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At least 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

GADRAS Batch Inject Tool User Guide

Gamma Detector Response and Analysis Software (GADRAS) is used by the radiation detection and emergency response community to perform modeling and spectral analysis for gamma detector systems. Built into GADRAS is the ability to define a detector, geometry, background characteristics and source composition to generate synthetic spectra for drills and exercises (injects). Consequence Management is currently in development of a sample result data simulator tool in which a deposition model is probed for source conditions at moments in time and locations in space. These values are used to generate realistic sample results for use in drills and exercises. In addition to sample results, there is a need to simulate the actual spectra that would be observed in the field by downlooking HPGe instruments given a deposition activity. This way, the FRMAC Gamma Spectroscopist can practice their process of generating quantified results from spectra on realistic data as well. Recognizing the decades of work done in GADRAS to accurately generate synthetic spectra, this team decided to build a link between the new simulator and GADRAS to generate these spectra quickly and easily. The simulator tool will generate a file that specifies the name of the spectra, its location, date/time of measurement, duration of measurement, height off the ground, and the deposition activity and age for every radionuclide in the simulation. Then, a new tool within the Inject Tab of GADRAS was developed to read in this file given a detector selection and generate In-Situ spectra for each row in the file in any file format the user chooses. This way, simulation cell staff can take these files and then upload them to the appropriate data system (RAMS or RadResponder) for use during drills and exercises. An advanced feature of this tool allows for generating any spectra given an appropriate model and mapping of source to model layer in the batch inject tool. This way, spectra from field sample counts, mobile laboratories, or even fixed laboratories can be generated in bulk given an estimate of the radioactivity concentration or total radioactivity in an import file. This expands the capabilities of this tool a great deal and will make it a more useful tool for CM and others to help estimate detector response for nearly any situation. This user guide will explain the steps needed to perform a batch inject file generation.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Evaluation of Thermolytic Hydrogen Generation Rate Models at High-Temperature/High-Hydroxide Regimes

This report describes the results of testing performed to extend the applicable ranges of temperature and hydroxide concentration for use within the Glycolate and Global Total Organic Carbon (TOC) Hydrogen Generation Rate (HGR) expressions. Seven experimental conditions (six simulants of the 242-25H Evaporator system chosen as a D-optimal set of experiments and a single test conducted at an elevated boiling point of 170 °C) were investigated in the presence of sodium glycolate and Xiameter TM AFE-1010. Glycolate was employed to study the extension of the Glycolate Thermolytic HGR expression while Xiameter TM AFE-1010 was employed to study the extension of the Global TOC Thermolytic HGR expression. The following conclusions were derived from this testing: The Glycolate Thermolytic HGR expression may be confidently used to predict thermolytic HGRs from glycolate at temperatures as high as 170 °C and hydroxide concentrations as high as 23 M.; The hydroxide and temperature-dependence predicted by the Global TOC Thermolytic HGR expression has been confirmed at temperatures as high as 170 °C and hydroxide concentrations as high as 23 M, suggesting that the Global TOC Thermolytic HGR expression may be used at these ranges.; Methane was observed from tests with Xiameter TM AFE-1010 at production rates higher than those observed for hydrogen. These rates were observed at temperatures higher than 100 °C.; Preliminary models suggest that increasing hydroxide/temperature causes an increase in Methane Generation Rate (MGR) from Xiameter TM AFE-1010. The following recommendations are based on this testing: The existing equations for thermolytic HGR from glycolate and non-glycolate organics should be used at Concentration, Storage, and Transfer Facilities (CSTF) storage and evaporation conditions, including temperatures and hydroxide concentrations exhibited in the 242-25H Evaporator.; Further investigation should be made into the influence of methylsilanes on CSTF flammability. This investigation should include: determination of the types of methylsilanes historically added to the CSTF, determination of methane formation rates from each type of methylsilane, and determination of the extent of degradation of methylsilanes in CSTF waste.; Characterization techniques should be developed by Savannah River National Laboratory (SRNL) to assist in the speciation of methylsilane-containing waste in the CSTF.; Additional testing with radioactive waste should be performed to determine the MGRs possible in radioactive waste and better inform model predictions made from testing with simulants.

