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Enqvist, Andreas

Publications and source records attributed to Enqvist, Andreas.

At least 19 records

ResNet and CycleGAN for pulse shape discrimination of He-4 detector pulses: Recovering pulses conventional algorithms fail to label unanimously

Pulse shape discrimination (PSD) capable detectors, such as He-4, that respond to neutron and gamma-ray 7 interactions have a threshold deposited energy value below which n/γ discrimination vanishes when using 8 conventional PSD algorithms. Recent attempts in applying supervised learning based artificial neural 9 networks for PSD use the pulses in the separated regions to train the networks so they can be used to classify 10 another set of separated pulses. In doing so, pulses previously indistinguishable are not recovered for 11 classification, which would have increased the number of neutron and gamma-ray pulses that could be used 12 for further analysis. Assuming the reason why conventional PSD algorithms have unseparated regions is 13 because the parameter space of the algorithms fail to capture the intrinsic (but subtle) distinguishing 14 behavior of some of the neutron and gamma-ray pulses, a cycle-consistent generative adversarial network 15 (CycleGAN) was trained to amplify those differences and extract well separated neutron and gamma-ray 16 clusters. Results show that, once the network is trained with pulses from separated and unseparated regions, 17 it was able to transform the pulses in the unseparated region to improve the PSD. Subsequent n/γ 18 classification was performed using deep residual network (ResNet) that takes pulses with 512 data points 19 as an input. Two different ResNets were explored – simple ResNet and modified ResNet which takes 20 segmented pulse inputs in the first layer and the corresponding time axis values in the last hidden layer. 21 The later approach enables the network to extract time correlated pulse features to enhance its ability to 22 capture the pulse behaviors relevant for PSD. Although it achieves slightly lower accuracy, 99.41% versus 23 99.89%, based on simply counting the number of correct n/γ labels assigned, compared to the simple 24 ResNet, the modified ResNets architecture was able to decreases the cross-entropy loss function by half, 25 which implies that the correct n/γ labels assigned are less likely to be accidental. PSD parameter 26 distributions based on n/γ classification by ResNet before and after transforming unseparated pulses using 27 CycleGAN show that by enhancing the separation between neutrons and gamma-rays, the transformation 28 helps improve the performance of classifier networks that are trained using labeled dataset. The 29 enhancement of neutron and gamma-ray separation by the CycleGAN increased the PSD figure of merit 30 (FOM) by up to 70% in some regions. Here, the results show that, if a given detector achieves clear separation 31 between neutron and gamma-ray pulses in any energy region, such neural network approaches can help 32 lower the energy threshold for the separation and increasing the number of neutron and gamma-ray pulses 33 that can be used for further analysis.

4He↗

Gamma-ray spectrum variations for surface measurements of uranium hexafluoride cylinders

Here, this work aims to characterize features that contribute to the spectral variations that appear in gamma-23 ray spectra of filled uranium hexafluoride (UF 6 ) cylinders. While the enrichment of the UF 6 in a 24 cylinder can greatly affect gamma-ray spectra at peaks such as the 185.7 keV peak and be used to 25 verify enrichment, many other characteristics of a filled UF 6 cylinder can affect the spectra. These 26 characteristics vary on a per-cylinder basis and can include: the time between when a cylinder was 27 filled and when it was measured; the previous fill history of a cylinder; how the cylinder was most 28 recently filled; the UF 6 geometric distribution within the cylinder; and the measurement location 29 along the cylinder. This work explores spectral variations seen in measurement data and uses 30 validated simulation models to find trends in these uncertainties. These trends are analyzed to observe 31 how each parameter can contribute to the overall uncertainty of UF 6 spectral measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Applications and Deployment of Neutron Scatter Cameras in Nuclear Safeguards Scenarios

