scholarly journals Methodology for Sensitivity Analysis of Homogenized Cross-Sections to Instantaneous and Historical Lattice Conditions with Application to AP1000® PWR Lattice

Energies ◽  
2021 ◽  
Vol 14 (12) ◽  
pp. 3378
Author(s):  
Dean Price ◽  
Thomas Folk ◽  
Matthew Duschenes ◽  
Krishna Garikipati ◽  
Brendan Kochunas
Keyword(s):  
Sensitivity Analysis ◽  
Cross Sections ◽  
Nuclear Reactor ◽  
The State ◽  
State Variables ◽  
Fuel Temperature ◽  
Step Method ◽  
Control Rod ◽  
And Control ◽  
The Impact

In the two-step method for nuclear reactor simulation, lattice physics calculations are performed to compute homogenized cross-sections for a variety of burnups and lattice configurations. A nodal code is then used to perform full-core analysis using the pre-calculated homogenized cross-sections. One source of uncertainty introduced in this method is that the lattice configuration or depletion conditions typically do not match a pre-calculated one from the lattice physics simulations. Therefore, some interpolation model must be used to estimate the homogenized cross-sections in the nodal code. This current study provides a methodology for sensitivity analysis to quantify the impact of state variables on the homogenized cross-sections. This methodology also allows for analyses of the historical effect that the state variables have on homogenized cross-sections. An application of this methodology on a lattice for the Westinghouse AP1000® reactor is presented where coolant density, fuel temperature, soluble boron concentration, and control rod insertion are the state variables of interest. The effects of considering the instantaneous values of the state variables, historical values of the state variables, and burnup-averaged values of the state variables are analyzed. Using these methods, it was found that a linear model that only considers the instantaneous and burnup-averaged values of state variables can fail to capture some variations in the homogenized cross-sections.

Design Sensitivity Analysis of Nonlinear Multidisciplinary Multibody Systems in the MIXEDMODELS Platform

2007 ◽  
Author(s):  
Shilpa A. Vaze ◽  
Prakash Krishnaswami ◽  
James DeVault
Keyword(s):  
Sensitivity Analysis ◽  
Multibody Systems ◽  
Design Sensitivity ◽  
Differential Algebraic Equations ◽  
The State ◽  
Design Sensitivity Analysis ◽  
Design Parameters ◽  
State Variables ◽  
Sensitivity Coefficients ◽  
Design Variables

Most state-of-the-art multibody systems are multidisciplinary and encompass a wide range of components from various domains such as electrical, mechanical, hydraulic, pneumatic, etc. The design considerations and design parameters of the system can come from any of these domains or from a combination of these domains. In order to perform analytical design sensitivity analysis on a multidisciplinary system (MDS), we first need a uniform modeling approach for this class of systems to obtain a unified mathematical model of the system. Based on this model, we can derive a unified formulation for design sensitivity analysis. In this paper, we present a modeling and design sensitivity formulation for MDS that has been successfully implemented in the MIXEDMODELS (Multidisciplinary Integrated eXtensible Engine for Driving Metamodeling, Optimization and DEsign of Large-scale Systems) platform. MIXEDMODELS is a unified analysis and design tool for MDS that is based on a procedural, symbolic-numeric architecture. This architecture allows any engineer to add components in his/her domain of expertise to the platform in a modular fashion. The symbolic engine in the MIXEDMODELS platform synthesizes the system governing equations as a unified set of non-linear differential-algebraic equations (DAE’s). These equations can then be differentiated with respect to design to obtain an additional set of DAE’s in the sensitivity coefficients of the system state variables with respect to the system’s design variables. This combined set of DAE’s can be solved numerically to obtain the solution for the state variables and state sensitivity coefficients of the system. Finally, knowing the system performance functions, we can calculate the design sensitivity coefficients of these performance functions by using the values of the state variables and state sensitivity coefficients obtained from the DAE’s. In this work we use the direct differentiation approach for sensitivity analysis, as opposed to the adjoint variable approach, for ease in error control and software implementation. The capabilities and performance of the proposed design sensitivity analysis formulation are demonstrated through a numerical example consisting of an AC rectified DC power supply driving a slider crank mechanism. In this case, the performance functions and design variables come from both electrical and mechanical domains. The results obtained were verified by perturbation analysis, and the method was shown to be very accurate and computationally viable.

