Improvement of dose calculation accuracy for BNCT dosimetry by the multi-voxel method in JCDS.

To carry out boron neutron capture therapy (BNCT) clinical trials based on accurate dosimetry of several dose components given to a patient, we had developed the JAERI computational dosimetry system (JCDS), which can determine the absorbed doses by numerical simulation. The verification results of initial version of JCDS indicated that JCDS causes characteristic discrepancy, when JCDS estimates a sharp change arising such as near the surface. The aim of this study is to improve the accuracy of the BNCT dosimetry efficiently. The multi-voxel calculation method that reconstructs the original voxel model by combining several voxel cell sizes such as 0.125, 1 and 8 cm(3) has been developed. To verify the accuracy of the method, the calculation results were compared with the phantom experimental data. Furthermore, to verify its practicality to BNCT, retrospective evaluation of an actual BNCT in JRR-4 was performed by the multi-voxel method. The results of the comparison with the phantom experiments demonstrated that the calculation accuracy for the distributions of the thermal neutron flux was improved by employing the multi-voxel method. The computing time using the multi-voxel method increased only approximately 33% compared to the conventional uniform 1cm(3) voxel method. These results proved that the multi-voxel calculation enables JCDS to more accurately estimate the absorbed doses to a patient by efficient calculations.     

Multistep Lattice-Voxel method utilizing lattice function for Monte-Carlo treatment planning with pixel based voxel model

Treatment planning for boron neutron capture therapy generally utilizes Monte-Carlo methods for calculation of the dose distribution. The new treatment planning system JCDS-FX employs the multi-purpose Monte-Carlo code PHITS to calculate the dose distribution. JCDS-FX allows to build a precise voxel model consisting of pixel based voxel cells in the scale of 0.4×0.4×2.0 mm³ voxel in order to perform high-accuracy dose estimation, e.g. for the purpose of calculating the dose distribution in a human body. However, the miniaturization of the voxel size increases calculation time considerably. The aim of this study is to investigate sophisticated modeling methods which can perform Monte-Carlo calculations for human geometry efficiently. Thus, we devised a new voxel modeling method “Multistep Lattice-Voxel method,” which can configure a voxel model that combines different voxel sizes by utilizing the lattice function over and over. To verify the performance of the calculation with the modeling method, several calculations for human geometry were carried out. The results demonstrated that the Multistep Lattice-Voxel method enabled the precise voxel model to reduce calculation time substantially while keeping the high-accuracy of dose estimation.     

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.     

基于人体体素模型的光子外照射事故剂量重建方法研究

目的 研究光子外照射事故下人体的剂量重建方法,并在局部剂量分布层面上验证方法的准确性。方法 基于开源蒙特卡罗代码Geant4,使用国际辐射防护委员会（ICRP）103号建议书推荐的人体体素模型,研究外照射事故照射条件下的剂量重建方法,实现全身平均吸收剂量、器官吸收剂量和局部剂量分布的评价。为了对建立的方法进行验证,使用组织等效的物理仿真模型ART;通过CT扫描,建立起其分辨率为1.57 mm×1.57 mm×10.00 mm的体素模型;在标准辐射场下进行一系列热释光剂量计（TLD）照射实验,比较实验和剂量重建模拟的结果。结果 实验测量值的综合相对不确定度为10.9%,剂量重建模拟值的综合相对不确定度在非组织交界面处为7.10%,在组织交界面处为16.6%。对451个测量点位进行统计分析,模拟值除以测量值的均值为0.972,标准差为0.083 8,在0.95~1.05,0.90~1.10和0.80~1.20范围内的比例分别为49.2%,79.4%和96.4%。结论 基于人体体素模型的蒙特卡罗剂量重建方法无论在全身或器官层面,还是在局部剂量分布层面都满足实际使用的精度要求,可作为外照射事故下对受照者进行剂量评估的有力工具,为诊断和救治提供支持。     

利用MRI三维男性人体模型对航天员所受空间辐射的估算

目的 建立空间辐射环境下,航天员器官所受辐射剂量及对其健康危险的计算方法.方法 利用符合航天员体征的人体MRI图像,建立三维男性人体模型及辐射数据库,并结合蒙特卡罗粒子输运程序GEANT4用于剂量计算.结果 我们得到了模拟各向同性抽样情况下,10 MeV到500 MeV单个质子对人体辐射敏感器官的吸收剂量及有效剂量.结论 在航天员体征三维人体模型及辐射数据库的基础上.利用空间舱内测量质子谱,得到了舱内累计剂量.计算的皮肤剂量为148.6 μGy/d.该值与美国和俄罗斯发表的数据100-300μGy/d比较接近.     

