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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

NCERC Capabilities [Slides]

NCERC is our nation’s only generalpurpose critical experiments facility and is one of only a few that remain operational throughout the world. The mission of the National Criticality Experiments Research Center (NCERC) is to conduct experiments and training with critical assemblies and fissionable material at or near criticality in order to explore reactivity phenomena, and to operate the assemblies in the regions from subcritical through delayed critical. One critical assembly, Godiva-IV, is designed to operate above prompt critical.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

DRiFT - Release 1.0.0 Organic Scintillators

DRiFT (a Detector Response Function Toolkit) is LANL-developed software that postprocesses output from the extensively validated radiation transport code, MCNP [1], and generates realistic nuclear instrumentation response. DRiFT is designed to be flexible, enabling users to specify detector type and many experimental settings, as well as accommodating the addition of their own desired features. Although DRiFT development has included scintillator [2], gas [3], and semiconductor features [4], the focus of this release is on organic scintillator and associated capabilities. Organic scintillators are widely used in the areas of nuclear safeguards and nuclear non-proliferation efforts [5, 6]. DRiFT has several diagnostic and detector physics features relevant to detailed scintillator simulations including: tracking source particle information, scintillation light production, the effects of PMT quantum efficiency and gain, and digitizer settings. Users can select responses from many scintillator and PMT types supported natively by DRiFT, or add their own by following the instructions in this document. We acknowledge that DRiFT is under active development, bug reports and general questions and comments should be directed to Madison Andrews, madison@lanl.gov. This manual is divided into four parts: I) An overview of DRiFT, including how to obtain and install the executable, II) A description of the detector physics related to scintillators available, III) a description of more general DRiFT features the user may find useful, and IV) a description of the test suite and examples made available with the code release.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Evaluation of Attila and MCNP computational methods for dose and exposure estimation

Radiation transport calculations are often used to estimate dose or exposure to components and personnel surrounding a radiation source. The sources for these calculations are decaying radionuclides within various nuclear materials. Historically, dose calculations use MCNP (Monte Carlo N-Particle) transport code as the primary particle transport tool without a secondary computational tool to validate the results from the MCNP simulations [1]. The goal of this study is to make an independent check of the Monte Carlo solution from MCNP6 Version 6.2.1 with the discrete ordinates solution from Attila 10.2.0 Beta 3. As an example problem for this study, water-filled, stainless-steel vessels, modeled with an unstructured mesh (UM) with both MCNP and Attila [2], are exposed to 252Cf and 60Co point sources. This report also includes a discussion of the limitations of unstructured mesh in a MCNP calculation.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Report on Field Test at INL Cask Farm of Single Detector Fast Neutron Spent Fuel Cask Verification System

Detecting diversion of spent fuel elements in dry storage casks is challenging due to the thick shielding used in cask construction. Measurements on top of the cask to map the underlying arrangement of the fuel elements and looking for anomalous changes over time has proven difficult to achieve using gamma rays due to the high scattering and attenuation from the thick steel structure, weakening information on the present or absence of fuel bundles. Simulations and laboratory experiments suggest that the high-energy neutron flux (>200 keV) measured directly above each fuel bundle is sufficient to produce a position map that enables detection of the present or absence of fuel bundles, and therefore diversion of a spent fuel bundle. A single-detector spent-fuel monitoring technique based on this principle was development at the Lawrence Livermore National Laboratory (LLNL). The INL Cask Farm in the INTEC technical area at Idaho National Laboratory (INL) offers the capability to test this technique on an MC-10 storage cask which has a distribution of full and empty fuel positions. An experimental test plan for the single-detector verification system was developed in consultation with INL personnel to be completed in FY2021. Due to travel advisories related to COVID-19, the experimental test plan was adapted to enable INL personnel to carry out the measurements in consultation with LLNL personnel following shipment of the LLNL system to INL. Field test measurements of the single detector verification system were successfully carried out at the INL cask farm on September 7-9, 2021. This document summarizes results of the field test.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

