Neutron Dose Distribution Research Articles

Radiation measurement in the environment of FLASH using passive dosimeters

Sophisticated electronic devices comprising sensitive microelectronic components have been installed in the close proximity of the 720 MeV superconducting electron linear accelerator (linac) driving the FLASH (Free Electron Laser in Hamburg), presently in operation at DESY in Hamburg. Microelectronic chips are inherently vulnerable to ionizing radiation, usually generated during routine operation of high-energy particle accelerator facilities like the FLASH. Hence, in order to assess the radiation effect on microelectronic chips and to develop suitable mitigation strategy, it becomes imperative to characterize the radiation field in the FLASH environment. We have evaluated the neutron and gamma energy (spectra) and dose distributions at critical locations in the FLASH tunnel using superheated emulsion (bubble) detectors, GaAs light emitting diodes (LED), LiF-thermoluminescence dosimeters (TLD) and radiochromic (Gafchromic EBT) films. This paper highlights the application of passive dosimeters for an accurate analysis of the radiation field produced by high-energy electron linear accelerators.

SU-FF-T-289: Measurement of Secondary Neutron Dose for Scattering Modes of Proton Radiotherapy in KNCC

Purpose: Recently, the secondary neutron dose for scattering modes of proton radiotherapy has been a big issue, since some study groups had reported that the neutron flux produced in scattering modes could be one hundred times of IMRT and 3D CRT. In this presentation, the neutron dose measured in proton therapy facility of KNCC (Korea National Cancer Center) is presented. Method and Materials: For the neutron dose measurement, a proton beam field with 20 g/cm2 of depth range and 9 g/cm2 of Spread Out Bragg Peak (SOBP) were produced. 6 Bonner spheres with various diameters ranging from 3.5cm to 25 cm were set up 270cm away from an isocenter. Neutron doses were measured using the CR-39 neutron detector installed by a fixed interval focusing into the isocenter in the treatment room. Results: As a result of neutron spectrum analysis, the average neutron energy was measured to be 2.09 MeV. The difference of doses measured using neutron detector and CR-39 was checked to be within ±15%. The measured neutron dose distribution shows a big population at 90°of beam direction, which agree with previously published papers. But, differently from Binns & Hough published results, the measured dose values is one order lower than the dose values for existing IMRT treatment. Conclusion: For the meantime, secondary neutron is generated a lot in the proton radiotherapy especially in scattering modes, but it is considered to reap a better treatment effect by showing the same or rather lower dose distribution compared to existing general radiation therapy. Conflict of Interest: This investigation was supported by a research grant from the National Cancer Center, Korea. Key words: secondary neutron, proton therapy.

Estimation of angular distribution of neutron dose using time-of-flight for [formula omitted] system at 110 MeV

We have reported measured angular and energy distributions of neutron dose from 110 MeV 19F projectiles bombarding a thick aluminum target. The measurements are carried out with BC501 liquid scintillator detector using the time-of-flight technique. We have measured neutron energy distributions at 0 ∘ , 30 ∘ , 60 ∘ , 90 ∘ , and 120 ∘ and converted them to dose distributions using the ICRP recommended fluence to ambient dose equivalent and absorbed dose conversion coefficients. Similar conversions to ambient dose equivalent are done for theoretically estimated distributions from the nuclear reaction model code EMPIRE-2.18. The experimental results are compared with calculated ambient dose equivalent from different empirical formulations proposed by earlier workers. Based on the comparison, we have attempted modifications of the parameters in these empirical expressions.

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.

Thick target neutron dose evaluation for 19F+Al system

Measured angular distribution of neutron dose from 110 MeV (19)F projectiles bombarding a thick aluminium target is reported. The measurements are carried out with a commercially available rem meter at 0 degrees, 30 degrees, 60 degrees and 90 degrees. The experimental results are compared with calculated dose from different empirical formulations proposed by earlier workers as well as with calculated dose from theoretically estimated energy distributions from the nuclear reaction model code EMPIRE-2.18.