Neutron scatter cameras (NSCs) are a type of directionally sensitive neutron detector that rely on two consecutive neutron scattering events to localize a source of neutrons. NSCs can be used to locate, image, and identify unknown neutron sources or verify the geometry and identity of known sources. Much technical progress has been made in improving NSC designs, but little literature exists exploring the full range of practical application of neutron scatter cameras. This paper seeks to identify scenarios related to nuclear security and non-proliferation where deployment of NSCs may be useful. These situations could include: limited searches for sources during cargo screening, counting nuclear warheads for treaty verification, verification of special nuclear material during inspections, imaging nuclear contamination, imaging nuclear reactor cores, searching for lost sources, and matching neutron images in shipper/receiver or inventory management scenarios. These scenarios are examined with respect to existing NSC designs and proposed designs in terms of usefulness and practicality of deployment when compared to currently used detection systems.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Dry cask radiation shielding validation and estimation of cask surface dose rate with MAVRIC during long-term storage

The dry storage cask became a semi-permanent solution for spent nuclear fuel since no permanent solution has been finalized over the past three decades. Considering the difficulty of deep geological repository projects in many aspects, prolonging the service life of dry casks longer than a few decades has become an option for spent nuclear fuel storage. It must be ensured that dry cask safety objectives, including radiation shielding, criticality safety, structure health, decay heat removal, and containment be fulfilled during the prolonged service life. This paper presents numerical modeling results of the radiation dose rate distribution on a TN-32 cask surface aged up to 300 years. The numerical modeling method was validated with measurement data, and the difference between the calculated results and measurement data in the total dose rate is 7%. The detailed radiation distribution variations over cask surfaces along the axial and radial directions for 300 years of storage were obtained, and their azimuthal angular dependence was established. The maximum total dose rate over the cask surfaces is less than 10 mrem/hour (9.258 ± 0.212 mrem/hour) after 40 years, and this outcome fulfills all dose rate requirements from 10 CFR 71.47(b). After 300 years of storage, the average total dose rates on the side surface and top surface are 0.150 ± 0.007mremmrem/hour and 0.057 ± 0.003mremmrem/hour, respectively, while the maxima are 0.601 ± 0.014 mrem/hour and 0.112 ± 0.006 mrem/hour, respectively. The variations in the fuel gamma ray, structure gamma ray, neutron and secondary gamma-ray contributions to the total dose rate are presented and analyzed in this work. The neutron and gamma-ray dose rate contributions by energy group are presented and analyzed at the cask top, mid-plane, and bottom at the selected storage times.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Used Fuel Storage Monitoring Using Helium-4 Scintillation Fast Neutron Detectors and Neutron Spectral Analysis

This project aimed to address the technology gap in the reliable methods to verify contents of sealed dry storage casks for spent nuclear fuel. Helium-4 gas scintillation neutron detectors was proposed to be developed into a neutron spectroscopy and imaging system for nuclear security and safeguards applications. 4 He gas is significantly more available than 3 He gas and it has unique advantages such as low gamma interaction probability and direct detection of fast neutrons. Due to the sheer amount of fissile materials in a single dry cask, there is a significant neutron fluence emitted whose energies range from thermal to uncollided high-energy neutrons suitable for the detection capability of the 4 He detectors. Our project demonstrated the first ever neutron spectroscopic capabilities using 4 He detectors, explored novel detector design and performed detector characterization. It further qualified the detectors for use in various applications including high-flux environments with significant gamma-fluence. Computer simulations were used to test various large cask measurement system designs and tests, beyond the compact prototype system and laboratory measurements exploring the feasibility of the neutron spectroscopic approach. Although some cask-characterization methods proved unfeasible, alternative characterization methods using the He-4 detectors were explored and generated promising results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Neutron spectroscopy and spectral unfolding with 4 He fast neutron scintillators

Time-of-flight measurements were performed to determine the 4 He pressurized fast neutron scintillation detectors response function and the ability of the detectors to be applied as a spectrometer. An experimentally validated neutron energy response function was used in conjunction with an iterative least squares unfolding technique to unfold a 252 Cf spontaneous fission spectrum, a PuBe (α, n) spectrum, and a deuterium-deuterium (D-D) fusion reaction spectrum. In conclusion, the ability of the detectors to retain incident neutron energies makes them suitable for applications such as nuclear non-proliferation and spent nuclear fuel monitoring.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Verification and validation of atmospheric transport models with the UF Training Reactor

