Engineering topics

Kamboj, Sunita

Publications and source records attributed to Kamboj, Sunita.

6 records

DOE OSTI2022

Assessment of doses in contaminated urban areas: modelling exercise based on Fukushima data

State-of-the-art dose assessment models were applied to estimate doses to the population in urban areas contaminated by the Fukushima Daiichi Nuclear Power Plant accident. Assessment results were compared among five models, and comparisons of model predictions with actual measurements were also made. Assessments were performed using both probabilistic and deterministic approaches. Predicted dose distributions for indoor and outdoor workers from a probabilistic approach were in good agreement with the actual measurements. In addition, when the models were applied to assess the doses to the representative person, based on a concept recommended by the International Commission on Radiological Protection and in the International Atomic Energy Agency Safety Standards, it was evident that doses to the representative person obtained with a deterministic approach were always higher than those obtained with a probabilistic approach using the same model.

dose assessment↗

DOE OSTI2022

Applying ALARA Principles in the Design of New Radiological Facilities

The application of ALARA (As Low As Reasonably Achievable) principles to the design of new radiological facilities at Argonne National Laboratory provides a consistent radiation safety basis for future facility operations. The Department of Energy Rule 10 CFR 835 specifies design objectives to be adopted during the design of new facilities for controlling personal radiation exposure. One is to keep exposure levels below 20% of the applicable standards in 10 CFR 835.202. For a radiation worker 20% of the standard corresponds to 5 µSv hr -1 for a 2,000 hour work year. For a member of the public the DOE design objective corresponds to 0.2 mSv in a calendar year. For a new facility worst case radionuclides and their source strengths are chosen. Local shielding is specified to reduce exposure rates to less than 50 µSv hr -1 at 30 cm from the shielding. The current version MCNP6 of the Los Alamos radiation shielding computer program MCNP is then used to calculate the exposure rates elsewhere. Design modifications are made to meet the criteria. The calculations and resulting facility design modifications are discussed for two new radiological facilities.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2022

Urban working groups in the IAEA’s model testing programmes: overview from the MODARIA I and MODARIA II programmes

The IAEA’s model testing programmes have included a series of Working Groups concerned with modelling radioactive contamination in urban environments. These have included the Urban Working Group of Validation of Environmental Model Predictions (1988–1994), the Urban Remediation Working Group of Environmental Modelling for Radiation Safety (EMRAS) (2003–2007), the Urban Areas Working Group of EMRAS II (2009–2011), the Urban Environments Working Group of (Modelling and Data for Radiological Impact Assessments) MODARIA I (2013–2015), and most recently, the Urban Exposures Working Group of MODARIA II (2016–2019). The overarching objective of these Working Groups has been to test and improve the capabilities of computer models used to assess radioactive contamination in urban environments, including dispersion and deposition processes, short-term and long-term redistribution of contaminants following deposition events, and the effectiveness of various countermeasures and other protective actions, including remedial actions, in reducing contamination levels, human exposures, and doses to humans. This paper describes the exercises conducted during the MODARIA I and MODARIA II programmes. These exercises have included short-range and mid-range atmospheric dispersion exercises based on data from field tests or tracer studies, hypothetical urban dispersion exercises, and an exercise based on data collected after the Fukushima Daiichi accident. We report improvement of model capabilities will lead to improvements in assessing various contamination scenarios (real or hypothetical), and in turn, to improved decision-making and communication with the public following a nuclear or radiological emergency.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2022

Best practices for predictions of radionuclide activity concentrations and total absorbed dose rates to freshwater organisms exposed to uranium mining/milling

