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DOE OSTI2013

Testing and Performance Validation of a Sensitive Gamma Ray Camera Designed for Radiation Detection and Decommissioning Measurements in Nuclear Facilities-13044

This paper describes the measurements, testing and performance validation of a sensitive gamma ray camera designed for radiation detection and quantification in the environment and decommissioning and hold-up measurements in nuclear facilities. The instrument, which is known as RadSearch, combines a sensitive and highly collimated LaBr{sub 3} scintillation detector with an optical (video) camera with controllable zoom and focus and a laser range finder in one detector head. The LaBr{sub 3} detector has a typical energy resolution of between 2.5% and 3% at the 662 keV energy of Cs-137 compared to that of NaI detectors with a resolution of typically 7% to 8% at the same energy. At this energy the tungsten shielding of the detector provides a shielding ratio of greater than 900:1 in the forward direction and 100:1 on the sides and from the rear. The detector head is mounted on a pan/tile mechanism with a range of motion of ±180 degrees (pan) and ±90 degrees (tilt) equivalent to 4 π steradians. The detector head with pan/tilt is normally mounted on a tripod or wheeled cart. It can also be mounted on vehicles or a mobile robot for access to high dose-rate areas and areas with high levels of contamination. Ethernet connects RadSearch to a ruggedized notebook computer from which it is operated and controlled. Power can be supplied either as 24-volts DC from a battery or as 50 volts DC supplied by a small mains (110 or 230 VAC) power supply unit that is co-located with the controlling notebook computer. In this latter case both power and Ethernet are supplied through a single cable that can be up to 80 metres in length. If a local battery supplies power, the unit can be controlled through wireless Ethernet. Both manual operation and automatic scanning of surfaces and objects is available through the software interface on the notebook computer. For each scan element making up a part of an overall scanned area, the unit measures a gamma ray spectrum. Multiple radionuclides may be selected by the operator and will be identified if present. In scanning operation the unit scans a designated region and superimposes over a video image the distribution of measured radioactivity. For the total scanned area or object RadSearch determines the total activity of operator selected radionuclides present and the gamma dose-rate measured at the detector head. Results of hold-up measurements made in a nuclear facility are presented, as are test measurements of point sources distributed arbitrarily on surfaces. These latter results are compared with the results of benchmarked MCNP Monte Carlo calculations. The use of the device for hold-up and decommissioning measurements is validated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Validation of Non-Invasive Waste Assay System (Gamma Box Counter) Performance at AECL Whiteshell Laboratories - 13136

Low-level radioactive waste (LLW) in solid form, resulting from decommissioning and operations activities at AECL's Whiteshell Laboratories (WL), is packaged in B-25 and B-1000 standard waste containers and characterized before it is shipped to an on-site interim storage facility, pending AECL decisions on long term management of its LLW. Assay of the waste packages before shipment contributes to an inventory of the interim storage facility and provides data to support acceptance at a future repository. A key characterization step is a gamma spectrometric measurement carried out under standard conditions using an automated, multi-detector Waste Assay System (WAS), purchased from Antech Corporation. A combination of ORTEC gamma acquisition software and custom software is used in this system to incorporate multiple measurements from two collimated high-resolution detectors. The software corrects the intensities of the gamma spectral lines for geometry and attenuation, and generates a table of calculated activities or limits of detection for a user-defined list of radioisotopes that may potentially be present. Validation of WAS performance was a prerequisite to routine operation. Documentation of the validation process provides assurance of the quality of the results produced, which may be needed one or two decades after they were generated. Aspects of the validation included setting up a quality control routine, measurements of standard point sources in reproducible positions, study of the gamma background, optimization of user-selectable software parameters, investigation of the effect of non-uniform distribution of materials and radionuclides, and comparison of results with measurements made using other gamma detector systems designed to assay bulk materials. The following key components of the validation process have been established. A daily quality control routine has been instituted, to verify stability of the gamma detector operation and the background levels. A test box containing plywood (specific gravity 0.5) with holes for reproducible positioning of known point sources has been constructed and used in several configurations. Gamma spectra have been processed multiple times with user-adjustable parameters set to a variety of values to gain agreement on the most appropriate settings. A similar optimization process was followed with the custom software provided by the instrument manufacturer to reach agreement on the most stable settings for reliable results. Measurements made with the Canberra In-Situ Object Counting System (ISOCS) instrument were compared with the WAS results as part of an estimate of instrument precision and accuracy. The validation report will conclude with recommendations for the configuration, appropriate use and limitations of the Waste Assay System. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

