Crack Propagation Under High Thermal Amplitude Cycles in a Ductile Material

An experimental setup was built in order to study the fatigue crack propagation in high stressed structure with plastified zone. This paper deals with results obtained in a first stage of the experiment program. Thermal cycles are applied on a simplified mock-up including four short pre-cracking defects. Local measurements (thermocouples, strain gages) and crack lengths are provided. A numerical interpretation of the test was then performed and the results were compared to the experiment.

The Proposal of Wall Thickness Management Criteria of T-Pipes

Although the reinforcement for openings is checked in accordance with design / construction standard when thinning was observed in T-pipes, this evaluation becomes too conservative or requires much time and effort. This paper describes additional parametric study results and proposes a guideline for thickness management of wall thinning T-pipes. On the other papers related to this project, the experiment and numerical simulation results are reported. This paper referred these results and performed further investigation.

Creep Failure Simulations for 316H at 550°C

In this work a method to simulate failure due to creep is proposed using finite element damage analysis. The creep damage model is based on the creep ductility exhaustion concept. Incremental damage is defined by the ratio of incremental inelastic (plastic & creep) strain and multi-axial ductility. A simple linear damage summation rule is applied. When accumulated damage becomes unity, element stresses are reduced to almost zero to simulate progressive crack growth. The model is validated through comparison with experimental data on various sized compact tension, C(T), specimens of 316H stainless steel at 550 °C. The influence of the inelastic strain rate on the uniaxial ductility is considered. Good agreement is found between the simulated results and the experimental data.

Design, Simulation and Testing of Overpack Encapsulation Vessels

The paper presents an overpack designed to contain nuclear product cans which may become pressurised or contaminated. The overpack provides a protective barrier to an inner product can, and due to the possibility of leakage of gas from the contents, the overpack must also function as a pressure vessel. Furthermore, the overpack is required to provide physical protection to the inner can and proof of containment was therefore necessary under a number of different impact scenarios, both pre-pressurised and also with the simulation of pressurisation at the moment of impact. Additionally, the inner product can was to be maintained in a central position during the deceleration at impact. This paper focuses on the analytical design and substantiation of the impact of the system which was performed using an explicit dynamic solver for a number of impact orientations. The design of the overpack to satisfy the relevant pressure vessel Code are not discussed in detail. The potential failure modes of the overpack during impact were assessed and design improvements made over a number of iterations. Following completion of the design and simulation phase, prototypes were built and tested to verify the engineering design and analysis. The testing showed that simulation driven design in conjunction with a pressure vessel design by rule approach was successful in creating a solution for the product can encapsulation. A comparison between the analytical simulation and high-speed video footage of the testing was also made.

A Fully Implicit, Lower Bound, Multi-Axial Solution Strategy for Direct Ratchet Boundary Evaluation: Implementation and Comparison

Ensuring sufficient safety against ratcheting is a fundamental requirement of pressure vessel design. However, determining the ratchet boundary using a full elastic plastic finite element analysis can be problematic and a number of direct methods have been proposed to overcome difficulties associated with ratchet boundary evaluation. This paper proposes a new approach, similar to the previously proposed Hybrid method but based on fully implicit elastic-plastic solution strategies. This method utilizes superimposed elastic stresses and modified radial return integration to converge on the residual state throughout, resulting in one Finite Element model suitable for solving the cyclic stresses (stage 1) and performing the augmented limit analysis to determine the ratchet boundary (stage 2). The modified radial return methods for both stages of the analysis are presented, with the corresponding stress update algorithm and resulting consistent tangent moduli. Comparisons with other direct methods for selected benchmark problems are presented. It is shown that the proposed method consistently evaluates a lower bound estimate of the ratchet boundary, which has not been demonstrated for the Hybrid method and is yet to be clearly shown for the UMY and LDYM methods. Limitations in the description of plastic strains and compatibility during the ratchet analysis are identified as being a cause for the differences between the proposed methods and other current upper bound methods.

