Experimental determination of the thermal neutron flux around two different types of high intensity 252Cf sources

Set-up and calibration of a triple ionization chamber system for dosimetry in mixed neutron/photon fields

The aim of this study is to introduce a triple ionization chamber system to separate dose components of mixed neutron/photon fields. Fast and thermal neutron dose components have a different biological effectiveness than gamma dose components. If boron neutron capture is used to enhance the dose in certain areas of a patient, the precise knowledge of the thermal neutron flux is essential. A tissue equivalent and two magnesium ionization chambers have been prepared for use in a triple chamber system for this purpose. One of the magnesium chambers is coated with (10)B on the inside to enhance its response to thermal neutrons. All three chambers have been calibrated at a cobalt source, medical linear accelerators and several neutron sources. The chambers have been studied in Monte Carlo simulations and the results are compared with measurements.

Measured Microdosimetric Spectra and Therapeutic Potential of Boron Neutron Capture Enhancement of252Cf Brachytherapy

Californium-252 is a neutron-emitting radioisotope used as a brachytherapy source for radioresistant tumors. Presented here are microdosimetric spectra measured as a function of simulated site diameter and distance from applicator tube 252Cf sources. These spectra were measured using miniature tissue-equivalent proportional counters (TEPCs). An investigation of the clinical potential of boron neutron capture (BNC) enhancement of 252Cf brachytherapy is also provided. The absorbed dose from the BNC reaction was measured using a boron-loaded miniature TEPC. Measured neutron, photon and BNC absorbed dose components are provided as a function of distance from the source. In general, the absorbed dose results show good agreement with results from other measurement techniques. A concomitant boost to 252Cf brachytherapy may be provided through the use of the BNC reaction. The potential magnitude of this BNC enhancement increases with increasing distance from the source and is capable of providing a therapeutic gain greater than 30% at a distance of 5 cm from the source, assuming currently achievable boron concentrations.

Miniature tissue-equivalent proportional counters for BNCT and BNCEFNT dosimetry

A dual miniature tissue-equivalent proportional counter (TEPC) system has been developed to facilitate microdosimetry for Boron Neutron Capture Therapy (BNCT). This system has been designed specifically to allow the analysis of the single event charged particle spectrum in phantom in high intensity BNCT beams and to provide this microdosimetric information with excellent spatial resolution. Paired A-150 and 10B-loaded A-150 TEPCs with 12.3 mm3 collecting volumes have been constructed. These TEPCs allow more accurate neutron dosimetry than current techniques, offer a direct measure of the boron neutron capture dose, and provide a framework for predicting the biological effectiveness of the absorbed dose. Design aspects and characterization of these detectors are reviewed, along with an exposition of the advantages of microdosimetry using these detectors over conventional dosimetry methods. In addition, the utility of this technique for boron neutron capture enhancement of fast neutron therapy (BNCEFNT) is discussed.

Neutron dosimetry for a general 252Cf brachytherapy source

This paper extends previous work to characterize neutron dosimetry in the vicinity of 252Cf brachytherapy sources. A general source is examined with an arbitrary length, diameter, and encapsulation using Monte Carlo methods. Fast neutron dosimetry and thermal neutron fluence rates were determined in a variety of clinically relevant media of varying dimensions. Applicator Tube, point source, high dose rate VariSource, and high dose rate muSelectron source geometries were analyzed. Fast neutron dosimetry was relatively independent of encapsulation thickness for an assortment of encapsulation materials less than 2 mm thick. Large variations in phantom size made minimal differences in the fast neutron dose close to the source. Specific source geometries were compared with dosimetry obtained from a simplified point model. The consequence of these results is a convenient means of accurately predicting clinical fast neutron dosimetry characteristics around a general 252Cf brachytherapy source in a variety of media without requiring neutron transport. Thermal neutron fluence rates were determined for a variety of source encapsulation materials, encapsulation thicknesses, and phantom sizes. At a distance of 3 cm from the source center, the thermal neutron fluence rate for a 30 cm diameter phantom was a 2.65 times greater than for a 10 cm diameter water phantom. These results demonstrate 252Cf thermal neutron fluence rate is relatively independent of encapsulation thickness and composition, yet highly dependent on hydrogen mass density and phantom size for phanta with diameters <30 cm.

