Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “radioactive”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 217 records · Page 12

Role of Human Intrusion in Decision-Making for Radioactive Waste Disposal - Results of the IAEA HIDRA Project - 16287

Disposal of radioactive waste aims to protect people and the environment from the potential hazards of the materials by removing the wastes from the human environment. This approach is considered desirable to dispersal of the waste in the environment. However, the waste becomes concentrated in one location, which poses potential hazards should the waste become disturbed at a future date. For radioactive waste, it is generally expected that in preparing safety cases for such disposal facilities, the possibility of the wastes being inadvertently disturbed at some point in the future and the potential consequences of such an intrusion need to be considered. Potential inadvertent human intrusion (IHI) has a role in the decision-making and the safety case throughout the lifecycle of a disposal facility. In 2013, the International Atomic Energy Agency (IAEA) formally initiated a project entitled 'Human Intrusion in the Context of Disposal of Radioactive Waste' (HIDRA) to provide insights regarding the role of IHI for these decisions and to provide a common framework for consideration of IHI as part of development of a safety case for a disposal facility. HIDRA was implemented with a series of annual plenary meetings involving participants from more than 20 countries organised into three working groups. The working groups addressed: societal factors, stylised scenarios, and protective measures. The project resulted in a general approach with specific suggestions for consideration of IHI during the lifecycle of a disposal facility. From a general perspective, it was emphasized that IHI is a contributing factor to the safety case, but does not outweigh considerations related to safety of operations and long-term performance for the normal evolution of the facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Performance Test of Removal of Gaseous Radioactive Methyl Iodide by TEDA-Impregnated Activated Carbon - 18049

During the operation of nuclear facility as well as decommissioning stage, various radioactive gaseous wastes are generated. It is generally recognized that radioiodine is potentially one of the most hazardous gas on the human body and environment. Iodine-131 release in the gaseous effluents at the nuclear facilities had been a critical environmental concern. Reducing iodine release is a critical issue in radioactive waste system evaluation, especially in their adherence to the 'as low as reasonably achievable' (ALARA) criteria because radioiodine among these gaseous wastes has radiological harm. Therefore, numerous studies for the removal of radioiodine from the off-gas stream of nuclear facilities have been performed with formation processes and chemical structure. This paper aims to evaluate the performance of removal efficiency of radioiodine gas by TEDA-impregnated activated carbon, which is used for radioiodine retention in nuclear facilities. The removal efficiencies of methyl iodide were evaluated by using a radioactive methyl iodide tracer at various temperatures up to 250 deg. C. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Savannah River Site High Level Radioactive Waste Tank Sample Thermolytic Hydrogen Generation - 20346

To support the revision of flammability calculations and controls for the Savannah River Site (SRS) High-Level Radioactive Waste (HLW) tanks, Savannah River National Laboratory (SRNL) conducted laboratory measurements with the goal of quantifying hydrogen produced in HLW by non-radiolytic chemical reactions (i.e., thermolysis). Testing was performed with HLW tank supernate samples that represented a cross section of types of waste at SRS, including tanks that store typical evaporator concentrate, dissolved salt-cake, dilute recycle stream from the Defense Waste Processing Facility (DWPF), evaporator concentrate of the recycle stream from DWPF, fresh waste from the canyon separations facility, and waste that has been through salt processing cesium removal. Non-radioactive simulant testing included the classes of organic compounds historically introduced into the HLW tanks that were recently shown to be the most active toward the thermolytic generation of hydrogen. The reaction rate equations developed for thermolytic hydrogen generation from each class of organic compound showed direct proportionality to organic compound concentration and hydroxide concentration. Simulant testing identified that highly concentrated waste with high hydroxide concentration had the highest rates of thermolytic hydrogen generation. Using the radioactive tank sample thermolysis measurements at high temperatures and the mechanistic salt dependence of the simulant models, a global model was developed for tank waste thermolytic hydrogen generation as a function of total organic carbon content of SRS HLW. Through the primary functionality of the global model (organic carbon and free hydroxide concentration), the relative reactivity of the organic carbon in the waste was quantified. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The DECOVALEX international collaboration on modeling of coupled subsurface processes and its contribution to confidence building in radioactive waste disposal

