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 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%.

Uncertainty Analysis for the Steady-Stated BEAVRS Benchmark Problem at ARO and ARI Situations

2017 ◽  
Author(s):  
Chenghui Wan ◽  
Liangzhi Cao ◽  
Hongchun Wu
Keyword(s):  
Uncertainty Analysis ◽  
Sampling Method ◽  
Nuclear Data ◽  
Multiplication Factor ◽  
Statistical Sampling ◽  
Benchmark Problem ◽  
Reactor Physics ◽  
The Core ◽  
Maximum Value ◽  
Fuel Assemblies

In this paper, the capability of uncertainty propagations of the nuclear-data to the reactor-physics calculations has been implemented in our home-developed code NECP-UNICORN based on the statistical sampling method (SSM). The “two-step” scheme has been applied in NECP-UNICORN to perform the uncertainty analysis for the reactor-physics calculations. For the lattice calculations, the nuclear-data uncertainties are propagated to the few-group constants; then for the core simulations, the uncertainties of the multiplication factor and power distributions introduced by the few-group constants’ uncertainties can be quantified. Applying the NECP-UNICORN code, uncertainty analysis has been performed to the BEAVRS benchmark problem at the Hot Zero Power (HZP) conditions, with situations of All Rod In (ARI) and All Rod Out (ARO). From the numerical results, it can be observed that for the multiplication factors of the core simulations, the relative uncertainties are about 5.1‰ for the ARO situation and 5.0‰ for the ARI situation, which are the same magnitude of the relative uncertainties of the fuel assemblies; for the radial power distributions, the relative uncertainties can up to be 4.27% as the maximum value and 2.08% as the RMS value for the ARO situation, and 6.03% as the maximum value and 2.37% as the RMS value for the ARI situation.

Research on Monte Carlo Perturbation Calculation Methods Applied in Reactor Physics

2009 ◽  
Author(s):  
Ze-guang Li ◽  
Kan Wang ◽  
Gang-lin Yu
Keyword(s):  
Monte Carlo ◽  
Differential Operator ◽  
Sampling Method ◽  
Perturbation Methods ◽  
Calculation Model ◽  
Perturbation Calculation ◽  
Reactor Physics ◽  
Tracking Method ◽  
Monte Carlo Techniques ◽  
Sensitivity Studies

In the reactor design and analysis, there is often a need to calculate the effects caused by perturbations of temperature, components and even structure of reactors on reactivity. And in sensitivity studies, uncertainty analysis of target quantities and unclear data adjustment, perturbation calculations are also widely used. To meet the need of different types of reactors (complex, multidimensional systems), Monte Carlo perturbation methods have been developed. In this paper, several kinds of perturbation methods are investigated. Specially, differential operator sampling method and correlated tracking method are discussed in details. MCNP’s perturbation calculation capability is discussed by calculating certain problems, from which some conclusions are obtained on the capabilities of the differential operator sampling method used in the perturbation calculation model of MCNP. Also, a code using correlated tracking method has been developed to solve certain problems with cross-section changes, and the results generated by this code agree with the results generated by straightforward Monte Carlo techniques.

scholarly journals ANSWERS® TOOLS FOR UNCERTAINTY QUANTIFICATION AND VALIDATION

2021 ◽  
Vol 247 ◽  
pp. 15015
Author(s):  
Paul N Smith ◽  
Dave Hanlon ◽  
Geoff Dobson ◽  
Richard Hiles ◽  
Tim Fry ◽  
...  
Keyword(s):  
Uncertainty Quantification ◽  
Sampling Method ◽  
Similarity Index ◽  
Nuclear Data ◽  
Covariance Matrices ◽  
Reactor Physics ◽  
Development Environment ◽  
Post Process ◽  
Manufacturing Tolerances ◽  
Validation Experiments

