Feedwater Nozzle Research Articles

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.

Influence on structure and fracture mechanics evaluations of a BWR feedwater nozzle following weld overlay repair

Abstract This present paper summarizes applying weld overlay repair on the feedwater nozzle in a boiling water reactor, and evaluates the influence on the structure and fracture mechanics of the component. The requirements for utilizing a full structural weld overlay, including the weld overlay design, effects of welding residual stresses (WRSs), impacts of weld shrinkage, prediction of potential flaw growth, and influence on ASME Code Section III design evaluations, are further defined in ASME Code Case N-504–2. In order to confirm the effectiveness of the weld overlay design of the nozzle, the WRSs following overlay installation were determined using engineering simulation software. The impacts of weld shrinkage from overlay welding are also addressed. The shrinkage stresses were evaluated via a numerical model of the attached piping system to ensure the design margins of the pipe structure. Since the weld overlay technique is well known to be effective in mitigating crack initiation or cracking to leakage, a fatigue crack growth prediction under ASME Code Section XI must be conducted to address the fatigue qualification of the affected component. The ASME Code Section III construction qualification of the whole nozzle region must consider the applicable thermal transient stresses, structural discontinuities, and bimetallic effects that are caused by the weld overlay. This investigation proves that the current stress analyses of the nozzle and the pipe structure are not significantly affected by the added overlay mass, so the structural integrity of the component can be ensured, meeting the requirements of ASME Code Case N-504–2.

A Simplified Analytical Model for Estimation of the Initial Waterhammer Pressure and Force on a Circular Tube

Abstract A simplified analytical model is developed for estimating the initial waterhammer pressure and force on a circular tube when a plane decompression wave, parallel to its axis, passes over it. The simplified analysis superimposes the solutions for a plane pressure wave traveling at sound speed in the surrounding water, and two-dimensional solution to the wave equation in cylindrical coordinates, with a nonflow boundary condition across the tube boundary at a fixed radius. The analytical method is compared to the computational fluid dynamics (CFD) approach by applying to predict the initial waterhammer pressure and force on the closest tube to the feedwater nozzle of a geometrically simplified nuclear steam generator (SG) analysis model, caused by a feedwater pipe break (FWPB). As the result, it is found that the simplified analytical model, while not matching results of the CFD calculations with precise accuracy, does confirm the nature of the waterhammer impact pressure loads on the SG tubes.

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.

Mathematical model for multivariate nonlinear prediction of SMD of X-type swirl pressure nozzles

X-type swirl pressure nozzle is a new type of nozzle with the spin core structure. Due to its superior atomization characteristics, it has been widely applied in the field of spraying dust suppression in recent years. Sauter mean diameter (hereinafter referred to as SMD) is an important indicator to evaluate the atomization characteristics and dust suppression performance of nozzles. In order to build the SMD prediction model for this type of nozzles, based on the independently designed spray experiment platform, the interaction law between the nozzle SMD and the four influence factors is obtained through the orthogonal design method. On this basis, a mathematical model for predicting the nozzle SMD is built using the multivariate nonlinear regression method. The results show that SMD increases with the increase of axial distance, absolute radial distance value and nozzle diameter, but decreases with the increase of feed water pressure. The primary and secondary sequence of the four factors affecting SMD is: feed water pressure > axial distance > nozzle diameter > radial distance. The fitting values of the multivariate nonlinear regression model are basically the same with the experimental values, with the relative error of less than 3.50% and the average relative error of only 1.35%, which can be used for the theoretical calculation of the SMD of the X-type swirl pressure nozzles. The multivariate nonlinear prediction model built in this study can provide guidance for the design and selection of such spraying dust suppression schemes as nozzle outlet diameter, feed water pressure and nozzle layout, etc. for the field engineering application of such nozzles.

