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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

Comparison of Modeled to Measured Spectra using MCNP and GADRAS to Benchmark and Contrast Modeling Limitations

The desire to improve nuclear material detection through portal monitors and other low resolution detectors has led to interest in benchmarking the performance of radiation transport codes. These codes can be used to generate a variety and quantity of spectra that may be cost prohibitive to measure directly. Particular characteristics of typical detection scenarios were isolated to compare the performance of radiation transport codes to laboratory experiments. These benchmark experiments were performed to validate simulations in a number of key areas. These experiments included high areal density configurations in both one- and three-dimensional configurations. In addition, off-angle measurements were conducted to investigate the impact of detector response assumptions on complex geometries, for example where an unknown source is not physically collocated with an apparent hot spot. Furthermore, by examining both simple shielding and backscatter configurations, isolation of the scatter emanating from the object and environmental background scatter contributions to the spectra could be elucidated. A series of experiments were performed in which the order of shielding layers and the thickness were evaluated to examine the ability of the simulations to address these variables. Experiments with increasing thickness of polyethylene around a neutron source also allows for examinations of the ability of the transport simulations to properly model n,γ reactions. Finally, depleted uranium (DU) measurements have more complicated spectra than the cobalt-60 and consequently allow for investigations into the ability of transport simulations to model self-shielding. In each case, the results of the experiments were compared to MCNP simulations to judge the performance of the code, as was done previously with GADRAS 18.7.9 simulations [7]. All of the benchmark measurements were taken with a liquid nitrogen cooled 140% high purity germanium (HPGe) detector with a bismuth collimator that had the front tin filter removed. The measurement location and detector configuration were subsequently used for all experimental configurations. The remainder of this report details the experimental measurements, Section 2 and the MCNP model, Section 3. Comparisons between the experimental results and MCNP-generated spectra are performed, Section 4. Section 5 compares the results obtained with MCNP with the previously reported results obtained with GADRAS [2]. Finally, Section 6 presents the conclusions.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Evaluating Precision of Measured Radionuclide Ambient Air Concentrations

According to AIRNET – Quality Assurance Project Plan for the Radiological Air Sampling Network, SOP-5140, Revision 2 (to become EPC-CP-PIP-5140): “The regulation 40 CFR Part 61.93 requires that the system be able to readily detect a dose of 1.0 mrem above background. Following statistical principles and assuming 95% confidence for such a measurement, we require that 95% of all measurements must fall within two standard deviations of the mean. This represents the minimum acceptable precision for decision making.” An independent assessment of the Rad-NESHAP program completed in early 2019 found that no documentation existed proving that the Airnet data meets the precision defined above. Therefore, to address this particular finding, Airnet concentration data was pulled from the Environmental Information Management (EIM), a cloud-based database maintained by Locus Technologies, for analysis.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

WIPP MLU Incoming Vehicles Survey Document

The following SNL document contains requested radiological survey information, as part of the documentation for the TRU-waste shipment being performed by the LANL MLU team. The surveys were performed in TA-5, on July 20th, 2020: Incoming WIPP TRU-Pact trailers survey: 1-20200720-4, and, Incoming MLU trailer and mobile crane survey: 1-20200720-8 The first survey was of 2 WIPP trailers carrying 2 empty TRU Half-Pacts each. The second survey was for a trailer with the MLU team's equipment, and for the contracted mobile crane that be used for loading Half-Pacts. All information contained was completed, reviewed, and approved by SNL RP personnel, and is released for receipt and use by the LANL MLU Team and WIPP personnel at their discretion.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

(U) Preliminary Notes on Bondarenko Self-Shielded Cross Section Sensitivities

The Bondarenko method for resonance self-shielding was recently implemented in the PARTISN multigroup discrete ordinates code. This report derives the sensitivities of self-shielded cross sections with respect to nuclide densities and the subsequent sensitivities of k eff with respect to nuclide densities. Numerical results are presented for two test problems.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

In-Device Delay: Introduction to Obscurant Physcial Protection System

The Office of Radiological Security (ORS) In-Device Delay (IDD) program has undertaken a project to research and develop a novel protection system for industrial irradiators that contain high-activity Co-60 sources. Based on adversary testing conducted by ORS, it is was determined that to successfully accomplish the theft of the target material, the adversary will require visual contact of the sources and source rack located at the bottom of the pool. Therefore, if a means of obscuring or visually hiding the sources in the pool can be achieved (while adhering to facility operations, safety, and regulatory requirements), then illicit source theft will be significantly hindered. This project aims to develop a low-cost, non-propriety obscurant that, when an adversary action is detected, the obscurant will be deployed into the pool quickly, rendering visual observation of the source problematic; however, this obscurant will not otherwise disturb the sources, source rack, and filtration system. The obscurant will remain in the pool until removed by another process.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Implementing an aerosol dynamic shape factor for cerium oxide powder

The shape factor (X B ) of bulk powder CeO 2 (cerium oxide) particles is estimated to be X B = 1.34 (dimensionless). This is defined in relation to the measured aerated bulk density (0.91 g cm -3 , Hosokawa Micron, Summit NJ, 2018) of the CeO 2 powder. This information is used to determine the RRF (respirable release factor), and other parameters during drop test experiments of nuclear material storage containers at Los Alamos National Laboratory. The calculation is based on an experimental shape factor (X P = 2.66 dimensionless) for CeO 2 particles of the same (chain aggregate) morphology studied by other researchers. In that work, the shape factor was defined in terms of the maximum possible (compacted) density of CeO 2 material (7.13 g cm -3 ).

