Design of an explosive detection system using Monte Carlo method.

Regardless the motivation terrorism is the most important risk for the national security in many countries. Attacks with explosives are the most common method used by terrorists. Therefore several procedures to detect explosives are utilized; among these methods are the use of neutrons and photons. In this study the Monte Carlo method an explosive detection system using a 241AmBe neutron source was designed. In the design light water, paraffin, polyethylene, and graphite were used as moderators. In the work the explosive RDX was used and the induced gamma rays due to neutron capture in the explosive was estimated using NaI(Tl) and HPGe detectors. When light water is used as moderator and HPGe as the detector the system has the best performance allowing distinguishing between the explosive and urea. For the final design the Ambient dose equivalent for neutrons and photons were estimated along the radial and axial axis.

Study of suitability of Fricke-gel-layer dosimeters for in-air measurements to characterise epithermal/thermal neutron beams for NCT.

The reliability of Fricke gel dosimeters in form of layers for measurements aimed at the characterization of epithermal neutron beams has been studied. By means of dosimeters of different isotopic composition (standard, containing (10)B or prepared with heavy water) placed against the collimator exit, the spatial distribution of gamma and fast neutron doses and of thermal neutron fluence are attained. In order to investigate the accuracy of the results obtained with in-air measurements, suitable MC simulations have been developed and experimental measurements have been performed utilizing Fricke gel dosimeters, thermoluminescence detectors and activation foils. The studies were related to the epithermal beam designed for BNCT irradiations at the research reactor LVR-15 (Rez). The results of calculation and measurements have revealed good consistency of gamma dose and fast neutron 2D distributions obtained with gel dosimeters in form of layers. In contrast, noticeable modification of thermal neutron fluence is caused by the neutron moderation produced by the dosimeter material. Fricke gel dosimeters in thin cylinders, with diameter not greater than 3mm, have proved to give good results for thermal neutron profiling. For greater accuracy of all results, a better knowledge of the dependence of gel dosimeter sensitivity on radiation LET is needed.

Neutron cross-sections for next generation reactors: new data from n_TOF.

In 2002, an innovative neutron time-of-flight facility started operation at CERN: n_TOF. The main characteristics that make the new facility unique are the high instantaneous neutron flux, high resolution and wide energy range. Combined with state-of-the-art detectors and data acquisition system, these features have allowed to collect high accuracy neutron cross-section data on a variety of isotopes, many of which radioactive, of interest for Nuclear Astrophysics and for applications to advanced reactor technologies. A review of the most important results on capture and fission reactions obtained so far at n_TOF is presented, together with plans for new measurements related to nuclear industry.

Measurements of neutron distribution in neutrons-gamma-rays mixed field using imaging plate for neutron capture therapy.

The imaging plate (IP) technique is tried to be used as a handy method to measure the spatial neutron distribution via the (157)Gd(n,gamma)(158)Gd reaction for neutron capture therapy (NCT). For this purpose, IP is set in a water phantom and irradiated in a mixed field of neutrons and gamma-rays. The Hiroshima University Radiobiological Research Accelerator is utilized for this experiment. The neutrons are moderated with 20-cm-thick D(2)O to obtain suitable neutron field for NCT. The signal for IP doped with Gd as a neutron-response enhancer is subtracted with its contribution by gamma-rays, which was estimated using IP without Gd. The gamma-ray response of Gd-doped IP to non-Gd IP is set at 1.34, the value measured for (60)Co gamma-rays, in estimating the gamma-ray contribution to Gd-doped IP signal. Then measured distribution of the (157)Gd(n,gamma)(158)Gd reaction rate agrees within 10% with the calculated value based on the method that has already been validated for its reproducibility of Au activation. However, the evaluated distribution of the (157)Gd(n,gamma)(158)Gd reaction rate is so sensitive to gamma-ray energy, e.g. the discrepancy of the (157)Gd(n,gamma)(158)Gd reaction rate between measurement and calculation becomes 30% for the photon energy change from 33keV to 1.253MeV.

[Particle radiotherapy for malignant gliomas].

Particle radiations using protons or carbons, and boron neutron capture therapy are emerging as a therapeutic modality for malignant gliomas. Some non-randomized prospective studies found favorable results; however, the advantageous method of advantage of conformal radiation using protons or carbon ions and tumor cell-selective radiation using boron neutron capture therapy have not been confirmed. In a recent clinical trial involving 20 newly diagnosed glioblastomas at the Proton Medical Research Center at Tsukuba, the median overall survival time and the 1- and 2-year survival rates were 21.6 months, 71.1% and 45.3%, respectively. In the clinical trial of boron neutron capture therapy of 15 newly diagnosed glioblastomas at Tsukuba, the median overall survival time and the 1- and 2-year survival rates were 25.7 month, 80.0% and 53.3%, respectively. The rationale, history, and clinical results of particle radiotherapy for glioblastoma were also discussed.

