SYMPOSIA PAPER Published: 01 January 1987
STP25691S

Comparison of Calculated Integral Values Using Measured and Calculated Neutron Spectra for Fusion Neutronics Analyses

Source

The kerma heat production density, tritium production density, and dose in a lithiumfluoride pile with a deuterium-tritium neutron source were calculated with a data processing code, UFO, from the pulse height distribution of a miniature NE213 neutron spectrometer, and compared with the values calculated with a Monte Carlo code, MORSE-CV. Both of the UFO and MORSE-CV values agreed within the statistical error (less than 6%) of the MORSE-CV calculations, except for the outer-most point in the pile. The MORSE-CV values were slightly smaller than the UFO values for almost all cases, and this tendency increased with increasing distance from the neutron source.

Author Information

Sekimoto, H
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo, Japan
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Details
Developed by Committee: E10
Pages: 761–768
DOI: 10.1520/STP25691S
ISBN-EB: 978-0-8031-5017-1
ISBN-13: 978-0-8031-0963-6