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.     

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.     

严重事故条件下压力容器完整性评价的研究进展

堆芯熔融物堆内滞留(In-Vessel Retention,IVR)是以AP1000为代表的第三代轻水反应堆严重事故管理的重要策略之一,也是严重事故条件下保证压力容器完整性(Reactor Vessel Integrity,RVI)的典型方法之一.该文综述了国外在严重事故条件下压力容器完整性试验研究和理论分析的现状,总...     

Prediction of crack coalescence of steam generator tubes in nuclear power plants

Prediction of failure pressures of cracked steam generator tubes of nuclear power plants is an important ingredient in scheduling inspection and repair of tubes. Prediction is usually based on nondestructive evaluation (NDE) of cracks. NDE often reveals two neighboring cracks. If the cracks interact, the tube pressure under which the ligament between the two cracks fails could be much lower than the critical burst pressure of an individual equivalent crack. The ability to accurately predict the ligament failure pressure, called “coalescence pressure,” is important. The failure criterion was established by nonlinear finite element model (FEM) analyses of coalescence of two 100% through-wall collinear cracks. The ligament failure is precipitated by local instability of the ligament under plane strain conditions. As a result of this local instability, the ligament thickness in the radial direction decreases abruptly with pressure. Good correlation of FEM analysis results with experimental data obtained at Argonne National Laboratory’s Energy Technology Division demonstrated that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for 100% through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments.     

Structural analysis of electrosleeved tubes under severe accident transients

A flow stress model was developed for predicting failure of electrosleeved PWR steam generator tubing under severe accident transients. The electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400°C during severe accidents because of grain growth. A grain growth model and the Hall–Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data, as well as high-temperature failure tests, on notched electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of electrosleeved tubes with throughwall and part-throughwall axial cracks in the parent tube during a postulated severe accident transient.     

蒸汽发生器完全丧失给水引发的压水堆严重事故研究

采用严重事故最佳估算程序SCDAP/RELAP5/MOD3.4,建立了美国Surry核电站的详细计算模型,对完全丧失给水(TLFW)引发的堆芯熔化事故进行了研究分析.为准确预测压力容器内堆芯熔化的进程,给二级PSA提供可信的初始条件,计算中考虑了一回路压力边界的蠕变破裂失效,并评价了人为干预对堆芯熔化进程及事故后果的影响.     

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.     

Iodine-steel reactions under severe accident conditions in light-water reactors

Owing to large surface areas, the reaction of volatile molecular iodine (I₂) with steel surfaces in the containment may play an important role in predicting the source term to the environment. Both wall retention of iodine and conversion of volatile into non-volatile iodine compounds at steel surfaces have to be considered. Two types of laboratory experiment were carried out at Siemens (KWU) in order to investigate the reaction of I₂ at steel surfaces representative for German power plants.

• 1. 

  (1) For steel coupons submerged in an I2 solution at T = 50, 90 or 140 °C the reaction rate of the I2−I− conversion was determined. No iodine loading was observed on the steel in the aqueous phase tests. I2 reacts with the steel components (Fe, Cr or Ni) to form metal iodides on the surface which are all immediately dissolved in water under dissociation into the metal and the iodide ions. From these experiments, the I2−I− conversion rate constants over the temperature range 50–140 °C as well as the activation energy were determined. The measured data are suitable to be included in severe accident iodine codes such as IMPAIR.

• 2. 

  (2) Steel tubes were exposed to a steam-I2 flow under dry air at T = 120 °C and steam-condensing conditions at T = 120 and 160 °C. In dry air, I2, was retained on the steel surface and a deposition rate constant was measured. Under steam-condensing conditions there is an effective conversion of volatile I2 to non-volatile I− which is subsequently washed off from the steel surface. The I2−T− conversion rate constants suitable for modelling this process were determined. No temperature dependence was found in the range 120–160 °C.

     

Stress identification in steam generator tubes from profile measurements

An identification method devoted to the determination of stresses in tubes, by means of profile measurements, available from on site non-destructive evaluations, is presented here. From the only furnished data (the radial displacement component w on the inner wall), the computation of the strain, and consequently the stresses in the elastic-plastic range, is made within the framework of the shell theory. For this purpose, we need to determine the associated curvature w″: this step is an ill-posed problem because of the lack of continuity with respect to the discrete data. This difficulty is overridden by means of an appropriate regularization procedure. The predictive ability of the method has been tested by comparison with direct simulations; we present an industrial application. This diagnosis tool has been applied successfully to PWR steam generator tubes (in the roll expansion transition zone) and vessel closure head penetrations.     

Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

ABSTRACT

To evaluate the fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident (LOCA) conditions, laboratory-scale integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73–85 GWd/t: M-MDA^TM, low-tin ZIRLO^TM, M5®, and Zircaloy-2 (LK3). In total eight integral thermal shock tests were performed for these fuel cladding tube specimens, simulating LOCA conditions including ballooning and rupture, oxidation, hydriding, and quenching. During the tests, the specimens were oxidized to 10%–30% equivalent cladding reacted (ECR) at approximately 1473 K and were quenched under axial restraint load of approximately 520–530 N. The effects of burnup extension and use of the advanced fuel cladding tubes on the ballooning and rupture, oxidation, and hydriding under LOCA conditions were inconsiderable. Further, the high-burnup advanced fuel cladding tube specimens did not fracture in the ECR values equal to or lower than the fracture limits of the unirradiated Zircaloy-4 cladding tube reported in previous studies. Therefore, it can be concluded that the fracture limit of fuel cladding tubes is not significantly reduced by extending the burnup to approximately 85 GWd/t and using the advanced fuel cladding tubes, though it slightly decreases with increasing initial hydrogen concentration.     

Leakage potential of LWR containment penetrations under severe accident conditions

This paper is an overview of a Sandia National Laboratories, Albuquerque (SNLA) study of the performance of mechanical penetrations in light-water reactor (LWR) containment buildings that are subjected to severe accident environments. The study is concerned with modes of failure as well as the magnitude of leakage. The following tests have been completed, are under way, or are planned: (a) seals and gaskets have been tested to register the effects of radiation aging, thermal aging, seal geometry, and squeeze on seal and gasket materials in severe accident environments; (b) the performance of a full-scale airlock will be evaluated at severe accident temperature and pressure levels; (c) personnel airlock and equipment hatch tests were made on a model of a steel containment building; and (d) tests of mechanical penetrations are planned as part of a test on a model of a reinforced concrete building. This program is part of an overall US Nuclear Regulatory Commission (USNRC) effort to evaluate the integrity of LWR containment buildings.     

External cooling of a reactor vessel under severe accident conditions

The TMI-2 accident demonstrated that a significant quantity of molten core debris could drain into the lower plenum during a severe accident. For such conditions, the Individual Plant Examinations (IPEs) and severe accident management evaluations, consider the possibility that water could not be injected to the RCS. However, depending on the plant specific configuration and the accident sequence, water may be accumulated within the containment sufficient to submerge the lower head and part of the reactor vessel cylinder. This could provide external cooling of the RPV to prevent failure of the lower head and discharge of core debris into the containment.This paper evaluates the heat removal capabilities for external cooling of an insulated RPV in terms of (a) the water inflow through the insulation, (b) the two-phase heat removal in the gap between the insulation and the vessel and (c) the flow of steam through the insulation. These results show no significant limitation to heat removal from the bottom of the reactor vessel other than thermal conduction through the reactor vessel wall. Hence, external cooling is a possible means of preventing core debris from failing the reactor, which if successful, would eliminate the considerations of ex-vessel steam explosions, debris coolability, etc. and their uncertainties. Therefore, external cooling should be a major consideration in accident management evaluations and decision-making for current plants, as well as a possible design consideration for future plants.     

Study on the structural analysis of the containment and the hydrogen detonation problem under severe accident conditions

A brief review of the possible failure modes of a reactor containment is made. And, in light of recent research achievements, the threat of loss of structural integrity of the containment due to gradual overpressure and hydrogen detonation is emphasized.With regard to the methods for assessing the ultimate capacity of a prestressed concrete containment, the simplified engineering method proposed by Bechtel Inc. is verified by comparing the results calculated by the Bechtel method with that obtained by more sophisticated methods and experimental investigations. The comparison demonstrates that the accuracy of the Bechtel method is quite satisfactory.As an extension of the idea of probabilistic evaluation of the occurrence of local hydrogen detonation in a containment, that was suggested by experts from Sandia Laboratory, a new approach based on the use of fuzzy mathematics is proposed. The fuzzy evaluation approach provides a flexible mathematical framework for systematical collection and processing of experimental data, and is a simple and quite effective tool for treating the problem of assessing the possibility of local hydrogen detonation within a containment.     

Simulation of stress corrosion crack growth in steam generator tubes

Assuming a small axial surface crack inside a steam generate (S/G) tube, stress corrosion crack growth is simulated by using finite element method. Pressure difference and residual stresses induced from the roll expansion are considered as applied forces and Scott's crack growth equation based on the stress intensity factor is used. Stress intensity factor distribution along crack front, variation of crack shape and crack growth rate are obtained during the crack growth. From the results, it is noted that for the given residual stress distribution, variation curve of the crack aspect ratio during the crack growth is uniquely determined. In addition, the curve shows nearly constant crack aspect ratio during the initial crack growth stage. When adjacently growing two small cracks are coalesced to form a longer crack, the growth rate of crack depth is increasing but that of crack length is decreasing, and the crack aspect ratio is converging to the original variation curve during the subsequent crack growth.     

