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

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.

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

2006 ◽  
Author(s):  
A. Gorzel
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressure Increase ◽  
Thermal Hydraulics ◽  
Short Term ◽  
Speed Reduction ◽  
Pump Speed ◽  
Transient Behaviour ◽  
Limit Value

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and impermissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second — much smaller — maximum that would occur around one second after the first one in the absence of a SCRAM.

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.

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.

Investigation of Fan Rotor Interaction With Pressure Disturbance Produced by Downstream Pylon

2013 ◽  
Author(s):  
Tomonori Enoki ◽  
Hidekazu Kodama ◽  
Shinya Kusuda
Keyword(s):  
Flow Field ◽  
Three Dimensional ◽  
Numerical Results ◽  
Rotor Blades ◽  
Flow Structures ◽  
Coupled Analysis ◽  
Unsteady Forces ◽  
Rotor Stator Interaction ◽  
The Impact ◽  
Potential Pressure

This paper presents an investigation of fan rotor interaction with potential pressure disturbances produced by a downstream pylon. Three-dimensional unsteady viscous analyses are performed for two fan rotor-stator-pylon configurations with different axial gaps between the stator and the pylon, and compared with the experimental results. To clarify the impact of the rotor-pylon interaction on the potential pressure flow field, a numerical analysis for the configuration in which a fan rotor is removed is also performed and compared with the numerical results with fan rotor. Actuator disk analyses are also performed to interpret the flow structures observed in the experiments and the numerical results. It is found that a fan rotor-stator interaction also exists in the fan flow field, and this may impact on the upstream propagating potential flow that dominates the unsteady forces acting on the rotor blades. A coupled analysis between fan rotor and stator is essential to accurately predict the unsteady blade force.

The Reliability Prediction of HTR’s Graphite Component in Various Temperature and Neutron Dose Levels

2013 ◽  
Author(s):  
Xiang Fang ◽  
Haitao Wang ◽  
Xingtuan Yang ◽  
Suyuan Yu
Keyword(s):  
High Temperature ◽  
Neutron Irradiation ◽  
Three Dimensional ◽  
Reactor Core ◽  
Neutron Dose ◽  
Element Analysis ◽  
The Core ◽  
Three Dimensional Finite Element ◽  
Dose Levels ◽  
The Impact

In high temperature gas-cooled reactors (HTRs), graphite is used as the main structure material. The side reflecter of the reactor core is composed by a pile of graphite bricks. In real operational condition of the reactor, both high temperature and fast neutron irradiation have great effect on the behavior of graphite components. The non-uniform distribution of temperature and neutron dose cause obvious stress accumulation, which greatly affects the security and reliability of the graphite components. In addition, high temperature and neutron irradiation make the properties of graphite change in evidence, and the changes are not linear. Such changes must be considered and simulated in the calculation, in order to predict the stress concentration condition and the reliability of the graphite brick correctly. A FORTRAN code based on user subroutines of MSC.MARC is developed in INET in order to perform three-dimensional finite element analysis of irradiated behavior of the graphite components for the HTRs. In this paper, the stress level and failure probability of graphite components are calculated and obtained under different in-core temperatures and neutron dose levels of the core side of brick. 400°C, 500°C, 600°C and 700°C are selected as the core side temperature, while the range of neutron dose is 0 to 1022n cm-2 (EDN). Different constitutive laws are used in stress analysis procedure. The impact of different temperature and neutron dose levels are discussed.

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.

A Simplified Two-Group Multipoint Kinetics Model

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Santosh K. Pradhan ◽  
K. Obaidurrahman ◽  
Kannan N. Iyer
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Reactor Core ◽  
Coupled System ◽  
Kinetics Model ◽  
Primary Circuit ◽  
Thermal Hydraulics ◽  
Time Step ◽  
Large Power

Abstract Detailed multiphysics modeling of nuclear power plants has become a necessity in the era of best-estimate analysis. For a number of transients with strong coupling between the neutronics in the reactor core and the fluid-dynamics in the primary circuit and overall heat transfer, it is required to carry out coupled system thermal hydraulics and core three-dimensional (3D) neutronics analysis. Point kinetics approach in the system thermal-hydraulics (TH) code RELAP5 limits its use for many reactivity-induced transients, which involve asymmetric core behavior. In a recent development, a simplified multipoint kinetics model has been coupled with system TH code RELAP5 to circumvent its inadequacy for the analysis of reactivity-induced transients involving asymmetric core behavior. The objective of this paper is to validate the simplified multipoint kinetics model against an asymmetric fast transient benchmark problem in a large power reactor. Time-step and nodalization sensitivity studies have been performed. It is demonstrated that the multipoint kinetics model results are in good agreement with the benchmark, advocating its applicability.

scholarly journals The Best Estimate Plus Uncertainty approach in nuclear reactor safety and licensing: Brief history and the elements after licensing

2019 ◽  
Vol 34 (3) ◽  
pp. 299-312
Author(s):  
Francesco D’Auria ◽  
Giorgio Galassi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Thermal Hydraulics ◽  
Reactor Safety ◽  
Two Phase ◽  
Best Estimate ◽  
Nuclear Reactor Safety ◽  
The Impact

The best estimate plus uncertainty is, at the same time, an approach, a procedure and a frame- work in nuclear thermal-hydraulics and nuclear reactor safety and licensing. The motivation at the basis of the best estimate plus uncertainty is the lack of knowledge in the areas of single and, mainly, two-phase transient thermal-hydraulics. In other terms and introducing some simplifications, the insufficient knowledge of turbulence imposes the design of roadmaps for the application of imperfect (thermal-hydraulic) models to the evaluation of design features and of safety for complex technological installations or systems like the nuclear power plants and, more specifically, the water cooled nuclear reactors. Furthermore, the legal counterpart of nuclear reactor safety, or the licensing, is concerned: therefore the best estimate plus uncertainty must account for rules and regulations derived from the fundamental radioprotection principle which imposes the minimization of the impact of radiations upon humans and the environment under any circumstance. In the present paper, the key elements of the approach are identified and characterized. These shall be seen as the support for a consistent application of thermal-hydraulics to the design and safety of water-cooled nuclear reactors.

Sign in / Sign up

Export Citation Format

Share Document