scholarly journals Experimental Investigation of the Coolant Flow in the VVER Reactor Core with TVSA Fuel Assemblies

2020 ◽  
Vol 2020 (4) ◽  
pp. 25-36
Author(s):  
Sergej Mihajlovich Dmitriev ◽  
Anton Vladimirovich Gerasimov ◽  
Alexander Alekseevich Dobrov ◽  
Denis Vladimirovich Doronkov ◽  
Aleksej Nikolaevich Pronin ◽  
...  
Keyword(s):  
Experimental Investigation ◽  
Reactor Core ◽  
Coolant Flow ◽  
Vver Reactor ◽  
Fuel Assemblies

scholarly journals Investigation of Coolant Local Hydrodynamics in the Mixed Core of the VVER Reactor

2020 ◽  
Vol 63 (2) ◽  
pp. 151-162
Author(s):  
S. M. Dmitriev ◽  
A. V. Gerasimov ◽  
A. A. Dobrov ◽  
D. V. Doronkov ◽  
A. N. Pronin ◽  
...  
Keyword(s):  
Cross Sections ◽  
Experimental Studies ◽  
Reactor Core ◽  
Cross Flow ◽  
Coolant Flow ◽  
Vver Reactor ◽  
Computational Hydrodynamics ◽  
Cross Section Area ◽  
Fuel Assemblies ◽  
Flow Pressure

The article presents the results of experimental studies of the local hydrodynamics of the coolant flow in the mixed core of the VVER reactor, consisting of the TVSA-T and TVSA-T mod.2 fuel assemblies. Modeling of the flow of the coolant flow in the fuel rod bundle was carried out on an aerodynamic test stand. The research was carried out on a model of a fragment of a mixed core of a VVER reactor consisting of one TVSA-T segment and two segments of the TVSA-T.mod2. The flow pressure fields were measured with a five-channel pneumometric probe. The flow pressure field was converted to the direction and value of the coolant velocity vector according to the dependencies obtained during calibration. To obtain a detailed data of the flow, a characteristic cross-section area of the model was selected, including the space cross flow between fuel assemblies and four rows of fuel rods of each of the TVSA fuel assemblies. In the framework of this study the analysis of the spatial distribution of the projections of the velocity of the coolant flow was fulfilled that has made it possible to pinpoint regularities that are intrinsic to the coolant flowing around spacing, mixing and combined spacing grates of the TVSA. Also, the values of the transverse flow of the coolant caused by the flow along hydraulically nonidentical grates were determined and their localization in the longitudinal and cross sections of the experimental model was revealed. Besides, the effect of accumulation of hydrodynamic flow disturbances in the longitudinal and cross sections of the model caused by the staggered arrangement of hydraulically non-identical grates was determined. The results of the study of the coolant cross flow between fuel assemblies interaction, i.e. between the adjacent TVSA-T and TVSA-T mod.2 fuel assemblies were adopted for practical use in the JSC of “Afrikantov OKB Mechanical Engineering” for assessing the heat engineering reliability of VVER reactor cores; also, they were included in the database for verification of computational hydrodynamics programs (CFD codes) and for detailed cell-based calculation of the reactor core.

scholarly journals Experimental investigation of the coolant flow in the VVER reactor core with TVSA fuel assemblies

2021 ◽  
Vol 7 (1) ◽  
pp. 59-54
Author(s):  
Sergey M. Dmitriyev ◽  
Anton V. Gerasimov ◽  
Aleksander A. Dobrov ◽  
Denis V. Doronkov ◽  
Aleksey N. Pronin ◽  
...  
Keyword(s):  
Fluid Dynamics ◽  
Axial Velocity ◽  
Velocity Vector ◽  
Reactor Core ◽  
Coolant Flow ◽  
Vver Reactor ◽  
Dynamics Simulation ◽  
Channel Calculation ◽  
Fuel Bundle ◽  
Fuel Assemblies

The paper presents the results of an experimental study to investigate the coolant interaction in adjoining fuel assemblies in the VVER reactor core composed of TVSA-T and upgraded TVSA FAs. The processes of the in-core coolant flow were simulated in a test wind tunnel. The experiments were conducted using models representing different portions of the VVER reactor core fuel bundle and consisted in measuring the radial and axial airflow velocities in representative areas within the FAs and in the interassembly space. The results of the experiments can be translated to the full-scale conditions of the coolant flow with the use of the fluid dynamics simulation theory. The measurements were performed using a five-channel pressure-tube probe. The coolant flow pattern in different portions of the fuel bundle is represented by distribution diagrams and distribution maps for the radial and axial velocity vector components in the representative areas of the models. An analysis for the spatial distribution of the radial and axial velocity vector components has made it possible to obtain a detailed pattern of the coolant flow about the FA spacer, mixing and combined spacer grids of different designs. The accumulated database for the coolant flow in FAs of different designs forms the basis for the engineering justification of the VVER reactor core reliability and serviceability. The investigation results for the coolant interaction in adjoining TVSA FAs of different designs have been adopted for the practical use at JSC Afrikantov OKBM to estimate the heat-engineering reliability of the VVER reactor cores and have been included in the database for verification of computational fluid dynamics (CFD) codes and detailed by-channel calculation codes.