Residual stresses due to cladding, buttering and dissimilar welding of the main feed water nozzle in a power plant reactor

A 100 MW AP1000 Reactor was considered and welding processes were employed in joining of the main feed water nozzle, in primary circuit of nuclear reactors were simulated by means of finite element method. Distribution and magnitude of the residual stresses created by thermal processes including cladding inside the nozzle, buttering prior to dissimilar welding, heat treatment and dissimilar welding of the safe-end austenitic stainless steel to the nozzle ferritic steel were investigated. Norton-Bailey creep model was used for heat treatment simulations. Experimental results obtained in previous studies and residual stress measurements were applied for verification of the developed model. It was revealed that residual stresses after cladding were approximately equal to yield strength of the materials. Also, buttering led to reduction of residual stresses in the nozzle free end. Heat treatment decreased residual stresses in the base metal significantly; whereas, it had minor effects on those in the cladding layer. Besides, circumferential residual stresses were invariably tensile at inner and outer surfaces of the joint. Further discussions were provided to explain the obtained result.

Transient Hydraulic Response of a Pressurized Water Reactor Steam Generator to a Feedwater Line Break Using the Nonflashing Liquid Flow Model

In this study, a computational fluid dynamics (CFD) analysis of the transient flow field inside the secondary side of a nuclear reactor steam generator (SG) during blowdown following a feedwater line break (FWLB) accident is performed to evaluate the transient hydraulic loading (pressure) on the SG internals and tubes. The nonflashing liquid flow is assumed for a conservative prediction of the transient blowdown loading. The CFD analysis results are illustrated in terms of the transient velocity and pressure disturbances at some selected monitoring points inside the SG secondary side and compared with those predictions obtained from the existing simple analytical model to examine the physical validity of the CFD analysis model. As a result, the existing simple analytical model cannot yield the transient velocity and pressure disturbances and results in underestimation during blowdown as compared to the CFD calculations. Based on the present CFD analysis results, it is seen that an FWLB may result in excessive disastrous transient hydraulic loading on the SG internal structures and tubes near the feedwater inlet nozzle due to the significant pressure changes (pressure wave with very high amplitude) and abruptly increased velocity of water near the feedwater nozzle.

Experimental study on the effect of feed water nozzles on non-equilibrium temperature difference and flash evaporation in a single-stage evaporator and an investigation of effect of process parameters on the liquid flashing in a LTTD desalination process

Low temperature thermal desalination (LTTD) process involves flash evaporation of a seawater at 28–29°C in a single-stage evaporator maintained at a vacuum of around 25–27 m bar (abs). The seawater was splashed inside an evaporator through a 0.1 m diameter upward facing nozzles of around 24 nos arranged evenly throughout the evaporator and generated vapour was condensed in the shell and tube condenser using cooling water available at 12–13°C sucked from the deep sea through a long HDPE pipe. The main objective of this study was to find out the effect of geometry of the upward facing nozzles on the flash evaporation rate as well as on the non-equilibrium temperature difference (NETD) of the flashing process. Two different spout nozzle geometries with 0.37–0.87 m height were used in the experiment. The study indicated that the flashing rate increased by 0.9% (average) and and the NETD (Two – Tsat) decreased by 0.7°C (average), respectively, when nozzle height was increased by 0.87 m. Mechanism that controlled these two factors were identified and discussed in this paper. Drawbacks of 0.37 m nozzle geometry was also discussed. It was reported in literature that 4% yield ratio was obtained for a nozzle injection pressure of 1 bar for a similar desalination process. But in the present study, a maximum of 1.12% yield ratio was obtained with a nozzle injection pressure of around 0.17 bar. In this work,the effect of the process parameters on the liquid flashing in a LTTD desalination process was investigated and discussed. In order to fine tune the evaporator design for the future LTTD plants, the experimental results of flash evaporation were compared with two mathematical models obtained from the literature. While comparing the results, it was observed that the model which used actual heat (Twi – Two) made a good agreement with the experimental data compared to the other model that used superheat (Twi – Tsat). From the experimental study, it was observed that the NETD (i.e. thermal loss) measured was found to be higher than the predicted value. The reason that caused the difference in the NETD value was investigated and discussed. Suitable suggestions to reduce these NETD in the flashing process were also presented.

