Dose calculation from a D–D-reaction-based BSA for boron neutron capture synovectomy

Monte Carlo simulations were carried out to calculate dose in a knee phantom from a D–D–reaction-based Beam Shaping Assembly (BSA) for Boron Neutron Capture Synovectomy (BNCS). The BSA consists of a D(d,n)-reaction-based neutron source enclosed inside a polyethylene moderator and graphite reflector. The polyethylene moderator and graphite reflector sizes were optimized to deliver the highest ratio of thermal to fast neutron yield at the knee phantom. Then neutron dose was calculated at various depths in a knee phantom loaded with boron and therapeutic ratios of synovium dose/skin dose and synovium dose/bone dose were determined. Normalized to same boron loading in synovium, the values of the therapeutic ratios obtained in the present study are 12–30 times higher than the published values.     

Resumption of JRR-4 and characteristics of neutron beam for BNCT

The clinical trials of Boron Neutron Capture Therapy (BNCT) have been conducted using Japan Research Reactor No. 4 (JRR-4) at Japan Atomic Energy Agency (JAEA). On December 28th, 2007, a crack of a graphite reflector in the reactor core was found on the weld of the aluminum cladding. For this reason, specifications of graphite reflectors were renewed; dimensions of the graphite were reduced and gaps of water were increased. All existing graphite reflectors of JRR-4 were replaced by new graphite reflectors. In February 2010 the resumption of JRR-4 was carried out with new graphite reflectors. We measured the characteristics of neutron beam at the JRR-4 Neutron Beam Facility. A cylindrical water phantom of 18.6 cm diameter and 24 cm depth was set in front of the beam port with 1 cm gap. TLDs and gold wires were inserted within the phantom when the phantom was irradiated. The results of the measured thermal neutron flux and the gamma dose in water were compared with that of MCNP calculation. The neutron energy spectrum of the calculation model with new reflector had little variation compared to that with old reflector, but intensities of the neutron flux and gamma dose with new reflector were rather smaller than those with old reflector. The calculated results showed the same tendency as that of the experimental results. Therefore, the clinical trials of BNCT in JRR-4 could be restarted.     

Analysis of neutron and gamma-ray streaming along the maze of NRCAM thallium production target room.

Study of the shield performance of a thallium-203 production target room has been investigated in this work. Neutron and gamma-ray equivalent dose rates at various points of the maze are calculated by simulating the transport of streaming neutrons, and photons using Monte Carlo method. For determination of neutron and gamma-ray source intensities and their energy spectrum, we have applied SRIM 2003 and ALICE91 computer codes to Tl target and its Cu substrate for a 145 microA of 28.5 MeV protons beam. The MCNP/4C code has been applied with neutron source term in mode n p to consider both prompt neutrons and secondary gamma-rays. Then the code is applied for the prompt gamma-rays as the source term. The neutron-flux energy spectrum and equivalent dose rates for neutron and gamma-rays in various positions in the maze have been calculated. It has been found that the deviation between calculated and measured dose values along the maze is less than 20%.     

Radiotherapy dose calculations in the presence of hip prostheses.

The high density and atomic number of hip prostheses for patients undergoing pelvic radiotherapy challenge our ability to accurately calculate dose. A new clinical dose calculation algorithm, Monte Carlo, will allow accurate calculation of the radiation transport both within and beyond hip prostheses. The aim of this research was to investigate, for both phantom and patient geometries, the capability of various dose calculation algorithms to yield accurate treatment plans. Dose distributions in phantom and patient geometries with high atomic number prostheses were calculated using Monte Carlo, superposition, pencil beam, and no-heterogeneity correction algorithms. The phantom dose distributions were analyzed by depth dose and dose profile curves. The patient dose distributions were analyzed by isodose curves, dose-volume histograms (DVHs) and tumor control probability/normal tissue complication probability (TCP/NTCP) calculations. Monte Carlo calculations predicted the dose enhancement and reduction at the proximal and distal prosthesis interfaces respectively, whereas superposition and pencil beam calculations did not. However, further from the prosthesis, the differences between the dose calculation algorithms diminished. Treatment plans calculated with superposition showed similar isodose curves, DVHs, and TCP/NTCP as the Monte Carlo plans, except in the bladder, where Monte Carlo predicted a slightly lower dose. Treatment plans calculated with either the pencil beam method or with no heterogeneity correction differed significantly from the Monte Carlo plans.     

