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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Unreviewed Disposal Question Evaluation: Disposal of the Oak Ridge National Laboratory Contaminated Debris at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Oak Ridge National Laboratory (ORNL) Contaminated Debris (ORNLDEBRIS001, Revision 11 [ORNL 2019a]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the waste stream meets all performance objectives in U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 2011). The ORNL Contaminated Debris waste stream is recommended for acceptance with conditions on the total activity disposed of tritium ( 3 H), niobium-94 ( 94 Nb), technetium-99 ( 99 Tc), cesium-137 ( 137 Cs), iridium-192 ( 192 mIr), radium-226 ( 226 Ra), uranium-234 ( 234 U), and 238 U. The total activity disposed shall not exceed the following limits: 3 H: 2E12 Bq, 94 Nb: 2E9 Bq, 99 Tc: 8E10 Bq, 137 Cs: 1E11 Bq, 192 mIr: 1E9 Bq, 226 Ra: 3E11 Bq, 234 U: 2E9 Bq, 238 U: 5E9 Bq.

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DOE OSTI2018

Unreviewed Disposal Question Evaluation: Disposal of the Transuranic Waste Processing Center Mixed Low-Level Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Transuranic Waste Processing Center (TWPC) Mixed Low-Level Waste (MLLW) (FWORCHMLLW103, Revision 11 [TWPC 2018]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the TWPC MLLW waste stream meets all performance objectives of U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P (DOE 1999). The TWPC MLLW waste stream is recommended for acceptance without conditions.

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DOE OSTI2018

UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Los Alamos National Laboratory Regulated Asbestos-Contaminated Low-Level Waste at the Area 5 Radioactive Waste Management Sited, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Los Alamos National Laboratory (LANL) Regulated Asbestos-Contaminated Low-Level Waste (LANL000000029, Revision 5 [LANL 2017]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the LANL Regulated Asbestos-Contaminated Low-Level Waste meets all U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P performance objectives (DOE 1999). The LANL Regulated Asbestos Contaminated Low-Level Waste is recommended for acceptance without conditions.

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DOE OSTI2016

Special Analysis for the Disposal of the Transuranic Waste Processing Center Mixed Low-Level Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This special analysis (SA) evaluates whether the Transuranic Waste Processing Center (TWPC) Mixed Low-Level waste stream is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the TWPC Mixed Low-Level waste was meets all U.S. Department of Energy (DOE) Manual DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P performance objectives. The TWPC Mixed Low-Level waste stream is recommended for acceptance without conditions.

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DOE OSTI2017

Treatment of Radioactive Sludge From Waste Water Treatment for Weight and Volume Reduction with FreezeTec-Technique - 17010

At the Studsvik site in Sweden, Studsvik Nuclear AB operates a water treatment facility that treat all the waste water from the nuclear facilities at the site. The facility contains several pools for waste water treatment. The standard procedure for treating the water consist of adding FeCl{sub 3} to precipitate any particulates. Historically this waste has been mixed with concrete and grouted into barrels which will increase the volume needed for disposal in final repository. The sludge that is generated from the water treatment facility has been subject to treatment with the Freeze-Tec method where the sludge is freeze dewatered and after thawing generating a solid phase with high dry substance and a clear water phase. The radioactive particles/nuclides tend to stay in the solid phase while the water phase has very low concentration of radioactive particles. 99.9% of the Co-60 and Cs-137 activity end up in the solids and a dry solid with smaller volume and weight is obtained after treatment. (authors)

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DOE OSTI2017

A Study on Korea Experience in Regulating the Transport of Low and Intermediate-level Radioactive Wastes - 17182

