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At least 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

2011 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses

The Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (National Security Technologies, LLC, 2007a) requires an annual review to assess the adequacy of the Performance Assessments (PAs) and Composite Analyses (CAs), with the results submitted annually to U.S. Department of Energy (DOE) Office of Environmental Management. The Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) also require that such reviews be made and that secondary or minor unresolved issues be tracked and addressed as part of the maintenance plan (DOE, 1999a; 2000).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Summer Student Jessica Halterman Report

This summer I worked on the Enhanced Radiological Nuclear Inspection and Evaluation (ERNIE) project with Dr. Simon Labov. Radiation screening of vehicles is performed at ports of entry and checkpoints. Vehicles carrying cargo containing radioactive material set off an alarm. Currently, trained experts are required to evaluate each alarm. This takes time and holds up traffic. Also, the accuracy of the threat evaluation is dependent on the training of the evaluator.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

AHTR Mechanical, Structural, and Neutronic Preconceptual Design

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming an option for commercial reactor deployment. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The reactor concept development remains at a preconceptual level of maturity. While the overall appearance of an AHTR design is anticipated to be similar to the current concept, optimized dimensions will differ from those presented here. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month two-batch cycle with 9 wt. % enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The report includes a preconceptual design of the manipulators, the fuel transfer system, and the used fuel storage system. The present design intent is for used fuel to be stored inside of containment for at least six months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents as well as multiple levels of radioactive material containment. Key building design elements include (1) below grade siting to minimize vulnerability to aircraft impact, (2) multiple natural circulation decay heat rejection chimneys, (3) seismic base isolation, and (4) decay heat powered back-up electricity generation.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2012

Phase II Characterization Survey of the USNS Bridge (T-AOE 10), Military Sealift Fleet Support Command, Naval Station, Norfolk, Virginia (Final Report)

In March 2011, the USNS Bridge was deployed off northeastern Honshu, Japan with the carrier USS Ronald Reagan to assist with relief efforts after the 2011 Tōhoku earthquake and tsunami. During that time, the Bridge was exposed to air-borne radioactive materials leaking from the damaged Fukushima I Nuclear Power Plant. The proximity of the Bridge to the air-borne impacted area resulted in the contamination of the ship’s air-handling systems and the associated components, as well as potential contamination of other ship surfaces due to either direct intake/deposition or inadvertent spread from crew/operational activities. Preliminary surveys in the weeks after the event confirmed low-level contamination within the heating, ventilation, and air conditioning (HVAC) ductwork and systems, and engine and other auxiliary air intake systems. Some partial decontamination was performed at that time. In response to the airborne contamination event, Military Sealift Fleet Support Command (MSFSC) contracted Oak Ridge Associated Universities (ORAU), under provisions of the Oak Ridge Institute for Science and Education (ORISE) contract, to assess the radiological condition of the Bridge. Phase I identified contamination within the CPS filters, ventilation systems, miscellaneous equipment, and other suspect locations that could not accessed at that time (ORAU 2011b). Because the Bridge was underway during the characterization, all the potentially impacted systems/spaces could not be investigated. As a result, MSFSC contracted with ORAU to perform Phase II of the characterization, specifically to survey systems/spaces previously inaccessible. During Phase II of the characterization, the ship was in port to perform routine maintenance operations, allowing access to the previously inaccessible systems/spaces.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2012

