Technical Basis for a Flaw Tolerance Option for Code Case N-770-1 for Large Diameter Cold Leg Piping to Main Coolant Pump Welds, With Obstructions

2010 ◽  
Author(s):  
Warren Bamford ◽  
Reddy Ganta ◽  
Gordon Hall ◽  
Matthew Kelley
Keyword(s):  
Large Diameter ◽  
Coolant Pump ◽  
Dissimilar Metal ◽  
Flaw Tolerance ◽  
Detectable Rate ◽  
Asme Code ◽  
Mitigation Techniques ◽  
Technical Basis ◽  
Flaw Type ◽  
Very High

An extensive series of evaluations have been performed on the Alloy 82/182 dissimilar metal butt welds located at the safe end regions of the CE designed reactor coolant pump suction and discharge nozzles. These nozzles present inspection coverage challenges, which hinder the likelihood of obtaining the required inspection coverage of MRP-139, and the successor document, ASME Code Case N-770. Furthermore, the geometry of the region also contributes to the difficulty of performing standard mitigation techniques. However, these nozzle regions operate at cold leg temperatures, nominally 550°F, and have a very high resistance to the potential for PWSCC, and a low predicted crack growth rate, if such a flaw were to exist in the region. This leads to the suggestion that the required inspection regimen may be too strong for these regions, and the study described herein was structured to investigate that possibility and develop a technical basis for proposing changes to inspection requirements consistent with the flaw tolerance of the region. Specifically, changes to Code Case N-770 are proposed herein, to take advantage of the flaw tolerance of the region. These proposed changes are described in this paper, and the technical basis for them is described in the remainder of the paper. The technical basis rests on three complementary findings: 1. The probability of a flaw existing or initiating in this region is very low; 2. There is a significant margin between the size flaw which would leak at a detectable rate, and the size flaw which would cause the pipe to fail. This provides a significant level of defense in depth for the region; and 3. The flaw tolerance of the region, for both axial and circumferential flaws, is very high, as measured by the size flaw which could grow to the Section XI allowable flaw size for either flaw type.

Comparison of International Codes for a Fatigue Crack Growth Flaw Tolerance Sample Problem

2021 ◽  
Author(s):  
Gary L. Stevens
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Large Diameter ◽  
Sample Problem ◽  
Pressurized Water ◽  
Typical Application ◽  
Flaw Tolerance ◽  
Asme Code ◽  
The Impact

Abstract As part of the development of American Society of Mechanical Engineers Code Case N-809 [1], a series of sample calculations were performed to gain experience in using the Code Case methods and to determine the impact on a typical application. Specifically, the application of N-809 in a fatigue crack growth analysis was evaluated for a large diameter austenitic pipe in a pressurized water reactor coolant system main loop using the current analytical evaluation procedures in Appendix C of Section XI of the ASME Code [2]. The same example problem was previously used to evaluate the reference fatigue crack growth curves during the development of N-809, as well as to compare N-809 methods to similar methods adopted by the Japan Society of Mechanical Engineers. The previous example problem used to evaluate N-809 during its development was embellished and has been used to evaluate additional proposed ASME Code changes. For example, the Electric Power Research Institute investigated possible improvements to ASME Code, Section XI, Nonmandatory Appendix L [3], and the previous N-809 example problem formed the basis for flaw tolerance calculations to evaluate those proposed improvements [4]. In addition, the ASME Code Section XI, Working Group on Flaw Evaluation Reference Curves continues to evaluate additional research data and related improvements to N-809 and other fatigue crack growth rate methods. As a part of these Code investigations, EPRI performed calculations for the Appendix L flaw tolerance sample problem using three international codes and standards to evaluate fatigue crack growth (da/dN) curves for PWR environments: (1) ASME Code Case N-809, (2) JSME Code methods [5], and (3) the French RSE-M method [6]. The results of these comparative calculations are presented and discussed in this paper.