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DOE OSTI2020

Estimating a protection factor from homemade facemasks for Los Alamos radioactive aerosols

The Los Alamos National Laboratory has required usage of facemasks “whenever interacting with or near others." The present document estimates the respiratory protection factor of homemade facemasks to be 1.4 and 1.2 for aerosol particles, below and above 0.3 μm diameter, respectively. This includes an estimated leakage of 70% and 52% for aerosol particles below and above 0.3 μm diameter, respectively.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Groundwater and Intruder Radionuclide Screening

Solid Waste Management (SWM) operates the E-Area Low-Level Waste Facility (ELLWF) where low level on-site and off-site solid waste streams are buried. The facility has been in operation since late 1994 and is currently projected to remain in operation until 2065. This facility can accommodate a broad range of waste forms resulting from the six different types of disposal unit options (i.e., varying degrees of engineered barriers → trenches to concrete vaults). This facility is currently operating under a Performance Assessment (PA) issued back in 2008, along with several subsequent supporting Special Analyses (SA). The Savannah River National Laboratory (SRNL) developed the prior PAs and SAs and has been tasked to update the facility’s PA. For operating the E-Area facility, a Waste Inventory Tracking System (e.g., WITS) is actively employed by waste generators where every radionuclide entering the facility (to be buried in one of its many disposal units1) must be either directly or indirectly accounted for. Since there is a large number of radionuclides in existence (>3,000), the International Commission on Radiological Protection (ICRP-107) has provided guidance on the subset of radionuclides requiring further assessment in landfills such as the E-Area LLWF. The ICRP-107 publication provides critical radiological information on 1,252 radionuclides of 97 elements. This database is the critical starting point for developing a consistent inventory limit system. The explicit measurement and tracking of all 1,252 ICRP-107 radionuclides can be reduced when process knowledge, burial history, and radiological aspects are factored into conservative groundwater and intruder screening processes. Across the DOE complex these screening processes have been historically performed using the methodology suggested by the National Council on Radiation Protection and Measurements (NCRP) as presented in their original report (NCRP 1984) and then refined in a later report (NCRP 1996). In the recommended screening models employed within this report the traditional NCRP models are updated to better handle progeny and better reflect the known characteristics of E-Area. This improved screening process is referred to as the “NCRP-like” method. Various upgrades to the traditional NCRP methodology have been used by others (e.g., NRC funded effort by Kennedy and Strenge 1992). The more detailed models by Kennedy and Strenge (1992) were considered in this effort and are compared to the results from the more traditional NCRP-like models. The groundwater and inadvertent intruder screening analyses presented in this report start with this 1,252 radionuclide list and reduces it down to more manageable lists that are applicable to the various disposal unit types contained within E-Area. In order to reduce this starting list, some level of exposure risk must be considered acceptable. Historically, a dose (or concentration level) has been compared with a screening criterion set to 1% of a performance measure (e.g., a betagamma dose not to exceed the 4 mrem/yr beta-gamma performance measure x 0.01 = 0.04 mrem/yr). Thus, if a radionuclide produced a bounding or screening-level dose (or concentrations) less than the screening criterion, it could be safely removed from further consideration. Several tiers of screening and bounding level analyses have been considered in this report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Evaluation of Detritiation Strategies for Mitigating Tritium Releases

Tritium confinement is performed using different barriers to minimize releases to the environment. Primary confinement is usually performed by process piping and components, secondary confinement typically by inerted gloveboxes connected to a continuously operating tritium stripper system. Some facility designs employ tertiary tritium confinement by using process room confinement with a tritium stripper system activated after an accident to further mitigate tritium releases to the environment. Secondary confinement of some tritium systems such as radiation-hardened or shielded-cell enclosures are not easily achieved using designs for inerted glovebox. These enclosures can be challenging to seal for secondary tritium confinement due to heating, ventilation, and air conditioning (HVAC) needs for temperature control, as well as feedthroughs for equipment and instrumentation, are potential pathways for tritium leaks or loses. Additional tritium leak paths can be created in secondary confinement enclosures due to a Design Basis Accident (DBA) seismic event. These new, larger tritium leak rates from the DBA event are usually not part of the stripper system design basis so all tritium released to the secondary confinement enclosure is assumed released to the environment. Detritiation strategies for shielded enclosures and/or gloveboxes with significant leak rates are necessary to minimize off-site radiological dose consequences. To address this, the relationship between the confinement system leak rate and stripper system (recirculating) flow rate was evaluated to meet prescribed maximum allowable environmental tritium emissions from the event. The simple analysis described assumes a constant leak rate from the confinement system (i.e. shielded enclosure or glovebox) while a recirculating stripper system strips tritium from the system– a competition between tritium release and tritium recovery. At a high level, the resulting expression summarizes the relationship between system leak rate (F L ) and stripper flow rate (F S ) relative to allowable tritium release (Q A ) and initial tritium release (Q 0 ): $\frac{F_L}{F_s}$ ≈ $\frac{Q_A}{Q_0}$. This report derives the relationship between these parameters, examines the impact of finite stripping times followed by purging of the stripped volume. The analysis provides example results for up to 30 gram tritium releases: the maximum tritium inventory of a Hazard Category III (Department of Energy) Nuclear Facility. The detritiation model presented represents a high-level analysis of tritium recovery versus losses for “leaky” confinement enclosures such as shielded cells after large tritium releases. Key parameters for doing the analyses are the system leak rate, stripper flow rate, initial tritium release, and allowable tritium release values. The analyses show a proportional decrease in releases with reduction in initial tritium release but a non-linear increase in tritium recovery with increased stripper flow rate or reduced leak-rate. The analyses also apply to glovebox confinement systems which could have significant increases in leak rates after a DBA. The analysis recognizes but does not include tritium absorption followed by re-emission from the walls of the confinement volume – an analysis which could be pursued in future studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development of FRET clusters for CBRN Detection

