Nuclear Measurement Laboratory Research Articles

Microdosimetry

The interest in the application of microdosimetry for assessing the radiation quality of hadrontherapy fields has grown in the last two decades. The reference detector for microdosimetry is the tissue-equivalent proportional counter (TEPC). Novel types of microdosimeters have been developed by the Nuclear Measurement Laboratory of the Politecnico di Milano in collaboration with the INFN National Legnaro Laboratories. This work describes the main features of an avalanche confinement TEPC, capable of simulating sites down to 25 nm and a silicon telescope detector. The main dosimetric and microdosimetric quantities are also given and discussed in the Introduction for the sake of clarity.

New measurements of cumulative photofission yields of 239Pu, 235U and 238U with a 17.5 MeV Bremsstrahlung photon beam and progress toward actinide differentiation

In the frame of a long-term research program on the characterization of large radioactive waste packages by photofission, the Nuclear Measurement Laboratory of CEA IRESNE has measured cumulative yields of 239Pu, 235U and 238U photofission products by using a Bremsstrahlung photon beam produced by a 17.5 MeV linear electron accelerator. A characterization of the energy of the Bremsstrahlung photon beam has been carried out by photon activation analysis with different samples of gold, nickel, uranium, zinc and zirconium. The contribution of neutron fission in the different samples has also been estimated by MCNP simulations in order to assess as precisely as possible the photofission yields. Finally, 26 cumulative photofission product yields are reported for 239Pu, 28 for 238U and 26 for 235U, with half-lives ranging from 14 min to more than 3 days, some of them being not recorded so far in the literature. Among these reported photofission product yields, 18 have been measured for all 3 actinides, which can thus be used for their discrimination. A differentiation criterion based on delayed gamma-ray ratios has been established to determine the most efficient photofission product couples to estimate the enrichment of a 235U/238U mixture or the fissile fraction (235U+239Pu)/actinide mass in a mixture of uranium and plutonium.

Measurement of cumulative photofission yields of 235U and 238U with a 16 MeV Bremsstrahlung photon beam

In the frame of a long-term research program on the characterization of large radioactive waste packages by photofission, the Nuclear Measurement Laboratory of CEA IRESNE, France, has measured cumulative photofission yields of 235U and 238U short-lived and long-lived fission products by using a Bremsstrahlung photon beam produced by a 16 MeV electron linear accelerator (LINAC). To this aim, a characterization of the Bremsstrahlung photon beam has been carried out by photon activation analysis with different samples of gold, nickel, uranium and zirconium. The residual neutron flux exiting the LINAC head (lead collimator, borated polyethylene and cadmium shield) has also been characterized by neutron activation analysis with indium samples to estimate the contribution of photoneutron fissions in the uranium samples used to assess the photofission yields. Finally, 49 fission product yields are reported for 238U and 26 for 235U, with half-lives ranging from 64 s to more than 3 days, some of them not recorded so far in the literature Some photofission products cumulative yields show significant differences between 235U and 238U, which confirms the possibility of an isotopic discrimination method based on delayed gamma-ray ratios analysis for radioactive waste characterization.

Design of MICADO advanced passive and active neutron measurement system for radioactive waste drums

In the frame of the MICADO H2020 project, a passive and active neutron measurement system is being developed to estimate the nuclear material mass inside legacy waste drums of low and intermediate radioactivity levels. Monte-Carlo simulations have been performed to design a new modular and transportable neutron system, with the main objective to reach a good tradeoff between the performances in passive mode, i.e. neutron coincidence counting, and in active interrogation mode with the Differential Die-away Technique. Different designs are compared, which mainly differ in their moderation materials, graphite and polyethylene. This parametric study allowed us to define a prototype taking into account practical constraints in view of its final implementation in a wide range of in-situ locations and nuclear facilities. The total neutron detection efficiency of the prototype is 6.75%, as calculated for an empty drum, i.e. without waste matrix. The detection limit in terms of nuclear material equivalent mass have also been estimated based on assumptions for a homogeneous distribution of nuclear materials inside the drum, filled with four types of matrices covering the range of nuclear waste drums defined in the frame of the project. The most favorable matrix is made of stainless steel in passive mode and of polyethylene in active mode, with an apparent density of 0.7 g cm−3 and 0.1 g cm−3, respectively. The calculated mass detection limits are respectively 68 mg of 240Pu, 62 mg of 235U and 39 mg of 239Pu. The most penalizing matrix is made of polyethylene with an apparent density of 0.7 g cm−3, which leads to a mass detection limit of 519 mg of 240Pu in passive mode, and 564 mg of 235U or 349 mg of 239Pu in active mode. Measurement time is 30 min for both passive and active modes. Next steps will be a complete investigation of matrix effects based on intensive Monte-Carlo calculations and an experimental design to figure out the appropriate corrections. Experiments will also be conducted at CEA Cadarache Nuclear Measurement Laboratory with the construction and the assembly of the neutron system prototype, and the measurement of mock-up drums filled with different matrices.

