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LH2TS enables to calculate the wasted H2 due to transfer and boil-off while using LH2 as a fuel for transportation. The supply chain from the liquefaction plant to the station can be modeled using real gas equations of state from REFPROP and heat transfer models.
Highlights: Black-Right-Pointing-Pointer We calculated hazardous medical waste generation rates (HMWGR) from 132 hospitals. Black-Right-Pointing-Pointer Based on a 22-month study period, HMWGR were highly skewed to the right. Black-Right-Pointing-Pointer The HMWGR varied from 0.00124 to 0.718 kg bed{sup -1} d{sup -1}. Black-Right-Pointing-Pointer A positive correlation existed between the HMWGR and the number of hospital beds. Black-Right-Pointing-Pointer We used non-parametric statistics to compare rates among hospital categories. - Abstract: Goal of this work was to calculate the hazardous medical waste unit generation rates (HMWUGR), in kg bed{sup -1} d{sup -1}, using data from 132 health-care facilities in Greece. The calculations were based on the weights of the hazardous medical wastes that were regularly transferred to the sole medical waste incinerator in Athens over a 22-month period during years 2009 and 2010. The 132 health-care facilities were grouped into public and private ones, and, also, into seven sub-categories, namely: birth, cancer treatment, general, military, pediatric, psychiatric and university hospitals. Results showed that there is a large variability in the HMWUGR, even among hospitals of the same category. Average total HMWUGR varied from 0.012 kg bed{sup -1} d{sup -1}, for the public psychiatric hospitals, to up to 0.72 kg bed{sup -1} d{sup -1}, for the public university hospitals. Within the private hospitals, average HMWUGR ranged from 0.0012 kg bed{sup -1} d{sup -1}, for the psychiatric clinics, to up to 0.49 kg bed{sup -1} d{sup -1}, for the birth clinics. Based on non-parametric statistics, HMWUGR were statistically similar for the birth and general hospitals, in both the public and private sector. The private birth and general hospitals generated statistically more wastes compared to the corresponding public hospitals. The infectious/toxic and toxic medical wastes appear to be 10% and 50% of the total hazardous medical wastes generated by the public cancer treatment and university hospitals, respectively.
The Hanford Tank Waste Treatment and Immobilization Plant (WTP) Project is currently transitioning its emphasis from an engineering design and construction phase toward facility completion, start-up and commissioning. With this transition, the WTP Project has initiated more detailed assessments of the requirements that must be met during the actual processing of the Hanford Site tank waste. One particular area of interest is the waste qualification program. In general, the waste qualification program involves testing and analysis to demonstrate compliance with waste acceptance criteria, determine waste processability, and demonstrate laboratory-scale unit operations to support WTP operations. The testing and analysis are driven by data quality objectives (DQO) requirements necessary for meeting waste acceptance criteria for transfer of high-level wastes from the tank farms to the WTP, and for ensuring waste processability including proper glass formulations during processing within the WTP complex. Given the successful implementation of similar waste qualification efforts at the Savannah River Site (SRS) which were based on critical technical support and guidance from the Savannah River National Laboratory (SRNL), WTP requested subject matter experts (SMEs) from SRNL to support a technology exchange with respect to waste qualification programs in which a critical review of the WTP program could be initiated and lessons learned could be shared. The technology exchange was held on July 18-20, 2011 in Richland, Washington, and was the initial step in a multi-phased approach to support development and implementation of a successful waste qualification program at the WTP. The 3-day workshop was hosted by WTP with representatives from the Tank Operations Contractor (TOC) and SRNL in attendance as well as representatives from the US DOE Office of River Protection (ORP) and the Defense Nuclear Facility Safety Board (DNFSB) Site Representative office. The purpose of the workshop was to share lessons learned and provide a technology exchange to support development of a technically defensible waste qualification program. The objective of this report is to provide a review, from SRNL's perspective, of the WTP waste qualification program as presented during the workshop. In addition to SRNL's perspective on the general approach to the waste qualification program, more detailed insight into the specific unit operations presented by WTP during the workshop is provided. This report also provides a general overview of the SRS qualification program which serves as a basis for a comparison between the two programs. Recommendations regarding specific steps are made based on the review and SRNL's lessons learned from qualification of SRS low-activity waste (LAW) and high-level waste (HLW) to support maturation of the waste qualification program leading to WTP implementation.
