Load Factor for Evaluating Non-Planar Flaws in Carbon Steel Piping Using Limit Load Methodology

Non-planar flaw such as local wall thinning flaw is a major piping degradation in nuclear power plants. Hundreds of piping components are inspected and evaluated for pipe wall loss due to flow accelerated corrosion and microbiological corrosion during a typical scheduled refueling outage. The evaluation is typically based on the original code rules for design and construction, and so often that uniformly thin pipe cross section is conservatively assumed. Code Case N-597-2 of ASME B&PV, Section XI Code provides a simplified methodology for local pipe wall thinning evaluation to meet the construction Code requirements for pressure and moment loading. However, it is desirable to develop a methodology for evaluating non-planar flaws that consistent with the Section XI flaw evaluation methodology for operating plants. From the results of recent studies and experimental data, it is reasonable to suggest that the Section XI, Appendix C net section collapse load approach can be used for non-planar flaws in carbon steel piping with an appropriate load multiplier factor. Local strain at non-planar flaws in carbon steel piping may reach a strain instability prior to net section collapse. As load increase, necking starting at onset strain instability leads to crack initiation, coalescence and fracture. Thus, by limiting local strain to material onset strain instability, a load multiplier factor can be developed for evaluating non-planar flaws in carbon steel piping using limit load methodology. In this paper, onset strain instability, which is material strain at the ultimate stress from available tensile test data, is correlated with the material minimum specified elongation for developing a load factor of non-planar flaws in various carbon steel piping subjected to multiaxial loading.

Study on Metallography Test of the Steel SPV490 for Atmospheric Storage Tank After Fire

The fire process was simulated by the heat treatment to the Steel SPV490 of atmospheric storage tank, thereby obtaining the metal specimens in different fire temperature, holding time, and cooling modes. And as the temperature increases, the microscopic structure of Steel SPV490 changes under different working conditions, which could be shown in optical microstructure pictures after doing the interception, inlay, polishing, finishing to the specimens. The result shows that, the mechanical properties of the Steel SPV490 for storage tank changes as the temperature rising from the microscopic view. Nodulizing of the cementite in pearlite occurs, and the strength decreases when the high strength steel SPV490 of large atmospheric storage tanks under air cooling condition below 700 °C, however, it equivalents to the normalizing process, as the sorbite occurs in the steel, and the strength increases a bit when the temperature is above 900 °C. The water-cooling of steel SPV490 above 900 °C equivalents to the process of quenching. The occurrence of martensitic substantially increases the strength and the brittleness, and the elongation decreases rapidly.

Technical Basis for Stress Levels Needed to Mitigate PWSCC in Alloy 82/182/600

The level of stress at which stress corrosion cracking can occur in nickel-based alloys is a key factor in assessing the effectiveness of methods that have been proposed to mitigate such cracking. As various mitigation methods have different approaches and capabilities, the stress limit specified for each case has varied. This paper will review the available data on this subject and develop a consistent basis for defining the stress levels that would be required to initiate stress corrosion cracking in these materials. This paper may be applied to improve the consistency of the requirements of the various mitigation methods involving stress improvement.

Bayesian Statistical Analysis on Chemical Composition Contributing to Irradiation Embrittlement at High Fluence Region

It has been accepted that neutron irradiation embrittlement of reactor pressure vessel is caused by irradiation-induced formation of solute clusters (SCs) and matrix damages (MDs). In the present study, to analyze the contribution of chemical composition contained in SCs to irradiation embrittlement at high fluence region, statistical analysis using the Bayesian nonparametric (BNP) method was performed for Japanese PWR surveillance data. The significance of P, Si and Mn contents, which are not necessarily included in embrittlement correlations unlike the Cu and Ni content, was evaluated. The BNP method can learn the complexity of the statistical model itself from the input data and infer the predicted data with individual probability distribution of predict condition. The result suggested that irradiation embrittlement was most affected by the Si content in three examined elements at high fluence region.

Overview of Proposed High Temperature Design Code Cases

Proposals for high temperature design methods have been developed for primary loads, creep-fatigue and strain limits. The methodologies rely on a common basis and assumption, that elastic, perfectly plastic analysis based on appropriate properties reflects the ability of loads and stress to redistribute for steady and cyclic loading for high temperature as well as for conventional design. The cyclic load design analyses rely on a further key property, that a cyclic elastic-plastic solution provides an upper bound to displacements, strains and local damage rates. The primary load analysis ensures that the design load is in equilibrium with the code allowable stress, taking into account: i) The stress state dependent (multi-axial) rupture criterion, ii) The limit to stress re-distribution defined by the material creep law. The creep-fatigue analysis is focused on the cyclic creep damage calculation, and uses conventional fatigue and creep-fatigue damage calculations. It uses a temperature-dependent pseudo “yield” stress defined by the material yield and rupture data to identify cycles which will not cause creep damage > 1 for the selected life. Similarly the strain limits analysis bounds cyclic strain accumulation. It also uses a temperature-dependent pseudo “yield” stress defined by the material yield and creep strain accumulation data to identify cycles which will not cause average (membrane) inelastic strain > 1% for the design life. The paper gives an overview of the background and justification of these statements, and examples.

