Core Mechanical Dynamics Experiment in the Phénix Reactor

2015 ◽  
Author(s):  
Daniel Broc ◽  
Jérome Cardolaccia ◽  
Laurent Martin ◽  
Jean Louis Portier
Keyword(s):  
Dynamic Behavior ◽  
Fast Reactor ◽  
Early Stage ◽  
Response Characteristic ◽  
Point Of View ◽  
Primary Coolant ◽  
The Core ◽  
Safety Studies ◽  
Numerical Tools ◽  
Dynamic Response Characteristic

ASTRID is a project for an industrial prototype of a 600 MWe sodium cooled Fast Reactor, led by CEA. A consequent program is in progress for the development and the validation of numerical tools for the simulation of the dynamic mechanical behavior of the Fast Reactor cores, with both experimental and numerical parts. The cores are constituted of Fuel Assemblies (or FA) and Neutronic Shields (or NS) immersed in the primary coolant (sodium), which circulates inside the Fluid Assemblies. The FA and the NS are slender structures, which may be considered as beams, from a mechanical point of view. The dynamic behavior of this system has to be understood, for design and safety studies. Two main movements have to be considered: global horizontal movements under the effect of a seismic excitation, and a radial opening of the core. The fluid presence leads to complex interactions between the structures at a distance. The dynamic behavior of the core is strongly influenced by contacts between the beams and by the interactions with the sodium, which both limit their relative displacements. Numerical methods and models are built to describe and simulate this dynamic behavior. The validation of the numerical tools is based on the results of different experimental programs, already performed or in progress. The paper presents the interpretation of tests performed in 2013 in the Phénix reactor. The French Phénix reactor was definitively shutdown in 2009 and is currently at an early stage of the decommissioning process. Before unloading the core, it has been decided to perform one last experimental campaign aimed at testing the mechanical dynamic behavior of the core. The interpretation of the tests highly contributes to the validation of the simulation methods. Relatively good comparisons have been obtained between the theoretical and experimental results, for the static excitation (stiffness of the bundle) and for the dynamic response (characteristic times). The tests confirm that the fluid leads to a significant decrease of the frequencies. Uncertainties remain on the significant damping which seems to be present, and may be due to the fluid or to the structures.

Physical and Numerical Methods for the Dynamic Behavior of the Fast Reactor Cores

2014 ◽  
Author(s):  
Daniel Broc ◽  
Jérome Cardolaccia ◽  
Laurent Martin
Keyword(s):  
Numerical Methods ◽  
Dynamic Behavior ◽  
Fast Reactor ◽  
Computer Code ◽  
Point Of View ◽  
Liquid Sodium ◽  
Primary Coolant ◽  
The Core ◽  
Safety Studies ◽  
Reactor Cores

In the frame of the GEN IV Forum and of the ASTRID Project, a program is in progress in the CEA (France) for the development and the validation of numerical tools for the simulation of the dynamic mechanical behavior of the Fast Reactor cores, with both experimental and numerical parts. The cores are constituted of Fuel Assemblies (of FA) and Neutronic Shields (or NS) immersed in the primary coolant (sodium), which circulates inside the Fluid Assemblies. The FA and the NS are slender structures, which may be considered as beams, form a mechanical point of view. The dynamic behavior of this system has to be understood, for design and safety studies. Two main movements have to be considered: global horizontal movements under a seismic excitation, and opening of the core. The dynamic behavior of the core is strongly influenced by contacts between the beams and by the sodium. The contacts between the beams limit the relative displacements. The fluid leads to complex interactions between the structures in the whole core. The paper presents the physical and numerical methods and tools used to describe and simulate the phenomena. A key point is the Fluid Structure Interaction (or FSI): the interactions between the beams and the liquid sodium. The fluid movement is assumed to be described by the equations of a perfect fluid. Simple and efficient homogenization methods may be used to reduce the size of the problem. These methods are integrated in a general computer code, CAST3M developed at the CEA Saclay. This computer code allows to take into account the impacts between the beams. Some applications are presented.

