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At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Developments of the Versa-Pac to Expand Use for the Industry - 17475

The Versa-Pac is a certified Type AF packaging (USA NRC Certificate no. USA/9342/AF-96), designed as a replacement for USA DOT 6M drum-type transportation packaging, which was retired approximately ten years ago. The Versa-Pac can be used directly or in conjunction with pails, drums, or inserts, and a variety of smaller containers and vessels. The Versa-Pac comes in two sizes: the VP-55 (i.e. 200-liter drum version) and VP-110 (i.e. 400-liter drum version). It features a patented design that, in combination with the drum exterior, provides enhanced structural protection to payloads under Normal Conditions of Transport and Hypothetical Accident Conditions. Since its introduction in 2010, the Versa-Pac has been utilized mainly as a one-time-use package for disposing of radioactive waste. However, the Versa-Pac package, which has the familiar simplicity of a drum, is a reusable transport package with the capability of shipping a variety of radioactive materials. In the spirit of simplicity yet versatility, the capability of the package has been improved several times in collaboration with our customers. The latest design improvements include a new configuration designated the VP-55HC (high capacity), which utilizes an internal safe geometry pipe container that increases the U-235 capacity. The pipe container is similar in design to the historic USA DOT 2R vessel. Other planned and proposed improvements include provisions to transport UF{sub 6} in 1S, 2S, and 5A/5B cylinders; design changes that further increase the amount of U-235 at enrichments less than 20 wt.%; increasing the capacity of the VP-110 with a VP-110-specific analysis; developing a new, lighter 113-liter version of the package; and introducing content limits specific to uranium oxides in multiple forms. This paper traces the engineering and design efforts involved in expanding the Versa-Pac's versatility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Characterization and Remediation of the Hanford 618-10 Burial Ground Steel-pipe Vertical Pipe Units - 18238

The 618-10 Burial Ground is approximately 485 feet by 470 feet and consisted of 12 trenches and 94 VPUs. It is located approximately 5 miles north of Richland, Washington in the 600 Area of the Hanford Site. The 618-10 Burial Ground received low- to high-activity waste (e.g., fission products and some plutonium-contaminated waste). Most of the waste originated at the 300 Area laboratory facilities. Lower activity radioactive material and shielded drums were disposed in trenches, while laboratory wastes with higher levels of radiation were disposed in Vertical Pipe Units (VPUs). The 618-10 Burial Ground contained approximately 94 VPUs (uncertain mix of first and second generation). There are different variants of VPUs. The first generation of VPUs was simply corrugated culverts or standard pipes, 14-in diameter and 10- to 15-ft long, buried vertically. The second generation of VPUs was 22-in diameter, 15-ft long waste receptacles constructed by welding five 55-gallon bottomless drums together end-to-end and burying them vertically, separated by approximately 10 feet on center in rows approximately 25 feet apart. A subset of the first generation VPUs consisted of the steel pipe VPUs represented by carbon-steel pipes, 10 to 24 in. in diameter with various wall thicknesses. This paper will present the remediation of the steel-pipe VPUs, which were the last of the VPUs remediated in this project. The remediation process developed for steel pipe VPUs included exposing a section of the VPU within an excavated trench, placing a mix box around the VPU, filling the mix box with grout, shearing the VPU and contents, and then removing the grout waste mixture into a waste box for disposal at the Hanford Environmental Restoration Disposal Facility (ERDF). Prior the start of remedial activities extensive mock-ups were performed on surrogate steel-pipe VPUs to ensure the effectiveness of the process before introducing the hazards of radioactive VPU waste. The critical steps in the process of size reducing steel-pipe VPUs and mixing the contents with grout was successfully demonstrated. This process was completed for each steel-pipe VPU in successive lifts until the entire set of steel-pipe VPUs were mixed in grout and removed from the excavation. The remediation of the steel-pipe VPUs proved be successful in that all the VPUs were removed and transported to ERDF without a release to the local environment. (authors)

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DOE OSTI2015

Simultaneous Thermal Analysis of WIPP and LANL Waste Drum Samples: A Preliminary Report

