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

Assessment of the nano-fluid effects on the thermal reactivity feedback coefficients in the VVER-1000 nuclear reactor with nano-fluid as a coolant using thermal hydraulic and neutronics analysis

Optimal design of a VVER-1000 nuclear reactor core with dual cooled annular fuel based on the reactivity temperature coefficients using Thermal hydraulic and neutronic analysis by implementing the genetic algorithms

Design of the CAREM nuclear reactor core with dual cooled annular fuel and optimizing the thermal-hydraulic, natural circulation, and neutronics parameters

Neutronic analysis of new generation TVS-2M fuel assemblies in VVER-1000 reactor type

VVER-Type Reactors of Russian Design

This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al2O3 nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

A new approach to nuclear reactor design optimization using genetic algorithms and regression analysis

Investigation of burnable poisons effects in reactor core design

Investigation of the nano fluid effects on heat transfer characteristics in nuclear reactors with dual cooled annular fuel using CFD (Computational Fluid Dynamics) modeling

Study of power distribution in the CZP, HFP and normal operation states of VVER-1000 (Bushehr) nuclear reactor core by coupling nuclear codes

Numerical study of single and two-phase models of water/Al2O3 nanofluid turbulent forced convection flow in VVER-1000 nuclear reactor

A coupled thermohydraulic/neutronic investigation of noise propagation in VVER-1000 reactor using[formula omitted] nanofluid coolant

The BIPR-8A executable module is a main component of the BIPR 2007 code. This module is intended for three-dimensional coarse mesh calculation of the VVER reactor cores. This paper presents the results of validation of the BIPR-8A executable module. Validation is based on comparison of the calculated results with the operating data of power plants with the VVER-440, VVER-1000 and VVER-1200 reactors, and with the results of calculations obtained by using Monte Carlo codes. The results of comparison of the following values at HZP (hot zero power) power level and operational power level are presented: critical concentration of boric acid, temperature reactivity coefficient, efficiency of the control rods system, relative power distribution of fuel assemblies and 3D power distribution, effective multiplication factor.

<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>

The implementation of nanofluid within a preferred volume fraction in base fluid is one of the innovative and inexpensive methods of enhancing the thermal and hydrodynamic behaviour in terms of increased heat transfer rate and reduced pumping power. In the present paper, forced convection heat transfer and pumping power were studied for a rough parallel plate subjected to constant heat flux under turbulent flow condition. The investigation was performed for a wide range of Reynolds number 10,000 to 30,000 with Al2O3 nanoparticles volume fraction 1% to 5% dispersed in base fluid water and three different rough surfaces (relative roughness: 0.001, 0.002, and 0.003) were considered. Heat transfer performance was substantially improved more by employing roughness at the wall of the boundary than the smooth parallel plate and also for the increased more volume fraction of nanofluid than base fluid water. Augmentation was found significant at higher surface roughness and volume fraction by the virtue of superior thermo-physical properties of nanofluid up to 36.9% and 26.1% for the rough surface. Roughness did not increase the pumping power as its effect was mitigated by the nanofluid. 2% volume fraction of Al2O3-water nanofluid on 0.003 relative roughness that showed the superior behaviour for heat transfer enhancement with minimum pumping power requirement.