Benchmark experiments for cyclotron-based neutron source for BNCT.

In the previous study, we found the feasibility of a cyclotron-based BNCT using the Ta(p,n) neutrons at 90 degrees bombarded by 50 MeV protons, and the iron, AlF(3), Al and (6)LiF moderators by simulations using the MCNPX code. In order to validate the simulations to realize the cyclotron-based BNCT, we measured the epithermal neutron energy spectrum passing through the moderators with our new spectrometer consisting of a (3)He gas counter covered with a silicon rubber loaded with (nat)B and polyethylene moderator and the depth distribution of the reaction rates of (197)Au(n,gamma)(198)Au in an acrylic phantom set behind the rear surface of the moderators. The measured results were compared with the calculations using the MCNPX code. We obtained the good agreement between the calculations and measurements within approximately 10% for the neutron energy spectra and within approximately 20% for the depth distribution of the reaction rates of (197)Au(n,gamma)(198)Au in the phantom. The comparison clarified a good accuracy of the calculation of the neutron energy spectrum passing through the moderator and the thermalization in a phantom. These experimental results will be a good benchmark data to evaluate the accuracy of the calculation code.     

BNCT dose calculation in irregular fields using the sector integration method.

Irregular fields for boron neutron capture therapy (BNCT) have been already proposed to spare normal tissue in the treatment of superficial tumors. This added dependence would require custom measurements and/or to have a secondary calculation system. As a first step, we implemented the sector-integration method for irregular field calculation in a homogeneous medium and on the central beam axis. The dosimetric responses (fast neutron and photon dose and thermal neutron flux), are calculated by sector integrating the measured responses of circular fields over the field boundary. The measurements were carried out at our BNCT facility, the RA-6 reactor (Argentina). The input data were dosimetric responses for circular fields measured at different depths in a water phantom using ionisation and activation techniques. Circular fields were formed by shielding the beam with two plates: borated polyethilene plus lead. As a test, the dosimetric responses of a 7x4 cm(2) rectangular field, were measured and compared to calculations, yielding differences less than 3% in equivalent dose at any depth indicating that the tool is suitable for redundant calculations.     

Geometry corrections for cylindrical neutron area survey meters

MCNP calculations have been performed to investigate the effects of beam divergence on the response of selected cylindrical neutron area survey meters irradiated by selected neutron sources. By comparing the results to calculations performed using plane-parallel beam irradiations, geometry correction factors have been calculated that can be applied to instrument calibrations. The results indicate that the effective centres of cylindrical detectors may not lie on the axis of symmetry, as previously assumed.     

Neutron fluence depth profiles in water phantom on epithermal beam of LVR-15 research reactor

Horizontal channel with epithermal neutron beam at the LVR-15 research reactor is used mainly for boron neutron capture therapy. Neutron fluence depth profiles in a water phantom characterise beam properties. The neutron fluence (approximated by reaction rates) depth profiles were measured with six different types of activation detectors. The profiles were determined for thermal, epithermal and fast neutrons.     

A preliminary investigation of the EBT2 radiochromic films response to low energy fast neutrons

EBT2 radiochromic films were used to study the relative dose distribution of the neutron field. The correlation between the beam current and the optical density showed good linear dependence with a correlation coefficient exceeding 98%. At any given beam energy, neutron dose rates can be changed by a factor of 40 without changing the neutron spectrum. This result is consistent with what was found by the Tissue Equivalent Proportional Counter measurements. The uniformity of the neutron field was inspected by the optical density profile of the exposed film.     