AHTR Mechanical, Structural, And Neutronic Preconceptual Design

This report provides an overview of the mechanical, structural, and neutronic aspects of the Advanced High Temperature Reactor (AHTR) design concept. The AHTR is a design concept for a large output Fluoride salt cooled High-temperature Reactor (FHR) that is being developed to enable evaluation of the technology hurdles remaining to be overcome prior to FHRs becoming a commercial reactor class. This report documents the incremental AHTR design maturation performed over the past year and is focused on advancing the design concept to a level of a functional, self-consistent system. The AHTR employs plate type coated particle fuel assemblies with rapid, off-line refueling. Neutronic analysis of the core has confirmed the viability of a 6-month 2-batch cycle with 9 weight-percent enriched uranium fuel. Refueling is intended to be performed automatically under visual guidance using dedicated robotic manipulators. The present design intent is for used fuel to be stored inside of containment for at least 6 months and then transferred to local dry wells for intermediate term, on-site storage. The mechanical and structural concept development effort has included an emphasis on transportation and constructability to minimize construction costs and schedule. The design intent is that all components be factory fabricated into rail transportable modules that are assembled into subsystems at an on-site workshop prior to being lifted into position using a heavy-lift crane in an open-top style construction. While detailed accident identification and response sequence analysis has yet to be performed, the design concept incorporates multiple levels of radioactive material containment including fully passive responses to all identified design basis or non-very-low frequency beyond design basis accidents. Key building design elements include: 1) below grade siting to minimize vulnerability to aircraft impact, 2) multiple natural circulation decay heat rejection chimneys, 3) seismic base isolation, and 4) decay heat powered back-up electricity generation. The report provides a preconceptual design of the manipulators, the fuel transfer system, and the salt transfer loops. The mechanical handling of the fuel and how it is accomplished without instrumentation inside the salt is described within the report. All drives for the manipulators reside outside the reactor top flange. The design has also taken into account the transportability of major components and how they will be assembled on site

Varma, Venugopal Koikal↗
DOE OSTI2013

3D Flow and Temperature Analysis of Filling a Plutonium Mold (Revision 1) [Poster]

The plutonium foundry at Los Alamos National Laboratory casts products for various special nuclear applications. However, plutonium’s radioactivity, material properties, and security constraints complicate the ability to perform experimental analysis of mold behavior. The Manufacturing Engineering and Technologies (MET-2) group previously developed a graphite mold to vacuum cast small plutonium disks to be used by the Department of Homeland Security as point sources for radiation sensor testing. A two-stage pouring basin consisting of a funnel and an angled cavity directs the liquid into a vertical runner. A stack of ten disk castings connect to the runner by horizontal gates. Volumetric flow rates were implemented to limit overflow into the funnel and minimize foundry returns. Models using Flow-3D computational fluid dynamics software are employed here to determine liquid Pu flow paths, optimal pour regimes, temperature changes, and pressure variations.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Nevada National Security Site Environmental Report 2012

This report was prepared to meet the information needs of the public and the requirements and guidelines of the U.S. Department of Energy (DOE) for annual site environmental reports. It was prepared by National Security Technologies, LLC (NSTec), for the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) (formerly designated as the Nevada Site Office [NNSA/NSO]). The new field office designation occurred in March 2013. Published reports cited in this 2012 report, therefore, may bear the name or authorship of NNSA/NSO. This and previous years’ reports, called Annual Site Environmental Reports (ASERs), Nevada Test Site Environmental Reports (NTSERs), and, beginning in 2010, Nevada National Security Site Environmental Reports (NNSSERs), are posted on the NNSA/NFO website at http://www.nv.energy.gov/library/publications/aser.aspx. This NNSSER was prepared to satisfy DOE Order DOE O 231.1B, “Environment, Safety and Health Reporting.” Its purpose is to (1) report compliance status with environmental standards and requirements, (2) present results of environmental monitoring of radiological and nonradiological effluents, (3) report estimated radiological doses to the public from releases of radioactive material, (4) summarize environmental incidents of noncompliance and actions taken in response to them, (5) describe the NNSA/NFO Environmental Management System and characterize its performance, and (6) highlight significant environmental programs and efforts. This NNSSER summarizes data and compliance status for calendar year 2012 at the Nevada National Security Site (NNSS) (formerly the Nevada Test Site) and its two support facilities, the North Las Vegas Facility (NLVF) and the Remote Sensing Laboratory–Nellis (RSL-Nellis). It also addresses environmental restoration (ER) projects conducted at the Tonopah Test Range (TTR) and the Nevada Test and Training Range (NTTR). Through a Memorandum of Agreement, NNSA/NFO is responsible for the oversight of these ER projects, and the Sandia Site Office of NNSA (NNSA/SSO) has oversight of all other TTR and NTTR activities. NNSA/SSO produces the TTR annual environmental report available at http://www.sandia.gov/news/publications/environmental/index.html.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2011

