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DOE OSTI2018

A Geant4 evaluation of the Hornyak button and two candidate detectors for the TREAT hodoscope

The performance of traditional Hornyak buttons and two proposed variants for fast-neutron hodoscope applications was evaluated using Geant4. The Hornyak button is a ZnS(Ag)-based device previously deployed at the Idaho National Laboratory's TRansient REActor Test Facility (better known as TREAT) for monitoring fast neutrons emitted during pulsing of fissile fuel samples. Past use of these devices relied on pulse-shape discrimination to reduce the significant levels of background Cherenkov radiation. Proposed are two simple designs that reduce the overall light guide mass (here, polymethyl methacrylate or PMMA), employ silicon photomultipliers (SiPMs), and can be operated using pulse-height discrimination alone to eliminate background noise to acceptable levels. Geant4 was first used to model a traditional Hornyak button, and for assumed, hodoscope-like conditions, an intrinsic efficiency of 0.35% for mono-directional fission neutrons was predicted. The predicted efficiency is in reasonably good agreement with experimental data from the literature and, hence, served to validate the physics models and approximations employed. Geant4 models were then developed to optimize the materials and geometries of two alternatives to the Hornyak button, one based on a homogeneous mixture of ZnS(Ag) and PMMA, and one based on alternating layers of ZnS(Ag) and PMMA oriented perpendicular to the incident neutron beam. For the same radiation environment, optimized, 5-cm long (along the beam path) devices of the homogeneous and layered designs were predicted to have efficiencies of approximately 1.3% and 3.3%, respectively. For longer devices, i.e., lengths larger than 25 cm, these efficiencies were shown to peak at approximately 2.2% and 5.9%, respectively. Furthermore, both designs were shown to discriminate Cherenkov noise intrinsically by using an appropriate pulse-height discriminator level, i.e., pulse-shape discrimination is not needed for these devices.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2019

ADVANCED NDE TECHNOLOGIES FOR DETECTING AND MANAGING AGING AND DEGRADATION IN REACTOR COMPONENTS

Recent research in nuclear power reactors has focused on the development and deployment of technologies for improving the economics of nuclear power plant (NPP) operation. Of interest, both from an economic and safety perspective, is the management and mitigation of aging-related degradation in critical components. In this paper, we describe potential techniques for nondestructive evaluation (NDE) that may be adapted for online monitoring of component aging in advanced non-light water reactors (ANLWR). The implications of the operating environment on the NDE measurement, especially when conducted continuously with the sensor placed in situ, are described. The design of sensors that can be used for monitoring component aging in situ and online are discussed, and results from laboratory testing at elevated temperature are outlined. Laboratory testing indicates the sensor can survive the operating temperatures in ANLWRs while providing a reasonable measurement signal-to-noise ratio. Ongoing research is focused on further testing and refining the sensor design and measurement procedures for deployment in ANLWR environments.

Ramuhalli, Pradeep↗

DOE OSTI2018

Performance of Advanced Signal Processing and Pattern Recognition Algorithms Using Raw Data from Ultrasonic Guided Waves and Fiber Optics Transducers

The Light Water Reactor Sustainability Program was initiated to evaluate technologies that could be used to perform online monitoring of piping and other secondary system structural components in commercial nuclear power plants. These online monitoring systems have the potential to identify when a more detailed inspection is needed using real-time measurements, rather than at a pre-determined inspection interval. This transition to condition-based, risk-informed automated maintenance will contribute to a significant reduction of operations and maintenance costs that account for most nuclear power generation costs. This report describes the current state of research related to ultrasonic guided wave testing and its application to detecting defects in commercial nuclear power plants. The report analyzes the applicability of the guided wave technology to secondary piping systems, as well as studying the potential for expanding the range of guided wave technology to include bent piping and other piping components. The ultrasonic guided waves can inspect long stretches of straight piping; however, more complex geometries such as elbows, welds, and tees are causing spurious reflections and coherent noise, which significantly decreases the sensitivity of the technique. To deal with these limitations, high-definition fiber optic sensors are applied to complex piping geometries, and advanced machine learning algorithms are used to detect deviations from healthy states. This report also analyzes the performance of advanced signal processing and machine learning-based pattern recognition algorithms in detecting defects in secondary structures. It is demonstrated on guided wave data collected at nuclear power plants that the independent component analysis can separate different coherent noise components and segregate them from useful signals. It also demonstrates that advanced machine learning techniques, such as neural networks, support vector machines, and deep learning networks, can detect minor defects present in inspected structures. Recommendation about the applicability of advanced machine learning techniques to online piping monitoring are also given.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Comparing Control Performance Between Simulation and Experiment using the Microreactor Automated Control System Testbed

