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 325 records · Page 18

Implementation of Treatment Systems for Low and Intermediate Radioactive Waste at Site Radwaste Treatment Facility (SRTF), PR China - 12556

The AP1000 reactors being built in the People's Republic of China require a waste treatment facility to process the low and intermediate radioactive waste produced by these nuclear power stations. Westinghouse Electric Germany GmbH was successful in being awarded a contract as to the planning, delivery and commissioning of such a waste treatment facility. The Site Radwaste Treatment Facility (SRTF) is a waste treatment facility that can meet the AP1000 requirements and it will become operational in the near future. The SRTF is situated at the location of Sanmen, People's Republic of China, next to one of the AP1000 and is an adherent building to the AP1000 comprising different waste treatment processes for radioactive spent filter cartridges, ion-exchange resins and radioactive liquid and solid waste. The final product of the SRTF-treatment is a 200 l drum with cemented waste or grouted waste packages for storage in a local storage facility. The systems used in the SRTF are developed for these special requirements, based on experience from similar systems in the German nuclear industry. The main waste treatment systems in the SRTF are: - Filter Cartridge Processing System (FCS); - HVAC-Filter and Solid Waste Treatment Systems (HVS); - Chemical Liquid Treatment Systems (CTS); - Spent Resin Processing Systems (RES); - Mobile Treatment System (MBS). (authors)

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↗

Radioactive Waste Disposal Strategies within the US Army Corps of Engineers FUSRAP - 18328

The U.S. Army Corps of Engineers' (USACE) highest priority is the protection of human health and the environment and public and worker safety during the cleanup, transportation, and disposal of Formerly Utilized Sites Remedial Action Program (FUSRAP) material. In addition to following all appropriate federal and state rules and regulations, the USACE has its own safety program that ensures that we are being protective. One aspect of the USACE FUSRAP is the radioactive waste disposal strategies that we must consider. The USACE will only dispose of FUSRAP material in a facility that is licenses/permitted to accept it. The estimated exposures to facility workers and members of the public from the permitted disposal of FUSRAP wastes is evaluated as part of this process. The USACE uses the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process to study and remediate FUSRAP sites. An evaluation is performed to decide whether it is best to manage the waste on-site or off-site and the requirements for disposal. The action selected must be protective of human health and the environment and must comply with regulations. When making the decision, there are several other factors taken into consideration including cost of the disposal and whether the public and state agencies support the action. Prior to determining disposal options, wastes must first be characterized. This involves an evaluation of both site history and analytical data. Determining the proper disposal requirements and subsequent disposal option involves: Characterizing the waste as stated above. This is important because there are several regulatory agencies which may oversee management of the waste. The US Nuclear Regulatory Commission, the US Environmental Protection Agency (US EPA), the US Department of Transportation, and state hazardous waste and/or radiation protection agencies may all have varying degrees of involvement. Identifying disposal requirements. Certain wastes must be disposed at licensed radioactive waste disposal facilities. Wastes containing US EPA regulated hazardous waste must be disposed at permitted hazardous waste disposal facilities. Certain wastes must be disposed at facilities licensed/permitted for both radioactive and hazardous waste. Some lower activity wastes may be accepted for disposal at other types of landfills. Considering acceptance criteria of the disposal facility. Radionuclides and/or hazardous constituents in the waste, as well as physical and chemical form of the waste, are evaluated to ensure they meet the acceptance criteria of the receiving facility's license and/or permit. If they don't, another disposal location must be used. Considering the accessibility and location of the disposal facility. Some facilities are able to accept shipments only by truck and others can accept material via railcar as well. The distance to a disposal facility impacts the risk of transportation accidents and efficient use of project resources such as the number of trucks required to perform the transportation. These factors are also important in evaluating whether material can be cost effectively shipped to a disposal facility. Evaluating the compliance status of the receiving facility. It is not assumed that just because a facility has a license or permit that it is acceptable for the USACE' waste. Prior to selecting a facility for disposal of FUSRAP waste, a site visit, audit and consultation with the US EPA is required to determine whether there are any known relevant violations at the facility. Facilities that have violations that negatively impact the USACE or US EPA approval of the facility are not used for disposal of FUSRAP waste. Once a facility is selected. The USACE procures contracts, or utilizes existing contracts, to facilitate the transportation and disposal action. Additionally, the USACE continues to consult with US EPA routinely to confirm the compliance status of the facility. Considering state acceptance of wastes to a facility. The USACE seeks to obtain written authorization from the facility state regulator indicating the action is compliant with existing licenses, permits, laws, and regulations. Each of these factors is considered by the USACE in selecting a FUSRAP disposal location. The goal is to provide safe, cost effective disposal which is protective of human health and the environment and which complies with all regulatory requirements. These steps all feed in to the disposal strategy consideration that USACE must go through for our FUSRAP waste streams. There are a number of challenges in the process to include In order to achieve the Congressional mandate of completing cleanup work under FUSRAP in a more cost effective, timely manner, while assuring protection of public health and safety. USACE must determine all disposal options that are available for the types of waste at each FUSRAP site. As a solution, completion of a disposal strategy document can help ensure this goal is met and will foster consistency across the FUSRAP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Legislative and Regulatory Histories Related to Greater-Than-Class-C Low-Level Radioactive Waste - 18602