Safety Analysis Report for Packaging Model 9977, Addendum 3: Justification for Small Gram Quantity Contents (Revision 5)

This Addendum establishes a new family of content envelopes consisting of small quantities of radioactive materials. These content envelopes and specific packing configurations are shown to be subcritical. However, the dose rates of some payloads must be measured and shown to comply with applicable radiation limits. Authorization for shipment of the content envelop requires acceptance of this Addendum by the DOE-HQ certifying official as a supplement to the 9977 SARP Revision 2 and DOE-HQ's subsequent revision of the CoC Revision 10 (which is based on SARP Addendum 2 and SARP Addendum 4) to authorize the additional content envelope. The Small Gram Quantity Content Envelopes and packing configurations will be incorporated in the next revision of the 9977 SARP.

42 ENGINEERING↗
DOE OSTI2013

Functional Bounding Content Envelope for Gamma and Neutron Emitting Isotopes in the Model 9977 Packaging

The contents approved for shipment in a Type B radioactive material transportation package have historically been descriptions of discrete items, or groupings of well-defined similar items, in the package safety basis documentation. The need for a comprehensive functional content envelope of both gamma and neutron emitting nuclides compliant with regulatory limits has become necessary as the DOE complex requires shipments of unique mixtures of radionuclides and impurities. Recent publications have presented a calculational model and a corresponding content envelope intended to be compliant with federal regulatory external radiation limits as well as design decay heat limits based on the Model 9977 Packaging. The methodology used to develop this content envelope consisted of determining the external radiation dose rates based on one gram of a given isotope combined, in the case of actinides, with various levels of light element impurities and determining the allowable mass to the regulatory limits based on these dose rates. The method of ratioing from these calculations to the masses that meet the regulatory limits fails in the case of some combinations of actinides and light element impurities because of the effect of subcritical multiplication, resulting in non-conservative and non-compliant dose rates in some cases. This is particularly true for many actinides combined with beryllium, boron, fluorine, lithium, and sodium. In this study, the source was modeled as a sphere with the appropriate dimensions based on the actinide or gamma source density and placed at the bottom of the containment vessel for maximum conservatism. The results presented in this report rectify the above-noted deficiencies by adjusting the previously determined mass limits so that the resulting dose rates are compliant with regulatory limits. Some amount of iteration was required in some instances to adjust the masses for dose compliance. In some instances the masses were increased since they were too conservative. In addition, the method of ratioing the masses when applied to several gamma emitting isotopes also resulted in either non-conservative or overly conservative data. The mass imits presented here are for bare sources and are complaint with the design decay heat limit of 19W. The masses are based on package surface dose rates (the limiting dose rate) that fall in the range 185-195 mrem/h, thus giving an additional margin of between 7.5% and 2.5% to the regulatory limit of 200 mrem/h. This report presents a revised set of mass limits (i.e., content envelope) for several neutron emitting actinides with varying levels of light element impurities compliant with both external radiation and design decay heat limits. In addition, revised limits for gamma sources are also presented. The neutron emitter mass limits are in Tables 1 through 9 and the gamma emitter mass limits are in Table 10. The revised content envelope covering a wide range of materials present in the DOE complex, was developed for the Model 9977 Packaging, providing conservative limits for other commonly-used Type B Packagings such as the Models 9975 and 9978. It should be noted that for the Model 9977 package, shipment is also possible with three shielded containers – lead or tungsten for gamma sources and polyethylene for neutron emitters. The design decay heat limits for these are as follows: 19W for tungsten, 6W for lead and 3W for polyethylene. The limits presented in this report would need to be checked against the lower decay heat limits for the lead and polyethylene and adjusted as required for compliance. Criticality safety is not factored into these limits and should be evaluated on a case by case basis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

