Structural integrity evaluation method for overheating rupture of FBR steam generator tube

This paper describes a structural integrity evaluation method for a SG tube of FBR in case of sodium–water reaction and creep rupture tests to obtain the strength of the tube material. In the SG of FBR, if intermediate size of water/steam leak (1–2 kg s⁻¹) would occur from a tube, it could cause overheating rupture of the multiple tubes surrounding the initially failed tube due to generated sodium–water reaction heat. In the ultra-high temperature condition, the creep strength of the material is one of the dominant factors for failure behavior. Accordingly, we tried to apply the creep failure criterion for the overheating rupture of the SG tube. The creep rupture tests have been performed at ultra-high temperature conditions ranging from 1223.2 to 1323.2 K. The test material is ‘Mod .9Cr–1Mo steel’ which is one of the candidate materials for the tubes of the future SG of FBR. The test results have shown that tube rupture depends on the creep strength of the material; hence, instantaneous rupture does not occur even if the stress exceeds the design value of ultimate tensile strength. The test data have been suitably expressed using the Larson–Miller Parameter, and a structural integrity evaluation method based on the sum of the use-fraction associated with the creep damage has been proposed. Based on this method, the structural integrity of the tube in the sodium–water reaction flame has been evaluated. The results show that it is important to detect the initial leak of the tube within a short period and to reduce the steam pressure more rapidly by SG blowdown.     

Probabilistic modelling of steam generator tube pitting corrosion

Pitting corrosion is a serious form of degradation in steam generator (SG) tubing of some nuclear stations. The nature and extent of the pitting process is assessed through inspection programs, typically using various eddy current (EC) techniques, while the impact of pitting is minimized through deposit removal maintenance activities such as water lancing and chemical cleaning of SGs. This paper presents a probabilistic model of SG tube pitting corrosion that incorporates trends observed from a large EC inspection database from a nuclear generating station. The pitting occurrence process is modelled as a stochastic Poisson process and the pit size is treated as a random variable. The model is statistically calibrated with the available EC inspection data. The model is applied to estimate the probability of tube leakage, forced outage rate and the distribution of the number of tubes plugged per SG in a given operating interval. The proposed model is useful in optimizing strategies for the life-cycle management of SGs.     

Development of stabilizers for steam generator tube repair

Fluidelastic stability, turbulence-induced and vortex-induced vibration analysis of different types of stabilizers for repairing steam generator tubes are presented. The performances of the different designs are compared with that of a common basis — the virgin tube. It was found that in addition to permitting the stabilized tubes to remain in operation, the sleeve has the additional merit of being the best performer of all the designs. In addition, it is adaptable to remote installation.     

Detection of steam generator tube leaks in pressurized water reactors

This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in pressurized water reactors (PWRs), formed the basis of study for the last year of the project.Four tasks are addressed in this study of the detection of steam tube leaks.

1. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks.

2. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks.

3. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above.

4. (4) Assessing the need for diagnostic data processing and display.

Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation.     

Prediction of structural integrity of steam generator tubes under severe accident conditions

Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating and design-basis accident conditions are reviewed. These rate-independent flow stress models are inadequate for predicting failure of steam generator tubes under severe accident conditions because the temperature of the tubes during such accidents can reach as high as 800°C where creep effects become important. Therefore, a creep rupture model for predicting failure was developed and validated by tests on unflawed and flawed specimens containing axial and circumferential flaws and loaded by constant as well as ramped temperature and pressure loadings. Finally, tests were conducted using pressure and temperature histories that are calculated to occur during postulated severe accidents. In all cases, the creep rupture model predicted the failure temperature and time more accurately than the flow stress models.     

Corrosion aspects of Ni-Cr-Fe based and Ni-Cu based steam generator tube materials

This paper reviews corrosion related issues of Ni-Cr-Fe based (in a general sense) and Ni-Cu based steam generator tube materials for nuclear power plants those have been dealt with for last more than four decades along with some updated information on corrosion research. The materials include austenitic stainless steels (SSs), Alloy 600, Monel 400, Alloy 800 and Alloy 690. Compatibility related issues of these alloys are briefly discussed along with the alloy chemistry and microstructure. For austenitic SSs, stress corrosion cracking (SCC) behaviour in high temperature aqueous environments is discussed. For Alloy 600, intergranular cracking in high temperature water including hydrogen-induced intergranular cracking is highlighted along with the interactions of material in various environments. In case of Monel 400, intergranular corrosion and pitting corrosion at ambient temperature and SCC behaviour at elevated temperature are briefly described. For Alloy 800, the discussion covers SCC behaviour, surface characterization and microstructural aspects of pitting, whereas hydrogen-related issues are also highlighted for Alloy 690.     

