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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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At least 1,693 records · Page 94
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

(U) Feynman-Y calculations using PARTISN

A prescription for computing the Feynman Y as a function of coincidence gate width using a deterministic multigroup neutron transport code has been published and the results compared favorably with measurements of the BeRP ball. In this paper, we report on our project to implement the method and reproduce the results. There are several clarifications and corrections of the published prescription. We show results using two multigroup cross section libraries compared with measurements and with Monte Carlo results. Deterministic simulations of the mean count rates compare very favorably with previously published Monte Carlo results, and deterministic simulations of the Feynman Y asymptote compare somewhat favorably. In Feynman beta plots, the deterministic simulations reached the asymptotic value much sooner than did a fit to the measured data.

42 ENGINEERING↗
DOE OSTI2016

Evaluation of the Fluence Conversion Factor for 32 P in Sulfur

When 32 S is exposed to neutrons it undergoes a 32 S(n,p) 32 P reaction with a neutron cross section as shown in Figure 1. This reaction may be used to characterize the neutron fluence for neutrons greater than 3 MeV.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

Beam line shielding calculations for an Electron Accelerator Mo-99 production facility

The purpose of this study is to evaluate the photon and neutron fields in and around the latest beam line design for the Mo-99 production facility. The radiation dose to the beam line components (quadrupoles, dipoles, beam stops and the linear accelerator) are calculated in the present report. The beam line design assumes placement of two cameras: infra red (IR) and optical transition radiation (OTR) for continuous monitoring of the beam spot on target during irradiation. The cameras will be placed off the beam axis offset in vertical direction. We explored typical shielding arrangements for the cameras and report the resulting neutron and photon dose fields.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Use of Sources and SNM in COE Facilities

This procedure describes how to use Special Nuclear Material (SNM) and sources in order to perform measurements in a safe manner. The hazards and controls associated with these activities are addressed in this document.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Aerosol Physics Considerations for Using Cerium Oxide CeO 2 as a Surrogate for Plutonium Oxide PuO 2 in Airborne Release Fraction Measurements for Storage Container Investigations

Cerium oxide (CeO2) dust is recommended as a surrogate for plutonium oxide (PuO2) in airborne release fraction experiments. The total range of applicable particle sizes for PuO2 extends from 0.0032 μm (the diameter of a single PuO2 molecule) to 10 μm (the defined upper boundary for respirable particles). For particulates with a physical particle diameter of 1.0 μm, the corresponding aerodynamic diameters for CeO2 and PuO2 are 2.7 μm and 3.4 μm, respectively. Cascade impactor air samplers are capable of measuring the size distributions of CeO2 or PuO2 particulates. In this document, the aerodynamic diameters for CeO2 and PuO2 were calculated for seven different physical diameters (0.0032, 0.02, 0.11, 0.27, 1.0, 3.2, and 10 μm). For cascade impactor measurements, CeO2 and PuO2 particulates with the same physical diameter would be collected onto the same or adjacent collection substrates. The difference between the aerodynamic diameter of CeO2 and PuO2 particles (that have the same physical diameter) is 39% of the resolution of a twelve-stage MSP Inc. 125 cascade impactor, and 34% for an eight-stage Andersen impactor. An approach is given to calculate the committed effective dose (CED) coefficient for PuO2 aerosol particles, compared to a corresponding aerodynamic diameter of CeO2 particles. With this approach, use of CeO2 as a surrogate for PuO2 material would follow a direct conversion based on a molar equivalent. In addition to the analytical information developed for this document, several US national labs have published articles about the use of CeO2 as a PuO2 surrogate. Different physical and chemical aspects were considered by these investigators, including thermal properties, ceramic formulations, cold pressing, sintering, molecular reactions, and mass loss in high temperature gas flows. All of those US national lab studies recommended the use of CeO2 as a surrogate material for PuO2.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2017

