Gamma Radiation Field Research Articles

Airborne gamma radiation field simulation method based on finite element analysis

Airborne gamma radiation field simulation method based on finite element analysis

NPTS-PK: A new point kernel code for fast calculation of 3D gamma radiation field

The point kernel integration method is commonly utilized for the analytical calculation of gamma radiation fields in the field of radiation protection and shielding design. This study introduces NPTS-PK, a program developed for rapid calculation of 3D gamma radiation fields based on the NPTS program and the point kernel integration method. Based on the geometric construction and input method of the NPTS program, NPTS-PK supports point kernel calculations for radiation sources and shielding structures of various complex shapes and materials. By calculating material attenuation coefficients using continuous energy cross-section parameters, combined with a more precise buildup factor calculation method, the accuracy of calculation results has been enhanced. Improvements in ray tracing processes and the implementation of the probability neighbor list method have accelerated geometric processing. To address the challenge of excessive computation time associated with large-scale grid counting in point kernel programs, NPTS-PK integrates several efficient acceleration techniques, elevating the speed of radiation field calculation by an order of magnitude. Tests on typical model and engineering scenario demonstrate that the deviation between NPTS-PK results and the reference values is within a few tens of percent, and the computational efficiency and accuracy are improved compared to the standard point kernel programs.

Study of the use of sodium thiocyanate as an indicator for three-dimensional dosimetry

Three-dimensional dosimetry allows the determination of dosimetric quantities with a volumetric gradient. With Fricke Gel it is possible to make measurements of this nature due to the restriction of ions in this dosimeter caused by the presence of a gelatinous matrix. When ion indicators are added to Fricke Gel, the diffusion of the ions is further hampered and, in addition, the use of optical measurement techniques for the determination of the absorbed dose becomes possible. In this study, a new ferric ion indicator possibility for Fricke Gel is evaluated. Sodium thiocyanate was added to the Fricke Gel matrix and the effect of this addition was analyzed in a gamma radiation field and in a high-energy electron beam. The obtained results demonstrated the possibility of using this indicator as a binding agent in the Fricke Gel dosimeter. In the gamma irradiation field, the dosimeter showed a linear response with the absorbed dose between 5 Gy and 150 Gy, in addition to presenting visual evidence of the spatial distribution of the radiation field to which it is subjected.

Measurement of tritium production in the helium cooled pebble bed test blanket module mock-up at JET during DTE2

Quite often, detectors for measuring nuclear performance and radiation quantities of relevance in fusion experiments are requested to withstand harsh working conditions due to intense neutron and gamma radiation fields. High temperature constitutes a further harsh element in some locations of the machine, where it is necessary to perform some on-line measurements, as expected in the breeding blanket. This is an essential component in future fusion power plants to provide tritium self-sufficiency and its performance must be continuously monitored. Some Test Blanket Modules (TBMs) will be installed in ITER to provide the first experimental data to validate the predictions on tritium production and recovery. In the meantime, within EUROfusion program, the mock-up of the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM), previously used for the TBM experiment at the Frascati Neutron Generator (FNG), had been installed at JET to test some detectors and for benchmarking numerical codes used for breeding blanket assessment during DTE2 campaign. A diamond detector, calibrated to measure the tritium production through neutron detection inside the HCPB TBM mock-up, was tested during some plasma pulses of the DTE2 campaign at JET. The main outcome is that, as far as neutron emission rate is below 1015 s−1, neutrons are properly detected along the plasma discharge evolution by TBM diamond detector, consistently with the JET neutron monitor KN1. Moreover, the amount of tritium measured (E) is 1.40 × 10–12 tritons per source neutron and the comparison with MCNP radiation transport simulation (C) gives a ratio C/E = 0.77. Such measurements, considered promising, and their comparison with calculations are discussed in the present work. Criticalities emerged are analyzed and some improvements proposed with the main purpose of speeding up signal processing to make the system capable of working at higher plasma neutron emission rates.

