On the optimization of the isotopic neutron source method for measuring the thermal neutron absorption cross section: Advantages and disadvantages of BF3 and 3He counters

Bolewski, Andrzej; Ciechanowski, M.; Dydejczyk, A.; Kreft, A.

doi:10.1016/j.apradiso.2007.10.010

Cited by 16 publications

(8 citation statements)

References 7 publications

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“…The row format of ptrac file is complicated and very difficult to analyze for beginners; however, it contains the complete information of primary protons and all secondary particles. [3839404142]…”

Section: Methodsmentioning

confidence: 99%

A fast Monte Carlo code for proton transport in radiation therapy based on MCNPX

Jabbari

Seuntjens

2014

J Med Phys

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An important requirement for proton therapy is a software for dose calculation. Monte Carlo is the most accurate method for dose calculation, but it is very slow. In this work, a method is developed to improve the speed of dose calculation. The method is based on pre-generated tracks for particle transport. The MCNPX code has been used for generation of tracks. A set of data including the track of the particle was produced in each particular material (water, air, lung tissue, bone, and soft tissue). This code can transport protons in wide range of energies (up to 200 MeV for proton). The validity of the fast Monte Carlo (MC) code is evaluated with data MCNPX as a reference code. While analytical pencil beam algorithm transport shows great errors (up to 10%) near small high density heterogeneities, there was less than 2% deviation of MCNPX results in our dose calculation and isodose distribution. In terms of speed, the code runs 200 times faster than MCNPX. In the Fast MC code which is developed in this work, it takes the system less than 2 minutes to calculate dose for 106 particles in an Intel Core 2 Duo 2.66 GHZ desktop computer.

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Section: Methodsmentioning

confidence: 99%

A fast Monte Carlo code for proton transport in radiation therapy based on MCNPX

Jabbari

Seuntjens

2014

J Med Phys

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“…The problem is to compute the energy of recoiled nucleus, and it involves applying the neutron diffusion equation. 11,12 Therefore, the maximum rate of 2 1 E E (or on the other hand minimum lost energy) occurs whenever θ=0 and the maximum lost energy will occur in the case of .…”

Section: Applying Mathematical Methodsmentioning

confidence: 99%

“…The low LET of gamma rays is resulting from the capture of thermal neutrons with tissue and also hydrogen atoms [ 1 H(n,γ) 2 H]. 13 In the laboratory system, because of having collision between a neutron and a nucleus, the amount of energy transferred to a recoiled nucleus is computed with Eq.8: 11 1 2 ( )…”

Section: Applying Mathematical Methodsmentioning

confidence: 99%

Applying mathematical equations and computer programming to obtain transferred energy into body tissue

Shirazi

2018

OAJMTP

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In this paper, a mathematical method is investigated for therapy by fast neutron. This method can be applied for all tissues and investigations in connection with cancer therapy. The mathematical method is considered using equations and computer programming. The amounts of Kerma are calculated by this mathematical method in a liver tissue for a wide range of energies as a result of collision between neutrons with nuclei of constituent elements of tissue that have various mass numbers. This method can be applied to estimate the energy of neutrons reached the appropriate energy range through deducing the incident neutron energy emitted from clinical neutron source for a wide spectrum of neutron energies.

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“…The polyethylene sphere is surrounded by a graphite shell which has 25 cm radius and 5 mm thickness (according to moderation ratio: ξ(Σ s /Σ a )). This layer serves as a reflector to reduce escaping the fast neutrons [6].…”

Section: Introductionmentioning

confidence: 99%

Design and Simulation of a New Model for Treatment by NCT

Shirazi

Sardari

2012

Science and Technology of Nuclear Installations

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In this investigation, neutron capture therapy (NCT) through high energy neutrons using Monte Carlo method has been studied. In this study a new method of NCT for a sample liver phantom has been defined, and interaction of 12 MeV neutrons with a multilayer spherical phantom is considered. In order to reach the desirable energy range of neutrons in accord with required energy in absence of eligible clinical neutron source for NCT, this model of phantom might be utilized. The neutron flux and the deposited dose in the all components and different layers of the mentioned phantom are computed by Monte Carlo simulation. The results of Monte Carlo method are compared with analytical method results so that by using a computer program in Turbo-Pascal programming, the deposited dose in the liver phantom has been computed.

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On the optimization of the isotopic neutron source method for measuring the thermal neutron absorption cross section: Advantages and disadvantages of BF3 and 3He counters

Cited by 16 publications

References 7 publications

A fast Monte Carlo code for proton transport in radiation therapy based on MCNPX

A fast Monte Carlo code for proton transport in radiation therapy based on MCNPX

Applying mathematical equations and computer programming to obtain transferred energy into body tissue

Design and Simulation of a New Model for Treatment by NCT

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