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

SRNL Training Courses for Packaging, Transportation, and Storage of Radioactive Material - 17597

The U.S. Department of Energy (DOE) Packaging Certification Program (PCP), Office of Packaging and Transportation has sponsored a suite of training courses that have been conducted since the early 1990s. Of these training courses, the Savannah River National Laboratory (SRNL) is currently leading the effort for two courses: (1) Management of SARP Preparation, and (2) Nuclear and Other Radioactive Materials Transport Security and Leak Testing. This paper provides an overview of both courses. (authors)

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DOE OSTI2015

Experience in environmental remediation of radioactively contaminated site in waste disposal area in Istanbul - 15011

In this paper specific environmental remediation examples of method applications are given. Radiological contaminated site was created by non-nuclear industry in Turkey. This is designated as a contaminated site and can ultimately lead to undesired health effects for residents. Appropriate actions have been taken to protect human and environment from contaminated site. As every site has its own characteristics, contaminated area was examined by sampling and surveying works. Before taken into action, remediation actions need to be justified and optimized. Radiation levels and homogeneity of contamination was investigated. Remediation program was established and taken into action in parallel two main works. First one covers all actions to the contamination itself. The site was isolated properly and all equipment was deployed. Working methods for segregation of contaminated material from the soil were determined. Besides, risk evaluation was carried out. Pathways between the contaminated soil and people were taken into account. Groundwater samples were taken and analyzed. According to these results, evacuation, area isolation or changing land use options were evaluated. Isolation of the site and removal of radioactive materials were determined as appropriate actions. (authors)

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DOE OSTI2015

A regulatory perspective on the response function method for determining allowable content of a radioactive materials transportation package or storage cask - 15023

In order to simplify shielding design, there has been an increasingly use of a so-called dose rate response function method for determining the allowable contents in radioactive materials (including spent fuel) transportation package and spent fuel storage cask designs. In essence, this approach uses a pre-calculated dose rate at a given location outside the storage cask or transportation package for a given source inside the cask. It assumes that there is a one-to-one relationship between a source inside the cask and the dose rate outside the cask, i.e., a dose rate response function of a source. Thus, the allowable content is obtained by simply dividing the targeted dose rate by the response function. This method converts the complex particle transport problem into a simple arithmetic calculation. If used correctly and prudently, this approach can greatly simplify the shielding design of casks. However, the use of this method must demonstrate the validity of the assumption, that is, the nuclear physics property of the media through which the particle transverses must be similar to that used in computing the response function. Because of the assumptions and approximations, either implicitly or explicitly, made in the models for dose rate response functions, to assure safety the users should also consider the errors in the models and the bias and uncertainties associated with the code used to calculate the response function. This paper reviewed the advantages and some of the major potential pitfalls of this method from the regulatory perspective. (authors)

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DOE OSTI2015

A new measurement method of tritiated water with no radioactive waste- for monitoring post - 15050

A new measurement method has been developed by using a conventional liquid scintillation counter (LSC) with plastic scintillator pellets (PS-pellets) which is used as an alternative material of liquid scintillator. The measurement efficiency (%=cpm/dpm x 100) of PS-pellets with a conventional LSC is as high as using a liquid scintillator. Measurement efficiency of tritiated water was more than 50% by using PS-pellets for LSC without liquid scintillator. Qualitative and quantitative analysis were possible by using PS-pellets for LSC, because the PS-pellets method showed good linearity between the activity and count rate. Also, an end point of a spectrum showed a maximum energy of a radionuclide when one channel estimated 0.5 keV. The PS-pellets were reusable by dipping them in 5% soapsuds of hot water for more than overnight except for a case of measuring for a long elapsed time. So, no radioactive waste of liquid scintillator and PS-pellets is produced with this method.We designed a new field monitor and a monitoring post. (authors)

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DOE OSTI2015

Dual solidification process of BN-350 liquid radioactive waste using high technology polymers and newly designed encapsulation techniques - 15198