Production of 177Lu at the new research reactor FRM-II: Irradiation yield of 176Lu(n,gamma)177Lu.

Due to its physical and chemical characteristics, 177Lu is a very attractive radionuclide for use in nuclear medicine. This paper introduces a method for a precise calculation of the irradiation yield of 177Lu produced by neutron activation of 176Lu in a nuclear reactor. The calculation is based on the Westcott convention which requires the knowledge of the neutron flux parameters. In this work, the neutron flux parameters of the new research reactor FRM-II (Garching, Germany) were determined and the stability of thermal neutron flux and thermal neutron flux temperature was monitored. The comparison of theoretically calculated and experimentally determined yield for Lu176(n,gamma)Lu177 reaction is presented.     

CR39中子个人剂量监测的不确定度评定

目的 通过对CR39中子个人剂量监测的不确定度评定,定量分析中子个人监测结果的可靠性,提高中子个人剂量监测的质量水平。方法 建立适当数学模型,通过对个人剂量计各个环节因素引入不确定度的计算,评定CR39中子个人剂量监测的扩展不确定度。结果 中子个人监测的A类不确定度主要来自测量值和本底,测量值为A类不确定度的最大分量,为11.68%。引入B类不确定度主要为能量响应为7.85%、剂量线性检验为4.0%、距离响应为1.68%和蚀刻条件为3.0%等。相对扩展不确定度约为32%,k=2。结论 放射性同位素场所不确定度评定中,对于某一较小能量范围的中子（即中子的能量分群）,应采用符合其能量响应特性的中子探测器。     

Methods for estimating radiation doses received by commercial aircrew

INTRODUCTION: Radiation doses received onboard aircraft are monitored in Europe to protect aircrew in accordance with a European Union directive. The French Aviation Authorities have developed a system called SIEVERT, using calculation codes to monitor effective radiation doses. METHODS: For the galactic cosmic ray component, a 3-D world map of effective dose rates is computed using available operational codes. Detailed flight plans are used to ensure sufficient precision. For the solar particle event component, a semi-empirical model called SiGLE has been developed to calculate a time-dependent map of effective dose rates in the course of the event. SiGLE is based on particle transport code results and measurements during solar particle events onboard Concorde airplanes. RESULTS: We present a comparison of the calculated effective radiation dose and measured dose equivalent for various flights onboard Air France aircraft. The agreement is within 15%, which is about the precision of the state-of-the-art dosimetric measurements. Meteorological effects on the dose calculation appear to be negligible. Preliminary results based on solar particle events observed since 1942 with ionization chambers and neutron monitors are given. CONCLUSIONS: The present analysis shows that for the galactic cosmic ray component, monthly world maps based on neutron monitor observations are sufficient to ensure a precision of about 20% on the dose estimate for each flight. For the past 40 yr, according to the model SiGLE, none of the solar events has given an effective radiation dose larger than 1 mSv for flights on the most exposed routes.     

(125)I liquid in an intracranial balloon: TG-43 formalism for an extended source

PURPOSE: To present a calculation formalism for spherical homogeneous liquid brachytherapy sources and to analyze the difficulties encountered in numerical integration over non-spherical source volumes. METHODS AND MATERIALS: TG-43 formalism provides a general definition of the geometric factor for distributed source distributions. Here the geometric factor and dose for a spherical source are computed for an (125)I balloon. The errors in numerical summation of dose from point sources are assessed by computing ratios of the geometric factors of point-source and extended-source voxel volumes. RESULTS: This sphere calculation agrees well with measured dose values, showing maximum and average differences of 4.8% and 0.5% at the balloon surface. Numerical integration gives errors greater than 1% within five voxel radii of a voxel source. CONCLUSIONS: A new spherical dose calculation technique is proposed for brachytherapy treatment planning systems. Numerical integration over point source voxels is not accurate near the balloon surface.     