On Friday, February 14, 2014, an incident in P7R7 of the WIPP underground repository released radioactive material into the environment. The direct cause of the event was a breached transuranic (TRU) waste container, subsequently identified as Drum 68660. Photographic and other evidence indicates that the breach of 68660 was caused by an exothermic event. Subsequent investigations (Britt, 2015; Clark and Funk, 2015; Wilson et al., 2015; Clark, 2015) indicate that the combination of nitrate salts, pH neutralizing chemicals, and organic-based adsorbent represented a potentially energetic mixture. The materials inside the breached steel drum consisted of remediated, 30- to 40-year old, Pu processing wastes from LANL. The contents were processed and repackaged in 2014. Processing activities at LANL included: 1) neutralization of acidic liquid contents, 2) sorption of the neutralized liquid, and 3) mixing of acidic nitrate salts with an absorber to meet waste acceptance criteria. The contents of 68660 and its sibling, 68685, were derived from the same parent drum, S855793. Drum S855793 originally contained ten plastic bags of acidic nitrate salts, and four bags of mixed nitrate and oxalate salts generated in 1985 by Pu recovery operations. These salts were predominantly oxalic acid, hydrated nitrate salts of Mg, Ca, and Fe, anhydrous Na(NO 3 ), and minor amounts of anhydrous and hydrous nitrate salts of Pb, Al, K, Cr, and Ni. Other major components include sorbed water, nitric acid, dissolved nitrates, an absorbent (Swheat Scoop®) and a neutralizer (KolorSafe®). The contents of 68660 are described in greater detail in Appendix E of Wilson et al. (2015)

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DOE OSTI2025

Commercialization of the Transportation-Security, Tracking, and Reporting System (T-STAR)

The Transportation-Security, Tracking, and Reporting System (T-STAR) was developed by the National Nuclear Security Administration, NA-21, Office of Radiological Security (ORS) to provide a transportation security system for detection and tracking during transport of Category 1 and Category 2 radiological material. Few off-the-shelf systems for conveyance tracking offer detection of a cargo compartment breach or a removal of the cargo. Systems that do offer this capability often require permanent installation through modifying of the conveyance itself. This is not sustainable in many countries where ORS is building use, storage, and transport security capacity. The development of T-STAR has moved from fielding robust prototypes deployed in countries ranging from North America, Latin America and Central Asia to a commercially produced product that can now be deployed to provide enhanced security during transit. Each prototype deployment resulted in important lessons learned, which informed the requirements for the final commercial product. T-STAR uses both cellular and Iridium satellite modems to provide redundant communications to provide the configuration, status, and alerts to a server monitoring the shipment, which is accessible using a multilanguage browser-based user interface. A wireless security system employing using Z-wave sensors for intrusion detection located in the conveyance provide low cost but effective solution for a wide range of conveyance types. Additional capabilities include the ability to monitor a vehicles’ CANBUS (Controller Area Network) system, an ethernet port for high throughput sensor information such as video cameras, and the ability to power and use advanced external sensor payloads. These features make the T-STAR a capable and expandable security gateway that can be deployed on a variety of conveyances from box trucks to open trailers. The ability to provide tracking, monitoring, and detection provide a key component in overall best practices designed to protect shipments of radioactive material.

Schultze, Michael [ORNL] (ORCID:0000000283205671)↗
DOE OSTI2018

Marine Sciences Laboratory Radionuclide Air Emissions Report for Calendar Year 2017

The U.S. Department of Energy Office of Science (DOE-SC) Pacific Northwest Site Office has oversight and stewardship duties associated with the Pacific Northwest National Laboratory Marine Sciences Laboratory (MSL), located on Battelle Land-Sequim. The facility has two buildings with the potential to emit low levels of radioactive materials. This report is prepared to document compliance with the Code of Federal Regulation, Title 40, Protection of the Environment, Part 61, National Emission Standards for Hazardous Air Pollutants, Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities” and Washington Administrative Code Chapter 246-247, Radiation Protection–Air Emissions. Compliance is indicated by comparing the estimated effective dose equivalent (EDE) to the maximally exposed individual (MEI) with the 10 millirem per year (mrem/yr) U.S. Environmental Protection Agency (EPA) standard. The MSL has only fugitive emissions sources. Despite the fact that the regulations are intended for application to point source emissions, fugitive emissions are included with regard to complying with the EPA standard. The EDE to the MSL MEI due to routine operations in 2017 was 1.6E-04 mrem (1.6E-06 mSv). No non-routine emissions occurred in 2017. The MSL is in compliance with the federal and state 10 mrem/yr standard.