Abstract The long-lived radiotoxicity of the high-level radioactive waste generated by nuclear power plants requires safe isolation from the biosphere for many hundreds of thousands of years. An international consensus has emerged that such isolation can best be provided by disposal in mined geologic repositories, a strategy that today is pursued by most countries dealing with radioactive waste. However, the need to predict the performance of such repositories over very long time periods generates large uncertainties that have to be accounted for in safety assessments. The findings from such safety assessments need to be conveyed to all stakeholders in a clear way, such that public confidence in geologic disposal solutions can be achieved. It is suggested here that close international collaboration on the technical aspects of geologic waste disposal has helped, and will continue to help, building trust and increasing confidence. This paper discusses a particular international collaboration initiative referred to as DECOVALEX, which brings together multiple teams and disciplines to collectively tackle complex experimental and modeling challenges related to geologic disposal. By describing how DECOVALEX works and by providing joint research examples, a case is made that such international collaboration contributes to knowledge transfer and confidence building in radioactive waste disposal science.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Feasibility of Using Steel Drums for the Confinement of Burned Low-Level Radioactive Waste Ashes

The selection of suitable materials to construct a facility for the final disposal of radioactive wastes (repository) is a challenge in materials science, since these facilities must be designed to provide confinement to the radionuclides for long periods of time. In the case of the low-level radioactive waste (LLRW) repository, the confinement of the species must be guaranteed for more than 300 years. In this context, the evaluation of the proposed materials must be performed under accelerated laboratory tests. One type of LLRW consists of ashes that result from the incineration of solid combustible radioactive waste. In Argentina, one of the proposed projects is to place these ashes inside steel drums and immobilized them by a cementation process. The issue to be resolved is whether the steel drums will undergo internal corrosion depending on the presence of certain deleterious species. The aim of the present work is to evaluate the corrosion susceptibility of steel drums containing simulated cemented incineration ashes. To this purpose, a special type of specimen was designed and the corrosion process was followed up for almost 4 years. It was found that, although the deleterious effect of chloride is undeniable, corrosion rates are not as high as to compromise the integrity of the steel drums for a design period of 300 years.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Scenario development for safety assessment in deep geologic disposal of high-level radioactive waste and spent nuclear fuel: A review

Radiation and radioactive substances result in the production of radioactive wastes which require safe management and disposal to avoid risks to human health and the environment. To ensure permanent safe disposal, the performance of a deep geological repository for radioactive waste is assessed against internationally agreed risk-based standards. Assessing postclosure safety of the future system's evolution includes screening of features, events, and processes (FEPs) relevant to the situation, their subsequent development into scenarios, and finally the development and execution of safety assessment (SA) models. Global FEP catalogs describe important natural and man-made repository system features and identify events and processes that may affect these features into the future. By combining FEPs, many of which are uncertain, different possible future system evolution scenarios are derived. Repository licensing should consider both the reference or “base” evolution as well as alternative futures that may lead to radiation release, pollution, or exposures. Scenarios are used to derive and consider both base and alternative evolutions, often through production of scenario-specific SA models and the recombination of their results into an assessment of the risk of harm. Furthermore, while the FEP-based scenario development process outlined here has evolved somewhat since its development in the 1980s, the fundamental ideas remain unchanged. A spectrum of common approaches is given here (e.g., bottom–up vs. top–down scenario development, probabilistic vs. bounding handling of uncertainty), related to how individual numerical models for possible futures are converted into a determination as to whether the system is safe (i.e., how aleatoric uncertainty and scenarios are integrated through bounding or Monte Carlo approaches).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

