Neutron Dose Research Articles

Characterization of the Radiation Contamination in an 18MV Linac Treatment Room Made of some Nanoparticles Mixture

Purpose: Shielding against radiation in radiotherapy and radiology requires deep knowledge of radiation physics and shielding design methods. The application of nanoparticles in the photon and neutron dose moderation has been proven in the literature.
 Materials and Methods: Effective neutron mass removal cross-section (ΣR/ρ) was for the ordinary concrete doped with 50nm in diameter nanosphers mixture of Fe2O3(5%), WO3(5%), B4H (5%), Pb2O3(5%) was estimated with MCNP5 Monte Carlo (MC) code and N-XCOM computational program. An 18MV linac room made of the nanoparticles dopped ordinarily simulated. Additionally, the room was considered with three legs in the maze and photoneutron and capture γ-ray Dose Equivalent (DE) were estimated at the modeled rooms maze.
 Results: Total ΣR/ρ with energies 100 keV-2000keV was estimated using MC and N-XCOM as 0.02802-0.02687 cm2/gand 0.02810- 0.02687 cm2/g, respectively. Total ΣR/ρ of the neutron for the pure ordinary concrete was estimated by MC code and N-XCOM with energies 100 keV-2000keV as 0.02802-0.02687 cm2/g and 0.02810- 0.02687 cm2/g. Borated Polyethylene (BPE) and Lead required thickness calculated as 7.43×10-06 mm and 4.73×10-06 mm for the capture γ-ray shielding.

Feasibility study of nanomaterials synthesis at MNSR research reactor through design and construction of a gamma irradiation cell

Herein, a gamma irradiation cell consisting of a borated polyethylene (BPE) cylinder with a cadmium inner wall was designed and constructed at Isfahan miniature neutron source reactor (MNSR) for synthesis of metal nanoparticles. The cell is placed in a dry channel irradiation site near the reactor core. The performance of this irradiation cell with appropriate shape and size was evaluated and the neutron flux (φ) and the neutron and gamma dose rates (Dn and Dg) inside it were calculated using MCNP code. The results showed that the designed cell could deliver an appropriate gamma dose rate to samples and reduce the thermal neutron flux from 5.82 × 1010 to 6.28 × 107n.cm−2.s−1, i.e. over 99.89% attenuation. Also, the gamma dose rate inside the designed neutron shield is calculated to be 3.83 kSv.h−1. To evaluate the designed and constructed cell, synthesis of Pd and Ag nanoparticles (NPs) on SiO2 supporter was done. Prepared nanocomposites were characterized by using XRD, FE-SEM, and EDXA techniques. Observed results showed metallic Pd and Ag NPs were produced successfully employing a designed gamma irradiation cell at the reactor via gamma-assisted route in nanometer sizes for the first time.

Assessing the impact of neutron relative biological effectiveness on all solid cancer mortality risks in the Japanese atomic bomb survivors

Purpose Risk analyses, based on relative biological effectiveness (RBE) estimates for neutrons relative to gammas, were performed; and the change in the curvature of the risk to dose response with increasing neutron RBE was analyzed using all solid cancer mortality data from the Radiation Effect Research Foundation (RERF). Results were compared to those based on incidence data. Materials and methods This analysis is based on RERF mortality data with separate neutron and gamma doses for colon doses, from which organ averaged doses could be calculated. A model for risk ratio variation with RBE was developed. Results The best estimate of the neutron RBE considering mortality data was 200 (95% confidence interval (CI): 50–1010) for colon dose using the weighted-dose approach and for organ averaged dose 110 (95% CI: 30–350). The ERR risk ratios for all solid cancers combined, for the best fitting neutron RBE estimate and the neutron RBE of 10 result in a ratio of 0.54 (95% CI: 0.17–0.85) for colon dose and 0.55 (95% CI: 0.18–0.87) for organ averaged dose. The risk to dose response curvature became significantly negative (concave down) with increasing RBE, at a neutron RBE of 170 using colon dose and at an RBE of 90 using organ averaged dose for males when fitting a linear-quadratic dose response. For females, the curvature decreased toward linearity with increasing neutron RBE and remained significantly positive until RBE of 80 and 40 using colon and organ averaged dose, respectively. For higher neutron RBEs, no significant conclusion could be drawn about the shape of the dose–response curve. Conclusions Application of neutron RBE values higher than 10 results in substantially reduced cancer mortality risk estimates and a significant reduction in curvature of the risk to dose responses for males. Using mortality data, the best fitting neutron RBE is much higher than when incidence data is used. The neutron RBE ranges covered by the overlap in the CIs from both the mortality and incidence analyses are 50–190 using colon dose and in all cases, the best fitting neutron RBE and lower 95% CI are higher than the value of 10 traditionally applied by the RERF. Therefore, it is recommended to consider uncertainties in neutron RBE values when calculating radiation risks and discussing the shape of dose responses using Japanese A-bomb survivors data.

