A Study on the Influences of Reactivity Feedback Mechanisms in China Experimental Fast Reactor Unprotected Transients

2013 ◽  
Author(s):  
Yuanyu Wu ◽  
Hong Yu ◽  
Lixia Ren ◽  
Wenjun Hu ◽  
Hongtao Qian
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Inherent Safety ◽  
Fuel Temperature ◽  
Feedback Mechanisms ◽  
Control Rod ◽  
Severe Accidents ◽  
Proper Design ◽  
China Experimental Fast Reactor

Inherent safety properties of reactor have always played an important role in severe accidents preventing and consequences mitigation. With proper design, reactivity feedback mechanisms can bring benign reactivity feedbacks to the reactor core during unprotected transients, thus contributing to the severe accidents mitigation. In overpower transients, the increasing power causes the fuel temperature to increase, which directly brings fuel Doppler feedback and core axial expansion feedback. In unprotected loss-of-flow accidents, as the flow rate decreases, the mismatch of power and flow causes the increase of coolant temperature, thus directly resulting in the coolant reactivity, core radial expansion as well as the control rod driveline expansion feedbacks. Through the simulation of China Experimental Fast Reactor (CEFR) unprotected transients, the influences of different reactivity feedback mechanisms have been investigated and analyzed. The coolant reactivity exhibits significant negative feedback and makes the dominant contribution in controlling the reactivity in both UTOP and ULOF transients.

scholarly journals Improved FAST Device for Inherent Safety of Oxide-Fueled Sodium-Cooled Fast Reactors

Energies ◽  
2021 ◽  
Vol 14 (15) ◽  
pp. 4610
Author(s):  
Ahmed Amin E. Abdelhameed ◽  
Chihyung Kim ◽  
Yonghee Kim
Keyword(s):  
Operating Temperature ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Inherent Safety ◽  
Oxide Fuel ◽  
Fast Reactors ◽  
Power Oscillation ◽  
Negative Reactivity ◽  
Oscillation Problem ◽  
Reactivity Coefficient

The floating absorber for safety at transient (FAST) was proposed as a solution for the positive coolant temperature coefficient in sodium-cooled fast reactors (SFRs). It is designed to insert negative reactivity in the case of coolant temperature rise or coolant voiding in an inherently passive way. The use of the original FAST design showed effectiveness in protecting the reactor core during some anticipated transients without scram (ATWS) events. However, oscillation behaviors of power due to refloating of the absorber module in FAST were observed during other ATWS events. In this paper, we propose an improved FAST device (iFAST), in which a constraint is imposed on the sinking (insertion) limit of the absorber module in FAST. This provides a simple and effective solution to the power oscillation problem. Here, we focus on an oxide fuel-loaded SFR that is characterized by a more negative Doppler reactivity coefficient and higher operating temperature than the metallic-loaded SFR cores. The study is carried out for the 1000 MWth advanced burner reactor with an oxide fuel-loaded core during postulated ATWS events that are unprotected transient over power, unprotected loss of flow, and unprotected loss of the heat sink. It was found that the iFAST device has promising potentials for protecting the oxide SFR core during the various studied ATWS events.

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.

Numerical Simulation and Analysis of Temperature Fluctuation in Core Outlet Zone of China Experimental Fast Reactor

2013 ◽  
Author(s):  
Huanjun Zhu ◽  
Yijun Xu
Keyword(s):  
Fast Reactor ◽  
Temperature Fluctuation ◽  
Mechanical Damage ◽  
Low Frequency ◽  
Normal Operation ◽  
Large Temperature ◽  
Control Rod ◽  
Sector Model ◽  
And Control ◽  
China Experimental Fast Reactor

As for the fast reactor, the temperature fluctuation in core outlet zone in normal operation was an important and typical thermal hydraulic phenomenon. In certain conditions, these temperature fluctuations can lead to thermal-mechanical damage to the Upper Core Structures (UCS). In this paper, the temperature fluctuation in Core Outlet Region of China Experimental Fast Reactor (CEFR) was numerically simulated and researched using the CFD software FLUENT. In the simulation process, LES turbulence method was used for the turbulent model selection. For the mesh generation, 1/4 sector model was built considering the layout of the symmetry and very small-size cell was used for satisfied with the Corant number. At the same time the core outlet temperatures and flows of the coolant in each subassembly under reactor rated power was used as the boundary conditions. The simulation shows that the temperature fluctuation was mainly concentrated in the zone between the fuel subassembly and control rod assembly for the large temperature difference of these subassemblies CEFR was operated in normal conditions. The largest fluctuation amplitude was 19°C and the remarkable frequency was below 5Hz. It belongs to typically low frequency fluctuation. The conclusion was useful for further experimental work and reactor operation.