Fluid Structure Interaction for Tubes Bundles: Different Homogenization Methods

2017 ◽  
Author(s):  
Daniel Broc ◽  
Gianluca Artini
Keyword(s):  
Fluid Flow ◽  
Dynamic Behavior ◽  
Fast Reactor ◽  
Free Fluid ◽  
Inertial Effects ◽  
The Core ◽  
Dissipative Effects ◽  
Numerical Tools ◽  
Reactor Cores ◽  
Theoretical Analyses

ASTRID is a project for an industrial prototype of a 600 MWe sodium cooled Fast Reactor, led by CEA. An important program is in progress for the development and the validation of numerical tools for the simulation of the dynamic mechanical behavior of the Fast Reactor cores, with both experimental and numerical parts. The cores are constituted of Fuel Assemblies (of FA) and Neutronic Shields (or NS) immersed in the primary coolant (sodium), which circulates inside the Fluid Assemblies. The FA and the NS are slender structures, which may be considered as beams, form a mechanical point of view. The dynamic behavior of this system has to be understood, for design and safety studies. Two main movements have to be considered: global horizontal movements under a seismic excitation, and opening of the core. The fluid leads to complex interactions between the structures in the whole core. The dynamic behavior of the core is also strongly influenced by contacts between the beams and by the sodium, which limit the relative displacements. Numerical methods and models are built to describe and simulate this dynamic behavior. The validation of the numerical tools is based on the results of different experimental programs, already performed or in progress. The paper is mainly devoted to the modeling of the Fluid Structure Interaction phenomena in the Fast Reactor cores. Tubes bundles immersed in a dense fluid are very common in the nuclear industry (reactor cores and steam generators). In the case of an external excitation (earthquake or shock) the presence of the fluid leads to “inertial effects” with lower natural frequencies, and “dissipative effects”, with higher damping. The geometry of a tubes bundle is complex, which may lead to very huge sizes for the numerical models. Many works have been made during the last decades to develop homogenization, in order to simplify the problem. Theoretical analyses are presented on different simplifications and assumptions which can be made in the homogenization approach. The accuracy of the different assumptions depends of the conditions of the system: fluid flow or fluid at rest, small or large displacements of the structure. In the general case, it is theoretically necessary to consider the Navier Stokes equations: the fluid flow is fully nonlinear. Models have been developed during the last years, based on the Euler linear equations, corresponding to a fluid at rest, with small displacements of the structure. Only the inertial effects are theoretically described but the dissipative effects may be taken into account by using a Rayleigh damping. Different theoretical analyses show that, even in the case of a nonlinear fluid flow, the linear potential flow models may be used as linear equivalent models. In the cases with an important head loss in the fluid flow through the tubes, the fluid movement is mainly driven by the important forces exchanged with the structure and by the pressure gradient. The global equations of the system are close to the equations used for porous media, like the Darcy equations. An important condition to get a relevant model is to describe globally the energy balance in the system. The energy given to the fluid by the solid correspond to a variation of kinetic energy in the fluid and to energy dissipation in the fluid. Attention will be paid to the cases where the tubes bundle is in interaction with free fluid, without tubes. The global equation of the system has to be accurate for the tubes bundle and for the free fluid also.

LBLOCA Analysis of CPR1000 NPP With Advanced Accumulator

2013 ◽  
Author(s):  
Hongwei Hu ◽  
Jianqiang Shan ◽  
Junli Gou ◽  
Bo Zhang ◽  
Haitao Wang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Low Pressure ◽  
Point Of View ◽  
Primary Coolant ◽  
Pressure Drops ◽  
The Core ◽  
Injection Pump ◽  
Model Combining ◽  
Large Flow

Large break LOCA (LBLOCA) is one of the limit design basic accidents in nuclear power plant. The large flow water in the advanced accumulator is injected into primary loop in early short time. When the vessel pressure drops and reactor core is re-flooded, the advanced accumulator provides a small injection flow to keep the reactor core in flooded condition. Thus, the startup grace time of the low pressure safety injection pump is extended, and the core still stays in a long-term cooling state. By deducing the original accumulator model in RELAP5 accident analysis code, a new model combining the advanced and the traditional accumulator is obtained and coupled into RELAP5/ MOD 3.3. Simulation results show that there is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, resulting in a large pressure drop and small flow. The phenomenon meets the demand of the advanced accumulator design and the simulation of the advanced accumulator is accomplished successfully. Based on this, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 is analyzed with the modified RELAP5 code. When the double ended cold leg guillotine accident with 200s delayed startup of the low pressure safety injection occurs, maximum cladding temperature in the core with traditional accumulator is 1860K which seriously exceeded the safety temperature (1477K)[1] prescribed limits while the maximum cladding temperature with advanced accumulator has the security temperature-1277K. From this point of view, adopting passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, security and economy of reactor system were improved.