2013 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses

The Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (National Security Technologies, LLC 2007a) requires an annual review to assess the adequacy of the performance assessments (PAs) and composite analyses (CAs), with the results submitted to the U.S. Department of Energy (DOE) Office of Environmental Management. The Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) also require that such reviews be made and that secondary or minor unresolved issues be tracked and addressed as part of the maintenance plan (DOE 1999a, 2000). The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office performed an annual review of the Area 3 and Area 5 RWMS PAs and CAs for fiscal year (FY) 2013. This annual summary report presents data and conclusions from the FY 2013 review, and determines the adequacy of the PAs and CAs. Operational factors (e.g., waste forms and containers, facility design, and waste receipts), closure plans, monitoring results, and research and development (R&D) activities were reviewed to determine the adequacy of the PAs. Likewise, the environmental restoration activities at the Nevada National Security Site (NNSS) relevant to the sources of residual radioactive material that are considered in the CAs, the land-use planning, and the results of the environmental monitoring and R&D activities were reviewed to determine the adequacy of the CAs. Important developments in FY 2013 include the following: • Development of a new Area 5 RWMS closure inventory estimate based on disposals through. FY 2013 • Evaluation of new or revised waste streams by special analysis. • Development of version 4.115 of the Area 5 RWMS GoldSim PA/CA model.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

2016 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analyses

The Maintenance Plan for the Performance Assessments and Composite Analyses for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (National Security Technologies, LLC 2007a) requires an annual review to assess the adequacy of the performance assessments (PAs) and composite analyses (CAs), with the results submitted to the U.S. Department of Energy (DOE) Office of Environmental Management. The Disposal Authorization Statements for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) also require that such reviews be made and that secondary or minor unresolved issues be tracked and addressed as part of the maintenance plan.The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) performed an annual review of the Area 3 and Area 5 RWMS PAs and CAs for federal fiscal year (FY) 2016. This annual summary report presents results and conclusions of the FY 2016 review, and determines the adequacy of the PAs and CAs. Operational factors (e.g., waste forms and containers, facility design, and waste receipts), closure plans, monitoring results, and research and development (R&D) activities were reviewed to determine the adequacy of the PAs. Likewise, the environmental restoration activities at the Nevada National Security Site (NNSS) relevant to the sources of residual radioactive material that are considered in the CAs, the land-use planning, and the results of the environmental monitoring and R&D activities were reviewed to determine the adequacy of the CAs. Important developments in FY 2016 include the following: Development of a new Area 5 RWMS closure inventory estimate based on disposals through FY 2016, Evaluation of new or revised waste streams by special analysis, and Development and acceptance of version 4.200 of the Area 5 RWMS GoldSim PA/CA model. The Area 3 RWMS has been in inactive status since July 1, 2006, with the last shipment received in April 2006. The FY 2016 review of operations, facility design, closure plans, monitoring results, and R&D results for the Area 3 RWMS indicates no changes that would impact PA validity. The conclusion of the annual review is that all performance objectives can be met and the Area 3 RWMS PA remains valid. There is no need to revise the Area 3 RWMS PA. Review of Area 5 RWMS operations, design, closure plans, monitoring results, and R&D activities indicates that no significant changes have occurred. The FY 2016 PA results, generated with the Area 5 RWMS v4.200 GoldSim PA model, indicate that there continues to be a reasonable expectation of meeting all performance objectives. All results remain less than 50% of their performance objectives. The results and conclusions of the Area 5 RWMS PA are judged valid, and there is no need to the revise the PA. A review of changes potentially impacting the CAs indicates that no significant changes occurred in FY 2016. A new Corrective Action Unit (CAU), CAU 576, with corrective action sites near the Area 3 RWMS and Area 5 RWMS was identified in FY 2016. The sites are currently under investigation and not expected to be important sources for the CA. The continuing adequacy of the CAs was evaluated with the new models, and no significant changes that would alter the CAs’ results or conclusions were found. The revision of the Area 3 RWMS CA, which will include the Yucca Flat Underground Test Area (CAU 97) source, is scheduled for FY 2024, following the completion of the Corrective Action Decision Document/Corrective Action Plan. Inclusion of the Frenchman Flat Underground Test Area (CAU 98) source in the Area 5 RWMS CA is planned pending the completion of the CAU 98 Closure Report in FY 2017. Near-term R&D efforts will focus on continuing development of the PA, CA, and inventory models for the Area 3 and Area 5 RWMSs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Design and Validation Testing of TruckScan to Assay Large Sacks of Fukushima Radioactive Debris on a Truck

