Neutron Dose Distribution Research Articles

Patient-specific voxel phantom dosimetry during the prostate treatment with high-energy linac

Dose due to neutrons resulting from high-energy linear accelerator is not detailed in routine treatment planning, though this information is potentially important for better estimates of health risks including secondary cancers. In this study, the neutron contaminations were evaluated in the patient-specific voxel phantom for the treatment of prostate cancer using MCNP code. The results showed that the neutron organ doses are independent from the distance between studied organs and treatment area, and unlike the photon dose, the neutron dose distribution is almost homogeneous in the patient body.

전립선암의 세기조절 방사선치료 시 조사문수별 중성자선량 평가

Abstract The aim of this study was provide basic information and establish the criteria in radiation therapy planningby measuring the absorbed neutron dose of normal tissues and lesions according to the number of portals. From September 2013 to January 2014, 20 patients who were diagnosed with prostate cancer and were previously treated with radiation therapy were replanned retrospectively to measure the absorbed neutron dose distribution according to the number of portals. The absorbed neutron dose was measured in each of the 5, 7 and 9 portals using a 15 MV energy, which meant a therapeutic dose of 220 cGy. The optical stimulation luminescence dosimeter was separated by 20cm and 60cm away from the center of the field of view. As a result, the average radiation dose in the abdomenappeared to have a positive relationship with the number of portals, which was statistically significant (p<.05). The average radiation dose was 4.34±1.08. The average radiation dose in the thyroid was 2.71±.37. Although it showeda positive relationship with the number of portals, it did not have statistical significance. The number of portals andthe neutron dose depending on the position showed a significant positive relationship, particularly in the abdomen. As a result of linear regression analysis, as the number of the portal increased in steps, the average volume of the neutrons increased significantly (0.416 times). In conclusion, efficient selection of the number of portals is needed considering the difference in the absorbed neutron dose in the normal tissues depending on the number of the portals.

Radiation shielding in SPring-8 linear accelerator tunnels for safety interlock system upgrade

To upgrade a safety interlock system, a restricted area was converted into an accessible area at an operating condition. To judge the necessity for an additional shield design, -ray and neutron dose distributions were measured in the linac, booster synchrotron, L3BT, and New SUBARU (NS) areas. In the linac, chicane was found to be the main source of rays. Among the new accessible areas, the -ray doses exceeded the limit at the electron transport openings of the synchrotron and NS areas. Similarly, the neutron dose surpassed the limit at the synchrotron opening. The beam-dump contribution to the neutron dose was estimated to be 20% by using the Monte Carlo code FLUKA.

Neutron spectral and dose distribution studies during fast reactor fuel fabrication

In mixed oxide (MOX) fuel fabrication facilities, the fabrication of MOX fuel involves various metallurgical operations such as mixing and milling of weighed quantities of uranium and plutonium oxides, precompaction, granulation of precompacts, final compaction of granules, sintering of green pellets, followed by fuel pin fabrication. All these operations are carried out in glove boxes, which have complex geometries due to housing of various equipments. Plutonium, a source of neutrons, is handled in large quantities in various forms such as powder, granules, pellets and pellets encapsulated in pins during these operations. It is obvious that extensive knowledge on the neutron spectral distributions in workplace is required from radiation protection and shielding point of view. In this paper, a brief introduction to the source of neutrons in MOX fuel handling facilities, studies that include experimental measurements of neutron spectra of various forms of MOX, contribution of neutron fluence in various energy groups and its dose equivalent, establishment of a simulation procedure for glove boxes handling Pu using FLUKA Monte Carlo codes, comparison of simulation results with the actual experimental measurements are presented. The results indicate that bare PuO 2 and MOX sources require 3 mm of lead shield to eliminate gamma interference in MICROSPEC with N-probe neutron spectrometer. Furthermore, the studies reveal that dose equivalent contribution from MOX pellets is very significant in the energy group of 1.0-1.5 MeV unlike PuO 2 powder and fuel pins, which exhibit significant contribution in the energy groups of both 1.0-1.5 MeV and 2.0-3.0 MeV. Also, the fuel pins show high neutron fluence in the energy group of 0.0-0.01 MeV, but they do not contribute significantly to dose equivalent. A good agreement between experimentally measured data and FLUKA simulated results has been observed.

