George J. Schuster

A system for high-resolution, nondestructive, ultrasonic imaging of weld grains

The purpose of nondestructive evaluation is to detect degradation so that corrective action can be taken before the degradation challenges the structural integrity of an industrial system or one of its components. Accurate characterization is required to distinguish progressive degradation from benign material conditions. In nondestructive evaluation, characterization includes quantification and description of location, dimensions, shape, orientation, and composition of an indication of degradation. An imaging system that uses synthetic aperture focusing is one choice for detection and characterization of degradation in welded assemblies. In this paper, the ultrasonic imaging of the intended weld microstructure is reported. New technology invented for this purpose is described. This paper reviews how an ultrasonic imaging system that uses a synthetic lens can have a resolution that approaches the diffraction limit. A constrained solution to the coherent summation problem is presented for near real-time performance in high-resolution synthetic aperture focusing. Data are included to show that nondestructive, ultrasonic imaging of weld grains is practical.

Developing a generalized flaw distribution for reactor pressure vessels

The US Nuclear Regulatory Commission (NRC) is re-evaluating the guidance and criteria in the code of federal regulations as it relates to reactor vessel integrity, specifically pressurized thermal shock (PTS). Recent ultrasonic examination of considerable vessel material at Pacific Northwest National Laboratory (PNNL) and industry experiences with Yankee Rowe have provided the NRC with a better understanding of PTS issues. The re-evaluation of PTS will consider a risk-informed approach to the PTS rule and also provide important benefits for licensees considering license renewal. Pressurized thermal shock transients can lead to reactor vessel failure. These transients have occurred at operating reactors but, to date, they have not resulted in vessel failure. To properly determine the potential or probability for vessel failure from a PTS event, an accurate estimate of fabrication flaws is necessary. The characteristics of the fabrication flaw are inputs to fracture mechanics structural calculations that will determine the probability of vessel failure during a PTS event. Also, the results will indicate the sizes and locations of flaws that are most likely to cause failures. This information is also an integral input to the overall pressure vessel safety program. In order to obtain an accurate estimate of fabrication flaws to address PTS events for all classes of reactors, a generic flaw distribution must be developed. An expert judgment process will be used in conjunction with empirical data from PNNL, reactor pressure vessel studies and modeling (RR- PRODIGAL Code) in developing generalized flaw distributions. This paper will demonstrate the important relationship between reactor vessel integrity and flaw distributions in reactor pressure vessel material, discuss the PNNL work to date on developing flaw density and distributions for domestic RPVs, and describe the expert judgment process that was used to verify that a generalized flaw distribution can be properly developed and then assist in developing a generalized flaw distribution.

Distributions of Fabrication Flaws in Reactor Pressure Vessels for Structural Integrity Evaluations

To reduce uncertainties in flaw-related inputs for probabilistic fracture mechanics (PFM) evaluations, the U.S. Nuclear Regulatory Commission (USNRC) has supported research at Pacific Northwest National Laboratory (PNNL) involving nondestructive and destructive examinations for fabrication flaws in reactor pressure vessel (RPV) material. Using these data, statistical distributions have been developed to characterize the flaws in regions of a RPV. The regions include the main seam welds, repair welds, base metal, and the cladding at the inner surface of the vessel. This paper summarizes the available data and describes the treatment of these data to estimate flaw densities, flaw depth distributions, and flaw aspect ratio distributions. The methodology has generated flaw-related inputs for PFM calculations that have been part of an effort to update pressurized thermal shock (PTS) regulations. Statistical treatments of uncertainties in the parameters of the flaw distribution functions are part of the inputs to the PFM calculations. The paper concludes with a presentation of some example input files that have supported evaluations by USNRC of the risk of vessel failures caused by PTS events.Copyright

Development of a Remotely Operated NDE System for Inspection of Hanford's Double Shell Waste Tank Knuckle Regions

This report documents work performed at the PNNL in FY01 to support development of a Remotely Operated NDE (RONDE) system capable of inspecting the knuckle region of Hanfords DSTs. The development effort utilized commercial off-the-shelf (COTS) technology wherever possible and provided a transport and scanning device for implementing the SAFT and T-SAFT techniques.

