3D Printed Lung Phantom for Individual Monitoring

ABSTRACT

The Human Monitoring Laboratory, Health Canada (HML), has used a 3D printer to re-engineer its Lawrence Livermore National Laboratory (LLNL) foam lung sets (manufactured by Radiology Support Devices, Inc., Long Beach, CA). The foam sets are currently the HML standard for calibrating and performance testing lung-counting systems in Canada. This paper describes the process of creating and validating new 3D-printed lung sets modeled from one of the HML's existing RSD foam sets. The existing sets were custom made, making them costly and difficult to obtain or replace. Also, after many years of use, the HML has found that they are prone to wear and tear. When used with planar inserts containing various isotopes, the blank sets can become contaminated and are difficult to clean. Using 3D printing, the HML has created new blank lung sets that are nearly identical copies of the originals and are inexpensive and easily manufactured. Measurements using natural uranium (Nat U), 241Am, and 152Eu planar lung inserts were performed to compare obtained efficiencies at a wide range of energies using the original RSD foam sets and the 3D-printed ones. Both the foam and the 3D-printed lung sets were counted using the LLNL chest phantom positioned in the same counting geometry in the lung counting system. Biases, all below 15%, were obtained between the foam and the 3D-printed sets for energies above 40 KeV. Based on these results, as well as cost benefits and ease of use, the HML has decided to replace its original RSD foam lung set with the 3D-printed version for its lung performance testing program.

Modelling DTPA therapy following Am contamination in rats

A major challenge in modelling the decorporation of actinides (An), such as americium (Am), with DTPA (diethylenetriaminepentaacetic acid) is the fact that standard biokinetic models become inadequate for assessing radionuclide intake and estimating the resulting dose, as DTPA perturbs the regular biokinetics of the radionuclide. At present, most attempts existing in the literature are empirical and developed mainly for the interpretation of one or a limited number of specific incorporation cases. Recently, several approaches have been presented with the aim of developing a generic model, one of which reported the unperturbed biokinetics of plutonium (Pu), the chelation process and the behaviour of the chelated compound An-DTPA with a single model structure. The aim of the approach described in this present work is the development of a generic model that is able to describe the biokinetics of Am, DTPA and the chelate Am-DTPA simultaneously. Since accidental intakes in humans present many unknowns and large uncertainties, data from controlled studies in animals were used. In these studies, different amounts of DTPA were administered at different times after contamination with known quantities of Am. To account for the enhancement of faecal excretion and reduction in liver retention, DTPA is assumed to chelate Am not only in extracellular fluids, but also in hepatocytes. A good agreement was found between the predictions of the proposed model and the experimental results for urinary and faecal excretion and accumulation and retention in the liver. However, the decorporation from the skeletal compartment could not be reproduced satisfactorily under these simple assumptions.

Biokinetic model analysis with DTPA administration for a case of accidental inhalation of actinides in Japan

Accidental inhalation intake of plutonium isotopes and 241Am occurred at a Pu research facility in Japan in 2017, and the five workers involved in this accident were treated by the administration of Ca/Zn-diethylenetriaminepentaacetic acid (DTPA). For the worker who was most internally exposed, the therapy was continued over 1 y after the accident. Urinary samples collected before and after each administration were subject to bioassay to evaluate the efficacy of the dose reduction. This study performed numerical analyses using a biokinetic model dealing with 241Am-DTPA with reference to the European Coordinated Network on Radiation Dosimetry approach, which assumes that the complex of actinides and Ca/Zn-DTPA is generated in the designated compartments in the biokinetic model. The results of the model prediction well captured the trend of the observed urinary excretion in the long-term bioassay and would be useful to evaluate the efficacy of the Ca/Zn-DTPA administration for the worker involved in the accident.

Source efficiency of alpha-emitters applied to the skin surface

Skin surface contamination by alpha-emitters is in itself not hazardous, but it would cause significant internal exposure in the case of injured skin as well as misjudgment in direct in vivo measurements (e.g. lung counting). The present study determined the source efficiency of alpha-emitters (241Am) applied to swine skin samples by analysing the observed alpha-particle energy spectra using advanced alpha-spectrometric simulation. Based on our results, the source efficiency was determined to be 0.365 (alpha-particle s-1 per Bq) on average (c.f. 0.5 in the case of no self-absorption in the source). The decrease in source efficiency would be attributed primarily to the radionuclide entering hair follicles or deep wrinkles. The degradation of the measured spectra from the skin samples indicates the penetration of some radionuclides into the upper layers of the stratum corneum. Although this study was limited to results obtained from swine skin samples, it suggests that irregularities in the skin surface may affect direct alpha measurements.

Development of a new integrated IN-VIVO counting system at the QST

A new in-vivo counting system that functions as both a whole-body counter (WBC) and a lung counter (LC) was developed at the QST to enhance its dose assessment capability. This paper presents an overview of this system and the results of its performance tests. For use of the system as a WBC, three high purity germanium (HPGe) detectors installed in a 20-cm-thick iron shielding chamber are linearly arrayed over a subject lying on the bed, whereas two of the three HPGe detectors are placed over the subject's chest from side to side when using the system as an LC. The new in-vivo system was calibrated using three de-facto phantoms owned by the QST: an adult-male BOttle Manikin ABsorption (BOMAB) phantom, a Lawrence Livermore National Laboratory (LLNL) phantom and a Japan Atomic Energy Research Institute (JAERI) phantom. Monte Carlo simulations were also performed to determine an optimum location for the three detector array in the WBC mode and revealed that the peak efficiency for the BOMAB phantom (662 keV) was little varied as long as the middle detector was placed above the thorax and abdomen parts of the phantom. The calculated peak efficiencies agreed well with the observed peak efficiencies for photons with energies over 100 keV. For lung counting, a tentative Minimum Detectable Activity of 241Am was evaluated as 9.5 Bq for a counting time of 30 minutes, and a Japanese male subject with an average chest wall thinness (2.27 cm). The developed system is now ready for use.

THE IMPACT OF NEW ICRU95 QUANTITIES AND SPECTRUM DATA ON THE DOSIMETRIC QUANTITIES OF THE NEUTRON REFERENCE FIELDS AT PTB

In 2020, the ICRU released a new report which includes the re-definition of the operational quantities used in radiation protection and new conversion coefficients from physical quantities to operational radiation protection quantities. An assessment of the ambient and personal dose conversion coefficients for the reference neutron fields of radionuclide sources at PTB is necessary based on these new definitions. In this work, a numerical estimation of the conversion coefficients of moderated and unmoderated 252Cf and 241Am-Be neutron sources based on ICRU57 and ICRU95 reports and using spectrum data available in the ISO 8529-1 standard and at PTB are discussed. Two numerical approaches are used for this estimation to ensure the reliability of the calculated values: a direct calculation using MCNP6, and cubic interpolation of conversion coefficients datasets written in Python. The results show large differences between the spectrum-averaged operational quantities for the current and new conversion coefficients of up to 23%. The choice of spectrum data affects conversion coefficient values by 6-8%.

