Experimental Study of Feed Water Level Decreasing Effect on VVER Steam Generator Model Operation in Condensation Mode

2013 ◽  
Author(s):  
A. V. Morozov ◽  
O. V. Remizov ◽  
A. S. Soshkina
Keyword(s):  
Water Level ◽  
Steam Generator ◽  
Vver Reactor ◽  
Test Facility ◽  
Primary Circuit ◽  
Feed Water ◽  
Steam Condensation ◽  
Design Basis ◽  
Condensation Mode ◽  
Design Basis Accident

The essential technological distinctions of the AES-2006 project with VVER-1200 reactor, equipped with passive safety systems, determine some specifics in the character of accidents with the coolant leaks from the reactor primary circuit and the refusal of active part of the emergency core cooling system. In the case of an accident with depressurization of the reactor primary circuit, the system of passive heat removal (PHRS) ensures the transition of steam generators (SG) into the mode of steam condensation. As a result, a condensate comes to the core, providing its additional cooling. But in the case of beyond design basis accident the rupture of second circuit pipelines or PHRS steam-condensation path is possible. At such type of accidents the feed water level in SG vessel will be decrease, what can results in the deterioration of its condensation power. For experimental investigation of the condensation mode of VVER steam generator, a large scale HA2M-SG test facility was constructed in IPPE. The test facility incorporates VVER reactor SG model with volumetric-power scale of piping is 1:46, PHRS heat exchanger imitator, cooling by process water and buffer tank, equipped by steam supply system from the IPPE heat power plant. The facility main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project. Experiments at the HA2M-SG test facility have been performed at the pressure 0.36 MPa, correspond to VVER reactor pressure at the last stage of the beyond design basis accident. On the base of the results of these experiments the correlation of condensation power from feed water level in SG model was obtained. The results of carried out tests make it possible to draw a conclusion about sufficiently large stability of VVER steam generator working in condensation mode to feed water level decreasing.

Effect of Pressure on Underdeveloped Nucleate Boiling on the Single Horizontal Tube Heated by Condensing Steam

2012 ◽  
Author(s):  
A. V. Morozov ◽  
O. V. Remizov ◽  
D. S. Kalyakin
Keyword(s):  
Steam Generator ◽  
Horizontal Tube ◽  
Nucleate Boiling ◽  
Vver Reactor ◽  
Test Facility ◽  
Tube Length ◽  
Primary Circuit ◽  
Empirical Correlations ◽  
Steam Generators ◽  
Condensation Mode

In new Russian NPP with VVER reactor in the event of LOCA, provision is made for the use of passive heat removal system for necessary core cooling. In the case of leakage in the primary circuit this system assures the transition of steam generators to operation in the mode of condensation of the primary circuit steam coming to steam generator piping from the reactor. As a result, the condensate from the steam generators arrives to the core providing its additional cooling. The steam generator off-design condensation mode has the following features: undeveloped nucleate boiling on the horizontal tubes heated by condensing steam; natural circulation processes in the both steam generator circuits; low heat fluxes and temperature differences. The experimental study of undeveloped nucleate boiling on the single horizontal tube heated by condensing steam has been carried out in the Institute for Physics and Power Engineering. The experiments have been carried out on the GROT test facility. The heart of the test facility is the single VVER steam generator tube (length l = 10.2 m, outer diameter D = 16 mm, wall thickness d = 1.5 mm). The tube is fabricated of the original stainless steel 08Cr18Ni10Ti. The length and geometry of the test tube corresponded to that of real steam generator. The test facility was equipped with thermocouples enabling the temperatures of primary and secondary facility circuits to be controlled. The experiments were carried out at three heating steam pressures Ps1: 0.21, 0.35, 0.55 MPa. The main task of the research was to study the pressure effect on the process of undeveloped nucleate boiling on the single horizontal tube. On the base of the results of these experiments the empirical correlations for prediction of heat transfer coefficient and heat flux were obtained. The generalizing empirical correlations obtained can be used for the substantiation of work heat-exchanging equipment of NPP with VVER reactor in the condensation mode, and also can be applied to the verification of computer codes.

The CFD Analysis of Hydrogen Behavior in HTR-PM During the Water-Ingress Accident Caused by Double-Ended Guillotine Break of Two Steam Generator Heating Tubes

2014 ◽  
Author(s):  
Ruidong Zhang ◽  
Ximing Sun ◽  
Yujie Dong
Keyword(s):  
Steam Generator ◽  
Radioactive Material ◽  
Primary Circuit ◽  
Preliminary Design ◽  
Chemical Corrosion ◽  
Water Ingress ◽  
Design Basis ◽  
Design Basis Accident ◽  
Computational Fluid Dynamics Cfd ◽  
Combustible Gases

