Experimental verification of beam characteristics for cyclotron-based epithermal neutron source (C-BENS)

A cyclotron-based epithermal neutron source has been developed for boron neutron capture therapy. This system consists of a cyclotron accelerator producing 1.1-mA proton beams with an energy of 30 MeV, a beam transport system coupled with a beryllium neutron production target, and a beam-shaping assembly (BSA) with a neutron collimator. In our previous work, the BSA was optimized to obtain sufficient epithermal neutron fluxes of using a Monte Carlo simulation code. In order to validate the simulation results, irradiation tests using multi-foil activation at the surface of a gamma-ray shield located behind the collimator and water phantom experiments using a collimated epithermal neutron beam were performed. It was confirmed experimentally that the intensity of the epithermal neutrons was 1.2×10⁹ cm⁻² s⁻¹.     

BINP accelerator based epithermal neutron source

Innovative facility for neutron capture therapy has been built at BINP. This facility is based on compact vacuum insulation tandem accelerator designed to produce proton current up to 10 mA. Epithermal neutrons are proposed to be generated by 1.915–2.5 MeV protons bombarding a lithium target using ⁷Li(p,n)⁷Be threshold reaction. In the article, diagnostic techniques for proton beam and neutrons developed are described, results of experiments on proton beam transport and neutron generation are shown, discussed, and plans are presented.     

The production by 72 MeV protons of keV neutrons for 10B neutron capture therapy

Fast neutrons, produced in a spallation process by 72 MeV protons hitting heavy nuclei, can be moderated down to keV energies, suitable for 10B capture therapy. Measurements of neutron flux from copper and lead targets, after moderation by water and hydrocarbons, have been performed, together with measurements of the fast neutron and gamma-backgrounds. The comparisons with Monte Carlo calculations show good agreement. As neutron fluences up to 10(12) to 10(13)n/cm2 can be produced within a few hours in about 0.5 m target distance the technique seems to be suitable for clinical experiments.     

Prompt gamma emissions in the reaction for neutrons around the 1.45 eV absorption resonance

The relative intensities of different gamma emissions produced after the reaction ¹¹⁵In(n,γ)¹¹⁶In were analyzed for the particular case of incident neutron energies around the 1.45 eV indium absorption resonance. For this purpose, a pulsed neutron source in combination with the time-of-flight method for selecting the incident neutron energy range was employed. For neutrons around the mentioned absorption resonance the prompt gamma spectrum was extended to energies below 273 keV, and the intensities of gamma emissions not reported in the literature for epithermal neutrons were determined.     

Near-threshold 7Li(p,n)7Be neutrons on the practical conditions using thick Li-target and Gaussian proton energies for BNCT

The near threshold ⁷Li(p,n)⁷Be neutrons generated by incident proton energy having Gaussian distribution with mean energies from 1.85 to 1.95 MeV, were studied as a practical neutron source for BNCT wherein an RFQ accelerator and a thick Li-target are used. Gaussian energy distributions with the standard deviation of 0, 10, 20 and 40 keV for mean proton energies from 1.85 to 1.95 MeV were surveyed in 0.01 MeV increments. A thick liquid Li-target whose dimensions were established in our previous experiments (i.e., 1 mm-thick with 50 mm width and 50 mm length) was considered in this study. The suitable incident proton energy and physical dimensions of Pb layer which serves as a gamma absorber and a Polyethylene layer which is used as a BDE were surveyed by means of the concepts of TPD. Dose distribution were calculated by using MCNP5. A proton beam with mean energy of 1.92 MeV and a Gaussian energy distribution with a standard deviation of 20 keV at a current of 10 mA was selected from the viewpoint of irradiation time and practically achievable proton current. The suitable thicknesses of Pb gamma absorber was estimated to be about 3 cm. The estimated thickness of the polyethylene BDE was about 24 mm for an ideal proton current of 13 mA, and was 18 mm for a practical proton current of 10 mA.     