The University of Florida Training Reactor (UFTR) is a 100-kW ARGONAUT type reactor with a unique design such that a large volume of air passes through the core, producing and releasing significant quantities of 41 Ar. Once released, the radioargon is dispersed throughout the atmosphere and decays by a 1294 keV gamma ray emission. This quantified, validated, and controlled 41 Ar plume can be utilized to verify and validate atmospheric transport models (ATMs). This is accomplished by coupling an ATM with MCNP and comparing the results to experimental measurements. For this study, an in-house developed Gaussian dispersion model and the EPA approved Gaussian model AERMOD were coupled with MCNP to simulate the results of a six hour measurement taken on May 8th, 2018. For the measurements and simulations, a grid of seven NaI detectors were placed in four different rooftop locations near the UFTR plume stack. With easterly winds dominating during the morning and early afternoon, both the measurement results and simulations displayed a higher 41 Ar photon count rate on the detectors west of the plume stack throughout the day. Overall, the AERMOD simulations produced better results when compared to the in-house model, particularly for the detectors placed east and west of the plume stack. While some of the AERMOD data showed a strong correlation with the measurements, the results remain inconsistent. Here, due to the varying accuracy of the simulation results, it is theorized that the relationship between measurement results and simulations will improve with the use of localized weather information or the utilization of non-Gaussian dispersion models.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

A comparison of Monte Carlo fission models for safeguards neutron coincidence counters

Monte Carlo simulation is a powerful tool used to model neutron coincidence detectors for international safeguards. The simulation has typically sampled properties such as the fission neutron multiplicity, energy, and direction, from independent probability density functions. However, multiplicity counters detect event-based neutron correlations and thus more accurate fission event modeling is needed. To respond to this need, the Fission Reaction Event Yield Algorithm (FREYA) and the Cascading Gamma-ray Multiplicity with Fission (CGMF) models were added in the newest version of MCNP, MCNP6.2. The models simulate individual fission events conserving momentum, energy, and angular momentum such that correlated particles are emitted. The effects of the new models on simulations of safeguards neutron coincidence counters were studied and compared to standard MCNP simulations. The MCNPX-PoliMi model was also included in the comparison. The properties of fission neutrons from safeguards relevant isotopes were compared to literature references. Then a hypothetical simplified detector was modeled to isolate the effects of specific differences between models. Experimental measurements from previous work were modeled and agreements were compared. Lastly, the probabilities of correlated events occurring in the experimental measurements were calculated with the different models. For example, the probability was calculated of detecting neutrons from both induced fission in uranium and spontaneous fission of Cf-252 in the same fission chain.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DE-NE0008277_PROTEUS final technical report 2018

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce such a comprehensive suite of experimental data to support any future GCFR endeavors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Radiation dose rate distributions of spent fuel dry casks estimated with MAVRIC based on detailed geometry and continuous-energy models

This study presents a detailed comparison of the dose rate distributions of a dry TN–32 fuel cask with two geometry models and two cross-sectional datasets. The accuracies of radiation dose rate estimation and computational efficiencies of each geometry model with two cross-sectional data-sets are compared. The use of automated variance reduction techniques can significantly improve the computational efficiency of such a realistic, deep penetration problem that involves radiation transport from different volumetric sources, thereby eliciting only a small statistical error. Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) is a computational sequence within the SCALE 6.2 code package based on consistent adjoint driven importance sampling (CADIS), a type of automated variance reduction technique. Homogenous and full fuel assembly models are built herein, and two nuclear cross-section libraries (V7–200N47G and continuous energy) are applied in this work. Based on the detailed comparisons, we found that neutron dose rate estimation is more dependent on geometry modeling than on cross-section data. For neutron-induced gamma rays, the dose rate distribution depends on both the spatial self-shielding effect and the cross-section library. The primary gamma rays respectively contribute to the total dose rate by ~91% and ~ 99% on side and top surfaces, and the dose rate accuracy is more dependent on the cross-section library than on geometry modeling. In terms of the computation efficiency and efforts spent on geometry modeling, the homogenous fuel assembly model with the MG library can produce an acceptable dose rate distribution, but the detailed fuel assembly model with the continuous-energy library is required for more precise dose rate estimation.

