Steam generator tube vibration tests

The buildup of magnetite in the steam generators of some pressurized water reactors had led operators to propose chemical cleaning to remove this product. In some cases, the volume of magnetite formed by the corrosion of the carbon steel has been sufficient to cause “denting” or reduction of the outer diameter of the tubes where they pass through the support plates. The US Nuclear Regulatory Commission has expressed concern that the resulting increased clearances may allow an increased level of flow-induced vibrations in chemically cleaned steam generators that could lead to high tube wear rates and unacceptable levels of tube failure.A project has been undertaken by Pacific Northwest Laboratory to address the effects of increased tube/tube-support clearances and to provide the NRC with criteria to evaluate licensees' specific proposals for chemical cleaning of steam generators. The first phase of the project consists of flow tests to establish the forcing boundary conditions, using clearances representing various conditions following chemical cleaning. The second phase consists of accelerated wear tests to determine the potential wear rates possible based on the vibrations characterized in the flow tests.     

Aerosol retention in the flooded steam generator bundle during SGTR

A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out.To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated.Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet-bundle interactions close to the tube break. Droplets were entrained from the water surface with high gas flow rates carrying aerosol particles with them. However, compared to particle retention in the water close to the tube break, the effect of droplet entrainment on particle transport was small.     

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.     

Effects of tube rupture modeling and the parameters on the analysis of multiple steam generator tube rupture event progression in APR1400

A multiple steam generator tube rupture (MSGTR) event in APR1400 has been investigated using the best estimate thermal hydraulic system code, MARS1.4. The effects of the parameters such as the number of ruptured tubes, rupture location, affected steam generator on the analysis of the MSGTR event in APR1400 are examined. In particular, tube rupture modeling methods, single tube modeling (STM) and double tube modeling (DTM), are compared. The APR1400 is found to have the capability of allowing more than 30 min to operators for the MSGTR event of five tubes. The effects of rupture location on the MSSV lift time is not significant in the case of STM, but the MSSV lift time for tube-top rupture is found to be 25.3% larger than that for rupture at the hot-leg side tube sheet in the case of DTM. The MSSV lift time for the cases that both steam generators are affected (4C5x, 4C23x) are found to be larger than that of the single steam generator cases (4A5x, 4B5x) due to a bifurcation of the primary leak flow. The discharge coefficient of Cd is found to affect the MSSV lift time only for a smaller value of 0.5. It is found that the most dominant parameter governing the MSSV lift time is the leak flow rate. Whether any modeling method is used, it gives the similar MSSV lift time if the leak flow rate is close, except in the case where both steam generators are affected. Therefore, the system performance and the MSSV lift time of the APR1400 are strongly dependent on the break flow model used in the best estimate system code.     

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.     

Evaluation of a steam generator tube rupture accident in an accelerator driven system with lead cooling

A postulated steam generator tube rupture (SGTR) accident in a lead cooled accelerator driven transmuter (ADT) is investigated. The design of the ADT without intermediate loops bears the risk of water/steam blasting into the primary coolant. As a consequence a nuclear power excursion could be triggered by steam ingress into the ADT core which has a significant positive void worth. A thermal coolant–coolant interaction (CCI) might initiate a local core voiding too and additionally could lead to sloshing of the lead pool with mechanical impact of the heavy liquid on structures. The steam formation will also lead to a pressurization of the cover gas. The problems related to an SGTR are identified and investigated with the SIMMER-III accident code.     

The multiple pressure tube rupture (MPTR) issue in RBMK safety technology

The RBMK core is constituted by more than one-thousand pressurized channels housed into stacked graphite blocks and connected at the bottom and at the top by small diameter (D) and long length (L) pipes (less than 0.01 and more than 10 m, respectively) that end up into headers and drum separators. Control valves are installed in the bottom lines. Due to the large L/D value and to the presence of valves and other geometric discontinuities along the lines connecting with the pressure channels, the Fuel Channel Blockage (FCB) event is possible and already occurred in two documented NPP events. Pressure tube rupture occurred in a third NPP event not originated by FCB. Previous investigations, have shown the relevance of these events for the safety technology, and the availability of proper computational technique for the analysis, see the first and the third companion paper in this journal issue, respectively.The occurrence of the FCB event remains undetected for a few tens of seconds because of the lack of full monitoring for the individual channels, fourth companion paper in this journal issue. Therefore, fission power continues to be produced in the absence of cooling. This brings in subsequent times to fuel rod overheating, pressure tube failure, damage of the neighbouring graphite brick and ejection of damaged fuel.Following the pressure tube rupture, reactor cavity pressurization, radioactivity release into the same area and change of fluid properties occur that allow the detection of the event and cause the reactor scram at a time of a few tens of seconds depending upon the channel working conditions and the severity of the blockage.Notwithstanding the [delayed] scram and the full capability of the reactor designed safety features to keep cooled the core, the multiple pressure tube rupture (MPTR) issue is raised. The question to be answered is whether the ‘explosion’ of the blocked pressure tube damages not only the neighbour graphite bricks but propagates to other channels causing the potential for several channel failure.In order to address the MPTR issue fuel channel thermal-hydraulics and three-dimensional (3D) neutron kinetics analyses have been performed, as well structural mechanics calculations for the graphite bricks and rings (graphite rings surround the pressure tube to accommodate for thermal and radiation induced expansions).The bases for the analysis and the results of the study are presented. The conclusion, not reported within a licensing based format, is that the MPTR consequences are not expected to be relevant for the safety of the RBMK installations. This is supported by the analysis of experiments performed at the TKR facility available at the EREC research Centre near Moscow.     