08 HYDROGEN↗
DOE OSTI2020

Crossroads Baker, "Helen of Bikini"

Test Baker, the underwater detonation of a Fat Man bomb nicknamed Helen of Bikini by an anonymous “Kilroy,” took place on July 25, 1946. Encased in a bathysphere fashioned from a submarine conning tower and suspended ninety feet below the surface of Bikini’s lagoon, Helen of Bikini was spectacular. Her energy burst through the lagoon surface at 11,000 ft. /sec pushing over two million cubic yards of radioactive seawater and sediment to a height of 4,300 feet within sixty seconds. Nine vessels were sunk and an additional five ships were physically destroyed, although still afloat. When it collapsed, the column of radioactive sediment and seawater spread over the surviving ships permanently contaminating them. Helen of Bikini destroyed the entire fleet of target vessels.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2020

Using the CAP88 Plume Model & Dose Assessment Code for the FTWC Project

The Environmental Protection & Compliance Division (EPC) uses the Clean Air Act Assessment Package – 1988 (CAP88) plume model and dose assessment code to evaluate measured emissions or projected emissions of airborne radionuclides from Los Alamos National Laboratory (LANL) facilities. This software package will use facility data (e.g., stack height and exit velocity), emissions parameters (chemical form and quantity), and environmental data (wind speed and direction) to predict downwind concentrations of radioactive material and subsequent dose to a member of the public at that location. The United States Environmental Protection Agency (EPA) has approved CAP88 for use in demonstrating compliance with the Radionuclide NESHAP2, that portion of the Clean Air Act addressing emissions of airborne radioactive material from Department of Energy facilities. EPC Division has two procedures which discuss the use of CAP88.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

CASTOR THTR/AVR Containment Review

The Lawrence Livermore National Laboratory (LLNL)Radioactive Material Packaging and Transportation Review Team (herein referred to as ‘LLNL Staff’) was tasked with the technical review and confirmatory analysis of the containment analysis and corresponding leakage test implementation procedure for the CASTOR®THTR/AVR radioactive material transportation package with the AVR contents.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Spectroscopy of Resonances in the Astrophysical r p –Process Path. Final Technical report

This report describes the results of a project to study the spectrum of proton-resonances which determine the energy release and timescale for the astrophysical rp-process path. Steps to achieve successful experiments include the commissioning of a the in-flight radioactive-beam facility RESOLUT , development of instrumentation and analysis techniques for experiments performed with it. Radioactive-beam experiments performed in the context of the grant included the reactions 25 Al(d,n) 26 Si, 19 O(d,p) 20 O, 19 Ne(d,n) 20 Na and 17 F(d,n) 18 Ne. The experimental program developed experiments and methods to characterize aspects of proton-capture reactions with respect to the in the context of the astrophysical rp-process.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Understanding Materials Characteristics of Palladium Lanthanide Cermet Wires as a Mimic of Californium Wire Production

Californium-252 (Cf-252) has been synthesized on a large scale (mg) since the 1970s for use as a portable neutron source. With a half-life of 2.645 years and a spontaneous fission fraction of 3.09%, Cf-252 emits 2.31 × 10 12 neutrons/gram second. Today most of the Cf-252 produced in the world is made at Oak Ridge National Laboratory (ORNL) using the High Flux Isotope Reactor (HFIR) and the Radiochemical Engineering Development Center (REDC). Because of its highly radioactive nature, all manipulations of Cf-252 need to be performed in hot cell facilities by highly trained personnel. The Cf-252 produced at ORNL is done at the REDC, which is conveniently located next to the HFIR. Cf-252 is sold as bulk wire, which has been described as palladium (Pd) metal incasing californium oxide (Cf 2 O 3 ). The process was developed by Mosley et al. at Savannah River Laboratory (SRL) where it was modeled after the thorium work done by Fuschillo et al. This process has been used since the 1970s to synthesize this unique wire, but its fundamental chemistry and materials characteristics were never openly reported. Here we aim to better understand the chemistry and the materials characteristics of Cf-252 wire production for process improvements and to better our fundamental understanding of the mid-late actinides. Though there are many challenges in working with radioactive elements such as Cf, the isotope Cf-252 is even more complicated due to its neutron emission. One challenge with neutron emission is the ability of neutrons to penetrate much deeper through materials, causing radiation damage. This makes using instrumentation to characterize the chemistry and materials properties of this process difficult. For this reason, this study will first focus on terbium (Tb) as a surrogate for Cf. Currently Tb is used as a carrier or filler in Cf production when small amounts of Cf are being processed. This, along with their similar size and charge, makes Tb a reasonable surrogate for this study.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Drop Analysis of the Advanced Test Reactor Fresh Fuel Shipping Container with Heavier Low-Enriched Uranium Fuel Contents