Effective heat conductivity of fuel element bundles and steam generator tube bundles

Effective heat conductivity of rod and tube bundles is one of thermophysical properties necessary for calculation of thermo hydraulic characteristics of heat producing devices, heat exchange devices and steam generators. This report introduces results of mathematical modeling of effective heat conductivity of transversally anisotropic rod bundles in solid conductive medium. The considered bundles represented cylindrical rods fitted in corners of stretched and compressed in direction of heat transfer rectangular and triangular grids. The calculated results were compared to analytical solutions and previous numerical results.     

Anisotropic modelling of tube bundle flow during steam generator wet layup

Tube bundle flow can be considered as a porous medium flow and a fluid continuum can be established by introducing the porosity which is a ratio of fluid volume to total volume. Darcy's flow regime applies for the tube bundle flow of low Reynolds number during steam generator wet layup circulation. A general three-dimensional formulation appears as a steady-state heat conduction equation with source term and anisotropic conductivities. Solution to such an equation with appropriate boundary conditions can be obtained by any finite element computer program which solves anisotropic heat conduction problems. Capability of anisotropic modelling has been demonstrated by a sample problem of axisymmetric tube bundle flow with orthotropic hydraulic conductivities which are derived according to the existing empirical correlations for friction factors.     

Characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation

The Canadian Nuclear Standard CSA N285.4 requires the periodic metallurgical examination of removed ex-service steam generator tubes. This paper describes the practices used for the characterization and structural integrity tests of ex-service steam generator tubes at Ontario Power Generation (OPG). It shows that there is no degradation of mechanical properties of Monel 400 tubes after 7-18 effective full power years (EFPY) of operation and Incoloy 800 tubes after more than 10 EFPY of operation.     

Stress analysis and fatigue life prediction for a U-bend steam generator tube

An analysis is carried out to determine the stresses in a steam generator tube that failed by fatigue. Using data available for the failed tube and for failures in two similar steam generators, the magnitudes of the alternating and mean stresses produced during operation are estimated. The cause for the early failure is shown to be the high mean stress caused by denting of the tube in the location where it passed through the tube sheet.     

套管式直流蒸汽发生器动态特性仿真研究

套管式直流蒸汽发生器是一种采用双面传热的新型蒸汽发生器.在中心管和环管外侧与环形通道流体间热流密度相等的假设基础上,合理选择集中参数并应用可动边界的处理方法对套管式直流蒸汽发生器传热管进行了动态仿真.仿真结果与热工水力定性机理分析结果及相关的试验结果相符,从而验证了仿真方法是有效的.     

Root causes of intergranular attack in an operating nuclear steam generator tube

This paper reports the secondary side intergranular attack of an Alloy 600 tube, which was located within sludge piles in the hot-leg side of an operating nuclear steam generator. Carbide distribution along the grain boundaries and chromium depletion were analyzed using optical microscopy and transmission electron microscopy. Local crevice chemistry in contact with the defect was also assessed from the hideout return test data and oxide film analysis results using energy dispersive spectroscopy. The main causes of this defect are discussed based on the microstructure, local chemistry and operation temperature.     

The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet.     

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant.The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions.Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated.The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.     

Calculation of tube degradation induced by dryout instability in sodium-heated steam generators

Analytical methods are adapted and presented for the calculation of thermal fluctuations and thermal stresses experienced by a tube wall in the region of departure from nucleate boiling (DNB) in a sodium-heated steam generator. Calculated results are presented using parameter ranges and geometry adopted from the Atomics International reference design for the steam generators of the Clinch River Breeder Reactor Plant. The physical phenomenon was modeled by subjecting the tube wall and the oxide scale layer on the water side of the tube (treated as either a powdery substance or a dense protective film) to periodic water heat transfer coefficient fluctuations simulating the effect of DNB in that area of the tube. The stresses obtained were employed in estimating (where possible) the most detrimental contribution to the life expectancy of the steam generator tube based on fatigue of the tube wall and repeated exfoliation of the oxide layers.     