4D in vivo dose verification for real‐time tumor tracking treatments using EPID dosimetry

The use of sophisticated techniques such as gating and tracking treatments requires additional quality assurance to mitigate increased patient risks. To address this need, we have developed and validated an in vivo method of dose delivery verification for real-time aperture tracking techniques, using an electronic portal imaging device (EPID)-based, on-treatment patient dose reconstruction and a dynamic anthropomorphic phantom. Using 4DCT scan of the phantom, ten individual treatment plans were created, 1 for each of the 10 separate phases of the respiratory cycle. The 10 MLC apertures were combined into a single dynamic intensity-modulated radiation therapy (IMRT) plan that tracked the tumor motion. The tumor motion and linac delivery were synchronized using an RPM system (Varian Medical Systems) in gating mode with a custom breathing trace. On-treatment EPID frames were captured using a data-acquisition computer with a dedicated frame-grabber. Our in-house EPID-based in vivo dose reconstruction model was modified to reconstruct the 4D accumulated dose distribution for a dynamic MLC (DMLC) tracking plan using the 10-phase 4DCT dataset. Dose estimation accuracy was assessed for the DMLC tracking plan and a single-phase (50% phase) static tumor plan, represented a static field test to verify baseline accuracy. The 3%/3 mm chi-comparison between the EPID-based dose reconstruction for the static tumor delivery and the TPS dose calculation for the static plan resulted in 100% pass rate for planning target volume (PTV) voxels while the mean percentage dose difference was 0.6%. Comparing the EPID-based dose reconstruction for the DMLC tracking to the TPS calculation for the static plan gave a 3%/3 mm chi pass rate of 99.3% for PTV voxels and a mean percentage dose difference of 1.1%. While further work is required to assess the accuracy of this approach in more clinically relevant situations, we have established clinical feasibility and baseline accuracy of using the transmission EPID-based, in vivo patient dose verification for MLC-tracking treatments.     

Evaluation of neutron sources for ISAGE—in-situ-NAA for a future lunar mission

For a future Moon landing, a concept for an in-situ NAA involving age determination using the ⁴⁰Ar–³⁹Ar method is developed. A neutron source ²⁵²Cf is chosen for sample irradiation on the Moon. A special sample-in-source irradiation geometry is designed to provide a homogeneous distribution of neutron flux at the irradiation position. Using reflector, the neutron flux is likely to increase by almost 200%. Sample age of 1 Ga could be determined. Elemental analysis using INAA is discussed.     

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.     

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.     

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%.     

An attempt to calculate correctly the region of influence in gauging moisture with neutrons

A neutron device, when employed for moisture gauging, samples only part of bulk material having dimensions large compared with the neutron range. In this paper, firstly, a new method is presented for calculating the size of this sample, i.e. the region of influence of the gauge. The method is not distorted by the neutron leakage (the large value of such an effect in the method availed by finite spheres is also pointed out here). In the present method the relative influence of a sample point, i.e. its weight, is calculated. For this the neutron flux and neutron importance (or, possibly, slowing-down density) can be used. The region of influence is then inside a surface of equiqeight points and it has the total weight of a certain percentage, e.g. 99%. Secondly, this method was applied to the subsurface gauge based on neutron slowing down. Three simple diffusion approximations of the neutron transport were then used, though the reliability of these approximations is in doubt. Values for the radius of a sphere as the region of 99% influence were calculated for a certain soil. Among these results those obtained by means of the 3-group diffusion are, especially at larger moisture values, preferred by the author. The age theory may be useful for nearly-dry substances.     

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.     

Characterization of differences in calculated and actual measured skin doses to canine limbs during stereotactic radiosurgery using Gafchromic film

