Fitness-for-Service Assessment for Pressure Equipment in Japan

Pressure equipment in refinery and petrochemical industries in Japan has been getting old, mostly more than 30 years in operation. Currently, the Japanese regulations for pressure equipment in service are the same as those in existence during the fabrication of the pressure equipment. Accordingly, there is an immediate need for an up to date more advanced “Fitness For Service” (FFS) evaluation requirements for pressure equipment. In order to introduce the latest FFS methodologies to Japanese industries, the High Pressure Institute of Japan (HPI) has organized two task groups. One is a working group for development of a maintenance standard for non-nuclear industries. Its prescribed code “Assessment procedure for crack-like flaws in pressure equipment” is for conducting quantitative safety evaluations of flaws detected in common pressure equipment such as pressure vessels, piping, storage tanks. The other is a special task group to study of API RP579 from its drafting stage as a member of TG579. The FFS Handbook, especially for refinery and petrochemical industries, has been developed based on API RP579 with several modifications to meet Japanese pressure vessel regulations on April 2001. [1] It is expected that both the Standard and FFS handbook will be used as an exemplified standard with Japanese regulations for practical maintenance. This paper presents concepts of “Assessment procedure for crack-like flaws in pressure equipment” HPIS Z101, 2001 [2].

Pressure Protection Evaluation Program at Savannah River Site

This paper describes the current pressure protection program at Savannah River Site (SRS), a Department of Energy chemical processing and nuclear material handling facility in Aiken, South Carolina. It gives a brief description of the design requirements based on ASME, API, CGA, and ASHRAE Codes. Equipment and systems requiring pressure protection at SRS are primarily pressure vessels, steam stations, process chemical systems, refrigerant and cryogenic systems and other air or gas systems. It is understood that any pressure protection program is built on five fundamental areas of responsibility: procurement, verification, registration, inspection, and repair. This paper focuses on the existing process of facility pressure protection evaluation for code compliance followed by identification of failure scenarios and system design requirements, valve selection and sizing, and verification record generation. Improvements to this process are recognized and discussed. They include the development of a computer program to perform pressure protection evaluation and generate verification records. The software would process all applicable pressure protection calculations using improved methodologies. All relevant data required would be accessible within the program. Pressure safety relief device attributes and system parameters would be displayed. The computer program would enhance design consistency, improve quality and plant safety, and make the pressure protection verification process more efficient and cost effective.

Defect Assessment Procedure for Low to High Temperature Range

There are many similarities between available procedures used for defect assessment. They have been developed as a result of experience gained from material-specific programs and have often been verified using the same data. One recently updated document covering life assessment procedures under creep and creep/fatigue crack growth conditions is BS 7910. This document takes into account some of the most recent developments in the subject, including some from the British Energy R5 Procedure. Future developments in defect assessment procedures will follow the route of simplified and unified codes covering defect behaviour in the low to high temperature range. In this paper, the relevance of the insignificant creep curves in RCC-MR for defect free structures and the creep exemption criteria in BS7910 are examined. Then, an overview is given of some European developments in defect assessment methods for Fitness-for-Service assessment, based on recent and current projects such as the EC thematic network FITNET.

API 579: A Comprehensive Fitness-for-Service Standard

An overview of API 579 Recommended Practice For Fitness-For-Service [1] is presented in this paper. This document was initially released in January of 2000 and since that time has become the de facto international fitness-for-service standard for the refining and petrochemical industry. Insights into the driving force to create API 579 and the activities of an MPC Joint Industry Project to initiate development of the new FFS technologies included in this publication are discussed. A detailed overview of API 579 is then given that covers applicability of the FFS assessment procedures, overall organization, the general assessment methodology used for all flaw and damage types, options for different assessment levels, remaining life and rerating issues, and the relationship with other existing FFS codes and standards. A discussion of the changes planned for the next release of API 579, efforts to validate the fitness-for-service technology incorporated in API 579, and a discussion of a new API document pertaining to damage mechanisms and the relationship to a fitness-for-service assessment are provided. Plans for a joint API and ASME standard and future directions of the API in-service inspection codes relative to API 579 and equipment integrity are also covered.

