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The purpose of this Packaging and Transportation Safety Functional Area (PTSFA) procedure is to describe the process in which the PTSFA Quality Assurance Engineer (QAE) or designee reviews and approves the documentation associated with the procurements of containers to be used for onsite transfer or off-site transportation of hazardous materials and waste.
For over a decade, the Town of Bedford's Energy Commission and the Department of Public Works has explored the viability of ground mounted solar development on town property.
This paper presents the design approaches and test facility operation methods used to successfully accomplish a 56-day test for Skylab to permit evaluation of selected Skylab medical experiments in a ground test simulation of the Skylab environment with an astronaut crew. The systems designed for this test include the two-gas environmental control system, the fire suppression and detection system, equipment transfer lock, ground support equipment, safety systems, potable water system, waste management system, lighting and power system, television monitoring, communications and recreation systems, and food freezer.
Sealed radioactive sources are utilized for a wide range of applications across nuclear facilities, universities, hospitals, and industry. When these sources reach the end of their serviceable life, they become waste. This radioactive waste then goes through a process of recapture and then transfer to long term storage. With the advancement of technology in conjunction with better accessibility of technology, industries are exploring the use of digital automation to enhance productivity, efficiency, and safety while minimizing operation and maintenance costs, health and environmental risks, and uncertainty in the project life cycles. One area for exploration and the use of Digital Twin for providing a robust, versatile, and safe solution for recapturing spent sources.
The well-known 80 - 20 principle, also known as the Pareto principle, describes the fact that usually 80 % of the result is achieved by 20 % of the effort, because many probability distributions are caused by a power-law. This principle is valued in the realm of project management and management in general because of the fast progress which generally takes place in the first fifth of the process. If the final result is not exactly specified or resources are spent wisely, 100 % of the planned result will be reached easily. However, in nuclear business safety is a principal parameter. Therefore precision is inevitable. In long-lasting projects the bad end of the Pareto-distribution is dominant: The very shallowly-inclined gradient at the end of the graph means that results at that point come at a very high effort and cost. The graph reaches the 100% only asymptomatically, meaning the initially projected results are never met. Especially if in earlier project phases the Pareto principle is applied very successfully, a lot of work is transferred to the future. For example, in decommissioning projects the waste often is badly characterized causing an enormous effort in the process of characterization and qualification. Especially if funding or time are limited, it is important to be aware of the asymmetric distribution of applied work and result. The 80 - 20 principle is a great advantage, if you start a project and respect the asymmetric distribution of applied resources and results in your project plan. Especially in long-lasting projects with changing perspectives, this asymmetric distribution may cause massive delays and expansion of costs, because small changes near the end cause large/big effort due to consequences on the existing data. If the proper goal of a project is not entirely known in the beginning, it is very probable that the last 5 - 10 % of the result will cause major financial threat and extensions of the estimated time frame. If funding or time are limited, it is almost impossible to readjust a project in the estimated frame of resources. In this case a complete restart of the project may be more reasonable than a long-lasting change management with uncertain results. (authors)
The Savannah River Site treats and disposes of legacy high-level waste, which resulted from 60 years of nuclear material production, medical research, and other nuclear related work. Complex mixtures of soluble and insoluble solids occupy tanks of which the South Carolina Department of Health and Environmental Control recognizes as the single largest environmental risk state-wide. The risk is reduced through removal, treatment, and solidification of decontaminated salt waste. Presently, SRS decontaminates the soluble salt component of the waste, a solution containing actinides and cesium, via implementation of the Actinide Removal Process and the Modular Caustic Side Solvent- Extraction Unit. Caustic solvent extraction is set to increase in scale by late 2018, with startup of the newly constructed Salt Waste Processing Facility. In conjunction with Salt Waste Processing Facility startup, Savannah River Remediation has been tasked by the Department of Energy to demonstrate modular ion exchange technology. The Tank Closure Cesium Removal project aims to demonstrate the feasibility of modular ion exchange processes for future Department of Energy applications. The Tank Closure Cesium Removal system is scheduled to begin operation in mid-2018. Dissolved salt solution, or influent, will be fed through a filter and ion exchange columns in lead-lag configuration, producing effluent. The effluent is fed to a separate high-level waste tank, for future disposition at Salt-stone Production Facility. Once the demonstration project is complete, the loaded ion exchange columns will be purged, dried, and placed in interim safe storage, awaiting final disposition. The ion exchange media selected for the Tank Closure Cesium Removal system is Honeywell UOP IONSIV{sup C} R-9120. The media is a prime candidate for separation of cesium from radioactive high-level waste. UOP R-9120{sup C} is inorganic and has high capacity, selectivity, and affinity for cesium. During operation, influent is fed to the ion exchange columns, contacting UOP R-9210{sup C} packed within the media vessel. Contact initiates loading of isotopes, and given ample time the solid phase (ion exchange media) and solution achieve thermodynamic equilibrium. UOP R-9120{sup