The history surrounding the legislative and regulatory developments regarding Greater-Than-Class C Low Level Radioactive Waste (GTCC LLW) is both interesting and complex. On October 25, 1978, the U.S. Nuclear Regulatory Commission (NRC) published an Advanced Notice of Proposed Rulemaking pertaining to the development of regulations for the land disposal of LLW. This regulation was one of the first undertaken by NRC since its inception pursuant to the Energy Reorganization Act of 1974. Given the problematic challenges that many States were facing with releases of radioactivity from waste disposal facilities at the time, the technical bases supporting this rulemaking were both forward-thinking and comprehensive. The rule included a waste classification system that was tiered in four waste classes: Class A, B, C and GTCC LLW. On December 27, 1982, the NRC finalized the requirements for disposing of these classes of radioactive waste in Title 10 of the Code of Federal Regulations, Part 61 (10 CFR 61). The NRC established the waste classification system to protect both members of the public and individuals that may someday inadvertently intrude in the waste placed in the disposal facility after expiration of the institutional control period. This waste classification was developed to ensure that radiation doses to members of the public and an inadvertent intruder would not exceed 0.25 mSv y{sup -1} (25 mrem y{sup -1}) or 5.0 mSv y{sup -1} (500 mrem y{sup -1}), respectively. At the time the regulation was promulgated, GTCC LLW was considered generally not suitable for near surface disposal. It was recognized that disposal of GTCC LLW would require disposal technologies having greater confinement capacity or protection than 'normal' near-surface disposal. In 1980 Congress enacted the Low-Level Waste Policy Act of 1980, as amended in 1985 (LLWPAA of 1985) that assigned responsibility of disposing of GTCC LLW generated in the commercial sector or owned or generated by the federal government to the U.S. Department of Energy (DOE). At the time States were having difficulties in developing disposal facilities within the time constraints mandated in the LLWPA of 1980. Therefore, Congress believed that if DOE was assigned responsibility of developing disposal facilities for GTCC LLW, then the States would be more likely to develop new LLW disposal facilities. Congress also assigned the NRC with the responsibility to license the disposal of GTCC LLW generated or owned by the federal government or by the commercial sector. Lastly, Congress changed the definition of LLW that had been defined in the LLWPA of 1980 removing the 'transuranic' exclusionary language. Four years later, on May 25, 1989, the NRC promulgated its regulation needed to implement the mandates specified in the LLWPAA of 1985. The NRC revised 10 CFR 61.55 specifying that the NRC was responsible for licensing the disposal of GTCC LLW. However, it inadvertently did not revise the definition of LLW needed to remove the 'transuranic' exclusionary language in this rulemaking consistent with the LLWPAA of 1985. This seemingly inconsequential omission was not discovered until a Petition for Rulemaking was filed in Texas in July 2014. Moreover, this omission has effectively orphaned waste with certain transuranic radionuclides with concentrations exceeding 100 nanocuries per gram (100 nCi/g). In December 2015, the NRC Commissioners directed its staff to move forward with a rulemaking to resolve this issue and to perform a regulatory analysis to determine whether or not an Agreement State can license the disposal of GTCC LLW pursuant to Section 274 of the Atomic Energy Act of 1954. The paper will discuss past and current legislation, rulemaking, and regulatory developments pertaining to resolving the transuranic waste discrepancy and providing a pathway to disposition GTCC LLW. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

