scholarly journals UNCERTAINTY ANALYSIS TO C5G7-TD BENCHMARK BASED ON THE COST METHOD

2021 ◽  
Vol 247 ◽  
pp. 15007
Author(s):  
Liangzhi Cao ◽  
Zhuojie Sui ◽  
Bo Wang ◽  
Chenghui Wan ◽  
Zhouyu Liu
Keyword(s):  
Uncertainty Analysis ◽  
Nuclear Reactor ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Reactor Physics ◽  
Oriented Sample ◽  
Neutron Kinetics ◽  
Minimal Sample ◽  
Physics Simulation ◽  
The Cost

A method of Covariance-Oriented Sample Transformation (COST) has been proposed in our previous work to provide the converged uncertainty analysis results with a minimal sample size. The transient calculation of nuclear reactor is a key part of the reactor-physics simulation, so the accuracy and confidence of the neutron kinetics results have attracted much attention. In this paper, the Uncertainty Quantification (UQ) function of the high fidelity neutronics code NECP-X has been developed based on our home-developed uncertainty analysis code UNICORN, building a platform for the UQ of the transient calculation. Furthermore, the well-known space-time heterogeneous neutron kinetics benchmark C5G7 and its uncertainty propagation from the nuclear data to the interested key parameters of the core have been investigated. To address the problem of “the curse of dimensionality” caused by the large number of input parameters, the COST method has been applied to generate multivariate normal-distribution samples in uncertainty analysis. As a result, the law of the assembly/pin normalized power and their uncertainty with respect to time after introducing an instantaneous perturbation has been obtained. From the numerical results, it can be observed that the maximum relative uncertainties for the assembly normalized power can up to be about 1.65% and the value for the pin-wise power distributions can be about 2.71%.

scholarly journals New features and improvements in the NEA nuclear data tool suite

2020 ◽  
Vol 239 ◽  
pp. 19003
Author(s):  
M. Fleming ◽  
I. Hill ◽  
J. Dyrda ◽  
L. Fiorito ◽  
N. Soppera ◽  
...  
Keyword(s):  
Relational Databases ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Data Types ◽  
Reactor Physics ◽  
Energy Agency ◽  
Order Of Magnitude ◽  
Data Files ◽  
Improved Performance ◽  
Data Libraries

The OECD Nuclear Energy Agency (NEA) has developed and maintains several products that are used in the verification and validation of nuclear data, including the Java-based Nuclear Data Information System (JANIS) and the Nuclear Data Sensitivity Tool (NDaST). These integrate other collections of the NEA, including the International Handbooks of benchmark experiments on Criticality Safety and Reactor Physics (ICSBEP and IRPhEP) and their supporting relational databases (DICE and IDAT). Recent development of the JANIS, DICE and NDaST systems have resulted in the ability to perform uncertainty propagation utilising Legendre polynomial sensitivities, calculation of case-to-case covariances and correlations, use of spectrum weighting in perturbations, calculation of statistical results with suites of randomly sampled nuclear data files and new command-line interfaces to automate analyses and generate XML outputs. All of the most recent, major nuclear data libraries have been fully processed and incorporated, along with new visualisation features for covariances and sensitivities, an expanded set of reaction channel definitions, and new EXFOR data types defined by the NRDC. Optimisation of numerical methods has also improved performance, with over order-of-magnitude speed-up in the case of sensitivity-uncertainty calculations.

scholarly journals AN IRSN CONTRIBUTION TO THE UAM PROJECT: THERMAL-HYDRAULIC AND NEUTRONIC UNCERTAINTIES PROPAGATION IN A ROD EJECTION, FIRST RESULTS

2021 ◽  
Vol 247 ◽  
pp. 07003
Author(s):  
A. Sargeni ◽  
E. Ivanov
Keyword(s):  
Cross Sections ◽  
Uncertainty Propagation ◽  
Direct Calculation ◽  
Nuclear Data ◽  
Reactor Physics ◽  
Mean Power ◽  
Flow Area ◽  
First Results ◽  
The Mean ◽  
Power Variance

