_version_ 1826216056561074176
author Olson, Arne Peter
author2 Massachusetts Institute of Technology. Department of Nuclear Engineering
author_facet Massachusetts Institute of Technology. Department of Nuclear Engineering
Olson, Arne Peter
author_sort Olson, Arne Peter
collection MIT
description "August 1967."
first_indexed 2024-09-23T16:41:37Z
format Technical Report
id mit-1721.1/89691
institution Massachusetts Institute of Technology
last_indexed 2024-09-23T16:41:37Z
publishDate 2014
publisher Cambridge, Mass. : Massachusetts Institute of Technology, Dept. of Nuclear Engineering, [1967]
record_format dspace
spelling mit-1721.1/896912019-04-10T10:15:46Z Computer simulation of neutron capture therapy Olson, Arne Peter Massachusetts Institute of Technology. Department of Nuclear Engineering Massachusetts General Hospital. Physics Research Laboratory TK9008.M41 N96 no.83 Neutrons -- Capture -- Computer simulation "August 1967." "Prepared for Physics Research Laboratory Massachusetts General Hospital Boston, Massachusetts." Also issued as an Sc. D. thesis, MIT, Dept. of Nuclear Engineering, 1967 Includes bibliographical references (pages 340-343) Analytical methods are developed to simulate on a large digital computer the production and use of reactor neutron beams f or boron capture therapy of brain tumors. The simulation accounts for radiation dose distributions in tissue produced by fast neutrons and by neutron capture reaction products such as gamma rays, C -particles, protons, and heavy particles. These techniques are applied to optimize the effectiveness of the M.I.T. Reactor Medical Therapy Facility through a survey of the effects of neutron filters and of modifications to the beam collimation system. Neutron beams reflected from thin slabs of hydrogenous materials are shown to have an improved ability to effectively irradiate a deep tumor without destroying normal tissue above it because relatively few fast neutrons are reflected. Considerable improvements in thermal neutron distribution in tissue are shown to result from surrounding the head with a neutron-reflecting annulus to reduce lateral neutron leakage. A new numerical solution is obtained for the problem of neutron transport in finite thickness slabs with isotropic scattering. Gaussian quadratures are used to evaluate the neutron transport integral equations, yielding transmission, absorption, and reflection probabilities, and fluxes, as a function of collision number. Collision history correlations are devised which use only five paraeters to predict the fate of neutrons incident on an infinite slab having arbitrary thickness and neutron cross sections. A very fast multigroup neutron spectrum calculation is developed by combining collision history correlations with single-collision group transfer probabilities to directly obtain transmission and reflection matrices for multi-slab shielding problems. 2014-09-16T23:30:04Z 2014-09-16T23:30:04Z 1967 Technical Report http://hdl.handle.net/1721.1/89691 856017412 MITNE ; no. 83 343 pages application/pdf Cambridge, Mass. : Massachusetts Institute of Technology, Dept. of Nuclear Engineering, [1967]
spellingShingle TK9008.M41 N96 no.83
Neutrons -- Capture -- Computer simulation
Olson, Arne Peter
Computer simulation of neutron capture therapy
title Computer simulation of neutron capture therapy
title_full Computer simulation of neutron capture therapy
title_fullStr Computer simulation of neutron capture therapy
title_full_unstemmed Computer simulation of neutron capture therapy
title_short Computer simulation of neutron capture therapy
title_sort computer simulation of neutron capture therapy
topic TK9008.M41 N96 no.83
Neutrons -- Capture -- Computer simulation
url http://hdl.handle.net/1721.1/89691
work_keys_str_mv AT olsonarnepeter computersimulationofneutroncapturetherapy