Variation in lunar neutron dose estimates

The radiation environment on the Moon includes albedo neutrons produced by primary particles interacting with the lunar surface. In this work, HZETRN2010 is used to calculate the albedo neutron contribution to effective dose as a function of shielding thickness for four different space radiation environments and to determine to what extent various factors affect such estimates. First, albedo neutron spectra computed with HZETRN2010 are compared to Monte Carlo results in various radiation environments. Next, the impact of lunar regolith composition on the albedo neutron spectrum is examined, and the variation on effective dose caused by neutron fluence-to-effective dose conversion coefficients is studied. A methodology for computing effective dose in detailed human phantoms using HZETRN2010 is also discussed and compared. Finally, the combined variation caused by environmental models, shielding materials, shielding thickness, regolith composition and conversion coefficients on the albedo neutron contribution to effective dose is determined. It is shown that a single percentage number for characterizing the albedo neutron contribution to effective dose can be misleading. In general, the albedo neutron contribution to effective dose is found to vary between 1-32%, with the environmental model, shielding material and shielding thickness being the driving factors that determine the exact contribution. It is also shown that polyethylene or other hydrogen-rich materials may be used to mitigate the albedo neutron exposure.     

High-energy neutron spectroscopy with thick silicon detectors

The high-energy neutron component of the space radiation environment in thick structures such as the International Space Station contributes to the total radiation dose received by an astronaut. Detector design constraints such as size and mass have limited the energy range of neutron spectrum measurements in orbit to about 12 MeV in Space Shuttle studies. We present a new method for high-energy neutron spectroscopy using small silicon detectors that can extend these measurements to more than 500 MeV. The methodology is based on measurement of the detector response function for high-energy neutrons and inversion of this response function with measured deposition data to deduce neutron energy spectra. We also present the results of an initial shielding study performed with the thick silicon detector system for high-energy neutrons incident on polyethylene.     

The response of a spherical tissue-equivalent proportional counter to iron particles from 200-1000 MeV/nucleon

The radiation environment on board the space shuttle and the International Space Station includes high-Z and high-energy (HZE) particles that are part of the galactic cosmic radiation (GCR) spectrum. Iron-56 particles are considered to be one of the most biologically important parts of the GCR spectrum. Tissue-equivalent proportional counters (TEPCs) are used as active dosimeters on manned space flights. These TEPCs are further used to determine the average quality factor for each space mission. A TEPC simulating a 1-microm-diameter sphere of tissue was exposed as part of a particle spectrometer to (56)Fe particles at energies from 200-1000 MeV/nucleon. The response of TEPCs in terms of mean lineal energy, y(F), and dose mean lineal energy, y(D), as well as the energy deposited at different impact parameters through the detector was determined for six different incident energies of (56)Fe particles in this energy range. Calculations determined that charged-particle equilibrium was achieved for each of the six experiments. Energy depositions at different impact parameters were calculated using a radial dose distribution model, and the results were compared to experimental data.     

Problems of component discrimination in space radiation dosimetry

Resolving the LET spectrum of environmental radiation in space for assessing dose equivalents creates special problems due to superposition effects. Three components of the radiation field in space, trapped protons, tissue disintegration stars, and neutrons, contribute the bulk of the total dose equivalent. While lack of discrimination of neutron recoil and trapped primary protons does not interfere with correct determination of the combined dose equivalent as such, the simultaneous bursts of several low-energy protons and alpha particles from tissue disintegration stars completely defy LET-resolution with conventional instrumentation. So far, the tissue star dose has been determined only semiquantitatively from nuclear emulsion data. The neutron spectrum in space shows a markedly higher relative fluence in the region beyond 5 MeV than the fission neutron spectrum. Therefore, its LET spectrum centers less heavily on LET values near the proton Bragg Peak. This would call for assigning a QF value of less than 10 to the neutron dose in space. Still more serious shortcomings exist with regard to LET interpretation of heavy primaries.     

Microdosimetry for boron neutron capture therapy.

