Demonstration of Fatigue for LTO License of NPP Borssele

Long Term Operation (LTO) of nuclear power plants (NPPs) requires an ageing management review and a revalidation of Time Limited Ageing Analyses (TLAAs) of structures and components important for nuclear safety. An important ageing effect to manage is fatigue. Generally, the basis for this is formed by the fatigue analyses of the safety relevant components. In this paper, the methodology for the revalidation of fatigue TLAAs is demonstrated for LTO of NPP Borssele in the Netherlands. The LTO demonstration starts with a scoping survey to determine the components and locations having relevant fatigue loadings. The scope was defined by assessment against international practice and guidelines and engineering judgment. Next, a methodical review was performed of all existing fatigue TLAAs. This also includes the latest international developments regarding environmental effects. In order to reduce conservatism, a comparison was made between the number of cycles in the analyses and the number of cycles projected to the end of the intended LTO period. The projected number of cycles is based on transient counting. The loading conditions used in the analyses were assessed by means of temperature measurements by the fatigue monitoring system (FAMOS). As a result of the review, further fatigue assessment or assessment of environmental effects was necessary for certain locations. New analyses were performed using state-of-the-art calculation and assessment methods. The methodology is demonstrated by means of an example of the surge line. The model includes the piping, as well as the nozzles on the pressurizer and the main coolant line. The thermal loadings for the fatigue analysis are based on temperature measurements. Fatigue management of the NPP is ensured by means of the fatigue concept where load monitoring, transient counting and fatigue assessment are coupled through an integrated approach during the entire period of LTO.

Creep Simulation of the Hydroprocessing Reactor Using a Physically-Based CDM Model

Creep failure is one of the most important failure modes in the design of hydroprocessing reactors at elevated temperatures, and the accurate prediction of the creep behavior in structural discontinuities is a critical issue for component design. A physically-based continnum damage mechanics (CDM) model was adopted to describe all three creep stages of 2.25Cr-1Mo-0.25V ferritic steel widely used in manufacturing modern hydroprocessing reactors. The material constants in the damage constitutive equations were identified using an efficient optimization scheme based on genetic algorithm (GA). The user-defined subroutine implementing the CDM model was developed using user programmable features (UPFs) in ANSYS. Three-dimensional finite element analysis of the hydroprocessing reactor was conducted to determine the critical regions, and the studies on the stress redistribution and the prediction of damage evolution in these regions during creep were carried out. The results show that FE modelling based on CDM theory can provide a good tool for creep design of complex engineering components.

Uncertainty Characterization of the CrCOD Circumferential Crack Opening Displacement Module for xLPR

The US NRC/EPRI xLPR (eXtremely Low Probability of Rupture) probabilistic pipe fracture analysis program uses deterministic modules as the foundation for the calculation of the probability of pipe leak or rupture as a consequence of active degradation mechanisms, vibration or seismic loading. The circumferential crack opening displacement module, CrCOD, estimates crack opening displacement (COD) at the inside pipe surface, at the mid-wall thickness location, and at the outside pipe surface using a combined tension/crack face pressure/bending GE/EPRI-like solution. Each module has an uncertainty beyond the uncertainty of the xLPR data inputs. This paper documents the uncertainty for CrCOD. Using 36 pipe fracture experiments, including: base metal, similar metal weld, and dissimilar metal weld experiments; bend only and pressure and bend loading; static and dynamic load histories; cracks that range from short to long, the uncertainty of the CrCOD methodology is characterized. Module uncertainty is presented in terms mean fit and standard deviation between prediction and experimental values.

Investigations on Crack Propagation Under Cyclical Isothermal and Thermo-Mechanical Loadings for a Type 304-L Stainless Steel Used for Pressurized Water Reactor

The integrity of structures exhibiting flaws in Pressurized Water Reactor (PWR) has to be assessed to meet safety criteria. This paper deals with crack-propagation under cyclic thermo-mechanical loadings, as encountered in class I austenitic pipes of PWR’s. To have a conservative and reliable assessment of the crack propagation due to the in-service loading, various codes and standards use simplified method. For example, RSE-M introduces a plastic correction depending on the proportion of the mechanical loading. An improvement of the current method requires additional investigations. Moreover, components loaded with transient or thermal fluctuations are not really in strength-controlled conditions. To this end, a new device called PROFATH was designed. The specimen is a pre-cracked thick-walled tube undergoing a set of thermal cycles and loaded with a static mechanical force. During the first part of the cycle, a high frequency induction coil heats the external wall. Then, the heating system stops and the specimen is cooled down by running water inside the tube. Finite element calculations show that only a region half-way along the tube should be heated to ensure adequate structural effect. In the heated zone, the machining of a sharp circumferential groove ensures the propagation of a unique crack. An electro-mechanical jack controls the level of the mechanical static load. To obtain a very precise thermal mapping, a specific specimen with dedicated instrumentation is used: 20 thermocouples are appropriately located on the outer surface and along the wall thickness. During the test, the crack-propagation is estimated through crack-opening measurements (compliance method). Now, three first tests have been carried out. These tests allow having an evaluation of the pertinence of the correction proposed by the RSE-M for a significant plasticity. Following tests are planned in order to have a confirmation or to propose an extension of the validity domain.

Effect of Bend Angle on Gross Plastic Deformation of Pipe Bends: A Finite Element Based Study

Bends are an integral part of a piping system. Because of the ability to ovalize and warp they offer more flexibility when compared to straight pipes. Piping Code ASME B31.3 [1] provides flexibility factors and stress intensification factors for the pipe bends. Like any other piping component, one of the failure mechanisms of a pipe bend is gross plastic deformation. In this paper, plastic collapse load of pipe bends have been analyzed for various bend parameters (bend parameter = tRbrm2) under internal pressure and in-plane bending moment for various bend angles using both small and large deformation theories. FE code ABAQUS version 6.9EF-1 has been used for the analyses.