scholarly journals CFD Analysis of Downcomer of Nuclear Reactor VVER 440

2016 ◽  
Vol 66 (2) ◽  
pp. 55-62
Author(s):  
Vladimír Kutiš ◽  
Jakub Jakubec ◽  
Juraj Paulech ◽  
Gálik Gálik ◽  
Tibor Sedlár
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Coolant Flow ◽  
Orifice Diameter ◽  
Cfd Analysis ◽  
Flow Conditions ◽  
Fuel Assemblies ◽  
Flow Through ◽  
Cfd Analyses

Abstract The paper is focused on CFD analyses of the coolant flow in the nuclear reactor VVER 440. The goal of the analyses is to investigate the influence of the orifice diameter on the mass flow through individual fuel assemblies in the reactor core. The diameter of orifice can be changed during the operation of a nuclear power plant. Considered boundary conditions in the investigated region of the coolant are based on nominal coolant flow conditions in the nuclear reactor VVER 440.

scholarly journals Hydrodynamics of coolant in VVER reactor core with fuel assemblies of various designs

2020 ◽  
Vol 1683 ◽  
pp. 022004
Author(s):  
S M Dmitriev ◽  
A V Gerasimov ◽  
A A Dobrov ◽  
D V Doronkov ◽  
V E Liskova ◽  
...  
Keyword(s):  
Reactor Core ◽  
Vver Reactor ◽  
Fuel Assemblies ◽  
Hydrodynamics Of Coolant

Computational and Experimental Investigation of Patterns of Coolant Flow in the Mixed Core of a VVER Reactor

2021 ◽  
Author(s):  
S. M. Dmitriev ◽  
A. A. Dobrov ◽  
D. V. Doronkov ◽  
D. S. Doronkova ◽  
M. A. Legchanov ◽  
...  
Keyword(s):  
Experimental Investigation ◽  
Coolant Flow ◽  
Vver Reactor

scholarly journals Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER

2021 ◽  
Vol 2 (2) ◽  
pp. 207-214
Author(s):  
Thinh Truong ◽  
Heikki Suikkanen ◽  
Juhani Hyvärinen
Keyword(s):  
Low Temperature ◽  
Conceptual Design ◽  
Nuclear Power ◽  
Water Pressure ◽  
Reactor Core ◽  
Electricity Production ◽  
Design Parameters ◽  
Experimental Reactor ◽  
Fuel Assemblies ◽  
Enriched Uranium

In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sector, which is a much more significant greenhouse gas emitter than electricity production in the Nordic countries. The main principle in the development of LUTHER is to simplify the core design and safety systems, which, along with using commercially available reactor components, would lead to lower fabrication costs and enhanced safety. LUTHER also features a unique design with movable individual fuel assembly for reactivity control and burnup compensation. Two-dimensional (2D) and three-dimensional (3D) fuel assemblies and reactor cores are modeled with the Serpent Monte Carlo reactor physics code. Different reactor design parameters and safety configurations are explored and assessed. The preliminary results show an optimal basic core design, a good neutronic performance, and the feasibility of controlling reactivity by moving fuel assemblies.

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.

scholarly journals THE DEVELOPMENT OF A THREE-DIMENSIONAL MODEL OF WWER-1000 CORE USING THE MONTE CARLO SERPENT CODE FOR NEUTRON-PHYSICAL MODELING

2019 ◽  
pp. 58-61
Author(s):  
V.I. Gulik ◽  
O.R. Trofymenko ◽  
V.V. Galchenko ◽  
D.V. Budik
Keyword(s):  
Monte Carlo ◽  
3D Modeling ◽  
Physical Modeling ◽  
Three Dimensional ◽  
Reactor Core ◽  
Dimensional Model ◽  
Three Dimensional Model ◽  
Fuel Assemblies

The article presents the use of the new Monte Carlo Serpent code for 3D modeling of the WWER-1000 reactor core. Core models for the first loading of RNPP4 and the 28th loading of SUNPP3, the fuel assemblies’ models of different manufacturers were developed and presented. Considerable attention was paid to the detailed modeling of the upper, lower and side reflectors. Validation calculations of the Monte Carlo Serpent code for the WWER-1000 reactor were performed on the basis of the first RNPP4 loading. For the 28th loading of SUNPP3, albedo coefficients for radial and axial reflectors were obtained.

scholarly journals Analysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors

2021 ◽  
Vol 7 (4) ◽  
pp. 311-318
Author(s):  
Artavazd M. Sujyan ◽  
Viktor I. Deev ◽  
Vladimir S. Kharitonov
Keyword(s):  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Negative Impact ◽  
Reactor Core ◽  
Resistance Coefficient ◽  
Coolant Flow ◽  
Water Reactor ◽  
Hydraulic Stability ◽  
Supercritical Water Reactor

The paper presents a review of modern studies on the potential types of coolant flow instabilities in the supercritical water reactor core. These instabilities have a negative impact on the operational safety of nuclear power plants. Despite the impressive number of computational works devoted to this topic, there still remain unresolved problems. The main disadvantages of the models are associated with the use of one simulated channel instead of a system of two or more parallel channels, the lack consideration for neutronic feedbacks, and the problem of choosing the design ratios for the heat transfer coefficient and hydraulic resistance coefficient under conditions of supercritical water flow. For this reason, it was decided to conduct an analysis that will make it possible to highlight the indicated problems and, on their basis, to formulate general requirements for a model of a nuclear reactor with a light-water supercritical pressure coolant. Consideration is also given to the features of the coolant flow stability in the supercritical water reactor core. In conclusion, the authors note the importance of further computational work using complex models of neutronic thermal-hydraulic stability built on the basis of modern achievements in the field of neutron physics and thermal physics.

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