Pressurized Water Reactor Steam Generator Research Articles

Elucidating the interaction mechanisms between polyacrylic acid and Alloy 800 oxide films under pressurized water reactor steam generator conditions

The interaction mechanisms between polyacrylic acid (PAA) and Alloy 800 were investigated. Alloy 800 formed a bilayer oxide structure in all conditions, with an outer layer of Fe-rich spinel and α-Ni(OH)2 and a Cr-rich inner layer. The chelating effect of PAA on Fe ions reduced the presence of Fe-rich spinel in the outer layer and increased the Cr content in the inner layer. However, thermal degradation of PAA (≥ 500 ppb) promoted dissolution of Cr in the initial inner layer, increasing the initial corrosion rate. The optimal PAA dosage in steam generators is determined to be approximately 250 ppb.

The Study of Multiaxial Loading and Damage to the Structure and Materials in the PWR Steam Generator of Nuclear Reactor

In Pressurized Water Reactor (PWR) Nuclear Power Plants (NPPs), the steam generator is crucial for transferring heat from the primary to secondary cooling systems, vital for steam production to drive turbines, and central to nuclear power safety. This study explores recent research on multi-axial loading, structural integrity, and material durability in PWR steam generators, shedding light on key factors affecting these systems. Common corrosion-related degradation in steam generators often arises from design, material, and water chemistry factors. However, the shift to All Volatile Treatment (AVT), the development of advanced material alloys, and enhanced water quality control in primary and secondary systems have significantly reduced instances of steam generator degradation. These findings promise to enhance the reliability and safety of steam generators in future nuclear applications.

Design of a laser welding plugging system for the steam generators of high-temperature gas-cooled reactor

As fourth-generation nuclear power equipment, high-temperature gas-cooled reactor (HTGR) is attracting more and more attention both in industry and academia. The core outlet temperature of the steam generator (SG) in HTGR is much higher than that of pressurized water reactor steam generators, which is more likely to cause damage. Till now, welding plugging is still one of the important means to prevent the leakage of damaged heat transfer tubes. However, how to improve the automation of welding plugging to reduce the short-range operation time in a radioactive environment is still difficult. This work proposes the design of a laser welding plugging system for the HTGR. Firstly, the design requirements are analyzed and the overall design scheme is given. The whole system consists of a laser generator, a multi-axis robot, and supporting devices. Secondly, the key supporting devices including a laser welding head, a pneumatic clamping jaw, and a single-axis motion platform are explained in detail. The structure simulation based on SolidWorks is adopted, which provides the basis for the design of the single-axis motion platform. Thirdly, the control system which takes the PLC control system as the main control unit and the industrial computer as the monitoring data processing unit is described. Finally, an experimental system is built and the laser welding experiment is carried out to test the feasibility of the designed laser welding plugging system.

Study on Performance and Operation Mechanism of a Separation Equipment for a PWR Steam Generator

Computational fluid dynamics (CFD) is adopted to calculate and analyze the performance and separation mechanism of a steam-water separation equipment for a pressurized water reactor (PWR) steam generator. The steam-water two-phase flow is simulated by the Euler two-fluid model, and several representative water droplet diameters are selected among 50 to 400 μm. The influence mechanism of water droplet diameter on the performance is investigated by analyzing the flow parameters such as phase volume fraction, velocity, and turbulent kinetic energy. The results show that the integrated calculation and analysis of the separation equipment can more truly reflect the flow state between the separators, improving the reliability of the performance prediction and mechanism analysis. With the increase in water droplet diameter, the separation efficiency of the separation equipment and each separator gradually increases, the outlet wetness gradually decreases, and the pressure loss first decreases and then stabilizes. When 400 μm is taken as the characteristic value of the actual droplet diameter distribution at the inlet of the separation equipment, the performance prediction is more accurate, the pressure loss of each separator is relatively close, and the pressure loss of the primary separator is less affected by the droplet diameter and smaller than that of the swirl-vane primary separator, which is conducive to achieving a higher circulation ratio (CR). Re-entrainment occurs at the perforations of the primary separator and the outlet of the secondary separator, and the corresponding structure is suggested to be optimized to further improve separation efficiency of the separation equipment.