A new Multi Leaf Collimator for the fission neutron therapy facility at FRM II

As a part of the general upgrade program of the fission neutron therapy facility MEDAPP at the research reactor FRM II, a newly designed, electronically operated Multi Leaf Collimator (MLC) was introduced. The design of the new collimator was optimized in order to increase dose output and to allow patient positioning closer to the collimator. The new collimator allows divergent and convergent beam shaping in order to maximize the effective area of the fission source used for irradiation. In comparison to the old MLC, the radiation dose for the staff is reduced by the motorized operation of the new MLC. The advanced beam shaping abilities of the new collimator are investigated by a comparison with depth dose curves and beam profiles of the old MLC. For the separation of neutron and gamma dose deposition in water, the two chamber dosimetry method recommended for mixed field dosimetry was applied. A re-calibration of the set of ionization chambers used for dosimetry in mixed neutron-gamma radiation fields at FRM II was performed. The calibration was done in a reference neutron field from 3H(p,n)3He-reactions in a low scattering environment. An overall increased dose output as intended by the design of the new MLC was shown. Also, improved dosimetric characteristics of the treatment field due to the optimized beam shaping abilities of the new MLC were demonstrated. In the calibration measurements of the magnesium ionization chamber used in the two chamber method, an unexpected increase of the neutron sensitivity by a factor of 8 was measured. The new MLC will enable more conformal dose delivery to the target volume in patients and therefore help to reduce healthy tissue damage by the application of neutron radiation.

Photo-neutron production of medical linear accelerator Elekta, precise model: A study with the Geant4 MC toolkit

We use Geant4 Monte Carlo (MC) toolkit to simulate a Precise Series Elekta linear accelerator. The simulation is validated for 6 and 15 MV X-rays with measured dose taken along the central axis of the radiation and off axis. The measurements are used in a clinical system, The neutrons generated from a (X, n) reaction when the 15 MV X-ray beam is on, are unwated byproducts that add up to the patient’s dose. Through the MC simulation we quantify the neutron dose contribution per delivered x-ray dose. Neutron fluence, in the units of neutron/cm2, contamination of the beam’s central axis was found to be proportionally dependent on the size of radiation field and decreases with the field size. This indicated that the number of neutrons produced in the head is constant, and when the field size opened wider, the neutrons distributed to a larger area, hence the fluence decreases. On the other hand, the neutron contamination outside the boundaries of the beam did not depend on its size and was found to be constant. This is intuitive, as those are the neutrons leaking from the head of the linac laterally, and as such are independent of the radiation beam size. Moreover, the maximum dose, dmax, for 15 MV x-rays from neutrons was found to be at a voxel 2 at the central axis at a depth of 2.5 cm inside the water phantom and it contributes an amount of 0.0003% to total dose.

A Novel, Rapid Response Renewable Biopolymer Neutron and Gamma Radiation Solid-State Detector for Dosimetry and Nuclear Reactor Flux-Power Mapping