Distributions of fallout 137Cs, 239+240Pu and 241Am in a soil core from South Central China

The source and vertical distribution of ¹³⁷Cs,^239+240Pu and ²⁴¹Am activity concentrations in a soil core from Hunan Province, China were investigated. The maximum ¹³⁷Cs and ^239+240Pu activity concentrations were 15.45 ± 0.76 mBq/g and 0.819 ± 0.066 mBq/g, respectively. While the maximum ²⁴¹Am activity concentration in samples obtained from the core was 0.341 ± 0.019 mBq/g. The ²⁴⁰Pu/²³⁹Pu atom ratio and the ¹³⁷Cs/^239+240Pu activity ratio were 0.183 ± 0.011 and 19.5 ± 1.8, respectively, and both were consistent with the characteristic value of global fallout. The integrated ²⁴¹Am/^239+240Pu activity ratio for global fallout was also re-estimated. The measured ²⁴¹Am/^239+240Pu activity ratio (average 0.43 ± 0.07) in the samples was very close to the estimated value (0.45), which suggested their ²⁴¹Am also came from the global fallout. Regarding the vertical distribution of ¹³⁷Cs, ^239+240Pu and ²⁴¹Am in these red soil samples, all these radionuclides had higher concentrations in upper layers of several centimeters of soil while they had slightly lower concentrations in lower soil layers down to 30 cm. Vertical distributions of ¹³⁷Cs/^239+240Pu and ²⁴¹Am/^239+240Pu activity ratios indicated the migration velocity was Am ≈ Pu > Cs. The intrinsic chemical properties of the radionuclides as well as soil type and properties (acidic, nutrient-deficient and low in organic matter and cation exchange capacity) might be reasons for the differences in their migration behaviors.

Characterization of Thermoluminescent Dosimeters for Neutron Dosimetry at High Altitudes

Neutrons constitute a significant component of the secondary cosmic rays and are one of the most important contributors to natural cosmic ray radiation background dose. The study of the cosmic ray neutrons' contribution to the dose equivalent received by humans is an interesting and challenging task for the scientific community. In addition, international regulations demand assessing the biological risk due to radiation exposure for both workers and the general population. Because the dose rate due to cosmic radiation increases significantly with altitude, the objective of this work was to characterize the thermoluminescent dosimeter (TLDs) from the perspective of exposing them at high altitudes for longtime neutron dose monitoring. The pair of TLD-700 and TLD-600 is amply used to obtain the information on gamma and neutron dose in mixed neutron-gamma fields due to the present difference in 6Li isotope concentration. A thermoluminescence dosimeter system based on pair of TLD-600/700 was characterized to enable it for neutron dosimetry in the thermal energy range. The system was calibrated in terms of neutron ambient dose equivalent in an experimental setup using a 241Am-B radionuclide neutron source coated by a moderator material, polyethylene, creating a thermalized neutron field. Afterward, the pair of TLD-600/700 was exposed at the CERN-EU High-Energy Reference Field (CERF) facility in Geneva, which delivers a neutron field with a spectrum similar to that of secondary cosmic rays. The dosimetric system provided a dose value comparable with the calculated one demonstrating a good performance for neutron dosimetry.

TISSUE-EQUIVALENCE OF H2 GAS FOR MICRODOSIMETRY IN NEUTRON FIELDS: A GEANT4 MONTE CARLO STUDY

Hydrogen (H2) as filling gas in Tissue-equivalent proportional counter (TEPC) for measurements of microdosimetric distributions in neutron fields is investigated using the Monte Carlo-based Geant4 toolkit. The neutron fields considered are monoenergetic neutrons (1 keV-14 MeV) and ISO reference neutron sources 241Am-Be, 241Am-B, 252Cf and 252Cf + D2O. Based on the energy deposited in the gas cavities (tissue-equivalent (TE) propane and H2) of the TEPC, density scaling correction, ${f}_{\rho }$, to be applied on H2 gas to achieve tissue-equivalence is calculated for the above sources at 1μm site size. The calculated value of ${f}_{\rho }$ for the ISO-neutron sources is 0.40, which suggests that energy deposition in the gas cavities is predominantly due to crossers. In the case of monoenergetic neutrons, depending upon the energy, ${f}_{\rho }$ is in the range of 0.11-0.45. The study shows that the TE propane and H2-based microdosimetric distributions are comparable when the density of H2 is scaled appropriately. Mean quality factor $\overline{Q}$ calculated based on microdosimetric distribution increases initially with neutron energy up to 100 keV and thereafter decreases with energy. Depending upon the neutron energy, TE propane and H2-based $\overline{Q}$ values are comparable to within 3-15%, and for the ISO sources, the comparison is within 5-8%. For the ISO neutron sources, $\overline{Q}$ values are in the range of ~10-15.

Spectroscopic study of the effect of low dose fast neutrons on the hemoglobin structure

Cosmic rays, nuclear accidents, and neutron therapy could be sources for exposure to low-dose fast neutrons. However, the study of low dose effects needs sentient techniques to detect slight alteration happen by this low dose. Herein, the effects of low-dose fast neutrons on the structure of hemoglobin (Hb) using spectroscopic techniques, namely, Fourier transform infrared (FTIR), Raman, and ultraviolet-visible (UV-Vis) spectroscopic. Forty (20 control/20 irradiated) female Wistar rats were used in this work. The irradiated rats were irradiated to low-dose at a total dose of 10 mGy from a fast neutron source (241Am-Be, 0.2 mGy/h). Multivariate analyses were applied to differentiate between the control and irradiated rats' Raman spectra. The erythrocytes samples were isolated from whole blood to explore the Hb structure. FTIR results revealed changes in the ν(S-H) bond of α-104 and β-93 cysteines by low-dose fast neutrons. Raman spectra showed changes in the spin state and oxidation state of the iron atom of the Hb. Besides, deformation in methine C-H was recorded. UV-Vis spectroscopy disclosed that the irradiated rats might be more susceptive to oxidation than control rats. The study deduced that the low dose fast neutron could cause tiny Hb structure changes by indirect effects. Besides, the spectroscopic techniques showed a potent ability to reveal tiny changes in the Hb structure that happened by a low dose of fast neutrons.