Model of occupational exposure of workers performing inspection of welded joints

A model of occupational exposure to gamma flaw detector operators working with portable flaw detectors in the field has been developed. The initial data for the development and verification of the model were the results of measurements of the characteristics of the gamma radiation field at the workplaces of flaw detectors and data from individual dosimetric monitoring. The relationships between the measured (H*(10), Hp(10)) and protection (effective dose) quantities (conversion coefficients) were determined using calculations and phantom experiments simulating three main operations of the full production cycle: transportation of the flaw detector to the place of X-raying of the product, installation of a flaw detector to perform x-raying and X-raying of the product. As a result of the study, it was found that more than 90% of the dose contribution to the readings of an individual dosimeter is due to the installation of the flaw detector in the working position and X-raying of the product. The values of the conversion coefficients for these operations in the form of the ratio of the effective dose values and the readings of dosimeters (Hp(10)) located on the worker's body at chest level (standard place) and abdominal level differ little for both positions of individual dosimeters. The use of maximum conversion coefficient value of 0.8 Sv/Sv corresponding to the operation of X-raying of the product will ensure conservatism in the assessment of the effective dose for the entire production cycle by no more than 15% and 25% for dosimeters located at the chest level and abdominal level, respectively.

Field Test of the MiniRadMeter Gamma and Neutron Detector for the EU Project CLEANDEM.

In the framework of the H2020 CLEANDEM project, a small robotic vehicle was equipped with a series of different sensors that were developed for the preliminary inspection of areas possibly contaminated by radiation. Such unmanned inspection allows to identify dangerous locations prior to the possible start of human operations. One of the developed devices, named the MiniRadMeter, is a compact low-cost sensor that performs gamma and neutron radiation field mapping in the environment. The MiniRadMeter was successfully tested in a simulated field mission with four "hidden" radioactive sources and a neutron generator. In this work, we describe the test procedure and the results, which were supported by the outcome of dedicated Monte Carlo simulations.

Dosimetric studies of e-, γ-, n- Irradiation fields of the M-30 microtron (18.5 mev)

The results of the implementation and study of the radiation characteristics of high-energy electron (e-), gamma (γ-), and neutron (n-) radiation fields of the M-30 microtron with a maximum energy of 18.5 MeV are presented. Dosimetric characteristics were studied using the method of thermally stimulated luminescence (TSL) and using samples TLD-100H (LiF: Mg,Cu,P) and TLD-500K (Al2O3:C). Three e-, γ-, and n-experiments were performed at the same M-30 electron flux; the dosimetric indicators Yγ,n;e = Dγ,n/De, where Dγ,n is the intensity of the dosimetric peaks of TSL after γ- or n-irradiation, and De is, respectively, e-irradiation. We discuss the relative stability of these dosimeters' Yγ,e parameters at gamma irradiations and the presence of a partial correlation of Yn,e values in neutron fields. The latter is associated with the peculiarity of radiation defect formation in the material of dosimeters when interacting with nuclear particles of different types. The research results confirm the general ideas about the physics of electron radiation conversion into beams of nuclear particles of different natures. The data obtained are important in developing methods for dosimetric support of radiation technologies that use beams of mixed high-energy nuclear radiation, particularly for radiation therapy.

A comparative study of different radial basis function interpolation algorithms in the reconstruction and path planning of γ radiation fields

Accurate reconstruction of radiation field and path planning are very important for the safety of operators in the process of dismantling nuclear facilities. Based on radial basis function (RBF) interpolation algorithm, this paper discussed the application of inverse multiquadric radial basis Function (IMRBF) interpolation method to the reconstruction of gamma radiation field, and proved the feasibility of reconstructing a radiation field with multiple γ sources. The average relative errors of IMRBF interpolation results were 4.28% and 8.76%, respectively, for the experimental scenarios with single and double gamma sources. After comparing the consistency between the simulated scene and the experimental scene, IMRBF method and Cubic Spline method were respectively used to reconstruct the gamma radiation field by Geant4 simulation data. The results showed that the interpolation accuracy of IMRBF method was superior to that of Cubic Spline method. Further, more RBF interpolation algorithms were used to reconstruct the multi-γ source radiation field, and then the Probabilistic Roadmap (PRM) algorithm was used to optimize the human walking path in the radiation field reconstructed by different interpolation methods. The optimal paths in radiation fields generated by multiple interpolation methods were compared. The results herein contribute to a comprehensive understanding of RBF interpolation methods in reconstructing γ radiation fields and their application in optimizing paths in radiation environments. The insights may provide valuable information for decision-making in radiation protection during the decommissioning of nuclear facilities.

