Using Phased Array Ultrasonic Testing in Lieu of Radiography for Acceptance of Carbon Steel Plate Welds

2014 ◽  
Author(s):  
Traci L. Moran ◽  
Michael T. Anderson ◽  
Anthony D. Cinson ◽  
Susan L. Crawford ◽  
Matthew S. Prowant ◽  
...  
Keyword(s):  
Carbon Steel ◽  
Ultrasonic Testing ◽  
Nuclear Power ◽  
Phased Array ◽  
Steel Plate ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Incomplete Fusion ◽  
Replacement Method

The Pacific Northwest National Laboratory (PNNL) is conducting studies for the U.S. Nuclear Regulatory Commission (NRC) to assess the capability, effectiveness, and reliability of ultrasonic testing (UT) as a replacement method for radiographic testing (RT) for volumetric examination of nuclear power plant (NPP) components. This particular study focused on evaluating the use of phased-array (PA) UT on carbon steel plate welds. Welding fabrication flaws included a combination of planar and volumetric types; for example, incomplete fusion, lack of penetration, cracks, porosity, and slag inclusions. The examinations were conducted using PAUT techniques applied primarily for detection and flaw type characterization. This paper will discuss the results of using PAUT in lieu of RT for detection and classification of fabrication flaws in carbon steel plate welds.

Evaluation of Ultrasonic Phased-Array for Detection of Planar Flaws in High-Density Polyethylene (HDPE) Butt-Fusion Joints

2016 ◽  
Author(s):  
Matthew S. Prowant ◽  
Kayte M. Denslow ◽  
Traci L. Moran ◽  
Richard E. Jacob ◽  
Trenton S. Hartman ◽  
...  
Keyword(s):  
High Density Polyethylene ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Phased Array ◽  
National Laboratory ◽  
Cooling Water ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
High Density

The desire to use high-density polyethylene (HDPE) piping in buried Class 3 service and cooling water systems in nuclear power plants is primarily motivated by the material’s high resistance to corrosion relative to that of steel alloys. The rules for construction of Class 3 HDPE pressure piping systems were originally published as an alternative to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPVC) in ASME Code Case N-755 and were recently incorporated into the ASME BPVC Section III as Mandatory Appendix XXVI (2015 Edition). The requirements for HDPE examination are guided by criteria developed for metal pipe and are based on industry-led HDPE research and conservative calculations. Before HDPE piping will be generically approved for use in U.S. nuclear power plants, the U.S. Nuclear Regulatory Commission (NRC) must have independent verification of industry-led research used to develop ASME BPVC rules for HDPE piping. With regard to examination, the reliability of volumetric inspection techniques in detecting fusion joint fabrication flaws against Code requirements needs to be confirmed. As such, confirmatory research was performed at the Pacific Northwest National Laboratory (PNNL) from 2012 to 2015 to assess the ability of phased-array ultrasonic testing (PAUT) as a nondestructive evaluation (NDE) technique to detect planar flaws, represented by implanted stainless steel discs, within HDPE thermal butt-fusion joints. All HDPE material used in this study was commercially dedicated, 305 mm (12.0 in.) nominal diameter, dimension ratio (DR) 11, PE4710 pipe manufactured with Code-conforming resins, and fused by a qualified and experienced operator. Thermal butt-fusion joints were fabricated in accordance with or intentionally outside the standard fusing procedure specified in ASME BPVC. The implanted disc diameters ranged from 0.8–2.2 mm (0.03–0.09 in.) and the post-fabrication positions of the discs within the fusion joints were verified using normal- and angled-incidence X-ray radiography. Ultrasonic volumetric examinations were performed with the weld beads intact and the PA-UT probes operating in the standard transmit-receive longitudinal (TRL) configuration. The effects of probe aperture on the ability to detect the discs were evaluated using 128-, 64-, and 32-element PA-UT probe configurations. Results of the examinations for each of the three apertures used in this study will be discussed and compared based on disc detection using standard amplitude-based signal analysis that would typically be used with the ultrasonic volumetric examination methods found in ASME BPVC.