As a result of the March 2011 earthquake and resulting tsunami in Japan, there was a serious accident at the Fukushima Dai-ichi Nuclear Power Plant. This accident has contaminated soil and vegetation in a wide area around the plant. Decontamination projects over the last 4 years have resulted in large numbers of 1 cubic meter canvas bags of debris, commonly called Super Sacks [SS]. These are currently stored nearby where they were generated, but starting in 2015, they will be moved to various Interim Storage Facilities [ISF]. Trucks will typically carry 8-20 of these SSs. When the trucks arrive at the ISF they need to be rapidly sorted into groups according to radioactivity level, for efficient subsequent processing. Canberra Industries, Inc. [CI] has designed a new truck monitoring system 'TruckScan' for use at these ISFs. The TruckScan system must measure the entire truck loaded with multiple closely packed SSs, and generate a nuclide specific assay report showing the radioactivity in each individual SS. The Canberra-Japan office, along with Obayashi Corporation have performed validation testing to demonstrate to the regulatory authorities that the proposed technique was sufficiently accurate. These validation tests were conducted at a temporary storage area in Fukushima prefecture. Decontaminated waste of various representative types and of various levels of radioactivity was gathered and mixed to create homogeneous volumes. These volumes were sampled multiple times and assayed with laboratory HPGe detectors to determine the reference concentration of each pile. Multiple SSs were loaded from each pile. Some of the SSs were filled 50% full, others 75% full, and others 100% full, to represent the typical loading configuration of the existing SSs in the field. The content of the SSs are either sand, soil, or vegetation with densities ranging from 0.3 g/cc - 1.6 g/cc. These SSs with known concentrations of Cs-134 and Cs-137 were then loaded onto trucks in a variety of configurations, typical of how they might be on the real trucks. A partial system was installed at the site and used to assay these trucks with the various loading configurations. Whereas the full system will have 8 collimated 3 x 3'' NaI detectors, the test system only had two detectors; therefore the truck was moved and counted 4 times. The data were acquired and analyzed with the Canberra Genie software to determine the peak counts for both Cesium nuclides. That data was then analyzed with a prototype version of a Maximum Entropy algorithm, to determine the individual SS activity. The goal of the validation tests was to demonstrate that the system could detect 1000 Bq/kg in 15 seconds, and to determine how accurately it could quantify individual SSs. The validation tests demonstrated that the product would perform as predicted. The TruckScan results were consistent with the sample assay results [y = 1.0029 x, R{sup 2} = 0.914]. The Total Propagated Uncertainty, including both uncertainties from these tests and others that were estimated but not tested was 16.6% percent. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

U.S. Department of Energy/Argonne National Laboratory Training Courses for Packaging, Transportation, and Storage of Radioactive Material - 17269

The U.S. Department of Energy (DOE) Packaging Certification Program (PCP), Office of Packaging and Transportation, Office of Environmental Management, has sponsored a suite of training courses that have been conducted since the early 1990s. Of these training courses, Argonne National Laboratory (Argonne) conducts three courses: (1) Quality Assurance (QA) for Radioactive Material Packaging, (2) American Society for Mechanical Engineers (ASME) Pressure Vessel Code for Nuclear Transport and Storage, and (3) Nuclear and Other Radioactive Materials Transport Security. Argonne has been working with the University of Nevada, Reno, to develop an accredited graduate-level program in nuclear packaging. The Graduate Certificate in Nuclear Packaging (GCNP) program provides an applied knowledge-based curriculum in packaging for nuclear and radioactive materials that complements other graduate programs at the University (for additional information, see https://www.unr.edu/degrees/nuclear-packaging/certificate). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Use of Ductile Iron for Radioactive Waste Packaging and Shielding in the United States - 17278