Study on Evaluation Procedure for Calculating the Stress Intensity Factor of Flaws Beneath RPV Cladding During Pressurised Thermal Shock Events by FE Analysis

Elastic-plastic finite element (FE) analysis is performed to determine the plastic behavior of the reactor pressure vessel (RPV) inner surface caused by rapid cooling during pressurized thermal shock (PTS) events. However, as the J-integral is not path-independent for elastic-plastic material in the unloading process, it is necessary to apply a suitable correction method using elastic material. In addition, it is also necessary to consider the effect of the welding residual stress appropriately. Therefore, we investigated the stress intensity factor derived from FE analysis based on a model consisting of elastic-plastic cladding and linear elastic low-alloy steel with subsequent plastic zone correction, since the stress level of low-alloy steel remains within the elastic region except the crack front during a PTS event. Furthermore, we examined whether the stress mapping method is applicable for reflecting the effect of welding residual stress in FE analysis, even though the plastic strain generated during welding is ignored.

Developments in Structural Integrity Safety Cases for Highest Reliability Components in the UK

Gross failure of certain components in nuclear power plant has the potential to lead to intolerable radiological consequences. For these components, UK regulatory expectations require that the probability of gross failure must be shown to be so low that it can be discounted, i.e. that it is incredible. For prospective vendors of nuclear power plant in the UK, with established designs, the demonstration of “incredibility of failure” can be an onerous requirement carrying a high burden of proof. Requesting parties may need to commit to supplementary manufacturing inspection, augmented material testing requirements, enhanced defect tolerance assessment, enhanced material specifications or even changes to design and manufacturing processes. A key part of this demonstration is the presentation of the structural integrity safety case argument. UK practice is to develop a safety case that incorporates the notion of ‘conceptual defence-in-depth’ to demonstrate the highest structural reliability. In support of recent Generic Design Assessment (GDA) submissions, significant experience has been gained in the development of so called “incredibility of failure” arguments. This paper presents an overview of some of the lessons learned relating to the identification of the highest reliability components, the development of the structural integrity safety arguments in the context of current GDA projects, and considers how the UK Technical Advisory Group on Structural Integrity (TAGSI) recommendations continue to be applied almost 15 years after their work was first published. The paper also reports the approach adopted by Horizon Nuclear Power and their partners to develop the structural integrity safety case in support of the GDA process to build the UK’s first commercial Boiling Water Reactor design.

On the Applicability of Neuber’s Rule for Low-Cycle Fatigue

Neuber’s rule is commonly applied in fatigue analysis to estimate the plasticity of purely elastic FEA results. In certain cases, this is more efficient than running elastic-plastic models. However, the applicability of Neuber’s rule is not well understood for complex models and may not always be appropriate. In this paper, the applicability of Neuber’s rule is investigated. The background of Neuber’s rule is discussed, theoretical limitations are derived, and algorithmic outlines of the procedures are presented. Neuber’s plasticity correction procedure is applied to both the Ramberg-Osgood elastic-plastic constitutive relation and the advanced Chaboche isotropic/kinematic nonlinear hardening relation. Throughout the manuscript, the aspects of each model are discussed from an educational perspective, highlighting each step of the implementation in sufficient detail for independent reproduction and verification. This level of detail is often absent from similar publications and, it is hoped, may lead to the wider dissemination of Neuber’s rule for plasticity correction. The final component of the paper presents a multiaxial correction of the Chaboche hardening model. To the best of the authors’ knowledge, this is the first published application of Neuber’s rule to the multiaxial plasticity correction of the Chaboche combined isotropic/kinematic hardening model. Examples are used to illustrate the behavior of the method and to present some of the commonly overlooked components when assessing the applicability of Neuber’s method.

A Comparison of Different Design Codes on Fatigue Life Assessment Methods

Different pressure vessel and piping design codes and standards have adopted different fatigue analysis methods. In order to make some contribution to current efforts to harmonize international design codes and standards, a review of fatigue analysis methods for a number of selected nuclear and non-nuclear design codes and standards has been carried out. The selected design codes and standards are ASME Boiler and Pressure Vessel Code Section III Subsection NB and Section VIII Division 2, EN 12952, EN 13445, EN 13480, PD 5500, RCC-M, RCC-MRx, JSME, PNAEG and R5. This paper presents the initial review results. The results of the study could be used as part of the on-going work of the Codes and Standards Task Force of the World Nuclear Association (WNA) Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group.

Numerical Investigation of Nozzle Aspect Ratio Effects on Hydrogen Jet Release Characteristics

As one of the carbon-free energies, hydrogen is considered as an important energy carrier in the 21st century. The minimum ignition energy of hydrogen is the lowest among flammable gases, hence, hydrogen leaking from pinholes, narrow gaps, or broken pipes can be ignited by ignition sources such as static electricity. Understanding of the characteristics of the hydrogen jet release is crucial for the better design and the applications. In this paper, the hydrogen under-expanded jet and flammable envelope were studied by simulation method. The effects of the nozzle shape and release pressure on the under-expanded jet, the downstream shock structure and flammable envelope were investigated. The simulation results showed that the nozzle geometry had great influence hydrogen under-expanded jet. And the maximum flammable length increased with the increasing of aspect ratio. For the split nozzle, with the increasing of pressure the hydrogen diffusion region of minor axes in the near-to-nozzle field increased but decreased of major axes.