Application of high technology polymers no.910 and no.960 manufactured by NOCHAR, USA, for treatment of liquid radioactive waste (LRW) have demonstrated good results at test sites worldwide. At the BN-350 site (Aktau, Kazakhstan), the tests were conducted successfully with contaminated oil, alkaline and salt containing LRW. This paper represents the results of immobilization of NOCHAR polymers, saturated with LRW, into solid matrix to comply with Kazakhstan regulatory requirements for long term storage and/or disposal. (authors)

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DOE OSTI2015

Hanford double-shell tank AY-102 radioactive waste leak investigation update - 15302

Tank AY-102 was the first of 28 double-shell radioactive waste storage tanks constructed at the U.S. Department of Energy's Hanford Site, near Richland, WA. The tank was completed in 1970, and entered service in 1971. In August, 2012, an accumulation of material was discovered at two sites on the floor of the annulus that separates the primary tank from the secondary liner. The material was sampled and determined to originate from the primary tank. This paper summarizes the changes in leak behavior that have occurred during the past two years, inspections to determine the capability of the secondary liner to continue safely containing the leakage, and the initial results of testing to determine the leak mechanism. (authors)

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DOE OSTI2015

In-service inspection of radioactive waste tanks at the Savannah River Site - 15410

Liquid radioactive wastes from the Savannah River Site (SRS) separation process are stored in large underground carbon steel tanks. The high level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks, will be required to be in service for times exceeding the initial intended life. Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. A structural integrity program is in place to maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. In-service inspection (ISI) is an essential element of a comprehensive structural integrity program for the waste tanks at the Savannah River Site (SRS). The ISI program was developed to determine the degree of degradation the waste tanks have experienced due to service conditions. As a result of the inspections, an assessment can be made of the effectiveness of corrosion controls for the waste chemistry, which precludes accelerated localized and general corrosion of the waste tanks. Ultrasonic inspections (UT) are performed to detect and quantify the degree of general wall thinning, pitting and cracking as a measure of tank degradation. The results from these inspections through 2013, for the 27 Type III/IIIA tanks, indicate no reportable in-service corrosion degradation in the primary tank (i.e., general, pitting, or cracking). The average wall thickness for all tanks remains above the manufactured nominal thickness minus 0.25 millimeter and the largest pit identified is approximately 1.70 millimeter deep (i.e., less than 10% through-wall). Improvements to the inspection program were recently instituted to provide additional confidence in the degradation rates. Thickness measurements from a single vertical strip along the accessible height of the primary tank have been used as a baseline to compare historical measurements. Changes in wall thickness and pit depths along this vertical strip are utilized to estimate the rate of corrosion degradation. An independent review of the ISI program methodology, results, and path forward was held in August 2009. The review recommended statistical sampling of the tanks to improve the confidence of the single strip inspection program. The statistical sampling plan required that SRS increase the amount of area scanned per tank. Therefore, in addition to the baseline vertical strip that is obtained for historical comparisons, four additional randomly selected vertical strips are inspected. To date, a total of 104 independent vertical strips along the height of the primary tank have been completed. A statistical analysis of the data indicates that at this coverage level there is a 99.5% confidence level that one of the worst 5% of all the vertical strips was inspected. That is, there is a relatively high likelihood that the SRS inspection program has covered one of the most corroded areas of any of the Type III/IIIA waste tanks. These data further support the conclusion that there are no significant indications of wall thinning or pitting. Random sampling will continue to increase the confidence that one of the worst 5% has been inspected. In order to obtain the additional vertical strips, and minimize budget and schedule impacts, data collection speed for the UT system was optimized. Prior to 2009, the system collected data at a rate of 32 square centimeters per minute. The scan rate was increased to 129 - 160 square centimeters per minute by increasing the scanner step and pixel sizes in the data acquisition set-up. Laboratory testing was utilized to optimize the scan index/pixel size such that the requirements for wall thinning and pit detection were still maintained. SRS continues to evaluate improvements to ultrasonic equipment. (authors)

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DOE OSTI2016

Effect of Radiation from Low-Level Radioactive Waste Leachate on Antioxidant Depletion in HDPE Geo-membrane - 16087