Preclinical studies for boron neutron capture therapy (BNCT) using epithermal neutrons are ongoing at several laboratories. The absorbed dose in tumor cells is a function of the thermal neutron flux at depth, the microscopic boron concentration, and the size of the cell. Dosimetry is therefore complicated by the admixture of thermal, epithermal, and fast neutrons, plus gamma rays, and the array of secondary high-linear-energy-transfer particles produced within the patient from neutron interactions. Microdosimetry can be a viable technique for determining absorbed dose and radiation quality. A 2.5-cm-diameter tissue-equivalent gas proportional counter has been built with 50 parts per million (ppm) 10B incorporated into the walls and counting gas to simulate the boron uptake anticipated in tumors. Measurements of lineal energy (y) spectra for BNCT in simulated volumes of 1-10 microns diameter show a dose enhancement factor of 4.3 for 30 ppm boron, and a "y" of 250 keV/microns for the boron capture process. Chamber design plus details of experimental and calculated linear energy spectra will be presented.     

Influence of neutron collimation on the energy deposition in a tissue equivalent sphere

Summary The influence of neutron collimation on the shape of energy deposition spectra was investigated using a spherical walled proportional counter. Experimental dose averaged lineal energies were obtained and compared with theoretical values. The fractions of total absorbed dose corresponding to various intervals of lineal energies were also deduced from the measured distributions.Contribution No. 1289 of the Biology Programme, Directorate General XII of the Commission of the European Communities     

Space radiation absorbed dose distribution in a human phantom.

The radiation risk to astronauts has always been based on measurements using passive thermoluminescent dosimeters (TLDs). The skin dose is converted to dose equivalent using an average radiation quality factor based on model calculations. The radiological risk estimates, however, are based on organ and tissue doses. This paper describes results from the first space flight (STS-91, 51.65 degrees inclination and approximately 380 km altitude) of a fully instrumented Alderson Rando phantom torso (with head) to relate the skin dose to organ doses. Spatial distributions of absorbed dose in 34 1-inch-thick sections measured using TLDs are described. There is about a 30% change in dose as one moves from the front to the back of the phantom body. Small active dosimeters were developed specifically to provide time-resolved measurements of absorbed dose rates and quality factors at five organ locations (brain, thyroid, heart/lung, stomach and colon) inside the phantom. Using these dosimeters, it was possible to separate the trapped-proton and the galactic cosmic radiation components of the doses. A tissue-equivalent proportional counter (TEPC) and a charged-particle directional spectrometer (CPDS) were flown next to the phantom torso to provide data on the incident internal radiation environment. Accurate models of the shielding distributions at the site of the TEPC, the CPDS and a scalable Computerized Anatomical Male (CAM) model of the phantom torso were developed. These measurements provided a comprehensive data set to map the dose distribution inside a human phantom, and to assess the accuracy and validity of radiation transport models throughout the human body. The results show that for the conditions in the International Space Station (ISS) orbit during periods near the solar minimum, the ratio of the blood-forming organ dose rate to the skin absorbed dose rate is about 80%, and the ratio of the dose equivalents is almost one. The results show that the GCR model dose-rate predictions are 20% lower than the observations. Assuming that the trapped-belt models lead to a correct orbit-averaged energy spectrum, the measurements of dose rates inside the phantom cannot be fully understood. Passive measurements using 6Li- and 7Li-based detectors on the astronauts and inside the brain and thyroid of the phantom show the presence of a significant contribution due to thermal neutrons, an area requiring additional study.     

Tests of shielding effectiveness of Kevlar and Nextel onboard the International Space Station and the Foton-M3 capsule

Radiation assessment and protection in space is the first step in planning future missions to the Moon and Mars, where mission and number of space travelers will increase and the protection of the geomagnetic shielding against the cosmic radiation will be absent. In this framework, the shielding effectiveness of two flexible materials, Kevlar and Nextel, were tested, which are largely used in the construction of spacecrafts. Accelerator-based tests clearly demonstrated that Kevlar is an excellent shield for heavy ions, close to polyethylene, whereas Nextel shows poor shielding characteristics. Measurements on flight performed onboard of the International Space Station and of the Foton-M3 capsule have been carried out with special attention to the neutron component; shielded and unshielded detectors (thermoluminescence dosemeters, bubble detectors) were exposed to a real radiation environment to test the shielding properties of the materials under study. The results indicate no significant effects of shielding, suggesting that thin shields in low-Earth Orbit have little effect on absorbed dose.     