Characterization of the X-ray spectrum of a linear accelerator

The Nuclear Measurement Laboratory (LMN) at CEA Cadarache in France uses a high-energy electron linear accelerator, LINAC (9-21 MeV), to characterize nuclear waste drums. It enables to explore new examination modalities, such as active photon interrogation or dualenergy CT to scan large concrete objects with diameters up to 140 cm. These techniques require precise awareness of the photon spectrum emitted by the LINAC. However, direct measure of this photon energy spectrum cannot be achieved because of the accelerator pulses causing detector saturation. During the last few years, a large number of indirect methods has been developed. From an experimental point of view, the simplest indirect method for spectrum estimation method is ransmission measurements. Because it can be set up easily and accurately using an ionization chamber as well as an appropriate screen. The obtained transmission curve depends on the photon energy spectrum, which can be estimated using inverse models. In this paper, we present the development of a numerical model to determine the energy spectrum from an attenuation curve via transmission measurements which combines two types of inverse models: a continue model and a discrete model. We validate this tool using a test spectrum and its transmission curve obtained via Monte-Carlo simulation. This qualification allowed us to determine its sensitivity (signal-to-noise ratio, SNR) in order to have a good convergence. We show that if the SNR is less than 4%, we have a good estimation of the photon energy spectrum. Then, it was experimentally tested with a transmission curve obtained at the laboratory.

Localization of nuclear materials in large concrete radioactive waste packages using photofission delayed gamma rays

The characterization of radioactive waste packages is mandatory for their transport, interim storage and final disposal. In this framework, the Nuclear Measurement Laboratory of CEA DES IRESNE Institute, at Cadarache, France, uses a high-energy electron linear accelerator (LINAC) to produce an interrogating bremsstrahlung beam with endpoint energies ranging from 9 to 21 MeV to perform X-ray imaging and high-energy photon interrogation on large concrete packages. In particular, highenergy photon beam induces photofission reactions in both fissile (235U, 239Pu, 241Pu) and fertile (238U, 240Pu, 232Th, etc.) actinides possibly present in the radioactive waste. In order to assess their mass, we use delayed gamma rays emitted by their photofission products, which are measured with a 50 % relative efficiency High-Purity Germanium (HPGe) detector. Actinide differentiation, which is important for the fissile mass estimation, is based on the ratios of gamma rays emitted by different photofission products and requires appropriate corrections for the gamma attenuation in concrete. To this aim, we report here a localization method of point-like nuclear materials in the concrete matrix, based on the differential attenuation of several gamma rays emitted by a same photofission product. We use here the 1435.9 and 2639.6 keV lines of 138Cs, with both experimental data and MCNP numerical simulations to determine the (r,θ) coordinates of nuclear materials. Then, the depth inside the concrete matrix, which is determined with a precision of a few percent, mainly depending on counting statistics on 1435.9 and 2639.6 keV net peak areas, is used to correct for the different gamma ratios used in the actinide identification method. Experimental tests with uranium samples have been performed to validate the localization method.

Design of a High-Energy and High-Resolution detector for X-ray computed tomography

The Nuclear Measurement Laboratory (LMN) at CEA Cadarache in France is developing a high-energy tomograph reaching energies up to 21 MeV with high dose rates. It allows performing tomographies on massive objects (5 tons, 140 cm diameter) with a multimillimetric spatial resolution. For the control of absence of cracks, bubbles or defects in the concrete coating of some CEA waste drums, the laboratory needs a “high-resolution” version of this tomograph. The purpose of this study is to design an imaging detector able to provide a high spatial resolution on large objects with high dose resistance and scanning times of a few hours. To achieve this design, several steps are necessary and will be presented: – A first step consists in characterizing the current detection elements of the tomograph, such as the camera and its lens with experimental measurements. This enables to compare performance of these elements with that available on the market and to consider a replacement. – Then, the scintillator is selected according to experimental measurements on few scintillator types and a state of the art. – Next, the detector configuration is optimized to achieve the higher spatial resolution. Throughout these steps, we present the design of a highresolution detector with a spatial resolution around 300 µm for a 10%contrast.

