| 1. |
- Bernitt Cartemo, Petty, 1984, et al.
(author)
-
Building a generic voxel phantom ofIRINA for Monte Carlo simulations
- 2014
-
Reports (other academic/artistic)abstract
- The human phantom IRINA, which is widely used for whole body counting calibrations, has been modelled using MATLAB. This document summa-rizes and explains the procedure that was applied for building voxel ver-sions of IRINA in standing position. All 6 sizes in standing positions were successfully modelled with the help of MATLAB and the files are ready to use for any MC simulation. The MC code GATE was used to verify the geometry of the IRINA phantom by comparing the placement of source tubes and scatterers in the Monte Carlo model to the original IRINA docu-mentation. The methodology can easily be used for building voxel phan-toms of IRINA in sitting and bending position, as well as any other geome-try that may be needed.
|
|
| 2. |
- Bernitt Cartemo, Petty, 1984, et al.
(author)
-
Comparison of computational phantoms and investigation of the effect of biodistribution on activity estimations
- 2016
-
In: Radiation Protection Dosimetry. - : Oxford University Press (OUP). - 0144-8420 .- 1742-3406. ; 171:3, s. 358-364
-
Journal article (peer-reviewed)abstract
- We have made a comparison between two computational phantoms, modeled from the unified phantom UPh-08T, for whole-body counting applications. One of these was further compared with the ICRP reference adult male computational phantom. The simulations that were performed for the comparison of all three voxel phantoms use various distributions of 60Co. The two voxel phantoms of the UPh-08T showed good agreement, despite different methods of phantom modeling. Also, effects on efficiency of the inhomogeneous distribution of a radionuclide in the computational UPh-08T phantom were studied, using the realistic biodistribution of 140La. Our results show that the activity estimation of radionuclides, which are inhomogeneously distributed in the human body, will be in error if a homogeneous distribution is assumed for the calibration of whole-body counting systems.
|
|
| 3. |
- Bernitt Cartemo, Petty, 1984, et al.
(author)
-
Depth Profiling with the Chalmers Pulsed Positron Beam
- 2014
-
In: International Journal of Nuclear Energy Science and Technology. - 1741-637X .- 1741-6361. ; 8:2, s. 105 - 116
-
Journal article (peer-reviewed)abstract
- Positron lifetime measurements are used as a tool for obtaining a deeper understanding of positron behaviour in solids and calibrating the Chalmers Pulsed Positron Beam. Experiments are executed for positron energies up to 13 keV corresponding to the maximum penetration depth in Au of roughly 400 nm and measurement results are compared to PENELOPE simulations. Reference towards Makhov modelling is made. The future goal of the study is to investigate the depth dependence of ion-induced radiation damage where positrons will be used to characterise sizes and intensities of vacancy-type lattice defects.
|
|
| 4. |
- Bernitt Cartemo, Petty, 1984, et al.
(author)
-
Sensitivity of the neutronic design of an Accelerator-Driven System (ADS) to the anisotropy of yield of the neutron generator and variation of nuclear data libraries
- 2013
-
In: Proceeding of ESARDA meeting 2013.
-
Conference paper (peer-reviewed)abstract
- The Accelerator-Driven System concept was chosen to be a basis for the Multi-purpose Hybrid Research Reactor for High-tech Applications (MYRRHA), which can operate in both sub-critical and critical mode. Therefore, the design studies in the scope of this project were varying from accelerator and material aspects to the demonstration of transmutation possibilities of the system. However, the sensitivity of neutron characteristics of the system to the anisotropy of yield of neutron generator and variation of nuclear data libraries may appear to be an important issue.In this study the corresponding sensitivity analysis was performed in order to evaluate the sensitivity of ADS neutronic design to the variation of nuclear data libraries and influence of the anisotropy of the neutron yield of the accelerator.
|
|
| 5. |
- Cartemo, Petty, et al.
(author)
-
Letter to the Editor.
- 2016
-
In: Radiation protection dosimetry. - : Oxford University Press (OUP). - 1742-3406 .- 0144-8420. ; 168:2
-
Journal article (other academic/artistic)
|
|
| 6. |
- Chernikova, Dina, 1982, et al.
