Neutronics Characteristics Study of Conceptual Space Heat-Pipe-Cooled Fast Reactor Core

TESTS ON AN ELECTRICALLY HEATED SODIUM HEAT PIPE FOR TEMPERATURE CONTROL OF IRRADIATION EXPERIMENTS IN THE UK PROTOTYPE FAST REACTOR

Proceeding of International Heat Transfer Conference 8 ◽

10.1615/ihtc8.3320 ◽

1986 ◽

Author(s):

J. C. Ralph ◽

G. Chaffey

Keyword(s):

Heat Pipe ◽

Temperature Control ◽

Fast Reactor ◽

Sodium Heat Pipe ◽

The Uk

Download Full-text

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

Download Full-text

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

Download Full-text

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.

Download Full-text

A Heat Pipe Cooled Modular Reactor Concept for Manned Lunar Base Application

Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors ◽

10.1115/icone21-16006 ◽

2013 ◽

Author(s):

Gu Hu ◽

Shouzhi Zhao ◽

Zhiyong Sun ◽

Chengzhi Yao

Keyword(s):

Power System ◽

Heat Pipe ◽

Fast Reactor ◽

Waste Heat ◽

Single Point ◽

Liquid Coolant ◽

Thermoelectric Conversion ◽

Low Mass ◽

Long Lifetime ◽

Base Application

A lithium heat pipe cooled modular fast reactor (HPCMR) power system concept has been developed for manned lunar base application. The system is designed to use the static thermoelectric conversion module to produce over 100kW electricity for up to ten years. Waste heat is rejected by potassium heat pipe radiator. This system has advantages of low mass, long lifetime, no pumped liquid coolant, and no single point of failure. Main parameters of the system are also given in this paper.

Download Full-text

Application of Covariances to Fast Reactor Core Analysis

Nuclear Data Sheets ◽

10.1016/j.nds.2008.11.009 ◽

2008 ◽

Vol 109 (12) ◽

pp. 2778-2784 ◽

Cited By ~ 7

Author(s):

M. Ishikawa

Keyword(s):

Fast Reactor ◽

Reactor Core ◽

Core Analysis

Download Full-text

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

Volume 9: Student Paper Competition ◽

10.1115/icone26-81626 ◽

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.

Download Full-text