Commissioning of intensity modulated neutron radiotherapy (IMNRT)

Intensity modulated neutron radiotherapy (IMNRT) has been developed using inhouse treatment planning and delivery systems at the Karmanos Cancer Center∕Wayne State University Fast Neutron Therapy facility. The process of commissioning IMNRT for clinical use is presented here. Results of commissioning tests are provided including validation measurements using representative patient plans as well as those from the TG-119 test suite. IMNRT plans were created using the Varian Eclipse optimization algorithm and an inhouse planning system for calculation of neutron dose distributions. Tissue equivalent ionization chambers and an ionization chamber array were used for point dose and planar dose distribution comparisons with calculated values. Validation plans were delivered to water and virtual water phantoms using TG-119 measurement points and evaluation techniques. Photon and neutron doses were evaluated both inside and outside the target volume for a typical IMNRT plan to determine effects of intensity modulation on the photon dose component. Monitor unit linearity and effects of beam current and gantry angle on output were investigated, and an independent validation of neutron dosimetry was obtained. While IMNRT plan quality is superior to conventional fast neutron therapy plans for clinical sites such as prostate and head and neck, it is inferior to photon IMRT for most TG-119 planning goals, particularly for complex cases. This results significantly from current limitations on the number of segments. Measured and calculated doses for 11 representative plans (six prostate∕five head and neck) agreed to within -0.8 ± 1.4% and 5.0 ± 6.0% within and outside the target, respectively. Nearly all (22∕24) ion chamber point measurements in the two phantom arrangements were within the respective confidence intervals for the quantity [(measured-planned)∕prescription dose] derived in TG-119. Mean differences for all measurements were 0.5% (max = 7.0%) and 1.4% (max = 4.1%) in water and virtual water, respectively. The mean gamma pass rate for all cases was 92.8% (min = 88.6%). These pass rates are lower than typically achieved with photon IMRT, warranting development of a planar dosimetry system designed specifically for IMNRT and∕or the improvement of neutron beam modeling in the penumbral region. The fractional photon dose component did not change significantly in a typical IMNRT plan versus a conventional fast neutron therapy plan, and IMNRT delivery is not expected to significantly alter the RBE. All other commissioning results were considered satisfactory for clinical implementation of IMNRT, including the external neutron dose validation, which agreed with the predicted neutron dose to within 1%. IMNRT has been successfully commissioned for clinical use. While current plan quality is inferior to photon IMRT, it is superior to conventional fast neutron therapy. Ion chamber validation results for IMNRT commissioning are also comparable to those typically achieved with photon IMRT. Gamma pass rates for planar dose distributions are lower than typically observed for photon IMRT but may be improved with improved planar dosimetry equipment and beam modeling techniques. In the meantime, patient-specific quality assurance measurements should rely more heavily on point dose measurements with tissue equivalent ionization chambers. No significant technical impediments are anticipated in the clinical implementation of IMNRT as described here.

Intensity modulated neutron radiotherapy optimization by photon proxy

Introducing intensity modulation into neutron radiotherapy (IMNRT) planning has the potential to mitigate some normal tissue complications seen in past neutron trials. While the hardware to deliver IMNRT plans has been in use for several years, until recently the IMNRT planning process has been cumbersome and of lower fidelity than conventional photon plans. Our in-house planning system used to calculate neutron therapy plans allows beam weight optimization of forward planned segments, but does not provide inverse optimization capabilities. Commercial treatment planning systems provide inverse optimization capabilities, but currently cannot model our neutron beam. We have developed a methodology and software suite to make use of the robust optimization in our commercial planning system while still using our in-house planning system to calculate final neutron dose distributions. Optimized multileaf collimator (MLC) leaf positions for segments designed in the commercial system using a 4 MV photon proxy beam are translated into static neutron ports that can be represented within our in-house treatment planning system. The true neutron dose distribution is calculated in the in-house system and then exported back through the MATLAB software into the commercial treatment planning system for evaluation. The planning process produces optimized IMNRT plans that reduce dose to normal tissue structures as compared to 3D conformal plans using static MLC apertures. The process involves standard planning techniques using a commercially available treatment planning system, and is not significantly more complex than conventional IMRT planning. Using a photon proxy in a commercial optimization algorithm produces IMNRT plans that are more conformal than those previously designed at our center and take much less time to create. The planning process presented here allows for the optimization of IMNRT plans by a commercial treatment planning optimization algorithm, potentially allowing IMNRT to achieve similar conformality in treatment as photon IMRT. The only remaining requirements for the delivery of very highly modulated neutron treatments are incremental improvements upon already implemented hardware systems that should be readily achievable.