40 CFR 61 Subpart H↗
DOE OSTI2019

Marine Sciences Laboratory Radionuclide Air Emissions Report for Calendar Year 2018

The U.S. Department of Energy Office of Science (DOE-SC) Pacific Northwest Site Office has oversight and stewardship duties associated with the Pacific Northwest National Laboratory Marine Sciences Laboratory (MSL), located on Battelle Land-Sequim. The facility has two buildings with the potential to emit low levels of radioactive materials. This report is prepared to document compliance with the Code of Federal Regulation, Title 40, Protection of the Environment, Part 61, National Emission Standards for Hazardous Air Pollutants, Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities” and Washington Administrative Code Chapter 246-247, Radiation Protection–Air Emissions. Compliance is indicated by comparing the estimated effective dose equivalent (EDE) to the maximally exposed individual (MEI) with the 10 millirem per year (mrem/yr) U.S. Environmental Protection Agency (EPA) standard. The MSL has only fugitive emissions sources. Despite the fact that the regulations are intended for application to point source emissions, fugitive emissions are included with regard to complying with the EPA standard. The EDE to the MSL MEI due to routine operations in 2018 was 4.5E-04 mrem (4.5E-06 mSv). No non-routine emissions occurred in 2018. The MSL is in compliance with the federal and state 10 mrem/yr standard.

40 CFR 61 Subpart H↗
DOE OSTI2020

Best Practices for Shielding Analyses of Activated Metals and Spent Resins from Reactor Operation

This report investigated best practices for performing shielding evaluations of Type B waste packages, as defined in 10 CFR Part 71 regulations on packaging and transportation of radioactive material, or packages for which the contents are not defined before loading and may include a broad range of nuclides, geometries, and non-fuel materials. The following non-fuel waste streams were analyzed: Activated metals from decommissioned commercial power reactors, including Type 304 stainless steel, reactor vessel steel, and Inconel, Control blades from boiling water reactors, Neutron-activated corrosion products on surfaces of activated metals, and Spent resins from power plant operations. Measured elemental compositions, including major constituents and impurities, for steel and Inconel samples from commercial power reactors were used in activation calculations to determine radionuclide inventories in activated metals. For a simplified cask model, 60 Co contribution to the total external package dose rate at 30 days after shutdown varied from approximately 60% to 95%, depending on the activated metal, initial cobalt impurity concentration in the metal, and the thickness of the overpack gamma shield. Its maximum contribution to the total external package dose rate of approximately 100% was reached within the time interval of 2 to 5 years after shutdown and was maintained for up to 45 to 60 years after shutdown, depending on material, initial cobalt impurity concentration, and shield thickness. Thereafter, the 60 Co contribution to external package dose rate decreased with increasing decay time. Cobalt-60 is primarily produced by neutron reactions with the cobalt impurity in steel and Inconel. Other important radionuclides in activated metals contributing to package external dose rate are radionuclides with relatively short decay times, including 51 Cr, 59 Fe, 58 Co, and 54 Mn. These radionuclides may be represented as an equivalent 60 Co activity/source because 60 Co gamma ray emissions are bounding in terms of source strength and energy to other important radionuclides identified in the analyzed activated metals. Approaches for modeling the neutron-activated corrosion products that may be attached to activated reactor components were analyzed in this report. It was demonstrated that a surface source is more conservative than a uniform volumetric source for the treatment of neutron-activated corrosion products with respect to external package dose rates. An analysis of the maximum radionuclide loadings reported on spent resins identified the radionuclides 137 Cs, 60 Co, 134 Cs, 65 Zn, and 58 Co as the primary contributors to external package dose rate. For a resin cooled for 3.08 years, the external package dose rate was entirely produced by the reported 137 Cs and 60 Co inventory. The neutron sources from actinides found on spent resins or activated metals produced negligible dose rates and may be ignored in dose rate analyses. Effects of idealized waste material, source geometry, and spatial material/source distributions on external package dose rates were determined based on dose rate results for a simplified cask model under normal conditions of transport. Type 304 stainless steel, zirconium, and aluminum with adjusted mass densities based on a maximum content weight were analyzed for material modeling effects on external package dose rate. These materials produced identical external package dose rates within the statistical uncertainties of the dose rate estimates. Among four different source geometry configurations with homogeneous material of different mass densities, uniform volumetric source distribution, and the same source strength, the geometry configuration with lowest mass density (i.e., minimum self-shielding effect among the four cases) was most conservative. Spatial source distributions that better represent localized peak 60 Co activity values were more conservative than a uniform volumetric source distribution, assuming the same weight and total source strength per package. The increase in external dose rate caused by localized source peaks can be as much as the ratio between source peak activity density to average activity density, depending on the location of the activated metal with peak activity density. Therefore, the shielding analysis may be simplified if localized peak activities and the average activity per package can be measured/determined and documented at the time of cask loading. For simplicity, external package dose rates may be determined based on average source activity and a uniform volumetric source distribution. The dose rate results from that calculation model multiplied by the ratio of peak activity density to average activity density will produce maximum dose rate values for conservative estimates.

