Steam Generator Secondary Side Research Articles

Thermal hydraulic analysis of a high-temperature gas reactor steam generator after loss of feed water accidents

The helical coil steam generator is widely employed in the energy conversion process of high-temperature gas-cooled reactors for its high thermal efficiency in producing superheated steam within the helical tubes. These tubes necessitate materials specifically designed to endure high temperatures, prevent ruptures, and ensure durability under various operational conditions. This study aims to identify the maximum temperatures arising from potential transients in the secondary system and to propose immediate safety measures for minimizing associated risks. The RELAP5/MOD3.4 code was used to simulate the behavior of the next generation nuclear plant steam generator's conceptual design using steady-state conditions. Following the validation of the model, the established steady-state conditions were utilized to simulate accident scenarios, including the feed water line break and the loss of feed water flow to the steam generator. The transient scenarios assumptions include closed flap valves and retained hot helium in the shell side. The simulation tasks were conducted both with and without activating the depressurization system, in order to assess its effect on the temperature of tubes on the wall side. The analysis results indicate that tube wall temperatures can reach 720 °C at the steam generator outlet region and 620 °C at the weld region. The immediate activation of the depressurization system with a controlled relief valve can effectively reduce these temperatures by 239 °C in the steam generator outlet and by 133 °C in the weld region. This also lowers the steam generator's secondary side pressure. As a result, an upgrade and enhancement of operational procedures is straightforward.

Validation of transient SYSTEM analysis code GINKGO based on a SLB TEST

Transient System analysis code is an important tool for reactor thermal hydraulic analysis. GINKGO is a transient system analysis in-house code developed by China General Nuclear Power Corporation (CGN). The code is primarily used for the analysis of Non-Loss of Coolant Accident (Non-LOCA)transients. Steam Line Break (SLB) accident is one of the representative accident condition of Non-LOCA transients. To validate the capability of GINKGO to simulate SLB accident scenario, a test of small non-isolatable main steam line break (10 %) at hot standby conditions, PKL III B5.1 test is selected. The characteristic of PKL facility, the PKL III B5.1 test sequence and the validation process of PKL III B5.1 test by GINKGO code are briefly introduced in the paper. In validation process, GINKGO simulates the steady state first and then simulates the transient process by setting reasonable initial conditions and boundary conditions. Then, the calculation results of key parameters including temperature in hot leg, temperature in cold leg, Steam Generator (SG) secondary side pressure and Pressurizer (PZR) pressure are compared with the experimental results. Through the rationality of the calculation results, it shows that GINKGO has the capability to simulate the key phenomena in a typical SLB accident scenario.

Structural Integrity Assessment and Fatigue Evaluation of Feed Water Nozzle of 700MWe PHWR Steam Generator

Steam Generator (SG) is a prime component in Indian Pressurized Heavy Water Reactors (PHWRs). The Steam Generator consists of an inverted vertical U-tube bundle in shell for primary flow (Heavy water). The shell side is designed for recirculation. Steam Generator is a Safety Class-1 component and designed & fabricated as per the requirements of ASME Section III, Subsection-NB.Secondary side of the steam generator consists of a shell, Feed Water Nozzle (FWN), Steam Outlet Nozzle, moisture separator, drier, tubes and shroud. The principal material of construction for secondary side pressure boundary is of low alloy carbon steel. Secondary side of steam generator is subjected various loadings (Dead weight, pressure, pipe reactions & thermal) during normal plant condition as well as transient conditions including severe transient like crash cool down where in the thermal gradients are much higher. To take care of these thermal transient loadings a special thermal sleeve type of design has been incorporated in the FWN.As mentioned above, FWN is subjected to thermal transients throughout its life and susceptible to high thermal gradient which makes it important to evaluate for structural integrity and fatigue failure. This paper covers the approach & results for structural integrity assessment (stress evaluation) and fatigue evaluation (cumulative fatigue damage factor) of Feed Water Nozzle of 700MWe PHWR SG for various Service levels A, B, D & Design. These service conditions comprise of thermal & pressure transients during operating and accidental conditions. Detailed integrity check performed for all conditions by stress categorization at critical locations and qualify primary & secondary stresses against ASME Section III allowable.