scholarly journals Effects of relaxation on anxiety and salivary IgA levels in puerperae

2008 ◽  
Vol 16 (1) ◽  
pp. 36-41 ◽  
Author(s):  
Cândida Caniçali Primo ◽  
Maria Helena Costa Amorim
Keyword(s):  
Statistical Tests ◽  
Immunoglobulin A ◽  
Maternity Hospital ◽  
The State ◽  
Relaxation Techniques ◽  
State Variables ◽  
State Trait Anxiety Inventory ◽  
Salivary Iga ◽  
And Control ◽  
Experimental Group

This experimental study aimed to evaluate the effect of relaxation techniques on anxiety levels, and the relation between anxiety and the concentration of Immunoglobulin A. The study was carried out in a maternity hospital in a city of the State of Espírito Santo, Brazil. The sample was composed of 60 puerperae. The information on the variables: age, education, marital status, type of childbirth, and parity were collected with a specific form; the trait and state of anxiety were based on the State Trait Anxiety Inventory (STAI/IDATE); and the level of salivary IgA was obtained through immunoturbidimetry. The application of the Mann-Whitney, Wilcoxon, and Pearson's correlation statistical tests showed a significant reduction in the levels of the state of anxiety in the experimental group (p = 0.01); there was no correlation between the trait and state variables of anxiety and the salivary IgA level; both groups (experimental and control) showed trait and state of medium-intensity anxiety.

How to Simulate the Microstructure Induced by a Nuclear Reactor with an Ion Beam Facility : DART

2009 ◽  
Vol 1215 ◽  
Author(s):  
Laurence Luneville ◽  
David Simeone ◽  
Gianguido Baldinozzi ◽  
Dominique Gosset ◽  
yves serruys
Keyword(s):  
Cross Sections ◽  
Nuclear Reactor ◽  
Nuclear Reactions ◽  
Ion Beam ◽  
Fast Neutrons ◽  
Relaxation Processes ◽  
Pressurized Water ◽  
Water Reactors ◽  
Displacement Per Atom ◽  
The Impact

AbstractEven if the Binary Collision Approximation does not take into account relaxation processes at the end of the displacement cascade, the amount of displaced atoms calculated within this framework can be used to compare damages induced by different facilities like pressurized water reactors (PWR), fast breeder reactors (FBR), high temperature reactors (HTR) and ion beam facilities on a defined material. In this paper, a formalism is presented to evaluate the displacement cross-sections pointing out the effect of the anisotropy of nuclear reactions. From this formalism, the impact of fast neutrons (with a kinetic energy En superior to 1 MeV) is accurately described. This point allows calculating accurately the displacement per atom rates as well as primary and weighted recoil spectra. Such spectra provide useful information to select masses and energies of ions to perform realistic experiments in ion beam facilities.

Experimental Characterization of Hydraulics in EPR™ Control Rod Guide Assembly: EDF and AREVA Complementary Approaches

2009 ◽  
Author(s):  
Philippe Dolleans ◽  
Thierry Muller ◽  
Benoiˆt Migot ◽  
Se´bastien Petitjean ◽  
Jean-Philippe Fontes ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Operating Conditions ◽  
Plant Design ◽  
Characterization Methods ◽  
Control Rod ◽  
Engineering Teams ◽  
Head Losses ◽  
Hydraulic Behavior ◽  
Drag Forces ◽  
The Impact

