Neutron Dose Rates Research Articles

Developing power plant materials using the life cycle lens.

The Spherical Tokamak for Energy Production (STEP) environment will include magnetic, thermal, mechanical and environmental loads far greater than those seen in the Joint European Torus campaigns of the past decade or currently contemplated for ITER. Greater still are the neutron peak dose rates of 10-6 displacements per atom, per second, which in-vessel materials in STEP are anticipated to be exposed to. Reduced activation and high-fluence resilience therefore dominate the materials strategy to support the STEP Programme. The latter covers the full life cycle from downselected compositions and new microstructural developments to irradiation-informed modelling and end-of-life strategies. This article discusses how the materials downselection is oriented in plant power trade-off space, outlines the development of an advanced ferritic-martensitic structural steel, describes the 'Design by Fundamentals' mesoscale modelling approach and reports some of the waste mitigation routes intended to make STEP operations as sustainable as possible.This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

Evaluation of neutron dosimetry capabilities with the MC-15 portable multiplicity counter

This work proposes a preliminary neutron dose rate estimation method for a neutron multiplicity detector through measurement- and simulation-based analyses. Uncharacterized neutron-emitting sources may be encountered in situations such as nuclear emergency response, safeguards, and treaty verification. These circumstances may present irradiation risk to personnel conducting field assay, search, and characterization measurements. It is therefore of interest to provide a field neutron dosimetry capability with the existing neutron multiplicity counting (NMC) capabilities. To date, no commercially-available neutron detection systems are capable of both accurate NMC and real-time neutron dosimetry. This work will focus on estimating dose rate using input from a single fielded NMC called the MC-15. The energy-dependent neutron detection efficiency response of the MC-15 was quantified in monoenergetic neutron simulations and evaluated in response to two neutron-emitting sources and to a polyethylene-moderated source. The results were compared to existing neutron dosimeters and established the proof of concept for further investigation of the MC-15 for dose estimation. Measurement results were also replicated in simulations; additional simulations were then conducted to expand upon the limited empirical data. The initial empirical results provided a conversion factor appropriate for use when measuring 252Cf neutrons that is independent of polyethylene shielding presence and thickness. The simulated data sets were then used to evaluate a fit equation allowing estimation of the neutron dose rate for less restricted geometric configurations and dependent only on the distance between the source and the detector. Additionally, the energy dependence of the efficiency response indicates that further empirical evaluations could provide energy-dependent conversion factors for broader neutron dosimetry capabilities with a wider range of neutron-emitting sources.

Neutron spectrum measurements near KAMINI reactor south beam vault door using nested neutron spectrometer.

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275-300μSv/h and 175-200μSv/h, respectively, when the reactor is operating at 10kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178keV. Apart from energy reduction, 10kW normalised neutron flux of south beam pit is also attenuated by the shield by 25000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated.

Photon and neutron dose evaluation at the Beam Test Facility of the INFN - National Laboratory of Frascati

Dose evaluation and direct measurements are fundamental for radiation protection in non-conventional accelerator facilities, both before and after the primary and secondary shielding. In this paper, we will report about the experimental setup, data acquisition and analysis, together with FLUKA modeling, of the dose measurements test carried out in the Beam Test Facility (BTF) of the INFN - Frascati’s National Laboratories (LNF), where an intense mixed field is produced and measured with thermoluminescent dosimeters. BTF is an extraction and transport line of DAΦNE LINAC (Buonomo et al. 2021; Mazzitelli et al. 2003). It is optimized for electrons and positrons production in a wide range of intensity, energy (30 MeV–800 MeV), beam spot dimensions and divergence, using both primary and secondary beam of the DAΦNE LINAC. Through the years, the BTF has gained an important role in particle detectors test and development with electron/positron beam. A small fraction of the BTF’s shifts have been dedicated to radiation damage test using LINAC electron primary beam up to 5×1010 e-/s. As radiation protection group of the LNF, we evaluated the dose when electrons impinging on a Pb target from: (i) photon Bremsstrahlung production; (ii) photoneutron production. Three dedicated tests with 503 MeV electrons impinging on a ∼16 cm thick Pb target have been carried out in February, June 2022 and in January 2023, using TLD700 and TLD600, measuring doses at several charge intervals. The aim of this study focuses on evaluating dosimetric quantities produced by the mixed field, air kerma for the photon component, and ambient dose equivalent for the neutron one, using thermoluminescence dosimeters calibrated with low-energy standards: Cs-137 and Am-Be. The approach adopted involves the use of Monte Carlo simulations of the experiment, both to benchmark against experimental measurements and to validate the results obtained for energies higher than those of calibration. The results of this comparison show excellent agreement between measured and simulated quantities in the forward direction, allowing us to conclude and confirm the validity of the calibrations themselves.