Performance Assessment of Amplification and Discrimination Electronic Devices for Passive Neutron Measurements

The assessment of fissile material mass is a key requirement to enhance radioactive waste management and to ensure a high level of safety in the nuclear industry. The assessment of plutonium fissile mass by passive coincidence measurements is usually obtained by detecting the neutrons generated by the spontaneous fissions of Pu isotopes. The detection and the quantification of a fissile mass in the radioactive waste can also be carried out using active neutron interrogation with a pulsed D-T neutron generator. The 14-MeV neutrons are moderated to induce fissions in the uranium and plutonium-contaminated waste. The emitted neutron signal can be analyzed by considering their detection time after the generator pulse. Data are measured according to the principles of the neutron measurement techniques. As proportional counters filled with 3He gas display high neutron detection efficiency and good gamma-ray immunity, 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 article 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 the Mirion 7820 current amplifier used in specific high-count rate applications. An optimization of MONACO settings would probably be necessary to adapt to 3He counters instead of fission chambers.

High-Resolution Gamma Spectrometry of a Plutonium Bearing Waste Drum With High-Energy Reaction-Induced Gamma Rays

In the framework of the radioactive waste drum characterization using neutron coincidence counting, the Nuclear Measurement Laboratory of French Alternative Energies and Atomic Energy Commission (CEA) Cadarache is studying plastic scintillators as an alternative to ideal but costly 3He gas proportional counters. Plastic scintillators are at least five times cheaper for the same detection efficiency, and in addition, they detect fast neutrons three orders of magnitude more quickly than 3He detectors. However, they are sensitive to gamma rays with no pulse shape discrimination abilities for large detection volumes, which implies the necessity to identify precisely gamma background sources that may affect the useful signal. This article presents a detailed analysis of the gamma-ray spectrum of a radioactive waste drum containing glove box filters contaminated by plutonium dioxide. Classical gamma emissions following alpha and beta disintegrations are identified, and also those accompanying inelastic scattering (n, $\text{n}^{\prime }$ ) or radiative capture (n, $\gamma$ ) reactions in the whole waste drum, and ( $\alpha $ , n) or ( $\alpha $ , p) reactions in the filtration media, which can lead to neutron-gamma coincidences parasitizing useful coincidences from plutonium spontaneous fissions.

Non-Destructive Examination Development for the JHR Material Testing Reactor

The Jules Horowitz Reactor (JHR) is a European material testing reactor (MTR) under construction at the CEA Cadarache centre. It will be dedicated to material and fuel irradiation tests, as well as to the production of medical isotopes. Gamma and X-Ray benches will be implemented in the reactor pool (RER), the irradiated component storage pool (EPI) and in a shielded hot cell for measuring either the whole underwater test device still containing the experimental sample or just the experimental sample before its extraction in the hot cell. The CEA/Cadarache Nuclear Measurement Laboratory (LMN) has been working in collaboration with VTT (Technical Research Centre in Finland Ltd.) since 2008 under a Finnish in-kind contribution agreement. This agreement focuses on the development of NDE systems implementing gamma-ray spectroscopy and high-energy X-ray imaging of the sample and irradiation device with the highest definition possible (resolution of 100 μm). The CEA-VTT technical specifications led to a European call for tenders launched by VTT. The contract was awarded to the Spanish company IDOM for the design, manufacturing, assembly and commissioning of: - Underwater gamma and X-ray (UGXR) mechanical benches and their associated gamma and X-ray collimation systems for the RER and EPI pools - Hot cell gamma and X-ray (HGXR) bench in the JHR NDE hot cell. The Final Design Reviews (FDR) of the UGXR and HGXR systems were completed in 2016. The design phase has been an iterative process in order to manage interfacing specifications and constraints: - Challenging experimental requirements, mainly to cover the wide diversity of sample shapes, sample activity levels and measurement processes, but also to achieve a level of mechanical accuracy to reach the ambitious geometrical resolution target in X-ray imaging, - Environmental constraints (immersion, radiation, compactness, limited accessibility for maintenance), - Nuclear safety constraints (seism, radiation protection). The whole design process has produced a number of elaborate and innovative mechatronic systems, which is rather unusual in nuclear applications since the resulting solutions have benefited from IDOM’s technological expertise in designing and commissioning large telescopes for the astronomy sector. Once the manufacturing phase and assembly finalised, the site acceptance tests for the UGXR and HGXR mechanical systems will be performed in 2019-2020 in the TOTEM facility at the CEA Cadarache center. The underwater benches will be tested in the CESARINE pool to check their requirements.