Structural Integrity Evaluation of a Reactor Vessel Outlet Nozzle Weld Inlay

Nickel-base weldments such as Alloy 82/182 dissimilar metal (DM) butt welds used in Pressurized Water Reactor (PWR) nuclear power plant components have experienced Primary Water Stress Corrosion Cracking (PWSCC), resulting in the need to repair/replace these weldments. The nuclear industry has been actively engaged in inspecting and mitigating these susceptible DM butt welds for the past several years. Full and Optimized Structural Weld Overlay as well as Mechanical Stress Improvement Process (MSIP®) are some of the mitigation/repair processes that have been implemented successfully by the nuclear industry to mitigate PWSCC. Three conditions must exist simultaneously for PWSCC to occur: high tensile stresses, susceptible material and an environment that is conducive to stress corrosion cracking. These mitigation/repair processes are effective in minimizing the potential for future initiation and crack propagation resulting from PWSCC by generating compressive residual stress at the inner surface of the susceptible DM weld. Weld inlay is an alternative mitigation/repair process especially for large bore nozzles such as reactor vessel nozzles. The weld inlay process consists of excavating a small portion of the susceptible weld material at the inside surface of the component and then applying a PWSCC resistant Alloy 52/52M repair weld layer on the inside surface of the component to isolate the susceptible DM weld material from the primary water environment. The design and analysis requirements of the weld inlay are provided in ASME Code Case N-766. This paper provides the structural integrity evaluation results for a typical reactor vessel outlet nozzle weld inlay performed in accordance with the ASME Code Case N-766 design and analysis requirements. The evaluation results demonstrate that weld inlay is also a viable PWSCC mitigation and repair process especially for large bore reactor vessel nozzles.

Creep Life Simulation and Assessment of High Temperature Piping Systems and Components

High temperature piping systems and associated components, elbows and bellows in particular, are vulnerable to damage from creep. The creep behavior of the system is simulated using finite element analysis (FEA). Material behavior and damage is characterized using the MPC Omega law, which captures creep embrittlement. Elbow elements provide rapid yet accurate modeling of pinching of piping, which consumes a major portion of the creep life. The simulation is used to estimate the remaining life of the piping system, evaluate the adequacy of existing bellows and spring can supports and explore remediation options.

Parametric Study of External Pressure on Steam Generator Tube Bends

Integral light-water reactor designs propose the use of steam generators located within the reactor vessel. Steam generator tubes in these designs must withstand external pressure loadings to prevent buckling, which is affected by material strength, fabrication techniques, chemical environment and tube geometry. Experience with fired tube boilers has shown that buckling in boiler tubes is greatly alleviated by controlling ovality in bends when the tubes are fabricated. Light water reactor steam generator pressures will not cause a buckling problem in steam generators with reasonable fabrication limits on tube ovality and wall thinning. Utilizing existing Code rules, there is a significant design margin, even for the maximum differential pressure case. With reasonable bend design and fabrication limits the helical steam generator thermodynamic advantages can be realized without a buckling concern. This paper describes a theoretical methodology for determining allowable external pressure for steam generator tubes subject to tube ovality based on ASME Section III Code Case N-759-2 rules. A parametric study of the results of this methodology applied to an elliptical cross section with varying wall thicknesses, tube diameters, and ovality values is also presented.

Consideration of Environmentally Assisted Fatigue in Austenitic Stainless Steel: Calculation and Practical Application

The influence of LWR coolant environment to the lifetime of materials has been discussed recent years. Nowadays the consideration of environmentally assisted fatigue is under consideration in Codes and Standards like ASME and the German KTA Rules (e.g. Standard No. 3201.2 and Standard No. 3201.4) by means of so called attention thresholds. Basic calculation procedures in terms of quantifying the influence of LWR coolant environment by the Fen correction factor were proposed by Higuchi and others and are given in NUREG/CR-6909. This paper deals with the application of the proposed assessment procedures of ANL and the application to plant conditions. Therefore conservative assessment procedures are introduced without assuming the knowledge of detailed stress and strain calculations or temperature transients. Additionally, detailed assessment procedures based on Finite-Element calculations, respecting in-service temperature measurements including thermal reference transients and complex operational loading conditions are carried out. Fatigue evaluation of a PWR primary circuit component is used in order to evaluate the influence of plant like conditions numerically. Conclusions regarding the practical application are drawn by means of comparing the ANL approach considering laboratory conditions, conservative assessment procedures for the determination of cumulative fatigue usage factors of plant components and detailed assessment procedures. Plant like loading conditions, complex component geometries, loading scenarios and reference temperature transients shall be taken into account. Practical issues like the determination of the mean temperature or the strain rate have to be considered adequately.

Evaluation of Fracture Behavior for Pressurized T-Pipe by Finite Element Analysis

T-pipe is one of the typical structural elements of LWR piping as well as T-joint. In the present situation, the evaluation of wall thinning for T-pipe is accomplished by assuring the sufficient strength around the opening area by using the design and construction code. This evaluation often assumes the replacement of a local wall thinning with a global wall thinning, which leads to excessively conservative prediction of burst pressures. In this study, three-dimensional finite element analysis was conducted to simulate the fracture behavior of a pressurized T-pipe test. The accuracy of the predicted burst pressures and appropriate modeling of the welded joint at the junction of the main and branch pipes were investigated. It was found that the burst pressure could be adequately predicted by applying a proper fracture criterion. Allocation of experimental tensile property of weld metal to the welded joint gave more accurate prediction; nevertheless, allocation of experimental tensile property of base metal to the welded joint enabled suitably conservative prediction.