硼中子俘获治疗的加速器中子源7Li(p,n)7Be的中子特性研究

目的 通过研究质子加速器 ⁷Li(p,n)⁷Be 反应的中子特性,为研究和制作适用于硼中子俘获治疗(BNCT)的加速器中子源提供基础数据。方法 加速质子使其轰击Li靶后产生中子;通过金属箔活化法,测量中子与In箔发生阈值反应后放出的γ射线;然后计算出In箔的放射性活度、加速器反应后放出中子的注量和反应的微分截面。结果 质子加速轰击Li靶后,在不同方向产生不同能量和注量的中子。加速器电压分别为3.0、2.8和2.6 MV,出射中子与入射质子束的方向一致时, ⁷Li(p,n)⁷Be 反应的微分截面约为50 mb/mr;夹角为60°时,反应的微分截面减小到30 mb/mr左右。由于部分中子与其他金属原子等发生弹性散射而射向后方,提高了这一范围内In箔的比放射性活度,影响了其微分截面的准确性。结论 用金属箔活化法测定中子简便易行,可同时测得多个方向的中子分布,但需对中子与其他金属弹性散射产生的影响进行进一步的研究; ⁷Li(p,n)⁷Be 反应后发射出的中子经慢化后,能得到适于BNCT治疗的热中子和超热中子;若作为BNCT的中子源,加速器的质子束流需达到10 mA。     

High-power electron beam tests of a liquid-lithium target and characterization study of 7Li(p,n) near-threshold neutrons for accelerator-based boron neutron capture therapy

A compact Liquid-Lithium Target (LiLiT) was built and tested with a high-power electron gun at Soreq Nuclear Research Center (SNRC). The target is intended to demonstrate liquid-lithium target capabilities to constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals. The lithium target will produce neutrons through the ⁷Li(p,n)⁷Be reaction and it will overcome the major problem of removing the thermal power >5 kW generated by high-intensity proton beams, necessary for sufficient therapeutic neutron flux.In preliminary experiments liquid lithium was flown through the target loop and generated a stable jet on the concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power densities of more than 4 kW/cm² and volumetric power density around 2 MW/cm³ at a lithium flow of ~4 m/s, while maintaining stable temperature and vacuum conditions. These power densities correspond to a narrow (σ=~2 mm) 1.91 MeV, 3 mA proton beam. A high-intensity proton beam irradiation (1.91–2.5 MeV, 2 mA) is being commissioned at the SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator.In order to determine the conditions of LiLiT proton irradiation for BNCT and to tailor the neutron energy spectrum, a characterization of near threshold (~1.91 MeV) ⁷Li(p,n) neutrons is in progress based on Monte-Carlo (MCNP and Geant4) simulation and on low-intensity experiments with solid LiF targets. In-phantom dosimetry measurements are performed using special designed dosimeters based on CR-39 track detectors.     

Dosimetric influence of secondary electrons during accelerator-produced fast neutron irradiation

1 MeV neutrons are produced by the ⁷Li(p, n)⁷Be nuclear reaction using 2.75 MeV protons and a target consisting of 3 × 10⁻⁴ g cm⁻² Li evaporated on 1 g cm⁻² Ta. The response due to secondary electrons was about 25% of the neutron response, but 1 mm polyethylene absorbs all these particles. This finding may be of particular interest for radiobiologists.     

Hydrogen effects in neutron activation analysis

Perturbations generated in two neutron fields, with different energy distributions, by scattering on hydrogen atoms are reported and discussed, as well as the influence of various parameters. The thermal neutron capture rates of several nuclei can be enhanced by about 18% in solutions in light water as compared to heavy water, for 20 cm³ cylindrical containers. This effect appears to be independent of the initial distribution of the epithermal neutrons, and can be accounted for by the scattering of non-Maxwellian, quasi thermalized neutrons to lower energies. The enhancement ratio is found to be about the same for a variety of nuclei (B, Hg, Gd, Eu, Sm and Cd), in spite of large differences in the variations of their cross sections with neutron energy. This can be explained by assuming that the incoming fluxes present a depression at the level of the Cd resonance, at about 0.18 eV.     

