Alterations in dose and lineal energy spectra under different shieldings in the Los Alamos high-energy neutron field

Nuclear interactions of space radiation with shielding materials result in alterations in dose and lineal energy spectra that depend on the specific elemental composition, density and thickness of the material. The shielding characteristics of materials have been studied using charged-particle beams and radiation transport models by examining the risk reduction using the conventional dose-equivalent approach. Secondary neutrons contribute a significant fraction of the total radiation exposure in space. An experiment to study the changes in dose and lineal energy spectra by shielding materials was carried out at the Los Alamos Nuclear Science Center neutron facility. In the energy range of about 2 to 200 MeV, this neutron spectrum is similar in shape within a factor of about 2 to the spectrum expected in the International Space Station habitable modules. It is shown that with a shielding thickness of about 5 g cm(-2), the conventional radiation risk increases, in some cases by as much as a factor of 2, but decreases with thicknesses of about of 20 g cm(-2). This suggests that care must be taken in evaluating the shielding effectiveness of a given material by including both the charged-particle and neutron components of space radiation.     

Neutron, proton, and photonuclear cross-sections for radiation therapy and radiation protection

I review recent work at Los Alamos undertaken to evaluate neutron, proton, and photonuclear cross-sections up to 150 MeV (to 250 MeV for protons), based on experimental data and nuclear model calculations. These data are represented in the ENDF format and can be used in computer codes to simulate radiation transport. They permit calculations of absorbed dose in the body from therapy beams, and through use of kerma coefficients allow absorbed dose to be estimated for a given neutron energy distribution. In radiation protection, these data can be used to determine shielding requirements in accelerator environments and to calculate neutron, proton, gamma-ray, and radionuclide production. Illustrative comparisons of the evaluated cross-section and kerma coefficient data with measurements are given.     

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.     

基于体素模型的中子外照射小鼠器官剂量分布

获得外照射条件下动物的器官剂量,为生物效应评价、剂量-效应关系研究提供准确的剂量信息和依据,是近年来辐射生物效应研究的热点.本文基于建立的一个质量为26.9g的小鼠体素模型（体素精度为0.2mm×0.2mm×0.2mm,体素数量为9424000）,利用蒙特卡罗方法计算了5种照射几何条件（左侧向、右侧向、腹背向、背腹向和各向同性）下,中子能量为10-9～20MeV共37个能量点的小鼠器官剂量转换系数,基于此套数据,在照射条件已知情况下,可以获得小鼠的器官剂量;分析讨论了照射几何条件、中子能量及器官的位置对小鼠器官剂量的影响.     

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.     

Solid cancer risk coefficient for fast neutrons in terms of effective dose

Cancer mortality risk coefficients for neutrons have recently been assessed by a procedure that postulates for the neutrons a linear dose dependence, invokes the excess risk of the A-bomb survivors at a gamma-ray dose D(1) of 1 Gy, and assumes a neutron RBE as a function of D(1) between 20 and 50. The excess relative risk (ERR) of 0.008/mGy has been obtained for R(1) = 20 and 0.016/mGy for R(1) = 50. To compare these results to the current ICRP nominal risk coefficient for solid cancer mortality (0.045/Sv for a population of all ages; 0.036/Sv for a working population), the ERR is translated into lifetime attributable risk and is then related to effective dose. The conversion is not trivial, because the neutron effective dose has been defined by ICRP not as a weighted genuine neutron dose (neutron kerma), but as a weighted dose that includes the dose from gamma rays that are induced by neutrons in the body. If this is accounted for, the solid cancer mortality risk for a working population is found to agree with the ICRP nominal risk coefficient for neutrons in their most effective energy range, 0.2 MeV to 0.5 MeV. In radiation protection practice, there is an added level of safety, because the effective dose, E, is-for monitoring purposes-assessed in terms of the operational quantity H*, which overestimates E substantially for neutrons between 0.01 MeV and 2 MeV.     

