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.

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.

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.

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.

scholarly journals Calculation of excore detector weighting functions for a sodium-cooled TRU burner mockup using MCNP5

2015 ◽  
Vol 5 (2) ◽  
pp. 15-25
Author(s):  
Viet Ha Pham Nhu ◽  
Min Jae Lee ◽  
Sunghwan Yun ◽  
Sang Ji Kim
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Detector Response ◽  
Weighting Functions ◽  
Control Rod ◽  
The Core ◽  
Power Regulation ◽  
Fuel Assemblies ◽  
Control Rods ◽  
Horizontal Layers

Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A.

Analysis of Seismic Experiment of Hexagonal Multi Bundle Model Using Core Assembly Mock-Up

2020 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Shinichiro Matsubara ◽  
Hidenori Harada ◽  
Tomohiko Yamamoto
Keyword(s):  
Fast Reactor ◽  
Three Dimensional ◽  
Reactor Core ◽  
Vertical Direction ◽  
Horizontal Displacement ◽  
Vertical Displacement ◽  
The Core ◽  
Vibration Behavior ◽  
Seismic Experiment ◽  
Core Elements

Abstract The fast reactor core is composed of hundreds of core elements that stand independently on the core support plate, but does not have support to constrain vertical displacement in order to avoid effects such as thermal elongation. When the earthquake occurs, the group vibration behavior is shown, including the rising of core elements in vertical direction, the collision with adjacent core elements in horizontal direction, and the fluid structure interaction. The three dimensional core group vibration analysis code (REVIAN-3D) was constructed to evaluate them. In the case of fast reactor cores in Japan, the horizontal displacement of core elements at the outermost periphery is restricted by the core former (core barrel). However, since there is no core former in fast reactors other than Japan and the boundary conditions are different from those in Japan, the vibration behavior also differs. In this study, to grasp and estimate the group vibration behavior with and without a core former under the earthquake motion, seismic experiment of hexagonal multi bundle model using core assembly mock-up was conducted [1]. These test results show that the horizontal displacements are larger and impact force between pads of core assembly mock-up is smaller without the core former. In this paper, the analysis was verified by group vibration tests with and without a core former.

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.

Experimental investigation on fluid–structure-coupled dynamic characteristics of the double fuel assemblies in a fast reactor

2013 ◽  
Vol 255 ◽  
pp. 180-184 ◽  
Author(s):  
Daogang Lu ◽  
Aiguo Liu ◽  
Chaohao Shang ◽  
junjie Dang ◽  
yang Hong ◽  
...  
Keyword(s):  
Experimental Investigation ◽  
Dynamic Characteristics ◽  
Fast Reactor ◽  
Fluid Structure ◽  
Fuel Assemblies

scholarly journals MODELLING ASTRID-LIKE SODIUM-COOLED FAST REACTOR WITH SERPENT-DYN3D CODE SEQUENCE

2021 ◽  
Vol 247 ◽  
pp. 02028
Author(s):  
Wojciech Rydlewicz ◽  
Emil Fridman ◽  
Eugene Shwageraus
Keyword(s):  
Monte Carlo ◽  
Fuel Assembly ◽  
Fast Reactor ◽  
Power Distribution ◽  
Fast Reactors ◽  
The Core ◽  
Fuel Assemblies ◽  
Complex Core ◽  
Nodal Diffusion ◽  
Full Core

This study explores the feasibility of applying the Serpent-DYN3D sequence to the analysis of Sodium-cooled Fast Reactors (SFRs) with complex core geometries, such as the ASTRIDlike design. The core is characterised by a highly heterogeneous configuration and was likely to challenge the accuracy of the Serpent-DYN3D sequence. It includes axially heterogeneous fuel assemblies, non-uniform fuel assembly heights and large sodium plena. Consequently, the influence of generation and correction methods of various homogenised, few-group crosssections (XS) on the accuracy of the full-core nodal diffusion DYN3D calculations is presented. An attempt to compare the approximate time effort spent on models preparation against the accuracy of the result is made. Results are compared to reference full-core Serpent MC (Monte Carlo) solutions. Initially, XS data was generated in Serpent using traditional methods (2D single assemblies and 2D super-cells). Full core calculations and MC simulations offered a moderate agreement. Therefore, XS generation with 2D fuel-reflector models and 3D single assembly models was verified. Super-homogenisation (SPH) factors for XS correction were applied. In conclusion, the performed work suggests that Serpent-DYN3D sequence could be used for the analysis of highly heterogeneous SFR designs similar to the studied ASTRID-like, with an only small penalty on the accuracy of the core reactivity and radial power distribution prediction. However, the XS generation route would need to include the correction with SPH factors and generation of XS with various MC models, for different core regions. At a certain point, there are diminishing returns to using more complex XS generation methods, as the accuracy of full-core deterministic calculations improves only slightly, while the time effort required increases significantly.

Safety Analysis of a Super Fast Reactor With Upward Flow Cooling in Two Pass at Supercritical Pressure

2013 ◽  
Author(s):  
Takayoshi Kamata ◽  
Haipeng Li ◽  
Yoshiaki Oka ◽  
Yuki Ishiwatari
Keyword(s):  
Fast Reactor ◽  
Safety Analysis ◽  
Supercritical Pressure ◽  
Safety Performance ◽  
Coolant Flow ◽  
Upward Flow ◽  
The Core ◽  
Flow Scheme ◽  
Fuel Assemblies ◽  
Loss Of Coolant

Safety characteristics of the supercritical-pressure light water-cooled fast reactor (Super FR) with upward flow core cooling in two pass is investigated for the abnormal transients and accidents at supercritical pressure. Upward flow cooling has advantage of simplifying the upper core structure in comparison with the downward flow scheme that part of the coolant flows downward in the blanket fuel assemblies from the top dome of reactor pressure vessel. It also has advantages that flow stagnation does not occur at loss of coolant flow events due to the buoyancy of the coolant. The coolant flow scheme of this design is the all blanket fuel assemblies and part of the seed fuel assemblies are cooled with upward flow first, the coolant flows radially above the core and flow downward in the gap between the core and the shroud to the lower mixing plenum and cools the rest of seed fuel assemblies with upward flow till the upper mixing plenum before core outlet. To evaluate the safety performance, eleven transients and four accidents at supercritical-pressure are analyzed. Safety analysis results show that the safety criteria are satisfied with large margins for all the selected transients and accidents. But in the total loss of coolant flow accident the MCST (maximum cladding surface temperature) is still high. Because of this flow scheme, it is found that the MCST is sensitive to the volume of the gap between two pass. Actuating depressurization valves with low flow single at total loss of flow events is effective to induce flow for once-through SCWR and therefore improves safety performance.

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