The paper presents our first results of the exercise III-I-2c from the OECD-NEA UAM-LWR benchmark intended to an elaboration of the methodology of uncertainty propagation. The considered case studied a full PWR core behavior in fast (~0.1 sec) rod ejection transient. According to the benchmark, the core represented a Hot Zero Power state. Authors used brute-force sampling propagating nuclear data and thermo-fluid uncertainties using 3D computational IRSN chain HEMERA. It couples the reactor physics code CRONOS and thermal-hydraulic core code FLICA4. The nuclear data uncertainties were represented in a form of cross sections standard deviations (in percentage of the mean cross sections values) supplied by the UAM team. In addition to the original benchmark, the study includes a case with an increased power peak by supplementary rod ejection, i.e. with higher reactivity. Both the results are similar to what we obtained in the mini-core rod ejection: the power standard deviation follows, in percentage of the mean power, the mean power curve. We split the variance with a direct calculation: once the cross sections are modified and the thermal-hydraulics inputs are kept constant, another time the contrary. The results show that uncertainties dues to nuclear data dominate over ones due to the thermal-flow area. Furthermore, the major contributors in peak-of-power variance lie in a fast group of cross sections.

scholarly journals FAST REACTOR MULTIPHYSICS AND UNCERTAINTY PROPAGATION WITHIN WIMS

2021 ◽  
Vol 247 ◽  
pp. 06002
Author(s):  
Ben Lindley ◽  
Brendan Tollit ◽  
Peter Smith ◽  
Alan Charles ◽  
Robert Mason ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Uncertainty Propagation ◽  
Predictive Modelling ◽  
Nuclear Data ◽  
Reactor Physics ◽  
Fuel Performance ◽  
Horizon 2020 ◽  
Calculation Methodology ◽  
International Projects ◽  
Fast Reactor Fuel

For liquid metal-cooled fast reactors (LMFRs), improved predictive modelling is desirable to facilitate reactor licensing and operation and move towards a best estimate plus uncertainty (BEPU) approach. A key source of uncertainty in fast reactor calculations arises from the underlying nuclear data. Addressing the propagation of such uncertainties through multiphysics calculations schemes is therefore of importance, and is being addressed through international projects such as the Sodium-cooled Fast Reactor Uncertainty Analysis in Modelling (SFR-UAM) benchmark. In this paper, a methodology for propagation of nuclear data uncertainties within WIMS is presented. Uncertainties on key reactor physics parameters are calculated for selected SFR-UAM benchmark exercises, with good agreement with previous results. A methodology for coupled neutronic-thermal-hydraulic calculations within WIMS is developed, where thermal feedback is introduced to the neutronic solution through coupling with the ARTHUR subchannel code within WIMS and applied to steady-state analysis of the Horizon 2020 ESFR-SMART project reference core. Finally, integration of reactor physics and fuel performance calculations is demonstrated through linking of the WIMS reactor physics code to the TRAFIC fast reactor fuel performance code, through a Fortran-C-Python (FCP) interface. Given the 3D multiphysics calculation methodology, thermal-hydraulic and fuel performance uncertainties can ultimately be sampled alongside the nuclear data uncertainties. Together, these developments are therefore an important step towards enabling propagation of uncertainties through high fidelity, multiphysics SFR calculations and hence facilitate BEPU methodologies.

scholarly journals The Impact of Fueling Operations on Full Core Uncertainty Analysis in CANDU Reactors

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
M. D. Tucker ◽  
D. R. Novog
Keyword(s):  
Standard Deviation ◽  
Uncertainty Analysis ◽  
Cross Sections ◽  
Nuclear Data ◽  
Analysis Framework ◽  
Reactor Physics ◽  
Candu Reactors ◽  
Fixed Set ◽  
The Impact ◽  
Full Core

Abstract Within emerging best-estimate-plus-uncertainty (BEPU) approaches, code output uncertainties can be inferred from the propagation of fundamental or microscopic uncertainties. This paper examines the propagation of fundamental nuclear data uncertainties though the entire analysis framework to predict macroscopic reactor physics phenomena, which can be measured in Canada Deuterium Uranium (CANDU) reactors. In this work, 151 perturbed multigroup cross sections libraries, each based on a set of perturbed microscopic nuclear data, were generated. Subsequently, these data were processed into few-group cross sections and used to generate full-core diffusion models in PARCS. The impact of these nuclear data perturbations leads to changes in core reactivity for a fixed set of fuel compositions of 4.5 mk. The impact of online fueling operations was simulated using a series of fueling rules, which attempted to mimic operator actions during CANDU operations such as studying the assembly powers and selecting fueling sites, which would minimize the deviation in power from some desirable reference condition or increasing or decreasing fueling frequency to manage reactivity. An important feature of this analysis was to perform long-transients (1–3 years) starting with each one of the 151 perturbed full core models. It was found that the operational feedback reduced the standard deviation in core reactivity by 99% from 0.0045 to 2.8 × 10−5. Overall, the conclusions demonstrate that while microscopic nuclear data uncertainties may give rise to large macroscopic variability during simple propagation, when important macrolevel feedback are considered the variability is significantly reduced.