Study on properties of 18MND5 steel forgings for PWR steam generator

In the pressurized water reactor (PWR) nuclear power plant, the shell material of steam generator is required to have good strength-toughness matching, anti-fatigue performance, and neutron radiation resistance to ensure the long-term safe and reliable service. The manufacturing requirements of Manganese-Nickel-Molybdenum alloy steel forgings of steam generator channel head for Hua-long Pressurized Reactor (HPR1000) and Advanced Passive PWR(AP1000) nuclear power plant were compared. The heat treatment process, chemical composition, mechanical properties, metallographic structure of 18MND5 steel forging and SA 508-III steel forging were analysed and studied. The results show that the performance heat treatment temperature of HPR1000 forging has more stringent controls and with lower element content of C, Mo, P, Si, V, Co than the AP1000 forging. By reducing the C content, HPR1000 forging got better toughness while the strength was ensured.

Numerical Study of the Deposition of Radioactive Nuclide (Aerosol) in Pressurized Water Reactor Steam Generator

Abstract This research presents a numerical study of deposition efficiency and decontamination factor of the radioactive nuclide in steam generator tubes of a typical 325 MWel pressurized water reactor (PWR). To find out the deposition of aerosol, the discrete phase model (DPM) has been used. The flow has been characterized as compressible, adiabatic, turbulent and wall-bounded. When steam generator tube gets ruptured, the radioactive nuclides can escape from primary side and create a radioactive field on the secondary side. This can be harmful at the personnel working at the plant. Therefore, in order to ensure the safety of the plant and personnel, it is important to study the particles deposition on the wall of steam generator tubes. In the present study, a computational fluid dynamics (CFD) methodology has been first developed and validated with the published results. After methodology validation, it has been applied to the U-tube of a typical PWR steam generator. It has been observed that due to the action of centrifugal force near the bent, the velocity magnitude is high toward the inner wall and the flow separates at the bent entrance. Furthermore, the flow inside the tube is rotational with vortices throughout the domain due to the presence of the bent. Finally, the deposition efficiency and decontamination factor have been calculated and it has been observed that both increase with the increase in particle size due to inertial effects.

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.

Numerical prediction of transient hydraulic loads acting on PWR steam generator tubes and supports during blowdown following a feedwater line break

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

Numerical Analysis of Subcooled Water Flashing Flow From a Pressurized Water Reactor Steam Generator Through an Abruptly Broken Main Feed Water Pipe

A computational fluid dynamics (CFD) analysis was performed to investigate the hydraulic response of the flow field inside the pressurized water reactor (PWR) steam generator (SG) secondary side and the connected part of main feed water pipe to an abrupt main feed water line break (FWLB) accident. To realistically analyze the transient flow field situation, the flow field was assumed to be occupied initially by highly compressed subcooled water except that the upper part of the SG secondary side where steam occupied as in the practical case and the break was assumed to occur at the circumferential weld line between the feed water nozzle and the main feed water pipe. This would result in a subcooled water flashing flow from the SG through the short-broken pipe end to the surrounding atmosphere, which was numerically simulated in this study. Typical results of the prediction in terms of the fluid transient velocity and pressure were illustrated and discussed. To examine the physical validity of the present numerical simulation of the subcooled water flashing flow, the transient mass flow rates predicted in this study were compared with the other previous numerical predictions based on the subcooled water nonflashing (no phase change) flow or saturated water flashing flow assumptions and the prediction by a simple analysis method.

Numerical prediction of a flashing flow of saturated water at high pressure

Abstract Transient fluid velocity and pressure fields in a pressurized water reactor (PWR) steam generator (SG) secondary side during the blowdown period of a feedwater line break (FWLB) accident were numerically simulated employing the saturated water flashing model. This model is based on the assumption that compressed water in the SG is saturated at the beginning and decompresses into the two-phase region where saturated vapor forms, creating a mixture of steam bubbles in water by bulk boiling. The numerical calculations were performed for two cases of which the outflow boundary conditions are different from each other; one is specified as the direct blowdown discharge to the atmosphere and the other is specified as the blowdown discharge to an extended calculation domain with atmospheric pressure on its boundary. The present simulation results obtained using the two different outflow boundary conditions were discussed through a comparison with the predictions using a simple non-flashing model neglecting the effects of phase change. In addition, the applicability of each of the non-flashing water discharge and saturated water flashing models for the confirmatory assessments of new SG designs was examined.

