This session is focused on research developments that are compatible with the characterisation of both operating and non-operating reactors, particularly with regard to novel developments in terms of sensors, systems and new analytical methods.
A Compact Calorimetric Sensor Suited to High Nuclear Dose Rate in MTRs: Experimental and Numerical Studies (#1071)
A. Volte1, C. Reynard-Carette1, A. Lyoussi2, J. Brun1, M. Carette1
1 Aix Marseille Univ, IM2NP, Marseille, France
Aix-Marseille University and the CEA into the framework of a joint laboratory called LIMMEX conduct research works on instrumentation and measurement methods for online quantification of nuclear and thermal parameters in Material Testing Reactors (MTRs). In particular, studies focus on the design and characterization of differential calorimeters used for the measurement of the energy deposition rate induced by interactions between nuclear rays and matter (nuclear dose rate). A complete approach coupling experiments (in the laboratory and under real conditions) with thermal numerical works (1D calculations and 3D simulations) is developed to propose various sensors targeting specific characteristics (sensitivity, range...).
At present, one major challenge is to quantify high nuclear dose rates (up to 20kGy.s-1) by a sensor with at least the same performances of the standard differential calorimeters used for lower dose rates because no sensor allowing the measurements of such rates exists. Consequently, a new configuration of a compact calorimeter called CALORRE has been developed recently to reach this purpose and will be presented in this paper. Firstly, the experimental characterizations of this original sensor under laboratory conditions without radiation will be shown for an energy deposition range never tested for other differential calorimeters by using an adapted heating system. The influence of the energy deposition for various fluid flow temperatures on its linear calibration curve will be explained by means of calculations to quantify heat transfer contributions (analytical calculations). The advantages of a linear response will be discussed. Secondly, the sensor behavior will be determined for real irradiation conditions (20W.g-1) with 3D numerical calculations (simulations validated thanks to comparisons between experimental and numerical results obtained for a previous non-linear CALORRE configuration tested in MARIA reactor (<1W.g-1)).
Improvement of a Non-destructive Nuclide Assay using a Self-indication Method (#1135)
J. - I. Hori1, T. Sano1, Y. Takahashi1, D. Ito1, J. Lee1, N. Abe1, K. Nakajima1
1 Kyoto University, Institute for Integrated Radiation and Nuclear Science, Osaka, Japan
To make a fast reactor system with trans-uranium (TRU) fuel containing minor actinides (MA) safer and more practical, the non-destructive quality control method for material accountancy is an important issue. Neutron Resonance Densitometry (NRD) is a non-destructive nuclide assay technique applicable to quantify nuclear materials in the fuel. To improve the rapidity and practicality in nuclide assay processing, the self-indication method was applied to NRD. In the self-indication method, an indicator consisting of the target nuclide is placed at the neutron beam downstream from a sample. The transmitted neutron thorough the sample can be measured indirectly by detecting the reaction products from the indicator with the neutron time-of-flight method. The areal density of target nuclide was able to be derived from the reduction ratio of counting rate around the neutron resonance energy. A 4p Bi4Ge3O12(BGO) spectrometer was installed to detect the capture gamma-rays from the indicator at the 46-MeV electron linear accelerator facility in Institute for Integral Radiation and Nuclear Science, Kyoto University. In the present work, we performed the quantitative examination for a nuclear material with highly enriched and depleted uranium-aluminum alloys. A natural uranium sheet was used as a self-indicator. As the results, the areal density of 238U was obtained successfully with the self-indication method for each uranium-aluminum alloy.
Development of gamma insensitive Silicon Carbide diagnostics to qualify intense thermal neutron fields at the e_LiBANS facility. (#1971)
M. Costa1, 2, R. Bedogni3, A. Pola4, 5, V. Monti1, 2, O. Sans-Planell1, 2, E. Durisi1, 2, L. Visca1, 2, D. Bortot4, 5, M. Treccani6, 4, J. M. Gomez-Ros7
1 university of torino, Department of Physics, torino, Italy
Radiation-resistant, gamma-insensitive, active, silicon carbide detectors were developed to qualify and monitor the thermal neutron fields at the e_LIBANS facility in Torino. Silicon carbide devices are sensitized to neutrons by means of 6LiF deposit process optimized to maximize the neutron capture probability and the subsequent detection of the alphas and tritons reaction products. This communication describes the study of the performances of these detectors, based on dedicated measurement campaigns resulting in a 2% response accuracy and shows how they can be made insensitive to gammas. Further a detailed description of their use to qualify the homogeneity of the field in the e_LiBANS cavity, where typical thermal neutron fluence rates are of the order of 2 x 106 cm-2s-1, is discussed. A matrix of 16 SiCs, 7 mm2 each, distributed on a 20 x 20 cm2 plane, have been used: parallel acquisition of the SiC's response allows to online elaborate the map of the thermal neutron field inside the cavity. The homogeneity has been proved to be better than 5% independent on the LINAC parameters, current and energy, that can be used to vary the absolute fluence rate in the cavity. The use of such a SiC's matrix for neutron beam monitor in other facilities is discussed.