Low-level radioactive waste (LLW) and mixed waste (MW) disposal facilities in the US are required to have a service life in excess of 1000 yr. Understanding the rate of degradation of high-density polyethylene (HDPE) geo-membranes (GMs) used in LLW and MW facilities is necessary to assess their service life when conducting a performance assessment for a disposal facility. In LLW and MW disposal facilities, GMs may be exposed to radionuclides that may accelerate antioxidant depletion and degradation. This paper evaluates how radiation dosage affects antioxidant depletion in HDPE GMs and makes inferences regarding the impact of radiation on GM service life. Specimens of 0.04-mm-, 0.1-mm-, and 2-mm-thick HDPE GMs were exposed to a sealed source of Am-241 to simulate alpha radiation from LLW leachate (e.g., uranium). The specimens were exposed to the sealed source for 1-50 h, followed by oxidative induction time (OIT) testing to determine depletion of antioxidants in response to exposure. Penetration of alpha particles in the HDPE GM exposed to the sealed source was evaluated with the program Geometry And Tracking (GEANT4). The GEANT4 modeling indicates that physical dose deposition for alpha particles largely occurs on the superficial layers and decreased monotonically with depth to a negligible dose (∼ 28 μm). For a 2-mm-thick HDPE GM, there was no change in OIT after exposure, whereas in much thinner GMs (e.g., 0.04 mm), significant antioxidant depletion occurred after exposure, reflecting the shallow penetration of alpha radiation at the surface of HDPE GM. (authors)

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DOE OSTI2016

New High-Throughput Measurement Systems for Radioactive Wastes Segregation and Free Release - 16166

The European Metrology Research Program (EMRP) is a coordinated research-and-development program that pursues the closer integration of national research programs. The EMRP Metrology for Decommissioning Nuclear Facilities 'MetroDecom' Joint Research Project ENV54 (JRP ENV54), a 36-month project started in September 2014, delivers research products addressing all aspects of the decommissioning process. This includes the characterization of waste materials present on the decommissioning site, pre-selection prior to free release or repository acceptance measurement, free-release measurement, the measurement of waste packages thermal power prior to the repository storage, and the monitoring of stored wastes and repositories in the very long term. This paper describes the track record of two JRP ENV54 working packages, which address the measurement facilities for pre-selection of waste materials prior to measurement for repository acceptance or possible free release (segregation measurement system); and free release (free release measurement system), based on a single standardized concept characterized by unique, patented lead-free shielding. The key objective is to improve the throughput, accuracy, reliability, modularity and mobility of segregation and free-release measurement. This will result in more reliable decision-making with regard to the safe release and disposal of radioactive wastes into the environment and, resulting in positive economic outcomes. (authors)

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DOE OSTI2016

A Study on Vitrification for LLW Radioactive Wastes - 16235

Ash and ion exchange resin wastes to be vitrified were considered among various radioactive wastes in the Japanese nuclear industry. The appropriate glass composition for the waste is calculated using the computer program. The glass composition is systematically studied using a ternary diagram, considering viscosity, electrical conductivity, and durability. The optimized glass composition based on calculated parameters of vitrification process and the convenience of mixing wastes is selected for crucible experiment. Since differences between calculated values and measured values may occur upon evaluation of glass formulations using glass formulation computer simulation program, the calculated glass formulations should be verified through experiments on a laboratory scale for more meticulous development of glass formulations. For the follow-up development of glass formulations, several glass formulations are planned to be developed. The candidate glasses for treatment of ash and ion exchange resin were melted and fabricated using reagents, respectively. The chemicals were weighed out to produce the desired batch compositions and mixed thoroughly and then put into clay crucibles for melting. Next, the crucibles were introduced into an electric furnace at 1,150 deg. C in the ambient atmosphere and pressure with occasional stirring by a quartz rod. A hold time of 1 hour was introduced to promote homogenization. Finally, the melted material was poured on a graphite plate to be rapidly cooled. The viscosities of all fabricated glasses were in the desired range. After rapid cooling, the micro-structure of the glasses was analyzed using the SEM. They exhibited amorphous structure with good homogeneity. (authors)