Thermal Hydraulic Modeling of Once-Through Steam Generator by Two-Fluid U-Tube Steam Generator Code

The THERMIT U-tube steam generator (THERMIT-UTSG) code was used for evaluation for the parametric study of a scaled once-through pressurized water reactor steam generator (OTSG) made by Babcock & Wilcox. The results of the code were compared to the experimental data of the 19-tube OTSG and a simple heat transfer code that was developed by Osakabe. The main calculated thermodynamic parameters were primary-secondary fluid temperatures, tube wall internal and external temperatures that were subjected to primary and the secondary fluid, and the secondary fluid vapor quality. The assessed code can be used for modeling the OTSGs with some modification. The results of THERMIT-UTSG were in agreement with the experimental results and the prediction of Osakabe’s numerical model.

Evaluation of a Numerical Analysis Model for the Transient Response of Nuclear Steam Generator Secondary Side to a Sudden Steam Line Break

This paper presents an evaluation of the applicability of a numerical analysis model to the transient thermal-hydraulic response of steam generator (SG) secondary side to blowdown following a steam line break (SLB) at a pressurized water reactor (PWR). To do this, the numerical analysis model was applied to simulate the same blowdown situation as in an available experiment which was conducted for a simplified SG blowdown model, and the numerical results were compared with the measurements. As a result, both are in reasonably good agreement with each other. Consequently, the present numerical analysis model is evaluated to have the applicability for numerical simulations of the transient phase change heat transfer and flow situations in PWR SGs during blowdown following a SLB.

Effects of a Venturi-Type Flow Restrictor on the Thermal-Hydraulic Response of the Secondary Side of a Pressurized Water Reactor Steam Generator to a Main Steam Line Break

A numerical analysis has been performed to simulate the transient thermal-hydraulic response to a main steam line break (MSLB) for the secondary side of a steam generator (SG) model equipped with a venturi-type SG outlet flow restrictor at a pressurized water reactor (PWR) plant. To investigate the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB, numerical calculation results for the SG model equipped with the flow restrictor are compared to those obtained for an SG model without the restrictor. Both analysis models contain internal structures. The present computational fluid dynamics (CFD) model has been examined by comparing to a simple analytical model. It is confirmed from the comparison that the CFD model simulates the transient response of the SG secondary to the MSLB physically plausibly and minutely. Based on the CFD analysis results for both cases with or without the restrictor, it is seen that the intensities of the steam velocity and dynamic pressure are considerably attenuated in the SG model equipped with the restrictor comparing to the case in the SG model without the restrictor.

PWR蒸気発生器蒸気量増大に関する研究（フィンおよびねじり板導入による伝熱改善）

Increasing the electric power of nuclear power plant is important for the economy and the environment. Pressurized water reactor (PWR) exchanges the heat of primary system with secondary system in the steam generator (SG) to generate vapor. In this study, the steam generator of PWR was focused on. A prediction method to estimate the vapor mass flow rate of steam generator has been developed. In the method, the heat transfer from the primary system to the secondary system can be categorized into the following types. (1) Heat transfer from the primary system to the heating tube. (2) Heat conduction of the heating tube. (3) Heat transfer between the heating tubes and the secondary system. The liquid in the primary system is in the single-phase, while the liquid in the secondary system is in both single-phase and two-phase. In this research, the Colburn's equation and the Chen's equation to calculate the heat transfer coefficient for both single-phase and two-phase were used, while using the Saha-Zuber's equation and the Clark-Rohsenow equation to calculate the vapor mass flow rate at the sub-cool region. In addition, vapor mass flow rate of the steam generator with the axial economizer is compared with that of the current SG to evaluate the merit of the axial economizer. Moreover, we added fins and twisted tape to the current steam generator with the axial economizer as an approach to increase the vapor mass flow rate. Then, the rate of heat transfer and the temperature difference between the primary system and the secondary system were compared to inspect the economizer, fin, and twisted tape. As a result, vapor mass flow rate has increased about 3% by the combination of the external fin and twisted tape.

