Measurement and Characterization of Nuclear Material at Idaho National Laboratory

A measurement plan and preliminary Monte Carlo simulations are presented for the investigation of well-defined mixed-oxide fuel pins. Measurement analysis including pulse-height distributions and time-dependent cross-correlation functions will be performed separately for neutrons and gamma rays. The utilization of Monte Carlo particle transport codes, specifically MCNP-PoliMi, is discussed in conjunction with the anticipated measurements. Four EJ-309 liquid scintillation detectors with an accurate pulse timing and digital, offline, optimized pulse-shape discrimination method will be used to prove the dependency of pulse-height distributions, cross-correlation functions, and material multiplicities upon fuel pin composition, fuel pin quantity, and detector geometry. The objective of the measurements and simulations is to identify novel methods for describing mixed-oxide fuel samples by relating measured quantities to fuel characteristics such as criticality, mass quantity, and material composition. This research has applications in nuclear safeguards and nonproliferation

Fast neutron spectrum unfolding for nuclear nonproliferation and safeguards applications

We present new neutron spectrum unfolding results obtained from measurements with Cf-252 and plutonium-oxide sources. The precise knowledge of the neutron energy spectrum provides information about the presence or absence of fissile material and about the characteristics of the material. We used a neutron spectrum unfolding technique based on a modification of the least-squares method. The main innovation is the use of a Krylov subspace iteration which performs better on ill-conditioned systems of linear equations than standard direct-solution methods. The proposed technique performed well in the unfolding of measured neutron pulseheight distributions from a Cf-252 neutron source and from plutonium-oxide samples and could be easily implemented in a portable neutron spectroscopy system for nuclear nonproliferation and safeguard applications

Neutron and gamma-ray cross-correlation measurements of plutonium oxide powder

a b s t r a c t For the first time, measurements of the time-dependent cross-correlation distributions of plutonium oxide have been made separately for neutrons and gamma rays. Six EJ-309 liquid scintillation detectors with a digital, offline pulse shape discrimination and pulse timing method were used to measure five different samples of varying mass and burnup. The number of (neutron, neutron) correlations were selectively analyzed versus plutonium mass and a clear, increasing trend was observed. Additionally, the measurement scenarios were modeled using the MCNP-PoliMi code and good agreement was observed between the measured and simulated cross-correlation functions

Neutron Emission Characteristics of Two Mixed-Oxide Fuels: Simulations and Initial Experiments

Simulations and experiments have been carried out to investigate the neutron emission characteristics of two mixed-oxide (MOX) fuels at Idaho National Laboratory (INL). These activities are part of a project studying advanced instrumentation techniques in support of the U.S. Department of Energy's Fuel Cycle Research and Development program and it's Materials Protection, Accounting, and Control for Transmutation (MPACT) campaign. This analysis used the MCNP-PoliMi Monte Carlo simulation tool to determine the relative strength and energy spectra of the different neutron source terms within these fuels, and then used this data to simulate the detection and measurement of these emissions using an array of liquid scintillator neutron spectrometers. These calculations accounted for neutrons generated from the spontaneous fission of the actinides in the MOX fuel as well as neutrons created via (alpha,n) reactions with oxygen in the MOX fuel. The analysis was carried out to allow for characterization of both neutron energy as well as neutron coincidences between multiple detectors. Coincidences between prompt gamma rays and neutrons were also analyzed. Experiments were performed at INL with the same materials used in the simulations to benchmark and begin validation tests of the simulations. Data was collected in these experiments using an array of four liquid scintillators and a high-speed waveform digitizer. Advanced digital pulse-shape discrimination algorithms were developed and used to collect this data. Results of the simulation and modeling studies are presented together with preliminary results from the experimental campaign

A scintillator‐based approach to monitor secondary neutron production during proton therapy

Peer Reviewedhttp://deepblue.lib.umich.edu/bitstream/2027.42/135048/1/mp3813.pdfhttp://deepblue.lib.umich.edu/bitstream/2027.42/135048/2/mp3813_am.pd