The Science of the Bomb

Uranium is a naturally occurring element with wo isotopes of interest – U238 and U235; U235 is useful for its ability to sustain a chain reaction; Chain reactions are the fundamental way that nuclear reactors work, as well as how bombs work; Enrichment of U235 is important for both nuclear energy as well as nuclear weapons; A safety focus is at the center of all nuclear operations in the United States

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Second Target Station (STS) Project: Analysis of Bunker Shielding–Preliminary Design

At the Department of Energy’s Oak Ridge National Laboratory, the Second Target Station (STS) beamline sources for preliminary design have been used to perform a shielding analysis of the bunker. The simulations in this report used the ST11 beamline source, primarily the ST11 neutron source, because that source provides conservative prompt total dose rate estimates for all beamlines. Prompt total dose rates (i.e., neutron plus photon dose rates when the beam is on) were calculated on top of the bunker roof, outside the bunker wall, and in the basement below the bunker and instrument floor. Design changes were made to ensure the prompt total dose rate in these areas does not exceed 0.25 mrem/h. This report documents the required shielding thicknesses to meet this dose rate limit. There is one exception to this: for a combination of open and closed beamlines, the prompt total dose rate outside the bunker across from the closed beamline, where there is less shielding due to the lack of beamline shielding, slightly exceeds 0.25 mrem/h. A future analysis will document the shielding modifications required to reduce the calculated prompt total dose rates for this configuration to less than 0.25 mrem/h once more details regarding the STS high density concrete density and composition are known. Finally, prompt total dose rates inside the bunker were simulated. The results of this simulation show that the majority of prompt total dose rate in the bunker is due to neutrons, and a few percent of the total is caused by secondary photons created by the neutrons. The contribution to prompt total dose rate inside the bunker by the primary photons that enter each beamline from the target/moderator area is at least two orders of magnitude less than the neutron contribution. This means prompt total dose rate contributions due to primary photons inside and outside the bunker can be neglected.

42 ENGINEERING↗
DOE OSTI2022

Materials and Fuels Complex (FY22 – FY26 Five-Year Investment Strategy)

The Department of Energy Office of Nuclear Energy (DOE-NE) vision is to “Advance nuclear energy science and technology to meet U.S. energy, environmental, and economic needs.” The Materials and Fuels Complex (MFC) serves as the foundation of a nuclear RD&D enabling test bed at Idaho National Laboratory (INL) and is an integral part of a National Reactor Innovation Center (NRIC) strategy. MFC facilities focus on developing and maintaining RD&D capabilities that can increase research throughput, reduce barriers to deployment, and facilitate commercialization of new ideas and technologies for clean and secure sources of energy.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Characterization of B255 R183 C-Cell Cobalt-60 Irradiator Pool

Characterization of the Radiation Calibration and Spectroscopy Laboratory (RCS) Co-60 source radiation pool was successfully completed. The characterization was done by performing several measurements of dose rate in the center of the Co-60 source radiation pool and at multiple heights from the bottom of the pool. The measurements were taken between November 2018 through January 2019. Opti-Chromic FWT-70-40M dosimeters were used for this measurement. A 7.3 cm inner diameter stainless steel cylinder (vessel) was used to lower the dosimeters to the bottom of the pool inside the 16 cm dimeter Co-60 source basket. The dose rate was measured at heights of 5.5, 12, 19, and 23 cm from the bottom of the pool. The report “Tour of the Standards and Calibration Laboratory” includes information about the Co-60 source radiation pool. The importance of performing the Co-60 source characterization is to confirm previously reported Co-60 source dose rate measurements and to provide a documented measuring procedure for researchers and others who want to use the Co-60 source to evaluate the effects of radiation on space-based equipment for other applications. The measured dose rate at the abovementioned heights was 0.65, 0.69, 0.53, 0.23 krad/min respectively with the highest coefficient of variation of less than 10%.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Health Physics Research Reactor Criticality Accident Alarm System Benchmark Overview [Slides]