Assessment of Creep and Creep-Fatigue Crack Growth Following the R5 Procedures

Parts of the R5 high temperature assessment procedures address creep and creep-fatigue crack growth. The procedures were developed some years ago as separate approaches for creep crack growth (R5 Volume 4) and creep-fatigue crack growth (R5 Volume 5). A major revision to these procedures has recently been completed. This unifies the separate approaches into a single procedure for creep-fatigue crack growth (R5 Volume 4/5). The revision restructures the procedure for easier application and includes a number of technical developments, including criteria for deciding when creep-fatigue interactions are important. This paper describes in outline the new R5 creep-fatigue crack growth procedure.

A Fracture Mechanics Evaluation of BWR Shroud Mid-Core Horizontal Weld to Justify Continued Operation

The Mid-core shroud weld (H4 weld) at a BWR plant was inspected during refueling outages in 1998 (RF06) and 2000 (RF07). A structural margin evaluation considering 2000 inspection results for this weld provided technical justification for continued operation to at least the fuel cycle ending in 2002. IGSCC mitigation measures were implemented during operation through 2002. Therefore, a factor of two improvement in the assumed crack growth rate in the depth direction has been applied for operation between 2002 and 2004. The objectives of this paper are to describe the structural evaluation methodology used and describe the results of the evaluation in support of continued operation of H4 weld to 2004. Structural margins for continued operation of H4 weld to 2004 were evaluated two ways. First, the limit load calculations were conducted for a configuration in which through-wall flaws were assumed in regions where the ID surface fluence in 2004 exceeded 3×1020 n/cm2. Since all of the areas taken credit for had a fluence less than the threshold value of 3×1020 n/cm2, a limit load evaluation constituted a complete structural margin evaluation and no linear elastic fracture mechanics (LEFM) or elastic plastic fracture mechanics (EPFM) evaluations were necessary. Secondly, LEFM and EPFM calculations were conducted for the assumed configuration in which through-wall cracking was assumed where the ID surface fluence exceeded 5×1020 n/cm2. The calculated safety factor for the nominal case was determined to be 4.98, which exceeds the required value of 2.77. To demonstrate structural margin in the LEFM regime, a configuration similar to the limit load nominal case was used except that through-wall cracking was assumed in regions where ID fluence exceeded 5×1020 n/cm2 instead of 3×1020 n/cm2. The calculated value of the highest stress intensity factor was 49.2 ksi√in that is less than the allowable value of 54.2 ksi√in. Additional evaluation with EPFM was also conducted to demonstrate higher available structural margins. The EPFM evaluation was conducted by first determining an equivalent single through-wall flaw to conservatively model the LEFM configuration. The applied J-integral values were calculated using the EPRI ductile fracture handbook. A conservative material J-T curve corresponding to a fluence level of 5×1020 n/cm2 was used in the evaluation. The EPFM evaluation showed the structural margin for this case to be 4.0, which exceeds the required value of 2.77. Based on the results of these limit load, LEFM and EPFM structural margin evaluations, it was concluded that the required structural margins will be maintained at the H4 weld for operation through year 2004.

Probabilistic Predictions for Piping Fatigue Including the Effects of Ultrasonic Examinations

This paper presents a methodology for estimating failure probabilities of piping welds that experience cyclic stresses and that are subject to ultrasonic examinations designed to detect growing fatigue cracks. Fatigue cracks can start as either preexisting fabrication flaws or as cracks initiated after an accumulation of stress cycles. Low levels of cyclic stresses and/or small numbers of cycles produce low failure probabilities, with the failures caused mainly by fabrication flaws. More severe cyclic stress conditions produce higher failure probabilities, with the failures caused mainly by fatigue cracks that initiate during the life of the component. Numerical results are presented to address both crack initiation and crack growth. The calculations cover both stainless and ferritic steels, inservice inspections with different inspection intervals, and stress states with and without high levels of through-wall stress gradients. It is shown that effective inspection programs can significantly reduce failure probabilities, and that such programs require suitable NDE sensitivities and adequate inspection frequencies.