scholarly journals The Dilution Dependency of Multigroup Uncertainties

2014 ◽  
Vol 2014 ◽  
pp. 1-14
Author(s):  
M. R. Ball ◽  
C. McEwan ◽  
D. R. Novog ◽  
J. C. Luxat
Keyword(s):  
Uncertainty Analysis ◽  
Cross Sections ◽  
Infinite Dilution ◽  
Nuclear Data ◽  
Monte Carlo Sampling ◽  
Reactor Physics ◽  
Cross Sectional ◽  
Rigorous Approach ◽  
Common Strategy ◽  
Potential Methods

The propagation of nuclear data uncertainties through reactor physics calculation has received attention through the Organization for Economic Cooperation and Development—Nuclear Energy Agency’s Uncertainty Analysis in Modelling (UAM) benchmark. A common strategy for performing lattice physics uncertainty analysis involves starting with nuclear data and covariance matrix which is typically available at infinite dilution. To describe the uncertainty of all multigroup physics parameters—including those at finite dilution—additional calculations must be performed that relate uncertainties in an infinite dilution cross-section to those at the problem dilution. Two potential methods for propagating dilution-related uncertainties were studied in this work. The first assumed a correlation between continuous-energy and multigroup cross-sectional data and uncertainties, which is convenient for direct implementation in lattice physics codes. The second is based on a more rigorous approach involving the Monte Carlo sampling of resonance parameters in evaluated nuclear data using the TALYS software. When applied to a light water fuel cell, the two approaches show significant differences, indicating that the assumption of the first method did not capture the complexity of physics parameter data uncertainties. It was found that the covariance of problem-dilution multigroup parameters for selected neutron cross-sections can vary significantly from their infinite-dilution counterparts.

scholarly journals Uncertainties of calculated coincidence-summing correction factors in gamma-ray spectrometry

2020 ◽  
Vol 239 ◽  
pp. 12003
Author(s):  
V. Semkova ◽  
N. Otuka ◽  
A.J.M. Plompen
Keyword(s):  
Point Source ◽  
Uncertainty Analysis ◽  
Hpge Detector ◽  
Gamma Ray ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Gamma Ray Spectrometry ◽  
Correction Factors ◽  
P Type ◽  
Coincidence Summing

Uncertainty propagation to the γ-γ coincidence-summing correction factor from the covariances of the nuclear data and detection efficiencies have been formulated. The method was applied in the uncertainty analysis of the coincidence-summing correction factors in the γ-ray spectrometry of the 134Cs point source using a p-type coaxial HPGe detector.

Covariance-oriented sample transformation: A new sampling method for reactor-physics uncertainty analysis

2019 ◽  
Vol 134 ◽  
pp. 452-463 ◽  
Author(s):  
Zhuojie Sui ◽  
Liangzhi Cao ◽  
Chenghui Wan ◽  
Xiaoyang Zou
Keyword(s):  
Uncertainty Analysis ◽  
Sampling Method ◽  
Reactor Physics ◽  
Oriented Sample

scholarly journals Cross sections Data Adjustment For KRITZ-2:13

2021 ◽  
Author(s):  
Abdulaziz Ahmed ◽  
◽  
H. Boukhal Boukhal ◽  
E. Chakir Chakir ◽  
S. EL Ouahdani ◽  
...  
Keyword(s):  
Cross Sections ◽  
Nuclear Reactor ◽  
Least Squares Method ◽  
Generalized Least Squares ◽  
Nuclear Data ◽  
Nuclear Model ◽  
Reactor Physics ◽  
Generalized Least Squares Method ◽  
Great Reliability ◽  
Data Adjustment