Ductile Cast Iron (DCI) has been used for packaging radioactive waste in Europe and the United Kingdom for many years. The ductile iron provides many favorable properties when compared with steel and offers many advantages including: - Monolithic structure increasing overall strength; - No welding required; no tempering or maintenance/inspection of the weld area is required; - Fast, serial production of similar containers reducing manufacturing costs; - Lower cost compared to steel or steel/lead composites; - Good corrosion resistance even when uncoated; - Can be coated or plated to provide suitable surface for decontamination; - Flexibility in design due to the manufacturing process permits a variety of shapes and very large size container possibilities; - Hazardous materials not needed or used for shielding. The flexibility in design will help generators find viable and more cost effective packaging options for radioactive waste. The package is conformed to the waste instead of conforming waste to the package. Commercial utilities often replace radioactive components which are too large for conventional packaging and have dose rates too high for flexible wraps while plant schedule and physical configuration won't allow for component segmentation. This is one situation where the use of ductile iron can be used to provide a packaging solution. By assessing the radiological characterization and component size and weight, ductile iron can be cast into certified containers to easily package, ship and dispose of the waste in a cost effective manner. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Mitigation of Radioactive Waste By Sampling and Analyzing The Affected Area and Efficient Use of Different Detectors - 17400

The paper summarizes the way by which the mitigation of the radioactive waste during events like decommissioning of nuclear power plants, or during a nuclear disaster can be done so that the rehabilitation process for flora and fauna can begin in a natural manner. The paper describes the problems faced during ER operations in the developing countries and need of international committees in these areas. The paper suggests marking of the radioactive areas which can be used for controlled personnel access and which can be useful for monitoring purposes, while also focusing on the need of proper knowledge while handling detectors in ER operations for successful site characterization. It also highlights the importance of bio-indicators in these operations and also gives a description of various detectors used widely and their related significance for ER operations. The paper gives an overview of the planning and site characterization stages that follow in the environmental remediation. Planning stage should consider: 1. The impact on the ecosystem and human beings and the risk of spreading of the radionuclides. 2. The financial aid in hand for detectors, machinery and skilled labor. As the planning stages develop, the very first thing is the survey of the site. Site investigation should be carried out in all aspects including meteorology, geology, and hydrology, seismicity, geomorphology and the other aspects depending on site for a certain period of time. A site characterization should take: 1. The amount of damage occurred to the environment in the nearby region. 2. The distribution, character and the extent to which radioactivity has been dispersed. 3. The amount of risk that will be involved for personnel working in the field. 4. The precision of data collected that is to be used as input for further processes. In the developing countries, the central governmental bodies play a pivotal role in providing funds internally or from international bodies. The paper shows the effect how political issues effect the work progress of ER operations, that hamper the process and make people frustrated, who really work on the field. Paper provides the suggestion for the proper use of detectors in these countries where surveyors lack knowledge on what kind of detectors to be used while taking precision as a parameter. The table beside gives just a small picture of the large variety of detectors available today and tries to summarizes the main detectors with applications. There should also be a proper marking of the areas into different levels of radiation that will be helpful for work access control and monitoring personnel health. The paper also suggests of using bio-indicators in ER operation more often. The paper proposes to use site-specific bio-indicators which are very helpful in knowing the proper effect of the event in the environment. There is no single paved path for every situation that can lead to a successful ER. The planning should be done in such a way that the future work is not disturbed due to technical problems. In developing countries, political issues can hamper the progress of work and time remains a critical issue in such areas. The lack of proper knowledge for the use of detectors can affect the outcome of site characterization. Areas of the survey should be marked with different levels of activity for efficient monitoring of personnel's. The use of bio-indicators should be used more often and should be chosen as per the site. As they are a helpful tool in evaluating the work progress. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Treatment of radioactive liquid waste by the forced-air exhaust system - 15190

A Natural Evaporation Facility (NEF) was constructed to treat low level (LL) radioactive liquid waste arising from a Radioactive Waste Treatment Facility (RWTF) at the Korea Atomic Energy Research Institute (KAERI). In order to process LL radioactive liquid waste, we investigated the influence of climate condition, mainly atmospheric humidity and temperature, on the amount of evaporation in a forced-air exhaust system in KAERI. Evaporation media were made of cotton and polyester. Air exhaust in the facility was forced by exhaust fans. The evaporation rate and the decontamination factor were calculated and reported as shown in the results section. The evaporation and air supply flow increased at the same rate. In addition, the liquid waste feed rate and temperature of supplied air increased. Regarding the relative humidity of supplied air, the evaporation rate increased as the relative humidity decreased. As the result of this study, the operating conditions for the NEF were optimized in the following areas: The air temperature above 8 deg. C, the air relative humidity below 70%, the air flow rate 1.14-1.47 m/sec, and the liquid waste feed rate 4.6 l/hr.m{sup 2}. The decontamination factor was 1.1x10{sup 3}. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Vadose zone monitoring at the radioactive waste management sites at the Nevada national security site - 15294