Gadolinium diethylenetriaminopentaacetic acid-loaded chitosan microspheres for gadolinium neutron-capture therapy.

In order to provide a suitable device that would contain water-soluble drugs, highly water-soluble gadolinium diethylenetriaminopentaacetic acid-loaded chitosan microspheres (CMS-Gd-DTPA) were prepared by the emulsion method using glutaraldehyde as a cross-linker and Span 80 as a surfactant for gadolinium neutron-capture therapy of cancer. The gadolinium content and the mass median diameter of CMS-Gd-DTPA were estimated. The size and morphology of the CMS-Gd-DTPA were strongly influenced by the initial applied weight ratio of Gd-DTPA:chitosan. FTIR spectra showed that the electrostatic interaction between chitosan and Gd-DTPA accelerated the formation of gadolinium-enriched chitosan microspheres. Sufficient amounts of glutaraldehyde and Span 80 were necessary for producing discrete CMS-Gd-DTPA. The CMS-Gd-DTPA having a mass median diameter 11.7microm and 11.6% of gadolinium could be used in Gd-NCT following intratumoral injection.

Comparison doses of secondary neutron with the heavy ions in a 75-Mev/n heavy ion beam.

The angular distributions for neutrons of energy >6 MeV that are induced by 75 MeV/n 12C6+ and 16O8+ ions were measured with the activation method of Al threshold detectors at the radiobiological terminal of HIRFL. The data were obtained by a high-purity Ge(HpGe) detector. The results show that the neutron angular distributions produced by heavy ion beams are strongly peaked in the forward direction and decreased exponentially with angles in experimental area. The experimental conditions for these measurements were similar to those for biological experiments, so the results should be representative of neutrons produced by heavy ions during the biological experiments and tumour therapy. Comparing with the neutron doses produced by the heavy ion beam, the heavy ion dose is the main factor in biological effects and tumour therapy response, so the contribution of neutron dose can be neglected.

Combination of boron and gadolinium compounds for neutron capture therapy. An in vitro study.

In neutron capture therapy, the therapeutic effect of the boron compound is based on alpha particles produced by the B(n, alpha) reaction while with the gadolinium compound the main radiation effect is from gamma rays derived from the Gd(n, gamma) reaction. The uptake and distribution within the tumor may be different among these compounds. Thus, the combination of the boron and gadolinium compounds may be beneficial for enhancing the radiation dose to the tumor. Chinese hamster fibroblast V79 cells were used. For the neutron targeting compounds, 10B (BSH) at 0, 5, 10, and 15 ppm, and 157Gd (Gd-BOPTA) at 0, 800, 1600, 2400, 3200, and 4800 ppm, were combined. The neutron irradiation was performed with thermal neutrons for 30 min. (neutron flux: 0.84 x 10(8) n/cm2/s in free air). The combination of the boron and gadolinium compounds showed an additive effect when the gadolinium concentration was lower than 1600 ppm. This additive effect decreased as a function of gadolinium concentration at 2400 ppm and resulted in no additive effect at more than 3200 ppm of gadolinium. In conclusion, the combination of the boron and gadolinium compounds can enhance the therapeutic effect with an optimum concentration ratio. When the gadolinium concentration is too high, it may weaken the boron neutron capture reaction due to the high cross-section of gadolinium compound against neutrons.

Application of semiconductors for dosimetry of fast-neutron therapy beam.

Two types of ion implanted miniature p-i-n diodes were tested in a d(48.5) + Be fast-neutron beam produced in the Detroit superconducting cyclotron. The increase in forward voltage drop caused by neutron-induced damage was correlated with neutron dose measured in a water phantom. The neutron and gamma dose components were predetermined using twin detector (Tissue-equivalent ion chamber paired with miniature Geiger-Müller counter) method. The increase in the voltage drop for 1 mA injection current was monitored together with the cyclotron beam target current, thus the differential voltage drop could be defined precisely for given radiation dose. The average neutron sensitivities of tested diodes were 1.284 +/- 0.014 and 0.528 +/- 0.058 mV per cGy. The miniature detectors can be utilised in characterisation of small radiation fields and in the regions of high dose gradient as well as for in vivo dosimetry of the patients undergoing fast-neutron therapy.

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.

A state-of-the-art epithermal neutron irradiation facility for neutron capture therapy.

