Dosimetry for californium-252 (252Cf) neutron-emitting brachytherapy sources and encapsulation, storage, and clinical delivery thereof

Abstract

The present invention discloses a methodology for the characterization and determination of mixed-field dosimetry for 252 Cf Applicator Tube (AT)-type medical sources, utilizing ionization chambers, GM counters, and Monte Carlo methods. Unlike the previous methodologies, the present invention discloses a specification of dose to muscle, rather than dose to water, for clinical dosimetry of 252 Cf medical sources. A dosimetry protocol, similar to that utilized for ICRU-45, with parameters determined specifically for 252 Cf brachytherapy is disclosed. Neutron isodose distributions and data necessary for clinical implementation of 252 Cf AT sources are also disclosed herein. Additionally, novel methods for the encapsulation, storage, and delivery/implantation of 252 Cf radionuclide sources are disclosed.

1 ) A method for the computation of neutron isodose distributions for a neutron-emitting radionuclide, by use of LINUX-based computer system, and wherein the values for the absorbed dose are relative to muscle, rather than water.  
   
   
       2 ) A method for the computation of low-LET photon and electron isodose distributions for a californium-252 by use of LINUX-based computer system, and wherein the values for the absorbed dose are relative to water as suggested by AAPM TG-43 or to muscle.  
   
   
       3 ) A method of calculating a biologically-effective dose of a neutron emitting radionuclide for use in brachytherapy by use of a modified ICRU 45 Protocol.  
   
   
       4 ) The method of  claim 3 , wherein the modification to said ICRU 45 Protocol comprises: using a mixed neutron-photon radiation field produced by the neutron-emitting radionuclide; 
 (a) calculating the neutron kerma for neutron-emitting radionuclides in materials by use of a networked parallel computer-based system, wherein said materials are selected from the group consisting of: the synthetic tissue-equivalent (TE) A-150 plastic, PMMA, brain, muscle, fat, pancreas, lung, bone, skin, and blood, and    (b) computing the neutron isodose distributions for a neutron-emitting radionuclide, by use of LINUX-based computer computation, and wherein the values of the absorbed dose are relative to muscle, rather than water.    
   
   
       5 ) The method of  claim 3 , wherein calculation of said biologically-effective dose comprises the following steps: 
 (a) measuring the total dose by use of a TE ion chamber, possessing equal detection sensitivity to both neutrons and photons;    (b) measuring the photon dose by use of a miniature Geiger-Muller (GM) counter, possessing a decreased detection sensitivity to neutrons in comparison to photons; and    (c) determining the neutron absorbed dose by computation using the photon dose measured in step (b) and said total dose measured in step (a).    
   
   
       6 ) A device for the encapsulation of radionuclides for use in brachytherapy, comprising: 
 a flexible wire, essentially circular in cross-section, having a first and second end, and an original configuration, said flexible wire being constructed of a material that is chemically- and thermally-stable to both neutron and gamma radiation and which can withstand flexation without permanent alteration in said flexible wire's original configuration;    an internal cavity with a contiguous inner surface located within said second end of said flexible wire; and    a means for emitting radioactivity, said radioactive means localized within said internal cavity of said flexible wire, wherein said second end of said flexible wire is sealed to prevent the release of said radioactive means.    
   
   
       7 ) The device of  claim 6 , wherein said radioactive means is a neutron-emitting radionuclide.  
   
   
       8 ) The device of  claim 6 , wherein said radioactive means is the neutron-emitting radionuclide californium-252.  
   
   
       9 ) The device of  claim 8 , wherein said californium-252 source is in the form of californium-252-containing californium oxide (cermet).  
   
   
       10 ) The device of claims  8  or  9 , wherein the concentration of said californium-252 source ranges from approximately 1 μg to 10 mg.  
   
   
       11 ) The device of  claim 6 , wherein said flexible wire is fabricated from a material selected from the group consisting of: titanium, stainless steel, and nickel/titanium alloy.  
   
   
       12 ) The device of  claim 6 , wherein the shape and total dimensions of said flexible wire is such that it is capable of passing through the interior of a hollow tube or needle, and wherein the interior diameter of said hollow tube or needle ranges from approximately 20 gauge to approximately 12 gauge.  
   
   
       13 ) The device of  claim 6  further comprising a hollow, generally-cylindrical, elongated tube having an exterior surface, an interior surface, and a first and second end, and an original configuration, said elongated tube constructed of a material that can withstand flexation without permanent alteration in said elongated tube's original configuration, and wherein said elongated tube has an internal diameter sufficient to allow passage of said radioactive means sealed within said flexible wire therethrough.  
   
   
       14 ) A method of brachytherapy treatment comprising administration of an encapsulated radionuclide source to an individual in need thereof, said encapsulated radionuclide source comprising: 
 a flexible wire, having a first and second end, and an original configuration, said flexible wire being constructed of a material that is chemically- and thermally-stable to both neutron and gamma radiation and which can withstand flexation without permanent alteration in said flexible wire's original configuration;    an internal cavity with a contiguous inner surface located within said second end of said flexible wire; and    a means for emitting radioactivity, said radioactive means localized within said internal cavity of said flexible wire, wherein said second end of said flexible wire is sealed to prevent the release of said radioactive means.    
   