Mean neutron energy and the neutron spectra from T(d,n)4He reaction

The mean neutron energy and energy spread of neutrons from the T(d,n)⁴He reaction have been calculated for deuteron beams of energy from 50 keV to 1 MeV incident on a thick tritiated titanium target. This has been done over the angular range from 0° to 180° and the results are presented graphically as a function of deuteron energy and neutron emission angle and are also tabulated.     

宽能中子辐射个人防护材料研究

目的 研究可用于制作宽能中子辐射个人防护用品的柔性屏蔽材料。方法 依据理论计算结果配比加工实验用材料样品,利用²⁵²Cf裂变中子和加速器单能中子开展中子屏蔽效率实验。结果 给出了1.5 cm厚含10％碳化硼(B₄C)的乙丙橡胶板对144 keV等6个单能中子和²⁵²Cf裂变中子的屏蔽效率,其中对²⁵²Cf裂变中子的屏蔽效率为31.02%,对144 keV中子的屏蔽效率可达到76.9％。结论 以乙丙橡胶为基材掺入B₄C研制的个人中子防护柔性材料对不同能量中子辐射均有一定屏蔽效果,可用于制作宽能中子个人防护用品。     

Development of a low-energy monoenergetic neutron source for applications in low-dose radiobiological and radiochemical research

The McMaster University 3 MV KN Van de Graff accelerator facility primarily dedicated to in vivo neutron activation measurements has been used to produce moderate dose rates of monoenergetic fast neutrons of energy ranging from 150 to 600 keV with a small energy spread of about 25 keV (1σ width of Gaussian) by bombarding thin lithium targets with 2.00–2.40 MeV protons. The calculated dose rate of the monoenergetic neutrons produced using thin lithium targets as functions of beam energy, target thickness, lab angle relative to beam direction, and the solid angle subtended by the sample with the target has also been reported.     

Optimization of the target of an accelerator-driven neutron source through Monte Carlo numerical simulation of neutron and gamma transport by the PRIZMA code

At Budker Institute of Nuclear Physics, epithermal neutron source for neutron-capture therapy was built and neutron generation was realized. Source is based on tandem accelerator and uses near-threshold neutron generation from the reaction ⁷Li(p,n)⁷Be. The paper describes target optimization through the numerical simulation of proton, neutron and gamma transport by Monte Carlo method (PRIZMA code). It is shown that the near-threshold mode attractive due low activation provides high efficiency of the dose and acceptable therapeutic ratio and advantage depth.     

Lithium neutron producing target for BINP accelerator-based neutron source.

Pilot innovative accelerator-based neutron source for neutron capture therapy is under construction now at the Budker Institute of Nuclear Physics, Novosibirsk, Russia. One of the main elements of the facility is lithium target, that produces neutrons via threshold (7)Li(p,n)(7)Be reaction at 25 kW proton beam with energies 1.915 or 2.5 MeV. In the present report, the results of experiments on neutron producing target prototype are presented, the results of calculations of hydraulic resistance for heat carrier flow and lithium layer temperature are shown. Calculation showed that the lithium target could run up to 10 mA proton beam before melting. Choice of target variant is substantiated. Program of immediate necessary experiments is described. Target design for neutron source constructed at BINP is presented. Manufacturing the neutron producing target up to the end of 2004 and obtaining a neutron beam on BINP accelerator-based neutron source are planned during 2005.     

Interaction between the biological effects of high- and low-LET radiation dose components in a mixed field exposure

Abstract

Purpose: The relative biological effectiveness of two epithermal neutron sources, a reactor based source at Studsvik, Sweden, and a proton accelerator-based source in Birmingham, UK, was studied in relation to the proportional absorbed dose distribution as a function of neutron energy. Evidence for any interactions between the effects of biological damage induced by high- and low-linear energy transfer (LET) dose components, in this ‘mixed field’ irradiation, was also examined

Materials and methods: Clonogenic survival in Chinese Hamster-derived V79 cells was used to assess biological effectiveness in this study. Cells were irradiated in suspension at 4°C at depths of 20, 35, 50 and 65 mm in a water phantom. This prevented the repair of sublethal damage, predominantly that produced by both incident and induced γ-rays in the field, over the variable periods of exposure required to irradiate cells with the same total absorbed dose. Cell survival, as a function of the absorbed radiation dose and depth in the phantom, was compared with Monte Carlo N-Particle (MCNP) calculations of the proportional absorbed dose distribution as a function of neutron energy for the two sources.