In the Republic of Korea, the transport of Low and Intermediate-Level Radioactive Wastes (LILW) from nuclear power plants (NPP), nuclear research reactor, and radioactive isotope storage facility to Gyeongju LILW disposal facility has increased since the operation license of facility was approved in December 2014. As a result, there has been an increase in public anxiety and concerns over damage or social disorder with LILW transport. So the necessity of strengthened safety confirmation on LILW has been constantly raised. In August 2016, the Enforcement Regulations of the Nuclear Safety Act were amended in relation to the reporting and inspection of LILW transport. The first section of this paper describes the domestic transport regulations including the recently revised contents. The second section shows that the status of domestic and inspection of LILW transport were introduced and transportation regulatory experience was shared in Korea. (authors)

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DOE OSTI2017

Chemical Compatibility of Conventional Geo-synthetic Clay Liners to Aggressive Low-Level Radioactive Waste Leachate - 17296

representative of leachates in low-level radioactive waste (LLW) disposal facilities affected the hydraulic conductivity of a conventional sodium-bentonite (Na-B) geo-synthetic clay liner (GCL) used in composite liner systems. One leachate was highly alkaline (H-pH) and the other had a low abundance of monovalent cations relative to divalent cations (L-RMD). Tests were also conducted with a typical radioactive synthetic leachate (RSL) and with deionized (DI) water as control. Na-B from the GCL had a swell index of 24 mL/2 g in RSL, 22 mL/2 g in H-pH leachate, and 16 mL/2 g in L-RMD leachate, whereas the swell index was 36 mL/2 g in DI water. Hydraulic conductivity tests were conducted in flexible-wall permeameters using the falling headwater with an effective confining stress at 20 kPa. Hydraulic conductivity of Na-B GCL was low when permeated with RSL (1.8 x 10-10 m/s) and H-pH leachate (2.4 x 10{sup -10} m/s), but higher than the hydraulic conductivity to DI water (2.5 x10{sup -11} m/s). The GCL was two orders of magnitude more permeable when permeated with L-RMD leachate (7.9 x 10{sup -8} m/s). The higher hydraulic conductivity of the GCL to the leachates is attributed to reduced swelling of the Na-B in leachates with higher ionic strength solution (H-pH, RSL) or a predominance of divalent cations (L-RMD). Increasing the effective confining stress from 20 to 450 kPa reduced the hydraulic conductivity approximately two orders of magnitude for the L-RMD and H-pH leachates. (authors)

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DOE OSTI2017

Comparison of U.S. Special Permit/Exemption/Authorization Processes for Licensing Type B Radioactive Material Packages - 17513

In the United States, there are three regulators (agencies) associated with licensing Type B radioactive material packages for shipping. They are the U.S. Department of Energy (DOE), the United States Nuclear Regulatory Commission (NRC), and the United States Department of Transportation (DOT). Each regulator has requirements and guidance which implement United States Code of Federal Regulations 10 CFR 71 - Packaging and Transportation of Radioactive Material and 49 CFR 171 through 180 - Hazardous Materials Regulations. Periodically, select packages may not clearly fit within the requirements or guidance. In such cases, the planned shipper can petition the applicable agency for a special permit/exemption/authorization approving the shipment with the specified deviations from the requirements or guidance. If equivalent safety is successfully demonstrated to the applicable regulator, such authorization may be obtained. This paper compares the: 1) DOE exemption process; 2) the NRC special package authorization process; and, 3) the DOT special permit process. (authors)

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DOE OSTI2015

An experimental program to characterize confined explosions of hydrogen/oxygen mixtures in the context of radioactive material transport between CEA nuclear facilities - 15125