Technical Area 54, at Los Alamos National Laboratory (LANL) is situated in the east-central portion of the Laboratory on the Mesita del Buey between Pajarito Canyon to the south and Cañada del Buey to the north. TA-54 includes four MDAs designated as G, H, J, and L; a waste characterization, container storage, and transfer facility; active TRU waste and MLLW waste storage and low-level waste (LLW) disposal operations at Area G; active hazardous and mixed low-level (MLLW) waste storage operations at Area L; and administrative and support areas. MDA J has previously under-gone closure. Area G is a waste management and disposal area, used for the disposal and storage of radioactive wastes since 1957. Since August 2015, Area G has been in warm standby and provides minimal operations to support safety, compliance, and nitrate salt remediation. Located within Area G, MDA G covers 63-acres. MDA G contains 334 active and inactive waste management units, which include 36 pits, 294 shafts, and 4 trenches. In 1971, Area G began use for the retrievable storage of TRU waste. There are two pits, four trenches and 60 shafts that contain retrievable TRU waste. Thirty-three of the shafts contain TRU waste that may present unique problems for retrieval. In 1986, segregation of MLLW was initiated at Area G for treatment and temporary storage or for off-site disposal. Area G is the only active LLW disposal facility at the Laboratory. Current operations at Area G include storage and characterization of TRU and mixed TRU waste destined for off-site disposal at the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico and the storage of MLLW destined for off-site treatment and/or disposal. Several above-ground container storage units (CSUs) are currently used for storage of containerized MLLW and/or mixed TRU wastes. These consist of asphalt pads and associated fabric domes or other structures. As defined by the Consent Order, MDA G contains 229 of the 334 subsurface waste management units at Area G. These MDA G disposal units include 32 pits, 193 shafts, and 4 trenches and contain LLW, MLLW and TRU waste. The remaining 105 solid waste management units (SWMUs) include RCRA-regulated landfill and storage units and DOE-regulated LLW disposal units. The TA-54 closure project must ensure that continuing waste operations at Area G and their transition to an interim or enduring facility are coordinated with closure activities.
To assure the engineered barrier of a high-level radioactive waste repository functions correctly, long-term monitoring inside the backfilled deposition holes and tunnels is planned. Cables are used to power monitoring equipment and send signals however complete isolation will not be possible. Previously developed monitoring systems send information wirelessly and used batteries. Due to the limited battery life, these systems cannot be used for long-term monitoring. In this paper a wireless power transfer method based on magnetic resonant coupling is presented. The necessary transmitter and receiver coils have been developed. The performance of the individual coils has been tested. The performance of the system of transmitter and receiver with both air and bentonite as medium was tested. The results showed that the system with bentonite performed similarly well as the system with air. Wireless power transfer based on magnetic resonant coupling is therefore a feasible method for high-level radioactive waste repository monitoring applications which could in principle work for a period of 100 year or longer. (authors)
In this memorandum, the chemical and radionuclide contaminant results from the Fourth Quarter Calendar Year 2015 (CY15) sample of Tank 50H salt solution are presented in tabulated form. The Fourth Quarter CY15 Tank 50H samples were obtained on October 29, 2015 and received at Savannah River National Laboratory (SRNL) on October 30, 2015. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Fourth Quarter Calendar Year 2015 (CY15) sampling of Tank 50H were requested by SRR personnel and details of the testing are presented in the SRNL Task Technical and Quality Assurance Plan.
In this memorandum, the chemical and radionuclide contaminant results from the Third Quarter Calendar Year 2016 (CY16) sample of Tank 50H salt solution are presented in tabulated form. The Third Quarter CY16 Tank 50H samples (a 200 mL sample obtained 6” below the surface (HTF-5-16-63) and a 1 L sample obtained 66” from the tank bottom (HTF-50-16-64)) were obtained on July 14, 2016 and received at Savannah River National Laboratory (SRNL) on the same day. Prior to obtaining the samples from Tank 50H, a single pump was run at least 4.4 hours, and the samples were pulled immediately after pump shut down. The information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering for the transfer of aqueous waste from Tank 50H to the Saltstone Production Facility, where the waste will be treated and disposed of in the Saltstone Disposal Facility. This memorandum compares results, where applicable, to Saltstone Waste Acceptance Criteria (WAC) limits and targets. Data pertaining to the regulatory limits for Resource Conservation and Recovery Act (RCRA) metals will be documented at a later time per the Task Technical and Quality Assurance Plan (TTQAP) for the Tank 50H saltstone task. The chemical and radionuclide contaminant results from the characterization of the Third Quarter CY16 sampling of Tank 50H were requested by Savannah River Remediation (SRR) personnel and details of the testing are presented in the SRNL TTQAP.
