Gamma-insensitive measurements of high neutron fluxes using solid-state composite scintillators
Not Available
36 MATERIALS SCIENCE↗
Engineering topics
Favalli, Andrea
Publications and source records attributed to Favalli, Andrea.
Solid, structured composite neutron detectors with high dynamic range capability
Neutron detectors are essential across disciplines such as fundamental science, nuclear security, safeguards, and civilian applications. While 3 He-filled gas proportional counters have long been revered for their efficacy in detecting thermal neutrons and praised for their efficiency, neutron/gamma discrimination, and stability, the scarcity of 3He has spurred a search for alternatives. Here, we explore a solid structured scintillating particle composite (SPC) consisting of 6 Li-containing scintillating glass particles within an acrylic matrix as a neutron detector for high dynamic range applications. We show for the first time that an SPC neutron detector can boast an intrinsic detection efficiency of 0.261% for pure 252 Cf fission neutrons and an overall neutron detection efficiency of (0.546 ± 0.003)% at the Neutron Free-in-Air facility while being able to function in an intense gamma-ray environment. We also show that the SPC neutron detector supports fast neutron capture times and enables a dual-readout scheme that extends the detector dynamic range to high incident neutron fluxes. A scalable fabrication process allows for tailoring the SPC detector properties to the requirements of specific applications. Good agreement is found between the experimental results taken with a National Institute of Standards and Technology traceable 252 Cf source and the coupled MCNP6 and optical-ray-tracing simulations.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
252 Cf yield calibration method for nuclear material accountancy and safeguards practitioners
Californium-252 is an isotope commonly encapsulated as a physically small but strong spontaneous fission neutron source for applications in industry, academia, and research. Within the nuclear safeguards community, a well-characterized 252 Cf source is often used as an accessible alternative for plutonium in calibrations measurements, which require the absolute source strength. Current methods of source strength quantification can provide an accurate estimate, approximately 1% uncertainty under optimal conditions, but are typically costly and time consuming. An alternative method was developed to determine 252 Cf source strength using passive neutron correlation counting (PNCC). The PNCC method enables institutions and facilities to perform 252 Cf source calibration measurements in-house using detection systems that are common within the nuclear safeguards community. This novel method was previously successfully demonstrated using neutron multiplicity well counters. However, the previous evaluation utilized detection systems with high neutron detection efficiencies, greater than 60%. The purpose of the current study is to extend the previous evaluation and demonstrate the feasibility of the PNCC 252 Cf calibration method for suite of coincidence counters with lower neutron detection efficiencies, between 15% and 35%, which are more commonly encountered in the field, as well as to provide a step-by-step guide to using the method. The neutron source used in this evaluation was previously certified by the National Institutes of Science and Technology (NIST), which provides a reference for the accuracy of the PNCC method. In conclusion, the present source yield calculations demonstrate excellent agreement with a NIST calibration technique and achieve a standard uncertainty below the target 1%.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
BAS-MS Image Plate Calibration
Abstract not provided.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
The calculation of light element impurity (α, n) yield curves in a PuO 2 matrix and associated specific yield coefficients: Influence of the reaction cross sections
Most of the Pu separated from irradiated commercial nuclear fuel is stored as PuO 2 . The primary quantitative nondestructive measurement technique used to verify the amount of Pu in storage containers is passive neutron correlation counting. An important physical property of the oxide material is the ratio, α, of the rate of (α, n) neutrons produced inside the item to the rate of neutrons produced by spontaneous fission. This ratio influences the precision of the correlated counting method and affects the interpretation of the data because of how it changes both the primary total neutron production rate and the rate of induced fission events taking place inside the item. In addition to the main O(α, n) contribution, additional contributions come from α-particle interactions with light element impurities that are inevitably present. In this work, we calculate specific (α, n) yield coefficients, expressed in units of neutrons per second per gram of α-emitting nuclide per part per million by mass of the specified impurity element distributed in a pure PuO 2 matrix, for some key α-emitting actinides commonly present in reprocessed Pu ( 238–242 Pu+ 241 Am). These coefficients are directly applicable to nuclear safeguards verification work in which the α ratio is often calculated from the Pu-isotopic composition and chemical information obtained by other means. Further, they also provide a convenient up-to-date reference set against which values generated by other methods can be compared. Results are presented for impurities with atomic number from 3 to 17 inclusive, plus K and Fe. In most cases, these coefficients are not expected to change by more than 5%–10% at any time in the future. However, as new data become available, changes as large as 20% may be needed for some targets (e.g., F). The present yield calculations are limited by the general shortage of quality experimental total (α, n) reaction cross section data, which, together with unexplained variation between determinations, means that an objective and coherent evaluation is not possible. The situation is even less satisfactory for the partial differential cross section needed to calculate neutron spectra.
