Commercialized fast reactor cycle systems and reactor core performance of the promising fast reactors

2005 ◽  
Vol 47 (1-4) ◽  
pp. 300-313 ◽  
Author(s):  
Shoji Kotake ◽  
Yoshihiko Sakamoto ◽  
Yutaka Sagayama
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Fast Reactors ◽  
Cycle Systems

scholarly journals Improved FAST Device for Inherent Safety of Oxide-Fueled Sodium-Cooled Fast Reactors

Energies ◽  
2021 ◽  
Vol 14 (15) ◽  
pp. 4610
Author(s):  
Ahmed Amin E. Abdelhameed ◽  
Chihyung Kim ◽  
Yonghee Kim
Keyword(s):  
Operating Temperature ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Inherent Safety ◽  
Oxide Fuel ◽  
Fast Reactors ◽  
Power Oscillation ◽  
Negative Reactivity ◽  
Oscillation Problem ◽  
Reactivity Coefficient

The floating absorber for safety at transient (FAST) was proposed as a solution for the positive coolant temperature coefficient in sodium-cooled fast reactors (SFRs). It is designed to insert negative reactivity in the case of coolant temperature rise or coolant voiding in an inherently passive way. The use of the original FAST design showed effectiveness in protecting the reactor core during some anticipated transients without scram (ATWS) events. However, oscillation behaviors of power due to refloating of the absorber module in FAST were observed during other ATWS events. In this paper, we propose an improved FAST device (iFAST), in which a constraint is imposed on the sinking (insertion) limit of the absorber module in FAST. This provides a simple and effective solution to the power oscillation problem. Here, we focus on an oxide fuel-loaded SFR that is characterized by a more negative Doppler reactivity coefficient and higher operating temperature than the metallic-loaded SFR cores. The study is carried out for the 1000 MWth advanced burner reactor with an oxide fuel-loaded core during postulated ATWS events that are unprotected transient over power, unprotected loss of flow, and unprotected loss of the heat sink. It was found that the iFAST device has promising potentials for protecting the oxide SFR core during the various studied ATWS events.

Advances in the analysis of fast reactor core and coolant circuit structures

1992 ◽  
Vol 134 (1) ◽  
pp. 37-58
Author(s):  
Y.W. Chang ◽  
D.T. Eggen ◽  
A. Imazu ◽  
M. Livolant
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Coolant Circuit

Prospects for fast reactors

1987 ◽  
Vol 92 (1-2) ◽  
pp. 57-71
Author(s):  
C. W. Blumfield
Keyword(s):  
Fast Reactor ◽  
High Energy ◽  
High Energy Density ◽  
Fast Reactors ◽  
Advantages And Disadvantages ◽  
Reactor Technology ◽  
European Collaboration ◽  
Fast Reactor Fuel ◽  
Fuel Reprocessing ◽  
Planned Activities

SynopsisThe background to recent major advances in the construction and operation of fast reactors is outlined with particular reference to the Dounreay Prototype Fast Reactor. The advantages and disadvantages of sodium as a coolant of the high energy density assembly are discussed and an account given of the consequences of a leak and the precautions taken against this eventuality. Attention is drawn to the safety aspects of the system. The economics of the plans for fuel reprocessing are explained and an account given of the progress in the fabrication of fast reactor fuel pins. Finally the environmental impact of present and planned activities on the Dounreay site is presented in the context of participation in the European Collaboration on Fast Reactor Technology and attention drawn to the importance of the planning inquiry in progress at Dounreay.

Concept of hydride fuel target subassemblies in a fast reactor core for effective transmutation of MA

1998 ◽  
Vol 271-273 ◽  
pp. 530-533 ◽  
Author(s):  
M Yamawaki ◽  
H Suwarno ◽  
T Yamamoto ◽  
T Sanda ◽  
K Fujimura ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core

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.

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

scholarly journals Burnup concept for a long-life fast reactor core using MCNPX.

2013 ◽  
Author(s):  
Holschuh, Thomas Vernon, ◽  
Lewis, Tom Goslee, ◽  
Parma, Edward J.,
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Long Life
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