Performance assessment of amplification and discrimination electronic devices for passive neutron measurements

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.

Study of gamma-ray background noise for radioactive waste drum characterization with plastic scintillators

In the framework of the radioactive waste drum characterization using neutron coincidence counting, the Nuclear Measurement Laboratory of CEA Cadarache is studying plastic scintillators as an alternative to ideal but costly 3He gas proportional counters. Plastic scintillators are at least 5 times cheaper for the same detection efficiency, and in addition, they detect fast neutrons about three orders of magnitude faster than 3He detectors. However, they are sensitive to gamma rays, which implies the necessity to identify precisely gamma background sources that may affect the useful signal. This paper presents a detailed analysis of the gamma-ray spectrum of a radioactive waste drum containing glove box filters contaminated by plutonium dioxide. Gamma emissions accompanying inelastic scattering (n,n’) and (α,n) reactions that can lead to neutron-gamma coincidences parasitizing useful coincidences from plutonium spontaneous fissions are identified. Some of these parasitic gamma rays having energies up to several MeV, we plan to reject high-energy scintillator pulses with an electronics rejection threshold above 1 MeV, which should preserve the major part of useful fission neutron pulses.

Simulation of delayed gamma rays from neutron-induced fissions using MCNP 6.1

As part of its R&xD activities in the fields of radioactivewaste drum storage and homeland security, the NuclearMeasurement Laboratory of CEA Cadarache has started studiesrelated to the detection of induced delayed fission gamma rays asa signature of U/Pu presence either in radioactive wastes or incargo containers and luggage. The study described in the presentpaper explores the feasibility of detecting fission delayed gammarays of nuclear materials interrogated by a pulsed neutrongenerator. For this purpose, Monte Carlo simulations have beenperformed with ACT, the MNCP6 Activation Control Card.Simulated results have been compared with experimental data tovalidate the numerical model. Samples of uranium andplutonium have been irradiated for 2 hours with a pulsed D-Tneutron generator delivering 14 MeV neutrons with an averageemission of 8.107n/s, which are thermalised in a graphite cellcalled REGAIN. At the end of irradiation, activated nuclearmaterials were placed in a low-background, high-resolutiongamma spectroscopy station in order to detect delayed gammarays emitted by fission products. Anomalies have been observedin the calculated time decay curve of fission delayed gamma rayswith MCNP6 ACT card, but the time behavior is correct for non-fission activated materials like aluminum or copper. On the otherhand, the number of counts recorded in the main simulatedgamma ray lines from activated nuclear material fission productsis consistent with the experimental results, thus validating thesimulation scheme in view of further studies on thecharacterization of radioactive waste drums or special nuclearmaterial detection in cargo containers.

Simulated Performances of Very High Energy Detectors for Nondestructive Computed Tomography Characterization of Large Objects

As part of its research and development activities on high-energy X-ray imaging for nondestructive characterization, the Nuclear Measurement Laboratory has started an upgrade of its imaging system currently implemented at the CEA-Cadarache Center. The goals are to achieve a submillimeter spatial resolution and the ability to perform tomographies on very large objects (more than 100-cm standard concrete or 40-cm steel). This paper presents the results on the detection part of the imaging system. The upgrade of the detection part needs a thorough study of the performance of two detectors: a series of CdTe semiconductor sensors and two arrays of segmented CdWO4 scintillators with different pixel sizes. This paper consists of a quantum accounting diagram analysis coupled with Monte Carlo simulations. The scintillator arrays should be able to detect the millimeter details through 140 cm of concrete but are limited to 120 cm for the smaller ones. CdTe sensors have lower but more homogeneous performance, with a 0.5-mm resolution for 90 cm of concrete. The choice of the detector then depends on the preferred characteristic: the spatial resolution or the use on large volumes. The combination of the features of the source and the studies on the detectors gives the expected performance of the whole equipment, in terms of signal-over-noise ratio, spatial resolution, and acquisition time. Finally, experimental validations are carried out and validate the simulated performances.

