Validation of THACS for Sodium Cooled Fast Reactor Based on Benchmark Analysis of EBR-II

2014 ◽  
Author(s):  
Yue Nina ◽  
Ma Zaiyong ◽  
Hu Benxue ◽  
Suizheng Qiu ◽  
Guanghui Su
Keyword(s):  
Fast Reactor ◽  
Channel Model ◽  
Flow Characteristics ◽  
Primary Loop ◽  
Reactor Kinetics ◽  
Multiple Channel ◽  
Benchmark Analysis ◽  
The Core ◽  
Intermediate Heat Exchanger ◽  
Transient Thermal

In this paper the thermal-hydraulic characteristics of the primary loop of the Experimental Breeder Reactor (EBR-II), including the temperature and the flow characteristics of the core, the intermediate heat exchanger (IHX) and the experiment subassembly XX09 and XX10, were analyzed with the transient thermal-hydraulic code THACS. The THACS code contains the core, the pumps, IHX, the sodium pool and some other modules, and each module could operate separately. All of the primary–loop components are simulated one-dimensional, and in the core calculation the incompressible model for the single phase. The multiple-channel model is applied to simulate the core subassemblies, including the average, hot, XX09, XX10, the reflector and the blanket channels. The neutron physics is calculated with the point reactor kinetics, and the reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of core are considered. Two tests, namely the SHRT-17 and SHRT-45R tests, are simulated to validate our tools and models. The THACS simulation results show that the EBR-II type sodium cooled fast reactor could shut down automatically relying on inherent negative feedbacks in the two tests.

Allegro: The European Gas Fast Reactor Demonstrator Project

2009 ◽  
Author(s):  
Christian Poette ◽  
Vale´rie Brun-Magaud ◽  
Franck Morin ◽  
Jean-Franc¸ois Pignatel ◽  
Richard Stainsby ◽  
...  
Keyword(s):  
High Temperature ◽  
Fast Reactor ◽  
Heat Removal ◽  
Reactor System ◽  
Large Temperature ◽  
The Core ◽  
Intermediate Heat Exchanger ◽  
Current Design ◽  
Ceramic Fuel ◽  
High Temperature Components

In the Gas Fast Reactor development plan, ALLEGRO is the first necessary step towards the electricity generating prototype GFR. The ALLEGRO start of operation is planned by 2020. This needs to define all design options in 2010 and to start detailed design studies in 2013. ALLEGRO is a low power Gas Cooled Fast Reactor studied in the European framework. It is a loop type, non electricity generating reactor. Its power is about 80 MW. Several objectives are assigned to ALLEGRO. At first, it will demonstrate the viability of the GFR reactor system, no reactor of this type having been built in the past. Most of the GFR architecture, materials and components features are considered at reduced scale in ALLEGRO, excluding the energy conversion system. ALLEGRO will rely on the same safety options as the reactor system. In addition, the ALLEGRO core will allow the progressive qualification of the GFR ceramic fuel, with the possibility to load some ceramic carbide or nitride sub-assemblies in a first MOX core, with SiC/SiCf cladding and wrappers. When such unit test will be considered convincing enough, the diagrid and circuits are designed to accept full high temperature ceramic cores. The core neutrons can also be used to irradiate structural materials with fast neutron spectrum and in a large temperature range. The core can also include innovative irradiation fuel devices (samples or full bundles) for other reactor systems. Finally, branches on the main intermediate heat exchanger will allow the testing and validation of high temperature components and processes. The pre-conceptual design of ALLEGRO is shared between European partners through the GCFR 6th R&D Framework Program. After recalling the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: • Core design and neutron performances, • The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the MOX core, • Fuel handling principles and solutions, • System design and global reactor architecture which is largely influenced by the Decay Heat Removal strategy (DHR) for depressurized accidents.

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.

Progressive Evolution of the Core of the Fast Breeder Test Reactor

2010 ◽  
Author(s):  
S. Varatharajan ◽  
K. V. Sureshkumar ◽  
K. V. Kasiviswanathan ◽  
G. Srinivasan
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Power Reactor ◽  
Test Bed ◽  
Fast Breeder Test Reactor ◽  
The Core ◽  
Linear Heat ◽  
Mox Fuel ◽  
Test Reactor ◽  
Carbide Fuel