A novel solid-state neutron and gamma radiation monitor-dosimeter based on biopolymer polylactic acid (PLA) is presented. The resulting detector (PLAD) technology takes advantage of property changes of the renewable PLA resin when subject to ionizing nuclear radiation. A simple yet rapid and accurate (±10%) low-cost (<$0.01/detector) mass loss upon dissolution (MLD) technique was successfully developed; MLD is based on a simple mass balance for discerning neutron and/or gamma doses using small (40 mg, ~4 mm diameter) ultra-low-cost (<$0.01) resin beads via dissolution in acetone. The GammaCellTM Co-60 irradiator, and the PUR-1 12 kW fission nuclear research reactor were utilized, respectively. Irradiation absorbed doses ranged from 1 to 100 kGy. Acetone bath temperature was varied from ~40 °C to ~54 °C. Results revealed a strong dependence of MLD on acetone bath temperature between neutron and gamma photon dose components; this allowed for the unique ability of PLAD to potentially perform as both a neutron-cum-gamma or as a gamma or neutron radiation dosimeter and intensity level detector. A linear trend is found for combined neutron and gamma radiation doses from 0 to 40 kGy when dissolution is conducted above 50 °C. The important potential ability to distinguish neutron from gamma radiation fields was scoped and found to be feasible by determining MLD at 45 °C. The potential was studied for simultaneous use as an in-core neutron and gamma monitor of an operating 3 GWt light-water reactor (LWR). Scoping tests were conducted with the pre-irradiated (@ 20 °C) PLAD resin beads followed by heating to in-core LWR coolant (300 °C) conditions for ~30 s corresponding to the time to reach ~40 kGy total doses in a typical 3 GWt LWR. MLD results were unaffected, indicating the exciting and unique potential for in situ (low-cost, accurate and rapid) simultaneous mapping of neutron and gamma radiation fluxes, related dosimetry, and fission power level monitoring.

Optimization design of the portable cylindrical water injection multilayer neutron spectrometer

The currently widely used multi-sphere neutron spectrometers still have many drawbacks, including complex design and processing, the need for multiple moderating spheres, high costs, large volumes, and complicated measurement procedures. This work proposes the portable cylindrical water injection multilayer neutron spectrometer (CWNS) as a promising alternative based on water pumping injection. The structure of CWNS consists of a central thermal neutron detector and a surrounding 6-layer of coaxial cylindrical water bags with varying diameters. During non-measurement periods, this CWNS is convenient to carry due to the absence of the need to inject moderating water. To optimize the CWNS design, we employed FLUKA simulation software to study and refine various parameters, including the thickness of the water bag, the material composition of the water bag, and the parameters of the supporting column. We finally achieved an optimized design. Specifically, the water bag of the CWNS is constructed using a 0.3 mm thick polyethylene film. The supporting column for the water bag is made of aluminum, providing stability and support to the overall structure. These optimized design parameters determine the specific size and configuration of the CWNS. The CWNS offers the benefits of convenient carrying, simplified processing, cost-effectiveness, and straightforward measurement. It has a promising potential use for the directional neutron dose monitoring.

28 Measurement of Neutron Dose for VMAT and SRS Treatment at Energies Between 6 and 10 MV

28 Measurement of Neutron Dose for VMAT and SRS Treatment at Energies Between 6 and 10 MV

Cladding Failure Modelling for Lead-Based Fast Reactors: A Review and Prospects

Lead-cooled fast reactors (LFRs) are considered one of the most promising technologies to meet the requirements introduced for advanced nuclear systems. LFRs have higher neutron doses, higher temperatures, higher burnup and an extremely corrosive environment. The failure studies of claddings play a vital role in improving the safety criteria of nuclear reactors and promoting research on advanced nuclear materials. This paper presented a comprehensive review of the extreme environment in LFRs based on the fuel performance analyses and transient analyses of reference LFRs. It provided a clear image of cladding failure, focusing on the underlying mechanisms, such as creep, rupture, fatigue, swelling, corrosion, etc., which are resulted from the motions of defects, the development of microcracks and accumulation of fission products to some extent. Some fundamental parameters and behavior models of Ferritic/Martensitic (F/M) steels and Austenitic stainless (AuS) steels were summarized in this paper. A guideline for cladding failure modelling was also provided to bridge the gap between fundamental material research and realistic demands for the application of LFRs.

Mixed field radiation measurement using MOSFET sensor

We report here a study performed on MOSFET (Metal Oxide Semiconductor Field Effect Transistor) sensor for the measurement of accumulated doses from a mixed field radiations (especially gamma and fast neutron) environment such as nuclear detonation. The MOSFET sensor has shown its capability towards fast neutron dose measurement along with its well known gamma dose measurement capabilities. The study results are helpful in the establishment of MOSFET as a Personal Dosimeter for mixed field radiation. The measured sensitivities for gamma and fast neutron dose measurements are 1.1 mV/cGy and 0.6 mV/cGy respectively.

