Machine Learning-Based Methodology for Assessment of Doppler Reactivity of Sodium-Cooled Fast Reactor

2021 ◽  
Vol 7 (4) ◽  
Author(s):  
Đorđe Petrović ◽  
Konstantin Mikityuk
Keyword(s):  
Machine Learning ◽  
Fast Reactor ◽  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Fourth Generation ◽  
Fuel Temperature ◽  
Horizon 2020 ◽  
Classical Methodology ◽  
Artificial Neural Network Ann

Abstract In order to close nuclear fuel cycle and address the problem of sustainability, advanced nuclear reactor systems of the fourth generation are in the focus of the research for many years. With a simple goal of supporting this research, machine learning-based methodology for the assessment of the Doppler reactivity has been developed and applied to the European Sodium Fast Reactor (ESFR) in the frame of the ESFR-Safety Measures Assessment and Research Tools (SMART) Horizon-2020 project. In the scope of this study, a database of the precise Monte Carlo (MC) calculations was prepared and used to train artificial neural network (ANN) as a surrogate model to assess the Doppler reactivity across the range of reactor conditions that could occur throughout the life of the reactor core, in fast, yet accurate manner. The database was generated for all the combinations of several core parameters carefully predefined in order to account for both operational and accidental states of the core. Subsequently, Doppler reactivity change as a function of the above-mentioned parameters was assessed by herein developed methodology, as well as by widely used logarithmic dependence of the Doppler reactivity on the fuel temperature and compared to the results of the precise MC simulations. This study proves that, if certain computational resources are allocated to the database generation and ANN training, newly developed methodology yields similar or even more accurate results compared to the classical methodology and at the same time provides a tool for parameterization and interpolation of Doppler reactivity not only on the fuel temperature but also on the other parameters characterizing core of the sodium-cooled fast reactor (SFR).

scholarly journals Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

Nukleonika ◽  
2015 ◽  
Vol 60 (3) ◽  
pp. 581-590 ◽  
Author(s):  
Przemysław Stanisz ◽  
Jerzy Cetnar ◽  
Grażyna Domańska
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Nuclear Fuel Cycle ◽  
Minor Actinide ◽  
Closed Nuclear Fuel Cycle ◽  
European Collaboration ◽  
Multiple Recycling

Abstract The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR) was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System) and LEADER (Lead-cooled European Advanced Demonstration Reactor) projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA) are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs), and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB) code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

Unprotected Loss of Flow Analysis of a Million Kilowatt Traveling Wave Reactor Core

2017 ◽  
Author(s):  
Wenjun Hu ◽  
Pengrui Qiao
Keyword(s):  
Fast Reactor ◽  
Traveling Wave ◽  
Nuclear Reactor ◽  
Flow Analysis ◽  
Reactor Core ◽  
Travelling Wave ◽  
Nuclear Nonproliferation ◽  
Utilization Rate ◽  
Technological Level ◽  
Long Lifetime

Traveling wave reactor (TWR) is an innovation concept nuclear reactor, through the once-through deep burning, the proliferation of fuel can be achieved and the utilization rate of Uranium can be increased. TWR has the characteristics of long lifetime, deep burn up and nuclear nonproliferation, because of its physical character, which makes it to be an attractive innovation concept fast reactor. The China institute of atomic energy (CIAE) has designed a million kilowatt TWR core based on a breeding and burn principle, which has considered the current technological level of sodium cooled fast reactor. In this paper, based on the TWR core design scheme, considered the design of fuel assembly, neutronics and thermal-hydraulic, analyzed the Unprotected loss of flow (ULOF) accident in the TWR core with the SAS4A code, through which research about the transient safety characteristics of a million kilowatt travelling wave reactor core has been done. Analysis shows that the peak temperature of fuel, cladding and coolant in the TWR core have a certain margin from the safety limits through the negative feedback of itself in the ULOF accident, the core of the million kilowatt TWR demonstrates a good safety performance.

A Procedure for the Pre-Conceptual Design of Fast Reactors and Application to a Gas-Cooled Sub-Critical Transmuter

2014 ◽  
Author(s):  
Carlo Fiorina ◽  
Konstantin Mikityuk ◽  
Jiři Křepel
Keyword(s):  
Conceptual Design ◽  
Fast Reactor ◽  
Power Distribution ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Maximum Pressure ◽  
Test Case ◽  
Pressure Losses ◽  
Design And Optimization ◽  
Inert Matrix

A C++ procedure has been developed for the design and optimization of Fast Reactor (FR) cores. It couples the ERANOS based EQL3D procedure developed at PSI for FR equilibrium fuel cycle analysis with a dedicated MATLAB script that evaluates the thermal-hydraulic characteristics of the reactor core. It is conceived to investigate reactors with both standard pins and annular pins. The procedure accepts as input the physical properties of the system, as well as a set of target core parameters presently consisting of core power, maximum fuel burnup, multiplication factor, inner pin diameter (for annular pins) or maximum pressure losses (for standard pins), and core height. It gives as a result a core design fulfilling these design objectives and meeting the constraints on maximum fuel and clad temperatures. In case of annular pins, it also equalizes the temperature rise inside and outside of the core average pin. The procedure considers the possibility of two-zone cores and adjusts the fuel composition in the two zones to achieve an optimal radial power distribution. Finally, it can evaluate safety parameters and fuel cycle characteristics both at beginning-of-life and at equilibrium. As a test case, the procedure has been used for the pre-conceptual design of a sub-critical Gas Fast Reactor core employing inert-matrix sphere-pac fuel and annular pins with SiC cladding.