The second stage of Indian nuclear programme envisages the deployment of fast reactors on a large scale for the effective use of India’s limited uranium reserves. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is a loop type, sodium cooled fast reactor, meant as a test bed for the fuels and structural materials for the Indian fast reactor programme. The reactor was made critical with a unique high plutonium MK-I carbide fuel (70% PuC+30%UC). Being a unique untested fuel of its kind, it was decided to test it as a driver fuel, with conservative limits on Linear Heat Rating and burn-up, based on out-of-pile studies. FBTR went critical in Oct 1985 with a small core of 23 MK-I fuel subassemblies. The Linear Heat Rating and burn-up limits for the fuel were conservatively set at 250 W/cm & 25 GWd/t respectively. Based on out-of-pile simulation in 1994, it was possible to raise the LHR to 320 W/cm. It was decided that when the fuel reaches the target burn-up of 25 GWd/t, the MK-I core would be progressively replaced with a larger core of MK-II carbide fuel (55% PuC+45%UC). Induction of MK-II subassemblies was started in 1996. However, based on the Post-Irradiation Examination (PIE) of the MK-I fuel at 25, 50 & 100 GWd/t, it became possible to enhance the burn-up of the MK-I fuel to 155 GWd/t. More than 900 fuel pins of MK-I composition have reached 155 GWd/t without even a single failure and have been discharged. One subassembly (61 pins) was taken to 165 GWd/t on trial basis, without any clad failure. The core has been progressively enlarged, adding MK-I subassemblies to compensate for the burn-up loss of reactivity and replacement of discharged subassemblies. The induction of MK-II fuel was stopped in 2003. One test subassembly simulating the composition of the MOX fuel (29% PuO2) to be used in the 500 MWe Prototype Fast Breeder Reactor was loaded in 2003. It is undergoing irradiation at 450 W/cm, and has successfully seen a burn-up of 92.5 GWd/t. In 2006, it was proposed to test high Pu MOX fuel (44% PuO2), in order to validate the fabrication and fuel cycle processes developed for the power reactor MOX fuel. Eight MOX subassemblies were loaded in FBTR core in 2007. The current core has 27 MK-I, 13 MK-II, eight high Pu MOX and one power reactor MOX fuel subassemblies. The reactor power has been progressively increased from 10.5 MWt to 18.6 MWt, due to the progressive enlargement of the core. This paper presents the evolution of the core based on the progressive enhancement of the burn-up limit of the unique high Pu carbide fuel.

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

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.

scholarly journals Swirling Flows Characteristics in a Cylinder Under Effect of Buoyancy

Mechanika ◽  
2021 ◽  
Vol 27 (3) ◽  
pp. 201-208
Author(s):  
Mustafa FEKHAR ◽  
Rachid SACI ◽  
Renée GATIGNOL
Keyword(s):  
Aspect Ratio ◽  
Thermal Stratification ◽  
Flow Behavior ◽  
Flow Characteristics ◽  
Swirling Flows ◽  
Viscous Effects ◽  
Thermal Buoyancy ◽  
State Diagrams ◽  
The Core ◽  
Flow Stagnation

Thermal buoyancy, induced by injection or by differential heating of a tiny rod is explored to control breakdown in the core of a helical flow driven by the lid rotation of a cylinder. Three main parameters are required to characterize numerically the flow behavior; namely, the rotational Reynolds number Re, the cavity aspect ratio and the Richardson number Ri. Warm injection/rod, Ri > 0, is shown to prevent on-axis flow stagnation while breakdown enhancement is evidenced when Ri < 0. Results revealed that a bubble vortex evolves into a ring type structure which may remain robust, as observed in prior related experiments or, in contrast, disappear over a given range of parameters (Λh, Re, Ri > 0). Besides, the emergence of such a toroidal mode was not found to occur under thermal stratification induced by a differentially heated rod. Moreover, three state diagrams were established which provide detailed flow characteristics under the distinct and combined effects of buoyancy strength, viscous effects and cavity aspect ratio.

scholarly journals Characteristics of a gas-cooled fast reactor with minor actinide loading

2021 ◽  
Vol 8 (2) ◽  
pp. 1-9
Author(s):  
Hoai Nam Tran ◽  
Yasuyoshi Kato ◽  
Van Khanh Hoang ◽  
Sy Minh Tuan Hoang
Keyword(s):  
Fast Reactor ◽  
Supercritical Co2 ◽  
Cycle Length ◽  
Minor Actinide ◽  
Co2 Gas ◽  
Control Rod ◽  
The Core ◽  
Thermal Output ◽  
Long Life ◽  
Gas Turbine Cycle

This paper presents the neutronics characteristics of a prototype gas-cooled (supercritical CO2-cooled) fast reactor (GCFR) with minor actinide (MA) loading in the fuel. The GCFR core is designed with a thermal output of 600 MWt as a part of a direct supercritical CO2 (S-CO2) gas turbine cycle. Transmutation of MAs in the GCFR has been investigated for attaining low burnup reactivity swing and reducing long-life radioactive waste. Minor actinides are loaded uniformly in the fuel regions of the core. The burnup reactivity swing is minimized to 0.11% ∆k/kk’ over the cycle length of 10 years when the MA content is 6.0 wt%. The low burnup reactivity swing enables minimization of control rod operation during burnup. The MA transmutation rate is 42.2 kg/yr, which is equivalent to the production rates in 7 LWRs of the same electrical output.

