Microdosimetric spectra of the THOR neutron beam for boron neutron capture therapy

A primary objective of the BNCT project in Taiwan, involving THOR (Tsing Hua Open Pool Reactor), was to examine the potential treatment of hepatoma. To characterise the epithermal neutron beam in THOR, the microdosimetry distributions in lineal energy were determined using paired tissue-equivalent proportional counters with and without boron microfoils. Microdosimetry results were obtained in free-air and at various depths in a PMMA phantom near the exit of the beam port. A biological weighting function, dependent on lineal energy, was used to estimate the relative biological effectiveness of the beam. An effective RBE of 2.7 was found at several depths in the phantom.     

Nanodosimetric measurements and calculations in a neutron therapy beam

A comparison of calculated and measured values of the dose mean lineal energy (y(D)) for the former neutron therapy beam at Louvain-la-Neuve is reported. The measurements were made with wall-less tissue-equivalent proportional counters using the variance-covariance method and simulating spheres with diameters between 10 nm and 15 microm. The calculated y(D)-values were obtained from simulated energy distributions of neutrons and charged particles inside an A-150 phantom and from published y(D)-values for mono-energetic ions. The energy distributions of charged particles up to oxygen were determined with the SHIELD-HIT code using an MCNPX simulated neutron spectrum as an input. The mono-energetic ion y(D)-values in the range 3-100 nm were taken from track-structure simulations in water vapour done with PITS/KURBUC. The large influence on the dose mean lineal energy from the light ion (A > 4) absorbed dose fraction, may explain an observed difference between experiment and calculation. The latter being larger than earlier reported result. Below 50 nm, the experimental values increase while the calculated decrease.     

Development of a portable high-energy neutron spectrometer

The design of a portable high-energy (20-800 MeV) neutron spectrometer based on CsI or BaF2 is described. The particle discrimination properties of these scintillators allow the light-ion spallation products (p, d, t and alpha) from neutron interactions to be identified uniquely. One or more of the resulting pulse-height spectra can be unfolded to reveal the incident neutron spectrum. Dosimetric quantities can then be calculated based on the unfolded spectrum. Due to the high stopping power of these scintillators, modest-sized crystals are suitable for this application. Combined with advances in electronics, a lightweight instrument capable of on-line particle discrimination with a real-time display of neutron-induced count rate is feasible. Preliminary experimental data are presented, and the importance of validating MCNPX-generated response functions is discussed. A brief discussion on future work follows.     

Neutron spectra and angular distributions of concrete-moderated neutron calibration fields at JAERI

The Facility of Radiation Standards of Japan Atomic Energy Research Institute has been equipped with concrete-moderated neutron calibration fields as simulated workplace neutron fields. The fields use an 241Am-Be neutron source placed in the narrow space surrounded by concrete bricks, walls and floor. The neutron spectra and the neutron fluence rates of the fields were measured with the Bonner multi-sphere spectrometer system (BMS), spherical recoil-proton proportional counters, and a liquid scintillation counter (NE-213). The results were compared with each other. The reference values of H*(10) were determined from the results of BMS. The angular distributions of neutron fluence were calculated using MCNP-4B2 to obtain the reference values of Hp(10). The calculated results show that the scattered neutrons have a wide range of incident angles. The reference Hp(10) values considered the angular distribution were found to be 10-18% smaller than those without consideration.     

Determination of the neutron fluence spectra in the neutron therapy room of KIRAMS

High energy proton induced neutron fluence spectra were determined at the Korea Institute of Radiological and Medical Sciences (KIRAMS) using an extended Bonner Sphere (BS) set from the Korea Atomic Energy Research Institute (KAERI) in a series of measurements to quantify the neutron field. At the facility of the MC50 cyclotron of KIRAMS, two Be targets of different thicknesses, 1.0 and 10.5 mm, were bombarded by 35 and 45-MeV protons to produce six kinds of neutron fields, which were classified according to the measurement position and the use or no use of a beam collimator such as the gantry of the neutron therapy unit. In order to obtain a priori information to unfold the measured BS data the MCNPX code was used to calculate the neutron spectrum, and the influence of the surrounding materials for cooling the target assembly were also reviewed through this calculation. Some dosimetric quantities were determined by using the spectra determined in this measurement. Dose equivalent rates of these neutron fields ranged from 0.21 to 5.66 mSv h(-1)nA(-1) and the neutron yields for a thick Be target were 3.05 and 4.77% in the case of using a 35 and a 45-MeV proton, respectively.     

