scholarly journals Flow Pattern and Pressure Fluctuation Characteristics on the 1/3 Scale Hot-Leg Piping Experiments of a Primary Circuit Hot-Leg Piping in a Sodium-Cooled Fast Reactor

2012 ◽  
Vol 78 (792) ◽  
pp. 1378-1382
Author(s):  
Hiromi SAGO ◽  
Tadashi SHIRAISHI ◽  
Hisato WATAKABE ◽  
Yang XU ◽  
Kousuke AIZAWA ◽  
...  
Keyword(s):  
Flow Pattern ◽  
Fast Reactor ◽  
Pressure Fluctuation ◽  
Primary Circuit

scholarly journals Effect of Swirl Inflow on Flow Pattern and Pressure Fluctuation onto a Single-Elbow Pipe in Japan Sodium-Cooled Fast Reactor

2012 ◽  
Vol 7 (3) ◽  
pp. 329-344
Author(s):  
Hidemasa YAMANO ◽  
Hiromi SAGO ◽  
Kazuo HIROTA ◽  
Satoshi HAYAKAWA ◽  
Yang XU ◽  
...  
Keyword(s):  
Flow Pattern ◽  
Fast Reactor ◽  
Pressure Fluctuation

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.

A Study and Development of Primary Circuit Simulation Code for China Experimental Fast Reactor

2013 ◽  
Author(s):  
Jin Wang ◽  
Donghui Zhang ◽  
Wenjun Hu ◽  
Lixia Ren
Keyword(s):  
Fast Reactor ◽  
Nuclear Energy ◽  
Circuit Simulation ◽  
Thermal Power ◽  
Physical Models ◽  
Primary Circuit ◽  
Cooling Channel ◽  
Simulation Code ◽  
Wide Range ◽  
China Experimental Fast Reactor

A fast reactor is one of recommended candidates of Generation IV nuclear energy systems, which would meet wide requirements such as sustainability, safety and economics for nuclear energy development. To be the China’s first fast reactor, China Experimental Fast Reactor (CEFR) typical technical options are following: 65 MW thermal power and 20 MW electric power, three circuits of sodium-sodium-water, integrated pool type structure for the primary circuit. To establish modular simulation system for sodium fast reactor, the code which simulated the thermal-hydraulic behavior of primary circuit was developed. The physical models include reactor core, reactor vessel cooling channel, pumps, protection vessel, intermediate heat exchangers, ionization chamber cooling channel, cold sodium pool, hot sodium pool, inlet plenum, and pipes, etc. The code could compute coolant pressures, flow rates, and temperatures in the primary circuit. This module was designed for analysis of a wide range of transients. Although based on CEFR, it can treat an arbitrary arrangement of components.

Numerical Analysis of Flow-Induced Vibration of Large Diameter Pipe With Short Elbow

2017 ◽  
Author(s):  
Shigeru Takaya ◽  
Tatsuya Fujisaki ◽  
Masaaki Tanaka
Keyword(s):  
Numerical Simulation ◽  
Numerical Analysis ◽  
Fast Reactor ◽  
Pressure Fluctuation ◽  
Cooling System ◽  
Large Diameter ◽  
Recirculation Region ◽  
Flow Induced Vibration ◽  
Scaled Model ◽  
Energy Agency

Japan Atomic Energy Agency is now conducting design study and R&D of an advanced loop-type sodium cooled fast reactor. The cooling system is planned to be simplified by employing a two-loop configuration and shortened piping with less elbows than a prototype fast reactor in Japan, Monju, in order to reduce construction costs and enhance economic performance. The design, however, increases flow velocity in the hot-leg piping and induces large flow turbulence around elbows. Therefore, flow-induced vibration (FIV) of a hot-leg piping is one of main concerns in the design. Numerical simulation is a useful method to deal with such a complex phenomenon. We have been developing numerical analysis models of the hot-leg piping using Unsteady Reynolds Averaged Navier-Stokes simulation with Reynolds stress model. In this study, numerical simulation of a 1/3 scaled-model of the hot-leg piping was conducted. The results such as velocity profiles and power spectral densities (PSD) of pressure fluctuations were compared with experiment ones. The simulated PSD of pressure fluctuation at the recirculation region agreed well with the experiment, but it was found some underestimation at other parts, especially in relatively high frequency range. Eigenvalue vibration analysis was also conducted using a finite element method. Then, stress induced by FIV was evaluated using pressure fluctuation data calculated by URANS simulation. The calculated stress generally agrees well the measurement values, which indicates the importance of precise evaluation of the PSD of pressure fluctuation at the recirculation region for evaluation of FIV of the hot-leg piping with a short elbow.

RBEC-M Lead-Bismuth Cooled Fast Reactor: Optimization of Conceptual Decisions

2002 ◽  
Author(s):  
K. Mikityuk ◽  
A. Vasiliev ◽  
P. Fomichenko ◽  
T. Schepetina ◽  
S. Subbotin ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Cooling System ◽  
The Other ◽  
Primary Circuit ◽  
Core Concept ◽  
The Core ◽  
Nitride Fuel ◽  
Reactor Optimization ◽  
Cost Parameters ◽  
Gas Lift

A concept of the RBEC lead-bismuth fast reactor is a synthesis, on one hand, experience in development and operation of fast sodium power reactors and reactors with Pb-Bi coolant, and, on the other hand, of large R&D activities on development of the core concept for modified fast sodium reactor. The paper presents improved project RBEC-M, characterized by a number of innovative decisions, which allow to improve safety and cost parameters compared to the basic RBEC project. These innovative decisions include application of nitride fuel based on 15N, two-circuit scheme without main reactor pumps, gas lift system in the primary circuit, passive reactor auxiliary cooling system, etc.

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