Evaluation of Migration Radiological Equivalence for Dual Component Nuclear Waste in a Deep Geological Repository

ABSTRACT

The paper is concerned with the issue of achieving the radiological equivalence (the equivalence of radiation risks) of radioactive waste of nuclear reactors and corresponding mass of natural uranium, taking into account the different migration ability of radionuclides in geological formations and soil. This migration radiological equivalence is being investigated for the deep burial of radioactive waste in the case of the development of a two-component nuclear power system with the concurrent use of thermal neutron reactors and fast neutron reactors. Calculations were performed of radiation doses and radiation risks of cancer death arising from consumption of drinking water from a well above a disposal site. The radiation risk relating to a two-component nuclear power system is lower than that from natural uranium; i.e., after reaching the radiological equivalence (100 y of storage) over the timescale of 109 y, the principle of migration radiological equivalence is satisfied. It would take 106 y after radioactive waste disposal to reach the migration radiological equivalence if only thermal reactors were operated. As regards consumption of well drinking water, the radiation risk does not exceed 10-5 y-1 for a two-component nuclear power system, while being 10-3 y-1 (socially unacceptable level) for a power system using only thermal reactors. Radionuclides 241Am, 239Pu, and 240Pu in drinking water make the main contribution to the doses and radiation risks of people for 104 y after the disposal of radioactive waste.

Inhalation of Soluble Plutonium: 53-year Follow-up of Manhattan Project Worker

ABSTRACT

This whole-body tissue donor to the United States Transuranium and Uranium Registries was occupationally exposed to plutonium nitrate-dioxide mixture via chronic inhalation. This individual was involved in the Manhattan Project operations and later participated in medical follow-up studies. Soft tissues and bones collected at autopsy were analyzed for 238Pu, 239+240Pu, and 241Am. Fifty-three years post-intake, 700±2 Bq of 239+240Pu were still retained in the skeleton, 661±11 Bq in the liver, and 282±3 Bq in the respiratory tract. Bioassay measurements and organ activities at the time of death were used to estimate the intake and radiation doses using the TAURUS internal dosimetry software. For this individual, an ICRP Publication 130 Human Respiratory Tract Model with case-specific particle size of 0.3 μm, ICRP Publication 100 Human Alimentary Tract Model, and ICRP Publication 141 Plutonium Systemic Model adequately described long-term plutonium retention and excretion. The total cumulative 239+240Pu intake of 31,716 Bq was estimated, of which 24,853 Bq (78.4%) were contributed by inhalation of plutonium nitrate and 6,863 Bq (21.6%) of plutonium dioxide. The committed equivalent doses to the red bone marrow, bone surface, liver, lungs, and brain were 0.71 Sv, 6.5 Sv, 8.3 Sv, 3.8 Sv, and 0.068 Sv, respectively. The committed effective dose was 1.22 Sv.

MEASUREMENT AND ANALYSIS OF THE NEUTRON ENERGY SPECTRUM AND ASSOCIATED OCCUPATIONAL EXPOSURE FROM A SHIELDED241AM-BE SOURCE

A set of five Bonner spheres was used to measure the ambient neutron H*(10) dose around an orphan 241Am-Be neutron source shielded with different arrangements and types of neutron-absorbing materials. These results were compared to measurements obtained using a portable radiation dosemeter. The Bonner sphere measurement results identified the presence of a large thermal and intermediate neutron component from the shielded 241Am-Be source that was not revealed using the portable instrument.

PHOTON FIELD OF ~100-200 KEV FOR ENVIRONMENTAL DOSEMETER CALIBRATION

As a reference photon field, several radionuclides have been used frequently, such as 241Am,137Cs and60 Co for calibration. These nuclides provide mono-energy photons for dosemeters covering few tens of keV-MeV. The main energy around 200 keV is important for both environmental and medical fields since the former should consider scattering photons and the later should measure photons from X-ray generator. In our previous work, a backscattered layout can provide a uniform photon field spectra and dose rate with an energy of 190 keV by using an affordable intensity 137 Cs gamma source. Several other quasi-monoenergetic photon fields in the range of 100-200 keV could be obtained by using several available gamma sources. Two calibrated environmental CsI(Tl) survey meters, Horiba PA-1000 and Mr. Gamma A2700, had been measured with the developed backscattered photon field to understand energy-dependent features in order to confirm dosemeter readings. Consequently, both scintillator instruments are sensitive for measurements of the relatively low dose rates at 190 keV.

EXPOSURE INHOMOGENEITY FROM 241AM AND 90SR/90Y SOURCES IN TERMS OF THE EYE LENS MONITORING IN THE NUCLEAR FACILITIES

This article highlights the issues of exposure inhomogeneity that are relative to eye lens monitoring for low-energy photons from 241Am and beta-rays from 90Sr/90Y including a personal protective equipment because eye lens exposure has been concerned more than before due to the proposed reduction of relevant dose limit. These nuclides are common and concerned sources in the nuclear industry. Our previous study presented a quantitative estimation of exposure inhomogeneity, which was applied to simple but typical exposure situations. For the present study, exposure inhomogeneity of 241Am was approximately within a factor of 1.6, implying a more homogeneous situation than expected. Regarding 90Sr/90Y exposure, estimation from both Hp(10) and Hp(0.07) on trunk would lead to an over- or underestimation by a factor of more than 10. In contrast, Hp(3) measurement on trunk will improve by up to a factor of 2. With respect to the personal protective equipment, lead apron and protective glasses are effective for the 60-keV photons for both anterior-posterior and rotational irradiations, while a full-face respirator can reduce the eye lens dose by approximately 17% for 90Sr/90Y betas. As a whole, this study demonstrated that the effect of protective equipment could be effectively incorporated into the homogeneity evaluation.

IMPROVEMENT OF GAMMA-RAY SUBTRACTION PROCEDURE FOR A CURRENT-MODE NEUTRON DETECTOR WITH A PAIR OF 6Li- AND 7Li-GLASS SCINTILLATORS

A current-mode neutron detector with a pair of 6Li- and 7Li-glass scintillators has been developed to measure high-flux neutrons in a boron neutron capture therapy field. Neutrons are basically measured by subtracting gamma-ray component using current outputs from the 7Li-glass scintillator. In the present study, the difference in the gamma-ray sensitivity between the 6Li- and 7Li-glass scintillators and the neutron sensitivity for the 7Li-glass scintillator due to the 6Li contamination were also considered to improve the gamma-ray subtraction precision. The gamma-ray subtraction procedure was experimentally investigated in thermal neutron fields with 252Cf and 241Am-Be neutron sources, which have different gamma-ray intensities per unit neutron fluence. A linear relation between neutron fluence and current output was obtained for the neutron detector in the two types of thermal neutron fields with different gamma-ray intensities. It was found that the gamma-ray subtraction procedure is useful for current-mode neutron detectors.

Development of a method for studying 241Pu/241Am activity ratio in the soil of the main Semipalatinsk test site areas

The methodology for determination of ²⁴¹Pu in soils of the main test sites of the Semipalatinsk Test Site (STS) has been developed. The results of ²⁴¹Am and ²⁴¹Pu activity determination in soil samples sampled at the main testing sites of the STS are presented. ²⁴¹Pu/²⁴¹Am activity ratio is calculated. In most cases, on local sites, ²⁴¹Pu/²⁴¹Am activity ratio is in a rather narrow range, which makes it possible to estimate ²⁴¹Pu activity based on the results of ²⁴¹Am gamma-spectrometric analysis. The accuracy of ²⁴¹Pu activity estimation using this method will be about 50%, which is quite enough in most cases.