(𝛼, n) reactions↗
High-yield and high-angular-fluence neutron generation from deuterons accelerated by laser-driven collisionless shock
Since a bright collimated neutron source is an essential tool for global security missions and fundamental scientific research. In this paper, we study a compact high-yield and high-angular-fluence neutron source particularly suitable for high-energy neutron applications utilizing the breakup reaction of laser-driven deuterons in a 9 Be converter. The neutron generation scaling from such a reaction is used to guide the choice and optimization of the acceleration process for bulk ions in a low density CD 2 foam. In particular, the collisionless shock acceleration mechanism is exploited with proper choice in the laser and target parameter space to accelerate these ions toward energies above the temperature of the distribution. Particle-in-cell and Monte Carlo simulations are coupled here to investigate this concept and possible adverse effects as well as the contribution from the surface ions accelerated and the optimal converter design. The simulation results indicated that our design can be a practical approach to increase both the neutron yield and angular fluence of laser-driven neutron sources, reaching >10 11 neutron/pulse (or >10 8 neutron/J) and >10 11 neutron/sr (or >10 8 neutron/sr/J) with present-day kJ-class high-power lasers. Such developments will advance fundamental neutron science, high precision radiography, and other global security applications with laser-driven sources.
72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
Software and Files Index
The purpose of this document is to provide an index of the software and files used to generate the results and findings documented in Section 4 of the scoping study report on Radiation Signatures relevant to Thorium Fuel Cycles.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Sensitivity of the simulation of passive neutron emission from UF 6 cylinders to the uncertainties in both 19 F(α,n) energy spectrum and thick target yield of 234 U in UF 6
Interest in safeguards verification measurements using passive thermal neutron counting to assay 235 U content in large 30B UF 6 canisters has grown in recent years. Here, the prohibitively high cost and impracticality of using reference 30B calibration cylinders extensively will likely make accurate simulations of increasing interest. Accuracy of the simulated response will define the confidence in the predicted response and the extent to which simulations can reasonably be relied upon. With 234 U driven 19 F(α, n) reactions being the main neutron source in low enriched UF 6 the uncertainties of the 19 F(α, n) energy spectrum and the thick target yield of 234 U in UF 6 propagate into the uncertainty in the predicted response and represent a major influence of basic nuclear data. Here sensitivity of the simulated total (Singles) and coincidence (Doubles) count rates are assessed for the Passive Neutron Enrichment Meter using six potential 19 F(α, n) neutron energy spectra over a range of enrichments and material distributions. The results indicate that variations in the Singles and Doubles due to simulated (α, n) neutron spectrum are less than 1.5% for this set of simulated neutron spectra, with dependence varying inversely with enrichment. Singles uncertainty is only slightly less than that of the thick target 19 F(α, n) yield corresponding to the primary neutron source, whereas the 19 F(α, n) yield dependence of the Doubles is reduced by the non-negligible 238 U(SF) coincident neutron emissions. Based on available thick target 19 F(α, n) yield estimates the uncertainty is on the order of 5%, establishing this as the main nuclear data limitation when simulating thermal neutron detectors response for 30B UF 6 storage cylinders. Based on these findings, it appears that the measurement and evaluation of the thick target 19 F(α, n) yield for uranium hexafluoride is due.
