Electromagnetic Conductivity and Magnetic Surveys for Characterization of an Abandoned Nuclear Reactor Facility

1996 ◽  
Author(s):  
Arthur J. Fromm ◽  
Robert W. Taylor ◽  
Kenneth Skinner ◽  
Joseph Litehiser ◽  
Paul Linley
Keyword(s):  
Nuclear Reactor ◽  
Reactor Facility

Electromagnetic Conductivity And Magnetic Surveys For Characterization Of An Abandoned Nuclear Reactor Facility

1996 ◽  
Author(s):  
Arthur J. Fromm ◽  
Robert W. Taylor ◽  
Kenneth Skinner ◽  
Joseph Litehiser ◽  
Paul Linley
Keyword(s):  
Nuclear Reactor ◽  
Reactor Facility

Determination of the Hydrogen Concentration in Zr-2.5Nb Alloy by a Resonant Ultrasound Spectroscopy

2006 ◽  
Vol 321-323 ◽  
pp. 1576-1579
Author(s):  
Yong Moo Cheong ◽  
Young Suk Kim
Keyword(s):  
Hydrogen Concentration ◽  
Zirconium Alloy ◽  
Nuclear Reactor ◽  
Zirconium Alloys ◽  
Resonant Ultrasound Spectroscopy ◽  
Heavy Water Reactor ◽  
Resonant Ultrasound ◽  
Non Destructive

Zirconium alloys are used for many applications in nuclear components, such as the pressure tube material in a pressurized heavy water reactor, nuclear fuel cladding, etc. One of the problems during the operation of a nuclear reactor is the degradation of the zirconium alloys, which is due to an increase of the hydrogen content in the zirconium alloy. Therefore a non-destructive determination of the hydrogen concentration in zirconium alloy is one of the important issues that need to be addressed. The resonant ultrasound spectroscopy (RUS) technique is evaluated for a characterization of the hydrogen concentration in Zr-2.5Nb alloy. Referring to the terminal solid solubility for dissolution (TSSD) of Zr-2.5Nb alloy, the plot of the mechanical damping coefficient (Q-1) versus the temperature or the deviation of the resonant frequency for the temperature (df/dT) versus the temperature was correlated for the hydrogen concentration in Zr-2.5Nb alloy. It was found that the temperature at an abrupt change of the slope can be correlated with the hydrogen concentration of the Zr-2.5Nb alloy.

Characterization of radiation hardened fibers in a research grade nuclear reactor

2021 ◽  
Author(s):  
Garry Berkovic ◽  
Shlomi Zilberman ◽  
Ehud Shafir ◽  
Yosef London ◽  
M. Dadon ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Radiation Hardened ◽  
Research Grade

Characterization of Changes in Properties and Microstructure of Glassy Polymeric Carbon Following Au Ion Irradiation

2009 ◽  
Vol 1215 ◽  
Author(s):  
Malek Amir Abunaemeh ◽  
Mohamed Seif ◽  
Young Yang ◽  
Lumin Wang ◽  
Yanbin Chen ◽  
...  
Keyword(s):  
Photoelectron Spectroscopy ◽  
Nuclear Reactor ◽  
Ion Irradiation ◽  
Pyrolytic Carbon ◽  
X Ray ◽  
Triso Fuel ◽  
Transmission Electron ◽  
Displacement Per Atom ◽  
Polymeric Carbon

AbstractThe TRISO fuel has been used in some of the Generation IV nuclear reactor designs [1,2]. It consists of a fuel kernel of UOx coated with several layers of materials with different functions. Pyrolytic carbon (PyC) is one of the materials in the layers. In this study we investigate the possibility of using Glassy Polymeric Carbon (GPC) as an alternative to PyC. In this work, we are comparing the changes in physical and microstructure properties of GPC after exposure to irradiation fluence of 5 MeV Au equivalent to a 1 displacement per atom (dpa) at samples prepared at 1000, 1500 and 2000°C. The GPC material is manufactured and tested at the Center for Irradiation Materials (CIM) at Alabama A&M University. Transmission electron microscopy (TEM), Rutherford backscattering spectroscopy (RBS), X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy were used for the analysis.

scholarly journals UNCERTAINTY CHARACTERIZATION FRAMEWORK FOR STEADY-STATE AND TRANSIENT NEUTRONICS SIMULATIONS OF A CANDU REACTOR

2021 ◽  
Vol 247 ◽  
pp. 15002
Author(s):  
Hany S. Abdel-Khalik ◽  
Alexandre Trottier ◽  
Dumitru Serghiuta ◽  
Dongli Huang
Keyword(s):  
Steady State ◽  
Cross Section ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Modeling Uncertainty ◽  
Group Format ◽  
Nuclear Reactor Core ◽  
Nodal Diffusion ◽  
Complete Framework

This paper reports on the development and testing of a comprehensive few-group cross section input uncertainty library for the NESTLE-C nodal diffusion-based nuclear reactor core simulator. This library represents the first milestone of a first-of-a-kind framework for the integrated characterization of uncertainties in steady-state and transient CANDU reactor simulations. The objective of this framework is to propagate, prioritize and devise a mapping capability for uncertainties in support of model validation of best-estimate calculations. A complete framework would factor both input and modeling uncertainty contributions. The scope of the present work is limited to the propagation of multi-group cross-section uncertainties through lattice physics calculations down to the few-group format, representing the input to the NESTLE-C core simulator, and finally to core responses of interest.

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