Assessment of Molten Fuel Relocation through Fuel Assembly Inner Duct System of Sodium-cooled Fast Reactor

2019 ◽  
Author(s):  
Prasant Kumar Panigrahi ◽  
K. Velusamy
Keyword(s):  
Fuel Assembly ◽  
Fast Reactor ◽  
Duct System

scholarly journals Comparative Study on Fuel Assembly of Modular Gas-cooled Fast Reactor using MCNP and OpenMC Code

2021 ◽  
Vol 1772 (1) ◽  
pp. 012031
Author(s):  
H. Raflis ◽  
M. Ilham ◽  
Z. Su’ud ◽  
A. Waris ◽  
D. Irwanto
Keyword(s):  
Comparative Study ◽  
Fuel Assembly ◽  
Fast Reactor

Component design and accident analysis of fuel assembly for prototype GEN-IV sodium-cooled fast reactor

2018 ◽  
Vol 340 ◽  
pp. 133-145 ◽  
Author(s):  
K.H. Yoon ◽  
H.-K. Kim ◽  
H.S. Lee ◽  
J.S. Cheon
Keyword(s):  
Fuel Assembly ◽  
Fast Reactor ◽  
Accident Analysis ◽  
Component Design ◽  
Gen Iv

scholarly journals Structural seismic characteristics assessment of LMFR fuel assembly hexagonal wrapper over operational temperature range

2019 ◽  
Vol 34 (4) ◽  
pp. 313-324
Author(s):  
M. Khizer ◽  
Zhang Yong ◽  
Yang Guowei ◽  
Wu Qingsheng ◽  
Wu Yican
Keyword(s):  
Fuel Assembly ◽  
Fast Reactor ◽  
Natural Frequencies ◽  
Mass Effect ◽  
Added Mass ◽  
Reactor Fuel ◽  
Temperature Range ◽  
Fast Reactor Fuel ◽  
Lead Bismuth Eutectic ◽  
Reactor Fuel Assembly

In this study, the structural integrity of liquid metal fast reactor fuel assembly has been established for different parameters considering the optimum fuel design. Analytical calculation of added mass effect due to lead bismuth eutectic and verification through previously presented theories, has been established. The integrity of the hexagonal wrapper of fuel assembly has been guaranteed over the entire operating temperature range. Effect of temperature on the density of lead bismuth eutectic, the subsequent change in added mass of lead bismuth eutectic, the effect on natural frequencies and effect on stresses on wrapper, has been studied in detail. A simple empirical relationship is presented for estimation of added mass effect for lead bismuth eutectic type fast reactors for any desired temperature. An approach for assessment of fast reactor fuel assembly performance has been outlined and calculated results are presented. Nuclear seismic rules require that systems and components which are important to safety, shall be capable of bearing earthquake effects and their integrity and functionality should be guaranteed. Mode shapes, natural frequencies, stresses on wrapper and seismic aspect has also been considered using ANSYS. Modal analysis has been compared in vacuum and lead bismuth eutectic using the calculated added mass.

scholarly journals Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly

2016 ◽  
Vol 298 ◽  
pp. 218-228 ◽  
Author(s):  
E. Merzari ◽  
P. Fischer ◽  
H. Yuan ◽  
K. Van Tichelen ◽  
S. Keijers ◽  
...  
Keyword(s):  
Fluid Flow ◽  
Liquid Metal ◽  
Fuel Assembly ◽  
Fast Reactor ◽  
Reactor Fuel ◽  
Flow Simulations ◽  
Fast Reactor Fuel ◽  
Reactor Fuel Assembly

Hydraulic experiment investigation on the plate-throttle entry tube flow characteristics of fast reactor fuel assembly

2019 ◽  
Vol 352 ◽  
pp. 110172
Author(s):  
Haiqi Qin ◽  
Daogang Lu ◽  
Shaohua Liu ◽  
Jiaxuan Tang ◽  
Yu Wang ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Fast Reactor ◽  
Flow Characteristics ◽  
Reactor Fuel ◽  
Tube Flow ◽  
Hydraulic Experiment ◽  
Fast Reactor Fuel ◽  
Reactor Fuel Assembly