Comparison between FLACS explosion simulations and experiments conducted in a PWR Steam Generator casemate scale down with hydrogen gradients

Abstract In case of hypothetic severe accident in a Pressurized Water Reactor (PWR), interaction of the melted core with the cooling water can generate large amounts of hydrogen. Hydrogen can also be generated by oxidation of metals present in the corium recovery pool or in the basemat during the molten corium–concrete interaction phase. Then hydrogen is dispersed into the containment by convection loops arising essentially from condensation of steam released via the Reactor Cooling System (RCS) break or during corium–concrete interaction. Distribution of hydrogen can be more or less homogeneous, depending on mixing in the containment atmosphere. If considerable hydrogen stratification exists, then local concentration of hydrogen may become substantial, and may exceed the lower flammability limit. In case of ignition, the subsequent overpressure may adversely affect the containment building, internal walls and equipment integrities. Evaluation of such scenarios needs CFD codes. However, a thorough validation process is necessary before using such codes with a high level of confidence. The original vertical facility ENACCEF, which represents a scale down of a PWR Steam Generator casemate, has been built to fulfil this objective. Flame acceleration has been largely studied for a homogeneous hydrogen distribution in a containment volume ( Yang et al., 1991 , Dorofeev et al., 2001 ). However, very few data are available on the behaviour of a hydrogen/air flame in a nonuniform mixture ( Whitehouse et al., 1996 , Sochet et al., 1997 ). The present work aims to validate the commercial code FLACS on ENACCEF flame acceleration tests characterized by hydrogen gradients ( Cheikhravat et al., 2007 ). Previous simulations for different experiments have been made using the in-house code TONUS ( Riviere and Bleyer, 2004 ). Positive and negative gradients are considered and experimental data are compared to 3D calculations.

New method to assess tube support plate clogging phenomena in steam generators of nuclear power plants

Pressurized water reactor steam generators (PWR SGs) are large components whose main function is to cool the fission reactor by extracting the thermal power conveyed by the primary coolant, and thus to produce steam for the turbine generator. Tube support plate (TSP) clogging phenomena may occur in the SG when iron oxide particles carried in the secondary feedwater get unavoidably deposited inside the SG structure, and specifically in two main regions: on the U-tube outer walls and on the quatrefoil sections of the plates that support the U-tubes. They may reduce its cooling efficiency and impact its dynamic behaviour, thus leading to possible safety issues. Methods based on local inspections of the SG have been designed and are already in use. They produce estimators of the local TSP clogging ratio, during the yearly outage of the plant for refuelling and maintenance. A new method to assess a global estimator of the TSP clogging ratio is presented. This method is based on a 1D physical model of the SG that reproduces the complex dynamics of the two-phase flow phenomena inside the SG. The model was developed in the Modelica language. This model is used to compute response curves of the SG characteristics to a particular transient that challenges the dynamics of the SG, when affected by TSP clogging. The estimator is obtained by comparing the computed response curves to real response curves measured on-site. The method is still under validation. However, first results show that it is able to give global estimators that are consistent with local ones. This new method is expected to improve the monitoring of possible TSP clogging phenomena by producing quantitative estimators with better accuracy, with a quarterly periodicity while the plant is in operation.

Development and application of an efficient method for performing modal analysis of steam generator tubes in nuclear power plants

A typical pressurized water reactor (PWR) steam generator has approximately 10,000 tubes. These tubes have different geometries, supporting conditions, and different material properties due to the non-uniform temperature distribution throughout the steam generator. Even though some tubes may have the same geometry and boundary conditions, the non-uniform distribution of coolant densities adjacent to the tubes causes them to have different added mass effects and dynamic characteristics. Therefore, for a reliable design of the steam generator, a separate modal analysis for each tube is necessary to perform the FIV (flow-induced vibration) analysis. However, the modal analysis of a tube including the finite element modeling is cumbersome and takes lots of time. And when a commercial finite element code is used, interfacing the modal analysis result, such as natural frequencies and mode shapes, with the FIV analysis procedure requires an additional significant amount of time and can possibly incur inadvertent error due to the complexity of data processing. It is therefore impossible to perform the complete FIV analysis for ten thousands of tubes when designing or maintaining a steam generator although it is necessary. Rather, to verify the safe design against the FIV, only a couple of tubes are chosen based on engineering judgment or past experience. In this paper, a computer program, PIAT-MODE, was developed which is able to perform modal analysis of all tubes of a PWR steam generator in a very efficient way. The geometries and boundary conditions of every tube were incorporated into PIAT-MODE using appropriate mathematical formulae. Material property data including the added mass effect was also included in the program. Once a specific tube is selected, the program automatically constructs the finite element model and generates the modal data very quickly. Therefore, modal analysis can be performed for every single tube in a straight way. When PIAT-MODE is coupled with the FIV analysis program, it gives an amazing benefit, which makes the FIV analysis of all tubes possible within a limited time during the design or maintenance period. Using the developed program, the stability ratio regarding the fluid-elastic instability and the amplitude of vibration resulting from the turbulence flow excitation can be calculated for all tubes according the standard ASME Code, therefore, much more reliable design of the steam generator against the FIV related failures can be achieved. For an operating plant, there is a requirement that every single tube must be quantitatively checked whether wear would be more than 40% in thickness during the next operation period. If yes, the tube must be plugged to prevent severe failure including the tube wall penetration. In order to decide the tube plugging, which leads the plant performance degradation, the wear prediction is required. The wear prediction analysis needs wear data by ultrasonic testing and modal property data of tubes. PIAT-MODE will be a power tool enabling the wear prediction in a limited maintenance period by supplying modal data for all tubes.