Nuclear scoping analysis of ITER bioshield top lid toward its preliminary design review

During ITER operations, electronics located in the crane hall, which is above the tokamak, will be exposed to neutron and photon fields from both the plasma and the activated water. To protect the electronics, the implementation of dedicated shielding on the crane hall platform and the bioshield top lid is required. The design demands optimisation attending to constructability, weight limits, and radiation shielding requirements. This work evaluates eight shielding configurations by assessment of the neutron flux and dose accumulated over 4700 h of operation at 500 MW for electronics protection. This corresponds to a neutron wall load of 0.3 MW a/m² as specified in the ITER Project Specification. An intermediate-source approach has been followed with SRC-UNED, considering all relevant radiation sources while minimising the computational time required. Results were presented at the top lid Conceptual Design Review aiming to support decision-making. Further optimisation has since been performed to reach a top lid proposal for its Preliminary Design Review. All outcomes show that radiation levels above the north and south crane hall platforms are compatible with the critical electronics requirements.

Global dose distributions of neutrons and gamma-rays on the Moon

Dose assessment on the lunar surface is important for future long-term crewed activity. In addition to the major radiation of energetic charged particles from galactic cosmic rays (GCRs), neutrons and gamma-rays are generated by nuclear interactions of space radiation with the Moon’s surface materials, as well as natural radioactive nuclides. We obtained neutron and gamma-ray ambient dose distributions on the Moon using Geant4 Monte Carlo simulations combined with the Kaguya gamma-ray spectrometer measurement dataset from February 10 to May 28, 2009. The neutron and gamma-ray dose rates varied in the ranges of 58.7–71.5 mSv/year and 3.33–3.76 mSv/year, respectively, depending on the lunar geological features. The lunar neutron dose was high in the basalt-rich mare, where the iron- and titanium-rich regions are present, due to their large average atomic mass. As expected, the lunar gamma-ray dose map was similar to the distribution of natural radioactive elements (238U, 232Th, and 40K), although the GCR-induced secondary gamma-ray dose was significant at ~ 3.4 mSv/year. The lunar secondary dose contribution resulted in an additional dose of 12–15% to the primary GCR particles. Global dose distributions on the lunar surface will help identify better locations for long-term stays and suggest radiation protection strategies for future crewed missions.

Tensile properties of modified 316 stainless steel (PNC316) after neutron irradiation over 100 dpa

ABSTRACT The effects of fast neutron irradiation on tensile properties of 20% cold-worked modified 316 austenitic stainless steel (PNC316), which was improved swelling resistance and creep strength, claddings and wrappers for fast reactors were investigated. PNC316 claddings and wrappers were irradiated in the experimental fast reactor Joyo at irradiation temperatures between 400°C and 735°C to fast neutron doses ranging from 21 to 125 dpa. The post-irradiation tensile tests were carried out at room and irradiation temperatures. Elongations of PNC316 measured by the tensile tests were maintained at an engineering level, although the material incurred significant irradiation hardening and softening depending on irradiation temperature. The maximum swelling of PNC316 wrappers was about 2.5 vol.% at irradiation temperature between 400°C and 500°C up to 110 dpa. Japanese 20% cold-worked austenitic steels, PNC316 and 15Cr-20Ni, had sufficient ductility and work-hardenability even after above 10 vol.% swelling, while they had very weak plastic instabilities.

Thermomechanical Properties of Neutron Irradiated Al3Hf-Al Thermal Neutron Absorber Materials.