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Massimo Sarotto ◽  
Gabriele Firpo ◽  
Anatoly Kochetkov ◽  
Antonin Krása ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Control Rod ◽  
The Core ◽  
Major Interest ◽  
Reactor Spectrum ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Sensitivity And Uncertainty Analysis

Abstract During the EURATOM FP7 project FREYA, a number of experiments were performed in a critical core assembled in the VENUS-F zero-power reactor able to reproduce the ALFRED lead-cooled fast reactor spectrum in a dedicated island. The experiments dealt with the measurements of integral and local neutronic parameters, such as the core criticality, the control rod and the lead void reactivity worth, the axial distributions of fission rates for the nuclides of major interest in a fast spectrum, the spectral indices of important actinides (238U, 239Pu, 237 Np) with respect to 235U. With the main aim to validate the neutronic codes adopted for the ALFRED core design, the VENUS-F core and its characterization measurements were simulated with both deterministic (ERANOS) and stochastic (MCNP, SERPENT) codes, by adopting different nuclear data libraries (JEFF, ENDF/B, JENDL, TENDL). This paper summarizes the main results obtained by highlighting a general agreement between measurements and simulations, with few discrepancies for some parameters that are discussed here. Additionally, a sensitivity and uncertainty analysis was performed with deterministic methods for the core reactivity: it clearly indicates that the small over-criticality estimated by the different codes/libraries resulted to be lower than the uncertainties due to nuclear data.

scholarly journals MULTI-PHYSICS TRANSIENT SIMULATIONS WITH TRIPOLI-4®

2021 ◽  
Vol 247 ◽  
pp. 07019 ◽  
Author(s):  
Margaux Faucher ◽  
Davide Mancusi ◽  
Andrea Zoia
Keyword(s):  
Monte Carlo ◽  
Neutron Transport ◽  
Coolant Temperature ◽  
Test Case ◽  
Monte Carlo Code ◽  
Thermal Hydraulics ◽  
Fuel Temperature ◽  
Control Rod ◽  
Transient Simulations ◽  
Neutron Power

In this work, we present the first dynamic calculations performed with the Monte Carlo neutron transport code TRIPOLI-4R with thermal-hydraulics feedback. For this purpose, the Monte Carlo code was extended for multi-physics capabilities and coupled to the thermal-hydraulics subchannel code SUBCHANFLOW. As a test case for the verification of transient simulation capabilities, a 3x3-assembly mini-core benchmark based on the TMI-1 reactor is considered with a pin-by-pin description. Two reactivity excursion scenarios initiated by control-rod movement are simulated starting from a critical state and compared to analogous simulations performed using the Serpent 2 Monte-Carlo code. The time evolution of the neutron power, fuel temperature, coolant temperature and coolant density are analysed to assess the multi-physics capabilities of TRIPOLI-4. The stabilizing e_ects of thermal-hydraulics on the neutron power appear to be well taken into account. The computational requirements for massively parallel calculations are also discussed.

Optimization of the TiO2 nanofluid as a coolant in the VVER-1000 nuclear reactor based on the thermal reactivity feedback coefficients via the genetic algorithm

Kerntechnik ◽  
2021 ◽  
Vol 86 (6) ◽  
pp. 419-436
Author(s):  
R. Kianpour ◽  
G. R. Ansarifar
Keyword(s):  
Neural Network ◽  
Nuclear Reactor ◽  
Volume Fraction ◽  
Reactor Core ◽  
Dynamical Analysis ◽  
Coolant Temperature ◽  
Optimal Size ◽  
Fuel Temperature ◽  
Volume Fractions ◽  
Reactivity Coefficients

Abstract The purpose of this study is to display the neutronic simulation of nanofluid application to reactor core. The variations of VVER-1000 nuclear reactor primary neutronic parameters are investigated by using different volume fraction of nanofluid as coolant. The effect of using nanofluid as coolant on reactor dynamical parameters which play an important role in the dynamical analysis of the reactor and safety core is calculated. In this paper coolant and fuel temperature reactivity coefficients in a VVER-1000 nuclear reactor with nanofluid as a coolant are calculated by using various volume fractions and different sizes of TiO2 (Titania) nanoparticle. For do this, firstly the equivalent cell of the hexagonal fuel rod and the surrounding coolant nanofluid is simulated. Then the thermal hydraulic calculations are performed at different volume fractions and sizes of the nanoparticle. Then, using WIMS and CITATION codes, the reactor core is simulated and the effect of coolant and fuel temperature changes on the effective multiplication factor is calculated. For doing optimization, an artificial neural network is trained in MATLAB using the observed data. The different sizes and various volume fractions are inputs, fuel and coolant temperature reactivity coefficients are outputs. The optimal size and volume fraction is determined using the neural network by implementing the genetic algorithms. In the optimization, volume fraction of 7% and size 77 nm are optimal values.