At the Massachusetts Institute of Technology (MIT) the first fission converter-based epithermal neutron beam (FCB) has proven suitable for use in clinical trials of boron neutron capture therapy (BNCT). The modern facility provides a high intensity beam together with low levels of contamination that is ideally suited for use with future, more selective boron delivery agents. Prescriptions for normal tissue tolerance doses consist of 2 or 3 fields lasting less than 10 min each with the currently available beam intensity, that are administered with an automated beam monitoring and control system to help ensure safety of the patient and staff alike. A quality assurance program ensures proper functioning of all instrumentation and safety interlocks as well as constancy of beam output relative to routine calibrations. Beam line shutters and the medical room walls provide sufficient shielding to enable access and use of the facility without affecting other experiments or normal operation of the multipurpose research reactor at MIT. Medical expertise and a large population in the greater Boston area are situated conveniently close to the university, which operates the research reactor 24 h a day for approximately 300 days per year. The operational characteristics of the facility closely match those established for conventional radiotherapy, which together with a near optimum beam performance ensure that the FCB is capable of determining whether the radiobiological promise of NCT can be realized in routine practice.

Neutron fibres and possible applications to NCT.

We summarize previous researches regarding neutron guides of small transverse cross-section (neutron fibres), smaller than those of the standard hollow guides and collimators employed currently. Those studies may not be widely known in the neutron capture therapy (NCT) community, but they may be interesting for it. Such neutron fibres could allow to deliver and concentrate neutron beams selectively in regions of size smaller than 1mm. We present new estimates and point out and discuss some new possible specific applications of those neutron fibres, which would not replace standard NCT but could supplement it. Thus, we entertain the possibility that neutron fibres could be useful for additional therapies (in typical NCT durations) of: (i) rather small tumours, (ii) thin borders of tumours. The use of these neutron fibres could reduce the undesirable delivery of radiation to healthy tissue around regions with malignant tissue.

Wedge factor dependence with depth and field size for fast neutron beams.

The dependence of the wedge factors (WFs) on field size (FS) and depth for a fast neutron beam has been investigated. In a previous study (Popescu et al 1999 Med. Phys. 26 541), a method was presented that allows a simple and accurate way of calculating the wedge-factor dependence on FS and depth in the case of a photon beam. The validity of a similar approach is tested in the present study for neutron beam dosimetry. The clinical neutron therapy system at the University of Washington (UW) has a flattening filter assembly consisting of two filters: a small field filter and a large field filter. Despite this complication, the approach presented in Popescu et al (1999 Med. Phys. 26 541) can be used to describe the WF dependence on FS and depth (d).

Reference dosimetry at the neutron capture therapy facility at Studsvik.

The purpose of this publication was to present and evaluate the methods for reference dosimetry in the epithermal neutron beam at the neutron capture therapy facility at Studsvik. Measurements were performed in a PMMA phantom and in air using ionization chambers and activation probes in order to calibrate the epithermal neutron beam. Appropriate beam-dependant calibration factors were determined using Monte Carlo methods for the detectors used in the present publication. Using the presented methodology, the photon, neutron and total absorbed dose to PMMA was determined with an estimated uncertainty of +/- 5.0%, +/- 25%, and +/- 5.5% (2 SD), respectively. The uncertainty of the determination of the photon absorbed dose was comparable to the case in conventional radiotherapy, while the uncertainty of the neutron absorbed dose is much higher using the present methods. The thermal neutron group fluence, i.e., the neutron fluence in the energy interval 0-0.414 eV, was determined with an estimated uncertainty of +/- 2.8% (2 SD), which is acceptable for dosimetry in epithermal neutron beams.

Fission reactor neutron sources for neutron capture therapy--a critical review.

The status of fission reactor-based neutron beams for neutron capture therapy (NCT) is reviewed critically. Epithermal neutron beams, which are favored for treatment of deep-seated tumors, have been constructed or are under construction at a number of reactors worldwide. Some of the most recently constructed epithermal neutron beams approach the theoretical optimum for beam purity. Of these higher quality beams, at least one is suitable for use in high through-put routine therapy. It is concluded that reactor-based epithermal neutron beams with near optimum characteristics are currently available and more can be constructed at existing reactors. Suitable reactors include relatively low power reactors using the core directly as a source of neutrons or a fission converter if core neutrons are difficult to access. Thermal neutron beams for NCT studies with small animals or for shallow tumor treatments, with near optimum properties have been available at reactors for many years. Additional high quality thermal beams can also be constructed at existing reactors or at new, small reactors. Furthermore, it should be possible to design and construct new low power reactors specifically for NCT, which meet all requirements for routine therapy and which are based on proven and highly safe reactor technology.