61 RADIATION PROTECTION AND DOSIMETRY↗

Data fusion for a vision-aided radiological detection system: Calibration algorithm performance

In order to improve the ability to detect, locate, track and identify nuclear/radiological threats, the University of Florida nuclear detection community has teamed up with the 3D vision community to collaborate on a low cost data fusion system. The key is to develop an algorithm to fuse the data from multiple radiological and 3D vision sensors as one system. The system under development at the University of Florida is being assessed with various types of radiological detectors and widely available visual sensors. A series of experiments were devised utilizing two EJ-309 liquid organic scintillation detectors (one primary and one secondary), a Microsoft Kinect for Windows v2 sensor and a Velodyne HDL-32E High Definition LiDAR Sensor which is a highly sensitive vision sensor primarily used to generate data for self-driving cars. Each experiment consisted of 27 static measurements of a source arranged in a cube with three different distances in each dimension. The source used was Cf-252. The calibration algorithm developed is utilized to calibrate the relative 3D-location of the two different types of sensors without need to measure it by hand; thus, preventing operator manipulation and human errors. The algorithm can also account for the facility dependent deviation from ideal data fusion correlation. Use of the vision sensor to determine the location of a sensor would also limit the possible locations and it does not allow for room dependence (facility dependent deviation) to generate a detector pseudo-location to be used for data analysis later. Using manually measured source location data, our algorithm-predicted the offset detector location within an average of 20 cm calibration-difference to its actual location. Calibration-difference is the Euclidean distance from the algorithm predicted detector location to the measured detector location. The Kinect vision sensor data produced an average calibration-difference of 35 cm and the HDL-32E produced an average calibration-difference of 22 cm. Using NaI and He-3 detectors in place of the EJ-309, the calibration-difference was 52 cm for NaI and 75 cm for He-3. In conclusion, the algorithm is not detector dependent; however, from these results it was determined that detector dependent adjustments are required.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on “fingerprinting” casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17 x 17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4 He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244 Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

High energy neutron transmission analysis of dry cask storage

Since the U.S. currently only approves of storing used nuclear fuel in pools or dry casks, the demand for dry cask storage is on the rise due to the continuous operation of currently existing nuclear plants which are reaching or have reached the capacity of their used fuel pools. With the rising demand comes additional pressure to ensure the integrity of dry cask systems. Visual inspection is costly and man-power intensive, so alternative nondestructive testing techniques are desired to insure the continued safe and effective storage of fuel. One such approach being investigated by the University of Florida is neutron based computed tomography. Simulations in MCNP are preformed where D-T energy neutrons are transmitted through the dry cask and measured on the opposite side. If the transmitted signal is clear enough, the interior of the cask can be reconstructed from the measurement of the alterations of neutron signal intensity using standard mathematical techniques developed for medical imaging. Preliminary efforts show a correlation between energy and number of scatters (which is an indication of retention of position information). Work is ongoing to quantify if the correlation is strong enough that an energy discriminator may be used as a filter in future image reconstruction. The calculated transmission probability suggests that an image could be reconstructed with a week of scanning.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Pulse shape discrimination of Cs 2 LiYCl 6 :Ce 3+ detectors at high count rate based on triangular and trapezoidal filters