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DOE OSTI2021

Saltstone Waste Characterization Analysis - Salt Waste Processing Facility (SWPF) Waste Stream (Q4 CY2020)

The Saltstone Production and Disposal Facility is designed and permitted by the state of South Carolina Department of Health and Environmental Control (SCDHEC) to treat and dispose of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site (SRS). Low-level waste (LEW) aqueous streams from the Effluent Treatment Project (ETP) and decontaminated solutions from the Tank Closure Cesium Removal Unit (TCCR) and the Salt Waste Processing Facility (SWPF) are stored in Tank 50 until the LEW can be transferred to the Saltstone Facility for treatment and disposal. In the past, decontaminated solution from the Modular Caustic Side Solvent Extraction Unit (MCU) was stored in Tank 50 until the LEW could be transferred to the Saltstone Facility for treatment and disposal. MCU is currently in a suspended operations state. LEW that meets the Waste Acceptance Criteria (WAC) can be transferred, stored, and treated in the Saltstone Production Facility (SPF) for subsequent disposal as saltstone grout in the Saltstone Disposal Facility (SDF). Sampling will be conducted as new waste streams are identified for treatment and disposal at the Saltstone Industrial Wastewater Treatment Facility (IWTF) and Z-Area Industrial Solid Waste Landfill (ISWLF), Facility ID# 025500-1603, General Condition B.9 or every six years in accordance with South Carolina (SC) Regulation 61-107.19 Parti C, “Solid Waste Management: Solid Waste Landfills and Structural Fill - General Requirements.”

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DOE OSTI2017

Marine Sciences Laboratory Radionuclide Air Emissions

The U.S. Department of Energy Office of Science (DOE-SC) Pacific Northwest Site Office has oversight and stewardship duties associated with the Pacific Northwest National Laboratory Marine Sciences Laboratory (MSL), located on Battelle Land-Sequim. The facility has two buildings with the potential to emit low levels of radioactive materials. DOE-SC contracted for exclusive use of its radiological operations effective October 1, 2012. This report is prepared to document compliance with the Code of Federal Regulation, Title 40, Protection of the Environment, Part 61, National Emission Standards for Hazardous Air Pollutants , Subpart H, “National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities” and Washington Administrative Code Chapter 246-247, Radiation Protection–Air Emissions . Compliance is indicated by comparing the estimated effective dose equivalent (EDE) to the maximally exposed individual (MEI) with the 10 millirem per year (mrem/yr) U.S. Environmental Protection Agency (EPA) standard. The MSL has only fugitive emissions sources. Despite the fact that the regulations are intended for application to point source emissions, fugitive emissions are included with regard to complying with the EPA standard. The EDE to the MSL MEI due to routine operations in 2016 was 5.7E-04 mrem (5.7E-06 mSv). No non-routine emissions occurred in 2016. The MSL is in compliance with the federal and state 10 mrem/yr standard.