19F(α,n) neutron spectrum↗
A new generation of uranium coincidence fast neutron collars for assay of LWR fresh fuel assemblies
The active uranium neutron coincidence collar provides a means of non-destructively assaying the fissile linear density of Light Water Reactor fresh fuel assemblies containing low enriched uranium. These neutron collars can operate in two modes: a thermal and a fast mode. In fast mode, a neutron collar has an added cadmium (Cd) liner in the sample cavity of the detector to reduce the impact of the burnable poison (thermal neutron absorber) on the detector signal (doubles). The main advantage for operating in fast mode is a detected signal that is less dependent of the burnable neutron poison content and thus less dependent on facility operator declarations. The drawback is that operating in fast mode requires a longer measurement time (~hour vs tens of minutes for thermal mode) to achieve the statistically needed precision in the measurements. The trend in the modern reactor fuel assemblies is moving to higher burnup by using higher initial enrichment and, consequently, a higher number of burnable poison rods to compensate the initial neutron reactivity. The increase of the burnable poison loading has motivated the development of a new generation of high efficiency fast neutron collars to allow practical measurements in-field by nuclear inspectors. This paper describes the development and performance evaluation of a new generation of neutron collars, for both boiling water reactor (BWR) and pressurized water reactor (PWR) fuels, jointly developed between the US Department of Energy, through Los Alamos National Laboratory, and the Euratom Safeguards Directorate of the European Commission. In this work, we present here calibrations with reference fuel assemblies at Los Alamos National Laboratory as well as the results of in-field measurement campaigns in fuel fabrication plants with modern commercial fuel assemblies. The experimental results show that a typical PWR verification can be made in a total time of 30 min with an uncertainty in the measured mass of 2% at one standard deviation (1σ). A BWR verification can be made in 47 min with an uncertainty in the measured mass of 1.9% at 1σ, or a total time of 20 min with 1σ uncertainty in the measured mass of 2.5%.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Safeguards Technology for Thorium Fuel Cycles: Research and Development Needs Assessment and Recommendations
This report presents Safeguards Technology for Thorium Fuel Cycles: Research and Development Needs Assessment and Recommendations prepared for the National Nuclear Security Administration (NNSA) Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D) Safeguards Program by a multilaboratory team from Oak Ridge National Laboratory, Los Alamos National Laboratory, and Y-12 National Security Complex. It documents key findings of a 2-year scoping study on “Safeguards Technology Needs Assessment for Leading Thorium Fuel Cycles” (project OR18-V-SG Tec Needs Th Fuel Cycles-PD1Lb).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Modernizing the active neutron collar poison rod correction using simulations to enhance PWR fresh fuel verification
The declared inventory of 235 U in fresh, low enriched uranium, nuclear fuel assemblies is routinely verified by nuclear safeguards inspectorates using the UNCL (Uranium Neutron Collar – Light Water Reactor Fuel) nondestructive assay instrument. The trend in modern fuels is towards higher initial enrichment, which in turn requires a larger number of burnable poison pins with higher Gadolinia concentration to hold down the initial reactivity. UNCL assay error for modern fuel assemblies typically exceeds 10%. The traditional algorithm used to correct the response for burnable poison content needs revision to achieve accuracy comparable to the $\sigma_\text{R}$ of 4.5% expected for poison-free assemblies. Here we review the problem and use a large set of 287 Monte Carlo simulations based on an experimentally benchmarked model corresponding to the 16×16 PWR array used in Brazilian Angra type II and III fuel. Simulated relative responses are used to evaluate the functional form of the poison correction and determine parameters optimal for 16×16 PWR assemblies. This update is conducted using between 4–24 Gadolinia pins ranging 2–11 wt% Gd 2 O 3 with assemblies having mean enrichments ranging 2.5%–5%. Experimental validation of the updated coefficients using seven measurements shows bias is reduced by an order of magnitude and $\sigma_\text{R}$R reduced from 10% to less than 2%. Two updated sets of coefficients are provided, the most accurate of which is directly useable in the existing analysis code (INCC) in use by inspectorate at fuel fabrication facilities.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Sensitivity of the active neutron coincidence collar response during simulated and experimental fresh fuel assay
Verification of the fissile (and fertile) content in fresh nuclear fuel assemblies is conducted by the IAEA to enforce the Nuclear Non-Proliferation Treaty using the UNCL (Uranium Neutron Collar — Light Water Reactor Fuel). The UNCL uses an uncorrelated AmLi neutron source to interrogate the fuel, producing a signal of coincident fission neutrons (Doubles rate) used to assay 235 U content of the fuel. The cost of producing calibration assemblies and limited availability of diverse commercial assemblies at any one time historically restricted the ability to explore the full parameter space experimentally. Monte Carlo simulations can overcome this, but introduce additional sources of uncertainty. In this work, the sensitivity of simulations and measurements to various parameters is assessed for a reference 1616 PWR assembly of uniform 3.2% enrichment. Uncertainty contributions in this evaluation include: simulated AmLi neutron source spectrum, AmLi neutron emission rate, AmLi anisotropicity, high density polyethylene (HDPE) density, the precise position of the fuel assembly within the detector, and experimentally the statistical uncertainty. The overall total systematic uncertainty estimate for the simulation of the absolute/relative Doubles rates responses are estimated to be approximately 2.0%/1.5%, and for experimental measurements systematic uncertainty reduces to 1.1%. This analysis supports further work using the relative Doubles rates in place of measurements for updating and extending the UNCL analysis methodology as systematic uncertainty is reasonably small.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Strengthening Technical Safeguards of Fresh Fuel Through International Cooperation