Measurements and calculations of thermal neutron fluence rate and neutron energy spectra resulting from moderation of 252Cf fast neutrons: applications for neutron capture therapy

252Cf is a neutron emitting radioisotope which has promise for both standard brachytherapy and neutron capture enhanced brachytherapy. In this study, experimental measurements and calculations were used to determine the thermal neutron fluence rate, phi(th) [n cm(-2) s(-1) mg(-1)], in the vicinity of 252Cf applicator tube (AT) type sources. Results of these measurements were confirmed with Monte Carlo calculations performed in a distributed manner on multiple workstations using MCNP. Three studies were executed: (1) relative phi(th) as a function of distance from a 252Cf AT source in an A-150 tissue equivalent plastic phantom using thermoluminescent dosimeters (TLDs) of varying 6Li/Li enrichment, (2) phi(th) measured with gold foils in a 114 liter water phantom 5 cm from two 252Cf AT sources, and (3) calculations of the impact of phantom material composition (e.g., A-150, water, brain, muscle) on phi(th) from moderated 252Cf fast neutrons. TLD results and Monte Carlo calculations in A-150 of relative phi(th) typically agreed within 1% and at most differed by 3% for distances from 1 to 6 cm. Foil measurements followed the ASTM E 262-86e protocol, and the ratio of activated plain and Cd encased gold foils (7.31) agreed well with the calculated ratio (7.26). Measured phi(th) at 5 cm (1.70+/-0.10 x 10(7) n cm(-2) s(-1) mg(-1)) was 10% greater than that determined using MCNP (1.55+/-0.12 x 10(7) n cm(-2) s(-1) mg(-1)), but was within the combined uncertainties. Compared with A-150 at a distance of 1 cm, phi(th) was 20%, 22%, and 32% less for water, brain, and muscle, respectively; these ratios decreased to 16%, 16%, and 24% less, respectively, at a distance of 5 cm from the source in a 15 cm diameter phantom. Comparisons of these results generally agreed with those in the literature for a value of 2 x 10(7) n cm(-2) s(-1) mg(-1) in water at 3 cm.

Paired Mg and Mg(B) ionization chambers for the measurement of boron neutron capture dose in neutron beams

The use of the boron neutron capture (BNC) reaction to provide a dose enhancement in fast neutron therapy is currently under investigation at the Gershenson Radiation Oncology Center of Harper Hospital in Detroit, MI. The implementation of this treatment modality presents unique challenges in dosimetry. In addition to the measurement of photon and neutron doses in the mixed field, a measure of the thermal neutron flux and the associated boron neutron capture dose throughout the treatment volume is desired. A pair of small-volume magnesium ionization chambers has been constructed with the aim of providing this information. One of the chambers, denoted the Mg(B) chamber, is lined with a boron-loaded foil. The ionization response of this chamber has been calibrated in terms of BNC dose per ppm loading of 10B. These paired chambers can be used to map the local BNC response in neutron beams. From this data and an estimation of the boron concentration in the tumor and normal tissue, the boron neutron capture enhancement may be evaluated.

Fission converter and metal-oxide-semiconductor field effect transistor study of thermal neutron flux distribution in an epithermal neutron therapy beam

The depth distribution of the thermal neutron flux is a major factor in boron neutron capture therapy (BNCT) in determining the efficiency of cell sterilization. In this paper the fission detector method is developed and applied to measure the in-phantom thermal neutron flux depth distribution. Advantages of the fission detector include small size, direct measurement of thermal neutron flux in a mixed radiation field of BNCT beam, self-calibration, and the possibility of on-line measurement. The measurements were performed at epithermal a BNCT facility. The experimental results were compared with the thermal neutron flux calculated by the Monte Carlo method and found to be in good agreement.