French Innovative Thermal Treatment Processes for The Management Radioactive Organic Liquid Waste - 19017

The nuclear industry as well as hospitals or some research laboratories produce Radioactive Organic Liquid Waste (ROLW) such as scintillation cocktails for medical imaging, labeled molecules for medical research or liquids from former activities. In order to be compatible with disposal facilities requirements, this waste must be inerted and stabilized in solid matrix. The route generally used is incineration in the Centraco facilities at the Marcoule site (Gard, France), ashes produced being cemented downstream. However, liquids whose chemical or radiological characteristics are not compatible with the incinerator operating rules (high level of {sup 14}C or halogen content, corrosive gas production etc.) cannot follow this path. The purpose of the MILOR project is then to propose alternative paths to treat all kinds of ROLW. Initiated in September 2017, this project has the technical objective of developing, over a four-year period, two pilot-scale mineralization processes. Two variants of complementary plasma technologies are evaluated: aerial plasma with the 'IDOHL' process and submerged plasma with the 'ELIPSE' process. Compared with other methods, thermal plasmas have unique characteristics such as rapid decomposition with high throughput, fast startup and shutdown, and high-energy delivery. The process called IDOHL consists of a process involving an induction plasma in which the liquids are directly introduced. It has been designed for organic liquids without mineral load and for treatment rates less than 0.5 l/h. This type of plasma allows the destruction of liquid organohalogens without any prior evaporation or treatment. Specific gas treatment systems have been developed to direct the gas produced to a dry or wet gas treatment system. One aim of the project is to subsequently install the process in a radioactive zone at Saclay DRF. The second called ELIPSE consists of a breakthrough process involving a submerged plasma at the heart of which organic liquids are also introduced. Current research has demonstrated that such a technique may enable the instantaneous and complete destruction of liquids with a wide variety of constituents, such as chlorine, fluorine, or phosphorus. The submersion solution offers many advantages: quenching and cleaning of combustion gases; filtering of the particles they contain; and cooling maintained for the entire process, which guarantees excellent corrosion control. An advantage of this type of design is that the gas treatment system can be reduced to a demister-condenser followed by a simple safety filter, thereby offering the additional advantage of an extremely compact treatment system. This design also allows the ELIPSE process to become by this way an embeddable process if required. The MILOR project, financed by the French State through investment program is a tripartite project bringing together Andra, in charge of radioactive waste management, the CEA in charge of leading scientific and technological researches to propose effective ways of treatment and the company A3i whose role is to ensure the subsequent industrial development. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Treatment of Foreign Radioactive Waste at the BELGOPROCESS Plant - 19345

Belgium has used a large number of nuclear technologies during the last 60 years. Power generation, reprocessing, fuel fabrication, isotope production and research activities have produced a large variety of radioactive waste streams. BELGOPROCESS is the operator of the centralised waste treatment facilities located in the Campine region in Flanders. Over the years BELGOPROCESS has developed the know-how and facilities to treat each one of these waste streams. BELGOPROCESS is a subsidiary of NIRAS, the governmental institution responsible for the safe management of radioactive waste and enriched fissile materials in Belgium. A number of these waste streams are only produced during a short period of time. After processing the existing stockpile the treatment facility is put into operational standby and eventually prepared for decommissioning. In order to treat the stockpile of some types of legacy waste within an acceptable time frame, the treatment facilities needed to be designed at a certain treatment capacity. Since the majority of the legacy waste is processed nowadays, a significant amount of treatment capacity becomes available. In June 2006 the Belgian government decided that this know-how and facilities should become available to foreign waste producers. This would allow BELGOPROCESS to fill up the available capacity of its facilities and consequently lower the cost for the Belgian waste producers. Nowadays two facilities are open to accept foreign radioactive waste. The CILVA-facility can be used for sorting, cutting, incineration, precompaction, supercompaction and cementation of low level solid waste. The PAMELA-facility - a former vitrification plant - can be used for the treatment of medium level and alpha-bearing waste. In this paper we will discuss the technical characteristics of the CILVA-incinerator and a showcase for the treatment of foreign waste at our facilities. In 2016 BELGOPROCESS started a cooperation with MATE-FIN for the incineration of low level burnable waste from the Cernavoda NPP in Romania. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of semivolatile organics in liquid radioactive residue using sorptive stir bar and solvent back-extraction