15MV医用直线加速器光中子蒙特卡罗模拟

目的 研究高能医用直线加速器运行过程中因光核反应所形成的光中子辐射场。方法 利用蒙特卡罗(MC)程序模拟Clinic 2300CD型医用电子加速器15 MV X射线模式下光中子污染,掌握机头内不同位置光中子能谱和不同照射野下等中心处中子周围剂量当量变化,分析光中子在等中心平面内剂量分布和水模体中剂量衰减。结果 准直器关闭时,加速器机头内靶、主准直器、均整器和多叶准直器下表面的光中子平均能量分别为1.08、1.20、0.35、0.30MeV;等中心处中子周围剂量当量随着照射野的增大先增大后减少,在30 cm × 30 cm照射野下达到最大;随着测点在水模体中的深度增加,中子通量先增加后减小,而中子剂量却在逐渐减小;不同照射野下,光中子剂量率在水模体深度20 cm处,基本都接近本底。结论 探究高能医用直线加速器机头光中子谱和剂量分布特点,以及光中子在水模体内剂量沉积规律,能为进一步研究高能医用直线加速器光中子污染对患者产生的附加剂量提供支持。     

Spatial and spectral characteristics of a compact system neutron beam designed for BNCT facility

The development of suitable neutron sources and neutron beam is critical to the success of Boron Neutron Capture Therapy (BNCT). In this work a compact system designed for BNCT is presented. The system consists of ²⁵²Cf fission neutron source and a moderator/reflector/filter/shield assembly. The moderator/reflector/filter arrangement has been optimized to maximize the epithermal neutron component which is useful for BNCT treatment of deep seated tumors with the suitably low level of beam contamination. The MCMP5 code has been used to calculate the different components of neutrons, secondary gamma rays originating from ²⁵²Cf source and the primary gamma rays emitted directly by this source at the exit face of the compact system. The fluence rate distributions of such particles were also computed along the central axis of a human head phantom.     

~(252)Cf裂变中子源在组织等效模体中的中子和γ辐射剂量分布的计算

目的　计算2 5 2 Cf裂变中子源的中子和γ辐射在组织等效模体内的剂量分布 ,为使用2 5 2 Cf裂变中子源进行中子放疗提供有用的剂量学参数。方法　建立2 5 2 Cf源和组织等效模体的三维几何计算模型 ,利用蒙特卡罗方法进行中子和γ辐射联合输运计算。结果　计算了两种医用2 5 2 Cf裂变中子源在水、血液、肌肉、皮肤、骨骼和肺组织等效材料构成的模体中距源不同距离点处的中子和γ辐射吸收剂量。结论　蒙特卡罗计算结果与文献数据以及使用双电离室实验测量的结果符合得较好。对2 5 2 Cf裂变中子源在 5种组织材料构成的模体中中子和γ辐射的剂量分布进行了比较 ,使用水作为组织等效材料对2 5 2 Cf裂变中子源在以肌肉、血液和皮肤构成的局部组织内的剂量分布进行模拟计算 ,可取得比较可靠的结果。     

Calculated DNA damage from gadolinium Auger electrons and relation to dose distributions in a head phantom

PURPOSE: To calculate the number of 157Gadolinium (157Gd) neutron capture induced DNA double strand breaks (DSB) in tumor cells resulting from epithermal neutron irradiation of a human head when the peak tissue dose is 10 Gy. To assess the lethality of these Gd induced DSB. MATRIALS AND METHODS: DNA single and double strand breaks from Auger electrons emitted during 157Gd(n,gamma) events were calculated using an atomistic model of B-DNA with higher-order structure. When combined with gadolinium neutron capture reaction rates and neutron and photon physical dose rates calculated from the radiation transport through a model of the human head with explicit tumors, peak tissue dose can be related to the number of Auger electron induced DSB in tumor cell DNA. The lethality of these DNA DSB were assessed through a comparison with incorporated 125I decay cell survival curves and second comparison with the number of DSB resulting from neutron and photon interactions. RESULTS: These calculations on a molecular scale (microscopic calculations) indicate that for incorporated 157Gd, each neutron capture reaction results in an average of 1.56 +/- 0.16 DNA single strand breaks (SSB) and 0.21 +/- 0.04 DBS in the immediate vicinity (approximately 40 nm) of the neutron capture. In an example case of Gd Neutron Capture Therapy (GdNCT), a 1 cm radius midline tumor, peak normal tissue dose of 10 Gy, and a tumor concentration of 1000 ppm Gd, result in a maximum of 140 +/- 27 DSBs per tumor cell. CONCLUSIONS: The number of DSB from the background radiation components is one order of magnitude lower than the Gd Auger electron induced DSB. The cell survival of mammalian cell lines with a similar amount of complex DSB induced from incorporated 125I decay yield one to two magnitudes of cell killing. These two points indicate that gadolinium auger electrons could significantly contribute to cell killing in GdNCT.     