DEMO toroidal field coil fast discharge unit integration studies

The Fast Discharge Units (FDUs) of the Superconducting (SC) Toroidal Field (TF) coils in the European demonstration fusion power plant DEMO warrant the machine integrity over its full lifetime against severe failure events, such as SC coil quenches or any other plant events requiring the safe TF magnet system discharge. A low (75 kA) and a high current (105 kA) configuration are under study for the TF coils for DEMO. The FDUs must be extremely reliable for the purpose of commutating in short time (∼1 s) the currents and to discharge the TF magnets safely into resistors outside of the tokamak building. Malfunctions of the FDUs must be avoided. The FDUs are considered as Safety Important Class (SIC) components that need to discharge high amounts of energy of about 161 GJ (@75 kA) resp. 118 GJ (@105 kA) stored in the DEMO TF coils.The TF FDUs Circuit Breakers (CBs) shall be installed in the lower level of the tokamak to minimize the length of the connecting busbars. The FDUs integration is challenging because of the high neutron and gamma radiation and stray magnetic fields of the tokamak.Since in DEMO the neutron fluence over lifetime is much higher than in ITER, the problems of using FDUs with electronic subsystems was expected to be more severe, so that their integration has been considered from the beginning of the DEMO project. Sufficient shielding or possible re-positioning of the whole FDUs or sensitive FDU components compared to ITER are being investigated, to reduce the neutron fluxes and neutron and gamma ray fluences. Alternative concepts, e.g., fully mechanical CBs are studied in the EUROfusion Work Package Plant Electrical System (WPPES) in parallel.This paper presents the CAD integration work on the DEMO TF FDUs supported by neutronics assessments. It is assumed the same FDU technology as in ITER. The magnet feeder´s integration is commenced at the same time.

The influence of neutron and gamma radiation on the reliability of magnetic and semiconductor memories

The miniaturization of computer facilities conditioned by the miniaturization of applied components makes them very sensitive to radioactive radiation. This is where neutron and electromagnetic radiation come to the fore. The reason for the particularly pronounced effects of this radiation is the fact that they do not interact with the Coulomb force, so they pass (leave) the reactor vessel unimpeded. This study examines the reliability of magnetic and semiconductor computer memories in the field of neutron and gamma radiation. This study experimental, and conducted under well-controlled conditions. The combined measurement uncertainty of the experimental procedure is less than 5 %. Sophisticated methods of mathematical statistics were used to process the stochastic results of measurements.

Estimation of Reference Gamma Radiation Field for Calibration of Radiation Monitors

Estimation of Reference Gamma Radiation Field for Calibration of Radiation Monitors

Measurement of fast neutron spectrum in the radial channel of VR-1 research reactor

Having knowledge of the spectrum of fast neutrons in the irradiation position of a reactor plays an important role in many experimental settings. If well described, it can be used as a reference neutron field for validating and testing measuring devices or detectors. This paper compares model calculations with measurements of the neutron spectrum in the research reactor VR-1. The fast neutron spectrum was measured in the radial channel of VR-1 reactor using a stilbene scintillation detector and the NGA-01 measuring device. The NGA-01 is a two-parameter spectrometric system for neutron and gamma radiation fields. The calculations were performed in a critical mode of the MCNP6 code using ENDF/B-VII.1 nuclear data library. Generally, a good agreement was achieved. However, a slight discrepancy in the neutron spectrum between the calculated and measured values was found in the 2–3 MeV region. In other regions, the differences between calculated and experimental values are comparable with uncertainties. Since the detector is distanced from the core, the neutron beam can be assumed parallel and therefore possible issues with stilbene anisotropy are eliminated. Thus, this experiment can be used for validation of a reactor leakage spectrum and nuclear cross sections in the energy region above 6 MeV, which is of interest for the nuclear data section of International Atomic Energy Agency.