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 D_c = 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.     

Level 2 analysis to estimate LERF risk from a thermally induced rupture of a steam generator tube defect

This paper outlines the Level 2 portion of a methodology for determining the incremental induced steam generator tube rupture large early release fraction caused by an actual through-wall defect. This defect was responsible for the minor steam generator tube leak that occurred in September 2002 at the Comanche Peak Steam Electric Station Unit 1. In order to quantify the performance of the defect over the operating cycle, a range of defect lengths were input to the PROBFAIL computer code [Kenton, M., 2001. PROBFAIL: A Computer Code for Evaluating the Likelihood of Steam Generator Tube Rupture in Severe Nuclear Power Plant Accidents, CREARE TM-2138], using appropriate boundary conditions derived from MAAP4 [Henry, R., et al., May 1994. MAAP4—Modular Accident Analysis Program for LWR Power Plants, Computer Code Manual, EPRI Research Project 3131-02] runs. From the analysis of the calculated times of burst for each assumed defect length, the minimum through-wall defect length necessary for tube burst to occur prior to hot leg or surge line creep rupture was calculated. The probability that the defect would actually have this length was then estimated by determining the fraction of the cycle for which the defect would be at least that long. The methodology development and implementation relied on MAAP4 runs, which are discussed extensively in connection with their role in: (1) guiding the construction of the accident progression event tree, (2) generating relevant information for probability assignments in the various underlying fault trees and (3) obtaining boundary conditions of pressure and temperature for use in PROBFAIL. The overall increment in LERF due to the existence of the defect was calculated to be approximately 4E−08.     

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feed-water. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data.     

CFD analysis of a CiADS fuel assembly during the steam generator tube rupture accident based on the LBEsteamEulerFoam

Steam generator tube rupture(SGTR)accident is an important scenario needed to be considered in the safety analysis of lead-based fast reactors.When the steam generator tube breaks close to the main pump,water vapor will enter the reactor core,resulting in a two-phase flow of heavy liquid metal and water vapor in fuel assemblies.The thermal-hydraulic prob-lems caused by the SGTR accident may seriously threaten reactor core's safety performance.In this paper,the open-source CFD calculation software OpenFOAM was used to encapsulate the improved Euler method into the self-developed solver LBEsteamEulerFoam.By changing different heating boundary conditions and inlet coolant types,the two-phase flow in the fuel assembly with different inlet gas content was simulated under various accident conditions.The calculation results show that the water vapor may accumulate in edge and corner channels.With the increase in inlet water vapor content,outlet coolant velocity increases gradually.When the inlet water vapor content is more than 15％,the outlet coolant temperature rises sharply with strong temperature fluctuation.When the inlet water vapor content is in the range of 5-20％,the upper part of the fuel assembly will gradually accumulate to form large bubbles.Compared with the VOF method,Euler method has higher computational efficiency.However,Euler method may cause an underestimation of the void fraction,so it still needs to be calibrated with future experimental data of the two-phase flow in fuel assembly.     

传热管破裂位置及根数对SGTR事故进程的影响

以一体化反应堆为研究对象,应用RELAP5/MOD3.4程序对套管式直流蒸汽发生器发生传热管破裂事故时,影响事故进程的一些因素进行了分析,其中包括破口在传热管轴向高度不同断裂位置,以及同时断裂多根传热管等。分析结果表明:不同断裂位置处的SGTR事故,其系统响应大致相同;不同破裂面积的SGTR事故,其破口处临界喷放流量与破口面积有着密切的联系。但总体来看,无论直流蒸汽发生器发生何种形式的SGTR,其一回路冷却剂通过破口处向二回路侧泄漏的积分流量大致相同,而且这个积分流量决定了一体化反应堆的瞬态响应。     