This presentation discusses the Health Physics Research Reactor (HPRR) or Fast Burst Reactor (FBR) and how available data from Health Physics Research Reactor (HPRR) operation to can be used to create a benchmark report for inclusion in the ICSBEP, as a Criticality Accident Alarm System (CAAS) shielding benchmark. The evaluation of experimental data is also presented, specifically four experimental candidates are considered of potential value for the benchmark: 1. Neutron source estimation from a HPRR pulse (energy spectrum, fission yield), 2. Threshold Detector Unit (TDU) measurements at different distances from a HPRR pulse, shielded and unshielded, 3. Sulfur pellet activation at different distances from a HPRR pulse, shielded and unshielded, and 4. Total neutron fluence from a HPRR pulse measured by Bonner Sphere Spectrometry, shielded and unshielded. Additionally, a benchmark model overview is discussed as are sample calculation results.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

IER 538: International Dosimetry Intercomparison Exercise with Godiva-IV [Slides]

Presentation on International Dosimetry Intercomparison Exercise with Godiva-IV. This presentation details Department of Energy (DOE) requirements and irradiation. Graphs show Neutron dose conversion factors, Neutron dose characterization, and Photon dose characterization. Results of 2m, 3m, and 4m tests are then compared to DOE standards. Findings show photon dose measurements were limited by instrumentation not available. Additionally, there has been a turnover of staff in dosimetry. However, funding is being provided for Universal Nuclear Accident Dosimeter (U-NAD).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Neural MUSE Analysis

Researchers at Oak Ridge National Laboratory (ORNL) created data as part of the MUSE (Multi-Agency Urban Search Experiment Detector and Algorithm Test Bed) project simulating illicit nuclear materials located in various buildings along a road. In the simulation, a truck containing a radiation detector drives down the road gathering listmode data (counting the and energy of incident gamma radiation). Building materials, source shielding, driving speed, truck direction, truck location on the road, source type, and source placement are all varied between runs of the data set. This data was created using deterministic neutron transport and Monte Carlo methods through a combination of SCALE, MAVRIC, MCNP, and GADRAS. As part of a follow-on NA-22 project, two Kaggle competitions were created to determine the best algorithms for finding and identifying gamma sources in this simulated urban environment. The winning algorithm was neural network-based and had a test accuracy of 76.4% accuracy for source identification. This work seeks to build upon this work and improve the results through the application of novel machine learning techniques. As a first step, the data was classified by a simple Convolutional Neural Network (CNN) To accomplish this, the data was first preprocessed into “waterfall plots.” These plots are composed of energy vs count plots that are stacked vertically to show progression in time. The horizontal axis indicating the particle energy incorporated user defined bin spacing with options for in linear-, logarithmic-, square root-, and user-spaced bins. The z or color dimension showed the number of counts corresponding the energy-time combination. This data was then used to generate more data, by generating a local estimate of the mean of the distribution for a bin and then randomly re-sampling that bin from a Poisson distribution. Once all of this data was generated, it was fed into a well-known CNN architecture, ResNet50. The output layer of this model was removed and replaced with layers corresponding to the shape desired isotope outputs. The provided training data was used to train the classifier and the remaining testing data was used to evaluate the model. Results are soon to be forthcoming.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Measurement of ambient radon daughter decay rates and energy spectra in liquid argon using the MicroBooNE detector