SIAL{sup R} is a proven, safe, cost-effective and licensed waste solidification technology that has been used successfully for 20 years in Slovakia and the Czech Republic for processing of various problematic radioactive waste streams. This abstract presents and summarises the most interesting performance records of the SIAL{sup R} geopolymer solidification technology, the applied mobile devices as well as the characteristics of final products. Radioactive waste solidification into matrix SIAL{sup R} is based on polycondensation of inorganic aluminosilicate components with chemical composition similar to average cement. The main components of matrix SIAL{sup R} are natural inorganic raw materials containing SiO{sub 2}, Al{sub 2}O{sub 3}, CaO, MgO, Na{sub 2}O, K{sub 2}O, Fe{sub 2}O{sub 3}, and TiO{sub 2}. The treatment/solidification process is modified for individual radioactive streams with different compositions. To determine the most appropriate composition, the SIAL{sup R} application technology is laboratory tested on real samples of each individual waste stream identified for solidification. Then an appropriate mixture of aluminosilicate composition and other inorganic compounds is defined for a given solidification application. When determining the most appropriate mixture, other requirements are taken into consideration: maximum dose rate on the drum with solidified product, mass, volume, alpha radionuclide content in the final product, need for a waste's dewatering before solidification, dry matter content in the solidification product as well as other specific waste acceptance criteria (i.e., compressive strength, leachability, diffusion coefficients, radiation stability, biodegradability of matrix, distribution and partition coefficients, influence of frost and water presence to the product stability including long term physical stability in water, dustiness, flammability, time dependence of compression strength and water reduction, product homogeneity and also radiolytic G-value for hydrogen release, GH{sub 2}). The Amec Foster Wheeler baseline approach to implementing the SIAL{sup R} waste treatment process is structured on three steps - waste characterization, waste retrieval and pre-treatment and at last the solidification of pre-treated waste itself. The equipment used for the solidification process is relatively small, lightweight and designed for the specific application on site, directly 'in situ' or local to the tanks storing the sludge's/resins/other waste streams. This mitigates the problems normally experienced by transferring sticky, tight waste through long pipes. The equipment is normally 'tailor-made' to take into account the real space conditions and other existing facility requirements. At Bohunice and Dukovany sites, the facilities for encapsulation were designed to occupy existing areas of the plants in close proximity to the tanks where the waste was being stored. This allowed for the removal of the waste using Remote Operated Vehicles and for the waste to be immediately treated on site without the need for transportation thus saving both, time and resource. To date, SIAL{sup R} has been used successfully to immobilize approximately 2 x 10{sup 6} kg of waste which includes sludge, resin, crystalline borates and contaminated organic waste from Bohunice and Mochovce, Slovakia and approximately 300 m{sup 3} of spent ion exchange resins and sludge from tanks on site at the Dukovany NPP (Nuclear Power Plant) in the Czech Republic. The technology now has a track record of over 20 years which includes on-going research and development. The International missions WANO (the World Association of Nuclear Operators) and OSART (the Operational Safety Review Team) evaluated the SIAL{sup R} matrix technology at Dukovany NPP as an example of good practice. (authors)
Within the U.S. Department of Energy’s (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in “tank farms”). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are “first-of-a-kind” and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE’s specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford’s Waste Treatment and Immobilization Plant (WTP) or Savannah River’s Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration.
Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R&D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the salt and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, while treatment of phosphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.
