scholarly journals AR-1 experimental model and facility preparation for the purpose of experimental investigation of sodium boiling in fuel subassembly mockup for new generation fast reactor safety justification

2014 ◽  
Vol 2014 (1) ◽  
pp. 77-87
Author(s):  
Ruslan Rashitovich Khafizov ◽  
Yurij Mikhaylovich Ashurko ◽  
Andrey Vikent’evich Volkov ◽  
Evgeny Fyodorovich Ivanov ◽  
Vitalij Vladimirovich Privezentsev ◽  
...  
Keyword(s):  
Experimental Investigation ◽  
Experimental Model ◽  
Fast Reactor ◽  
Reactor Safety ◽  
New Generation

scholarly journals Experimental investigation of sodium boiling heat exchange in fuel subassembly mockup for perspective fast reactor safety justification

2015 ◽  
Vol 2015 (3) ◽  
pp. 85-96
Author(s):  
Ruslan Rashitovich Khafizov ◽  
Vladimir Mikhaylovich Poplavsky ◽  
Valery Ivanovich Rachkov ◽  
Aleksandr Pavlovich Sorokin ◽  
Yurij Mikhaylovich Ashurko ◽  
...  
Keyword(s):  
Experimental Investigation ◽  
Heat Exchange ◽  
Fast Reactor ◽  
Reactor Safety

scholarly journals Experimental investigation of sodium boiling heat exchange in fuel subassembly mockup for perspective fast reactor safety substantiation

2015 ◽  
Vol 1 (2) ◽  
pp. 147-152 ◽  
Author(s):  
R.R. Khafizov ◽  
V.M. Poplavsky ◽  
V.I. Rachkov ◽  
A.P. Sorokin ◽  
Yu.M. Ashurko ◽  
...  
Keyword(s):  
Experimental Investigation ◽  
Heat Exchange ◽  
Fast Reactor ◽  
Reactor Safety

scholarly journals Fast reactor safety and the fire at Beloyarsk

Nature ◽  
1980 ◽  
Vol 284 (5751) ◽  
pp. 10-10
Author(s):  
R.D. SMITH
Keyword(s):  
Fast Reactor ◽  
Reactor Safety

scholarly journals MELT-III: a neutronics, thermal-hydraulics computer program for fast reactor safety analysis. Volume II

1974 ◽  
Author(s):  
A.E. Waltar ◽  
W.L. Partain ◽  
D.C. Kolesar ◽  
L.D. O'Dell ◽  
A. Jr. Padilla ◽  
...  
Keyword(s):  
Computer Program ◽  
Fast Reactor ◽  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Reactor Safety

New Reactor Safety Measures for the European Sodium Fast Reactor - Part I: Conceptual Design

2021 ◽  
Author(s):  
Joel Guidez ◽  
Janos Bodi ◽  
Konstantin Mikityuk ◽  
Enrico Girardi ◽  
Bernard Carluec
Keyword(s):  
Fast Reactor ◽  
Heat Removal ◽  
Thermal Inertia ◽  
Lessons Learned ◽  
Reactor Safety ◽  
Fukushima Accident ◽  
Safety Level ◽  
Safety Measures ◽  
Sodium Fast Reactor ◽  
High Thermal Inertia

Abstract Following up the previous CP-ESFR project, the ESFR-SMART project considers the safety objectives envisaged for Generation-IV reactors, taking into account the lessons learned from the Fukushima accident, in order to increase the safety level of the European Sodium Fast Reactor (ESFR). In accordance with these objectives, guidelines have been defined to drive the ESFR-SMART developments, mainly simplifying the design and using all the positive features of Sodium Fast Reactors (SFR), such as low coolant pressure, efficiency of natural convection, possibility of decay heat removal (DHR) by atmospheric air, high thermal inertia and long grace period before a human intervention is needed. In this paper, a set of new ambitious safety measures is introduced for further evaluation within the project. The proposed set aims at consistency with the main lines of safety evolutions since the Fukushima accident, but it does not yet constitute the final comprehensive safety analysis. The paper gives a first review of the new propositions to enhance the ESFR safety, leading to a simplified reactor, forgiving and including a lot of passivity. This first version is supported by the various project tasks in order to assess the relevance of the whole design in comparison to the final safety objectives.

Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

2016 ◽  
Vol 63 (2) ◽  
pp. 130-139 ◽  
Author(s):  
V. M. Alipchenkov ◽  
A. M. Anfimov ◽  
D. A. Afremov ◽  
V. S. Gorbunov ◽  
Yu. A. Zeigarnik ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Current State ◽  
New Generation

Fuel behavior aspects of the interpretation of the SCARABEE fast reactor safety experiments

1980 ◽  
Vol 59 (2) ◽  
pp. 275-286 ◽  
Author(s):  
F. Schmitz ◽  
J.R. Matthews ◽  
H. Elbel ◽  
L. Roche ◽  
M. deFranco ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Safety ◽  
Fuel Behavior

Experimental Modelling of a Molten Salt Reactor Concept

2013 ◽  
Author(s):  
Bogdán Yamaji ◽  
Attila Aszódi
Keyword(s):  
Molten Salt ◽  
Experimental Model ◽  
Fast Reactor ◽  
Measurement Data ◽  
Internal Heat ◽  
Working Fluid ◽  
Experimental Modelling ◽  
The Core ◽  
Single Region ◽  
Piv Measurement

Based on the MSFR (Molten Salt Fast Reactor) reactor concept proposed within the framework of the EVOL (Evaluation and Viability of Liquid Fuel Fast Reactor System, EU FP7) international research project a scaled and segmented experimental model of the MSFR and first measurement result will be presented in the paper. MSFR is a single region, homogeneous liquid fuelled fast reactor concept. The reactor uses fluoride-based molten salts as fuel and coolant, with fissile uranium and/or thorium and other heavy nuclei content with the purpose of applying the thorium cycle and the burn-up of transuranic elements. The concept has a single region cylindrical core with sixteen radial inlet and outlet nozzles located at the bottom and top of the core. The external circuit (internal heat exchanger, pump, pipes) is broken up in sixteen identical modules distributed around the core. A scaled and segmented experimental model of the MSFR concept was designed and built in order to carry out Particle Image Velocimetry (PIV) measurements. Purpose of the experimental mock-up is to provide measurement data for validation and benchmarking of CFD simulations, and also to study specific problems or phenomena related to the MSFR, such as design of inlet geometry, effects of internal structures, coolant mixing. The experimental model uses water as working fluid with 50 μm polyamide seeding particles added for PIV measurement. Geometrical scaling was applied in order to reduce size and necessary pumping power and the geometry represents a 90 degree segment of the original cylindrical geometry. It was not possible to maintain the nominal value of the Reynolds-number (∼1E+06 for the core) however a highly turbulent flow (Re>1E+05) can be reproduced in the system. Final design of the scaled and segmented plexiglas model will be presented, capabilities and limitations of the measurement assembly will be discussed together with the presentation of first measurements results.

Sign in / Sign up

Export Citation Format

Share Document