LLNL HEPA Filter Support of Nuclear Facility Safety R&D (FY12)

The ceramic HEPA filter project has been a successful and innovative project. This report contains information from the work from FY12 funding, as well as remaining FY11 funding. Of note, a total of 12 patents for the ceramic HEPA filter program have been filed, and an additional record of invention has been prepared with a provisional patent to follow. This project supports development of a ceramic HEPA filter technology to benefit DOE nuclear facilities by providing lower life-cycle costs and reducing or eliminating costs associated with Safety Class and Safety Significant systems in nuclear facilities. The satisfactory performance of ceramic filters in a fire could significantly reduce costs of support systems associated with mitigating a release, such as fire suppression, fire detection and alarm, and internal building structure. In nuclear safety, the best way to provide protection to public, workers and the environment is to provide a barrier between the hazard and the receptors. A passive engineered design feature is the best barrier for that purpose. Advanced ceramic HEPA filters provide an excellent barrier between radioactive material in a nuclear facility and everything downwind. Both university (Cal Poly) student teams did very well. As summarized in this report, they successfully added and upgraded instrumentation and controls to the High Temperature Test Unit (HTTU) for the ceramic filter project. They also succeeded in adding viewing ports, remotely controlled closed circuit television, a flame impingement system and a helium filter gasket leak detection system. Details can be found in Ref. 3 - 6. The HTTU has flexible capabilities and can also be used to test other types of filters or filter components in the future, if so desired. At LLNL, research to improve filter media coating (Fig. 1) is being conducted to maintain filtration efficiency while improving performance (reducing pressure drop). Improvements in the individual filter element are necessary to maintain filtration efficiency while decreasing pressure drop. The path forward for an optimal ceramic HEPA filter meeting DOE’s needs is to improve the filtration media. This task utilized the LLNL/DOE developed innovations along with progress in previous activities to prepare multiple ceramic filter elements and refine the technical details for depositing advanced filtration materials to the ceramic filter to meet desired filter performance. The ceramics filter elements were tested for pressure drop at the required air flow rate using an air flow test system. The filtration efficiency and pressure drop of the ceramic filter elements will be measured as part of the development process.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