A thermal neutron absorber material composed of Al3Hf particles in an aluminum matrix is under development for the Advanced Test Reactor. This metal matrix composite was fabricated via hot pressing of high-purity aluminum and micrometer-size Al3Hf powders at volume fractions of 20.0, 28.4, and 36.5%. Room temperature tensile and hardness testing of unirradiated specimens revealed a linear relationship between volume fraction and strength, while the tensile data showed a strong decrease in elongation between the 20 and 36.5% volume fraction materials. Tensile tests conducted at 200 °C on unirradiated material revealed similar trends. Evaluations were then conducted on specimens irradiated at 66 to 75 °C to four dose levels ranging from approximately 1 to 4 dpa. Tensile properties exhibited the typical increase in strength and decrease in ductility with dose that are common for metallic materials irradiated at ≤0.4Tm. Hardness also increased with neutron dose. The difference in strength between the three different volume fraction materials was roughly constant as the dose increased. Nanoindentation measurements of Al3Hf particles in the 28.4 vol% material showed the expected trend of increased hardness with irradiation dose. Transmission electron microscopy revealed oxygen at the interface between the Al3Hf particles and aluminum matrix in the irradiated material. Scanning electron microscopy of the exterior surface of tensile tested specimens revealed that deformation of the material occurs via plastic deformation of the Al matrix, cracking of the Al3Hf particles, and to a lesser extent, tearing of the matrix away from the particles. The fracture surface of an irradiated 28.4 vol% specimen showed failure by brittle fracture in the particles and ductile tearing of the aluminum matrix with no loss of cohesion between the particles and matrix. The coefficient of thermal expansion decreased upon irradiation, with a maximum change of -6.3% for the annealed irradiated 36.5 vol% specimen.

Embrittlement of WCLL blanket and its fracture mechanical assessment

In the European fusion programme, the Water Cooled Lithium Lead breeding blanket (WCLL BB) uses EUROFER as a structural material cooled with water at temperatures between 295 °C–328 °C and a pressure of 155 bar. The WCLL BB will be significantly irradiated (>2 dpa), while some parts will not receive significant heat loads, e.g. the sidewalls or the back-supporting structures. The irradiation, together with the irradiation temperature of EUROFER below 350 °C, produces a shift of the ductile-to-brittle-transition temperature (DBTT) to levels above room temperature at neutron doses, causing material damage as low as 2–3 dpa. Even though the DBTT does not reach the operating temperature level, brittle/non-ductile fracture is a concern during in-vessel maintenance when the BB temperature is below the DBTT. Two loading scenarios were identified as severe in this respect: (i) re-pressurization of the WCLL BB cooling loop after in-vessel maintenance, and (ii) dead weight loads during lifting of the BB segment. The embrittlement of the WCLL BB was investigated by quantifying the local DBTT shift in its parts based on current knowledge of the embrittlement behaviour of EUROFER under neutron irradiation. Therefore, a suitable, not overly conservative procedure was derived considering dpa damage and transmuted helium effects. The results demonstrate the ability to identify the 3D spread of the severely embrittled zones in the structure whose impact on the structural integrity was assessed considering the risk of brittle/non-ductile fracture. Thereby, the fracture mechanics approach established in nuclear codes was applied assuming its applicability to EUROFER. The embrittled zones in the first wall (FW) and its sidewalls pass the criteria when assessing the relatively low stresses resulting from the coolant pressure. The assessment was then continued considering stresses appearing in the FW during maintenance, in particular, when lifting the BB segment and transporting it out of the vacuum vessel. In this context, the maximum tolerable flaw sizes were determined in a parameter study considering designs of the FW with different cooling channel wall thicknesses.

Measurements of photoneutron dose rate for 15 MV photon beam from medical linear accelerator using neutron survey meter

High-energy medical linear accelerators (>10MV) are increasingly used in the medical field to treat cancer patients depending on the treatment organ and patient physic. Medical linear accelerators (Linacs) operating above 10 MV photon beam produce unwanted neutrons (photoneutron) by means of photonuclear reaction (γ, n) between bremsstrahlung photon beams and constituent materials of Linac head. This study involved the measurement of the neutron dose rate from photon beam of medical linear accelerator (Elekta Synergy) operated at 15 MV photon. For the measurement of dose rate, a neutron survey meter LB 6411 probe was used. Neutron dose rate was obtained as a function of delivered dose, field size and detector position. In the study neutron dose rate was found 13.14 mSv/h at a field size of 10 × 10 cm2 at position (0,90,0) cm for 350 cGy and 9.26 mSv/h at 5 × 5 cm2 field size at position (0,80,0) cm for 200 cGy photon dose delivery. The dose rate varied with respect to detector position in an irregular manner; 2.95 mSv/h at position (-100,0,0) cm -on the side of maze entrance and 4.85 mSv/h at position (100,0,0) cm -off side of maze entrance.

Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor

The present study investigated the release of tritium from beryllium irradiated at 323 K to a neutron fluence of 4.67 × 1026 m−2 (E > 1 MeV), corresponding up to 22,000 appm helium and 2000 appm tritium productions. The TPD tests revealed a single tritium release peak during thermal desorption tests, irrespective of the heating mode employed. The tritium release peaks occurred at temperatures ranging from 1031–1136 K, depending on the heating mode, with a desorption energy of 1.6 eV. Additionally, the effective tritium diffusion coefficient was found to vary from 1.2 × 10−12 m2/s at 873 K to 1.8 × 10−10 m2/s at 1073 K. The evolution of beryllium microstructure was found to be dependent on the annealing temperature. No discernible differences were observed between the as-received state and after annealing at 473–773 K for 5 h, with a corresponding porosity range of 1–2%. The annealing at temperatures of 873–1373 K for 5 h resulted in the formation of large bubbles, with porosity increasing sharply above 873 K and reaching 30–60%.

Effects of neutron radiation generated in deep space-like environments on food resources

The impact of deep space cosmic rays on food resources is as important as the risks of cosmic rays to the human body. This study demonstrates the potential for neutrons as secondary radiation in deep space spacecraft to cause meat activation and oxidative modification of proteins and lipids. We conducted a series of experiments such as the neutron irradiation experiment, the radioactivation analysis and the biochemical analysis. Neutrons with energies from 1 to 5 MeV with doses from 0.01 Gy to 4 Gy were irradiated by the RIKEN accelerated-driven neutron source (RANS). Radioactive nuclei, 24Na, 42K, and 38Cl, were detected in the neutron-irradiated meat. The modification products of the proteins by oxidative nitration, 6-nitrotryptophan (6NO2Trp), and by a lipid peroxidation, 4-hydroxy-2-nonenal (4-HNE), were detected in several proteins with neutron dose dependent. The proteome analysis showed that many oxidative modifications were detected in actin and myosin which are major proteins of myofibrils. This study is of crucial importance not only as risk factors for human space exploration, but also as fundamental effects of radiation on the components of the human body.

Proof-of-concept study on a water phantom-based neutron spectrometer: Experimental test with 252Cf and 241Am-Be sources

Boron neutron capture therapy (BNCT) is a promising cancer treatment that uses energetic ions released from 10B(n, α)7Li reactions. Accurate assessment of neutron energy spectra is important for simulation-based evaluation of neutron doses during BNCT. In this study, a proof-of-concept study was conducted for a neutron spectrometry technique that involves the use of a water phantom, which is commonly used for quality assurance in BNCT, as a moderator. The technique involves applying unfolding to the count rate distribution of the thermal neutron counter measured within the phantom to derive the energy spectrum. We performed experiments using a spherical 3He proportional counter in neutron fields generated by 252Cf and 241Am-Be sources. The results demonstrated that the spectrometer reasonably reproduced neutron spectra and showed the potential of using a water phantom as a moderator for such a technique.

Hydrothermal crystal growth of tetralithium octafluoridozirconate (Li4ZrF8) and its physicochemical properties

A single crystal of tetralithium octafluridozirconate (Li4ZrF8) has been obtained by using a hydrothermal process. The crystal phase was analyzed using a powder X-ray diffraction pattern and matched with standard reference data. The high-resolution X-ray diffraction curve indicates that the crystal has good crystalline perfection. The thermal behavior of the crystal sample was investigated by TGA/DSC analysis. The optical bandgap of the crystal sample was calculated from the Tauc plot and found to be 5.8 eV. The crystal sample exhibits intrinsic X-ray excited luminescence with a peak maximum of 265 nm. The thermoluminescence measurement shows two peaks, one is at the high-temperature region of 545 K and another one is at the low-temperature region at 383 K. TL kinetic parameters such as activation energy (E) and frequency factor (s) for the peak at 545 K was estimated as 1.20 eV and 2.06 × 10−10 respectively. The sample has good TL sensitivity to the neutron dose in the order of µsv. The low-temperature thermoluminescence measurements were carried out to study the trap levels and found two major glow peaks at 108 and 130 K. Scintillation properties were analyzed under α-rays from a 241Am source. The decay time profile was well fitted with three components and the fitted values show the fast component of 500 ns (9 %), the medium component of 2.4 μs (24 %), and a slow component of 12.8 μs (67 %).