scholarly journals NEUTRON NOISE PATTERNS FROM COUPLED FUEL-ASSEMBLY VIBRATIONS

2021 ◽  
Vol 247 ◽  
pp. 02015
Author(s):  
M. Viebach ◽  
C. Lange ◽  
M. Seidl ◽  
Y. Bilodid ◽  
A. Hurtado
Keyword(s):  
Neutron Flux ◽  
Fuel Assembly ◽  
Cross Sections ◽  
Three Dimensional ◽  
Time Dependent ◽  
Radial Dependence ◽  
Feedback Parameter ◽  
Fuel Assemblies ◽  
Neutron Noise ◽  
The Impact

The neutron flux fluctuation magnitude of KWU-built PWRs shows a hitherto unexplained correlation with the types of loaded fuel assemblies. Also, certain measured long-range neutron flux fluctuation patterns in neighboring core quadrants still lack a closed understanding of their origin. The explanation of these phenomena has recently revived a new interest in neutron noise research. The contribution at hand investigates the idea that a synchronized coolant-driven vibration of major parts of the fuel-assembly ensemble leads to these phenomena. Starting with an assumed mode of such collective vibration, the resulting effects on the time-dependent neutron-flux distribution are analyzed via a DYN3D simulation. A three-dimensional representation of the time-dependent bow of all fuel assemblies is taken into account as a nodal DYN3D feedback parameter by time-dependent variations of the fuel-assembly pitch. The impact of its variation on the cross sections is quantified using a cross-section library that is generated from the output of corresponding CASMO5 calculations. The DYN3D simulation qualitatively reproduces the measured neutron-flux fluctuation patterns. The magnitude of the fluctuations and its radial dependence are comparable to the measured details. The results imply that collective fuel-assembly vibrations are a promising candidate for being the key to understand long-known fluctuation patterns in KWU built PWRs. Further research should elaborate on possible excitation mechanisms of the assumed vibration modes.

scholarly journals MODELLING AND ANALYSIS OF FUEL ASSEMBLY VIBRATIONAL MODES IN PWRS USING SIMULATE-3K

2021 ◽  
Vol 247 ◽  
pp. 21008
Author(s):  
V. Verma ◽  
D. Chionis ◽  
A. Dokhane ◽  
H. Ferroukhi
Keyword(s):  
Cross Sections ◽  
Natural Oscillation ◽  
Time Dependent ◽  
Reactor Dynamics ◽  
Vibration Model ◽  
Oscillation Modes ◽  
Fuel Assemblies ◽  
Neutron Noise ◽  
Vibration Pattern ◽  
Mode Vibration

Some of the KWU pre-KONVOI PWRs operating across Europe saw a systematic increase in the neutron noise levels over several cycles in the last decade, and subsequently, core internals’ movements, especially vibrations of fuel assemblies with specific designs were identified as one of the plausible causes. Therefore, it is important to develop computational methods that can allow to investigate and predict the reactor noise response to fuel assemblies vibrations. To this aim, the 3D nodal reactor dynamics code SIMULATE-3K is used at PSI with a special module called the ‘assembly vibration model’ that imitates time-dependent motions of fuel assemblies by dynamically modifying the water-gaps surrounding the laterally moving fuel assemblies. The varying water-gaps are represented by the variation in the corresponding two-group macroscopic cross sections generated using the lattice code CASMO-5 in 2D. The studies conducted so far to assess the methodology for full core noise simulations were based on assuming vibrations of a clamped-free cluster of fuel assemblies that are unsupported from both ends. However, as this represents a non-physical movement, further developments were made at PSI to allow simulating more realistic movements of fuel assemblies such as the cantilevered mode vibration. The updated methodology, along with evaluations of the simulated noise response to realistic vibration modes, is presented in this paper. Results show that, as expected, the radial and axial neutron noise behaviour follow the vibration pattern of the imposed time-dependent axial functions corresponding to the natural oscillation modes of the fuel assemblies, thereby providing confidence in the application of the developed methodology for numerical neutron noise analyses of the PWR cores.

