Identification of Failure Modes Under Design Extension Conditions

As the important lessons learned from the Fukushima-nuclear power plant accident, mitigation of failure consequences and prevention of catastrophic failure became essential against severe accident and excessive earthquake conditions. To improve mitigation measures and accident management, clarification of failure behaviors with locations is premise under design extension conditions such as severe accidents and earthquakes. Design extension conditions induce some different failure modes from design conditions. Furthermore, best estimation for these failure modes are required for preparing countermeasures and management. Therefore, this study focused on identification of failure modes under design extension conditions. To observe ultimate failure behaviors of structures under extreme loadings, new experimental techniques were adopted with simulation materials such as lead and lead-antimony alloy, which has very small yield stress. Postulated failure modes of main components under design extension conditions were investigated according three categories of loading modes. The first loading mode is high temperature and internal pressure. Under this mode, ductile fracture and local failure were investigated. At the structural discontinuities, local failure may become dominant. The second is high temperature and external pressure loading mode. Buckling and fracture were investigated. Buckling occurs however hardly break without additional loads or constraints. The last loading is excessive earthquake. Ratchet deformation, collapse, and fatigue were investigated. Among them, low-cycle fatigue is dominant.

Design of Lateral Tee for Steam Pipe in Power Plants

The purpose of this study is to evaluate the design verification of the welded type 45° lateral tee for the steam pipe in power plants. For it, first, the stress analysis was carried out under design condition in accordance with ASME Sec. VIII Div. 2 in order to evaluate the possible occurrence of plastic collapse and local failure. And next, the creep-fatigue damage analysis was performed under the normal operating condition in accordance with ASME Sec. III Subsection NH considering the service temperature of 566°C. From the results, it was found that the welded type 45° lateral tee satisfies the design criteria corresponding to the plastic collapse and the local failure. However, it has a probability of creep rupture during the design life due to the high stress localized in the crotch region. Therefore, a welded type 90° lateral tee was also evaluated with the same analysis procedures to consider the influence of the geometry at the crotch region. Based on the results, the welded type 90° lateral tee satisfies the design criteria of the plastic collapse, local failure and the creep-fatigue strength. This result indicated that an optimal shape design of the crotch region shall be required in order to secure the creep strength of the welded type 45° lateral tee having high service temperature.

Further Improvements to ASME Section XI Code Case N-806 for Evaluation of Metal Loss in Class 2 and 3 Metallic Piping Buried in a Back-Filled Trench

ASME Section XI Code Case N-806, for evaluation of metal loss in Class 2 and 3 metallic piping buried in a back-filled trench, was first published in 2012. This Code Case has been prepared by the ASME Section XI Task Group on Evaluation Procedures for Degraded Buried Pipe. The Code Case addresses the nuclear industry need for evaluation procedures and acceptance criteria for the disposition of metal loss that is discovered during the inspection of metallic piping buried in a back-filled trench. A number of additional improvements have been proposed for Code Case N-806. These include expanded guidance for the determination and validation of a corrosion rate and other clarifications to improve ease of use. This paper presents an update of details of the proposed revisions to Code Case N-806 and their technical basis.

The Investigation of Significance of the Geometrical Nonlinearity for 1-D Crack Stress Intensity Factor Calculation in Slightly Distorted Thin-Walled Pressurized Pipe

The consideration of a geometrical nonlinearity is a common practice for the thin-walled structures. The relevance here are two well-known cases treated in ASME codes. First one is accounting for reduction of the pipe bends flexibility due to the inner pressure. The second one is the retarded increasing (and subsequent saturation) of additional local bending stress with increasing of inner pressure in a pipe with initial cross section form distortion. In both cases the rerounding effect and decreasing of local flexibilities take place. The crack can be treated as the concentrated flexibility and it is quite natural to expect that the stress intensity factor should grow nonlinearly with applied load. Two cases of SIF calculation for 1-D long axial surface crack in a pipe loaded by inner pressure are considered here: a) cross section has an ideal circular form: b) the form has a small distortion and crack is located in the place of maximal additional bending stresses. The theoretical analysis is based on: a) the well known crack compliance method [1] and b) analytical linearized solution obtained for deformation of the curved beam in case of action of fixed circumferential stress due to pressure written in the form convenient for transfer matrix method application. It was shown that for moderately deep crack (crack depth to the wall thickness ratio is 0.5 and bigger) and typical dimensions of pipes used for oil and gas transportation (radius to thickness ratio is 25–40) and loading which can reach up to 200 to 300 MPa, the effect investigated can be quite noticeable and can lead to 5–15 percent reduction of calculated SIF as compared with linear calculation. The analytical results are supported by nonlinear FEM calculation.