Monte Carlo simulation of neutron dose equivalent by photoneutron production inside the primary barriers of a radiotherapy vault

PurposeTo evaluate the neutron dose equivalent produced by photoneutrons inside the primary barriers of a radiotherapy vault.MethodsMonte Carlo simulations were performed for investigating the production of photoneutrons as well as neutron shielding requirements. Two photon beams of 15 and 18 MV struck sheets of steel and lead, and the neutron doses were calculated at the isocenter (P_iso) and at a distance of 50 cm from the inside wall (P_wall) while delivering 1 Gy to the patient. The proper thicknesses of borated polyethylene (BPE) and concrete were simulated to reduce neutron contamination.ResultsWhen the primary barrier consisted of a concrete alone, the neutron doses at P_iso were 0.5 μSv/Gy and 12.8 μSv/Gy for 15- and 18-MV, respectively. At P_wall, the neutron doses were 15.8 μSv/Gy and 318.4 μSv/Gy for 15- and 18-MV, respectively. When 15 MV photons interacted with metal sheets, the neutron doses were 0.4–22.2 μSv/Gy at P_iso and 15.8–812.5 μSv/Gy at P_wall, depending on the thickness and material of the metal sheets and neutron shielding. In the case of 18 MV photons with the same configuration, the neutron doses were 0.9–59.5 μSv/Gy and 73.9–5006.1 μSv/Gy for P_iso and P_wall, respectively. The neutron dose delivered to the patient was reduced to the level of the dose delivered with a concrete barrier by including a 10-cm-thick BPE for each beam.ConclusionsWhen the primary barrier shielding is designed with a metal sheet inside for high energy, proper neutron shielding should be constructed to avoid undesirable photoneutron dose.     

Induction of micronuclei in human fibroblasts across the Bragg curve of energetic heavy ions

The space environment consists of a varying field of radiation particles including high-energy ions, with spacecraft shielding material providing the major protection to astronauts from harmful exposure. Unlike low-LEpsilonTau gamma or X rays, the presence of shielding does not always reduce the radiation risks for energetic charged-particle exposure. The dose delivered by the charged particle increases sharply as the particle approaches the end of its range, a position known as the Bragg peak. However, the Bragg curve does not necessarily represent the biological damage along the particle path since biological effects are influenced by the track structures of both primary and secondary particles. Therefore, the "biological Bragg curve" is dependent on the energy and the type of the primary particle and may vary for different biological end points. Here we report measurements of the biological response across the Bragg curve in human fibroblasts exposed to energetic silicon and iron ions in vitro at two different energies, 300 MeV/nucleon and 1 GeV/nucleon. A quantitative biological response curve generated for micronuclei per binucleated cell across the Bragg curve did not reveal an increased yield of micronuclei at the location of the Bragg peak. However, the ratio of mono- to binucleated cells, which indicates inhibition of cell progression, increased at the Bragg peak location. These results confirm the hypothesis that severely damaged cells at the Bragg peak are more likely to go through reproductive death and not be evaluated for micronuclei.     

Tungsten-based material as promising new lead-free gamma radiation shielding material in nuclear medicine

PurposeThe main objective of this study was to evaluate the efficacy of tungsten carbide as new lead-free radiation shielding material in nuclear medicine by evaluating the attenuation properties.Materials and methodsThe elemental composition of tungsten carbide was analysed using Field-Emission Scanning Electron Microscopy (FESEM) with energy dispersive X-ray (EDX). The purity of tungsten carbide was 99.9%, APS: 40–50 µm. Three discs of tungsten carbide was fabricated with thickness of 0.1 cm, 0.5 cm and 1.0 cm. Three lead discs with similar thickness were used to compare the attenuation properties with tungsten carbide discs. Energy calibration of gamma spectroscopy was performed by using ¹²³I, ¹³³Ba, ¹⁵²Eu, and ¹³⁷Cs. Gamma radiation from these sources were irradiated on both materials at energies ranging from 0.160 MeV to 0.779 MeV. The experimental attenuation coefficients of lead and tungsten carbide were compared with theoretical attenuation coefficients of both materials from NIST database. The half value layer and mean free path of both materials were also evaluated in this study.ResultsThis study found that the peaks obtained from gamma spectroscopy have linear relationship with all energies used in this study. The relative differences between the measured and theoretical mass attenuation coefficients are within 0.19–5.11% for both materials. Tungsten carbide has low half value layer and mean free path compared to lead for all thickness at different energies.ConclusionThis study shows that tungsten carbide has high potential to replace lead as new lead-free radiation shielding material in nuclear medicine.     