Two lysimeter facilities continuously monitor the vadose zone at the radioactive waste disposal sites at the Nevada National Security Site. The time-series data collected at these facilities support the performance assessments of the radioactive waste disposal sites, the closure cover designs, and the long-term monitoring decisions for these sites. The Weighing Lysimeter Facility at the Area 5 Radioactive Waste Management Site (RWMS) consists of a bare-soil weighing lysimeter and a vegetated weighing lysimeter. The upper surfaces of both weighing lysimeters are exposed to precipitation. The Drainage Lysimeter Facility at the Area 3 RWMS includes bare-soil lysimeters that receive precipitation or precipitation supplemented with irrigation equal to double precipitation and vegetated lysimeters that have the same upper boundary conditions as the bare-soil lysimeters. Lysimeter sensors are connected to data-loggers that are accessed remotely for data acquisition. Lysimeter water balance measurements show the arid climate and native vegetation at the NNSS provide a natural sustainable system to effectively prevent transport of contaminants from the RWMSs through the vadose zone to groundwater. Transport by infiltration and percolation of present day precipitation at the RWMSs is prevented. The native desert vegetation is adapted to extract moisture from very dry soils and maintains negative matric potentials at the base of the root zone effectively buffering the deep vadose zone from episodic precipitation events at the surface. The very negative matric potential at the base of the root zone effectively intercepts all infiltration from precipitation and draws moisture upward from the vadose zone below the root zone. This system is sustainable because it does not rely on engineered features and mimics the stable landscapes and no groundwater recharge of the current condition surrounding the RWMSs. The system depends on establishing native vegetation in the covers above the waste cells and a stable climate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Radiometric characterization process for locating radioactivity hold-up and measuring non-uniformly distributed sources - 15342

A novel approach to facility radiometric characterization for non-uniformly distributed radiation sources has been developed by ANTECH and is currently being employed at the Argonne National Laboratory. It is being applied to the measurement of a wide range of objects and the interiors and contents of laboratories and buildings. The integrated non destructive assay measurement process involves the use of a scanning spectroscopic gamma ray camera as well as high purity Germanium detectors with different collimators and hence differing detector fields of view to measure and quantify non-uniformly distributed sources of radioactivity in laboratory equipment, glove boxes, storage tanks, pipework and general laboratory objects and process components. The purpose of the radiometric characterization is to quantify radiation sources and radioactive contamination levels in facilities containing nuclear material and to ensure compliance with facility characterization requirements and for decontamination and decommissioning planning and execution. The integrated measurement process also involves automated data capture and processing software and a database with which both measurement instrument operators and subject matter expert analysts can interact. An appropriate measuring instrument (gamma camera, scintillation detector, high purity Germanium detector system, neutron detector or alpha-beta counter) records the measurement data and the measurement site is recorded in a photograph. All of the data relevant to the measurement (or set of measurements for a particular object) are recorded in the database including measured gamma ray spectra, equipment description and relevant calibration, measurement site photograph and measurement descriptive data. Once data analysis protocols are established for specific geometries and measurement situations they can be applied to the measured data in a partially automated manner, reducing analysis and transcription errors. In addition to measurements and data analysis the process also utilises radiometric modelling computer codes including MicroShield and Monte Carlo Neutrons Photons in order to perform model calculations where appropriate. In a typical measurement, the ANTECH RadSearch spectroscopic gamma camera is utilized to find areas of concentrated activity so that these areas can be measured and calculated separately from areas with diffuse contamination. The use of the gamma camera in scanning mode helps to ensure that the assay does not overestimate the detected radiation source, which is a frequent consequence of using standard far field geometry and attenuation analysis codes that assume uniformity of both density and radioactive source distribution. Once the hot spots are identified, high purity Germanium detector measurements are made using various collimators with wide or narrow fields of view as required. These measurements are then processed through a semi-automatic calculation process assuming an appropriate source configuration as identified by RadSearch. These source configurations include point sources, line sources or surface sources. The Generalized Geometry Hold-up algorithms are employed to determine an estimate of the activity of the various sources that are identified in a measurement. The ANTECH integrated measurement process provides an automated calculation procedure where the analyst can select the source type, geometry and appropriate input data, and then the software calculates the activity by radionuclide for all sources identified in the measurement. The measurement and analysis process is controlled under the ANTECH ISO9001: 2008 Quality Assurance System and managed by a site specific Quality Assurance Plan. Both Argonne National Laboratory and Department of Energy technical reviewers have vetted the Integrated Measurement and Analysis Process. The application of the measurement and analysis process is described and measurement examples are provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Extending the lifetime of mixer paddles used in the production of a low-level radioactive cementitious waste form - 15624