Silicate coating to prevent leaching from radiolabeled surrogate far-field fallout in aqueous environments

Recent characterization of radioactive particles indicate that a large percentage of the radioactivity observed during the Fukushima Daiichi nuclear meltdown was insoluble ¹³⁷Cs bound within silica microparticles. Therefore, much of the decontamination research performed prior to the Fukushima incident that used either soluble radionuclides deposited onto wet surfaces or large (∼100 μm) particles characteristic of nuclear weapons fallout do not accurately represent the characteristics of potential contamination. Thus, the common practice of extrapolating radioactive decontamination methods generically to all radioactive release events is, at best, suspect. In response, a method to produce chemically-inert, radiolabeled silica particles was developed. Binding ¹⁵²Eu within a sodium silicate coating required proper temperature control and ethanol was beneficial as a volatile dispersant to limit residues. In the end, a step-wise method, which first deposited ¹⁵²Eu or ²⁴¹Am as a nitrate salt, decomposed the salt to a sesquioxide, and finally coated the surface with sodium silicate led to dispersed particles of the desired 2 or 0.5 μm diameters. Dynamic light scattering and scanning election microscopy confirmed the particle size was unchanged. Leaching studies into several common decontaminants were performed to ensure particle inertness. Our approach allows for substitution of other radionuclides making it a robust, simple, and novel method to produce inert particle surrogates for a release event that allows direct comparison of decontamination techniques and contaminant fate studies, greatly aiding the development of response and recovery plans.

Genome-wide DNA methylation changes in two Brassicaceae species sampled alongside a radiation gradient in Chernobyl and Fukushima

The long-term radiological impact to the environment of the nuclear accidents in Chernobyl and Fukushima is still under discussion. In the course of spring of 2016 we sampled two Brassicacea plants, Arabidopsis thaliana and Capsella bursa-pastoris native to Ukraine and Japan, respectively, alongside a gradient of radiation within the exclusion and difficult to return zones of Chernobyl (CEZ) and Fukushima (FEZ). Ambient dose rates were similar for both sampling gradients ranging from 0.5 to 80 μGy/h at plant height. The hypothesis was tested whether a history of several generations of plants growing in enhanced radiation exposure conditions would have led to changes in genome-wide DNA methylation. However, no differences were found in the global percentage of 5-methylated cytosines in Capsella bursa pastoris plants sampled in FEZ. On the other hand a significant decrease in whole genome methylation percentage in Arabidopsis thaliana plants was found in CEZ mainly governed by the highest exposed plants. These data support a link between exposure to changed environmental conditions and changes genome methylation. In addition to methylation the activity concentration of different radionuclides, ¹³⁷Cs, ⁹⁰Sr, ²⁴¹Am and Pu-238,239,240 for CEZ and ^{137, 134}Cs for FEZ, was analysed in both soil and plant samples. The ratio of 5.6 between ¹³⁷Cs compared to ¹³⁴Cs was as expected five years after the FEZ accident. For CEZ ¹³⁷Cs is the most abundant polluting radionuclide in soil followed by ⁹⁰Sr. Whereas ²⁴¹Am and Pu-isotopes are only marginally present. In the plant tissue, however, higher levels of Sr than Cs were retrieved due to a high uptake of ⁹⁰Sr in the plants. The ⁹⁰Sr transfer factors ranged in CEZ from 5 to 20 (kg/kg) depending on the locality. Based on the activity concentrations of the different radionuclides the ERICA tool was used to estimate the total dose rates to the plants. It was found that for FEZ the doses was mainly contributable to the external Cs-isotopes and as such estimated total dose rates (0.13-38 μGy/h) were in the same range as the ambient measured dose rates. In strong contrast this was not true for CEZ where the total dose rate was mainly due to high uptake of the ⁹⁰Sr leading to dose rates ranging from 1 to 370 μGy/h. Hence our data clearly indicate that not taking into account the internal contamination in CEZ will lead to considerable underestimation of the doses to the plants. Additionally they show that it is hard to compare the two nuclear accidental sites and one of the main reasons is the difference in contamination profile.

COMPARISON OF THE NEUTRON ENERGY RESPONSE OF TWO DIFFERENT TLD ALBEDO DOSEMETERS

Albedo dosemeters remain the most used dosemeters in neutron individual monitoring. In Brazil, most of the neutron occupational fields are from radionuclide sources, often without any moderation, where albedo dosemeters have poor energy response. The purpose of this work is to compare the HP(10) energy response of the IRD and ALNOR TLD albedo dosemeter systems, calculated by their modelling with Monte Carlo code MCNPX. Their energy responses are similar, as expected, but the IRD system is about five times more sensitive than the ALNOR one. IRD albedo system can measure the Brazilian monthly recording level of 0.2 mSv, even for bare 252Cf and 241Am-Be neutron fields. On the other hand, the ALNOR system can measure values higher than 0.2 mSv only after huge moderation of theses sources. These results show that IRD TLD albedo is more suitable than the ALNOR one to measure low doses at occupational fields from radionuclide sources.

Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS)

The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi-Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were found to have excellent agreement with the reference data. Also, the unfolded energy spectra of the neutron sources as obtained using ANFIS were more accurate than the results reported from calculations performed using artificial neural networks in previously published papers.

A new approach to the analysis of thermoluminescence glow-curve of TLD-600 dosimeters following Am-241 alpha particles irradiation

A new approach to the analysis of thermoluminescence (TL) glow-curves of LiF:Mg, Ti (TLD-600) following alpha particles irradiation has been presented using general expressions derived from the one-trap-one recombination (OTOR)-level model. It has been found that the results of the deconvolution in the case of using these expressions are different from the case of using the other TL expressions, especially for the high-temperature region. The advantage of using the OTOR expressions is that, unlike the usual analysis, a good fit and reasonable kinetic parameter values could be obtained with a lower number of deconvoluted peaks. The kinetic parameters and the dosimetric properties of each peak have been investigated over different absorbed doses and new findings have been found. Improvement of the accuracy and the precision of the dose measurements at low-dose levels has been achieved using the glow-curve deconvolution algorithm. In addition, theoretical interpretations of some observed behaviors have been presented using the interactive multiple trap system model.

STUDYING THE STRUCTURAL, OPTICAL, CHEMICAL AND ELECTROCHEMICAL ETCHING CHANGES OF CR-39 FOR DOSEMETRIC APPLICATIONS

The present work shows the induced modification of the structural, optical, chemical etching and electrochemical etching parameters of CR-39 irradiated with alpha-particles. CR-39 polymer track detectors were irradiated with different fluences (1.62 × 106, 2.72 × 106, 3.82 × 106 and 5.21 × 106 particles/cm2) of alpha-particles using 241Am source. The structural and optical properties were measured by FT-IR spectroscopy, X-ray diffraction and UV/Vis spectroscopy, respectively. The FT-IR spectra reveal that no major changes in the typical functional groups of irradiated polymer detectors. The X-ray diffraction patterns show that a broad band in the region of 12° < 2θ > 27°, which refers to the presence of the combination of amorphous and crystalline phases. UV/Vis responses of irradiated polymer track detectors exhibit a single absorption band in the range of 254-352 nm that is correlated to the occurrence of electronic transition. Also, the changes in the chemical and electrochemical parameters due to alpha-irradiation are examined and thoroughly discussed.