Detailed MCNP Simulations of Gamma-Ray Spectroscopy Measurements With Calibration Blocks for Uranium Mining Applications

AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration evaluation by means of gamma-ray measurements. This paper reports gamma-ray spectra, recorded with two high-purity germanium detectors, on standard cement blocks with increasing uranium content, and the corresponding MCNP simulations. The detailed MCNP model of detectors and experimental setup has been validated by calculation versus experiment comparisons. An optimization of detector MCNP models is presented in this paper, as well as a comparison of different nuclear data libraries to explain missing or exceeding peaks in the simulation. Energy shifts observed between the fluorescence X-rays produced by MCNP and atomic data are also investigated, as well as gamma backscattering and beta radiation effects. The validated numerical models will be used in further studies to develop new gamma-ray spectroscopy approaches aiming at reducing acquisition times, especially for ore samples with low uranium content.

Fast High-Energy X-Ray Imaging for Severe Accidents Experiments on the Future PLINIUS-2 Platform

The future PLINIUS-2 platform of the Commissariat a l’Energie Atomique (CEA) of Cadarache will be dedicated to the study of corium interactions in severe nuclear accidents, and will host innovative large-scale experiments. The nuclear measurement laboratory of CEA Cadarache is in charge of real-time high-energy X-ray imaging setups, for the study of the corium–water and corium–sodium interaction, and of the corium stratification process. Imaging such large and high-density objects requires a 15 MeV linear electron accelerator coupled to a tungsten target creating a high-energy bremsstrahlung X-ray flux, with corresponding dose rate about 100 Gy/min at 1 m. The signal is detected by phosphor screens coupled to high-framerate scientific CMOS cameras. The imaging setup is established using an experimentally validated home-made simulation software High-Energy MODeling for RAdiography and TOmography (MODHERATO). The code computes quantitative radiographic signals from the description of the source, object geometry and composition, detector, and geometrical configuration (magnification factor, etc.). It accounts for several noise sources (photonic and electronic noises, and swank and readout noise), and for image blur due to the source spot-size and to the detector unsharpness. In a view to PLINIUS-2, the simulation has been improved to account for the scattered flux, which is expected to be significant. This paper presents the scattered flux calculation using the Monte-Carlo N-particle transport code, and its integration into the MODHERATO simulation. Then, the validation of the improved simulation is presented, through confrontation to real measurement images taken on a small-scale equivalent setup on the PLINIUS platform. Excellent agreement is achieved. This improved simulation is therefore being used to design the PLINIUS-2 imaging setups (source, detectors, and cameras). Simulation will be presented for different configurations, as well as preliminary tests of alternative detectors considered for PLINIUS-2.

Improving Gross Count Gamma-Ray Logging in Uranium Mining With the NGRS Probe

AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration measurement by means of gamma-ray logging. The determination of uranium concentration in boreholes is performed with the Natural Gamma Ray Sonde (NGRS) based on a NaI(Tl) scintillation detector. The total gamma count rate is converted into uranium concentration using a calibration coefficient measured in concrete blocks with known uranium concentration in the AREVA Mines calibration facility located in Bessines, France. Until now, to take into account gamma attenuation in a variety of boreholes diameters, tubing materials, diameters and thicknesses, filling fluid densities, and compositions, a semiempirical formula was used to correct the calibration coefficient measured in Bessines facility. In this paper, we propose to use Monte Carlo simulations to improve gamma attenuation corrections. To this purpose, the NGRS probe and the calibration measurements in the standard concrete blocks have been modeled with Monte Carlo N-Particles (MCNP) computer code. The calibration coefficient determined by simulation 5.3 $\text{s}^{-1}\cdot \text {ppm}_{U}^{-1}$ with 10% accuracy is in good agreement with the one measured in Bessines (and for which no uncertainty was provided), 5.2 $\text{s}^{-1}\cdot \text {ppm}_{U}^{-1}$ . The calculations indicate that the concrete blocks used for measuring the calibration coefficients measured in Bessines are underestimated by about 10%. Based on the validated MCNP model, several parametric studies have been performed. For instance, the rock density and chemical composition proved to have a limited impact on the calibration coefficient. However, gamma self-absorption in uranium leads to a nonlinear relationship between count rate and uranium concentration beyond approximately 1% of uranium weight fraction, the underestimation of the uranium content reaching more than a factor 2.5 for a 50% uranium weight fraction. Parametric studies have also been performed with different tubing materials, diameters, and thicknesses, as well as different borehole filling fluids representative of real measurement conditions, in view to validate gamma attenuation corrections based on the semiempirical formula. In addition, a multilinear analysis approach has been tested to further improve accuracy on uranium concentration determination, leading to only a few percent uncertainties on a large range of configurations.