Study on diffusion and migration characteristics of impurities in secondary side of simplified u-tube steam generator

During the operation of nuclear power systems, the occurrence of leakage accidents of primary and secondary condensers are a major concern due to the potential threat to safe operation. The entry of seawater into the secondary loop from the condenser, along with impurities such as Cl-, can accumulate near the heat transfer tube, causing wear and corrosion, which can further aggravate the situation. Therefore, it is essential to accurately understand the migration and diffusion process of impurities in the secondary side of the steam generator. This study aims to investigate the three-dimensional multi-tube physical model of fluid–structure coupling heat transfer in the primary and secondary loops of the steam generator, including the heat exchange tube, to gain insights into the migration and diffusion of Cl- and other impurities. The distribution of important thermal parameters, such as volume vapor fraction, temperature, and velocity, in a U-shaped tube bundle were analyzed by numerical simulation of heat transfer and fluid flow. The diffusion and migration of Cl- in a two-phase flow were also studied. Results of this study reveal that impurities concentration is influenced by the effective diffusion coefficient, which is further influenced by the boiling intensity and flow velocity of the coolant. The combined effect of the boiling intensity and flow velocity of the coolant has a synergistic effect on the migration and diffusion of Cl-. In conclusion, this study provides valuable insights into the flow heat transfer impurity migration under a multi-tube coupled heat transfer condition of the steam generator, which can be used as a scientific basis for the design optimization and operation of steam generators.

Development of three-dimensional simulation method for two-phase flow in square-pitch tube bundle in secondary side of steam generators based on porous drift-flux model

ABSTRACT In designing a steam generator in nuclear power plant, it is necessary to evaluate the flow-induced vibration of a U-shaped tube bundle from the viewpoint of safety engineering. CFD technology is used for predicting velocity and void fraction distributions along the U-tubes to evaluate the force acting on the tube from the surrounding two-phase flow. This study validated a newly developed thermal-hydraulic simulation method based on the drift-flux model to simulate various operating conditions under flexible power operations, including partial load conditions. Constitutive equations for two-phase flow were implemented into the platform of a general-use commercial code, ANSYS fluent, through user-defined functions. Experimental data measured in the slice model U-tube bundle apparatus with boiling Freon two-phase flow, which was a simulant of steam-water under the prototypic pressure and thermal conditions, was utilized to validate the developed simulation method. The predicted void fraction and velocity distribution were compared with the experimental results. In the cold-side region, a low void fraction occurred in a partial load condition. It is considered to be caused by the liquid downward flow outside the bundle. It was confirmed that the developed simulation method could reproduce such phenomena in the secondary side of a steam generator.

Sludge Deposit Mapping for Steam Generators

During plant operation, corrosion products from feedwater, drain and condensate system accumulate on the secondary side of steam generator in form of scaled deposits around the tubes and sludge piles on top of the tubesheet. These deposits increase potential for corrosion, affect fluid flow and reduce heat transfer efficiency of the steam generator. Using data obtained during periodical eddy current examination of steam generator tubing, information on sludge deposit location and thickness is extracted from low frequency absolute channels. Sludge deposits can be detected over the whole length of the inspected tube. Using automatic analysis deposit results are provided within 36 hours after the inspection. Results are imported in Sludge Mapping software and presented in 3D visualization of steam generator. Both PWR and VVER steam generator visualization is supported. Sludge mapping software provides visual information on sludge deposit thickness using color code and by calculation position of sludge indication, software draws the indication on the location within the steam generator where indication was found. Sludge mapping software provides information on heavy sludge loading areas of steam generator and can help with tracking of sludge build-up over time that can be used for optimizing steam generator maintenance. This paper presents INETEC’s Sludge Mapping solution, its functionality and features for visualization of sludge deposit location and thickness within the steam generator.

Research on Thermal Stratification Phenomenon at the Secondary Side of Steam Generator in Isolated Loop

Research on Thermal Stratification Phenomenon at the Secondary Side of Steam Generator in Isolated Loop

Validation of three-dimensional simulation method for two-phase flow in triangular-pitch tube bundle in secondary side of steam generators on porous two-fluid model

ABSTRACT In the design process of a steam generator, a porous-media-approach computational code is utilized for simulating three-dimensional two-phase flow behavior in the secondary side of steam generators. For the design of next-generation steam generators, the advanced thermal-hydraulic analysis code based on the two-fluid model has been developed by implementing constitutive equations into the ANSYS fluent. For code validation, experiments using two-phase simulant fluids were performed to collect data in a simulated steam generator with a triangular tube array. The simulant two-phase flow system selected in the experiment was an adiabatic sulfur hexafluoride gas and liquid ethanol system, which allowed us to achieve the density ratio of prototypic steam-water flow with low pressure (=0.68 MPa). In the experiment, void fraction and gas–liquid interfacial velocity distributions along U-tubes were measured. The code validation was conducted by analyzing code predictions for measured distributions of void fraction and interfacial velocity under prototypic full load and partial load conditions, i.e., 50–80% flow rate conditions. Their bias and random errors were evaluated. The reasonable random errors demonstrated the validity of the newly developed code based on the two-fluid model in terms of predictions of void fraction and interfacial velocity in the secondary side of steam generators.