MPACT Fast Neutron Multiplicity System Design Concepts

This report documents work performed by Idaho National Laboratory and the University of Michigan in fiscal year (FY) 2012 to examine design parameters related to the use of fast-neutron multiplicity counting for assaying plutonium for materials protection, accountancy, and control purposes. This project seeks to develop a new type of neutron-measurement-based plutonium assay instrument suited for assaying advanced fuel cycle materials. Some current-concept advanced fuels contain high concentrations of plutonium; some of these concept fuels also contain other fissionable actinides besides plutonium. Because of these attributes the neutron emission rates of these new fuels may be much higher, and more difficult to interpret, than measurements made of plutonium-only materials. Fast neutron multiplicity analysis is one approach for assaying these advanced nuclear fuels. Studies have been performed to assess the conceptual performance capabilities of a fast-neutron multiplicity counter for assaying plutonium. Comparisons have been made to evaluate the potential improvements and benefits of fast-neutron multiplicity analyses versus traditional thermal-neutron counting systems. Fast-neutron instrumentation, using for example an array of liquid scintillators such as EJ-309, have the potential to either a) significantly reduce assay measurement times versus traditional approaches, for comparable measurement precision values, b) significantly improve assay precision values, for measurement durations comparable to current-generation technology, or c) moderating improve both measurement precision and measurement durations versus current-generation technology. Using the MCNPX-PoliMi Monte Carlo simulation code, studies have been performed to assess the doubles-detection efficiency for a variety of counter layouts of cylindrical liquid scintillator detector cells over one, two, and three rows. Ignoring other considerations, the best detector design is the one with the most detecting volume. However, operational limitations guide a) the maximum acceptable size of each detector cell (due to PSD performance and maximum-acceptable per-channel data throughput rates, limited by pulse pile-up and the processing rate of the electronics components of the system) and b) the affordability of a system due to the number of total channels of data to be collected and processed. As a first estimate, it appears that a system comprised of two rows of detectors 5" Ø ? 3" would yield a working prototype system with excellent performance capabilities for assaying Pu-containing items and capable of handling high signal rates likely when measuring items with Pu and other actinides. However, it is still likely that gamma-ray shielding will be needed to reduce the total signal rate in the detectors. As a first step prior to working with these larger-sized detectors, it may be practical to perform scoping studies using small detectors, such as already-on-hand 3" Ø ? 3" detectors

A compact fast-neutron producing target for high resolution cross section measurements

A proper knowledge of neutron cross sections is very important for the operation safety of various nuclear facilities. Reducing uncertainties in the neutron cross sections can lead to an enhanced safety of present and future nuclear power systems. Accurate neutron cross sections also play a relevant role in many other disciplines such as astrophysics, medicine, and security. Therefore it is essential to have at disposal tools to measure the neutron cross sections at required resolution. The measurement accuracy required to extract properly resonance parameters of the cross sections can only be obtained at time-of-flight facilities specially designed to have a high resolution in energy. Among the other neutron TOF facilities available in the world, the bremsstrahlung-based Geel Electron LINear Accelerator (GELINA) facility of the Joint Research Centre of the European Commission in Belgium is the one with the best energy resolution. The main goal of this thesis was to investigate the possibilities to improve even further the capabilities of this neutron data measurement facility. The thesis describes a design study with the purpose to further enhance the quality of the GELINA facility by proposing a new neutron producing target. It is demonstrated that there is a potential for an improved target to allow GELINA users to measure neutron cross sections with even higher accuracy. Therefore an effort was made to optimize the size, shape, and material composition of such a target design in view of the optimal neutron source characteristics, while providing an adequate solution for target cooling. The final design consists of seven 3 x 3 cm2 U-Mo plates with well-defined thicknesses in the direction of the beam. A plate-by-plate optimization has been carried out in order to maximize the neutronics properties of the target, while keeping the maximum plate temperatures reasonably low. A tantalum cladding coats each U-Mo plate to provide the required containment of radioactive material, and to avoid the direct contact of the U-Mo alloy with the Hg coolant. Seven Hg channels located in between the plates cool the target.Applied Science

Measurement and simulation of neutron/gamma-ray cross-correlation functions from spontaneous fission

For the first time, a technique is presented for the measurement of total and separate neutron and gamma-ray cross-correlation functions from a spontaneous fission source. The cross-correlation functions are unique for given material-geometry configuration, thus represent signatures that can be used for the identification of radioactive materials. The measurement technique allows for the collection of fast coincidences within a time window of the order of a few tens of nanoseconds. A digital pulse shape discrimination technique is used, which allows for the accurate acquisition of the coincidences in all particle combinations. Specifically, separate neutron–neutron, neutron–gamma-ray, gamma-ray–neutron, and gamma-ray–gamma-ray coincidences are acquired with two liquid scintillation detectors. The measurements are compared to results obtained with the MCNP-PoliMi code, which simulates neutron and gamma-ray coincidences from a source on an event-by-event basis. This comparison leads to relatively good qualitative agreement.The measurements and simulations of the separate neutron and gamma-ray contributions to the total cross-correlations provide new signatures that can be obtained using existing experimental systems built to accurately identify nuclear materials. This research has direct applications in the areas of nuclear nonproliferation and homeland security

Initial Evaluation for a Combined Neutron and Gamma 1 Ray Multiplicity Counter

Multiplicity counters for neutron assay have been extensively used in materials control and accountability for nonproliferation, and nuclear safe- guards. Typically, neutron coincidence counters are utilized in these fields. In this paper we present a measurement system that makes use not only of neutron (n) multiplicity counting but also of gamma ray (γ) multiplicity counting and the combined higher-order multiples containing both neutrons and gamma rays. The benefit of this approach is in using both particle types available from the sample, leading to a reduction in measurement time when using more measurables. We present measurement results of n, γ, nn, nγ, γγ, nnn, nnγ, nγγ and γγγ multiples emitted by a 252Cf source and a 239Pu-Be19 source. A dual radiation measuring system proposed here could use extra measurables either to improve the statistics when compared to a neutron-only system, or alternatively be used for extended analysis and interpretation of sample parameters.The study presented here provides an answer to the questions regarding the ability to measure both neutron and gamma ray coincidences at once. The results show that the measurements system proposed in this study has a potential to be valuable in the area of nuclear nonproliferation and homeland security