scholarly journals Simulation of the Thermal Hydraulic Processes in the Horizontal Steam Generator with the Use of the Different Interfacial Friction Correlations

2011 ◽  
Vol 2011 ◽  
pp. 1-9
Author(s):  
Vladimir Melikhov ◽  
Oleg Melikhov ◽  
Yury Parfenov ◽  
Alexey Nerovnov
Keyword(s):  
Steam Generator ◽  
Three Dimensional ◽  
Test Facility ◽  
Interfacial Friction ◽  
Water Mixture ◽  
Pressurized Water ◽  
Design And Optimization ◽  
Water Flows ◽  
Horizontal Steam Generator ◽  
Water Reactors

The horizontal steam generator (SG) is one of specific features of Russian-type pressurized water reactors (VVERs). The main advantages of horizontal steam generator are connected with low steam loads on evaporation surface, simple separation scheme and high circulation ratio. The complex three-dimensional steam-water flows in the steam generator vessel influence significantly the processes of the steam separation, distribution, and deposition of the soluble and nonsoluble impurities and determine the efficiency and reliability of the steam generator operation. The 3D code for simulation of the three-dimensional steam-water flows in the steam generator could be effective tool for design and optimization of the horizontal steam generator. The results of the code calculations are determined mainly by the set of the correlations describing interaction of the steam-water mixture with the inner constructions of the SG and interfacial friction. The results obtained by 3D code STEG with the usage of the different interfacial friction correlations are presented and discussed in the paper. These results are compared with the experimental ones obtained at the experimental test facility PGV-1500 constructed for investigation of the processes in the horizontal steam generator.

Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

2009 ◽  
Author(s):  
Yuriy V. Parfenov ◽  
Oleg I. Melikhov ◽  
Vladimir I. Melikhov ◽  
Ilya V. Elkin
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Evolutionary Development ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tube Bundles ◽  
Horizontal Steam Generator ◽  
Calculation Results ◽  
Steam Separation

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

2008 ◽  
Author(s):  
Mitsuhiro Suzuki ◽  
Takeshi Takeda ◽  
Hideo Nakamura
Keyword(s):  
Time Delay ◽  
Large Scale ◽  
Simulation Experiment ◽  
Three Dimensional ◽  
Test Facility ◽  
Large Temperature ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Accident Management

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary sides in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.

U.S. Operating Experience With Thermally Treated Alloy 600 Steam Generator Tubes

2002 ◽  
Author(s):  
April Smith ◽  
Kenneth J. Karwoski
Keyword(s):  
Steam Generator ◽  
Operating Experience ◽  
Alloy 600 ◽  
Steam Generators ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Water Reactors ◽  
1960S And 1970S ◽  
The 1960S ◽  
Thermally Treated

Steam generators placed in service in the 1960s and 1970s were primarily fabricated from mill-annealed Alloy 600. Over time, this material proved to be susceptible to stress corrosion cracking in the highly pure primary and secondary water chemistry environments of pressurized-water reactors. The corrosion ultimately led to the replacement of steam generators at numerous facilities, the first U.S. replacement occurring in 1980. Many of the steam generators placed into service in the 1980s used tubes fabricated from thermally treated Alloy 600. This tube material was thought to be less susceptible to corrosion. Because of the safety significance of steam generator tube integrity, this paper evaluates the operating experience of thermally treated Alloy 600 by looking at the extent to which it is used and recent results from steam generator tube examinations.

Validation of Trio_U: Numerical Simulations of a ROCOM Buoyancy Driven Test Case

2004 ◽  
Author(s):  
T. Ho¨hne ◽  
U. Bieder ◽  
S. Kliem ◽  
H.-M. Prasser
Keyword(s):  
Turbulent Flows ◽  
Natural Circulation ◽  
Density Difference ◽  
Test Facility ◽  
Test Case ◽  
Pressurized Water ◽  
Constant Flow Rate ◽  
Sensor Position ◽  
Nuclear Systems ◽  
Water Reactors

A generic investigation of the influence of density differences between the primary loop inventory and the ECC water on the mixing in the downcomer was made at the ROCOM Mixing Test Facility at Forschungszentrum Rossendorf (FZR)/Germany. ROCOM is designed for experimental coolant mixing studies over a wide variety of possible scenarios. It is equipped with advanced instrumentation, which delivers high-resolution information characterizing either temperature or boron concentration fields in the investigated pressurized water reactor. For the validation of the Trio_U code an experiment with 5% constant flow rate in one loop (magnitude of natural circulation) and 10% density difference between ECC and loop water was taken. Trio_U is a CFD code developed by the CEA France, aimed to supply an efficient computational tool to simulate transient thermal-hydraulic single-phase turbulent flows encountered in the nuclear systems as well as in the industrial processes. For this study a LES approach was used for mesh sizes according to between 300000–2 million control volumes. The results of the experiment as well as of the numerical calculations show, that a streak formation of the water with higher density is observed. At the upper sensor, the ECC water covers a small azimuthal sector. The density difference partly suppresses the propagation of the ECC water in circumferential direction. The ECC water falls down in an almost straight streamline and reaches the lower downcomer sensor position directly below the affected inlet nozzle. Only later, coolant containing ECC water appears at the opposite side of the downcomer. The study showed, that density effects play an important role during natural convection with ECC injection in pressurized water reactors. Furthermore it was important to point out, that Trio_U is able to cope the main flow and mixing phenomena.

