Surgery – Diagnostic testing – Detecting nuclear – electromagnetic – or ultrasonic radiation
Reexamination Certificate
1998-04-21
2001-01-16
Lateef, Marvin M. (Department: 3737)
Surgery
Diagnostic testing
Detecting nuclear, electromagnetic, or ultrasonic radiation
C128S920000, C378S065000
Reexamination Certificate
active
06175761
ABSTRACT:
BACKGROUND OF THE INVENTION
1. The Field of the Invention
The present invention relates generally to radiation therapy and specifically to the analytical computations for the dosimetric planning thereof. More specifically, the present invention relates to the macrodosimetry planning for specific radiotherapies such as neutron capture therapy, especially boron neutron capture therapy (BNCT), and fast neutron therapy having a BNCT component. Even more specifically, the present invention relates to methods and computer executable instructions for geometrically modeling a treatment volume planned for irradiation during various therapies and to calculating simulated particle transports through the model with simulation methods such as the Monte Carlo stochastic simulation method.
2. Copyrighted Materials
A portion of the disclosure of this patent document contains materials to which a claim of copyright protection is made. The copyright owner has no objection to the reproduction by anyone of the patent document or the patent disclosure as it appears in the Patent and Trademark Office patent file or records, but reserves all other rights with respect to the copyrighted work.
3. The Relevant Technology
Because significant improvements in radiobiological knowledge and encouraging human clinical results have recently been reported neutron radiotherapy modalities in the treatment of certain presently intractable malignancies are gradually becoming accepted practice.
Although application of neutrons for radiotherapy of cancer has been a subject of considerable clinical and research interest since the discovery of the neutron by Chadwick 1932, the earliest of clinical trials produced conspicuous failures. Since then, however, neutron radiotherapy has successfully evolved into a viable modality for treating inoperable salivary gland tumors and has emerged as a promising alternative for treating advanced prostate cancer, various lung tumors and certain other malignancies. Neutron capture therapy (NCT), a somewhat modified form of neutron radiotherapy, has entered clinical trials as a modality in treating glioblastoma multiforme and metastic malignant melanoma.
The basic physical process involved in past neutron therapy and neutron capture therapy differs in several respects. In fast neutron therapy, neutrons having relatively high energy (approximately 30 to 50 MeV) are generated by a suitable neutron source and used directly for irradiation of the treatment volume, as is done with standard photon (X-ray) therapy. In NCT, a neutron capture agent is introduced into a patient and is selectively taken into the malignant tissue. The administration of a pharmaceutical containing the neutron capture agent is preferably accomplished by a directed administration into the blood stream of the patient. At an appropriate time after administration of the neutron capture agent, the treatment volume (i. e., the anatomical structure to be treated) is exposed to a field of thermal neutrons produced by application of an external neutron beam. Because boron-10 has a large cross sectional area particularly suited for the capture of thermal neutrons (neutrons having energy less than 0.5 eV), boron-10 has preferentially become the capture agent of choice. Thus, the technology is commonly referred to as boron neutron capture therapy or BNCT.
BNCT is based on the nuclear reaction that occurs when boron 10, a non-radioactive isotope that accounts for approximately one-fifth of natural occurring boron, is irradiated with and absorbs or “captures” neutrons. Because the thermal neutrons that it captures are of such low energy they cause little tissue damage as compared with other forms of radiation such as protons, gamma rays, and fast neutrons. When an atom of boron 10 captures a neutron, an unstable isotope, boron 11, forms. The boron 11 instantly fissions yielding lithium-7 nuclei and energetic alpha particles (helium-4). The average total energy of this charged particle pair is about 2.35 MeV which is a highly lethal form of radiation. Yet, since these alpha particles have a path length of about ten microns (about 1 cell diameter), BNCT offers the advantage of cancer cell neutralization with only limited damage to nearby tissues. Moreover, BNCT does not rely on the capture of numerous neutrons because it only takes a few particles releasing their energy within a cancer cell to destroy it.
Another form of neutron radiotherapy is also the subject of current research interest. This other form is essentially a hybrid that combines the features of fast neutron therapy and NCT. In this type of therapy, a neutron capture agent is introduced into a patient, preferably into the malignant tissue only. This treatment is prior to the administration of standard fast neutron therapy. Because a small fraction of the neutrons in fast neutron therapy will be thermalized in the irradiation volume, it is possible to obtain a small incremental absorbed dose from the neutron capture interactions that result. Thus, based on current radiobiological research, improved tumor control appears to be likely when using this augmentation concept.
No matter which radiotherapy modality is used, some basic elements exist in the analytical computation of the macroscopic dosimetry, or macrodosimetry, thereof. With respect to BNCT therapy, for example, the essence of all analytical macrodosimetry is found by solving the three dimensional Boltzmann transport equation for neutral particles given certain descriptive information as input. This descriptive information consists generally of: (i) a geometric model of the irradiation volume; (ii) a mathematical description of the treatment neutron beam; and (iii) and a complete set of coupled neutron and photon interaction cross section data for all significant elements within the irradiation volume.
The geometric model of the patient and simulation of the irradiation will be discussed subsequently. As for the neutron beam, the description consists primarily of the neutron and photon spectra, angular distributions, and spatial intensity distributions in a defined plane of incidence (usually the exit port of the treatment beam collimator). This description is ordinarily constructed from a combination of calculational and experimental data pertinent to the beam of interest. Neutron and photon cross sections are typically taken from standard data collections such as the Evaluated Nuclear Data File (ENDF) and pre-processed into an appropriate multi-group or continuous-energy format.
Given the neutron and photon fluxes throughout the treatment volume (i.e., the complete space and energy-dependent solution of the Boltzmann equation for the situation of interest), the corresponding macroscopic absorbed dose distribution is ordinarily constructed by: (i) multiplying the calculated energy-dependent flux data by appropriate energy-dependent flux-to-dose conversion factors for each radiation dose component; (ii) integrating over the neutron or photon energy range as appropriate; and (iii) summing all components.
Two fundamentally different methods are available for computing the neutron and photon fluxes throughout the treatment volume. The first method is the Monte Carlo Stochastic Simulation Method which is presently preferred since the complex geometries that are characteristic of biological systems can be very accurately represented. The second method is a deterministic method based on direct numerical solution of the transport equation. Although the methods can be complimentary in terms of detailed spatial dose-distributions and dose-volume histogram information, both methods have their own appropriate uses, to which, only the stochastic methods will be used to describe the problem herein.
The stochastic methods for determining neutron and photon fluxes throughout the treatment volume are conceptually very simple, but the actual mathematical implementations for each particular method have attained high degrees of theoretical sophistication.
For BNCT purposes, the basic idea is to solve the fixed-source form of th
Frandsen Michael W.
Wessol Daniel E.
Wheeler Floyd J.
Bechtel BWXT Idaho LLC
Lateef Marvin M.
Shaw Shawna J
Workman & Nydegger & Seeley
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