Dose-rate scaling factor estimation of THOR BNCT test beam.

In 1998, an epithermal neutron test beam was designed and constructed at the Tsing Hua Open-Pool Reactor (THOR) for the purpose of preliminary dosimetric experiments in boron neutron capture therapy (BNCT). A new epithermal neutron beam was designed at this facility, and is currently under construction, with clinical trials targeted in late 2004. Depth dose-rate distributions for the THOR BNCT test beam have been measured by means of activation foil and dual ion chamber techniques. Neutron and structure-induced gamma spectra measured at the test beam exit were configured into a source function for the Monte Carlo-based treatment planning code NCTPlan. Dose-rate scaling factors (DRSFs) were determined to normalize computationally derived dose-rate distributions with experimental measurements in corresponding mathematical and physical phantoms, and to thus enable accurate treatment planning using the NCTPlan code. A similar approach will be implemented in characterizing the new THOR epithermal beam in preparation for clinical studies. This paper reports the in-phantom calculated and experimental dosimetry comparisons and derived DRSFs obtained with the THOR test beam.     

In-phantom characterisation studies at the Birmingham Accelerator-Generated epIthermal Neutron Source (BAGINS) BNCT facility.

A broad experimental campaign to validate the final epithermal neutron beam design for the BNCT facility constructed at the University of Birmingham concluded in November 2003. The final moderator and facility designs are overviewed briefly, followed by a summary of the dosimetric methods and presentation of a small subset of the results from this campaign. The dual ionisation chamber technique was used together with foil activation to quantify the fast neutron, photon, and thermal neutron beam dose components in a large rectangular phantom exposed to the beam with a 12 cm diameter beam delimiter in place. After application of a normalisation factor, dose measurements agree with in-phantom MCNP4C predictions within 10% for the photon dose, within 10% for thermal neutron dose, and within 25% for the proton recoil dose along the main beam axis.     

BNCT dose distribution in liver with epithermal D-D and D-T fusion-based neutron beams.

Recently, a new application of boron neutron capture therapy (BNCT) treatment has been introduced. Results have indicated that liver tumors can be treated by BNCT after removal of the liver from the body. At Lawrence Berkeley National Laboratory, compact neutron generators based on (2)H(d,n)(3)He (D-D) or (3)H(t,n)(4)He (D-T) fusion reactions are being developed. Preliminary simulations of the applicability of 2.45 MeV D-D fusion and 14.1 MeV D-T fusion neutrons for in vivo liver tumor BNCT, without removing the liver from the body, have been carried out. MCNP simulations were performed in order to find a moderator configuration for creating a neutron beam of optimal neutron energy and to create a source model for dose calculations with the simulation environment for radiotherapy applications (SERA) treatment planning program. SERA dose calculations were performed in a patient model based on CT scans of the body. The BNCT dose distribution in liver and surrounding healthy organs was calculated with rectangular beam aperture sizes of 20 cm x 20 cm and 25 cm x 25 cm. Collimator thicknesses of 10 and 15 cm were used. The beam strength to obtain a practical treatment time was studied. In this paper, the beam shaping assemblies for D-D and D-T neutron generators and dose calculation results are presented.     

Dose calculation and in-phantom measurement in BNCT using response matrix method

In-phantom measurement of physical dose distribution is very important for Boron Neutron Capture Therapy (BNCT) planning validation. If any changes take place in therapeutic neutron beam due to the beam shaping assembly (BSA) change, the dose will be changed so another group of simulations should be carried out for dose calculation. To avoid this time consuming procedure and speed up the dose calculation to help patients not wait for a long time, response matrix method was used. This procedure was performed for neutron beam of the optimized BSA as a reference beam. These calculations were carried out using the MCNPX, Monte Carlo code. The calculated beam parameters were measured for a SNYDER head phantom placed 10 cm away from beam the exit of the BSA. The head phantom can be assumed as a linear system and neutron beam and dose distribution can be assumed as an input and a response of this system (head phantom), respectively. Neutron spectrum energy was digitized into 27 groups. Dose response of each group was calculated. Summation of these dose responses is equal to a total dose of the whole neutron/gamma spectrum. Response matrix is the double dimension matrix (energy/dose) in which each parameter represents a depth–dose resulted from specific energy. If the spectrum is changed, response of each energy group may be differed. By considering response matrix and energy vector, dose response can be calculated. This method was tested for some BSA, and calculations show statistical errors less than 10%.     