Measurement of neutron dose distribution for a passive scattering nozzle at the Proton Medical Research Center (PMRC)

For a passive scattering nozzle at the Proton Medical Research Center (PMRC), the distribution of the neutron dose equivalent in the direction transverse to the proton beam line was measured for 200 MeV protons using a tissue equivalent proportional counter REM500. The neutron dose equivalent per Gray of proton-absorbed dose at the isocenter was approximately 2.3 mSv/Gy. The neutron dose outside of the treatment field was approximately 2 mSv/Gy, at maximum and it should be reduced as low as possible according to the ALARA principle. The aperture size of the multi-leaf collimator (MLC), which is a pre-collimator for shielding unwanted protons and mounted at a position upstream from the patient bolus and collimator, was varied in order to investigate its reduction effect on the neutron dose equivalent outside of the treatment field. We confirmed that the neutron dose equivalent outside of the treatment field can be limited depending on the aperture size of the MLC. This could be considered in future treatment planning optimization and patient treatment. MCNPX calculations showed good agreement with the measurements within 10%.

SU-FF-T-338: Comparison Between Proton and Neutron Dose Distributions From Single-Scattering and Dual-Scattering Systems in Ocular Proton Therapy

Purpose: Typically, ocular proton treatment nozzles use a single‐scattering (SS) flattening filter design to achieve lateral spreading of a raw pencil‐beam to create a flat field. This approach has the advantage of simplicity but at the cost of proton efficiency, η. It is expected, though never demonstrated, that a more efficient, dual‐scattering (DS) flattening filter design may possess several advantages. Method and Materials: A proton nozzle with the SS filter design was modeled in Monte Carlo(MC)radiation transport software. Simulations of this nozzle provided in‐field proton absorbed dose distributions in water, D, and the neutrondose equivalent values, H, outside of the field, which were subsequently benchmarked to published measurements. Then, a proposed nozzle was modeled that uses a DS filter design. Simulations of the SS and DS nozzles were conducted to investigate differences in several figures‐of‐merit including the distal 80%–20% falloff l D80–20 and neutrondose equivalence per therapy Gray (H/D) distributions. Other figures‐of‐merit presented will include the field uniformity U, therapeuticdose rate Ddot, and the 80–20% lateral penumbral width l L80–20. Results: The simulations and measurements of the proton absorbed dose distributions from the SS nozzle agreed to within 2% or 0.1 mm. The shape of the measured and simulated H/D values as a function of distance from isocenter perpendicular to the beam agreed within 3%. The distal falloff width was expectedly narrower from the DS nozzle by 2.5 mm. The simulations revealed that the DS design yields substantially lower H/D values (between 0.3–0.6 times the SS values). This is partially attributed to the DS design's increased peak dose per proton.Conclusion: The DS flattening filter design may offer clinical advantages when compared to the SS filter design, including a sharper distal falloff, Ddot, and decreased H/D values.

SU-FF-T-349: Validation of DSAR Algorithm for Intensity Modulated Neutron Radiotherapy

Purpose: To validate the differential scatter air ratio dose calculation algorithm for Intensity Modulated Neutron Radiotherapy (IMNRT). Method and Materials: Our current treatment planning system uses a differential Scatter Air Ratio (dSAR) technique to calculate neutron dose distributions from a d(48.5) + Be cyclotron. An empirical function was used to fit the profiles at different depths. Intensity modulated beams are delivered using the segmental MLC (sMLC) technique. Application of the dSAR method to small beam segments used for IMNRT has been investigated. Beam profiles for rectangular fields ranging from 2 × 2 cm2 to 10 × 10 cm2 were measured using a 0.3 cc TE chamber and compared to that predicted by the empirical profile function. This chamber has a diameter of 6 mm and a wall thickness of 2.5 mm. Point dose measurements using the dual ionization chamber technique and individual profile measurements were performed to validate the dose calculation accuracy for small irregular segments. Results: For field sizes ranging from 2 × 2 cm2 to 10 × 10 cm2, measured profiles agreed with the dSAR calculations to within 5% at 2.5 cm depth, and to within 3% at 10 cm depth, inside the field. The pdd's matched to within 2% of that calculated by the treatment planning system. For irregularly shaped segments, absolute point dose measurements matched to within 2% along the central axis and to within 3% for off-axis points. For some irregular segment profiles, large deviations were observed in the penumbral region, particularly at shallower depths. This was due primarily to partial volume effects resulting from the large diameter of the ionization chamber used for measuring these profiles. However, all measured profiles agreed with the calculated profiles to within 5% inside the segments. Conclusions: The dSAR dose calculation algorithm has been validated for IMNRT.