ANDI-03: a genetic algorithm tool for the analysis of activation detector data to unfold high-energy neutron spectra

The thresholds of (n,xn) reactions in various activation detectors are commonly used to unfold the neutron spectra covering a broad energy span, i.e. from thermal to several hundreds of MeV. The saturation activities of the daughter nuclides (i.e. reaction products) serve as the input data of specific spectra unfolding codes, such as SAND-II and LOUHI-83. However, most spectra unfolding codes, including the above, require an a priori (guess) spectrum to starting up the unfolding procedure of an unknown spectrum. The accuracy and exactness of the resulting spectrum primarily depends on the subjectively chosen guess spectrum. On the other hand, the Genetic Algorithm (GA)-based spectra unfolding technique ANDI-03 (Activation-detector Neutron DIfferentiation) presented in this report does not require a specific starting parameter. The GA is a robust problem-solving tool, which emulates the Darwinian Theory of Evolution prevailing in the realm of biological world and is ideally suited to optimise complex objective functions globally in a large multidimensional solution space. The activation data of the 27Al(n,alpha)24Na, 116In(n,gamma)116mIn, 12C(n,2n)11C and 209Bi(n,xn)(210-x)Bi reactions recorded at the high-energy neutron field of the ISIS Spallation source (Rutherford Appleton Laboratory, UK) was obtained from literature and by applying the ANDI-03 GA tool, these data were used to unfold the neutron spectra. The total neutron fluence derived from the neutron spectrum unfolded using GA technique (ANDI-03) agreed within +/-6.9% (at shield top level) and +/-27.2% (behind a 60 cm thick concrete shield) with the same unfolded with the SAND-II code.     

Measurement of the neutron fluence and dose spectra using an extended bonner sphere and a tissue-equivalent proportional counter

A conventional Bonner Sphere (BS) set consisting of six polyethylene spheres was modified to enhance its response to a high-energy neutron by putting a lead shell inside a polyethylene moderator. The response matrix of an extended BS was calculated using the MCNPX code and calibrated using a 252Cf neutron source. In order to survey the unknown photon and neutron mixed field, a spherical tissue equivalent proportional counter (TEPC) was constructed and assembled as a portable measurement system. The extended BS and the self-constructed TEPC were employed to determine the dosimetric quantities of the neutron field produced from the thick lead target bombarded by the 2.5 GeV electron beam of Pohang Accelerator Laboratory (PAL) and the neutron calibration field of Korea Atomic Energy Research Institute (KAERI).     

Self-shielding effects in neutron spectra measurements for neutron capture therapy by means of activation foils

The design and optimisation of a neutron beam for neutron capture therapy (NCT) is accompanied by the neutron spectra measurements at the target position. The method of activation detectors was applied for the neutron spectra measurements. Epithermal neutron energy region imposes the resonance structure of activation cross sections resulting in strong self-shielding effects. The neutron self-shielding correction factor was calculated using a simple analytical model of a single absorption event. Such a procedure has been applied to individual cross sections from pointwise ENDF/B-VI library and new corrected activation cross sections were introduced to a spectra unfolding algorithm. The method has been verified experimentally both for isotropic and for parallel neutron beams. Two sets of diluted and non-diluted activation foils covered with cadmium were irradiated in the neutron field. The comparison of activation rates of diluted and non-diluted foils has demonstrated the correctness of the applied self-shielding model.     