Welding residual stress analysis for weld overlay on a BWR feedwater nozzle

This paper is to perform the welding residual stress analysis for weld overlay on a dissimilar metal weld of feedwater nozzle at a Taiwan domestic boiling water reactor. To mitigate the degradation of nozzle weld, a weld overlay repair was implemented for assuring its structural integrity. Based upon ASME Code Case N-504, the weld overlay repair has to be analyzed to improve residual stress at the inside surface of original weld. Using ANSYS finite element analysis software, an axis-symmetric model is specifically developed for the feedwater nozzle in this work. Field welding procedures and detailed welding simulations are both taken into account to predict welding residual stress during weld overlay repair. The residual compressive stress can thus be obtained at the inside surface of original weld. The present numerical models are also compared with those in the public technical reports to show their accuracy. The computed welding compressive stress would be helpful for the resistance of degraded feedwater nozzle with overlay to stress corrosion cracking and further crack growth during subsequent operation.

CFD Analysis of Thermally Stratified Flow and Conjugate Heat Transfer in a PWR Pressurizer Surgeline

Temperature gradients in the thermally stratified fluid flowing through a pipe may cause undesirable excessive thermal stresses at the pipe wall in the axial, circumferential, and radial directions, which can eventually lead to damages such as deformation, support failure, thermal fatigue, cracking, etc., to the piping systems. Several nuclear power plants have so far experienced such unwelcome mechanical damages to the pressurizer surgeline, feedwater nozzle, high pressure safety injection lines, or residual heat removal lines at a pressurized water reactor (PWR). In this regard, determining with accuracy the transient temperature distributions in the wall of a piping system subjected to internally thermal stratification is the essential prerequisite for the assessment of the structural integrity of such a piping system. In this study, to realistically predict the transient temperature distributions in the wall of an actual PWR pressurizer surgeline with a complex geometry of three-dimensionally bent piping, three-dimensional transient computational fluid dynamics (CFD) calculations involving the conjugate heat transfer analysis are performed for the PWR pressurizer surgeline subjected to either out- or in-surge flows using a commercial CFD code. In addition, the wall temperature distributions obtained by taking into account the existence of wall thickness are compared with those by neglecting it to identify some requirements for a realistic and conservative thermal analysis from a safety viewpoint.

Welding overlay analysis of dissimilar metal weld cracking of feedwater nozzle

Inspection of the weld between the feedwater nozzle and the safe end at one Taiwan BWR showed axial indications in the Alloy 182 weld. The indication was sufficiently deep that continued operation could not be justified considering the crack growth for one cycle. A weld overlay was decided to implement for restoring the structural margin. This study reviews the cracking cases of feedwater nozzle welds in other nuclear plants, and reports the lesson learned in the engineering project of this weld overlay repair. The overlay design, the FCG calculation and the stress analysis by FEM are presented to confirm that the Code Case structural margins are met. The evaluations of the effect of weld shrinkage on the attached feedwater piping are also included. A number of challenges encountered in the engineering and analysis period are proposed for future study.

Reconciliation of fatigue usage factors for class-1 components considering plant operation history

The evaluation of metal fatigue for the period of extended operation requires a time-limited aging analysis (TLAA) and is an important task for Nuclear Power Station. Both cumulative usage factors of components and the actual numbers of design transient cycles have to be reviewed. The factors of environment are also imposed on metal fatigue by NRC as a multiplier. In reviewing the cycles of a generic BWR pressure vessel design transients, the number of cycles of some design transients have no sufficient margin for current license basis even for period of extended operation. The operating margins of technical specification, operating cycles of design transients and cumulative usage factors of components, are subjected to severe challenge under the review of license renewal. The objective of this study aims to relieve such limits by using design transients reconciliation based on existing plant operating history. The predicted fatigue usage factors of Feed water nozzle of the reactor vessel are shown as an example using the adjusting cycles of design transients.

On the use of a monitoring system for fatigue usage calculations

Fatigue is a limiting factor in structural component life. Design calculations are performed according to conservative methodologies that shorten component life. Fatigue Monitoring Systems are an alternate that register the component service conditions. This work studies the application of one of those systems to determine the fatigue usage of the safe end of the feedwater system nozzle of the Santa Marı́a de Garoña nuclear power plant. The transients recorded are compared to the ones evaluated in the design specifications and the fatigue usage is compared with the one provided in the stress report. The main conclusion is that the fatigue monitoring system calculates a fatigue usage that is significantly lower than the design one, this is due to two main reasons: actual transients are less severe than design ones, and the number of occurrences of actual transients is lower than specified.

