scholarly journals SWANLAKE, a computer code utilizing ANISN radiation transport calculations for cross section sensitivity analysis

1973 ◽  
Author(s):  
D. E. Bartine ◽  
F. R. Mynatt ◽  
E. M. Oblow
Keyword(s):  
Sensitivity Analysis ◽  
Cross Section ◽  
Radiation Transport ◽  
Computer Code

scholarly journals SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross-Section Sensitivity-Uncertainty Analysis

2017 ◽  
Vol 2017 ◽  
pp. 1-16 ◽  
Author(s):  
Ivan A. Kodeli ◽  
Slavko Slavič
Keyword(s):  
Cross Section ◽  
Radiation Transport ◽  
Computer Code ◽  
Data Bank ◽  
Code System ◽  
Input And Output ◽  
Complete Chain ◽  
Driven System ◽  
Accelerator Driven System ◽  
User Friendly

A Windows interface XSUN-2017 facilitating the deterministic radiation transport and cross-section sensitivity-uncertainty (S/U) calculation is presented. The package was developed to assist the users in the preparation of input cards, rapid modification, and execution of the complete chain of codes including TRANSX, PARTISN, and SUSD3D, all available from the OECD/NEA Data Bank and RSICC. The objective of this work was to make the input and output handling for these codes as user-friendly as possible, passing information among codes internally. XSUN-2017 allows a user-friendly viewing of results obtained from the PARTISN and SUSD3D programs. The first version of the Windows interface XSUN-2013 was developed in 2013 and submitted to OECD/NEA Data Bank Computer Code Collection and RSICC in early 2014. An updated version, XSUN-2017, will be released in 2017. The package includes also the new version of the SUSD3D code. The XSUN-2017 and SUSD3D code systems and recent improvements and updates are described. Examples of the use and validation are presented, including the S/U intercomparison exercise using the SNEAK-7 benchmarks involving the XSUN-2017 code system comparison with the codes such as TSUNAMI, SERPENT, and MCNP6, and the S/U analysis of the keff and βeff parameters for the MYRRHA accelerator driven system (ADS).

Sobol′ Sensitivity Analysis Using a Neural Network Model of a LB-LOCA in the ZION Nuclear Power Plant With CATHARE-2 V2.5 Code

2013 ◽  
Author(s):  
Fabrice Fouet ◽  
Pierre Probst
Keyword(s):  
Neural Network ◽  
Sensitivity Analysis ◽  
Nuclear Power ◽  
Uncertainty Propagation ◽  
Correlation Coefficients ◽  
Computer Code ◽  
Numerical Application ◽  
Loss Of Coolant Accident ◽  
Input Parameters ◽  
Artificial Neural Network Ann

In nuclear safety, the Best-Estimate (BE) codes may be used in safety demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters, which comes mainly from the lack of knowledge of the input parameters, is evaluated by estimating the 95% percentile with a high degree of confidence. IRSN, technical support of the French Safety Authority, developed a method of uncertainty propagation. This method has been tested with the BE code used is CATHARE-2 V2.5 in order to evaluate the Peak Cladding Temperature (PCT) of the fuel during a Large Break Loss Of Coolant Accident (LB-LOCA) event, starting from a large number of input parameters. A sensitivity analysis is needed in order to limit the number of input parameters and to quantify the influence of each one on the response variability of the numerical model. Generally, the Global Sensitivity Analysis (GSA) is done with linear correlation coefficients. This paper presents a new approach to perform a more accurate GSA to determine and to classify the main uncertain parameters: the Sobol′ methodology. The GSA requires simulating many sets of parameters to propagate uncertainties correctly, which makes of it a time-consuming approach. Therefore, it is natural to replace the complex computer code by an approximate mathematical model, called response surface or surrogate model. We have tested Artificial Neural Network (ANN) methodology for its construction and the Sobol′ methodology for the GSA. The paper presents a numerical application of the previously described methodology on the ZION reactor, a Westinghouse 4-loop PWR, which has been retained for the BEMUSE international problem [8]. The output is the first maximum PCT of the fuel which depends on 54 input parameters. This application outlined that the methodology could be applied to high-dimensional complex problems.

scholarly journals Sensitivity analysis of free torsional vibration frequencies of thin-walled laminated beams under axial load

2019 ◽  
Vol 32 (5) ◽  
pp. 1347-1356 ◽  
Author(s):  
Czesław Szymczak ◽  
Marcin Kujawa
Keyword(s):  
Sensitivity Analysis ◽  
Cross Section ◽  
Torsional Vibration ◽  
Volume Fraction ◽  
Fibre Volume Fraction ◽  
Vibration Frequencies ◽  
Cross Sectional ◽  
First Order ◽  
Thin Walled ◽  
Frequency Changes