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DOE OSTI2016

Siting Radioactive Waste Management Facilities: Successes and Failures - 16392

Our assessment in 2015 of 39 attempts to site nuclear waste facilities in democratic countries over the past 40 years identified common inherent threads of both success and failure. In the successes, including those efforts still ongoing, the common threads include a) providing the host communities the opportunity to withdraw at any point, b) the physical presence of a significant project team in the host community, and c) the establishment of credibility and trust in the project team, the regulator and, where possible, a neutral commenting party. In the failures, two dominant causes are present. The first is the failure to properly address the legitimate concerns of indigenous people. This has caused failures in the US and Australia. The second and more common cause is lack of political consensus amongst the host community. In nearly every case when the facility proponent selects the site and attempts to gain host approval, the effort has failed. Within the failed cases, two topics appear to be common. The first is that the definition of 'host community' is initially too narrow by the proponents. In many situations, the very local community is supportive of radioactive waste projects but the larger community is not. The second factor is a failure on the part of the proponent to accurately communicate the risks associated with the project, driven significantly by differences in the manner in which the general public and experts process risk communications. (authors)

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DOE OSTI2016

Determination of Erosion/Corrosion Rates in Hanford Tank Farms Radioactive Waste Transfer System Pipelines - 16395

The twenty-eight double-shell underground radioactive waste storage tanks at the U.S. Department of Energy's Hanford Site near Richland, WA are interconnected by the Waste Transfer System network of buried steel encased pipelines and pipe jumpers in below-grade pits. The pipeline material is stainless steel or carbon steel in 51 mm to 152 mm (2 in. to 6 in.) sizes. The pipelines carry slurries ranging up to 20 volume percent solids and supernatants with less than one volume percent solids at velocities necessary to prevent settling. The pipelines, installed between 1976 and 2011, were originally intended to last until the 2028 completion of the double-shell tank storage mission. The mission has been subsequently extended. In 2010 the Tank Operating Contractor began a systematic evaluation of the Waste Transfer System pipeline conditions applying guidelines from API 579-1/ASME FFS-1 (2007), Fitness-For- Service. Between 2010 and 2014 Fitness-for-Service examinations of the Waste Transfer System pipeline materials, sizes, and components were completed. In parallel, waste throughput histories were prepared allowing side-by-side pipeline wall thinning rate comparisons between carbon and stainless steel, slurries and supernatants and throughput volumes. The work showed that for transfer volumes up to 6.1 E+05 m{sup 3} (161 million gallons), the highest throughput of any pipeline segment examined, there has been no detectable wall thinning in either stainless or carbon steel pipeline material regardless of waste fluid characteristics or throughput. The paper describes the field and laboratory evaluation methods used for the Fitness-for-Service examinations, the results of the examinations, and the data reduction methodologies used to support Hanford Waste Transfer System pipeline wall thinning conclusions. (authors)

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DOE OSTI2016

Behavior of Radioactive Cesium in Municipal Solid Waste of Thermal Treatment Plants after Fukushima Nuclear Power Plant Accident - 16626

Municipal solid waste (MSW) contaminated by radioactive cesium (r-Cs) has been treated by various thermal treatment facilities in Japan. However, the behavior of r-Cs in their plants is scarcely known. Therefore, we investigated 4 types of the thermal treatment plants to understand the r-Cs behavior during different thermal treatments of the MSW. The distribution of r-Cs between residue discharged from bottom of furnace (bottom ash, non-combustibles or slag) and fly ash was affected by furnace type. The leaching rate of r-Cs from the fly ash and the bottom residues were measured. The leaching rate from the former was high and ranged from 40 to 100%. In contrast, the most of r-Cs in the latter did not dissolve into water. Furthermore, we evaluated whether baghouse (an air-pollution control equipment) successfully removed r-Cs compound in flue gas. R-Cs was efficiently removed regardless of thermal treatment type. (authors)

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DOE OSTI2018

Innovation in Contracting Strategy for a Master Services Agreement for Remediation of Low Level Radioactive Waste - 18381