Optimized inverse distance weighted interpolation algorithm for γ radiation field reconstruction

The inversion of radiation field distribution is of great significance in the decommissioning sites of nuclear facilities. However, the radiation fields often contain multiple mixtures of radionuclides, making the inversion extremely difficult and posing a huge challenge. Many radiation field reconstruction methods, such as Kriging algorithm and neural network, can not solve this problem perfectly. To address this issue, this paper proposes an optimized inverse distance weighted (IDW) interpolation algorithm for reconstructing the gamma radiation field. The algorithm corrects the difference between the experimental and simulated scenarios, and the data is preprocessed with normalization to improve accuracy. The experiment involves setting up gamma radiation fields of three Co-60 radioactive sources and verifying them by using the optimized IDW algorithm. The results show that the mean absolute percentage error (MAPE) of the reconstruction result obtained by using the optimized IDW algorithm is 16.0%, which is significantly better than the results obtained by using the Kriging method. Importantly, the optimized IDW algorithm is suitable for radiation scenarios with multiple radioactive sources, providing an effective method for obtaining radiation field distribution in nuclear facility decommissioning engineering.

Comparison of cell casted and 3D-printed plastic scintillators for dosimetry applications.

Currently, the most used methods of plastic scintillator (PS) manufacturing are cell casting and bulk polymerisation, extrusion, injection molding, whereas digital light processing (DLP) 3D printing technique has been recently introduced. For our research, we measured blue-emitting EJ-200, EJ-208, green-emitting EJ-260, EJ-262 cell cast and two types of blue-emitting DLP-printed PSs. The light output of the samples, with the same dimension of 10mm × 10mm × 10mm, was compared. The light output of the samples, relative to the reference EJ-200 cell-cast scintillator, equals about 40-49 and 70-73% for two types of 3D-printed, and two green-emitting cell-casted PSs, respectively. Performance of the investigated scintillators is sufficient to use them in a plastic scintillation dosemeter operating in high fluence gamma radiation fields.

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

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

Air kerma rate from radionuclides distributed in forest ecosystem

This study evaluates the air kerma rate in radioactively contaminated forests. The air kerma rates created by plane sources of monoenergetic photons in the energy range 0.02–3 MeV located at different depths in soil up to 50 g cm−2 and at different heights in the forest medium from 0.05 to 50 m were calculated using numeric solution of the transport (Boltzmann) equation. To simplify the practical use of the results obtained by solving the Boltzmann equation, the study additionally includes approximation formulae for calculating air kerma rate separately from contaminated soil, crowns and trunks of trees in a forest ecosystem for 20 radionuclides - fission products that significantly contribute to the external dose. Biomaterial of trunks and crowns was modeled as uniformly distributed in corresponding layers and homogeneously mixed with air. Different distributions of radionuclides in soil were considered including plane source located at different depths, exponential and uniform distribution. Based on the results, the effect of forest biomass presence as an absorbing and scattering medium on the air kerma rate at 1 m above soil was evaluated. The estimated relative difference in air kerma rate at 1 m above soil in the forest medium and in free air for monoenergetic photon sources with energies 0.1 MeV, 0.66 MeV and 3 MeV did not correlate significantly with the energy of photons. Its maximum value in forest medium with biomass density of 5 kg m−3 was 15–20% for the source at soil depth ∼0.3 g cm−2, decreasing to less than 5% when it is at soil depth greater than 7 g cm−2. An example calculation of the air kerma rate dynamics is presented for the initial period after radioactive fallout considering weathering processes (rainfall and wind action) that contribute to the transfer of activity from the canopies to the forest floor. The differences in air kerma rate values, as an integral characteristic of the gamma radiation field from a radioactive cloud in the forest and in the open area, were evaluated.