Water ingress into the primary circuit of HTR-PM (High Temperature Gas-cooled Reactor Pebble-bed Modular) is one of the beyond-design-basis accidents, which will lead to chemical corrosion of graphite and metal material in the reactor, resulting in the generation of combustible gases such as hydrogen and carbonic oxide. Besides, the rise of pressure in the reactor containment caused by the incoming of abundant of gas will threaten to the safety of radioactive material releasing. Therefore, it is essential to analyze the gas behavior in the containment of HTR-PM under the condition of water ingress. Based on the preliminary design of 250MW HTR-PM, the paper analyzes the transport and distribution of hydrogen, helium and other kinds of gas in the containment under the condition of water ingress accident resulted from a double-ended guillotine break of two steam generator heating tubes which is a typical beyond design basis accident of HTR-PM with the multi-dimensional computational fluid dynamics CFD code GASFLOW. In addition, the density and distribution of hydrogen will be investigated in relation to possible hydrogen explosion.

Investigation of Operation of VVER Steam Generator in Condensation Mode at the Large-Scale Test Rig

2008 ◽  
Author(s):  
A. D. Efanov ◽  
S. G. Kalyakin ◽  
A. V. Morozov ◽  
O. V. Remizov ◽  
A. A. Tsyganok ◽  
...  
Keyword(s):  
Steam Generator ◽  
Large Scale ◽  
Test Procedure ◽  
Heat Removal ◽  
Primary Circuit ◽  
Test Rig ◽  
Steam Generators ◽  
Main Equipment ◽  
Mode Of Operation ◽  
Condensation Mode

In new Russian NPP with VVER-1200 reactor (V-392M reactor plant) in the event of an accident being due to the rupture of the reactor primary circuit and accompanied by the loss of a.c. sources, provision is made for the use of passive safety systems for necessary core cooling. Among these is passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam coming to SG piping from the reactor. As a result, the condensate from steam generators arrives at the core providing its additional cooling. To experimental investigation of the condensation mode of operation of VVER steam generator, a large scale HA2M-SG test rig was constructed. The test rig incorporates: tank-accumulator, equipped by steam supply system; SG model with volumetric-power scale is 1:46; PHRS heat exchanger simulator, cooling by process water. The rig main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project. The rig maximum operating parameters: steam pressure – 1.6 MPa, temperature – 200 Celsius degrees. Experiments at the HA2M-SG test rig have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond basis accident. The report presents the test procedure and the basic obtained test results.

Non-Condensable Gases Effect on Heat Transfer Processes Between Condensing Steam and Boiling Water in Heat Exchanger With Multirow Horizontal Tube Bundle

2010 ◽  
Author(s):  
A. V. Morozov ◽  
O. V. Remizov ◽  
A. A. Tsyganok
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Tube Bundle ◽  
Horizontal Tube ◽  
Boiling Water ◽  
Test Facility ◽  
Primary Circuit ◽  
Outer Diameter ◽  
Steam Condensation ◽  
Experimental Heat

The experimental investigations of non-condensable gases effect on the steam condensation inside multirow horizontal tube bundle of heat exchanger under heat transfer to boiling water were carried out at the large-scale test facility in the Institute for Physics and Power Engineering (IPPE). The experiments were carried out for natural circulation conditions in primary and secondary circuits of the facility at primary circuit steam pressure of Ps1 = 0.34 MPa. The experimental heat exchanger’s tube bundle consists of 248 horizontal coiled tubes arranged in 62 rows. Each row consists of 4 stainless steel tubes of 16 mm in outer diameter, 1.5 mm in wall thickness and of 10.2 m in length. The experimental heat exchanger was equipped with more than 100 thermocouples enabling the temperatures of primary and secondary facility circuits to be controlled in both tube bundle and in the inter-tubular space. The non-condensable gases with different density — nitrogen and helium were used in the experiments. The volumetric content of gases in tube bundle amounted to ε = 0.49. The empirical correlation for the prediction of the relative heat transfer coefficient k/k0 = f (ε) for steam condensation in steam-gas mixture was obtained.

Experimental Investigation of Non-Condensable Gases Effect on Novovoronezh NPP-2 Steam Generator Condensation Power Under the Condition of Passive Safety Systems Operation

2009 ◽  
Author(s):  
I. I. Kopytov ◽  
S. G. Kalyakin ◽  
V. M. Berkovich ◽  
A. V. Morozov ◽  
O. V. Remizov
Keyword(s):  
Steam Generator ◽  
Flow Rate ◽  
Computational Simulation ◽  
Heat Removal ◽  
Test Facility ◽  
Passive Safety ◽  
Primary Circuit ◽  
Water Radiolysis ◽  
Passive Safety Systems ◽  
Safety Systems