Analysis of Radiation Streaming Experiment through a Labyrinth at TIARA HIR1

In order to estimate the accuracy of calculation methods such as empirical formulae and Monte Carlo codes for neutron fluxes, neutron dose distributions, and gamma-ray dose distributions in a room with a Cu target irradiated by 67-MeV-proton and a connected labyrinth, experimental analyses were carried out. As the results of the analyses, the neutron energy spectra, neutron dose distributions, and gamma-ray dose distributions showed the Monte Carlo methods to be applicable to streaming calculations at the proton accelerator facility in the energy region of several tens MeV, although thermal neutron flux distributions were slightly overestimated.

Measurement of radiation skyshine with D–T neutron source

The D–T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of ∼1.7×10 11 n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9×0.9 m 2 was open during the experimental period. The radiation dose rate outside the target room was measured a maximum distance of 550 m from the D–T target point with a spherical rem-counter. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation counters. The highest neutron dose was about 9×10 −22 Sv/(source neutron) at a distance of 30 m from the D–T target point and the dose rate was attenuated to 4×10 −24 Sv/(source neutron) at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 230 m. The line source model agrees well with the experimental results within the distance of 350 m.

D-T Neutron Skyshine Experiments at JAERI/FNS.

The D-T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of ∼1.7×1011n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9 × 0.9 m2 was open during the experimental period.The radiation dose rate outside the target room was measured as far as about 550 m away from the D-T target point with a spherical rem-counter. The highest neutron dose was about 0.5 μSv/hr at a distance of 30 m from the D-T target point and the dose rate was attenuated to 0.002 μSv/hr at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 250 m. The neutron spectra were evaluated with a 3He detector with different thickness of polyethylene neutron moderators. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation detectors.

A computational study into the use of polyacrylamide gel and A-150plastic as brain tissue substitutes for boron neutron capture therapy

A precise evaluation of the dosimetric performance of epithermal neutron beamsdesigned for boron neutron capture theory of brain tumours requires the use ofa phantom material that closely matches brain tissue. The aim of this studywas to investigate how well polyacrylamide gel (or PAG) and A-150 plasticperformed as substitutes for brain tissue compared with standard phantommaterials such as water and polymethyl-methacrylate (or PMMA). Thermal neutronfluence, photon dose and epithermal neutron dose distributions were calculatedfor the epithermal neutron beam available at the University of Birmingham. Theresults presented in this paper show that the PAG provides a good simulationof radiation transport in the brain with differences from the real brain of + 9.4%, -10.8% and + 5.1% at a depth of 50 mm for thermal neutronfluence, gamma dose and epithermal neutron dose distributions respectively.The polyacrylamide gel presented is therefore a promising substitute for braintissue that can, as a dosimeter, provide a three-dimensional map of theabsorbed dose delivered by the epithermal neutron beam. However, this studydoes not investigate the agreement between doses derived from magneticresonanceand physical doses for such gels. A-150 plastic was shown to be a bettersubstitute for brain tissue than PMMA, with differences from brain of-1.9%, -12.4% and -13.2% at a depth of 50 mm for thermal neutronfluence, gamma dose and epithermal neutron dose distributions respectively,against + 21.1%, -16.2% and + 19.2% for PMMA. A-150 plastic shouldtherefore be the material of choice for solid phantoms.