Plasma-Facilitated Synthesis of Amidoxime/Carbon Nanofiber Hybrids for Effective Enrichment of 238U(VI) and 241Am(III)

Plasma- and chemical-grafted amidoxime/carbon nanofiber hybrids (p-AO/CNFs and c-AO/CNFs) were utilized to remove ²³⁸U(VI) and ²⁴¹Am(III) from aqueous solutions, seawater, and groundwater. Characteristic results indicated more nitrogen-containing groups in p-AO/CNFs compared to c-AO/CNFs. The maximum adsorption capacities of p-AO/CNFs at pH 3.5 and T = 293 K (588.24 mg of ²³⁸U(VI)/g and 40.79 mg of ²⁴¹Am(III)/g from aqueous solutions, respectively) were significantly higher than those of c-AO/CNFs (263.18 and 22.77 mg/g for ²³⁸U(VI) and ²⁴¹Am(III), respectively), which indicated that plasma-grafting was a highly effective, low-cost, and environmentally friendly method. Adsorption of ²³⁸U(VI) on AO/CNFs from aqueous solutions was significantly higher than that of ²³⁸U(VI) from seawater and groundwater; moreover, AO/CNFs displayed the highest effective selectivity for ²³⁸U(VI) compared to the other radionuclides. Adsorption of ²³⁸U(VI) onto AO/CNFs created inner-sphere complexes (e.g., U-C shells) as shown by X-ray absorption fine structure analysis, which was supported by surface complexation modeling. Three inner-sphere complexes gave excellent fits to pH-edge and isothermal adsorption of ²³⁸U(VI) on the AO/CNFs. These observations are crucial for the utilization of plasma-grafted, AO-based composites in the preconcentration and immobilization of lanthanides and actinides in environmental remediation.

How Do Radionuclides Accumulate in Marine Organisms? A Case Study of Europium with Aplysina cavernicola

In the ocean, complex interactions between natural and anthropogenic radionuclides, seawater, and diverse marine biota provide a unique window through which to examine ecosystem and trophic transfer mechanisms in cases of accidental dissemination. The nature of interaction between radionuclides, the marine environment, and marine species is therefore essential for better understanding transfer mechanisms from the hydrosphere to the biosphere. Although data pertaining to the rate of global transfer are often available, little is known regarding the mechanism of environmental transport and uptake of heavy radionuclides by marine species. Among marine species, sponges are immobile active filter feeders and have been identified as hyperaccumulators of several heavy metals. We have selected the Mediterranean sponge Aplysina cavernicola as a model species for this study. Actinide elements are not the only source of radioactive release in cases of civilian nuclear events; however, their physicochemical transfer mechanisms to marine species remain largely unknown. We have targeted europium(III) as a representative of the trivalent actinides such as americium or curium. To unravel biological uptake mechanisms of europium in A. cavernicola, we have combined radiometric (γ) measurements with spectroscopic (time-resolved laser-induced fluorescence spectroscopy, TRLIFS, and X-ray absorption near-edge structure, XANES) and imaging (transmission electron microscopy, TEM, and scanning transmission X-ray microscopy, STXM) techniques. We have observed that the colloids of NaEu(CO₃)₂·nH₂O formed in seawater are taken up by A. cavernicola with no evidence that lethal dose has been reached in our working conditions. Spectroscopic results suggest that there is no change of speciation during uptake. Finally, TEM and STXM images recorded at different locations across a sponge cross section, together with differential cell separation, indicate the presence of europium particles (around 200 nm) mainly located in the skeleton and toward the outer surface of the sponge.

CREATION OF FEMALE COMPUTATIONAL PHANTOMS FOR CALIBRATION OF LUNG COUNTERS

Plutonium isotopes are of high concern because they lead to high doses. In case of contamination, the activity burden inside the lungs should be assessed accurately. Many studies showed that the presence of breasts has a substantial influence on lung counting efficiencies. Currently, the calibration of most lung counting systems is done by means of physical phantoms representing only male chests. A set of female computational phantoms has been developed in order to provide gender-specific efficiency calibrations for the (241)Am gamma emission (59.54 keV). The phantoms were created starting from a library of female chest phantoms provided by Institut de radioprotection et de sûreté nucléaire (IRSN) (Farah, J. Amélioration des mesures anthroporadiamétriques personnalisées assistées par calcul Monte Carlo: optimisation des temps de calculs et méthodologie de mesure pour l'établissement de la répartition d'activite. PhD Thesis, 2011). While the IRSN phantoms represent a supine measurement position, the SCK•CEN lung counter set-up requires the persons to be sitting in a chair. Using open-source software, the breast shapes of the original phantoms have been recreated to simulate the drooping of breasts in vertical sitting position. A Monte Carlo approach was chosen for calculating calibration coefficients for female lung counting. The results obtained with MCNPx 2.7 simulations showed a significant decrease in the detection efficiency. For bigger bust and breast sizes, the detection efficiency showed to be up to 10 times lower than the ones measured with the Livermore male torso phantom.

Slănic-Prahova low background calibration facility

The reduced background of 2 nSv h(-1) of the Slănic-Prahova Low Background Radiation Laboratory allowed installing a calibration stand for low-dose-rates dosimetry. The stand is provided with (60)Co, (137)Cs and (241)Am low activity sources. A Physikalisch-Technische Bundesanstalt-calibrated AUTOMESS 6150 AD-6 dose rate meter with a 6150 AD-b/H external probe was used to check to what extent this stand could serve as a low background calibration facility. A detailed analysis of possible uncertainties in measuring dose rates evidenced an extended uncertainty related to the certified calibration as well as instrument readings of about 3 % for a confidence level of 95 %. In these conditions, the experimentally determined dose rates for all three gamma ray sources and for source-to-probe distances varying between 0.3 and 2.0 m confirmed a good correlation between the calculated and measured dose rates. At the same time, dose rates perfectly obey to an inverse squared distances law.

Radiosensitivity of Prostate Cancer Cell Lines for Irradiation from Beta Particle-emitting Radionuclide ¹⁷⁷Lu Compared to Alpha Particles and Gamma Rays

AIM

The purpose of the present study was to investigate the radiosensitivity of the prostate cancer cell lines LNCaP, DU145, and PC3 when irradiated with beta particles emitted from (177)Lu, and to compare the effect with irradiation using alpha particles or gamma rays.