Density and distribution of fabrication flaws in the Shoreham reactor vessel

The Pacific Northwest National Laboratory (PNNL) has been conducting a multi-year program for the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research on the effectiveness of NDE for inspection of nuclear power plants. One task of this program concerns the development of generic flaw density and distribution functions for fabrication flaws in reactor pressure vessels. Reactor pressure vessel material from the cancelled Shoreham nuclear power plant was obtained as a part of a joint program between the NRC, Baltimore Gas and Electric and the Electric Power Research Institute. PNNL has conducted NDE inspections of this material and has estimated the density and distributions of fabrication flaws in this material. This paper discusses these inspections and the results of analyzing this inspection data. More than 4,000 fabrication flaws were detected and are described in this paper.

SAFT inspections for developing empirical database of fabrication flaws in nuclear reactor pressure vessels

The Pacific Northwest National Laboratory (PNNL) is developing a methodology for estimating the size and density distribution of fabrication flaws in U.S. nuclear reactor pressure vessels. This involves the nondestructive evaluation (NDE) of reactor pressure vessel materials and the destructive validation of the flaws found. NDE has been performed on reactor pressure vessel material made by Babcock & Wilcox and Combustion Engineering. A metallographic analysis is being performed to validate the flaw density and size distributions estimated from the 2500 indications of fabrication flaws that were detected and characterized in the very sensitive SAFT-UT (synthetic aperture focusing technique for ultrasonic testing) inspection data from the Pressure Vessel Research User Facility (PVRUF) vessel at Oak Ridge National Laboratory. Research plans are also described for expanding the work to include other reactor pressure vessel materials.

NDE Assessment of PWSCC in Control Rod Drive Mechanism Housings

Studies being conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington are focused on assessing the effectiveness of nondestructive examination (NDE) techniques for inspecting control rod drive mechanism (CRDM) nozzles and J-groove weldments. The primary objective of this work is to provide information to the United States Nuclear Regulatory Commission (US NRC) on the effectiveness of NDE methods as related to the in-service inspection of CRDM nozzles and J-groove weldments, and to enhance the knowledge base of primary water stress corrosion cracking (PWSCC) through destructive characterization of the CRDM assemblies. In describing two CRDM assemblies removed from service, decontaminated, and then used in a series of NDE measurements, this paper will address the following questions: 1) What did each technique detect?, 2) What did each technique miss?, and 3) How accurately did each technique characterize the detected flaws? Two CRDM assemblies including the CRDM nozzle, the J-groove weld, buttering, and a portion of the ferritic head material were selected for this study. One contained suspected PWSCC, based on in-service inspection data and through-wall leakage; the other contained evidence suggesting through-wall leakage, but this was unconfirmed. The two CRDMs used in this study were cut from a pressure vessel head that has since been replaced. The selected NDE measurements follow standard industry techniques for conducting in-service inspections of CRDM nozzles and the crown of the J-groove welds and buttering. In addition, laboratory based NDE methods were employed to conduct inspections of the CRDM assemblies, with particular emphasis on inspecting the J-groove weld and buttering. This paper will also describe the NDE methods used and discuss the NDE results. Future work will involve using the results from these NDE studies to guide the development of a destructive characterization plan to reveal the crack morphology and a comparison of the degradation found by the destructive evaluation with the recorded NDE responses.© 2006 ASME

Effects of Weld Repairs on Calculated Failure Probabilities for Reactor Pressure Vessels and Piping

Section III of the ASME Boiler and Pressure Vessel Code for nuclear power plant components requires radiographic examinations (RT) of welds and repairs for RT indications that exceed code acceptable sizes. This paper describes research that generated data on welding flaws, which indicated that the largest flaws occur in repaired welds. Evidence suggests that repairs are often for small and benign RT indications at locations buried within the vessel or pipe wall. Probabilistic fracture mechanics calculations are described in this paper to predict the increases in vessel and piping failure probabilities caused by the repair-induced flaws. Calculations address failures of embrittled vessel welds for pressurized thermal shock (PTS) transients. In this case small flaws, which are relatively common, can cause brittle fracture such that the rarely encountered repair flaws of large sizes gave only modestly increased failure probabilities. The piping calculations simulated fatigue crack growth of fabrication flaws. These calculations showed that only relatively large fabrication flaws can fail piping because of the ductile nature of the piping material. The large repair flaws significantly increased the failure probabilities. The paper recommends the use of more discriminating UT examinations in place of RT examinations along with repair criteria based on a fitness-for-purpose approach that minimizes the number of unjustified repairs.© 2003 ASME