In the advanced reactor domain, a flexible and scalable software/hardware infrastructure is crucial for integrating and validating various control technologies. This study used the Microreactor Automated Control System (MACS) hardware platform as a testbed. MACS was originally designed to mirror Idaho National Laboratory (INL)'s Microreactor Applications Research Validation and Evaluation (MARVEL), a 85-kW thermal fission microreactor. It features control drums for simulated reactivity control; lights that function as a surrogate reactor core, with the brightness being proportional to the reactor power; and light sensors that emulate neutron detectors. To transform MACS into a physical twin of MARVEL for evaluating control methods, the Control and Optimization Modular Modeling Application for Nuclear Deployment (COMMAND) software was employed. This software integrated the hardware with two models of the MARVEL core, based on Reactor Excursion and Leak Analysis Program (RELAP5-3D) and Monte Carlo N-Particle (MCNP) models. The study aimed to demonstrate the gap between control theory and actual practice—a gap that often necessitates empirical adjustments such as control gain retuning, filters, time discretization, and integrator anti-windup measures. Controllers were developed based on increasingly complex simulations without hardware, starting from the base MARVEL model and then introducing actuator saturation constraints and sensor noise. The final control strategy was then tested using MACS, and a comparative performance analysis was conducted.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

DOE OSTI2012

Development of coherent germanium neutrino technology (CoGeNT) for reactor safeguards

Antineutrinos are extremely penetrating elementary particles that have unique features of interest for safeguards at nuclear reactors. Current antineutrino detectors systems have large size and often use hazardous materials, presenting concerns to the safeguards agencies. We propose a new antineutrino detector, based on HPGe detector technology, which is much smaller and safer and therefore more likely to find widespread acceptance as a monitoring tool. The proposed system should be sensitive to a universally predicted but as-yet undetected antineutrino signature, the coherent neutrino-nucleus scattering, for which an unprecedentedly low noise threshold is required. Based on the noise analysis of an existing HPGe detector and the results of noise tests, a new system was designed and fabricated in collaboration with LBNL, though a full noise optimization has not been possible. Shielding design has been analyzed in the context of other deployment results, and an anticoincidence veto with high neutron efficiency has been tested.

Cabrera-Palmer, Belkis↗

DOE OSTI2012

Nondestructive Examination (NDE) Detection and Characterization of Degradation Precursors, Technical Progress Report for FY 2012

The overall objective of this project was to investigate the effectiveness of nondestructive examination (NDE) technology in detecting material degradation precursors by initiating and growing cracks in selected materials and using NDE methods to measure crack precursors prior to the onset of cracking. Nuclear reactor components are subject to stresses over time that are not precisely known and that make the life expectancy of components difficult to determine. To prevent future issues with the operation of these plants because of unforeseen failure of components, NDE technology is needed that can be used to identify and quantify precursors to macroscopic degradation of materials. Some of the NDE methods being researched as possible solutions to the precursor detection problem are magnetic Barkhausen noise, nonlinear ultrasonics, acoustic emission, eddy current measurements, and guided wave technology. In FY12, the objective was to complete preliminary assessment of advanced NDE techniques for sensitivity to degradation precursors, using prototypical degradation mechanisms in laboratory-scale measurements. This present document reports on the deliverable that meets the following milestone: M3LW-12OR0402143 – Report detailing an initial demonstration on samples from the crack-initiation tests will be provided (demonstrating acceleration of the work).

Nondestructure examination, degradation, precursor↗

DOE OSTI2014

Displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor

The experiments of displacement damage effects on CMOS APS image sensors induced by neutron irradiation from a nuclear reactor are presented. The CMOS APS image sensors are manufactured in the standard 0.35 μm CMOS technology. The flux of neutron beams was about 1.33 × 10{sup 8} n/cm{sup 2}s. The three samples were exposed by 1 MeV neutron equivalent-fluence of 1 × 10{sup 11}, 5 × 10{sup 11}, and 1 × 10{sup 12} n/cm{sup 2}, respectively. The mean dark signal (K{sub D}), dark signal spike, dark signal non-uniformity (DSNU), noise (V{sub N}), saturation output signal voltage (V{sub S}), and dynamic range (DR) versus neutron fluence are investigated. The degradation mechanisms of CMOS APS image sensors are analyzed. The mean dark signal increase due to neutron displacement damage appears to be proportional to displacement damage dose. The dark images from CMOS APS image sensors irradiated by neutrons are presented to investigate the generation of dark signal spike.