40 CFR 61 Subpart H↗
DOE OSTI2025

Stabilization of Preternatural Barium Oxidation States as an Unexpected Byproduct of β-Decay: Discovery of a New Halide Semiconductor Alloy

137 Cs has a wide range of roles in the nuclear industry. The solid material, safely encapsulated in CsCl as 137 CsCl, is stored as fission product waste from nuclear power production and legacy waste from nuclear weapons production; it has also served as a radiation source in food and sewage irradiators as well as medical devices. However, because of the solubility of the chloride salt and the relatively high specific activity of 137 Cs, damaged or broken capsules can lead to severe radiological accidents. Safe capsule design and material recycling are complicated by the unclear structural evolution during β-decay, which remains ambiguous due to the differing oxidation states of Cs (1+) and Ba (2+). Here, in this study, we use first-principles calculations to investigate the evolving structure–property relationships of Cs 1–x Ba x Cl during β-decay. Despite the well-established 2+ formal oxidation state of alkali-earth metals, we find that Ba 1+ can be stabilized in the form of a mixed-valence alloy at low concentrations. Specifically, we identify three regimes for the β-decay of 137 Cs into CsCl: Ba-doped CsCl (Ba ≤ 14%), wherein Ba has the expected 2+ oxidation state; Cs–Ba–Cl alloys, where Ba has a mix of the usual Ba 2+ and highly unusual Ba 1+ oxidation state in the form of a quasi-disordered mixed-valence alloy (Ba = 25%); and phase separation into a CsCl + BaCl 2 + Ba (m) mechanical mixture, where Ba reverts to its expected 2+ oxidation state (Ba > 25%). Surprisingly, the Cs 0.75 Ba 0.25 Cl mixed-valence alloy is a narrow indirect band gap semiconductor (1.05 eV) despite the insulating nature of both CsCl and BaCl 2 . It also exhibits strongly excitonic polarized optical properties, has glass-like ultralow thermal conductivity (directional average of 0.21 W/mK at 300 K), and shows greater resistance to deformation under both tensile and volumetric strengths compared with the original CsCl structure (e.g., shear and Young’s modulus of 9.04 and 31.62 GPa, respectively). These findings imply that transmutation of 137 Cs leads to highly unusual chemical bonding that stabilizes Ba 1+ in local regions of the quasi-disordered Cs 0.25 Ba 0.75 Cl, resulting in anomalous physical properties. Moreover, this discovery provides valuable insight for safe nuclear waste capsule design, which can aid in preventing environmental or human exposure to radioactive materials.

Fuhr, Addis S. [Oak Ridge National Laboratory (ORN↗
DOE OSTI2013

Fukushima Media Involvement: Lessons Learned and Challenges

Only days after the Fukushima nuclear reactor disaster on March 11, 2011, the Department of Energy’s Pacific Northwest National Laboratory, or PNNL, found itself in a maelstrom of media attention following announcement of the detection of minute levels of radioactive material originating from the damaged reactors 4,500 miles away. Within days, PNNL had set up a technical team in support of the U.S. government’s efforts to assist the devastated country. While a vast amount of information began to flow from Japan, the situation rapidly evolved challenging PNNL scientists and engineers, as well as news media and the general public to deal with a complex and often conflicting information stream. Over the course of about three weeks, PNNL’s News & Media Relations staff and its scientists and engineers responded to more than one hundred requests for information, and engaged in dozens of personal interviews with international, national, regional and local media. While many of the interviews and resulting stories were accurate and well done, to say that all communication went flawless would be far from the truth. In the midst of chaos and confusion, which are part of any significant crisis, hiccoughs are sure to occur. Even when communications guidelines are established and agreed-to ahead of time, and spokespeople promise to stay on message and within their areas of expertise, there is no guarantee the ball will not be dropped on occasion. Addressed here is “the rest of the story.”