Scoping Studies for a Small Modular Lead-Cooled Fast Reactor

2004 ◽  
Author(s):  
Chenggang Yu ◽  
Michael A. Smith ◽  
Earl E. Feldman ◽  
Won Sik Yang ◽  
James J. Sienicki
Keyword(s):  
Carbon Dioxide ◽  
Fast Reactor ◽  
Natural Circulation ◽  
Power Conversion ◽  
Power Converter ◽  
Working Fluid ◽  
Brayton Cycle ◽  
Primary Coolant ◽  
The Core ◽  
Scoping Studies

A scoping design study has been carried out of the feasibility of a small, 25 MWt (∼10 MWe), modular lead-cooled fast reactor coupled to an advanced power converter consisting of a gas turbine Brayton cycle that utilizes supercritical carbon dioxide as the working fluid. Major constraints of the study are an ultralong 20 year core lifetime, near zero reactivity burnup swing over the core lifetime, Pb primary coolant natural circulation heat transport, road transportability of plant modular assemblies including the reactor and guard vessels, and high Brayton cycle power conversion efficiency. It is found that the goal of a near zero reactivity burnup swing implies a low core power density that results in an unacceptably low discharge burnup.

Concept of Pressurized Water Lead-Bismuth-Cooled Fast Reactor (PLFR)

2014 ◽  
Author(s):  
Minoru Takahashi
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Point Of View ◽  
Steam Generators ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Lead Bismuth Eutectic ◽  
Innovative Concept

The innovative concept of pressurized water lead-bismuth-cooled fast reactor (PLFR) has been proposed and studied based on the previous LFR concept: PBWFR. Primary pumps and steam generators that contact lead-bismuth coolant are eliminated. A feedwater is directly injected into the primary coolant of hot lead-bismuth eutectic (LBE) at the outlet of the reactor core under the pressure of 14 MPa as PWR. The specifications of PLFR system are discussed and presented from thermal-hydraulic point of view.

Fast Reactor Cores: Seismic Excitation in the Vertical Direction — Numerical Methods

2018 ◽  
Author(s):  
Daniel Broc ◽  
Gianluca Artini ◽  
Jérome Cardolaccia ◽  
Laurent Martin
Keyword(s):  
Numerical Methods ◽  
Dynamic Behavior ◽  
Nonlinear Model ◽  
Fast Reactor ◽  
Vertical Direction ◽  
Seismic Excitation ◽  
Fluid Structure ◽  
The Core ◽  
Fuel Assemblies ◽  
Reactor Cores

In the frame of the GEN IV Forum and of the ASTRID Project, a program is in progress in the CEA (France) for the development and the validation of numerical tools for the simulation of the dynamic mechanical behavior of the Fast Reactor cores, with both experimental and numerical parts. The cores are constituted of Fuel Assemblies (or FA) and Neutronic Shields (or NS) immersed in the primary coolant (sodium), which circulates inside the Fuel Assemblies. The FA and the NS are slender structures, inserted in a grid plate, which may be considered as beams form a mechanical point of view. The dynamic behavior of this system has to be understood, for design and safety studies. This dynamic behavior of the core is strongly influenced by the sodium and by contacts between the beams at the pads level and at the top. The fluid leads to complex interactions between the structures in the whole core. The contacts between the beams limit the relative displacements. Two main movements have been considered so far: global horizontal movements under a seismic excitation, and opening of the core. Physical and numerical methods and tools have been developed to describe and simulate the dynamic behavior. These methods are integrated in CAST3M, general computer code developed at the CEA Saclay. The assemblies are modeled as beams. The impacts at the pads between the assemblies are taken into account by using a nonlinear model. The Fluid Structure Interaction is taken into account by using homogenization methods. This paper is devoted to the improvement of these methods to take into account the vertical component of a seismic excitation. The key points are: - the fluid structure coupling in the vertical direction, - the modification of the description of the impacts to take into account the vertical displacements of the assemblies, - the modification of the boundary condition at the foot of the assembly, in order to take into account the uplift with a nonlinear model.

Fluid Structure Interaction for Tubes Bundles: Presentation of a Linear Equivalent Model

2017 ◽  
Author(s):  
Daniel Broc ◽  
Gianluca Artini
Keyword(s):  
Dynamic Behavior ◽  
Fast Reactor ◽  
Stokes Equations ◽  
Nuclear Industry ◽  
Equivalent Model ◽  
Free Fluid ◽  
Inertial Effects ◽  
Primary Coolant ◽  
Dissipative Effects ◽  
Reactor Cores