Simulated Performances of a Very High Energy Tomograph for Non-Destructive Characterization of large objects

As part of its R&D activities on high-energy X-ray imaging for non-destructive characterization, the Nuclear Measurement Laboratory has started an upgrade of its imaging system currently implemented at the CEA-Cadarache center. The goals are to achieve a sub-millimeter spatial resolution and the ability to perform tomographies on very large objects (more than 100-cm standard concrete or 40-cm steel). This paper presentsresults on the detection part of the imaging system. The upgrade of the detection part needs a thorough study of the performance of two detectors: a series of CdTe semiconductor sensors and two arrays of segmented CdWO4 scintillators with different pixel sizes. This study consists in a Quantum Accounting Diagram (QAD) analysis coupled with Monte-Carlo simulations. The scintillator arrays are able to detect millimeter details through 140 cm of concrete, but are limited to 120 cm for smaller ones. CdTe sensors have lower but more stable performance, with a 0.5 mm resolution for 90 cm of concrete. The choice of the detector then depends on the preferred characteristic: the spatial resolution or the use on large volumes. The combination of the features of the source and the studies on the detectors gives the expected performance of the whole equipment, in terms of signal-over-noise ratio (SNR), spatial resolution and acquisition time.

Improving gross count gamma-ray logging in uranium mining with the NGRS probe

AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration measurement by means of gamma ray logging. The determination of uranium concentration in boreholes is performed with the Natural Gamma Ray Sonde (NGRS) based on a NaI(Tl) scintillation detector. The total gamma count rate is converted into uranium concentration using a calibration coefficient measured in concrete blocks with known uranium concentration in the AREVA Mines calibration facility located in Bessines, France. Until now, to take into account gamma attenuation in a variety of boreholes diameters, tubing materials, diameters and thicknesses, filling fluid densities and compositions, a semi-empirical formula was used to correct the calibration coefficient measured in Bessines facility. In this work, we propose to use Monte Carlo simulations to improve gamma attenuation corrections. To this purpose, the NGRS probe and the calibration measurements in the standard concrete blocks have been modeled with MCNP computer code. The calibration coefficient determined by simulation, 5.3 s-1.ppmU-1 ± 10%, is in good agreement with the one measured in Bessines, 5.2 s-1.ppmU-1. Based on the validated MCNP model, several parametric studies have been performed. For instance, the rock density and chemical composition proved to have a limited impact on the calibration coefficient. However, gamma self-absorption in uranium leads to a nonlinear relationship between count rate and uranium concentration beyond approximately 1% of uranium weight fraction, the underestimation of the uranium content reaching more than a factor 2.5 for a 50 % uranium weight fraction. Next steps will concern parametric studies with different tubing materials, diameters and thicknesses, as well as different borehole filling fluids representative of real measurement conditions.

Fast high-energy X-ray imaging for Severe Accidents experiments on the future PLINIUS-2 platform

The future PLINIUS-2 platform of CEA Cadarache will be dedicated to the study of corium interactions in severe nuclear accidents, and will host innovative large-scale experiments. The Nuclear Measurement Laboratory of CEA Cadarache is in charge of real-time high-energy X-ray imaging set-ups, for the study of the corium-water and corium-sodium interaction, and of the corium stratification process. Imaging such large and high-density objects requires a 15 MeV linear electron accelerator coupled to a tungsten target creating a high-energy Bremsstrahlung X-ray flux, with corresponding dose rate about 100 Gy/min at 1 m. The signal is detected by phosphor screens coupled to high-framerate scientific CMOS cameras. The imaging set-up is established using an experimentally-validated home-made simulation software (MODHERATO). The code computes quantitative radiographic signals from the description of the source, object geometry and composition, detector, and geometrical configuration (magnification factor, etc.). It accounts for several noise sources (photonic and electronic noises, swank and readout noise), and for image blur due to the source spot-size and to the detector unsharpness. In a view to PLINIUS-2, the simulation has been improved to account for the scattered flux, which is expected to be significant. The paper presents the scattered flux calculation using the MCNP transport code, and its integration into the MODHERATO simulation. Then the validation of the improved simulation is presented, through confrontation to real measurement images taken on a small-scale equivalent set-up on the PLINIUS platform. Excellent agreement is achieved. This improved simulation is therefore being used to design the PLINIUS-2 imaging set-ups (source, detectors, cameras, etc.).