Study on the Modeling and Simulation of the Horizontal Steam Generator in VVER-1000

2021 ◽  
Author(s):  
Ru Zhang ◽  
Junyan Qing ◽  
Xiaolong Bi ◽  
Guanfu Jiang ◽  
Peiwei Sun ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Tube Bundle ◽  
Natural Circulation ◽  
Thermal Power ◽  
Secondary Circuit ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Horizontal Steam Generator ◽  
Secondary Side

Abstract Steam Generator (SG) is one of the main components of the power cycle in pressurized water reactor (PWR), and it is the hub of primary coolant circuit and secondary circuit, so the thermal hydraulic analysis of the SG is crucial in the system design and safety analysis of the PWR. The horizontal steam generator (HSG) is one of the main types SG in the PWR nuclear power plant (NPP), and its advantages are that it has more secondary side water capacity and good safety and reliability. The VVER-1000 is a PWR with a thermal power of 3000 MW, and has four HSGs for four loops. The RELAP5 has been used to model the VVER-1000’s HSG and performs the analysis described in this paper. The HSG tube bundle is modeled by three horizontal channels, and the steam control volumes above the heat transfer tube bundle are modeled with three volumes. The steam space is modeled as a steam separator and the steam reception shield is the dryer. The HSG secondary side downcomers are represented with a separate component to provide the power of the natural circulation. To verify the accuracy of the model, three different typical conditions are simulated. The simulation results show that the model built in this paper can correctly simulate the operation of the HSG in VVER-1000.

Influence of Steam Generator Tube Bundle Vibrations on the Operating Diagram of a Nuclear Plant During Stretch Out

2006 ◽  
Author(s):  
Andre´ Adobes ◽  
Joe¨l Pillet ◽  
Franck David ◽  
Michae¨l Gaudin
Keyword(s):  
Steam Generator ◽  
Cooling System ◽  
Tube Bundle ◽  
Thermal Power ◽  
Nuclear Plant ◽  
Coolant Temperature ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Normal Cycle ◽  
System Pressure

During the normal cycle of a pressurized water reactor, boron concentration is reduced in the core until fuel burns up. A stretch out of the normal cycle is however possible afterwards, provided primary coolant temperature is reduced. In those stretch out periods, nuclear operators want to keep constant thermal power exchanged in the steam generator, in order to preserve its performances. Under that constraint, the required reduction in primary coolant temperature involves both a decrease of secondary cooling system pressure and an increase of tube bundle vibrations. Since neither pressure nor vibrations should exceed some given thresholds in order to preserve component integrity, the reduction of primary coolant temperature has to be limited. Nuclear plant operators thereafter need an operating diagram, i.e. a diagram that provides minimum allowed primary coolant temperature versus power rate. In that context, we propose a method to derive such a diagram, by combining, on the one hand a code for simulating primary and secondary fluid flows in steam generators and, on the other hand, a software that allows one to predict fluid elastic tube bundle instabilities. That method allows one to take into account both tube fouling and plugging. It is now used by French utility “Electricite´ De France”, in order to check or supplement the analysis that are provided by steam generator manufacturers.

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.

The simulation research on the natural circulation operation characteristic of FNPP in rolling and inclined condition

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 4-16
Author(s):  
R. Li ◽  
M. Peng ◽  
G. Xia ◽  
H. Li
Keyword(s):  
Nuclear Power ◽  
Natural Circulation ◽  
Rapid Development ◽  
Rolling Motion ◽  
Stable Operation ◽  
Circulation Flow ◽  
Operation Condition ◽  
Pressurized Water ◽  
The Core ◽  
Secondary Side

Abstract Recently, the FNPP (Floating Nuclear Power Plant) has got more and more attention and rapid development due to very wide prospect application on remote areas or islands. In general, the IPWR (Integral Pressurized Water Reactor) is adopted to meet the requirements of the limited space, the nuclear safety and the maneuverability in marine. The IPWR could depend on natural circulation operation to remove the residual heat of core under accident or low load operation condition. Because the driving head is low, the natural circulation flow is likely to be influenced by rolling and inclined condition. To clarify the natural circulation flow characteristics of the core in FNPP rolling motion and inclined condition, based on the modified THEATRe code by adding the ocean motion module and spatial coordinate convert module, the main thermal-hydraulic parameters variation in rolling and inclined condition were obtained. The effect of inclined angle, rolling amplitude and period on the natural circulation flow were discussed. The natural circulation flow in the core fluctuates periodically with rolling motion. And the inclination and rolling will also cause the degree of steam superheat of OTSG secondary side fluctuate, which could impact on the stable operation of secondary side system.

Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)

Kerntechnik ◽  
2020 ◽  
Vol 85 (1) ◽  
pp. 54-67
Author(s):  
A. Hamedani ◽  
O. Noori-Kalkhoran ◽  
R. Ahangari ◽  
M. Gei
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Axial Direction ◽  
Pressurized Water Reactors ◽  
Steam Generators ◽  
Pressurized Water ◽  
Heated Channel ◽  
Water Reactors ◽  
Secondary Side

Abstract Steam generators are one of the most important components of pressurized-water reactors. This component plays the role of heat transfer and pressure boundary between primary and secondary side fluids. The Once Through Steam Generator (OTSG) is an essential component of the integrated nuclear power system. In this paper, steady-state analysis of primary and secondary fluids in the Integral Economizer Once Through Steam Generator (IEOTSG) have been presented by Single Heated Channel (SHC) and subchannel modelling. Models have been programmed by MATLAB and FORTRAN. First, SHC model has been used for this purpose (changes are considered only in the axial direction in this model). Second, the subchannel approach that considers changes in the axial and also radial directions has been applied. Results have been compared with Babcock and Wilcox (B&W) 19- tube once through steam generator experimental data. Thermal- hydraulic profiles have been presented for steam generator using both of models. Accuracy and simplicity of SHC model and importance of localization of thermal-hydraulic profiles in subchannel approach have been proved.

The PSI Artist Project: Aerosol Retention and Accident Management Issues Following a Steam Generator Tube Rupture

2002 ◽  
Author(s):  
Salih Gu¨ntay ◽  
Abdel Dehbi ◽  
Detlef Suckow ◽  
Jon Birchley
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Severe Accident ◽  
Pressurized Water ◽  
Steam Generator Tube ◽  
Full Size ◽  
Management Actions ◽  
Accident Management ◽  
Dry Conditions ◽  
Secondary Side

Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes’ models for aerosol retention in the secondary side of a steam generator (SG) have not been assessed against any experimental data, which means that the uncertainties in the source term following an unisolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (AeRosol Trapping In a Steam generaTor), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2) the near field close to break under dry conditions, 3) the bundle far-field under dry conditions, 4) the separator and dryer under dry conditions, 5) the bundle section under wet conditions, 6) droplet retention in the separator and dryer sections and 7) the overall SG (integral tests). Prototypical test parameters are selected to cover the range of conditions expected in severe accident as well as DBA scenarios. This paper summarizes the relevant issues and introduces the ARTIST facility and the provisional test program which will run between 2003 and 2007.

Natural Circulation Characteristics of an Integral Pressurized Water Reactor

2006 ◽  
Author(s):  
Junli Gou ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
Dounan Jia
Keyword(s):  
Steady State ◽  
Steam Generator ◽  
Natural Circulation ◽  
Heat Removal ◽  
Primary Circuit ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Secondary Side ◽  
Circulation Characteristics

Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation.

Modifications of Generic Pressurized-Water Reactor Units to Increase Cycle Efficiency

2016 ◽  
Author(s):  
Paul Ponomaryov ◽  
Yifeng Zhou ◽  
Cristina-Maria Mazza ◽  
Igor Pioro
Keyword(s):  
Moisture Content ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Thermal Power ◽  
Mechanical Efficiency ◽  
Moisture Removal ◽  
Pressurized Water ◽  
Water Reactors ◽  
Lp Turbine ◽  
Secondary Side

Currently, Pressurized Water Reactors (PWRs), Boiling Water Reactors (BWRs) and Pressurized Heavy Water Reactors (PHWRs) have the lowest thermal efficiencies compared to those of other nuclear-power reactors and thermal power plants. Therefore, the objective of this paper is to propose modifications to a generic PHWR to yield an increase in overall plant thermal efficiency. The focus of this paper is primarily on the secondary side of a Nuclear Power Plant (NPP) and is directly dealing with wet-steam High-Pressure (HP) and Low-Pressure (LP) turbine stages and a Moisture Separator and Reheater (MSR). Modifications of the HP and LP turbine stages are based on utilizing moisture removal stages (having up to 60% removal efficiency), which reduce the moisture content as the steam passes through those turbine stages. Reduced energy losses and an increase in mechanical efficiency due to lesser moisture content results in an increase in thermal efficiency. Furthermore, implementing moisture-removal stages in the LP turbine gives the ability to eliminate the reheater in the MSR, thus resulting in an increase of thermal efficiency due to both, the higher mechanical efficiency of an LP turbine and the redirection of live-steam previously used by the MSR to a HP turbine. To be able to show an increase in thermal efficiency based on these modifications of a generic PHWR, the Pickering CANDU-6 nuclear-reactor parameters were used as a reference case in the software, called DE-TOP. The modifications suggested in this paper can be applied to any NPP that uses a Rankine steam-turbine cycle on the secondary side (PWR, PHWR and/or BWR) and recommended for implementation during planned replacement of LP- and/or HP-turbine rotors for new construction of PWRs and PHWRs.