Effective dose of A-bomb radiation in Hiroshima and Nagasaki as assessed by chromosomal effectiveness of spectrum energy photons and neutrons

The effective dose of combined spectrum energy neutrons and high energy spectrum γ-rays in A-bomb survivors in Hiroshima and Nagasaki has long been a matter of discussion. The reason is largely due to the paucity of biological data for high energy photons, particularly for those with an energy of tens of MeV. To circumvent this problem, a mathematical formalism was developed for the photon energy dependency of chromosomal effectiveness by reviewing a large number of data sets published in the literature on dicentric chromosome formation in human lymphocytes. The chromosomal effectiveness was expressed by a simple multiparametric function of photon energy, which made it possible to estimate the effective dose of spectrum energy photons and differential evaluation in the field of mixed neutron and γ-ray exposure with an internal reference radiation. The effective dose of reactor-produced spectrum energy neutrons was insensitive to the fine structure of the energy distribution and was accessible by a generalized formula applicable to the A-bomb neutrons. Energy spectra of all sources of A-bomb γ-rays at different tissue depths were simulated by a Monte Carlo calculation applied on an ICRU sphere. Using kerma-weighted chromosomal effectiveness of A-bomb spectrum energy photons, the effective dose of A-bomb neutrons was determined, where the relative biological effectiveness (RBE) of neutrons was expressed by a dose-dependent variable RBE, RBE(γ, D _n), against A-bomb γ-rays as an internal reference radiation. When the newly estimated variable RBE(γ, D _n) was applied to the chromosome data of A-bomb survivors in Hiroshima and Nagasaki, the city difference was completely eliminated. The revised effective dose was about 35% larger in Hiroshima, 19% larger in Nagasaki and 26% larger for the combined cohort compared with that based on a constant RBE of 10. Since the differences are significantly large, the proposed effective dose might have an impact on the magnitude of the risk estimates deduced from the A-bomb survivor cohort.     

Radiation dose around a PET scanner installation: Comparison of Monte Carlo simulations,analytical calculations and experimental results

PurposeMonte Carlo study of radiation transmission around areas surrounding a PET room.MethodsAn extended population of patients administered with ¹⁸F-FDG for PET-CT investigations was studied, collecting air kerma rate and gamma ray spectra measurements at a reference distance. An MC model of the diagnostic room was developed, including the scanner and walls with variable material and thickness. MC simulations were carried out with the widely used code GEANT4.ResultsThe model was validated by comparing simulated radiation dose values and gamma ray spectra produced by a volumetric source with experimental measurements; ambient doses in the surrounding areas were assessed for different combinations of wall materials and shielding and compared with analytical calculations, based on the AAPM Report 108.In the range 1.5–3.0 times of the product between the linear attenuation coefficient and thickness of an absorber (μ x), it was observed that the effectiveness of different combinations of shielding is roughly equivalent. An extensive tabulation of results is given in the text.ConclusionsThe validation tests performed showed a satisfactory agreement between the simulated and expected results. The simulated dose rates incident on, and transmitted by the walls in our model of PET scanner room, are generally in good agreement with analytical estimates performed using the AAPM Publication No. 108 method. This provides an independent confirmation of AAPM's approach. Even in this specific field of application, GEANT4 proved to be a relevant and accurate tool for dosimetry estimates, shielding evaluation and for general radiation protection use.     