The medical-irradiation characteristics for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

At the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor, the mix irradiation of thermal and epi-thermal neutrons, and the solo irradiation of epi-thermal neutrons are available additionally to the thermal neutron irradiation, and then the neutron capture therapy (NCT) at this facility became more flexible, after the update in 1996. The estimation of the depth dose distributions in NCT clinical irradiation, were performed for the standard irradiation modes of thermal, mixed and epi-thermal neutrons, from the both sides of experiment and calculation. On the assumption that the 10B concentration in tumor part was 40 ppm and the ratio of tumor to normal tissue was 3.5, the advantage depth were estimated to 5.4, 6.0, and 8.0, for the respective standard irradiation modes. It was confirmed that the various irradiation conditions can be selected according to the target-volume conditions, such as size, depth, etc. Besides, in the viewpoint of the radiation shielding for patient, it was confirmed that the whole-body exposure is effectively reduced by the new clinical collimators, compared with the old one.

Controllability of depth dose distribution for neutron capture therapy at the Heavy Water Neutron Irradiation Facility of Kyoto University Research Reactor.

The updating construction of the Heavy Water Neutron Irradiation Facility of the Kyoto University Research Reactor has been performed from November 1995 to March 1996 mainly for the improvement in neutron capture therapy. On the performance, the neutron irradiation modes with the variable energy spectra from almost pure thermal to epi-thermal neutrons became available by the control of the heavy-water thickness in the spectrum shifter and by the open-and-close of the cadmium and boral thermal neutron filters. The depth distributions of thermal, epi-thermal and fast neutron fluxes were measured by activation method using gold and indium, and the depth distributions of gamma-ray absorbed dose rate were measured using thermo-luminescent dosimeter of beryllium oxide for the several irradiation modes. From these measured data, the controllability of the depth dose distribution using the spectrum shifter and the thermal neutron filters was confirmed.

Simulations of silicon microdosimetry measurements in fast neutron therapy.

Experimental silicon microdosimetry measurements were performed at a Fast Neutron Therapy facility. Monte Carlo based calculations of these measurements were made using the GEANT4 toolkit. Reasonable agreement between theoretical and experimental results was obtained and the contribution of elastic and inelastic reaction products to the final microdosimetric spectrum was determined.

Dose distributions in a human head phantom for neutron capture therapy using moderated neutrons from the 2.5 meV proton-7Li reaction or from fission of 235U.

The feasibility of neutron capture therapy (NCT) using an accelerator-based neutron source of the 7Li(p,n) reaction produced by 2.5 MeV protons was investigated by comparing the neutron beam tailored by both the Hiroshima University radiological research accelerator (HIRRAC) and the heavy water neutron irradiation facility in the Kyoto University reactor (KUR-HWNIF) from the viewpoint of the contamination dose ratios of the fast neutrons and the gamma rays. These contamination ratios to the boron dose were estimated in a water phantom of 20 cm diameter and 20 cm length to simulate a human head, with experiments by the same techniques for NCT in KUR-HWNIF and/or the simulation calculations by the Monte Carlo N-particle transport code system version 4B (MCNP-4B). It was found that the 7Li(p,n) neutrons produced by 2.5 MeV protons combined with 20, 25 or 30 cm thick D20 moderators of 20 cm diameter could make irradiation fields for NCT with depth-dose characteristics similar to those from the epithermal neutron beam at the KUR-HWNIF.

Enhancement of a 252Cf-based neutron beam via subcritical multiplication for neutron capture therapy.

Previous studies indicated that an epithermal-neutron beam based on bare 252Cf is not feasible for neutron capture therapy (NCT). It was reported that a clinically useful epithermal-neutron beam requires a minimum of 1.0 g of 252Cf, which is more than twice the US current annual supply. However, it was reasoned that the required quantity of 252Cf could be dramatically reduced when used with a subcritical multiplying assembly (SMA). This reasoning is based on the assumption that the epithermal-neutron beam intensity for NCT is directly proportional to the fission neutron population, and that the neutron multiplying factor of the SMA can be estimated by 1/(1 - k(eff)). We have performed detailed Monte Carlo calculations to investigate the validity of the above reasoning. Our results show that 1/(1 - k(eff)) grossly overestimates the beam enhancement factor for NCT. For example, Monte Carlo calculations predict a beam enhancement factor of 6.0 for an optimized SMA geometry with k(eff) = 0.968. This factor is much less than 31 predicted by 1/(1 - k(eff)). The overestimation is due to the fact that most of the neutrons produced in the SMA are self-shielded, whereas self-shielding is negligible in a bare 252Cf source. Since the beam intensity of a 0.1 g 252Cf with the optimized SMA enhancement is still more than an order of magnitude too low compared to the existing reactor beams, we conclude that the enhancement via an SMA for a 252Cf-based epithermal-neutron beam is inadequate for NCT.