CFD Simulation of Slug Mixing in VVER-1000 Reactor

2006 ◽  
Author(s):  
Ladislav Vyskocil
Keyword(s):  
Boron Concentration ◽  
Cfd Simulation ◽  
Cold Water ◽  
Reactor Core ◽  
Time History ◽  
Scale Model ◽  
Forced Circulation ◽  
Coolant Pump ◽  
Design Bureau ◽  
The Core

Recently, the safety analyses of VVER and PWR reactors have dealt with the possibility of reactivity-induced accidents related to the penetration of a water slug with low boron concentration into the reactor core. Loop seals at the reactor coolant pump (RCP) suction are the most likely places for the formation of these slugs. The slug is formed in the loop when there is neither natural nor forced circulation. When the circulation is restored, the slug travels towards the reactor and causes an insertion of positive reactivity in the core. This report deals with a CFD simulation of the most dangerous event—the start-up of the first RCP. Only several seconds are needed for slug to reach the core and the operator has no time for corrective action. Mixing of slug on its way to the core can reduce the danger of core recriticality. The primary objective of this study was to find out whether the FLUENT 6 CFD code is capable of predicting the mixing in the cold leg, downcomer and lower plenum as the slug moves toward the reactor core. Numerical simulations were based on mixing tests performed on 1:5 scale model of VVER-1000 reactor at the Gidropress Design Bureau, Russia. In the physical mixing tests, temperature was substituted for Boron concentration through the use of hot and cold water. The time history of core inlet average temperature was calculated by FLUENT and was found to be in good qualitative agreement with experimental data. This work was carried out as part of the EU project FLOMIX-R, Work Package 4.

Analysis and Research on Natural Circulation Capacity of HFETR

2016 ◽  
Author(s):  
Taozhong Xu ◽  
Caiyu Deng ◽  
Yuxin Xiang
Keyword(s):  
Pressure Vessel ◽  
Cold Water ◽  
Flow Direction ◽  
Natural Circulation ◽  
Reactor Power ◽  
Hot Water ◽  
Forced Circulation ◽  
The Core ◽  
Establishment Process ◽  
Residual Heat

Natural circulation is being used as an important circulation to remove reactor residual heat. In the core of High Flux Engineering Trial Reactor of China (HFETR), the coolant is driven by pumps normally and flows from upside to downside in the core. When HFETR is shut down or runs in low power, the natural circulation between the hot water in the core and the cold water in the reflector inside the pressure vessel is established to cool down the core. Since the natural circulation processed only in the pressure vessel, the accident pumps need to be turned on periodically to remove reactor residual heat. The inversion of flow direction in HFETR and internal natural circulation lead to a different natural circulation establishment process from traditional reactor in which coolant flows form down to top normally. In this paper the transition between the natural circulation and forced circulation is analyzed by RELAP5/MOD3 code. The results showed that the accident pump could be turned off in the power of 850kW; The time, at which the accident pump needs to be turned on to transit the natural circulation to forced circulation, is decided by the temperature of the water in top of pressure vessel, and a formula between temperature of the water in the top of pressure vessel and the reactor power was obtained. The research results have theoretical and practical value to the full use of the natural circulation ability, as well as the safety of the engineering reactors or similar test facilities.

Analysis of Dynamic Stress in the Reactor Core Barrel due to the Pressure Pulsations

2015 ◽  
Author(s):  
Igor Orynyak ◽  
Iaroslav Dubyk ◽  
Anatolii Batura
Keyword(s):  
Nuclear Power ◽  
Dynamic Stress ◽  
Reactor Core ◽  
Pressure Pulsations ◽  
Power Station ◽  
Coolant Pump ◽  
The Core ◽  
Calculation Results ◽  
Full Power ◽  
Reactor Pressure