Savannah River Remediation (SRR) is the current Liquid Waste contractor at the U.S. Department of Energy's (DOE) Savannah River Site (SRS), located near Aiken, South Carolina, and operates the Defense Waste Processing Facility (DWPF), along with the other SRS liquid waste facilities. The DWPF, the nation's only operating vitrification facility, completed 20 years of radioactive operations in 2016. During this period, DWPF has safely treated the high-level liquid waste at this DOE facility since radioactive operations began with the first transfer of sludge waste feed to DWPF on March 12, 1996. A little over a month later, on April 29, 1996, the first radioactive canister was poured. DWPF poured its 4,000. canister of vitrified waste on December 31, 2015. As of September 30, 2016, DWPF has removed approximately 59.9 million curies from the liquid waste system at SRS and placed them in 4,106 canisters. SRS has operationally closed 8 high-level waste tanks. The DWPF melter has poured nearly 7.14 million kg of molten glass, and the facility has used only two melters in its lifetime. Melter 1 poured 1,339 canisters (1996-2002), while melter 2 to date (September 30, 2016) has poured 2,767 canisters (2003-present). Melter 2 continues to safely and efficiently operate. In order to achieve these accomplishments, the facility has chronicled a history of continued operational improvements such as the installation of bubblers in the melter to increase production, incorporation of glass compositions to achieve higher waste loadings, and improvements to melter design to extend melter life. As SRR moves forward to the next 20 years, continued improvements as well as maintaining the aging infrastructure through a proactive system health program are key priorities. This paper chronicles 20 years of challenges, improvements and accomplishments, and discusses future operations and infrastructure upgrades including implementation of a key flowsheet change to significantly reduce hydrogen generation in the Chemical Process Cell. (authors)
Savannah River Mission Completion is the Liquid Waste (LW) contractor at the Savannah River Site (SRS). The LW mission is tasked with treating and disposing of legacy nuclear waste. There are multiple facilities involved in this work, including the Concentration, Storage, and Transfer Facilities (CSTF), the Defense Waste Processing Facility (DWPF), the Salt Waste Processing Facility (SWPF), and the Saltstone Production Facility (SPF). The CSTF includes 43 underground waste tanks used to store and support processing of radioactive liquid waste. Waste removal activities, such as salt dissolution campaigns and sludge agitation, are conducted within the CSTF waste tanks to convert the waste into a form that allows for downstream processing at other LW facilities. While performing these waste removal campaigns, camera inspections are performed to assess the quantity and distribution of the remaining waste within the waste tank (i.e. saltcake or sludge). Understanding the quantity and distribution of the salt/sludge within the waste tanks allows for improved waste removal strategies (e.g. mixing pump operation) and refined safety controls. Typically, several camera inspections are performed during a waste removal transfer to verify the elevation of the visible salt/sludge mounds against the known elevation of the liquid surface. The camera inspection footage must then be interpreted by a trained engineer who will develop a 2-D map that depicts the waste distribution at various elevations within the waste tank. This tank mapping is then used in conjunction with conservative assumptions to evaluate the volume of saltcake or sludge that is present within the waste tank.
The Resource Conservation and Recovery Act (RCRA) was passed to set up a framework for the proper management of waste. The 40CFR260-273 regulation established the regulatory framework to handle solid and hazardous waste, from the point of generation to its disposal; or cradle-to-grave management. Per these regulations, a hazardous waste generator is responsible for its waste from initial generation through its ultimate disposal and beyond. Nothing will relinquish a generator of this responsibility. There is no expiration date or time limit, and hiring a someone else to transport and dispose of waste does not transfer responsibility. If an incident occurs at any point during the life cycle of a waste, the generator can be held liable. The first step to compliance with RCRA is proper characterization of the waste. Waste may either be listed specifically or may show a characteristic as defined in the regulation: Is my waste hazardous? Why is it Hazardous? What should I do as a generator? Plan chemical use: - Create a chemical inventory for your research project. - Use existing chemicals if available instead of ordering new ones. - Brainstorm methods that will help to reduce toxicity and minimize waste. Waste Inventory: - Label all waste bottles with: ID no., What is it, Whose is it, Hazards, and Date. - Add waste containers into the inventory log. - Initiate/complete the container sheet to describe the waste. Satellite Areas: - Store waste in containers compatible with the waste with lids on tight. - Store waste containers properly and segregate them according to their specific hazards. - Use secondary containment for liquids. Store acids and bases separately; apart from cyanides or sulfides. Acids should never be put into steel containers. Water-reactive, strong acids such as organic acid halides, organic acid anhydrides, inorganic acid anhydrides, and strong acidic salts must be kept apart from both alkalis and water. Oxidizing agents must be kept apart from reducing agents and organic compounds. Water-reactive agents must be stored apart from water, aqueous solutions, and acids. Air-reactive materials must be packed in containers that are sealed off from the atmosphere. Explosive and shock-sensitive materials present risks that require special handling.