MATERIALS AND METHODS

Cells were irradiated with beta particles emitted from (177)Lu, alpha particles from (241)Am, or gamma rays from (137)Cs. A non-specific polyclonal antibody was labeled with (177)Lu and used to irradiate cells in suspension with beta particles. A previously described in-house developed alpha-particle irradiator based on a (241)Am source was used to irradiate cells with alpha particles. External gamma-ray irradiation was achieved using a standard (137)Cs irradiator. Cells were irradiated to absorbed doses equal to 0, 0.5, 1, 2, 4, 6, 8, or 10 Gy. The absorbed doses were calculated as mean absorbed doses. For evaluation of cell survival, the tetrazolium-based WST-1 assay was used. After irradiation, WST-1 was added to the cell solutions, incubated, and then measured for level of absorbance at 450 nm, indicating the live and viable cells.

RESULTS

LNCaP, DU145, and PC3 cell lines all had similar patterns of survival for the different radiation types. No significant difference in surviving fractions were observed between cells treated with beta-particle and gamma-ray irradiation, represented for example by the surviving fraction values (mean±SD) at 2, 6, and 10 Gy (SF2, SF6, and SF10) for DU145 after beta-particle irradiation: 0.700±0.090, 0.186±0.050 and 0.056±0.010, respectively. A strong radiosensitivity to alpha particles was observed, with SF2 values of 0.048±0.008, 0.018±0.006 and 0.015±0.005 for LNCaP, DU145, and PC3, respectively.

CONCLUSION

The surviving fractions after irradiation using beta particles or gamma rays did not differ significantly at the absorbed dose levels and dose rates used. Irradiation using alpha particles led to a high level of cell killing. The results show that the beta-particle emitter (177)Lu as well as alpha-particles are both good candidates for radionuclide-therapy applications in the treatment of prostate cancer.

Comparison of different PADC materials for neutron dosimetry

Investigations on track density and track size distributions of different PADC (poly allyl diglycol carbonate) materials have been performed. The PADC used for the tests has been produced by Thermo Electron (USA), Track Analysis System Limited (UK), Chiyoda Technol Corporation (Japan) and Intercast srl (Italy). For each PADC material 120 detectors were randomly selected out of 2 sheets: 60 detectors from one sheet have been irradiated with a personal dose equivalent of 3 mSv in the field of a (241)Am-Be source at the calibration laboratory of PSI, whilst the other 60 detectors from the other sheet have been used as background samples. All detectors have been processed according to an identical etching procedure and have been analysed with TASLImage scanning system. For each set of detectors the value of the average background signal, the average neutron sensitivity and the detection limit with respect to a personal dose equivalent measured with a dosemeter based on PADC have been determined. The results of the investigations allowed a comparison of the neutron sensitivity and background signal behaviours of PADC materials from different manufacturers and the assessment of the variation of neutron sensitivity and background signal over a single sheet.

Design and verification of the shielding around the new Neutron Standards Laboratory (LPN) at CIEMAT

The construction of the new Neutron Standards Laboratory at CIEMAT (Laboratorio de Patrones Neutrónicos) has been finalised and is ready to provide service. The facility is an ∼8 m×8 m×8 m irradiation vault, following the International Organization for Standardization 8529 recommendations. It relies on several neutron sources: a 5-GBq (5.8× 10(8) s(-1)) (252)Cf source and two (241)Am-Be neutron sources (185 and 11.1 GBq). The irradiation point is located 4 m over the ground level and in the geometrical centre of the room. Each neutron source can be moved remotely from its storage position inside a water pool to the irradiation point. Prior to this, an important task to design the neutron shielding and to choose the most appropriate materials has been developed by the Radiological Security Unit and the Ionizing Radiations Metrology Laboratory. MCNPX was chosen to simulate the irradiation facility. With this information the walls were built with a thickness of 125 cm. Special attention was put on the weak points (main door, air conditioning system, etc.) so that the ambient dose outside the facility was below the regulatory limits. Finally, the Radiation Protection Unit carried out a set of measurements in specific points around the installation with an LB6411 neutron monitor and a Reuter-Stokes high-pressure ion chamber to verify experimentally the results of the simulation.

The analysis of complex mixed-radiation fields using near real-time imaging

A new mixed-field imaging system has been constructed at Lancaster University using the principles of collimation and back projection to passively locate and assess sources of neutron and gamma-ray radiation. The system was set up at the University of Manchester where three radiation sources: (252)Cf, a lead-shielded (241)Am/Be and a (22)Na source were imaged. Real-time discrimination was used to find the respective components of the neutron and gamma-ray fields detected by a single EJ-301 liquid scintillator, allowing separate images of neutron and gamma-ray emitters to be formed. (252)Cf and (22)Na were successfully observed and located in the gamma-ray image; however, the (241)Am/Be was not seen owing to surrounding lead shielding. The (252)Cf and (241)Am/Be neutron sources were seen clearly in the neutron image, demonstrating the advantage of this mixed-field technique over a gamma-ray-only image where the (241)Am/Be source would have gone undetected.

An attempt to decrease anisotropic emissions of neutrons from a cylindrical 241Am-Be-encapsulation source

An attempt to decrease the anisotropic emissions of neutrons from a cylindrical (241)Am-Be-encapsulated X3 source was conducted with Monte Carlo calculations and experiments. The influence of metal materials and shapes of the external casing to the anisotropy factor were focussed on. Results obtained by calculations using MCNP4C implied that a light and spherical-shaped external casing decreases the anisotropic emission of neutrons. Experimental results using the spherical-shaped aluminium protection case also revealed that the anisotropy factor was close to 1.0 with wide zenith angle ranges.

Sensitivity and uncertainty in the measurement of H*(10) in neutron fields using an REM500 and a multi-element TEPC

The REM500 is a commercial instrument based on a tissue-equivalent proportional counter (TEPC) that has been successfully deployed as a hand-held neutron monitor, although its sensitivity is regarded by some workers as low for nuclear power plant radiation protection work. Improvements in sensitivity can be obtained using a multi-element proportional counter design in which a large number of small detecting cavities replace the single large volume cavity of conventional TEPCs. In this work, the authors quantify the improvement in uncertainty that can be obtained by comparing the ambient dose equivalent measured with a REM500, which utilises a 5.72 cm (2(1/4) inch) diameter Rossi counter, with that of a multi-element TEPC designed to have the sensitivity of a 12.7 cm (5 inch) spherical TEPC. The results obtained also provide some insight into the influence of other design features of TEPCs, such as geometry and gas filling, on the measurement of ambient dose equivalent.

Long counter and its application for the calibration of the neutron irradiators

The Korea Research Institute of Standards and Science (KRISS) has constructed a new long counter that can be disassembled in parts and reassembled. This counter can be easily transported and used as a transfer standard instrument for neutron fluence measurements. The response function and the effective centre of the counter are investigated by calculating neutron energies from thermal to 30 MeV using MCNPX. By carrying out measurements using a (252)Cf source in the KRISS irradiation room, the accuracy of the evaluated effective centre position is confirmed. The 'distance variation method' is adopted to eliminate the effect of inscatter neutrons. This method is effective and used for the experimental determination of the effective centre. The neutron emission rates determined by the neutron fluence measurements using the long counter developed are compared with those measured by a manganese sulphate bath, and show good agreement within 3 %.

