scholarly journals Application of the FAVOR-OCI Fracture Mechanics Computer Program to ASME Code Section XI, IWB-3610 Flaw Acceptance Criteria Evaluations

2018 ◽  
Author(s):  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass ◽  
Hilda B. Klasky
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Critical Values ◽  
Initiation Toughness ◽  
Acceptance Criteria ◽  
Oak Ridge ◽  
The Us ◽  
Asme Code

This paper presents an overview of added features in a new version of the FAVOR (Fracture Analysis of Vessels Oak Ridge) computer code called FAVOR-OCI. The original FAVOR code was developed at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission (NRC). FAVOR is applied by US and international nuclear power industries to perform deterministic and probabilistic fracture mechanics analyses of commercial nuclear reactor pressure vessels (RPVs). Applications of FAVOR are focused on insuring that the structural integrity of aging, and increasingly embrittled, RPVs is maintained throughout their licensed service life. Based on the final ORNL release of FAVOR, v16.1, FAVOR-OCI extends existing deterministic features of FAVOR while preserving all previously-existing probabilistic capabilities of FAVOR. The objective of this paper is to describe new deterministic features in FAVOR-OCI that can be applied to analytical evaluations of planar flaws. These evaluations are consistent with the acceptance criteria of ASME Code, Section XI, Article IWB-3610, including Subarticles IWB-3611 (acceptance based on flaw size) and IWB-3612 (acceptance based on applied stress intensity factor). The linear elastic fracture mechanics (LEFM) capabilities of FAVOR-OCI also incorporate the analytical procedures presented in the Nonmandatory Appendix A, Analysis of Flaws, Article A-3000, Method of KI Determination, for both surface and subsurface (embedded) flaws. The paper describes a computational methodology for determining critical values of fracture-related parameters that satisfy ASME Code Section XI acceptance criteria for flaws exposed to multiple thermal-hydraulic transients. These compute-intensive analyses can be carried out with a single execution of FAVOR-OCI. The new methodology determines critical values by solving for either the point of tangency or point of intersection between applied KI versus time histories and a user-selected cleavage initiation toughness material property (e.g., ASME KIc, FAVOR Weibull KIc, or Master Curve Weibull KJc) for surface or subsurface flaws. Situations where warm prestress conditions apply can also be addressed. The paper highlights a need for this new capability via applications to a recent independent review of safety cases for RPVs in two Belgian nuclear power plants (NPPs). That review required ASME Section XI assessments of several thousand embedded, quasi-laminar flaws in the wall of each RPV Analysis results provided by the new capability contributed to the technical bases compiled from several sources by the Belgian nuclear regulatory agency (FANC) and eventually used by FANC to justify the restart of these NPPs.

Alternative RTNDT for Linde 80 Weld Materials

2002 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Screening Criteria ◽  
Master Curve Method ◽  
Asme Code ◽  
Curve Method

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.

2004 Edition of Japanese Fitness-for-Service Code for Nuclear Power Plants

2006 ◽  
Author(s):  
Koichi Kashima ◽  
Tomonori Nomura ◽  
Koji Koyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Industrial Safety ◽  
National Rule ◽  
Class 1 ◽  
Asme Code ◽  
Service Inspection ◽  
Core Shroud

JSME (Japan Society of Mechanical Engineers) published the first edition of a FFS (Fitness-for-Service) Code for nuclear power plants in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection. Individual inspection rules were prescribed for specific structures, such as the Core Shroud and Shroud Support for BWR plants, in consideration of aging degradation by Stress Corrosion Cracking (SCC). Furthermore, JSME established the third edition of the FFS Code in December 2004, which was published in April 2005, and it included requirements on repair and replacement methods and extended the scope of specific inspection rules for structures other than the BWR Core Shroud and Shroud Support. Along with the efforts of the JSME on the development of the FFS Code, Nuclear and Industrial Safety Agency, the Japanese regulatory agency approved and endorsed the 2000 and 2002 editions of the FFS Code as the national rule, which has been in effect since October 2003. The endorsement for the 2004 edition of the FFS Code is now in the review process.

Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL

2017 ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments

Probabilistic fracture mechanics (PFM) has been recognized as a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency (JAEA) to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against non-ductile fracture. On the other hand, unlike deterministic analysis codes, the verification of PFM analysis codes is not easy. A series of activities has been performed to verify the applicability of PASCAL. In this study, as a part of the verification activities, a working group was established in Japan, with seven organizations from industry, universities and institutes voluntarily participating as members. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group including the verification plan, approaches and results.

Technical Basis for Cases N-629 and N-631 as an Alternative for RTNDT Reference Temperature

2007 ◽  
Author(s):  
J. G. Merkle ◽  
K. K. Yoon ◽  
W. A. VanDerSluys ◽  
W. Server
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Threshold Level ◽  
Nuclear Regulatory Commission ◽  
New Approach ◽  
The Us ◽  
Asme Code ◽  
Regulatory Commission ◽  
Technical Basis

ASME Code Cases N-629/N-631, published in 1999, provided an important new approach to allow material specific, measured fracture toughness curves for ferritic steels in the code applications. This has enabled some of the nuclear power plants whose reactor pressure vessel materials reached a certain threshold level based on overly conservative rules to use an alternative RTNDT to justify continued operation of their plants. These code cases have been approved by the US Nuclear Regulatory Commission and these have been proposed to be codified in Appendix A and Appendix G of the ASME Boiler and Pressure Vessel Code. This paper summarizes the basis of this approach for the record.

Machining Test Specimens From Harvested Zion RPV Segments

2015 ◽  
Author(s):  
Thomas M. Rosseel ◽  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Compact Tension ◽  
Department Of Energy ◽  
Oak Ridge ◽  
Damage Models ◽  
The Us ◽  
Machining Test

The decommissioning of the Zion Nuclear Generating Station (NGS) in Zion, Illinois, presents a special and timely opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, an international nuclear services company, the selective procurement of materials, structures, components, and other items of interest from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), cutting these segments into blocks from the beltline and upper vertical welds and plate material and machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for microstructural (TEM, SEM, APT, SANS and nano indention) characterization. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models [1].

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

2003 ◽  
Author(s):  
Douglas A. Scarth ◽  
Gery M. Wilkowski ◽  
Russell C. Cipolla ◽  
Sushil K. Daftuar ◽  
Koichi K. Kashima
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Limit Load ◽  
Acceptance Criteria ◽  
Linear Elastic ◽  
Evaluation Procedures ◽  
Piping Systems ◽  
Asme Code ◽  
American Society ◽  
Plant Evaluation

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.

Validation of Pipe Flaw Evaluation Procedures in Section XI of the ASME Code Against Pipe Fracture Experiments

2004 ◽  
Author(s):  
Phuong H. Hoang ◽  
Gery M. Wilkowski
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Limit Load ◽  
Acceptance Criteria ◽  
Linear Elastic ◽  
Evaluation Procedures ◽  
Asme Code ◽  
Elastic Fracture ◽  
American Society ◽  
Plant Evaluation

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 piping 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. 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. A brief overview of the pipe flaw evaluation procedures published in the 2002 Addenda are provided in the paper. The evaluation procedures that were published in the 2002 Addenda have been validated against the results of a large number of pipe fracture experiments. The results of this validation exercise are summarized in this paper.

Uncertainties in Probabilistic Fracture Mechanics Calculations

2010 ◽  
Author(s):  
F. A. Simonen
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Flaw Detection ◽  
Pressure Vessels ◽  
Operating Conditions ◽  
Statistical Distributions ◽  
Probabilistic Fracture Mechanics ◽  
Crack Growth Rates ◽  
Failure Probabilities

This paper addresses uncertainties in probabilistic fracture mechanics (PFM) calculations for pressure boundary components at commercial nuclear power plants. Such calculations can predict the probability that a component will have failed after a specified period of operation, but with large uncertainties that are difficult to quantify. PFM models only approximate details of as-built components as well as actual operating conditions over the lifetime of the component. Statistical distributions used as inputs to the calculations are subject to uncertainties, which also results in large uncertainties in calculated failure probabilities. This paper describes from the author’s perspective various uncertainties that are associated with PFM calculations. Efforts to quantify PFM uncertainties are described along with their impacts on calculated failure probabilities. Many uncertainties are explicitly addressed by statistical distributions for input parameters to the PFM models (e.g. crack growth rates, material strengths, probabilities of flaw detection, etc.). Other calculations have gone further by estimating uncertainties in the parameters of these statistical distributions along with uncertainties in parameters treated as deterministic inputs to the PFM models. Examples from the author’s experience with uncertainty analyses for pressure vessels and piping components are described.