Accurate calculation of absorbed dose to the skin, especially the superficial and radiosensitive basal cell layer, is difficult for many reasons including, but not limited to, the build-up effect of megavoltage photons, tangential beam effects, mixed energy scatter from support devices, and dose interpolation caused by a finite resolution calculation matrix. Stereotactic body radiotherapy (SBRT) has been developed as an alternative limb salvage treatment option at Colorado State University Veterinary Teaching Hospital for dogs with extremity bone tumors. Optimal dose delivery to the tumor during SBRT treatment can be limited by uncertainty in skin dose calculation. The aim of this study was to characterize the difference between measured and calculated radiation dose by the Varian Eclipse (Varian Medical Systems, Palo Alto, CA) AAA treatment planning algorithm (for 1-mm, 2-mm, and 5-mm calculation voxel dimensions) as a function of distance from the skin surface. The study used Gafchromic EBT film (International Specialty Products, Wayne, NJ), FilmQA analysis software, a limb phantom constructed from plastic water? (fluke Biomedical, Everett, WA) and a canine cadaver forelimb. The limb phantom was exposed to 6-MV treatments consisting of a single-beam, a pair of parallel opposed beams, and a 7-beam coplanar treatment plan. The canine forelimb was exposed to the 7-beam coplanar plan. Radiation dose to the forelimb skin at the surface and at depths of 1.65 mm and 1.35 mm below the skin surface were also measured with the Gafchromic film. The calculation algorithm estimated the dose well at depths beyond buildup for all calculation voxel sizes. The calculation algorithm underestimated the dose in portions of the buildup region of tissue for all comparisons, with the most significant differences observed in the 5-mm calculation voxel and the least difference in the 1-mm voxel. Results indicate a significant difference between measured and calculated data extending to average depths of 2.5 mm, 3.4 mm, and 10 mm for the 1-mm, 2-mm, and 5-mm dimension calculation matrices, respectively. These results emphasize the importance of selecting as small a treatment planning software calculation matrix dimension as is practically possible and of taking a conservative approach for skin treatment planning objectives. One suggested conservative approach is accomplished by defining the skin organ as the outermost 2–3 mm of the body such that the high dose tail of the skin organ dose-volume histogram curve represents dose on the deep side of the skin where the algorithm is more accurate.     

~(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。     

基于深度学习的乳腺癌保乳术后调强放疗剂量分布预测

目的研究基于深度学习的方法预测乳腺癌保乳术后调强放疗(IMRT)剂量分布, 并评估其预测精度。方法回顾性分析2018年1月至2023年3月在上海国际医学中心接受IMRT的110例左侧乳腺癌保乳术后患者的调强放疗数据, 随机固定选择80例作为训练集, 随机固定10例作为验证集, 剩余20例作为测试集。首先将患者的计算机体层成像(CT)图像、感兴趣区、体素与靶区距离和对应的剂量分布四通道特征作为输入数据, 然后使用U-net网络进行训练得到预测模型, 利用该模型对测试集进行剂量预测, 验证体素与靶区距离特征在剂量预测中的影响, 并将剂量预测结果与实际手动计划剂量进行比较。结果加入体素与靶区距离特征的模型使预测精度更高, 测试集中20例患者的剂量评分和剂量体积直方图(DVH)评分分别为2.10±0.18和2.28±0.08, 与手动计划剂量分布更加接近(t=2.52、2.40, P<0.05)。靶区和危及器官(OAR)的剂量预测结果与手动计划剂量的偏差在4%以内, 健侧乳腺平均剂量增加了13 cGy, 均在临床可接受范围内。除PTV60的D2、D98(Di为i%的PTV体积接受的剂量)...     

Validation of a commercial software dose calculation for Y-90 microspheres

PURPOSESeveral new commercial software packages have become available that can calculate the tumor and normal tissue dose distributions from post-treatment PET-CT scans for Y-90 microsphere treatments of liver lesions. This work seeks to validate the MIM SurePlan Liver Y90 software by comparing its results to a previously developed Monte Carlo derived voxel dose kernel calculation method.METHODSWe analyzed 10 patients who had treatments for metastatic liver cancer and created contours on post Y-90 treatment PET-CT images. We then performed dose calculations using three methods and compared the results. The first two methods calculated the dose using MIM SurePlan Liver Y90’s LDM (Local Deposition Method) and the VSV (Voxel S Value) algorithms. The third method calculated the dose using a publicly available Fluka Monte Carlo-derived dose kernel (MCK) calculation (used as ground truth). We investigated 3D Gamma passing rates and several dosimetric parameters.RESULTSA total of 3%/3 mm 3D gamma passing rates averaged 99.3% for the VSV and 78.9% for LDM. Compared to the MCK distribution, the differences for combined target GTV V_70Gy and normal liver and/or lobe mean doses were small. Larger differences were seen in GTV mean doses and D₉₅, likely due to large dose gradients in the treated regions combined with differences in dose kernel, dose grid and finite volume effects.CONCLUSIONSThe MIM SurePlan Liver Y90 VSV algorithm agreed well with the MCK calculation for patients treated with Y-90 microspheres based on the gamma analysis and several dosimetric parameters. Larger dosimetric differences in lesion mean doses and D₉₅ suggests that these metrics are less robust to changes in calculation grid location and finite volume effects for small lesions.