Computation of neutron dose distribution in a cylindrical phantom of tissue equivalent materials

This work gives the total dose and the spectral distribution at any specific point inside a tissue equivalent cylindrical phantom due to Pu–Be neutron source. The absorbed dose through the material are coming from 17 reactions between the neutrons and the phantom constituents: H 2,0 2,N 2. and C. The changes in the neutron spectra are due to energy loss as a result of neutron slowing down. A computer program is used to carry out these calculations using the Monte-Carlo technique.

A fast and accurate treatment planning method for boron neutron capture therapy

Background and purpose: The aim of this study was to test the applicability of conventional semi-empirical algorithms for the treatment planning of boron neutron capture therapy (BNCT). Materials and methods: Beam data of a clinical epithermal BNCT beam obtained in a large cuboid water phantom were introduced into a commercial treatment planning system (TPS). For the calculation of thermal neutron fluence distributions, the Gaussian pencil beam model of the electron beam treatment planning algorithm was used. A simple photon beam algorithm was used for the calculation of the gamma-ray and fast neutron dose distribution. The calculated dose and fluence distributions in the central plane of an anthropomorphic head phantom were compared with measurements for various field sizes. The calculation time was less than 1 min. Results: At the normalization point in the head phantom, the absolute dose and fluence values agreed within the measurement uncertainty of approximately 2–3% (1 SD) with those at the same depth in a cuboid phantom of approximately the same size. Excellent agreement of within 2–3% (1 SD) was obtained between measured and calculated relative fluence and dose values on the central beam axis and at most off-axis positions in the head phantom. At positions near the phantom boundaries, generally in low dose regions, local differences of approximately 30% were observed. Conclusions: A fast and accurate treatment planning method has been developed for BNCT. This is the first treatment planning method that may allow the same interactive optimization procedures for BNCT as applied clinically for conventional radiotherapy.

Simultaneous macro and micro dosimetry with MOSFETs

The application of MOSFET dosimeters in complicated mixed radiation fields for measurement of absorbed dose distribution in tissue equivalent phantoms has been studied. The spectra of secondary charged particles have been measured simultaneously with average absorbed doses by the same MOSFET dosimeter. A good correlation has been observed between neutron depth dose distribution in a water phantom obtained using MOSFETs in integral mode and a tissue equivalent (T.E.) ionisation chamber. Such MOSFET dosimeters are a promising tool for micro-macro dosimetry in Boron Neutron Capture Therapy (BNCT) and Fast Neutron Therapy (FNT). Paired MOSFETs with one of the dosimeters covered by /sup 10/B have been applied for measuring of average boron dose distribution and microdosimetric spectra due to alpha particles and /sup 7/Li ions throughout a perspex phantom exposed in the epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR).

A comparative study of the attenuation of reactor thermal neutrons in different types of concrete

This study was carried out to assess the distribution of thermal neutrons emitted directly from the core of the ET-RR-1 reactor in ordinary concrete, ilmenite concrete and ilmenite-limonite concrete shields. Measurements were carried out by using a direct beam and a cadmium filtered beam of reactor neutrons. The neutron dose distributions were measured using Li 2B 4O 7: Mn thermoluminescent dosimeters. The data obtained show that ilmenite concrete is better for slow and thermal neutron attenuation than both ordinary and ilmenite-limonite concretes. Also it was concluded that thermal neutrons emitted directly from the reactor core are highly absorbed within the first few centimeters of each type of concrete. The thickness of ilmenite concrete required to attenuate the doses of neutrons to a certain value along the beam axis for a direct reactor beam estimated to be about 75 and 57% of the shield thickness made from ordinary and ilmenite-limonite concretes, respectively. Empirical formulae were derived to calculate the neutron dose distributions in ordinary, ilmenite and ilmenite-limonite concrete shields both along and perpendicular to the beam axis for both the direct reactor neutrons and the reactor thermal neutrons.