Comparison of neutron organ and effective dose coefficients for PIMAL stylized phantom in bent postures in standard irradiation geometries

Neutron dose coefficients for standard irradiation geometries have been reported in International Commission on Radiological Protection (ICRP) Publication 116 for the ICRP Publication 110 adult reference phantoms. In the present work, organ and effective dose coefficients have been calculated for a receptor in both upright and articulated (bent) postures representing more realistic working postures exposed to a mono-energetic neutron radiation field. This work builds upon prior work by Dewji and co-workers comparing upright and bent postures for exposure to mono-energetic photon fields. Simulations were conducted using the Oak Ridge National Laboratory’s articulated stylized adult phantom, “Phantom wIth Moving Arms and Legs” (PIMAL) software package, and the Monte Carlo N-Particle (MCNP) version 6.1.1 radiation transport code. Organ doses were compared for the upright and bent (45° and 90°) phantom postures for neutron energies ranging from 1 × 10^??9 to 20 MeV for the ICRP Publication 116 external exposure geometries—antero-posterior (AP), postero-anterior (PA), and left and right lateral (LLAT, RLAT). Using both male and female phantoms, effective dose coefficients were computed using ICRP Publication 103 methodology. The resulting coefficients for articulated phantoms were compared to those of the upright phantom. Computed organ and effective dose coefficients are discussed as a function of neutron energy, phantom posture, and source irradiation geometry. For example, it is shown here that for the AP and PA irradiation geometries, the differences in the organ coefficients between the upright and bent posture become more pronounced with increasing bending angle. In the AP geometry, the brain dose coefficients are expectedly higher in the bent postures than in the upright posture, while all other organs have lower dose coefficients, with the thyroid showing the greatest difference. Overall, the effective dose estimated for the upright phantom is more conservative than that for the articulated phantom, which may have ramifications in the estimation or reconstruction of radiation doses.     

Effective dose rate coefficients for exposure to contaminated soil

The Oak Ridge National Laboratory Center for Radiation Protection Knowledge has undertaken calculations related to various environmental exposure scenarios. A previous paper reported the results for submersion in radioactive air and immersion in water using age-specific mathematical phantoms. This paper presents age-specific effective dose rate coefficients derived using stylized mathematical phantoms for exposure to contaminated soils. Dose rate coefficients for photon, electron, and positrons of discrete energies were calculated and folded with emissions of 1252 radionuclides addressed in ICRP Publication 107 to determine equivalent and effective dose rate coefficients. The MCNP6 radiation transport code was used for organ dose rate calculations for photons and the contribution of electrons to skin dose rate was derived using point-kernels. Bremsstrahlung and annihilation photons of positron emission were evaluated as discrete photons. The coefficients calculated in this work compare favorably to those reported in the US Federal Guidance Report 12 as well as by other authors who employed voxel phantoms for similar exposure scenarios.     

Comparison of ITO and ZnO ternary glassy composites in terms of radiation shielding properties by Monte Carlo N-particle transport code and BXCOM

In the present study, radiation shielding properties of two glassy composite materials that are widely used in electronics, photovoltaic applications, and sensor technology, were investigated in the photon energy range from 15 keV to 15 MeV. The materials chosen were (ITO)/V2O5/B2O3 and ZnO/V2O5/B2O3 including various concentrations of B2O3. Radiation interaction was simulated and shielding parameters calculated by means of the MCNP and BXCOM codes. More specifically, buildup factors, effective electron density ($$N_{\text{eff}}$$) and effective atomic number ($$Z_{\text{eff}}$$) were calculated with BXCOM, while mass attenuation coefficients ($$\mu /\rho$$), half-value layer (HVL) and tenth-value layer (TVL) values were calculated with MCNP. The results were compared with those obtained with the WinXCOM code, for validation. Acceptable and preferable results were obtained for both composites as alternative to other glassy shielding materials. The composite including ITO showed better shielding properties than the composite including ZnO. In terms of radiation shielding, both composites turned out to be better than concrete and close to lead.     

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.     

Effective dose conversion coefficients for radionuclides exponentially distributed in the ground

In order to provide fundamental data required for dose evaluation due to environmental exposures, effective dose conversion coefficients, that is, the effective dose rate per unit activity per unit area, were calculated for a number of potentially important radionuclides, assuming an exponential distribution in ground, over a wide range of relaxation depths. The conversion coefficients were calculated for adults and a new-born baby on the basis of dosimetric methods that the authors and related researchers have previously developed, using Monte Carlo simulations and anthropomorphic computational phantoms. The differences in effective dose conversion coefficients due to body size between the adult and baby phantoms were found to lie within 50?%, for most cases; however, for some low energies, differences could amount to a factor of 3. The effective dose per unit source intensity per area was found to decrease by a factor of 2–5, for increasing relaxation depths from 0 to 5?g/cm², above a source energy of 50?keV. It is also shown that implementation of the calculated coefficients into the computation of the tissue weighting factors and the adult reference computational phantoms of ICRP Publication 103 does not significantly influence the effective dose conversion coefficients of the environment. Consequently, the coefficients shown in this paper could be applied for the evaluation of effective doses, as defined according to both recommendations of ICRP Publications 103 and 60.     