Study on Phased Array Ultrasonic Testing Technique During Repair of Carbon Steel Plate using CFRP Materials

2021 ◽  
Vol 41 (2) ◽  
pp. 126-132
Author(s):  
Jung Bin Kim ◽  
Kwang Woon Jung ◽  
Dong chul Choi ◽  
Chang Hun Kim ◽  
Youngman Kim
Keyword(s):  
Carbon Steel ◽  
Ultrasonic Testing ◽  
Phased Array ◽  
Steel Plate ◽  
Testing Technique

Results of a Virtual Round Robin Study to Estimate Probability of Detection for Dissimilar Metal Welds

2021 ◽  
Author(s):  
Ryan M. Meyer ◽  
Aimee E. Holmes ◽  
Romarie Morales ◽  
Iikka Virkkunen ◽  
Thiago Seuaciuc-Osorio ◽  
...  
Keyword(s):  
Pacific Northwest ◽  
Nuclear Power ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Round Robin ◽  
Probability Of Detection ◽  
Pod Analysis ◽  
Regulatory Commission

Abstract This paper presents efforts to overcome challenges with empirical probability of detection (POD) estimations in the nuclear power industry through the utilization of a novel virtual flaw method. A virtual round robin (VRR) study was conducted under the Program for Investigation Of NDE by International Collaboration (PIONIC), organized by the United States Nuclear Regulatory Commission (NRC) utilizing data generated by the virtual flaw method. Analysis of results from the VRR was performed by teams from Pacific Northwest National Laboratory (PNNL), Electric Power Research Institute (EPRI), and Aalto University. Empirically derived POD estimations are presented, and challenges associated with obtaining these estimations are discussed. The virtual flaw method is introduced and some details of its implementation for the VRR activity are described. Results from POD analysis of the VRR data by PNNL, EPRI, and Aalto University are presented and a discussion regarding differences in analysis results is provided. Finally, potential future efforts to improve the application of the virtual flaw method and its estimation of POD are discussed.

An Ultrasonic Phased Array Evaluation of Intergranular Stress Corrosion Crack (IGSCC) Detection in Austenitic Stainless Steel Piping Welds

2010 ◽  
Author(s):  
Aaron A. Diaz ◽  
Michael T. Anderson ◽  
Anthony D. Cinson ◽  
Susan L. Crawford ◽  
Stephen E. Cumblidge
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Pacific Northwest ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Intergranular Stress Corrosion ◽  
Water Reactor ◽  
Nondestructive Examination ◽  
Steel Piping

Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light water reactor (LWR) components and challenging material/component configurations. This study assessed the effectiveness of far-side inspections on wrought stainless steel piping with austenitic welds, as found in thin-walled, boiling water reactor (BWR) component configurations, for the detection and characterization of intergranular stress corrosion cracks (IGSCC).

Enhanced VVER-1000 Fuel Technology and Performance

2009 ◽  
Author(s):  
H. Shah ◽  
R. Latorre ◽  
G. Raspopin ◽  
J. Sparrow
Keyword(s):  
Pacific Northwest ◽  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
The United States ◽  
Department Of Energy ◽  
United States Department ◽  
Safety Level ◽  
And Performance

The United States Department of Energy, through the Pacific Northwest National Laboratory (PNNL), provides management and technical support for the International Nuclear Safety Program (INSP) to improve the safety level of VVER-1000 nuclear power plants in Central and Eastern Europe.

scholarly journals An Approach for Assessing Structural Uplifting Using Blast Motions

2008 ◽  
Author(s):  
Jinsuo Nie ◽  
Jim Xu ◽  
Charles H. Hofmayer ◽  
Syed A. Ali
Keyword(s):  
Coal Mine ◽  
Nuclear Power ◽  
Large Scale ◽  
Ground Motions ◽  
National Laboratory ◽  
Field Tests ◽  
Field Measurements ◽  
Nuclear Regulatory Commission ◽  
Brookhaven National Laboratory ◽  
Ground Acceleration

When a nuclear power plant (NPP) structure is subjected to beyond-design-basis seismic motions, a localized nonlinear effect on the soil-structure system is attributed to separations between the structure and the surrounding soils such as basemat uplift. Experiments involving field tests for real seismic events are usually difficult because of the low probability for large earthquakes at any particular site. To this end, the magnitudes of blast-induced ground motions at a coal mine have been found to be predicatable and can reach very large values. An approach has been developed to investigate whether the strong ground motions recorded at this coal mine can be used to evaluate the basemat uplift effect. This approach involves the use of a scaled ground motion to establish the relationship between the basemat uplift and the peak ground acceleration (PGA). This paper summarizes the field measurements for the ground motions at a coal mine by the Japan Nuclear Safety Organization (JNES) and a method using large scale finite element analyses for basemat uplift assessment performed by Brookhaven National Laboratory for the US Nuclear Regulatory Commission.

scholarly journals An Assessment of Simplified vs. Detailed Methodologies for SSI Analyses of Deeply Embedded Structures

2004 ◽  
Author(s):  
J. Xu ◽  
C. Miller ◽  
C. Hofmayer ◽  
H. Graves
Keyword(s):  
Performance Assessment ◽  
Nuclear Power ◽  
National Laboratory ◽  
Response Spectra ◽  
Nuclear Regulatory Commission ◽  
Brookhaven National Laboratory ◽  
Beam Model ◽  
Simplified Approach ◽  
Embedded Structures ◽  
Different Depths