Advanced sensor capabilities for the simultaneous on-site detection of specific chemical, biological, and radiological/nuclear (CBRN) threats is important to significantly limit the risk of exposure to personnel and allow the rapid collection of essential scientific data and critical evidence. Commercially available sensor capabilities are generally complex, and/or require highly specific and ultra-sensitive methods that are often power demanding or require offsite post-analysis for positive detection, leaving personnel vulnerable. The goal of this project is to develop a portable sensor capable of simultaneously detecting CBRN signatures using a multiplexed Förster Resonance Energy Transfer (FRET) based sensor. FRET sensors are tailorable, specific, and highly dependent on the donor-acceptor distance. In this project, nanoparticles were functionalized with aptamers that were designed for the detection of methylphosphonate, a sarin metabolite. FRET was measured between quantum dot donors and dye and metal nanoparticle acceptors using optical spectroscopy.

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DOE OSTI2020

Final Status of HEPA Filter 10-Year Lifetime Evaluation

High-efficiency particulate air (HEPA) filters are widely employed by nuclear facilities to remove radiological particulate matter from their effluent exhaust streams. The purpose of this study is to evaluate the relationships between the 10-year HEPA filter lifespan and its other performance indicators. The 10-year-long endeavor to collect and analyze data regarding the lifetime of HEPA filters at the Pacific Northwest National Laboratory began in 2010. Forty-nine HEPA filters were selected and have been surveyed and analyzed at least annually to verify compliance with permit conditions. The study suggests the frequency of filter replacement should be based on the actual operational requirements, such as fume hood face velocity and/or efficiency test results, instead of on the prescribed filter “age limit” of 10 years from the date of manufacture (e.g., birth date) when operating under dry conditions. Over the past decade, the study has now been completed, and all forty-nine HEPA filters have been replaced, due to either technical issues listed in this report, or the previously recommended filter “age limit” of 10 years.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Validation of Spent Fuel Dose Rate Calculations Using NDA Measurement Campaign Data

The predicted radiation gamma dose rate is a key parameter to determine if a spent nuclear fuel assembly is self-protected from misuse as well as an indicator for worker/inspector safety. Recent studies showed that using state-of-the-art techniques for dose rate calculations, including the one adopted for this study, the numerical calculations of the gamma dose rates are three times lower than those from the most commonly-cited reference values for spent LWR fuel used by domestic and international agencies. With only a few publically available experimental data points worldwide for this type of benchmark study, the gamma flux measurements from the Spent Fuel Non Destructive Assay (NDA) measurement campaign can be used to help validate the physics approach used in calculating gamma dose rates. Specifically, this data would validate the gamma flux calculation from spent fuel, which is a key component of these multi-physics techniques for dose rate assessments; dose rate measurements were not part of the NDA campaign. These data are primarily the passive gamma measurements of the spent nuclear fuel assemblies from the Swedish Nuclear Fuel and Waste Management Company (SKB).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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

West Valley Demonstration Project Annual Site Environmental Report (CY2019)

The report, prepared for the U.S. Department of Energy West Valley Demonstration Project office (DOE-WVDP), summarizes the environmental protection program at the WVDP for calendar year (CY) 2019. Monitoring and surveillance of the facilities used by the DOE are conducted to verify protection of public health and safety and the environment. The report is a key component of DOE’s effort to keep the public informed of environmental conditions at the WVDP. The quality assurance protocols applied to the environmental monitoring program ensure the validity and accuracy of the monitoring data. In addition to demonstrating compliance with environmental laws, regulations, and directives, evaluation of data collected in 2019 continued to indicate that WVDP activities pose no threat to public health or safety, or to the environment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

2019 Nevada National Security Site Environmental Report Summary

This document is a summary of the full 2019 Nevada National Security Site Environmental Report (NNSSER) prepared by the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office. This summary provides an abbreviated and more readable version of the full NNSSER.

54 ENVIRONMENTAL SCIENCES↗