Characteristics of the new THOR epithermal neutron beam for BNCT.

A characterization of the new Tsing Hua open-pool reactor (THOR) epithermal neutron beam designed for boron neutron capture therapy (BNCT) has been performed. The facility is currently under construction and expected in completion in March 2004. The designed epithermal neutron flux for 1 MW power is 1.7x10(9)n cm(-2)s(-1) in air at the beam exit, accompanied by photon and fast neutron absorbed dose rates of 0.21 and 0.47 mGys(-1), respectively. With (10)B concentrations in normal tissue and tumor of 11.4 and 40 ppm, the calculated advantage depth dose rate to the modified Snyder head phantom is 0.53RBE-Gymin(-1) at the advantage depth of 85 mm, giving an advantage ratio of 4.8. The dose patterns determined by the NCTPlan treatment planning system using the new THOR beam for a patient treated in the Harvard-MIT clinical trial were compared with results of the MITR-II M67 beam. The present study confirms the suitability of the new THOR beam for possible BNCT clinical trials.     

Effective dose evaluation for BNCT brain tumor treatment based on voxel phantoms

For BNCT treatments, in addition to tumor target doses, non-negligible doses will result in all the remaining organs of the body. This work aims to evaluate the effective dose as well as the average absorbed doses of each of organs of patients with brain tumor treated in the BNCT epithermal neutron beam at THOR. The effective doses were evaluated according to the definitions of ICRP Publications 60 and 103 for the reference male and female computational phantoms developed in ICRP Publication 110 by using the MCNP5 Monte Carlo code with the THOR-Y09 beam source. The effective dose acquired in this work was compared with the results of our previous work calculated for an adult hermaphrodite mathematical phantom. It was found that the effective dose for the female voxel phantom is larger than that for the male voxel phantom by a factor of 1.2–1.5 and the effective dose for the voxel phantom is larger than that for the mathematical phantom by a factor of 1.3–1.6. For a typical brain tumor BNCT, the effective dose was calculated to be 1.51 Sv and the average absorbed dose for eye lenses was 1.07 Gy.     

A feasibility study of the Tehran research reactor as a neutron source for BNCT

Investigation on the use of the Tehran Research Reactor (TRR) as a neutron source for Boron Neutron Capture Therapy (BNCT) has been performed by calculating and measuring energy spectrum and the spatial distribution of neutrons in all external irradiation facilities, including six beam tubes, thermal column, and the medical room. Activation methods with multiple foils and a copper wire have been used for the mentioned measurements. The results show that (1) the small diameter and long length beam tubes cannot provide sufficient neutron flux for BNCT; (2) in order to use the medical room, the TRR core should be placed in the open pool position, in this situation the distance between the core and patient position is about 400 cm, so neutron flux cannot be sufficient for BNCT; and (3) the best facility which can be adapted for BNCT application is the thermal column, if all graphite blocks can be removed. The epithermal and fast neutron flux at the beginning of this empty column are 4.12×10⁹ and 1.21×10⁹ n/cm²/s, respectively, which can provide an appropriate neutron beam for BNCT by designing and constructing a proper Beam Shaping Assembly (BSA) structure.     

Changes in relative biological effectiveness with depth of the Clatterbridge neutron therapy beam

We have measured the biological equivalence of the Clatterbridge neutron therapy beam [p(62)-Be] and the Hammersmith neutron therapy beam [d(16)-Be] using the mouse intestinal crypt assay. The ratio (NDR) of Clatterbridge neutron (n + gamma) dose relative to Hammersmith neutron dose (n + gamma) was found to be 1.2-1.13 over a dose/fraction range of 1.8-9 Gy at 2 cm deep in a Perspex phantom. It is shown that the effectiveness of the Clatterbridge beam was reduced with penetration into the phantom because of hardening of the beam to a maximum reduction of 11% at 12 cm deep in the phantom. The hardening of the beam with depth of penetration will need to be taken into account by clinicians in assessing the tumour dose and tissue tolerance. Relative biological effectiveness values for the Clatterbridge and Hammersmith neutron beams were also measured. All neutron doses for both Hammersmith and Clatterbridge beams are total doses (n + gamma) which comply with the European protocol for neutron dosimetry and include the gamma-ray component of dose.     