Lessons From an Operational Small Leakage at Tsuruga Nuclear Power Station

Operational small leakage is occasionally observed in a nuclear power plant, and the leak forces an operator to decide whether to shut down the plant or not. Even if the leakage is just a little, it might draw the considerable attention in the society, so that the operator sometimes gets into the situation to judge more severely than technical judgment. Furthermore, at the time of plant restart and the system leak test just after maintenance, even the operator doesn’t accept any leakage considering the long management for the leakage up to the next outage. On the other hand, once the operator shut down the plant, it sometimes takes long time to restart again because of the difficulty to obtain new pipes and valves in short time. The temporary repair techniques referred to the JSME code might be able to be applied to maintain the plant operation, however some difficulties exist in a practical process. One of the authors has faced with many cases in which the operational small leakage had to be dealt at Tsuruga nuclear power station. This paper shows some cases of them and discusses lessons which are related to the codes and standards.

Study of Joule-Thomson Cooling Effect During Leakage With Application to Compressed Natural Gas Containment on Marine Vessels

As the localized temperature drop induced by the Joule-Thomson cooling effect in a leak causes a reduction in the fracture toughness at the crack front, leak-before-break approach that does not take this effect into account may be unconservative. In this paper, argon was selected as the experimental gas used in the experiment due to its incombustibility and similar Joule-Thomson coefficient to that of methane. An experimental pressure vessel with a design pressure of 250 bar was designed and fabricated. Liquid nitrogen cracking method was employed to fabricate a realistic through-thickness crack in a test plate. Under the condition of ambient temperature of 30 °C and maximum internal pressure of 91 bar, the temperature of argon at the exit of the crack and the measured lowest temperature of metal near the crack are −9.2 °C and 7.9 °C, respectively.

Failure Analysis of a Cracked Outlet Elbow in Coal Gasification Unit

A crack was observed on an outlet elbow of the pre-converter in coal gasification unit during operation. This paper details the investigation into the failure and highlights the most probable cause of failure based on available documents and experimental analysis. Visual examination, chemical components analysis, energy spectrum analysis, fracture analysis, metallurgical analysis, mechanical properties test and residual stress measurement were performed. The experimental results show that the primary crack initiated from inside and propagated to outer surface of the elbow. The content of titanium element was lower than the requirement in GB/T 14976-2002. Corrosion products were rich in O and S elements. Amounts of secondary cracks and strain induced martensite were observed. Furthermore, the residual stress on the inner surface near the crack tip was extremely high. According to the experiment results and the analysis of operating condition and history, the failure mechanism of the elbow is stress corrosion cracking. Sensitization of the stainless steel due to low Ti content and the faulty heat treatment contributed to the intergranular stress corrosion cracking.

Probabilistic Fracture Mechanics Evaluations That Consider Nozzles in the Extended Beltline Region of Reactor Pressure Vessels