A Mobile Automated Tomographic Gamma Scanning System - 13231

Canberra Industries have recently designed and built a new automated Tomographic Gamma Scanning (TGS) system for mobile deployment. The TGS technique combines high-resolution gamma spectroscopy with low spatial resolution 3-dimensional image reconstruction to provide increased accuracy over traditional approaches for the assay of non-uniform source distributions in low-to medium-density, non-heterogeneous matrices. Originally pioneered by R. Estep at Los Alamos National Laboratory (LANL), the TGS method has been further developed and commercialized by Canberra Industries in recent years. The present system advances the state of the art on several fronts: it is designed to be housed in a standard cargo transport container for ease of transport, allowing waste characterization at multiple facilities under the purview of a single operator. Conveyor feed, drum rotator, and detector and collimator positioning mechanisms operated by programmable logic control (PLC) allow automated batch mode operation. The variable geometry settings can accommodate a wide range of waste packaging, including but not limited to standard 220 liter drums, 380 liter overpack drums, and smaller 20 liter cans. A 20 mCi Eu-152 transmission source provides attenuation corrections for drum matrices up to 1 g/cm{sup 3} in TGS mode; the system can be operated in Segmented Gamma Scanning (SGS) mode to measure higher density drums. To support TGS assays at higher densities, the source shield is sufficient to house an alternate Co-60 transmission source of higher activity, up to 250 mCi. An automated shutter and attenuator assembly is provided for operating the system with a dual intensity transmission source. The system's 1500 kg capacity rotator turntable can handle heavy containers such as concrete lined 380 liter overpack drums. Finally, data acquisition utilizes Canberra's Broad Energy Germanium (BEGE) detector and Lynx MCA, with 32 k channels, providing better than 0.1 keV/channel resolution to support both isotopic analysis with the MGA/MGAU software and a wide 3 MeV dynamic range. The calibration and verification of the system is discussed, and quantitative results are presented for a variety of drum types and matrices. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Handling Radioactive Waste from the Proton Accelerator Facility at the Paul Scherrer Institut (PSI) - Always Surprising? - 13320

The Paul Scherrer Institut (PSI) is the largest national research centre in Switzerland. Its multidisciplinary research is dedicated to a wide field in natural science and technology as well as particle physics. In this context, PSI is operating, amongst others, a large proton accelerator facility since more than 30 years. In two cyclotrons, protons are accelerated to high speeds and then guided along roughly 100 m of beam line to three different target stations to produce secondary particles like mesons and neutrons for experiments and a separately beam line for UCN. The protons induce spallation processes in the target materials, and also at other beam loss points along the way, with emission of protons, neutrons, hydrogen, tritium, helium, heavier fragments and fission processes. In particular the produced neutrons, due to their large penetration depth, will then interact also with the surrounding materials. These interactions of radiation with matter lead to activation and partly to contamination of machine components and the surrounding infrastructures. Maintenance, operation and decommissioning of installations generate inevitably substantial amounts of radioactive operational and dismantling waste like targets, magnets, collimators, shielding (concrete, steel) and of course secondary waste. To achieve an optimal waste management strategy for interim storage or final disposal, radioactive waste has to be characterized, sorted and treated. This strategy is based on radiation protection demands, raw waste properties (size, material, etc.), and requirements to reduce the volume of waste, mainly for legal and economical reasons. In addition, the radiological limitations for transportation of the waste packages to a future disposal site have to be taken into account, as well as special regulatory demands. The characterization is a task of the waste producer. The conditioning processes and quality checks for radioactive waste packages are part of an accredited waste management process of PSI, especially of the Section Dismantling and Waste Management. Strictly proven and accepted methods needed to be developed and enhanced for safe treatment, transport, conditioning and storage. But in the field of waste from research activities, individual and new solutions have to be found in an increasingly growing administrative environment. Furthermore, a wide variety of components, with a really large inventory of radioactive nuclides, has to be handled. And there are always surprising challenges concerning the unusual materials or the nuclide inventory. In case of the operational and dismantling radioactive accelerator waste, the existing conditioning methods are in the process of a continuous enhancement - technically and administratively. The existing authorized specifications of conditioning processes have to be extended to optimize and fully describe the treatment of the inevitably occurring radioactive waste from the accelerator facility. Additional challenges are the changes with time concerning the legal and regulatory requirements - or do we have to consider it as business as usual? This paper gives an overview of the current practices in radioactive waste management and decommissioning of the existing operational accelerator waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