A novel approach to coupling the fluid and structural analysis of a boiler nozzle

Boiler feedwater nozzles are often protected from steep thermal gradients by a fluid annulus that separates the feedwater pipe from the inside surface of the boiler. Experimental results have confirmed that hot water from the boiler is transported along the length of the annulus by a thermosyphoning flow, which is driven by natural convection effects within the annulus. This represents a highly complex flow field consisting of both laminar and turbulent flow combined with both global and local natural convection effects. This paper describes a quasi-steady thermofluid analysis of a feedwater nozzle operating at different transient conditions, which was used to predict the variation of peak stresses within the nozzle. This method represents a fast, effective approach for obtaining justifiable results for a large number of different operating transients. A steady-state FLUENT model was constructed and validated against measured data from an experiment on a similar nozzle design and the results demonstrated good agreement with the experimental data. The FLUENT model of the boiler feedwater nozzle was solved at a fixed operating condition to provide a three-dimensional profile of the heat transfer coefficients and bulk temperatures on the inner and outer surfaces of the fluid annulus. These data were imported into the finite-element analysis (FEA) code as a surface boundary condition. The FEA model was then used to predict the nozzle temperatures and stresses during the operating transient as part of a comprehensive fatigue and fracture assessment. This work has demonstrated that the use of advanced computational fluid clynamic (CFD) methods can form an integral part of the life assessment process. As a result, the boiler operator was able to significantly extend the predicted service life of the component.

Metallurgical Evaluation of a Feedwater Nozzle to Safe-End Weld

Weld cracks in safety class systems are a serious concern, because these systems are part of the primary barrier providing containment of radioactive coolant. Loss of weld integrity yields leaks, or, under catastrophic failure, can be the basis for a severe loss of coolant accident. A circumferential indication was found by ultrasonic examination (UT) in the N4A-2 inlet feedwater nozzle to safe-end weld during the second refueling outage of River Bend Station Unit 1 in March 1989. The indication, approximately 15cm (6in) long with a reported maximum depth of 0.5cm (0.1in), was located in the Alloy 182 weld butter on the safe-end side of the weld. (The safe-end base metal was ASME SA 508 Class 1 carbon steel.) The reported characteristics of the UT indication were indicative of intergranular stress corrosion cracking. This indication was reexamined during the second and third fuel cycles in March 1990 and September 1991, respectively, and during the third refuel outage in November 1990. Crack growth was reported during each examination. The safe-end was replaced during the fourth refueling outage in the summer of 1992. The US Nuclear Regulatory Commission (NRC) subsequently contracted with Brookhaven National laboratory (BNL) to conduct a confirmatory investigation to establishmore » the failure mode and determine the root causes of cracking in the safe-end weld.« less

Detection and repair of a crack in a BWR feed water nozzle safe end weld : Pers-Anderson, E.-B. Proceedings of the 10th International Conference on NDE in the Nuclear and Pressure Vessel Industries, Glasgow (Scotland), 11–14 Jun. 1990. pp. 271–275. Edited by M.J. Whittle, J.E. Doherty and K. Iida. ASM International, (1990)

Detection and repair of a crack in a BWR feed water nozzle safe end weld : Pers-Anderson, E.-B. Proceedings of the 10th International Conference on NDE in the Nuclear and Pressure Vessel Industries, Glasgow (Scotland), 11–14 Jun. 1990. pp. 271–275. Edited by M.J. Whittle, J.E. Doherty and K. Iida. ASM International, (1990)

Framatome Antistratification Device for Steam Generator Feedwater Nozzle: Effectiveness, Qualification Test, and On-Site Measurements

Thermal stratification may be responsible for the development of severe cracks in the feedwater line of steam generators in pressurized water reactors, leading Framatome to carry out an experimental program especially dedicated to this problem.This developmental program analyzes the mechanisms and the driving parameters of thermal stratification. It develops and qualifies an antistratification device to prevent thermal stratification at the steam generator feedwater nozzle location. The program also compares on-site measurements with mockup results.The outcome of the experimental program is a qualified helical antistratification device to be installed in the thermal sleeve of the steam generator feedwater nozzle. As required, this device significantly reduces thermal stratification effects in the feedwater system, even in very low feedwater flow conditions.