Over the past years, the cross-sections reaction data has been re-evaluated several times, in order to approximate the nuclear model measurements with the predictions with great reliability. In our work, uncertainty analysis caused by the data on the neutron factor (Keff) and the reactivity temperature coefficient (RTC), in addition to nuclear data adjustment related to the nuclear reactor physics have been done for KRITZ-2:13 reactor, with ENDF/B - VII.1, ENDF/B - VIII.0 and JENDL - 4.0 evaluations by the nuclear code MCNP6.1. Our analysis detects that the greatest uncertainty on Keff and RTC in the studied libraries comes from the capture and fission reaction contributions respectively, for U-238 and U-235. The previous reactions and their covariances were adjusted using the generalized least squares method (GLLSM), in order to contribute to improve the data needed for neutron simulation of experiments and to ensure the installations safety, where Keff and RTC represent neutron parameters reflecting the modification effects in the data.

scholarly journals Nuclear data sensitivity analysis and uncertainty propagation in the KYADJ whole-core transport code

2020 ◽  
Vol 239 ◽  
pp. 22012
Author(s):  
Qu Wu ◽  
Xingjie Peng ◽  
Guanlin Shi ◽  
Yingrui Yu ◽  
Qing Li ◽  
...  
Keyword(s):  
Sensitivity Analysis ◽  
Uncertainty Analysis ◽  
Neutron Transport ◽  
Uncertainty Propagation ◽  
Reactor Core ◽  
Nuclear Data ◽  
Data File ◽  
Neutron Transport Equation ◽  
Transport Code ◽  
Data Sensitivity

Nuclear data sensitivity analysis and uncertainty propagation have been extensively applied to nuclear data adjustment and uncertainty quantification in the field of nuclear engineering. Sensitivity and Uncertainty (S&U) analysis is developed in the KYADJ whole-core transport code in order to meet the requirement of advanced reactor design. KYADJ aims to use two-dimension Method of Characteristic (MOC) and one-dimension discrete ordinate (SN) coupled method to solve the neutron transport equation and achieve one-step direct transport calculation of the reactor core. Developing sensitivity and uncertainty analysis module in KYADJ can minimize deviations caused by modeling approximation and enhance calculation efficiency. This work describes the application of the classic perturbation theory to the KYADJ transport solver. In order to obtain uncertainty, a technique is proposed for processing a covariance data file in 45-group energy grid instead of 44-group SCALE 6.1 covariance data which is extensively used in various codes. Numerical results for Uncertainty Analysis in Modelling (UAM) benchmarks and the SF96 benchmark are presented. The results agree well with the reference and the capability of S&U analysis in KYADJ is verified.

scholarly journals Perturbation-Theory-Based Sensitivity and Uncertainty Analysis with CASMO-4

2012 ◽  
Vol 2012 ◽  
pp. 1-11 ◽  
Author(s):  
Maria Pusa
Keyword(s):  
Perturbation Theory ◽  
Uncertainty Analysis ◽  
Adjoint System ◽  
Nuclear Data ◽  
Theoretical Background ◽  
Reactor Physics ◽  
Deterministic Transport ◽  
Practical Guidelines ◽  
Rate Ratios ◽  
Sensitivity And Uncertainty Analysis

The topic of this paper is the development of sensitivity and uncertainty analysis capability to the reactor physics code CASMO-4 in the context of the UAM (Uncertainty Analysis in Best-Estimate Modelling for Design, Operation, and Safety Analysis of LWRs) benchmark. The sensitivity analysis implementation is based on generalized perturbation theory, which enables computing the sensitivity profiles of reaction rate ratios efficiently by solving one generalized adjoint system for each response. Both the theoretical background and the practical guidelines for modifying a deterministic transport code to compute the generalized adjoint solutions and sensitivity coefficients are reviewed. The implementation to CASMO-4 is described in detail. The developed uncertainty analysis methodology is deterministic, meaning that the uncertainties are computed based on the sensitivity profiles and covariance matrices for the uncertain nuclear data parameters. The main conclusions related to the approach used for creating a covariance library compatible with the cross-section libraries of CASMO-4 are presented. Numerical results are given for a lattice physics test problem representing a BWR, and the results are compared to the TSUNAMI-2D sequence in SCALE 6.1.

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