A Novel and Cost Effective Approach to the Decommissioning and Decontamination of Legacy Glove Boxes - Minimizing TRU Waste and Maximizing LLW Waste - 13634

This paper describes the process of decommissioning two gloveboxes at the Argonne National Laboratory (ANL) that were employed for work with plutonium and other radioactive materials. The decommissioning process involved an initial phase of clearing tools and materials from the glove boxes and disconnecting them from the laboratory infrastructure. The removed materials, assessed as Transuranic (TRU) waste, were packaged into 55 gallon (200 litre) drums and prepared for ultimate disposal at the Waste Isolation Pilot Plant (WIPP) at Carlsbad New Mexico. The boxes were then sampled to determine the radioactive contents by means of smears that were counted with alpha and beta detectors to determine the residual surface contamination, especially in terms of alpha particle emitters that are an indicator of TRU activity. Paint chip samples were also collected and sent for laboratory analysis in order to ascertain the radioactive contamination contributing to the TRU activity as a fixed contamination. The investigations predicted that it may be feasible to reduce the residual surface contamination and render the glovebox structure low level waste (LLW) for disposal. In order to reduce the TRU activity a comprehensive decontamination process was initiated using chemical compounds that are particularly effective for lifting and dissolving radionuclides that adhere to the inner surfaces of the gloveboxes. The result of the decontamination process was a reduction in the TRU surface activity on the inner surfaces of the gloveboxes by four orders of magnitude in terms of disintegrations per unit area (DPA). The next phase of the process involved a comprehensive assay of the gloveboxes using a combination of passive neutron and gamma ray scintillation detectors and a shielded and collimated high purity Germanium (HPGe) gamma ray detector. The HPGe detector was used to obtain gamma ray spectra for a variety of measurement positions within the glovebox. The spectra were used to determine the TRU content of the boxes by assessing the activity of Am-241 (59 keV) and Pu-241 (414 keV). Using the data generated it was possible for qualified subject matter experts (SME) to assess that the gloveboxes could be consigned for disposition as LLW and not as TRU. Once this determination was assessed and accepted the gloveboxes were prepared for final disposition to the Nevada National Security Site (NNSS) - formerly the Nevada Test Site (NTS). This preparation involved fixing any remaining radioactive contamination within the gloveboxes by filling them with a foam compound, prior to transportation. Once the remaining contamination was fixed the gloveboxes were removed from the laboratory and prepared for transported by road to NNSS. This successful glovebox decontamination and decommissioning process illustrates the means by which TRU waste generation has been minimized, LLW generation has been maximized, and risk has been effectively managed. The process minimizes the volume of TRU waste and reduced the decommissioning time with significant cost savings as the result. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Neutron Calibration Studies for the SuperCDMS SNOLAB Experiment

My time at Fermilab has been dedicated to advancing the field of dark matter detection through my contributions to the SuperCDMS collaboration. The first project involved the development and testing of a delivery system for a Cf-252 neutron source to calibrate the SuperCDMS silicon and germanium detectors. Subsequently, I focused on parametrizing the detectors' instrumental response by optimizing a plastic scintillator backing array to measure the scattering angle of a collimated neutron beam. The Monte Carlo simulation I designed provided insights into the detection efficiency of the array, and the best-fit design I arrived at for the scattering counters will be implemented into the final array. Additionally, I assembled a muon detector with legacy hardware to enhance the ability of the calibration experiment to reduce dark current and false signals. As my work nears completion, I am confident that these contributions will significantly enhance the precision and reliability of the SuperCDMS experiment, pushing the boundaries of our understanding of dark matter and contributing to the broader scientific community's quest to unravel the mysteries of the universe.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2015

Performance Study of an aSi Flat Panel Detector for Fast Neutron Imaging of Nuclear Waste