Modified Svensson’s Formula for More Accurate Burst Pressure Predictions of Thin Cylindrical Shells With Small Shell Length to Diameter Ratios

For process safety management purposes, it is often required in industries to perform burst pressure calculations for pressurized equipment subjected to internal deflagrations. In predictions of burst pressures of vessels with short cylindrical shell sections, the stiffening effects of heads are often overlooked, which may lead to underestimated results. In this paper, a new improved method is proposed for predictions of burst pressures of thin cylindrical shells with small shell length to diameter ratios (L/D) at ambient temperature. The idea is to use the Svensson’s formula, modified by a factor that accounts for the effects of small L/D ratios, head types, and materials for more accurate predictions. Tables of such factors are developed based on study results from a series of elastic-plastic Finite Element Analyses including large displacement effect for different L/D ratios, thicknesses, materials, and head types. Some example problems are presented to validate the method.

Fatigue Crack Initiation and Growth Observation for 316 Stainless Steel Subjected to Equi-Biaxial Cyclic Loading

Fatigue accumulation is one of the ageing phenomena considered in the plant design and maintenance. The degree of fatigue damage is evaluated by cumulative usage factor using design fatigue curve, which is determined from results of uniaxial fatigue tests. The stress caused by thermal transients is generally equibiaxial, not uniaxial. Fluid temperature fluctuation due to changes in plant conditions, such as plant start-up and shutdown, is the primary cause of fatigue damage. For accurate fatigue damage evaluation, it is important to be conducted under equi-biaxial condition. In this study, pressurized disc fatigue test was conducted in order to simulate the cyclic equi-biaxial stress. In order to clarify how the crack initiates and grows under the equi-biaxial stress condition. Cracking behavior was examined by replica observation method. The crack growth rates were identified by the change in the crack length. It was shown that the fatigue crack growth rate under equi-biaxial stress was faster than that under uniaxial stress for the same equivalent stress intensify factor. It was concluded that the reduction in the fatigue life under equi-biaxial stress was brought about by the accelerated crack growth.

A Rational Methodology for Determination of Service Life for a Delayed Coker Drum

Delayed unit coker drums operate in a severe service environment that precludes long term reliability due to excessive shell bulging and cracking of shell joint and shell to skirt welds. Thermal fatigue is recognized as the leading damage mechanism and past work has provided an idealized description of the thermo-mechanical mechanism via local hot and cold spot formation to quantify a lower bound life estimate for shell weld failure. The present work extends this idealized thermo-mechanical damage model by evaluating actual field data to determine a potential upper bound life estimate. This assessment also provides insight into practical techniques for equipment operators to identify design and operational opportunities to extend the service life of coke drums for their specific service environments. A modern trend of specifying higher chromium and molybdenum alloy content for drum shell material in order to improve low cycle fatigue strength is seen to be problematic; rather, the use of lower alloy materials that are generally described as fatigue tough materials are better suited for the high strain-low cycle fatigue service environment of coke drums. Materials such as SA 204 C (C – ½ Mo) and SA 302 B (C – Mn – ½ Mo) or SA 302 C (C – Mn – ½ Mo – ½ Ni) are shown to be better candidates for construction in lieu of low chromium alloy steel materials such as SA 387 grades P11 (1¼ Cr – ½ Mo), P12 (1 Cr – ½ Mo), P22 (2¼ Cr – 1 Mo) and P21 (3 Cr – 1 Mo).

Structural Assessment of Main Steam Line and Containment Building Under Steam Explosion Conditions

The steam explosion is a fuel-coolant interaction process where the heat transfer from the melt to water is quite intense and rapid. This phenomenon may threat integrity of nuclear components and structures. For instance, the dynamic loads on the reactor cavity and the reactor lower plenum could potentially lead to failure of the main steam lines connected to the steam generators. In addition, since the main steam line extends to the containment wall, failure of the containment building may occur. The object of present study is to examine characteristics of Main Steam Line (MSL) and containment building under the steam explosion conditions. In this context, the corresponding FE models were generated and previously determined displacements of penetration piping were used as loading conditions. Subsequent FE analyses were conducted for the main steam line and containment building to calculate stresses and crack evaluations. Finally, structural assessment of nuclear component and structure combined with concrete failure criteria was performed and their results were discussed.

Validation of Linear Elastic Fracture Mechanics in Predicting the Fracture Toughness of Polyethylene Pipe Materials

The main purpose of the present paper is to compare between the fracture toughness based on linear elastic fracture mechanics (GIC), and that based on nonlinear fracture mechanics (JIC). The material of the investigated pipe is a high-density polyethylene (HDPE), which is commonly used in natural gas piping systems. The welds at the pipe junction are produced by butt-fusion (BF), welding. Curved three-point bend (CTPB), fracture specimens are used. The crosshead speed ranged from 5 to 500 mm/min and specimen thickness ranged from 9 to 45mm for both welded and unwelded specimens at room temperature Ta, equal 23°C. The study reveals that the crosshead speed has a significant effect on the fracture toughness of both welded and unwelded specimens. The results of GIC for different specimen thickness and crosshead speed found previously by the authors [1] have been compared with JIC under the same operating conditions [2]. The comparison between welded and unwelded specimens revealed that in the welded specimens there is a marginal difference between fracture toughness measured using linear elastic fracture mechanics LEFM and elastic plastic fracture mechanics EPFM, for both crosshead speeds.