The new EPR™ Evolutionary high-Power nuclear Reactor is based on the best design concepts of French and German experiences over the past thirty years in power plant design construction and operation. In the framework of detailed engineering phase of plant construction in Finland (Olkiluoto 3) and in France (Flamanville 3), various investigations were led to improve the knowledge of hydraulics in innovative Reactor Pressure Vessel internals. A specific attention has been paid to hydraulics in Control Rod Guide Assembly (CRGA), which is an innovative component derived from the KONVOI design but adapted to the EPR™ design. In order to assess the hydraulic behavior of the EPR™ CRGA, a complementary experimental approach was developed by EDF and AREVA. Two mock-ups, representative of the full-scale EPR™ Control Rod Guide Assembly, were designed: MAGALY mock-up using real prototype of CRGA within AREVA’s engineering teams and PHACETIE mock-up at same scale but dedicated to a more analytic approach designed by EDF. The aim of the paper is to present the tests performed on two mock-ups to characterize the hydraulic behavior of this innovative EPR™ CRGA. At first, a comparison of the two mock-ups was done. The benchmark showed that the two mock-ups feature similar hydraulic behavior. The head losses evaluated along the experimental devices are practically identical. Moreover, the results showed that the global hydraulic forces applied to the Rod Cluster Control Assembly (RCCA) just up to the Upper Core Plate (UCP) are in a very good agreement. The good comparison between the two mock-ups strengthens the hydraulic characterization methods of the hydraulics in EPR™ CRGA. Specific studies were also led on the two mock-ups. The MAGALY mock-up was used to characterize the impact of the EPR™ operating conditions on the CRGA hydraulics behavior and especially on the vibrations and the drag forces applied to RCCA. Complementary measurements by using Laser Velocimetry techniques (PIV and LDV) were carried out on the PHACETIE mock-up allowing an evaluation of the velocity field inside the CRGA. The results obtained thanks to the two mock-ups allowed AREVA and EDF to have a good knowledge of the hydraulics behavior in the EPR™ CRGA. In a next step, further studies will be performed in order to deepen the knowledge of the vibratory behavior of the RCCA and the control rod drop time.

scholarly journals SAFETY ANALYSIS OF NEUTRON INTERACTION WITH MATERIAL PRACTICUM MODULE FOR THE KARTINI INTERNET REACTOR LABORATORY

2020 ◽  
Vol 22 (3) ◽  
pp. 104
Author(s):  
Prasetyo Haryo Sadewo ◽  
Puradwi Ismu Wahyono
Keyword(s):  
Radiation Protection ◽  
Nuclear Reactor ◽  
Power Level ◽  
Safety Analysis ◽  
Reactor Physics ◽  
Control Rod ◽  
Reactor Kinetics ◽  
Additional Equipment ◽  
Activation Products ◽  
And Control

Kartini Research Reactor, which is situated in Yogyakarta, is a 100 kW TRIGA (Training, Research, and Isotope Production by General Atomic)-type reactor mainly used for educational and training purposes. A system for remote learning on nuclear reactor physics named the Internet Rector Laboratory has been developed and is fully operational since 2019. To enrich its curriculum, a new practicum module has been developed, that can be immediately implemented and does not require any additional equipment or materials. To ensure safety in reactor kinetics and radiation protection, a safety analysis on the implementation of the practicum module has been conducted using MCNP and ORIGEN utilizing the current conditions of the reactor regarding its fuel burnup and control rod positions at a certain power level. Based on the results of the analysis, the practicum is safe to perform from a neutronic and radiation protection perspective. Given the long half-life and the large amount of radiation exposure that comes from activation products of iron, it is recommended that only cadmium, boron, graphite, and aluminum are allowed to be irradiated during the practicum.Keywords: Internet Reactor Laboratory, Activation Product, Radiation Protection, Reactor Safety

scholarly journals Influence of the coronavirus pandemic Covid-19 on the activities of the armed forces of the Russian Federation and foreign armies to ensure the defense of the state