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Nevada National Security Site Environmental Report 2016

This Nevada National Security Site Environmental Report (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 National Nuclear Security Administration Nevada Field Office (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 2016 at the Nevada National Security Site (NNSS) and its two Nevada-based 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). NNSA/NFO directs the management and operation of the NNSS and six sites across the nation. In addition to the NNSA itself, the six sites include two in Nevada (NLVF and RSL-Nellis) and four in other states (RSL-Andrews in Maryland, Livermore Operations in California, Los Alamos Operations in New Mexico, and Special Technologies Laboratory in California). Los Alamos, Lawrence Livermore, and Sandia National Laboratories are the principal organizations that sponsor and implement the nuclear weapons programs at the NNSS. National Security Technologies, LLC (NSTec), is the current Management and Operating contractor accountable for the successful execution of work and ensuring that work is performed in compliance with environmental regulations. The six sites all provide support to enhance the NNSS as a location for its multiple missions. The three major NNSS missions include National Security/Defense, Environmental Management, and Nondefense. The major programs that support these missions are Stockpile Stewardship and Management, Nonproliferation and Counterterrorism, Nuclear Emergency Response, Strategic Partnership Projects, Environmental Restoration, Waste Management, Conservation and Renewable Energy, Other Research and Development, and Infrastructure. The major facilities that support the programs include the U1a Facility, Big Explosives Experimental Facility (BEEF), Device Assembly Facility, Dense Plasma Focus Facility, Joint Actinide Shock Physics Experimental Research Facility, Radiological/Nuclear Countermeasures Test and Evaluation Complex, Nonproliferation Test and Evaluation Complex (NPTEC), Radiological/Nuclear Weapons of Mass Destruction Incident Exercise Site, the Area 5 Radioactive Waste Management Complex (RWMC), and the Area 3 Radioactive Waste Management Site (RWMS).

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DOE OSTI2018

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on “fingerprinting” casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17 x 17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4 He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244 Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Modelling and mitigating dose to firefighters from inhalation of radionuclides in wildland fire smoke

Firefighters responding to wildland fires where surface litter and vegetation contain radiological contamination will receive a radiological dose by inhaling resuspended radioactive material in the smoke. This may increase their lifetime risk of contracting certain types of cancer. Using published data, we modelled hypothetical radionuclide emissions, dispersion and dose for 70th and 97th percentile environmental conditions and for average and high fuel loads at the Savannah River Site. We predicted downwind concentration and potential dose to firefighters for radionuclides of interest ( 137 Cs, 238 Pu, 90 Sr and 210 Po). Predicted concentrations exceeded dose guidelines in the base case scenario emissions of 1.0 × 10 7 Bq ha –1 for 238 Pu at 70th percentile environmental conditions and average fuel load levels for both 4- and 14-h shifts. Under 97th percentile environmental conditions and high fuel loads, dose guidelines were exceeded for several reported cases for 90 Sr, 238 Pu and 210 Po. Potential for exceeding dose guidelines was mitigated by including plume rise (>2 m s –1 ) or moving a small distance from the fire owing to large concentration gradients near the edge of the fire. As a result, our approach can quickly estimate potential dose from airborne radionuclides in wildland fire and assist decision-making to reduce firefighter exposure.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Disposal of Greater Than Class C Low-Level Radioactive Waste in Andrews County, Texas - 16408