Compact fusion blanket using plasma facing liquid Li-LiH walls and Pb pebbles

Liquid plasma facing walls allow for increased neutron-wall loading expanding the design space of fusion power-plants and experimental devices towards compact high-field reactors. This study presents the design of a compact radial build blanket for fusion devices composed of variable quantities of Lead (Pb) and Lithium-Lithium Hydride (Li-LiH). A tank-like cylindrical neutronic model of the early design of the stellarator reactor proposed by Renaissance Fusion is implemented in OpenMC (neutron transport and dose rate analyses). The reactor's radial build composition and blanket layer thicknesses are varied to fulfill the requirements on tritium breeding ratio (TBR), nuclear heat extraction, radiation shielding (for the coils, internal structures and external environment) for a stellarator-based power-plant. The analyses suggest that a radial build lower than a meter thick between the plasma and coils would be sufficient to allow for a TBR ∼ 1.60, an energy multiplication factor of ∼ 1.07, to capture ≥ 90% of the nuclear heat, limit the neutron fluence at the coils below 1019 n/cm2, and limit the structural damage on the liquid metal vessel and magnet structure. In particular, a blanket composed of 32 cm of Pb and Li-LiH, 54 cm of a heavy metal hydride such as vanadium hydride (VH2), along with a 1.3 m of concrete bioshield, would minimize the radial build of the stellarator reactor while fulfilling tritium breeding, shielding and heat extraction requirements.

Effect of fibers and boron carbide on the radiation shielding properties of limestone and magnetite aggregate concrete

Research on new composite materials in the field of radiation shielding materials requires adequate means and meticulousness in their manufacture and characterization. In this study, concretes that use elements that theoretically improve their shielding properties are developed, such as fibers and boron-rich additions. Both magnetite and limestone concrete have been manufactured, incorporating steel fibers to improve the linear attenuation coefficient and polyvinyl alcohol to increase neutron scattering cross section. The incorporation of boron carbide is intended to increase the neutron absorption cross section. 5 mixes have been manufactured that have been tested with several gamma radiation sources (570–1280 keV) and with an Am–Be neutron source (111 GBq). The experimental results have been used to validate simulations carried out with MAVRIC® and have led to the modelling of a simplified spent nuclear fuel cask. Boron carbide produces a decrease of more than 50% on the neutron dose rate in contact with the cask. Incorporating PVA fibers supposes gain in neutron attenuation capacity of up to 13%, while fibers have a negligible effect on the photon dose rate.

Neutron irradiation-induced shutdown dose rate estimation in EU DEMO divertor

The EU DEMO fusion reactor, which is part of the EUROfusion PPPT program and the successor to ITER, aims to facilitate the transition of fusion from the science laboratory to industrial use. Its key objective is net energy production. Its critical component, the divertor, manages power exhaust, ash, and helium removal. The key elements of the divertor include Plasma Facing Components (PFCs), shielding liners, reflector plates, divertor cassettes, and a vacuum vessel locking system. It is imperative that the shutdown dose rates are assessed to ensure workers' radiation protection during maintenance, plan the disposal of radioactive material, and ensure the overall safety of the plant. The shutdown dose rate tool employed in the present study is R2Smesh, which couples MCNP for radiation transport and FISPACT for neutron activation to calculate neutron-induced shutdown dose rates, utilizing an advanced, semi-heterogeneous EU DEMO divertor model, presenting a challenging task in neutron transport calculations.