Modeling of eddy current probe response for steam generator tubes

Sample calculations were performed with a three-dimensional (3D) finite-element model to describe the response of an eddy current (EC) probe to defects in steam generator (SG) tubing. Such calculations could be very helpful in understanding and interpreting EC probe response to complex tube/defect geometries associated with the inservice inspection (ISI) of SG tubes. The governing field equations are in terms of coupled magnetic vector and electric scalar potentials in conducting media and of total or reduced scalar potentials in nonconducting regions. To establish the validity of the model, comparisons of the theoretical and experimental responses of an absolute bobbin probe are given for two types of calibration standard defects. Simulation results are also presented on the effect of ligament size in axial cracks on bobbin probe response.     

Flow-induced vibration and fretting-wear predictions of steam generator helical tubes

This paper addresses the potential flow-induced vibrations and fretting-wear of helically coiled tubes of the once-through steam generator employed at an integral type nuclear reactor, where the tubes are subjected to liquid cross-flow externally and multi-phase flow internally. The thermal-hydraulic conditions of both tube side and shell side flow fields are predicted using a general purpose computational fluid dynamics code using the finite volume element modeling. To get the natural frequency and corresponding mode shape of the helically coiled tubes with various conditions, a finite element analysis code is used. Based on the results of both the thermal-hydraulic analysis of helically coiled tube steam generator and the modal analysis of the tubes, predictions of turbulence-induced vibration, fluidelastic instability and fretting-wear of the helically coiled tubes are performed. In the predictions, special emphasis is placed on determining the effects of the number of supports, coil diameter and helix pitch on the natural vibration mode, turbulence vibration amplitude, fluidelastic instability and fretting-wear characteristics of the tubes. The results provide the technical information and bases needed by designers and regulatory reviewers for evaluating the design.     

Simulation of the fluid-structure-interaction of steam generator tubes and bluff bodies

The accuracy of computational fluid dynamics in simulating the cross-flow around a steam generator and the feasibility of a full scale coupled CFD/FEA fluid-structure-interaction (FSI) analysis is examined through successive validations.The study begins with a comparison between experiment and computation of flow within a stationary tube bank. Results from the simulation of an individual tube experiencing two-degree-of-freedom flow-induced vibration (at a Reynolds number of 3800) are then shown to compare favorably to experimental results. Finally, free vibration of a single cantilevered hydrofoil is simulated with comparison of mean square acceleration at resonant and non-resonant velocities, respectively. The magnitudes and frequencies of vibration are shown to be accurately captured.     

Coalescence evaluation of collinear axial through-wall cracks in steam generator tubes

To maintain the structural integrity of steam generator tubes, usually, 40% of wall thickness plugging criterion has been adopted. However, since the criterion is applicable only for the steam generator tube containing a single crack, the interaction effect of multiple cracks cannot be considered. In this paper, the coalescence pressure of tube with dual cracks is evaluated based on detailed three-dimensional elastic–plastic finite element analyses. In terms of the crack configuration, collinear axial through-wall cracks with various length, distance and ratio between individual cracks are selected. The applicability of failure pressure prediction models recently proposed by the authors was verified by comparing the finite element analyses results with corresponding experimental data for tubes with two identical cracks. Further, in order to quantify the effect of crack length ratio on failure behavior, the failure pressure prediction model was used expansively for tubes containing different-sized cracks and a coalescence evaluation diagram was developed.     

Hydrogen distribution tests under severe accident conditions at the large-scale HDR-facility

The E11 tests series covered a combination of important issues dominating the physical phenomena controlling the hydrogen distribution mechanisms, namely: large-scale, multi-compartment, geometry with large-sized dome volume, high gas release rates, multiple steam and gas injection phases at different axial positions and examinations of the efficiencies of mitigative system features including the impact of external sprays at the top of the dome.The test series consisted of a total of eight different experiments covering all aspects of the H₂-distribution and potential mitigation features.A total of 700 sensors were applied during these experiments.The paper outlines experimental and computational results of tests E11.2 and E11.4 which were chosen for two computational PHDR-Benchmark Exercises in the context of blind posttest predictions with broad international participation applying the majority of known computer codes. In addition test E11.2 was selected as an open post-test, OECD International Standard Problem No. 29 (H. Karwat, Distribution of hydrogen within the HDR-containment under severe accident conditions — Task specifications (July 1990)) which is presently in progress.     

Lower head integrity under steam explosion loads

Lower head integrity under steam explosion loads in an AP600-like reactor design is considered. The assessment is the second part of an evaluation of the in-vessel retention idea as a severe accident management concept, the first part (DOE/ID-10460) dealing with thermal loads. The assessment is conducted in terms of the risk oriented accident analysis methodology (ROAAM), and includes the comprehensive evaluation of all relevant severe accident scenarios, melt conditions and timing of release from the core region, fully three-dimensional mixing and explosion wave dynamics, and lower head fragility under local, dynamic loading. All of these factors are brought together in a ROAAM probabilistic framework to evaluate failure likelihood. The conclusion is that failure is ‘physically unreasonable’.