36 MATERIALS SCIENCE↗

DOE OSTI2017

Linear Array Ultrasonic Testing Of A Thick Concrete Specimens For Non-Destructive Evaluation

The University of Minnesota and Oak Ridge National Laboratory are collaborating on the design and construction of a concrete specimen with sufficient reinforcement density and cross-sectional size to represent a light water reactor (LWR) containment wall with various defects. The preliminary analysis of the collected data using extended synthetic aperture focussin technique (SAFT) reconstruction indicated a great potential of the ultrasound array technology for locating relatively shallow distresses. However, the resolution and reliability of the analysis is inversely proportional to the defect depth and the amount of reinforcement between the measurement point and the defect location. The objective of this round of testing is to evaluate repeatability of the obtained reconstructions from measurements with different frequencies as well as to examine the effect of the duration of the sending ultrasound signal on the resulting reconstructions. Two series of testing are performed in this study. The objective of the first series is to evaluate repeatability of the measurements and resulting reconstructed images. The measurements use three center frequencies. Five measurements are performed at each location with and without lifting the device. The analysis of the collected data suggested that a linear array ultrasound system can produce reliably repeatable reconstructions using 50 kHz signals for relatively shallow depths (less than 0.5 m). However, for reconstructions at the greater depths the use of lower frequency and/or signal filtering to reduce the effect of signal noise may be required. The objective of the second series of testing is to obtain measurements with various impulse signal durations. The entire grid on the smooth surface is tested with four different various impulse signal durations. An analysis of the resulting extended SAFT reconstructions suggested that Kirchhoff-based migration leads to easier interpreting reconstructions when shorter duration impulse is used. Longer duration impulses may provide useful information for model-based reconstructions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques

A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming algorithms and codes for both design and safety analysis. In particular, the new generation of system analysis codes aim to embrace several phenomena such as thermo-hydraulic, structural behavior, and system dynamics, as well as uncertainty quantification and sensitivity analyses. The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic approach to uncertainty quantification. Dynamic methodologies in PRA account for possible coupling between triggered or stochastic events through explicit consideration of the time element in system evolution, often through the use of dynamic system models (simulators). They are usually needed when the system has more than one failure mode, control loops, and/or hardware/process/software/human interaction. Dynamic methodologies are also capable of modeling the consequences of epistemic and aleatory uncertainties. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. The major challenges in using MC and DET methodologies (as well as other dynamic methodologies) are the heavier computational and memory requirements compared to the classical ET analysis. This is due to the fact that each branch generated can contain time evolutions of a large number of variables (about 50,000 data channels are typically present in RELAP) and a large number of scenarios can be generated from a single initiating event (possibly on the order of hundreds or even thousands). Such large amounts of information are usually very difficult to organize in order to identify the main trends in scenario evolutions and the main risk contributors for each initiating event. This report aims to improve Dynamic PRA methodologies by tackling the two challenges mentioned above using: 1) adaptive sampling techniques to reduce computational cost of the analysis and 2) topology-based methodologies to interactively visualize multidimensional data and extract risk-informed insights. Regarding item 1) we employ learning algorithms that aim to infer/predict simulation outcome and decide the coordinate in the input space of the next sample that maximize the amount of information that can be gained from it. Such methodologies can be used to both explore and exploit the input space. The later one is especially used for safety analysis scopes to focus samples along the limit surface, i.e. the boundaries in the input space between system failure and system success. Regarding item 2) we present a software tool that is designed to analyze multi-dimensional data. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Unpacking model inadequacy: The quantification of silver release from TRISO fuel by considering empirical and mechanistic approaches