CEA, the French Alternative Energies and Atomic Energy Commission, carries out a variety of nuclear research programs in many research reactors and nuclear facilities, decommissioning and dismantling operations that require multiple transports of different type of radioactive material such as radioactive waste or experimental fuel rods using a wide range of transportation packages. Packages are designed, constructed and operated to meet safety and regulatory requirements and must be approved by the French Safety Authority. Application for package approval (e.g. type B package) requires a complete and thorough safety demonstration in which flammable gas generation like radiolysis has become a major issue in the last few years. To face these challenges, the CEA unit in charge of transport package construction and transport operations (CEA/STMR, Cadarache Center, France), has developed a many-sided approach based on one hand on a better knowledge of both hydrogen gas generation and oxygen consumption in waste materials, and on the other hand, on design of transport packages or waste containers capable of withstanding dynamic pressure loads due to an inner hydrogen/air mixture explosion. This paper describes the experimental program that CEA/STMR has decided to implement in order to better characterize confined space explosions of hydrogen/oxygen mixtures. It presents how pressure profiles were experimentally measured in various pressure and temperature conditions which are representative of normal and accident transport conditions. It describes how different experimental set-ups were designed and improved progressively so as to obtain reliable and reproducible detonation conditions, investigate the influence of water vapor and characterize pressure profiles (peak pressure, residual pressure, peak duration). This experimental data will be used to qualify computational models and tools that are used to calculate the structural response of waste containers or fuel canisters, and finally demonstrate the safety of transport packages following an inner hydrogen explosion. (authors)

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DOE OSTI2015

Tracking radioactivity content in the spent fuel pool in PWRS - 15150

Over the last decade, significant progress has been made in managing radioisotopes generated in the PWR primary systems. These include limiting the use of cobalt containing materials, along with improved fuel performance and zinc injection to passivate the corrosion layers to limit the accumulation of Co-58 and Co-60 on primary coolant surfaces. Attention here is directed to the spent fuel pool where accumulation of spent fuel and activated material represents a growing source of mobile radioactivity supplemented with periodic exchanges of water from the reactor cooling system. Characterization of this source is increasingly important to effective management of disposal options. This paper examines the radioactive resin source term as determined from routine sampling of the coolant and fuel pool liquids. Targeted resin stream sampling and radiochemical analyses conducted at Diablo Canyon serve to inform and provide context for this examination. This examination is supplemented with computer modelling using the DW James Consulting computer program, 3R-SCAN. At most operating plants daily reactor coolant samples are counted using gamma spectroscopy and recorded. These samples serve to track fuel performance and maintain water chemistry. A secondary use for this data is to maintain a day to day tracking of releases of radioisotopes important to waste disposal. Given the coolant concentrations, plant operating history, and corroborative mathematical models and sampling, accumulation of these radioisotopes in coolant cleanup system ion-exchangers can be predicted. Fuel pool sampling is less frequent, but at the same time, things change more slowly there. The fuel pool contains about ten times the volume of the reactor coolant but its cleanup system process rate is comparable to reactor coolant letdown, i.e. 5.047 - 8.833 liters per second [lps] (80-140 gallons per minute [gpm]). A spent fuel pool clean-up demineralizer flow rate is typically about 5.047 lps (80 gpm). Concentrations of longer lived radioisotopes in the spent fuel pool can be comparable to those in reactor coolant. Releases of corrosion products and fission products from the stored fuel occur at a low rate but given the overall inventory can be significant. Recognized scaling factor ratios, particularly those to Co-60, are skewed by the aging of the material in the pool. An additional consideration stems from refueling operations themselves. Prior to refueling, it is common practice to conduct a forced oxidation of the corrosion films within the reactor. This spikes the activity within the reactor coolant. Most of this activity is removed by the reactor coolant purification system prior to removing the reactor head. This paper examines the accumulation of activity in the fuel pool purification ion exchange resins and how it impacts the overall classification of those ion exchange resins. The following discussion focuses on the transport of nickel and cobalt isotopes particularly as it relates to determination of Ni-63 which has become increasingly prominent in disposal classification. (authors)

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DOE OSTI2015

The future of RADCALC- the radioactive material and waste transportation software program - 15154