The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is a rectangular stainless steel container used for shipping radioactive material. The container is described in the ATR FFSC Safety Analysis Report (SAR). Per the ATR FFSC SAR, the ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.4, and was originally designed to transport high enriched uranium (HEU) reactor fuel elements for the Advanced Test Reactor (ATR), the Advanced Test Reactor Critical (ATRC) facility, the Massachusetts Institute of Technology Reactor (MITR), and the University of Missouri Research Reactor (MURR). The Department of Energy, National Nuclear Security Administration’s (NNSA), Office of Material Management and Minimization (M3) is working with the Idaho National Laboratory (INL) to develop and qualify new low enriched uranium (LEU) fuels and technologies for use in the ATR, ATRC, MITR, and MURR reactors. The LEU fuel elements will weigh significantly more than the current HEU designs and, combined with their associated Fuel Handling Enclosures for packaging, some configurations will exceed the 50 lbf used in the ATR FFSC qualifying drop tests. There are LEU versions of MITR, MURR, and ATR fuel elements. However, for this evaluation, drop analysis of the ATR FFSC with only the heavier ATR Low Enrichment (LOWE) fuel element is considered in this evaluation because the LOWE fuel element is the heaviest of the considered LEU fuel elements. The ATR HEU fuel element and the ATR LOWE fuel element are identical in every design aspect except for the fuel meat inside the 19 fuel plates. The LEU fuel meats are made using a U-10Mo high-density foil rather than uranium dispersed in aluminum in the HEU fuel elements. The high density of the uranium in the LEU fuel meat increases the LOWE fuel element weight to just under 44 lbf (versus the 22.1 lbf weight of the tested ATR HEU fuel element). ATR fuel elements are placed in a thin-gauge aluminum weldment called a "Fuel Handling Enclosure" during packaging. The Fuel Handling Enclosure is used to cover and protect the element during loading and unloading operations. The ATR Fuel Handling Enclosure weighs about 15 lbf per the drawings in the ATR FFSC SAR and the weight is accounted for in this evaluation. Transporting the heavier LEU fuel elements require evaluation of two issues. The first is the effect of the increased mass of the LEU fuel elements on the survivability of the ATR FFSC package following the requisite drop qualifications. The second is the effect of the increased mass of the fuel plates on the fuel element during the same drops. The ATR FFSC containing an ATR HEU fuel element in an ATR Fuel Handling Enclosure was physically dropped multiple times to qualify the container as a Type AF-96 package. The ATR FFSC SAR describes the drop tests performed with an actual ATR HEU fuel element weighing 22.1 lbf contained in a 14.3 lbf Fuel Handling Enclosure for a total payload of 36.4 lbf. Those drop tests showed that the ATR FFSC maintained containment of the ATR HEU fuel element and the fuel element was not significantly damaged. (Containment herein is not defined as a leak tight but is retention of the radioactive contents.) The purpose of the evaluation is to analytically show that, for a similar set of tests, the ATR FFSC maintains containment of the heavier ATR LOWE fuel element and to assess the damage to the fuel element during the drops. The approach was to create finite element analysis (FEA) models that produce the same results as the physical drops. Those models were then used as the benchmarks for the follow-on analyses using the heavier contents. FEA models of the drops of ATR FFSC using up to a 115 lbf fuel element were run and evaluated. Likewise, drops of a LOWE fuel element weighing 44 lbf in the ATR FFSC were run and evaluated. It is important to note that this report was done at the quality level necessary to be included in a nuclear facility safety basis. However, it is not the intent of this report to conclude the suitability of the ATR FFSC for transporting the heavier payloads. This report only describes the results of the FEA as related to the required drop scenarios. Incorporation of the FEA into the safety basis will be evaluated by the ATR FFSC design authority. The physical drop tests of the HEU fuel element and FEA drop analysis for the LOWE fuel element showed noteworthy damage to the fuel plates. An aluminum protective block was conceived to mitigate the damage. The concept requires the blocks to be placed in the fuel element between the end boxes and fuel plates. Additional FEA drop analyses were performed using the protective block. The addition of the blocks is primarily intended to mitigate the damage to the LOWE fuel element fuel plates. However, FEA drop analyses of the ATR HEU fuel element with the blocks were also performed and included for information.