Current methods for semivolatiles analysis in radioactive samples can produce large volumes of radioactive solvent residue. A method utilizing stir-bar sorptive extraction has been explored in this work for its applicability to radioactive waste samples. This low solvent analytical method may accelerate remediation, minimize hazardous solvent waste, and reduce exposure risk to workers. Organic compounds (polyaromatic hydrocarbons, chlorinated aromatics, and phenolic compounds) were chosen as surrogates for common Liquid Waste System (LWS) contaminants at Savannah River Site (Aiken, SC). Stir-bar extraction parameters (extraction time, matrix modification, and effective pH range) and solvent back extraction parameters (solvent type, volume, and extraction time) were optimized experimentally for the chosen compounds. Affinity of the stir-bar extraction polymer to radionuclides Cs-137 and Am-241 was observed to determine radionuclide concentration effects. The stir-bar method achieved mean recovery of 100 ± 0.7% (1σ), relative to 114 ± 7% using solvent extraction, while reducing weekly method hands-on time by 93.4% and solvent volume consumption by 99.3%. Sensitivity was improved by 378% in simulated tank waste and 278% in real-world LWS matrix, relative to solvent extraction. This work has produced a safe and optimized method for the low solvent analysis of organics in legacy radioactive tank waste by stir-bar sorptive extraction.

GC-MS↗

Correlation between Asian Dust and Specific Radioactivities of Fission Products Included in Airborne Samples in Tokushima, Shikoku Island, Japan, Due to the Fukushima Nuclear Accident

Radioactive fission product {sup 131}I released from the Fukushima Daiichi Nuclear Power Plants (FD-NPP) was first detected on March 23, 2011 in an airborne aerosol sample collected at Tokushima, Shikoku Island, located in western Japan. Two other radioactive fission products, {sup 134}Cs and {sup 137}Cs were also observed in a sample collected from April 2 to 4, 2011. The maximum specific radioactivities observed in this work were about 2.5 to 3.5 mBq×m{sup -3} in a airborne aerosol sample collected on April 6. During the course of the continuous monitoring, we also made our first observation of seasonal Asian Dust and those fission products associated with the FDNPP accident concurrently from May 2 to 5, 2011. We found that the specific radioactivities of {sup 134}Cs and {sup 137}Cs decreased drastically only during the period of Asian Dust. And also, it was found that this trend was very similar to the atmospheric elemental concentration (ng×m{sup -3}) variation of stable cesium ({sup 133}Cs) quantified by elemental analyses using our developed ICP-DRC-MS instrument.

54 ENVIRONMENTAL SCIENCES↗

Indirect methods in nuclear astrophysics with relativistic radioactive beams

Reactions with radioactive nuclear beams at relativistic energies have opened new doors to clarify the mechanisms of stellar evolution and cataclysmic events involving stars and during the big bang epoch. Numerous nuclear reactions of astrophysical interest cannot be assessed directly in laboratory experiments. Ironically, some of the information needed to describe such reactions, at extremely low energies (e.g., keVs), can only be studied on Earth by using relativistic collisions between heavy ions at GeV energies. In this contribution, we make a short review of experiments with relativistic radioactive beams and of the theoretical methods needed to understand the physics of stars, adding to the knowledge inferred from astronomical observations. We continue by introducing a more detailed description of how the use of relativistic radioactive beams can help to solve astrophysical puzzles and several successful experimental methods. State-of-the-art theories are discussed at some length with the purpose of helping us understand the experimental results reported. The review is not complete and we have focused most of it to traditional methods aiming at the determination of the equation of state of symmetric and asymmetric nuclear matter and the role of the symmetry energy. Whenever possible, under the limitations of our present understanding of experimental data and theory, we try to pinpoint the information still missing to further understand how stars evolve, explode, and how their internal structure might be. We try to convey the idea that in order to improve microscopic theories for many-body calculations, nuclear structure, nuclear reactions, and astrophysics, and in order to constrain and allow for convergence of our understanding of stars, we still need considerable improvements in terms of accuracy of experiments and the development of new and dedicated nuclear facilities to study relativistic reactions with radioactive beams.

79 ASTRONOMY AND ASTROPHYSICS↗