Radiation shielding properties of composites of TiZrNbHfTa refractory high entropy alloy reinforced with TiZrNbHfTaOx high-entropy oxide

Radiation shielding materials play a critical role in advancing the next generation of nuclear reactors. Among these, high-entropy materials with refractory properties emerge as a new category for radiation shielding applications. This study introduces TiZrNbHfTa-TiZrNbHfTaOx composites tailored for such purposes. Initially, a TiZrNbHfTa refractory high-entropy alloy (RHEA) is synthesized via mechanical alloying, followed by the oxidation of the alloy to obtain its refractory high-entropy oxide (RHEO), TiZrNbHfTaOx. Composites are then prepared by blending 5%, 10%, and 20% weight fractions of RHEO with RHEA using mechanical alloying. Comparative analysis of shielding properties reveals a decrease in mass attenuation coefficient and effective atomic number with increasing RHEO content. While RHEA exhibits superior shielding performance against gamma photons and neutrons, the composite containing 20% RHEO emerges as the optimal shield against alpha and proton irradiations. The equivalent neutron dose rate and removal cross-section values are obtained 49.8% and 0.142 for the composite with 20% RHEO, respectively. These findings, supported by computational simulations in this study, contribute valuable insights for the advancement of novel shielding materials crucial for future nuclear reactor technologies.

Radiation shielding for inertial electrostatic confinement fusion system utilizing concrete and water

In this investigation, the impact of water thickness and various types of concrete materials, each characterized by different densities and elemental compositions, is examined for their role in shielding a radiation generator based on Inertial Electrostatic Confinement Fusion (IECF) device. The study focuses on assessing several vital parameters, including the effective removal cross-section (∑Rt) for fast neutrons, total mass attenuation coefficients (μmt), linear attenuation coefficients (μl), half-value layer (HVL), and mean free path (MFP) for X-rays across different concrete types. Different water thicknesses around the IECF chamber, ranging from 0.5 to 10.0 cm, are investigated, and five concrete types are evaluated: Ilmenite-magnetite Concrete (IMC), Ordinary Concrete-1 (OC1), Barite-containing Concrete (BC), Ordinary Concrete-2 (OC2), and Serpentine-containing Concrete (SC). The results indicate that, among these materials, SC requires the least thickness to attenuate 2.45 MeV generated from IECF to 1/100th of its initial intensity across varying water thicknesses (6.5 cm in case of 5 cm water thickness). The values of μmt, HVL, and MFP are also calculated for different water thicknesses and X-ray energies (ranging from 0.2 to 3.0 MeV). These calculations highlight BC as the material requiring the least thickness to attenuate X-rays to 1/100th of their initial intensity. Moreover, neutron dose rate measurements are conducted on a commercial IECF system shielded with 50 cm of water and operated at a neutron intensity of 105 ns−1, which was ∼12 nSv/h on average, approximately 0.002% of the initial intensity. This underscores the efficacy of water shielding in attenuating the outcomes of IECF.

Neutron spectra from photonuclear reactions: Performance testing of Monte-Carlo particle transport simulation codes