Dosimetric shield evaluation with tungsten sheet in 4, 6, and 9 MeV electron beams

In electron radiotherapy, shielding material is required to attenuate beam and scatter. A newly introduced shielding material, tungsten functional paper (TFP), has been anticipated to become a very useful device that is lead-free, light, flexible, and easily processed, containing very fine tungsten powder at as much as 80% by weight. The purpose of this study was to investigate the dosimetric changes due to TFP shielding for electron beams. TFP (thickness 0–15 mm) was placed on water or a water-equivalent phantom. Percentage depth ionization and transmission were measured for 4, 6, and 9 MeV electron beams. Off-center ratio was also measured using film dosimetry at depth of dose maximum under similar conditions. Then, beam profiles and transmission with two shielding materials, TFP and lead, were evaluated. Reductions of 95% by using TFP at 0.5 cm depth occurred at 4, 9, and 15 mm with 4, 6, and 9 MeV electron beams, respectively. It is found that the dose tend to increase at the field edge shaped with TFP, which might be influenced by the thickness. TFP has several unique features and is very promising as a useful tool for radiation protection for electron beams, among others.     

Benchmark studies of the effectiveness of structural and internal materials as radiation shielding for the international space station

Accelerator-based measurements and model calculations have been used to study the heavy-ion radiation transport properties of materials in use on the International Space Station (ISS). Samples of the ISS aluminum outer hull were augmented with various configurations of internal wall material and polyethylene. The materials were bombarded with high-energy iron ions characteristic of a significant part of the galactic cosmic-ray (GCR) heavy-ion spectrum. Transmitted primary ions and charged fragments produced in nuclear collisions in the materials were measured near the beam axis, and a model was used to extrapolate from the data to lower beam energies and to a lighter ion. For the materials and ions studied, at incident particle energies from 1037 MeV/nucleon down to at least 600 MeV/nucleon, nuclear fragmentation reduces the average dose and dose equivalent per incident ion. At energies below 400 MeV/nucleon, the calculation predicts that as material is added, increased ionization energy loss produces increases in some dosimetric quantities. These limited results suggest that the addition of modest amounts of polyethylene or similar material to the interior of the ISS will reduce the dose to ISS crews from space radiation; however, the radiation transport properties of ISS materials should be evaluated with a realistic space radiation field.     

On the need for massive additional shielding of a catheterization laboratory for the implementation of high dose rate 192Ir intravascular brachytherapy

Purpose: There is a widespread belief in the cardiology and radiation oncology community that high dose rate 192Ir intravascular brachytherapy cannot be implemented without massive additional shielding of the conventional catheterization labs. The purpose of this work is to show that this is a myth, which is not based on sound radiation protection principles.Methods: Exposure rates in air were calculated for a variety of point and line sources of 192Ir. Exposures per treatment at different distances from the source were calculated for a typical intravascular brachytherapy treatment of a 15-Gy dose at a radial distance of 2 mm from the source and for source lengths in the range of 0 to 10 cm. Additionally, exposure rates outside the catheterization lab were calculated for various lead shielding thicknesses typical of conventional X-ray facilities. These rates were used along with the NCRP recommendations on radiation facility design to assess shielding requirements.Results: For a treatment dose of 15 Gy at 2 mm, the occupational exposure per treatment at 2 m in air without any tissue attenuation or shielding was 7.8 mR for a lesion length of 3.0 cm. This exposure/treatment is independent of the dose rate or the activity of the source. However, it increases as lesion length is increased, increasing from 5.4 to 24.9 mR as lesion length increased from 2 to 10 cm. Exposures in unrestricted areas outside the catheterization lab using the NCRP shielding rationale can be kept below 2 mR per treatment and using appropriate workload, use, and occupancy factors below 2 mR per week.Conclusions: The feasibility of implementing a high dose rate 192Ir intravascular brachytherapy program in a catheterization laboratory is totally independent of the dose rate or the activity of the source. If it is feasible to implement 192Ir brachytherapy in a conventional catheterization lab using low activity 192Ir seeds, then it is also feasible to do so with a high activity 192Ir afterloader.     