Scaling analysis of the thermal hydraulic behavior in the secondary side of steam generator

Steam generator is one of the key facilities in nuclear power plant. Due to its complex structure and huge volume, the experimental and numerical research of steam generator needs a large amount of costs and computational resources. One feasible method is to replace the prototype steam generator with a model by the scaling analysis. In this work, a systematical scaling analysis for the thermal hydraulic phenomenon in the secondary side of steam generator in ocean motions is carried out. The scaling analysis is decomposed into three processes, natural circulation, forced convection and lateral flow. A series of scaling criteria are obtained. It is found that the similarity of heat transfer could be achieved by adjusting the parameters of the primary side. It is also revealed that the conventional scaling method could not guarantee the similarity of transverse flow due to the improper similarity of lateral resistance. Besides that, the scaling criteria for the amplitude and period are also obtained, which could be used in the subsequent experiments and numerical simulation.

Liquid spray modeling under sodium fire accidents

The Sodium-Cooled Fast Reactor (SFR) system, which features a fast-spectrum, sodium-cooled reactor and a closed fuel cycle for efficient management of actinides and conversion of fertile uranium, is rather prominent in the nuclear reactor design in Generation IV (Gen IV), and likely to be widespreadly used and well-constructed all over the world. However, molten sodium may be sprayed into the ambient or secondary side of steam generator, induced by the pressure drop, in case of an inevitable break in the primary system. This kind of leakage can result in sodium spray fire accident, as a result of the chemical activity of sodium, significantly damaging the equipment and working staffs nearby. Sodium spray fire, which occurs not only in the nuclear industry, but also in solar industry, is a complex process consisting of several specific phenomena, such as the spray dynamics (e.g. droplet particle size/velocity distribution, particle collision and agglomeration), droplets evaporation, sodium aerosol diffusion, thus being drawn many experts’ attention. Based on International Atomic Energy Agency (IAEA)’s effort on phenomena identification and ranking table (PIRT) for sodium spray fire, prediction of droplet size distribution and the mean diameter in particular, are supposed to be in the very first place at level 2, due to the lack of available experimental data for both of model development and validation, and code inadequacy, as well. In this paper, sodium droplet size distribution model is theoretically derived, regarding both of mass and momentum equations as constraints, based on the maximum entropy principle, which can quantitatively describe the effects of two main influencing factors, including initial injection velocity and thermophysical properties of molten sodium. This analytical model is implemented in three typical conditions, including two sodium droplets size distribution experiments and one water droplet experiment. Validations against both of the available test data and predictions by an empirical model, named of Nukiyama-Tanasawa correlation, are conducted with the relative error of median diameter ranging from 17.3% to 18.2%, which ensures the reliability and feasibility of the new model. Moreover, sensitivity analyses are carried out, in terms of typical test conditions of sodium spray fire, which demonstrates that the median diameter of droplets’ size decreases with both of the increase of initial injection velocity and sodium temperature, thus probable triggering positive effect on the propagation of the accident, due to rapid interaction of violent chemical reaction and sodium spray. Besides, the present work can provide a reliable analytical model for the development of safety analysis codes for SFRs.

Difference analysis of in-service hydrostatic test for second and third generation PWR

Hydraulic test of nuclear power plant is an important part of in-service inspection of nuclear power plant, and it is one of the important technical means to test the pressure bearing capacity of equipment. In this paper, the technical requirements of the second generation PWR based on safety classification and the third generation reactor based on radioactive release classification are discussed from the aspects of equipment classification, primary circuit hydrostatic test, steam generator secondary side hydraulic test, hydrostatic test of vessel and pipeline, as well as the difficulty of implementation in hydrostatic test.

Investigation of the equalization capability of submerged perforated sheets forWWER-1500 horizontal steam generator

Non-uniform heat transfer occurs at the steam generator secondary side, thus inducing a non-uniform steam generation and a non-uniform steam flow rate along the evaporation surface. In order to equalize the steam flow rate, submerged perforated sheet is installed in the horizontal SGs An analysis of the equalization of the non-uniform steam load in the WWER-1500 horizontal steam generator is performed using the STEG calculation code. Variant calculations of various options for the perforation of sheets allowed us to determine the design of submerged perforated sheet, which very well equalizes the initially non-uniform steam load.