Operating data monitoring and fatigue evaluation systems and findings for boiling water reactors in Japan

Abstract The reactor pressure vessel (RPV) is one of the most critical components of a boiling water reactor (BWR) when utilities think about plant life extension (PLEX). Design stress analysis sometimes reports very high fatigue usage factors for such portions of RPVs as stud bolt, feedwater nozzle and support skirt. In order to evaluate design margin and to eliminate excessive conservatism in this design analysis to pave the way for PLEX, Japanese BWR utilities jointly with BWR manufacturers in Japan established a programme (1) to acquire plant operational data on line for specific parameters used in stress analysis, (2) to evaluate margin in the design using measured plant data best estimate boundary conditions for stress analysis, and (3) to establish a simplified fatigue analysis method for BWR RPV. A plant data acquisition system, named OPEDAS, has been developed and installed in Tokyo Electric Power Company's 1100 MWe BWR at its Kashiwazaki-Kariwa nuclear power plant. Best estimate stress analysis using measured in-plant data has been carried out and the results show considerable margin in fatigue usage factor over the design. A simplified fatigue analysis method using in-plant data has been developed with the Green's function, although some limitations have been identified for its use.

Application of advanced ultrasonic test techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles

As a result of feedwater nozzle cracking observed in Boiling Water Reactor (BWR) plants, several design modifications were implemented to eliminate the thermal cycling that led to crack initiation. BWR plants with these design changes have successfully operated for over ten years without any recurrence of cracking. To provide further assurance of this, the U.S. Nuclear Regulatory Commission (NRC) issued NUREG-0619, which established periodic ultrasonic testing (UT) and liquid penetration testing (PT) requirements. While these inspections are useful in confirming structural integrity, they are time consuming and can lead to significant radiation exposure to plant personnel. In particular, the PT requirement poses problems since it is difficult to perform the inspections with the feedwater sparger in place and also leads to additional personnel exposure. Clearly, an inspection and monitoring program that eliminates the PT examination and still verifies the absence of surface cracking would be extremely valuable in limiting costs as well as radiation exposure. This paper describes a program involving the application of advanced UT techniques coupled with fatigue and leakage monitoring to assure integrity of BWR feedwater nozzles. The inspection methods include: (1) scanning with optimized transducers and techniques from the outside vessel wall surface to inspect the nozzle inner radius region, and (2) scanning from the nozzle forging outside-diameter to inspect the nozzle bore region. Methods of analyzing the data using 3-D graphics displays have been developed that show crack location, size, and maximum depth of penetration into the nozzle inner surface. These techniques have been developed to the point where they are now considered a reliable alternative to the liquid penetrant requirements of NUREG-0619. An important supplement to the UT program is the use of automated fatigue, leakage and crack growth monitoring to verify the absence of cracking. This approach provides for a continuous assessment of the integrity of the nozzle structure by tracking the actual fatigue duty, measuring thermal sleeve bypass leakage and performing crack growth predictions based on actual thermal duty. Collectively, the monitoring and inspection program provides technically sound assurance of nozzle integrity and a firm basis for plant operational planning.

Use of On-Line Fatigue Monitoring of Nuclear Reactor Components as a Tool for Plant Life Extension

The application of an on-line fatigue monitoring system for tracking fatigue usage in operating power plants is described. The system, like several others which have been developed, uses the influence function approach, operates on a microcomputer, and determines component stresses using temperature, pressure, and flow rate data that are typically available from plant computers. Using plant-unique influence functions developed specifically for each component location, the system calculates stresses as a function of time and computes the fatigue usage. Stress values are calculated at time intervals defined by the user and the fatigue values are saved on files for use at a later time. The application of the GE Fatigue Monitoring System (GEFMS) to calculate fatigue usage in the feedwater nozzle of a GE Boiling Water Reactor is described in this paper. Stress predictions using GEFMS for a sample temperature transient show excellent agreement with results from finite element thermal and stress analysis performed on a mainframe computer. Fatigue usage calculations for a simulated 24-hr temperature record confirm that the system provides accurate results at a cost that is significantly lower than similar analysis done on mainframe computer systems. The system, which has been installed in a boiling water reactor plant, provides the technical basis to evaluate actual reactor conditions and justify plant life extension.