The production of neutrons in photon-induced nuclear reactions in the giant-dipole-resonance energy domain remains a topic of high interest for various applications, including the activation and decommissioning of electron accelerator facilities, the detection of illicit materials for homeland security, and the evaluation of neutron dose received by patients during radiotherapy treatments. General-purpose Monte-Carlo (MC) simulation codes for particle transport are intensively used to account for photoneutron production in these applications. However, due to the current scarcity of measured photoneutron energy spectra in the literature, experimental validation of MC-simulated photoneutron energy distributions is not always feasible. Therefore, a critical benchmark among simulation results from various MC codes presently appears as the only option to systematically assess their capabilities in accurately simulating photoneutron production for nuclear reactions of interest. In this work, neutron energy spectra from several targets under irradiation by 20 MeV photons are simulated, employing various state-of-the-art MC codes (FLUKA, Geant4, MCNP6, and PHITS) in their default or generally employed settings. A detailed analysis of the simulated neutron spectra allows one to not only assess the performance of various MC codes in applications such as those mentioned above, but also to partially gauge the incurred systematic uncertainty, and to highlight the present need for more comprehensive evaluated nuclear data in this domain. Ultimately, this work suggests that more prudence is required when using MC codes for applications where photonuclear reactions play a dominant role and where not only the production rate but also the energy spectrum of the emitted neutrons matters.

Experimental evaluation of transition rate of sapphire crystal for thermal and fast neutrons using MNSR vertical neutron beam line

Using a perfect single crystal as a neutron filter allows us to have a thermal neutron beam with almost no background of fast neutrons. Single crystals of Al2O3 (sapphire) have proven to be effective filters for fast neutrons and are incorporated into neutron instruments. The present work would experimentally investigate c-axis neutron transmission rate by using different crystal thicknesses. In fact, the optimal thickness for sapphire filter is the one that maximizes the transmission of low energy neutrons and minimizes the transmission of fast neutrons, if there is no significant decrease in thermal neutron flux. In addition, neutron-filtering power of a-axis and c-axis sapphire crystals were compared with each other using different tests on a 2.5 cm slab of the sapphire crystals. The experimental tests were carried out by means of the available neutron flux top of the vertical neutron beam line of the Isfahan Miniature Neutron Source Reactor (MNSR) in two methods of foil activation and flux monitoring. In addition, the thermal and fast neutron dose rate reduction was discussed by using different thicknesses of the c-axis crystal.

Sub-background radiation exposure at the LNGS underground laboratory: dosimetric characterization of the external and underground facilities

Radiobiological studies conducted in Deep Underground Laboratories allow to improve the knowledge of the biological effects induced by ionizing radiation at low doses/dose rates. At the Gran Sasso National Laboratory of the Italian Institute of Nuclear Physics we can study the possible differences in behavior between parallel biological systems, one maintained in a Reference-Radiation Environment (RRE, external) and the other maintained in an extremely Low-Radiation Environment (LRE, underground), in the absence of pressure changes, the RRE and LRE laboratories being at the same altitude. For these investigations, it is mandatory to evaluate the dose rate values at RRE and LRE. The aim of our work is to provide a comprehensive dosimetric analysis for external and underground laboratories. Measurements of the different low Linear Energy Transfer (LET) components at RRE and LRE were performed using different detectors. Gamma dose rates were 31 nSv/h at RRE and 27 nSv/h at LRE respectively. The muon dose rate was 47 nSv/h at RRE and negligible at LRE (less than pGy/h). Dosimetric measurements were also carried out to characterize the devices used to modulate the gamma dose rate, namely, a gamma source irradiator (to increase the dose rate by about 90 nSv/h) and shields (of iron at LRE and lead at RRE). Using the iron shield at LRE a dose reduction factor of about 20, compared to the RRE, was obtained for the low LET components; inside the lead shield at RRE the gamma component was negligible compared to the muonic component. Radon activity concentrations were approximately of 20 Bq/m3 at both LRE and RRE. The intrinsic contribution of radioactivity in the experimental set up was of 0.25 nGy/h, as evaluated with a GEANT4-simulation, using as input the measured activity concentrations. GEANT4 simulations were also performed to calculate the neutron dose rate at RRE, yielding a value of 1.4 nGy/h, much larger than that at LRE (which is less than pGy/h). In conclusion, RRE and LRE are currently characterized and equipped to perform radiobiological studies aimed at understanding the involvement of the different low LET components in determining the response of biological systems.