Gadolinium neutron capture in glioblastoma multiforme cells

Purpose: A proof of principle for cell killing by Gadolinium (Gd) neutron capture in Magnevist® preloaded Glioblastoma multiforme (GBM) cells is provided.

Materials and methods: U87cells were pre-loaded with 5 mg/ml Magnevist® (Gd containing compound) and irradiated using an enhanced neutron beam developed at NIU Institute for Neutron Therapy at Fermilab. These experiments were possible because of an enhanced fast neutron therapy assembly designed to use the fast neutron beam at Fermilab to deliver a neutron beam containing a greater fraction of thermal neutrons and because of the development of improved calculations for dose for the enhanced neutron beam. Clonogenic response was determined.

Results: U87 cell survival after γ irradiation, fast neutron irradiation and irradiation with the enhanced neutron beam in the presence or absence of Magnevist® were determined.

Conclusions: U87 cells were the least sensitive to γ radiation, and increasingly sensitive to fast neutron irradiation, irradiation with the enhanced neutron beam and finally a significant enhancement in cell killing was observed for U87 cells preloaded with Magnevist®. The sensitivity of U87 cells pre-loaded with Magnevist® and then irradiated with the enhanced neutron beam can at least in part be attributed to the Auger electrons emitted by the neutron capture event.     

The neutron therapy facility (DT, 14 MeV) at the radiotherapy department of the university hospital Hamburg-Eppendorf

The neutron therapy facility at the Radiotherapy Department of the University Hospital Hamburg-Eppendorf is described. This unit has been developed for clinical purposes according to the initiative and conception of the radiotherapist by AEG/Fed. Rep. of Germany and RDI/USA since 1969. The installation was completed at the beginning of 1974. Special treatment head and bed systems allow isocentric treatment and arc or multiple port therapy. For routine work operation conditions of 8 to 12 mA total beam current and 500 kV accelerating voltage are used. The neutron output at 12 mA is about 3.5 x 10(12) n/s giving a phantom dose rate of more than 20 rad/min for a field size of 17.8 x 17.8 cm2 at 80 cm source-skin distance. Technical installations for improvement of dose rate and half-life of the target are planned. Results of physical measurements about neutron energy distributions, contributions from neutrons and gamma-rays to the total absorbed dose, build-up effect, axial and lateral dose distributions as well as isodose profiles for different field sizes in a homogeneous phantom are presented. Bewteen February 1976 and November 1977 up to 180 patients have been treated.     

Verification of eye lens dose in IMRT by MOSFET measurement

The eye lens is recognized as one of the most radiosensitive structures in the human body. The widespread use of intensity-modulated radiotherapy (IMRT) complicates dose verification and necessitates high standards of dose computation. The purpose of this work was to assess the computed dose accuracy of eye lens through measurements using a metal–oxide–semiconductor field-effect transistor (MOSFET) dosimetry system. Sixteen clinical IMRT plans of head and neck patients were copied to an anthropomorphic head phantom. Measurements were performed using the MOSFET dosimetry system based on the head phantom. Two MOSFET detectors were imbedded in the eyes of the head phantom as the left and the right lens, covered by approximately 5-mm-thick paraffin wax. The measurement results were compared with the calculated values with a dose grid size of 1 mm. Sixteen IMRT plans were delivered, and 32 measured lens doses were obtained for analysis. The MOSFET dosimetry system can be used to verify the lens dose, and our measurements showed that the treatment planning system used in our clinic can provide adequate dose assessment in eye lenses. The average discrepancy between measurement and calculation was 6.7 ± 3.4%, and the largest discrepancy was 14.3%, which met the acceptability criterion set by the American Association of Physicists in Medicine Task Group 53 for external beam calculation for multileaf collimator-shaped fields in buildup regions.