The effect of BaTiO3 on the mechanical, thermal and radiation shielding properties of the borate glass with the composition of (60-x)B2O3 + 10MgO + 15Na2O + 15CaO+(x)BaTiO3 (x: 0,2.5,5,7.5 and ...10 mol%) were synthesized by the melt quenching method. The physical, thermal, mechanical and radiation protection properties of the glasses produced with this designed composition were carefully investigated. As the BaTiO3 amount increased from 0 % to 10 %, the density of the glass increased from 2.571 to 2.870 g/cm3 and the glass thermal stability (ΔT) also enhanced from 51 to 63. The LAC value at 276 keV, 0.2831 cm−1 for 0 % BaTiO3 and enhanced linearly with increasing BaTiO3 concentration and reach to 0.3475 cm−1 at 10 % BaTiO3. The experimentally obtained MAC values are in agreement with the theoretically calculated MAC values with XCOM and MCNP6.2. It was concluded that the increase in the BaTiO3 value of the glass positively affects the radiation shielding properties of the glasses.
Ion cyclotron resonance heating (ICRH) is an important auxiliary heating method applied to the China Fusion Engineering Test Reactor, which can effectively heat the ions and electrons in plasma. ...Owing to the harsh nuclear environment, neutronic analyses are required to verify tritium self-sufficiency and neutron-shielding requirements. In this study, a neutronics analysis of the ICRH antenna was conducted using the COre and System integrated engine for Reactor Monte Carlo (cosRMC) code to estimate the neutron flux, radiation damage, nuclear heating, gas generation rate of key components, and tritium breeding ratio (TBR), providing data support for the subsequent optimization of the shielding design. In addition, the neutron flux of the coils around the antenna was calculated to prevent the entry of neutrons that damage the magnetic field coils through the gaps between the port plugs and antenna, and the shielding effects of the port-plug antenna on the surrounding components were analyzed. Finally, the results obtained using the cosRMC and MCNP codes were compared, which and presented good agreement, thus verifying the reliability of the neutronic analysis using the cosRMC code.
MCNP radiation transport output is post-processed by DRiFT, a Detector Response Function Toolkit to simulate detailed nuclear instrumentation response. DRiFT can be used to assess the performance and ...potential limitations of scintillator, gas, and semiconductor detectors under a variety of simulated conditions not easily achievable in a laboratory setting. This work describes new updates in DRiFT for scintillator simulations which focus on the capability to simulate scintillators in current mode, an expansion of trigger options, and the ability to customize individual detector properties in a simulation. These improvements are designed to facilitate the ability to model large arrays of scintillator detectors with higher fidelity than was previously possible and are demonstrated in three examples. The first shows the difference between operating DRiFT in current and pulse mode. In the second example, which is intended to demonstrate deviations in individual detector performance, each detector has properties (PMT gain, optical transport, scintillation yield, etc.) that vary between detectors and are specified in DRiFT. A final example examines how DRiFT could be used to optimize digitizer settings in high rate measurements with split signals using the new common trigger option.
•The ability to simulate current mode experiments has been implemented in DRriFT.•New features include common triggers and the ability to assign individual scintillator, PMT, and digitizer properties.•New features are demonstrated in several examples with scintillator arrays.
A series of borate glass (80-y)B2O3-10ZnO-10CdO-yBaO, where 10 ≤ y ≤ 30, was synthesized by the melt quench method. The durability and optical features were explored to study the structural features. ...The sample S2 appears the highest durability than other samples. Several optical parameters were determined, such as bandgap, refractive index, reflection loss, metallization, and cutoff wavelength; these parameters show a good relationship with durability. Moreover, comprehensive radiation shielding properties, including gamma, neutron, and proton, were computed theoretically. The mass attenuation coefficient (MAC)theoretical was defined by using the XCOM program, the (MAC)theoretical was compared with MCNP values, and the acquired results show excellent compatibility. The sample S5 gives the best sheilding properties for gamma and proton radiations. While, S1 sample provides the highest neutron shielding features. Finally, it can be concluded that the fabricated glass samples can be used as a shielding substance for proton, neutron, and gamma-ray.
The MCNP code was utilized to assess the operational status of an N-type high-purity germanium (HPGe) detector and to evaluate its potential for a collimated and focused Prompt Gamma Neutron ...Activation Analysis (PGNAA) detection system. In order to calibrate the detector, a narrow beam γ collimator was designed to scan the detector in both axial and radial directions. The curvature of the crystal front edge and the non-uniform description of the detector dead layer (DL) were taken into account when adjusting the effective crystal size. A collimated 137Cs γ source was used to assess the detection efficiency in the scanning experiments. Monte Carlo studies were performed using 0.662 MeV γ rays, and separately, a uniform γ-ray energy distribution from 0 to 1.6 MeV. Lastly, the effects of a collimation aperture with different sizes and different incident directions were considered to discuss the relationship between the spatial resolution and suitable measurement scenarios. Therefore, a detailed calibration method for non-uniform dead layers was introduced through a comparison between experiments and Monte Carlo simulations. This work reevaluates the performance of HPGe detectors and provides valuable data support for a detailed system design and γ spectroscopy analysis in future.