Alanine and TLD coupled detectors for fast neutron dose measurements in neutron capture therapy (NCT)

A method was investigated to measure gamma and fast neutron doses in phantoms exposed to an epithermal neutron beam designed for neutron capture therapy (NCT). The gamma dose component was measured by TLD-300 [CaF2:Tm] and the fast neutron dose, mainly due to elastic scattering with hydrogen nuclei, was measured by alanine dosemeters [CH3CH(NH2)COOH]. The gamma and fast neutron doses deposited in alanine dosemeters are very near to those released in tissue, because of the alanine tissue equivalence. Couples of TLD-300 and alanine dosemeters were irradiated in phantoms positioned in the epithermal column of the Tapiro reactor (ENEA-Casaccia RC). The dosemeter response depends on the linear energy transfer (LET) of radiation, hence the precision and reliability of the fast neutron dose values obtained with the proposed method have been investigated. Results showed that the combination of alanine and TLD detectors is a promising method to separate gamma dose and fast neutron dose in NCT.     

Polarization of a neutron beam using a superconductor

Based on the London theory of superconductivity and using the quasistatic approximation, we have calculated the interaction energy between a superconducting plane and a dipole which is moving toward or away from the plane. Using the decoupling approximation, we have investigated the displacement and the angle flipping of a neutron, when the neutron moves toward or away from the superconducting plane. The theoretical model can, in principle, be applied to construct an apparatus to polarize the beam from a thermal neutron reactor.     

Model for forecasting the dose distributions in a low atomic-number material when irradiated with a beam of gamma photons

A mathematical model for forecasting the dose distributions in the volume of a low atomic-number material when irradiated with a beam of γ-photons using the data of measurements in the plane of the central section of the beam at several levels from the surface to the depth is developed. The results of modeling of the dose distributions are presented.     

Patient neutron dose equivalent exposures outside of the proton therapy treatment field

A large fraction of dose to healthy tissue located outside of the treatment field during proton therapy is attributable to neutrons produced in the beam-delivery apparatus. In this work, the neutron dose equivalent (H) per therapeutic proton absorbed dose (D) was estimated for typical treatment conditions as a function of range modulation width, angle with respect to the incident proton beam, and the distance from the isocentre at the Harvard Cyclotron Laboratory's (Cambridge, MA) passively spread treatment field using Monte Carlo simulations. For a beam with 16 cm penetration (depth) and a 5 x 5 cm2 lateral field size at the patient location along the incident beam direction at 100 cm from the isocentre, the predicted H/D values are 0.35 and 0.60 mSv Gy(-1) from the simulations and measurements, respectively. At all locations, the predicted H/D values are within a factor of 2 and 3 of the measured result for no modulation and 8.2 cm of modulation, respectively.     

Development of a neutron personal dose equivalent detector

A new neutron-measuring instrument that is intended to measure a neutron personal dose equivalent, H(p)(10) was developed. This instrument is composed of two parts: (1) a conventional moderator-based neutron dose equivalent meter and (2) a neutron shield made of borated polyethylene, which covers a backward hemisphere to adjust the angular dependence. The whole design was determined on the basis of MCNP calculations so as to have response characteristics that would generally match both the energy and angular dependencies of H(p)(10). This new instrument will be a great help in assessing the reference values of neutron H(p)(10) during field testing of personal neutron dosemeters in workplaces and also in interpreting their readings.     

Preliminary design of a Gd-NCT neutron beam based on compact D-T neutron source

Gadolinium has been recently proposed, as neutron capture agent in NCT (Neutron Capture Therapy), due to both the nuclide high neutron capture cross section, and the remarkable selective uptake inside tumour tissue that Gd-loaded compounds, can provide. When a neutron external source is supplied, different Gd nuclear reactions, and the generated Auger electrons in particular, cause a high local energy deposition, which results in a tumour cell inactivation. Preliminary micro- as well as macrodosimetric Monte Carlo computational investigations show that the tumour-to-healthy tissue biological damage ratio is in close relation to the neutron beam energy spectrum. The results points out that the optimum neutron spectrum, to be used for Gd-NCT, seems to lie in the 1 to 10 keV energy range. In order to 'tailor' such spectra, an original, accelerator-driven, neutron source and spectrum shaping assembly for hospital-based Gd-NCT are presented and preliminary results are reported.     