scholarly journals Analysis of the Accelerator-Driven System Fuel Assembly during the Steam Generator Tube Rupture Accident

Materials ◽  
2021 ◽  
Vol 14 (8) ◽  
pp. 1818
Author(s):  
Di-Si Wang ◽  
Bo Liu ◽  
Sheng Yang ◽  
Bin Xi ◽  
Long Gu ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Steam Generator ◽  
Metal Flow ◽  
Maximum Temperature ◽  
Steam Generator Tube ◽  
Driven System ◽  
Fuel Assemblies ◽  
Accelerator Driven System ◽  
The Impact ◽  
Steam Bubble

China is developing an ADS (Accelerator-Driven System) research device named the China initiative accelerator-driven system (CiADS). When performing a safety analysis of this new proposed design, the core behavior during the steam generator tube rupture (SGTR) accident has to be investigated. The purpose of our research in this paper is to investigate the impact from different heating conditions and inlet steam contents on steam bubble and coolant temperature distributions in ADS fuel assemblies during a postulated SGTR accident by performing necessary computational fluid dynamics (CFD) simulations. In this research, the open source CFD calculation software OpenFOAM, together with the two-phase VOF (Volume of Fluid) model were used to simulate the steam bubble behavior in heavy liquid metal flow. The model was validated with experimental results published in the open literature. Based on our simulation results, it can be noticed that steam bubbles will accumulate at the periphery region of fuel assemblies, and the maximum temperature in fuel assembly will not overwhelm its working limit during the postulated SGTR accident when the steam content at assembly inlet is less than 15%.

Water Entry of a Wedge by Hydroelastic Orthotropic Plate Theory

1999 ◽  
Vol 43 (03) ◽  
pp. 180-193 ◽  
Author(s):  
Odd M. Faltinsen
Keyword(s):  
Impact Velocity ◽  
Cross Sections ◽  
Orthotropic Plate ◽  
Plate Theory ◽  
Three Dimensional ◽  
Water Entry ◽  
Natural Period ◽  
Cross Sectional ◽  
Flow Effects ◽  
The Impact

Water entry of a hull with wedge-shaped cross sections is analyzed. The stiffened platings between two transverse girders on each side of the keel are separately modeled. Orthotropic plate theory is used. The effect of structural vibrations on the fluid flow is incorporated by solving the two-dimensional Laplace equation in the cross-sectional fluid domain by a generalized Wagner's theory. The coupling with the plate theory provides three-dimensional flow effects. The theory is validated by comparison with full-scale experiments and drop tests. The importance of global ship accelerations is pointed out. Hydrodynamic and structural error sources are discussed. Systematic studies on the importance of hydroelasticity as a function of deadrise angle and impact velocity are presented. This can be related to the ratio between the wetting time of the structure and the greatest wet natural period of the stiffened plating. This ratio is proportional to the deadrise angle and inversely proportional to the impact velocity. A small ratio-means that hydroelasticity is important and a large ratio means that hydroelasticity is not important.

scholarly journals possis: predicting spectra, light curves, and polarization for multidimensional models of supernovae and kilonovae

2019 ◽  
Vol 489 (4) ◽  
pp. 5037-5045 ◽  
Author(s):  
M Bulla
Keyword(s):  
Radiation Transport ◽  
Three Dimensional ◽  
Polar Axis ◽  
Light Curves ◽  
Time Dependent ◽  
Opening Angle ◽  
Monte Carlo Code ◽  
Lanthanide Elements ◽  
The Impact ◽  
Multidimensional Models