In-phantom neutron dose distribution for bladder cancer cases treated with high-energy photons

This work presents an estimation of the neutron dose distribution for common bladder cancer cases treated with high-energy photons of 15 MV therapy accelerators. Neutron doses were measured in an Alderson phantom, using TLD 700 and 600 thermoluminescence dosimeters, resembling bladder cancer cases treated with high-energy photons from 15 MV LINAC and having a treatment plan using the four-field pelvic box technique. Thermal neutron dose distribution in the target area and the surrounding tissue was estimated. The sensitivity of all detectors for both gamma and neutrons was estimated and used for correction of the TL reading. TLD detectors were irradiated with a Co60 gamma standard source and thermal neutrons at the irradiation facility of the National Institute for Standards (NIS). The TL to dose conversion factor was estimated in terms of both Co60 neutron equivalent dose and thermal neutron dose. The dose distribution of photo-neutrons throughout each target was estimated and presented in three-dimensional charts and isodose curves. The distribution was found to be non-isotropic through the target. It varied from a minimum of 0.23 mSv/h to a maximum of 2.07 mSv/h at 6 cm off-axis. The mean neutron dose equivalent was found to be 0.63 mSv/h, which agrees with other published literature. The estimated average neutron equivalent to the bladder per administered therapeutic dose was found to be 0.39 mSv Gy−1, which is also in good agreement with published literature. As a consequence of a complete therapeutic treatment of 50 Gy high-energy photons at 15 MV, the total thermal neutron equivalent dose to the abdomen was found to be about 0.012 Sv.

Imaging of gamma and neutron dose distributions at LVR-15 epithermal beam by means of FGLDs

Gamma and fast neutron dose spatial distributions have been measured at the collimator exit of the epithermal neutron beam of LVR-15 reactor (Řež). Measurements were performed by means of optically analyzed Fricke-gel-layer detectors. The separation of the two dose contributions has been achieved by suitable pixel-to-pixel elaboration of the light transmittance images of Fricke-gel-layer detectors prepared with water and heavy water.

Dose estimation in B16 tumour bearing mice for future irradiation in the thermal column of the TRIGA reactor after B/Gd/LDL adduct infusion

To test the efficacy of a new (10)B-vector compound, the B/Gd/LDL adduct synthesised at Torino University, in vivo irradiations of murine tumours are in progress at the TRIGA Mark II reactor of the Pavia University. A localised B16 melanoma tumour is generated in C57BL/6 mice and subsequently infused with the adduct. During the irradiation, the mouse will be put in a shield to protect the whole body except the tumour in the back-neck area. To optimise the treatment set-up, MCNP simulations were performed. A very simplified mouse model was built using MCNP geometry capabilities, as well as the geometry of the shield made of 99% (10)B enriched boric acid. A hole in the shield is foreseen in correspondence of the back-neck region. Many configurations of the shield were tested in terms of neutron flux, dose distribution and mean induced activity in the tumour region and in the radiosensitive organs of the mouse. In the final set-up, up to five mice can be treated simultaneously in the reactor thermal column and the neutron fluence in the tumour region for 10 min of irradiation is of about 5×10(12) cm(-2).

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%.

Neutron dose distribution from 12C induced reactions on Ti and Ag using proton recoil scintillator

Abstract We measured the double differential neutron yield at 0°, 30°, 60° and 90° from 12 C 5+ induced reactions on thick targets of Ti and Ag at 12 MeV/amu at the Cyclotron at National Institute of Radiological Sciences, Inage, Japan, with 5″ × 5″ proton recoil scintillation detectors BC-501. The measured neutron spectra were unfolded using pulse height unfolding algorithm and energy and angular distribution obtained. Energy distribution of neutron ambient dose equivalent, H*(10) and absorbed dose, D at different angles was determined from double differential neutron spectra using ICRP recommended fluence to dose conversion coefficients.

Detailed dose distribution prediction of Cf-252 brachytherapy source with boron loading dose enhancement

The purpose of this work is to evaluate the dose rate distribution and to determine the boron effect on dose rate distribution for 252Cf brachytherapy source. This study was carried out using a Monte Carlo simulation. To validate the Monte Carlo computer code, the dosimetric parameters were determined following the updated TG-43 formalism and compared with current literature data. The validated computer code was then applied to evaluate the neutron and photon dose distribution and to illustrate the boron loading effect.

GEANT4 Transport Calculations for Neutrons and Photons Below 15 MeV

Neutron transport calculations are needed in many fields of radiation protection and medical physics. Neutron transport is often simulated with a Monte Carlo code, which allows the calculation of neutron dose distributions in different geometries, for example in voxel phantoms. So far little information is available about neutron transport below 15 MeV in tissue-like matter performed with the Monte Carlo code GEANT4. Therefore, we compared GEANT4 results of the ambient dose equivalent for neutron energies from 10 meV to 15 MeV with MCNP results and values published in ICRU 57. The GEANT4 code was found to be appropriate to calculate neutron and photon dose at energies below 15 MeV correctly. Furthermore, a method how to include a dose-assessment factor into the calculation was developed and tested.