CFD Pre-Test Analysis and Design of the NACIE-UP BFPS Fuel Pin Bundle Simulator

2016 ◽  
Author(s):  
R. Marinari ◽  
I. Di Piazza ◽  
M. Tarantino ◽  
F. Magugliani ◽  
A. Alemberti ◽  
...  
Keyword(s):  
Heat Flux ◽  
Fuel Assembly ◽  
Fast Reactor ◽  
Test Section ◽  
Wall Heat Flux ◽  
Test Analysis ◽  
Important Indicator ◽  
Analysis And Design ◽  
Fuel Pin ◽  
Flow Blockage

In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel sub-assembly is considered one of the main issues to be addressed and one of the most important and realistic accident for Lead Fast Reactors (LFR) fuel assembly. The blockage in a fast reactor Fuel Assembly (FA) may have serious effects on the safety of the plant leading to the FA damaging or melting. The temperature of the coolant leaving the FA is considered an important indicator of the health of the FA (i.e. the effective heat removal) and is usually monitored via a dedicated, safety-related system (e.g. thermocouple). The external or internal blockage of the FA may impair the correct cooling of the fuel pins, be the root cause of anomalous heating of the cladding and of the wrapper and potentially impact also fuel pins not directly located above or around the blocked area. In order to model the temperature and velocity field inside a wrapped FA under unblocked and blocked conditions, detailed experimental campaign as well as 3D thermal hydraulic analyses of the FA is required. The present paper is focused on the CFD pre-test analysis and design of the new experimental facility ‘Blocked’ Fuel Pin bundle Simulator (BFPS) that will be installed into the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The BFPS test section will be installed into the NACIE-UP loop facility aiming to carry out suitable experiments to fully investigate different flow blockage regimes in a 19 fuel pin bundle providing experimental data in support of the development of the ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) LFR DEMO. In particular, the ‘Blocked’ Fuel Pin bundle Simulator (BFPS) cooled by lead bismuth eutectic (LBE), was conceived with a thermal power of about 250 kW and a uniform wall heat flux up to 0.7 MW/m2, relevant values for a LFR. It consists of 19 electrical pins placed on a hexagonal lattice with a pitch to diameter ratio of 1.4 and a diameter of 10 mm. The geometrical domain of the fuel pin bundle simulator was designed to reproduce the geometrical features of ALFRED, e.g. the external wrapper in the active region and the spacer grids. Pre-tests calculations were carried out by applying accurate boundary conditions; the conjugate heat transfer in the clad is also considered. The numerical simulation test matrix covered the envisioned experimental range in terms of mass flow rate; the wall heat flux was imposed in order to have a fixed temperature difference across the BFPS in unblocked conditions. The blockages investigated are internal blockages of different extensions and in different locations (central subchannel blockage, corner sub-channel blockage, edge subchannel blockage, one sector blockage, two sector blockage). High resolution RANS simulations were carried out adopting the ANSYS CFX V15 commercial code with the laminar sublayer resolved by the mesh resolution. The loci of the peak temperatures and their width as predicted by the CFD simulations are used for determining the location of the pin bundle instrumentation. The CFD pre-test analysis allowed also investigating the temperature distribution in the clad to operate the test section safely.

scholarly journals Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

2014 ◽  
Vol 2014 ◽  
pp. 1-14 ◽  
Author(s):  
Franck Dechelette ◽  
Franck Morin ◽  
Guy Laffont ◽  
Gilles Rodriguez ◽  
Emmanuel Sanseigne ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Fuel Assembly ◽  
Fast Reactor ◽  
Route Optimization ◽  
Load Factor ◽  
The Third ◽  
Technological Improvement ◽  
Investment Cost ◽  
Innovation Study ◽  
Design Simplicity

The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth) but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

Experimental studies of heat exchange for sodium boiling in the fuel assembly model: Safety substantiation of a promising fast reactor

2017 ◽  
Vol 64 (1) ◽  
pp. 6-14 ◽  
Author(s):  
R. R. Khafizov ◽  
V. M. Poplavskii ◽  
V. I. Rachkov ◽  
A. P. Sorokin ◽  
A. A. Trufanov ◽  
...  
Keyword(s):  
Heat Exchange ◽  
Fuel Assembly ◽  
Fast Reactor ◽  
Experimental Studies ◽  
Assembly Model
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