ANSWERS® is developing a set of uncertainty quantification (UQ) tools for use with its major physics codes: WIMS/PANTHER (reactor physics), MONK (criticality and reactor physics) and MCBEND (shielding and dosimetry). The Visual Workshop integrated development environment allows the user to construct and edit code inputs, launch calculations, post-process results and produce graphs, and recently uncertainty quantification and optimisation tools have been added. Prior uncertainties due to uncertainties in nuclear data or manufacturing tolerances can be estimated using the sampling method or using the sensitivity options in the physics codes combined with appropriate covariance matrices. To aid the user in the choice of appropriate validation experiments, the MONK categorisation scheme and/or a similarity index can be used. An interactive viewer has been developed which allows the user to search through, and browse details of, over 2,000 MONK validation experiments that have been analysed from the ICSBEP and IRPhE validation sets. A Bayesian updating approach is used to assimilate the measured data with the calculated results. It is shown how this process can be used to reduce bias in calculated results and reduce the calculated uncertainty on those results. This process is illustrated by application to a PWR fuel assembly.

Enhancement of Stochastic Sampling Capability in RMC Code

2018 ◽  
Author(s):  
Guanlin Shi ◽  
Yishu Qiu ◽  
Kan Wang
Keyword(s):  
Uncertainty Analysis ◽  
Sampling Method ◽  
Nuclear Data ◽  
Data Uncertainty ◽  
Monte Carlo Code ◽  
Tabular Data ◽  
Stochastic Sampling ◽  
First Order ◽  
Fission Spectrum ◽  
Sensitivity And Uncertainty Analysis

As people pay more attention to nuclear safety analysis, sensitivity and uncertainty analysis has become a research hotspot. In our previous research, we had developed an integrated, built-in stochastic sampling module in the Reactor Monte Carlo code RMC [1]. Using this module, we can perform nuclear data uncertainty analysis. But at that time the uncertainty of fission spectrum was not considered. So, in this work, the capability of computing the uncertainty of keff induced by the uncertainty of fission spectrum, including tabular data form and formula form, is implemented in RMC code based on the stochastic sampling method. The algorithms and capability of computing keff uncertainty induced by uncertainty of fission spectrum in RMC are verified by comparison with the results calculated by the first order uncertainty quantification method [2].

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.

Evaluation of the Esfr End of Equilibrium Cycle State: Spatial Distributions of Reactivity Coefficients

2021 ◽  
Author(s):  
Una Baker ◽  
Marat Margulis ◽  
Eugene Shwageraus ◽  
Emil Fridman ◽  
Antonio Jiménez-Carrascosa ◽  
...  
Keyword(s):  
Uncertainty Analysis ◽  
Operating Conditions ◽  
Reactor Physics ◽  
Horizon 2020 ◽  
Decay Heat ◽  
Continuous Energy ◽  
Reactivity Coefficients ◽  
Sensitivity And Uncertainty Analysis ◽  
Commercial Size ◽  
Good Agreement

Abstract The Horizon 2020 ESFR-SMART project investigates the behaviour of the commercial-size European Sodium-cooled Fast Reactor (ESFR) throughout its lifetime. This paper reports work focused on the End of Equilibrium Cycle (EOEC) loading of the ESFR, including neutronic analysis, core- and zone-wise reactivity coefficients, and more detailed local mapping of important safety-relevant parameters. Sensitivity and uncertainty analysis on these parameters have also been performed and a detailed investigation into decay heat mapping carried out. Due to the scope of this work the results have been split into three papers. The nominal operating conditions and both zone-wise and local mapping of reactivity coefficients are considered in this paper; the sensitivity and uncertainty analysis are detailed in Margulis et al. [1]; and the decay heat mapping calculations are reported in Jimenez-Carrascosa et al. [2]. The work was performed across four institutions using both continuous-energy Monte Carlo and deterministic reactor physics codes. A good agreement is observed between the methods, verifying the suitability of these codes for simulation of large, complicated reactor configurations; and giving confidence in the results for the most limiting ESFR EOEC core state for safety analysis. The results from this work will serve as basis for the transient calculations planned for the next stage of work on the ESFR, allowing for more in-depth studies to be performed on the multiphysics behaviour of the reactor.

Sign in / Sign up

Export Citation Format

Share Document