The U.S. Nuclear Regulatory Commission (NRC) staff has recommended approval of an initiative to permit the Texas Commission on Environmental Quality (TCEQ) to license the disposal of commercial and co-mingled Greater Than Class C Low-Level Radioactive Waste (GTCC LLW) in Andrews County, Texas. Such a framework is clearly articulated in the Section 274 of the Atomic Energy Act of 1954, as amended. The NRC staff also recommended approval to proceed with a rulemaking to align the Waste Classification Tables in Title 10 of the Code of Federal Regulations (CFR), Part 61.55 with the paragraph 61.2 definition of 'waste' to provide a disposal pathway for waste containing certain alpha-emitting transuranic elements with concentrations exceeding 100 nanocuries per gram (nCi/g). A pathway for the disposal of commercial and co-mingled GTCC LLW, including waste with certain alpha emitting transuranic elements exceeding 100 nCi/g (TRU waste), might emerge if the initiative described in the Commissions' July 22, 2015 document, SECY-15-0094, Historical and Current Issues Related to Disposal of GTCC LLW, is approved. The NRC staff recommendations envisioned the NRC and TCEQ working together to review a site-specific performance assessment that would be submitted by Waste Control Specialists LLC (WCS) as part of an amendment to its existing Radioactive Materials License No. R04100. The TCEQ's recommendations were prompted by a petition for rulemaking filed by WCS to allow the disposal of GTCC waste in Texas. On September 10, 2014, the TCEQ Commissioners directed their staff to engage stakeholders and request input from the NRC and the U.S. Department of Energy (DOE) regarding actions needed to proceed with any future rulemaking. On January 30, 2015, the TCEQ requested clarification from the NRC regarding their legal authority and jurisdiction to regulate GTCC, GTCC-like, and TRU waste. The deliberations that have transpired over the past year have raised important legal, policy, and technical matters. This initiative could provide a safe disposal pathway for certain waste streams that have been orphaned over the past 30 years. One important realization that has emerged is the conservative assumptions used to establish the upper limits for Class C LLW when Part 61 was promulgated in 1981 would be much too restrictive when applied to a modern, highly engineered disposal facility located in arid west Texas. Waste that was not suitable for near surface disposal in 1981 could certainly be demonstrated to be safely disposed at the WCS Federal Waste Disposal Facility in Andrews County, Texas. This paper will address the legal, policy, and technical matters supporting this important topic currently under consideration both nationally and within the State of Texas. (authors)

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

Deep Isolation-Development of the Safety Case for Disposal of Radioactive Wastes in Horizontal Boreholes - 20028

Deep Isolation has developed a safe, secure, and permanent geological disposal method for high-level waste, including spent nuclear fuel as well as sealed sources and other highly radioactive materials. The method leverages well established directional drilling technology to create horizontal repositories deep underground. The combination of great depth (1-3 km) and the ability to precisely position repositories in a horizontal orientation provides access to geologic strata that are inaccessible for typical mined and deep vertical borehole repositories. A number of potential direct and indirect safety benefits accrue including:1) greater depth below regional freshwater aquifers; 2) increased flexibility and more geologic options for siting; 3) access to deep formations with persistent and sustained reducing conditions. In addition there are a number of safety elements related specifically to the horizontal repository geometry including: 1) passive direction of thermally driven fluid and radionuclide movement away from the vertical access hole and toward the 'dead end' portion of the repository; 2) mitigation of seismic hazards by orienting repositories parallel to local and regional fault structures. In this paper we explore and discuss some of the key technologic, geologic and hydrologic elements that support the deep horizontal borehole safety case. The stalemate seen across the globe on the disposal of spent nuclear fuel and high-level waste can be broken. Deep Isolation offers a novel option for safe, secure, and permanent deep geological disposal of nuclear waste that can be developed as centralized repositories or adapted to smaller regional or site-specific repositories located near waste sources. (authors)

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

Safety Enhancements for TRU Waste Handling - 12258

For years, proper Health Physics practices and 'As Low As Reasonably Achievable' (ALARA) principles have fostered the use of glove boxes or other methods of handling (without direct contact) high activities of radioactive material. The physical limitations of using glove boxes on certain containers have resulted in high-activity wastes being held in storage awaiting a path forward. Highly contaminated glove boxes and other remote handling equipment no longer in use have also been added to the growing list of items held for storage with no efficient method of preparation for proper disposal without creating exposure risks to personnel. This is especially true for wastes containing alpha-emitting radionuclides such as Plutonium and Americium that pose significant health risks to personnel if these Transuranic (TRU) wastes are not controlled effectively. Like any good safety program or root cause investigation PFNW has found that the identification of the cause of a negative change, if stopped, can result in a near miss and lessons learned. If this is done in the world of safety, it is considered a success story and is to be shared with others to protect the workers. PFNW believes that the tools, equipment and resources have improved over the past number of years but that the use of them has not progressed at the same rate. If we use our tools to timely identify the effect on the work environment and immediately following or possibly even simultaneously identify the cause or some of the causal factors, we may have the ability to continue to work rather than succumb to the start and stop-work mentality trap that is not beneficial in waste minimization, production efficiency or ALARA. (authors)

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