At Hanford, an extensive network of pipelines traversing several miles is used to transfer the high level radioactive nuclear waste from tanks to the treatment facilities. During transfer operations, however, there is a potential risk for the transport lines to plug, causing significant delays, increasing operation costs and creating hazardous conditions for personnel and the environment. In this study we investigate the phenomena of plug formation in the pipelines due to solids depositing during the waste transfer process. A three-dimensional (3-D) Computational Fluid Dynamics (CFD) model has been developed using Comsol Multiphysics 4.3b by Florida International University (FIU) to evaluate the factors that can contribute to pipeline plug formation. Several virtual scenarios are developed simulating the settling mechanics as a function of particle size, solids volume fraction and solids density. The numerical results are validated with the available experimental results and the established critical velocity correlations. The 3-D numerical results were also compared with the previous 2-D numerical studies to understand the trade-off between the two studies in terms of computing speed and numerical accuracy. (authors)
Models have been developed to simulate the thermal characteristics of crystalline silicotitanate ion exchange media fully loaded with radioactive cesium either in a column configuration or distributed within a waste storage tank. This work was conducted to support the design and operation of a waste treatment process focused on treating dissolved, high-sodium salt waste solutions for the removal of specific radionuclides. The ion exchange column will be installed inside a high level waste storage tank at the Savannah River Site. After cesium loading, the ion exchange media may be transferred to the waste tank floor for interim storage. Models were used to predict temperature profiles in these areas of the system where the cesium-loaded media is expected to lead to localized regions of elevated temperature due to radiolytic decay. Normal operating conditions and accident scenarios (including loss of solution flow, inadvertent drainage, and loss of active cooling) were evaluated for the ion exchange column using bounding conditions to establish the design safety basis. The modeling results demonstrate that the baseline design using one central and four outer cooling tubes provides a highly efficient cooling mechanism for reducing the maximum column temperature. In-tank modeling results revealed that an idealized hemispherical mound shape leads to the highest tank floor temperatures. In contrast, even large volumes of CST distributed in a flat layer with a cylindrical shape do not result in significant floor heating.
According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license, granted by the competent authority in this field, which has been the Federal Office of Radiation Protection (BfS) until July 2016 and currently is the Federal Office for the Safety of Nuclear Waste Management (BfE). A first license was granted in 1983 for storage of spent nuclear fuel (SNF) and vitrified high level waste (HAW) at the interim storage facility in Gorleben. This license was reviewed in 1995. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since that time at each site of an operational NPP storage facilities were licensed and erected. Thus additionally to the three central storage facilities (Ahaus, Gorleben and Rubenow) there are now 12 storage facilities onsite the NPPs. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2034. The German 'Programme for the safe and responsible management of spent fuel and radioactive waste' expects to take the disposal facility into operation in the year 2050. Thus necessary preparations for interim storage till that time have to be taken on the side of the regulator. In the German Atomic Energy Act precautions have already been taken and a consulting of the German parliament (Bundestag) is mandatory if inevitable events would require an extension of currently licensed interim storage period. In the year 2016, the regulatory body in Germany was given a new structure. The BfS was formerly competent authority in the field of radiation protection since 1989. It had competencies for ionizing and non-ionizing radiation protection (e.g. mobile phone radiation, high voltage power lines and the effects of electric fields), licensing storage of spent nuclear fuel and as the operator of facilities for final disposal. Now there is a new structure with two Federal Offices (BfS and BfE) and a federal company for final disposal (BGE). The BfE is responsible for regulation in the field of nuclear disposal since 30.07.2016. So the package design approval for DPCs and the storage licenses for SNF are now granted by this office. The search for a repository site will be coordinated there as well. The site selection, construction and operation of the repository will be in the responsibility of BGE. The parts of the former BfS dealing with recovery of the waste from Asse (a former salt mine used for disposal research) and the construction of the repository Konrad (for low and medium level radioactive waste) will be transferred into that company. It is planned to incorporate the German company for construction and management of waste disposal facilities (DBE) which is owned by the German utilities as well. The BfS will focus on the scientific foundations and further advancement of radiation protection. (authors)
