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.

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.

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%.

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.

Representative neutronic characteristics calculations for the VVER-1000 reactors using SRAC and MCNP5

2016 ◽  
Vol 6 (2) ◽  
pp. 21-30
Author(s):  
Huu Tiep Nguyen ◽  
Viet Phu Tran ◽  
Tuan Khai Nguyen ◽  
Vinh Thanh Tran ◽  
Minh Tuan Nguyen
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Multiplication Factor ◽  
Benchmark Problem ◽  
Fuel Assemblies ◽  
Effective Multiplication Factor ◽  
Neutronic Characteristics ◽  
Effective Multiplication ◽  
Good Agreement

This paper presents the results of neutronic calculations using the deterministic and Monte-Carlo methods (the SRAC and MCNP5codes) for the VVER MOX Core Computational Benchmark Specification and the VVER-1000/V392 reactor core. The power distribution and keff value have been calculated for a benchmark problem of VVER core. The results show a good agreement between the SRAC and MCNP5 calculations. Then, neutronic characteristics of VVER-1000/V392 such as power distribution, infinite multiplication factor (k-inf) of the fuel assemblies, effective multiplication factor keff, peaking factor and Doppler coefficient were calculated using the two codes.

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.

Mixed Reload Design Using MOX and UOX Fuel Assemblies

2002 ◽  
Author(s):  
J. Ramo´n Rami´rez Sa´nchez ◽  
R. T. Perry
Keyword(s):  
Conceptual Design ◽  
Cycle Length ◽  
Multiplication Factor ◽  
The Core ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
Current Cycle ◽  
Full Power ◽  
Control Rods ◽  
Mixed Load

As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10×10 BWR fuel assemblies but different fisil material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fisil plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO2 fresh fuel were tested to verify the shutdown margin, the UO2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper.

Analysis of the Effects Caused by the Core Compaction Into LFR Reactor

2016 ◽  
Author(s):  
Rosa Lo Frano ◽  
Pugliese Giovanni
Keyword(s):  
Fast Reactor ◽  
Initial Conditions ◽  
Vertical Axis ◽  
Contact Condition ◽  
Time Interval ◽  
The Core ◽  
Half Height ◽  
Maximum Value ◽  
Fuel Assemblies ◽  
Metal Coolant

Due to the high inertia of the metal coolant, the safety concerns of the next generation LMRs (e.g. the Advanced Lead Fast Reactor European Demonstrator - ALFRED) have some connections with the core compaction phenomenon when severe earthquake occurs. In this paper the effects on the fuel assemblies (FAs) are numerically analyzed (by FEM code) taking into account suitable boundary and initial conditions. To characterize the interaction between the internal components, surface-to-surface contact condition has been implemented. The results indicate that the annular area neighboring the piping penetration ovalizes and so a circumferential buckling occurs. The FAs undergo bending deformation especially in correspondence of the half height of the elements. The displacement varies along the vertical axis (direction of maximum flexibility) reaching, in some time interval, the maximum value of about 9 cm. Vibration phenomenon also appeared.

Total Uncertainty Analysis for PWR Assembly Based on the Statistical Sampling Method

2016 ◽  
Vol 183 (3) ◽  
Author(s):  
Tiejun Zu ◽  
Chenghui Wan ◽  
Liangzhi Cao ◽  
Hongchun Wu ◽  
Wei Shen
Keyword(s):  
Uncertainty Analysis ◽  
Sampling Method ◽  
Statistical Sampling ◽  
Total Uncertainty
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