scholarly journals On the importance of target accuracy assessments and data assimilation for the co-development of nuclear data and fast reactors: MYRRHA and ESFR

2021 ◽  
Vol 161 ◽  
pp. 108416
Author(s):  
P. Romojaro ◽  
F. Álvarez-Velarde ◽  
O. Cabellos ◽  
N. García-Herranz ◽  
A. Jiménez-Carrascosa
Keyword(s):  
Data Assimilation ◽  
Nuclear Data ◽  
Fast Reactors ◽  
Accuracy Assessments

The Use of Nuclear Data as Nuisance Parameters in the Integral Data Assimilation of the PROFIL Experiments

2016 ◽  
Vol 182 (3) ◽  
pp. 377-393 ◽  
Author(s):  
E. Privas ◽  
P. Archier ◽  
C. De Saint Jean ◽  
G. Noguere ◽  
J. Tommasi
Keyword(s):  
Data Assimilation ◽  
Nuclear Data ◽  
Nuisance Parameters ◽  
Integral Data

scholarly journals Impact of nuclear data evaluations on data assimilation for an LFR

2020 ◽  
Vol 239 ◽  
pp. 13007
Author(s):  
Pablo Romojaro ◽  
Francisco Álvarez-Velarde
Keyword(s):  
Data Assimilation ◽  
Fast Reactor ◽  
Nuclear Reactor ◽  
Critical Mass ◽  
Nuclear Data ◽  
Operational Experience ◽  
Operation Conditions ◽  
Nuclear Data Library ◽  
Safety Parameters ◽  
The Impact

The Lead-cooled Fast Reactor is one of the three technologies selected by the Sustainable Nuclear Energy Technology Platform that can meet future European energy needs. The main drawbacks for the industrial deployment of LFR are the lack of operational experience and the impact of uncertainties. In nuclear reactor design the uncertainties mainly come from material properties, fabrication tolerances, operation conditions, simulation tools and nuclear data. The uncertainty in nuclear data is one of the most important sources of uncertainty in reactor physics simulations. Furthermore, it is known that the uncertainties in reactor criti-cality safety parameters are severely dependent on the nuclear data library used to estimate them. However, the impact of using different evaluations while performing data assimilation to constraint the uncertainties in the criticality parameters has not been properly assessed yet. In this work, a data assimilation for the main isotopes contributing to the uncertainty in keff of the ALFRED lead-cooled fast reactor has been performed with the SUMMON system using JEFF-3.3, ENDF/B-VIII.0 and JENDL-4.0u2 state-of-the-art nuclear data libraries, together with critical mass experiments from the International Criticality Safety Benchmark Evaluation Project that are representative of ALFRED, in order to assess the impact of using different evaluations for data assimilation.

scholarly journals Trends on major actinides from an integral data assimilation

2020 ◽  
Vol 239 ◽  
pp. 13002
Author(s):  
Gerald Rimpault ◽  
Gilles Noguère ◽  
Cyrille de Saint Jean
Keyword(s):  
Data Assimilation ◽  
Energy Range ◽  
Cross Section ◽  
Capture Cross Section ◽  
Critical Mass ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Capture Cross ◽  
Plateau Region ◽  
Integral Data

The objective of this work is to revisit integral data assimilation for a better prediction of the characteristics of SFR cores. ICSBEP, IRPhE and MASURCA critical masses, PROFIL irradiation experiments and the FCA-IX experimental programme (critical masses and spectral indices) with well-mastered experimental technique have been used. As calculations are performed without modelling errors (with as-built geometries) and without approximations with the TRIPOLI4 MC code, highly reliable C/E are achieved. Assimilation results suggest a 2.5% decrease for 238U capture from 3 keV to 60 keV, and a 4-5% decrease for 238U inelastic in the plateau region. For this energy range, uncertainties are respectively reduced to 1-2% and to 2-2.5% for 238U capture and 238U inelastic respectively. The increase trends on 239Pu capture cross section of around 3% in the [2 keV-100 keV] energy range come from a low PROFIL 240Pu/239Pu ratio C/E. For 240Pu capture cross section, the increase trend of around 4% in the [3 keV-100 keV] energy range goes in the same direction as the recent ENDF/B.VIII evaluation though at a much lower level. The nuclear data uncertainty associated to SFR ASTRID critical mass is reduced to 470 pcm.

Uncertainty Assessment for Fast Reactors Based on Nuclear Data Adjustment

2014 ◽  
Vol 118 ◽  
pp. 592-595 ◽  
Author(s):  
T. Ivanova ◽  
E. Ivanov ◽  
F. Ecrabet
Keyword(s):  
Nuclear Data ◽  
Uncertainty Assessment ◽  
Fast Reactors ◽  
Data Adjustment

scholarly journals Data assimilation of post-irradiation examination data for fission yields from GEF

2020 ◽  
Vol 6 ◽  
pp. 52
Author(s):  
Daniel Siefman ◽  
Mathieu Hursin ◽  
Henrik Sjostrand ◽  
Georg Schnabel ◽  
Dimitri Rochman ◽  
...  
Keyword(s):  
Data Assimilation ◽  
Marginal Likelihood ◽  
Fission Products ◽  
Nuclear Data ◽  
Optimization Method ◽  
Model Parameters ◽  
Spent Fuel ◽  
Post Irradiation ◽  
Relative Standard ◽  
Fission Yields

Nuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel which can exceed engineering target accuracies. Herein, we present a new framework that extends data assimilation methods to burnup simulations by using post-irradiation examination experiments. The adjusted fission yields lowered the bias and reduced the uncertainty of the simulations. Our approach adjusts the model parameters of the code GEF. We compare the BFMC and MOCABA approaches to data assimilation, focusing especially on the effects of the non-normality of GEF’s fission yields. In the application that we present, the best data assimilation framework decreased the average bias of the simulations from 26% to 14%. The average relative standard deviation decreased from 21% to 14%. The GEF fission yields after data assimilation agreed better with those in JEFF3.3. For Pu-239 thermal fission, the average relative difference from JEFF3.3 was 16% before data assimilation and after it was 12%. For the standard deviations of the fission yields, GEF’s were 100% larger than JEFF3.3’s before data assimilation and after were only 4% larger. The inconsistency of the integral data had an important effect on MOCABA, as shown with the Marginal Likelihood Optimization method. When the method was not applied, MOCABA’s adjusted fission yields worsened the bias of the simulations by 30%. BFMC showed that it inherently accounted for this inconsistency. Applying Marginal Likelihood Optimization with BFMC gave a 2% lower bias compared to not applying it, but the results were more poorly converged.

scholarly journals FES2014 global ocean tides atlas: design and performances

2020 ◽  
Author(s):  
Florent H. Lyard ◽  
Damien J. Allain ◽  
Mathilde Cancet ◽  
Loren Carrère ◽  
Nicolas Picot
Keyword(s):  
Data Assimilation ◽  
Unstructured Grid ◽  
Primary Objective ◽  
Global Ocean ◽  
Open Boundary ◽  
Attraction Effects ◽  
Tidal Constituents ◽  
Accuracy Assessments ◽  
Data Assimilation Technique ◽  
By Products

Abstract. Since the mid-1990’s, a series of Finite Element Solution (FES) global ocean tidal atlases has been produced and released with the primary objective to provide altimetry missions with tidal de-aliasing correction at the best possible accuracy. We describe the underlying hydrodynamic and data assimilation designs for the last FES2014 release (finalized in early 2016), and some accuracy assessments especially for the altimetry de-aliasing purposes. The FES2014 atlas shows extremely significant improvements compared to the FES2004 (Lyard et al. 2006) and (intermediary) FES2012 atlases, in all ocean regions, especially in shelf and coastal seas; these advances are due to the unstructured grid flexible resolution, recent progress in the (prior to assimilation) hydrodynamic tidal solutions and to the use of an ensemble data assimilation technique. Compared to earlier releases, the FES2014 available tidal constituents spectrum has been significantly extended, the overall resolution augmented; some new additional scientific by-products have been derived from the atlas and are available, including the loading and self-attraction effects, energy diagnostics or the lowest astronomical tides . Compared to the other available global ocean tidal atlases, FES2014 clearly shows improved de-aliasing performances in most of the global ocean areas. It has consequently been integrated in satellite altimetry and gravimetry data processing, and adopted in recently renewed ITRF standards. It also provides very accurate open boundary tidal conditions for regional and coastal modelling.

NEUTRONIC PERFORMANCE OF (U, Pu)C FUEL IN A LATTICE OF GFR USING SCALE 6.0

2016 ◽  
Vol 1814 ◽  
Author(s):  
A. A. P. Macedo ◽  
Carlos E. Velasquez ◽  
C. A. M. da Silva ◽  
C. Pereira
Keyword(s):  
Computer Analysis ◽  
Fast Reactor ◽  
Nuclear Data ◽  
Multiplication Factor ◽  
Assembly Model ◽  
Fast Reactors ◽  
Full Power ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Neutronic Performance

ABSTRACTThis paper studies the performance of (U, Pu)C fuel in a hexagonal assembly of a GFR (Gas Fast Reactor). The SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation version 6.0) code was used in the calculation. The goal is to evaluate the behavior of the infinite multiplication factor (kinf) for a heterogeneous assembly model using four nuclear data libraries: V6-238, V7-238, ENDF/B-VI.8 and ENDF/B-VII.0. The burnup of (U, Pu)C was performed by the TRITON-6 module, and the isotopic concentrations were evaluated during the cycle. The present work comprises calculations at Zero Power and Full Power condition. This study intends to achieve more information about different Fast Reactors.

scholarly journals Analysis of the EBR-II SHRT-45R Unprotected Loss of Flow Experiment with ERANOS and RELAP

2015 ◽  
Vol 2015 ◽  
pp. 1-14 ◽  
Author(s):  
W. F. G. van Rooijen ◽  
H. Mochizuki
Keyword(s):  
Liquid Metal ◽  
Fast Reactor ◽  
Flow Distribution ◽  
Nuclear Data ◽  
Measured Data ◽  
Outlet Temperature ◽  
Fast Reactors ◽  
Flow Experiment ◽  
Neutronic Parameters

This paper presents the results of the analysis of the Unprotected Loss of Flow (ULOF) experiment SHRT-45R performed in the EBR-II fast reactor. These experiments are being analyzed in the scope of a benchmark exercise coordinated by the IAEA. The SHRT-45R benchmark contains a neutronic and a thermal-hydraulic part and results are presented for both. Neutronic calculations are performed with the ERANOS2.0 code in combination with various sets of nuclear data. The thermal-hydraulic evaluation is done with RELAP5-3D. The results are that the major neutronic parameters are well predicted with error margins on the order of 1%. The thermal-hydraulic results are less favourable: a consistent overestimation of the outlet temperature occurs in combination with erroneous flow distribution. Observed differences with measured data cannot be explained easily. The work presented in this paper was undertaken to investigate and validate the effectiveness of the calculational tools and data that are commonly used in our lab for the design and analysis of liquid metal cooled fast reactors.

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