C} is a prime candidate for separation of cesium and dissolved fission products. Although the properties of UOP R-9120{sup C} are favorable for separation, the same properties lead to accumulation of a heat source in the ion exchange bed. Cesium-137 and its daughter product, barium-137 m, generate a clear majority of heat (i.e. ∼ 94%). During operation salt solution flows through each column and transports heat from the ion exchange bed. Once operation is terminated, the effluent contained within the porous bed volume becomes essentially static. In the absence of engineering controls, the heat generated may boil the effluent. Rapid vaporization of aqueous salt solution would lead to bed over pressurization. Over pressurization and an associated release of high-level waste would result in adverse impacts to human health and the environment. This paper focuses on modeling the heat transfer associated with the Tank Closure Cesium Removal system ion exchange column. The model evaluates the maximum heat generation and the associated radionuclide loadings that prevent boiling within of the media vessel, under no flow conditions. A safety significant control, maintaining loadings less than the established loading limit ensures associated impacts to human health and the environment are prevented. (authors)
Highlights: Black-Right-Pointing-Pointer The goal is to compare collection costs for residual household waste. Black-Right-Pointing-Pointer We have clustered all municipalities in order to find mutual comparable pairs. Black-Right-Pointing-Pointer Each pair consists of one private and one public operating waste collection program. Black-Right-Pointing-Pointer All cases show that private service has lower costs than public service. Black-Right-Pointing-Pointer Municipalities were contacted to identify the deeper causes for the waste management program. - Abstract: The rising pressure in terms of cost efficiency on public services pushes governments to transfer part of those services to the private sector. A trend towards more privatizing can be noticed in the collection of municipal household waste. This paper reports the findings of a research project aiming to compare the cost between the service of private and public collection of residual household waste. Multiple case studies of municipalities about the Flemish region of Belgium were conducted. Data concerning the year 2009 were gathered through in-depth interviews in 2010. In total 12 municipalities were investigated, divided into three mutual comparable pairs with a weekly and three mutual comparable pairs with a fortnightly residual waste collection. The results give a rough indication that in all cases the cost of private service is lower than public service in the collection of household waste. Albeit that there is an interest in establishing whether there are differences in the costs and service levels between public and private waste collection services, there are clear difficulties in establishing comparisons that can be made without having to rely on a large number of assumptions and corrections. However, given the cost difference, it remains the responsibility of the municipalities to decide upon the service they offer their citizens, regardless the cost efficiency: public or private.
For the management of radioactive waste, either during nuclear facility operation or decommissioning, one must consider a diverse array of nuclear waste forms of various types, volumes and activities. Examples are: activated metallic waste; legacy waste; sludge; resins; and orphan waste. Customers are frequently faced with the obligation of undertaking multiple, costly waste management operations. These operations activities include handling, reconditioning or material transfers between packages, as in the case when moving material from on-site storage to transportation, and then from transportation to final disposal. To address this issue, a new - highly flexible up to reversible - cask system, the TN{sup R} MW, is being developed as an 'All in One Solution'. TN{sup R} MW has a total weight of 10 metric tons and is compliant with the 2012 International Atomic Energy Agency (IAEA) regulations. It was developed to provide flexibility and designed with the capability to be adaptable to the various needs of the nuclear industry. From IP-2 and Type A containers to Type B(U) and B(U)F casks, this cask offers different options to sufficiently address the requirements of each step of the waste stream from on-site and international transport to long-term interim storage and final disposal. Licensing and manufacturing of a number of packagings in this TN{sup R} MW family are underway. (authors)
The Savannah River National Laboratory (SRNL) has supported the Saltstone Production Facility (SPF) since its conception. However, bench scaled tests have not always provided process or performance data related to the mixing, transfer, and other operations utilized in the SPF. A need was identified to better understand the SPF processes and to have the capabilities at SRNL to simulate the SPF unit operations to support an active low-level radioactive waste (LLW) processing facility. At the SPF, the dry premix is weighed, mixed and transferred to the Readco '10-inch' continuous mixer where it is mixed with the LLW salt solution from the Salt Feed Tank (SFT) to produce fresh Saltstone slurry. The slurry is discharged from the mixer into a hopper. The hopper feeds the grout pump that transfers the slurry through at least 457.2 meters of piping and discharges it into the Saltstone Disposal Units (SDU) for permanent disposal. In conjunction with testing individual SPF processes over several years, SRNL has designed and fabricated a scaled Saltstone Facility. Scaling of the system is primarily based on the volume capacity of the mixer and maintaining the same shear rate and total shear at the wall of the transfer line. At present, SRNL is utilizing the modular capabilities of the scaled Saltstone Facility to investigate the erosion issues related to the augers and paddles inside the SPF mixer. Full implementation of the scaled Saltstone Facility is still ongoing, but it is proving to be a valuable resource for testing alternate Saltstone formulations, cleaning sequences, the effect of pumping Saltstone to farther SDU's, optimization of the SPF mixer, and other operational variables before they are implemented in the SPF. (authors)