Steam generator mock-up for assessment of inservice inspection technology

A steam generator mock-up has been assembled for round-robin (RR) studies of the effectiveness of currently practiced inservice inspection technology for detection of current-day flaws. The mock-up will also be used to evaluate emerging inspection technologies. The 3.66 m (12 ft)-tall mock-up contains 400 tube openings, each consisting of nine test sections that can be used to simulate current-day field-induced flaws and artifacts. Included in the mock-up are simulations of tube support plate (TSP) intersections and the tube sheet (TS). Cracks are present at the TSP, TS, and in the free span sections of the mock-up. For initial evaluation of the RR results, various eddy current methods, as well as multivariate models for data analysis techniques are being used to estimate the depth and length of defects in the mock-up. To ensure that the RR is carried out with procedures as close as possible to those implemented in the field, input was obtained from industry experts on the protocol and procedures to be used for the exercise. One initial assembly of the mock-up with a limited number of flaws and artifact has been completed and tested. A second completed configuration with additional flaw and artifacts simulations will be used for the round robin.     

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.     

Steam generator transient studies using a simplified two-fluid computer code

A simplified two-fluid computer code has been used to simulate reactor-side (or primary-side) transients in a PWR steam generator. The disturbances are modelled as ramp inputs for pressure, internal energy and mass flow-rate for the primary fluid. The CPU time for a transient duration of 4 s is approx. 10 min on a DEC-1090 computer system. The results are thermodynamically consistent and encouraging for further studies.     

Overview of steam generator tube degradation and integrity issues

The degradation of steam generator tubes in pressurized water nuclear reactors, and, in particular, the stress corrosion cracking (SCC) of Alloy 600 tubes, continues to be a serious problem. Primary water SCC is commonly observed at the roll transition zone (RTZ), at U-bends and tube denting locations, and occasionally in plugs and sleeves. Outer-diameter SCC (ODSCC) and intergranular attack (IGA) commonly occur near tube support plate (TSP) crevices, near the tube sheet in crevices, or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of axial and circumferential cracking at the RTZ on both the primary and secondary sides. Outer-diameter stress corrosion cracking in TSP crevices, commonly consisting of segmented axial cracks with interspersed uncracked ligaments, is also becoming more common. Despite recent advances in inservice inspection (ISI) technology, a clear need still exists for quantifying and improving the reliability of ISI methods with respect to the probability of detection of the various types of flaws and their accurate sizing. These improvements are necessary to permit an accurate assessment of the consequences of leaving degraded tubes in service over the next reactor operating cycle. Degradation modes such as circumferential cracking, intergranular attack, and ODSCC at the TSP have affected a large number of tubes. New regulatory guidance is being developed that requires the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes will perform the required safety function over the next operating cycle. In formulating new guidance for the implementation of alternate repair criteria, the U.S. Nuclear Regulatory Commission is also evaluating the contribution to overall plant risk from severe accidents. Preliminary analyses are being performed for a postulated severe-accident scenario that involves station blackout and loss of primary feedwater, in order to determine the probability of failure for degraded tubes.     

Stress corrosion cracking susceptibility of steam generator tube materials in AVT (all volatile treatment) chemistry contaminated with lead

Alloy 600 steam generator tubing has shown a high susceptibility to stress corrosion degradation at the operation conditions of pressurized water reactors. Several contaminants, such as lead, have been postulated as being responsible for producing the secondary side stress corrosion cracking that has occurred mainly at the location where these contaminants can concentrate. An extensive experimental work has been carried out in order to better understand the effects of lead on the stress corrosion cracking susceptibility of steam generator tube materials, namely Alloys 600, 690 and 800. This paper presents the experimental work conducted with a view to determining the influence of lead oxide concentration in AVT (all volatile treatment) conditions on the stress corrosion resistance of nickel alloys used in the fabrication of steam generator tubing.     

Preliminary steps toward assessing aerosol retention in the break stage of a dry steam generator during severe accident SGTR sequences

Severe accidents SGTR sequences are identified as major contributors to risk of PWRs. Their relevance lies in the potential radioactive release from reactor coolant system to the environment. Lack of knowledge on the source term attenuation capability of the steam generator has avoided its consideration in probabilistic safety studies and severe accident management guidelines. This paper describes a research program presently under way on the aerosol retention on the tubes surrounding the breach within the secondary side of the steam generator in the absence of water. Its development has been internationally framed within the EU-SGTR and the ARTIST program. Experimental activities are focused on setting up a reliable database in which the influence of gas mass flow rate, breach configuration and particle nature in the aerosol retention are properly considered. Theoretical activities are aimed at developing a predictive tool (ARISG) capable of assessing source term attenuation in the scenario with reasonable accuracy. Given the major importance of jet aerodynamics, 3D CFD analyses are being conducted to assist both test interpretation and model development.     

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.