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.

Ultrasonic Phased Array: Detection of Detrimental Conditions in High-Density Polyethylene Butt-Fusion Joints

2010 ◽  
Author(s):  
Caleb J. Frederick
Keyword(s):  
High Density Polyethylene ◽  
Nuclear Power ◽  
Power Plants ◽  
Phased Array ◽  
Cold Fusion ◽  
High Density ◽  
Visual Examination ◽  
Full Spectrum ◽  
Ultrasonic Phased Array ◽  
Joint Integrity

Today, commercial nuclear power plants are installing High-Density Polyethylene (HDPE) in non-safety-related and safety-related applications. While this material has numerous advantages over the carbon steel pipes that historically have been used for the same applications, developing a way to accurately inspect for joint integrity in HDPE has become increasingly important to utilities and the U.S. Nuclear Regulatory Commission (USNRC). This paper will investigate the ability to quantify the levels of detection of flaws and detrimental conditions using ultrasonic phased array, in butt-fusion joints throughout the full spectrum of applicable HDPE pipe diameters and wall-thicknesses. Perhaps the most concerning joint condition is that of “Cold Fusion”. A cold-fused joint is created when molecules along the fusion line do not fully entangle or co-crystallize. Once the fusion process is complete, during visual examination, there is the appearance of a good quality joint. However, the joint does not have the strength needed, as the required co-crystallization along the pipe faces has not occurred. Performing a visual examination of the bead, as required by the current revision of ASME Code Case N-755, does not provide adequate guarantee of joint integrity. Therefore, volumetric examination is of special concern to the USNRC to safeguard against this type of detrimental condition. Factors addressed will include pipe diameter, wall-thickness, fusing temperature, interfacial pressure, dwell (open/close) time, and destructive verification of ultrasonic data.

Quantifying Detection Capabilities of Detrimental Conditions in High-Density Polyethylene Butt-Fusion Joints

2010 ◽  
Author(s):  
Caleb J. Frederick
Keyword(s):  
High Density Polyethylene ◽  
Nuclear Power ◽  
Power Plants ◽  
Cold Fusion ◽  
Nuclear Regulatory Commission ◽  
High Density ◽  
Full Spectrum ◽  
Asme Code ◽  
Regulatory Commission ◽  
Joint Integrity

Today, commercial nuclear power plants are installing High-Density Polyethylene (HDPE) piping in non-safety-related and safety-related systems. HDPE has been chosen in limited quantity to replace carbon steel piping as it does not support rust, rot, or biological growth. However, due to its relatively short nuclear service history, developing a way to accurately evaluate joint integrity in HDPE is crucial to utilities and the U.S. Nuclear Regulatory Commission (USNRC). This paper will investigate using ultrasonic Phased Array to quantify detection of flaws and detrimental conditions in butt-fusion joints throughout the full spectrum of applicable HDPE pipe diameters and wall-thicknesses. Currently the most concerning joint condition is that of “Cold Fusion”. A cold-fused joint is created when molecules along the fusion line do not fully entangle or co-crystallize. Once the fusion process is complete, there is the appearance of a good, quality joint. However, the fabricated joint does not have the required strength as the co-crystallization along the pipe faces has not occurred. Therefore, performing a visual examination of the bead, as required by the current revision of ASME Code Case N-755, does not provide adequate assurance of joint integrity. As a potential solution, volumetric examination is being considered by the USNRC to safeguard against this and other types of detrimental conditions. Factors addressed will include pipe diameter, wall-thickness, fusing temperature, interfacial pressure, dwell (open/close) time, and destructive correlation with ultrasonic data.

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.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Incorporation of Risk Insights in the Regulatory Treatment of Nuclear Power Plant Structures, Systems, and Components

2002 ◽  
Author(s):  
Eugene Imbro ◽  
Thomas G. Scarbrough
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Special Treatment ◽  
Regulatory Commission ◽  
High Level

The U.S. Nuclear Regulatory Commission (NRC) has established an initiative to risk-inform the requirements in Title 10 of the Code of Federal Regulations (10 CFR) for the regulatory treatment of structures, systems, and components (SSCs) used in commercial nuclear power plants. As discussed in several Commission papers (e.g., SECY-99-256 and SECY-00-0194), Option 2 of this initiative involves categorizing plant SSCs based on their safety significance, and specifying treatment that would provide an appropriate level of confidence in the capability of those SSCs to perform their design functions in accordance with their risk categorization. The NRC has initiated a rulemaking effort to allow licensees of nuclear power plants in the United States to implement the Option 2 approach in lieu of the “special treatment requirements” of the NRC regulations. In a proof-of-concept effort, the NRC recently granted exemptions from the special treatment requirements for safety-related SSCs categorized as having low risk significance by the licensee of the South Texas Project (STP) Units 1 and 2 nuclear power plant, based on a review of the licensee’s high-level objectives of the planned treatment for safety-related and high-risk nonsafety-related SSCs. This paper discusses the NRC staff’s views regarding the treatment of SSCs at STP described by the licensee in its updated Final Safety Analysis Report (FSAR) in support of the exemption request, and provides the status of rulemaking that would incorporate risk insights into the treatment of SSCs at nuclear power plants.

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.

Overview of Seismic Issues in Design Certification and Combined License Application Reviews

2009 ◽  
Author(s):  
Jim Xu ◽  
Sujit Samaddar
Keyword(s):  
Seismic Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Response Spectra ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Site Specific ◽  
License Application ◽  
Standard Design

The U.S. Nuclear Regulatory Commission (NRC) established a new process for licensing nuclear power plants under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” which provides requirements for early site permits (ESPs), standard design certifications (DCs), and combined license (COL) applications. In this process, an application for a COL may incorporate by reference a DC, an ESP, both, or neither. This approach allows for early resolution of safety and environmental issues. The COL review will not reconsider the safety issues resolved by the DC and ESP processes. However, a COL application that incorporates a DC by reference needs to demonstrate that pertinent site-specific parameters are confined within the safety envelopes established by the DC. This paper provides an overview of site parameters related to seismic designs and associated seismic issues encountered in DC and COL application reviews using the 10 CFR Part 52 process. Since DCs treat the seismic design and analysis of nuclear power plant (NPP) structures, systems, and components (SSC) as bounding to future potential sites, the design ground motions and associated site parameters are often conservatively specified, representing envelopes of site-specific seismic hazards and parameters. For a COL applicant to incorporate a DC by reference, it needs to demonstrate that the site-specific hazard in terms of ground motion response spectra (GMRS) is enveloped by the certified design response spectra of the DC. It also needs to demonstrate that the site-specific seismic parameters, such as foundation-bearing capacities, soil profiles, and the like, are confined within the site parameter envelopes established by the DC. For the noncertified portion of the plant SSCs, the COL applicant should perform the seismic design and analysis with respect to the site-specific GMRS and associated site parameters. This paper discusses the seismic issues encountered in the safety reviews of DC and COL applications. Practical issues dealing with comparing site-specific features to the standard designs and lessons learned are also discussed.

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