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.     

Evaluation of conversion coefficients from measurable to risk quantities for external exposure over contaminated soil by use of physical human phantoms

Conversion coefficients from measurable quantities such as air kerma free-in-air or personal dose equivalent to effective dose were determined by phantom experiments. Heterogenic anthropomorphic phantoms representing children of one and five years age, and a Rando phantom representing an adult were exposed in the open field contaminated by different levels of radiocesium in the upper soil layer, in a forest site and inside a wooden house. LiF thermoluminescent (TL) detectors were used inside the phantoms for the estimation of organ doses and effective dose. Personal dosimeters similar to those used in radiation protection for individual dose measurements were placed onto the phantom surface (chest area). The ratios of dose values in separate organs to air kerma free-in-air varied from 0.69 to 1.15 for the children phantoms, and from 0.55 to 0.94 for the adult phantom, respectively, when irradiated in the open field. Body size (weight) was found to be the most important factor influencing the values of the conversion coefficients. The differences observed can reach approximately 40% when comparing conversion factors from air kerma free-in-air to effective dose for adults and newborns. For conversion coefficients from personal dose to effective dose, these differences can reach approximately 15%. The dependences of the various conversion coefficients on body mass were quantified by regression analysis. The results were compared with those calculated for a plane mono-energetic photon source having an energy of 700 keV and being located in the ground at a depth of 0.5 g cm⁻². Calculated and measured conversion coefficients from air kerma free-in-air to effective dose agreed within 12%.     

落叶阔叶林反照率的时间变化与影响因素

反照率原位测量对生态系统能量收支及其遥感应用至关重要,但目前坡面地形反照率的测量方式有局限且可见光与近红外波段反照率时间变化的差异尚不清楚。本研究以东北地区帽儿山森林生态站的落叶阔叶林为例,探究入射和反射太阳辐射(SR,300～2800 nm)、光合有效辐射(PAR,400～700 nm)、近红外辐射(NIR,700～2800 nm)的反照率时间变化特征及其影响因子,同时分析了两种辐射表安装方式反照率的差异。结果表明: 晴天SR和NIR反照率日变化呈上下午不对称的U型曲线,但PAR从早到晚递增;阴天反照率均先急剧下降后趋于稳定。平行于坡面测量增大了反照率的日均值,但缓和了SR、NIR反照率日不对称的现象。从整个生长季来看,SR、NIR与PAR反照率水平测量时最大值分别为0.16、0.27和0.11,最小值分别为0.07、0.11和0.03。SR和NIR反照率季节变化均为先增大后减小(7月为峰值),PAR则相反,SR反照率主要受NIR而不是PAR控制。各波段反照率季节变化的影响因子按照贡献率排序为宽带归一化植被指数(61.7%～78.5%,可表征叶面积指数)>太阳高度角(15.4%～36.9%)>晴空指数(0.4%～36.9%)。     

A review on photoneutrons characteristics in radiation therapy with high-energy photon beams

In radiation therapy with high-energy photon beams (E > 10 MeV) neutrons are generated mainly in linacs head thorough (γ,n) interactions of photons with nuclei of high atomic number materials that constitute the linac head and the beam collimation system. These neutrons affect the shielding requirements in radiation therapy rooms and also increase the out-of-field radiation dose of patients undergoing radiation therapy with high-energy photon beams. In the current review, the authors describe the factors influencing the neutron production for different medical linacs based on the performed measurements and Monte Carlo studies in the literature.     

