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.

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.

High Breeding Core of a Supercritical-Pressure Light Water Cooled Fast Reactor

2013 ◽  
Author(s):  
Taishi Yoshida ◽  
Yoshiaki Oka
Keyword(s):  
Volume Fraction ◽  
Supercritical Pressure ◽  
Collision Probability ◽  
Zirconium Hydride ◽  
Outlet Temperature ◽  
Water Cooling ◽  
Developed Country ◽  
Light Water ◽  
The Core ◽  
Fuel Assemblies

Breeding of plutonium with light water cooling has been studied for many years, but high breeding to meet growing demand for electricity in a developed country has not been accomplished. The purpose of this study is to investigate a high breeding core of Super FBR (supercritical pressure light water cooled fast breeder reactor) with new fuel assemblies consisting of tightly packed fuel rods without gaps, which leads to low coolant to fuel volume fraction. The plant system of a Super FBR is once-through coolant cycle with high head pumps. The coolant flow rate is low due to the high enthalpy rise in the core. It is compatible with the high pressure drop of the new fuel assemblies. Both neutronic and thermal hydraulic design of the core is considered. The challenge of high breeding with light water cooling is to satisfy negative coolant void reactivity, high breeding and low enrichment simultaneously. The core with new assemblies has been designed with the average coolant density of 248 kg/m3. It is achieved by setting 380C inlet and 500C outlet temperature. For satisfying negative void reactivity, a solid moderator layer composed of zirconium hydride (ZrH) rods are adopted in some blanket assemblies. Cross sections of the blanket fuel assemblies with ZrH rods are prepared with assembly-wise calculation, because the pin-wise collision probability calculation overestimates the breeding. MOX fuel is used for seed fuel assemblies. Three types of core layouts with “radially heterogeneous”, “radiating” and “scattered” seed assemblies have been considered, and “radiating” layout shows best breeding characteristics among them. The seed assemblies in a “radiating” layout are not radially separated so that more numbers of blanket assemblies can be placed in high neutron flux region of a core. Fraction of blanket fuel assemblies with ZrH rods is selected for high breeding. Super FBR using the new fuel assemblies achieved both negative void and high plutonium breeding.

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 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.

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.

scholarly journals Safety Analysis of Small Modular Natural Circulation Lead-Cooled Fast Reactor SNCLFR-100 Under Unprotected Transient

2021 ◽  
Vol 9 ◽  
Author(s):  
Chao Guo ◽  
Pengcheng Zhao ◽  
Jian Deng ◽  
Hongxing Yu
Keyword(s):  
Fuel Assembly ◽  
Flow Rate ◽  
Fast Reactor ◽  
Optimization Design ◽  
Natural Circulation ◽  
Coolant Flow ◽  
Coolant Flow Rate ◽  
Design Stage ◽  
Large Power ◽  
The Core

SNCLFR-100 is a small modular natural circulation lead-cooled fast reactor, developed by University of Science and Technology of China, aiming at taking full advantage of the good economics and inherent safety of lead-cooled fast reactors to develop miniaturized, lightweight and multi-purpose special nuclear reactor technology. SNCLFR-100 is still in the conceptual design stage, in order to fully evaluate the safety features of the reactor and provide reference for the optimization design of the next stage, three typical transients are selected based on the analysis of the SNCLFR-100 initiating events by using the code Analysis of Thermal-hydraulics of Leaks and Transients (ATHLET), which are unprotected transient overpower (UTOP), unprotected loss of heat sink (ULOHS) and unprotected partial blockage in the hottest fuel assembly. For UTOP, the unexpected positive reactivity insertion of 0.7$ in 15s led to two large power peaks in the core quickly, and then the core power began to decrease and gradually stabilized under the action of various of negative feedbacks of the reactor, the peak temperatures of fuel and cladding rose rapidly with the increase of core power and eventually stabilized at a higher temperature. For ULOHS, as the reactor were driven by natural circulation, the coolant mass flow rate continued to decline after the transient, both core and cladding temperatures rose quickly and the temperature rise were smaller than that of UTOP transient, the reactor shutdown by itself and the peak temperatures of fuel and cladding were smaller than the safety limit. For unprotected partial blockage in the hottest fuel assembly, with the increase of the blockage rate of the hottest fuel assembly inlet, the coolant flow rate, the peak temperatures of coolant, fuel and cladding increased significantly, when the blockage rate increased to 0.9, the coolant flow rate of the hottest fuel assembly dropped to about 12.6% of the normal value, and the cladding peak temperature would exceed the cladding melting point, measures should be taken to avoid the happening of severe accident.

Safety Analysis of a High Breeding Fast Reactor Cooled by Supercritical-Pressure Light Water: Accidents and Abnormal Transients at Supercritical Pressure

2014 ◽  
Author(s):  
Rui Guo ◽  
Yoshiaki Oka
Keyword(s):  
Fast Reactor ◽  
Doubling Time ◽  
Safety Analysis ◽  
Supercritical Pressure ◽  
Compound System ◽  
System Configuration ◽  
Fuel Rod ◽  
Negative Reactivity ◽  
Breeding Ratio ◽  
Reactivity Coefficients

The high breeding core of a supercritical water cooled fast reactor has been designed with tightly packed fuel rod assemblies to obtain a high breeding ratio and negative reactivity coefficients. A high breeding capability of less than 50 years of compound system doubling time was reported by the conceptual design study of the reactor at ICONE21. The present paper describes the safety analysis of the reactor for accidents and abnormal transients at supercritical pressure. The safety principle, safety system configuration and types of the abnormal transients and accidents are the same as those of the Super FR [1]. Safety criteria such as cladding temperature, pressure and fuel enthalpy are similar to those of Super FR. Results indicate that all safety criteria are satisfied.

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.

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