ABSTRACT We present possis, a time-dependent three-dimensional Monte Carlo code for modelling radiation transport in supernovae and kilonovae. The code incorporates wavelength- and time-dependent opacities, and predicts viewing-angle dependent spectra, light curves, and polarization for both idealized and hydrodynamical explosion models. We apply the code to a kilonova model with two distinct ejecta components, one including lanthanide elements with relatively high opacities and the other devoid of lanthanides and characterized by lower opacities. We find that a model with total ejecta mass $M_\mathrm{ej}=0.04\, \mathrm{M}_\odot$ and half-opening angle of the lanthanide-rich component Φ = 30° provides a good match to GW 170817/AT 2017gfo for orientations near the polar axis (i.e. for a system viewed close to face-on). We then show how crucial is the use of self-consistent multidimensional models in place of combining one-dimensional models to infer important parameters, such as the ejecta masses. We finally explore the impact of Mej and Φ on the synthetic observables and highlight how the relatively fast computation times of possis make it well-suited to perform parameter-space studies and extract key properties of supernovae and kilonovae. Spectra calculated with possis in this and future studies will be made publicly available.

scholarly journals Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

Nukleonika ◽  
2015 ◽  
Vol 60 (3) ◽  
pp. 531-536 ◽  
Author(s):  
Igor P. Królikowski ◽  
Jerzy Cetnar
Keyword(s):  
Cross Sections ◽  
Nuclear Reactions ◽  
Nuclear Reactors ◽  
Input Parameter ◽  
Three Dimensional ◽  
Reactor Core ◽  
Coupled Analysis ◽  
Thermal Hydraulics ◽  
Power Profile ◽  
The Impact

Abstract Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD) code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

The influence of designs of protective uniforms on firefighters’ performance during moderate physical exercises

2017 ◽  
Vol 88 (17) ◽  
pp. 1979-1991 ◽  
Author(s):  
Izabela Ciesielska-Wróbel ◽  
Emiel DenHartog ◽  
Roger Barker
Keyword(s):  
Cross Sections ◽  
Three Dimensional ◽  
Physiological Parameters ◽  
Subjective Perception ◽  
Body Scanning ◽  
Physical Exercises ◽  
Subjective Assessments ◽  
Movement Performance ◽  
Physiological Tests ◽  
The Impact

The aim of this study was to verify whether the minor differences in the design of uniforms and their fit can be quantified in terms of their impact on firefighters’ cardiorespiratory parameters and subjective perception of these uniforms. The impact of minor design improvements compared to the existing designs of personal protective clothing (PPC) is still relatively difficult to quantify due to the lack of sensitive devices used in smart measuring methodologies; however, the perception of these slight differences is reported by PPC users. The impact of these design differences in PPC on firefighters was studied via physiological tests based on occupation-related activities in which cardiorespiratory parameters were monitored and three-dimensional (3D) silhouette scanning was performed on the firefighters. Apart from heart rate (beats/min), none of the other measured physiological parameters, for example, oxygen consumption (VO2, ml/min) demonstrated statistically significant differences when firefighters were testing uniforms: ergonomic (ER), standard (ST), bulky (BU), and reference outfit (RO), the latter being T-shirt and shorts. A statistically significant correlation was found between parameters measured via 3D body scanning and selected cross-sections of the silhouettes as well as subjective assessments of easiness of specific movement performance during the physiological test and assessment of bulkiness of the uniforms. There is a limited influence of the minor design differences between firefighters’ uniforms on the selected physiological parameters of the subjects wearing them. The outcome of the study can be utilized when performing the test on subjects and improving designs of PPC.

Validation of HELIOS Neutron Cross-Section Library for RBMK Reactors Against the Data From the Critical Facility Experiments

2002 ◽  
Author(s):  
Audrius Jasiulevicius ◽  
Bal Raj Sehgal
Keyword(s):  
Neutron Flux ◽  
Cross Section ◽  
Cross Sections ◽  
Measurement Errors ◽  
Reactor Core ◽  
Rbmk Reactor ◽  
Cross Section Data ◽  
Section Data ◽  
Fuel Assemblies ◽  
Calculation Results