Tabulated Neutron Emission Rates for Plutonium Oxide

This work tabulates neutron emission rates for 80 plutonium oxide samples as reported in the literature. Plutonium-­238 and plutonium-­239 oxides are included and such emission rates are useful for scaling tallies from Monte Carlo simulations and estimating dose rates for health physics applications.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Technical Evaluation of Electrically Cooled Detectors for use with the LLNL Whole Body Scanning Bed System

Lawrence Livermore National Laboratory (LLNL) is replacing a liquid nitrogen cooled high purity germanium detection system with an electrically cooled high purity germanium system for its scanning bed whole body counting system. The new detectors offer a safer environment for workers while they are measured inside the shielded facility at LLNL and provide highly stable cooling of the detectors. The positioning of the electrically cooled detectors is slightly different from the liquid nitrogen cooled detectors due to the electrical cooling units attached to the detectors. The technical evaluations and results described in this document demonstrate that the new electrically cooled system provides results that comply with internal dosimetry requirements

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

MN471017, Safety Basis Manual Appendix NF-01. Nuclear Facility Hazard Categorization

The purpose of this appendix is to provide a consistent approach for establishing the hazard category for a nuclear facility as required in 10 CFR 830, Nuclear Safety Management, Subpart B, "Safety Basis Requirements," Section 202 (b)(3). As defined, this approach is consistent with DOE-STD-1027-92 Change Notice No. 1, Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports (hereafter DOE-STD-1027-92), and facilitates the use of updated dosimetry and release fractions as provided in NNSA SD G 1027 Admin Change 1, Guidance on Using Release Fraction and Modern Dosimetric Information Consistently with DOE STD 1027-92, Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports, Change Notice No. 1 (hereafter NNSA SD G 1027).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Blind Intercomparison of Nuclear Accident Dosimetry using the Flattop Reactor at NCERC

The purpose of this exercise was to continue United States, United Kingdom, and France collaborative activities performing Nuclear Accident Dosimetry intercomparison. IER 148 was the inaugural intercomparison exercise using the Godiva reactor at the DAF. In contrast to IER 148, in the current exercise participants were “blind” to the known doses during the exercise, to simulate the dosimetry response to a real criticality incident. This document is the Final Exercise (CED-x) Report for IER-253, “International Intercomparison Exercise for Nuclear Accident Dosimetry at the DAF Using Flattop.” The report describes the structure and results of the exercise consisting of two irradiations. The details of all dosimetry elements and their placement in proximity to Flattop on support stands or phantoms was unknown to participating laboratories. The participating laboratories in this report are deidentified and the report summarizes participant results from the intercomparison. The exercise occurred during the week of May 21, 2018.

42 ENGINEERING↗
DOE OSTI2019

Technical Equivalency Evaluation of Lynx Electronics and Apex-InVivo software for the LLNL Whole Body Scanning Bed System

Lawrence Livermore National Laboratory (LLNL) Whole Body Counter (WBC) is replacing the following components for its scanning bed counting system: NIM electronics with modern Lynx electronics, The Alpha Workstation running VMS with a PC running Windows , and Abacos-Plus in-vivo counting software with the new Apex-InVivo software. These new components provide a compact design, extend the lifecycle of the system as a whole, reduce maintenance and user time, an improve overall ease of use. The technical evaluations and results described in this document demonstrate that the total system, incorporating these new components, provides results that comply with Internal Dosimetry requirements.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Preliminary Design of a New Pulsed-Neutron Die-Away Experimental Testbed for Thermal Scattering Law Benchmarks (PNDA) (IER-501 CED-1)

This report presents the preliminary design (CED-1) of IER-501, a new testbed for pulsed-neutron die-away (PNDA) experiments at Lawrence Livermore National Laboratory (LLNL). The PNDA experiments will be high-quality benchmarks that serve to optimize and validate thermal scattering laws (TSLs), which are high-priority nuclear data for the Department of Energy’s Criticality Safety Program (NCSP). PNDA experiments can be used to generate nuclear data validation benchmarks quickly and cheaply, as they do not require fissile material or a nuclear facility to conduct them.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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.

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