Measurement of the 250Cf component in a 252Cf neutron source at KRISS

Neutron emission rate measurements have been carried out at the Korea Research Institute of Standards and Science using a manganese sulphate bath system for (252)Cf and (241)Am-Be sources since 2004. The relative measurement method was chosen in 2012, and the neutron emission rates agreed with those by the absolute measurement method within uncertainties. The neutron emission rate of an old (252)Cf source has been measured three times: in 2004, 2009 and 2012. The (250)Cf component was fitted to a double-exponential function of (252)Cf+(250)Cf, and the ratio of the (250)Cf component to the (252)Cf component was estimated to be 7.8 % in 2004 and 46.8 % in 2012. Underestimation of the neutron emission rates of old (252)Cf sources can be corrected if the neutron emission rate of the (250)Cf component is taken into account.

Neutron detection by a CR-39 detector and analysis of proton tracks etched in the same and opposite directions

A program code to simulate neutron interactions with a CR-39 detector and calculate parameters describing the induced etched proton tracks in the CR-39 material was previously developed(( 1)). This code was used to understand the mechanisms involved during interactions with neutrons in the CR-39 material and the influence of the etching process, enabling an improvement in the efficiency of the CR-39 detector. Due to neutron interaction with atoms of the detector material, the created protons are emitted in different directions and their latent tracks are oriented randomly within the detector. The aim of this paper is to show differences between the number of visible tracks etched in the same and opposite directions from both sides of the detector. The efficiency of neutron detection was analysed as a function of the removed layer and neutron energy for both sides of detector.

Correction and verification of AECL Bonner Sphere response matrix based on mono-energetic neutron calibration performed at NPL

The AECL Bonner Sphere Spectrometer (BSS) was taken to National Physical Laboratory (NPL) for calibration in mono-energetic neutron fields and bare (252)Cf neutron fields. The mono-energetic radiations were performed using ISO-8529 prescribed neutron energies: 0.071, 0.144, 0.565, 1.2, 5 and 17 MeV. A central SP9 proportional counter was also evaluated at the NPL thermal neutron calibration facility in order to assess an effective pressure of (3)He inside the counter, i.e. number density of (3)He atoms. Based on these measurements and methods outlined by Thomas and Soochak, a new BSS response matrix was generated. The response matrix is then verified by unfolding spectra corresponding to various neutron fields. Those are NPL bare (252)Cf source, National Institute of Standards and Technology bare and heavy water moderated (252)Cf source and (241)AmBe calibration source located at National Research Council. A good agreement was observed with expected neutron fluence rates, as well as derived dosimetric quantities, such as International Commission on Radiological Protection-74 ambient dose equivalent.

Thermal neutron calibration channel at LNMRI/IRD

The Brazilian Metrology Laboratory of Ionizing Radiations (LNMRI) standard thermal neutron flux facility was designed to provide uniform neutron fluence for calibration of small neutron detectors and individual dosemeters. This fluence is obtained by neutron moderation from four (241)Am-Be sources, each with 596 GBq, in a facility built with blocks of graphite/paraffin compound and high-purity carbon graphite. This study was carried out in two steps. In the first step, simulations using the MCNPX code on different geometric arrangements of moderator materials and neutron sources were performed. The quality of the resulting neutron fluence in terms of spectrum, cadmium ratio and gamma-neutron ratio was evaluated. In the second step, the system was assembled based on the results obtained on the simulations, and new measurements are being made. These measurements will validate the system, and other intercomparisons will ensure traceability to the International System of Units.

Study of neutron scattering contribution on Hp(10) and H*(10) calibration in the Brazilian National Low Scattering Laboratory

The neutron scattering at the Low Scattering Laboratory of the Brazilian National Neutron Laboratory has been studied using three different methods. The measurements have been done with a traceable standard (241)Am-Be from source-to-detector distances of 0.52-3.00 m. The obtained results with the variation distance methods are in agreement. Measurements with a large shadow cone are not worth for larger distances due to overshadowing. As the quantity required in a calibration is the response of the device being calibrated to the scattered neutron component in order to subtract this from the total response, for these purposes, the distance variation method must be used for each device. To quantify absolutely the scattering contribution on the quantity rates of fluence, Hp(10) and H*(10) in irradiation procedures, a Bonner sphere spectrometer with the shadow cone was employed. The evaluated scattering correction factor value may be employed for a distance of 1.00 m.

New PTB thermal neutron calibration facility: first results

A new thermal neutron calibration facility based on a moderator assembly has been set up at PTB. It consists of 16 (241)Am-Be radionuclide sources mounted in a graphite block, 1.5 m wide, 1.5 m high and 1.8 m deep. The sources are distributed to eight different positions, at a mean distance of ∼1.25 m from the front face of the moderator. The neutron field at the reference position, 30 cm in front of the moderator device and 75 cm above the floor, has been characterised using calculations, Bonner sphere measurements and gold foil activation. First results are shown. The field is highly thermalised: 99 % in terms of fluence. It is quite homogenous within a 20 cm×20 cm area, but the absolute value of the thermal neutron fluence rate is small and yields an ambient dose equivalent rate of 3 µSv h(-1).

Neutron metrology laboratory facility simulation

The Neutron Low Scattering Laboratory in Brazil has been completely rebuilt. Evaluation of air attenuation parameters and neutron component scattering in the room was done using Monte Carlo simulation code. Neutron fields produced by referenced neutron source were used to calculate neutron scattering and air attenuation.

Fast neutron dosemeter using pixelated detector Timepix

A Timepix detector covered with polyethylene convertors of different thicknesses is presented as a fast neutron real-time dosemeter. The application of different weighting factors in connection with the position of a signal in a Timepix detector enables one to obtain an energy-dependent signal equal to neutron dose equivalents. A simulation of a Timepix detector covered with polyethylene convertors using monoenergetic neutrons is presented. The experimental set-up of a dosemeter was also produced. The first results of detector response using different fast neutron sources are presented.

Automatic neutron dosimetry system based on fluorescent nuclear track detector technology

For the first time, the authors are describing an automatic fluorescent nuclear track detector (FNTD) reader for neutron dosimetry. FNTD is a luminescent integrating type of detector made of aluminium oxide crystals that does not require electronics or batteries during irradiation. Non-destructive optical readout of the detector is performed using a confocal laser scanning fluorescence imaging with near-diffraction limited resolution. The fully automatic table-top reader allows one to load up to 216 detectors on a tray, read their engraved IDs using a CCD camera and optical character recognition, scan and process simultaneously two types of images in fluorescent and reflected laser light contrast to eliminate false-positive tracks related to surface and volume crystal imperfections. The FNTD dosimetry system allows one to measure neutron doses from 0.1 mSv to 20 Sv and covers neutron energies from thermal to 20 MeV. The reader is characterised by a robust, compact optical design, fast data processing electronics and user-friendly software.