A 3D model to solve U-tube steam generator secondary side thermal-hydraulics with coupled primary-to-secondary side heat transfer

The U-tubes of steam generators (SG) are concerned by vibration issues. The vibration behavior is driven by the local secondary flow (velocity, void fraction, two-phase pattern). Fluid-elastic instability (FEI) can be considered as the most severe vibration phenomenon, since it features high vibration amplitudes and tube-to-tube contacts, which can rapidly cause tube rupture. In order to predict if the fluid forces exerted along the tube bundle can trigger a FEI, numerical tools able to calculate the secondary side of a SG are needed. In the frame of its mission of assessing methods adopted by utilities for demonstrating the safe operation of nuclear facilities, the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) has developed such a numerical model. The model is based on a 3D homogeneous drift-flux formulation of the SG steam-water mixture, together with a porous media approach to account for the tube bundle. The model is implemented in the ANSYS Fluent CFD code. The SG primary side is solved using a dedicated meshing. The secondary side and the primary side meshes communicate through heat transfer, which is fundamental to obtain consistent numerical results. The details of the numerical approach are described in the first part of this paper. Then, numerical versus experimental results comparisons in terms of heat transfer are provided for a SG test facility: numerical and experimental results are in good agreement. The model was then applied to a real SG: numerical results are in good agreement with nominal operating parameters.

Study on Flow Interference of Multi-group Parallel Jets between Heat Transfer Tubes

In order to find an approach to optimize the hydraulic cleaning equipment of steam generator secondary side of CPR1000 reactor, the Computational Fluid Dynamics (CFD) method was used to simulate the flow of multiple groups of parallel jets. In this simulation, the RNG k-ε turbulence model and the gas-liquid two-phase flow model is employed to distribute the jet flow between heat transfer tubes. The results show that with the increase of the jet angle, the pressure aggregation on the tube sheet decreases obviously, which is caused by the deflection force due to the pressure difference between the two sides of the jet. With the jets distance increasing, the interaction between jets decreases obviously. When the distance is up to more than nine times of tube spacing, the force that causes deflection of the jet will almost disappear. With this distance, the jets can maintain a good linearity and obtain a satisfied cleaning effect on the heat transfer tubes.

Numerical and analytical predictions of nuclear steam generator secondary side flow field during blowdown due to a feedwater line break

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

Analysis of fission product reduction strategy in SGTR accident using CFVS

In order to reduce risks from the Steam Generator Tube Rupture (SGTR) accident and to meet safety targets, various measures have been analyzed to minimize the amount of fission product (FP) release. In this paper, we propose an introduction of a Containment Filtered Venting System (CFVS) connected to the steam generator secondary side, which can reduce the amount of FP release while minimizing adverse effects identified in the previous studies. In order to compare the effect of new equipment with the existing strategy, accident simulations using MELCOR were performed. As a result of simulations, it is confirmed that CFVS operation lowers FP release into the environment, and the release fractions are lower (minimum 0.6% of the initial inventory for Cs) than that of the strategy which intends to depressurize the primary system directly (minimum 15.2% for Cs). The sensitivity analyses identify that refill of the CFVS vessel is a dominant contributor reducing the amount of FP released. As the new strategy has the possibility of hydrogen combustion and detonation in CFVS, the installation of an igniter inside the CFVS vessel may be considered in reducing such hydrogen risk.

Investigation of two-phase flow hydrodynamics under SGTR severe accident conditions

The international EU-SGTR and ARTIST projects investigated the transport of fission products in the form of aerosols during SGTR severe accidents. The major finding of the two projects was that there was significant retention of aerosols in the steam generator secondary side, and that the retention was increased by more than an order of magnitude if the secondary side of the steam generator was flooded with water. Furthermore, the experiments with the flooded secondary side showed that the aerosol particle retention was significantly increased due to the presence of submerged structures, i.e., the tube bundle, as compared to an empty pool. The increased aerosol retention was attributed to the interactions of the high velocity gas jet discharged from the tube break with the dense bundle of the steam generator tubes. Under these conditions, the two-phase flow is very complex due to the high gas velocities and complicated geometry of the steam generator secondary side.To determine the effect of the tube bundle on aerosol retention, hydrodynamic characteristics of an empty pool and a flooded steam generator secondary side were measured in a facility equipped with wire-mesh sensors. The facility was equipped with either a tube bundle consisting of 221 steam generator tubes, or with a single tube in the center of the facility. The flow development could be followed by making the measurements at different distances between the gas injection and the measurement point, and using different flow rates. The facility was operated at close to ambient conditions.Void fraction, bubble size distributions, gas phase velocity as well as interfacial area concentration were determined based on the wire-mesh sensor data. In addition, the penetration depth of the initial large gas bubble into the channel was studied for the closest break-sensor distances. The investigations show distinct differences between the flow characteristics in the single tube geometry and in the tube bundle. As compared to the single tube, the flow was more confined in the tube bundle due to the interactions of the flow with the tubes. The interfacial area between the liquid and gas phase is larger with the tube bundle than with the single tube and also the bubble size distributions show distinct differences between the two geometries.