This paper provides a thorough review of RADCALC, the U.S. Department of Energy (DOE)-developed software tool to assist DOE sites in making radioactive material and waste packaging and transportation determinations in compliance with applicable Federal regulations and DOE Orders and Standards. There are more than 400 registered users of the RADCALC tool across the DOE complex. DOE has also authorized a limited number of users from certain universities and foreign countries. DOE periodically upgrades the software consistent with changing domestic and international requirements, as well as needs unique to DOE. Due to recent changes in regulatory requirements by the Department of Transportation (DOT), it is necessary that the current version of RADCALC should be updated in a timely manner. Prior to investing funds in software upgrades, DOE decided to evaluate the nature of changes required and level of operational support and looked for alternatives for meeting sites' operational needs. DOE formed a joint review team consisting of experts from Energy Facility Contractors Group (EFCOG) sites, commercial waste treatment facilities and the National Laboratories. This team evaluated the RADCALC software, commercially available software packages, and internal DOE site-developed programs against an established set of criteria. The team also reviewed and prioritized regulatory changes and user-recommended enhancements. Based on extensive review, the team recommended that DOE continue to support RADCALC noting that RADCALC is a cost-effective, technically robust tool that can perform complicated radioactive material shipping determinations. The tool promotes accuracy, consistency, and completeness between shipments by a single shipper, between shippers at a given site, and between sites. The isotopic coverage, decay algorithm and the range of calculations are without compare either within DOE or in industry in a tool of this kind. The team also recommended that DOE should actively promote RADCALC use throughout the DOE complex and consider cooperative arrangements with the Department of Defense or other government entities that may have similar needs. (authors)

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DOE OSTI2015

Setting an optimal procedure for transferring intermediate radioactive effluents: contribution of laboratory experiments - 15234

The INB35 nuclear facility at the CEA Saclay center is in charge of collecting and treating radioactive effluents generated from site's research activities. The first step of the process consists of an evaporation. The concentrate is an IL radioactive suspension with high salt concentration and suspended matter content. Until the 1990's, the resulting effluents were bitumen coated and since 2012 they have been cemented. During the 1990's, concentrates were stored in 7 tanks, named MA50X with X from 1 to 7. With today's changes to safety regulations, the effluents must now be transferred before tank dismantling. According to the site records, the compositions of the liquid wastes MA501, MA502, MA503, MA504, MA506 and MA507 are similar. From 2013 to 2014, effluents contained in MA502, MA503, MA504 tanks have been transferred. Several difficulties have been encountered: - The tanks were historically different, especially manhole tank access - Since the 1990's, effluent composition has changed due to water evaporation. The salt concentration has increased and insoluble elements have formed a deposit in the bottom of the tanks resulting in high composition heterogeneity - In some tanks, a wall breeze-blade exists dividing the tank into two zones, which complicates the interventions. Major challenges have been met and will be described in the presentation. - To obtain an effluent which is as homogeneous as possible, the duration of mixing has been optimized. - To pump the liquid, it has been necessary to treat the effluent with nitric acid. This step has been optimized thanks to experiments in laboratory and hydrodynamic modelling. - A mobile workshop has been designed in order to perform the effluent transfer. It is set up directly above the tanks and can be moved from one tank to another on rails. Laboratory experiments have been carried out to characterize the suspension and sludge in order to understand its chemical behavior. - The treatment of the effluent has been studied in the laboratory and is now mastered. - The analytical strategy has been optimized. The close collaboration between the analytical laboratory and the dismantling project was a real advantage. Thanks to this, the duration of the last transfer on MA 504 was reduced by half, compared to the first operations. Transfer operations will be described and the differences and similarities between laboratory and in-situ observations will be pointed out. (authors)

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DOE OSTI2015

Design and testing of a sensitive heat-flow calorimeter for measuring radioactive waste in 55-gallon (US) drums - 15334