Changing the Frequency of Inspections for PWSCC Susceptible Welds at Cold Leg Temperatures

2011 ◽  
Author(s):  
Nathan A. Palm ◽  
Warren H. Bamford ◽  
Craig Harrington
Keyword(s):  
Water Stress ◽  
Stress Corrosion Cracking ◽  
Large Diameter ◽  
Reactor Vessel ◽  
Detailed Model ◽  
Service Experience ◽  
Dissimilar Metal ◽  
Flaw Tolerance ◽  
Excellent Service ◽  
Primary Water

A project has been completed under the sponsorship of the EPRI Materials Reliability Program to evaluate the acceptability of returning to an inservice inspection (ISI) frequency of ten years for the large diameter cold leg pipes (525 to 580F), with Alloy 82/182 dissimilar metal (DM) welds. This effort addresses alternative inspection requirements with a frequency of 7 years that have recently been imposed in order to address the potential for service induced Primary Water Stress Corrosion Cracking (PWSCC) of these welds. Careful review of the service experience shows that cracking has only been observed in the hot leg piping locations with DM welds, and the cold leg locations continue to exhibit very reliable service. There are a number of technical and practical arguments in favor of making this change, even beyond the excellent service experience, and these arguments are summarized in this paper. • Pulling the reactor vessel (RV) core barrel is a serious activity which can entail many risks, so additional pulls should be avoided. Inspection at a frequency of less than 10 years involves additional core barrel pulls. • The flaw tolerance of these large diameter cold leg pipes is very good, and example calculations show that reasonably large flaws are acceptable for ten years. • The probability of cracks initiating in cold leg piping is significantly lower than that for piping at hotter temperatures, and a detailed model has been developed to demonstrate this. Actions are underway to revise the relevant inspection requirements, back to a more typical Section XI ten-year interval, using this technical work as a basis.

Technical Basis for Revision of Inspection Requirements for Regenerative and Residual Heat Exchangers

2006 ◽  
Author(s):  
Warren Bamford ◽  
Bruce Bishop ◽  
Richard Haessler ◽  
Mark Bowler
Keyword(s):  
Heat Exchanger ◽  
Heat Exchangers ◽  
Ultrasonic Inspection ◽  
Regenerative Heat ◽  
Flaw Tolerance ◽  
Nuclear Plants ◽  
History Of ◽  
Technical Basis ◽  
Very High ◽  
Residual Heat

Section XI imposed volumetric inservice inspection requirements on heat exchangers in nuclear plants after most of this equipment was designed and installed. Consequently the equipment was not designed for ultrasonic examination, and in some cases such volumetric examination is not justified. The man-rem dose received from the ultrasonic inspection of some of these components is very high, and there are no known mechanisms of degradation; thus, the volumetric inspection serves no useful purpose. With the use of the newly approved code case, N706, volumetric and surface inspection of the regenerative and residual heat exchangers in PWR plants may be replaced with a visual inspection. These two heat exchangers have high irradiation fields, and both have a number of complicated weld geometries that are difficult to inspect. The regenerative heat exchanger provides preheat for the normal charging water going into the reactor coolant system (RCS). The residual heat exchanger is designed to cool the RCS during plant shut down operations. The technical basis for changing these inspection requirements was derived from four fundamental arguments: 1. The heat exchangers were carefully constructed to nuclear quality requirements. 2. They were inspected during construction, and then during service, and there is no history of degradation. 3. The flaw tolerance of the components is very high, since their duty cycle is mild, and they are constructed of stainless steel. 4. The risk is not significantly changed by replacement of the examinations with visual examinations. This paper will describe in detail the technical arguments under each of these topics, which together form the basis for the code case.

Structural Integrity Code and Regulatory Issues in the Design of High Temperature Reactors

2008 ◽  
Author(s):  
William J. O’Donnell ◽  
Amy B. Hull ◽  
Shah Malik
Keyword(s):  
High Temperature ◽  
Elevated Temperature ◽  
Structural Integrity ◽  
Cyclic Creep ◽  
Creep Rupture ◽  
Creep Fatigue ◽  
Asme Code ◽  
High Temperature Reactors ◽  
Gen Iv ◽  
Very High