Combined anticancer effects of neutron radiation and genistein on prostate cancer cells

Prostate cancer is the most frequent cancer in men and the second largest cause of cancer-related mortality after lung cancer. Today, studies on neutron radiation therapy show that this method alone is not able to prevent the progression of the disease in most patients with prostate cancer, so it is better to combine neutron radiation with an auxiliary antitumor such as genistein. We found that genistein has anticancer properties and few antioxidant capabilities and could cause the death of prostate cancer cells, mostly through triggering apoptosis. Here the purpose of this study was to evaluate the effect of very low dose rate neutron irradiation emitted from 241Am-Be neutron source as well as the effect of the combination of neutron irradiation and adding the optimal concentration of genistein (480 μM) on the attenuation of prostate cancer PC3 cells in 3D culture environment. MTT assay, Neutral red assay, NO assay, Catalase activity, Reduced glutathione (GSH), Cytochrome c releasing, and Comet assay were used to assess the effect and type of mortality caused by neutron radiation and also neutron and genistein combination in prostate cancer cells. The results of cytotoxicity test reveal that very low dose rate neutron radiation (0.6 mGy/h) with or without genistein at 5, 10, 15, and 20h reduced the survival of prostate cancer cells. Moreover, the rate of necrosis due to the neutron effect at different irradiation times enhanced with the increasing time. Despite significant anticancer effect of Genistein, it cannot be used simultaneously with neutron radiation for prostate cancer treatment.

Отдаленные последствия γ, n-облучения мышей: снижение длины теломер и развитие опухолей

Purpose: To investigate the telomere length (TL) of bone marrow and thymus cells as a marker of replicative aging late after the prolonged γ, n-irradiation of mice at low and moderate doses and analysis of the appearance of tumors by the end of the experiment − after 14 months. Material and methods: C57Bl/6 and CBA mice were irradiated at doses of 10–500 mGy at the OR-M facility using Pu-Be radionuclide sources at a total absorbed dose rate of neutrons and gamma rays of 2.13 mGy/h, 75 % of which – 1.57 mGy/h – accounted for neutrons with an average energy of 3.5 MeV. Absolute TL in bone marrow and thymus cells was determined using real-time PCR 2 months and 1 year 2 months after irradiation, and the mean TL was calculated. Tumors found during the mice organs examination after autopsy were subjected to histological examination. Results: It was shown that the TL in bone marrow and thymus cells of control CВA mice was 2 times higher than the TL observed in C57Bl/6 mice. Prolonged γ, n-irradiation of C57Bl/6 mice led to a dose-dependent decrease in TL in bone marrow cells 14 months after exposure, which was statistically significant at doses of 100 and 500 mGy. A decreased TL in the thymus was found only at a dose of 500 mGy. During this period, TL in bone marrow cells of CBA mice was reduced in dose-independent manner, starting from as low as 10 mGy, but no statistically significant decrease in TL was found in the thymus. The results obtained indicate the acceleration of replicative senescence of bone marrow cells in mice in the long term period after γ,n-irradiation already at low doses, and in thymus cells only at a dose of 500 mGy. Twenty-four hours after irradiation at doses of 100 and 500 mGy the number of leukocytes in mice of both lines was reduced, which was recovered in C57Bl/6 mice after a week, and in CBA mice – after two weeks. In 14 months after γ, n-irradiation, the appearance of tumors was found in mice of both studied lines: in CBA mice, lung adenocarcinoma at a dose of 50 mGy (in 1 out of 10) and uterine carcinosarcoma at a dose of 500 mGy (in 1 out of 10); in C57Bl/6 mice, keratinizing squamous cell carcinoma of the uterus at a dose of 500 mGy (2 out of 10) was seen in the absence of tumors in control mice. Histological examination of the liver of CBA mice after γ, n-irradiation at a dose of 500 mGy revealed deep dystrophic changes, the causes of which are not clear. Conclusion: The results obtained indicate a high biological hazard of prolonged γ, n-irradiation at doses above 10 mGy, since after irradiation at this dose, an acceleration of replicative senescence of bone marrow cells in the long-term period was found, and the possibility of tumor formation increases after irradiation at a dose of 50 mGy and higher.