A sensitive large volume calorimeter for measuring the thermal power generated by the radioactive decay of heat producing materials in 55-gallon drums has been developed and tested. The technology is applicable to the measurement of all radionuclides that decay by alpha or beta decay such as waste or product material containing plutonium, tritium and americium. It is also relevant to the measurement of drums where gamma rays from radioactive decay are captured locally in attenuating materials and the resulting rate of thermal energy deposition (sample thermal power) is of sufficient magnitude for a calorimetry measurement. The initial application of the drum calorimeter is for measuring and characterizing plutonium-bearing waste drums. The calorimeter design has optimised the two characteristics of high thermal sensitivity and reduced measurement times. In order to achieve these characteristics for large volume drums, the heat-flow method of operation was chosen in a single measurement chamber configuration. Using a drum-lifting device, the waste drum is positioned on a sliding platform and subsequently moved by the operator inside an octagonal inner chamber, which acts as the heat sink. The heat-sink chamber has significant thermal inertia and remains effectively at a constant temperature throughout the duration of a drum measurement. The octagonal inner measurement chamber is constructed inside a multi-layer external thermal enclosure, which provides both thermal insulation and significant thermal inertia. The external thermal enclosure isolates the measurement process from the effects of changes in the external ambient temperature. During a measurement, heat flows from the drum being measured to the heat sink. This heat-flow is measured by a series of thermopile sensor assemblies, which are positioned between the waste drum and the heat sink. A high sensitivity of greater than 200 μV/mW (microvolts per milliwatt) has been achieved using multiple close-coupled thermopile sensors in each assembly. Reduced measurement times are achieved by positioning the thermopile sensor assemblies close to the drum surface in order to reduce the thermal transport delay. The instrument incorporates a calibrated precision power supply, which requires annual recalibration and is traceable to national standards. It is used to power a calibration heater that has been built into a calibration and test drum. The calibration and test drum is used both to calibrate the instrument and provide periodic confirmation of correct operation. Measurements of calorimeter performance are reported for a range of drum thermal powers. Long-term temperature drift measurements are used to provide estimates of the minimum detectable drum thermal power and hence the minimum detectable activity for both plutonium and tritium. (authors)

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DOE OSTI2015

Ecological risk assessment software for radioactive waste management facilities - 15464

Environmental Risk Assessment (ERA) is a systematic process which can be used to identify, quantify and characterize the risk posed by contaminants and physical stressors in the environment on biological receptors, including human and non-human biota. Human receptors can be addressed through a Human Health Risk Assessment and non-human biota can be addressed through an Ecological Risk Assessment (EcoRA). In 2012 the Canadian Standards Association (CSA) issued a CSA standard N288.6 Environmental Risk Assessment at Class I Nuclear Facilities and Uranium Mines and Mills. The standard applies to radioactive waste management facilities, as well as nuclear power plants. The ERA can also be used retrospectively, to estimate the effect that a contaminant or stressor has already had on the environment; and can be used to assist in determining possible remedial or risk management options to return the quality of the affected environment to a predetermined state. While industry-standard tools for estimating radiological doses to humans are readily available, there is a need for a tool to implement the CSA N288.6 standard for Ecological Risk Assessment. To address this need, AMEC developed a computer code called -Assessment of Impact of Contaminants on Ecological Receptors (AICER)-. The code was developed to be consistent with the CSA standard N288.6, by assessing the impact of radiological contamination on non-human biota for normal operations. The code can also be used to screen for the potential effects of conventional contamination on the environment. In addition, the code is also applicable for abnormal events and accidents with the appropriate modification of some parameters used in the code. The AICER code was qualified in accordance with the requirements of the Canadian quality standard CSA N286.7 -Quality Assurance of Analytical, Scientific, and Design Computer Programs for Nuclear Power Plants-. The functionality of the code is illustrated with screen shots in the paper, following the introduction of the theoretical model for the dose calculation. An inter-comparison with RESRAD-BIOTA and ERICA Assessment tool at the conceptual level was conducted. In addition, the practical application of the code for the assessment of the impacts related to an existing radioactive waste management facility is provided. In particular, analysis was conducted to evaluate radiological doses to non-human biota resulting from a fire accident scenario involving an existing waste storage and processing facility. Analysis included all feasible waterborne and airborne exposure pathways. Doses were evaluated for a range of species representing valued ecological components at various trophic levels and compared with benchmark values for indicator species. (authors)