Combination Criterion for Multiple Laminar Flaws in Steel Components

2017 ◽  
Vol 741 ◽  
pp. 63-69
Author(s):  
Valéry Lacroix ◽  
Vratislav Mareš ◽  
Bohumír Strnadel ◽  
Kunio Hasegawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Rolling Direction ◽  
Separation Distance ◽  
Pressure Vessels ◽  
New Combination ◽  
Mechanical Interaction ◽  
Single Plane ◽  
Asme Code ◽  
American Society

A laminar flaw is a planar subsurface flaw parallel to the rolling direction of the plate, where the applied stress is typically parallel to the rolling direction. The laminar flaw oriented within 10 degree of a plane parallel to the component surface is defined as a laminar flaw, in accordance with the definition of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel (B&PV) Code Section XI. The ASME Code provides combination criterion for multiple laminar flaws. If there are two or more laminations, these laminations are projected to a single plane and, if the separation distance of the projected laminations is less than or equal to 25.4 mm, the laminations shall be combined into a single large laminar flaw in the assessment. The combination criterion was established on the basis of the non-destructive examination capabilities in the 1970’s. However, this methodology did not consider the offset distance of the laminations nor the mechanical interaction between the flaws. Therefore that combination methodology is not suited in case of a large number of laminar flaws. This may occur e.g. in case of hydrogen flaking in steel forging components. Actually, when multiple discrete laminar flaws are close to each other, interaction between the flaws has to be taken into account and these flaws shall be combined to a single laminar flaw for assessment. Stress intensity factor interactions for inclined laminar flaws were analyzed in the frame of hydrogen flaking issue in reactor pressure vessels of Doel 3 and Tihange 2 Belgian nuclear power plants. Based on the mechanical interaction between flaws, new combination criterion was developed and was presented in this paper.

Nuclear Regulatory Commission Staff View of Progress in the Nondestructive Testing of Cast Austenitic Steel Components

2014 ◽  
Author(s):  
Stephen E. Cumblidge
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Degradation Mechanism ◽  
Austenitic Steels ◽  
Inspection Method ◽  
The Past ◽  
Asme Code

Welds in cast austenitic steels (CASS) are very challenging to inspect using the current American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements. Supplement 9 of ASME Boiler and Pressure Vessel Code Section XI, Appendix VIII is still in the course of preparation, requiring inspectors to use ASME Code Section XI, Appendix III, which provides prescriptive ultrasonic testing (UT) requirements that are significantly less rigorous than UT techniques that have been demonstrated under Appendix VIII. The inability of licensees to demonstrate that the welds in CASS components meet ASME Code requirements has been an ongoing area of concern for the NRC staff. The lack of a reliable inspection method for welds in CASS materials has led to hundreds of relief requests over the past four decades. While no degradation mechanism has been found in CASS components to date, there is no guarantee that a new degradation mechanism affecting CASS welds will not emerge as nuclear power plants go beyond forty years of operation. Licenses need qualified procedures and personnel for the inspection of welds in CASS materials in order to put licensees into compliance with ASME Code, meet federal regulations, reduce the number of needed relief requests, and ensure the structural integrity of their welds. Over the past decade there have been significant developments in nondestructive examination (NDE) technology. The use of encoded phased array techniques using low frequency ultrasound has been shown to be able to reliably find flaws greater than 30% through wall in CASS materials with a variety of microstructures. Additionally, an improved understanding of the fracture mechanics of CASS components is being developed that shows the flaw sizes that can be tolerated in CASS components. These advances in NDE techniques and fracture mechanics theory are converging on a path to allow for qualifications of procedures and personnel for the ultrasonic inspections of welds in CASS components. Recent developments in ASME Code includes Code Case N-824, which provides guidance on the examination of CASS materials based on the advances in NDE technology and an improved understanding of the NDE techniques capable of finding flaws in CASS components as well as Code Case N-838 for flaw tolerance evaluations of CASS piping components. Finally, work on ASME Code Section XI Supplement 9 is progressing, with several important issues still to be addressed. The NRC staff sees a clear path forward and is working to ensure that qualified inspections of welds in CASS materials will be possible in the future.

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