SU‐FF‐T‐400: Monte Carlo Simulations of Compact Plasma Accelerators for Proton Radiotherapy

Purpose: One common characteristic of the compact plasma proton accelerators is a broad energy spectrum which can not be used directly for radiotherapy. The purpose of this study is to analyze the effect of the geometrical parameters of the treatment head components, their materials, and the magnetic field configuration, on the proton and neutron dose distributions to identify an optimal design for the plasma proton accelerator's treatment head and provide important feedback to ongoing plasma proton experiments. Method and Materials: Calculations of proton dose distributions and the neutron dose equivalent were performed using Monte Carlo simulations by irradiating a water phantom with protons in the energy range of 175 to 250 MeV. The simulations were performed using the Geant4 (version 4.9.2) toolkit. The input spectra were experimental energy spectra scaled to energies relevant for radiotherapy, having a broad Gaussian distribution. Results: We have designed a collimating system, magnetic field region, energy selector, and shielding components for a plasma accelerated proton therapy system. The collimating and energy selector systems are needed in order to achieve proton dose distributions superior to the conventional X‐ray radiotherapy distributions. It is shown that a great number of unwanted electrons and photons is stopped by a sequence of tungsten absorbing disks placed inside the collimating and energy selector systems. Also, a sequence of three boronated polyethylene layers is shown to substantially capture the secondary neutrons. Dose deposited by neutrons in the water phantom is found to be 6 orders of magnitude lower than the dose deposited by protons, thus giving negligible contribution to the integral dose. Conclusion: We have studied the design of the plasma proton accelerator's treatment head. It is shown that the proposed system is capable of delivering clinically relevant proton dose distributions. The implemented shielding components ensure extremely low neutron dose distributions.

Radiation release characterization of PWR spent fuel assemblies generated from Korean nuclear power plants

Spent nuclear fuel should be kept under safe management until it is disposed of permanently. Because of this, it is important to understand its radiation release characteristics. In this paper, the Monte Carlo method is applied to evaluate the radiation release characteristics of two types of PWR spent fuel assembly generated from the operating plants in Korea: Westinghouse and Korea Standard Nuclear Power Plant. The source terms were calculated using ORIGEN-ARP. The neutron and photon (or gamma) dose distributions along the vertical and horizontal directions of each spent fuel assembly were evaluated using MCNPX code. Compared with the two dose distributions, the photon dose was found to be about 105 times higher than the neutron dose.

SU-GG-T-12: Neutron Energy and Dose Considerations of Cf-252 Brachytherapy Sources for Various Geometries

Purpose: To examine the effect of the 252 Cf source size to the neutronenergy spectrum and dose contribution in a tissue equivalent material, with and without boronneutron capture enhancement. Method and Materials: Version 2.5.0 of the MCNPXcomputer code (Pelowitz, 2005) was used in this study to calculate the neutron energy spectra and dose distribution with and without 10B loading for 252Cf sources of various geometries. A spherical phantom geometry with a centrally positioned 252 Cf point‐source was first implemented in order to verify our simulation code by reproducing existing data in the literature. The neutronenergy flux was calculated for various 10 B loadings and various distances from the source. Cylindrical sources of different sizes were simulated in a 30×30×30 cm3 water phantom and the neutronenergy flux and energy deposition in the medium was calculated on cylindrical surfaces enclosing each source at various distances from each source surface. The boron enhancement of 30 ppm was also studied in this case. The neutronenergy spectrum was modeled as an isotropic Watt distribution and all calculated spectra were normalized assuming the same amount of 252 Cf distributed uniformly inside the source volume. Results: Calculated neutron energy spectra for the 252 Cf point‐source geometry showed very good agreement with existing literature, verifying our simulation model. For the cylindrical 252 Cf sources, preliminary results showed increased fast neutron contribution from the compact source, as compared to the conventional one, especially at small distances from the source surface. Conclusion: Smaller size 252 Cf sources can be more beneficial for brachytherapy treatments, not only due to localized dose distribution, but also due to higher dose contribution from fast neutrons of high RBE at small distances from the source.