Biochar has been reported as an effective additive to improve methane production during anaerobic digestion (AD). However, the mechanism for such a stimulatory impact remains unclear. Here we investigated the capability of three pyrolytic biochars with distinctive electrochemical properties (BC300, BC500, BC700) to promote methanogenic performance of anaerobic digesters treating waste activated sludge (WAS). The cumulative methane production and the maximum methane production rate were increased by 46.9% and 181.6%, respectively, with BC300 amendment. By characterizing the electrochemical properties of biochar, our research team found out that methane production in AD of sludge was positively correlated to biochar’s electron-donating capacity (EDC) rather than its bulk electrical conductivity. These results indicate that the electron transfer mediated by the redox-active functional groups may prevail over the direct electron migration as the predominant mechanism to facilitate interspecies electron transfer in syntrophic communities for enhanced methanogenesis. Microbial community analysis suggests that biochar enriched Methanosarcina and Methanobacterium. Overall, this study shows that functional groups-mediated electron transfer contributes greatly to the improved methane production in WAS digester with biochar amendment.
Savannah River Remediation (SRR) manages and operates the liquid waste facilities at Savannah River Site (SRS) for the Department of Energy (DOE). Stored liquid waste is a complex mixture of insoluble solids (sludge) and soluble salts in an alkaline solution. SRR has deployed the Tank Closure Cesium Removal (TCCR) system, a tank-side ion exchange process, to remove radioactive cesium from salt waste and enable onsite disposal of the resulting decontaminated salt solution as low-level waste at the Saltstone facilities. The TCCR system consists of two prefilters, four ion exchange (IX) columns, one resin trap, and a ventilation system. The IX process uses a form of inorganic crystalline silicotitanate (CST), which has a high affinity for cesium and other alkali metals, strontium, and actinides. This process is currently deployed utilizing salt feed from Tank 10, with future plans to dissolve solid salt in Tank 9 and transfer the salt solution to Tank 10 for processing through TCCR. The feed for TCCR must be created from salt-cake in Tank 10 through a dissolution process. Once enough salt has been dissolved, a qualification process is entered. This process characterizes the feed and ensures the cesium loading on the columns will not cause boiling of waste within the columns during or after processing. Once the batch has been qualified, salt waste is fed to the TCCR system through a transfer pump in the center of the tank. The waste is filtered through a set of two shielded, dead-end prefilters that prevent solids buildup in the columns. The filtered salt solution then travels to the shielded IX columns, which can be operated individually or in series, where the cesium is sorbed on the CST media. The decontaminated salt solution (DSS) then travels through a resin trap and out of the module to Tank 11. TCCR has successfully processed approximately 795,000 L of Tank 10H radioactive salt waste over two batches to date. There has not yet been a system induced shutdown. The prefilters performed as expected with only minor degradation in recovery of differential pressure after a backflush sequence. The time between backflushes decreased as each batch reached the end of processing. The hydraulics in the IXCs mostly performed as expected at all flow rates, except for one IXC that will be further investigated during Batch 3 processing. The TCCR system has shown some opportunities for more efficient processing during the length of the demonstration so far. For future processing of material from Tank 9H through Tank 10H and the TCCR unit, TCCR 1A will implement changes to the prefilters and the IXCs. The prefilters will have an increased surface area and a new filter media in an effort to increase time between filter swaps and improve backwashing cleaning capability. The IXCs will have a reduced diameter to allow for increased heat transfer out of the column and increased loading of Cs-137. Additionally, a new form of CST with an increased kinetic performance is being investigated for use during TCCR 1A operation. (authors)
The Saltstone Production Facility (SPF) immobilizes and disposes of low-level radioactive and hazardous liquid waste (salt solution) remaining from the processing of radioactive material at the Savannah River Site (SRS). Low-level waste (LLW) streams from processes at SRS are stored in Tank 50 until the LLW can be transferred to the SPF for treatment and disposal. The Salt Feed Tank (SFT) at the Saltstone Production Facility (SPF) holds approximately 6500 gallons of low level waste from Tank 50 as well as drain water returned from the Saltstone Disposal Facility (SDF) vaults. Over the past several years, Saltstone Engineering has noted the accumulation of solids in the SFT. The solids are causing issues with pump performance, agitator performance, density/level monitoring, as well as taking up volume in the tank. The tank has been sounded at the same location multiple times to determine the level of the solids. The readings have been 12, 25 and 15 inches. The SFT is 8.5 feet high and 12 feet in diameter, therefore the solids account for approximately 10 % of the tank volume. Saltstone Engineering has unsuccessfully attempted to obtain scrape samples of the solids for analysis. As a result, Savannah River National Laboratory (SRNL) was tasked with developing a soft core sampler to obtain a sample of the solids and to analyze the core sample to aid in determining a path forward for removing the solids from the SFT. The source of the material in the SFT is the drain water return system where excess liquid from the Saltstone disposal vaults is pumped back to the SFT for reprocessing. It has been shown that fresh grout from the vault enter the drain water system piping. Once these grout solids return to the SFT, they settle in the tank, set up, and can't be reprocessed, causing buildup in the tank over time. The composition of the material indicates that it is potentially toxic for chromium and mercury and the primary radionuclide is cesium-137. Qualitative measurements show that the material is not cohesive and will break apart with some force.