Experimental and theoretical study of the neutron dose produced by carbon ion therapy beams

High-energy (12)C ions offer favourable conditions for the treatment of deep-seated local tumours. Several facilities for the heavy ion therapy are planned or under construction, for example the new clinical ion-therapy unit HIT at the Radiological University Clinics in Heidelberg. In order to improve existing treatment planning models, it is essential to evaluate the secondary fragment production and to include these contributions to the therapy dose with higher accuracy. Secondary neutrons are most abundantly produced in the reactions between (12)C beams and tissues. The dose contribution to tissues by a neutron is fairly small compared with the projectile and the other charged fragments due to no ionisation and the small reaction cross-sections; however, it distributes in a considerably wider region beyond the bragg-peak because of the strong penetrability. Systematic data on energy spectra and doses of secondary neutrons produced by (12)C beams using water targets of different thicknesses for various detection angles have therefore been measured in this study at GSI Darmstadt.     

Differential absorbed dose distributions in lineal energy for neutrons and gamma rays at the mono-energetic neutron calibration facility

Absorbed dose distributions in lineal energy for neutrons and gamma rays of mono-energetic neutron sources from 140 keV to 15 MeV were measured in the Fast Neutron Laboratory at Tohoku University. By using both a tissue-equivalent plastic walled counter and a graphite-walled low-pressure proportional counter, absorbed dose distributions in lineal energy for neutrons were obtained separately from those for gamma rays. This method needs no knowledge of energy spectra and dose distributions for gamma rays. The gamma-ray contribution in this neutron calibration field >1 MeV neutron was <3%, while for <550 keV it was >40%. The measured neutron absolute absorbed doses per unit neutron fluence agreed with the LA150 evaluated kerma factors. By using this method, absorbed dose distributions in lineal energy for neutrons and gamma rays in an unknown neutron field can be obtained separately.     

Thermoluminescence measurements of neutron dose around a medical linac

The photoncutron ambient dose around a 18 MV medical electron lineal accelerator has been measured with LiF:Mg,Ti chips of 3 x 3 x 1 mm inside moderating spheres. During the measurements a water phantom was irradiated in a field of 40 x 40 cm2. Two methods have been considered for comparison. In the first, a TLD-600/TLD-700 pair at the centre of a 25 cm diameter paraffine sphere was used, with the system behaving as a rem meter. In the second method, TLD-600/TLD-700 pairs, bare and at the centre of 7.6, 12.7, 20.3, 25.4, and 30.5 cm diameter polyethylene Bonner spheres were used to obtain the neutron spectrum. This was unfolded using the BUNKIUT code with the SPUNIT algorithm and the UTA4 and ARKI response functions. The neutron dose was followed by multiplying the unfolded neutron spectrum by the ambient dose equivalent to neutron fluence conversion factors. Both methods result in 0.5 mSv x Gy(-1) m away from the isocentre.     

Measurements of secondary neutron dose from 15 MV and 18 MV IMRT

Secondary neutron dose-equivalents were determined for conventional and intensity modulated radiation therapy (IMRT) prostate treatments for 15 and 18 MV X-ray beams. Conventional and IMRT treatment plans were generated to deliver 45 Gy to the prostate, seminal vessicles and external and internal iliac lymph nodes. Neutron spectra were determined by unfolding measurements from a TLD-based Bonner sphere system. Treatments using 18 MV IMRT and conventional plans result in neutron ambient dose-equivalents of 687 and 112 mSv, respectively. Delivery of the 15 MV IMRT and conventional plans results in neutron ambient dose-equivalents of 327 and 52 mSv, respectively. The data illustrate that using lower photon energies for IMRT reduces the secondary neutron dose, while still achieving comparable treatment volume coverage and sparing critical normal tissue.