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

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.

2021 ◽  
Vol 21 (2) ◽  
pp. 39-48
Author(s):  
V. I. Borysenko ◽  
◽  
V. V. Goranchuk ◽  

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.


Author(s):  
Chi Wang ◽  
Xuebei Zhang ◽  
Jingchao Feng ◽  
Muhammad Shehzad Khan ◽  
Minyou Ye ◽  
...  

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.


2021 ◽  
Vol 11 (1) ◽  
pp. 9-15
Author(s):  
Van Khanh Hoang ◽  
Vinh Thanh Tran ◽  
Dinh Hung Cao ◽  
Viet Ha Pham Nhu

This work presents the neutronic analysis of fuel design for a long-life core in a pressurized water reactor (PWR). In order to achieve a high burnup, a high enrichment U-235 is traditionally considered without special constraints against proliferation. To counter the excess reactivity, Erbium was selected as a burnable poison due to its good depletion performance. Calculations based on a standard fuel model were carried out for the PWR type core using SRAC code system. A parametric study was performed to quantify the neutronically achievable burnup at a number of enrichment levels and for a numerous geometries covering a wide design space of lattice pitch. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are also investigated. It was found that it is possible to achieve sufficient criticality up to 100 GWd/tHM burnup without compromising the safety parameters.


Author(s):  
Motoo Fumizawa ◽  
Yuta Kosuge ◽  
Hidenori Horiuchi

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.


Author(s):  
Yuanyu Wu ◽  
Hong Yu ◽  
Lixia Ren ◽  
Wenjun Hu ◽  
Hongtao Qian

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.


Author(s):  
Xiaosheng Li ◽  
Linsen Li ◽  
Lianghui Peng ◽  
Xiaosong Chen ◽  
Zhaocan Meng ◽  
...  

The pressure and coolant temperature of Heating-reactor of Advanced low-Pressurized and Passive safetY system (HAPPY200) is significantly lower than PWR of the NPP, the core design and analysis were completed according to the design parameters and features of HAPPY200. The fuel assembly and its feature was firstly designed and studied based on the investigation of different types of fuel assemblies. Then the core configuration was studied and optimized according to the design parameters of HAPPY200; Eventually, neutronics calculation of the core was performed and key parameters were obtained including cycle length, power distribution, control rod worth, reactivity coefficients and etc. The study shows that with the core design HAPPY200 can be operated for 18 months in full power and reactivity control system can maintain criticality of the core in the full cycle. Due to the non-soluble boron design of the reactivity control scheme, moderator temperature coefficient and isothermal temperature coefficient are both negative, the Doppler temperature coefficients and power coefficients in different phase of the lifetime and in different power levels are also negative, therefore, the reactivity safety of the reactor core can be ensured.


KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Rokh Madi

<p>Doppler coefficient is defined as a relation between fuel temperature changes and reactivity changes in the nuclear reactor core. Doppler reactivity coefficient needs to be known because of its relation to the safety of reactor operation. This study is intended to determine the safety level of the  typical PWR-1000 core by calculating the Doppler reactivity coefficient in the core with UO<sub>2</sub> and MOX fuels. The  typical PWR-1000 core  is similar to the PWR AP1000 core designed by Westinghouse but without Integrated Fuel Burnable Absorber (IFBA) and Pyrex. Inside the core, there are  UO<sub>2</sub> fuel elements with 3.40 % and 4.45 % enrichment, and MOX fuel elements with 0.2 % enrichment. By its own way, the presence of Plutonium in the MOX fuel will contribute to the change in core reactivity. The calculation was conducted using MCNPX code with the ENDF/B- VII nuclear data. The reactivity change was investigated at various temperatures. The calculation results show that the core reactivity coefficient of both UO<sub>2</sub> and MOX fuel are negative, so that the reactor is operated safely.</p>


2021 ◽  
Vol 7 (4) ◽  
Author(s):  
Đorđe Petrović ◽  
Konstantin Mikityuk

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).


2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Edyene Oliveira ◽  
Victor Castro ◽  
Carlos Velasquez ◽  
Claubia Pereira

An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reloading. The main bounding parameter of the modeling was the neutron multiplication factor, keff. The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. As for the artificial neural network, the ANN Feedforward Multi_Layer_Perceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). The artificial neural network has implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core.


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