用于硼中子俘获治疗的超热中子束理论设计

目的 设计用于硼中子俘获治疗(BNCT)的超热中子束理论方案。方法 基于清华大学试验核反应堆,以其1号孔道为材料布放孔道,设计了由慢化材料、热中子吸收材料、γ屏蔽材料组成,但材料布放位置具有差异的5种理论方案;利用蒙特卡罗(MC)模拟方法,分别计算5种方案束出口处的中子注量率、剂量率及γ剂量率值,通过与BNCT技术指标对比,从5种方案中选择一种合适的方案。结果 得到了一个符合BNCT各项技术指标的超热中子束理论方案,其慢化材料厚度为53.5 cm、热中子吸收材料厚度为2 mm、γ屏蔽材料厚度为9 cm。结论 本研究给出的超热中子束理论方案为基于反应堆实现BNCT提供一定的理论参考。     

A preliminary study on using the radiochromic film for 2D beam profile QC/QA at the THOR BNCT facility

The GAFCHROMIC^® EBT2 dosimetry film has been studied as a rapid QC/QA tool for 2D dose profile mapping in the BNCT beam at THOR. The pixel values of the EBT2 film image were converted to the 2D dose profile using a dose calibration curve obtained by 6-MV X-ray. The reproducibility of the 2D dose profile measured using the EBT2 film in the PMMA phantom was preliminarily found to be acceptable with uncertainties within about ±2 to ±3.5%. It is found that the EBT2 measured dose profile consisted of both gamma-ray components and neutron contributions. Therefore, the dose profile measured using the EBT2 film is significantly different from the neutron flux profile measured using the indirect neutron radiography method. Further study of the influence of neutrons to the response of the EBT2 film is indispensible for the absolute dose profile determination in a BNCT beam.     

MCNP study for epithermal neutron irradiation of an isolated liver at the Finnish BNCT facility.

A successful boron neutron capture treatment (BNCT) of a patient with multiple liver metastases has been first given in Italy, by placing the removed organ into the thermal neutron column of the Triga research reactor of the University of Pavia. In Finland, FiR 1 Triga reactor with an epithermal neutron beam well suited for BNCT has been extensively used to irradiate patients with brain tumors such as glioblastoma and recently also head and neck tumors. In this work we have studied by MCNP Monte Carlo simulations, whether it would be beneficial to treat an isolated liver with epithermal neutrons instead of thermal ones. The results show, that the epithermal field penetrates deeper into the liver and creates a build-up distribution of the boron dose. Our results strongly encourage further studying of irradiation arrangement of an isolated liver with epithermal neutron fields.     

Dosimetry of Small Circular Fields for 6-MeV Electron Beams

Small field electron beams used in the clinic present complex dosimetry. This investigation reports dosimetric characteristics of small 6-MeV electron field measurements and Monte Carlo calculations, and examines their impact on patient dosimetry. Radiographic film and ionization chamber were used for dosimetric measurements for a 6-MeV electron beam from a Varian 21EX linac. A set of circular cerrobend cutouts with diameters ranging from 2 to 3.0 cm was made. A clinical Monte Carlo algorithm was used to calculate dosimetric parameters. As the cutout diameter decreased, the build-up portion of the percentage depth dose (PDD) curves shifted toward the phantom surface, the depth of maximum dose, d_max, decreased from 1.2 cm to 0.6 cm, but the practical range, R_p, remained constant at about 3.1 cm. The absorbed dose rate, ?_r, decreased linearly with decreasing diameter. Profile coverage ratios CR (ie, the ratio of a given isodose line to the cutout diameter) at d_max, for 90% and 80%, CR₉₀, CR₈₀, are approximately 55% and 73%, respectively. As the cutout diameter decreased, the 90% to 10% penumbra to diameter ratio, PR, increased from 0.49 to 0.56 for 3- to 2-cm cutouts. The total 90% to 10% penumbra was about the same size as the cutout diameter. The measured output factors were in good agreement with Monte Carlo calculation within 2.2%. Accurate small electron field measurements were performed using parallel plate ion chamber and film. The data were well supported by Monte Carlo calculations. These data facilitate routine clinical treatments for small cutouts as d_max, ?_r, and other data can be quickly obtained instead of performing labor-intensive individual patient measurements.     

Design and construction of shoulder recesses into the beam aperture shields for improved patient positioning at the FiR 1 BNCT facility.

Improvements have been made at the FiR 1 BNCT facility to ease the positioning of the patient with a tumor in the head and neck region into a lateral neutron beam. Shoulder recesses were constructed horizontally on both sides of the beam aperture. When shoulder recesses are not needed, they are filled with neutron attenuating filling blocks. MCNP simulations using an anthropomorphic human model BOMAB phantom showed that the main contribution to the increase in the effective dose to the patient's body due to the shoulder recesses was from the neutron dose of the arm. In a position when one arm is inside the shoulder recess, the maximal effective dose of the patient was estimated to be 0.7Sv/h. Dose measurements using the twin ionization chamber technique showed that the neutron dose increased on the sides as predicted by the MCNP model but there was no noticeable change in the gamma doses. When making the recesses into the lithium containing neutron shield material tritium contamination was confined using an underpressurized glove box and machine tools with local exhaust. The shoulder recesses give space for more flexible patient positioning and can be considered as a significant improvement of the Finnish BNCT facility.     

