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Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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Latest News
College students help develop waste-measuring device at Hanford
A partnership between Washington River Protection Solutions (WRPS) and Washington State University has resulted in the development of a device to measure radioactive and chemical tank waste at the Hanford Site. WRPS is the contractor at Hanford for the Department of Energy’s Office of Environmental Management.
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.