Intentional Back Flow Effects on Ruptured Steam Generator Cooldown During a SGTR Event for KSNP

2004 ◽  
Author(s):  
Cheol Woo Kim ◽  
Seok Jeong Park ◽  
Chul Jin Choi ◽  
Jong Tae Seo
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Cooling System ◽  
Initial Conditions ◽  
Natural Circulation ◽  
Radioactive Material ◽  
Steam Condensation ◽  
Back Flow ◽  
Operation Conditions ◽  
Secondary Side

For an optimum recovery from a steam generator tube rupture (SGTR) event, the operators are directed to isolate the steam generator (SG) with ruptured tube as early as possible to minimize the radioactive material release. However, the reactor coolant system (RCS) cooldown and depressurization to the shutdown cooling system (SCS) operation conditions using the intact SG only are hard to achieve unless the ruptured SG is properly cooled since the ruptured SG, which is isolated by operator, remains at high temperature even though the RCS has been cooled down. The effects of intentional back flow from the SG secondary side to the RCS through the ruptured U-tube on the the ruptured SG cooldown were evaluated for the pressurized light water reactor, especially for the Korean standard nuclear power plant (KSNP). In order to evaluate the back flow effect, a series of analyses was conducted using the RELAP5/MOD3 computer code. For the first stage of the analysis, the cooldown process by natural circulation in the SG secondary side was simulated for the initial conditions of the ruptured SG cooldown. In the next analysis stage, two methods of the ruptured SG cooldown by using back flow after RCS cooldown were evaluated. The one is a tube uncovery, which utilizes the steam condensation on the uncovered U-tube surface, and the other is a SG drain and fill. In the former method, SG tubes are exposed to the steam space by draining SG secondary water into the RCS in order to condense the steam directly onto the uncovered tubes. This method showed that the steam condensation decreased SG secondary pressure and temperature rapidly, demonstrating its effectiveness for cooling. However, this process has a limited applicability in cases of that the rupture is located at the lower region. The latter method, draining by back flow and filling using the feedwater system was also found to be effective in ruptured SG cooldown and depressurization even if the rupture occurred at the top of the U-tube. However, the actuation of the feedwater system is a burden to operator since the makeup of cold feedwater was required to complete cooldown by one cycle of the draining and filling. It is concluded that utilization of the intentional back flow from the SG secondary side to the RCS is very effective for rapid cooling of the RCS to the SCS entry conditions.

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

2018 ◽  
Author(s):  
Katarzyna Skolik ◽  
Anuj Trivedi ◽  
Marina Perez-Ferragut ◽  
Chris Allison
Keyword(s):  
Natural Circulation ◽  
System Modeling ◽  
Thermal Power ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Circulation System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Steam

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break1

2015 ◽  
Vol 137 (4) ◽  
Author(s):  
Jong Chull Jo ◽  
Frederick J. Moody
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Analysis Model ◽  
Pressurized Water ◽  
Line Pipe ◽  
Short Period ◽  
Main Steam Line ◽  
Main Steam ◽  
Transient Thermal ◽  
Secondary Side

This paper presents a multidimensional numerical analysis of the transient thermal-hydraulic response of a steam generator (SG) secondary side to a double-ended guillotine break of the main steam line attached to the SG at a pressurized water reactor (PWR) plant. A simplified analysis model is designed to include both the SG upper space, which the steam occupies and a part of the main steam line between the SG outlet nozzle and the pipe break location upstream of the main steam isolation valve. The transient steam flow through the analysis model is simulated using the shear stress transport (SST) turbulence model. The steam is treated as a real gas. To model the steam generation by heat transfer from the primary coolant to the secondary side coolant for a short period during the blow down process following the main steam line break (MSLB) accident, a constant amount of steam is assumed to be generated from the bottom of the SG upper space part. Using the numerical approach mentioned above, calculations have been performed for the analysis model having the same physical dimensions of the main steam line pipe and initial operational conditions as those for an actual operating plant. The calculation results have been discussed in detail to investigate their physical meanings and validity. The results demonstrate that the present computational fluid dynamics (CFD) model is applicable for simulating the transient thermal-hydraulic responses in the event of the MSLB accident including the blowdown-induced dynamic pressure disturbance in the SG. In addition, it has been found that the dynamic hydraulic loads acting on the SG tubes can be increased by 2–8 times those loads during the normal reactor operation. This implies the need to re-assess the potential for single or multiple SG tube ruptures due to fluidelastic instability for ensuring the reactor safety.

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