The salt-stone Production Facility (SPF) at the Savannah River Site (SRS) immobilizes low-level radioactive salt solution as a cementitious waste form known as salt-stone. The mixer used in the SPF is a '10-inch' Readco-Kurimoto continuous mixer with paddles fabricated from a high toughness steel. salt-stone is processed daily at a rate of 9.46 E-03 m{sup 3}/s (150 gallons per minute) in a caustic and radioactive environment. The paddles have degraded much quicker than expected which limits production rates and increases outage times. Due to the high cost of the mixer and the limited maintenance due to the radioactive environment, a combination of material properties and paddle configuration must be employed to extend the lifetime of the paddles. This paper will discuss two wear tests conducted by the Savannah River National Laboratory (SRNL) to determine the optimum paddle configuration to reduce wear on the mixer paddles. The first wear test used the mixer design operated up to December 2011 as the baseline and the second test used the mixer design installed after December 2011 as its baseline. As a result of these tests, the SPF will be implementing a new paddle configuration in the mixer currently installed in the SPF process room. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of a Wide Range Segmented Gamma Scanner for Measuring Radioactive Wastes Arising in the Chinese Nuclear Industry - 16261

There is a requirement to measure a variety of challenging radioactive waste forms generated by the Chinese nuclear industry. These range from low activity waste to high activity waste with a high content of Cs-137 and waste in high density 200 litre drums. This paper describes the design, development and testing of an updated and automated Wide Range Segmented Gamma Scanner (WR-SGS) specifically configured for Chinese nuclear industry radioactive waste measuring requirements. The instrument retains the capabilities of the original SGS developed by Parker and Martin at the Los Alamos National Laboratory and now widely used for radioactive waste assay. It also incorporates additional features to improve measurement efficiency and reduce cost. These features include continuous helical scanning of the waste drum, and separate lifting pillars to lift the detector and the transmission source. In place of a transmission source shutter mechanism, the source is placed in a shielded safe when not in use. The design also incorporates a variable aperture collimator, and tungsten filters that can be deployed for the measurement of high activity, high dose-rate drums. Updated motion control electronics, based on the use of a programed logic controller have been incorporated to address potential future obsolescence. The WR-SGS employs a 40% high purity Germanium detector with electro-mechanical cooling and state of the art digital spectroscopy counting electronics for gamma ray spectrum data acquisition. Testing of the unit has been performed with both point sources and rod sources, which are traceable to international standards. Typical test drums, each with six distributed re-entrant tubes and with different density matrices, have been employed with the rod sources to validate both the calibration and the measurement performance of the WR-SGS for drums of different density. The measurement results have been compared with the reference decay corrected activity values and very good agreement is achieved, confirming the measurement performance of the instrument. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Roll Compaction of Powdered Radioactive Wastes - 18251