AbstractThe paper addresses sensitivity analysis of free torsional vibration frequencies of thin-walled beams of bisymmetric open cross section made of unidirectional fibre-reinforced laminate. The warping effect and the axial end load are taken into account. The consideration is based upon the classical theory of thin-walled beams of non-deformable cross section. The first-order sensitivity variation of the frequencies is derived with respect to the design variable variations. The beam cross-sectional dimensions and the material properties are assumed the design variables undergoing variations. The paper includes a numerical example related to simply supported I-beams and the distributions of sensitivity functions of frequencies along the beam axis. Accuracy is discussed of the first-order sensitivity analysis in the assessment of frequency changes due to the fibre volume fraction variable variations, and the effect of axial loads is discussed too.

scholarly journals Uncertainty and Sensitivity Analysis of Void Reactivity Feedback for 3D BWR Assembly Model

2017 ◽  
Vol 2017 ◽  
pp. 1-9
Author(s):  
Andrius Slavickas ◽  
Raimondas Pabarčius ◽  
Aurimas Tonkūnas ◽  
Eugenijus Ušpuras
Keyword(s):  
Sensitivity Analysis ◽  
Cross Section ◽  
Input Data ◽  
Statistical Approach ◽  
Assembly Model ◽  
Uncertainty And Sensitivity Analysis ◽  
Geometrical Dimensions ◽  
Basic Input ◽  
Propagation Of Uncertainties ◽  
Selection Of

Uncertainty and sensitivity analysis of void reactivity feedback for 3D BWR fuel assembly model is presented in this paper. Uncertainties in basic input data, such as the selection of different cross section library, manufacturing uncertainties in material compositions, and geometrical dimensions, as well as operating data are considered. An extensive modelling of different input data realizations associated with their uncertainties was performed during sensitivity analysis. The propagation of uncertainties was analyzed using the statistical approach. The results revealed that important information on the code predictions can be obtained by analyzing and comparing the codes estimations and their associated uncertainties.

scholarly journals CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

2018 ◽  
Vol 4 ◽  
pp. 10 ◽  
Author(s):  
Guillaume Ritter ◽  
Romain Eschbach ◽  
Richard Girieud ◽  
Maxime Soulard
Keyword(s):  
Nuclear Fuel ◽  
Cross Section ◽  
Cross Sections ◽  
Power Plants ◽  
Fuel Cycle ◽  
Nuclear Data ◽  
Computer Code ◽  
Test Case ◽  
Industrial Facilities ◽  
Transport Code

CESAR stands in French for “simplified depletion applied to reprocessing”. The current version is now number 5.3 as it started 30 years ago from a long lasting cooperation with ORANO, co-owner of the code with CEA. This computer code can characterize several types of nuclear fuel assemblies, from the most regular PWR power plants to the most unexpected gas cooled and graphite moderated old timer research facility. Each type of fuel can also include numerous ranges of compositions like UOX, MOX, LEU or HEU. Such versatility comes from a broad catalog of cross section libraries, each corresponding to a specific reactor and fuel matrix design. CESAR goes beyond fuel characterization and can also provide an evaluation of structural materials activation. The cross-sections libraries are generated using the most refined assembly or core level transport code calculation schemes (CEA APOLLO2 or ERANOS), based on the European JEFF3.1.1 nuclear data base. Each new CESAR self shielded cross section library benefits all most recent CEA recommendations as for deterministic physics options. Resulting cross sections are organized as a function of burn up and initial fuel enrichment which allows to condensate this costly process into a series of Legendre polynomials. The final outcome is a fast, accurate and compact CESAR cross section library. Each library is fully validated, against a stochastic transport code (CEA TRIPOLI 4) if needed and against a reference depletion code (CEA DARWIN). Using CESAR does not require any of the neutron physics expertise implemented into cross section libraries generation. It is based on top quality nuclear data (JEFF3.1.1 for ∼400 isotopes) and includes up to date Bateman equation solving algorithms. However, defining a CESAR computation case can be very straightforward. Most results are only 3 steps away from any beginner's ambition: Initial composition, in core depletion and pool decay scenario. On top of a simple utilization architecture, CESAR includes a portable Graphical User Interface which can be broadly deployed in R&D or industrial facilities. Aging facilities currently face decommissioning and dismantling issues. This way to the end of the nuclear fuel cycle requires a careful assessment of source terms in the fuel, core structures and all parts of a facility that must be disposed of with “industrial nuclear” constraints. In that perspective, several CESAR cross section libraries were constructed for early CEA Research and Testing Reactors (RTR’s). The aim of this paper is to describe how CESAR operates and how it can be used to help these facilities care for waste disposal, nuclear materials transport or basic safety cases. The test case will be based on the PHEBUS Facility located at CEA − Cadarache.

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