Predictions of radionuclide dose rates to freshwater organisms can be used to evaluate the radiological environmental impacts of releases from uranium mining and milling projects. These predictions help inform decisions on the implementation of mitigation measures. The objective of this study was to identify how dose rate modelling could be improved to reduce uncertainty in predictions to non-human biota. For this purpose, we modelled the activity concentrations of 210 Pb, 210 Po, 226 Ra, 230 Th, and 238 U downstream of uranium mines and mills in northern Saskatchewan, Canada, together with associated weighted absorbed dose rates for a freshwater food chain using measured activity concentrations in water and sediments. Differences in predictions of radionuclide activity concentrations occurred mainly from the different default partition coefficient and concentration ratio values from one model to another and including all or only some 238 U decay daughters in the dose rate assessments. Consequently, we recommend a standardized best-practice approach to calculate weighted absorbed dose rates to freshwater biota whether a facility is at the planning, operating or decommissioned stage. At the initial planning stage, the best-practice approach recommend using conservative site-specific baseline activity concentrations in water, sediments and organisms and predict conservative incremental activity concentrations in these media by selecting concentration ratios based on species similarity and similar water quality conditions to reduce the uncertainty in dose rate calculations. At the operating and decommissioned stages, the best-practice approach recommends relying on measured activity concentrations in water, sediment, fish tissue and wholebody of small organisms to further reduce uncertainty in dose rate estimates. This approach would allow for more realistic but still conservative dose assessments when evaluating impacts from uranium mining projects and making decision on adequate controls of releases.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗

DOE OSTI2020

Peer Review Comments on the Authorized Limits Technical Document (OREM-19-2573) for Disposal of Waste Containing Residual Radioactive Material at the U.S. Department of Energy Industrial Landfill IV, Y-12 National Security Complex, Oak Ridge, Tennessee

Argonne National Laboratory is assigned responsibility by DOE/EM-4 to conduct a peer review of the proposed Authorized Limits (ALs) for disposal of waste containing residual radioactive material at the U.S. Department of Energy (DOE) Industrial Landfill IV, Y-12 National Security Complex, Oak Ridge, Tennessee. The ALs document package (Document Number OREM-19-2573, dated May 2020) was sent to Argonne by DOE on June 1, 2020 (Oak Ridge 2020a). Argonne started the review immediately and held several BlueJeans meetings with Oak Ridge and DOE to discuss review comments. The first BlueJeans meeting was June 17, 2020. Oak Ridge presented a briefing of the AL package, followed by discussion of the assumptions, methodology, and parameters used in the analysis. A follow-up BlueJeans meeting was held on June 24, 2020, at which Argonne presented both general and specific review comments in PowerPoint slides. This set of slides - Review Comments on the “Authorized Limits for Disposal of Waste Containing Residual Radioactive Material at DOE Industrial Landfill IV, Y-12 National Security Complex, Oak Ridge, Tennessee” - was then emailed to Oak Ridge and DOE after the BlueJeans meeting on June 24, 2020. A PDF version of this slide set is included as Appendix A.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Independent Review of Paducah Gaseous Diffusion Plant Freon Release Authorized Limits and Evaluation of Corresponding Radiological Doses

The Portsmouth/Paducah Project Office (PPPO) proposes to use the American National Standard Institute (ANSI)/Health Physics Society (HPS) N13.12 (ANSI 2013) volumetric screening levels as Authorized Limits (ALs) for the release of cryofluorane R-114 (Freon) at the Paducah Gaseous Diffusion Plant (PGDP) in Paducah, Kentucky. A review of the proposed ALs document indicated that potential exposure scenarios from the release of Freon are adequately discussed therein. However, questions arose regarding the assumptions made for the ANSI N13.12 clearance values, and whether or not all pathway doses were adequately considered in the Authorized Limits Request. For example, PPPO used the inverse square rule to estimate dose to the truck driver. A more appropriate evaluation would have considered the material as a cylindrical source for this pathway. In addition, the inverse square rule was inadequately applied to the inhalation and ingestion pathways. Furthermore, while the ANSI N13.12 volumetric clearance values are used in the ALs, questions arose concerning whether the values are appropriate for materials of greater than 1 metric ton, and whether or not the ANSI N13.12 standard could be used for liquids that are sent to an incinerator. For these reasons, dose modeling was conducted to demonstrate that using ANSI/HPS N13.12 screening levels as ALs for release of PGDP Freon meets the dose criteria set in U.S. Department of Energy (DOE) Order 458.1. This report provides a bounding dose analysis using scenarios described in PPPO document. Realistic parameters, such as the incinerator characteristics data, are used when available.

61 RADIATION PROTECTION AND DOSIMETRY↗