Changes in prescribed doses for the Seattle neutron therapy system

From the beginning of the neutron therapy program at the University of Washington Medical Center, the neutron dose distribution in tissue has been calculated using an in-house treatment planning system called PRISM. In order to increase the accuracy of the absorbed dose calculations, two main improvements were made to the PRISM treatment planning system: (a) the algorithm was changed by the addition of an analytical expression of the central axis wedge factor dependence with field size and depth developed at UWMC. Older versions of the treatment-planning algorithm used a constant central axis wedge factor; (b) a complete newly commissioned set of measured data was introduced in the latest version of PRISM. The new version of the PRISM algorithm allowed for the use of the wedge profiles measured at different depths instead of one wedge profile measured at one depth. The comparison of the absorbed dose calculations using the old and the improved algorithm showed discrepancies mainly due to the missing central axis wedge factor dependence with field size and depth and due to the absence of the wedge profiles at depths different from 10 cm. This study concludes that the previously reported prescribed doses for neutron therapy should be changed.

Influences of Neutron Source Spectrum and Thermal Neutron Scattering Law Data on the MCNPX Simulation of a Cyclotron-Based Neutron Field for Boron Neutron Capture Therapy

In our previous study, the simulation of a cyclotron-based neutron field for boron neutron capture therapy (BNCT) using a (p,n) spallation source with the MCNPX code was validated through measurements of the neutron energy spectrum behind the moderator assembly and the thermal neutron distribution in an acrylic phantom using reaction rates of 198Au. These validations showed that the simulation generally well reproduced the measurements. However, some discrepancies between the measurements and the calculation remained for clinical trials. In this paper, we investigated the influences of neutron source spectrum and thermal neutron scattering law data in the simulation to resolve those discrepancies. We also compared measured and calculated neutron doses behind the moderator assembly with results obtained using a tissue equivalent proportional counter. We clarified that the neutron source spectrum calculated using the LA150 data led to the overestimation of high-energy neutrons in a phantom, but this overestimation did not significantly affect the neutron dose distribution in a phantom, because a dominant part of the absorbed dose is due to neutrons of energies below 1MeV. The study of the influence of neutron scattering law data in a phantom also indicated that the use of selected S(α,β) data led to an improvement in the simulation of thermal neutron behavior.

Estimation of photon and neutron dose distributions in the THOR BNCT treatment room using dual TLD method

Dual detector dosimetry using the paired detectors, TLD-600 and TLD-700 chips, were selected in this study to distinguish the doses of neutrons and photons in Tsing Hua Open-pool Reactor (THOR) boron neutron capture therapy (BNCT) preclinical test. Since the neutron response of TLD is dependent on the neutron spectrum, responses of TLD-600 and TLD-700 for neutrons and photons were estimated by irradiation methods and in addition the effect of the neutron spectrum on the TLD response was studied by Monte Carlo simulations in present work. A sectional ART head phantom was used to be irradiated with the THOR BNCT beam. TLD-600 and TLD-700 chips were placed inside the phantom to measure and distinct the doses of neutrons from photons. Besides, dual TLD chips were placed in plane of the room space area to estimate the spatial dose distributions in the THOR BNCT treatment room. A cube water phantom was used to consider the deviation between the two algorithms used in this work. In the results, neutron and photon dose distributions in phantom and the spatial dose distributions in THOR BNCT treatment room were estimated; characteristics and the treatment indexes for BNCT were also assessed.

Intensity Modulated Neutron Radiotherapy for the Treatment of Adenocarcinoma of the Prostate

This study investigates the enhanced conformality of neutron dose distributions obtainable through the application of intensity modulated neutron radiotherapy (IMNRT) to the treatment of prostate adenocarcinoma. An in-house algorithm was used to optimize individual segments for IMNRT generated using an organ-at-risk (OAR) avoidance approach. A number of beam orientation schemes were investigated in an attempt to approach an optimum solution. The IMNRT plans were created retrospectively for 5 patients previously treated for prostate adenocarcinoma using fast neutron therapy (FNT), and a comparison of these plans is presented. Dose distributions and dose-volume histograms (DVHs) were analyzed and plans were evaluated based on percentage volumes of rectum and bladder receiving 95%, 80%, and 50% (V(95), V(80), V(50)) of the prescription dose, and on V(60) for both the femoral heads and GM(muscle) group. Plans were normalized such that the IMNRT DVHs for prostate and seminal vesicles were nearly identical to those for conventional FNT plans. Use of IMNRT provided reductions in rectum V(95) and V(80) of 10% (2-27%) and 13% (5-28%), respectively, and reductions in bladder V(95) and V(80) of 12% (3-26%) and 4% (7-10%), respectively. The average decrease in V(60) for the femoral heads was 4.5% (1-18%), with no significant change in V(60) for the GM(muscle) group. This study provides the first analysis of the application of intensity modulation to neutron radiotherapy. The IMNRT technique provides a substantial reduction in normal tissue dose in the treatment of prostate cancer. This reduction should result in a significant clinical advantage for this and other treatment sites.