scholarly journals Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

2017 ◽  
Vol 153 ◽  
pp. 07031
Author(s):  
Georgy Tikhomirov ◽  
Mikhail Ternovykh ◽  
Ivan Saldikov ◽  
Peter Fomichenko ◽  
Alexander Gerasimov
Keyword(s):  
Isotopic Composition ◽  
Nuclear Fuel ◽  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Nuclear Fuel Cycle ◽  
Closed Nuclear Fuel Cycle

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.

Void Reactivity Effect Research on Liquid Metal Cooled Fast Reactor

2017 ◽  
Author(s):  
Yang Lyu ◽  
Xiao Liang
Keyword(s):  
Liquid Metal ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Reactor Core ◽  
Nuclear Industry ◽  
Fourth Generation ◽  
Reactivity Effect ◽  
The Core ◽  
Reactivity Coefficient

In the fourth generation of advanced nuclear power systems, the liquid metal cooled fast reactor plays a more and more important role, such as SFR, LFR and ADS system with LBE coolant. Void reactivity effect means bubbles produced in the core area will induce the change of reactivity. And this reactivity will affect the safety of the reactor. Through investigation and comparison of several liquid metal cooled fast reactors in the nuclear industry, this paper studies bubbles in different positions and partial voiding of the active zone inside the core and fuel assemblies with Monte Carlo core physics calculation method and then concludes the main influencing factors of void reactivity coefficient. The results can provide reference for the follow-up research and development of new type liquid metal fast reactor core design.

Unprotected Overpower Transient Analysis of a Million Kilowatt Traveling Wave Reactor Core

2017 ◽  
Author(s):  
Pengrui Qiao ◽  
Jian Zhang ◽  
Chao Lin
Keyword(s):  
Fast Reactor ◽  
Traveling Wave ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Travelling Wave ◽  
Nuclear Nonproliferation ◽  
Utilization Rate ◽  
Fuel Cladding ◽  
Technological Level ◽  
Long Lifetime

Traveling wave reactor (TWR) is an innovation concept nuclear reactor, through the once-through deep burning, the proliferation of fuel can be achieved and the utilization rate of Uranium can be increased. TWR has the characteristics of long lifetime, deep burn up and nuclear nonproliferation, because of its physical character, which makes it to be an attractive innovation concept fast reactor. The China institute of atomic energy (CIAE) has designed a million kilowatt TWR core based on a breeding and burn principle, which has considered the current technological level of sodium cooled fast reactor. In this paper, based on the TWR core design scheme, considered the design of fuel assembly, neutronics and thermal-hydraulic, analyzed the unprotected overpower transient (UTOP) accident in the TWR core with the SAS4A code, through which research about the transient safety characteristics of a million kilowatt travelling wave reactor core has been done. Analysis shows that the peak temperature of fuel, cladding and coolant in the TWR core have a certain margin from the safety limits through the negative feedback of itself in the UTOP accident, the core of the million kilowatt TWR demonstrates a good safety performance.

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.

Preliminary Core Design of the Long Life S-CO2 Cooled Fast Reactor With 300WMth

2014 ◽  
Author(s):  
Baolin Liu ◽  
Hongchun Wu ◽  
Youqi Zheng ◽  
Liangzhi Cao ◽  
Xianbao Yuan
Keyword(s):  
Fast Reactor ◽  
Nuclear Power ◽  
Power Plants ◽  
Single Channel ◽  
Reactor Core ◽  
Fuel Temperature ◽  
The Core ◽  
Long Life ◽  
High Conversion Ratio ◽  
Thermal Hydraulic Analysis

Gas cooled fast reactors are one of the Generation 4 nuclear power plants with hard neutron spectrum and high conversion ratio. In the study a long life Supercritical CO2 (S-CO2) cooled fast reactor core design with 300 MWth is presented. Physical calculation was carried out based on Dragon and CITATION, and thermal hydraulic analysis was performed based on the single channel code. The MOX fuel was utilized in the core design, and the tube-in-duct (TID) assembly was chosen for its excellent characteristics. According to the physical and thermal hydraulic coupling calculation, the reactor in the study can be operated with 300MWth for 20Ys without shuffling or refueling. Through the core life power peaking was kept relatively low, and the fuel temperature was kept below the 1800 degree centigrade.

scholarly journals A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle

2013 ◽  
Vol 2013 ◽  
pp. 1-10 ◽  
Author(s):  
Hangbok Choi ◽  
Robert W. Schleicher ◽  
Puja Gupta
Keyword(s):  
Fast Reactor ◽  
Nuclear Power ◽  
Fuel Cycle ◽  
Matrix Material ◽  
Reactor Core ◽  
Operation Time ◽  
Fuel Load ◽  
Fissile Material ◽  
Reactor Physics

In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2), which is a compact gas-cooled fast reactor (GFR). The EM2augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.

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