When processing High Level Waste (HLW) glass, the Defense Waste Processing Facility (DWPF) cannot wait until the melt or waste glass has been made to assess its acceptability, since by then no further changes to the glass composition and acceptability are possible. Therefore, the acceptability decision is made on the upstream feed stream, rather than on the downstream melt or glass product. This strategy is known as 'feed forward statistical process control.' The DWPF depends on chemical analysis of the feed streams from the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME) where the frit plus adjusted sludge from the SRAT are mixed. The SME is the last vessel in which any chemical adjustments or frit additions can be made. Once the analyses of the SME product are deemed acceptable, the SME product is transferred to the Melter Feed Tank (MFT) and onto the melter. The SRAT and SME analyses have been analyzed by the DWPF laboratory using a 'Cold Chemical' method but this dissolution did not adequately dissolve all the elemental components. A new dissolution method which fuses the SRAT or SME product with cesium nitrate (CsNO{sub 3}), germanium (IV) oxide (GeO{sub 2}) and cesium carbonate (Cs{sub 2}CO{sub 3}) into a cesium germanate glass at 1050 C in platinum crucibles has been developed. Once the germanium glass is formed in that fusion, it is readily dissolved by concentrated nitric acid (about 1M) to solubilize all the elements in the SRAT and/or SME product for elemental analysis. When the chemical analyses are completed the acidic cesium-germanate solution is transferred from the DWPF analytic laboratory to the Recycle Collection Tank (RCT) where the pH is increased to {approx}12 M to be released back to the tank farm and the 2H evaporator. Therefore, about 2.5 kg/yr of GeO{sub 2}/year will be diluted into 1.4 million gallons of recycle. This 2.5 kg/yr of GeO{sub 2} may increase to 4 kg/yr when improvements are implemented to attain an annual canister production goal of 400 canisters. Since no Waste Acceptance Criteria (WAC) exists for germanium in the Tank Farm, the Effluent Treatment Project, or the Saltstone Production Facility, DWPF has requested an evaluation of the fate of the germanium in the caustic environment of the RCT, the 2H evaporator, and the tank farm. This report evaluates the effect of the addition of germanium to the tank farm based on: (1) the large dilution of Ge in the RCT and tank farm; (2) the solubility of germanium in caustic solutions (pH 12-13); (3) the potential of germanium to precipitate as germanium sodalites in the 2H Evaporator; and (4) the potential of germanium compounds to precipitate in the evaporator feed tank. This study concludes that the impacts of transferring up to 4 kg/yr germanium to the RCT (and subsequently the 2H evaporator feed tank and the 2H evaporator) results in <2 ppm per year (1.834 mg/L) which is the maximum instantaneous concentration expected from DWPF. This concentration is insignificant as most sodium germanates are soluble at the high pH of the feed tank and evaporator solutions. Even if sodium aluminosilicates form in the 2H evaporator, the Ge will likely substitute for some small amount of the Si in these structures and will be insignificant. It is recommended that the DWPF continue with their strategy to add germanium as a laboratory chemical to Attachment 8.2 of the DWPF Waste Compliance Plan (WCP).
Los Alamos National Laboratory (LANL) Technical Area 55 (TA 55) utilizes several different container types to store and transfer special nuclear material and waste for numerous programs. The Nuclear Process Infrastructure - Infrastructure Operations Group (NPI-2) analyzes and certifies containers to meet the specifications for distinct processes and storage. Engineering analysis performed on these containers include but are not limited to: leak testing, filter efficiency testing, pressure drop test, drop testing, water penetration testing, polymer (o-ring) hardness testing, and thermal steady state testing. Of all of these tests, thermal steady-state testing is most crucial to ensure critical components for specific containers do not fail under an internal heat payload. This testing provides data for setting heat payload limits for different containers. The Pipe Over-pack Container (POC) is a vented carbon steel container with a removable lid designed for storing or transferring nuclear waste. Additionally, TA 55 has been tasked to manage and store all of its Transuranic (TRU) waste for the foreseeable future. Being able to place more material into a single container conserves the physical storage space available at TA 55. The Pipe Over pack Handling and Operations Manual requires each user to ensure that the maximum temperature of each component in the POC is not exceeded. Conducting a thermal steady-state test with a combustible waste matrix on the POC can benefit stakeholders on increasing the material limits of the container. The results of these measurements will establish new wattage limits to meet LANL's transportation requirements. The current LANL Transportation Safety Document (TSD) limits the amount of heat source plutonium to 10 grams per POC, corresponding to approximately 5 Watts of heat. The result of this test was provided to the Packaging and Transportation (OS-PT) group to evaluate new wattage limits that can satisfy the needs of TA 55, while maintaining a safe packaging configuration. (authors)