Since the 1980s, the ASME Code has made numerous improvements in elevated-temperature structural integrity technology. These advances have been incorporated into Section II, Section VIII, Code Cases, and particularly Subsection NH of Section III of the Code, “Components in Elevated Temperature Service.” The current need for designs for very high temperature and for Gen IV systems requires the extension of operating temperatures from about 1400°F (760°C) to about 1742°F (950°C) where creep effects limit structural integrity, safe allowable operating conditions, and design life. Materials that are more creep and corrosive resistant are needed for these higher operating temperatures. Material models are required for cyclic design analyses. Allowable strains, creep fatigue and creep rupture interaction evaluation methods are needed to provide assurance of structural integrity for such very high temperature applications. Current ASME Section III design criteria for lower operating temperature reactors are intended to prevent through-wall cracking and leaking and corresponding criteria are needed for high temperature reactors. Subsection NH of Section III was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid-Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC, DOE and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. The design lifetime of Gen IV Reactors is expected to be 60 years. Additional materials including Alloy 617 and Hastelloy X need to be fully characterized. Environmental degradation effects, especially impure helium and those noted herein, need to be adequately considered. Since cyclic finite element creep analyses will be used to quantify creep rupture, creep fatigue, creep ratcheting and strain accumulations, creep behavior models and constitutive relations are needed for cyclic creep loading. Such strain- and time-hardening models must account for the interaction between the time-independent and time-dependent material response. This paper describes the evolving structural integrity evaluation approach for high temperature reactors. Evaluation methods are discussed, including simplified analysis methods, detailed analyses of localized areas, and validation needs. Regulatory issues including weldment cracking, notch weakening, creep fatigue/creep rupture damage interactions, and materials property representations for cyclic creep behavior are also covered.

The Effect of Temperature Dependent Yield Strength on Upper Bounds for Creep Ratcheting

2006 ◽  
Author(s):  
Timothy E. McGreevy ◽  
Frederick A. Leckie ◽  
Peter Carter ◽  
Douglas L. Marriott
Keyword(s):  
Yield Strength ◽  
Upper Bounds ◽  
Effect Of Temperature ◽  
Strain Accumulation ◽  
Core Concept ◽  
Temperature Dependent ◽  
Inelastic Analysis ◽  
Asme Code ◽  
Inelastic Design ◽  
Very High

The Bree model and the elastic core concept have been used as the foundation for the simplified inelastic design analysis methods in the ASME Code for the design of components at elevated temperature for nearly three decades. The methodology provides upper bounds for creep strain accumulation and a physical basis for ascertaining if a structure under primary and secondary loading will behave elastically, plastically, shakedown, or ratchet. Comparisons of the method with inelastic analysis results have demonstrated its conservatism in stainless steel at temperatures representative of those in LMBR applications. The upper bounds on creep accumulation are revisited for very high temperatures representative of VHTR applications, where the yield strength of the material is strongly dependent upon temperature. The effect of the variation in yield strength on the evolution of the core stress is illustrated, and is shown to extend the shakedown regions, and affects the location of the boundaries between shakedown, ratcheting, and plasticity.

Effects of Cladding on Reactor Pressure Vessel Integrity Based on the Revised PTS Rule (10 CFR 50.61)

2009 ◽  
Author(s):  
Vikram Marthandam ◽  
Timothy J. Griesbach ◽  
Jack Spanner
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Historical Perspective ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Potential Benefits ◽  
Vessel Integrity ◽  
Technical Basis ◽  
Reactor Pressure

This paper provides a historical perspective of the effects of cladding and the analyses techniques used to evaluate the integrity of an RPV subjected to pressurized thermal shock (PTS) transients. A summary of the specific requirements of the draft revised PTS rule (10 CFR 50.61) and the role of cladding in the evaluation of the RPV integrity under the revised PTS Rule are discussed in detail. The technical basis for the revision of the PTS Rule is based on two main criteria: (1) NDE requirements and (2) Calculation of RTMAX-X and ΔT30. NDE requirements of the Rule include performing volumetric inspections using procedures, equipment and personnel qualified under ASME Section XI, Appendix VIII. The flaw density limits specified in the new Rule are more restrictive than those stipulated by Section XI of the ASME Code. The licensee is required to demonstrate by performing analysis based on the flaw size and density inputs that the through wall cracking frequency does not exceed 1E−6 per reactor year. Based on the understanding of the requirements of the revised PTS Rule, there may be an increase in the effort needed by the utility to meet these requirements. The potential benefits of the Rule for extending vessel life may be very large, but there are also some risks in using the Rule if flaws are detected in or near the cladding. This paper summarizes the potential impacts on operating plants that choose to request relief from existing PTS Rules by implementing the new PTS Rule.