Radioactive waste must be characterized to check its conformance for intermediate storage and final disposal according to national regulations. For the determination of radio-toxic and chemo-toxic contents of radioactive waste packages non-destructive analytical techniques are preferentially used. Fast neutron imaging is a promising technique to assay large and dense items providing, in complementarity to photon imaging, additional information on the presence of structures in radioactive waste packages. Therefore the feasibility of a compact Neutron Imaging System for Radioactive waste Analysis (NISRA) using 14 MeV neutrons is studied in a cooperation framework of Forschungszentrum Juelich GmbH, RWTH Aachen University and Siemens AG. However due to the low neutron emission of neutron generators in comparison to research reactors the challenging task resides in the development of an imaging detector with a high efficiency, a low sensitivity to gamma radiation and a resolution sufficient for the purpose. The setup is composed of a commercial D-T neutron generator (Genie16GT, Sodern) with a surrounding shielding made of polyethylene, which acts as a collimator and an amorphous silicon flat panel detector (aSi, 40 x 40 cm{sup 2}, XRD-1642, Perkin Elmer). Neutron detection is achieved using a general propose plastic scintillator (EJ-260, Eljen Technology) linked to the detector. The thermal noise of the photodiodes is reduced by employing an entrance window made of aluminium. Optimal gain and integration time for data acquisition are set by measuring the response of the detector to the radiation of a 500 MBq {sup 241}Am-source. Detector performance was studied by recording neutron radiography images of materials with various, but well known, chemical compositions, densities and dimensions (Al, C, Fe, Pb, W, concrete, polyethylene, 5 x 8 x 10 cm{sup 3}). To simulate gamma-ray emitting waste radiographs in presence of a gamma-ray sources ({sup 60}Co, {sup 137}Cs, {sup 241}Am) were performed. A homemade algorithm was developed to determine a value which is related to the neutron absorption of the sample with the analysis of the raw detector data. The detector was placed 42 cm away from the neutron source. Distance between detector and the samples was 0.5 cm. At the sample position the fast neutron flux was estimated to 9x10{sup 3} n cm{sup -2} s{sup -1} for a neutron emission of 10{sup 8} n s{sup -1}. The acquisition time was 15 minutes. First neutron radiographs were successfully recorded despite the low detector efficiency and low neutron emission. Analysis of the data shows a correlation between the measured signal and determined neutron absorption. Thus discrimination between different materials of same thicknesses may be achieved. The measurements and results will be presented and discussed in details.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Implementation of a new gamma spectrometer on the MERARG loop: Application to the volatile fission products release measurement

The MERARG facility initially aims at the annealing of irradiated fuel samples to study the gaseous fission products release kinetics. In order to complete the evaluation of the source term potentially released during accidental situation, the MERARG experimental circuit has been enhanced with a new gamma spectrometer. This one is directly sighting the fuel and is devoted to the fission products release kinetics. Because of the specificities of the fuel measurements, it has been dimensioned and designed to match the specific requirements. The acquisition chain and the collimation system have been optimized for this purpose and a first set of two experiments have shown the good functioning of this new spectrometry facility. (authors)

42 ENGINEERING↗
DOE OSTI2015

Design and Validation Testing of TruckScan to Assay Large Sacks of Fukushima Radioactive Debris on a Truck

As a result of the March 2011 earthquake and resulting tsunami in Japan, there was a serious accident at the Fukushima Dai-ichi Nuclear Power Plant. This accident has contaminated soil and vegetation in a wide area around the plant. Decontamination projects over the last 4 years have resulted in large numbers of 1 cubic meter canvas bags of debris, commonly called Super Sacks [SS]. These are currently stored nearby where they were generated, but starting in 2015, they will be moved to various Interim Storage Facilities [ISF]. Trucks will typically carry 8-20 of these SSs. When the trucks arrive at the ISF they need to be rapidly sorted into groups according to radioactivity level, for efficient subsequent processing. Canberra Industries, Inc. [CI] has designed a new truck monitoring system 'TruckScan' for use at these ISFs. The TruckScan system must measure the entire truck loaded with multiple closely packed SSs, and generate a nuclide specific assay report showing the radioactivity in each individual SS. The Canberra-Japan office, along with Obayashi Corporation have performed validation testing to demonstrate to the regulatory authorities that the proposed technique was sufficiently accurate. These validation tests were conducted at a temporary storage area in Fukushima prefecture. Decontaminated waste of various representative types and of various levels of radioactivity was gathered and mixed to create homogeneous volumes. These volumes were sampled multiple times and assayed with laboratory HPGe detectors to determine the reference concentration of each pile. Multiple SSs were loaded from each pile. Some of the SSs were filled 50% full, others 75% full, and others 100% full, to represent the typical loading configuration of the existing SSs in the field. The content of the SSs are either sand, soil, or vegetation with densities ranging from 0.3 g/cc - 1.6 g/cc. These SSs with known concentrations of Cs-134 and Cs-137 were then loaded onto trucks in a variety of configurations, typical of how they might be on the real trucks. A partial system was installed at the site and used to assay these trucks with the various loading configurations. Whereas the full system will have 8 collimated 3 x 3'' NaI detectors, the test system only had two detectors; therefore the truck was moved and counted 4 times. The data were acquired and analyzed with the Canberra Genie software to determine the peak counts for both Cesium nuclides. That data was then analyzed with a prototype version of a Maximum Entropy algorithm, to determine the individual SS activity. The goal of the validation tests was to demonstrate that the system could detect 1000 Bq/kg in 15 seconds, and to determine how accurately it could quantify individual SSs. The validation tests demonstrated that the product would perform as predicted. The TruckScan results were consistent with the sample assay results [y = 1.0029 x, R{sup 2} = 0.914]. The Total Propagated Uncertainty, including both uncertainties from these tests and others that were estimated but not tested was 16.6% percent. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute'