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.

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.

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.

Performance evaluation of a currently in-use dry storage cask design for spent accident tolerant fuel loading case under normal operation condition

Abstract The aim of this study is to assess the performance of an existing dry storage cask design for accident tolerant fuel loading case and to examine the compliance with the safety limits applied currently for dry storage. In the study, for a dry storage cask design currently in use, criticality calculations, dose rate evaluation and thermal analyses are performed in case of loading with the accident tolerant fuel discharged from a PWR. Firstly, for an accident tolerant fuel selected among the concepts proposed for use in light water reactors, burnup analyses are performed by utilizing the Serpent continuous energy code and spent fuel characteristics are determined. Then, criticality analyzes are carried out by using the Serpent Monte Carlo code for the case of loading the accident tolerant fuel into the selected dry storage cask design. Gamma and neutron dose rates at the outer surface and close distances of the storage cask are determined with the Serpent code. To evaluate the thermal performance of the storage cask, thermal analyzes are performed by using the ANSYS Fluent computational fluid dynamics code. The analysis results are compared with the nuclear safety criteria applied to dry storage casks. Results of the analysis show that the dry storage cask design currently in-use does not exceed the criticality, dose rate and maximum surface temperature limits when loaded with spent accident tolerant fuel.

Radiation characterizations of two isotopic neutron sources merging in one irradiator for experimental applications in the laboratory

Neutron sources are utilized for different aims, such as studying the material’s internal structure, investigating the materials’ crystal structures, and neutron therapy. Using more than one isotopic neutron source in some aims could be beneficial. Herein, a neutron irradiator consisting of two 241Am-9Be neutron sources has been designed. Radiation characterizations, including the neutron fluence and dose rate, were investigated. The Monte Carlo simulation code Monte Carlo N-particle transport code version 5 (MCNP5) was used to conduct all the required investigations to achieve an optimal design. Two factors were considered; the first was the dose rate due to both neutron and gamma radiations to meet radiation protection requirements; MCNP5 was used to determine the adequate thickness of the shielding material. The second parameter is the neutron fluence. Two irradiation sites have been proposed, one for fast neutron irradiation and the other for thermal neutron irradiation. By utilizing MCNP5, we have determined the radiation characterization at the two sites to be used as the irradiation sites. After the optimal design had been produced, the thermal and the epithermal neutron fluences inside the two irradiation sites were determined experimentally. The experimental results showed a perfect agreement with the simulation ones. The neutron irradiator designed from merged two isotopic neutron sources could be a benefit in the neutron radiation applications in the labs.

Neutron gamma analysis of soil carbon: Post-irradiation physicochemical effects

In-situ soil carbon measurements would be beneficial when assessing carbon (C) sequestration practices and associated C credits. Neutron gamma analysis, which registers gamma rays that appear due to neutron irradiation, is a tool that can be used for such assessments. However, questions regarding post-effects of neutron irradiation on soil remain unexplored. Temporal post-irradiation effects (neutron flux 2⋅107 neutron/s, neutron energy 14 MeV) on soil chemical and physical attributes were investigated using a previously constructed pulsed fast-thermal neutron-gamma analysis system. Neutron and gamma dose rate distributions during stationary irradiation using this system and Monte-Carlo computer simulations were found to be in agreement; a stationary 1 h irradiation period (conservative or worse-case scenario) was utilized in these scenarios. Physical effects of activating new radioactive isotopes were experimentally determined by assessing changes in the post-irradiated soil gamma spectra (“hot background”) over time; additional soil radioactivity decreased to natural background levels within ∼1 h. Radiolytic decomposition estimates of soil water and soil organic material (primarily cellulosic residue), based on received dose loads and known radiation-chemical yields of water and organic material, were practically negligible. Results indicate that adsorbed radiation doses in scanning mode would be 500 to 1000 times less than in static mode. Thus, neutron gamma analysis does not impact physicochemical aspects of soil health and can be used for soil elemental content determinations without additional radiation safety concerns.