For many years, ASME Section XI committees have discussed the assessment of nozzle penetrations in various flaw evaluations for reactor pressure vessels (RPVs). As summarized in Reference [1], linear elastic fracture mechanics (LEFM) solutions for nozzle penetrations have been in use since the 1970s. In 2013, one of these solutions was adopted into ASME Code, Section XI, Nonmandatory Appendix G (ASME App. G) [2] for use in developing RPV pressure-temperature (P-T) operating limits. That change to ASME App. G was based on compilation of past work [3] and additional evaluations of fracture driving force [4][5]. To establish the P-T limits on RPV operation, consideration should be given to both the RPV shell material with the highest reference temperature as well as regions of the RPV (e.g., nozzles, flange) that contain structural discontinuities. In October 2014, the U.S. Nuclear Regulatory Commission (NRC) highlighted these requirements in Regulatory Issue Summary (RIS) 2014-11 [6]. Probabilistic fracture mechanics (PFM) analyses performed to support pressurized thermal shock (PTS) evaluations using the Fracture Analysis Vessels Oak Ridge (FAVOR) computer code [7] currently evaluate only the RPV beltline shell regions. These evaluations are based on the assumption that the PFM results are controlled by the higher embrittlement characteristic of the shell region rather than the stress concentration characteristic of the nozzle, which does not experience nearly the embrittlement of the shell due to its greater distance from the core. To evaluate this assumption, the NRC and the Oak Ridge National Laboratory (ORNL) performed PFM analyses to quantify the effect of these stress concentrations on the results of the RPV PFM analyses. This paper summarizes the methods and evaluates the results of these analyses.

Options for Defining the Upper Shelf Transition Temperature (Tc) for Ferritic Pressure Vessel Steels

This paper evaluates current guidance concerning conditions under which the analyst is advised to transition from a linear-elastic fracture mechanics (LEFM) based analysis to an elastic-plastic fracture mechanics (EPFM) based analysis of pressure vessel steels. Current guidance concerning the upper-temperature (T>c) for LEFM-based analysis can be found in ASME Section XI Code Case N-749. Also, while not explicitly stated, an upper-limit on the KIc value that may be used in LEFM-based evaluations is sometimes taken to be 220 MPa√m (a value herein referred to as KLIM). Evaluations of Tc and KLIM were performed using a recently compiled collection of toughness models that are being considered for incorporation into a revision to ASME Section XI Code Case N-830; those models provide a complete definition of all toughness metrics needed to characterize ferritic steel behavior from lower shelf to upper shelf. Based on these evaluations, new definitions of Tc and KLIM are proposed that are fully consistent with the proposed revisions to Code Case N-830 and, thereby, with the underlying fracture toughness data. Formulas that quantify the following values over the ranges of RTTo and RTNDT characteristic of ferritic RPV steels are proposed: • For Tc, two values, Tc(LOWER) and Tc(UPPER), are defined that bound the temperature range over which the fracture behavior of ferritic RPV steels transitions from brittle to ductile. Below Tc(LOWER), LEFM analysis is acceptable while above Tc(UPPER) EPFM analysis is recommended. Between Tc(LOWER) and Tc(UPPER), the analyst is encouraged to consider EPFM analysis because within this temperature range the competition of the fracture mode combined with the details of a particular analysis suggest that the decision concerning the type of analysis is best made on a case-by-case basis. • For KLIM, two values, KLIM(LOWER) and KLIM(UPPER), are defined that bound the range of applied-K over which ductile tearing will begin to occur. At applied-K values below KLIM(LOWER), ductile tearing is highly unlikely, so the use of the KIc curve is appropriate. At applied-K values above KLIM(UPPER), considerable ductile tearing is expected, so the use of the KIc curve is not appropriate. At applied-K values in between KLIM(LOWER) and KLIM(UPPER), some ductile tearing can be expected, so it is recommended to give consideration to the possible effects of ductile tearing as they may impact the situation being analyzed. These definitions of Tc and KLIM better communicate important information concerning the underlying material and structural behavior to the analyst than do current definitions.

Development of Probabilistic Evaluation Models of Fracture Toughness KIc and KIa for Japanese RPV Steels

Probabilistic fracture mechanics (PFM) analysis code PASCAL3 has been developed to apply the PFM analysis to the structural integrity assessment of domestic reactor pressure vessels (RPVs). In this paper, probabilistic evaluation models of fracture toughness KIc and KIa which have the largest scatter among the associated factors based on the database of Japanese RPV steels are presented. We developed probabilistic evaluation models for KIc and KIa based on the Weibull and lognormal distributions, respectively. The models are compared with the existing lower bound of fracture toughness in the Japanese code and probabilistic model in USA. As the results, the 5% confidence limits of the models established in present work corresponded to lower bounds of fracture toughness in the Japanese code. The comparison in the models between present work and USA showed significant differences that may have an influence on fracture probability of RPV.