A new LMR steam generator free from SWR with a double tube bundle configuration

This paper describes the activities made at KAERI to develop an advanced liquid metal reactor (LMR) steam generator which is free from a sodium water reaction (SWR) to resolve the concern of the SWR possibility and improve the economic features of the LMR. The steam generator design houses two tube bundles that are functionally different and its tube bundle region is radially or vertically divided into two. The SG is equipped with hot and cold fluid tube bundles, a medium fluid and a pump. It prevents the occurrence of the sodium water reaction while sodium is still used as the coolant for the primary heat transport system. The feasibility of using the SG with a double tube bundle for an actual use in a LMR plant is evaluated by setting up the skeleton of the NSSS for various possible configurations of the SG tube bundles.Analysis was made for various types of the new steam generator with a double tube bundle. Since the heat transfer in the SG is made among three different fluids, it has unique heat transfer characteristics. The analysis showed the possibility of the occurrence of an undesirable reversed heat transfer not only in the integrated single-region bundle type but also in the integrated double-region bundle type. The performance analysis revealed practical performance characteristics for the various bundle configurations. Also the feasibility study for the various NSSS configurations with the new SG is described.     

Improvement of steam generator tube failure propagation analysis code LEAP for evaluation of overheating rupture

As a part of safety assessment or design of steam generators of sodium-cooled fast reactors, it is necessary to evaluate the water leak rate under sodium–water reaction accident. The computer code called LEAP-II calculating a design basis water leak rate during long-time event progress including self-wastage, target-wastage, wastage-type failure propagation, water leak detection, and water/steam blowdown was developed for the prototype fast reactor in the past studies. In this study, a numerical analysis method to predict occurrence of overheating tube rupture was constructed and integrated into this code to expand its application range. The newly constructed method consists of the elemental analysis models for temperature distribution formed by a reacting jet, water-side thermal hydraulics, heat transfer at the tube wall, temperature and stress of the tube, and failure of the tube. Applicability of the method was investigated through the numerical analysis of the experiment on water vapor discharging into liquid sodium pool under the actual condition of the steam generator. The numerical analysis demonstrated that the method could provide the appropriately conservative result on the overheating-rupture-type failure propagation.     

Potential steam generator tube rupture in the presence of severe accident thermal challenge and tube flaws due to foreign object wear

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with counter-current natural circulating high temperature gas in the hot leg and SG tubes. The considered SG tube flaws are caused by foreign object wear, which in recent years has emerged as a major inservice degradation mechanism for the new generation tubing materials. The first step develops the statistical distributions for the flaw frequency, size, and the flaw location with respect to the tube length and the tube's tubesheet position, based on data of hundreds of flaws reported in numerous SG inservice inspection reports. The next step performs thermal-hydraulic analysis using the MELCOR code and recent CFD findings to predict the thermal challenge to the degraded tubes and the tube-to-tube difference in thermal response at the SG entrance. The final step applies the creep rupture models in the Monte Carlo random walk to test the potential for the degraded SG to rupture before the surge line. The mean and range of the SG tube rupture probability can be applied to estimate large early release frequency in probabilistic safety assessment.     

Friction and wear of Inconel 690 and Inconel 600 for steam generator tube in room temperature water

In the steam generators of nuclear power plants, the flow of cooling water can cause the tubes to vibrate, resulting in fretting wear damage due to contacts between these tubes and their supports. The tubes are made of Inconel 690 and Inconel 600 and the supports are made of STS 304. In this paper, fretting wear tests in water were performed using the materials Inconel 690 and Inconel 600 in contact with STS 304. Fretting tests using a cross-cylinder type set up were conducted under various vibrating amplitudes and applied normal loads in order to measure friction forces and wear volumes. Also, conventional sliding tests using a pin-on-disk type set up were carried out to compare these test results.In the fretting tests, friction force was found to be strongly dependent on normal load and vibrating amplitude. Coefficients of friction decreased with an increase in the normal load and a decrease in the vibrating amplitude applied. Also, the wear of Inconel 600 and Inconel 690 was predicted using a work rate model. Depending on the normal load and vibrating amplitude applied, distinctively different wear mechanisms and often drastically different wear rates occurred. It was found that the fretting wear coefficients for Inconel 600 and Inconel 690 were 9.3×10⁻¹⁵ and 16.2×10⁻¹⁵ Pa⁻¹, respectively. This study shows that Inconel 690 can result in lesser friction forces and exhibits less wear resistance than Inconel 600 in room temperature water.