2020 ◽  
Vol 2020 (3) ◽  
pp. 30-36
Author(s):  
V. Tolstykh ◽  
S. Yampolsky
Keyword(s):  
Control System ◽  
Russian Federation ◽  
Armed Forces ◽  
The State ◽  
Military Command ◽  
Coronavirus Infection ◽  
The Russian Federation ◽  
And Control ◽  
The Impact ◽  
Command And Control System

In the article, the authors disclose the impact of the crisis caused by the pandemic of the coronavirus COVID-19 on the activities of the armed forces of foreign armies, including the Armed Forces of the Russian Federation to carry out defense tasks. A detailed analysis of the ability of the armed forces to withstand a new coronavirus infection and to help neutralize it both domestically and in the international arena is presented. Separately, the issue of managing forces, means and resources under these conditions was considered and a list of additional requirements for the state and military command and control system in the interests of neutralizing the pandemic was proposed.

Analysis of Major Group Structures Used for Nuclear Reactor Simulations

2018 ◽  
Author(s):  
Marco Di Filippo ◽  
Jiri Krepel ◽  
Konstantin Mikityuk ◽  
Horst-Michael Prasser
Keyword(s):  
Cross Sections ◽  
Nuclear Reactor ◽  
Group Structure ◽  
Major Group ◽  
Computational Time ◽  
Sensitivity Threshold ◽  
Energy Structure ◽  
Reactor Spectrum ◽  
Group Structures ◽  
The Impact

Nuclear reactor simulation is often based on multi-group cross-section libraries. The structure and resolution of these libraries have a strong influence on the accuracy and computational time; hence, number of groups and energy structure must be carefully considered. The relationship between group structures and how they impact generated cross-sections can be a critical parameter. Common energy boundaries shared among major group structures were identified and the relative kinship among those was reconstructed in an effort to build a family tree of major group structures. Stochastic code Serpent2 [1] was employed to generate cross-sections of selected isotopes at different reactor compositions and conditions, using the investigated energy group structures. The impact on their generation was quantified by spectral weighted deviation. The 35 major energy structures were divided into three basic families. The key parameters distinguishing them were their applicability to thermal or fast reactors and their applicability in neutronic or multiphysics investigations. A sensitivity threshold of the generated cross-sections over the group structure resolution was investigated. The aim was to identify a group structure with very low dependency on the actual reactor spectrum.

A Probabilistic Theory for Balance Dynamics

2008 ◽  
Vol 65 (9) ◽  
pp. 2949-2960 ◽  
Author(s):  
Gregory J. Hakim
Keyword(s):  
Gravity Wave ◽  
Potential Vorticity ◽  
Joint Probability ◽  
The State ◽  
State Variables ◽  
Control Variables ◽  
Height Control ◽  
Gaussian Statistics ◽  
And Control ◽  
Balance Dynamics

Abstract Balance dynamics are proposed in a probabilistic framework, assuming that the state variables and the master, or control, variables are random variables described by continuous probability density functions. Balance inversion, defined as recovering the state variables from the control variables, is achieved through Bayes’ theorem. Balance dynamics are defined by the propagation of the joint probability of the state and control variables through the Liouville equation. Assuming Gaussian statistics, balance inversion reduces to linear regression of the state variables onto the control variables, and assuming linear dynamics, balance dynamics reduces to a Kalman filter subject to perfect observations given by the control variables. Example solutions are given for an elliptical vortex in shallow water having unity Rossby and Froude numbers, which produce an outward-propagating pulse of inertia–gravity wave activity. Applying balance inversion to the potential vorticity reveals that, because potential vorticity and divergence share well-defined patterns of covariability, the inertia–gravity wave field is recovered in addition to the vortical field. Solutions for a probabilistic balance dynamics model applied to the elliptical vortex reveal smaller errors (“imbalance”) for height control compared to potential vorticity control. Important attributes of the probabilistic balance theory include quantification of the concept of balance manifold “fuzziness,” and clear state-independent definitions of balance and imbalance in terms of the range of the probabilistic inversion operators. Moreover, the theory provides a generalization of the notion of balance that may prove useful for problems involving moist physics, chemistry, and tropical circulations.