The design substantiation of the heat removal efficiency from Novovoronezh NPP-2 (NPP-2006 project with VVER-1200 reactor) reactor core in the event of primary circuit leaks and operation of passive safety systems only (among these are the systems of hydroaccumulators of the 1st and 2nd stages and passive heat removal system) has been performed based on computational simulation of the related processes in the reactor and containment. The computational simulation has been performed with regard to the detrimental effect of non-condensable gases on steam generator (SG) condensation power. Nitrogen arriving at the circuit with the actuation of hydroaccumulators of the 1st stage and products of water radiolysis are the main sources of non-condensable gases in the primary circuit. The feature of Novovoronezh NPP-2 passive safety systems operation is that during the course of emptying of the 2nd stage hydroaccumulators system (HA-2) the gas-steam mixture spontaneously flows out from SG cold headers into the volume of HA-2 tanks. The flow rate of gas-steam mixture during the operation of HA-2 system is equal to the volumetric water discharge from hydroaccumulators. The calculations carried out by different integral thermal hydraulic codes revealed that this volume flow rate of gas-steam mixture from SG to the HA-2 system would suffice to eliminate the “poisoning” of SG piping and to maintain necessary condensation power. In support of the calculation results, the experiments were carried out at the HA2M-SG test facility constructed at IPPE. The test facility incorporates a VVER steam generator model of volumetric-power scale of 1:46. Steam to the HA2M-SG test facility is supplied fed from the IPPE heat power plant. Gas addition to steam coming to the SG model is added from high pressure gas cylinders. Nitrogen and helium are used in the experiments for simulating hydrogen. The report presents the basic results of experimental investigations aimed at the evaluation of SG condensation power under the inflow of gas-steam mix with different gases concentration to the tube bundle, both under the simulation of gas-steam mixture outflow from SG cold header to the HA-2 system and without outflow. As a result of the research performed at the HA2M-SG test facility, it has been substantiated experimentally that in the event of an emergency leak steam generators have condensation power sufficient for effective heat removal from the reactor provided by PHR system.

Water-Ingress Accident of the 250MW Pebble-Bed Modular High Temperature Gas-Cooled Reactor

2009 ◽  
Author(s):  
Zheng Yanhua ◽  
Shi Lei
Keyword(s):  
High Temperature ◽  
Safety Valve ◽  
Primary Circuit ◽  
Fuel Temperature ◽  
Pebble Bed ◽  
Water Ingress ◽  
Design Basis ◽  
Design Basis Accident ◽  
Water Gas ◽  
High Temperature Gas

Water-ingress accident, caused by the steam generator heating tube rupture of a high temperature gas-cooled reactor, will introduce a positive reactivity to lead the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Increase of the primary circuit pressure may result in the opening of the safety valve, which will cause the release of radioactive isotopes and flammable water gas. The analysis of such an important and particular accident is significant for verifying the inherent safety characteristics of the pebble-bed modular high temperature gas-cooled reactor. Based on the preliminary design of the 250MW Pebble-bed Modular High Temperature Gas-cooled Reactor (HTR-PM), the design basis accident of double-ended guillotine break of a heating tube has been analyzed by using TINTE, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature and primary loop pressure, the graphite corrosion inventory, the water gas releasing amount, as well as the natural convection influence under the condition of the failure of the blower flaps shut down, have been studied in detail. The calculation result of the design basis accident indicates that, the maximal possible water ingress amount is less than 600 kg and the maximal fuel temperature keeps far below the design limitation of 1620°C. The result also shows that the slight amount of graphite corrosion will not damage the reactor structure and the fuel element, and there is no potential explosive risk caused by the opening of the safety valve.

A balance procedure for calculating the model fuel assemblies reflooding during design basis accident and its verification on PARAMETER test facility

2013 ◽  
Vol 60 (5) ◽  
pp. 323-330 ◽  
Author(s):  
S. S. Bazyuk ◽  
D. N. Ignat’ev ◽  
N. Ya. Parshin ◽  
E. B. Popov ◽  
D. M. Soldatkin ◽  
...  
Keyword(s):  
Test Facility ◽  
Design Basis ◽  
Model Fuel ◽  
Fuel Assemblies ◽  
Design Basis Accident ◽  
Parameter Test

scholarly journals Optimal feed-water level control for steam generator in nuclear power plant based on meta-heuristic optimization

2020 ◽  
Vol 13 (1) ◽  
pp. 468-484
Author(s):  
Nehad Ali Demerdash ◽  
Mohamed A. El-Hameed ◽  
Ezzat A. Eisawy ◽  
Mahdy M. El-Arini
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Level ◽  
Steam Generator ◽  
Nuclear Power ◽  
Heuristic Optimization ◽  
Feed Water ◽  
Level Control ◽  
Optimal Feed

An investigation of the survivability of fuel-rod claddings during the first stage of the maximum design basis accident at a nuclear power station with a VVER reactor

2006 ◽  
Vol 53 (1) ◽  
pp. 19-25
Author(s):  
G. V. Karetnikov ◽  
Yu. A. Bezrukov ◽  
A. S. Bogdanov ◽  
A. M. Trushin ◽  
V. P. Onshin ◽  
...  
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Vver Reactor ◽  
Power Station ◽  
Fuel Rod ◽  
Design Basis ◽  
Design Basis Accident ◽  
Maximum Design
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