Determination of the thermal neutron flux in a fast neutron beam by use of a boron‐coated ionization chamber

The thermal neutron distribution in slow and fast neutron beams is usually determined using the foil activation method. In this work a small magnesium walled ionization chamber, in which the inner surface of the wall has been coated with 10B to increase the sensitivity for thermal neutrons, is used to estimate the thermal neutron component of the beam. After calibration and determination of the directional response in a thermal neutron beam a comparison with foil activation at different depths in water was performed to investigate the reliability of the ionization measurements. The chamber was used in a computer controlled water phantom to measure the depth and lateral distribution of the thermal neutron dose. With this arrangement two-dimensional scans of the thermal neutrons could be performed quickly and with high accuracy.

Neutron radiation distribution sensor using flexible plastic scintillating fiber combined with the time-of-flight technique

As a promising neutron radiation distribution sensor, a long fiber system has been made using two silica fibers connected to both ends of the plastic scintillating fiber (PSF). Its basic properties were investigated to detect the position of neutron incidence with the time-of-flight (TOF) method. From the experimental results for the fission neutron source, the two linear relations were confirmed (1) between the incident position of the neutrons and the time difference and (2) between the neutron flux level and the counting rate. The spatial resolution and the counting efficiency of this system were about 60 cm (FWHM) and about 1.0/spl times/10/sup -4/ [counts/incident neutron] respectively using 1 mm diameter fiber of 100 m length. A neutron dose distribution measured by this system was compared with the mapping result using a conventional neutron Sievert(Sv) counter. After some correction for the spatial resolution by an unfolding method, both results showed good agreement.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Description of a 3D conformal neutron and photon radiotherapy technique for prostate cancer

Several methods have been described to improve the therapeutic ratio of radiotherapy in patients with locally advanced carcinoma of the prostate (LACaP). Studies have shown improved survival and local control in patients treated with neutron irradiation as opposed to photons alone. However, an increased complication rate was observed when conformal field design was not utilized. A 3D conformal mixed neutron/photon technique for LACaP has been developed at this institution. Field shaping for the neutron component is achieved with a multirod collimator. Prior to treatment, all patients undergo a conventional simulation followed by a treatment planning computed tomography (CT) scan. Contours of target and normal structures are entered into the 3D treatment planning system. Beam apertures are designed utilizing beam's eye view display. The photon dose is given in 10 fractions at 1.8 Gy each to the prostate and seminal vesicles (PSVs) and pelvic lymph node (PLN) volumes. The neutron dose is given in 15 fractions at 1.0 Gy per fraction to the PSV volume and 0.6 Gy to the PLN volume. The neutron portion includes a partial transmission block to achieve this dose differential. The treatment planning process, including beam weighting and custom block design, is described. Using these techniques, the neutron dose distributions to the target volumes, bladder, and rectum were comparable to those seen with four-field conformal photon irradiation.

Impact of a multileaf collimator on treatment morbidity in localized carcinoma of the prostate

Purpose : To evaluate the effectiveness of variable multileaf collimation, three-dimensional treatment planning, and computer-controlled conformal radiation therapy of prostate cancer. Methods and Materials : Two hundred and forty-five patients with locally advanced prostate cancer have completed treatment over a 9-year time span using a multileaf collimator and conformal treatment techniques on the University of Washington cyclotron. All patients had three-dimensional treatment planning with computed tomography scans in the treatment position, and had treatment fields individually shaped to the target volume with a continuously variable multileaf collimator. Treatment was delivered under computer control with network transfer of the multileaf collimator settings from the treatment planning computer to the cyclotron control system. Results : The multileaf collimator combined with three-dimensional treatment planning results in elegant dose distributions. These neutron dose distributions resulted in a reduced local/regional tumor failure rate with no increase in complications when compared to control treatment with photons in a randomized trial. Neutron treatment delivered at other institutions without conformal beam shaping resulted in the same improvement in local-regional tumor control rates, but was associated with a significantly higher normal tissue complication rate than seen with conformal neutron beam delivery techniques (grade 3 and 4 cumulative late normal tissue toxicity rates of 39% vs. 10%, p = 0.0007). Conclusions : Conformal treatment of prostate cancer using a multileaf collimated neutron beam results in increased local/regional tumor control rates with low normal tissue toxicities. This experience is directly applicable to the conformal treatment of prostate cancer with photons.