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This document is a resource for ongoing work. This describes calculations for error propagation and test uncertainty of the LANL FTS Filter Test System for nuclear storage containers.
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.
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).
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.
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.
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.
Objectives of this presentation are: 1. Explain the difference between radiation safety vs. criticality safety, 2. Define criticality, 3. Explain the three types of chain reactions, 4. Define two terms scientists use to characterize criticality, 5. Explain what occurs during a criticality accident, 6. Identify prevention measures for avoiding a criticality accident, and 7. Identify the parameters that effect criticality of a system.
This data package summarizes the exposure assumptions, equations, and methods used to calculate radionuclide-specific unit dose factors and the radiological doses for both groundwater and atmospheric pathways as a part of the updated Hanford Site Composite Analysis (CA). An All Pathways Representative Person exposure scenario is considered to evaluate exposure via both groundwater and atmospheric transport pathways. The radiological dose assessments for both groundwater and atmospheric pathways are included in the performance assessments (PAs) for various Waste Management Areas at the Hanford Site. This data package calculates exposure route-specific and total unit dose factors for CA-specific radionuclides of concern based on the exposure assumptions used in the CA and PAs. This data package also presents the results and comparison of the radionuclide-specific unit dose factors based on the exposure assumptions used in the revised Hanford Site Composite Analysis and Hanford Site performance assessments for low-level waste disposal facilities.
The Federal Radiological Monitoring and Assessment Center (FRMAC) Assessment Manual is the tool used to organize and guide activities of the FRMAC Assessment Division. The mission of the FRMAC Assessment Division in a radiological emergency is to interpret radiological data and predict worker and public doses. This information is used by Decision Makers to recommend protective actions in accordance with Protection Action Guides (PAGs) issued by government agencies. This manual integrates many health physics tools and techniques used to make these assessments.
This document combines and supersedes earlier reports (e.g. LA-UR-19-32358 and LA-UR-20- 22216) that discussed MCNP-calculated conversion factors (count rate per unit activity per unit area) for collimated 2”x2” NaI detectors (Ludlum model 44-10 and Thermo model SPA3) in contact with various surfaces of interest to the University of Washington 137 Cs release event. These surfaces include: Exterior surface of round and rectangular ducts, Exterior surface of outdoor brick work, Exterior surface of cast iron pipes, Exterior surface of an O 2 filter assembly For each surface, the calculated conversion factors allowed inferences on 137 Cs activity on interior surfaces based on exterior contact measurements of the 662 keV photopeak count rate. The MCNP calculations included the effect of attached collimators (i.e. either a Pb/Sb-based Ludlum or a W-based “Rees” model) on detector response. As a consequence, the conversion factors only apply to the appropriate collimated 2”x2” NaI detector operated in pulse height analysis (PHA) mode. Other assumptions affecting the calculations and valid application of the conversion factors will be outlined below.
Washington River Protection Solutions (WRPS) tested four types of chemical cartridges for use in air-purifying respirators (APR) and powered air-purifying respirators (PAPR). These tests were undertaken to determine the period of time that the cartridges would provide adequate performance1 for APRs and PAPRs used to protect workers when exposed to a mixture of Chemicals of Potential Concern (COPC) from vapors exiting the Hanford AP tank farm exhauster slipstream. The Occupational Safety and Health Administration (OSHA) considers cartridge testing to be a valid approach for establishing cartridge change schedules. Testing commonly is applied in situations where mixtures of COPCs exist and where other approaches, such as manufacturer recommendations and modeling, are less reliable. The tests were designed and conducted to assure measurement and/or control of the key variables OSHA identified as important to estimate the cartridge service life, including temperature, humidity, COPC concentration, breathing rate, and cartridge adsorption capacity. Cartridge testing on a slipstream from the Hanford AP tank exhauster was conducted on March 23-24, 2018. This testing focused on both APR and PAPR cartridges. Previous testing of APR cartridges was conducted on the AP exhauster in June of 2016. However, an AP exhauster upgrade was completed in September 2016. In the most recent testing, slipstream vapors from the new AP exhauster were fed to two respirator cartridge test stands, one for the PAPR respirator cartridges and the other for the APR respirator cartridges. Both the APR and PAPR test stands were developed by WRPS in collaboration with HiLine Engineering (Richland, Washington). Multipurpose APR cartridges—SCOTT 7422-SD1 and SCOTT 7422-SC1 (SCOTT Safety, Monroe, North Carolina)—were assessed on separate days using the APR cartridge test stand. Multipurpose PAPR cartridges—MSA OptiFilter TL (MSA Safety Inc., Pittsburgh, Pennsylvania) and 3M FR-57 (3M Company, Maplewood, Minnesota)—were also tested consecutively over the same two days as the APR cartridge tests, using the PAPR cartridge test stand. Sample media (i.e., sorbent tubes) were used to collect samples of the vapor stream entering and exiting the respirator cartridges, and the samples then were analyzed for COPC concentrations. Pacific Northwest National Laboratory was tasked with conducting an independent analysis of the analytical results and making recommendations based on the results for respiratory cartridge performance and service life. The key conclusions from the analysis are described below.
