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Remembering Charles E. Till
Charles E. Till
Charles E. Till, an ANS member since 1963 and Fellow since 1987, passed away on March 22 at the age of 89. He earned bachelor’s and master’s degrees from the University of Saskatchewan and a Ph.D. in nuclear engineering from Imperial College, University of London. Till initially worked for the Civilian Atomic Power Department of the Canadian General Electric Company, where he was the physicist in charge of the startup of the first prototype CANDU reactor in Canada.
Till joined Argonne National Laboratory in 1963 in the Applied Physics Division, where he worked as an experimentalist in the Fast Critical Experiments program. He then moved to additional positions of increasing responsibility, becoming division director in 1973. Under his leadership, the Applied Physics Division established itself as one of the elite reactor physics organizations in the world. Both the experimental (critical experiments and nuclear data measurements) and nuclear analysis methods work were internationally recognized. Till led Argonne’s participation in the International Nuclear Fuel Cycle Evaluation (INFCE), and he was the lead U.S. delegate to INFCE Working Group 5, Fast Breeders.
Chad L. Pope, Michael J. Lineberry
Nuclear Technology | Volume 182 | Number 3 | June 2013 | Pages 335-348
Technical Paper | Radiation Transport and Protection/Radioisotopes | doi.org/10.13182/NT13-A16983
Articles are hosted by Taylor and Francis Online.
This paper compares measured results with simulation results of neutron beam transmission through an irradiated fuel assembly. The main objective of the comparison is to establish the technical foundation for using Monte Carlo simulation to evaluate the feasibility of using neutron computed tomography for irradiated fuel assembly inspection. The measured results were obtained from an irradiated fuel assembly from the Experimental Breeder Reactor II (EBR-II), and the neutron beam was produced by the Argonne National Laboratory Neutron Radiography Reactor (NRAD). The measurements consist of a projection profile representing the relative neutron beam attenuation at a specific fuel assembly axial elevation obtained from digitized neutron radiography film. Simulation of the neutron beam and fuel assembly was performed using the Monte Carlo code MCNP5. Results presented include the measured beam attenuation projection profile, simulated neutron beam attenuation projection profiles, parametric study of simulation results, and comparison of the projection results. Comparison of the radiography-based measurement with the simulation results shows good agreement, thereby confirming that Monte Carlo simulation of neutron transmission through an irradiated fuel assembly using MCNP5 is a reliable method for evaluating the use of neutron computed tomography as a means of inspecting irradiated fuel assemblies.