   
       15 ) The method of  claim 14 , wherein said radioactive means is a neutron-emitting radionuclide.  
   
   
       16 ) The method of  claim 14 , wherein said radioactive means is the neutron-emitting radionuclide californium-252.  
   
   
       17 ) The method of  claim 14 , wherein said californium-252 source is in the form of californium-252-containing californium oxide (cermet).  
   
   
       18 ) The method of claims  16  or  17 , wherein the concentration of said californium-252 source ranges from approximately 1 μg to 10 mg.  
   
   
       19 ) The method of  claim 14 , wherein said flexible wire is fabricated from a material selected from the group consisting of: titanium, stainless steel, and nickel/titanium alloy.  
   
   
       20 ) The method of  claim 14 , wherein the shape and total dimensions of said flexible wire is such that it is capable of passing through the interior of a hollow tube or needle, and wherein the interior diameter of said hollow tube or needle ranges from approximately 20 gauge to approximately 12 gauge.  
   
   
       21 ) The method of  claim 16  further comprising a hollow, generally-cylindrical, elongated tube having an exterior surface, an interior surface, and a first and second end, and an original configuration, said elongated tube constructed of a material that can withstand flexation without permanent alteration in said elongated tube's original configuration, and wherein said elongated tube has an internal diameter sufficient to allow passage of said radioactive means sealed within said flexible wire therethrough.  
   
   
       22 ) A device for the storage and containment of a radionuclide source used in brachytherapy, said device comprising: 
 an inner lead container of sufficient area to completely enclose said radionuclide source;    an outer container, having an inner and outer wall, fabricated from an alloy which is thermally- and chemically-inert to both neutron and gamma radiation, wherein said inner and outer walls of said outer container form an internal cavity, said cavity being completely filled with borated polystyrene or an aqueous, neutron-attenuating solution such as a saturated water solution of boric acid;    sheets of hydrogenous material such as polyethylene, borated-polyethylene, polystyrene, polyester, water-extended polyester, acrylic, nylon or rubber of sufficient size and shape so as to completely enclose said outer container in all dimensions; and    outer sheets of high-Z material such as lead, iron, stainless steel, tungsten, bismuth, or depleted uranium sufficient size and shape so as to completely enclose said hydrogenous material-enclosure, outer container in all dimensions;    wherein one surface of said inner high-Z material container, one surface of said outer container, one surface of said hydrogenous material, and one surface of said outer high-Z material possesses an aperture, said apertures being arranged in such a manner so as to form a continuous, generally cylindrical passageway of sufficient diameter to allow passage of a hollow, elongated tube, generally cylindrical in diameter, said elongated hollow tube having an exterior surface, an interior surface, and a first and second end, wherein said first end terminates within said inner lead container and said second end terminates external to the outer lead sheets.    
   
   
       23 ) The device of  claim 22 , wherein said hydrogenous material is at-least approximately 10 cm thick in cross-section and said high-Z material is at-least approximately 10 mm thick in cross-section.  
   
   
       24 ) The device of  claim 22 , wherein said outer container is fabricated from a stainless steel alloy.  
   
   
       25 ) The device of  claim 22 , wherein said aqueous, neutron-attenuating solution completely filling said internal cavity of the outer container is a saturated solution of boric acid or borated polystyrene.  
   
   
       26 ) The device of  claim 22 , which further comprising a stepping motor.  
   
   
       27 ) A device for the implantation of a radionuclide source into a patient in need thereof, said device comprising a shielded, stationary enclosure, which functions to contain the radioactive emissions produced by said radionuclide source, possessing an aperture through which is passed a hollow, elongated, and flexible catheter, internally-containing said radionuclide source attached to a flexible wire; wherein said wire-attached radionuclide source is transported through the internal cavity of said catheter to the selected site of implantation within the patient by the action of a stepping motor.  
   
   
       28 ) A means of calculating dosimetric enhancement by materials possessing relatively high neutron capture cross-sections such as  10 B,  157 Gd,  3 He,  133 Xe, or  135 Xe defined herein as neutron capture therapy (NCT) agents added to enhance the clinical application of brachytherapy using radioactive sources.  
   
   
       29 ) The method of  claim 28 , wherein said radioactive means is a neutron-emitting radionuclide.  
   
   
       30 ) The method of  claim 28 , wherein said radioactive means is the neutron-emitting radionuclide californium-252.  
   
   
       31 ) The method of  claim 28 , wherein said calculations are performed using Monte Carlo calculations using said methodology.  
   
   
       32 ) A means of administering said NCT agents where their presence enhances clinical outcomes compared to those obtained with only a radioactive brachytherapy source present.  
   
   
       33 ) The method of  claim 32 , wherein said radioactive means is a neutron-emitting radionuclide.  
   
   
       34 ) The method of  claim 32 , wherein said radioactive means is the neutron-emitting radionuclide californium-252.