This procedure provides methods for determining the appropriate level of worker protection from potential or actual airborne radioactivity, either through engineered controls (workplaces for radionuclides) or respiratory protection.
The presentation discusses what is involved in having radiation control as a career.
I use crater dimensions to estimate the yield of the August 4th, 2020 Beirut explosion to be equivalent to approximately 1.4 kilotons of TNT with a lower bound of about 0.7 kilotons. Based on the amount of ammonium nitrate reported to have been stored at the Beirut harbor, I assume an upper bound for the yield of 2.75 kilotons. However, it is highly likely that the yield was less than 2.75 kilotons, since reported values for TNT equivalence of ammonium nitrate are typically much less than one hundred percent. The crater-size based yield estimates are based on crater radius estimates from satellite imagery and empirical curves and data for scaled crater radius from past chemical and nuclear explosions. I present evidence that suggests that the relatively large crater radius is due to a high degree of coupling of shock wave energy to the surrounding medium and a reduction of the effective stress because of a high level of saturation of the geologic media beneath the explosion. I provide yield estimates based on seismic body-wave magnitude and crater depth as corroborating evidence. I compare preliminary estimates for the maximum debris cloud height, based on cell phone videos/images, with predicted maximum heights for this yield range from empirical formulas and numerical cloud-rise models. Based on a preliminary analysis of cell phone footage, the observed maximum cloud height appears to be approximately 1600 m. This is much lower than that predicted using standard empirical formulas and buoyant cloud rise models. I present results from a modified buoyant cloud rise model that more accurately predicts the maximum cloud height by allowing for the inclusion of a fixed amount of air and/or water into the fireball at the start of cloud rise. The amount of mass that needs to be added at the start, to reproduce the observed maximum cloud height, is relatively small compared to the total mass entrained during cloud rise. A much greater amount of dry air or debris is required, relative to water, for an equivalent reduction in maximum cloud height. The ammonium nitrate is one possible source for water in the fireball since it was being stored in a very humid environment and ammonium nitrate is known to be hygroscopic. The ground beneath the explosion, especially if it were saturated, and the nearby harbor could also have been sources for water or debris in the fireball.
Bladewerx TM LLC (Rio Rancho, NM) manufactures instrumentation, neutron shielding and activation foils for the radiation protection industry. Specializing in portable alpha/beta air monitors and sample counters, Bladewerx is the source of Speclon TM PTFE filter media that they recommend for high-resolution alpha spectroscopy. Los Alamos National Laboratory (LANL) utilizes Speclon TM filter material in CAM (Continuous Air Monitor) samplers for workplace air monitoring. The LANL Aerosol Engineering Facility received air filter material from Bladewerx, referred to as “New Speclon 5” in this document, in order to distinguish from filter material that was previously received (referred to as “Speclon 5” in this document). In this document, the aerosol collection efficiency and airflow resistance (pressure drop) were measured for the New Speclon 5 filter material.