42 ENGINEERING↗
DOE OSTI2020

Decovalex-2019 (Executive Summary)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. The overall aim of DECOVALEX-2019 was to increase the understanding of various thermo-hydro-mechanical-chemical processes of importance for radionuclide release and transport from a repository to the biosphere and how they can be described and modelled using mathematical models. The scientific and technical objectives are: to increase the basic understanding of T-H-M-C coupled processes in fractured rocks (crystalline, sedimentary, argillaceous) and buffer materials; to investigate the predictive capabilities of different codes to field experiments and to perform verification of codes; to exchange experimental data, and improve the understanding of the constitutive behavior of crystalline and argillaceous rock masses and buffer materials; and to perform THMC calculations in a performance/safety assessment context.

58 GEOSCIENCES↗
DOE OSTI2021

Update to the Finite Cloud Dose Correction Factors in MACCS

In WASH - 1400, external exposure from the finite radioactive cloud (cloudshine) is calculated by assuming that the cloud is semi-infinite, the concentration of radioactive material is uniform, and by using a correction factor to account for these approximations. This correction factor is originally based upon formulations by Healy and depends on the effective size of the plume and the distance from the plume center to the receptor. The range of the finite cloud dose correction factor table from WASH - 1400 developed using Healy formulations can be exceeded in certain situations. When the range of the table is exceeded, no extrapolation is performed; rather interpolation at the edge of the table is performed per WASH - 1400. The tabulated values of these finite cloud dose correction factors from WASH - 1400 and the interpolation at the edge of the table have been used in MACCS since its creation. An expanded table of finite cloud dose correction factors is one way to reduce the need of using interpolation at the edge of the table. The generation of an expanded finite cloud dose correction factor table for future use in MACCS is documented in this report.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Nuclear Forensics National Frameworks [Slides]

Nuclear forensics is the examination of nuclear or other radioactive material, or of evidence contaminated with radionuclides, in the context of legal proceedings under international or national law related to nuclear security. The goal of forensic science is to discover linkages among people, places, materials, and events. Nuclear forensics is an essential component of national response plans to events involving nuclear or other radioactive materials out of regulatory control and informs prevention, detection, and response.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Analysis of the MST-16 crimp process on cold 304L stainless steel

The purpose of this report is to assess the quality of the MST-16 crimp process through a series of cold experiments and metallographic analyses. Common practice in MST-16 uses a mechanical crimp on 304L stainless steel tubes to prevent radioactive contamination from escaping the tube interior for a variety of experiments. The crimped tubes must pass a leak check to demonstrate their radiological safety. Cold (i.e. non-radioactive samples) 304L tube samples were tested using two different approved crimping devices to assess the quality of the crimps. Each crimp was leak checked, followed by destructive metallographic analysis to determine the quality of both the crimp equipment and procedures used by MST-16 personnel.

42 ENGINEERING↗
DOE OSTI2021

Advanced Low-Temperature Chlorination of Zirconium

Recovery of Zr from nuclear fuel is of high interest because the Zr cladding material comprises up to 50% of the high-level radioactive waste (HLW) that requires disposition. An ideal Zr-recovery process would not only recover the bulk Zr from the fuel or cladding but would also provide decontamination from the components that give rise to the HLW designation. A simplified decontamination pathway providing recovered Zr in compliance with low-level radioactive waste acceptance criteria could significantly reduce the burden of HLW requiring geologic disposition. A newly developed advanced low-temperature chlorination process presents an opportunity to recover a ZrCl4 product that is effectively decontaminated from the elements of concern. This advanced low-temperature chlorination process uses a mixture of the sulfur-containing chlorination compounds sulfur monochloride (S 2 Cl 2 ) and thionyl chloride (SOCl 2 ).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

IRAD-MPE Radiological and Nuclear Security Threats (Day 1) [Slides]

This module is designed to enhance awareness regarding radiological and nuclear security threats and the types of radioactive materials that can be used for criminal purposes or terrorist acts. The goals are to: (1) Review the Threat Assessment Process for MPEs, (2) Recognize Potential Radiological and Nuclear Security Threats, (3) Understand the basics of Radiological Dispersal Devices (RDD) and Radiological Exposure Devices (RED), (4) Learn about Medical, Commercial, and Naturally Occurring Radioactive Materials (NORM) sources, and (5) Understand the term Material Out of Regulatory Control (MORC).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

General Employee Radiological Training: 24 Month Refresher (Course #47968)