Tank 15 is one of 43 active underground waste storage tanks in the Liquid Waste System at SRS. After receiving and storing waste containing insoluble solids (sludge), receipts stopped, and the free supernate evaporated beginning in the 1980's. Tank 15 remained a dry sludge tank for almost 30 years. In 2015, the Tank 15 sludge was rehydrated as the first step of waste removal. Submersible Mixer Pumps (SMPs) were then installed in four risers, as well as other equipment upgrades, to facilitate sludge solids removal. SMPs were successfully used in prior waste removal campaigns, but this was the first deployment of four in a tank. Automated controls were also added to better monitor and protect the SMPs during operation in the more challenging H-area sludge. Waste removal plans consisted of multiple campaigns where sludge was suspended, transferred to Tank 13 and supernate decanted back to Tank 15 for reuse after allowing solids to settle. During the first campaign, sludge solids did not settle at an adequate rate to support the operating plan, and flammability evaluations were impacted. The SMPs also experienced operational challenges. New control logic for the SMPs was useful for early identification of issues; however, some equipment failures still occurred. These challenges were overcome using a combination of additional analyses, operating plan and procedure changes, and tank mapping techniques resulting in the successful removal of the bulk of the sludge from Tank 15. (authors)
Hardware queues are import in many applications, such as data transfer, synchronization of concurrent modules with the need of mutual exclusion constructs. State of the art bounded (of a fixed size) lock free circular queues are implemented either by read/write atomic operations, or barrier conditions, or by separating dequeue and enqueue operations. However, these queues always require an unused element at all the times to safe-guard the front and rear pointers of the queue, so as to avoid data race conditions, which leads to the waste of memory. The waste of memory is especially disadvantageous in applications such as I/O data transfer, and image transfer between processing filters, when large element size is needed, We propose a lock-free solution of the bounded circular queue through read/write atomic operations, but without the need of an extra element in the queue. The proposed solution is implemented and verified in both Verilog and ’C’ languages. We also demonstrate its effectiveness by comparing its area and delay metrics with the implementations of other existing designs of queue.
