Development of a Dual Alpha-Gamma Camera for Radiological Characterization Applications (#1280)
C. Mahe1, J. Venara1, S. Mitra1
1 CEA, DEN/MAR/DE2D/SDTC/LSTD, BAGNOLS/CEZE, France
The localization and visualization of radionuclides are some of the most important aspects during decommissioning of nuclear facilities. In order to reduce contamination risks among the concerned personnel, it is very important to spot the contaminated areas and especially the “hot spots” in the dismantling plant. The CEA nuclear measurement laboratory has developed its expertise since 1980’s on gamma imaging. Several kinds of gamma imaging devices has been deployed on real application cases since the first gamma camera prototype called ALADIN. More recently on that topic, the CEA is still fully implicated in the development and use of new types of gamma cameras. Later, special attention was given to alpha localization and more recently the concept of dual camera was established on the basis of the newly developed (patented) method of combining the technology of alpha and gamma contamination detection. The basic aim of this paper is to explain the mechanism of a dual alpha-gamma camera to locate and image both radiations with a single tool at a nuclear dismantling facility. The new design of the dual camera will reduce the expenditure and also the space requirements previously needed by the two separate cameras and will have several applications in nuclear safety, security and safeguards for nuclear maintenance, research reactors as well as small modular reactors. A first prototype of a dual alpha/gamma camera has been patented and tested since 2015. The technology of this new radiological tool will be described and the main results will be discussed.
Keywords: Dual alpha-gamma camera, pinhole design, scintillators, UV lens, UV Filters, Nuclear Plant Decommissionning
Real-time determination of Rossi-α distribution, active fast neutron multiplicity, neutron angular distribution and neutron spectrum using organic liquid scintillators (#2434)
R. Sarwar1, V. Astromskas1, C. H. Zimmerman2, S. Croft3, M. J. Joyce1
1 Lancaster University, Engineering Department, Lancaster, Lancashire, United Kingdom of Great Britain and Northern Ireland
Neutron assay of fissile materials has often used 3He proportional counters because they offer high efficiency to thermalized neutrons, excellent discrimination against g radiation, and simple and stable long-term operation. However, 3He is in short supply and expensive, and potentially useful information is lost or obscured when compared to fast neutron detector alternatives because of the time lag between the neutron emission from fission and the 3He (n, p) detection step due to the need for thermalisation. This can limit the use of 3He-based detectors, particularly for neutron spectroscopy and high-order multiplicity assessment. With organic scintillators, it is possible to overcome these deficiencies because they detect fast neutrons and thus have very short die-away characteristics relative to 3He. Previously, a field-programmable gate array (FPGA) system was introduced utilizing organic scintillators to determine the multiplicity, Rossi-α, and Feynman-Y distributions with a novel event-triggered algorithm giving the size of neutron clusters of incoming events rather than the reduced factorial moment. In this paper, a new semi-empirical model is described for the characterization of Rossi-α distributions using this system which separates the incident radiation and environmental scatter. The results imply that: a gatewidth of just 20 ns is satisfactory for 99.7% fast neutrons from a fission event; the relationship between neutron multiplicity and the fissile mass can be obtained via active neutron coincidence counting for UOx samples (depleted through to high enrichments); a dipolar trend for the neutron angular distributions is observed for 252Cf and the neutron energy spectrum of 252Cf can be measured via a photon-triggered neutron time-of-flight method. In presenting these results the flexibility and potential of organic scintillator based fast neutron detectors for safety and safeguards applications will be discussed along with the remaining R&D challenges.
Keywords: Rossi-α distribution, active fast neutron multiplicity, angular distribution, neutron spectroscopy, liquid scintillants, FPGA
A new Reduced-Height Calorimeter for the Quantification of the Nuclear Energy Deposition Rate inside MTRs: Advantages during an Irradiation Campaign. (#2502)
A. Volte1, C. Reynard-Carette1, J. Brun1, A. Lyoussi2, M. Carette1
1 Aix Marseille Univ, Université de Toulon, CNRS, IM2NP, Marseille, France
The improvement of materials and fuels for nuclear reactors requires accurate measurements during in-pile experiments realized inside research reactors such as Material Testing Reactors (MTRs). Consequently, Aix-Marseille University and CEA into the framework of a joint laboratory, called LIMMEX, are involved in research programs focusing on the design of new instrumentation, sensors, and measurement methods. In particular, works are dedicated to the study of calorimeters used for the quantification of the energy deposition rate induced by interactions between nuclear rays and matter (called nuclear heating). A validated complete approach coupling experiments (under laboratory conditions out-of-nuclear-radiation and under real conditions inside MTR irradiation channels) with numerical works (thermal simulations) allows the design of different heat-flow differential calorimeters with specific targeted characteristics. The aim of this paper is to present a new configuration of the CALORRE calorimeter with a reduced-height (divided by two) and to give its advantages. The reduced-height calorimetric cell of CALORRE will be detailed. Then its response curve will be given in the case of a preliminary laboratory calibration step by simulating the nuclear heating thanks to heat sources with fixed fluid-flow temperature and forced convection regime. Moreover, a theoretical model of the sensor response will be applied to define the contribution of each heat transfer mode during calibration. Calibration curves will be used to estimate the temperatures reached inside experimental irradiation channels for a nuclear heating < 1 W/g. These results will be compared to experimental data obtained with a usual differential calorimeter (CARMEN) and a first prototype of CALORRE (full-height configuration) during an irradiation campaign in MARIA reactor in November 2015. This comparison will lead to the discussion of the advantages of this new configuration.