Surface Analysis of Pressurized Water Reactor Steam Generator Tubing Specimens

Extended abstract of a paper presented at Microscopy and Microanalysis 2010 in Portland, Oregon, USA, August 1 – August 5, 2010.

De-agglomeration mechanisms of TiO2 aerosol agglomerates in PWR steam generator tube rupture conditions

A steam generator tube rupture (SGTR) in a pressurized water reactor (PWR) might be a major source of accidental release of radioactive aerosols into the environment during severe accident due to its potential to by-pass the reactor containment. In the ARTIST program, tests were carried out at flow conditions typical to SGTR events to determine the retention of dry aerosol particles inside a steam generator tube. The experiments with TiO2 agglomerates showed that for high velocities in the range of 100–350 m/s, the average particle size at the outlet of the tube was significantly smaller than at the inlet due to particle de-agglomeration. Earlier, particle de-agglomeration has not been considered significant in nuclear reactor severe accidents. However, the tests in ARTIST program have shown that there is a possibility that TiO2 aerosol particles de-agglomerate inside a tube and in the expansion zone after the tube exit under SGTR conditions. In this investigation, we measured TiO2 aerosol de-agglomeration in the tube with very high flow velocities with two different TiO2 aerosols. The de-agglomeration was determined by measuring the size of the agglomerates at the inlet and outlet of the test section. The test section was composed of tubes with three different lengths, 0.20, 2.0 and 4.0 m, followed by an expansion zone. The main results were: (i) the de-agglomerate process was relatively insensitive to the initial particle size distribution, (ii) the agglomerates were observed to de-agglomerate in all the tubes, and the resulting particle size distributions were similar for both TiO2 aerosols, (iii) at high flow rates, increasing the gas mass flow rate did not produce further de-agglomeration, and (iv) the agglomerates did not de-agglomerate to primary particles. Instead, after de-agglomeration the particles had a median outer diameter Dc = 1.1 μm. Based on analysis using computational fluid dynamics (CFDs), the de-agglomeration was caused by the turbulent shear stresses due to the fluid velocity difference across the agglomerates in the viscous subrange of turbulence. It has to be noted that the particles used in this investigation were TiO2 agglomerates, and not prototypical nuclear aerosols with significantly different characteristics. Therefore, the results of this investigation cannot be directly used to determine whether the nuclear aerosol particles may de-agglomerate in SGTR sequences. However, this investigation highlights the possibility of particle de-agglomeration under SGTR conditions, and identifies the mechanism of the de-agglomeration inside the broken tube and when the aerosol is discharged to an open space.

RELAP5-3D modeling of PWR steam generator condensation experiments at the Oregon State University APEX facility

This report describes modeling using RELAP5-3D of a series of six steam generator U-tube steam condensation (without non-condensable gas) tests conducted at the Oregon State University Advanced Plant Experiment Test Facility from 2005 through 2007. These tests were designed to evaluate steam condensation rates in a scaled pressurized water reactor steam generator at various primary and secondary side pressures and inlet steam mass flow rates. Comparisons between the experimental data and the RELAP5-3D model results are made to quantify the effectiveness of RELAP5-3D in handling steam condensation in U-tube steam generators. RELAP5-3D tends to over predict the condensation rate and heat transfer coefficient when compared against the experimental data when the code uses the laminar Nusselt correlation to determine the heat transfer coefficient. When RELAP5-3D results are used with the Shah correlation the comparison between the heat transfer coefficients is much improved.