Neutron spectrometry and dosimetry using a multi-shell Single Sphere Neutron Spectrometer with thermo-luminescent and optically stimulated luminescent detectors

A portable neutron spectrometer with concentric multi-shell moderating assembly and TL, OSL detectors has been validated with standard neutron reference fields to generate the neutron spectra and estimate the ambient dose equivalents, H*(10). Further, the system has been used in the mixed neutron–gamma radiation fields at accelerator facility. The neutron fluence and the ambient dose equivalent estimates were verified using the standard reference neutron fields, viz. 241Am-Be, 252Cf, D-T neutron sources. The system was then employed in the accelerator radiation environment to measure the neutron spectral yield from the interactions of 15 and 20 MeV protons with a thick natural Ta target. The results were compared with those obtained from FLUKA simulations. A conventional neutron dose equivalent meter was also used to verify the ambient dose equivalent estimated from the spectrometric measurements. The reproducibility of the neutron spectra and dose values were found to be satisfactory, thereby encouraging the implementation of the portable single sphere neutron spectrometer in complex radiation environment around particle accelerators. Present studies demonstrate the measurements with the in-house built spectrometer for neutron energies up to 20 MeV.

Formation of Waveguide Layers on the Surface of K8 Glass Produced by Thermoradiation Ion Exchange

The low-temperature heat treatment of K8 glass is carried out at temperatures of 350–500°C in alkaline melts of NaNO3 and CsNO3 salts in the field of gamma radiation from the 60Co source at the dose rate of 3000 R/s and also outside the field. Under the influence of thermoradiation treatment due to the @ ↔ @ ion-exchange diffusion, mechanical compressive stresses are created in the surface layer of the glass, which lead to the formation of a waveguide layer of the given thickness and an increase in the refractive index (RI) increment, the number of waveguide modes, and the depth of the waveguide layer.

The Gamma and Neutron Sensor System for Rapid Dose Rate Mapping in the CLEANDEM Project.

The decommissioning of nuclear installations, as well as the possible necessary accident remediations, requires the physical presence of human operators in potentially radiologically hostile environments. The number of active nuclear reactors worldwide is greater than 400, and most of them are 40 to 50 years old, thus implying that soon they will have to be dismantled. In the framework of the H2020 CLEANDEM project, a small robotic vehicle is being developed that is equipped with a series of different sensors for areas that are significantly contaminated by radiation. In this work, we describe the MiniRadMeter system, a compact low-cost sensor capable of being used to perform quick gamma and neutron radiation field mapping of environments prior to the possible start of human operations. The miniature gamma sensor is a 1 cm3 scintillator counter with moderate spectroscopic features read out by means of a 6 × 6 mm2 SiPM, whereas neutrons are detected by means of a silicon diode coupled to a layer of 6LiF and placed inside a 6 × 6 × 6 cm3 polyethylene box. The front-end and data acquisition electronics were developed based on a Raspberry Pi4 microcomputer. In this paper, the system performance and the preliminary test results are described.

Validation of the neural network for 3D photon radiation field reconstruction under various source distributions

Introduction: This paper proposes a five-layer fully connected neural network for predicting radiation parameters in a radiation space based on detector readings.Methods: The network is trained and tested using gamma flux values from individual detector positions as input, and is used to predict the gamma radiation field in 3D space under different source term distributions. The method is evaluated using the mean percentage change error (PCT) for the test set under different source term distributions.Results: The results show that the neural network method can accurately predict radiation parameters with an average PCT error range of 0.53% to 3.11%, within the given measurement input error range of ± 10%. The method also demonstrates its ability to directly reconstruct the 3D radiation field with some simple source terms.Discussion: The proposed method has practical value in real operations within radiation spaces, and can be used to improve the accuracy and efficiency of predicting radiation parameters. Further research could explore the use of more complex source term distributions and the integration of other types of sensors for improved accuracy.