Feasibility study of gamma-ray medical radiography

This research explores the feasibility of using gamma-ray radiography in medical imaging. We will show that gamma-ray medical radiography has the potential to provide alternative diagnostic medical information to X-ray radiography. Approximately one Ci Am-241 radioactive source which emits mono-energetic 59.5 keV gamma rays was used. Several factors that influence the feasibility of this study were tested. They were the radiation source uniformity, image uniformity, and image quality parameters such as contrast, noise, and spatial resolution. In addition, several gamma-ray and X-ray images were acquired using humanoid phantoms. These images were recorded on computed radiography image receptors and displayed on a standard monitor. Visual assessments of these images were then conducted. The Am-241 radioactive source provided relatively uniform radiation exposure and images. Image noise and image contrast were mainly dependent on the exposure time and source size, whereas spatial resolution was dependent on source size and magnification factor. The gamma-ray humanoid phantom images were of lower quality than the X-ray images mainly due to the low radioactivity used and not enough exposure time. Nevertheless, the gamma-ray images displayed most of the main structures contained in the humanoid phantoms. Higher exposure rates and thus lower exposure times were estimated for different pure Am-241 source sizes that are hypothesized to provide high quality images similar to X-ray images. For instance, a 10mm source size of pure Am-241 with 7s exposure time should produce images similar in contrast and noise to X-ray images. This research paves the way for the production and usage of a highly radioactive Am-241 source with the potential to lead to the feasibility of acceptable quality medical gamma-ray radiography.

Characterisation of a setup for mixed beam exposures of cells to 241Am alpha particles and X-rays

Exposure of humans to mixed fields of high- and low-linear energy transfer (LET) radiation occurs in many situations-for example, in urban areas with high levels of indoor radon as well as background gamma radiation, during airplane flights or certain forms of radiation therapy. From the perspective of health risk associated with exposure to mixed fields, it is important to understand the interactions between different radiation types. In most cellular investigations on mixed beams, two types of irradiations have been applied sequentially. Simultaneous irradiation is the desirable scenario but requires a dedicated irradiation facility. The authors have constructed a facility where cells can be simultaneously exposed to (241)Am alpha particles and 190-kV X-rays at 37°C. This study presents the technical details and the dosimetry of the setup, as well as validates the performance of the setup for clonogenic survival in AA8 Chinese hamster ovary cells. No significant synergistic effect was observed. The relative biological effectiveness of the alpha particles was 2.56 for 37 % and 1.90 for 10 % clonogenic survival.

Investigation of activation rate uniformity in a prompt-gamma rays IVNAA facility

The preliminary design studies for developing a prompt-gamma rays in vivo neutron activation analysis facility are in construction at Neutron Activation Research Centre of Ferdowsi University of Mashhad. In this work, activation rate distribution is studied in different bilateral configurations in order to reduce the CV of activation rate distribution throughout the body to less than 10%. It means that the maximum RMS deviation from the mean value of distribution must be smaller than 0.5%. The best case includes two pairs of (241)Am-Be neutron sources. Each pair (with 70 cm distance between two sources) is positioned within up and down 40 x 110 cm(2) area graphite collimators. Also, four moderator/reflector objects are added to transversal body surfaces with 5 cm thick water and 3 cm thick graphite. The proper thickness of polyethylene pre-moderators is chosen 8 mm. The RMS deviation of distribution is 0.3% with the mean activation rate of 1.85 x 10(-6) cm(-3).

Improving the safety of a body composition analyser based on the PGNAA method

The 252Cf radioisotope and 241Am-Be are intense neutron emitters that are readily encapsulated in compact, portable and sealed sources. Some features such as high flux of neutron emission and reliable neutron spectrum of these sources make them suitable for the prompt gamma neutron activation analysis (PGNAA) method. The PGNAA method can be used in medicine for neutron radiography and body chemical composition analysis. 252Cf and 241Am-Be sources generate not only neutrons but also are intense gamma emitters. Furthermore, the sample in medical treatments is a human body, so it may be exposed to the bombardments of these gamma-rays. Moreover, accumulations of these high-rate gamma-rays in the detector volume cause simultaneous pulses that can be piled up and distort the spectra in the region of interest (ROI). In order to remove these disadvantages in a practical way without being concerned about losing the thermal neutron flux, a gamma-ray filter made of Pb must be employed. The paper suggests a relatively safe body chemical composition analyser (BCCA) machine that uses a spherical Pb shield, enclosing the neutron source. Gamma-ray shielding effects and the optimum radius of the spherical Pb shield have been investigated, using the MCNP-4C code, and compared with the unfiltered case, the bare source. Finally, experimental results demonstrate that an optimised gamma-ray shield for the neutron source in a BCCA can reduce effectively the risk of exposure to the 252Cf and 241Am-Be sources.

The applicability of MGA method for depleted and natural uranium isotopic analysis in the presence of actinides (232Th, 237Np, 233Pa and 241Am)

The multi-group analysis (MGA) method for the determination of uranium isotopic abundances in depleted uranium (DU) and natural uranium (NU) samples is applied in this study. A set of non-destructive gamma-ray measurements of DU and NU samples were performed using a planar Ge detector. The relative abundances of 235U and 238U isotopes were compared with the declared values of the standards. The relative abundance for 235U obtained by MGA for a "clean" DU or NU sample with a content of uranium>1wt% is determined with an accuracy of about +/-5%. However, when several actinides such as 232Th, 237Np, 233Pa and 241Am are present along with uranium isotopes simulating "dirty" DU or NU, it has been observed that MGA method gives erroneous results. The 235U abundance results for the samples were 6-25 times higher than the declared values in the presence of above-mentioned actinides, since MGA is utilized the X-ray and gamma-ray peaks in the 80-130 keV energy region, covering XKalpha and XKbeta regions. After the least-squares fitting of the spectra, it is found that the increases in the intensities of the X-ray and gamma-ray peaks of uranium are remarkably larger in the complex 80-130 keV region. On the other hand, it is observed that the interferences of the actinide peaks are relatively less dominant in the higher gamma-ray region of 130-300 keV. The results imply the need for dirty DU and NU samples that the MGA method should utilize the higher energy gamma-rays (up to 1001 keV of (234m)Pa) combined with lower energies of the spectra, which may be collected in a two detector mode (a planar Ge and a high efficient coaxial Ge).

Semiconductor quantum dot scintillation under gamma-ray irradiation

We recently demonstrated the ability of semiconductor quantum dots to convert alpha radiation into visible photons. In this letter, we report on the scintillation of quantum dots under gamma irradiation and compare the energy resolution of the 59 keV line of americium-241 obtained with our quantum dot-glass nanocomposite to that of a standard sodium iodide scintillator. A factor 2 improvement is demonstrated experimentally and interpreted theoretically using a combination of energy-loss and photon-transport models.