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Moore, Michael E.

Publications and source records attributed to Moore, Michael E..

6 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Applied radiation tags for warhead dismantlement transparency

Here, this study explores the feasibility of applying radiation tags as a Chain of Custody tool to confirm the dismantlement of individual warheads by investigating time dependent emissions from special nuclear material. Precedents of photon and neutron sources that have been used to irradiate warhead components are presented. Measurements of irradiated Highly Enriched Uranium (HEU) are used to benchmark simulations with the CINDER90 library for MCNP. Delayed gamma emissions are simulated as a function of time since irradiation of moderated HEU and Weapons Grade Plutonium with photon and neutron sources. Models of commercial 9 MV and 15 MV linacs were found to induce ~10 9 fissions/g and ~10 11 fissions/g, respectively, creating 3 to 4 orders of magnitude more delayed gammas than neutron sources, modeled at 10 10 n/s. However, a DT neutron source was found to induce sufficient fissions, ~10 7 fissions/g, to create quantifiable signatures relative to intrinsic emissions. Applied radiation tagging is determined to be technically feasible, but its operational context and practical implementation requires significant development.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

Neutron resonance transmission analysis prototype system for thorium fuel cycle safeguards

Emerging thorium-based reactor designs and fuel cycles present challenges to traditional non-destructive assay techniques used in international safeguards. Specifically, assaying the masses of 233 U and 235 U when they are present together in samples with high gamma ray backgrounds is difficult because of similar passive neutron signatures and relatively weak gamma-ray emissions of 233 U. The Pacific Northwest National Laboratory (PNNL) and the Massachusetts Institute of Technology (MIT) are developing a compact neutron resonance transmission analysis (NRTA) system as one potential solution to these challenges. The NRTA technique provides isotopic information for a sample via neutron time-of-flight (TOF) measurements that exploit a sample’s epithermal neutron resonance cross-sections. A recently developed portable NRTA system uses a commercially available, pulsed deuterium-tritium neutron generator with a ~2 m flight path and a GS20 lithium glass scintillator detector. Finally, this paper describes the prototype NRTA system design, a refined radiation transport model of the system, preliminary measurements with thorium and uranium sources, and demonstration of a quantitative isotopic estimation algorithm.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

New NDA Methods for Thorium Fuel Cycle Safeguards (Mid-Project Report)

This project is developing new Non-Destructive Assay (NDA) safeguards techniques for emergent thorium fuel cycles based on Neutron Resonance Transmission Analysis (NRTA), which can assay 233 U and 235 U when they are present together in a sample with potentially high gamma-ray backgrounds from fission products and 232 U. Existing passive techniques face large challenges for this task so new active interrogation methods are needed. This effort is also exploring how gamma-ray signatures can complement NRTA for enhanced assayed performance. This project leverages an NRTA system being developed at PNNL in collaboration with MIT, which uses a commercially available deuterium-tritium neutron generator at short standoff (~2 m). We aim to assess the feasibility and performance of these new NDA techniques for the range of relevant samples in thorium fuel cycles. Key advancements described in this mid-project report include NRTA system design for thorium safeguards measurements, characterization of 233 U oxide powder at PNNL, a survey of detector technologies suitable for NRTA in high gamma background environments, successful preliminary demonstration of a quantitative isotopic estimation algorithm, and first NRTA measurements of a thorium sample.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Configuration and impurity quantification of AmLi sources using radiography and gamma spectroscopy

This paper presents the non-destructive assay of two AmLi sources used for detector characterization and active interrogation at PNNL, identified as MRC-101 and MRC-103. First, a detailed description of the internal configuration of 2724-BT encapsulated MRC sources is established. X-ray radiographs of the AmLi sources are reported for the first time to verify the dimensions and orientation of the encapsulation. Notably, a density variation was observed as a contrast change across the height of the source cavities. Contrary to commonly made assumptions in literature, it is indirectly observed that the AmLi source/target material only occupies a portion of the source cavity. The intensity of the prompt gamma-rays resulting from the inelastic scatter of alpha particles on 7Li was used to calculate the mass ratios of LiH:AmO2. The gamma-ray spectra revealed the presence of numerous impurities in both source and target material (243Am, 237Np, 154Eu, 23Na, and 9Be). Neutron-gamma production ratios were applied to the intensity of the Doppler-broadened peaks from alpha-induced reactions with 7Li, 9Be, and 23Na to calculate neutron rates for both sources. A neutron/alpha production yield (derived in this work) was then related to the atomic concentration of these isotopes to find the approximate masses within each source. Ultimately, the AmLi source density was estimated with prompt gamma analysis based on partial source volumes as 0.45±0.08 g/cm^3.

DTRA, AmLi, neutron↗
DOE OSTI2021

Characterization of Two Extraordinary AmLi Neutron Sources

Two legacy AmLi sources, MRC-101 and MRC-103, are routinely used for neutron detector characterization and testing campaigns that Pacific Northwest National Laboratory manages for the Defense Threat Reduction Agency within the U.S. Department of Defense. However, applicability of source measurement data has been substantially limited due to the lack of original manufacturer documentation and cross comparison to other sources. This paper summarizes results of the measurement campaign conducted to better quantify source radionuclide content, neutron energy spectrum and emission rate of the AmLi sources. The source characterization study consisted of three major consecutive measurement exercises. First, gamma spectrometry measurements using a high purity germanium detector were used to estimate the radionuclide content and potential impurities of the AmLi sources. Then neutron spectral measurements were taken to determine the unique neutron energy distribution. Finally, assessments of the total neutron yields of the sources were made with measurements in a neutron well counter. The neutron detection efficiency for these measurements was determined via detailed Monte Carlo modeling based on the neutron energy distribution measured with the rotating neutron spectrometer. The most exciting and unexpected finding of the study was a significant contamination of the source material with beryllium, contributing up to 17% to the total neutron emission rate and correspondingly changing the neutron energy spectrum. Also, the obtained lithium-only portion of the measured energy distribution of the sources was compared against several AmLi spectra previously reported by different researchers.

AWCC, AmLi, beryllium contamination, neutron spect↗