Flaw Evaluation Procedures and Acceptance Criteria for Nuclear Components in ASME Code Section XI

The primary objective of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI is to provide the rules and requirements for maintaining pressure boundary integrity of components, piping, and equipment during the life of a nuclear power plant. Pressure boundary integrity in terms of assuring resistance to sudden and catastrophic failure has been an essential objective of the ASME Code since its inception in 1914. These objectives are especially important in ASME Section XI since maintaining pressure boundary integrity of components has a crucial role in ensuring safe and reliable operation in nuclear operating plants. The purpose of this paper is to describe the evaluation procedures, methods, and acceptance criteria for flaws detected in plant components during implementation of in-service inspection surveillance program. For nuclear plant components, pressure boundary integrity includes both leak integrity (no leakage from the reactor coolant system) and structural integrity (no rupture or burst of the pressure boundary). The evaluation requirements in ASME Section XI provide specific rules for assessing the acceptance limits for flaw indications that may be detected during the service life of a nuclear component. In addition to describing current flaw evaluation procedures, details of recent Code developments and improvements are discussed.

A French Guideline for Defect Assessment at Elevated Temperature and Leak Before Break Analysis

A large program is performed in France in order to develop, for the design and operating FBR plants, defect assessment procedures and Leak-Before-Break methods (L.B.B.). The main objective of this A16 guide is to propose analytical solutions at elevated temperature coherent with those proposed at low temperature by the RSE-M (RSE-M, 1997). The main items developed in this A16 guide for laboratory specimen, plates, pipes and elbows are the following: • Evaluation of ductile crack initiation and crack propagation based on the J parameter and material characteristics as JR-Δa curve or Ji / Gfr. Algorithms to evaluate the maximum endurable load under increasing load for through wall cracks or surface cracks are also proposed. • Determination of fatigue or creep-fatigue crack initiation based on the σd approach calculating stress and strain at a characteristic distance d from the crack tip. • Evaluation of fatigue crack growth based on da/dN-ΔKeff relationship with a ΔKeff derived from a simplified estimation of ΔJ for the cyclic load. • Evaluation of creep-fatigue crack growth adding the fatigue crack growth and the creep crack growth during the hold time derived from a simplified evaluation of C*. • Leak-Before-Break procedure. The fracture mechanic parameters determined in the A16 guide (KI, J, C*) are derived from handbooks and formula in accordance with those proposed in the RSE-M document for in service inspection. Those are: • The KI handbook for a large panel of surface and through-wall defects in plates, pipes and elbows. • Elastic stress and reference stress formula. • Analytical Js and Cs* formulations for mechanical and through thickness thermal load. The main part of the formula and assessment methodologies proposed in the A16 guide are included in a software, called MJSAM, developed under the MS Windows environment in support of the document. This allows a simple application of the analysis proposed in the document.

Flaw Evaluation Procedures and Acceptance Criteria for Nuclear Piping in ASME Code Section XI

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code provides rules and requirements for maintaining pressure boundary integrity of components, piping, and equipment during the life of a nuclear power plant. Evaluation procedures and acceptance criteria for the evaluation of flaws in nuclear piping in Section XI of the ASME Code were first published in 1983 and have been under revision for the past several years. This paper provides an overview of the procedures and acceptance criteria for pipe flaw evaluation in Section XI. Both planar and nonplanar flaws are addressed by Section XI. The evaluation procedures and acceptance criteria cover: failure by plastic collapse as characterized by limit load analysis; fracture due to ductile tearing prior to attainment of limit load, as characterized by elastic-plastic fracture mechanics (EPFM) analysis; and brittle fracture as characterized by linear elastic fracture mechanics (LEFM) analysis. A major revision to the evaluation procedures and acceptance criteria was published in the 2002 Addenda to Section XI. Evaluation procedures and acceptance criteria in the 2001 Edition, as well as the revisions in the 2002 Addenda, are described in this paper. Code Cases that address evaluation of wall thinning in piping systems, as well as temporary acceptance of flaws in moderate energy piping systems, are also described.