ASTRID is a project for an industrial prototype of a 600 MWe sodium cooled Fast Reactor, led by CEA. An important program is in progress for the development and the validation of numerical tools for the simulation of the dynamic mechanical behavior of the Fast Reactor cores, with both experimental and numerical parts. The cores are constituted of Fuel Assemblies (of FA) and Neutronic Shields (or NS) immersed in the primary coolant (sodium), which circulates inside the Fluid Assemblies. The FA and the NS are slender structures, which may be considered as beams, form a mechanical point of view. The analysis of the dynamic behavior of tubes bundles immersed in a dense fluid is a major challenge in the nuclear industry (reactor cores, steam generators). In some cases, the excitation is given by the fluid flow, with a complex behavior which may lead to instabilities. The paper only considers the case of an external excitation (earthquake or shock). The fluid leads to two main effects: “inertial effects” with lower vibration frequencies and “dissipative effects” with a higher damping. In the general case the fluid has to be described by the Navier-Stokes equations. It is possible to use the Euler linear equations in the case of vibrations of the tubes in a globally stagnant fluid. In all cases the modeling of the system could lead to huge numerical problems if each tube is described explicitly. Homogenization technics allow to limit the size of the problem. Homogenization methods taking into account the Euler equations for the fluid have been developed, and widely used for analyses of the dynamic behavior of reactor cores. Only the inertial effects are theoretically described but the dissipative effects may be roughly taken into account by using a Rayleigh damping. The paper presents an improvement of the method, allowing a better description of the dissipative effects, with a more general form of the expression of the forces exchanged between the fluid and the tubes. The theoretical basis of the numerical model are presented, as well as illustrations of the interest of the method: a better physical description is obtained for the dynamic behavior of the tubes bundle, particularly in the case of interactions with free fluid, without tubes.

Transient Analysis of Crossduct Break Scenarios Using the CATHARE2 Code for the 75MW ALLEGRO Demonstrator

2014 ◽  
Author(s):  
Gusztáv Mayer ◽  
Fabrice Bentivoglio
Keyword(s):  
Fast Reactor ◽  
Transient Analysis ◽  
Early Stage ◽  
Thermal Power ◽  
The Road ◽  
The Core ◽  
Nitrogen Injection ◽  
Stage Of Development ◽  
Wide Range ◽  
On The Road

The helium cooled Gas Fast Reactor (GFR) is one of the six reactor concepts selected in the frame of the Generation IV International Forum. Since no gas cooled fast reactor has ever been built, a medium power demonstrator reactor — named ALLEGRO — is necessary on the road towards the 2400MWth GFR power reactor. The French CEA completed a wide range of studies on the early stage of development of ALLEGRO, and later the ALLEGRO reactor have been developed in several European Union projects in parallel with the GFR2400. The 75 MW thermal power ALLEGRO is recently being developed in the frame of European ALLIANCE project. As a result of the collaboration between CEA and MTA EK new improvements were done in the CATHARE modeling of ALLEGRO. In particular, the capability of simulation of breaks located in the crossduct (concentrically arranged pipes with the hotduct located inside the colduct) has been developed. A first scenario of hotduct break has been simulated, that does not lead to the depressurization of the system because of the crossduct technology. Nevertheless this transient leads to a high bypass of the core. Then a scenario of full rupture of the hotduct and the colduct has been tested, leading to beyond design state with depressurized situation combined with a large bypass of the core. However this study shows that the peak cladding temperature can be kept below the cladding melting point using nitrogen injection. In this paper the CATHARE model implemented for the crossduct rupture scenario and the results of the simulation are presented and discussed.

Observation of Fatigue Damage to Gear Tooth Using Image Processing

2007 ◽  
Author(s):  
Tomoya Masuyama ◽  
Takuya Ikeda ◽  
Satoshi Yoshiizumi ◽  
Katsumi Inoue
Keyword(s):  
Image Processing ◽  
Acoustic Emission ◽  
Strain Distribution ◽  
Cyclic Load ◽  
Early Stage ◽  
Gear Tooth ◽  
Digital Camera ◽  
Correlation Method ◽  
Point Of View ◽  
Strain Diagram

The detection of damage in early stage of fatigue is important for a reliable evaluation of gear life and strength. From this point of view, the variation of strain distribution in a tooth due to cyclic load contains useful information because the fatigue crack will initiate as a result of the accumulation of plastic strain. Meanwhile, digital image equipments are widely used in our life and the performance is in progress. We took digital pictures of cyclic loaded tooth by the digital camera and compared with the picture of no load to find displacement. The strain distribution of tooth is calculated by the correlation method using those pictures. The initiation of a micro crack is observed by the method. It is also confirmed by the detection of acoustic emission wave with higher energy. The variation of stress-strain diagram in fatigue process is presented, and this illustrates the increase of strain in the final stage of fatigue.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

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