The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spent nuclear fuel from storages was removed for processing to 'Mayk'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ- spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Underwater characterization of control rods for waste disposal using SMOPY

Storage of spent fuel assemblies in cooling ponds requires careful control of the geometry and proximity of adjacent assemblies. Measurement of the fuel burnup makes it possible to optimise the storage arrangement of assemblies taking into account the effect of the burnup on the criticality safety margins ('burnup credit'). Canberra has developed a measurement system for underwater measurement of spent fuel assemblies. This system, known as 'SMOPY', performs burnup measurements based on gamma spectroscopy (collimated CZT detector) and neutron counting (fission chamber). The SMOPY system offers a robust and waterproof detection system as well as the needed capability of performing radiometric measurements in the harsh high dose - rate environments of the cooling ponds. The gamma spectroscopy functionality allows powerful characterization measurements to be performed, in addition to burnup measurement. Canberra has recently performed waste characterisation measurements at a Nuclear Power Plant. Waste activity assessment is important to control costs and risks of shipment and storage, to ensure that the activity level remains in the range allowed by the facility, and to declare activity data to authorities. This paper describes the methodology used for the SMOPY measurements and some preliminary results of a radiological characterisation of AIC control rods. After describing the features and normal operation of the SMOPY system, we describe the approach used for establishing an optimum control rod geometric scanning approach (optimum count time and speed) and the method of the gamma spectrometry measurements as well as neutron check measurements used to verify the absence of neutron sources in the waste. We discuss the results obtained including {sup 60}Co, {sup 110m}Ag and {sup 108m}Ag activity profiles (along the length of the control rods) and neutron results including Total Measurement Uncertainty evaluations. Full self-consistency checks were performed and these demonstrate the validity of the techniques. The results are described and analysed in the context of the measurement performance of the equipment. Different casks were fully characterized using a 60 mm{sup 3} CZT detector, to determine the total activities and spatial profiles. A total activity range measurement of 1x10{sup 8} - 1x10{sup 13} Bq/cm was found to be achievable. Finally, comments are made, based on our measurements, on the ability of this equipment for performing in-situ characterisation of wastes in the harsh environments typical of fuel assembly and waste storage ponds and silos. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

