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Accuracy of ultrasonic flaw sizing techniques for reactor pressure vessels: Final report

TLDR
In this article, the accuracy of several ultrasonic flaw sizing techniques was measured using thick-walled mock-ups simulating typical reactor pressure vessel weld and clad configurations, which contained intentional flaws representing three major classes of defects: cracks under the cladding, embedded defects in welds, and lack-of-fusion defects in nozzle-to-shell welds.
Abstract
The accuracy of several ultrasonic flaw sizing techniques was measured using thick-walled mock-ups simulating typical reactor pressure vessel weld and clad configurations. The evaluated techniques were: backward-scattering tip-diffraction, time-of-flight diffraction (TOFD), dB-drop (with three different amplitude thresholds), and large-diameter focused probes. For the amplitude-based dB-drop technique, amplitude thresholds of 6 dB below peak, 50 percent DAC, and 20 percent DAC were used, and the effect of correcting for the spreading of the ultrasonic beam was also evaluated. The mock-ups contained intentional flaws representing three major classes of defects: cracks under the cladding, embedded defects in welds, and lack-of-fusion defects in nozzle-to-shell welds. Approximately 200 measurements were made. The results showed that for sizing near-surface and embedded defects in thick welds, TOFD and the backward-scattering tip-diffraction techniques were far more accurate than the amplitude-based techniques. Beam-spread corrections reduced the mean sizing errors for the 20 percent DAC sizing method, but made negligible reduction in the scatter of the data. Even after beam-spread corrections were made, the sizing errors of the amplitude-based method remained much greater than the tip-diffraction measurement error. For small-bore nozzles, focused probe measurements were substantially more accurate than unfocused probe measurements. 21 refs., 43 figs., 2 tabs.

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ReportDOI

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

TL;DR: In this paper, a very high temperature Reactor (VHTR) with helium as the coolant has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases and the U.S. Department of Energy has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years.
ReportDOI

Evaluations of Structural Failure Probabilities and Candidate Inservice Inspection Programs

TL;DR: In this paper, the authors apply probabilistic structural mechanics models to predict the reliability of nuclear pressure boundary components and evaluate the effectiveness of alternative programs for inservice inspection to reduce these failure probabilities.
References
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ReportDOI

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

TL;DR: In this paper, a very high temperature Reactor (VHTR) with helium as the coolant has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases and the U.S. Department of Energy has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years.
ReportDOI

Evaluations of Structural Failure Probabilities and Candidate Inservice Inspection Programs

TL;DR: In this paper, the authors apply probabilistic structural mechanics models to predict the reliability of nuclear pressure boundary components and evaluate the effectiveness of alternative programs for inservice inspection to reduce these failure probabilities.
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