFF, JENDL and ENDF evaluations.

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Massimo Sarotto ◽  
Gabriele Firpo ◽  
Anatoly Kochetkov ◽  
Antonin Krása ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Control Rod ◽  
The Core ◽  
Major Interest ◽  
Reactor Spectrum ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Sensitivity And Uncertainty Analysis

Abstract During the EURATOM FP7 project FREYA, a number of experiments were performed in a critical core assembled in the VENUS-F zero-power reactor able to reproduce the ALFRED lead-cooled fast reactor spectrum in a dedicated island. The experiments dealt with the measurements of integral and local neutronic parameters, such as the core criticality, the control rod and the lead void reactivity worth, the axial distributions of fission rates for the nuclides of major interest in a fast spectrum, the spectral indices of important actinides (238U, 239Pu, 237 Np) with respect to 235U. With the main aim to validate the neutronic codes adopted for the ALFRED core design, the VENUS-F core and its characterization measurements were simulated with both deterministic (ERANOS) and stochastic (MCNP, SERPENT) codes, by adopting different nuclear data libraries (JEFF, ENDF/B, JENDL, TENDL). This paper summarizes the main results obtained by highlighting a general agreement between measurements and simulations, with few discrepancies for some parameters that are discussed here. Additionally, a sensitivity and uncertainty analysis was performed with deterministic methods for the core reactivity: it clearly indicates that the small over-criticality estimated by the different codes/libraries resulted to be lower than the uncertainties due to nuclear data.

scholarly journals Integral data assimilation of the MERCI-1 experiment for the nuclear data associated with the PWR decay heat computation

2019 ◽  
Vol 211 ◽  
pp. 07004
Author(s):  
J. Huyghe ◽  
C. De Saint-Jean ◽  
D. Lecarpentier ◽  
C. Reynard-Carette ◽  
C. Vaglio-Gaudard ◽  
...  
Keyword(s):  
Reactor Core ◽  
Nuclear Data ◽  
Nuclear Decay ◽  
Crucial Issue ◽  
Decay Heat ◽  
Uncertainty Calculation ◽  
Experimental Values ◽  
The Impact ◽  
Fission Yields ◽  
Integral Data

Nuclear decay heat is a crucial issue for PWR in-core safety after reactor shutdown and back-end cycle. It is a dimensioning parameter for safety injection systems (SIS) to avoid a dewatering of the reactor core. The decay heat uncertainty needs to be controlled over the largest range of applications. The assimilation of the MERCI-1 experiment was studied to provide feedbacks on nuclear data. This experiment consisted in the measurement of the decay heat of a PWR UOX fuel sample irradiated in the OSIRIS reactor, for cooling times between 45 minutes and 42 days. More specifically, the consideration of several experimental values of MERCI-1 at different cooling times was tested. This raised issues about correlations to consider between different measurements. Besides, the impact of considering correlations between independent fission yields in covariance matrices on the decay heat uncertainty calculation and on the feedbacks on nuclear data is discussed.

scholarly journals Methods and Models for the Coupled Neutronics and Thermal-Hydraulics Analysis of the CROCUS Reactor at EFPL

2015 ◽  
Vol 2015 ◽  
pp. 1-9 ◽  
Author(s):  
A. Rais ◽  
D. Siefman ◽  
G. Girardin ◽  
M. Hursin ◽  
A. Pautz
Keyword(s):  
Power Distribution ◽  
Reactor Core ◽  
Transient Behavior ◽  
Nuclear Data ◽  
Safety Analysis ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Control Rod ◽  
Nuclear Data Library ◽  
And Control

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

scholarly journals Neutronic Calculations for Certain Americium Mixed Fuels and Clads in a Boiling Water Reactor

2021 ◽  
Vol 9 ◽  
Author(s):  
Mehtap Düz
Keyword(s):  
Monte Carlo ◽  
Spent Nuclear Fuel ◽  
Reactor Core ◽  
Nuclear Data ◽  
Boiling Water ◽  
Square Lattice ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Constant Pitch ◽  
Nuclear Data Library

In this study, a Boiling Water Reactor (BWR) design was made using the Monte Carlo (MCNPX) method. The reactor core in the designed BWR system was divided into an 8 × 8 square lattice with a constant pitch of 30.48 cm. In this study, americium (Am), which is found in the minor actinidine (MA) of spent nuclear fuel known as nuclear waste from existing reactors, was used as fuel with the addition of oxygen and fluorine. In this study, AmO2 and AmF3 fuels at the rate of 0.02–0.1% were used as Americium Mixed Fuels, and Zircaloy-2 (Zr-2), SiC, and VC were used as clad. Neutronic calculations for certain Americium Mixed Fuels and clads were compared in the designed BWR system. In the BWR system designed in the study; keff, neutron flux, fission energy, heating, and depleted Am were calculated. The three-dimensional (3-D) modeling of the designed BWR system was performed by using MCNPX-2.7.0 Monte Carlo method and the ENDF/B-VII.0 nuclear data library.

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