This report documents the findings of an assessment of the Turbo FRMAC software's ability to implement International Atomic Energy Agency (IAEA) guidance for calculating Operational Intervention Levels (OIL) 1 & 2 for nuclear and radiological emergencies. The IAEA OIL and U.S. Federal Radiological Monitoring and Assessment Center (FRMAC) Derived Response Level methodology and implementation in respective tools were compared, as demonstrated through benchmarking activities for a nuclear power plant source term and potential radionuclides of concern for radiological dispersal devices. This comparison revealed some shortcomings in Turbo FRMACs ability to perform IAEA OIL calculations and resulted in recommended software modifications to be considered for future development.
In support of the E-Area Performance Assessment, a two-dimensional model of water flow and radionuclide transport through the E-Area Intermediate Level Vault (ILV) and local vadose zone has been developed using the PORFLOWTM software. The purpose of the model is to calculate flux to the water table for radionuclides eluted from the ILV during its operational life, the period of institutional control, and times following site closure. Results of model calculations will be used by a three-dimensional PORFLOW model of transport through the aquifer to determine radionuclide concentrations at a hypothetical 100 meter well and at the site boundary where contaminated groundwater could be assessed by members of the public following site closure. While newly developed, the model structure closely follows that used in the 2008 PA while incorporating a refined computational mesh, updated material properties for the vadose zone soil and vault concrete, revised infiltration rates, and new solid-liquid distribution coefficients. The model also addresses degradation of concrete hydraulic properties by blending soil and concrete water retention curves over a 500 year period. That is, the hydraulic properties of ILV concrete start out as fresh concrete at the time of site closure and degrade to soil properties over a 500 year time period. This approach has not been used previously in E-Area PA’s. The model has been used to calculate water flow through the ILV vadose zone and, on a trial basis, transport of a limited number of radionuclides from the waste region inside the ILV to the water table. Figure ES-1 provides a view of the model and an example of water flow after the concrete has substantially degraded. Figure dimensions are in centimeters. The red area in the left hand side figure is the waste disposal region within the ILV. Colored bands in the figures indicate different soil and concrete regions. Figure ES-2 shows a calculation of flux to the water table (mol/yr) for U-234 and its one-year half-life decay chain (U-234 → Th-230 → Ra-226 → Pb-210). Pb-210 flux is too small to show on the same scale used for Th-230 and Ra-226 in Figure ES-2. The model has also been used to make a preliminary evaluation of the release of tritium from TPBAR disposal containers that will be place in the ILV. A separately calculated tritium release rate as a function of time was used as a source in the PORFLOW model. The TPBAR disposal schedule and tritium release rate differ from the values used in the 2008 PA. Results from these initial trial runs are compared to calculations made for the 2008 PA ILV analysis. In brief, a new PORFLOW model of flow and transport through the ILV vadose zone has been developed and tested. The model is ready to be used for PA calculations.
The Sri Lanka Atomic Energy Regulatory Commission (AERC) and the United States Department of Energy (U.S. DOE) co-hosted the 6th Regional Review Meeting on Radiological Security involving representatives from over 20 countries, the International Atomic Energy Agency (IAEA), the International Criminal Police Organization (INTERPOL), and the World Institute for Nuclear Security. The purpose of the event was to discuss the implementation of, and plans for, high-activity radioactive source security (RSS). The U.S. DOE’s National Nuclear Security Administration (NNSA) Office of Radiological Security (ORS) fully sponsored this review meeting. Participants were welcomed to Colombo, Sri Lanka, and the meeting was formally opened by Nirmali Karunarathna of the Sri Lanka Atomic Energy Regulator Council who emphasized the strong partnerships among the participants. The opening Ceremony and Lamp Lighting included dignitaries from the sponsoring countries. Robert Hilton, Deputy Chief of Mission at the U.S. Embassy in Colombo, and Kristin Hirsch of the ORS gave other opening remarks that highlighted the social benefit from radiological sources in medicine, industry, and agriculture, while stressing the importance of addressing the risks associated with the malicious use of radiological sources. Emphasis was placed on the importance of partnerships among all the participants to help ensure the success of securing radiological materials throughout the world and the opportunity to share information and experiences. AERC was acknowledged with special thanks for hosting this event. A participant list is included as Attachment A, and the meeting agenda is provided as Attachment B. All presentations were made available to participants. The following sections summarize the meeting’s presentations, discussions, issues, suggestions, and recommendations.