Break Size Effects on CET Response in an Upper Head SBLOCA Transient

2012 ◽  
Author(s):  
S. Gallardo ◽  
A. Querol ◽  
G. Verdú
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Upward Direction ◽  
Water Reactors ◽  
Accident Conditions

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced. This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS). The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.

scholarly journals ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water

2016 ◽  
Vol 2016 ◽  
pp. 1-15
Author(s):  
Takeshi Takeda ◽  
Akira Ohnuki ◽  
Daisuke Kanamori ◽  
Iwao Ohtsu
Keyword(s):  
Steam Generator ◽  
Safety System ◽  
Injection System ◽  
Test Facility ◽  
Gas Release ◽  
Dissolved Gas ◽  
Nitrogen Gas ◽  
Two Phase ◽  
Pressurized Water ◽  
Secondary Side

Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility). The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG) secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC) water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

A Simplified Evaluation of Stress Intensity Factors for a Small Diameter Pipe Penetrating a Thick Plate

2003 ◽  
Author(s):  
John M. Emery ◽  
Katsumasa Miyazaki ◽  
Anthony R. Ingraffea
Keyword(s):  
Stress Intensity ◽  
Stress Intensity Factors ◽  
Complex Geometry ◽  
Three Dimensional ◽  
Small Diameter ◽  
Boiling Water Reactors ◽  
Intensity Factors ◽  
Pressurized Water ◽  
Diameter Pipe ◽  
Water Reactors

Recent trouble with stress corrosion cracking in the internal parts of boiling water reactors and similar issues found in pressurized water reactors has prompted interest in developing simplified methods to determine stress intensity factors for such cracks. Currently, there are many practical and accurate simplified methods to calculate stress intensity factors for a surface crack in plates and pipes. However, there are none that deal with the complex geometry that can arise within the reactors. The complex geometry found within the vessels often entails reentrant comers, welds, holes, and other stress amplifiers. This paper sets forth a means by which some commonly known and accepted simplified solutions to cracks in pipes and plates can be modified to improve the accuracy of stress intensity factors when applied to this complex geometry. The effort to do so included axisymmetric and fully three-dimensional numerical modeling of both the cracked and uncracked body with a variety of assumed surface flaws. It was confirmed that the simplified methods lead to exceedingly conservative estimates for the stress intensity factors of the complex geometry. Finally, a correction factor based on the axisymmetric analyses was applied to the three-dimensional results to improve the accuracy of the simplified solutions.

An Automated System for Detection of Through-Wall Cracks in Racks in Steam Generator Tubes

2005 ◽  
Vol 297-300 ◽  
pp. 2071-2076 ◽  
Author(s):  
Sung Jin Song ◽  
Chang Hwan Kim ◽  
Deok Hyun Lee ◽  
Myung Sik Choi ◽  
Do Haeng Hur ◽  
...  
Keyword(s):  
Steam Generator ◽  
Automated System ◽  
Pressurized Water ◽  
Promising Tool ◽  
Primary Water ◽  
Water Reactors ◽  
The Difference ◽  
Human Operators ◽  
Shape Curve ◽  
Steam Generator Tubes

Through-wall axial cracks occurred by primary water stress corrosion are one of the serious defects in steam generator (SG) tubes (made of alloy 600) in pressurized water reactors. Therefore, it is necessary to detect and size them by eddy current testing (ECT) conducted during in-service inspection of SG tubes. To address this issue, it has been recently proposed an effective method, namely „M-shape profile“ approach, which relies on the difference in the amplitude between the pancake and plus point coils in a MRPC probe. Even though the M-shape curve approach is straightforward in principle, it requires time-consuming data processing if performed by human operators. In order to get rid of this tedious task, an automated system is developed in the present work. This paper addresses the principle of the M-shape approach together with the automated system and its performances for the detection of natural axial cracks in SG tubes. The results observed in the present work demonstrate the high potential of the developed system as a very promising tool for detecting through-wall cracks in many practical field applications.

scholarly journals Micro-Galvanic Corrosion of Steam Generator Materials within Pores of Magnetite Flakes in Alkaline Solutions

Metals ◽  
2018 ◽  
Vol 8 (11) ◽  
pp. 899 ◽  
Author(s):  
Soon-Hyeok Jeon ◽  
Geun Song ◽  
Do Hur
Keyword(s):  
Steam Generator ◽  
Galvanic Corrosion ◽  
Area Ratio ◽  
Corrosion Products ◽  
Alkaline Solutions ◽  
Galvanic Coupling ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Alloy 690 ◽  
Water Reactors

In secondary coolant system of the pressurized water reactors, the reduced corrosion products such as metallic Cu and Pb particles were accumulated in the pores of the magnetite flakes and electrically contacted to the steam generator materials. The micro-galvanic corrosion behavior of steam generator materials (steam generator tube materials: Alloy 600 and Alloy 690, steam generator tube sheet materials: SA508 Gr.3) contacted to the corrosion products (magnetite, Cu, and Pb) was investigated in an alkaline solution. The steam generator materials considered in this study were all the anodic elements of the galvanic pair because their corrosion potentials were lower than those of the corrosion products. The corrosion rate of the steam generator materials was increased by the galvanic coupling with the each corrosion products, and was more accelerated with increasing the area ratio of the corrosion products to the steam generator materials. Among the corrosion products, Cu has the largest galvanic effect on steam generator materials in the pores when area ratio of cathode to anode is 10.

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