A summary of fission-product-transport phenomena during SGTR severe accidents

A steam-generator-tube rupture (SGTR) in a pressurized-water reactor is a design-basis accident with which plants have been designed to cope. Its specific significance is in its potential to bypass the containment thereby providing a pathway for radioactivity release to the environment. In combination with significant core damage, uncontrolled SGTR may lead to accidental release of radioactivity in the form of aerosol particles and gaseous compounds.In this paper, we describe the phenomena identification and ranking tables (PIRT) prepared as an international collaboration for uncontrolled SGTR in connection with core damage. This PIRT is limited to fission-product transport and retention in the steam generator and release to the auxiliary building, i.e., the source term. The phenomena affecting the fission-product retention in the steam generator and release to the environment were divided into those relevant in a dry steam-generator secondary side, and those relevant in a flooded secondary side. Rankings were given to 17 phenomena in the dry and 16 phenomena in the flooded secondary side regarding their: i) impact on the source term, ii) the availability of experimental data applicable to SGTR severe-accident conditions, and iii) the availability of codes/models to describe the phenomenon under the conditions of SGTR severe accidents.The SGTR boundary conditions and the considered phenomena are briefly described. The results are given in the form of ranking tables separately for the dry and flooded steam generator secondary side. The most important phenomena and their interactions with other phenomena are discussed with references to existing data and models. The most important uncertain phenomena in the dry steam generator scenario are particle bounce and fragmentation; whereas inertial impaction and retention by droplets in the jet region are the most uncertain ones in the flooded configuration. The need for improved knowledge is, though, tempered by overall high aerosol retention in the latter scenario. The applicability of the existing data and models for SGTR accident analysis is considered in the discussion.

Effects of pH control agents on magnetite deposition on steam generator tubes

Magnetite deposition on the secondary side of steam generators is a major concern in nuclear power plants. Therefore, it is necessary to mitigate the deposition using optimized water chemistry control. The purpose of this study is to investigate the effects of three pH control agents, ammonia, morpholine, and ethanolamine, on the magnetite deposition onto an Alloy 690TT tube under conditions similar to operating steam generators. The deposited magnetite particles were polyhedral or spherical in shape and numerous micropores were clearly observed in the deposits regardless of pH agents. The amount of deposited magnetite was lowest when ammonia was used, and increased approximately 1.7 times and 2.6 times when using morpholine and ethanolamine, respectively. The effects of the pH agents were discussed based on the difference of zeta potentials between magnetite particles and the bare surface of the Alloy 690TT tube, particle size, and magnetite solubility.

Local effect model development for the steam generator three dimensional thermal hydraulics analysis code

The thermal hydraulic characteristics of steam generator secondary side in pressurized water reactor has been always a research hot point in the literatures. The detailed three dimensional thermal hydraulic parameters distribution in steam generator is very important for the reactor safety and economy. A three-dimensional two-phase analysis program STAF for the secondary side of steam generator has been developed by NuTheL at Xi'an Jiaotong University. The program is based on porous medium model and four-equation drift flow model. It can simulate the flow and heat transfer characteristics of steam generator. However, due to the limitation of porous medium model, STAF are not functional to obtain the local velocity distribution near the supporting plate and anti-vibration bars, which are very important for fouling deposition and flow vibration analysis. In this paper, the local detailed thermal hydraulic characteristics near steam generator supporting plate and anti-vibration bars are studied. The local flow field is obtained and the influence range and degree of local components on the velocity distribution is analyzed. The supporting plate and anti-vibration bars analysis model was developed and implemented into the STAF code. Results show that the improved code could predict the local velocity variations compared with the original code. This work is meaningful for the function promotion of steam generator thermal hydraulic code STAF.