p（35）Be快中子辐射场生物剂量参数的测定

报道了ｐ（３５）Ｂｅ快中子、γ混合辐射场生物剂量参数，说明了采用双电离室方法测量ｐ（３５）Ｂｅ快中子时，剂量计算中有关因子和参数的确定原则和估算结果；给出了辐射场的中子／γ比，照射野内组织比释动能率分布，以及由比释动能率计算小鼠全身与局部照射和血液样品照射时的吸收剂量转换系数等剂量常参数；最后还对剂量测量的误差问题做了分析。     

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.     

A review of boron neutron capture therapy (BNCT) and the design and dosimetry of a high-intensity, 24 keV, neutron beam for BNCT research

This paper reviews the development of boron neutron capture therapy (BNCT) and describes the design and dosimetry of an intermediate energy neutron beam, developed at the Harwell Laboratory, principally for BNCT research. Boron neutron capture therapy is a technique for the treatment of gliomas (a fatal form of brain tumour). The technique involves preferentially attaching 10B atoms to tumour cells and irradiating them with thermal neutrons. The thermal neutron capture products of 10B are short range and highly damaging, so they kill the tumour cells, but healthy tissue is relatively undamaged. Early trials required extensive neurosurgery to exposure the tumour to the thermal neutrons used and were unsuccessful. It is thought that intermediate-energy neutrons will overcome many of the problems encountered in the early trials, because they have greater penetration prior to thermalization, so that surgery will not be required. An intermediate-energy neutron beam has been developed at the Harwell Laboratory for research into BNCT. Neutrons from the core of a high-flux nuclear reactor are filtered with a combination of iron, aluminium and sulphur. Dosimetry measurements have been made to determine the neutron and gamma-ray characteristics of this beam, and to monitor them throughout the four cycles used for BNCT research. The beam is of high intensity (approximately 2 x 10(7) neutrons cm-2 s-1, equivalent to a neutron kerma rate in water of 205 mGy h-1) and nearly monoenergetic (93% of the neutrons have energies approximately 24 keV, corresponding to 79% of the neutron kerma rate).     

A phantom experiment for the evaluation of whole body exposure during BNCT using cyclotron-based epithermal neutron source (C-BENS)

At Kyoto University Research Reactor Institute (KURRI), cyclotron-based epithermal neutron source was installed in December 2008, and the supplementary construction works have been performed. As of December 2010, the various irradiation characteristics important for BNCT were mostly evaluated. The whole body exposure during BNCT medical irradiation is one of the important characteristics.In this article, measurements of absorbed dose for thermal and fast neutrons and gamma-ray at ten positions corresponding to important organs are reported.     

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.     

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。     

The TL analysis methods used to determine absorbed gamma doses in vivo for the BNCT patients treated at FiR 1.

The gamma dose determination using thermoluminescent (TL) dosimeters in mixed neutron-gamma fields, such as in boron neutron capture therapy (BNCT), is difficult due to the thermal neutron sensitivity of the detectors; especially when equipment capable of glow curve analysis is not available. The two TL analysis methods used previously in Finnish BNCT to correct the measured TL signal to obtain absorbed gamma dose in vivo were studied and compared, and an enhanced method was introduced. The three TL methods were found surprisingly consistent despite, e.g. the rough estimate made in the first method.     

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.     

Investigation of irradiation conditions for recurrent breast cancer in JRR-4

Clinical trials of boron neutron capture therapy (BNCT) for recurrent breast cancers are considered at Japan Research Reactor No. 4 (JRR-4). In this study, the irradiation technique for a total mastectomy patient with recurrent cancer was optimized by dosimetric calculations using JAEA computational dosimetry system (JCDS). The evaluation was performed using an en face technique and a tangents technique with thermal neutron beam mode at JRR-4. The results revealed that equivalent doses of lung, heart, liver and skin were similar in each irradiation technique due to the isotropic scattering of thermal neutrons in the body. On the other hand, the irradiation time with the tangents technique was a few times longer than with the en face technique. We concluded that the en face technique was an optimal irradiation technique for recurrent breast cancers using thermal neutron beam mode in terms of shorter irradiation time and easier patient positioning.     

~(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。     

A shielding design for an accelerator-based neutron source for boron neutron capture therapy.

Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a (7)Li target, a target heat removal system (HRS), a moderator assembly, and a treatment room. The intent of this paper is to demonstrate the advantages of a shielded moderator assembly design, in terms of material requirements necessary to adequately protect radiation personnel located outside a treatment room for BNCT, over an unshielded moderator assembly design.