MARIA reactor is an open-pool research reactor what gives the chance to install uranium fission converter on the periphery of the core. It could be installed far enough not to induce reactivity of the core but close enough to produce high flux of fast neutrons. Special design of the converter is now under construction. It is planned to set the research stand based on such uranium converter in the near future: in 2015 MARIA reactor infrastructure should be ready (preparation started in 2013), in 2016 the neutron beam starts and in 2017 opening the stand for material and biological research or for medical training concerning BNCT. Unused for many years, horizontal channel number H2 at MARIA research rector in Poland, is going to be prepared as a part of unique stand. The characteristics of the neutron beam will be significant advantage of the facility. High flux of neutrons at the level of 2x10{sup 9} cm{sup -2}s{sup -1} will be obtainable by uranium neutron converter located 90 cm far from the reactor core fuel elements (still inside reactor core basket between so called core reflectors). Due to reaction of core neutrons with converter U{sub 3}Si{sub 2} material it will produce high flux of fast neutrons. After conversion neutrons will be collimated and moderated in the channel by special set of filters and moderators. At the end of H2 channel i.e. at the entrance to the research room neutron energy will be in the epithermal energy range with neutron intensity at least at the level required for BNCT (2x10{sup 9} cm{sup -2}s{sup -1}). For other purposes density of the neutron flux could be smaller. The possibility to change type and amount of installed filters/moderators which enables getting different properties of the beam (neutron energy spectrum, neutron-gamma ratio and beam profile and shape) is taken into account. H2 channel is located in separate room which is adjacent to two other empty rooms under the preparation for research laboratories (200 m2). It is planned to create fully equipped complex facility possible to perform various experiments on the intensive neutron beam. Epithermal neutron beam enables development across the full spectrum of materials research for example shielding concrete tests or electronic devices construction improvement. Due to recent reports on the construction of the accelerator for the Boron Neutron Capture Therapy (BNCT) it has the opportunity to become useful and successful method in the fight against brain and other types of cancers not treated with well known medical methods. In Europe there is no such epithermal neutron source which could be used throughout the year for training and research for scientist working on BNCT what makes the stand unique in Europe. Also our research group which specializes in mixed radiation dosimetry around nuclear and medical facilities would be able to carry out research on new detectors and methods of measurements for radiological protection and in-beam (therapeutic) dosimetry. Another group of scientists from National Centre for Nuclear Research, where MARIA research reactor is located, is involved in research of gamma detector systems. There is an idea to develop Prompt-gamma Single Photon Emission Computed Tomography (Pg- SPECT). This method could be used as imaging system for compounds emitting gamma rays after nuclear reaction with thermal neutrons e.g. for boron concentration in BNCT. Inside the room, where H2 channel is located, there is another horizontal channel - H1 which is also unused. Simultaneously with the construction of the H2 stand it will be possible to create special pneumatic horizontal mail inside the H1 channel for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. It might expand the scope of research at the planned neutron station. Secondly it is planned to equip both stands with moveable positioning system, video system and facilities to perform animal experiments (anaesthesia, vital signs control, imaging devices, positioning). These all above make constructed station unique in the world (uranium fission converter-based beam) and the only one of such intense neutron beam in the Europe. Moreover implementation of the station would allow the development of research on a number of issues for researchers from all over the Europe. One of very important advantages of the station is undisturbed exploitation of the reactor and other vertical and horizontal channels. MARIA reactor operates 6000 hours per year and that amount of time will be achievable for research on the neutron station. It have to be underlined that new neutron station will work parallel to all another ventures. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Container inspection in the port container terminal by using 14 MeV neutrons

A proposal for an autonomous and flexible ship container inspection system is presented. This could be accomplished by the incorporation of inspection system on the container transportation devices (straddle carriers, yard gentry cranes automated guided vehicles, trailers). This configuration is terminal specific and it will be decided by container terminal operator. In such a way no part of port operational area will be used for inspection. The inspection scenario will include container transfer from ship to transportation device with inspection unit mounted on it, inspection during container movement to the container location. A neutron generator without associated alpha particle detection will be used. This will allow the use of higher neutron intensity (5x10{sup 9} - 10{sup 10} n/s in 4π). The inspected container will be stationary in the 'inspection position' on the transportation device while the 'inspection unit' will move along its side. Following analytical methods will be used simultaneously: neutron radiography, X-ray radiography, neutron activation analysis, (n,γ) and (n,n'γ) reactions, neutron absorption, and scattering, X-ray backscattering, Neutron techniques will take the advantage of using 'smart collimators' for neutrons and gammas, both emitted and detected. The inspected voxel will be defined by intersections/union of neutron generator and detectors solid angles. The container inspection protocol will be based on identification of discrepancies between its cargo manifest and its elemental 'fingerprint' and radiography profiles. In addition, the information on container weight will be obtained during the container transport and foreseen screening from the measurement of density of material in the container. (authors)