Results: In terms of the dose-related reduction in clonogenic cell survival, the epithermal neutron source at Studsvik was more biologically effective than the Birmingham source at all depths considered in the phantom. Although the contribution from the high-LET dose component was greater for the Studsvik source at 20 mm depth in the phantom, at greater depths the dose contribution from the high-LET dose component at Studsvik overlap with those for the Birmingham source. However, the most striking difference is in the fast neutron component to the dose of the two sources, neutron energies > 1 MeV were only associated with the Studsvik source. The relative biological effectiveness (RBE) of both sources declined slightly with depth in the phantom, as the total high-LET dose component declined. The maximum source RBE for Studsvik was 2.70 ± 0.50 at 20 mm; reduced to 2.10 ± 0.35 at depths of 50 and 65 mm. The corresponding values for Birmingham were 1.68 ± 0.25 and 1.31 ± 0.19, all values relate only to the surviving fraction of V79 cells at 37%, since RBE values are only applicable to the selected endpoint. Based on a dose reduction factor (DRF) of 1.0 for the total low-LET component to the absorbed dose, the RBE values for the high-LET dose component (fast neutrons and induced protons from the nitrogen capture reaction) was 14.5 and 7.05 for the Studsvik and Birmingham neutron sources, respectively. This is well outside the range of RBE historically reported values for V79 cells for the same level of cell survival for fast neutrons. The calculation of RBE values, based on the proportional absorbed dose distribution as a function of neutron energy, from historical data, and using a RBE of 1.8 for the dose from the nitrogen capture reaction, suggests RBE values for the total high-LET dose component of 3.1–2.8 and 2.5–2.0 for Studsvik and Birmingham, respectively, values again declining with depth in the phantom.

Conclusions: The overall biological effectiveness of the mixed field irradiation from an epithermal neutron sources depends on the composition and quality of the different dose components. The experimentally derived RBE values for the total high-LET dose components in these ‘mixed field’ irradiations are well in excess of historical data for fast neutrons. The difference between the historically expected and the observed RBE values is attributed to the interactions between the damage produced by high- and low-LET radiation.     

Oncogenic transformation in C3H10T1/2 cells by low-energy neutrons

PURPOSE: Occupational exposure to neutrons typically includes significant doses of low-energy neutrons, with energies below 100 keV. In addition, the normal-tissue dose from boron neutron capture therapy will largely be from low-energy neutrons. Microdosimetric theory predicts decreasing biological effectiveness for neutrons with energies below about 350 keV compared with that for higher-energy neutrons; based on such considerations, and limited biological data, the current radiation weighting factor (quality factor) for neutrons with energies from 10 keV to 100 keV is less than that for higher-energy neutrons. By contrast, some reports have suggested that the biological effectiveness of low-energy neutrons is similar to that of fast neutrons. The purpose of the current work is to assess the relative biological effectiveness of low-energy neutrons for an endpoint of relevance to carcinogenesis: in vitro oncogenic transformation. METHODS: Oncogenic transformation induction frequencies were determined for C3H10T1/2 cells exposed to two low-energy neutron beams, respectively, with dose-averaged energies of 40 and 70 keV, and the results were compared with those for higher-energy neutrons and X-rays. RESULTS: These results for oncogenic transformation provide evidence for a significant decrease in biological effectiveness for 40 keV neutrons compared with 350 keV neutrons. The 70 keV neutrons were intermediate in effectiveness between the 70 and 350 keV beams. CONCLUSIONS: A decrease in biological effectiveness for low-energy neutrons is in agreement with most (but not all) earlier biological studies, as well as microdosimetric considerations. The results for oncogenic transformation were consistent with the currently recommended decreased values for low-energy neutron radiation weighting factors compared with fast neutrons.     