A portable neutron spectrometry system was designed based on thermal neutron detectors embedded in concentric polyethylene spherical shells. The system is flexible and can accommodate the use of ...either active or passive neutron detectors in different configurations. In this work, the response matrix of the system with In-115 foil detectors was calculated with MCNP5 v.1.6. Activation foils were chosen as an ideal detector for the planned use of the system in medical accelerator environments. Calculations were performed using ENDF/B VII.0 and ENDF/B VIII.0 data libraries. The response functions calculated with the two libraries differ by as much as 11.6% in the thermal energy region for the largest moderator. A sensitivity analysis was also performed to evaluate the effect of main design parameters on the response matrix.
•A portable neutron spectrometry system (PNSS) that can accommodate neutron detectors in various configurations is designed.•The current work focused on indium activation foils due to its advantages as a passive thermal neutron detector.•Response functions calculated with ENDF/B VIII.0 and ENDF/B VII.0 deviate by as much as 11.6% for the largest moderator.•The effect of relevant design parameters on the PNSS response matrix is negligible within a certain range.•PNSS provides an alternative flexible configuration for Bonner sphere spectrometers.
A closed-water activation loop is being built at the Jožef Stefan Institute TRIGA reactor in Slovenia to serve as a well-defined and stable source of high-energy gamma rays and neutrons. The radial ...piercing port, which penetrates the graphite reflector and touching the reactor core was chosen for the installation of the closed-water loop due to the high neutron flux and favourable shielding conditions of the surrounding concrete bioshield. The main objective of this work is to assess the neutron and gamma dose field outside the port to obtain important details for the final design of the inner part of the irradiation facility and to assess the background noise for the detectors. The main part of the work consists of the design of the shielding plugs and the construction of a detailed MCNP model of the whole irradiation facility. The dose field calculations were performed with a two-step hybrid transport approach using ADVANTG for variance reduction and MCNP for particle transport. Such deep penetration and shielding calculations are challenging and computationally intensive. The results showed that the dose rate using shielding plugs is more than 7 orders of magnitude lower compared to an empty open port. To reduce the computational uncertainty, further optimisation of the ADVANTG input is essential. The final design of the shielding plugs is described. Additional lead shielding blocks will be added outside the port afterwards if needed.
A combined empirical and theoretical characterisation approach was taken in order to investigate the activity distribution and inventory in the Dounreay Fast Reactor (DFR) vault, so as to aid with ...the decommissioning and waste management strategy of the cooling circuits located within the shielded vault. The DFR vault is a no-man access, shielded void, which surrounds the reactor core and the graphite shield.
Empirical data was collected through various Non-Destructive Assay methods, utilising all available accesses into the reactor vault. This collection of information was then used to make a first assumption about the activity distribution within the vault; this was then included into a MCNP model of the vault as the source terms.
The model was run and a dose rate map of the vault obtained, and compared to the empirical dose rate data previously gathered in the accessible penetrations in the vault; the model was altered until a satisfactory agreement was obtained between theoretical and measured data.
Once this was obtained then, the model was used to provide an estimate of the activity inventory within the vault, and specifically within the various reactor components.
In this paper, the reactor core, irradiating beam tubes, and radiological shields of a 5 MW open pool type Material Testing Reactor (MTR) are simulated in detail by using the MCNPX 2.6 code and its ...default library (i.e. mostly made of ENDF/B VI and VII). A safety assessment is performed and discussed based on the related health physics issues. Independently, dosimetry parameters are measured by using the Berthold LB 6411 neutron dose rate monitor and Berthold LB 123 Gamma dose rate monitor. Experimental results are used to benchmark the modeling and calculations especially regarding dosimetry, shielding, and health physics problems. Results are fairly appropriate for further calculations to be validated but some aging problems could be raised for a 50 years old research reactor particularly due to secondary gamma rate of the activated components. Then major beam tubes are characterized for high qualified irradiating applications. Moreover, external dose rates are estimated for empty beam tubes whenever radiological shields fail such as large break of beam tubes, hazardous seismic conditions, or any accident that can remove the plugs. Then, results are very important for the safety of the reactor operator to determine and establish emergency zones (i.e. yellow zones) and planning, respectively.
•Regarding and discussing about both of the relevant radiation applications and safety issues coherently.•Complete nuclear reactor modeling by using MCNP code in detail including all of radiological shield.•Benchmarking results against experimental data.•Developing a practical map for emergency planning.
The purpose of this study is dose evaluation induced by Auger electrons emitted from 119Sb in a tumor and surrounding healthy tissues. Dose evaluation was done by using the MCNP6 code in the tumor ...and healthy tissue (in thyroid follicles) and the calculated absorbed dose was significant. Most of the Auger electron energy emitted from 119Sb deposits in the decay position, so the effectiveness of Auger electrons should not be ignored also an accurate target therapy can be planned by them.
•The absorbed dose per Auger electron emitted from 119Sb are significant.•The absorbed dose was calculated in a cylinder phantom with different diameters.•Auger electrons emitted from 119Sb deposit at distance less than 80 μm.
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