The radioprotective properties of fungal melanin are a function of its chemical composition, stable radical presence and spatial arrangement

Melanized microorganisms are often found in environments with very high background radiation levels such as in nuclear reactor cooling pools and the destroyed reactor in Chernobyl. These findings and the laboratory observations of the resistance of melanized fungi to ionizing radiation suggest a role for this pigment in radioprotection. We hypothesized that the radioprotective properties of melanin in microorganisms result from a combination of physical shielding and quenching of cytotoxic free radicals. We have investigated the radioprotective properties of melanin by subjecting the human pathogenic fungi Cryptococcus neoformans and Histoplasma capsulatum in their melanized and non-melanized forms to sublethal and lethal doses of radiation of up to 8 kGy. The contribution of chemical composition, free radical presence, spatial arrangement, and Compton scattering to the radioprotective properties of melanin was investigated by high-performance liquid chromatography, electron spin resonance, transmission electron microscopy, and autoradiographic techniques. Melanin protected fungi against ionizing radiation and its radioprotective properties were a function of its chemical composition, free radical quenching, and spherical spatial arrangement.     

A comparison of depth dependence of dose and linear energy transfer spectra in aluminum and polyethylene

A set of four tissue-equivalent proportional counters (TEPCs), with their detector heads at the centers of 0 (bare), 3, 7 and 9-inch-diameter aluminum spheres, were flown on Shuttle flight STS-89. Five such detectors at the centers of polyethylene spheres were flown 1 year earlier on STS-81. The results of dose-depth dependence for the two materials convincingly show the merits of using material rich in hydrogen to decrease the radiation exposure to the crew. A comparison of the calculated galactic cosmic radiation (GCR) absorbed dose and dose-equivalent rates using the radiation transport code HZETRN with nuclear fragmentation model NUCFRG2 and the measured GCR absorbed dose rates and dose-equivalent rates shows that they agree within root mean square (rms) error of 12.5 and 8.2%, respectively. However, there are significant depth-dependent differences in the linear energy transfer (LET) spectra. A comparison for trapped protons using the proton transport code BRYNTRN and the AP-8 MIN trapped-proton model shows a systematic bias, with the model underpredicting dose and dose-equivalent rates. These results show the need for improvements in the radiation transport and/or fragmentation models.     

Retrospective dosimetry after criticality accidents using low-frequency EPR: a study of whole human teeth irradiated in a mixed neutron and gamma-radiation field

In the context of accidental or intentional radiation exposures (nuclear terrorism), it is essential to separate rapidly those individuals with substantial exposures from those with exposures that do not constitute an immediate threat to health. Low-frequency electron paramagnetic resonance (EPR) spectroscopy provides the potential advantage of making accurate and sensitive measurements of absorbed radiation dose in teeth without removing the teeth from the potential victims. Up to now, most studies focused on the dose-response curves obtained for gamma radiation. In radiation accidents, however, the contribution of neutrons to the total radiation dose should not be neglected. To determine how neutrons contribute to the apparent dose estimated by EPR dosimetry, extracted whole human teeth were irradiated at the SILENE reactor in a mixed neutron and gamma-radiation field simulating criticality accidents. The teeth were irradiated in free air as well as in a paraffin head phantom. Lead screens were also used to eliminate to a large extent the contribution of the gamma radiation to the dose received by the teeth. The EPR signals, obtained with a low-frequency (1.2 GHz) spectrometer, were compared to dosimetry measurements at the same location. The contribution of neutrons to the EPR dosimetric signal was negligible in the range of 0 to 10 Gy and was rather small (neutron/gamma-ray sensitivity in the range 0-0.2) at higher doses. This indicates that the method essentially provides information on the dose received from the gamma-ray component of the radiation.     

Dose estimation for astronauts using dose conversion coefficients calculated with the PHITS code and the ICRP/ICRU adult reference computational phantoms

Absorbed-dose and dose-equivalent rates for astronauts were estimated by multiplying fluence-to-dose conversion coefficients in the units of Gy.cm² and Sv.cm², respectively, and cosmic-ray fluxes around spacecrafts in the unit of cm⁻² s⁻¹. The dose conversion coefficients employed in the calculation were evaluated using the general-purpose particle and heavy ion transport code system PHITS coupled to the male and female adult reference computational phantoms, which were released as a common ICRP/ICRU publication. The cosmic-ray fluxes inside and near to spacecrafts were also calculated by PHITS, using simplified geometries. The accuracy of the obtained absorbed-dose and dose-equivalent rates was verified by various experimental data measured both inside and outside spacecrafts. The calculations quantitatively show that the effective doses for astronauts are significantly greater than their corresponding effective dose equivalents, because of the numerical incompatibility between the radiation quality factors and the radiation weighting factors. These results demonstrate the usefulness of dose conversion coefficients in space dosimetry.     

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.