Fast Reactor Core with Invariable Reactivity and Power Distribution

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.

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A Procedure for the Pre-Conceptual Design of Fast Reactors and Application to a Gas-Cooled Sub-Critical Transmuter

Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition ◽

10.1115/icone22-30387 ◽

2014 ◽

Author(s):

Carlo Fiorina ◽

Konstantin Mikityuk ◽

Jiři Křepel

Keyword(s):

Conceptual Design ◽

Fast Reactor ◽

Power Distribution ◽

Fuel Cycle ◽

Reactor Core ◽

Maximum Pressure ◽

Test Case ◽

Pressure Losses ◽

Design And Optimization ◽

Inert Matrix

A C++ procedure has been developed for the design and optimization of Fast Reactor (FR) cores. It couples the ERANOS based EQL3D procedure developed at PSI for FR equilibrium fuel cycle analysis with a dedicated MATLAB script that evaluates the thermal-hydraulic characteristics of the reactor core. It is conceived to investigate reactors with both standard pins and annular pins. The procedure accepts as input the physical properties of the system, as well as a set of target core parameters presently consisting of core power, maximum fuel burnup, multiplication factor, inner pin diameter (for annular pins) or maximum pressure losses (for standard pins), and core height. It gives as a result a core design fulfilling these design objectives and meeting the constraints on maximum fuel and clad temperatures. In case of annular pins, it also equalizes the temperature rise inside and outside of the core average pin. The procedure considers the possibility of two-zone cores and adjusts the fuel composition in the two zones to achieve an optimal radial power distribution. Finally, it can evaluate safety parameters and fuel cycle characteristics both at beginning-of-life and at equilibrium. As a test case, the procedure has been used for the pre-conceptual design of a sub-critical Gas Fast Reactor core employing inert-matrix sphere-pac fuel and annular pins with SiC cladding.

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

Nuclear Engineering and Design ◽

10.1016/0029-5493(92)90006-h ◽

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

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

Journal of Alloys and Compounds ◽

10.1016/s0925-8388(98)00144-3 ◽

1998 ◽

Vol 271-273 ◽

pp. 530-533 ◽

Cited By ~ 18

Author(s):

M Yamawaki ◽

H Suwarno ◽

T Yamamoto ◽

T Sanda ◽

K Fujimura ◽

...

Keyword(s):

Fast Reactor ◽

Reactor Core

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

Kerntechnik ◽

10.1515/kern-2019-0054 ◽

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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