Gamma logging for uranium exploration are currently based on total counting with Geiger Müller gas detectors or NaI (TI) scintillators. However, the total count rate interpretation in terms of uranium concentration may be impaired in case of roll fronts, when the radioactive equilibrium of the natural 238U radioactive chain is modified by differential leaching of uranium and its daughter radioisotopes of thorium, radium, radon, etc. Indeed, in case of secular equilibrium, more than 95 % of gamma rays emitted by uranium ores come from 214Pb and 214Bi isotopes, which are in the back-end of 238U chain. Consequently, these last might produce an intense gamma signal even when uranium is not present, or with a much smaller activity, in the ore. Therefore, gamma spectroscopy measurements of core samples are performed in surface with high-resolution hyper-pure germanium HPGe detectors to directly characterize uranium activity from the 1001 keV gamma ray of 234mPa, which is in the beginning of 238U chain. However, due to the low intensity of this gamma ray, i.e. 0.84 %, acquisitions of several hours are needed. In view to characterize uranium concentration within a few minutes, we propose here a method using both the 92 keV gamma ray of 234Th and the 98.4 keV uranium X-ray. This last is due to uranium self-induced fluorescence caused by gamma radiations of 214Pb and 214Bi, which create a significant Compton scattering continuum acting as a fluorescence source and resulting in the emission of uranium fluorescence X-rays. The comparison of the uranium activity obtained with the 92 keV and 98.4 keV lines allows detecting a uranium heterogeneity in the ore. Indeed, in case of uranium nugget, the 92 keV line leads to underestimated uranium concentration due to gamma self-absorption, but on the contrary the 98.4 keV line leads to an overestimation because of increased fluorescence. In order to test this new approach, several tens of uranium ore samples have been measured with a handheld HPGe FALCON 5000 detector.
Radioactive waste has to undergo a process of quality checking in order to check its conformance with national regulations prior to its transport, intermediate storage and final disposal. Within the quality checking of radioactive waste packages non-destructive assays are required to characterize their radio-toxic and chemo-toxic contents. The Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety of the Forschungszentrum Jülich develops in the framework of cooperation nondestructive analytical techniques for the routine characterization of radioactive waste packages at industrial-scale. During the phase of research and development Monte Carlo techniques are used to simulate the transport of particle, especially photons, electrons and neutrons, through matter and to obtain the response of detection systems. The radiological characterization of low and intermediate level radioactive waste drums is performed by segmented γ-scanning (SGS). To precisely and accurately reconstruct the isotope specific activity content in waste drums by SGS measurement, an innovative method called SGSreco was developed. The Geant4 code was used to simulate the response of the collimated detection system for waste drums with different activity and matrix configurations. These simulations allow a far more detailed optimization, validation and benchmark of SGSreco, since the construction of test drums covering a broad range of activity and matrix properties is time consuming and cost intensive. The MEDINA (Multi Element Detection based on Instrumental Neutron Activation) test facility was developed to identify and quantify non-radioactive elements and substances in radioactive waste drums. MEDINA is based on prompt and delayed gamma neutron activation analysis (P&DGNAA) using a 14 MeV neutron generator. MCNP simulations were carried out to study the response of the MEDINA facility in terms of gamma spectra, time dependence of the neutron energy spectrum, neutron flux distribution. The validation of the measurements simulations with Mont-Carlo transport codes for the design, optimization and data analysis of further P&DGNAA facilities is performed in collaboration with LMN CEA Cadarache. The performance of the prompt gamma neutron activation analysis (PGNAA) for the nondestructive determination of actinides in small samples is investigated. The quantitative determination of actinides relies on the precise knowledge of partial neutron capture cross sections. Up to today these cross sections are not very accurate for analytical purpose. The goal of the TANDEM (Trans-uranium Actinides’ Nuclear Data – Evaluation and Measurement) Collaboration is the evaluation of these cross sections. Cross sections are measured using prompt gamma activation analysis facilities in Budapest and Munich. Geant4 is used to optimally design the detection system with Compton suppression. Furthermore, for the evaluation of the cross sections it is strongly needed to correct the results to the self-attenuation of the prompt gammas within the sample. In the framework of cooperation RWTH Aachen University, Forschungszentrum Jülich and the Siemens AG will study the feasibility of a compact Neutron Imaging System for Radioactive waste Analysis (NISRA). The system is based on a 14 MeV neutron source and an advanced detector system (a-Si flat panel) linked to an exclusive converter/scintillator for fast neutrons. For shielding and radioprotection studies the codes MCNPX and Geant4 were used. The two codes were benchmarked in processing time and accuracy in the neutron and gamma fluxes. Also the detector response was simulated with Geant4 to optimize components of the system.
The knowledge of the fissile material mass is a key challenge to enhance radioactive waste management and to ensure a high level of safety in nuclear industry. Data is analyzed according to the principles of the neutron measurement techniques. As proportional counters filled with 3He gas display high neutron detection efficiency and a good gamma-ray discrimination, they are the reference detector for passive neutron coincidence counting. A charge preamplifier or a current amplifier, depending on applications, collects the electric pulse produced by neutron interaction in the 3He gas and a threshold discriminator produces a logic pulse used for neutron counting. This paper describes the performance assessment of different commercially available electronics from Mirion Technologies, Precision Data Technology (PDT), Mesytec, as well as MONACO electronics originally developed by CEA LIST for fission chamber measurements in experimental reactors. Comparative passive neutron measurements are carried out with these electronics at CEA/DEN Nuclear Measurement Laboratory in Cadarache. Overall, PDT and Mesytec electronics show similar detection efficiency as the ACH-NA98 charge amplifier, which is commonly used in our laboratory for such applications. However, MONACO electronics have a lower detection efficiency, similar to Mirion 7820 current amplifier used in specific high-count rate applications. An optimisation of MONACO settings would probably be necessary to adapt to 3He counters instead of fission chambers.
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