DOE 2012 Occupational Radiation Exposure October 2013

The U.S. Department of Energy (DOE) Office of Analysis within the Office of Health, Safety and Security (HSS) publishes the annual DOE Occupational Radiation Exposure Report to provide an overview of the status of radiation protection practices at DOE (including the National Nuclear Security Administration [NNSA]). The DOE 2012 Occupational Radiation Exposure Report provides an evaluation of DOE-wide performance regarding compliance with Title 10, Code of Federal Regulations (C.F.R.), Part 835, Occupational Radiation Protection dose limits and as low as reasonably achievable (ALARA) process requirements. In addition, the report provides data to DOE organizations responsible for developing policies for protection of individuals from the adverse health effects of radiation. The report provides a summary and an analysis of occupational radiation exposure information from the monitoring of individuals involved in DOE activities. Over the past 5-year period, the occupational radiation exposure information is analyzed in terms of aggregate data, dose to individuals, and dose by site. As an indicator of the overall amount of radiation dose received during the conduct of operations at DOE, the report includes information on collective total effective dose (TED). The TED is comprised of the effective dose (ED) from external sources, which includes neutron and photon radiation, and the internal committed effective dose (CED), which results from the intake of radioactive material into the body. The collective ED from photon exposure decreased by 23% between 2011 and 2012, while the neutron dose increased by 5%. The internal dose components of the collective TED decreased by 7%. Over the past 5-year period, 99.99% of the individuals receiving measurable TED have received doses below the 2 roentgen equivalent in man (rems) (20 millisievert [mSv]) TED administrative control level (ACL), which is well below the DOE regulatory limit of 5 rems (50 mSv) TED annually. The occupational radiation exposure records show that in 2012, DOE facilities continued to comply with DOE dose limits and ACLs and worked to minimize exposure to individuals. The DOE collective TED decreased 17.1% from 2011 to 2012. The collective TED decreased at three of the five sites with the largest collective TED. u Idaho Site – Collective dose reductions were achieved as a result of continuing improvements at the Advanced Mixed Waste Treatment Project (AMWTP) through the planning of drum movements that reduced the number of times a container is handled; placement of waste containers that created highradiation areas in a centralized location; and increased worker awareness of high-dose rate areas. In addition, Idaho had the largest decrease in the total number of workers with measurable TED (1,143 fewer workers). u Hanford Site (Hanford) – An overall reduction of decontamination and decommissioning (D&D) activities at the Plutonium Finishing Plant (PFP) and Transuranic (TRU) retrieval activities resulted in collective dose reductions. u Savannah River Site (SRS) – Reductions were achieved through ALARA initiatives employed site wide. The Solid Waste Management Facility used extended specialty tools, cameras and lead shield walls to facilitate removal of drums. These tools and techniques reduce exposure time through improved efficiency, increase distance from the source of radiation by remote monitoring, shield the workers to lower the dose rate, and reduce the potential for contamination and release of material through repacking of waste. Overall, from 2011 to 2012, there was a 19% decrease in the number of workers with measurable dose. Furthermore, due to a slight decrease in both the DOE workforce (7%) and monitored workers (10%), the ratio of workers with measurable doses to monitored workers decreased to 13%. Another primary indicator of the level of radiation exposure covered in this report is the average measurable dose, which normalizes the collective dose over the population of workers who actually received a measurable dose. The average measurable TED increased by 3% from 2011 to 2012. Additional analyses show that the dose distribution in 2012 was similar to the distribution in 2011. In 2012, 13% of the monitored workers received a measurable TED and the average measurable TED, 0.069 rem, was less than 2% of the DOE limit. From 2011 to 2012, the collective TED and the number of individuals with measurable TED decreased 17.1% and 19%, respectively. These decreases were mainly due to an overall reduction of D&D activities at the PFP and TRU retrieval activities at Hanford; a 78% decrease in the number of targeted waste drums that were processed at the Idaho Site’s Accelerated Retrieval Project (ARP) from 5,566 drums in 2011 to a total of 1,211 drums processed in 2012; and ALARA initiatives employed site wide at SRS. In addition, the decreases were the result of decreased American Recovery and Reinvestment Act (ARRA) activities and continuing D&D, particularly at the DOE sites that comprise the majority of DOE collective dose. Over the past 5 years, the size of the monitored workforce has remained at a fairly stable level (within 12%), while the collective dose has varied up to 37%. No reported doses exceeded the DOE occupational limit of 5 rems TED in 2012 and no reported doses exceeded the DOE ACL of 2 rems TED.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

Recommended Parameter Values for GENII Modeling of Radionuclides in Routine Air and Water Releases

The GENII v2 code is used to estimate dose to individuals or populations from the release of radioactive materials into air or water. Numerous parameter values are required for input into this code. User-defined parameters cover the spectrum from chemical data, meteorological data, agricultural data, and behavioral data. This document is a summary of parameter values that reflect conditions in the United States. Reasonable regional and age-dependent data is summarized. Data availability and quality varies. The set of parameters described address scenarios for chronic air emissions or chronic releases to public waterways. Considerations for the special tritium and carbon-14 models are briefly addressed. GENIIv2.10.0 is the current software version that this document supports.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Characterization of Bond Strength of U-Mo Fuel Plates Using the Laser Shockwave Technique: Capabilities and Preliminary Results