Wave Propagation and Power Flow Analysis of Sandwich Structures With Internal Absorbers

2014 ◽  
Vol 136 (4) ◽  
Author(s):  
J. S. Chen ◽  
R. T. Wang
Keyword(s):  
Wave Propagation ◽  
Power Flow ◽  
Wave Attenuation ◽  
Sandwich Structures ◽  
Flow Analysis ◽  
Flow Characteristics ◽  
Continuum Models ◽  
Sandwich Beams ◽  
The Core ◽  
In Series

This study examines wave attenuation and power flow characteristics of sandwich beams with internal absorbers. Two types of absorbing systems embedded in the core are considered, namely, a conventional spring-mass-dashpot system having a mass with a spring and a dashpot in parallel, and a relaxation system containing an additional relaxation spring added in series with the dashpot. Analytical continuum models used for interpreting the attenuation behavior of sandwich structures are presented. Through the analysis of the power flowing into the structure, the correlation of wave attenuation and energy blockage is revealed. The reduction in the power flow indicates that some amount of energy produced by the external force can be effectively obstructed by internal absorbers. The effects of parameters on peak attenuation, bandwidth, and power flow are also studied.

Effects of Core Flow Swirl on the Flow Characteristics of a Scalloped Forced Mixer

2011 ◽  
Author(s):  
Zhijun Lei ◽  
Ali Mahallati ◽  
Mark Cunningham ◽  
Patrick Germain
Keyword(s):  
Pressure Ratio ◽  
Velocity Ratio ◽  
Flow Characteristics ◽  
Pressure Probe ◽  
Streamwise Vortices ◽  
Core Flow ◽  
The Core ◽  
Flow Swirl ◽  
Swirl Angle ◽  
Oil Flow

This paper presents a detailed experimental investigation of the influence of core flow swirl on the mixing and performance of a scaled turbofan mixer with 12 scalloped lobes. Measurements were made downstream of the mixer in a co-annular wind tunnel. The core-to-bypass velocity ratio was set to 2:1, temperature ratio to 1.0, and pressure ratio to 1.03, giving a Reynolds number of 5.2 × 105, based on the core flow velocity and equivalent hydraulic diameter. In the core flow, the background turbulence intensity was raised to 5% and the swirl angle was varied using five vane geometries from 0° to 30°. Seven-hole pressure probe measurements and surface oil flow visualization were used to describe the flowfield and the mixer performance. At low swirl angles, additional streamwise vortices were generated by the deformation of normal vortices due to the scalloped lobes. With increased core swirl, greater than 10°, the additional streamwise vortices were generated mainly due to radial velocity deflection, rather than stretching and deformation of normal vortices. At high swirl angles, stronger streamwise vortices and rapid interaction between various vortices promoted downstream mixing. Mixing was enhanced with minimal or no total pressure and thrust losses for the inlet swirl angles less than 10°. However, the reversed flow downstream of the center-body was a dominant contributor to the loss of thrust at the maximum core flow swirl angle of 30°.

scholarly journals Thorium application in the medium-sized sodium-cooled fast reactor

2019 ◽  
Vol 137 ◽  
pp. 01030
Author(s):  
Eeshu Raaj Saasthaa Arumuga Kumar ◽  
Piotr Darnowski ◽  
Mihir Kiritbhai Pancholi ◽  
Aleksandra Dzido
Keyword(s):  
Fast Reactor ◽  
Waste Incineration ◽  
Computer Code ◽  
Fuel Cycles ◽  
Mixed Oxide Fuel ◽  
The Core ◽  
Long Time ◽  
The Impact ◽  
Thorium 232 ◽  
Full Core

The report presents an analysis of the medium-sized Sodium-Cooled Fast Reactor (SFR) core with Thorium-based Mixed-Oxide fuel. The introduction of Transuranics (TRU) to the fuel was to allow long-lived nuclear waste incineration. The studied core is based on the modified Advanced Burner Reactor (ABR) 1000MWth core design, which was analysed in the OECD/NEA “Benchmark for Neutronic Analysis of Sodium-Cooled Fast Reactor Cores with Various Fuel Types and Core Sizes”. The full-core simulations with SERPENT 2.1.31 Monte Carlo computer code and ENDF library were performed, including static criticality and fuel burnup calculations for five fuel cycles. The core inventories at the Beginning of Cycle (BOC) and End of Cycle (EOC) were studied, and the impact of thorium fuel was assessed. The proposed core design is a burner reactor which uses thorium fuel. The excess core reactivity stays positive for long time despite large net consumption of transuranic elements as new fissile Uranium 233 is constantly breed from Thorium 232. Breeding of uranium allows longer fuel cycles.

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