Flaw Tolerance Evaluation of CASS Piping Materials

2009 ◽  
Author(s):  
Timothy J. Griesbach ◽  
Vikram Marthandam ◽  
Haiyang Qian ◽  
Patrick O’Regan
Keyword(s):  
Fracture Toughness ◽  
Weld Metal ◽  
Thermal Aging ◽  
Prolonged Exposure ◽  
Delta Ferrite ◽  
Acceptance Criteria ◽  
Centrifugally Cast ◽  
Reactor Coolant ◽  
Flaw Tolerance ◽  
Asme Code

Prolonged exposure of cast austenitic stainless steels (CASS) to reactor coolant operating temperatures has been shown to lead to some degree of thermal aging embrittlement (reduction in fracture toughness of the material as a function of time). The fracture toughness data for the most severely aged CASS materials were found to be similar to those reported for some austenitic stainless steel weld metal, in particular weld metal from submerged arc welds (SAW). Such similarity offers the possibility for applying periodic inservice inspection flaw acceptance criteria, currently referenced in the ASME Code Section XI, Subsection IWB, for SAW and shielded metal arc weld (SMAW), to CASS component inservice inspection results. This paper presents the data to support both the proposed screening criteria (based on J-R crack growth resistance) for evaluation of the potential significance of the effects of thermal aging embrittlement for Class 1 reactor coolant system and primary pressure boundary CASS components, for those situations where the effects of thermal aging embrittlement are found to be potentially significant. The fitness for continued service is based on the comparison of the limiting fracture toughness data for Type 316 SAW welds and the lower-bound fracture toughness data reported for high-molybdenum, high delta-ferrite, statically and centrifugally-cast CASS materials. These comparisons and the associated flaw acceptance criteria can be used to justify exemptions from current ASME Code Section XI inservice inspection requirements through flaw tolerance evaluation (e.g., see ASME Nuclear Code Case N-481).

Application of ASME Code Section XI Appendix L Using 3-D Finite Element Analyses

2020 ◽  
Author(s):  
Charles Fourcade ◽  
Minji Fong ◽  
James Axline ◽  
Do Jun Shim ◽  
Chris Lohse ◽  
...  
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Crack Growth ◽  
Cyclic Stress ◽  
Finite Element Analyses ◽  
Element Analysis ◽  
Aspect Ratios ◽  
Flaw Tolerance ◽  
Asme Code ◽  
Stress Ratios

Abstract As part of a fatigue management program for subsequent license renewal, a flaw tolerance evaluation based on ASME Code, Section XI, Appendix L may be performed. The current ASME Code, Section XI, Appendix L flaw tolerance methodology requires determination of the flaw aspect ratio for initial flaw size calculation. The flaw aspect ratios listed in ASME Section XI, Appendix L, Table L-3210-2, for austenitic piping for example, are listed as a function of the membrane-to-gradient cyclic stress ratio. The Code does not explicitly describe how to determine the ratio, especially when utilizing complex finite element analyses (FEA), involving different loading conditions (i.e. thermal transients, piping loads, pressure, etc.). The intent of the paper is to describe the methods being employed to determine the membrane-to-gradient cyclic stress ratios, and the corresponding flaw aspect ratios (a/l) listed in Table L-3210-2, when using finite element analysis methodology. Included will be a sample Appendix L evaluation, using finite element analysis of a pressurized water reactor (PWR) pressurizer surge line, including crack growth calculations for circumferential flaws in stainless steel piping. Based on this example, it has been demonstrated that, unless correctly separated, the membrane-to-gradient cyclic stress ratios can result in extremely long initial flaw lengths, and correspondingly short crack growth durations.

Growth of Large Diameter 6H SI and 4H n+ SiC Single Crystals

2010 ◽  
Vol 1246 ◽  
Author(s):  
Avinash Gupta ◽  
Ping Wu ◽  
Varatharajan Rengarajan ◽  
Xueping Xu ◽  
Murugesu Yoganathan ◽  
...  
Keyword(s):  
Single Crystals ◽  
Growth Process ◽  
Large Diameter ◽  
Commercial Production ◽  
High Resistivity ◽  
Steady Growth ◽  
Vanadium Doping ◽  
Low Capacitance ◽  
Very High

AbstractSiC single crystals are grown at II-VI by the seeded sublimation technique. The process has been scaled up and optimized to support commercial production of high-quality 100 mm diameter, Semi-Insulating (SI) 6H substrates and 100 mm 4H n+ substrates. The growth process incorporates special elements aimed at achieving uniform sublimation of the source, steady growth rate, uniform doping and reduced presence of background impurities.Semi-insulating 6H substrates are produced using precise vanadium compensation. Vanadium doping is optimized to yield SI material with very high resistivity and low capacitance.Crystal quality of the substrates is evaluated using a wide variety of techniques. Specific defects, their interaction and evolution during growth are described with emphasis on micropipes and dislocations. The current quality of the 6H SI and 4H n+ crystals grown at II-VI is summarized.

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