This report summarizes work conducted to-date on the implementation of new laser-based capabilities for characterization of bond strength in nuclear fuel plates, and presents preliminary results obtained from fresh fuel studies on as-fabricated monolithic fuel consisting of uranium-10 wt.% molybdenum alloys clad in 6061 aluminum by hot isostatic pressing. Characterization involves application of two complementary experimental methods, laser-shock testing and laser-ultrasonic imaging, collectively referred to as the Laser Shockwave Technique (LST), that allows the integrity, physical properties and interfacial bond strength in fuel plates to be evaluated. Example characterization results are provided, including measurement of layer thicknesses, elastic properties of the constituents, and the location and nature of generated debonds (including kissing bonds). LST provides spatially localized, non-contacting measurements with minimum specimen preparation, and is ideally suited for applications involving radioactive materials, including irradiated materials. The theoretical principles and experimental approaches employed in characterizing nuclear fuel plates are described, and preliminary bond strength measurement results are discussed, with emphasis on demonstrating the capabilities and limitations of these methods. These preliminary results demonstrate the ability to distinguish bond strength variations between different fuel plates. Although additional development work is necessary to validate and qualify the test methods, these results suggest LST is viable as a method to meet fuel qualification requirements to demonstrate acceptable bonding integrity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

RADTRAN 6 technical manual.

This Technical Manual contains descriptions of the calculation models and mathematical and numerical methods used in the RADTRAN 6 computer code for transportation risk and consequence assessment. The RADTRAN 6 code combines user-supplied input data with values from an internal library of physical and radiological data to calculate the expected radiological consequences and risks associated with the transportation of radioactive material. Radiological consequences and risks are estimated with numerical models of exposure pathways, receptor populations, package behavior in accidents, and accident severity and probability.

Weiner, Ruth F.↗
DOE OSTI2014

Recommended values for the distribution coefficient (Kd) to be used in dose assessments for decommissioning the Zion Nuclear Power Plant

ZionSolutions is in the process of decommissioning the Zion Nuclear Power Plant. The site contains two reactor Containment Buildings, a Fuel Building, an Auxiliary Building, and a Turbine Building that may be contaminated. The current decommissioning plan involves removing all above grade structures to a depth of 3 feet below grade. The remaining underground structures will be backfilled. The remaining underground structures will contain low amounts of residual licensed radioactive material. An important component of the decommissioning process is the demonstration that any remaining activity will not cause a hypothetical individual to receive a dose in excess of 25 mrem/y as specified in 10CFR20 SubpartE.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2014

Nearest Neighbor Averaging and its Effect on the Critical Level and Minimum Detectable Concentration for Scanning Radiological Survey Instruments that Perform Facility Release Surveys.

Through the SNL New Mexico Small Business Assistance (NMSBA) program, several Sandia engineers worked with the Environmental Restoration Group (ERG) Inc. to verify and validate a novel algorithm used to determine the scanning Critical Level (L c ) and Minimum Detectable Concentration (MDC) (or Minimum Detectable Areal Activity) for the 102F scanning system. Through the use of Monte Carlo statistical simulations the algorithm mathematically demonstrates accuracy in determining the L c and MDC when a nearest-neighbor averaging (NNA) technique was used. To empirically validate this approach, SNL prepared several spiked sources and ran a test with the ERG 102F instrument on a bare concrete floor known to have no radiological contamination other than background naturally occurring radioactive material (NORM). The tests conclude that the NNA technique increases the sensitivity (decreases the L c and MDC) for high-density data maps that are obtained by scanning radiological survey instruments.

Fournier, Sean Donovan↗
DOE OSTI2014

GADRAS Detector Response Function.

The Gamma Detector Response and Analysis Software (GADRAS) applies a Detector Response Function (DRF) to compute the output of gamma-ray and neutron detectors when they are exposed to radiation sources. The DRF is fundamental to the ability to perform forward calculations (i.e., computation of the response of a detector to a known source), as well as the ability to analyze spectra to deduce the types and quantities of radioactive material to which the detectors are exposed. This document describes how gamma-ray spectra are computed and the significance of response function parameters that define characteristics of particular detectors.

Mitchell, Dean J.↗