We report measurements of radon daughters in liquid argon within the MicroBooNE time projection chamber (LArTPC). The presence of radon in MicroBooNE’s 85 metric tons of active liquid argon bulk is probed with newly developed charge-based low-energy reconstruction tools and analysis techniques to detect correlated 214 Bi- 214 Po radioactive decays. Special datasets taken during periods of active radon doping enable new demonstrations of the calorimetric capabilities of singlephase neutrino LArTPCs for β and α particles with electron-equivalent energies ranging from 0.1 to 3.0 MeV. By applying 214 Bi- 214 Po detection algorithms to beam-external physics data recorded over a 46-day period, no statistically significant presence of radon is detected, corresponding to a limit of < 0.38 mBq/kg at the 95% confidence level. The obtained radon radiopurity limit – the first ever reported for a noble element detector incorporating liquid-phase purification – is well below the target value of the future DUNE neutrino detector.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

SCALE 6.3 Validation: Radiation Shielding

Safe and reliable use of scientific and engineering computer codes requires validation for the types of applications in which they will be used. An example in the nuclear reactor engineering and licensing field is radiation transport employed in shielding analyses. The validity of computer codes for shielding applications is demonstrated in this report for SCALE version 6.3.0. Representative benchmarks corresponding to shielding analyses are selected for the validation study. Typical measurement results analyzed from these benchmarks include neutron fluxes, detector count rates, detector energy response functions, neutron and gamma dose rates, neutron activation rates and activities, neutron leakage fluxes, and skyshine dose rates. Thousands of points of comparison between measurement and calculation are presented in this work. Other than rare outliers typically explained by either a lack of information or large uncertainties in the experiment conditions, material, or dimensions, the Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) radiation transport computer code with built-in variance reduction methods distributed with the SCALE computer code system agrees well with the measurement results. In selected benchmarks, MAVRIC is also compared to Monte Carlo N- Particle® (MCNP® ) 1 calculations. Both computer codes generally agree well within the estimated uncertainties. With the release of SCALE 6.3.0, Shift was integrated as an alternative transport solver in MAVRIC, denoted MAVRIC-Shift. Although the traditional MAVRIC using Monaco was used primarily in this validation study, many results have also been generated using MAVRIC-Shift. Agreement between MAVRIC-Monaco and MAVRIC-Shift is generally very good. The benchmarks presented in this report were obtained from reliable sources such as the International Criticality Safety Benchmark Evaluation Project Handbook, the Shielding Integral Benchmark Archive & Database, and other shielding validation work found in the literature. Additional datapoints and benchmarks will be added to future versions of this report to expand the shielding validation suite.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

COG Software Architecture Design Description Document

This COG Software Architecture Design Description Document describes the organization and functionality of the COG Multiparticle Monte Carlo Transport Code for radiation shielding and criticality calculations, at a level of detail suitable for guiding a new code developer in the maintenance and enhancement of COG. The intended audience also includes managers and scientists and engineers who wish to have a general knowledge of how the code works. This Document is not intended for end-users.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Neutron production and dosimetry with D(D,n) 3 He, 12 C(D,n) 13 N and 7 Li(p,n) 7 Be reactions at the Sandia Ion Beam Laboratory

Neutrons are produced at the Sandia Ion Beam Laboratory (IBL) using a beam of 280 keV D 2 + ions onto an erbium deuteride target. At test locations in the forward direction these neutrons have energies of about 2.9 MeV and a maximum flux on the flange of the test chamber of about 10 10 n/cm 2 /hour. Dosimetry is done by counting associated charged particles and by nickel foil activation dosimetry. Neutrons with higher energies, up to 6 MeV are produced using higher energy deuterium, up to 3 MeV, from the Pelletron ion accelerator at the IBL on a deuteride target. This report provides detailed information on the methods used for the production and dosimetry of neutrons by the D(D,n) 3 He nuclear reaction at the IBL. Neutron production by the 7 Li(p,n) 7 Be reaction with a proton beam from the Pelletron on a lithium target, such as LiF is also discussed. This produces neutrons with lower energy, in the range from about 0.1 to 1.3 MeV. Neutron production by the 12 C(D,n) 13 N reaction is also discussed. The purpose of this report is to provide information needed to test the effects of neutron irradiation on semiconductor materials and devices at the Sandia IBL.

36 MATERIALS SCIENCE↗