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DOE OSTI2018

Issues Affecting the Assessment of Impacts of Disposal of Radioactive and Hazardous Waste - 18177

BIOPROTA is an international collaborative forum that seeks to address the key uncertainties in long term assessments of radionuclide and other contaminant releases into the environment arising from solid radioactive waste disposal. (see www.bioprota.org). Two workshops have been organized through the BIOPROTA forum to consider the non-radiological post-disposal impacts of radioactive waste disposal. In these workshops, it was noted that, ideally, the assessment of radionuclides and hazardous materials should be such that consistent assumptions and criteria are used in risk evaluation and that in cases where this is not possible, a clear understanding of the reasons why needs to be provided. Accordingly, a detailed review was organized through BIOPROTA to provide information to support development of a consensus on how to address this issue. The study suggests that there is a continuing need for some common measure of hazard that supports identification of risk management priorities for mixed hazardous waste. Without it, effective optimization of management of these mixed hazardous wastes will be restricted. (authors)

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DOE OSTI2019

Recovery of Buried Discrete Radioactive Objects by Channel-Specific Analysis of Gamma Scan Data - 19110

During radiological clean-up activities, gamma scan survey data collected from across an industrial site revealed spatially varying concentrations in radioactivity related to naturally occurring soil constituents or dissimilar fill materials that may have been placed at the site. A three-step process was developed to assess the data in order to identify radiation anomalies that most likely could represent small, buried discrete radioactive objects (ROs). The three-step process consists of: (1) define the radiological composition of the site isotopically, (2) spatially parse the data into more homogeneous subgroupings based on the site's radiological composition, and (3) perform analysis of the data subgroupings to identify statistical and spatial population outliers (i.e., radiation anomalies) that otherwise may be masked by the overall variation in the data. Using this process, locations containing buried low-specific activity ROs or higher activity ROs whose radiation is attenuated by soil can be identified and investigated. (authors)

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DOE OSTI2019

Revisit of Phase Separation of Borosilicate Glass for Vitrification of High Level Radioactive Waste - 19166

Phase separation of glasses for vitrification of high level radioactive waste has been experimentally studied to know uncertainty of the phase separation conditions; temperature and glass compositions as well as to propose possibility of recycling of heavy metals on a longer time scale. Our current results at 700 and 750 deg. C for ternary borosilicate glass system showed that the phase separation region at 700 deg. C spreads along with a direction of increasing Na{sub 2}O composition when compared with the data in literature. On the other hand, at the glass composition employed in the reprocessing plant in Japan, no segregation was observed. Concerning the application of phase separation phenomena, we had previously reported recovery of almost all elements from the vitrified borosilicate glass containing simulated fission products through phase separation of the borosilicate glass and this suggests that we could propose that after vitrification of high-level radioactive waste heavy metals may be retrieved as energy or materials resources when they are required to utilize on a longer time scale. From this standpoint, phase separation of lead borosilicate glass was also studied to seek economical and versatile recycling. (authors)

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DOE OSTI2019

Overview of Novel Canadian Approaches for Decontamination of Solid Radioactive and Mixed Waste - 19476

The availability of decontamination-for-release technology options provide a path forward to economically decontaminate radioactive metallic waste and enable its disposal as clean scrap metal, to decontaminate tools/equipment for unrestricted refurbishment or reuse, and to carry out inspection of in-service components. Projects involving management of solid mixed waste pose complex technical challenges that require the development and application of novel decontamination processes. The objective of this paper is to present project case studies that provide an overview of Kinectrics' experience with the decontamination of solid radioactive and mixed waste. Novel approaches to decontamination, recent developments, key results and lessons learned will also be discussed. (authors)

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