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. It is currently one of the largest environmental remediation projects in Canada. The safety of workers, the public and the environment while delivering the PHAI on time and within budget is CNL's number one priority. The PHAI is being carried out as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.2 million cubic metres of historic waste from several locations, including the Port Hope Harbour, a closed municipal landfill and approximately 800 residential properties. This complex project requires contracting the services of qualified, skilled and experienced environmental remediation contractors to meet the diverse needs of the myriad tasks that comprise the remediation and restoration of all properties. This paper will highlight how the CNL contracting services team developed an innovative approach to a contracting strategy designed to deliver the safe execution of work and keep the project on time and within budget. The approach addresses the complexities of a scope of work that consists of multiple work packages to be executed both concurrently and in tandem, and that require highly specialized trades/skills, as well as high-level project management expertise to ensure seamless coordination among the various contractors. The author will discuss CNL's strategy to engage the market with a Request for Expression of Interest to identify potential qualified suppliers and gauge their level of interest, issue a Request for Prequalification (RFPQ) to determine eligible contractors and, ultimately, issue a Request for Proposal for a Master Services Agreement (MSA) that would competitively evaluate each contractor's direct experience, technical expertise, quality management programs, safety systems compliance and track record. Using CNL's experience as a case study, the paper will highlight how an innovative approach can positively affect project timelines by describing how CNL addressed the challenge associated with the limited time available to implement procurement processes and deliver a series of executed agreements while ensuring that work would begin on time and key project milestones would be met. The author will summarize how CNL's strategy put the project back on schedule, without compromising the process guideline/requirements, and ensured that a qualified and experienced set of contractors, committed to a strong safety culture, will be available to perform the work required when the cleanup in the Port Hope community begins in 2018. (authors)

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DOE OSTI2019

Removal of Historic Low-Level Radioactive Waste from the Highland Drive South and Alexander Street Ravines - 19155

In Port Hope, the Highland Drive South Ravine (HDSR) is the discharge area for groundwater originating from the closed Highland Drive Landfill site. The presence of low-level radioactive waste (LLRW) in the HDSR is the result of a contaminated groundwater plume flowing from the landfill site. The contaminant plume contains elevated concentrations of uranium and arsenic and extends from the landfill south to discharge groundwater to shallow soils in a wet discharge area within the ravine. The groundwater discharges to Hunt's Ponds and Highland Drive South Creek, which are immediately downstream of the wet discharge area and will be included in the remediation design. Remediation and environmental management plans for HDSR have been developed within the framework of the Port Hope Project. Recent subsurface investigations have identified LLRW at greater depths than previously anticipated. Other investigations of surrounding properties have identified LLRW that is primarily from filling activities in backyards. The steep topography of the site and the presence of the creek and seepage face present remediation challenges. Also in Port Hope, the LLRW in the Alexander Street Ravine Area is the result of end dumping, filling of private backyards and stack deposition from the former refinery. The original site extent was focused on the deeper contamination due to end dumping and included a portion of Alexander Creek. However, recent investigations have confirmed other fill areas and the extent of stack deposition that are generally contiguous with the end dumping area. Restricted access to the site will result in all remediation in this area taking place via a single access point along the narrow existing Hayward Street. While both of these sites are ravines in Port Hope containing LLRW, the origins of contamination and remediation approaches are very different. This paper will compare these two challenging remediation sites. (authors)

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DOE OSTI2019

Inventory and Characterization of Categories 3-5 Disused Sealed Radioactive Sources - 19284