This article presents vibrations analysis of the reactor core barrel caused by pressure pulsations induced by the main coolant pump. For this purpose, the calculations of the pressure distribution in the annulus between the core barrel and the reactor pressure vessel, bounded above by a separating ring were performed. Using transfer matrix method is obtained the solution of two-dimensional problem of pressure pulsations in the annulus between reactor core barrel and reactor vessel. The calculation results are compared with the pulsation pressure measurements made at commissioning unit 2 of the South Ukraine Nuclear Power Station. The distribution of pressure over the height of core barrel was obtained, which makes possible to estimate its strength for variant deformation of the core barrel as a beam, and in the case of deformation of the core barrel as a shell. The calculation results are used to assess the reliability of core barrel pre-load, which clamps the core barrel flange in place at the top, at full power operating.

scholarly journals Design and Calibration of a Hall Effect System for Measurement of Six-Degree-of-Freedom Motion within a Stacked Column

Sensors ◽  
2021 ◽  
Vol 21 (11) ◽  
pp. 3740
Author(s):  
Olafur Oddbjornsson ◽  
Panos Kloukinas ◽  
Tansu Gokce ◽  
Kate Bourne ◽  
Tony Horseman ◽  
...  
Keyword(s):  
Hall Effect ◽  
Degrees Of Freedom ◽  
Reactor Core ◽  
Degree Of Freedom ◽  
Life Extension ◽  
Scale Model ◽  
Design Development ◽  
Seismic Safety ◽  
Six Degrees Of Freedom ◽  
Six Degree Of Freedom

This paper presents the design, development and evaluation of a unique non-contact instrumentation system that can accurately measure the interface displacement between two rigid components in six degrees of freedom. The system was developed to allow measurement of the relative displacements between interfaces within a stacked column of brick-like components, with an accuracy of 0.05 mm and 0.1 degrees. The columns comprised up to 14 components, with each component being a scale model of a graphite brick within an Advanced Gas-cooled Reactor core. A set of 585 of these columns makes up the Multi Layer Array, which was designed to investigate the response of the reactor core to seismic inputs, with excitation levels up to 1 g from 0 to 100 Hz. The nature of the application required a compact and robust design capable of accurately recording fully coupled motion in all six degrees of freedom during dynamic testing. The novel design implemented 12 Hall effect sensors with a calibration procedure based on system identification techniques. The measurement uncertainty was ±0.050 mm for displacement and ±0.052 degrees for rotation, and the system can tolerate loss of data from two sensors with the uncertainly increasing to only 0.061 mm in translation and 0.088 degrees in rotation. The system has been deployed in a research programme that has enabled EDF to present seismic safety cases to the Office for Nuclear Regulation, resulting in life extension approvals for several reactors. The measurement system developed could be readily applied to other situations where the imposed level of stress at the interface causes negligible material strain, and accurate non-contact six-degree-of-freedom interface measurement is required.

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.

Heat balance precedes stabilization of body temperatures during cold water immersion

2003 ◽  
Vol 95 (1) ◽  
pp. 89-96 ◽  
Author(s):  
Peter Tikuisis
Keyword(s):  
Heat Balance ◽  
Cold Water ◽  
Water Immersion ◽  
Heat Content ◽  
Cold Water Immersion ◽  
The Body ◽  
Body Temperatures ◽  
Mean Body Temperature ◽  
The Core ◽  
Region Temperature

Certain previous studies suggest, as hypothesized herein, that heat balance (i.e., when heat loss is matched by heat production) is attained before stabilization of body temperatures during cold exposure. This phenomenon is explained through a theoretical analysis of heat distribution in the body applied to an experiment involving cold water immersion. Six healthy and fit men (mean ± SD of age = 37.5 ± 6.5 yr, height = 1.79 ± 0.07 m, mass = 81.8 ± 9.5 kg, body fat = 17.3 ± 4.2%, maximal O2 uptake = 46.9 ± 5.5 l/min) were immersed in water ranging from 16.4 to 24.1°C for up to 10 h. Core temperature (Tco) underwent an insignificant transient rise during the first hour of immersion, then declined steadily for several hours, although no subject's Tco reached 35°C. Despite the continued decrease in Tco, shivering had reached a steady state of ∼2 × resting metabolism. Heat debt peaked at 932 ± 334 kJ after 2 h of immersion, indicating the attainment of heat balance, but unexpectedly proceeded to decline at ∼48 kJ/h, indicating a recovery of mean body temperature. These observations were rationalized by introducing a third compartment of the body, comprising fat, connective tissue, muscle, and bone, between the core (viscera and vessels) and skin. Temperature change in this “mid region” can account for the incongruity between the body's heat debt and the changes in only the core and skin temperatures. The mid region temperature decreased by 3.7 ± 1.1°C at maximal heat debt and increased slowly thereafter. The reversal in heat debt might help explain why shivering drive failed to respond to a continued decrease in Tco, as shivering drive might be modulated by changes in body heat content.

scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

A Research Reactor Source Term and Dose Results in the Event of a LOCA

2014 ◽  
Author(s):  
Charalampos Pappas ◽  
Andreas Ikonomopoulos ◽  
Athanasios Sfetsos ◽  
Spyros Andronopoulos ◽  
Melpomeni Varvayanni ◽  
...  
Keyword(s):  
Research Reactor ◽  
Source Term ◽  
Dispersion Model ◽  
Ventilation System ◽  
Atmospheric Dispersion ◽  
Reactor Core ◽  
Forecast Model ◽  
Scale Model ◽  
Continuous Reactor ◽  
Reactor Building

The present study discusses the source term derivation and dose result calculation for a hypothetical accident sequence in the Greek Research Reactor – 1 (GRR-1). A loss-of-coolant accident (LOCA) has been selected as a credible accident sequence. The source term derivation has been based on the GRR-1 confinement performance where the inventory has been computed assuming continuous reactor operation. A core damage fraction of 30% has been considered for the calculations while conservative core release fractions have been employed. The radionuclides released from the reactor core to the confinement atmosphere have been subjected to natural decay, deposition on and resuspension from various internal surfaces before being led to the release pathway. It has been assumed that an emergency shutdown is initiated immediately after the beginning of the accident sequence and the emergency ventilation system is also activated. Subsequently, the source term has been derived comprising of noble gases, iodine and aerosol. The JRODOS computational software for off-site nuclear emergency management has been utilized to estimate the dose results from the LOCA-initiated source term that is released in its entirety from the reactor stack at ambient temperature. The Local Scale Model Chain in conjunction with the DIPCOT atmospheric dispersion model that is embedded in JRODOS have been used with proper parameterization of the calculation settings. Five weather scenarios have been selected as representative of typical meteorological conditions at the reactor site. The scenarios have been assessed with the use of the Weather Research and Forecast model. Total effective, skin, thyroid, lung and inhalation doses downwind of the reactor building and up to a distance of 10 km have been calculated for each weather scenario and are presented. The total effective gamma dose rate at a fixed distance from the reactor building has been assessed. The radiological consequences of the dose results are discussed.

scholarly journals Experimental study on dynamic response of model shield tunnel induced by moving-axle loads of subway train

2018 ◽  
Vol 14 (10) ◽  
pp. 155014771880278
Author(s):  
Mengxi Zhang ◽  
Xiaoqing Zhang ◽  
Lei Li ◽  
Chengyu Hong
Keyword(s):  
Time History ◽  
Vertical Displacement ◽  
Wheel Speed ◽  
Dynamic Responses ◽  
Scale Model ◽  
Shield Tunnel ◽  
Laboratory Model ◽  
Cover Depth ◽  
Axle Loads ◽  
Subway Train

A new testing method was introduced to apply moving-axle loads of a subway train on a track structure. In order to investigate the dynamic responses of the shield tunnel subjected to moving-axle loads, a series of laboratory model tests were conducted in a 1/40 scale model tunnel. The influences of the axle load, the wheel speed, and the cover depth of the shield tunnel on the vertical displacement and acceleration of the lining were presented and discussed. Parametric studies revealed that the vertical displacement–time history of the lining presents a “W” shape due to the combined action of two axles of a bogie. The peak value of the vertical displacement increased with the axle load linearly, while it decreased with the increase in the cover depth. Moreover, response time of the displacement decreased with the increase in the wheel speed, but the peak values remained stable at the same level. Finally, a three-dimensional dynamic finite element model was adopted to simulate the movement of the axle loads and calculate the responses of the lining. The numerical results analysis agrees well with experimental results.

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