Keywords: Calorimeter, Nuclear heating, Material Testing Reactor, Calibration, Heat transfers, Irradiation campaign
Spectroscopic Gamma Ray Imaging of the High Flux Australian Reactor (#2684)
D. Boardman1, M. C. Guenette1, A. Sarbutt1, D. Prokopovich1, A. Flynn1, C. Hughes1
1 Australian Nuclear Science And Technology Organisation, Nuclear Stewardship, Lucas Heights, NSW, Australia
A specialized spectroscopic gamma ray imaging system has been developed to image the High Flux Australian Reactor (HIFAR) reactor vessel, in support of the characterization phase of the decommissioning process. The ability to remotely locate, identify, and quantify gamma emitting radionuclides in high dose rate environments can provide crucial information for the development of an efficient decommissioning plan. The imaging methodology is based on a new compressive sensing technique, which allows for a complete image to be formed from a highly under sampled set of measurements. The imaging system has been designed to operate within HIFAR’s harsh radiation environment of ~10 Sv/h, and provides a 2-pi field of view imaging capability for gamma-ray energies between 40 keV and 1.5 MeV. The imaging system makes use of a single spectroscopic CdZnTe detector and a set of dual, rotating, nested hemispherical masks, which have been 3D printed in tungsten. Gamma-ray images, from preliminary tests of the imaging system, have been generated with single and multiple 241Am point sources, and 152Eu rods arranged in more complex patterns. The first spectroscopic imaging results from within a high dose rate reactor vessel environment will also be presented.
Keywords: gamma-ray imaging, reactor imaging, decommissioning, compressed sensing
TRItium detection By ElectroChemical Assisted radiometrics (TRIBECA) (#3077)
G. Berhane1, C. Boxall1, M. J. Joyce1, J. Pates1
1 Lancaster University, Lancaster, United Kingdom of Great Britain and Northern Ireland
Tritium (T) is made during the operation of nuclear reactors. This can give rise to waterborne T (as tritiated water HTO) in spent fuel cooling ponds, processing & waste treatment facilities – potential sources of leaks to ground. HTO behaves identically to H2O and so is highly mobile in the environment and human tissue, with resultant health risks. Thus, there are pressing needs for fast, accurate & precise measurement of T on & around nuclear sites. Tritium decays to He-3 by soft b decay of low average energy 5.7keV, making detection difficult. The UK Environment Agency recommended technique for waterborne tritium detection is liquid scintillation counting (LSC). This has a number of drawbacks. Compounding problems due to the low b energy is the fact that LSC is subject to interferences from e.g. unwanted emissions from other radionuclides in the sample. Too, environmental T levels are typically low. Thus samples require substantial pre-treatment before analysis rendering LSC laborious, expensive and inappropriate for in situ or automated monitoring of T in e.g. groundwaters. Solid scintillators are an alternative which, whilst being faster & cheaper, are considered less efficient than LSC because they rely on surface detection whilst LSC involves volume detection. One way to obviate this is to pre-concentrate T from the sample at the solid scintillator surface. Tritium has identical chemistry to hydrogen, including its interaction with palladium (Pd), a metal unique in its capacity for absorbing H into its crystal matrix. Accordingly, we have been studying the possibility of coupling nanoporous Pd layers directly to the solid scintillators to give a means by which T could be pre-concentrated at a scintillator surface prior to analysis by PMT-based scintillation counting – the first time that this has been done. This has yielded a novel radiometric instrument for T detection potentially offering fast, interference free, in situ detection & monitoring.
Keywords: waterborne tritium detection, solid scintillation counting for tritium
Gamma Emission Tomography for the Inspection of Spent Nuclear Fuel (#3395)
A. R. Lebrun1, M. Mayorov1, T. A. White1, J. Brutscher2, J. Keubler2, A. Birnbaum2, V. Ivanov3, T. Honkamaa4, J. Dahlberg5
1 International Atomic Energy Agency, Department of Safeguards, Vienna, Austria
A Passive Gamma Emission Tomography system (PGET)  was developed for the IAEA Safeguards for verification of dismountable nuclear fuel assemblies (SFAs). In 2015-2016 PGET underwent significant re-design and its performance has been tested on multiple SFA types. The re-designed PGET features the functionality of traditional non-destructive assay systems commonly used for spent fuel verification: total neutron counting (Fork Detector, FDET), medium-resolution gamma spectrometry (Irradiated Item Attribute Tester, IRAT or Spent Fuel Attribute Tester, SFAT) and spent fuel assembly’s lattice image (Digital Cherenkov Viewing Device, DCVD). Two 10B neutron detectors and one-hundred-seventy-four collimated CdZnTe detectors are grouped in two arrays on a rotary baseplate inside a watertight stainless steel enclosure. A SFA is lowered through the center of the enclosure and held stationary to perform an underwater measurement. Detector arrays are rotated on a baseplate in the horizontal plane around vertical axis of symmetry to obtain gamma sinogram and neutron count data simultaneously, typically in 3-5 min per assembly. Additionally, medium resolution spectra from all gamma detectors can be collected and recorded. Functional, technical and operational performance of the PGET was tested at four nuclear reactors on mockup, PWR17×17, BWR and WWER—440 SFAs. Measurements have been performed on fuel with burnup in the range 5.7-58GWd/tU and cooling times from 1.9 to 27years. Lateral pin structure of the SFAs could be reconstructed for any tested fuel design in the above range of cooling times and burnups. Missing or replaced pins in all fuel types could be clearly visualized in the reconstructed images; spectrometric information (134Cs/137Cs peak ratio) and neutron counting rates were found to be consistent with declared fuel radiation history.
This paper describes details of the PGET hardware and electronics and presents some results of performance evaluation.
Keywords: gamma-ray emission tomography, neutron detection, safeguards, spent nuclear fuel