Brazil has one of the biggest inventories of SRS in the world, not only because of the large industrial park, the well-developed health care infrastructure and the numerous mining operations, but also by the use of radioactive sources in consumer products, such as smoke detectors, lightning arresters, surge protectors and even cardiac pacemakers. Presently, the inventory of DSRS at the WML is of more than 10,000; if the sources from lightning arresters, smoke detectors and Ir-192 with activities above the clearance levels are included, the estimate is over 150,000. The Waste Management Laboratory (WML) of the Nuclear and Energy Research Institute (IPEN/CNEN-SP) is responsible for the management and treatment of these sources. An important aspect to consider is that Brazil does not have a defined strategy or a protocol for screening and inventorying these sources, whose management was limited only to storage. The WML started the development of methods and operational work, to characterize and reduce the total volume of this inventory for the borehole-type final disposal facility yet to be constructed, where the DSRS will be disposed of. The information available from these sources is in a database created in 1980, when the sources began to be received. Several records on the FRSD that were checked are incomplete or simply incorrect, since there is no correspondence between the dates of entry of these sources in the storage and their physical characteristics. Some reasons can be listed to try to understand the divergences: the inspection when receiving the sources was not possible, no inspection procedure was defined, there were technical difficulties to open the packages. The objective of this work is to organize the inventory in a standardized way, control access, traceability, reduction of the necessary time to locate and characterize the sources upon reception, and of the repetitive work in the operational routine, and finally release of space in the storage. (authors)

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DOE OSTI2019

The Development of Legacy Waste Scaling Factors for Radioactively Contaminated Soils at Chalk River Laboratories - 19333

The Chalk River Laboratories (CRL) began nuclear operations in 1947. Beginning in 1952 after the NRX accident, the management of excess contaminated liquids from the accident, and later in 1953, other waste liquids, were intermittently dispersed to trenches and pits in the local sand. The local sands proved effective at immobilization of actinides, and most activation and fission products. However there was very limited retention of some more commonly mobile radionuclides. Sand filtration performance and radionuclide migration was studied beginning in 1952. Monitoring data indicated slow migration of most radioactive contaminants away from the trenches and pits due to the adsorbing capacity of the sand substrate. The radionuclide migration continues to be studied today in frequent plume monitoring studies. Because many hard-to measure (HTM) radionuclides may also present as long term dose contributors in post remediation scenarios, or in disposal site performance assessments, the quantification of these HTM radionuclides is important. The use of scaling factors between HTM radionuclides and easy-to-measure (ETM) radionuclides (gamma emitters) for packaged soil wastes would alleviate significant effort in sampling individual waste containers. As part of plume monitoring studies and the environmental remediation process, significant characterization of soils have been performed at the site. The soil sample data from multiple historic waste management areas across site is combined and relationships between HTM radionuclides and ETM or key radionuclides are examined. Additional (secondary) relationships between ETM key nuclides are also explored to evaluate secondary scaling relationships between ETM key radionuclides (e.g., Co- 60/Cs-137), because of the significant decay of the activation product key radionuclide Co-60, (predominantly 10-20 half-lives). The evaluation is yielding consistent scaling factors for legacy soil wastes in the historic waste management areas. Given the hydrogeology of the site, a sandy shallow unconfined aquifer, and the annual precipitation of about 0.85 m/a, this evaluation reveals some expected relationships between fission products and actinides and some unanticipated relationships between more mobile radionuclides and key gamma emitters. (authors)

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DOE OSTI2019

Immobilization of Radioactive Wastes Using Geopolymers - 19483

This work explores an encapsulation and immobilization technology for spent radioactive ion exchange (IX) resins using geopolymer materials as a matrix. Geopolymer-IX resin composites with different IX resin loadings were prepared by mixing an aluminosilicate material (metakaolin), sodium hydroxide and inactive IX resin (Amberlite IRN 150) and curing the mixture at room temperature with or without γradiation applied using an external radiation source. The metakaolin and sodium hydroxide reacted to form a geopolymer framework encapsulating the IX resin beads. The IX resin loading capacity and the encapsulation efficiency of the geopolymer material were investigated by examining the morphology and distribution of resins within the geopolymer framework using optical microscopy, and by measuring the compressive strengths of the geopolymer-IX resin composites. These properties of the geopolymer-IX resin composites were compared with those of ordinary Portland cement (OPC)-IX resin composites. The geopolymer matrix allowed for a higher IX resin loading capacity than OPC. The geopolymer-IX resin composite was stable under γ-radiation for 28 d. In leaching tests performed with encapsulated Sr-loaded IX resins, the ability of the geopolymer to prevent leaching of Sr{sup 2+} into solution was greatly superior to that of OPC. (authors)

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