Technical Justification for Increasing Temperbead Welding Area Limits on Low Alloy Steel

2007 ◽  
Author(s):  
Marcos L. Herrera ◽  
Shu S. Tang ◽  
Artie Peterson
Keyword(s):  
Ambient Temperature ◽  
Alloy Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Low Alloy Steel ◽  
Weld Overlay ◽  
Gas Tungsten Arc ◽  
Asme Code ◽  
Area Limit ◽  
Inspection Techniques

This paper presents the results of the analytical evaluation supporting the technical justification of increasing the amount of temperbead welding, currently limited to an area of 100 in2, that can be performed on low alloy steel (LAS) nuclear power plant components. The need to expand the application area limitations is increasing for ambient temperature Gas Tungsten Arc Weld (GTAW) temperbead weld overlay repairs on LAS components. As nuclear power plants age and as inspection techniques continue to improve increasing the area limit becomes increasingly important since more indications are being identified. Existing limitations of temperbead welding area of 100 in2 imposed in the ASME Code and in Code Cases 606 and 638 for ambient temperature temper bead welding are arbitrary and overly conservative. This paper presents the analyses supporting: 1) a weld overlay repair greater than 100 in2 on a Reactor Pressure Vessel (RPV) nozzle and 2) a weld cavity repair on an RPV of 500 in2 vertical shell weld. Based on the results of these cases, conclusions regarding temperbead welding in excess of the current 100 in2 limit are made.

Prediction of Plastic Collapse Moments for Circumferential Cracked Pipes With Weld Overlays

2009 ◽  
Author(s):  
Kunio Hasegawa ◽  
Yinsheng Li ◽  
Gery M. Wilkowski ◽  
Arthur F. Deardorff
Keyword(s):  
Experimental Data ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Plastic Collapse ◽  
Analytical Evaluation ◽  
Weld Overlay ◽  
Asme Code ◽  
Weld Overlays

Weld overlay (WOL) is one of the useful repair methods for cracked piping that has been successfully applied for piping in many nuclear power plants. In addition, ASME Boiler and Pressure Vessel Code Section XI provides a WOL method in Non-mandatory Appendix Q and a number of Code Cases. Currently, an analytical evaluation method for predicting failure stresses for WOL piping is under discussion in a working group of ASME Code Section XI. This paper proposes an approach for predicting the plastic collapse moment for WOL piping using a net-section collapse stress approach. In addition, the predicted collapse moments are compared with experimental data.

scholarly journals Effects of Induction Heat Bending and Heat Treatment on the Boric Acid Corrosion of Low Alloy Steel Pipe for Nuclear Power Plants

2016 ◽  
Vol 54 (11) ◽  
pp. 817-825
Author(s):  
Young-Sik Kim ◽  
Ki-Tae Kim ◽  
Min-Chul Shin ◽  
Hyun-Young Chang ◽  
Heung-Bae Park ◽  
...  
Keyword(s):  
Heat Treatment ◽  
Boric Acid ◽  
Alloy Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Acid Corrosion ◽  
Steel Pipe ◽  
Low Alloy Steel ◽  
Induction Heat

A Review of Welding Research Within the New Nuclear Manufacturing (NNUMAN) Programme

2017 ◽  
Author(s):  
Mike C. Smith ◽  
Anastasia N. Vasileiou ◽  
Dinesh W. Rathod ◽  
John Francis ◽  
Neil M. Irvine ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Post Weld Heat Treatment ◽  
Low Alloy Steel ◽  
Gas Tungsten Arc ◽  
Research Themes ◽  
Before And After ◽  
Manufacturing Technologies ◽  
Welding Processes ◽  
Submerged Arc

The New Nuclear Manufacturing (NNUMAN) Programme is a five year UK research project developing manufacturing technologies for next generation nuclear power plants. One of its research themes is advanced joining. Within this theme NNUMAN has manufactured, characterised, and modelled benchmark narrow groove weldments in SA508 Gr 3 Cl 1 low alloy steel over a range of thicknesses up to 130 mm thick using four welding processes: submerged arc (SAW), gas-tungsten arc (GTAW), Electron beam (EB) and cold wire laser (CWL). This paper describes the manufacture and characterisation of the weldments, including diverse residual stress measurements both before and after post weld heat treatment (PWHT).

Proposed Improvements of the ASME B&PV Code and Future Directions of CIWG

2013 ◽  
Author(s):  
Jinquan Yan ◽  
Yinbiao He ◽  
Gang Li ◽  
Hao Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Third Generation ◽  
Future Directions ◽  
The Past ◽  
Development Direction ◽  
Asme Code ◽  
Set Up ◽  
Near Future ◽  
First Time

The ASME B&PV Code, Section III, is being used as the design acceptance criteria in the construction of China’s third generation AP1000 nuclear power plants. This is the first time that the ASME Code was fully accepted in Chinese nuclear power industry. In the past 6 years, a few improvements of the Code were found to be necessary to satisfy the various requirements originated from these new power plant (NPP) constructions. These improvements are originated from a) the stress-strain curves needed in elastic-plastic analysis, b) the environmental fatigue issue, c) the perplexity generated from the examination requirements after hydrostatic test and d) the safe end welding problems. In this paper, the necessities of these proposed improvements on the ASME B&PV code are further explained and discussed case by case. Hopefully, through these efforts, the near future development direction and assignment of the ASME B&PV-III China International Working Group can be set up.

Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

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.

Consideration of Environmental Fatigue in the ASME Code for Carbon and Low-Alloy Steel Components

2003 ◽  
Author(s):  
Stan T. Rosinski ◽  
Arthur F. Deardorff ◽  
Robert E. Nickell
Keyword(s):  
Fatigue Life ◽  
Alloy Steel ◽  
Water Environment ◽  
Operating Experience ◽  
Nuclear Industry ◽  
Operating Conditions ◽  
Low Alloy Steel ◽  
Reactor Water ◽  
Asme Code ◽  
The Impact

The potential impact of reactor water environment on reducing the fatigue life of light water reactor (LWR) piping components has been an area of extensive research. While available data suggest a reduction in fatigue life when laboratory samples are tested under simulated reactor water environments, reconciliation of this data with plant operating experience, plant-specific operating conditions, and established ASME Code design processes is necessary before a conclusion can be reached regarding the need for explicit consideration of reactor water environment in component integrity evaluations. U.S. nuclear industry efforts to better understand this issue and ascertain the impact, if any, on existing ASME Code guidance have been performed through the EPRI Materials Reliability Program (MRP). Based on the MRP activities completed to date there is no need for explicit incorporation of reactor water environmental effects for carbon and low-alloy steel components in the ASME Code. This paper summarizes ongoing MRP activities and presents the technical arguments for resolution of the environmental fatigue issue for carbon and low-alloy steel locations.

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