42 ENGINEERING↗
DOE OSTI2015

Optimization of {sup 6}LiF:ZnS(Ag) Scintillator Light Yield Using Geant4

Neutrons provide an effective tool to probe materials structure. Neutron diffraction is a method to determine the atomic and magnetic structure of a material based on neutron scattering. By this method a collimated incident beam of thermal neutrons heat the examined sample and based on the obtained diffraction pattern information on the structure of the material is provided. Research for developing a novel cold neutron detector for Chromatic Analysis Neutron Diffractometer Or Reflectometer (CANDOR) is underway at the NIST center for neutron research. The system unique design is aimed to provide over ten times fold faster analysis of materials than conventional system. In order to achieve the fast analysis a large number of neutron detectors is required. A key design constraint for this detector is the thickness of the neutron sensitive element. This is met using {sup 6}LiF:ZnS(Ag) scintillation material with embedded wavelength shifting (WLS) fibers conducting scintillation light to silicon photomultiplier photo-sensors. The detector sensitivity is determined by both the neutron capture probability ({sup 6}Li density) and the detectable light output produced by the ZnS(Ag) ionization, the latter of which is hindered by the fluorescence absorption of the scintillation light by the ZnS. Tradeoffs between the neutron capture probability, stimulated light production and light attenuation for determining the optimal stoichiometry of the {sup 6}LiF and ZnS(Ag) as well as the volume ratio of scintillator and fiber. Simulations performed using the GEANT4 Monte Carlo package were made in order to optimize the detector design. GEANT4 enables the investigation of the neutron interaction with the detector, the ionization process and the light transfer process following the nuclear process. The series of conversions required for this detector were modelled: - A cold neutron enters the sensor and is captured by {sup 6}Li in the scintillator mixture ({sup 6}Li (n,α) {sup 3}H reaction). The study of investigating the capture process probability for neutron energy of 5.1 meV to 2.27 meV (4 - 6 A) is presented. - Alpha particles and tritons travel for a few microns in the scintillation material (α ∼0.007 mm, T ∼0.04 mm) losing energy and ionizing the ZnS. The mean free path of the two particles in each of the component materials and the complete compound was investigated. - The ionization of the ZnS(Ag) scintillation material produces blue light photons with luminescence wavelength of 450 nm. The amount of light output produced for different grain sizes of ZnS is discussed. - A large portion of the scintillation photons are reabsorbed during their passage through the scintillation material. - The blue photons that reach the WLS fibers are absorbed by fluorescent dye and are re-emitted as green photons, conducted by the fiber to the SiPM photo-sensor. This work presents the CANDOR unique design and its design constrains, the results measured by the ultra-thin {sup 6}LiF:ZnS(Ag)-based neutron detector versus the simulation results for several binder concentrations. The light measurement attenuation results along with the measured stopping power were utilized to predict the sensitivity results of configuration with different ZnS grain size, weight ratios and fibers geometry (number and location). The simulations enable to optimize the final sensor design. This design successfully achieved both the high gamma rejection with a sensitive and accurate neutron event detection of 80 percent. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Inspection of a LOCA test rod at the Halden reactor project using gamma emission tomography

The LOCA test series IFA-650 conducted at the OECD Halden Reactor Project (HRP) has provided unique data on the performance of fuel rods during LOCA transients. One focus of the current investigations is the performance of the fuel in the ballooning stage of the LOCA transient. In this stage, relocation of fuel material is a possibility, in which case pellet fragments fall down to fill the void introduced by the increased volume of the ballooned cladding. This increases the heat load in that region, further promoting corrosion of the cladding. A special concern in the case of high-burnup fuels is the increasing number of small fuel fragments, which may be expected to cause a higher packing fraction in the ballooned region. In this work, a novel technique is presented for assessing the average density of the fuel material in the ballooned region of LOCA test rods. The investigation is based on non-destructive gamma emission tomography measurements, using the dedicated instrument recently developed at the HRP in collaboration with Westinghouse (Sweden) and Uppsala University. In this approach, the gamma radiation field surrounding the test rod has been measured with a narrowly collimated HPGe detector. Tomographic reconstruction of the data was performed, providing the radial gamma-ray source distribution within the measured volume, which reveals the fuel fragment distribution. From this, the density of the fuel in the measured volume (i.e., the packing fraction) may be calculated. The technique has been used to investigate a LOCA test rod of the Halden Reactor Project LOCA series. The LOCA experiment was carried out about one month prior to the gamma tomography examination. The results show that the distribution of the relocated fuel can be imaged using gamma rays from fission products. The reconstructions of the 662 keV rays from {sup 137}Cs and 1596 keV from {sup 140}Ba/La are demonstrated. In addition, the peaks of activation products offer valuable information on the location of the test rig structures, which may be utilized in a quantitative tomographic reconstruction to assess the spatially resolved packing fraction. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Materials Identification System for 200 liter/55 gallon Waste Drums using Computer Tomographic Techniques - 17162