Increasing adoption of the proposed tristructural isotropic (TRISO) particle fuel for both advanced and existing reactors makes it critical to assess and address any uncertainties and inadequacies of TRISO fission product release models. Model inadequacy stems from simplifications made to the computational model when compared to the experiments. The modeling and simulation efforts conducted using the BISON fuel performance code, along with the experimental campaigns carried out under the Advanced Gas Reactor Fuel Development and Qualification Program, afford a unique opportunity to conduct a rigorous modeling inadequacy assessment within the Bayesian uncertainty quantification (UQ) framework. Here, this study compares the standard Bayesian framework against the Kennedy-O'Hagan (KOH) framework, which explicitly represents modeling inadequacy, in regard to UQ for TRISO silver release models. For this purpose, both the traditional Arrhenius equation fitted to experimental data and the more advanced lower-length-scale (LLS)-informed model, which considers microstructure information, are independently considered. Applying the inverse UQ process on the AGR-2 and -3/4 datasets revealed modeling inadequacy to be the most dominant source of uncertainty. Experimental noise uncertainty is also significant; however, model parameter uncertainty can be considered negligible. Interestingly, both the Arrhenius equation and the LLS-informed model demonstrated similar levels of modeling inadequacy. For the forward predictive UQ, the KOH framework improved both the accuracy and quality of quantified uncertainties in comparison to the standard Bayesian framework. This is true for both the Arrhenius equation and the LLS-informed model. In comparing these modeling approaches, both demonstrated similar performance at the engineering scale, while the LLS-informed model expectedly outperformed the Arrhenius equation at the mesoscale. These conclusions highlight the importance of explicitly accounting for modeling inadequacy in the UQ process, and reinforce the need for continuous refinement of physics-based models in order to address the modeling inadequacy.

Advanced reactors↗

DOE OSTI2021

Aftershock Identification Using a Paired Neural Network Applied to Constructed Data

This report is intended to detail the findings of our investigation of the applicability of machine learning to the task of aftershock identification. The ability to automatically identify nuisance aftershock events to reduce analyst workload when searching for events of interest is an important step in improving nuclear monitoring capabilities and while waveform cross - correlation methods have proven successful, they have limitations (e.g., difficulties with spike artifacts, multiple aftershocks in the same window) that machine learning may be able to overcome. Here we apply a Paired Neural Network (PNN) to a dataset consisting of real, high quality signals added to real seismic noises in order to work with controlled, labeled data and establish a baseline of the PNN's capability to identify aftershocks. We compare to waveform cross - correlation and find that the PNN performs well, outperforming waveform cross - correlation when classifying similar waveform pairs, i.e., aftershocks.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Radiation Testing for High-Resolution Radiation-Hardened Camera System (Final CRADA Report)

The nuclear energy industry needs higher-image quality and higher rad-hard vision systems for refueling and inspection operations that are required every 12-18 months for all operating commercial reactors. During refueling operations, the serial numbers of fuel assemblies need to be visually verified in a challenging high radiation environment with dose rates on the order of 1kGy/hr. at the top of the core (approximately 10 days after shutdown). Vega Wave Systems has developed and built a radiation hard vision system that: 1. has high radiation hardness, 2. demonstrated best in the industry image quality, 3. has small size and weight, and 4. does not suffer from radiation-induced noise. ANL-1060 (02/13/2023) The proprietary Enduray vision system from Vega Wave Systems, Inc. is designed to withstand more than 400x the radiation level of the Vidicon-based systems, currently the highest radiation tolerant system in use today. The Van De Graaff (VDG) electron accelerator at Argonne National Laboratory is ideally suited for testing of Enduray vision system. VDG provides a broad band (up to 3 MeV) x-ray source that reasonably approximates the energy and spectrum of the gamma radiation expected at the top of a nuclear core 10 days after shutdown (~1kGy/hr.), is easily controlled for variable tests, and will not activate the parts under radiation for ease of timely and close inspection and characterization of the system in case of operational issues.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Screening Fuels for Autoignition with Small-Volume Experiments and Gaussian Process Classification

Partially reacting candidate fuels under highly dilute conditions across a range of temperatures provides a means to classify the candidates based on traditional ignition characteristics using much lower quantities (sub-mL) than the full octane tests. Using a classifier based on a Gaussian Process model, synthetic species profiles obtained by plug flow reactor simulations at seven temperatures are used to demonstrate that the configuration can be used to classify 95% of the samples correctly for autoignition sensitivity exceeding a threshold (S = 8) and 100%of the samples correctly for research octane number exceeding a threshold (RON = 90). Molecular beam mass spectrometry (MBMS) experimental data at four temperatures is then used as the model input in a real-world test. Despite the nontrivial relationship between the MBMS measurements and speciation as well as experimental noise it is still possible to classify 95% of the samples correctly for RON and 85% of the samples correctly for S in a 'leave-one-out' cross validation exercise. The test data set consists of 45 fuels and includes a variety of primary reference fuels, ethanol blends and other oxygenates.