Over time, nuclear fuel designs have shifted towards having higher initial enrichments and a greater number of burnable poison rods. This enables increased burnup for commercial reasons. These changes have made safeguards measurements of the 235 U content in modern fresh fuel more challenging.This is addressed here through a re-evaluation of the UNCL (Uranium Neutron Collar–Light Water Reactor Fuel)poison rod correction factor. Coefficients of the poison rod correction factor were updated by simulating response of AngraII PWR fuel with a wide range of both burnable poison rods and Gd 2 O 3 contentper rod. Benchmark comparisons are made to experiments performed at Resende (Brazil) and CNEN/LASAL in Brazil. By updating these coefficients, while maintaining the same mathematical form to allow INCC to be used without software modification,it was possible to reduce error in experimental 235 U linear density assay by an order of magnitude.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Spent fuel nondestructive assay integrated characterization from active neutron, passive neutron, and passive gamma
Spent nuclear fuel comprises a wide range of irradiated isotopic material compositions, and characterization through nondestructive measurements is beneficial in verifying declared parameters before the fuel is placed in storage, final disposal, and/or reprocessed. This paper discusses results from three nondestructive assay instruments, including passive gamma, passive neutron, and active neutron methods, that measured fifty spent fuel assemblies at the Clab interim storage facility in Sweden. Here, integrated analysis of the measurements from the three different instruments allowed parametric assessments of cooling time, burnup, neutron multiplication, fissile mass, initial enrichment, and decay heat of each individual fuel assembly. Passive gamma measurements were found to be the most beneficial in predicting cooling time, passive neutron for determining burnup, active neutron in estimating initial enrichment, and both passive and total neutron for multiplication correlations. Fissile mass was best estimated using any combination of any two of the instruments such that corrections for isotopic changes in the fuel could be accounted for with the first set of measurements and multiplication of the assembly in the second. In conclusion, the nondestructive assay technologies demonstrated through this effort enhance the characterization of spent nuclear fuel assemblies.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Application of the Approximate Bayesian Computation Algorithm to Gamma-Ray Spectroscopy
Radioisotope identification (RIID) algorithms for gamma-ray spectroscopy aim to infer what isotopes are present and in what amounts in test items. RIID algorithms either use all energy channels in the analysis region or only energy channels in and near identified peaks. Because many RIID algorithms rely on locating peaks and estimating each peak’s net area, peak location and peak area estimation algorithms continue to be developed for gamma-ray spectroscopy. This paper shows that approximate Bayesian computation (ABC) can be effective for peak location and area estimation. Algorithms to locate peaks can be applied to raw or smoothed data, and among several smoothing options, the iterative bias reduction algorithm (IBR) is recommended; the use of IBR with ABC is shown to potentially reduce uncertainty in peak location estimation. Extracted peak locations and areas can then be used as summary statistics in a new ABC-based RIID. ABC allows for easy experimentation with candidate summary statistics such as goodness-of-fit scores and peak areas that are extracted from relatively high dimensional gamma spectra with photopeaks (1024 or more energy channels) consisting of count rates versus energy for a large number of gamma energies.
Burr, Tom↗
Nuclear Data Scoping Studies for Nonproliferation - paper
Explore the source record for details and available documents.
Romano, Catherine↗
Action Sheet 55 Measurement Campaign Summary
Neutron collars have long been used by international inspectorates to determine the 235 U loading of fresh light water reactor (LWR) fuel in nuclear fuel fabrication facilities. Burnable poisons are strong thermal neutron absorbers that are frequently used to increase the lifetime of LWR fuel in a reactor. These poisons have made the traditionally thermal neutron collar measurements challenging because they force inspectors to rely on operator declarations of the burnable poison content of the fuel in order to correct for the decreased count rate caused by the added thermal neutron absorption of the poison. Work has been conducted internationally to try to eradicate the reliance on operator declarations. The Euratom Fast Collar for BWR Fuel (EFCB) is one promising technique to manage this issue.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
EURATOM Fast Collar for BWR (EFCB) Field Calibration Exercise
A neutron collar for the measurement of fresh BWR fuel assemblies (EFCB) has been designed and built by Los Alamos National Laboratory (LANL) and the Euratom Safeguards Directorate (Euratom). The design was similar to a previous collar designed for the measurement of fresh PWR assemblies. A key feature of both of these collars is that they operate in fast mode (preventing the return of thermal neutrons into the assembly by means of a cadmium liner) to reduce the effect of burnable poisons on the measurement. The current report describes the results of the field calibration exercise that was carried out at the ENUSA Fuel Fabrication Plant in Juzbado, Spain from May 28 to May 31, 2019.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