Most of low- and intermediate-level radioactive wastes must be solidified by a proper binding matrix and disposed at the centralized disposal site. So it is desirable to achieve a high loading of radwastes in a drum to reduce the disposal fee while meeting both the national regulations and the waste acceptance criteria of the disposal site. In general, the radioactive wastes (liquid, dried powder and sludge) were solidified by mixing with a proper binding matrix (cement, polymer, asphalt, etc.), but their solidified volume inevitably increases due to the limitation of incorporating of radwaste into a binding matrix. To solve this problem, some Korean NPPs operate both a CTS (Concentrate Treatment System, evaporation and granulation) and a polymer solidification system. The size distribution of granulated wastes produced by CTS is so broad (diameter; several μm ∼ several cm) that a polymer cannot impregnate the void space between particles due to their small size. However, if the powdered radwaste can be compacted into high density pellets, and the pellets can be loaded in a drum with maximum waste loading, then the binding matrix can fill the void between pellets, it might be more desirable to keep the size distribution of pellets constant to reduce the solidified waste volume. This compaction methodology can greatly reduce the disposal fee due to the reduction of the solidified waste volume, and can facilitate polymer solidification due to the constant size of the radwaste pellets. This methodology is also applicable to solidify powdered wastes such a depleted uranium used as a catalyst, ash and sludge. Loading radwaste compacted with bentonite pellets versus particles into a container serves, as an engineered or natural barrier in a deep geologic disposal system. Roll compaction is a typical agglomeration process for dry powders and depends on the device parameters and the material properties. This study, firstly, was focused on the design of roll compactor (briquetting machine), the shape and size of pellet, and secondly the loading method of pellets into a drum, their solidification in polymer, and evaluating the overall volume reduction. We used bentonite (particle size; 85 ∼ 100 μm) as a material to be compacted instead of radwaste powders (concentrates and depleted uranium wastes) without the addition of agglomeration agents. In the design of roll, we considered the diameter (200 mm) and width (41 mm) of roll, two types of pocket shapes (< 9 mm, 0.275 ml), and the surface of the pocket. Fig. 2 shows two pellet shapes, tetragonal and circular. The pressure of roll compactor, one of the main process variables, is > 320 kg f/cm{sup 2}, and the roll gap is 0.5 mm. The powders are compacted in a compaction zone between two counter-rotating rolls according to the complicated mechanisms. As a result of the state and the apparent density of the pellets obtained by changing the pressure from 0 to 286 kg f/cm{sup 2}, the pellets were unstable at pressure=0, but above 148 kg f/cm{sup 2} the pellets were more stable, and the apparent density of the pellets was 2.63 g/cm{sup 3} at 148 kg f/cm{sup 3}, and 2.68 g/cm{sup 3} at 286 kg f/cm{sup 2}. The roller speed (rpm) can determine the capacity (kg/h) of a compactor, and depends on both the compaction pressure and the feeding rate. By increasing the roll speed from 2 to 4 rpm, the apparent density of pellets was decreased gradually from 2.74 g/cm{sup 3} to 1.45 g/cm{sup 3}. We designed the special feeding system in order to feed powders uniformly to the roll gripping zone and to supply powders into the narrowest part of the roll gap. At a constant roll speed (3 rpm) and a constant compaction pressure (286 kg f/cm{sup 2}), the apparent density of the pellets was linearly proportional to the feeding rate from 2.18 to 2.74 g/cm{sup 3} by changing the feeding rate of powders from 8 to 12 rpm. In the polymer solidification of the obtained pellets, we used epoxy resin as a solidification agent, and showed their solidified waste forms in Fig. 5. The incorporation ratio (wt.%) of powders/polymer was about 71.7/28.3 in tetragonal pellets, 71.9/28.1 in circular pellets. We calculated the volume reduction in an aspect of powders itself and the solidified waste form. In powder case, through the comparison of the apparent density of the obtained pellets (2.74 g/cm{sup 3}) and raw powders (0.93 g/cm{sup 3}), the volume reduction was calculated as about 2.74/0.93 = 2.95 times. This means that powders can be compacted to 1/2.95 times. And in waste form case, the weight ratio of powder/polymer in a mixing method (workable) and a compaction method is about 0.42, 2.56, individually. Therefore the volume reduction was calculated as about 2.56/0.42 = 6.01 times. We found a compaction methodology for the volume reduction of powdered radioactive wastes, and their polymer solidification will be more effective to reduce the volume of radwastes to be disposed. After a characterization of the solidified waste form incorporated pellets (including sintered pellets), this methodology will be applied to treat depleted uranium wastes used as catalyst, and to solve the problem of NPPs' polymer solidification system, as mentioned above. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