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2024

Performance Test of Mini LVDT - ELVIS

The Institute for Energy (IFE) Technology has been a pioneer in the development of Linear Variable Differential Transformers (LVDTs) for in-pile testing, deploying over 2,200 units in various reactor environments with less than a 10% failure rate after five years of operation. This report focuses on the performance testing of IFE’s Mini LVDT, a compact sensor ideal for material test reactor experiments. The Mini LVDT, with a limited range of +/- 1.5 mm, offers excellent performance comparable to larger LVDTs, making it valuable in space-constrained applications. The development of an Enhanced Linear Variable Intrinsic Sensor (ELVIS) with internal temperature monitoring capabilities represents a significant advancement, addressing the critical need for real-time, accurate measurements in high-radiation and high-temperature environments. Two ELVIS prototypes were evaluated in terms of both temperature and displacement, showcasing promising results, though challenges with noise during temperature measurements were identified. This report summarizes the rigorous testing performed at Idaho National Laboratory (INL) and highlights the potential applications of Mini LVDTs and ELVIS in nuclear and other high-precision industries.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Characterization of the molten salt FMgNaK through ab initio molecular dynamics and experimental density measurements

The selection of a salt species for use in a molten salt reactor MSR is a key part of any MSR design. However, many salts have sparse or no thermophysical property data sets, especially those with higher melting points. One such salt is the eutectic mixture of (NaF)0.345(KF)0.59(MgF 2 )0.065, or FMgNaK, and is explored here through ab initio molecular dynamics (AIMD) simulation (1023-1273K) as well as experimental measurement of the liquid density using the Archimedean method (973-1223K). Predicted densities from AIMD are similar to experimentally-measured densities, though with more noise, suggesting that the small scale necessitated by AIMD simulations may be problematic for simulating FMgNaK in particular. Here the coefficient of thermal expansion is predicted from simulation, and salt structure is characterized. Mg-F-Mg chaining is observed in the salt network, though the low concentration of Mg inhibits chaining on the scale that is observed in (LiF) 0.67 (BeF 2 ) 0.33 , or FLiBe.

36 MATERIALS SCIENCE↗

DOE OSTI2019

A new method for in situ structural investigations of nano-sized amorphous and crystalline materials using mixed-flow reactors

Structural investigations of amorphous and nanocrystalline phases forming in solution are historically challenging. Few methods are capable of in situ atomic structural analysis and rigorous control of the system. A mixed-flow reactor (MFR) is used for total X-ray scattering experiments to examine the short- and long-range structure of phases in situ with pair distribution function (PDF) analysis. The adaptable experimental setup enables data collection for a range of different system chemistries, initial supersaturations and residence times. The age of the sample during analysis is controlled by adjusting the flow rate. Faster rates allow for younger samples to be examined, but if flow is too fast not enough data are acquired to average out excess signal noise. Slower flow rates form older samples, but at very slow speeds particles settle and block flow, clogging the system. Proper background collection and subtraction is critical for data optimization. Overall, this MFR method is an ideal scheme for analyzing the in situ structures of phases that form during crystal growth in solution. As a proof of concept, high-resolution total X-ray scattering data of amorphous and crystalline calcium phosphates and amorphous calcium carbonate were collected for PDF analysis.

36 MATERIALS SCIENCE↗

DOE OSTI2018

A combined phase contrast imaging and heterodyne interferometer system for multiscale fluctuation measurements (invited)

In this work, a novel combined diagnostic capable of measuring multiscale density fluctuations that extend from magnetohydrodynamic (MHD) to the lower range of electron temperature gradient turbulence has been designed, installed, and operated at DIII-D. The combined diagnostic was constructed by adding a heterodyne interferometer to the pre-existing phase contrast imaging (PCI) system, both of which measure line-integrated electron density fluctuations. The port-space footprint is minimized via use of a single 10.6 μm probe beam. With temporal bandwidths in excess of 1 MHz, the PCI measures high-k (1.5 cm -1 < |k R | ≤ 25 cm -1 ) fluctuations with sensitivity 3×10 13 m -2 /$\sqrt{kHz}$, while the interferometer simultaneously measures low-k (|k R | < 5 cm -1 ) fluctuations with sensitivity 3×10 14 m -2 /$\sqrt{kHz}$. The intentional mid-k overlap has been empirically verified with sound-wave calibrations and allows quantitative investigation of multiscale effects that are predicted to be significant in the reactor-relevant T e ~ T i regime. Lastly, via correlation with the primary DIII-D interferometer, the toroidal mode numbers of core-localized MHD can be measured.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

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↗