Cs 2 LiYCl 6 :Ce 3+ (CLYC) detectors have demonstrated the capability to simultaneously detect γ-rays and thermal and fast neutrons with medium energy resolution, reasonable detection efficiency, and substantially high pulse shape discrimination performance. In this work, a disadvantage of CLYC detectors is the long scintillation decay times, which causes pulse pile-up at moderate input count rate. Pulse processing algorithms were developed based on triangular and trapezoidal filters to discriminate between neutrons and γ-rays at high count rate. The algorithms were first tested using low-rate data. They exhibit a pulse-shape discrimination performance comparable to that of the charge comparison method, at low rate. Then, they were evaluated at high count rate. Neutrons and γ-rays were adequately identified with high throughput at rates of up to 375 kcps. The algorithm developed using the triangular filter exhibits discrimination capability marginally higher than that of the trapezoidal filter based algorithm irrespective of low or high rate. The algorithms exhibit low computational complexity and are executable on an FPGA in real-time. Finally, they are also suitable for application to other radiation detectors whose pulses are piled-up at high rate owing to long scintillation decay times.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Variations in AmLi source spectra and their estimation utilizing the 5 Ring Multiplicity Counter

Active-mode assay systems are widely used for the safeguards of uranium items to verify compliance with the Non-Proliferation Treaty. Systems such as the Active-Well Coincidence Counter (AWCC) and the Uranium Neutron Coincidence Collar (UNCL) use americium-lithium (AmLi) neutron sources to induce fissions which are measured to determine the sample mass. These systems have historically relied on calibrations derived from well-defined standards. Recently, restricted access to standards or more difficult measurements have resulted in a reliance on modeling and simulation for the calibration of systems, which introduces potential simulation biases. Furthermore, the AmLi source energy spectra commonly used in the safeguards community do not accurately represent measurement results and the spectrum uncertainty can represent a large contribution to the total modeling uncertainty in active-mode systems.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Measuring the scintillation decay time for different energy deposited by γ-rays and neutrons in a Cs 2 LiYCl 6 :Ce 3+ detector

In nuclear safeguards and homeland security applications, it is greatly beneficial to simultaneously detect γ-rays, thermal neutrons, and fast neutrons using a single detector with reasonable pulse shape discrimination capability, energy resolution comparable with or even better than NaI(Tl) detectors, and high detection efficiency. Cs 2 LiYCl 6 :Ce 3+ detector (CLYC) scintillation detectors have been proven to be one promising candidate to meet these requirements. In this work, the decay time and fraction of each scintillation component for different energy deposition and incident particle type (γ-ray, thermal neutron, and fast neutron) were investigated based on fitting the PMT anode output with exponential functions. For γ-rays, four components were determined with ultrafast decay time of less than one nanosecond and slow time in the order of magnitude of microsecond. It was found that the dependence on the energy deposited by γ-rays of the fraction as well as the decay time of the three slow components was small. However, significant dependence was observed for the ultrafast component. Two or three components were determined for thermal neutrons and fast neutrons without observing a component with fast decay time. To verify the approach used it was first applied to scintillation pulses induced by γ-rays in a NaI(Tl) detector. We found the results were consistent with well-known data already published in the literature.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Improved fission neutron energy discrimination with 4 He detectors through pulse filtering

Our paper presents experimental and computational techniques implemented for 4 He gas scintillation detectorsfor inducedfission neutron detection. Fission neutrons are produced when natural uranium samples are activelyinterrogated by 2.45 MeV deuterium-deuterium fusion reaction neutrons. Fission neutrons of energies greaterthan 2.45 MeV can be distinguished by their different scintillation pulse height spectra since 4 He detectorsretain incident fast neutron energy information. To enable the preferential detection of fast neutrons up to10 MeV and suppress low-energy event counts, the detector photomultiplier gain is lowered and triggerthreshold is increased. Pile-up and other unreliable events due to the interrogating neutronflux and backgroundradiation arefiltered out prior to the evaluation of pulse height spectra. With these problem-specific calibrationsand data processing, the 4 He detector's accuracy at discriminatingfission neutrons up to 10 MeV is improvedand verified with 252 Cf spontaneousfission neutrons. Given the 4 He detector's ability to differentiate fastneutron sources, this proof-of-concept active-interrogation measurement demonstrates the potential of specialnuclear materials detection using a 4 He fast neutron detection system.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