The Relative Biological Effectiveness of Different Kinds of Radiations on Chromosome Aberrations in Nigella Damascena Seed

Summary

In Nigella damascena the relative biological effectiveness (RBE) of different kinds of radiations was measured for chromosome breaks induced in G₁ and scored at first mitosis after germination of seeds. The effects of γ-rays from ¹³⁷Cs were used for reference to compute RBE's for 2·8 GeV protons, mono-energetic neutrons, fission neutrons, and α-particles from the reactions ¹⁰B(n, α)⁷Li and ⁶Li(n, α)³H following irradiation with thermal neutrons. For 2·8 GeV protons the RBE was 2·1 for total aberrations. The RBE's for mono-energetic neutrons ranged from 44 (1·8 MeV) to 84 (0·43 MeV) for one-break, and 79 (1·8 MeV) to 154 (0·43 MeV) for two-break aberrations: for fission neutrons the RBE was 78 for total aberrations. The low RBE for α-particles from thermal neutron irradiation of seeds enriched with ¹⁰B and ⁶Li was attributed in part to the non-homogeneous distribution of these atoms in the seeds.     

Measurements of isomeric cross sections for the (n,α) reaction on the 142Nd isotope at approximately 14 MeV neutrons

In this study, the activation cross sections were measured for ¹⁴²Nd(n,α)^139mCe reaction at four neutron energies between 13.57 and 14.83 MeV, which were produced by a neutron generator through ³H(²H,n)⁴He reaction. The production of short-lived activity and the spectra accumulation were performed by the cyclic activation technique. Induced gamma-ray activities were measured using a high resolution gamma ray spectrometer equipped with a high-purity Germanium (HpGe) detector. In the cross section measurements, corrections were made regarding the effects of the gamma-ray attenuation, the dead-time, the fluctuation of the neutron flux, and low energy neutrons. The measured cross sections were compared with the published literature and the results of the model calculation (TALYS 1.4).     

Measurements of neutron distribution in neutrons–γ-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 ¹⁵⁷Gd(n,γ)¹⁵⁸Gd 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 γ-rays. The Hiroshima University Radiobiological Research Accelerator is utilized for this experiment. The neutrons are moderated with 20-cm-thick D₂O 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 γ-rays, which was estimated using IP without Gd. The γ-ray response of Gd-doped IP to non-Gd IP is set at 1.34, the value measured for ⁶⁰Co γ-rays, in estimating the γ-ray contribution to Gd-doped IP signal. Then measured distribution of the ¹⁵⁷Gd(n,γ)¹⁵⁸Gd 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 ¹⁵⁷Gd(n,γ)¹⁵⁸Gd reaction rate is so sensitive to γ-ray energy, e.g. the discrepancy of the ¹⁵⁷Gd(n,γ)¹⁵⁸Gd reaction rate between measurement and calculation becomes 30% for the photon energy change from 33 keV to 1.253 MeV.     

An optimized neutron-beam shaping assembly for accelerator-based BNCT.

Different materials and proton beam energies have been studied in order to search for an optimized neutron production target and beam shaping assembly for accelerator-based BNCT. The solution proposed in this work consists of successive stacks of Al, polytetrafluoroethylene, commercially known as Teflon, and LiF as moderator and neutron absorber, and Pb as reflector. This assembly is easy to build and its cost is relatively low. An exhaustive Monte Carlo simulation study has been performed evaluating the doses delivered to a Snyder model head phantom by a neutron production Li-metal target based on the (7)Li(p,n)(7)Be reaction for proton bombarding energies of 1.92, 2.0, 2.3 and 2.5 MeV. Three moderator thicknesses have been studied and the figures of merit show the advantage of irradiating with near-resonance-energy protons (2.3 MeV) because of the relatively high neutron yield at this energy, which at the same time keeps the fast neutron healthy tissue dose limited and leads to the lowest treatment times. A moderator of 34 cm length has shown the best performance among the studied cases.