The Tank Closure Cesium Removal (TCCR) process at the Savannah River Site (SRS) is currently processing waste from Tank 10H, generating decontaminated salt solution that is sent to Tank 11H. Transfers out of Tank 11H are then disposed of on site as a grout wasteform in the SRS Saltstone Processing Facility. This is a full-scale demonstration of the use of Crystalline Silicotitanate (CST) for the decontamination of aqueous nuclear waste supernate. CST primarily removes cesium from the high sodium alkaline feed, but also adsorbs strontium, actinides, and some other trace metals. The TCCR unit is an at-tank set of four ion exchange columns, with two typically being operated in series during waste processing. Tank 10H is serving a dual function as both the salt dissolution tank as well as the feed tank for the TCCR system. Prior to operation of TCCR, Tank 10H must undergo dissolution campaigns, dissolving the salt cake to form an aqueous salt solution (supernate). After each dissolution campaign, the supernate created must be qualified prior to processing through the TCCR system. Qualification includes detailed characterization of the supernate as well as in-tank batch contact tests to determine the equilibrium loading of Cs-137 on the CST ion exchange media (IONSIV{sup TM} R9120-Ba). In support of the in-tank batch contact testing, Savannah River National Laboratory (SRNL) developed a CST sample holder, also referred to as a 'tea-bag', to hold a measured amount (∼0.1 grams) of pretreated CST between stainless steel screens. The tea-bag fits within a standard stainless steel sample vial that has been modified to allow the flow of supernate to the CST. This sample holder is then immersed in the waste tank allowing free contact between the CST beads and the surrounding liquid, at an effectively infinite liquid-to-solid phase ratio. A pair of tea-bags are deployed in the tank for a period of 10 days for each batch, after which the CST is recovered, digested, and analyzed to determine the loading of cesium isotopes on the CST. This result then supports the TCCR column thermal loading limits for processing as part of the TCCR safety basis. This paper will discuss the design and testing of the CST sample holder, as well as results from the first few batches of waste qualified and subsequently processed through the TCCR system. In addition, results from analyses of Tank 11H samples, showing decontamination of the product will be included. (authors)
Advances in simulation of coupled heat and mass transfer in granular salt using the Finite Element Heat and Mass Transfer Code (FEHM) is presented in this paper. The modeling focuses on reproducing the results of an in-situ subsurface evaporation experiment beginning during the summer of 2015 carried out in the underground at the Waste Isolation Pilot Plant (WIPP). Run-of-mine (granular) salt was placed in a metal pan, and sample mass, as well as ambient temperature and humidity, were measured continuously on a data logger through May 2016. Changes in moisture content of the salt are interpreted from the change in mass, and these data provide insight into evaporation and water retention characteristics from salt in the WIPP underground. To allow simulation of the experimental data, FEHM was modified to include a variable humidity boundary condition. The new humidity boundary condition can adjust streams of inflowing air to reflect fractions of air and water at prescribed temperature and pressure conditions that in turn represent relative humidity. Attributes for the flow such as temperature, air mass fraction, and humidity can be specified for inflowing air. The in-situ experiment provided data for validation of FEHM code development to accurately compute water-vapor pressure lowering due to dissolved salt and capillary pressure for pore water stored granular WIPP salt. Using this new boundary condition capabilities, a numerical representation of the pan experiment was created. The numerical domain is a 2-dimensional square grid of 30 by 30 cm axes. Salt and air properties are assigned based on measured values. A low flow rate of air is specified across the pile of salt to represent drift ventilation airflow. Humidity and temperature of the flowing air are fixed to measured values, and the simulated pile of salt is allowed to take up or release water based only on gradients in humidity and water-vapor pressure. Mass changes of the modeled salt pan were compared to measured mass changes in the bench-scale experiment. Directionality of the modeled changes tracks well with measured values, with good correlation between peak and valley inflection points. This shows that in situations with airflow where the relative humidity boundary can apply, FEHM can produce accurate values if humidity, air flow, and initial conditions are well established. Finally, this work supports phased-field-thermal testing in salt planned for the underground at WIPP that is expected to begin in early 2017. (authors)
Abstract Nanoscale zero‐valent iron (nZVI) loaded on tea residue biochar activated with phosphoric acid composites was synthesized. This adsorbent can efficiently adsorb multiple heavy metal ions such as Pb (II), Cr (III), Cu (II), and Ni (II) simultaneously. The adsorption process was found to follow the pseudo‐second‐order kinetic model. Equilibrium adsorption capacities reached 288.18, 162.34, 242.72, and 267.22 mg·g −1 for Pb(II), Cr(III), Cu(II), and Ni(II), respectively. The common interfering ions exhibited lower influence on its adsorption capacity. The adsorption mechanism includes reduction, ion exchange, surface complexation and precipitation. Furthermore, the waste adsorbent was converted into a catalyst in situ, which could catalyze the degradation of Congo red by establishing an electron transfer pathway. This research provides a new concept for the sustainable utilization of industrial tea residues and the harmless treatment of waste nZVI composite adsorbents.
Before a sample of the Slurry Mix Evaporator (SME) can be taken, the SME product sampling procedure requires that the agitator power be stable between 20 and 30 kW for at least one hour. The SME transfer to Melter Feed Tank (MFT) procedure also requires the SME agitator power be stabilized between 20 and 30 kW prior to transfer. These requirements are specified to ensure samples are homogeneous, as discussed in the Waste Form Qualification Report. During SME Batch 804, the agitator power dropped to 19 kW and struggled to achieve and maintain 20 kW. It was later discovered that the cause of the power drop was due to one of the bottom blades of the agitator breaking off. Both the sample and the transfer occurred without the power stabilizing between 20 and 30 kW. Therefore, the quality of SME Batch 804 is indeterminate, and homogeneity was questionable.