scholarly journals Sensitivity Study on Availability of I&C Components Using Bayesian Network

2013 ◽  
Vol 2013 ◽  
pp. 1-10 ◽  
Author(s):  
Rahman Khalil Ur ◽  
Jinsoo Shin ◽  
Muhammad Zubair ◽  
Gyunyoung Heo ◽  
Hanseong Son
Keyword(s):  
Sensitivity Analysis ◽  
Bayesian Network ◽  
Circuit Breaker ◽  
Sensitivity Study ◽  
High Impact ◽  
System Risk ◽  
Pressure Transmitter ◽  
And Control ◽  
Reactor Protection System ◽  
The Impact

The objective of this study is to find out the impact of instrumentation and control (I&C) components on the availability of I&C systems in terms of sensitivity analysis using Bayesian network. The analysis has been performed on I&C architecture of reactor protection system. The analysis results would be applied to develop I&C architecture which will meet the desire reliability features and save cost. RPS architecture unavailabilityP(x=0)and availabilityP(x=1)were estimated to6.1276E-05and9.9994E-01for failure (0) and perfect (1) states, respectively. The impact of I&C components on overall system risk has been studied in terms of risk achievement worth (RAW) and risk reduction worth (RRW). It is found that circuit breaker failure (TCB), bi-stable processor (BP), sensor transmitter (TR), and pressure transmitter (PT) have high impact on risk. The study concludes and recommends that circuit breaker bi-stable processor should be given more consideration while designing I&C architecture.

Detection and Sizing of Defects in Control Rod Drive Mechanism Penetrations Using Eddy Current and Ultrasonics

1996 ◽  
Vol 118 (3) ◽  
pp. 301-307 ◽  
Author(s):  
G. M. Light ◽  
J. L Fisher ◽  
R. F. Tennis ◽  
J. S. Stolte ◽  
G. J. Hendrix
Keyword(s):  
Defect Detection ◽  
Eddy Current ◽  
Nuclear Reactor ◽  
Region Of Interest ◽  
Pressure Vessels ◽  
Ultrasonic Transducers ◽  
Detection Sensitivity ◽  
Inside Surface ◽  
Control Rod ◽  
And Control

Over the last two years, concern has been generated about the capabilities of performing nondestructive evaluation (NDE) of the closure-head penetrations in nuclear-reactor pressure vessels. These penetrations are primarily for instrumentation and control rod drive mechanisms (CRDMs) and are usually thick-walled Inconel tubes, which are shrink-fttted into the steel closure head. The penetrations are then welded between the outside surface of the penetration and the inside surface of the closure head. Stress corrosion cracks initiating at the inner surface of the penetration have been reported at several plants. Through-wall cracks in the CRDM penetration or CRDM weld could lead to loss of coolant in the reactor vessel. The CRDM penetration presents a complex inspection geometry for conventional NDE techniques. A thermal sleeve, through which pass the mechanical linkages for operating the control rods, is inserted into the penetration in such a way that only a small annulus (nominally 3 mm) exists between the thermal sleeve and inside surface of the penetration. Ultrasonic (UT) and eddy current testing (ET) techniques that could be used to provide defect detection and sizing capability were investigated. NDE techniques that could be applied inside and outside the annulus were investigated, but the major goal was to provide high defect-detection sensitivity without requiring removal of the thermal sleeve. As a result of this investigation, both ET and UT techniques for the CRDM penetrations have been developed and evaluated. Long, thin probes were designed to fit into the annulus to carry both eddy current coils and irrigated ultrasonic transducers into the region of interest. The eddy current probes were used primarily to detect cracks in the penetration while the ultrasonic transducers were used to provide an estimate of the remaining wall thickness. This paper describes the ET and UT techniques, the probes developed, and the results obtained using these probes and techniques on CRDM penetration mock-ups.

Sign in / Sign up

Export Citation Format

Share Document