In 2014 a thermal runaway event within a drum in-situ at the Waste Isolation Pilot Plant (WIPP) facility near Carlsbad, NM, resulted in the drum being breached and activity released. This emphasized the need for a better determination of drum contents and identification of material combinations and in-drum conditions. Enhanced characterization is required to identify and quantify the presence of absorbent materials, determine whether they are organic or inorganic, locate their position in the drum with respect to other materials, and quantify how much liquid is being held in the absorbent material. As part of the waste characterization process to ensure compliance with the WIPP waste acceptance criteria, drums are currently inspected by Real Time Radioscopic (RTR) x-ray to identify prohibited items such as free liquids and un-punctured aerosol cans, but that technique cannot satisfy these newly identified requirements, and a new approach to non-destructive examination was required. This approach complements the current RTR inspection with Cone-Beam Area Detector Computed Tomography (CT) scanning of the whole drum, followed by more-quantitative CT scanning of specific areas of concern, using a highly collimated Linear Detector Array. The density of representative waste drums required x-ray energies of up to 9 MeV, which meant that sorting of materials by their photo-electric absorption is not viable as this occurs in the 80 - 180 keV range. However, a notable characteristic of higher energy imaging in this context is the general correspondence between material density and total attenuation in the 1.5-4.0 MeV range due to Compton Scatter. Development of the imaging chain, comprising the x-ray source and the detectors was critical to achieving the required system performance. Theoretical analysis of the energy spectra of the 9 MeV beam, and filtration of the beam to remove the lower energies which cause beam hardening artifacts in the image was required. Theoretical evaluation of the required efficiency and image acquisition performance of area detectors and linear array detectors, was carried out to complete the selection of equipment. A system of density calibration rods was designed and, together with the selected equipment, was configured and used to acquire cone-beam CT data and line-scan CT data by scanning a 55 gallon drum containing surrogate waste. From the acquired data, CT reconstruction was performed and the basis of algorithms to identify materials were developed. Following this, a series of blind tests were done using a range of surrogate materials the contents of which were unknown to the system operators. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

High Resolution Scanning Configurations for the Canberra Tomographic Gamma Scanner - 17323

The Canberra Tomographic Gamma Scanner (TGS) is a Non-Destructive Assay (NDA) system that combines high resolution gamma ray spectroscopy (HRGS) with low-spatial-resolution imaging for the measurement of heterogeneous drummed waste with matrix densities in the range of 0 to 1 g/cc. The TGS methodology uses a two-pass scanning protocol; a 10-15 μCi {sup 152}Eu transmission source is used to interrogate density and composition of the matrix in the first pass, and passive gamma ray spectroscopy is used to collect emission spectra in the second, while the drum is rotated and translated to vary the view path. In standard implementations, the transmission and emission data are used to reconstruct three dimensional images of the matrix and activity distribution inside the drum on a 10x10x16 grid of discrete volume elements or 'voxels'. These images allow attenuation corrections to be calculated and applied on a view by view, and voxel by voxel basis, reducing the over- and under-correction biases obtained by assay techniques which assume homogeneous distributions of material. The most recent TGS system deployed by Canberra featured, among other new developments, the ability to perform higher resolution image re-construction, which may prove beneficial for the assay of more highly mixed matrices or oddly shaped assay items within a container. These additional scanning protocols reconstruct the image data into 14x14 and 20x20 voxel grids, using a smaller collimator aperture size. The hardware and software implementations of these higher resolution scans are discussed, and the calibration and initial evaluations of performance are presented. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

The Impacts of Source Location in Tomographic Gamma Scanning - 17324

Canberra Industries has recently designed and built a Tomographic Gamma Scanning (TGS) system for the deployment in a nuclear facility. The TGS technique combines high-resolution gamma spectroscopy with low spatial resolution 3-dimensional image reconstruction to provide increased accuracy over traditional approaches for the assay of non-uniform source distributions in low-to medium-density containers. The TGS system has been optimized to measure 120 liter drums and has been equipped to use various collimators and attenuators to assay containers under a wide range of operational conditions. In this paper, we report on experimental work using TGS systems aimed at evaluating the effects of source locations at the limits of volume elements, or voxels, defined in the image reconstruction. These include sources placed at voxel boundaries within a single vertical layer and sources placed in the lowest and highest possible vertical positions within the containers. Building upon previous studies, the accuracy of the TGS system was also assessed by varying the radial position of a point source within a vertical segment. The robustness of the TGS image reconstruction algorithm to location of the sources location was also analyzed as a function of matrix density and gamma-ray energy. Methods to reduce the biases from source location at extreme positions within the container were implemented and tested. The results from this work have been used to better understand the total measurement uncertainty (TMU) and analysis methodology used in TGS systems currently deployed and under development. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

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