Measurements of secondary neutrons from cosmic radiation with a Bonner sphere spectrometer at 79°N

Air crew members and airline passengers are continuously exposed to cosmic radiation during their flights. Particles ejected by the sun during so-called solar particle events (SPEs) in periods of high solar activity can contribute to this exposure. In rare cases the dose from a single SPE might even exceed the annual dose limit of 1 mSv above which dose monitoring of air crews is legally required in Germany. Measurements performed by means of neutron monitors have already shown that the relative intensity of secondary neutrons from cosmic radiation is enhanced during an SPE, particularly at regions close to the magnetic poles of the Earth where shielding of the cosmic radiation by the geomagnetic field is low. Here we describe a Bonner sphere spectrometer installed at the Koldewey station at 79°N, i.e. about 1,000 km from the geographic North pole, which is designed to provide first experimental data on the time-dependent energy spectrum of neutrons produced in the atmosphere during an SPE. This will be important to calculate doses from these neutrons to air crew members. The system is described in detail and first results are shown that were obtained during quiet periods of sun activity.     

Hiroshima-like neutrons from A-bomb replica: physical basis for their use in biological experiments

The lineal energy distribution and several other dosimetric parameters were measured for the neutrons emitted from a replica of the Hiroshima bomb to determine their usefulness in biological experiments designed to estimate the effectiveness of actual Hiroshima neutrons. The "Little-Boy" replica (LBR) was constructed at the Los Alamos National Laboratory in support of the recent atomic-bomb dose reevaluation and was made of identical materials and had nearly identical dimensions and geometry as the Hiroshima bomb. However, the LBR was operated as a steady-state nuclear reactor, which permitted measurements under controlled conditions. Detailed dosimetric measurements and calculations were made at distances of up to 2.1 m from the center of the LBR uranium core. At these distances, the in-air kerma was approximately 97% from neutrons and kerma rates were shown to be particularly useful for biological experiments (up to approximately 7 Gy/h was possible). Quantitative intercomparisons of neutron energy spectra, lineal energy distributions, and measured cytogenetic results for several fission-neutron sources indicate that Hiroshima and LBR neutrons should be of similar biological effectiveness. Based on these evaluations, and cytogenetic results for LBR neutrons reported in a companion paper (this issue), it is estimated that Hiroshima neutrons were 20 to 30% more effective than the fission neutrons commonly used in radiobiology.     

Chromosome aberrations induced in human lymphocytes by neutron irradiation.

In vitro dose--response curves of unstable chromosome aberrations in human lymphocytes have been obtained for neutron spectra of mean energies 0-7, 0-9, 7-6 and 14-7 MeV. The aberration yields have been fitted to the quadratic function Y = alphaD + betaD2, which is consistent with the single-track and two-track model of aberration formation. However with high-LET radiation, the linear component of yield, corresponding to damage caused by single tracks, predominants, and this term becomes more dominant with increasing LET, so that for fission spectrum neutrons the relationship is linear, Y = alphaD. At low doses, such as those recieved by radiation workers, limiting r.b.e. values between 13 and 47 are obtained relative to 60Co gamma-radiation. At higher doses, as used in radiotherapy, the values are much lower; ranging from 2-7 to 8 at 200 rad of equivalent gamma-radiation. Both sets of r.b.e. values correlate well with track-averaged LET but not with dose-averaged LET. When the numbers of cells without aberrations are plotted against radiation dose, curves are obtained which are similar in shape to those for conventional cell-survival experiments with comparable neutron spectra. The Do values obtained in the present study are close to those from other cell system.     

Calculation of dose components in head phantom for boron neutron capture therapy.

Application of neutrons to cancer treatment has been a subject of considerable clinical and research interest since the discovery of the neutron by Chadwick in 1932 (3). Boron neutron capture therapy (BNCT) is a technique of radiation oncology which is used in treating brain cancer (glioblastoma multiform) or melanoma and that consists of preferentially loading a compound containing 10B into the tumor location, followed by the irradiation of the patient with a beam of neutron. Dose distribution for BNCT is mainly based on Monte Carlo simulations. In this work, the absorbed dose spatial distribution resultant from an idealized neutron beam incident upon ahead phantom is investigated using the Monte Carlo N-particles code, MCNP 4B. The phantom model used is based on the geometry of a circular cylinder on which sits an elliptical cylinder capped by half an ellipsoid representing the neck and head, both filled with tissue-equivalent material. The neutron flux and the contribution of individual absorbed dose components, as a function of depths and of radial distance from the beam axis (dose profiles) in phantom model, is presented and discussed. For the studied beam the maximum thermal neutron flux is at a depth of 2 cm and the maximum gamma dose at a depth of 4 cm.