The RBMK reactors are channel type, water-cooled and graphite moderated reactors. The first RBMK type electricity production reactor was put on-line in 1973. Currently there are 13 operating reactors of this type. Two of the RBMK-1500 reactors are at the Ignalina NPP in Lithuania. Experimental Critical Facility for RBMK reactors, located at Kurchiatov Institute, Moscow was designed to carry out critical reactivity experiments on assemblies, which imitate parts of the RBMK reactor core. The facility is composed of Control and Protection Rods (CPR’s), fuel assemblies with different enrichment in U-235 and other elements, typical for RBMK reactor core loadings, e.g. additional absorber assemblies, CPR imitators, etc. A simulation of a set of the experiments, performed at the Experimental Critical Facility, was carried out at the Royal Institute of Technology (RIT), Nuclear Power Safety Division, using CORETRAN 3-D neutron dynamics code. The neutron cross sections for assemblies were calculated using HELIOS code. The aim of this work was to evaluate capabilities of the HELIOS code to provide correct cross section data for the RBMK reactor. The calculation results were compared to the similar CORETRAN calculations, when employing WIMS-D4 code generated cross section data. For some of the experiments, where calculation results with CASMO-4 code generated cross sections are available, the comparison is also performed against CASMO-4 results. Eleven different experiments were simulated. Experiments differ in size of the facility core (number of assemblies loaded): from simple core loadings, composed only of a few fuel assemblies, to complicated configurations, which represent a part of the RBMK reactor core. Diverse types of measurements were carried out during these experiments: reactivity, neutron flux distributions (both axial and radial), rod reactivity worth and the voiding effects. Results of the reactivity measurements and relative neutron flux distributions were given in the Experiment report [1] as parameters, to be obtained using static calculations, i.e. the reported results were already processed numerically using the facility equipment, e.g. the reactimeter. The reported measurement errors consist only of instrumentation errors, i.e. measurement method errors and the influence from the space–time effects were not included in the error evaluation.

scholarly journals Comparative Analysis of Isotopic Composition of VVER-1000 Westinghouse and TVEL Spent Fuel

2016 ◽  
pp. 33-37
Author(s):  
Yu. Kovbasenko
Keyword(s):  
Comparative Analysis ◽  
Isotopic Composition ◽  
Boric Acid ◽  
Fuel Assembly ◽  
Spent Fuel ◽  
Operational Conditions ◽  
Manufacturing Tolerance ◽  
Fuel Assemblies ◽  
The Impact

The paper considers the impact of different operational conditions on VVER-1000 spent fuel isotopic composition. Such operational conditions include the presence or absence of absorber rods in guide tubes of fuel assemblies, changing the concentration of boric acid dissolved in the moderator (water) during the campaign, fuel and/or moderator temperature. Moreover, the impact caused by manufacturing tolerance applied while manufacturing fuel assembly was analyzed by weight of fuel and by its enrichment. Calculations were made for reactor cells of fuel assemblies for VVER-1000. They were composed of the new fuel assemblies of Westinghouse Company and the typical fuel assemblies of Russian TVEL suppliers.

scholarly journals ADVANCED HOMOGENIZATION METHODS FOR PRESSURIZED WATER REACTORS

2018 ◽  
Vol 19 ◽  
pp. 14 ◽  
Author(s):  
Pavel Suk
Keyword(s):  
Fuel Assembly ◽  
Cross Sections ◽  
Fuel Loading ◽  
Pressurized Water ◽  
Testing Method ◽  
Infinite Lattice ◽  
Fuel Assemblies ◽  
Macroscopic Cross Section ◽  
Water Reactors ◽  
Homogenization Methods

Macroscopic cross section generation is key part of core calculation. Commonly, the data are prepared independently without a knowledge of fuel loading pattern. The fuel assemblies are simulated in infinite lattice (with mirror boundary conditions). Rehomogenization method is based on combination of actual neutron flux in fuel assembly with macroscopic data from infinite lattice. Rehomogenization method was implemented into the macrocode Andrea and tested on a reference cases. Cases consist of fuel cases, cases with strong absorber, cases with absorption rods, or cases with reflector assemblies. Testing method is based on a comparisons of homogenized and rehomogenized macroscopic cross sections and later on a comparisons of relative power of each fuel assembly. Above that there is comparison of eigenvalue.

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