(author)
-
A direct method for evaluating the concentration of boric acid in a fuel pool using scintillation detectors for joint-multiplicity measurements
- 2013
-
In: Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment. - : Elsevier BV. - 0168-9002. ; 714, s. 90-97
-
Journal article (peer-reviewed)abstract
- The present investigations are aimed at the development of a direct passive non-intrusive method for determining the concentration of boric acid in a spent fuel pool using scintillation detectors with the purpose of correcting joint-multiplicity measurement results. The method utilizes a modified relation between two gamma lines with energy of 480 keV and 2.23 MeV, respectively. The gamma line at 480 keV belongs to the thermal neutron capture in boron. The 2.23 MeV gamma line characterizes the capture of thermal neutrons in hydrogen. Thus, the relation between them can reveal the concentration of the boron in the fuel pool. In order to test this method, first MCNPX and MCNP-PoliMi simulations were performed. Then, based on the results of Monte Carlo simulations, the method was verified by an experimental study with a 241Am-Be source and EJ-309 scintillation detectors. The concentration of boron in water varied from 1550 ppm to 4000 ppm. The results of these tests are provided in the paper and they show that the spectral ratio between these two lines can in principle be used to determine the boron content.
|
|
| 7. |
|
|
| 8. |
- Chernikova, Dina, 1982, et al.
(author)
-
A potential alternative/complement to the traditional thermal neutron based counting in Nuclear Safeguards and Security
- 2016
-
In: Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment. - : Elsevier BV. - 0168-9002. ; 810, s. 164-171
-
Journal article (peer-reviewed)abstract
- A new concept for thermal neutron based correlation and multiplicity measurements is proposed in this paper. The main idea of the concept consists of using 2.223 MeV gammas (or 1.201 MeV, DE) originating in the (1) H (n,gamma)D-2-reaction instead of using traditional thermal neutron counting. Results of investigations presented in this paper indicate that gammas from thermal neutron capture reactions preserve the information about the correlation characteristics of thermal (fast) neutrons in the same time scale. Therefore, instead of thermal neutron detectors (or as a complement) one may use traditional and inexpensive gamma detectors, such as Nal, BGO, CdZnTe or any other gamma detector. In this work we used D8 x 8 cm(2) Nal scintillator to test the concept. Thus, the new approach helps to address the problem of replacement of He-3-counters and problems related to the specific measurements of spent nuclear fuel directly in the spent fuel pool. It has a particular importance for Nuclear Safeguards and Security. Overall, this work represents the proof of concept study and reports on the experimental and numerical evidence that thermal neutron capture gammas may be used in the context of correlation and multiplicity measurements. Investigations were performed using a (252)-Cf-correlated neutron source and an Am-241-Be-random neutron source. The related idea of the Gamma Differential Die-Away approach is investigated numerically in this paper as well, and will be tested experimentally in future work.
|
|
| 9. |
|
|
| 10. |
- Chernikova, Dina, 1982, et al.
(author)
-
Novel passive and active tungsten-based identifiers for maintaining the continuity of knowledge of spent nuclear fuel copper canisters
- 2015
-
In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 75, s. 219-227
-
Journal article (peer-reviewed)abstract
- A new approach to provide a long-term safeguards identification of spent nuclear fuel containers, in particular copper canisters, is presented in this paper. The approach proposes the use of a tungsten insert marked with a binary code and placed inside the container. The insert is read with a combination of two independent techniques, radiation and ultrasonic measurements, in order to get a unique identification of the cask. Passive and active versions of the tag are considered. The passive version makes use of the radiation coming from the spent nuclear fuel itself. The active version of the tag is based on the use of an artificially introduced mixture of α-emitting isotopes, such as 241Am with materials, 11B and 23Na, which easily undergo α-induced reactions with emission of specific γ-lines, 2313 keV and 1809 keV, respectively. The paper discusses results of the radiation and ultrasonic measurements and Monte-Carlo evaluations as the first proof of the concept. The results of the investigations show the strong potential for this concept to maintain the continuity of knowledge of spent nuclear fuel copper canisters for a time scale up to a few thousands years without compromising the environmental safety of the casks.
|
|