Sponsored by the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) is carrying out a research program to develop a technical basis to support the safety evaluation of deeply embedded and/or buried (DEB) structures as proposed for advanced reactor designs. In this program, the methods and computer programs established for the assessment of soil-structure interaction (SSI) effects for the current generation of light water reactors are evaluated to determine their applicability and adequacy in capturing the seismic behavior of DEB structures. This paper presents an assessment of the simplified vs. detailed methodologies for seismic analyses of DEB structures. In this assessment, a lump-mass beam model is used for the simplified approach and a finite element representation is employed for the detailed method. A typical containment structure embedded in a soil profile representative of a typical nuclear power plant site was utilized, considering various embedment depths from shallow to full burial. BNL used the CARES program for the simplified model and the SASSI2000 program for the detailed analyses. The calculated response spectra at the key locations of the DEB structure are used for the performance assessment of the applied methods for different depths of burial. Included in the paper are: 1) the description of both the simplified and detailed models for the SSI analyses of the DEB structure, 2) the comparison of the analysis results for the different depths of burial between the two methods, and 3) the performance assessment of the analysis methodologies for SSI analyses of DEB structures. The resulting assessment from this study has indicated that simplified methods may be capable of capturing the seismic response for much deeper embedded structures than would be normally allowed by the standard practice.

An Improved Method for Postulating Fabrication Flaws in Reactor Pressure Vessels for Structural Integrity Evaluation

2002 ◽  
Author(s):  
F. A. Simonen ◽  
T. L. Dickson
Keyword(s):  
Fracture Mechanics ◽  
Structural Integrity ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Inner Surface ◽  
Reactor Pressure

This paper presents an improved model for postulating fabrication flaws in reactor pressure vessels (RPVs) and for the treatment of measured flaw data by probabilistic fracture mechanics (PFM) codes that are used for structural integrity evaluations. The model used to develop the current pressurized thermal shock (PTS) regulations conservatively postulated that all fabrication flaws were inner-surface breaking flaws. To reduce conservatisms and uncertainties in flaw-related inputs, the United States Nuclear Regulatory Commission (USNRC) has supported research at Pacific Northwest National Laboratory (PNNL) that has resulted in data on fabrication flaws from non-destructive and destructive examinations of actual RPV material. Statistical distributions have been developed to characterize the number and sizes of flaws in the various material regions of a vessel. The regions include the main seam welds, repair welds, base metal of plates and forgings, and the cladding that is applied to the inner surface of the vessel. Flaws are also characterized as being located within the interior of these regions or along the weld fusion lines that join the regions. Flaws are taken that occur at random locations relative to the embrittled inner region of the vessel. The probabilistic fracture mechanics model associates each of the simulated flaw types with the fracture properties of the region being addressed.

A Foundation for Stressor-Based Prognostics for Next Generation Systems

2002 ◽  
Author(s):  
Don Jarrell ◽  
Daniel Sisk ◽  
Leonard Bond
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Pacific Northwest National Laboratory ◽  
Department Of Energy ◽  
Residual Lifetime ◽  
Dynamic Force ◽  
Next Generation ◽  
Accurate Identification ◽  
Degradation Rates

Pacific Northwest National Laboratory (PNNL) scientists are performing research under the Department of Energy Nuclear Energy Research Initiative (NERI) program, to develop a methodology for accurate identification and prediction of equipment faults in critical machinery. The 3-year project, on-line intelligent self-diagnostic monitoring system (SDMS) for next generation nuclear power plants is scheduled for completion at the end of FY 2002. The research involves running machinery to failure in the Laboratory by the introduction of intentional faults. During testing, advanced diagnostic/prognostic sensors and analysis systems monitor the equipment stressor levels, correlate them with expected degradation rates, and predict the resulting machinery performance levels and residual lifetime. Application of a first principles physics-based approach is expected to produce prognostic methodologies of significantly higher accuracies than are currently available. This paper reviews the evolution and current state of the maintenance art. It presents a key measurement philosophy that results from the use of condition based maintenance (CBM) as a fundamental investigative precept, and explains how this approach impacts degradation and failure measurement and prediction accuracy. It then examines how this measurement approach is applied in sensing and correlating pump stressors with regard to degradation rate and time to equipment failure. The specifics are examined on how this approach is being applied at PNNL to cavitation and vibration phenomena in a centrifugal pump. Preliminary vibration analysis results show an excellent correspondence between the (laser) motor position indication, the vibration response, and the dynamic force loading on the bearings. Orbital harmonic vibratory motion of the pump and motor appear to be readily correlated through the FFTs of all three sensing systems.

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