GERT is required by the 10 CFR 835, the Federal Regulation on Radiation Protection that we follow here at LANL. It’s the training for non Rad Workers who may enter Radiological Controlled Areas (RCAs) and Radioactive Material Areas (RMAs) in an unescorted manner. If your work assignment requires access to areas posted for higher radiological hazards like Radiological Buffer Areas (RBAs), Radiation Areas, Contamination Areas, and others you are required to have Rad Worker training. Once you complete this training you will be qualified to access RCAs and RMAs at LANL without an escort. However, you are not qualified to operate radiation producing devices, work with radioactive material, or do work where you may receive greater than 100 mrem of exposure in a year.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Investigation Support Scenario-based Policy Discussion Development [Slides]

Nuclear forensics is the examination of nuclear or other radioactive material, or of evidence contaminated with radionuclides, in the context of legal proceedings under international or national law related to nuclear security. The goal of forensic science is to discover linkages among people, places, materials and events. Nuclear forensics is an essential component of national response plans to events involving nuclear or other radioactive materials out of regulatory control and informs prevention, detection and response.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Engineering Evaluation of Barium Buildup in a Decayed CsCl Sealed Source and Potential Impact for Cesium Release from a Breached Source

For more than 50 years, radioactive 137 Cs has been a major source material for radioactive sealed sources, usually constructed as cesium chloride (CsCl) salt loaded into double-walled, stainless-steel capsules. A complication develops as 137 Cs decays to 137 Ba since this process creates a strongly reducing environment inside the capsule. A potential hazard exists if the capsule is breached and air ingress induces rapid exothermic oxidation, and this mechanism is suspected to be responsible for the well-known contamination incident at the Harbor View facility in Washington state. For this study, many thermodynamic evaluations were performed to assess the internal state of capsules after several decades of decay and to describe the potential oxidation if the capsule contents were to suddenly be exposed to air. Results suggest that reduction of impurities such as Cu, Fe, Pb, and Cr to metal will occur. If these impurities are lacking, it is possible that even Ba metal will form. In most cases, rapid oxidation can occur, and the exothermic reactions are sufficient to vaporize portions of the contents, which would include the remaining 137 Cs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

2021 Annual Site Environmental Report

This report provides the U.S. Department of Energy (DOE) and the public with information on the level of radioactive and non-radioactive pollutants (if any) that are added to the environment as a result of Princeton Plasma Physics Laboratory’s (PPPL) operations. The results of PPPL’s 2021 environmental surveillance and monitoring program are presented and discussed. The report also summarizes environmental initiatives, assessments, and community involvement programs that were undertaken in 2021. PPPL’s on-site operations were significantly curtailed in 2021 due to the global coronavirus pandemic. PPPL has engaged in fusion energy research since 1951. The Laboratory’s mission is to develop the scientific knowledge and advanced engineering to enable fusion to power the U.S. and the world, and to developing the understanding of plasmas from the nano- to the astrophysical scale. PPPL’s primary experiment, the National Spherical Torus Experiment-Upgrade (NSTX-U) is a collaboration among national laboratories, universities, and national and international research institutions and is a major element in the US Fusion Energy Sciences Program. Its design tests the physics principles of spherical torus (ST) plasmas, playing an important role in the development of smaller, more economical fusion reactors. Due to previous operational issues, NSTX-U did not operate in 2021. PPPL is engaged in a project to replace key NSTX-U components and systems to enable operation of this international fusion user facility. In 2021, PPPL’s radiological environmental monitoring program measured tritium in the air at the NSTX-U Stack and at onsite sampling stations. Using highly sensitive air monitors, PPPL is capable of detecting small changes in the ambient levels of tritium. The operation of an in-stack monitor located on D-site is used to demonstrate compliance with the National Emission Standard for Hazardous Air Pollutants (NESHAPs) regulations. Also included in PPPL’s radiological environmental monitoring program, are water monitoring – ground, surface, and waste waters. PPPL’s radiological monitoring program characterized the background levels of tritium in the environment and those data are presented in this report. Ground water monitoring continued under New Jersey Department of Environmental Protection’s (NJDEP) Site Remediation Program regulations. PPPL monitored for non-radiological contaminants, mainly volatile organic compounds (components of common degreasing solvents). In 2021, PPPL complied with permit limits for surface and sanitary discharges. PPPL was honored with awards for EPEAT-certified electronics purchasing and use of peracetic acid as an alternative water treatment chemical in its non-potable process water systems an NJDEP recycling award.

54 ENVIRONMENTAL SCIENCES↗