Thermal Evaluation for a Ultra High Temperature Reactor With Pebble Type Fuels

2009 ◽  
Author(s):  
Motoo Fumizawa ◽  
Yuta Kosuge ◽  
Hidenori Horiuchi
Keyword(s):  
High Temperature ◽  
Power Density ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Small Scale ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
System Pressure ◽  
High Temperature Reactor ◽  
Ultra High Temperature

This study presents a predictive thermal-hydraulic analysis with packed spheres in a nuclear gas-cooled reactor core. The predictive analysis considering the effects of high power density and the some porosity value were applied as a design condition for an Ultra High Temperature Reactor (UHTR). The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 °C was achieved in the case of an UHTREX at LASL, which was a small scale UHTR using hollow-rod as a fuel element. In the present study, the fuel was changed to a pebble type, a porous media. Several calculation based on HTGR-GT300 through GT600 were 4.8 w/cm3 through 9.6 w/cm3 respectively. As a result, the relation between the fuel temperature and the power density was obtained under the different system pressure and coolant outlet temperature. Finally, available design conditions are selected.

scholarly journals Core Design of a Small Pressurized Water Reactor with AP1000 Fuel Assembly Using SRAC and COBRA-EN Codes

2020 ◽  
Vol 2020 ◽  
pp. 1-7
Author(s):  
Van Khanh Hoang
Keyword(s):  
Safety Margin ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Control Rod ◽  
The Core ◽  
Small Modular Reactor ◽  
Active Core ◽  
And Performance

This paper presents the core design and performance characteristics of a 300 MWt small modular reactor (SMR) with fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system with the JENDL-4.0 data library and the CORBRA-EN code. The calculated temperature coefficients including fuel temperature, coolant temperature, and isothermal temperature coefficient provide adequate negative reactivity feedbacks. The thermal-hydraulic analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel and clad surface temperature. A safety analysis using the CORBRA-EN code shows that the core will remain covered during the entire transient procedure of the fast transient of remarkably increasing power that would be caused by the ejection of control rod. The analysis results indicate that the core with a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

scholarly journals Model of Reactivity Accident of the RBMK-1000 of the Chornobyl NPP 4th Power Unit

2021 ◽  
Vol 21 (2) ◽  
pp. 39-48
Author(s):  
V. I. Borysenko ◽  
◽  
V. V. Goranchuk ◽  
Keyword(s):  
Power Unit ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Fuel Temperature ◽  
Reactor Accident ◽  
Reactivity Coefficients ◽  
Accident Scenarios ◽  
Control Rods ◽  
Reactivity Accident

The article presents the results of modeling of the reactivity accident, which resulted in the destruction of reactor RBMK-1000 of the 4th power unit of the Chornobyl NPP on April 26, 1986. The RBMK-1000 reactivity accident model was developed on the basis of the kinetics of the nuclear reactor, taking into account the change in the reactivity of the reactor. Reactivity changes as a result of both external influence (movement of control rods; change in the reactor inlet coolant temperature (density)) and due to the action of reactivity feedback by the parameters of the reactor core (change in the fuel temperature, coolant temperature, concentration of 135Хе, graphite stack temperature, etc.). A similar approach was applied by the authors of the article for the study of transient processes with the operation of accelerated unit unloading mode on VVER-1000, and the validity of such model is confirmed. The study of the reactivity accident on RBMK-1000 was carried out for various combinations of values of the effectiveness of control rods; reactivity coefficients of the coolant temperature and fuel temperature; changes in the temperature of the coolant at the inlet to the reactor. In most of the studied RBMK-1000 reactor accident scenarios, the critical values of fuel enthalpy, at which the process of fuel destruction begins, are reached first. An important result of the research is the conclusion that it is not necessary to reach supercriticality on instantaneous neutrons, supercriticality on delayed neutrons is also sufficient to initiate fuel destruction.

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