Keywords: Thermal neutron fields, Silicon Carbide devices, Active Thermal Neutron detectors
Flexible silicon-based alpha particle detector (#1489)
1 University of York, Department of Physics, York, United Kingdom
A recognised issue in nuclear decommissioning, such as at the Sellafield site in the UK, is the contamination of the interior of pipework with alpha-only emitters such as Pu-239. Detectors are required that can sensitively identify very low levels of contamination without the need to cut up the pipe and inspect it manually. Given the short range of alpha particles in air, the curved nature of the pipe wall and the likely presence of high levels of background radiation, it would be advantageous to have a detector which was essentially insensitive to anything other than alpha radiation and which could conform to the geometry of the pipe wall. We have shown that silicon wafers of around 50-um thickness can be bent and conformed to a cylindrical geometry compatible with the interior of a typical 2" diameter pipe. Alpha particle detectors have been formed from such wafers following appropriate doping and preparation. A prototype pipeline inspection gauge based on such flexible silicon detectors has been demonstrated an a UK patent filed. Applications may also be foreseen for nuclear and particle physics where a cylindrical geometry is desirable.
Keywords: alpha particle detection, silicon detectors
Medical Radioisotopes production at a high-brilliance 14 MeV neutron source: the study case on 99Mo (#1902)
M. Capogni1, A. Pietropaolo1, L. Quintieri1, 2
1 ENEA, Roma, Italy
Single Photon Emission Computed Tomography, used worldwide for the diagnosis of a variety of pathological conditions, utilizes the gamma-emitting radionuclide 99mTc. This is obtained from 99Mo/99mTc generators as pertechnetate ion. 99Mo for such generators is currently produced in nuclear fission reactors as a by-product of 235U fission. Here we investigate an alternative production route of 99Mo, by irradiating a natural metallic Molybdenum powder at an accelerator-driven 14 MeV neutron source. Efficient isolation and purification of the 99mTc-pertechnetate is performed by solvent extraction, achieving in traceable metrological conditions, a level of radionuclide purity complying the pharmaceutical quality standards. Monte Carlo simulations enable at reliably predicting the99Mo production rates for a newly designed 14 MeV neutron source. We show that this source, may potentially supply an appreciable fraction of the current 99Mo global demand, accomplishing the request for a robust and safe solution to secure the long-term supply of 99Mo.
Keywords: medical radioisotopes, Fusion neutrons, medical applications of fusion sources
High-Sensitivity Fiber Optic Radiation Monitor for Design Basis Accident Environments (#2001)
S. Hatakeyama1, K. Ueno1, Y. Ueno1, T. Tadokoro1, K. Kamada2, R. Murakami2, A. Yoshikawa2
1 Hitachi, Ltd., Research & Development Group, Hitachi, Japan
Fiber optic radiation monitors are a promising dosimetry tool for remote and real time radiation monitoring in nuclear power plant facilities. Recently, we have been developing a high-sensitivity fiber optic radiation monitor for use during normal operation and design basis accidents of such facilities. In order to improve the sensitivity, we selected GdTaO4 (GTO):Eu and GTO:Tb single crystal as the detection material. In this work, we evaluated the temperature dependence and the afterglow properties of the prototype high-sensitivity fiber optic radiation monitor. According to the temperature dependence evaluation, the count rate of GTO:Tb satisfied the decision criterion within ±30% (JIS Z4324) up to 150℃ and achieved the target operating temperature of higher than 110℃. In addition, we considered that the electrons captured inside the defect level of the crystal were effectively applied to the 5d-4f electronic transition and they depressed the afterglow properties in GTO:Tb.
Keywords: Radiation monitor, Optical fiber, Single photon counting method, Temperature dependence, Afterglow property